ML100351157
| ML100351157 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 09/21/2009 |
| From: | Division of Reactor Safety III |
| To: | |
| References | |
| Download: ML100351157 (224) | |
Text
QF-1030-02 Rev. 4 WRITTEN/ORAL EXAMINATION COVERSHEET Trainee Name:
Employee Number: Site: DAEC Examination Number/Title: Series A, Rev. 0, 2009 NRC Reactor Operator Initial License Exam Training Program: Initial License Training Course/Lesson Plan Number(s): 50007 / Various Total Points Possible: 75 PASS CRITERIA: 80% Grade: /75 = %
Graded by: Date:
Co-graded by (if necessary): Date:
EXAMINATION RULES
- 1. References may not be used during this examination, unless otherwise stated.
- 2. Read each question carefully before answering. If you have any questions or need clarification during the examination, contact the examination proctor.
- 3. Conversation with other trainees during the examination is prohibited.
- 4. Partial credit will not be considered, unless otherwise stated. Show all work and state all assumptions when partial credit may be given.
- 5. Rest room trips are limited and only one examinee at a time may leave.
- 6. For exams with time limits, you have 360 (6 Hours) minutes to complete the examination.
- 7. Feedback on this exam may be documented on QF-1040-13, Exam Feedback Form. Contact Instructor to obtain a copy of the form.
EXAMINATION INTEGRITY STATEMENT Cheating or compromising the exam will result in disciplinary actions up to and including termination.
I acknowledge that I am aware of the Examination Rules stated above. Further, I have not given, received, or observed any aid or information regarding this examination prior to or during its administration that could compromise this examination.
Examinees Signature: Date:
REVIEW ACKNOWLEDGEMENT I acknowledge that the correct answers to the exam questions were indicated to me following the completion of the exam. I have had the opportunity to review the examination questions with the instructor to ensure my understanding.
Examinees Signature: Date:
Retention: 6 years Retain in: Training Records 2009 RO NRC Master 8-10-09.doc
1 Point
- 1. Following a normal plant shutdown, a Shutdown Cooling (SDC) system startup has commenced IAW OI 149, RHR System.
Which one of the following describes a constraint associated with the SDC startup?
- a. The RHR loop temperature should be cooled to 100°F to minimize thermal shock to the Recirc-RHR loop intertie.
- b. The Recirc Pump Discharge Valves are normally left open to keep the Recirc loop warm.
- c. The Recirc Pump Discharge BYPASS Valves are normally left open to keep the Recirc loop warm and need not be closed.
- d. A Recirc Pump Suction Valve may be closed instead of the respective pumps Discharge BYPASS Valve if Reactor Coolant temperature is <212°F.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 1 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 205000 K1.03 Importance Rating 3.4 Knowledge of the physical connections and/or cause- effect relationships between SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and the following:
Recirculation loop temperature Proposed Question: RO Question # 1 Proposed Answer: C A: Incorrect - This would be a misconception of the relationship. The RHR loop temperature must be warmed to 150 degrees to minimize shock B: Incorrect - Per the OI, during SDC startup, the Recirc Pump discharge valves in both Recirc loops shall remain closed to prevent SDC flow from bypassing the core.
C: Correct - Per OI 149 Note at step 5.5 (48) (a) - The Recirc Pump discharge valves in both Recirc loops shall remain closed to prevent SDC flow from bypassing the core.
The Recirc Pump discharge bypass valves are normally left open to keep the Recirc loop warm and need not be closed. A Recirc Pump suction valve may be closed instead of the respective pumps discharge valve if Reactor coolant temperature is
<212°F.
D: Incorrect - A Recirc Pump Suction Valve may be closed instead of the respective pumps Discharge Valve if Reactor Coolant temperature is <212°F Technical Reference(s): OI 149 rev 110 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam:
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 2 Exam Series A
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3, 10 55.43 (3) Mechanical components and design features of the reactor primary system.
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 3 Exam Series A
1 Point
- 2. The plant is operating at near full power with all systems in a normal lineup.
The equalizing valve on the level transmitter selected for input to the feedwater level control system vibrates open.
With no operator action, what will be the response of the plant and why?
- a. Reactor water level will rise to 211 inches causing a Main Turbine trip and reactor scram.
This is due to the level transmitter inputting a low level signal to the feedwater level control system causing the FRVs to open and increase flow.
- b. Reactor water level will increase but stabilize at less than 211 inches due to feed flow/steam flow mismatch.
- c. Reactor Vessel level will decrease and the reactor will scram at 170 inches. This is due to the level transmitter inputting a high level signal to the feedwater level control system causing the FRVs to close and decrease flow.
- d. Reactor water level will decrease but stabilize at greater than 170 inches due to feed flow/steam flow mismatch.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 4 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 259002 K1.01 Importance Rating 3.8 Knowledge of the physical connections and/or cause- effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: RPS Proposed Question: RO Question # 2 Proposed Answer: C A: Incorrect - The equalizing valve opening will cause the level input to the FW level control system to sense a high level. This would cause actual level to lower.
B: Incorrect - The equalizing valve opening will cause the level input to the FW level control system to sense a rising high level. This would cause level to lower.
C: Correct - SD 644, page 14,Opening the equalizing valve on the transmitter, a DP cell, would cause the transmitter to sense a high level and provide that input to the FRVs thru the Master level Controller summer circuit. Level would continue to lower until the RPS scram setpoint was reached (170 inches)
D: Incorrect - With the system continuing to sense a high level, vessel level would continue to lower and would not stabilize GFES Comp Chap 7, pg 20 Technical Reference(s): (Attach if not previously provided)
SD 644, Rev.9, page 39 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 5 Exam Series A
55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 6 Exam Series A
1 Point
- 3. Which one of the following describes the power supply arrangement for the ADS Logic?
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 7 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 218000 K2.01 Importance Rating 3.1 Knowledge of electrical power supplies to the following: ADS logic Proposed Question: RO Question # 3 Proposed Answer: B A: Incorrect - ADS SRV Logic A has NO backup power supply B: Correct - SD 183.1, page 22 - Normal and backup 125 VDC power for the ADS logic circuits and operation of the Safety/Relief Valves is provided from the two plant 125 VDC battery systems. 125 VDC battery 1D1 normally supplies power for ADS logic A and for all Safety/Relief Valves except LLS valve PSV-4407. 125 VDC battery 1D2 normally supplies power for ADS logic B and for LLS valve PSV-4407. Except for ADS logic A, loss of the normal 125 VDC power supply will deenergize a relay and automatically shift to the other 125 VDC supply as a backup. ADS logic A does not have a backup 125 VDC supply.
C: Incorrect - ADS SRV Logic A has NO backup power supply D: Incorrect - ADS SRV Logic B normal power supply is backed up by LLS SRV Logic A normal power supply ( same as ADS SRV Logic A normal power supply)
SD-183.1 Rev 7. page 22 Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam: 2007 NRC Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 8 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 9 Exam Series A
1 Point
- 4. Which one of the following power supply loss or losses would cause BOTH an APRM "C" INOP trip and a loss of the APRM "C" trip indication on Panel 1C05?
A loss of power from ______.
- a. 120 VAC from RPS Bus A".
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 10 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215005 K2.02 Importance Rating 2.6 Knowledge of electrical power supplies to the following: APRM channels Proposed Question: RO Question # 4 Proposed Answer: B A: Incorrect - 120 VAC Instrument AC Control Power provides trip indication at the panel B: Correct - 120 VAC from RPS Bus A supplies APRM Channels A, C, and E, LPRM Group B and Flow Units A and C. 120 VAC power from the Instrument AC Control Power system is used for LPRM meter lamps; for the four-rod display; trip, bypass, and inoperative indication and two IRM/APRM recorders at 1C-05.
C: Incorrect - 120 VAC Uninterruptible AC Control Power supplies recorder power for the 2 recorders not powered by Instrument AC D: Incorrect - 120 VAC from RPS Bus A" causes an APRM INOP trip.
SD-878.3, Rev.11, page 41 Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 11 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 12 Exam Series A
1 Point
- 5. The plant is at full power with all equipment in a normal lineup.
A partial loss of GSW occurs. No operator actions have been taken.
Which one of the following describes:
(1) loads DIRECTLY cooled by GSW, AND (2) conditions under which AOP 411-GSW Abnormal Operation, requires a Fast Power Reduction
- b. (1) Drywell Coolers (2) Drywell Cooler outlet temperature is rising faster than GSW inlet temperature is rising.
- c. (1) Main Generator Hydrogen Coolers (2) Main Generator Hydrogen Cooler temperature is rising and cannot be maintained.
- d. (1) Recirc Pump Motor Coolers (2) Recirc Pump Motor Cooler outlet temperatures reach 250°F.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 13 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 400000 K3.01 Importance Rating 2.9 Knowledge of the effect that a loss or malfunction of the CCWS will have on the following:
Loads cooled by CCWS Proposed Question: RO Question # 5 Proposed Answer: C A: Incorrect - This would require a reactor scram B: Incorrect - GSW does not provide drywell cooling C: Correct - Per AOP 411 page 22, - Follow up step 11 D: Incorrect - GSW does not directly provide cooling to the Recirc Pump Motor Coolers Technical Reference(s): AOP 411, rev 22 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: (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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4, 7, 10 55.43 (4) Secondary coolant and auxiliary systems that affect the facility.
(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 14 Exam Series A
1 Point
- 6. The plant is operating at 100% when annunciator 1C03A(C-7) SRV BELLOWS FAILURE is activated. At 1C21, the Bellows Integrity light is OFF for PSV-4405 ADS Relief Valve.
Which one of the following statements describes how PSV-4405 functions to control reactor pressure with this condition?
- a. PSV-4405 will NOT open if ADS is initiated but will open on its safety setpoint of 1140 psig.
- b. PSV-4405 will NOT open on its safety setpoint of 1140 psig but will open if ADS is initiated.
- c. PSV-4405 will NOT open on either its safety setpoint of 1140 psig or if ADS is initiated.
- d. PSV-4405 will open on its safety setpoint of 1140 psig or if ADS is initiated.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 15 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 239002 K3.01 Importance Rating 3.9 Knowledge of the effect that a loss or malfunction of the RELIEF/SAFETY VALVES will have on following: Reactor pressure control Proposed Question: RO Question # 6 Proposed Answer: B A: Incorrect - The ADS function will still be operable B: Correct - Per ARP 1C03A(C-7), Bellows failure does not necessarily mean the valve is leaking; however, it does make the safety (mechanical) actuation portion of the relief valve inoperative.
C: Incorrect - The ADS function will still be operable D: Incorrect - The safety relief function will not operate Technical Reference(s): ARP 1C03A(C-7) rev 48 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 16 Exam Series A
55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 17 Exam Series A
1 Point
- 7. The plant is at 100% power. All equipment is operable.
At T = 0 minutes - A grid disturbance occurs resulting in the diesel generator start logics sensing an Essential Bus undervoltage signal of <65% of rated bus voltage.
At T = 2 minutes - All Offsite Power is lost.
At T = 4 minutes - Drywell Pressure reaches 2.0 psig.
Which one of the following describes when the Diesel Generators, RHR pumps and Core Spray pumps will start?
- a. The Diesel Generators will start at T=0 The RHR and Core Spray pumps will sequence on after T=4
- b. The Diesel Generators will start at T=0 The RHR and Core Spray pumps will all start at T=4
- c. The Diesel Generators will start at T=2 The RHR and Core Spray pumps will sequence on after T=4
- d. The Diesel Generators will start at T=2 The RHR and Core Spray pumps will all start at T=4 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 18 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 264000 K4.05 Importance Rating 3.2 Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide for the following: Load shedding and sequencing Proposed Question: RO Question # 7 Proposed Answer: A A: Correct - Per SD 149 page 20, If a LOCA signal exists, the RHR and Core Spray pumps will automatically sequence onto the 1A3[1A4] once power is restored. Power will be restored before the LOOP occurs and before the LPCI signal occurs Per SD 324, page 33 - DGS will auto start on Loss of Offsite Power (LOOP) signals of
<65% of rated voltage on the secondaries of both the Startup and Standby transformers or degraded essential bus voltage of <92.5% of rated voltage for 8-8.5 seconds.
B: Incorrect - The RHR and Core Spray pumps will sequence on and not start immediately C:
Incorrect - The DGs will on start on the essential bus undervoltage at T=0, They will already be running when and loaded on the bus when all Offsite Power is lost.
D: Incorrect - The DGs will on start on the essential bus undervoltage at T=0, They will already be running when and loaded on the bus when all Offsite Power is lost.
The RHR and Core Spray pumps will sequence on and not start immediately SD 324 Rev 10 Technical Reference(s): (Attach if not previously provided)
SD 149 Rev 11 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam:
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 19 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 20 Exam Series A
1 Point
- 8. With the Reactor Mode Switch in REFUEL, which one of the following conditions will result in Source Range Monitors producing a Reactor Scram?
- a. ONE SRM Channel indicating >5 x 105 counts per second with the Shorting Links removed.
- b. ONE SRM Channel indicating >5 x 105 counts per second with the Shorting Links installed.
- c. At least 2 SRM Channels indicating >1 x 105 counts per second with the Shorting Links removed.
- d. At least 2 SRM Channels indicating >1 x 105 counts per second with the Shorting Links installed.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 21 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215004 K4.02 Importance Rating 3.4 Knowledge of SOURCE RANGE MONITOR (SRM) SYSTEM design feature(s) and/or interlocks which provide for the following: Reactor SCRAM signals Proposed Question: RO Question # 8 Proposed Answer: A A: Correct - The SRM upscale trip occurs if any channel counts exceed 5x105 cps. A light is energized on the SRM drawer and signals are sent to a 1C05 indicator. The upscale trip unit also generates a scram signal for use by the Reactor Protective System. This scram signal is only functional during initial fuel loading and low power physics testing.
At other times, this scram is removed by the installed shorting links.
B: Incorrect - the shorting links must be removed C: Incorrect - 1 x 105 Is the upscale alarm and not the scram setpoint D: Incorrect - the shorting links must be removed, 1 x 105 Is the upscale alarm and not the scram setpoint SD 878.1 rev 6 page 18 Technical Reference(s): (Attach if not previously provided)
SD 358 rev 7 page 13 & 14 Proposed References to be provided to applicants during examination: NONE Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 22 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 23 Exam Series A
1 Point
- 9. HPCI is being started for the quarterly full flow test surveillance. HPCI has reached the 2000 gpm flow rate when the ramp generator fails to its low limit.
Which one of the following describes the response of the HPCI System?
- a. HPCI speed and flow lower.
- b. HPCI trips due to a loss of reference signal.
- c. HPCI will be unaffected while in automatic ONLY.
- d. HPCI remains at the same speed and flow.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 24 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 206000 K5.06 Importance Rating 2.6 Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM : Turbine speed measurement: BWR-2,3,4 Proposed Question: RO Question # 9 Proposed Answer: A A: Correct - Per SD 152, page 10 - The outputs of the flow controller and ramp generator are applied to a low value signal selector, which passes the lower of the two signals.
Because the ramp generator output is less than the flow controller output, the HPCI turbine speed will be controlled solely by the ramp generator during startup.
Throughout the entire startup transient, the flow controller output calls for speed consistent with the flow controller setting until such time that pump flow reaches the setpoint of the flow controller and the ramp function signal exceeds the flow controller signal.
B: Incorrect - The reference signal is not lost. The signal now going through the LVG will cause speed and flow to decrease C: Incorrect - The ramp generator and the flow error signals feed into a LVG. The lower of the two signals is passed to the turbine control valve. If the ramp generator fails low, this low signal will pass to the TCV and it will close causing decreased speed and flow whether in manual or automatic.
D: Incorrect - Both speed and flow will decrease.
Technical Reference(s): SD 152 Rev 10 page 10 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Question Source: Bank # X - HC 2007 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 25 Exam Series A
10 CFR Part 55 Content: 55.41 5, 7 55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 26 Exam Series A
1 Point
- 10. Complete the following statements:
The IRM detectors operate in the ___(1)___ region of the gas amplification curve. In addition, a decrease in IRM detector argon gas pressure will cause the IRM detectors to be ___(2)___ sensitive.
- a. (1) proportional (2) more
- b. (1) proportional (2) less
- c. (1) ionization (2) more
- d. (1) ionization (2) less Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 27 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215003 K5.01 Importance Rating 2.6 Knowledge of the operational implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM) SYSTEM : Detector operation Proposed Question: RO Question # 10 Proposed Answer: D A: Incorrect: Proportional region is incorrect. In addition, argon is used as a detector ionization gas, so reduced argon gas pressure will yield less ionization events; therefore, the detector is less sensitive.
Plausible: The Fuel Loading Chambers used with SRMs for initial loading operated in the proportional region, but are no longer used. In addition, argon could be confused for a quench gas, and if it did act as a quench gas, then the detector would be more sensitive with less quench gas.
B: Incorrect: Proportional region is incorrect.
Plausible: The Fuel Loading Chambers used with SRMs for initial loading operated in the proportional region, but are no longer used. In addition, the second half of the answer is correct.
C: Incorrect: Argon is used as a detector ionization gas, so reduced argon gas pressure will yield less ionization events; therefore, the detector is less sensitive.
Plausible: Ionization region is correct. In addition, argon could be confused for a quench gas, and if it did act as a quench gas, then the detector would be more sensitive with less quench gas.
D: Correct: IRM detectors operate in the ionization region. Argon is used as a detector ionization gas, so reduced argon gas pressure will yield less ionization events; therefore, the detector is less sensitive.
Technical Reference(s): SD 878.1, Rev 6, page 11 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: IRM 78.1.1.4 (As available)
Question Source: Bank # X - Pilgrim 2007 Modified Bank # (Note changes or attach parent)
New Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 28 Exam Series A
Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2, 5, 7 55.43 (2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.
(5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 29 Exam Series A
1 Point
- 11. The plant is in normal full power operation with no equipment out of service.
Which of the following describes the plant response if all 250 VDC were to be lost?
- a. Both Instrument AC buses would transfer their respective Regulating Transformers and the plant would remain stable.
- b. Both Instrument AC buses would be lost and the plant would scram due to loss of all Instrument AC power.
- c. The Uninterruptible AC bus would transfer to the Regulating Transformer and the plant would remain stable.
- d. The Uninterruptible AC bus would be lost and the plant would scram due to loss of all Uninterruptible AC power.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 30 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 262002 K6.02 Importance Rating 2.8 Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) : D.C. electrical power Proposed Question: RO Question # 11 Proposed Answer: C A: Incorrect - Instrument AC would be unaffected. The Inst AC Regulating transformers are supplied by 480 VAC.
B: Incorrect - Instrument AC would be unaffected. The Inst AC Regulating transformers are supplied by 480 VAC.
C: Correct - The Uninterruptable AC regulating transformers would supply the bus thru the static switch upon loss of all 250 VDC (see SD 357 Figure page 5)
D: Incorrect - Uninterruptible AC power would not be lost . It would be supplied via 480 VAC thru Uninterruptable AC Regulating transformer.
SD -357, Rev 7, Figure 1 Technical Reference(s): (Attach if not previously provided)
AOP 388, Rev 18, pg 3 (AutoAct)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X - DAEC Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 31 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 32 Exam Series A
1 Point
- 12. A plant event occurred resulting in the need to use Standby Liquid Control (SBLC) as an injection source.
A loss of Instrument AC 1Y11 occurs before the operator attempts to place SBLC in service.
How will the SBLC system and indications respond?
- a. Both pumps will start, the squib valves will fire and flow indication will be zero.
- b. Both pumps will start but flow indication will be zero because the squib valves will not fire.
- c. Both pumps will start, the squib valves will fire and flow indication will be as expected for 2 pump operation.
- d. Neither pump will start due to a loss of control power and the squib valves loss of continuity lights will be illuminated.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 33 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 211000 K6.03 Importance Rating 3.2 Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY LIQUID CONTROL SYSTEM : A.C. power Proposed Question: RO Question # 12 Proposed Answer: A A: Correct - Per SD 153, page 26 - The Instrument AC System provides power for several of the SBLC instruments: SBLC Storage Tank Level (LI-2600A), SBLC Pump Discharge Pressure (PI-2605), SBLC System Flow (FI-2620), Injection Valve Position (V26-0032).
Therefore, flow indication will fail to zero. The pumps & squib valves have not lost power (powered from 1B34 and 1B44)
B: Incorrect - The pumps & squib valves have not lost power (powered from 1B34 and 1B44)
C: Incorrect - flow indication will be zero D: Incorrect - Both pumps will start, the squib valves will fire.
Technical Reference(s): SD 153, Rev 7, page 26 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 34 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 35 Exam Series A
1 Point
- 13. The plant was operating at full power when an event occurred which resulted in a manual scram.
- The CRS determined that RCIC was not required, ordered it to be tripped, and the operator performed the RCIC QRC actions for tripping the RCIC turbine.
- With RPV level at 175 inches and slowly lowering, the operators attempted to re-start a RFP and HPCI but were not successful.
Which one of the following describes actions that must be performed to restart RCIC injection?
- a. MO-2405, RCIC TURBINE STOP VALVE must be taken to CLOSE, then to OPEN AND MO-2404, TURBINE STEAM SUPPLY must be OPENED.
- b. MO-2405, RCIC TURBINE STOP VALVE must be locally reset AND MO-2404, TURBINE STEAM SUPPLY must be OPENED.
- c. HS-2482, RCIC INITIATION SEALED IN RESET pushbutton must be depressed AND MO-2405, RCIC TURBINE STOP VALVE must be locally reset.
- d. HS-2482, RCIC INITIATION SEALED IN RESET pushbutton must be depressed AND MO-2405, RCIC TURBINE STOP VALVE must be taken to CLOSE, then to OPEN.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 36 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 217000 A1.01 Importance Rating 3.7 Ability to predict and/or monitor changes associated with operating the REACTOR CORE ISOLATION COOLING SYSTEM: RCIC flow Proposed Question: RO Question # 13 Proposed Answer: A A: Correct - Per SD 150, page 32 - If RCIC is in operation, and reactor water level reaches 211" MO-2404 shuts. The valve automatically closes at an RPV water level of 211". The MO-2404 must then be reopened for injection unless a RCIC system initiation signal (119.5) is received, in which case it automatically opens.
The MO 2405 was closed when RCIC was tripped per the QRC and must be re-opened.
B: Incorrect - The RCIC turbine did not trip on mechanical overspeed therefore the stop valve need not be locally reset. (See SD-150 page 17)
C: Incorrect - The initiation logic does not need a reset to permit RCIC injection AND The RCIC turbine did not trip on mechanical overspeed therefore the stop valve need not be locally reset. (See SD-150 page 17)
D: Incorrect - The initiation logic does not need a reset to permit RCIC injection AND the MO -2404 must also be opened.
SD 150, Rev 6, pages 17 and 32 Technical Reference(s): (Attach if not previously provided)
OI 150 QRC 1 Proposed References to be provided to applicants during examination: NONE Learning Objective: RCIC 3.8.1.2 (As available)
Question Source: Bank # X - 20379 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 37 Exam Series A
10 CFR Part 55 Content: 55.41 5, 7 55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 38 Exam Series A
1 Point
- 14. The Div 1 125 VDC battery charger is being operated in the equalize mode.
Which one of the following describes: (1) the voltage relationship between the charger and the batteries and (2) the design rating of the batteries if a loss of AC power occurred?
- a. (1) In equalize, the charger output to the battery will be a higher voltage than when in the float mode (2) The 125 VDC batteries are sized to supply emergency power for a 4-hour time period.
- b. (1) In equalize, the charger output to the battery will be a lower voltage than when in the float mode (2) The 125 VDC batteries are sized to supply emergency power for a 4-hour time period.
- c. (1) In equalize, the charger output to the battery will be a higher voltage than when in the float mode (2) The 125 VDC batteries are sized to supply emergency power for an 8-hour time period.
- d. (1) In equalize, the charger output to the battery will be a lower voltage than when in the float mode (2) The 125 VDC batteries are sized to supply emergency power for an 8-hour time period.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 39 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 263000 A1.01 Importance Rating 2.5 Ability to predict and/or monitor changes associated with operating the D.C. ELECTRICAL DISTRIBUTION controls including: Battery charging/discharging rate Proposed Question: RO Question # 14 Proposed Answer: A A: Correct - Per SD 375, pages 7 and 17. Each charger can be placed in the equalizing mode with a switch on the battery charger. When in the equalizing mode, the charger output will be a higher voltage than when in the float mode.
The Plant 125v DC Power Supply System consists of two 125 VDC batteries each provided with its own charger and sized to supply emergency power for a 4-hour time period.
B: Incorrect - in equalize the charger output to the battery will be a higher (not lower) voltage than when in the float mode C: Incorrect - design is for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D: Incorrect - design is for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and - in equalize the charger output to the battery will be a higher (not lower) voltage than when in the float mode Technical Reference(s): SD 375, Rev 7, pages 7 and 17 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 40 Exam Series A
10 CFR Part 55 Content: 55.41 5, 7 55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 41 Exam Series A
1 Point
- 15. The plant is operating at 100% power. All systems are operable.
Instrument Air Dryer 1T-265A was in service when control power to the dryers was momentarily lost while it automatically switched between 1L150 and 1L21.
Which one of the following describes:
(1) the effect on instrument air header pressure and (2) actions required, if any, IAW OI 518.1 Instrument, Service and Breathing Air Systems in regard to continued operation of the air dryer.
(2) Dispatch an operator to reset the dryer logic locally.
- b. (1) Instrument air header pressure may fluctuate slightly.
(2) Dispatch an operator to reset the dryer logic locally.
(2) No action is required because the dryer power handswitch HS-3046A was in ON.
- d. (1) Instrument air header pressure may fluctuate slightly.
(2) No action is required because the dryer power handswitch HS-3046A was in ON.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 42 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 300000 A2.01 Importance Rating 2.9 Ability to predict the impacts of the following on the INSTRUMENT AIR SYSTEM: and use procedures to correct, control, or mitigate the consequences of those abnormal operation: Air dryer and filter malfunctions Proposed Question: RO Question # 15 Proposed Answer: B A: Incorrect- Service Air header isolates at 82 psig. In this situation the loss of power places both drying chambers in service and pressure should not significantly lower B: Correct - Per SD 518, page 20 The air dryer would shut down when control power is lost. If control power to dryers will be momentarily lost while the control power automatically switches between 1L150 to 1L21 or vice versa, de-energize and then re-energize affected dryer by placing dryer power handswitch HS-3046A[B] in OFF for 5 seconds and then return to ON. This will reset dryer logic. The air dryer would shut down when control power is lost and header pressure would lower with no flowpath.
SD 518 page 20 - The Bypass Valve will automatically open on two conditions. If Instrument Air pressure downstream of the dryers falls to 85 psig, the dryer Bypass Valve CV-3026 automatically opens and control room annunciator 1C07B (B-10)
INSTRUMENT AIR DRYERS 1T-265A/B LO DISCH PRESSURE is activated. If differential pressure across the dryer units reaches 15 psid, the dryer Bypass Valve CV-3026 automatically opens and control room annunciator 1C07B (C-10) INSTRUMENT AIR DRYERS 1T-265A/B HIP is activated.
C: Incorrect - Service Air header isolates at 82 psig. In this situation the loss of power places both drying chambers in service and pressure should not significantly lower.
Action is required to reset the dryer logic.
D: Incorrect - Action is required to reset the dryer logic.
Technical Reference(s): SD 518, Rev 8, page 20 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 43 Exam Series A
New X Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 7 55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 44 Exam Series A
1 Point
- 16. The plant was operating at 100% when a recirc line break occurred.
- Reactor pressure is at 410 psig and stable
- Drywell Pressure is at 3.4 psig and rising slowly
- Reactor level is at 60 inches and rising slowly
- Core Spray pumps are running
- Core Spray Valves INBD INJECT MO-2117 and MO-2137 are closed
- Core Spray MIN FLOW BYPASS VALVES MO-2104 and MO-2124 are open Which one of the following describes the response of the Core Spray System and actions required, if any, in regard to INBD INJECT VALVES and MIN FLOW BYPASS VALVES?
- a. The Core Spray Inboard Injection Valves should have opened and must be manually opened.
The Core Spray Min Flow Bypass Valves will auto-close ONLY when the Injection Valves are fully open.
- b. The Core Spray Inboard Injection Valves are closed and will open once reactor pressure lowers to below the shut off head of the Core Spray pumps.
The Core Spray Min Flow Bypass Valves will auto-close when Core Spray system flow reaches 600 gpm.
- c. The Core Spray Inboard Injection Valves should have opened and must be manually opened.
The Core Spray Min Flow Bypass Valves will auto-close when Core Spray system flow reaches 600 gpm.
- d. The Core Spray Inboard Injection Valves are closed and will open once reactor pressure lowers to below the shut off head of the Core Spray pumps.
The Core Spray Min Flow Bypass Valves will auto-close ONLY when the Injection Valves are fully open.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 45 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 209001 A2.08 Importance Rating 3.1 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 operation: Valve openings malfunctions Proposed Question: RO Question # 16 Proposed Answer: C A: Incorrect - The min flow bypass valve will close when system flow reaches 600 gpm B: Incorrect - The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating toverify they open.
C: Correct - OI 151, pages 6 and 7, steps 4.0 (2) and (3). When system flow reaches 600 gpm, as indicated on (A[B] CORE SPRAY PUMP) INJECT/TEST FLOW indicator FI-2110 [FI-2130] on Panel 1C03, verify MIN FLOW BYPASS MO-2104 [MO-2124]
valve CLOSES.
When reactor vessel pressure drops below the low pressure permissive setpoint of 450 psig, verify that the INBD INJECT MO-2117 [MO-2137] valves OPEN to inject to the reactor vessel.
The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating toverify they open.
D: Incorrect - The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating toverify they open.
The min flow bypass vlv will close when system flow reaches 600 gpm.
OI 151 Rev 57 steps 4.0 (2) and Technical Reference(s): (Attach if not previously provided)
(3)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam:
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 46 Exam Series A
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 7 55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 47 Exam Series A
1 Point
- 17. While operating at 75% power, the PB-5831A, "A" SBGT TEST push-button is depressed on SBGT control panel 1C24A.
Which of the following would automatically occur as a result of the above action?
- a. If the SBGT Mode Select Switch was in AUTO, the A SBGT would start, a secondary containment isolation would occur and normal Reactor Building Ventilation would isolate.
- b. If the SBGT Mode Select Switch was in AUTO, the A SBGT would start, a secondary containment isolation would NOT occur and Reactor Building Ventilation would NOT isolate.
- c. If the SBGT Mode Select Switch was in MANUAL, the A SBGT would start, a secondary containment isolation would occur and normal Reactor Building Ventilation would isolate.
- d. If the SBGT Mode Select Switch was in MANUAL, the A SBGT would start, a secondary containment isolation would NOT occur and Reactor Building Ventilation would NOT isolate.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 48 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 261000 A3.03 Importance Rating 3.0 Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including
- Valve operation Proposed Question: RO Question # 17 Proposed Answer: B A: Incorrect - a secondary containment isolation would NOT occur, Reactor Building Ventilation would NOT isolate.
B: Correct - Per SD 170, page 16 - The SBGT trains can also be manually initiated without causing any isolation using SBGT TEST pushbuttons PB-5831A (B) with the mode switch in AUTO. Use of the SBGT TEST pushbutton will initiate the associated SBGT train, however, will not initiate secondary containment isolation.
C: Incorrect - With the SBGT Mode Select Switch in MANUAL, no actions occur when the test PB is depressed.
D: Incorrect - With the SBGT Mode Select Switch in MANUAL, no actions occur when the test PB is depressed.
Technical Reference(s): SD 170, Rev 10 page16 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: SBGT 7.2.1.1 (As available)
Question Source: Bank # X - 19287 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 49 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 50 Exam Series A
1 Point
- 18. The plant is operating at full power. All systems are operable.
Then, the Fuel Pool Exhaust rad monitors alarm and both are reading 8.5 mr/hr.
Which one of the following describes PCIS Group(s) and valve(s) affected by this alarm?
- a. PCIS Group 3. The Well Water Drywell Cooling Water Supply and Return Valves close causing a rise in drywell temperatures.
- b. PCIS Groups 2 & 3. The drywell and torus sample supply valves isolate preventing any liquid or gaseous samples from being taken from the drywell or torus.
- c. PCIS Group 3. The drywell and torus sample supply valves isolate preventing any liquid or gaseous samples from being taken from the drywell or torus.
- d. PCIS Group 2. The RHR sample valves isolate which prevents sampling of torus water when in torus cooling or reactor water when in shutdown cooling.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 51 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 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 Proposed Answer: C A: Incorrect - This alarm produces a Group 3 isolation only. The jet pump sample valves isolate on a Group 3 B: Incorrect - This alarm produces a Group 3 isolation only. RWCU is a Group 5 isolation C: Correct - Per SD 959-1, page 10 table 1 - Fuel Pool Exhaust High Radiation, 8 mr/hr or Inop is a Group 3 isolation.
Group 3 includes a large number of valves, dampers, and fans. As direct isolations, this group includes the Primary Containment Atmosphere Control Valves for Drywell and Torus Ventilation and Purge; the Containment Nitrogen Compressor Suction and Discharge Valves; the Jet Pump Sample, Liquid Return, and Sample Station Exhaust PASS Valves; and Recirculation Pump Seal Purge Valves.
D: Incorrect - This alarm produces a Group 3 isolation only SD 959-1 Rev 8, page 10 table 1 Technical Reference(s): (Attach if not previously provided) and page 22 Proposed References to be provided to applicants during examination: NONE Learning Objective: PCIS 76.1.1.7 (As available)
Question Source: Bank # X - 20170 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 52 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 53 Exam Series A
1 Point
- 19. Which one of the following describes how the Scram Discharge Volume High Level trip may be bypassed?
- a. It must be bypassed by placing the keylocked High Water Level Bypass switch in BYPASS.
The mode switch may be in either the SHUTDOWN or REFUEL position.
- b. It must be bypassed by placing the keylocked High Water Level Bypass switch in BYPASS.
The mode switch must be in the REFUEL position ONLY.
- c. It is automatically bypassed when RPS is reset.
The mode switch may be in either the SHUTDOWN or REFUEL position.
- d. It is automatically bypassed when RPS is reset.
The mode switch must be in the REFUEL position ONLY.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 54 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 212000 A4.04 Importance Rating 3.9 Ability to manually operate and/or monitor in the control room: Bypass SCRAM instrument volume high level SCRAM signal Proposed Question: RO Question # 19 Proposed Answer: A A: Correct - Per ARP 1C05 - E1, The CRD Scram Discharge Volume High Water Level Scram trip is bypassed using a keylocked High Water Level Bypass switch if the Reactor Mode switch is in SHUTDOWN or REFUEL.
B: Incorrect - The mode switch may be in SHUTDOWN or REFUEL.
C: Incorrect - The keylock switch must in BYPASS.
D: Incorrect - The keylock switch must in BYPASS. The mode switch may be in SHUTDOWN or REFUEL.
Technical Reference(s): ARP 1C05B (E-1) Rev 81 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 7 55.43 Design, components, and function of control and safety systems, including instrumentation, Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 55 Exam Series A
signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 56 Exam Series A
1 Point
- 20. The A SBDG is being paralleled to 1A3.
- Incoming voltage is slightly higher than running voltage
- The synchroscope is rotating slowly in the clockwise direction The SBDG output breaker is closed when the synchroscope pointer reaches the 12 o'clock position.
Which one of the following will occur immediately after the breaker is closed?
- a. The SBDG will supply only MWe to the grid.
- b. The SBDG will supply both MWe and MVAR to the grid.
- c. The SBDG output breaker will trip open on reverse power.
- d. The SBDG output breaker will trip open on overcurrent.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 57 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 262001 A4.04 Importance Rating 3.6 Ability to manually operate and/or monitor in the control room: Synchronizing and paralleling of different A.C. supplies Proposed Question: RO Question # 20 Proposed Answer: B A: Incorrect - MVARs are also being supplied B: Correct - OI 304.2 Rev 77 & Theory portion of electrical fundamentals - current flows from the DG to the grid. Both MWE and MVARs will be supplied C: Incorrect - the output breaker will not trip open unless the difference between running (grid) and incoming (DG) exceeds specified values or the synchroscope is not close to the 12 oclock position. This answer is correct if the synch scope is running the counterclockwise direction D: Incorrect - the output breaker will not trip open unless the difference between running (grid) and incoming (DG) exceeds specified values or the synchroscope is not close to the 12 oclock position. This answer is correct if voltages are largely mismatched.
OI 304.2 Rev 77 Technical Reference(s): (Attach if not previously provided)
GFES Comp Ch 5, pg 51 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # ID 18809 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 58 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 59 Exam Series A
1 Point
- 21. The plant is operating at full power with all equipment operable. RHR Loop A is then placed in torus cooling in preparation for a surveillance test.
Which ONE of the following Technical Specification LCO(s) is required?
The LCO for _____ being inoperable.
- b. RHR Suppression Pool Spray
- c. RHR Suppression Pool Spray AND Low Pressure Coolant Injection
- d. RHR Suppression Pool Cooling AND Low Pressure Coolant Injection Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 60 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 203000 2.2.40 Importance Rating 3.4 Equipment Control:: Ability to apply Technical Specifications for a system.
Proposed Question: RO Question # 21 Proposed Answer: A A: Correct - Per OI 149 Continuous Recheck Statement at 5.4 - IF Torus Cooling is operating when LPCI is required to be Operable, THEN LPCI shall be declared inoperable and the Technical Specifications for ECCS-Operating and ECCS-Shutdown complied with.
B: Incorrect - PER OI 149 - LPCI is required to be declared inoperable C: Incorrect - ONLY the LPCI LCO must be entered D: Incorrect - ONLY the LPCI LCO must be entered Technical Reference(s): OI 149 Rev 110, Section 5.4 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # WTS Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 61 Exam Series A
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 62 Exam Series A
1 Point
- 22. The plant is at 30% power with a power ascension in progress. All equipment is operable.
Then, the following alarm annunciates:
- 1C05A (E-2) APRM FLOW UNIT UPSCALE, INOP OR COMPARE ERROR Investigation determines the alarm is due to a COMPARATOR error.
Which one of the following describes:
(1) the plant response to the alarm and (2) the back panel 1C37 indication(s)?
- a. (1) ONLY a Rod Block has occurred.
(2) At least 2 flow units would indicate a COMPARATOR alarm at back panel 1C37.
- b. (1) A Rod Block and a Half Scram have occurred.
(2) At least 2 flow units would indicate a COMPARATOR alarm at back panel 1C37.
- c. (1) ONLY a Rod Block has occurred.
(2) ONLY the flow unit with the flow mismatch would indicate a COMPARATOR alarm at back panel 1C37.
- d. (1) A Rod Block and a Half Scram have occurred.
(2) ONLY the flow unit with the flow mismatch would indicate a COMPARATOR alarm at back panel 1C37.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 63 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215005 2.4.46 Importance Rating 4.2 Emergency Procedures / Plan: Ability to verify that the alarms are consistent with the plant conditions. (APRMs)
Proposed Question: RO Question # 22 Proposed Answer: A A: Correct - Per 1C05A (E-2), A rod withdraw block occurs in all Modes of Operation. A flow signal mismatch which produces a compare error in one flow unit will normally produce a compare error in another (unbypassed) flow unit.
B: Incorrect - ONLY a Rod Block occurs C: Incorrect - A flow signal mismatch which produces a compare error in one flow unit will normally produce a compare error in another (unbypassed) flow unit.
D: Incorrect - ONLY a Rod Block occurs . A flow signal mismatch which produces a compare error in one flow unit will normally produce a compare error in another (unbypassed) flow unit.
Technical Reference(s): 1C05A (E-2) Rev 64 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 10 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 64 Exam Series A
55.43 (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 65 Exam Series A
1 Point
- 23. Which one of the following describes the relationship between HPCI and ventilation exhaust?
- a. The HPCI barometric condenser vacuum pump exhausts directly to the Offgas Stack, where any release is monitored. The OG STACK EXH FAN, 1V-EF-18A[B], can be off when HPCI is running.
- b. The HPCI barometric condenser vacuum pump exhausts directly to the Offgas Stack, where any release is monitored. At least one OG STACK EXH FAN, 1V-EF-18A[B], must be on when HPCI is running.
- c. The HPCI barometric condenser vacuum pump exhausts through the SBGT trains, where gases are treated before leaving the plant via the stack. Both SBGT EXHAUST FANs, 1V-EF-15A[B], can be off when HPCI is running.
- d. The HPCI barometric condenser vacuum pump exhausts through the SBGT trains, where gases are treated before leaving the plant via the stack. At least one SBGT EXHAUST FAN, 1V-EF-15A[B], must be on when HPCI is running.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 66 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 261000 K1.06 Importance Rating 3.0 Knowledge of the physical connections and/or cause effect relationships between STANDBY GAS TREATMENT SYSTEM and the following: High pressure coolant injection system; Plant-specific Proposed Question: RO Question # 23 Proposed Answer: C A: Incorrect - HPCI vacuum pump exhaust is directed to the SBGT system and 1V-EF-18A[B]-fan must be running.
B: Incorrect - HPCI vacuum pump exhaust is directed to the SBGT system C:
Correct - Per SD 152, page 16 - A vacuum pump maintains system vacuum by transferring non-condensible gases to the Standby Gas Treatment System. Per STP 3.5.1-05 rev 45 Section 6.0 (prereqs), at least one SBGT system has to be aligned for AUTO and one OG STACK EXH FAN, 1V-EF-18A[B], must be on when HPCI is running.
D: Incorrect - SBGT is not required to be in operation if HPCI is running Technical Reference(s): SD 152 Rev 10 page 16 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X - 20154 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 67 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 68 Exam Series A
1 Point
- 24. You have indications of an SRV leaking.
IAW AOP 683 Abnormal Safety Relief Valve Operation, which one of the following tailpipe temperatures would first indicate that the SRV is OPEN?
When tailpipe temperature reaches ______.
- a. 212 degrees F.
- b. 251 degrees F.
- c. 312 degrees F.
- d. 544 degrees F.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 69 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 218000 A1.01 Importance Rating 3.4 Ability to predict and/or monitor changes in parameters associated with operating the AUTOMATIC DEPRESSURIZATION SYSTEM controls including: ADS valve tail pipe temperatures Proposed Question: RO Question # 24 Proposed Answer: B A: Incorrect - This would indicate a leaking SRV.
B: Correct - Per AOP 683 NOTE page 3, An SRV or SV is considered OPEN based on the following: At 1C21- Tailpipe temperature at TR-4400 above 250ºF.
C: Incorrect - This would be the approximate temperature at the tailpipe pressure but IAW the AOP the SRV is considered open at >250 degrees F.
D: Incorrect - This temperature is not achievable given tailpipe pressure. It is the coolant temperature at 1000 psig.
Technical Reference(s): AOP 683 Rev 8 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: Steam Tables Learning Objective: (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 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 70 Exam Series A
55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 71 Exam Series A
1 Point
- 25. The plant was operating at 80% power when a condensate pump tripped. The plant was scrammed and RCIC is being placed in service manually.
As RCIC reaches rated flow, the following annunciators alarm:
- 1C04C-A8 - RCIC B LOGIC MAN/AUTO ISOL INITIATED What is the response of the RCIC system to this alarm?
- a. RCIC Turbine Stop Valve MO-2405 will close.
RCIC Turbine Steam Supply Valve MO-2404 remains open.
RCIC Inboard Steam Line Isolation Valve MO-2400 will close.
RCIC Outboard Steam Line Isolation Valve MO-2401 will close.
- b. RCIC Turbine Stop Valve MO-2405 will close.
RCIC Turbine Steam Supply Valve MO-2404 will close.
RCIC Inboard Steam Line Isolation Valve MO-2400 remains open.
RCIC Outboard Steam Line Isolation Valve MO-2401 remains open.
- c. RCIC Turbine Stop Valve MO-2405 remains open.
RCIC Turbine Steam Supply Valve MO-2404 remains open.
RCIC Inboard Steam Line Isolation Valve MO-2400 will close.
RCIC Outboard Steam Line Isolation Valve MO-2401 will close.
- d. RCIC Turbine Stop Valve MO-2405 remains open.
RCIC Turbine Steam Supply Valve MO-2404 will close.
RCIC Inboard Steam Line Isolation Valve MO-2400 remains open.
RCIC Outboard Steam Line Isolation Valve MO-2401 remains open.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 72 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 217000 A4.03 Importance Rating 3.4 (RCIC) Ability to manually operate and/or monitor in the control room: System valves Proposed Question: RO Question # 25 Proposed Answer: A A: Correct - Per SD 150, this is a RCIC trip condition which closes the valves listed except for the MO-2404 which remains open B: Incorrect - MO-2404 remains open, the MO-2400 & 2401 close C: Incorrect - MO-2405 closes D: Incorrect - MO-2404 remains open, the MO-2400 & 2401 close, MO-2405 closes SD 150 Rev 6 pages 18 & 23 Technical Reference(s): (Attach if not previously provided) 1C04C-B8 Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 73 Exam Series A
1 Point
- 26. The plant is shutdown and RHR Loop A is in Shutdown Cooling with the A pump running.
RPV water level lowers to 50 inches.
How do the RHR pumps automatically respond to the signal?
- a. The A pump remains in Shutdown Cooling.
All other pumps start and operate on min flow.
- b. The A pump trips then restarts on min flow.
All other pumps start and operate on min flow.
- c. The A pump trips and does not restart.
C pump starts and then trips.
B and D pumps auto start.
- d. The A pump remains in Shutdown Cooling.
C pump starts and then trips.
B and D pumps auto start.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 74 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 205000 A3.02 Importance Rating 3.2 Ability to monitor automatic operations of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) including: Pump trips Proposed Question: RO Question # 26 Proposed Answer: C A: Incorrect - The A pump trips. All others start but C trips B: Incorrect - The A pump trips and does not restart. All others start but C trips C: Correct - Per SD 149, page 22 - In the event a LOCA occurs when the RHR System is in the shutdown cooling mode, the RHR System will not automatically realign itself for LPCI injection. Operator actions required to initiate the LPCI mode of RHR include resetting the Group 4 Isolation Seal-In, restoring torus suction flowpath to the RHR pumps, and manually restarting the RHR pumps that have tripped.
Additionally, the SDC suction valves close on the LPCI signal (PCIS Group 4). The RHR Pumps will trip on a no suction path condition to prevent pump damage (SD-149 page 12).
D: Incorrect - The A pump trips Technical Reference(s): SD 149 Rev 11 pages 12 & 22 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 75 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 76 Exam Series A
1 Point
- 27. Which one of the following describes the relationships between the RWCU system and its associated component cooling systems?
Well Water DIRECTLY cools the RWCU pump coolers ONLY under Cold Shutdown conditions.
Well Water DIRECTLY cools the RWCU pump coolers ONLY under Cold Shutdown conditions.
Well Water is NOT DIRECTLY used for any cooling medium in the RWCU system.
Well Water is NOT DIRECTLY used for any cooling medium in the RWCU system.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 77 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 204000 K1.04 Importance Rating 2.9 Knowledge of the physical connections and/or cause- effect relationships between REACTOR WATER CLEANUP SYSTEM and the following: Component cooling water systems Proposed Question: RO Question # 27 Proposed Answer: C A: Incorrect - Per SD 414 page 12, Well water may be used only under cold shutdown conditions to cool RBCCW, not RWCU.
B: Incorrect - Per SD 261 - RBCCW supplies cooling water to the shell side of the HX.
Well water may be used only under cold shutdown conditions to cool RBCCW, not RWCU.
C: Correct - Per SD 261 page 10 - Water from the Reactor Building Closed Cooling Water System is circulated through the shell sides of the Non-Regenerative Heat Exchangers.
No other CCW system connects to RWCU.
D: Incorrect - RBCCW supplies cooling water to the shell side of the HX.
SD 261 Rev 6, page 10 Technical Reference(s): (Attach if not previously provided)
SD 414 Rev 8 Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 78 Exam Series A
10 CFR Part 55 Content: 55.41 4 55.43 (4) Secondary coolant and auxiliary systems that affect the facility Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 79 Exam Series A
1 Point
- 28. Given the following plant conditions:
- The plant is in Mode 5
- A RHR is in Shutdown Cooling
- B RHR is in Torus Cooling Then, annunciator 1C08B A-6 4KV Bus 1A4 Lockout Trip alarms.
SBDG 1G-21 auto starts and _____.
- a. loads Bus 1A4. Shutdown Cooling is lost.
- b. does NOT load Bus 1A4. Torus Cooling using B RHR is lost.
- c. loads Bus 1A4. Torus Cooling using B RHR remains in service.
- d. does NOT load Bus 1A4. Shutdown Cooling remains in service.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 80 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 233000 K2.02 Importance Rating 2.8 Knowledge of electrical power supplies to the following: RHR pumps Proposed Question: RO Question # 28 Proposed Answer: B A: Incorrect - With a bus lockout, the DG output breaker will not close.
B: Correct - IAW ARP 1C08B-A-6 this alarm causes a bus load shed, DG start however the output breaker will not close C: Incorrect - With a bus lockout, the DG output breaker will not close. B RHR will be lost D: Incorrect - SDC is lost due to the 1A4 loss causing a trip of B RPS and subsequent Group 4 partial isolation.
Technical Reference(s): ARP-1C08B (A-6) Rev 82 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # WTS Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 81 Exam Series A
signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 82 Exam Series A
1 Point
- 29. The plant is at 100% power near the end of cycle with all control rods fully withdrawn. At this point the SCRAM INLET VALVE fails OPEN for control rod 18-27.
Which of the following describes the Effects on the following over the next five (5) minutes?
(1) Effect on reactor power AND (2) Effect on Scram Discharge Volume (SDV)
- a. (1) Reactor power will remain at 100% power.
(2) There will be NO flow into the SDV.
- b. (1) Reactor power will be LOWER, but the plant will continue to operate at power.
(2) There will be NO flow into the SDV.
- c. (1) Reactor power will be LOWER, but the plant will continue to operate at power.
(2) There will be flow into the SDV, but it will be within the capacity of the SDV drain.
- d. (1) Reactor power will remain at 100%.
(2) There will be flow into the SDV, but it will be within the capacity of the SDV drain.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 83 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 201003 K3.01 Importance Rating 3.2 Knowledge of the effect that a loss or malfunction of the CONTROL ROD AND DRIVE MECHANISM will have on following: Reactor Power Proposed Question: RO Question # 29 Proposed Answer: B A: Incorrect - Power will lower Correct - Per SD 255, Figure on page 24 - Accumulator pressure acting on the B: CRD mechanism under piston area will cause the rod to insert and power to lower. Since the scram outlet valve is closed, no flow will go to the SDV C: Incorrect - There will be no flow to the SDV D: Incorrect - The rod will insert and there will be no flow to the SDV Technical Reference(s): SD 255 Rev 8 page 24 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X - 19990 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6, 7 55.43 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 84 Exam Series A
(6) Design, components, and functions of reactivity control mechanisms and instrumentation.
(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 85 Exam Series A
1 Point
- 30. The plant is operating at 100% power. Both Control Building intake radiation monitors, RM-6101A&B, have tripped. The Standby Filter Unit (SFU) Lockout Relays have tripped and all automatic actions have occurred as expected.
The control room operator has just placed the B SFU train in STANDBY following the auto initiation.
Which one of the following describes how the B SFU train will function?
The B SFU train will _____.
- a. auto initiate if the A SFU train flow rate lowers to 800 scfm or less.
- b. NOT auto initiate. Manual manipulations must be performed to place the B SFU train in operation.
- c. auto initiate ONLY if the high radiation condition clears and then occurs again.
- d. auto initiate immediately if the B SFU Mode Select HS-7316B is taken back to the AUTO position.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 86 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 290003 K4.01 Importance Rating 3.1 Knowledge of CONTROL ROOM HVAC design feature(s) and/or interlocks which provide for the following: System initiations/reconfiguration: Plant-Specific Proposed Question: RO Question # 30 Proposed Answer: A A: Correct - Per SD 730, page 28 - After both trains initiate, approximately 1000 scfm TOTAL is flowing through the system, one train can be manually transferred to standby mode by turning the mode switch to MAN and then back to AUTO. In this condition, mode switch in AUTO, lockout relay tripped and flow >800 scfm, the heater & exhaust fans are off and intake & discharge valves AV-7301A(B) and AV-7318A(B) are shut.
The standby train will auto initiate, start the fan, energize the heaters, and open supply and discharge valves, if system flow drops to 800 scfm or less.
B: Incorrect - It would still auto initiate on low flow C: Incorrect - It would also auto initiate on low flow D: Incorrect - If placed in manual then back in auto, a low flow condition is required to start the fan Technical Reference(s): SD 730 Rev 9, page 28 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X - 22687 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 87 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 88 Exam Series A
1 Point
- 31. OI 644 "Condensate and Feedwater Systems" directs that the Condensate/Feed system be filled and vented prior to the FIRST Condensate Pump start.
Which of the following is the reason for this action?
- a. To prevent pump damage due to exceeding pump vibration limits during pump startup.
- b. To prevent Condensate Pump vortex limits from being exceeded and vapor binding of the pump.
- c. To reduce the risk of water hammer and the system damage that could result.
- d. To prevent pump run out conditions in the Condensate Pump which will cause winding degradation.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 89 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 259001 K5.02 Importance Rating 2.5 Knowledge of the operational implications of the following concepts as they apply to REACTOR FEEDWATER SYSTEM : Water hammer Proposed Question: RO Question # 31 Proposed Answer: C A: Incorrect - Not the reason per the OI. Although vibration may rise slightly during startup, the concern per the OI is water hammer in the system.
B: Incorrect - Not the reason per the OI. This would be a concern with a saturated liquid C: Correct - Per OI 644 NOTE in Section 3.2 after step (9) - To reduce the risk of water hammer D: Incorrect - Not the reason per the OI. This condition would occur if pump resistance to flow were zero.
Technical Reference(s): OI 644 Rev 109 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
X - DAEC NRC Question Source: Bank #
2002 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam: 2002 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 90 Exam Series A
(5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 91 Exam Series A
1 Point
- 32. A plant event has occurred. The following conditions exist.
- Reactor level is 110 and lowering
- Reactor pressure is 550 psig and lowering
- Drywell pressure is 9.0 psig and rising
- Containment sprays are in service Which one of the following conditions would cause an isolation of Containment Sprays?
(Assume no operator actions)
- a. One Wide Range Yarway reference leg break.
- b. Reactor water level has lowered to 64 inches.
- c. One Fuel Zone instrument variable leg break.
- d. Loss of 1Y23, Uninterruptible AC power.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 92 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 226001 K6.08 Importance Rating 2.7 Knowledge of the effect that a loss or malfunction of the following will have on the RHR/LPCI:
CONTAINMENT SPRAY SYSTEM MODE : Nuclear boiler instrumentation Proposed Question: RO Question # 32 Proposed Answer: C A: Incorrect - The wide range level instruments do not span the 2/3 core coverage level.
B: Incorrect - This is the LPCI initiation level setpoint C: Correct - Per SD 880 page 39 - Containment spray cooling initiation logic receives a permissive signal from each of two reactor vessel level instruments. These switch contacts actuate when reactor vessel level is above -39" to allow containment spray cooling to be manually initiated. A variable leg break would give a false low level signal.
Therefore, the 2/3 core coverage permissive for containment spray would not be met.
The logic is shown in Figure 11 of SD 149 D: Incorrect - Uninterruptible AC power provides indication and recorder power for the GEMAC level instruments and would not affect the 2/3 core coverage containment spray interlock SD 149 Rev 11 page 54 Technical Reference(s): (Attach if not previously provided)
SD 880 Rev 13 page 39 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X - 19015 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 93 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 94 Exam Series A
1 Point
- 33. A reactor startup is in progress. Reactor power is at 35% and the operators have just finished bypassing rod 02-19 on the Rod Worth Minimizer. There are now eight rods bypassed.
The control room operator selects rod 02-19 and attempts to withdraw the control rod in accordance with the pull sheet.
How will the control rod respond to the withdraw signal and what is the reason for that response?
- a. The control rod will NOT withdraw. Since rod 02-19 has been bypassed, it can only be inserted.
- b. The control rod will withdraw. Since rod 02-19 has been bypassed, the RWM is incapable of enforcing a rod block.
- c. The control rod will NOT withdraw. This is due to having the maximum number of rods bypassed in the RWM.
- d. The control rod will withdraw, as long as the RWM-CC keylock mode switch is in OPER.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 95 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 201006 A1.02 Importance Rating 3.4 Ability to predict and/or monitor changes in parameters associated with operating the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) controls including: Status of control rod movement blocks; P-Spec(Not-BWR6)
Proposed Question: RO Question # 33 Proposed Answer: A A: Correct - Per SD 878.8, page 16 - Rods that are bypassed are only allowed to be inserted. If the selected rod is a bypassed rod, then a withdraw block is applied.
B: Incorrect - a withdraw block is enforced C: Incorrect - It will not withdraw because a withdraw block is enforced when the rod is bypassed in the RWM D: Incorrect - a withdraw block is enforced Technical Reference(s): SD 878.8 Rev 7 page 16 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X -19377 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 96 Exam Series A
10 CFR Part 55 Content: 55.41 5, 7 55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 97 Exam Series A
1 Point
- 34. With the plant operating at full power, an unisolable leak in the Main Turbine lube oil system occurs.
The operators trip the Main Turbine with the Emergency Trip PB and turbine lube oil is secured.
Which one of the following describes the recommended method of disconnecting the Main Generator from the grid IAW OI 698 Main Generator and any additional actions that are required in regard to the loss of turbine lube oil?
- a. Verify that GENERATOR OUTPUT H BREAKER (OCB 0220) and GENERATOR OUTPUT I BREAKER (OCB 4290) have OPENED, then verify that the GENERATOR EXCITER FIELD BREAKER has tripped.
Reduce generator hydrogen gas pressure to approximately 10 to 15 PSIG until the Hydrogen Seal Oil System is secured.
- b. Verify that GENERATOR OUTPUT H BREAKER (OCB 0220) and GENERATOR OUTPUT I BREAKER (OCB 4290) have OPENED, then verify that the GENERATOR EXCITER FIELD BREAKER has tripped.
Secure the Hydrogen Seal Oil System, then reduce generator hydrogen gas pressure to approximately 10 to 15 PSIG.
- c. Manually trip the GENERATOR EXCITER FIELD BREAKER, then verify that GENERATOR OUTPUT H BREAKER (OCB 0220) and GENERATOR OUTPUT I BREAKER (OCB 4290) have OPENED.
Reduce generator hydrogen gas pressure to approximately 10 to 15 PSI until the Hydrogen Seal Oil System is secured.
- d. Manually trip the GENERATOR EXCITER FIELD BREAKER, then verify that GENERATOR OUTPUT H BREAKER (OCB 0220) and GENERATOR OUTPUT I BREAKER (OCB 4290) have OPENED.
Secure the Hydrogen Seal Oil System, then reduce generator hydrogen gas pressure to approximately 10 to 15 PSI.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 98 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 245000 A2.01 Importance Rating 3.7 Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Turbine trip Proposed Question: RO Question # 34 Proposed Answer: A A: Correct - Per OI 698 - Section 5.2.2 (a) (Recommended method) trip the turbine with the TURBINE EMERGENCY TRIP pushbutton and verify GENERATOR OUTPUT H BREAKER (OCB 0220) and GENERATOR OUTPUT I BREAKER (OCB 4290) are OPEN.
Step 5.2.7 - Verify GENERATOR EXCITER FIELD BREAKER tripped and GENERATOR FIELD BREAKER BACKUP green indicating light ON at 1C08.
AOP 693 - Followup Action 4 - If turbine lube oil is secured, reduce generator hydrogen gas pressure to approximately 10 to 15 PSI until the Hydrogen Seal Oil System is secured.
B: Incorrect - H2 pressure must be lowered prior to securing the seal oil system.
C: Incorrect - the exciter field breaker trips automatically and is verified tripped after verifying the output breakers open.
D: Incorrect - the exciter field breaker trips automatically. and is verified tripped after verifying the output breakers open. H2 pressure must be lowered prior to securing the seal oil system.
AOP 693 Rev 12 Technical Reference(s): (Attach if not previously provided)
OI 698 Rev 70 page 26 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 99 Exam Series A
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 100 Exam Series A
1 Point
- 35. The plant is operating at full power with a TIP trace in progress. The system is being operated in the automatic mode. Status of the TIPs is as follows:
- Channel A: Position 2 selected and inserting
- Channel C: Position 2 selected and withdrawing (tracing)
At this point a transient occurs that results in RPV level lowering to 165.
Which one of the following describes the response, if any, of the TIP drives to this event?
- a. No response since RPV level is >119.5.
- b. Both drives reverse and withdraw from the core.
The ball valves automatically close when associated detectors are at the in shield position.
- c. Channel "A" continues insertion to the top of core, then withdraws to the in shield position without tracing.
Channel "C" continues withdrawing to the in shield position.
The operator then closes both ball valves by placing the MAN VALVE CONTROL to CLOSE.
- d. Channel "A" continues insertion to the top of core, then withdraws to the in shield position without tracing.
Channel "C" continues withdrawing to the in shield position.
The ball valves automatically close when associated detectors are at the in shield position.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 101 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 215001 A3.03 Importance Rating 2.5 Ability to monitor automatic operations of the TRAVERSING IN-CORE PROBE including: Valve operation: Not-BWR1 Proposed Question: RO Question # 35 Proposed Answer: B A: Incorrect - The TIP probes withdraw when RPV level reaches 170 B: Correct - Per SD 878.6 page 29 - The TIP System response to a Group 2 Containment Isolation (170 inches) is to retract to the shield any detectors that are inserted in the Reactor core. Then close the TIP Ball Valves.
C: Incorrect - Both probes immediately withdraw and the ball valves auto close D: Incorrect - Both probes immediately withdraw Technical Reference(s): SD 878.6 Rev 6 page 29 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: RO-83.03.01.05-02 (As available)
Question Source: Bank # X - 20210 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 102 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 103 Exam Series A
1 Point
- 36. The plant is operating at 90% power.
The CRS directs you to place Torus Cooling in service IAW OI 149 RHR System in advance of a RCIC quarterly surveillance test.
Which one of the following describes:
(1) How Torus temperature is controlled during this evolution AND (2) How RHRSW or RHR flows would be affected if a valid LPCI initiation signal occurred due to a small leak in the drywell that does not depressurize the RPV?
(2) RHRSW flow through the shell side of the RHR Heat Exchanger would continue unchanged.
(2) RHR flow through the tube side of the RHR Heat Exchanger would be automatically secured.
(2) RHRSW flow through the tube side of the RHR Heat Exchanger tube side would be automatically secured.
(2) RHR flow through the shell side of the RHR Heat Exchanger shell side would continue unchanged.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 104 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 219000 A4.05 Importance Rating 3.4 RHR/LPCI: Torus/Pool Cooling Mode - Ability to manually operate and/or monitor in the control room: Heat exchanger cooling flow Proposed Question: RO Question # 36 Proposed Answer: C A: Incorrect - RHR flow is thru or around the shell side of the RHR HX On a LPCI signal RHRSW pumps trip B: Incorrect - On a LPCI signal, RHR flow is throttled around the RHR HX C: Correct - Per OI 149, Section 5.4 step 11 - Close MO-2030 [1940] A[B] HEAT EXCH BYPASS valve if required.
Per SD 149 - The LPCI initiation signal overrides all modes of the RHR System (except shutdown cooling). The intent is to direct maximum system effort toward restoring and maintaining the reactor vessel water level, i.e., all pumps are started, all non-LPCI modes secured, and motor-operated valves positioned to direct the maximum amount of flow into the reactor vessel.
D: Incorrect - LPCI signal would shut the Torus Cooling valves, which would secure flow through the RHR heat exchanger.
OI 149 Rev 110 page 37 Technical Reference(s): (Attach if not previously provided)
SD 149 Rev 11 page 20 Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 105 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 106 Exam Series A
1 Point
- 37. A plant event has occurred.
Which one of the following lists instrumentation required by Technical Specifications to provide Post Accident Monitoring (PAM) indication in the control room?
(1) Drywell Pressure (2) RPV Fuel Zone Level (3) RPV Wide Range Level (4) RPV Narrow Range Level (5) Reactor Building Vent Shaft Rad Monitors ONLY______.
- a. (1), (2) and (5)
- b. (1), (3) and (4)
- c. (2), (4) and (5)
- d. (1), (2) and (3)
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 107 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 216000 2.4.3 Importance Rating 3.7 (Nuclear Boiler instrumentation) Emergency Procedures / Plan: Ability to identify post-accident instrumentation.
Proposed Question: RO Question # 37 Proposed Answer: D A: Incorrect - Reactor Building Vent Shaft Rad Monitor is not a PAM instrument B: Incorrect - narrow range level is not a PAM instrument C: Incorrect - Narrow Range Level and Reactor Building Vent Shaft Rad Monitors are not PAM instruments D: Correct -IAW TS Table 3.3.3.1.1 Technical Reference(s): TS 3.3.3.1. Table 1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 7 55.43 (7) Design, components, and function of control and safety systems, including instrumentation, Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 108 Exam Series A
signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 109 Exam Series A
1 Point
- 38. The plant is operating at 90% power with a power ascension in progress. All systems are operable.
Then, annunciator 1C06B (B-1), CONDENSATE DEMIN INLET/OUTLET HI P, alarms.
Which one of the following describes:
(1) an action required as a result of this alarm AND (2) what is the concern due to the alarm condition?
- a. (1) Verify that MO-1708 CONDENSATE DEMIN BYPASS has automatically opened and that demin inlet/outlet P is lowering to <40 psid.
(2) Resin breakthrough can occur resulting in decreasing condensate water conductivity.
- b. (1) Verify that MO-1708 CONDENSATE DEMIN BYPASS has automatically opened and that demin inlet/outlet P is lowering to <40 psid.
(2) Resin breakthrough can occur resulting in additional chlorides and sulfates in the condensate system.
- c. (1) Throttle open MO-1708 CONDENSATE DEMIN BYPASS to maintain system dP
<40 psid.
(2) Resin breakthrough can occur resulting in decreasing condensate water conductivity.
- d. (1) Throttle open MO-1708 CONDENSATE DEMIN BYPASS to maintain system dP
<40 psid.
(2) Resin breakthrough can occur resulting in additional chlorides and sulfates in the condensate system.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 110 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 256000 A1.08 Importance Rating 2.7 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CONDENSATE SYSTEM controls including: System water quality Proposed Question: RO Question # 38 Proposed Answer: D A: Incorrect - MO-1708 must be manually opened, Conductivity levels would increase not decrease B: Incorrect - MO-1708 must be manually opened C: Incorrect - Conductivity levels would increase not decrease D: Correct - Per ARP 1C06B (B-1) - Operator Action step 3.2 - Maintain system dP <40 psid by throttling open MO-1708 CONDENSATE DEMIN BYPASS with handswitch HS-1708 at 1C06 or reducing reactor power as necessary to clear the alarm.
Per AOP 639 - Exceeding the 40 psid total F/D system P as indicated at PDI-1707 at Panel 1C06 or PDI-1708 at Panel 1C80 or exceeding the 25 psid individual F/D bed P limits as indicated by PDI-1727A[ B, C, D E] at Panel 1C80 may cause septa damage and resin breakthrough.
AOP 639 Rev 29 page 9 Technical Reference(s): (Attach if not previously provided)
ARP 1C06B (B-1) Rev 45 Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 111 Exam Series A
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 112 Exam Series A
1 Point
- 39. The plant is operating at full power. All systems are operable.
Then, annunciator 1C08A (A-9), 125 VDC SYSTEM 1 TROUBLE, activates.
As the BOP operator you recognize that 125 VDC bus 1D10 is de-energized.
Which one of the following describes the effect on the associated bus normal supply breaker?
- a. 4KV breaker overcurrent and undervoltage protection is lost. The 4KV supply breaker has tripped and cannot be reclosed from the control room.
- b. 4KV breaker overcurrent and undervoltage protection is lost. The 4KV supply breaker remains closed.
- c. 4KV breaker undervoltage protection is lost; however, overcurrent protection remains available. The 4KV supply breaker has tripped and cannot be reclosed from the control room.
- d. 4KV breaker undervoltage protection is lost; however, overcurrent protection remains available. The 4KV supply breaker remains closed.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 113 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295004 AK1.05 Importance Rating 3.3 AK1.05 - Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER : Loss of breaker protection Proposed Question: RO Question # 39 Proposed Answer: B A: Incorrect - The breakers remain closed without breaker protection B: Correct - Per AOP 302.1 - Automatic actions list. Loss of 1A1/1A3 breaker control occurs. Additionally all breaker protection is lost on loss of DC C: Incorrect - The breakers will not trip. Control power to trip the breakers was lost.
Additionally, all breaker protection is lost.
D: Incorrect - All breaker protection is lost.
Technical Reference(s): AOP 302.1 Rev 48 page 3 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 8 55.43 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 114 Exam Series A
(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(8) Components, capacity, and functions of emergency systems Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 115 Exam Series A
1 Point
- 40. Which one of the following describes an operational requirement in regard to high suppression pool water temperature and the reason for the requirement?
Scramming the reactor before Suppression Pool average temperature FIRST reaches
__ (1)__ is required to ensure __(2)__.
- a. (1) 110°F (2) complete steam condensation following a Loss of Coolant Accident prior to reaching a Suppression Pool temperature of 170°F.
- b. (1) 120 °F (2) complete steam condensation following a Loss of Coolant Accident prior to reaching a Suppression Pool temperature of 170°F.
- c. (1) 110°F (2) complete steam condensation following a LLS Safety Relief Valve actuation.
- d. (1) 120 °F (2) complete steam condensation following a LLS Safety Relief Valve actuation.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 116 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295026 EK1.02 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE : Steam condensation Proposed Question: RO Question # 40 Proposed Answer: A A: Correct - Per TS bases background 3.6.2.1 - The technical concerns that lead to the development of suppression pool average temperature limits are as follows:
- a. Complete steam condensation the original limit for the end of a LOCA blowdown was 170°F, based on the Bodega Bay and Humboldt Bay Tests; and, b.Primary containment peak pressure and temperature design pressure is 56 psig and design temperature is 281°F (Ref. 1).
This information is also available from FSAR Chapter 6.2, page 6.2-42 and EOP Breakpoints Basis (page 11) and EOP Curves and Limits (page 18).
B: Incorrect - A depressurization within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is required above this temperature C: Incorrect - Per TS bases 3.6.2.2 - Initial suppression pool water level affects suppression pool temperature response calculations, calculated drywell pressure during vent clearing for a DBA, calculated pool swell loads for a DBA LOCA, and calculated loads due to SRV discharges. Suppression pool water level must be maintained within the limits specified so that the safety analysis of Reference 1 remains valid.
D: Incorrect - A depressurization within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is required above this temperature. The SRV concern is based on SP level.
TS bases 3.6.2.1, p 3.6-49 EOP Breakpoints (p11)
Technical Reference(s): (Attach if not previously provided)
EOP Curves and Limits (p18)
FSAR Chapter 6.2 (p6.2-42)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # WTS Modified Bank # (Note changes or attach parent)
New Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 117 Exam Series A
Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9, 10 55.43 (9) Shielding, isolation, and containment design features, including access limitations.
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 118 Exam Series A
1 Point
- 41. The plant was operating at 90% power when the following occurred:
- 1P201B, B RECIRC PUMP has tripped.
- Power has stabilized at 58% thermal power.
Which of the following describes how this event affects the reactor fuel?
Fuel failures _______.
- a. are more likely due to decreased radial peaking factor.
- c. are more likely due to a decrease in Average Planar Linear Heat Generation Rate.
- d. are less likely due to increased margin in the linear heat generation rate.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 119 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295001 AK1.03 Importance Rating 3.6 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Thermal limits Proposed Question: RO Question # 41 Proposed Answer: D A: Incorrect - Increased ( not decreased) radial peaking factors provide less margin to fuel failures - not more margin (related to APLHGR).
B: Incorrect - Pellet-clad interactions (PCI) type of fuel failures are enhanced by the production of embrittling agents (from fission) of Iodine and cadmium. As power drops, these isotopes are produced less and more margin to this type of failure is gained - not less margin.
C: Incorrect - APLHGR is a concern during LOCA conditions D: Correct - As power drops core wide from reduced recirculation flow, the amount of energy produced per linear foot of fuel drops. This provides more margin to the LHGR limit.
GFES Thermo, Chap 8, pg 26 Technical Reference(s): (Attach if not previously provided)
GFES Thermo, Chap 9, pg 4 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # WTS Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 120 Exam Series A
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 (2) 55.43 (2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 121 Exam Series A
1 Point
- 42. The plant is operating at 50% power.
Then, low EHC pressure causes a Main Turbine trip.
Which one of the following describes how the plant is affected?
- a. The Turbine Stop Valves Fast Close The Turbine Control Valves Fast Close Both Reactor Recirc Pumps Trip
- b. ONLY the Turbine Control Valves Fast Close Both Reactor Recirc Pumps Trip
- c. The Turbine Stop Valves Fast Close The Turbine Control Valves Fast Close Both Reactor Recirc Pumps remain running
- d. ONLY the Turbine Control Valves Fast Close Both Reactor Recirc Pumps remain running Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 122 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295005 AK2.03 Importance Rating 3.2 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: Recirculation system Proposed Question: RO Question # 42 Proposed Answer: A A: Correct - Per SD 693.2a - EHC low pressure results in a fast closure of the Turbine Control Valves & Stop valves. This will result in a reactor scram due to Turbine Control Valve Fast Closure signal (800# RETS pressure).
Per SD 693.1, page 19 - Turbine Control Valve Fast Closure is used to produce a Reactor Scram and End Of Cycle Recirculation Pump Trip (EOC-RPT) signal B: Incorrect - ONLY the Turbine Control Valves Fast Closure would occur if a power to load unbalance turbine trip occurred.
C: Incorrect - Turbine Control Valve Fast Closure is used to produce a Reactor Scram and End Of Cycle Recirculation Pump Trip (EOC-RPT) signal D: Incorrect - EOC-RPT from the TCV fast closure causes both reactor recirc pumps to trip which would occur if the turbine trip was caused by the power to load unbalance circuitry.
SD 693.1 Rev 9 page 19 Technical Reference(s): (Attach if not previously provided)
SD 693.2a Rev 5 table A page 37 Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 123 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 124 Exam Series A
1 Point
- 43. The plant is at full power. All systems are operable.
A local alarm horn sounds at panel 1C179 Cardox Fire Protection Control Panel and control room annunciator 1C40 (F-6) CARDOX PRE-INITIATION ALARM actuates.
Which one of the following describes actions, if any, that will occur if only ONE of the sixteen spot type heat detectors in the cable spreading room has reached its setpoint of 140°F?
- a. No additional actions will occur. At least two heat detectors must reach their alarm setpoint before any additional actions occur.
- b. The Cable Spreading Room A/C unit and exhaust fan will trip. Then, after a 24-second time delay the CO2 will inject into the room.
- c. After a 24-second time delay, the Cable Spreading Room A/C unit and exhaust fan will trip and CO2 will inject into the room.
- d. The Cable Spreading Room A/C unit and exhaust fan trip immediately. CO2 will inject into the room with NO time delay.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 125 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 600000 AK2.01 Importance Rating 2.6 Knowledge of the interrelations between PLANT FIRE ON SITE and the following: Sensors/
detectors, valves Proposed Question: RO Question # 43 Proposed Answer: B A: Incorrect - CO2 system actuates, fans trip.
B: Correct - Per SD 513, pages 16-18, The automatic initiation occurs if ONE of the 16 spot type heat detectors located in the Cable Spreading Room reaches 140°F. At this temperature an electrical signal is sent to the Cardox Control Panel 1C-179. This panel then sends signals to perform the automatic initiation sequence.
The following actions take place upon an automatic initiation:
- Local alarm horn sounds at panel 1C179
- Control room annunciator 1C40 (F-6) CARDOX PRE-INITIATION ALARM actuates
- A 24 second pre-discharge time delay starts
- Cable Spreading Room A/C Unit (1V-AC-32) trips
- Cable Spreading Room Exhaust Fan (1V-EF-33) trips After the 24 second pre-discharge time delay times out, the following occurs:
CO2 discharges into the Cable Spreading Room.
C: Incorrect - The A/C unit and fan trips immediately.
D: Incorrect - There is a 24 second time delay before CO2 discharges.
SD 513 Rev 11 pages 16-18 Technical Reference(s): (Attach if not previously provided) 1C40 (F-6) Rev 64 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X - 19105 Modified Bank # (Note changes or attach parent)
New Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 126 Exam Series A
Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 127 Exam Series A
1 Point
- 44. Which one of the following describes how the Reactor Protection System functions to allow control rod insertion on a valid low RPV level scram signal?
- a. RPS energizes the ARI solenoids which repositions the ARI valves to permit venting of the scram air header.
- b. RPS trips and back-up scram valves de-energize. This repositions the back-up scram valves to permit venting of the scram air header.
- c. RPS trips and scram pilot valves de-energize and reposition to vent air from the scram inlet and outlet valves. This causes the scram inlet and outlet valves to reposition.
- d. RPS de-energizes causing a loss of power to the scram inlet and outlet valves. This causes the scram air header to depressurize.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 128 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295006 AK2.01 Importance Rating 4.3 Knowledge of the interrelations between SCRAM and the following: RPS Proposed Question: RO Question # 44 Proposed Answer: C A: Incorrect - ARI is independent of RPS in regard to venting the scram air header.
B: Incorrect - When RPS trips the back-up scram valves are energized C: Correct - Per SD 358, page 28 - There are two scram pilot valves for each control rod, arranged as shown in the figure above. Each scram pilot valve is solenoid operated with the solenoids normally energized. The scram pilot valves control the air supply to the respective scram valves for each control rod. With either scram pilot valve energized, air pressure holds the scram valves closed. The scram valves control the supply and discharge paths for control rod scram water. RPS trip system A controls one of the scram pilot valves for each control rod. RPS trip system B controls the other scram pilot valve for each control rod.
D: Incorrect - Scram inlet and outlet valves are air operated. Their position is affected by the scram pilot valve.
Technical Reference(s): SD 358 Rev 7 page 28 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 129 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 130 Exam Series A
1 Point
- 45. One of the refueling interlocks prevents bridge movement when the refuel platform is over or near the core, all control rods are not in, and the grapple is loaded.
Which one of the following is the reason for this interlock?
- a. To prevent mechanical damage to the control rod blades during refueling.
- b. To prevent movement of unanalyzed loads over the core when control rods are withdrawn.
- c. To prevent fuel bundle insertion into the reactor and the potential for criticality when control rods are withdrawn.
- d. To prevent personnel from operating the refueling bridge over the core when control rods are withdrawn to reduce exposure.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 131 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 295023 AK3.02 Importance Rating 3.4 Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS : Interlocks associated the fuel handling equipment Proposed Question: RO Question # 45 Proposed Answer: C A: Incorrect - This is not the reason stated in TS bases or the FSAR but is plausible when refueling is in progress B: Incorrect - This is not the reason stated in TS bases or the FSAR but is plausible when refueling is in progress C: Correct - IAW TS Bases 3.9 - Refueling equipment interlocks restrict the operation of the refueling equipment or the withdrawal of control rods to serve as a backup to procedural core reactivity controls to prevent the reactor from achieving criticality during refueling. The refueling interlock circuitry senses the conditions of the refueling equipment and the control rods. Depending on the sensed conditions, interlocks are actuated to prevent the operation of the refueling equipment or the withdrawal of control rods.
D: Incorrect - This is not the reason stated in TS bases or the FSAR but is plausible when refueling is in progress Technical Reference(s): TS bases 3.9 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # WTS - NMP 2007 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 132 Exam Series A
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.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 133 Exam Series A
1 Point
- 46. Which one of the following is the reason EOP 2, Primary Containment Control, step DWT/4 directs spraying the drywell before drywell temperature reaches 280°F and entering EOP 1?
- a. Reduce RPV Level instrument inaccuracies.
- b. Prevent containment structural failure due to overheating.
- c. Prevent exceeding the environmental qualification of the MSIV solenoids.
- d. Limit the condensation effect of drywell sprays on drywell pressure.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 134 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295028 EK3.03 Importance Rating 3.6 Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE : Drywell spray operation: Mark-I&II Proposed Question: RO Question # 46 Proposed Answer: B A: Incorrect - RPV Saturation Temperature Graph 1 addresses this concern B: Correct - Per EOP 2 bases document, page 34 - When drywell temperature is approaching structural design limits in spite of previous temperature control actions, energy release to the drywell is reduced by entering EOP 1 and scramming the reactor.
C: Incorrect - Per The EOP bases, the concern is not the MSIV solenoids.
D: Incorrect - The Drywell Spray Initiation Limit curve addresses this concern Technical Reference(s): EOP 2 bases Rev 13 page 34 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # WTS Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 (9) 55.43 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 135 Exam Series A
(9) Shielding, isolation, and containment design features, including access limitations.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 136 Exam Series A
1 Point
- 47. Which one of the following describes the reason that spraying the drywell is permitted only within the limits of the Drywell Spray Initiation Limit Curve?
- a. It could result in an evaporative cooling pressure drop large enough to deinert the primary containment atmosphere through the reactor building-to-torus vacuum breakers.
- b. It could result in an evaporative cooling pressure drop large enough to deinert the primary containment atmosphere through the torus-to-drywell vacuum breakers.
- c. To prevent excessive cycling of the reactor building-to-torus vacuum breakers and challenge of the primary containment pressure suppression capability.
- d. To ensure that Suppression Chamber Pressure can be restored below the Torus Spray Initiation Pressure.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 137 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295024 EK3.08 Importance Rating 3.7 Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE : Containment spray: Plant-Specific Proposed Question: RO Question # 47 Proposed Answer: A A: Correct - EOP 2 bases page 53 discussion of step PC/P Drywell sprays may only be initiated if drywell temperature and pressure are within the unshaded region of the Drywell Spray Initiation Limit (discussion of the basis for the limit is in the EOP Curves and Limits Bases Document). Initiation of sprays from within the shaded region of the curve could result in an evaporative cooling pressure drop large enough to deinert the primary containment atmosphere through the reactor building-to-torus vacuum breakers or challenge the primary containment negative pressure capability.
B: Incorrect - The concern is with deinerting the primary containment atmosphere through the reactor building-to-torus vacuum breakers.
C: Incorrect - Cycling of the breakers is not a concern in the shaded area of the curve.
D: Incorrect - Restoring drywell pressure below the Torus Spray Initiation Pressure may occur but is not the reason for question asked.
Technical Reference(s): EOP 2 bases rev 13 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: none Learning Objective: (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 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 138 Exam Series A
10 CFR Part 55 Content: 55.41 (9) 55.43 (9) Shielding, isolation, and containment design features, including access limitations.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 139 Exam Series A
1 Point
- 48. A plant transient has occurred. EOP 1 has been entered and the level control leg is being implemented.
The only low-pressure injection systems available were "A" CS and the "A" RHR pump.
When RPV level lowered to +15", the crew decided that an ED was necessary.
The following plant conditions currently exist:
- "A" Core Spray is injecting at 2800 gpm
- "A" RHR pump is injecting at 4800 gpm
- RPV pressure is 50 psig and lowering
- RPV level is -35" and stable Which one of following states whether adequate core cooling is assured and the reason for that determination?
- a. Adequate core cooling is NOT assured because RPV level is below -25.
- b. Adequate core cooling is NOT assured because BOTH the "A" and "B" Core Spray pumps must be injecting at >3000 gpm.
- c. Adequate core cooling is assured because sufficient low pressure ECCS injection is occurring.
- d. Adequate core cooling is assured because the current RPV is the lowest level at which steam flow through the SRVs is sufficient to remove all decay heat from the core.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 140 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295031 EA1.03 Importance Rating 4.4 Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL : Low pressure core spray Proposed Question: RO Question # 48 Proposed Answer: A A: Correct - PER EOP 1 bases page 46 - If RPV water level cannot be restored and maintained above -25 inches, adequate core cooling cannot be ensured and SAG entry is required.
B: Incorrect - ONLY one Core Spray Pump at >3000 gpm is required at this RPV level if level is >-39 C: Incorrect - With Core Spray Pump at 2800 gpm and RPV level is below -25, adequate core cooling is NOT assured and a SAG entry is required.
D: Incorrect - Reactor pressure is at 50 psig. Steam flow thru the SRVs will not assure adequate core cooling Technical Reference(s): EOP 1 bases Rev 14 (Attach if not previously provided)
Do NOT provide the EOP 1 or ATWS Proposed References to be provided to applicants during examination:
level control legs with setpoints Learning Objective: (As available)
Question Source: Bank # 20766 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 141 Exam Series A
10 CFR Part 55 Content: 55.41 5, 10 55.43 (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 142 Exam Series A
1 Point
- 49. Plant conditions are as follows:
- High Pressure Coolant Injection (HPCI) is operating for RPV Level control as directed by EOP 1.
Torus Water Level is reported to be 10.1 feet and LOWERING.
Which ONE of the following identifies the Torus Water Level at which HPCI must be secured and the reason it must be secured?
HPCI must be secured when Torus Water Level reaches____.
- a. 7.1 feet, to prevent violating Vortex Limits.
- b. 7.1 feet, to prevent direct pressurization of the Torus by the HPCI exhaust.
- c. 5.8 feet, to prevent violating Vortex Limits.
- d. 5.8 feet, to prevent direct pressurization of the Torus by the HPCI exhaust.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 143 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295030 EA1.05 Importance Rating 3.5 Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCI Proposed Question: RO Question # 49 Proposed Answer: D A: Incorrect - This is a vortex level limit UNLESS directed to use HPCI in EOPs B: Incorrect - This is above the level that will result in direct pressurization of the torus by the HPCI exhaust.
C: Incorrect - EOP 1 overrides vortex concerns D: Correct - Per EOP 2 bases step T/L A torus level of 5.8 feet corresponds to the HPCI turbine exhaust elevation. Direction here attempts to maintain the availability of HPCI should it be needed as an injection source or alternate method of depressurizing the RPV.
Operation of the HPCI system with its exhaust device not submerged will directly pressurize the torus.
Technical Reference(s): EOP 2 bases Rev 13 page 13 (Attach if not previously provided)
Do NOT provide EOP 2 Torus Level Proposed References to be provided to applicants during examination:
legs with any setpoints filled in.
Learning Objective: (As available)
Question Source: Bank # DAEC Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 144 Exam Series A
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8, 10 55.43 (8) Components, capacity, and functions of emergency systems.
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 145 Exam Series A
1 Point
- 50. The plant was at full power, with A loop of Torus Cooling in service, when a Design Basis Earthquake occurred.
- All control rods were fully inserted.
- A ESW pump remains in service.
- Operators are inspecting for flooding damage in the Turbine and Reactor Buildings.
- The plant is currently being cooled down in preparation for going to cold shutdown.
In accordance with AOP-901, EARTHQUAKE, which one of the following actions is required in regard to component cooling water systems?
- a. Shutdown General Service Water until a system walkdown to assess damage is complete.
- b. RHRSW must be isolated to and from Well Water because the Well Water system is NOT seismically qualified.
- c. ESW must be isolated to and from Well Water because the Well Water system is NOT seismically qualified.
- d. RHRSW/ESW return must be realigned to the dilution structure because the Circ Water System is NOT seismically qualified.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 146 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295018 AA1.03 Importance Rating 3.3 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Affected systems so as to isolate damaged portions Proposed Question: RO Question # 50 Proposed Answer: C A: Incorrect - This is not required unless flooding observed.
B: Incorrect - RHRSW does not tie in to well water C: Correct - Per AOP 901 - Continuous Recheck Statement requires isolation of Well Water and calls the Chill water portion of the Well Water System "non-seismic". ESW operability is still required in Mode 3 D: Incorrect - Circ Water is also non-seismic. This action is directed in case of flooding in the Turbine Bldg. The stem states there is no flooding in the turbine building Technical Reference(s): AOP 901 Rev 17 page 5 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # DAEC Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 147 Exam Series A
10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 148 Exam Series A
1 Point
- 51. The plant is operating at full power. All systems are operable.
The River Water Supply (RWS) system is in operation with BOTH of the B RWS Subsystem Pumps, 1P-117B, B RWS Pump, and 1P-117D, D RWS Pump, running in AUTO.
RWS HSS-2911B "LOAD SHED AUTO START RWS PUMP SELECT" is selected to 1P-117B.
Then, a loss of offsite power occurs and the B SBDG restores power to bus 1A4 as designed.
Which of the following describes the response of the B and D RWS Pumps to this transient?
- a. Immediately after bus 1A4 restoration, the B RWS Pump will automatically start.
D RWS Pump will not automatically start.
- b. 2 minutes after bus 1A4 restoration, both the B and D RWS Pumps will automatically start.
- c. Immediately after bus 1A4 restoration, the B RWS Pump will automatically start.
2 minutes after bus 1A4 restoration, the D RWS Pump will automatically start.
- d. 2 minutes after bus 1A4 restoration, the B RWS Pump will automatically start.
D RWS Pump will not automatically start.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 149 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295003 AA2.04 Importance Rating 3.5 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER : System lineups Proposed Question: RO Question # 51 Proposed Answer: D A: Incorrect - B Pump will auto start after a 2 minute time delay.
B: Incorrect - Even though it is in AUTO, the D pump will not auto start.
C: Incorrect - B Pump will auto start after a 2 minute time delay. Even though it is in AUTO, the D pump will not auto start.
D: Correct - Per SD 410, page 8, There are two Load Shed Auto Start RWS Pump Select switches. HSS-2911A is a selector switch for Pump 'A' or Pump 'C' and HSS-2911B is a selector switch for Pump 'B' or Pump 'D'. The non-operating RWS pump in each subsystem is normally selected. The selected pump will automatically restart following a loss of offsite power. If the selected pump was not running prior to the power loss, it will start immediately when power is regained from respective emergency diesel. If the selected pump was previously running, it will restart after the 2 minute time delay is satisfied.
Technical Reference(s): SD 410 Rev 8 page 8 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # 20625 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 150 Exam Series A
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 151 Exam Series A
1 Point
- 52. A plant scram from full power was required. After the operator inserted a manual scram using the pushbuttons on 1C05, the operator questioned the status of control rods.
- The Mode Switch is in SHUTDOWN.
- The scram is NOT reset.
- All green FULL IN lights are ON.
- The rod select power switch was taken to OFF, and then returned to ON.
Which one of the following describes the status of:
(1) White Refuel Select Permissive light, AND (2) Rod position indication on four rod display?
- a. (1) ON (2) ALL rods indicate 00
- b. (1) ON (2) No rod indication is present
- c. (1) OFF (2) ALL rods indicate 00
- d. (1) OFF (2) No rod indication is present Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 152 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295037 EA2.05 Importance Rating 4.2 EA2.05 - Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Control rod position Proposed Question: RO Question # 52 Proposed Answer: D A: Incorrect - A rod must be selected for rod indication to be available and the light to be ON B: Incorrect - A rod must be selected for the light to be ON C: Incorrect - A rod must be selected for rod indication to be available D: Correct - Per SD 856-1, pages 9 & 23, The ROD SELECT POWER Switch is a two position, maintained contact switch located on Panel 1C05. The switch positions are ON and OFF. The switch provides + 28 VDC from RMCS power supplies to the Rod Select Matrix.
Selecting REFUEL on the Reactor Mode Switch energizes K21. Contacts of K21 contribute to the refuel mode one rod permissive relay circuit (K23A and B).
If no rods are selected, and RPIS indicates all rods are fully inserted, the REFUEL SELECT PERMISSIVE white light will be lit if the mode switch is in REFUEL.
Technical Reference(s): SD 856-1 Rev 5 pages 9 & 23 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # 20168 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 153 Exam Series A
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 154 Exam Series A
1 Point
- 53. The plant is operating at 90% power.
Due to thunderstorms in the area causing grid instability conditions, the load dispatcher requests increasing Megavars from 100 to 250.
Given the attached Generator Reactive Capability Curve and the following information:
- Hydrogen Pressure = 30 psig
- Generator Megavars Lagging = 100
- Megawatts = 600 Determine what, if any, curve limitation will be exceeded if the Megavars are increased as requested.
- a. NO curve limitation will be exceeded.
- b. The curve limitation for Field heating will be exceeded.
- c. The curve limitation for Armature heating will be exceeded.
- d. The curve limitation for Armature Core End heating will be exceeded.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 155 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 700000 AA2.01 Importance Rating 3.5 Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Operating point on the generator capability curve.
Proposed Question: RO Question # 53 Proposed Answer: B A: Incorrect - The curve limitation for Field heating will be exceeded.
B: Correct - Per Generator Reactive Capability Curve in OI 698 App.1, The curve limitation for Field heating will be exceeded.
C: Incorrect - The curve limitation for Field heating will be exceeded.
D: Incorrect - The curve limitation for Field heating will be exceeded.
Technical Reference(s): OI 698 App.1, Rev 70 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: OI 698 App.1 Learning Objective: (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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 156 Exam Series A
55.43 (4) Secondary coolant and auxiliary systems that affect the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 157 Exam Series A
1 Point
- 54. The plant was operating at full power when a loss of feedwater heating occurred.
The plant has been scrammed. Indications of fuel damage exist.
The following annunciators are in alarm:
- 1C05B (C-2) MAIN STEAM LINE HI HI RAD / INOP TRIP
- 1C05B (D-2) MAIN STEAM LINE HI RAD
- 1C03A (A-4) OFFGAS VENT PIPE RM-4116A/B HI-HI RAD The CRS has entered AOP 672.2 Offgas Radiation/Reactor Coolant Activity High.
Which one of following describes actions that will automatically occur and other required manual actions?
- a. All MSIVs, the Main Steam Line Drain Valves and the Recirc Sample CVs will automatically close.
SBGT will automatically start.
- b. The Main Steam Line Drain Valves and the Recirc Sample CVs will automatically close.
The MSIVs will NOT automatically close and must be manually closed.
SBGT will NOT automatically start.
- c. The Main Steam Line Drain Valves and the Recirc Sample CVs will automatically close.
SBGT will automatically start.
The MSIVs will NOT automatically close and must be manually closed.
- d. The Main Steam Line Drain Valves will automatically close.
SBGT will NOT automatically start.
The MSIVs will NOT automatically close and must be manually closed.
The Recirc Sample CVs will NOT automatically close.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 158 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295038 2.1.23 Importance Rating 4.3 (High Offsite release rate) Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Proposed Question: RO Question # 54 Proposed Answer: C A: Incorrect - The MSIVs must be closed manually, they do not auto close on steam line high rad.
B: Incorrect - SBGT auto starts on a Group 3 signal C: Correct - Per AOP 672.2 automatic action section and followup action step 6. The Main Steam Line Drain Valves and the Recirc Sample CVs should have closed. SBGT should have started. The MSIVs must be manually closed.
D: Incorrect - The Recirc Sample CVs auto close. SBGT auto starts on a Group 3 signal Technical Reference(s): AOP 672.2 Rev 33 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 7,10 55.43 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 159 Exam Series A
(7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 160 Exam Series A
1 Point
- 55. The plant was operating at full power when a fire in the control room required the control room to be abandoned.
You have been dispatched to 1C208 to operate RCIC to control reactor water level.
Which one of the below describes RCIC operations that can be performed at panel 1C208 and the location where RPV level may be monitored?
- a. Control of RCIC TURB SPEED CONTROLLER HIC-2440 with no local speed or flow indication.
RPV may be monitored on panel 1C388, Remote Shutdown Panel.
- b. Control of RCIC INJECTION VALVE MO-2512 and RCIC TURB SPEED CONTROLLER HIC-2440 with no local speed or flow indication.
RPV may be monitored on panel 1C388, Remote Shutdown Panel.
- c. Control of TURBINE STEAM SUPPLY MO-2404 and RCIC TURB SPEED CONTROLLER HIC-2440 with local flow indication ONLY.
RPV level may be monitored on RCIC Panel 1C208.
- d. Control of RCIC TURB SPEED CONTROLLER HIC-2440 with no local speed or flow indication.
RPV level may be monitored on RCIC Panel 1C208.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 161 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295016 AA1.06 Importance Rating 4.0 Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT: Reactor Water Level Proposed Question: RO Question # 55 Proposed Answer: A A: Correct - Per AOP 915 page TAB2 note prior to step 2, RCIC can be controlled from 1C208 with HIC-2440 but no flow or speed indication is available.
B: Incorrect - ONLY RCIC turbine speed with HIC-2440 can be controlled at the remote panel.
C: Incorrect - RCIC can be controlled from 1C208 with HIC-2440 ONLY. There is no panel 1C208 indication for RCIC speed, flow or RPV level.
D: Incorrect - RCIC can be controlled from 1C208 with HIC-2440 ONLY. There is no panel 1C208 indication for RCIC speed, flow or RPV level.
Technical Reference(s): AOP 915 Rev 39 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # 19072 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 162 Exam Series A
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 163 Exam Series A
1 Point
- 56. During a prolonged loss of the instrument and service air system, the operator is directed to maintain reactor water level in the normal operating band using a specific method stated in AOP 518, Failure of Instrument and Service Air.
What is the method and the reason for its use?
- a. Throttle the A and B FEEDLINE BLOCK valves MO-1636 and MO-1592 if the Feed Reg Valves lockup.
- b. Throttle the A and B FEEDLINE BLOCK valves MO-1636 and MO-1592 due to the possibility of the Feed Reg Valves drifting CLOSED.
- c. Throttle the Feed Reg Valves because the FEEDLINE BLOCK valves MO-1636 and MO-1592 will NOT properly operate due to excessive differential pressure across the valves with a Feedwater pump running.
- d. Throttle the Feed Reg Valves because the FEEDLINE BLOCK valves MO-1636 and MO-1592 will have locked up.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 164 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295019 2.1.20 Importance Rating 4.6 Conduct of Operations: Ability to interpret and execute procedure steps.
Proposed Question: RO Question # 56 Proposed Answer: A A: Correct - Per AOP 518 Note & Caution prior to step 10. - During a prolonged loss of air casualty the Feed Reg valves may drift open.
If the Feed Reg Valves lock up, Maintain Reactor water level in the normal operating band by throttling A and B FEEDLINE BLOCK valves MO-1592 and MO-1636 B: Incorrect - the FRVs would drift OPEN C: Incorrect - This is not specified in the AOP. The block valves may not REOPEN due to excessive DP with a Feedwater pump running.
D: Incorrect - The Feedline Block valves are not air operated and will not lockup Technical Reference(s): AOP 518 Rev 31 page 4 & 5 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
19511 Modified Bank # (Note changes or attach parent) 19512 New Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 165 Exam Series A
ORIGINAL QUESTION Due to lowering Instrument Air Pressure, AOP-518, FAILURE OF INSTRUMENT AND SERVICE AIR is being executed.
1C05A, F-1, A or B FEED REG VALVE POSITION LOCKED is activated.
If Instrument Air pressure cannot be restored, how are Feedwater Regulating Valves, CV-1579 and CV-1621, affected; and what procedural action(s) is (are) required?
Feedwater Regulating Valves will fail:
_____ A. SHUT; it is required to reduce Reactor Power to control RPV Water Level.
_____ B. SHUT; it is required to use the FEEDWATER STARTUP CONTROL VALVE, CV-1622 to control RPV Water Level.
_____ C. OPEN; it is required to THROTTLE A AND B FEEDLINE BLOCK Valves, MO-1592 and MO-1636 to control RPV Water Level.
_____ D. OPEN; it is required to completely SHUT A FEEDLINE BLOCK Valve, MO-1592, OR B FEEDLINE BLOCK Valve, MO-1636, to control RPV Water Level Question History: Last NRC Exam: 2007 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 166 Exam Series A
1 Point
- 57. The reactor was shutdown six days ago for a Refueling Outage. Reactor Coolant System Temperature is 150°F. Core Alterations have NOT been performed.
The cavity is flooded up and the fuel pool gates are removed.
With NO Decay Heat Removal, how long will it take for Reactor Coolant System Temperature to reach 200°F using the provided reference?
- a. <2 hours
- b. 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
- c. 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
- d. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 167 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295021 2.1.25 Importance Rating 3.9 Ability to interpret reference materials, such as graphs, curves, tables, etc.
Proposed Question: RO Question # 57 Proposed Answer: C A: Incorrect - would be correct if Appendix 2 was used in error B: Incorrect - would be correct if Appendix 3 was used in error C: Correct - Using AOP-149 Appendix 1 Curve, 5°F / hr heatup rate should be obtained.
200°F - 150°F = 50°F 50°F / 5 °F / hr = 10 hr using Appendix 1 of AOP 149, a 5 degree per hour heatup rate is obtained. To rise from 150 to 200 degrees is a 50 degree change. At 5 degrees per hour that would take 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> D: Incorrect - would be correct if initial temperature was 100 degrees and Appendix 1 was used in error Technical Reference(s): AOP 149 Rev 31 (Attach if not previously provided)
AOP 149 Proposed References to be provided to applicants during examination: Appendices 1,2 and 3
Learning Objective: (As available)
Question Source: Bank # X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 168 Exam Series A
10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 169 Exam Series A
1 Point
- 58. The reactor scrammed and the MSIVs have closed due to a small break in the piping from the Main Steam Line Equalizing Header.
SRVs are now being cycled to control reactor pressure. Suppression Pool level has risen to 13.8 feet.
If Suppression Pool level cannot be restored and maintained below 13.8 feet, Emergency Depressurization is required because _____.
- a. Suppression Pool level is approaching the Safety Relief Valve Tailpipe Vacuum Breakers.
- b. the containment spray ring header is completely submerged and containment integrity may be compromised.
- c. continued SRV operation may cause tailpipe damage and directly pressurize containment.
- d. a large break LOCA will result in drywell pressure exceeding design due to Suppression Pool level approaching the Torus-to-Drywell vacuum breaker level.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 170 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295025 EA2.04 Importance Rating 3.9 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Suppression pool level Proposed Question: RO Question # 58 Proposed Answer: C A: Incorrect - These are located in the drywell B: Incorrect - The level of the bottom of the ring header is 13.83 ft C: Correct - Per EOP 2 Bases page 16 - the SRV Tail Pipe Level Limit is the highest torus water level at which opening of an SRV will not result in exceeding the code allowable stresses in the SRV tail pipe, tail pipe supports, T-quencher, or T-quencher supports.
The SRV Tail Pipe Level Limit is a function of torus water level and RPV pressure. SRV operation with torus water level above the SRV Tail Pipe Level Limit could damage the SRV discharge lines. This, in turn, could lead to containment failure from direct pressurization and damage to equipment inside the containment (ECCS piping, RPV water level instrument runs, torus-to-drywell vacuum breakers, etc.) from pipe-whip and jet-impingement loads.
D: Incorrect - This level is at 13.5 ft and Emergency Depress is not required at that point Technical Reference(s): EOP 2 Bases Rev 13 page 16 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # WTS 2468 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 171 Exam Series A
Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3 55.43 (3) Mechanical components and design features of the reactor primary system.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 172 Exam Series A
1 Point
- 59. The plant is at full power. A loss of the running CRD pump has occurred.
Which one of the following describes the motive force to insert the control rods if accumulator pressures fall too low to accomplish the task?
- a. A ball check valve, located in the CRDM cylinder flange inlet port, repositions and allows reactor water pressure to act on the CRDM under piston area.
- b. Drive header pressure unseats a ball check valve, located in the CRDM cylinder flange inlet port, creating a flow path for reactor water through the drive header to the CRDM under piston area.
- c. A ball check valve, located in the CRDM cylinder flange withdraw port, repositions and allows reactor water pressure to act on the CRDM under piston area.
- d. Ball check valves, located in the CRDM cylinder flange inlet and withdraw ports, reposition and allow reactor water pressure to act on the CRDM under piston area.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 173 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295022 AK1.01 Importance Rating 3.3 Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS: Reactor pressure vs. rod insertion capability Proposed Question: RO Question # 59 Proposed Answer: A A: Correct - Per SD 255, page 46 - As water is forced from the accumulator, accumulator pressure falls below reactor pressure and causes the ball check valve (located in the insert port) to shift its position. This admits reactor pressure into the under piston area, completing the scram stroke.
B: Incorrect - Reactor pressure unseats the ball check valve C: Incorrect - the ball check is located in the cylinder flange inlet port D: Incorrect - there is only one ball check in the CRDM Technical Reference(s): SD 255 Rev 8 page 46 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 10.07.01.06-07 (As available)
Question Source: Bank # 19991 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 174 Exam Series A
55.43 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 175 Exam Series A
1 Point
- 60. The plant was operating at full power when an event occurred. The following conditions exist:
- Reactor Level is 58 and lowering slowly
- Reactor pressure is 750 psig and lowering slowly
- Drywell Pressure is 3.5 psig and rising slowly
- Drywell temperature is 152°F and rising slowly
- All Low Pressure ECCS pumps are running
- Well Water is in service Which one of the following actions, if any, is required to maximize drywell cooling IAW EOPs?
- a. The Drywell Cooling Loop A and B MODE SELECT hand switches HS-5718A and B must be taken to INOPERATIVE and returned to START. This will reset the 2 psig seal-in and allow the fans to shift to fast speed.
- b. Install Defeat 4. Well Water must be shutdown then restarted.
- c. Install Defeat 4. Well water does NOT need to be secured.
- d. No switch manipulations are required. Drywell Cooling Fans will have shifted to fast speed on the high drywell pressure signal.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 176 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295012 AK2.02 Importance Rating 3.6 Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following:
Drywell cooling Proposed Question: RO Question # 60 Proposed Answer: B A: Incorrect - Defeat 4 must be installed due to the PCIS Group 7 isolation on lo-lo-lo RPV level B: Correct - due to the PCIS Group 7 isolation on lo-lo-lo RPV level Defeat 4 must be installed and well water must be secured prior to installation to prevent water hammer.
(See Defeat 4 installation document)
C: Incorrect - Well water must be secured D: Incorrect - Defeat 4 must be installed Technical Reference(s): Defeat 4 Rev 8 page 2 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 177 Exam Series A
(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 178 Exam Series A
1 Point
- 61. A plant event resulted in a steam leak into secondary containment and rising secondary containment ventilation radiation levels and release rates.
The CRS has entered several EOPs including EOP 3 Secondary Containment Control.
What purpose does a reactor scram and emergency depressurization achieve as it relates to EOP 3?
- a. It allows establishment of adequate core cooling using low pressure ECCS pumps.
- b. It reduces the energy in the RPV before reaching conditions where the primary containment will not accommodate an SRV opening.
- c. It places the primary system in a low energy condition to reduce the driving head of the leak.
- d. It places the RPV in a low energy condition before reaching conditions where a loss of coolant accident could not be adequately quenched in the primary containment.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 179 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295034 EK3.05 Importance Rating 3.6 Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION : Manual SCRAM and depressurization:
Plant-Specific Proposed Question: RO Question # 61 Proposed Answer: C A: Incorrect - This is not the purpose of ED for this event. This would be correct in the event of a LOCA and lowering level.
B: Incorrect - Containment parameters such as increasing drywell pressure are not an issue in the described event. Accommodating SRV openings is not an issue for this event.
C: Correct - Scramming the reactor reduces the energy that the RPV may be discharging to the secondary containment to decay heat levels. If the RPV is the source of energy, radiation or water being released to secondary containment, scramming the reactor should greatly reduce any further release and may prevent the need for the more severe action of emergency depressurizing the RPV.
RPV depressurization places the primary system in its lowest possible energy state, rejects heat to the torus in preference to outside the containment, and reduces the driving head and flow of primary systems that are unisolated and discharging into the secondary containment.
D: Incorrect - The concern is not a LOCA in EOP 3.
EOP 3 bases Rev 10, pages 18 &
Technical Reference(s): (Attach if not previously provided) 19 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # DAEC NRC 2002 Modified Bank # (Note changes or attach parent)
New Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 180 Exam Series A
Question History: Last NRC Exam: 2002 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 181 Exam Series A
1 Point
- 62. A plant event has occurred.
With HPCI operating in the Pressure Control Mode with the Flow Indicating Controller (FIC) in automatic, Reactor Pressure is at 1050 psig and slowly rising.
Which ONE of the following will cause Reactor Pressure to lower?
(Assume the ONLY effect on RPV pressure is HPCI operation)
- a. Shut MO-2202, TURBINE STEAM SUPPLY VALVE.
- b. Adjust CV-2315, TEST BYPASS VALVE from 47% to 55% open.
- c. Adjust CV-2315, TEST BYPASS VALVE from 47% to 40% open.
- d. Adjust FIC-2309, HPCI FLOW CONTROL from 3000 gpm to 2600 gpm.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 182 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295007 AA1.02 Importance Rating 3.5 Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE :
HPCI Proposed Question: RO Question # 62 Proposed Answer: C A: Incorrect - this would reduce steam flow to the HPCI turbine and not lower RPV pressure B: Incorrect - This would cause HPCI to pump at a lower discharge head requiring less energy. Therefore RPV pressure would rise C: Correct - Throttling shut CV-2315, TEST BYPASS VALVE from 47% to 40%
Open, will cause HPCI to pump the same recirculation flow at a higher discharge head, which consumes MORE energy from the HPCI Turbine, requiring more Main Steam flow to the HPCI Turbine, which lowers Reactor Pressure.
D: Incorrect - lowering speed will reduce the energy required to operate the HPCI turbine.
Wth less energy required for HPCI, RPV pressure would rise.
OI 152 Rev 93, discussion - QRC Technical Reference(s): (Attach if not previously provided) 1 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 183 Exam Series A
10 CFR Part 55 Content: 55.41 7 55.43 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 184 Exam Series A
1 Point
- 63. The plant is shutdown for refueling with the following conditions:
- Core Alterations are in progress.
- The Refueling Supervisor reports that a fuel bundle has been loaded into the wrong reactor core location.
- The Control Room operator observes that Source Range count indication for the SRM in that quadrant, has increased and stabilized at a higher value.
As a result of this event, Shutdown Margin (SDM) has __(1)__ and the reactor __(2)__.
- a. (1) increased (2) remains subcritical
- b. (1) increased (2) is super-critical
- c. (1) decreased (2) remains subcritical
- d. (1) decreased (2) is super-critical Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 185 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295014 AA2.01 Importance Rating 4.1 Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION : Reactor power Proposed Question: RO Question # 63 Proposed Answer: C Explanation: The KA matches because any inadvertent reactivity addition and its relation to reactor power is contingent on reactor theory knowledge A: Incorrect - SDM decreases anytime fuel is added to the core.
B: Incorrect - SDM decreases anytime fuel is added to the core. Counts would be increasing if the reactor was supercritical.
C: Correct - SDM decreases anytime fuel is added to the core. Since count stabilized in the SR the reactor is still sub-critical D: Incorrect - Counts would be increasing if the reactor was supercritical.
TS Definition Technical Reference(s): GFES Rx Theory Chap 2 pg 20 (Attach if not previously provided)
GFES Rx Theory Chap 3 pg 8 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 1 55.43 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 186 Exam Series A
(1) Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 187 Exam Series A
1 Point
- 64. The plant has been operating at 90% thermal power for several days. Over the last several hours the following conditions have changed:
- Condenser Backpressure has risen from 2.5 Hg to 3.5 Hg
- MWE lower by 10 MW
- There have been NO alarms received Which one of the following could be the cause of these indications?
- a. Air leak into the condenser.
- b. Offgas premature recombination.
- c. One or more blown Offgas loop seals.
- d. Cooling tower outlet temperature increased.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 188 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295002 2.4.47 Importance Rating 4.2 (Loss of Main Condenser Vacuum) Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
Proposed Question: RO Question # 64 Proposed Answer: D A: Incorrect - An air leak would cause Backpressure to increase but would also cause Offgas flow to increase.
B: Incorrect - Premature recombination would cause Offgas flow to decrease but not Backpressure increase. There would also be several alarms associated with this problem.
C: Incorrect - Blown loop seals would cause Offgas flow to decrease but not Backpressure increase. There would also be radiation alarms associated with this problem.
D: Correct - The rise in cooling tower outlet temperature causes less condensate subcooling, the result is lower plant efficiency causing the lowering of MWE and higher circ water temp will cause less condensing of steam resulting in a higher condenser backpressure. Less vacuum in the condenser will cause less air inleakage lowering the Offgas system flow.
ARP 1C07B (D-9) rev 69 Technical Reference(s): (Attach if not previously provided)
AOP 672.3 Rev 12 Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 189 Exam Series A
10 CFR Part 55 Content: 55.41 7 55.43 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 190 Exam Series A
1 Point
- 65. Technical Specifications require closing the RWCU Primary Containment Isolation Valves if a RWCU area high temperature were to occur.
What is the reason for this requirement?
- a. To ensure that the release of radioactive material to the environment will be consistent with the assumptions used in the final safety analyses.
- b. To minimize moisture buildup and overheating in the Standby Gas Treatment System charcoal beds.
- c. To prevent exceeding the Environmental Qualification temperature limits on the electrical buses in the Turbine Building required for safe shutdown.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 191 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295032 EK3.03 Importance Rating 3.8 Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE : Isolating affected systems Proposed Question: RO Question # 65 Proposed Answer: A A: Correct - Per TS bases 3.3.6.1. - The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs). Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA.
B: Incorrect - This would be a concern but is not the reason associated with the TS required PCIV isolation.
C: Incorrect - the reason is to limit the environmental release and not for EQ concerns.
D: Incorrect - the reason is to limit the environmental release.
Technical Reference(s): TS bases 3.3.6.1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 10.07.01.06-07 (As available)
Question Source: Bank # WTS 1326 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 192 Exam Series A
10 CFR Part 55 Content: 55.41 7 55.43 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 193 Exam Series A
1 Point
- 66. Which ONE (1) of the following correctly completes the statement below?
In accordance with ACP 110.1, Conduct of Operations, when an operating crew correctly employs a conservative decision-making policy, all crewmembers understand that when faced with unexpected or uncertain plant conditions, they must __(1)__, and that all decisions must be based on maintaining __(2)__.
- a. (1) place the plant in a safe condition (2) nuclear and industrial safety
- b. (1) place the plant in a safe condition (2) nuclear safety ONLY
- c. (1) reduce power or scram the reactor (2) nuclear and industrial safety
- d. (1) reduce power or scram the reactor (2) nuclear safety ONLY Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 194 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # 2.1.39 Importance Rating 3.6 Conduct of Operations: Knowledge of conservative decision making practices.
Proposed Question: RO Question # 66 Proposed Answer: A Explanation (Optional): The KA is matched because the operator must demonstrate knowledge of conservative decision making practices, specifically, its principle (decisions are made with regard to both Nuclear and Industrial Safety) and expectations (the goal is to place the plant in a safe operating condition).
A. Correct - According to SOER 94-1 (p26; Rev 1), which is referenced by LP 50007-93.16 (p5; Rev 16), the key message in conservative decision-making is the expectation that operators, when faced with unexpected or uncertain conditions, must place the plant in a safe condition, and must not hesitate to reduce power or scram the reactor.
DAEC, using Attachment 5 of ACP 110.1, has expanded the message and the process to include both Nuclear and Industrial Safety. According to ACP 110.1 (p5; Rev 17)
Step 5.1.1, plant operations shall be conducted in a manner that establishes nuclear and personal safety as the highest priority while employing a conservative decision making process. According to ACP 110.1 (p20-21; Rev 17) Attachment 5, personal safety is stated as industrial safety. The principle statement of conservative decision-making policy identifies that Nuclear and industrial safety is maintained at the forefront of all decisions. Based on stated expectations of Attachment 5, the overall philosophy of conservative decision-making must be summed up by stating that when unexpected or uncertain plant conditions, operators must place the plant in a safe condition, rather than the more narrow approach of reducing power or scramming the reactor, although these strategies may be a part of such as policy. For instance, LP 50007-94.24 (p7; Rev 7) instructs the operator to consider the conservative decision-making policy of ACP-110.1, Attachment 5, in regard to sending individuals to perform outside Inspections during severe weather events such as a tornado or a thunderstorm. This clearly indicates that the DAEC policy is more broad based than SOER 94-1 and includes a focus on both nuclear and industrial safety.
B. Incorrect - 1st part correct, 2nd part wrong. This is plausible because SOER 94-1 was written to emphasize nuclear safety.
C. Incorrect - 1st part wrong, 2nd part correct. This is plausible because SOER 94-1 specifically states that the operators when making conservative decisions with respect to nuclear safety should not hesitate to reduce power or scram the reactor.
D. Incorrect - 1st part wrong, 2nd part wrong. This is plausible because SOER 94-1 specifically states that the operators when making conservative decisions with respect to nuclear safety should not hesitate to reduce power or scram the reactor, and SOER 94-1 was written to emphasize nuclear safety.
SOER 94-1 (p26; Rev 1)
Technical Reference(s): LP 50007-93.16 (p5; Rev 16) (Attach if not previously provided)
ACP 110.1 (p5, 20-21; Rev 17)
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 195 Exam Series A
LP 50007-94.24 (p7; Rev 7)
Proposed References to be provided to applicants during examination: No LP 50007 96.07, Objectives Learning Objective: 96.07.04.01 & 96.07.04.03 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: NA Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 196 Exam Series A
1 Point
- 67. The plant is operating at full power. All systems are operable.
Which one of the following describes how the Offgas & Recombiner System functions to create a substantial reduction in the release of radioactive materials to the environment?
- a. The system reduces the volume of Offgas flow; AND Delays the release of Hydrogen and Oxygen to the environment.
- b. The system reduces the volume of Offgas flow; AND Delays the release of Xenon and Krypton to the environment.
- c. The system recombines short-lived radioactive gases; AND Delays the release of Hydrogen and Oxygen to the environment.
- d. The system recombines short-lived radioactive gases; AND Delays the release of Xenon and Krypton to the environment.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 197 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # 2.1.27 Importance Rating 3.9 Conduct of Operations: Knowledge of system purpose and/or function.
Proposed Question: RO Question # 67 Proposed Answer: B A. Incorrect. 1 part correct, 2nd part wrong. This is plausible because the operator may st incorrectly believe that it is the Hydrogen and Oxygen that undergoes a delayed release.
B. Correct. 1st part correct, 2nd part correct. According to SD-672 (p5-7; Rev 13), the Offgas and Recombiner System is functionally two subsystems: the recombiner subsystem and the charcoal adsorber subsystem. The recombiner reduces the total volume of the offgas flow by recombining radiolytically dissociated hydrogen and oxygen to produce water vapor. After recombination, the offgas flow consists of small volume amounts of fission product and activation gases carried in the airflow arising out of inleakage to the condenser. This offgas stream is delayed for decay of short lived radioactive isotopes, and then conditioned to a low moisture content and the proper temperature for maximum delay in the charcoal adsorber system. The long holdup time produced by the charcoal adsorbers permits the xenon and krypton gases to decay to particulate daughter products, which either remain on the charcoal or are removed by high efficiency particulate (HEPA) filters. The composite effect of the reduction in system gas volume and the delay produced by charcoal adsorption results in a substantial reduction in the release of radioactive materials to the environment.
C. Incorrect. 1st part wrong, 2nd part wrong. This is plausible because the operator may incorrectly believe that it is the short lived activation gases (i.e. N-16, O-19, and N-13, which are all gases expected to be within the system) that are recombined, and that it is the Hydrogen and Oxygen that undergoes a delayed release.
D. Incorrect. 1st part wrong, 2nd part correct. This is plausible because the operator may incorrectly believe that it is the short lived activation gases (i.e. N-16, O-19, and N-13, which are all gases expected to be within the system) that are recombined.
SD-672 (p5-7; Rev 14)
Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: No LP 50007 47.0 Objective 47.00.00.01 Learning Objective: (As available)
Question Source: Bank # 22304 Modified Bank # (Note changes or attach parent)
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 198 Exam Series A
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 199 Exam Series A
1 Point
- 68. While performing a step in a Surveillance Test Procedure, it has been determined that a normally open, motor operated Primary Containment Isolation Valve will not stroke in the closed direction, as required by the procedure.
Which one of the following identifies when the Technical Specification LCO action time is started?
The LCO time would start_____.
- a. when the surveillance is logged as complete.
- b. as soon as the valve failure was recognized.
- c. when the start of the surveillance was logged on.
- d. at the time the surveillance was last satisfactorily completed.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 200 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # 2.2.12 Importance Rating 3.7 Equipment Control: Knowledge of surveillance procedures.
Proposed Question: RO Question # 68 Proposed Answer: B A. Incorrect - the LCO is entered upon discovery of the issue B. Correct - IAW TS 3.0.2 - Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.
C. Incorrect - the LCO is entered upon discovery of the issue D. Incorrect - the LCO is entered upon discovery of the issue Technical Reference(s): TS LCO 3.0.2. (Attach if not previously provided)
Proposed References to be provided to applicants during examination: No 1.07.03.04 Learning Objective: (As available)
Question Source: Bank # X - NMP 2005 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 201 Exam Series A
1 Point
- 69. The plant is operating at full power.
At the start of your shift (0700) the following conditions exist.
- Annunciator 1C-06A (A-1) - "A" RWS PIT LO LEVEL is in alarm and is LIT due to scheduled ongoing preventive maintenance. The annunciator became LIT when work began 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago. The work will be completed in 6 more hours.
- Annunciator 1C-06A (B-5) - "A" COOLING TOWER HI/LO LEVEL alarmed at 0000 and is LIT due a failed instrument. Maintenance will complete the work, including retests, in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Which one of the following describes how these alarms are tracked IAW ACP 1410.12, Operator Burden Program?
(Assume that maintenance will complete the work and retests as scheduled)
- a. Neither annunciator is required to be entered in the Operator Burden Database.
- b. BOTH annunciators are required to be entered in the Operator Burden Database by 1200.
- c. ONLY the "A" COOLING TOWER HI/LO LEVEL annunciator is required to be entered in the Operator Burden Database by 1200.
- d. ONLY the "A" RWS PIT LO LEVEL annunciator is required to be entered in the Operator Burden Database immediately.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 202 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # 2.2.43 Importance Rating 3.0 Equipment Control: Knowledge of the process used to track inoperable alarms.
Proposed Question: RO Question # 69 Proposed Answer: C A. Incorrect - The "A" COOLING TOWER HI/LO LEVEL must be entered because it was due to a failed instrument B. Incorrect - The "A" RWS PIT LO LEVEL is not required to be entered due to work being done for preventive maintenance C. Correct - Per the ACP, The "A" COOLING TOWER HI/LO LEVEL must be entered because it was due to a failed instrument D. Incorrect - The "A" RWS PIT LO LEVEL is not required to be entered due to work being done for preventive maintenance ACP 1410.12 Rev 16 - Definition Technical Reference(s): (Attach if not previously provided) 3 Proposed References to be provided to applicants during examination: No Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: None Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 203 Exam Series A
1 Point
- 70. Which one of the following activities requires that a plant area be evacuated and a normally unlocked room be locked to prevent plant personnel from being over-exposed to radiation during the activity?
- a. Conducting TIP traces while at power.
- b. Conducting a planned cold start of the HPCI Pump.
- c. Isolating RWCU while at power.
- d. Shifting the Spent Fuel Cooling Filter Demineralizer from the Filter to the Hold Mode.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 204 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # 2.3.12 Importance Rating 3.2 Radiation Control: Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
Proposed Question: RO Question # 70 Proposed Answer: A A. Correct. According to SD-878.6 (p15; Rev 6), due to exposure to the high neutron flux in the core during detector operations, the TIP becomes highly radioactive through activation reactions. To provide protection for personnel, the probe is retracted for storage in a shielded chamber. After the use of the TIP during reactor power operation, radiation levels in the vicinity of the chamber shield may be in the range of 200-500 mRem/hr. According to OI 878.6 (p8; Rev 39), Step 4.1.2, prior to conducting TIP traces at power, the operator must contact Health Physics to ensure there are no personnel in the TIP Machine Room and close and lock the door or ensure the TIP Drive Machine Room Door is under control of Health Physics.
B. Incorrect. According to OI 152 (7; Rev 93), access to the HPCI Room shall be controlled during a planned cold start until steady state operating conditions are achieved. Personnel may be present in the room, if necessary, to 1) locally manipulate valves or controls; or 2) perform troubleshooting which cannot be performed in a practical manner from outside the room. Otherwise, prior authorization from the Control Room Supervisor is required. While the access to the room is controlled during planned cold pump starts, it is NOT controlled to minimize radiation exposure, but rather for industrial safety reasons. According to LP 50007_5.0 (p21; Rev 10), during a quarterly surveillance test of the HPCI system at the Quad Cities Station, the exhaust line rupture disc on the Unit One HPCI turbine burst, releasing steam into the HPCI room, burning, and slightly contaminating, five workers. At the DAEC, access to the HPCI and RCIC rooms is controlled during planned cold starts. If an individual must be present in the room during a planned cold start, that individual must contact the OSS for any special precautions which must be taken prior to gaining access to the room. Planned starts of equipment are routinely announced over the plant paging system. Operators must evaluate each task for its possible consequences with respect to personal safety.
C. Incorrect. According to OI 261 (13; Rev 79), when RWCU is isolated, the alternate sample line dose rates raise by a factor of 10. This line is located near the entrance to the RWCU Pump room. The increased dose rates affect the Security Post stationed near that sample line. Because of this, when the RWCU System is isolated Health Physics must be notified that dose rates near RWCU pump room door will be elevated and to coordinate with Security to relocate officers, as allowed. However, the room is NOT isolated to control exposure to individuals, also RWCU HX room normally locked.
D. Incorrect. According to OI 435 (p24; Rev 55), when backwashing the SPC Filter Demineralizer in accordance with Section 7.7, The HP Shift Technician may restrict access to the Jungle Room during Fuel Pool bed backwashes due to radiation levels, the room is not locked. While this procedure includes going to the Hold Mode of Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 205 Exam Series A
operation as a prerequisite, it does NOT by itself require that access to specific areas of the plant be restricted for the personal exposure control.
SD-878.6 (p15; Rev 6)
OI 878.6 (p8; Rev 39), Step 4.1.2 OI 152 (7; Rev 93)
Technical Reference(s): LP 50007_5.0 (p21; Rev 10) (Attach if not previously provided)
OI 261 (13; Rev 79)
OI 435 (p24; Rev 55)
Proposed References to be provided to applicants during examination: No 50007_83.0, Objective 83.01.01.02 Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: New Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 12 55.43 (12) Radiological safety principles and procedures.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 206 Exam Series A
1 Point
- 71. DAEC is in a refueling outage.
- The refueling cavity is flooded
- The fuel pool gates are removed
- The cask pool gate is removed If the reactor building to drywell refueling bellows were to fail and the reactor cavity completely drained, which one of the following would pose the initial serious radiological hazard?
- a. The exposed reactor vessel head studs.
- b. The irradiated fuel remaining in the fuel pool.
- c. Irradiated fuel remaining in the reactor vessel.
- d. Irradiated components set on the floor in the cask pool.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 207 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # 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 Proposed Answer: B A. Incorrect - These are normally exposed before the cavity is filled and would not present a problem.
B. Correct - Of the given choices, the irradiation fuel in the fuel pool would have the highest location, so it would be the FIRST concern for radiological hazard.
C. Incorrect - The height of the reactor head flange provides sufficient shielding for the irradiated fuel in the reactor D. Incorrect - Components stored on the floor of the cask pool are kept below the top of the irradiated fuel in the fuel pool, so they would not be the first concern for radiological hazard.
Technical Reference(s): Drawing No. BECH-M009 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: No Learning Objective: (As available)
Question Source: Bank # WTS-Pilgrim 2009 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 12 55.43 (12) Radiological safety principles and procedures.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 208 Exam Series A
1 Point
- 72. With the plant at full power, the following events occur:
- A fire occurs at Bus 1A4.
- Smoke is entering the Control Room.
- The crew implements AOP 913, Fire.
- The crew selects Safe Shutdown Path CB2.
Which one of the following describes the actions required IAW AOP 913?
- a. Place Control Room Ventilation in the Recirc Mode, evacuate the Control Room and establish plant control at the Remote Shutdown Panel.
- b. Evacuate unnecessary personnel from the control room, don SCBAs and place Control Room Ventilation in the Fresh Air Mode.
- c. Place Control Room Ventilation in the Fresh Air Mode, evacuate the Control Room and establish plant control at the Remote Shutdown Panel.
- d. Evacuate unnecessary personnel from the control room, don SCBAs and place Control Room Ventilation in the Recirc Mode.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 209 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # 2.4.27 Importance Rating 3.4 Emergency procedures/Plan: Knowledge of fire in the plant procedures.
Proposed Question: RO Question # 72 Proposed Answer: B A. Incorrect - with a fire at 1A4 bus, power may not be available at the RSP. Only unnecessary personnel must be evacuated from the control room and the ventilation must be placed in Fresh Air Mode B. Correct - Per AOP 913, CB2 attachment Step1. If smoke is detected in the control room. Then -
- a. Evacuate unnecessary personnel from the Control Room
- b. Direct the operating crew to don SCBAs
- c. Immediately halt all maintenance or surveillance testing that could cause a plant trip or transient.
Per the following NOTE: Step 2 is performed to aid in removing/preventing smoke in the Control room by going to Fresh Air Mode.
C. Incorrect - with a fire at 1A4 bus, power may not be available at the RSP. Evacuating the CR is not an AOP requirement D. Incorrect - Fresh Air Mode is required.
ACP 913 Rev 56 (CB2 Technical Reference(s): attachment) (Attach if not previously provided)
Proposed References to be provided to applicants during examination: No LP 50007_94.25 Learning Objective: Objective 94.25.01.02 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: New Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 210 Exam Series A
10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 211 Exam Series A
1 Point
- 73. The plant was operating at full power when an immediate control room evacuation was required due to toxic gases entering the area. There is NO fire present.
IAW AOP 915, Shutdown Outside the Control, which one of the following describes action(s) required, if possible, prior to exiting the control room?
ONLY a ______.
- a. scram of the reactor is required.
- c. scram of the reactor AND trip of the main turbine are required.
- d. scram of the reactor, trip of the main turbine AND start of both SBDGs are required.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 212 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # 2.4.12 Importance Rating 4.0 Emergency procedures/Plan: Knowledge of general operating crew responsibilities during emergency operations.
Proposed Question: RO Question # 73 Proposed Answer: A A. Correct - Per AOP 915 Step 2 - If an immediate evacuation of the control room is required for a non-fire event - only a scram of the reactor is required B. Incorrect -this would be correct for an evacuation due to a fire (AOP 915 Step 1)
C. Incorrect - A main turbine trip is not required per the AOP D. Incorrect - A main turbine trip and start of the SBDGs is not required per the AOP AOP 915 Rev 39 - step 2 Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: No Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: New Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 213 Exam Series A
1 Point
- 74. Which one of the following describes conditions under which a safety limit would be violated?
- a. With reactor steam dome pressure at 1000 psig and core flow at 11% of rated, thermal power is 22% of rated thermal power.
- b. With reactor steam dome pressure at 900 psig and core flow at 9% of rated, thermal power is 22% of rated thermal power.
- c. With one recirc loop in operation, reactor steam dome pressure at 990 psig and core flow at 20% of rated, MCPR = 1.30.
- d. With two recirc loops in operation, reactor steam dome pressure at 1000 psig and core flow at 50% of rated, MCPR = 1.20.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 214 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # 2.2.22 Importance Rating 4.0 Knowledge of limiting conditions for operations and safety limits.
Proposed Question: RO Question # 74 Proposed Answer: B A. Incorrect - Per TS 2.0 - the safety limit is met B. Correct - 2.1.1.1 Fuel Cladding Integrity - With the reactor steam dome pressure <785 psig or core flow < 10% rated core flow: THERMAL POWER shall be 21.7% RTP.
C. Incorrect - Per TS 2.0 - the safety limit is met D. Incorrect - Per TS 2.0 - the safety limit is met Technical Reference(s): TS 2.0 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (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 Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 215 Exam Series A
(5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 216 Exam Series A
1 Point
- 75. With the plant at full power, the following conditions exist:
- You are an NSOE coming in to start your first day of work.
- You last stood watch 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> ago.
- You expect to be working in the Work Control Center, unassigned to the shift.
Four hours into the shift:
One of the Control Room NSOEs is required to attend a briefing, and you have been asked to relieve him for about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Which one of the following identifies:
(1) how far back you are required to read the Station Log prior to taking the watch AND (2) whether or not the NG-016K, NSOE and ANSOE Turnover Form, must be used?
- a. (1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (2) The Turnover form must be used.
- b. (1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (2) The Turnover form is NOT required.
- c. (1) 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (2) The Turnover form must be used.
- d. (1) 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (2) The Turnover form is NOT required.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 217 Exam Series A
Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # 2.1.3 Importance Rating 3.7 Equipment Control: Knowledge of shift or short-term relief turnover practices.
Proposed Question: RO Question # 75 Proposed Answer: A A. Correct. 1 part correct, 2nd part correct. According to ACP 110.1 (p44; Rev 22), the st on-coming watchstander shall review the Station Log back to the last time the individual stood the watch or three days (whichever is less). Since the operator last stood watch 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> ago, the requirement is to review back for ONLY 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. According to ACP 1410.10 (p9; Rev 22), in a section of the procedure entitled Relieving Crew Members During the Shift, it is stated that the appropriate Turnover Form shall be used any time a crew member is relieved. In the event that only one position (Shift Manager, CRS, STA, NSOE, or ANSOE) is relieved, then an N/A should be placed where appropriate on the shift turnover form for the non-relieved crew members.
B. Incorrect. 1st part correct, 2nd part wrong. This is plausible because the operator may incorrectly believe that the form does not need to be used because the relief is in the middle of a shift and of a temporary nature.
C. Incorrect. 1st part wrong, 2nd part correct. This is plausible because the operator last stood the watch 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> ago, and four hours have elapsed since the start of the work day. The operator may incorrectly believe that they are required to read the log back to the time that they last held the shift.
D. Incorrect. 1st part wrong, 2nd part wrong. This is plausible because the operator may incorrectly believe that they are required to read the log back to the time that they last held the shift; and because the operator may incorrectly believe that the form does not need to be used because the relief is in the middle of a shift and of a temporary nature.
ACP 110.1 (p44; Rev 22)
Technical Reference(s): (Attach if not previously provided)
ACP 1410.10 (p9; Rev 22)
Proposed References to be provided to applicants during examination: No LP 50007_1.08, Learning Objective: Objective 1.11.01.02 (As available)
DAEC Q10 1.8.1.2 Question Source: Bank # P327 of RO Exam Bank Modified Bank # (Note changes or attach parent)
New Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 218 Exam Series A
Question History: Last NRC Exam: 1999 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
Course: 50007 Rev. 0 Topic: Final 2009 RO NRC Master 8-10-09.doc Page 219 Exam Series A
Page 1 of 1 APPENDIX 1 ESTIMATED CAPABILITY CURVES OI 698 Page 40 of 42 Rev. 70
AOP APPENDIX 1 APPENDIX 1 HEATUP RATE CURVE - RPV FLOODED NOTE The RPV Flooded condition is defined as the RPV head removed, Spent Fuel Pool Gates removed, and the RPV and refuel cavity flooded up so that cavity level equals Spent Fuel Pool level. Spent Fuel Pool level is within the normal band.
Vessel Water Heatup Rate (Floodup Condition) 8.00 7.00 Heatup Rate (Deg. F/hr.)
6.00 5.00 4.00 3.00 2.00 1.00 0.00 1 7 13 19 25 31 37 43 49 55 61 67 73 79 85 91 97 T ime S ince S hutdown (D a ys)
CAUTION The initial heatup rate in the vessel may be higher than the calculated value when RHR or Fuel Pool Cooling is removed from service. The calculation used to generate the heatup rate curves assumes instantaneous mixing and heat transport from the fuel area to the remainder of the system volume. In addition, the calculated heatup rates reflect bulk temperatures not local temperatures. Under natural circulation conditions and the resulting time delay in heat transport, considerable differences in temperature may exist between the vessel and upper levels of the cavity or in the spent fuel pool. In some cases local boiling may occur but bulk boiling will not occur as long as cooling is restored within the calculated time-to-boil period.
AOP APPENDIX 1 Page 15 of 19 Rev. 31
AOP APPENDIX 1 APPENDIX 2 HEATUP RATE CURVE - RPV LEVEL AT 200" Vessel Water Heatup Rate (Water Level =200")
50.00 45.00 Heatup Rate (Deg. F/hr.)
40.00 35.00 30.00 25.00 20.00 15.00 10.00 5.00 0.00 1 7 13 19 25 31 37 43 49 55 61 67 73 79 85 91 97 T ime S ince S hutdown (D a ys)
CAUTION The initial heatup rate in the vessel may be higher than the calculated value when cooling is removed from service. The calculation used to generate the heatup rate curves assumes instantaneous mixing and heat transport from the fuel area to the remainder of the system volume. In addition, the calculated heatup rates reflect bulk temperatures not local temperatures. Under natural circulation conditions and the resulting time delay in heat transport, considerable differences in temperature may exist between the fuel area and upper levels of vessel. In some cases local boiling may occur but bulk boiling will not occur as long as cooling is restored within the calculated time-to-boil period.
AOP APPENDIX 1 Page 16 of 19 Rev. 31
AOP APPENDIX 1 APPENDIX 3 LOSS OF FUEL POOL COOLING HEATUP RATE CURVE Fuel Pool Isolated Water Heatup Rate (Full Core Offload and Previous Spent Fuel) 18.00 16.00 Heatup Rate (Deg. F/hr.)
14.00 12.00 10.00 8.00 6.00 4.00 2.00 0.00 1 7 13 19 25 31 37 43 49 55 61 67 73 79 85 91 97 T ime S ince S hutdown (D a ys)
CAUTION The initial heatup rate in the spent fuel pool may be higher than the calculated value when cooling is removed from service. The calculation used to generate the heatup rate curves assumes instantaneous mixing and heat transport from the fuel area to the remainder of the system volume. In addition, the calculated heatup rates reflect bulk temperatures not local temperatures. Under natural circulation conditions and the resulting time delay in heat transport, considerable differences in temperature may exist between the fuel area and measured temperatures in fuel pool cooling heat exchanger inlets.
AOP APPENDIX 1 Page 17 of 19 Rev. 31