ML14084A035
| ML14084A035 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/21/2014 |
| From: | NRC/RGN-II |
| To: | Duke Energy Corp |
| References | |
| 50-269/13-302, 50-270/13-302, 50-287/13-302 50-269/13-302, 50-270/13-302, 50-287/13-302 | |
| Download: ML14084A035 (152) | |
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FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION EPE007 EK1.06 - Reactor Trip 1 1 B Knowledge of the operational implications of the following concepts as they apply to the reactor trip: (CFR 41.8 / 41.10 / 45.3)
Relationship of emergency feedwater flow to S/G and decay heat removal following reactor trip ............................
Given the following plant conditions:
Unit 1 has just reached 100% power following a refueling outage Unit 2 is at 100% power with 93 EFPD Which ONE of the following will result in the highest amount of Emergency Feedwater flow required to stabilize RCS temperature 5 minutes following the trip?
A. Both Main Feedwater Pumps ONLY trip on Unit 1 B. Both Main Feedwater pumps ONLY trip on Unit 2 C. Loss of Offsite Power on Unit 1 D. Loss of Offsite Power on Unit 2 Friday, October 04, 2013 Page 1 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 1 1 B General Discussion Answer A Discussion Incorrect. FDW pump trip vs LOOP does required more EFDW since RCP's will still be operating. It is plausible to believe that the lower EFPD would result in more decay heat since there is more unused fuel early in core life vs late in core life and that could lead to a misunderstanding of decay heat loads following a trip.
Answer B Discussion Correct.
Design Basis Scenarios - FSAR
Requires the highest initial post trip EFDW flow of all analysis, therefore constitutes the design basis transient for post trip EFDW.
Answer C Discussion Incorrect. Plausible since this would be considered a much more serious event since there is a loss of much more equipment than just the Main Feedwater pumps however for the same EFPD the loss of the RCP's results in a lower EFDW flow requirement to stabilize RCS temps.
Answer D Discussion Incorrect. Plausible since this would be considered a much more serious event since there is a loss of much more equipment than just the Main Feedwater pumps however for the same EFPD the loss of the RCP's results in a lower EFDW flow requirement to stabilize RCS temps.
Basis for meeting the KA This question requires an understanding of the impact of length of core operation to amount of decay heat and the relationship of decay heat load to EFDW flow requirements. Since EFDW flow is to the SG's, the relationship of EFDW flow to S/G and DHR is established.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided CF-EF Obj R55 CF-EF Rule 7 EPE007 EK1.06 - Reactor Trip Knowledge of the operational implications of the following concepts as they apply to the reactor trip: (CFR 41.8 / 41.10 / 45.3)
Relationship of emergency feedwater flow to S/G and decay heat removal following reactor trip ............................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 2 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE008 AK1.01 - Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) 2 2 B Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: (CFR 41.8 / 41.10 /
45.3)
Thermodynamics and flow characteristics of open or leaking valves .......................................................
Given the following Unit 1 conditions:
Reactor power = 100%
1RC-66 (PORV) is leaking past its seat Pressurizer temperature = 648 0F Quench tank pressure = 5 psig Which ONE of the following describes the expected tailpipe temperature (°F) downstream of 1RC-66?
A. 162 B. 228 C. 272 D. 648 Friday, October 04, 2013 Page 3 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 2 2 B General Discussion Answer A Discussion Incorrect. Plausible because this will be the answer if 5 psig is not converted to psia.
Answer B Discussion CORRECT: The enthalpy for the steam leaving the pressurizer at 648 0F will be the same at 5 psig (20psia) - 1124 BTU/lb. This enthalpy at 20 psia constitutes a wet vapor with a temperature of 228 0F. Throttling processes are constant enthalpy processes and energy remains approximately the same on both sides of a throttling process.
Answer C Discussion Incorrect: Plausible if one thinks that the throttling process is a constant entropy process and looks for the same entropy as at 6480F - 1.27 BTU/R/lb Answer D Discussion Incorrect: Plausible with the same misconception made at TMI which was assuming constant temperature across the valve due to throttling process.
Basis for meeting the KA Requires knowledge of pzr vapor space accident (leaking PORV) on tailpipe temp by applying thermodynamic flow characteristics of a leaking valve Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ILT40 NRC Exam Q#2 Development References Student References Provided PNS-PZR R34 APE008 AK1.01 - Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)
Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: (CFR 41.8 / 41.10 /
45.3)
Thermodynamics and flow characteristics of open or leaking valves .......................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 4 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION EPE009 EK2.03 - Small Break LOCA 3 3 C Knowledge of the interrelations between the small break LOCA and the following: (CFR 41.7 / 45.7)
S/Gs ...........................................................
Given the following Unit 1 conditions:
Reactor tripped from 100% power due to SBLOCA 1A HPI Pump failed Subcooling Margin = 0°F stable Which ONE of the following is the reason the EOP directs increasing SG levels to the Loss of Subcooling Margin Setpoint level?
A. Establish a large secondary side inventory in support of a rapid RCS cooldown.
B. Establish a large secondary side inventory to ensure that a loss of coupling will NOT occur if a momentary loss of EFDW occurs.
C. Ensure a secondary water level higher than the primary water level inside the SG tubes to establish boiler condenser mode heat transfer D. Ensure a secondary side level sufficient to minimize the consequences of a total loss of feedwater during boiler condenser mode heat transfer Friday, October 04, 2013 Page 5 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 3 3 C General Discussion Answer A Discussion Incorrect: Plausible since SG heat transfer would assist in RCS cooldown and depressurization and SG heat transfer is credited for heat removal for certain break sizes and locations of SBLOCA's. The EOP does perform rapid RCS cooldown and depressurization only under other circumstances ( If HPI were further degraded). With the RCS saturated, the higher the SG level the more boiler condenser type heat transfer can occur as there would be more steam coming in contact with tubes that have secondary water on the other side. This means that it is plausible to deduce that I could perform a more rapid cooldown by increasing the SG levels.
Answer B Discussion Incorrect. Plausible since an increased inventory would help mitigate a momentary loss of EFDW during normal single phase natural circulation and once SG levels reach the LOSCM setpoint, momentary losses of EFDW flow would not stop heat transfer as long as secondary side water level is above primary side water level during boiler condenser heat transfer. Additionally, the EOP does increase SG levels to help mitigate a loss of feed availability to the SG's however that strategy is specific to a TB flood.
Answer C Discussion Correct. Establishing LOSCM setpoint ensures that the secondary water level is higher than the primary side water level inside the tubes thus allowing the steam in the primary side of the tubes to be condensed at locations where the secondary side water level exists thereby ensuring boiler condenser mode of heat transfer.
Answer D Discussion Incorrect. Plausible since increasing the secondary side to ensure heat transfer is not lost if FDW is lost is a mitigation strategy employed by the EOP however the strategy is used during a Turbine Building Flood in anticipation of the loss of feed pumps in the basement.
Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2009B Q26 Development References Student References Provided EAP-LOSCM Obj R6 EAP-LOSCM Att. 01 EPE009 EK2.03 - Small Break LOCA Knowledge of the interrelations between the small break LOCA and the following: (CFR 41.7 / 45.7)
S/Gs ...........................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 6 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE015/017 2.4.45 - Reactor Coolant Pump (RCP) Malfunctions 4 4 C APE015/017 GENERIC Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12)
Given the following Unit 1 conditions:
Reactor power = 50% stable 1B2 RCP is OFF Which ONE of the following would require immediate entry into AP/1/A/1700/016 (Abnormal Reactor Coolant Pump Operation)?
A. OAC point O1A0061 (RCP 1A1 MTR INPUT POWER) in HI alarm B. OAC point O1A1579 (RCP 1A2 MTR LOWER AIR TEMP) in HI alarm C. 1SA-15/A5 (RC PUMP MOTOR 1B1 OIL POT LOW LEVEL) in alarm D. 1SA-6/D5 (PUMP 1B2 CAVITY PRESS HI/LOW) in alarm Friday, October 04, 2013 Page 7 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 4 4 C General Discussion Answer A Discussion Incorrect. Plausible since there are many OAC alarms that will require entry into AP/16 and it is reasonable to believe that excessive input power could lead to issues related to motor and stator temps.
Answer B Discussion Incorrect. Plausible since there are many OAC alarms that will require entry into AP/16 and it is reasonable to believe that excessive air temperatures in the motor are indicative of problems with the pump and require it to be secured or actions taken to reduce air temps. AP/16 would be the most logical place for that guidance to be.
Answer C Discussion Correct. Any oil pot alarm is an entry condition for AP/16.
Answer D Discussion Incorrect. Plausible since this condition could exist even with the 1B2 RCP shutdown. It is plausible to believe that AP/16 could contain guidance to minimize loss of RCS water even with the pump secured. Also plausible because it would be correc if the pump were running.
Basis for meeting the KA Requires demonstrating the knowledge required to prioritize several alarms related to RCP malfunctions. "Prioritize and interpret" is being met in that each alarm must be interpreted to determine the impact on plant operation and only one of the alarms is the highest priority in that it is the only one requireing immediate entry into AP/16.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EAP-APG Obj R9 EAP-AP/16 APE015/017 2.4.45 - Reactor Coolant Pump (RCP) Malfunctions APE015/017 GENERIC Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 8 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION EPE011 EA1.14 - Large Break LOCA 5 5 B Ability to operate and monitor the following as they apply to a Large Break LOCA: (CFR 41.7 / 45.5 / 45.6)
Subcooling margin monitors .......................................
Given the following Unit 1 conditions:
Initial conditions:
Reactor power = 100%
Current conditions:
LPI Flow Train A = 1800 gpm stable LPI Flow Train B = 1780 gpm stable Rule 2 (Loss of SCM) in progress.
IMAs complete
- 2) In accordance with Rule 2, performance of Rule 3 (Loss of Main or Emergency FDW) __ (2) __ required.
Which ONE of the following completes the statements above?
A. 1. LOSCM
- 2. is B. 1. LOSCM
- 2. is NOT C. 1. ICC
- 2. is D. 1. ICC
- 2. is NOT Friday, October 04, 2013 Page 9 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 5 5 B General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because it would be correct if total LPI flow were less than 3400 gpm.
Answer B Discussion Correct. Transfer to LOSCM is required because any SCM is zero. Rule 3 is not performed because total LPI flow is greater than 3400 gpm.
Answer C Discussion Incorrect. First part is plausible because one SCM indication is super heated. However the Core must indicate super heat to require a transfer to the ICC tab. Second part is plausible because it would be correct if total LPI flow were less than 3400 gpm.
Answer D Discussion Incorrect. First part is plausible because one SCM indication is super heated. However the Core must indicate super heat to require a transfer to the ICC tab. Second part is correct.
Basis for meeting the KA At ONS there is no "operation" of the subcooling monitors. This question does require the ability to monitor the subcooled margins monitors following a LBLOCA. Since the determination of whether the entry conditions to the LOSCM and ICC tab are met are based on indications provided by the subcooleing margin monitors, determining which EOP tab entry conditions are met demonstrates the ability to monitor the subcooling margin monitors.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ILT41 Q3 Development References Student References Provided EOP-SA EOP Rule 2 EOP LOSCM EPE011 EA1.14 - Large Break LOCA Ability to operate and monitor the following as they apply to a Large Break LOCA: (CFR 41.7 / 45.5 / 45.6)
Subcooling margin monitors .......................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 10 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE025 AK1.01 - Loss of Residual Heat Removal System (RHRS) 6 6 A Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System: (CFR 41.8 /
41.10 / 45.3)
Loss of RHRS during all modes of operation .........................
Given the following Unit 1 conditions:
Initial conditions:
Normal LPI decay heat removal in service Current conditions:
Loss of offsite power occurs Power restored via CT-4 1A and 1B LPI Pumps NOT available Which ONE of the following describes the requirements to start the 1C LPI Pump?
Manual reset of Load Shed is __(1)___ and starting of 1C LPI Pump is allowed after a MINIMUM of ___(2)__ seconds.
A. 1. NOT required
- 2. 5 B. 1. required
- 2. 5 C. 1. NOT required
- 2. 30 D. 1. required
- 2. 30 Friday, October 04, 2013 Page 11 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 6 6 A General Discussion Answer A Discussion Correct: Pushing the Control Room MFB monitor RESET pushbuttons is not required because the signal for the 1C LPI Pump is removed 5 seconds after the Load Shed actuated.
Answer B Discussion Incorrect: First part is incorrect but plausible because load shed reset is required for many other components (seeEL-PSL). Second part is correct.
Answer C Discussion Incorrect: First part is correct. Second part is incorrect but plausible if confused with the Load Shed operation of X6 and X7 which automatically re-energize after 30 seconds.
Answer D Discussion Incorrect: First part is incorrect but plausible because reset is required for many other components. Second part is also incorrect but plausible if confused with the Load Shed operation of X6 and X7 which automatically re-energize after 30 seconds.
Basis for meeting the KA Requires knowledge of actions required to restore core decay heat removal following a failure of the LPI/DHR Pumps Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ILT39 Q6 Development References Student References Provided EL-PSL Obj R6 EL-PSL APE025 AK1.01 - Loss of Residual Heat Removal System (RHRS)
Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System: (CFR 41.8 /
41.10 / 45.3)
Loss of RHRS during all modes of operation .........................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 12 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE026 AK3.03 - Loss of Component Cooling Water (CCW) 7 7 D Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: (CFR 41.5,41.10 / 45.6 / 45.13)
Guidance actions contained in EOP for Loss of CCW ..................
Given the following Unit 1 conditions:
Reactor trip has just occurred Total RCP seal injection flow = 0 gpm Running Component Cooling pump tripped Standby CC pump did not start Which ONE of the following describes the procedure whose performance is directed by the EOP and why?
Initiate A. AP/20 (Loss of CC) to restore Component Cooling B. AP/20 (Loss of CC) to ensure letdown is isolated C. AP/25 (SSF EOP) to align an alternate letdown flowpath D. AP/25 (SSF EOP) to align an alternate source of seal injection Friday, October 04, 2013 Page 13 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 7 7 D General Discussion Answer A Discussion Incorrect: Plausible since the entry conditions for AP/20 are met and the EOP does direct entry into APs in other conditions (Ex. AP/11, AP/25). The EOP does not direct entry into AP/20 nor actions to restore CC. Seal injection flow is re-established via the RCMUP since both CC and SI have been lost.
Answer B Discussion Incorrect: Plausible since the entry conditions for AP/20 are met and AP/20 does ensure that letdown is isolated if letdown temp is >130 degrees which would normally be true if CC is lost. The EOP does not direct entry into AP/20 however the EOP does direct entry into APs in other conditions (Ex. AP/11, AP/25).
Answer C Discussion Incorrect: Plausible since the first part is correct in that AP/25 is directed by IMAs. Since the stem tells us that CC is unavailable, 1HP-5 would be closed on high letdown temperature. The fact that letdown has been isolated due to the loss of CC makes aligning the alternate letdown flowpath plausible. Additionally, an alternate letdown path is established while running the RCMUP however the RCMUP is started to provide an alternate source of SI and the alternate letdown path is a consequence of that.
Answer D Discussion CORRECT: If BOTH CC and HPI Seal injection are not available then RCP seal injection must be established from the SSF RCMUP via AP/25.
These directions are part of EOP Immediate Manual Actions performed by the RO.
Basis for meeting the KA Requires knowledge of reason EOP IMAs direct initiating AP/25 when RCP seal injection and CC have been lost Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2009A Q6 Development References Student References Provided EAP-IMA Obj R6 EAP-IMA APE026 AK3.03 - Loss of Component Cooling Water (CCW)
Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: (CFR 41.5,41.10 / 45.6 / 45.13)
Guidance actions contained in EOP for Loss of CCW ..................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 14 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE027 AK2.03 - Pressurizer Pressure Control System (PZR PCS) Malfunction 8 8 B Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: (CFR 41.7 / 45.7)
Controllers and positioners ........................................
Given the following Unit 1 conditions:
Initial conditions:
Reactor power = 90%
1B Main Feedwater pump trips Current conditions:
Reactor power = 70% decreasing RCS pressure = 2165 psig slowly decreasing Pressurizer level = 228 inches slowly decreasing Pressurizer temperature = 640°F slowly decreasing Pressurizer heater bank 1 (Group A and K) is ON Pressurizer heater banks 2, 3, and 4 are in AUTO and are OFF The pressurizer is ___(1)____ AND the pressurizer saturation circuit ___(2)____.
Which ONE of the following completes the statement above?
A. 1. subcooled
- 2. is responding as expected B. 1. subcooled
- 2. has failed C. 1. saturated
- 2. is responding as expected D. 1. saturated
- 2. has failed Friday, October 04, 2013 Page 15 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 8 8 B General Discussion Answer A Discussion Incorrect: First part is correct. Second part is plausible since parameters given are reasonable for the post runback condition. Normal pressurizer spray valve RC-1 would open at 2205 psig and not closed until pressure reaches 2155 psig. The decreasing RCS pressure could be explained by the decreasing Pzr level as it returns to setpoint after FDWP trip.
Answer B Discussion CORRECT: Saturation temp for 2165 psig is approximately 648 degrees. With the Pzr at 640 degrees it is clearly subcooled. Regarding the pressurizer level saturation circuitry, Psat must be 20 psig below actual RCS pressure before Bank 2 will energize and will not de-energize until Psat and RCS pressure (NR Med-selected RCS Pressure) are within 15 psig (5 psig dead band). With RCS pressure at 2165, pressurizer temp should be about 648F (saturation for 2165). Saturation for actual pzr temp of 640F is about 2045 psig therefore Bank 2 should be energized.
2205 psig.
Answer C Discussion Incorrect: First part is plausible since conditions in Pzr are consistent with the loss of FDWP runback. Decreasing RCS pressure is occurring concurrently with decreasing Pzr level which is a normal response if the Pzr is saturated. Second part is plausible since parameters given are reasonable for the post runback condition. Normal pressurizer spray valve RC-1 would open at 2205 psig and not closed until pressure reaches 2155 psig. The decreasing RCS pressure could be explained by the decreasing Pzr level as it returns to setpoint after FDWP trip.
Answer D Discussion Incorrect: First part is plausible since conditions in Pzr are consistent with the loss of FDWP runback. Decreasing RCS pressure is occurring concurrently with decreasing Pzr level which is a normal response if the Pzr is saturated. Second part is correct. Second part is also plausible if you believe the Pzr to be saturated based on a misconception regarding which Pzr heaters are part of the saturation circuit.
Basis for meeting the KA Requires knowledge of how controllers for Pzr saturation circuit function and the ability to diagnose a malfunction of circuitry.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2009A Q7 Development References Student References Provided PNS-PZR Obj R5, R7, R29 PNS-PZR APE027 AK2.03 - Pressurizer Pressure Control System (PZR PCS) Malfunction Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: (CFR 41.7 / 45.7)
Controllers and positioners ........................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 16 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION BWE04 EK3.1 - Inadequate Heat Transfer 9 9 A Knowledge of the reasons for the following responses as they apply to the (Inadequate Heat Transfer)
(CFR: 41.5 / 41.10, 45.6, 45.13)
Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.
Given the following Unit 2 conditions:
Loss of all sources of Feedwater has occurred RCS Pressure = 2250 psig increasing Pressurizer level = 294 inches increasing ALL SCMs = 24°F slowly decreasing What is the:
- 1) lowest RCS pressure (psig) that will require Rule 4 (Initiation of HPI Forced Cooling) to be performed?
- 2) PRIMARY reason for reducing the number of operating RCPs in accordance with Rule 4?
A. 1. 2300
- 2. Reduce the heat input to the RCS B. 1. 2300
- 2. Provide the ability to recover from HPI forced cooling and re-establish a Pressurizer bubble.
C. 1. 2450
- 2. Reduce the heat input to the RCS D. 1. 2450
- 2. Provide the ability to recover from HPI forced cooling and re-establish a Pressurizer bubble.
Friday, October 04, 2013 Page 17 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION General Discussion 9 9 A Answer A Discussion Correct.
Even with SCM > 0, IAAT NO SGs can be fed with FDW (Main/CBP/Emergency),
AND any of the following exists:
RCS pressure reaches 2300 psig OR NDT limit Pzr level reaches 375 [340 acc]
HPI FC should be initiated per Rule 4 The number of operating RCPs should be reduced to one to decrease the heat being added to the RCS Answer B Discussion Incorrect. First part is correct. Second part is plausible based on a NOTE in Rule 4 and would be correct if asked why the 1A1 RCP (vs. one of the other 3) is the preferred pump to leave in operation.
Answer C Discussion Incorrect. First part is plausible since it is the setpoint for the PORV and the PORV is opened during HPI forced cooling therefore it would be plausible to believe that the PORV opening setpoint would be the threshold for initiating HPI FC. Second part is correct.
Answer D Discussion Incorrect. First part is plausible since it is the setpoint for the PORV and the PORV is opened during HPI forced cooling therefore it would be plausible to believe that the PORV opening setpoint would be the threshold for initiating HPI FC. Second part is plausible based on a NOTE in Rule 4 and would be correct is asked why the 1A1 RCP (vs. one of the other 3) is the preferred pump to leave in operation.
Basis for meeting the KA Requires knowledge of the reason for operating limitations (number of running RCP's) that are a function of operating characteristics (HPI FC initiated based on increasing RCS pressure during a LOHT) during an Inadequate Heat Transfer condition.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-LOHT Att 4 Obj R3, EAP-LOHT Obj 2 EAP LOHT Att 4, EAP-LOHT BWE04 EK3.1 - Inadequate Heat Transfer Knowledge of the reasons for the following responses as they apply to the (Inadequate Heat Transfer)
(CFR: 41.5 / 41.10, 45.6, 45.13)
Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 18 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION EPE038 2.2.38 - Steam Generator Tube Rupture (SGTR) 10 10 A EPE038 GENERIC Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)
Given the following Unit 3 conditions:
Reactor power = 100%
Which ONE of the following will result in a Tech Spec LCO being NOT met?
A. 3A SGTL rate = 160 gpd B. 3B Core Flood Tank level = 12.69 feet C. 3B Core Flood Tank pressure = 622 psig D. 4 gpm RCS leak identified as being through valve stem packing of 3HP-1 Friday, October 04, 2013 Page 19 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 10 10 A General Discussion Answer A Discussion Correct. The TS 3.4.13 LCO limit on SG tube leakage is 150 gpd through any one SG.
Answer B Discussion Incorrect. Plausible since this is below the low level alarm setpoint of 12.7 feet. Still above the TS required level of 12.66 feet.
Answer C Discussion Incorrect. Plausible since this pressure is above the high pressure alarm setpoint of 615 psig, Still below the max TS pressure of 625 psig.
Answer D Discussion Incorrect. The TS 3.4.13 LCO limit on identified leakage is 10 gpm. Plausible based on confusing the unidentified and identified leakage limits.
Basis for meeting the KA Required knowledge of limitations on various forms of RCS Operational Leakage established in Tech Spec 3.4.13.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Admin-ITS Obj R8 TS 3.4.13 PNS-CF EPE038 2.2.38 - Steam Generator Tube Rupture (SGTR)
EPE038 GENERIC Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 20 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE054 AA1.04 - Loss of Main Feedwater (MFW) 11 11 A Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW):(CFR 41.7 / 45.5 / 45.6)
HPI, under total feedwater loss conditions ...........................
Given the following Unit 1 conditions:
ALL sources of feedwater have been lost Rule 4 (Initiation of HPI Forced Cooling) is complete with outstanding IAATs 1A HPI pump has failed HPI flow parameters are as indicated below In accordance with Rule 4, __(1)__ RCP(s) is/are operating and HPI flow __(2)__
required to be throttled.
Which ONE of the following completes the statement above?
A. 1. 1
- 2. is B. 1. 1
- 2. is NOT C. 1. 2
- 2. is D. 1. 2
- 2. is NOT Friday, October 04, 2013 Page 21 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 11 11 A General Discussion Answer A Discussion Correct. Rule 4 directs securing all but one RCP. HPI flow limits with 1 HPI pump/hdr is 475 gpm including seal injection for the A header.
Answer B Discussion Incorrect. First part is correct. Second part is plausible since it would be correct if the A HPI pump had not failed or it would be correct if you do not include seal injection flow for the A HPI header.
Answer C Discussion Incorrect. First part is plausible since reducing RCP's to one pump per loop is the guidance provided in the LOHT tab of the EOP and therefore would be correct for a LOHT if conditions had not degraded to the point where rule 4 had been implemented (RCS pressure reaching 2300 psig and no feed available to SG). Second part is correct.
Answer D Discussion Incorrect. First part is plausible since reducing RCP's to one pump per loop is the guidance provided in the LOHT tab of the EOP and therefore would be correct for a LOHT if conditions had not degraded to the point where rule 4 had been implemented (RCS pressure reaching 2300 psig and no feed available to SG). Second part is plausible since it would be correct if the A HPI pump had not failed or it would be correct if you do not include seal injection flow for the A HPI header.
Basis for meeting the KA Requires demonstrating the ability to monitor HPI flow parameters under conditions where all FDW has occurred.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-LOHT Obj R28 EAP-LOHT Att. 4 APE054 AA1.04 - Loss of Main Feedwater (MFW)
Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW):(CFR 41.7 / 45.5 / 45.6)
HPI, under total feedwater loss conditions ...........................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 22 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION EPE055 EA2.03 - Loss of Offsite and Onsite Power (Station Blackout) 12 12 C Ability to determine or interpret the following as they apply to a Station Blackout : (CFR 43.5 / 45.13)
Actions necessary to restore power .................................
Given the following Unit 1 Conditions:
Initial conditions:
Reactor Power = 100%
ACB-4 closed Current conditions:
Reactor trip CT-1 Locks out KHU-2 Emergency Lockout occurs Which ONE of the following describes how power will be restored to Unit 1 MFBs?
A. Automatically through ACB-3 B. Automatically through SL1 and SL2 C. Manually through ACB-3 D. Manually through SL1 and SL2 Friday, October 04, 2013 Page 23 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 12 12 C General Discussion Answer A Discussion Incorrect. Plausible since Zone Overlap protection will automatically close ACB-3 under certain conditions.
Answer B Discussion Incorrect. Plausible since this would be correct if the SBB's were already energized from CT-5.
Answer C Discussion Correct. With ACB-4 open due to the KHU-2 lockout, EOP Encl. 5.38 will direct the operator to close ACB-3 to restore power to the MFB from KHU-1.
Answer D Discussion Incorrect. Plausible since this would be a path used in Encl 5.38 to restore power if Closing ACB-3 did not result in restoration of power.
Basis for meeting the KA Requires determining the actions directed by Encl. 5.38 to restore power to MFB's following a blackout.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-BO Obj R6 EAP-BO EAP-BO Att 1 EPE055 EA2.03 - Loss of Offsite and Onsite Power (Station Blackout)
Ability to determine or interpret the following as they apply to a Station Blackout : (CFR 43.5 / 45.13)
Actions necessary to restore power .................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 24 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE056 AA2.43 - Loss of Offsite Power 13 13 D Ability to determine and interpret the following as they apply to the Loss of Offsite Power: (CFR: 43.5 / 45.13)
Occurrence of a turbine trip .......................................
Given the following Unit 1 conditions:
Unit shutdown in progress Reactor power = 38% slowly decreasing LOOP (Switchyard Isolation) occurs Which ONE of the following:
- 1) describes the status of the Main Turbine 5 minutes following the LOOP?
- 2) is used by ICS to control the Turbine Bypass Valves anytime the Main Turbine is tripped?
A. 1. tripped
- 2. Turbine Header Pressure B. 1. tripped
- 2. Steam Generator Outlet Pressure C. 1. NOT tripped
- 2. Turbine Header Pressure D. 1. NOT tripped
- 2. Steam Generator Outlet Pressure Friday, October 04, 2013 Page 25 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 13 13 D General Discussion Answer A Discussion Incorrect. First part is plausible since it is always true if Rx power is > 70%. Second part is plausible since it is the controlling signal when the Main Turbine ICS station is in Auto which is its normal position once the Main Turbine is brought online and loaded.
Answer B Discussion Incorrect. Incorrect. First part is plausible since it is always true if Rx power is > 70%. Second part is correct.
Answer C Discussion Incorrect. First part is correct. Second part is plausible since it is the controlling signal when the Main Turbine ICS station is in Auto which is its normal position once the Main Turbine is brought online and loaded.
Answer D Discussion Correct. For power levels below 40% power, a LOOP does not result in a Rx trip unless auxiliaries are being powered from the CT transformer.
During a unit shutdown this does not occur until about 25% power therefore the LOOP would not result in a Rx trip. During normal ops with the Main Turbine on-line, the ICS turbine master is maintained in Auto and Turbine Header Pressure is the controlling signal for the TBV's. If the Turbine station is in Manual (which occurs on a Turbine Trip) then the controlling signal is swapped to Steam Generator Outlet Pressure.
Basis for meeting the KA Question required the ability to determine if the turbine has tripped following a loss of offsite power (switchyard isolation).
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided IC-RPS Obj. R3, 4, 23, SAE-L216 Obj R3 IC-RPS SAE-L216 APE056 AA2.43 - Loss of Offsite Power Ability to determine and interpret the following as they apply to the Loss of Offsite Power: (CFR: 43.5 / 45.13)
Occurrence of a turbine trip .......................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 26 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE057 AA1.06 - Loss of Vital AC Electrical Instrument Bus 14 14 C Ability to operate and / or monitor the following as they apply to the Loss of Vital AC Instrument Bus: (CFR 41.7 / 45.5 / 45.6)
Manual control of components for which automatic control is lost .......
Given the following Unit 1 conditions:
Initial conditions Reactor Power = 100%
SASS in Manual while SPOC repairs Pressurizer Level 3 level transmitter 1HP-120 in AUTO selected to Pressurizer Level 1 Current conditions:
Vital Power to ICCM Train A fails Which ONE of the following describes Pressurizer level control with 1HP-120?
A. Selecting Pressurizer Level 2 and depressing the AUTO pushbutton on 1HP-120 are required to restore automatic control at setpoint B. Selecting Pressurizer Level 2 ONLY will restore automatic control at setpoint C. Manual control using 1HP-120 Bailey controller is all that is available D. Additional actions are NOT required since Automatic control at setoint is retained Friday, October 04, 2013 Page 27 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 14 14 C General Discussion Answer A Discussion Incorrect. ICCM Train A feeds both Pzr level 1 & 2. ICCM Train B feeds Pzr level 3. It is plausible to believe that since ICCM Train A feeds Pzr level 1 then ICCM Train B feeds Pzr level 2. Under this misconception it is plausible to believe that 1HP-120 would trip to Hand when power is lost to Pzr level 1 since there are multiple bailey control stations that trip to hand under various conditions.
Answer B Discussion Incorrect. ICCM Train A feeds both Pzr level 1 & 2. ICCM Train B feeds Pzr level 3. It is plausible to believe that since ICCM Train A feeds Pzr level 1 then ICCM Train B feeds Pzr level 2 which would lead choosing this as the correct answer.
Answer C Discussion Correct. ICCM Train A feeds both Pzr level 1 & 2. ICCM Train B feeds Pzr level 3. With Pzr level 3 unavailable, if ICCM Train A fails, all auto control is lost therefore only using 1HP-120 in hand would be correct.
Answer D Discussion Incorrect. Plausible since this would be correct if SASS were in Auto.
Basis for meeting the KA Requires the ability to both manually operate and monitor manual control of 1fHP-120 following a loss of Vital Power to ICCM Train A.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided PNS-PZR Obj R31, R35 PNS-PZR APE057 AA1.06 - Loss of Vital AC Electrical Instrument Bus Ability to operate and / or monitor the following as they apply to the Loss of Vital AC Instrument Bus: (CFR 41.7 / 45.5 / 45.6)
Manual control of components for which automatic control is lost .......
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 28 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE065 AA2.05 - Loss of Instrument Air 15 15 B Ability to determine and interpret the following as they apply to the Loss of Instrument Air: (CFR: 43.5 / 45.13)
When to commence plant shutdown if instrument air pressure is decreasing Given the following Unit 1 conditions:
Initial conditions:
Reactor power = 100%
Instrument Air pressure = 85 psig decreasing AP/22 (Loss of Instrument Air) has been initiated Which ONE of the following is the higher Instrument Air pressure (psig) that would require an immediate manual Reactor trip in accordance with AP/22?
A. 70 B. 65 C. 40 D. 30 Friday, October 04, 2013 Page 29 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 15 15 B General Discussion Answer A Discussion Incorrect. Plausible since there are automatic actions that happen at 70 psig IA pressure which are detailed in AP/22.
Answer B Discussion Correct. AP/22 informs the operator that FDW control valves fail "As Is" at 65 psig and there is an IAAT step directing a manual trip of the Main FDW pumps and Rx if FDW flow becomes uncontrollable. With a runback in progress, FDW flow would be uncontrollable as soon as FDW valves fail "as is".
Answer C Discussion Incorrect. Plausible since the RCW pressure switch on compressor unit will prevent compressor operation if no RCW is supplied to cooling system or if pressure drops below 40 psig. Also, Indication will drop to 35-40 psig if the air receiver/oil sump check valve is leaking adding additional plausibility to 40 psig.
Answer D Discussion Incorrect. Plausible since this IA pressure is a threshold pressure discussed in a NOTE in AP/22 however this is the pressure that SF level indications become inaccurate. Additionally, 30 psig is the pressure at which most pneumatic valves reach fully closed and therefore they lose all ability to control flows and pressures.
Basis for meeting the KA Requires knowledge of when a reactor trip is required based on decreasing IA pressure.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EAP-APG Obj R9 EAP-APG SAE-L035 SSS-IA AP/22 APE065 AA2.05 - Loss of Instrument Air Ability to determine and interpret the following as they apply to the Loss of Instrument Air: (CFR: 43.5 / 45.13)
When to commence plant shutdown if instrument air pressure is decreasing 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 30 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE077 AK2.03 - Generator Voltage and Electric Grid Disturbances 16 16 C Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: (CFR: 41.4, 41.5, 41.7, 41.10 /
45.8)
Sensors, detectors, indicators......................................................
Given the following Unit 1 conditions:
Initial conditions:
AP/34 (Degraded Grid) in progress Generator output = 850 MWe and 450 MVARs Generator Hydrogen Pressure = 60 psig Generator Output Voltage = 18.2 KV
- 1) The Generator output __ (1) __ within the limits of the Generator Capability Curve.
- 2) If the generator exceeds the Underfrequency Maximum Allowable Time given in AP/34 (Degraded Grid) the Main Turbine __ (2) __.
Which ONE of the following completes the statements above?
REFERENCE PROVIDED A. 1. is NOT
- 2. will automatically trip B. 1. is NOT
- 2. requires a manual trip C. 1. is
- 2. will automatically trip D. 1. is
- 2. requires a manual trip Friday, October 04, 2013 Page 31 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 16 16 C General Discussion Answer A Discussion Incorrect. First part is plausible since it would be correct if power factor were leading or if Gen H2 pressure were lower. Second part is correct.
Answer B Discussion Incorrect. First part is plausible since it would be correct if power factor were leading or if Gen H2 pressure were lower. Second part plausible because the AP does have the operator monitor how long a low frequency conditions lasts and trip the unit if it does not.
Answer C Discussion Correct. Since MVARS are positive, power factor is lagging and using the upper portion of the Gen Capacity Curve, this value is acceptable. The Digital T/G control system monitors how long the unit operates in a low frequency condition and will trip the unit if the time limit is exceeded.
Answer D Discussion Incorrect. First part is correct. Second part plausible because the AP does have the operator monitor how long a low frequency conditions lasts and trip the unit if it does not.
Basis for meeting the KA Requires the ability to use the Generator Capacity Curve that is applicable during degraded grid conditions and determine if Genertor output is accetpable during a grid disturbance. Also required the ability to utiilize frequency indicators and predict plant response based on those indications.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ILT41 Q16 Development References Student References Provided CP05 Obj 5, EAP-APG Obj R9 AP/34 Gen Capacity Curve AP/34 lesson and AP CP05 APE077 AK2.03 - Generator Voltage and Electric Grid Disturbances Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: (CFR: 41.4, 41.5, 41.7, 41.10 /
45.8)
Sensors, detectors, indicators......................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 32 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION BWE02 2.2.3 - Vital System Status Verification 17 17 D BWE02 GENERIC (multi-unit license) Knowledge of the design, procedural, and operational differences between units. (CFR: 41.5 / 41.6 / 41.7 / 41.10 / 45.12)
Given the following Unit 3 conditions:
A brief loss of power has occurred Unit auxiliaries are being supplied from the switchyard via CT-3 Subsequent Actions tab in progress
- 1) Subsequent Actions directs restarting __(1)__.
- 2) The __(2)__ RCP will provide the best Pressurizer Spray.
Which ONE of the following completes the statements above?
A. 1. one RCP ONLY
- 2. 3A1 B. 1. one RCP ONLY
- 2. 3B1 C. 1. one RCP per loop
- 2. 3A1 D. 1. one RCP per loop
- 2. 3B1 Friday, October 04, 2013 Page 33 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 17 17 D General Discussion Answer A Discussion Incorrect. First part is plausible since there are times when the EOP directs to have only 1 RCP operating. Second part is plausible since the A1 RCP is the spray pump on Unit 1.
Answer B Discussion Incorrect. First part is plausible since there are times when the EOP directs to have only 1 RCP operating. Second part is correct.
Answer C Discussion Incorrect. First part is correct. Second part is plausible since the A1 RCP is the spray pump on Unit 1.
Answer D Discussion Correct. Subsequent actions directs starting 1RCP/loop if available. There is a NOTE that informs the operator that the 3B1 RCP will provide the best Pzr spray flow.
Basis for meeting the KA Requires knowledge of procedural and design differences between Unit 1 and Unit 3 relative to guidance provided in the Subsequent Actions tab (Vital Systems Status Verification) and design differences. On Unit 1 the A1 RCP provides the best spray flow and on Unit 3 it is the B1 pump that provides the best spray flow. This is due to which cold leg the Pzr spray line taps off of.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EAP-SA Obj R39 EOP-SA Unit 3 SA BWE02 2.2.3 - Vital System Status Verification BWE02 GENERIC (multi-unit license) Knowledge of the design, procedural, and operational differences between units. (CFR: 41.5 / 41.6 / 41.7 / 41.10 / 45.12) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 34 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION BWE05 EK3.2 - Excessive Heat Transfer 18 18 D Knowledge of the reasons for the following responses as they apply to the (Excessive Heat Transfer)
(CFR: 41.5 / 41.10, 45.6, 45.13)
Normal, abnormal and emergency operating procedures associated with (Excessive Heat Transfer).
Which ONE of the following instruments should be used when initially stabilizing RCS temperature following a Main Steam Line Break and why?
A. Tcold is used since Tech Specs specifies that Tcold is RCS temperature B. Tcold is used since it is the coldest temperature and therefore most indicative of PTS issues C. CETCs are used since the resultant RCS cooldown may result in Tcold being off scale low D. CETCs are used since they are qualified instruments and are therefore more reliable in the hostile containment environment Friday, October 04, 2013 Page 35 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 18 18 D General Discussion Answer A Discussion Incorrect. Plausible since Tech Spec does specify that Tcold is RCS temperature however the EOP gives specific direction to use CETC's when stabilizing the RCS following a MSLB. Tcold would be correct for other events.
Answer B Discussion Incorrect. Plausible since Tcold would be the colder temperature and PTS due to decreasing RCS temperature during a MSLB are a concern.
Answer C Discussion Incorrect. CETC's are correct. The reason is plausible since the statement is a true statement. Tcold uses narrow range Temperature instruments and go off scale low at 520 degrees. Cooldown below 520 is well within the scope of a MSLB making this choice plausible.
Answer D Discussion Correct. CETC's are environmentally qualified instruments where Tc's are not. This ensures valid temperature instruments are being used to stabilize RCS following a MSLB inside the RB.
Basis for meeting the KA Requires knowledge of the reason for procedural guidance to use CETC's to stabilize RCS temperatures following a MSLB.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EAP-EHT Obj R8 EAP-EHT BWE05 EK3.2 - Excessive Heat Transfer Knowledge of the reasons for the following responses as they apply to the (Excessive Heat Transfer)
(CFR: 41.5 / 41.10, 45.6, 45.13)
Normal, abnormal and emergency operating procedures associated with (Excessive Heat Transfer).
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 36 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE076 2.3.5 - High Reactor Coolant Activity 19 19 B APE076 GENERIC Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 / 41.12 / 43.4 / 45.9)
Given the following Unit 1 conditions:
Initial conditions:
Time = 1200 Reactor power = 100%
1A steam generator tube leak = 2.1 gpd stable RCS activity = 0.25 Ci/ml DEI increasing Current conditions:
Time = 1400 NO change in 1A SG tube leak rate RCS activity = 0.65 Ci/ml DEI increasing Which ONE of the following describes the response of the radiation monitors between 1200 and 1400?
A. 1RIA-59 (N-16 monitor) and 1RIA-40 (CSAE Off-gas) increased.
B. 1RIA-16 (Main Steam Line Monitor) and 1RIA-40 increased.
C. 1RIA-59 increased while1RIA-40 remained constant.
D. 1RIA-16 increased while 1RIA-40 remained constant.
Friday, October 04, 2013 Page 37 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 19 19 B General Discussion Answer A Discussion Incorrect. RIA-40 will be affected by the fuel failure, whereas RIA 59 (N-16 detectors) will not. Plausible since RIA-59 & 60 are Main Steam Line monitors and activity that leaks to the secondary side will pass by the RIAs on the way to the Main Turbine however since they are N16 monitiors, the increase in activity will not impact their readings.
Answer B Discussion Correct: RIA-16 and 40 will respond to ALL activity, therefore an increase in RCS activity, which the stem provides with a degrading fuel failure, would cause both to increase.
Answer C Discussion Incorrect. RIA-40 will be affected by the fuel failure, whereas RIA 59 (N-16 detectors) will not. Plausible since RIA-40 is reading Air Ejector off gas flow and not directly monitoring the RCS. Since it is not directly monitoring the RCS water this would be a correct choice for increasing RCS activity without the presence of a SGTL and is therefore plausible as a choice.
Answer D Discussion Incorrect. RIA-16 is correct however RIA-40 will be affected by the fuel failure as described in A. Plausible since RIA-40 is reading Air Ejector off gas flow and not directly monitoring the RCS. Since it is not directly monitoring the RCS water this would be a correct choice for increasing RCS activity without the presence of a SGTL and is therefore plausible as a choice.
Basis for meeting the KA Demonstrates the ability to use radiation monitors during high activity in the RCS by being able to predict the proper response based on changes in RCS activity.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2009B Q24 Development References Student References Provided RAD-RIA Obj R2 RAD-RIA APE076 2.3.5 - High Reactor Coolant Activity APE076 GENERIC Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 / 41.12 / 43.4 / 45.9) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 38 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE032 AK3.01 - Loss of Source Range Nuclear Instrumentation 20 20 C Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR 41.5,41.10 /
45.6 / 45.13)
Startup termination on source-range loss ............................
Given the following Unit 2 condition:
Initial conditions:
Time = 0900 Reactor Startup in progress NI 1 & 2 = 370 cps NI 3 & 4 = 0 cps (out of service)
ALL WR NIs = ~ 2.7 E-4%
Current conditions:
Time = 0901 NI 1 & 2 are inoperable Which ONE of the following describes:
- 1) immediate actions required by Tech Spec 3.3.9 (Source Range Neutron Flux)?
- 2) the reason for the actions described above?
A. 1. Insert Control Rods to Group 1 at 50% withdrawn
- 2. Prevents power increases when the primary power indication available to the operator is not available.
B. 1. Insert Control Rods to Group 1 at 50% withdrawn
- 2. 2 dpm Startup Rate Control Rod Out Inhibit is no longer available C. 1. Fully insert all Control Rods
- 2. Prevents power increases when the primary power indication available to the operator is not available.
D. 1. Fully insert all Control Rods
- 2. 2 dpm Startup Rate Control Rod Out Inhibit is no longer available Friday, October 04, 2013 Page 39 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 20 20 C General Discussion Answer A Discussion Incorrect. First part is plausible since there are procedural requirements in the startup procedure that will direct the operator to insert rods to Group 1 to 50% when the startup is delayed. Second part is correct.
Answer B Discussion Incorrect. First part is plausible since there are procedural requirements in the startup procedure that will direct the operator to insert rods to Group 1 to 50% when the startup is delayed. Second part is plausible since there is a 2 dpm startup rate Control Rod Out Inhibit that is relied on during startups to prevent excessive startup rates primarily to prevent entering the POAH at too high a rate. This inhibit is provided by Wide Range NI's and is therefore still available.
Answer C Discussion Correct. TS 3.3.9 directs (among other things) to immediately insert all control rods and the bases explains that it is because the Source Range is the primary indication of reactor power in this condition and it has been lost.
Answer D Discussion Incorrect. First part is correct. Second part is plausible since there is a 2 dpm startup rate Control Rod Out Inhibit that is relied on during startups to prevent excessive startup rates primarily to prevent entering the POAH at too high a rate. This inhibit is provided by Wide Range NI's and is therefore still available.
Basis for meeting the KA Requires knowledge of the reason TS 3.3.9 directs inserting all control rods and therefore terminates the startup when source range is lost.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory MODIFIED 2007 Q20 Development References Student References Provided IC-CRI Obj R32, ADM-TSS Obj R4 TS 3.3.9 IC-CRI APE032 AK3.01 - Loss of Source Range Nuclear Instrumentation Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR 41.5,41.10 /
45.6 / 45.13)
Startup termination on source-range loss ............................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 40 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE037 AA1.11 - Steam Generator (S/G) Tube Leak 21 21 D Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: (CFR 41.7 / 45.5 / 45.6)
PZR level indicator ..............................................
Given the following Unit 1 conditions:
Reactor power = 92% decreasing Unit shutdown in progress per the SGTR tab
- 1) In accordance with the SGTR tab and Enclosure 5.5 (Pzr and LDST Level Control),
RCS makeup and letdown will be adjusted to maintain Pressurizer level between
__ (1) __ inches.
- 2) The reason for this Pzr level band is to provide adequate inventory to __ (2) __.
Which ONE of the following completes the statements above?
A. 1. 140 - 180
- 2. ensure Pzr heaters will remain covered if a subsequent reactor trip occurs B. 1. 140 - 180
- 2. accommodate system shrinkage during shutdown/cooldown from 18%
power C. 1. 220 - 260
- 2. ensure Pzr heaters will remain covered if a subsequent reactor trip occurs D. 1. 220 - 260
- 2. accommodate system shrinkage during shutdown/cooldown from 18%
power Friday, October 04, 2013 Page 41 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 21 21 D General Discussion Answer A Discussion Incorrect. First part is plausible because this is the correct level if the reactor is tripped.. Second part is plausible because Pzr level is maintained greater than 100 inches post trip in part to ensure Pzr heater are still available.
Answer B Discussion Incorrect. First part is plausible because this is the correct level if the reactor is tripped.. Second part is correct..
Answer C Discussion Incorrect. First part is correct. Second part is plausible because Pzr level is maintained greater than 100 inches post trip in part to ensure Pzr heater are still available.
Answer D Discussion Correct. Since the reactor has not been tripped, Pzr level is maintained 220 - 260 inches early in the shutdown to provide sufficient inventory to accommodate for the system shrinkage that will occur during the later stages of the shutdown/cooldown from 18% power.
Basis for meeting the KA Requires the ability to monitor and maintain the correct Pzr level during a SGTR shutdown.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ILT41 Q21 Development References Student References Provided EAP-SGTR Obj R4 SGTR tab EAP-SGTR APE037 AA1.11 - Steam Generator (S/G) Tube Leak Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: (CFR 41.7 / 45.5 / 45.6)
PZR level indicator ..............................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 42 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION BWA04 AA1.2 - Turbine Trip 22 22 A Ability to operate and / or monitor the following as they apply to the (Turbine Trip)
(CFR: 41.7 / 45.5 / 45.6)
Operating behavior characteristics of the facility.
Given the following Unit 1 conditions:
Reactor power = 100%
Which ONE of the following will result in an AUTOMATIC trip of the Main Turbine?
A. Bearing Oil Pressure = 5.5 psig B. Main Turbine speed = 1955 RPM C. Loss of both Active Turbine Speed signals D. EITHER Steam Generator Level = 93% OR Friday, October 04, 2013 Page 43 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 22 22 A General Discussion Answer A Discussion Correct. The Main Turbine bearing oil pressure trip is at 8 psig. Plausible as incorrect since the FDWP low bearing oil pressure trip setpoint is 4 psig.
Answer B Discussion Incorrect. Plausible since rated Turbine speed is 1800 rpm and 1955 rpm is significantly greater than rated speed.
Answer C Discussion Incorrect: Plausible since there are only two Active speed signals and there are automatic actions that occur on loss of both active speed signals however it takes a loss of all speed signals (2 active and 1 passive) to result in a Main Turbine trip on loss of speed signals.
Answer D Discussion Incorrect: Plausible since this level is above the high level limit setpoint of 86% OR Basis for meeting the KA Incorrect. Plausible since 93% is above the high level limit setpoint of Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided STG-EHC Obj R24 STG-EHC STG-ICS BWA04 AA1.2 - Turbine Trip Ability to operate and / or monitor the following as they apply to the (Turbine Trip)
(CFR: 41.7 / 45.5 / 45.6)
Operating behavior characteristics of the facility.
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 44 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION APE060 2.2.36 - Accidental Gaseous-Waste Release 23 23 B APE060 GENERIC Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR:
41.10 / 43.2 / 45.13)
Given the following Unit 1 conditions:
Initial conditions:
Reactor power = 100%
1A GWD tank release in progress 1RIA-38 OOS Current conditions:
Maintenance activities in the area result in an inadvertent loss of power to RM-80 skid of 1RIA-37 1SA8/B9 RM PROCESS MONITOR RADIATION HIGH in alarm 1SA8/B10 RM PROCESS MONITOR FAULT in alarm
- 1) 1GWD-4 (A GWD TANK DISCHARGE) will __(1)__.
- 2) The required Completion Time in SLC 16.11.3 (Radioactive Effluent Monitoring Instrumentation) for securing this release pathway if both 1RIA-37 and 1RIA-38 become inoperable is __(2)__.
Which ONE of the following completes the statements above?
A. 1. remain open
- 2. immediately B. 1. automatically close
- 2. immediately C. 1. remain open
- 2. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D. 1. automatically close
- 2. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Friday, October 04, 2013 Page 45 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 23 23 B General Discussion Answer A Discussion Incorrect, 1GWD-4 will close. Remaining open is plausible because the HIGH setpoint was not actually reached since the alarms were due to loss of power. Additionally, it is logical to assume the valve would fail "as is" since there is a loss of power under the assumption that the valve would lose power as well. Second part is correct.
Answer B Discussion Correct, if a loss of power to the RM80 skid for an RIA occurs, any interlocks for that RIA will occur as if a HIGH ALARM had occurred therefore 1GWD-4 would automatically close. SLC 16.11.3 completion time for securing releases from this pathway is immediately.
Answer C Discussion Incorrect, 1GWD-4 will close. Remaining open is plausible because the HIGH setpoint was not actually reached since the alarms were due to loss of power. Additionally, it is logical to assume the valve would fail "as is" since there is a loss of power under the assumption that the valve would lose power as well. Second part is plausible because 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is a common TS completion time. Additionally, specific to completion times in SLC 16.11.3, releases are allowed to continue for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for planned outages of the RIA's. Since there are conditions where the SLC allows continuing the release for up to 1 hr with no operable RIS it is a plausible distractor for this question.
Answer D Discussion Incorrect. First part is correct. Second part is plausible because 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is a common TS completion time. Additionally, specific to completion times in SLC 16.11.3, releases are allowed to continue for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for planned outages of the RIA's. Since there are conditions where the SLC allows continuing the release for up to 1 hr with no operable RIS it is a plausible distractor for this question.
Basis for meeting the KA Requires the ability to analyze a loss of power to RIA's affiliated with a GWR and determine the status of SLC requirements as a result. Ties to accidental gas release in that it requires knowledge that the tank discharge valve will automatically close to prevent an accidental (i.e.
unmonitored) release. Although the stem does not specifically state that the loss of power is due to maintenance activities, knowledge of the system response and the requirements of the associated SLC would apply.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2009B Q50 Development References Student References Provided RAD-RIA Obj R2, R15, ADMIN-TSS Obj R3 RAD-RIA, SLC-16.11.3 APE060 2.2.36 - Accidental Gaseous-Waste Release APE060 GENERIC Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR:
41.10 / 43.2 / 45.13) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 46 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION BWA03 AK1.3 - Loss of NNI-Y 24 24 B Knowledge of the operational implications of the following concepts as they apply to the (Loss of NNI-Y)
(CFR: 41.8 / 41.10 / 45.3)
Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of NNI-Y)
Given the following Unit 1 conditions:
Reactor power = 2%
1SA2/B11 (ICS AUTO POWER FAILURE) actuated 1SA2/B13 (ICS HAND POWER FAILURE) actuated Which ONE of the following describes:
- 1) the level at which SGs will be maintained?
- 2) how decay heat removal from the core is controlled?
A. 1. 25 inches SUR
- 2. ADVs B. 1. 30 inches XSUR
- 2. ADVs C. 1. 25 inches SUR
- 2. TBVs D. 1. 30 inches XSUR
- 2. TBVs Friday, October 04, 2013 Page 47 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 24 24 B General Discussion Answer A Discussion Incorrect. First part is plausible since it would be correct if Main FDW pumps did not trip when both Hand and Auto power are lost. Second part is correct.
Answer B Discussion Correct. Both Main FDW pumps will trip if both ICS Hand and Auto power are lost therefore EFDW will start and feed SG's while 1FDW-316
& 316 will control at 30" XSUR level. With BOTH Hand and Auto power lost, the TBV's will be failed closed and cannot be operated from the ASDP therefore the ADV's will be used to control decay heat removal.
Answer C Discussion Incorrect. First part is plausible since it would be correct if Main FDW pumps did not trip when both Hand and Auto power are lost. Second part is plausible since there is a condition where the TBV's are failed closed in the control room however they are still operable in manual from the ASDP (loss of vacuum). Since the TBV's are failed closed from the control room here it is plausible that they are still operable in manual from the ASDP.
Answer D Discussion Incorrect. First part is correct. Second part is plausible since there is a condition where the TBV's are failed closed in the control room however they are still operable in manual from the ASDP (loss of vacuum). Since the TBV's are failed closed from the control room here it is plausible that they are still operable in manual from the ASDP.
Basis for meeting the KA Requires knowledge of the operational implication of annunciators associated with loss of KI and KU (NNI-Y) as well as manual actions required following the loss of NNI-Y.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2007 Q25 Development References Student References Provided STG-ICS R33 STG-ICS Intro STG-ICS Chptr 8 STG-ICS Chptr 3 BWA03 AK1.3 - Loss of NNI-Y Knowledge of the operational implications of the following concepts as they apply to the (Loss of NNI-Y)
(CFR: 41.8 / 41.10 / 45.3)
Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of NNI-Y) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 48 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION BWA05 AA2.1 - Emergency Diesel Actuation 25 25 D Ability to determine and interpret the following as they apply to the (Emergency Diesel Actuation)
(CFR: 43.5 / 45.13)
Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
Given the following Unit 1 conditions:
Initial conditions:
Switchyard isolation occurs Current conditions:
Shutdown of KHUs is desired Which ONE of the following states:
- 1) if a Load Shed has occurred?
- 2) the procedure that will be used to perform a remote shutdown of the KHUs?
A. 1. Yes
- 2. OP/0/A/2000/041 (Keowee Modes of Operations)
B. 1. No
- 2. OP/0/A/2000/041 (Keowee Modes of Operations)
C. 1. Yes
- 2. OP/0/A/1106/019 (Keowee Hydro At Oconee)
D. 1. No
- 2. OP/0/A/1106/019 (Keowee Hydro At Oconee)
Friday, October 04, 2013 Page 49 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 25 25 D General Discussion Answer A Discussion Incorrect. First part is plausible since a load shed would occur if either ES had actuated or there was a CT transformer lockout. Second part is plausible since it would be correct if there were an ES actuation and shutdown of KHU's were directed from Encl. 5.41 (ES Recovery).
Answer B Discussion Incorrect. First part is correct. Second part is plausible since it would be correct if there were an ES actuation and shutdown of KHU's were directed from Encl. 5.41 (ES Recovery).
Answer C Discussion Incorrect. First part is plausible since a load shed would occur if either ES had actuated or there was a CT transformer lockout. Second part is correct.
Answer D Discussion Correct. Without either an ES actuation or loss of normal source (CT lockout) there would NOT be a load shed signal. AP/11 directs the RO to use OP/1106/19 to shutdown the KHU's when desired.
Basis for meeting the KA Requires selection of appropriate procedure to shutdown the KHU's following an emergency start signal. Since ONS has no Emergency Diesels and the KHU's are used as emergency power sources, KHU's are used to match the KA.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-APG Obj R9. EL-PSL Obj R5 EL-PSL, AP/11 Encl 5.41 BWA05 AA2.1 - Emergency Diesel Actuation Ability to determine and interpret the following as they apply to the (Emergency Diesel Actuation)
(CFR: 43.5 / 45.13)
Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 50 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION BWE08 EK3.1 - LOCA Cooldown 26 26 A Knowledge of the reasons for the following responses as they apply to the (LOCA Cooldown)
(CFR: 41.5 / 41.10, 45.6, 45.13)
Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.
Given the following Unit 1 conditions:
ES 1-8 have actuated LOCA CD tab in progress RCS pressure = 423 psig slowly decreasing 1A LPI Pump operating in the Piggyback alignment Which ONE of the following describes the:
- 1) operational limitations on the operating LPI pump?
- 2) pump(s) being protected by the above limitation?
A. 1. Maximized to < 3100 gpm
- 2. LPI B. 1. Maximized to < 3100 gpm
- 2. HPI C. 1. Maximized to < 2900 gpm
- 2. LPI D. 1. Maximized to < 2900 gpm
- 2. HPI Friday, October 04, 2013 Page 51 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION General Discussion 26 26 A Answer A Discussion Correct. With only one LPI pump operating in the Piggyback mode LPI flow is maximized to < 3100 gpm to protect the LPI pump from runout.
Answer B Discussion Incorrect. First part is correct. Second part is plausible since the LPI pump is supplying suction to the HPI pumps in this alignment and other conditions place strict flow limits on the HPI pumps to protect them from damage.
Answer C Discussion Incorrect. First part is plausible since 2900 gpm is a flow limit applicable when only one LPI train is operating however it is the LPI flow that transitions the mitigation strategy to a LBLOCA from a SBLOCA or allows securing HPI pumps following a SBLOCA.. Second part is correct.
Answer D Discussion Incorrect. First part is plausible since 2900 gpm is a flow limit applicable when only one LPI train is operating however it is the LPI flow that transitions the mitigation strategy to a LBLOCA from a SBLOCA or allows securing HPI pumps following a SBLOCA..Second part is plausible since the LPI pump is supplying suction to the HPI pumps in this alignment and other conditions place strict flow limits on the HPI pumps to protect them from damage.
Basis for meeting the KA Requires knowledge of the reasons for the operating limitations on LPI pump flow during the LOCA CD tab when only one LPI pump is supplying suction to HPI pumps.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-LCD Obj R6 EAP-SA Obj R17 EAP-LCD BWE08 EK3.1 - LOCA Cooldown Knowledge of the reasons for the following responses as they apply to the (LOCA Cooldown)
(CFR: 41.5 / 41.10, 45.6, 45.13)
Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 52 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION BWE03 EK2.1 - Inadequate Subcooling Margin 27 27 C Knowledge of the interrelations between the (Inadequate Subcooling Margin) and the following:
(CFR: 41.7 / 45.7)
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Given the following Unit 1 conditions:
Reactor trip from 100% due to a SBLOCA Reactor building pressure has peaked at 1.7 psig Subcooled margins are stable as indicated below Which ONE of the following describes how Feedwater will be used to mitigate this event?
Steam Generator levels will be controlled at ________?
A. 240 inches using Emergency Feedwater B. 240 inches using Main Feedwater C. Loss of Subcooling Margin setpoint using Emergency Feedwater D. Loss of Subcooling Margin setpoint using Main Feedwater Friday, October 04, 2013 Page 53 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 27 27 C General Discussion Answer A Discussion Incorrect. 240 inches is plausible since it would be the correct level if all RCP's were secured but SCM was still intact. Emergency Feedwater is correct.
Answer B Discussion Incorrect. 240 inches is plausible since it would be the correct level if all RCP's were secured but SCM was still intact. Using Main Feedwater is plausible since there is no indication of anything that would have caused a trip of the Main Feedwater pumps however Rule 2 directs tripping the MFDW pumps and using EFDW.
Answer C Discussion Correct. If any SCM reaches 0 degreees, Both SG levels must me manually increased to the LOSCM setpoint. Rule 2 directs doing this using Emergency feedwater.
Answer D Discussion Incorrect. LOSCM setpoint is the correct level. Using Main Feedwater is plausible since there is no indication of anything that would have caused a trip of the Main Feedwater pumps however Rule 2 directs tripping the MFDW pumps and using EFDW.
Basis for meeting the KA Requires knowledge of the relationship between a loss of subcooling and Emergency Feedwater controls and instrumentation as it relates to achieveing and maintaining the proper SG levels.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided CF-EF Obj R37 CF-EF Rule 7 BWE03 EK2.1 - Inadequate Subcooling Margin Knowledge of the interrelations between the (Inadequate Subcooling Margin) and the following:
(CFR: 41.7 / 45.7)
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 54 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS003 K5.02 - Reactor Coolant Pump System (RCPS) 28 28 A Knowledge of the operational implications of the following concepts as they apply to the RCPS: (CFR: 41.5 / 45.7)
Effects of RCP coastdown on RCS parameters ........................
Which ONE of the following describes:
- 1) the effect of extending RCP coast down time with the flywheel?
- 2) an expected core delta T (degrees) 30 minutes following a lockout of 1TA and 1TB?
A. 1. Helps prevent the core from reaching DNBR limits
- 2. 35 B. 1. Helps prevent the core from reaching DNBR limits
- 2. 47 C. 1. Reduces the likelihood of a Reactor trip following a RCP trip at power
- 2. 35 D. 1. Reduces the likelihood of a Reactor trip following a RCP trip at power
- 2. 47 Friday, October 04, 2013 Page 55 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 28 28 A General Discussion Answer A Discussion Correct. Coastdown of the RCP's following their trip provides 1-2 minutes of forced flow before the pump has completely stopped and therefore helps prevent the core from reaching or exceeding the DNBR limit. 30-40 degrees delta T is the expected delta T from a 100% power Rx trip after Natural Circulation flow has been established (10-15 minutes).
Answer B Discussion Incorrect. First part is correct. Second part is plausible since it is the expected delta T at 100% power.
Answer C Discussion Incorrect. First part is plausible since it is a true statement in that extending forced RCS flow conditions following a RCP trip would result in the ability to survive a loss of a RCP at a higher power level without reaching the RPS trip setpoint for flux/flow-imbalance. Second part is correct.
Answer D Discussion Incorrect. First part is plausible since it is a true statement in that extending forced RCS flow conditions following a RCP trip would result in the ability to survive a loss of a RCP at a higher power level without reaching the RPS trip setpoint for flux/flow-imbalance.Second part is plausible since it is the expected delta T at 100% power.
Basis for meeting the KA Requires knowledge of the effect that RCP coastdown has on RCS flow.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided PNS-CPM Obj R8 , TA-AM! Obj R3 PNS-CPM TA-AM1 SYS003 K5.02 - Reactor Coolant Pump System (RCPS)
Knowledge of the operational implications of the following concepts as they apply to the RCPS: (CFR: 41.5 / 45.7)
Effects of RCP coastdown on RCS parameters ........................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 56 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS004 A1.06 - Chemical and Volume Control System 29 29 C Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: (CFR: 41.5 / 45.5)
VCT level .......................................................
The Letdown Storage Tank:
- 1) contains approximately __(1)__ gallons of water per inch of level.
Which ONE of the following completes the statements above?
A. 1. 24
- 2. <40 B. 1. 24
- 2. <55 C. 1. 31
- 2. <40 D. 1. 31
- 2. <55 Friday, October 04, 2013 Page 57 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 29 29 C General Discussion Answer A Discussion Incorrect. First part is plausible since it would be correct for the Pressurizer. Second part is correct.
Answer B Discussion Incorrect. First part is plausible since it would be correct for the Pressurizer. Second part is plausible since it is the Lo Lo level alarm setpoint for the LDST.
Answer C Discussion Correct. The LDST is approximately 31.3 gal/in and the setpoint for the LDST level interlock is <40 inches.
Answer D Discussion Incorrect. First part is correct. Second part is plausible since it is the Lo Lo level alarm setpoint for the LDST.
Basis for meeting the KA Requires ability to predict automatic operation of the HPI system controls as a function of VCT (LDST) level.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided PNS-PZR Obj R2, PNS-HPI Obj R8 PNS-PZR, PNS-HPI SYS004 A1.06 - Chemical and Volume Control System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: (CFR: 41.5 / 45.5)
VCT level .......................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 58 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS005 K6.03 - Residual Heat Removal System (RHRS) 30 30 D Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: (CFR: 41.7 / 45.7)
RHR heat exchanger ..............................................
Given the following Unit 2 conditions:
RCS Cooldown in progress 2B LPI cooler isolated due to cooler leak Which ONE of the following states the LPI Decay Heat Removal mode that will be used for the INITIAL transition to LPI cooling?
A. Series B. Normal C. Switchover D. High Pressure Friday, October 04, 2013 Page 59 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 30 30 D General Discussion Answer A Discussion Incorrect. Plausible since the Series mode is one of the LPI cooler modes and it only uses one LPI pump however it uses both coolers..
Answer B Discussion Incorrect. Plausible since this would be correct for Unit 3. Additionally, the Normal mode is one of the LPI modes and it only uses one LPI pump however it uses both LPI coolers. Additionally, due to design restrictions the Normal MODE of LPI is not used for the initial transition to LPI cooling.
Answer C Discussion Incorrect. Plausible since this would be correct if the 2A cooler were not available.
Answer D Discussion Correct. High Pressure mode only uses one cooler and it is the A cooler.
Basis for meeting the KA Requires knowledge of the effect that a loss of one of the DHR coolers will have on available DHR alignments.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED 2009B Q31 Development References Student References Provided PNS-LPI Obj 13, 35 PNS-LPI SYS005 K6.03 - Residual Heat Removal System (RHRS)
Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: (CFR: 41.7 / 45.7)
RHR heat exchanger ..............................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 60 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS006 2.4.45 - Emergency Core Cooling System (ECCS) 31 31 D SYS006 GENERIC Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12)
Given the following Unit 1 conditions:
Rule 3 initiated Loss of Heat Transfer tab in progress Efforts underway to re-establish Steam Generator cooling 1SA-18/D1 (RC SYSTEM APPROACHING SATURATED CONDTIONS) in alarm 1SA-2/D3 (RC PRESS HIGH/LOW) in alarm Pressurizer level = 380 slowly increasing RCS pressure = 2240 psig slowly increasing SCM = 0°F Which ONE of the following states which additional EOP Rules (if any) should be initiated?
A. NO additional rules required B. Rule 2 (Loss of SCM) ONLY C. Rule 4 (Initiation of HPI Forced Cooling) ONLY D. Rule 2 AND Rule 4 Friday, October 04, 2013 Page 61 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 31 31 D General Discussion Answer A Discussion Incorrect. Not running Rule 2 is plausible since this is an LOHT scenario. During an LOHT, while efforts are underway to establish SG cooling you do NOT transfer to the LOSCM tab if SCM is lost due to the heatup. Since the transfer to LOSCM tab does not occur it is plausible to believe that Rule 2 would not be initiated. Not running Rule 4 is correct.
Answer B Discussion Incorrect. Although a transfer to the LOSCM tab is not made, Rule 2 is still required to be initiated and HPI flow established. Criteria for Rule 4 is also met in that Pzr level is > 375 inches and SCM = 0.
Answer C Discussion Incorrect. Plausible since Rule 4 criteria is met. Not running Rule 2 is plausible since this is an LOHT scenario. During an LOHT, while efforts are underway to establish SG cooling you do NOT transfer to the LOSCM tab if SCM is lost due to the heatup. Since the transfer to LOSCM tab does not occur it is plausible to believe that Rule 2 would not be initiated.
Answer D Discussion Correct. Rule 2 is performed due to the loss of subcooling and rule 4 is performed based on SCM and Pzr level.
Basis for meeting the KA Requires interpreting the significance of the RC System Approaching Saturated Conditions statalarm. Once the alarm has actuated, the significance of the alarm is established by monitoring SCM. In this case, the significance of the alarm is that Rule 2must be initiated which by definition is establishing HPI cooling to the core and therefore this is tied to ECCS since HPI is one of the ECCS systems.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-LOHT Obj R23 LOHT tab of EOP SYS006 2.4.45 - Emergency Core Cooling System (ECCS)
SYS006 GENERIC Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 62 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS007 2.2.36 - Pressurizer Relief Tank/Quench Tank System (PRTS) 32 32 D SYS007 GENERIC Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR:
41.10 / 43.2 / 45.13)
Given the following Unit 2 condition:
Initial conditions:
Unit startup in progress RCS temperature = 310°F slowly increasing Maintenance in progress in the area of 2DIB panelboard Current conditions:
2DIB breaker #24 (2RC-66 Pilot Valve DC solenoid power supply) is inadvertently opened Which ONE of the following describes:
- 1) a Tech Spec Limiting Condition of Operation that is NOT met?
- 2) the position of 2RC-66?
A. 1. 3.4.9 (Pressurizer)
- 2. Open B. 1. 3.4.9 (Pressurizer)
- 2. Closed C. 1. 3.4.12 (LTOP)
- 2. Open D. 1. 3.4.12 (LTOP)
- 2. Closed Friday, October 04, 2013 Page 63 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 32 32 D General Discussion Answer A Discussion Incorrect, First part is plausible since the PORV is attached to the Pzr. That means it would be reasonable to believe that the Pzr TS is what contains the requirments for the PORV to be operable since the Pzr could not perform its safety function without the PORV being closed.
Second part is plausible since there are components that fail to the Open position on loss of motive force. Ex: CRD breakers, 1HP-31, etc.
Answer B Discussion Incorrect, First part is plausible since the PORV is attached to the Pzr. That means it would be reasonable to believe that the Pzr TS is what contains the requirments for the PORV to be operable since the Pzr could not perform its safety function without the PORV being closed.
Second part is correct Answer C Discussion Incorrect. First part is correct. Second part is plausible since there are components that fail to the Open position on loss of motive force. Ex:
Answer D Discussion Correct. The LCO of TS 3.4.12 requires the PORV to be Operable. 2RC-66 pilot valve DC solenoid and indicating lights are powered from 2DIB Panelboard breaker #24. If this power is lost, the PORV will close and will NOT open under any conditions until power is restored..
Basis for meeting the KA Chief Examiner said OK to ask about Pzr RV's that relieve to QT. This question requires the ability to analyze the effect of a loss of a power source due to maintenance activities on the LCO to TS 3.4.12 (LTOP).
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided ADMIN-ITS Obj R8, PNS-PZR Obj R30 TS 3.4.12, 3.4.9 PNS-PZR SYS007 2.2.36 - Pressurizer Relief Tank/Quench Tank System (PRTS)
SYS007 GENERIC Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR:
41.10 / 43.2 / 45.13) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 64 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS007 K3.01 - Pressurizer Relief Tank/Quench Tank System (PRTS) 33 33 C Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: (CFR: 41.7 / 45.6)
Containment ....................................................
Given the following Unit 1 conditions:
Initial conditions Loss of all Feedwater HPI forced cooling initiated Quench Tank pressure = 40 psig increasing RCS activity indicates no fuel failures present Current conditions Quench Tank pressure = 3 psig stable Which ONE of the following describes the:
- 1) reactor building RIAs response to the above conditions?
- 2) valves that will automatically close anytime 1RIA-49 reaches its HIGH alarm setpoint?
A. 1. increases
- 2. 1LWD-2 ONLY D. 1. remains constant
- 2. 1LWD-2 ONLY Friday, October 04, 2013 Page 65 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 33 33 C General Discussion Since the second part of the question asks about "IF" 1RIA-49 reaches its setpoint, the second part is a valid question whether you assume RB RIA's are increasing as a result of plant conditions or not.
Answer A Discussion Incorrect. First part is correct. Second part is plausible since it would be correct for an ES 1&2 actuation.
Answer B Discussion Incorrect. First part is plausible under the assumption that failed fuel is the only source of RCS activity. Also plausible if the source of QT pressure rise is due to in-leakage from B Bleed (OE). Also plausible under the assumption that the rupture disc relieves to the component drain header. Second part is plausible since it would be correct for an ES 1&2 actuation.
Answer C Discussion Correct. Decrease in Quench Tank pressure indicates the Rupture Disk has blown. Inventory from the Quench Tank will go to the RBNS causing a level increase. RCS activity in the inventory will result in the RB RIA's increasing. Once 1RIA-49 HIGH alarm setpoint is reached, 1LWD-2 (ONLY) is interlocked to close.
Answer D Discussion Incorrect. First part is plausible under the assumption that failed fuel is the only source of RCS activity. Also plausible if the source of QT pressure rise is due to in-leakage from B Bleed (OE). Also plausible under the assumption that the rupture disc relieves to the component drain header. Second part is correct.
Basis for meeting the KA Requires knowledge of impact of discharge from PORV to the Quench Tank and indications of failed/blown rupture disk and the impact of the failure on containment parameters (loss of QT) and systems (RBNS flow path isolation).
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED 2009 Q32 Development References Student References Provided PNS-CS Obj R7, PNS-PZR RAD-RIA SYS007 K3.01 - Pressurizer Relief Tank/Quench Tank System (PRTS)
Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: (CFR: 41.7 / 45.6)
Containment ....................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 66 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS008 A2.04 - Component Cooling Water System (CCWS) 34 34 A Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunc- tions or operations : (CFR: 41.5 / 43.5 / 45.3 / 45.13)
PRMS alarm ....................................................
Given the following Unit 1 conditions:
1SA-08/B-9 (PROCESS MONITOR RADIATION HIGH) 1RIA-50 in HIGH alarm CC Surge Tank level increasing
- 1) The CC Surge tank __(1)__.
- 2) If the RCS leakage threatens to overflow the associated waste tank, AP/1/A/1700/002 (Excessive RCS Leakage) will direct __(2)__.
Which ONE of the following completes the statements above?
A. 1. will overflow to the LAWT
- 2. tripping the Reactor B. 1. will overflow to the LAWT
- 2. initiating a shutdown using AP/1/A/1700/029 (Rapid Unit Shutdown)
C. 1. will overflow to a floor drain which drains to the MWHUT
- 2. tripping the Reactor D. 1. will overflow to a floor drain which drains to the MWHUT
- 2. initiating a shutdown using AP/1/A/1700/029 (Rapid Unit Shutdown)
Friday, October 04, 2013 Page 67 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 34 34 A General Discussion Answer A Discussion Correct. A Note for step 4.17 in AP/2 says that the CC Surge Tank is hard piped to overflow to LAWT. Step 4.18 of AP/2 directs tripping the Rx if LAWT threatens to overflow.
Answer B Discussion Incorrect. First part is correct. Second part is plausible since there are several instances in AP/2 where the procedure directs using AP/29 to perform a Rapid Unit Shutdown.
Answer C Discussion Incorrect. First part is plausible since the MWHUT is one of the waste tanks located on the Primary side of the plant and is a collection point for various primary sources of waste water. Second part is correct.
Answer D Discussion Incorrect. First part is plausible since the MWHUT is one of the waste tanks located on the Primary side of the plant and is a collection point for various primary sources of waste water. Second part is plausible since there are several instances in AP/2 where the procedure directs using AP/29 to perform a Rapid Unit Shutdown.
Basis for meeting the KA Describes the impact of the CC leakage which is indicated by RIA alarms and uses AP/2 to mitigate the consequences of the leakage.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EAP-APG R9 AP/2 SYS008 A2.04 - Component Cooling Water System (CCWS)
Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunc- tions or operations : (CFR: 41.5 / 43.5 / 45.3 / 45.13)
PRMS alarm ....................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 68 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS010 K4.03 - Pressurizer Pressure Control System (PZR PCS) 35 35 A Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
Over pressure control ............................................
Which ONE of the following states the automatic OPEN setpoints (psig) for 1RC-1 (Pzr Spray) and 1RC-66 (PORV) in Mode 1?
1RC-1 1RC-66 A. 2205 2450 B. 2205 2500 C. 2255 2450 D. 2255 2500 Friday, October 04, 2013 Page 69 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 35 35 A General Discussion Answer A Discussion Correct: 1RC-1 (Pzr Spray) setpoint is 2205 psig and 1RC-66 (PORV) is 2450 psig when in HIGH (Mode 1)
Answer B Discussion Incorrect: 1RC-1 (Pzr Spray) setpoint is correct. 1RC-66 (PORV) setpoint is incorrect. Plausible as 2500 psig is the Pzr Safety Valve setpoint.
Answer C Discussion Incorrect: 1RC-1 (Pzr Spray) setpoint is incorrect. Plausible as 2255 psig is the Pzr High pressure alarm setpoint. 1RC-66 (PORV) setpoint is correct.
Answer D Discussion Incorrect: 1RC-1 (Pzr Spray) setpoint is incorrect. Plausible as 2255 psig is the Pzr High pressure alarm setpoint. 1RC-66 (PORV) setpoint is incorrect as noted above.
Basis for meeting the KA Requires knowledge of Pzr PCS setpoints for automatic pressure control.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2009 Q34 Development References Student References Provided PNS-PZR R5, R9 PNS-PZR SYS010 K4.03 - Pressurizer Pressure Control System (PZR PCS)
Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
Over pressure control ............................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 70 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS012 A2.05 - Reactor Protection System (RPS) 36 36 C Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)
Faulty or erratic operation of detectors and function generators .........
Given the following Unit 1 conditions:
Reactor power = 100%
1D RPS channel in Manual Bypass 1A RPS Thot RTD fails Which ONE of the following describes:
- 1) ALL RPS trips affected by the failure?
- 2) the actions directed in accordance with OP/1/A/1105/014 (Control Room Instrumentation Operation And Information)?
A. 1. RCS High Outlet Temperature ONLY
- 2. Place MANUAL TRIP Keyswitch in "TRIP".
B. 1. RCS High Outlet Temperature ONLY
C. 1. RCS High Outlet Temperature and RCS Variable Low Pressure
- 2. Place MANUAL TRIP Keyswitch in "TRIP".
D. 1. RCS High Outlet Temperature and RCS Variable Low Pressure
Friday, October 04, 2013 Page 71 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 36 36 C General Discussion Answer A Discussion Incorrect: First part is plausible since it is the only trip function in RPS with high temperature in its name. Second part is correct.
Answer B Discussion Incorrect: First part is plausible since it is the only trip function in RPS with high temperature in its name. Second part is plausible since it would be correct if the 1D RPS channel were not already in Manual Bypass.
Answer C Discussion Correct: The High Outlet Temperature trip uses Thot directly to determine if the trip setpoint has been reached. The Variable Low Pressure trip uses Thot in the formula to calculate the low pressure trip: 11.14Thot - 4706 With the 1D RPS channel in Manual Bypass, all functions in the 1A RPS channel are "required" and therefore OP/1105/014 directs tripping the RPS channel as described.
Answer D Discussion Incorrect. First part is correct. Second part is plausible since it would be correct if the 1D RPS channel were not already in Manual Bypass.
Basis for meeting the KA Requires ability to predict the impact of a detector malfunction and the ability to use the procedure to determine the correct actions to take based on the failure.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED ILT40 Q38 Development References Student References Provided Admin-ITS Obj R8, , Admin-PIS Obj R3 IC-RPS Obj R3,4,23 1105/014 IC-RPS SYS012 A2.05 - Reactor Protection System (RPS)
Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)
Faulty or erratic operation of detectors and function generators .........
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 72 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS012 K4.09 - Reactor Protection System (RPS) 37 37 B Knowledge of RPS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
Separation of control and protection circuits .........................
Which ONE of the following describes how RCS Pressure signals are used to provide control signals to the Integrated Control System?
A. Median Selected from two channels of RPS narrow range pressure (A and B) and one wide range pressure B. Median Selected from three channels of RPS narrow range pressure (A, B, and E)
C. 2nd Max Selected from RPS narrow range pressures (A, B, C, & D)
D. 2nd Min Selected from RPS narrow range pressures (A, B, C, & D)
Friday, October 04, 2013 Page 73 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 37 37 B General Discussion Answer A Discussion Incorrect. Plausible since this is how the Thot signal to ICS is generated.
Answer B Discussion Correct. RPS NR channel A, B, and E RCS pressures are median selected as a control signal for ICS.
Answer C Discussion Incorrect. Plausible since this is how RPS uses NR RCS pressure to determine if a high RCS pressure trip is required.
Answer D Discussion Incorrect. Plausible since this is how RPS uses NR RCS pressure to determine if a low RCS pressure trip is required.
Basis for meeting the KA Requires knowledge of design features of RPS which provide for separation between how RCS pressure signals are used for protection circuits and how the same signals are used for control circuits.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided IC-RCI Obj R61 Th ppt IC-RCI SYS012 K4.09 - Reactor Protection System (RPS)
Knowledge of RPS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
Separation of control and protection circuits .........................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 74 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS013 A3.02 - Engineered Safety Features Actuation System (ESFAS) 38 38 C Ability to monitor automatic operation of the ESFAS including: (CFR: 41.7 / 45.5)
Operation of actuated equipment ...................................
Given the following Unit 2 conditions:
Initial conditions:
Reactor power = 100%
Current conditions:
MSLB occurs RCS pressure = 1580 psig slowly increasing RB peak pressure = 2.8 psig Which ONE of the following describes a valve that has received a signal to CLOSE?
A. 2CC-7 B. 2HP-24 C. 2LWD-2 D. 2LPSW-1062 Friday, October 04, 2013 Page 75 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 38 38 C General Discussion Answer A Discussion Incorrect: Plausible since it would be correct if RB pressure had reached the ES 1-6 setpoint of 3 psig.
Answer B Discussion Incorrect. Plausible since this valve did receive an ES signal however it was a signal to open.
Answer C Discussion CORRECT: 2LWD-2 is on ES channel 2. With RCS pressure below the ES channel 1 actuation setpoint for RCS pressure (1600 psig) ES 1 will have actuated and sent a close signal to 2LWD-1 for non essential containment isolation.
Answer D Discussion D. Incorrect: Plausible since 2LPSW-1062 does receive a closed signal from ES actuation however it is not from either Channel 1 or 2. This answer would be correct if ES channel 6 had actuated which would occur at 3 psig RB pressure.
Basis for meeting the KA Requires knowledge of ES actuation setpoints, as well as what components are operated from which ES digital channels. This would demonstrate the ability to monitor automatic operation of actuated equipment.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED 2009A Q39 Development References Student References Provided IC-ES R14, R18 IC-ES SYS013 A3.02 - Engineered Safety Features Actuation System (ESFAS)
Ability to monitor automatic operation of the ESFAS including: (CFR: 41.7 / 45.5)
Operation of actuated equipment ...................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 76 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS064 K1.04 - Emergency Diesel Generator (ED/G) System 39 39 C Knowledge of the physical connections and/or cause-effect relationships between the ED/G system and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
DC distribution system ...........................................
Given the following plant conditions:
Time = 1200 Unit 1 Reactor power = 100%
Unit 2 Reactor power = 100%
ACB-4 closed Time = 1201 LOCA occurs on Unit 1 Switchyard Isolation occurs Which ONE of the following states the source of power being used to energize 1DIA at Time = 1202?
A. Control Batteries B. KHU-1 C. KHU-2 D. CT-5 Friday, October 04, 2013 Page 77 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 39 39 C General Discussion Answer A Discussion Incorrect. This is plausible for a couple of reasons:
- 1. This would be correct if asking between the Rx trip and when the MFB is energized which takes about 30 seconds.
- 2. Plausible since there are loads that have a delay before they re-energize following a LOCA/LOOP to protect bus voltage. Since the loads powered from DIA would be powered by the batteries during the "down" time it is reasonable to believe that the battery chargers would be fed from a load center that delays re-energizing following a LOOP since there would be no loss of power to the supplied components if that were true.
Answer B Discussion Incorrect. Plausible since this is the correct answer for 2DIA. Since Unit 2 did not have a loca it would energize its MFB from the overhead power path which would mean KHU-1.
Answer C Discussion Correct. With a LOCA/LOOP occurring, the MFB would re-energize from the underground power path which means if would energize from KHU-2. 1TC would energize from the MFB which would energize 1X8 which would energize 1XS1 which is the power supply for the 1CA battery Charger. Since the battery charger has a higher output voltage than the battery bank, 1DIA would be energized from the charger.
Answer D Discussion Incorrect. Plausible since this would be the correct answer if the standby bus were energized from Central or Lee prior to time = 1201.
Basis for meeting the KA Requires knowledge of the cause-effect relationship of the KHU emergency operation (ED/G system) and the source of power to one of the DC distribution panel boards.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EL-DCD Obj 06, EL-PSL Obj R24 EL-DCD EL-PSL SYS064 K1.04 - Emergency Diesel Generator (ED/G) System Knowledge of the physical connections and/or cause-effect relationships between the ED/G system and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
DC distribution system ...........................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 78 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS073 A1.01 - Process Radiation Monitoring (PRM) System 40 40 A Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRM system controls including: (CFR: 41.5 / 45.7)
Radiation levels .................................................
Given the following Unit 1 conditions:
1A GWD tank release in progress 1RIA-37 HIGH alarm actuates 1SA-8/B9 (Process Monitor Radiation High) actuates Which ONE of the following describes the:
- 1) automatic actions that will occur?
- 2) procedure that contains actions that must be performed prior to re-initiating the release?
A. 1. Closes the GWD tank outlet valves and stops the Waste Gas Exhauster but does NOT trip the running GWD compressors
- 2. OP/1-2/A/1104/018 (GWD System)
B. 1. Closes the GWD tank outlet valves, stops the Waste Gas Exhauster, AND trips running GWD compressors
- 2. OP/1-2/A/1104/018 (GWD System)
C. 1. Closes the GWD tank outlet valves and stops the Waste Gas Exhauster but does NOT trip the running GWD compressors
- 2. AP/18 (Abnormal Release of Radioactivity)
D. 1. Closes the GWD tank outlet valves, stops the Waste Gas Exhauster, AND trips running GWD compressors
- 2. AP/18 (Abnormal Release of Radioactivity)
Friday, October 04, 2013 Page 79 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 40 40 A General Discussion Answer A Discussion Correct: A HIGH alarm from RIA-37 will close all of the GWD tank outlet valves and stop the Waste Gas Exhauster. The associated ARG will direct going to OP/1-2/A/1104/018 (GWD System) to provide additional guidance on what to do with the release that has now been terminated.
The entry conditions for AP/18 are not met.
Answer B Discussion Incorrect: First part is plausible since it is partially correct in that a HIGH alarm from RIA-37 will close all of the GWD tank outlet valves and isolate the Waste Gas Exhauster. Tripping the GWD compressors is plausible under the misconception that it is the GWD compressors that are providing the driving force for the tank release. Second part is correct.
Answer C Discussion Incorrect: First part is correct. Second part is plausible since for both RIA-54 (Turbine Building Sump) and RIA-45 (RB Purge), there are actions in AP/18 that must be performed prior to going to the associated OP to take actions to resume the release.
Answer D Discussion Incorrect: First part is plausible since it is partially correct in that a HIGH alarm from RIA-37 will close all of the GWD tank outlet valves and isolate the Waste Gas Exhauster. Tripping the GWD compressors is plausible under the misconception that it is the GWD compressors that are providing the driving force for the tank release. Second part is plausible since for both RIA-54 (Turbine Building Sump) and RIA-45 (RB Purge), there are actions in AP/18 that must be performed prior to going to the associated OP to take actions to resume the release.
Basis for meeting the KA Question requires the ability to monitor changes in parameters (Radiation Levels) associated with RIA's to prevent exceeding design limits.
Verification that the correct automatic actions have occurred to isolate the release on high rad levels is demonstrating the ability to prevent exceeding design limits associated with Process RIA's and radiation levels.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ILT40 Q73 Development References Student References Provided EAP-APG Obj R9 , RAD-RIA Obj R2 AP/18, RAD-RIA 1SA-8/B9 /ARG SYS073 A1.01 - Process Radiation Monitoring (PRM) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRM system controls including: (CFR: 41.5 / 45.7)
Radiation levels .................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 80 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS073 K3.01 - Process Radiation Monitoring (PRM) System 41 41 B Knowledge of the effect that a loss or malfunction of the PRM system will have on the following: (CFR: 41.7 / 45.6)
Radioactive effluent releases ......................................
Given the following Unit 1 conditions:
Reactor is in MODE 5 RB Purge in progress Unit 1 vent activity increasing 1RIA-45 HIGH alarm fails to actuate at setpoint
- 1) Automatic termination of RB Purge operation due to increasing activity __(2)__
available?
- 2) Purge operation __(1)__ be allowed if the unit were in MODE 4.
Which ONE of the following completes the statements above?
A. 1. is
- 2. would B. 1. is
- 2. would NOT C. 1. is NOT
- 2. would D. 1. is NOT
- 2. would NOT Friday, October 04, 2013 Page 81 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 41 41 B General Discussion Answer A Discussion Incorrect First part is correct. Second part is plausible since all Containment isolation valves except the RB purge valves are allowed to be operated above MODE 5 under administrative controls IAW Tech Spec 3.6.3.. Also, IAW OP/1102/14 L&P's operation of Purge valves is allowed ONLY in Modes 5, 6, & NO MODE..
Second part is correct Answer B Discussion
- Correct, In case of a failure of RIA-45 HIGH alarm, then RIA-46 HIGH alarm (via the switchover function) will actuate the required interlock functions.
IAW OP/1102/14 L&P's operation of Purge valves is allowed ONLY in Modes 5, 6, & NO MODE..
Answer C Discussion Incorrect. First part is plausible since 1RIA-45 provides the normal means of automatic isolation of RB purge based on increasing activity therefore it would be plausible to assume that if 1RIA-45 did not auto terminate RB Purge then manual termination would be required.
Second part is plausible since all Containment isolation valves except the RB purge valves are allowed to be operated above MODE 5 under administrative controls IAW Tech Spec 3.6.3.. Also, IAW OP/1102/14 L&P's operation of Purge valves is allowed ONLY in Modes 5, 6, & NO MODE..
Second part is correct Answer D Discussion Incorrect. First part is plausible since 1RIA-45 provides the normal means of automatic isolation of RB purge based on increasing activity therefore it would be plausible to assume that if 1RIA-45 did not auto terminate RB Purge then manual termination would be required.
Second part is correct.
Basis for meeting the KA Requires knowledge of the effect that a loss of RIA's will have on Radioactive Effluent releases that are in progress.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory MODIFIED ILT39 Q51 Development References Student References Provided RAD-RIA R2 OP/1102/14 SYS073 K3.01 - Process Radiation Monitoring (PRM) System Knowledge of the effect that a loss or malfunction of the PRM system will have on the following: (CFR: 41.7 / 45.6)
Radioactive effluent releases ......................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 82 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS076 K3.01 - Service Water System (SWS) 42 42 A Knowledge of the effect that a loss or malfunction of the SWS will have on the following: (CFR: 41.7 / 45.6)
Closed cooling water .............................................
Given the following Unit 1 conditions:
1A LPSW Pump trips Standby LPSW pump fails to start Which ONE of the following will begin to increase in temperature?
ASSUME NO MANUAL ACTIONS ARE TAKEN A. Letdown B. Spent Fuel Pool C. Main Feedwater Pump oil temperature D. Primary Instrument Air Compressor discharge air temperature Friday, October 04, 2013 Page 83 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 42 42 A General Discussion Answer A Discussion Correct. LPSW cools the CC coolers which in turn cool letdown therefore degraded LPSW flow would result in increasing letdown temperatures.
Answer B Discussion Incorrect. Plausible since it would be correct for degraded RCW flows.
Answer C Discussion Incorrect. Plausible since it would be correct for degraded RCW flows.
Answer D Discussion Incorrect. Plausible since it would be correct for degraded HPSW flows.
Basis for meeting the KA To determine that letdown temperatures would increase on degraded LPSW the student must have knowledge of the effect of a loss of LPSW on the CC system since LPSW is the cooling medium for the CC coolers and then CC is the cooling medium for letdown.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided PNS-CC Obj 03 PNS-CC, SSS-IA, SSS-HPW, SSS-RCW SYS076 K3.01 - Service Water System (SWS)
Knowledge of the effect that a loss or malfunction of the SWS will have on the following: (CFR: 41.7 / 45.6)
Closed cooling water .............................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 84 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS078 K2.02 - Instrument Air System (IAS) 43 43 C Knowledge of bus power supplies to the following: (CFR: 41.7)
Emergency air compressor ........................................
Which ONE of the following is the power supply for the Unit 2 Auxiliary Instrument Air System compressor?
A. 2XD B. 2XF C. 2XP D. 2XS1 Friday, October 04, 2013 Page 85 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 43 43 C General Discussion Auxiliary Instrument Air System compressors are powered from non-load shed power supplies 1,2,3XP.
Answer A Discussion Incorrect, plausible because 1XD supplies the "A" Worthington compressor (Backup IA Compressor).
Answer B Discussion Incorrect, plausible because 2XF supplies the "C" Worthington compressor (Backup IA Compressor).
Answer C Discussion Correct, Auxiliary Instrument Air System compressors are power from non-load shed power supplies 2XP Answer D Discussion Incorrect, plausible because this is a ES safety related bus.
Basis for meeting the KA Requires knowledge of the power supply for the AIA compressors.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2007 Q49 Development References Student References Provided SSS-IA SYS078 K2.02 - Instrument Air System (IAS)
Knowledge of bus power supplies to the following: (CFR: 41.7)
Emergency air compressor ........................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 86 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS103 A4.04 - Containment System 44 44 A Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
Phase A and phase B resets ........................................
Given the following Unit 1 conditions:
Initial conditions:
Reactor Power = 100%
1A MSLB inside containment Current conditions:
Core SCM = 18°F stable RB Pressure = 17 psig slowly decreasing Which ONE of the following sets of actions is required by Enclosure 5.1 (ES Actuation)
A. Take ES Channel 1 to manual AND open 1HP-20 B. Take ES Channel 1 to manual AND open 1HP-3 C. Override Odd Voters AND open 1HP-20 D. Override Odd Voters AND open 1HP-3 Friday, October 04, 2013 Page 87 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 44 44 A General Discussion Answer A Discussion Correct. ES channels 1 and 2 are taken to manual and since RCP's would still be operating (SCM has not been lost) Encl 5.1 will direct restoring seal return by opening 1HP-20 and 21.
Answer B Discussion Incorrect. Plausible since:
1HP-3 is on ES channel 1 ES Channel 1 is taken to manual Restoring letdown is desired AND directed by the EOP however it is done by Encl 5.5 1HP-20 is not opened in all cases adding to the plausibility of 1HP-3 being correct.
Answer C Discussion Incorrect. Plausible since this would be correct if ES Channel 1 was unable to be placed in Manual.
Answer D Discussion Incorrect. Plausible since placing the Odd voter in Override is a correct action if ES-1 cannot be placed in Manual. 1HP-3 is plausible since:
Restoring letdown is desired AND directed by the EOP however it is done by Encl 5.5 1HP-20 is not opened in all cases adding to the plausibility of 1HP-3 being correct.
Basis for meeting the KA Requires the ability to manually operate Containment Isolation valves after being closed due to ES actuation. Phase A and Phase B correlate to essential and non-essential RB isolation here at ONS therefore determining if taking ES channel to manual or Overriding the associated ES Voters is required to reset the associated logic prior to being able to re-open a containment isolation valve following ES actuation demonstrates the ability to manually operate "Phase A and phase B resets".
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-ESA Obj R5 EAP-ESA EOP Encl 5.1 SYS103 A4.04 - Containment System Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
Phase A and phase B resets ........................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 88 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS016 A3.02 - Non-Nuclear Instrumentation System (NNIS) 45 45 A Ability to monitor automatic operation of the NNIS, including: (CFR: 41.7 / 45.5)
Relationship between meter readings and actual parameter value .........
Given the following Unit 1 conditions:
Initial conditions:
Reactor Power = 100%
Current conditions:
1TA and 1TB lockout occurs BOTH Main Feedwater pumps trip Which ONE of the following describes:
- 1) the Steam Generator levels that will be automatically maintained?
- 2) actions required (if any) to ensure desired SG level is maintained if Abnormal Containment conditions were to develop?
A. 1. 240 XSUR
- 2. manually increase SG level B. 1. 240 XSUR
- 2. no actions required C. 1. 50% OR
- 2. no actions required Friday, October 04, 2013 Page 89 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 45 45 A General Discussion Answer A Discussion Correct. 1TA and 1TB lockout result in a loss of all RCP's which would cause a Rx trip. Since both Main FDW pumps trip, EFDW will actuate and automatically control SG levels at 240" XSUR level. If ACC conditions were to develop the RO would be required to take manual control of EFDW and raise indicated SG levels to 270" XSUR to ensure the desired level of 240" is maintained.
Answer B Discussion Incorrect. First part is correct. Second part is plausible since it would be correct if SG levels were being controlled by Main FDW at 50% OR since the OR is temperature compensated and therefore does not require adjusting for degraded containment.
Answer C Discussion Incorrect. First part is plausible since it would be correct if either Main FDW pump were still in operation. Second part is correct.
Answer D Discussion incorrect. First part is plausible since it would be correct if either Main FDW pump were still in operation. Second part is plausible since it would be correct if SG levels were being controlled by Main FDW at 50% OR since the OR is temperature compensated and therefore does not require adjusting for degraded containment.
Basis for meeting the KA Requires demonstrating the ability to monitor for proper automatic operation of SG level control system following a loss of RCP's as well as demonstrating an understanding of the impact that abnormal containment conditions will have on indicated SG level by demonstrating the ability to maintain desired SG level when abnormal containment conditions develop.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided CF-EF R37 CF-EF Rule 7 SYS016 A3.02 - Non-Nuclear Instrumentation System (NNIS)
Ability to monitor automatic operation of the NNIS, including: (CFR: 41.7 / 45.5)
Relationship between meter readings and actual parameter value .........
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 90 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS015 2.2.12 - Nuclear Instrumentation System (NIS) 46 46 A SYS015 GENERIC Knowledge of surveillance procedures. (CFR: 41.10 / 45.13)
Given the following Unit 1 conditions:
Reactor power = 50% slowly decreasing OAC Unavailable Computer Reactor Calculation Package NOT running Which ONE of the following is:
- 1) the HIGHER power level (% Power) where Tech Spec limits on Reactor Power Imbalance do NOT apply?
- 2) directed to be used by OP/1/A/1105/014 (Control Room Instrumentation Operation And Information) to determine if Imbalance limits specified in the COLR have been exceeded?
A. 1. 35
- 2. CR gages for Power Range NIs and formula provided in OP/1/A/1105/014 B. 1. 35
- 2. PT/1/A/1103/019 (Backup Incore Detector System)
C. 1. 15
- 2. CR gages for Power Range NIs and formula provided in OP/1/A/1105/014 D. 1. 15
- 2. PT/1/A/1103/019 (Backup Incore Detector System)
Friday, October 04, 2013 Page 91 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 46 46 A General Discussion Answer A Discussion Correct. Tech Spec 3.2.2 Mode of Applicability is > 40%. If the Reactor Calculation Package is not running then Outcore detectors are used for Imbalance and the formula for calculating and the direction to use the CR gages are in step 3.2.7.
Answer B Discussion Incorrect. First part is correct. Second part is plausible as it would be correct if any one of the NI's were inoperable. It is also plausible from the perspective that if Incores is the highest priority (which is correct) then it would be reasonable to believe that the backup Incores would be the second highest priority.
Answer C Discussion Incorrect. First part is plausible since it would be correct if asking about TS 3.2.3 (Quadrant Power Tilt). Second part is correct.
Answer D Discussion Incorrect. First part is plausible since it would be correct if asking about TS 3.2.3 (Quadrant Power Tilt). Second part is plausible as it would be correct if any one of the NI's were inoperable. It is also plausible from the perspective that if Incores is the highest priority (which is correct) then it would be reasonable to believe that the backup Incores would be the second highest priority.
Basis for meeting the KA Requires knowledge of surveillance procedures as part of performing surveillances associated with nuclear instrumentation.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided ADM-PIS Obj R5 ADM-PIS OP/1/A/1105/014 Encl. 4.13 SYS015 2.2.12 - Nuclear Instrumentation System (NIS)
SYS015 GENERIC Knowledge of surveillance procedures. (CFR: 41.10 / 45.13) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 92of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS034 A4.02 - Fuel Handling Equipment System (FHES) 47 47 C Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
Neutron levels ...................................................
Given the following Unit 1 conditions:
Initial conditions:
Mode 6 REFUELING is in progress All four SR NIs in service SR 1NI-1 and SR 1NI-3 are the designated NIs for Fuel Handling Current conditions:
Power supply to SR 1NI-1 fails (0 vdc)
Which ONE of the following describes the impact on refueling activities in accordance with OP/1/A/1502/007 (Operations Defueling/Refueling Responsibilities)?
A. Allowed to continue because two other SR NIs remain in service B. Allowed to continue because SR NI-3 is still in service C. Required to be stopped until another SR NI is designated because other NIs are procedurally allowed to be designated D. Required to be stopped and cannot be resumed until SR 1NI-1 is returned to service because other NIs are NOT procedurally allowed to be designated Friday, October 04, 2013 Page 93 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 47 47 C General Discussion Harder this way since the LP talks about core alterations, not defeueling or refueling.
Answer A Discussion Incorrect: Plausible because the two SR NIs remain in service therefore IAW 1502/07 this would be the correct answer for defueling.
Answer B Discussion Incorrect and plausible. There are 4 Source Range NI's available, It is reasonable to conclude that we would have one more than is required while shutdown if one of the Source Range NI's failed (Incorrectly applying the minimum degree of redundancy concept). However the two must be selected by Reactor Engineering per OP/3/A/1502/007 (Encl 4.1 Step 4.3).
Answer C Discussion Correct: Procedure requires movement to be stopped until 2 NIs used to monitor core reactivity can be designated.
Answer D Discussion Incorrect but plausible since it is one of the "designated" NI's that has failed. It would be reasonable to infer that which NI's were "designated" were directed by the referenced procedure and if that were the case then this would be the correct answer.
Basis for meeting the KA Requires demonstrating the ability to monitor neutron levels during fuel handling activities as required by plant procedures. Knowing what the requirments for operable NI's during fuel movement is integral to the ability to montor neutron levels in the Control Room.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2009B Q21 Development References Student References Provided Obj. FH-FHS R20 FH-FHS OP/1/A/1502/007 SYS034 A4.02 - Fuel Handling Equipment System (FHES)
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
Neutron levels ...................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 94 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS041 K4.11 - Steam Dump System (SDS)/Turbine Bypass Control 48 48 C Knowledge of SDS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
T-ave./T-ref. program .............................................
Given the following Unit 1 conditions:
Time = 1200 Reactor Power = 40% stable following an instrument failure Turbine Header Pressure = 860 psig stable Feedwater, Reactor, and Main Turbine in Manual Time = 1300 ICS in Automatic Turbine Header Pressure = 860 psig stable Time = 1301 Reactor Trips prior to any Turbine Header Pressure setpoint adjustments Which ONE of the following is the pressure (psig) where the Turbine Bypass Valves will automatically control Steam Generator pressure?
A. 885 B. 910 C. 985 D. 1010 Friday, October 04, 2013 Page 95 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 48 48 C General Discussion Answer A Discussion Incorrect. Plausible since this is the normal THP setpoint.
Answer B Discussion Incorrect. Plausible since this is setpoint plus 50 psig and is therefore correct prior to the Rx trip.
Answer C Discussion Correct. On a Rx trip the TBV setpoint shifts from setpoint plus 50 psig to setpoint plus 125 psig to limit the RCS cooldown following a trip.
Answer D Discussion Incorrect. Plausible since this would be correct if THP setpoint were at its normal value of 885 psig.
Basis for meeting the KA Required knowledge of the interlock associated with the Turbine Bypass valves that shifts the controlling SG pressure following a Rx trip in order to control Tave at a higher value that would otherwise occur and thereby limit the RCS shrink following a Rx trip. The question requires knowledge of design features of the Turbine Bypass system that provide for Tave control following a Rx trip.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided STG-ICS R10 STG-ICS chapter 3 SYS041 K4.11 - Steam Dump System (SDS)/Turbine Bypass Control Knowledge of SDS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
T-ave./T-ref. program .............................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 96 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS056 A2.04 - Condensate System 49 49 D Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
Loss of condensate pumps .........................................
Given the following Unit 1 conditions:
Time = 1200:00 Reactor power = 80% stable 1A and 1B CBP operating Time = 1201:00 1A CBP trips Feedwater Pump suction pressure = 225 psig slowly decreasing Time = 1203:00 Feedwater Pump suction pressure = 220 slowly increasing Which ONE of the following describes the:
- 1) runback rate (%/min) inserted at Time = 1201:00 to ICS?
- 2) procedure that will be directed by the CRS at Time = 1203:00?
A. 1. 15
- 2. AP/1/A/1700/001 (Unit Runback)
B. 1. 15
- 2. EOP C. 1. 20
- 2. AP/1/A/1700/001 (Unit Runback)
D. 1. 20
- 2. EOP Friday, October 04, 2013 Page 97 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 49 49 D General Discussion Answer A Discussion Incorrect. First part is plausible since there are ICS runbacks that incorporate the 15%/min runback rate. Second part is plausible since it would be correct for the first 90 seconds of the transient.
Answer B Discussion Incorrect. First part is plausible since there are ICS runbacks that incorporate the 15%/min runback rate. Second part is correct, Answer C Discussion Incorrect. First part is correct. Second part is plausible since it would be correct for the first 90 seconds of the transient.
Answer D Discussion Correct. With FDWP suction pressure < 235 psig, an ICS runback is initiated. The runback rate is 20%/min to a power level of 15% or until the low suction pressure clears. After 90 seconds, if FDWP suction pressure is still < 235 psig the FDWP's will trip which will trip the Rx and require entry into the EOP to mitigate the loss of main feedwater.
Basis for meeting the KA Requires knowledge of the impact of a loss of Condensate Booster Pump and knowledge of the procedure that will be used to mitigate the event.
Basis for Hi Cog Requires analyzing plant data to determine the Unit response and the procedure that will be used to mitigate the event.
Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ILT40 Q62 Development References Student References Provided Obj STG-ICS R3 EAP-SA R21, R24 EAP-SA STG-ICS Intro & Chptr 2 SYS056 A2.04 - Condensate System Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
Loss of condensate pumps .........................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 98 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS068 K6.10 - Liquid Radwaste System (LRS) 50 50 D Knowledge of the effect of a loss or malfunction on the following will have on the Liquid Radwaste System : (CFR: 41.7 / 45.7)
Radiation monitors ...............................................
Given the following Unit 1 conditions:
Reactor power = 100%
Primary to Secondary leakage of 10 gpd has just been detected AP/1/A/1700/031 (Primary to Secondary Leakage) has been initiated
- 1) In accordance with AP/31, opening the Turbine Building Sump (TSP) pump breakers prior to being ready to hang White Tags on the TBS pump breakers
__(1)__ allowed.
__(2)__.
Which ONE of the following completes the statements above?
A. 1. is NOT
- 2. after a 2 minute timer B. 1. is NOT
- 2. immediately C. 1. is
- 2. after a 2 minute timer D. 1. is
- 2. immediately Friday, October 04, 2013 Page 99 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 50 50 D General Discussion Answer A Discussion Incorrect. First part is plausible for two reasons:
- 1. Normally, tags are prepared and carried to the compont so that they can be hung as soon as the component in question is placed in the position required by the tag,
- 2. Since there is a SGTL in progress it would be plausible to believe that procedure required hanging the tags as soon as the breakers were opened so that there would be no chance of someone closing the breakers back in with activity from the tube leak in the sump.
Second part is plausible because there is a 2 minute timer associated with low sample pump flow that provides for an automatic backwash of the strainer on high strainer DP. It is plausible to believe it applies here since most SLC requirements for RIA's have a provision to allow in progress releases to continue on loss of the associated RIA's which makes a 2 minute timer to allow power to be restored additionally plausible.
Answer B Discussion Incorrect. First part is plausible for two reasons:
- 1. Normally, tags are prepared and carried to the compont so that they can be hung as soon as the component in question is placed in the position required by the tag,
- 2. Since there is a SGTL in progress it would be plausible to believe that procedure required hanging the tags as soon as the breakers were opened so that there would be no chance of someone closing the breakers back in with activity from the tube leak in the sump.
Second part is correct, Answer C Discussion Incorrect. First part is correct. Second part is plausible because there is a 2 minute timer associated with low sample pump flow that provides for an automatic backwash of the strainer on high strainer DP. It is plausible to believe it applies here since most SLC requirements for RIA's have a provision to allow in progress releases to continue on loss of the associated RIA's which makes a 2 minute timer to allow power to be restored additionally plausible.
Second part is plausible because there is a 2 minute timer associated with low sample pump flow that provides for an automatic backwash of the strainer on high strainer DP. It is plausible to believe it applies here since most SLC requirements for RIA's have a provision to allow in progress releases to continue on loss of the associated RIA's which makes a 2 minute timer to allow power to be restored additionally plausible.
Answer D Discussion Correct.
A note in AP/31 informs the reader that the white tags can be created and hung after the TBS pumjp breakers are opened.
A loss of power to RIA=54 will automatically trip both TBS pump breakers.
Basis for meeting the KA Required knowledge of the effect of a loss of power to RIA-54 will have on Liquid Waste Releases from the Turbine Building Sumps.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Rad-RIA Obj R2 Rad-RIA SYS068 K6.10 - Liquid Radwaste System (LRS)
Knowledge of the effect of a loss or malfunction on the following will have on the Liquid Radwaste System : (CFR: 41.7 / 45.7)
Radiation monitors ...............................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 100 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS071 K3.04 - Waste Gas Disposal System (WGDS) 51 51 B Knowledge of the effect that a loss or malfunction of the Waste Gas Disposal System will have on the following: (CFR: 41.7 / 45.6)
Ventilation system ...............................................
Given the following Unit 1 conditions:
Initial conditions:
Time = 1200 1A GWD tank pressure = 68 psig stable Current conditions:
Time = 1205 1A GWD tank pressure = 18 psig rapidly decreasing Various Aux Building RIAs in alarm 1RIA-1 (Control Room Monitor) in HIGH alarm 1RIA-39 (CNTL RM Gas) in HIGH alarm AP/1/A/1700/018 (Abnormal Release of Radioactivity) in progress A and B Outside Air Booster Fans have been started Which ONE of the following:
- 1) states if 1RIA-1 has a local alarm (do not count associated statalarm(s))?
- 2) describes the areas being provided outside air via the Outside Air Booster Fans?
A. 1. Yes
- 2. Control Room ONLY B. 1. No
- 2. Control Room ONLY C. 1. Yes
- 2. Control Room, Cable Rooms, and the Equipment Rooms D. 1. No
- 2. Control Room, Cable Rooms, and the Equipment Rooms Friday, October 04, 2013 Page 101 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 51 51 B General Discussion Answer A Discussion Incorrect. First part is plausible because many of the Area RIA's do have local alarms that sound when setpoints are reached. Second part is correct.
Answer B Discussion Correct. 1RIA-1 has no local horn that sounds to alert the operator other than Statalarm annunciators. The Outside Air Booster Fans provide outside air to the Control Room only via existing AHU supply lines.
Answer C Discussion Incorrect. First part is plausible because many of the Area RIA's do have local alarms that sound when setpoints are reached. Second part is plausible since it is correct for the CRACS system operation but not the Outside Air booster fans.
Answer D Discussion Correct. 1RIA-1 has no local horn that sounds to alert the operator other than Statalarm annunciators. Second part is plausible since it is correct for the CRACS system operation but not the Outside Air booster fans.
Basis for meeting the KA Requires knowledge of the effect that a Malfunction of the GWD system (ruptured gas tank) will have on ventilation systems (Control Room Ventilation via the Outside Air Booster Fans). Specifically it required knowledge of how the ventilation system works once the OABF's are started during a malfunction of the GWD system, Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided RAD-RIA Obj R2, BPS-DPR Obj 01 RAD-RIA, AP/18 SYS071 K3.04 - Waste Gas Disposal System (WGDS)
Knowledge of the effect that a loss or malfunction of the Waste Gas Disposal System will have on the following: (CFR: 41.7 / 45.6)
Ventilation system ...............................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 102 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS072 A1.01 - Area Radiation Monitoring (ARM) System 52 52 B Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ARM system controls including: (CFR: 41.5 / 45.5)
Radiation levels .................................................
1RIA-59 setpoints are set by __(1)__ and the MINIMUM power level at which 1RIA-59 is used to determine SGTL rate is __(2)__ (% power) in accordance with the EOP.
Which ONE of the following completes the statement above?
A. 1. I&E
- 2. 20 B. 1. I&E
- 2. 40 C. 1. ROs
- 2. 20 D. 1. ROs
- 2. 40 Friday, October 04, 2013 Page 103 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 52 52 B General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible since 20% power is the level described in the lesson plan as the power level where the SGTR leak rates become accurate.
Answer B Discussion Correct. Unlike most other RIA's, I&E has to set the setpoints for RIA's 59/60 due to compatibility issues with the RM80's. The EOP directs only using these RIS's if at or greater than 40% power.
Answer C Discussion Incorrect. First part is plausible since it would be correct for most every other RIA. Second part is plausible since 20% power is the level described in the lesson plan as the power level where the SGTR leak rates become accurate.
Answer D Discussion Incorrect. First part is plausible since it would be correct for most every other RIA. Second part is correct.
Basis for meeting the KA RIA-59/60 are Area monitors used under certain conditions to determine the magnitude of SGTL present. Knowing the threshold power level for using the RIA's is integral in the ability to monitor changes in Radiation levels that correlate to SGTL rate. The KA does not require that the RIA controls actually be operated however setting the setpoint at which the RIA's come into alarm would be considered operating the controls and being able to determine when the alarms are valid would be :associated with "operating the controls" Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided RAD-RIA Obj R2, R15 RAD-RIA SGTR SYS072 A1.01 - Area Radiation Monitoring (ARM) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ARM system controls including: (CFR: 41.5 / 45.5)
Radiation levels .................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 104 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS075 K2.03 - Circulating Water System 53 53 D Knowledge of bus power supplies to the following: (CFR: 41.7)
Emergency/essential SWS pumps ...................................
Which ONE of the following states all of the switchgear that can supply power to the B LPSW pump?
A. 1TD ONLY B. 2TC ONLY C. 1TC or 2TD D. 1TD or 2TD Friday, October 04, 2013 Page 105 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 53 53 D General Discussion Answer A Discussion Incorrect. Plausible because it can supply power to the B LPSW pump but 2TD can also.
Answer B Discussion Incorrect. Plausible because it it would be correct for the C LPSW pump.
Answer C Discussion Incorrect. Plausible because they both supply power to the LPSW pumps; A and C respectively.
Answer D Discussion Correct. The B LPSW pump can be supplied power from 1TD or 2TD.
Basis for meeting the KA Question requires knopwledge of the power supply for one of the LPSW pumps.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK ILT41 Q50 Development References Student References Provided SSS-LPW R11 SSS-LPW SYS075 K2.03 - Circulating Water System Knowledge of bus power supplies to the following: (CFR: 41.7)
Emergency/essential SWS pumps ...................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 106 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS079 K1.01 - Station Air System (SAS) 54 54 D Knowledge of the physical connections and/or cause-effect relationships between the SAS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
IAS ............................................................
Based on the graph above, which ONE of the following describes the EARLIEST time at which SA-141 (SA to IA Controller) will automatically open?
A. 1207 B. 1210 C. 1212 D. 1215 Friday, October 04, 2013 Page 107 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 54 54 D General Discussion Answer A Discussion Incorrect: Plausible since 93 psig is the pressure at which the Backup IA compressors will start.
Answer B Discussion Incorrect: Plausible since 90 psig is the pressure at which the Diesel Air Compressors will start Answer C Discussion Incorrect: Plausible sine 88 psig is the pressure at which the AIA compressors will start Answer D Discussion CORRECT: SA to IA Controller (SA-141) valve senses the IA system pressure and opens at 85 psig to allow service air into the IA system.
Basis for meeting the KA Requires knowledge of automatic cross-connect between Service air and Instrument air systems.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK ILT40 Q65 Development References Student References Provided Obj SSS-IA R52 SYS079 K1.01 - Station Air System (SAS)
Knowledge of the physical connections and/or cause-effect relationships between the SAS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
IAS ............................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 108 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.1 2.1.26 - GENERIC - Conduct of Operations 55 55 B Conduct of Operations Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (CFR: 41.10 / 45.12)
Which ONE of the following describes what should be used in the case of a large Hydrogen leak to maintain Hydrogen concentration below the lower flammability limit in accordance with OP/1/A/1106/017 (Hydrogen System)?
A. CO2 B. Water C. Halon D. Foam fire retardant Friday, October 04, 2013 Page 109 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 55 55 B General Discussion Answer A Discussion Incorrect. Plausible since CO2 is a common agent used in fire prevention/extinguishing and the addition of CO2 would decrease the concentration of Hydrogen.
Answer B Discussion Correct. L&P 2.4 of 1106/017 says that in case of large Hydrogen leaks, water flow should be admitted to the leak to disperse the gas.
Answer C Discussion Incorrect. Plausible since Halon is a commonly used fire suppression agent and if used it would dilute the concentration of H2 in air.
Addtionally, Halon is an extinguishing agent that is used on site (simulator areas, document control, etc.)
Answer D Discussion Incorrect. Plausible since foam fire retardant is used to prevent fires during flammable liquid spills.
Basis for meeting the KA Requires knowledge of industrial safety procedure directed by procedure to mitigate effects of a large Hydrogen leak.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2009B Q67 Development References Student References Provided SSS-AGS Obj R15 SSS-AGS GEN2.1 2.1.26 - GENERIC - Conduct of Operations Conduct of Operations Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (CFR: 41.10 / 45.12) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 110 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.1 2.1.34 - GENERIC - Conduct of Operations 56 56 C Conduct of Operations Knowledge of primary and secondary plant chemistry limits. (CFR: 41.10 / 43.5 / 45.12)
Which ONE of the following is the LOWER limit on RCS activity that would require entry into AP/21 (RCS Activity)?
A. Xe-133 = 0.25 µCi/gm B. Xe-133 = 1.0 µCi/gm C. DEI = 0.25 µCi/gm D. DEI = 1.0 µCi/gm Friday, October 04, 2013 Page 111 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 56 56 C General Discussion Answer A Discussion Incorrect. Plausible since this would be correct if the element were I-131. Xe is plausible since it is repeatedly referenced in AP/21.
Answer B Discussion Incorrect. Plausible since 1.0 is a threshold value referenced several times in AP/21. Also, Xe is plausible since it is referenced repeatedly in AP/21.
Answer C Discussion Correct 0.25 micro Ci/gm is the threshold for entry level into AP/21.
Answer D Discussion incorrect. Plausible since 1.0 is a threshold value referenced several times in AP/21.
Basis for meeting the KA Requires knowledge of chemistry limit that is the entry condition for AP/21 (RCS Activity).
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EAP-APG Obj R9 AP/21 GEN2.1 2.1.34 - GENERIC - Conduct of Operations Conduct of Operations Knowledge of primary and secondary plant chemistry limits. (CFR: 41.10 / 43.5 / 45.12) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 112 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.1 2.1.37 - GENERIC - Conduct of Operations 57 57 C Conduct of Operations Knowledge of procedures, guidelines, or limitations associated with reactivity management. (CFR: 41.1 / 43.6 / 45.6)
Which ONE of the following activities complies with guidance contained in SOMP 1-2 (Reactivity Management)?
A. Manual rod withdrawal during a Feedwater transient to stop a temperature decrease caused by an instrument failure B. Manually increasing Feedwater flow to stop an RCS pressure increase caused by an RCS temperature increase C. Manually raising one Loop FDW demand while lowering the other Loop FDW demand to control Tcold following an RCP trip D. Manually increasing turbine demand to adjust RCS temperature Friday, October 04, 2013 Page 113 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 57 57 C General Discussion Answer A Discussion Incorrect: Manual rod withdrawal is not permitted.
Answer B Discussion Incorrect: Increasing FDW Flow is not permitted.
Answer C Discussion Correct: The sequence given is permitted as there is no intent to raise FDW Flow.
Answer D Discussion Incorrect: Increase in Turbine demand is only allowed if intent is to stabilize Turbine Header Pressure not to reduce pressure or RCS temperature.
Basis for meeting the KA Demonstrates an understanding of guidelines associated with reactivity management provided in SOMP 1-02.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2009 Q68 Development References Student References Provided ADM-OMP Obj R23 TA-PTR Obj R1 TA-PTR GEN2.1 2.1.37 - GENERIC - Conduct of Operations Conduct of Operations Knowledge of procedures, guidelines, or limitations associated with reactivity management. (CFR: 41.1 / 43.6 / 45.6) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 114 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.2 2.2.13 - GENERIC - Equipment Control 58 58 A Equipment Control Knowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13)
Which ONE of the following tags would be used ONLY for configuration control of 1HP-409 in accordance with NSD-500 (Red Tags/Configuration Control Tags)?
A. White Tag B. MORT Tag C. OORT Tag D. CORT Tag Friday, October 04, 2013 Page 115 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 58 58 A General Discussion Answer A Discussion Correct. White tags are used for configuration control of components and systems Answer B Discussion Incorrect. Plausible since MORT tags are Safety tags used in the field during equipment maintenance and are addressed by NSD 500. MORT tags are used when Chenmistry or Operations assigans a component which their group has ownership to Maintenance.
Answer C Discussion Incorrect. Plausible since OORT tags are Safety tags used in the field during equipment maintenance and are addressed by NSD 500. OORT tags are used to re-assign operations control of a component that is owned by Chemistry to Operations. Since the component in question is owned by Operations, an OORT tag is plausible since it begins with an "O" (for Operations).
Answer D Discussion Incorrect. Plausible since CORT tags are Safety tags that are use in the field during maintenance activities and are addressed by NSD 500. A CORT tag would be used to re-assign operational control of a component owned by operations to Chemistry. It is plausible to believe that a CORT tag is for configuration control since CORT tags are used on components where Operations is the Owner Control Group and Operations is the owner control group for HP-409.
Basis for meeting the KA Requires generic knowledge of the tagging process defined by NSD 500 Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK ILT40 Q70 Development References Student References Provided Obj ADM-SD R6 NSD 500 GEN2.2 2.2.13 - GENERIC - Equipment Control Equipment Control Knowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 116 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.2 2.2.12 - GENERIC - Equipment Control 59 59 B Equipment Control Knowledge of surveillance procedures. (CFR: 41.10 / 45.13)
Given the following Unit 1 condition:
Reactor in MODE 1 Which ONE of the following is the MINIMUM Pressurizer level (inches) that would require declaring Tech Spec 3.4.9 (Pressurizer) LCO NOT met in accordance with PT/1/A/0600/001 (Periodic Instrument Surveillance)?
A. 240 B. 260 C. 285 D. 340 Friday, October 04, 2013 Page 117 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 59 59 B General Discussion Answer A Discussion Incorrect. Plausible since this value is below the TS required value of 285 therefore it is plausible to believe it to be an instrument corrected value. Also, 240 inches is the hi level alarm setpoint for the OAC alarm. Additional plausibility from the fact that this is a fairly common level value however it is the SG level required for natural circ when on EFDW.
Answer B Discussion Correct. PT.600/01 corrects the TS required 285" for allowable instrument error and uses 260" as the threshold value.
Answer C Discussion Incorrect. Plausible since this is the analytical value provided in Tech Spec 3.4.9 for maximum level.
Answer D Discussion Incorrect. Plausible since this is a value associated with the pressurizer however this is the maximum Pzr level allowed for RCP restart with abnormal containment conditions.
Basis for meeting the KA Required knowledge of PT/600/01 surveillance requirements for Pzr level.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Adm-ITS Obj R8 TS 3.4.9 PT/600/01 GEN2.2 2.2.12 - GENERIC - Equipment Control Equipment Control Knowledge of surveillance procedures. (CFR: 41.10 / 45.13) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 118 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.2 2.2.37 - GENERIC - Equipment Control 60 60 D Equipment Control Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12)
Given the following Unit 1 conditions:
Reactor trip due to loss of both Main FDW pumps Instrument Air pressure = 0 psig Auxiliary Instrument Air pressure= 0 psig Which ONE of the following describes the status of 1FDW-315 and 1FDW-316?
A. Available for Manual operation ONLY once the air supply was lost B. Will be available for Automatic operation for a MINIMUM of 30 minutes from the loss of air supply C. Will be available for Automatic operation for a MINIMUM of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from the loss of air supply D. Will be available for Automatic operation for a MINIMUM of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from the loss of air supply Friday, October 04, 2013 Page 119 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 60 60 D General Discussion Answer A Discussion Incorrect. Plausible since both IA and AIA have been lost however there is a N2 backup supply to the valves.
Answer B Discussion Incorrect. Plausible since there is N2 backup to both valves and 30 minutes is a fairly common completion time for Time Critical Actions at Oconee. That makes it plausible that the N2 backup would only be good for 30 minutes minimum.
Answer C Discussion Incorrect. Plausible since 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is the credited time for battery backup following a loss of AC power. Additional plausibility comes from the "AC Independence" of EFDW and the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> battery backup.
Answer D Discussion Correct. The valves have a N2 backup credited for a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> backup following a loss of air supply.
Basis for meeting the KA Requires the ability to determine how long auto operation of FDW-315/316 is available following a loss of IA & AIA.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2007 Audit Q 54 Development References Student References Provided CF-EF Obj R39 CF-EF GEN2.2 2.2.37 - GENERIC - Equipment Control Equipment Control Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 120 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.3 2.3.11 - GENERIC - Radiation Control 61 61 C Radiation Control Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10)
Given the following Unit 2 conditions:
Initial conditions:
Time = 1200 RCS temperature = 92°F stable RB Purge in progress 2RIA-45 HIGH alarm setpoint = 1520 cpm 2RIA-45 = 1342 cpm stable Current conditions:
Time = 1205 2RIA-45 = 1520 cpm increasing Which ONE of the following describes:
- 1) ALL valves that will CLOSE?
- 2) 2RIA-46 reading (cpm) at time = 1200?
- 2. 1342 Friday, October 04, 2013 Page 121 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 61 61 C General Discussion Answer A Discussion Incorrect. First part is plausible since these are all valves that are in the RB purge flowpath and they do all have a function to automatically close however the signal that closes all 6 of the valves originates as part of RB isolation on ES actuation. Second part is correct.
Answer B Discussion Incorrect. First part is plausible since these are all valves that are in the RB purge flowpath and they do all have a function to automatically close however the signal that closes all 6 of the valves originates as part of RB isolation on ES actuation. Second part is plausible since 2RIA-45 and 2RIA-46 are actually monitoring the same activity. Under normal operating circumstances, when RIA 45/46 are both in service, the RIA 45 readings would increase to the high alarm setpoint and actuate the interlock. RIA 46 would continue to read zero on the RIA view screens while all this occurs. At this point, the interlock is NOT actuated by RIA 46. RIA 46 could actually be seeing some value (less than the 'switchover acceptance range setpoint'). Only when the 'switchover acceptance range setpoint' is reached will the RIA indicate a value.
Answer C Discussion Correct. 2RIA-45 HIGH alarm will close 2PR-2-5. Under normal operating circumstances, when RIA 45/46 are both in service, the RIA 45 readings would increase to the high alarm setpoint and actuate the interlock. RIA 46 would continue to read zero on the RIA view screens while all this occurs. At this point, the interlock is NOT actuated by RIA 46. RIA 46 could actually be seeing some value (less than the 'switchover acceptance range setpoint'). Only when the 'switchover acceptance range setpoint' is reached will the RIA indicate a value.
Answer D Discussion Incorrect. First part is correct. Second part is plausible since 2RIA-45 and 2RIA-46 are actually monitoring the same activity. Under normal operating circumstances, when RIA 45/46 are both in service, the RIA 45 readings would increase to the high alarm setpoint and actuate the interlock. RIA 46 would continue to read zero on the RIA view screens while all this occurs. At this point, the interlock is NOT actuated by RIA
- 46. RIA 46 could actually be seeing some value (less than the 'switchover acceptance range setpoint'). Only when the 'switchover acceptance range setpoint' is reached will the RIA indicate a value.
Basis for meeting the KA Requires demonstrating the ability to control radiation releases as a result of the RB Purge operation by demonstrating an understanding of how the associated RIA's and auto purge termination are designed to function to control releases.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided RAD-RIA Obj R2 RAD-RIA GEN2.3 2.3.11 - GENERIC - Radiation Control Radiation Control Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 122 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.3 2.3.7 - GENERIC - Radiation Control 62 62 A Radiation Control Ability to comply with radiation work permit requirements during normal orabnormal conditions. (CFR: 41.12 / 45.10)
Given the following plant conditions:
Venting the 1C LPI Pump in progress using the following RWP information:
o Dose Alarm : 25 mrem o Dose Rate Alarm: 200 mrem/hr o Dose Alarm: Stop work - Exit Area - Notify RP o Unanticipated Dose Rate Alarm: Stop Work - Exit Area - Notify RP Which ONE of the following states the MAXIMUM time work can continue before complying with the RWP will require exiting the area?
SEE PLAN VIEW PROVIDED Do NOT consider dose received while traveling to or from the job.
A. 15 minutes B. 30 minutes C. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> D. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Friday, October 04, 2013 Page 123 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 62 62 A General Discussion Answer A Discussion Correct. 15 minutes in a 100 mr/hr field will result in 25 mrem which is the limit provided with the RWP.
Answer B Discussion Incorrect. Plausible since this would be correct for the 1A LPI pump Answer C Discussion Incorrect. Plausible if using the Dose RATE setpoint with the 1C LPI pump Answer D Discussion Incorrect. Plausible if using the Dose RATE setpoint instead of the Dose alarm setpoint Basis for meeting the KA Requires demonstrating the ability to comply with RWP requirements.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Plan View Plan view GEN2.3 2.3.7 - GENERIC - Radiation Control Radiation Control Ability to comply with radiation work permit requirements during normal orabnormal conditions. (CFR: 41.12 / 45.10) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 124 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.4 2.4.22 - GENERIC - Emergency Procedures / Plan 63 63 A Emergency Procedures / Plan Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations. (CFR: 41.7 / 41.10 / 43.5 / 45.12)
Of the two tabs below, the __(1)__ tab of the EOP has a higher priority because
__(2)__.
Which ONE of the following completes the statement above?
A. 1. LOSCM
- 2. as long as the RCS remains subcooled, adequate core cooling is assured.
B. 1. LOSCM
- 2. a Reactor trip with a SGTR results in a direct release path for radionuclides to the environment D. 1. SGTR
- 2. actions to depressurize RCS to minimize SCM during a SGTR is a Time Critical Action that may not otherwise be met Friday, October 04, 2013 Page 125 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 63 63 A General Discussion Answer A Discussion Correct. As long as the RCS remains subcooled, adequate core cooling is assured. As soon as a loss of SCM occurs actions must be taken to ensure adequate core cooling. For this reason the loss of SCM has top priority requiring treatment ahead of other abnormal heat transfer symptoms or SGTR.
Answer B Discussion Incorrect. First part is correct. Second part is plausible since there is much focus on ensuring RCP's are secured within 2 minutes of a LOSCM.
Although it is the intent to get the secured before pump damage occurs, the concern is NOT that they will be unavailable but that there is the possibility of phase separation when RCP trips and therefore potential of core uncovery that is the concern.
Answer C Discussion Incorrect. It is plausible to believe that the SGTR tab would be a higher priority since it there is a direct path for radionuclides to reach the environment anytime the MSRV's are opened with a SGRT in progress. The second part is plausible because it is correct.
Answer D Discussion Incorrect. It is plausible to believe that the SGTR tab would be a higher priority since it there is a direct path for radionuclides to reach the environment anytime the MSRV's are opened with a SGRT in progress. The second part is plausible since there is a 22 minute Time Critical Action to begin depressurizing RCS to reduce SCM.
Basis for meeting the KA Chief Examiner said to ask question based on priority of EOP tabs. This question meets the KA as it requires knowledge of the bases behind the LOSCM tab being the higher priority tab vs. SGTR tab.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-LOSCM Obj R12 EAP-LOSCM EAP-SGTR GEN2.4 2.4.22 - GENERIC - Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations. (CFR: 41.7 / 41.10 / 43.5 / 45.12) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 126 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION GEN2.4 2.4.4 - GENERIC - Emergency Procedures / Plan 64 64 B Emergency Procedures / Plan Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6)
Given the following Unit 1 conditions:
Reactor Power = 70%
Which ONE of the following would require entry into the EOP?
A. Condenser vacuum = 22.3 hg B. 1RIA-59 = 31.4 gpm C. 1B Main FDW pump trips D. 1A1 RCP trips Friday, October 04, 2013 Page 127 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 64 64 B General Discussion Answer A Discussion Incorrect. Plausible since this would be correct if < 21.75". Additionally plausible since this would be low enough to meet entry conditions for the Loss of Vacuum AP.
Answer B Discussion
.Correct. Steam Generator tube leakage of > 25 gpm requires entry into the EOP.
Answer C Discussion Incorrect. Plausible since this could be correct at rated power since it is possible to trip on high RCS pressure if a FDW pump trips at 100%. It could also be correct at power levels below 50% since it it likely only one FDWP would be operating and therefore if it tripped the Rx would trip.
Answer D Discussion Incorrect. Plausible since it would be correct for higher power levels.
Basis for meeting the KA Requires the ability to recognize an abnormal system parameter that is an entry conditions for the Emergency Operating Procedure.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided RAD-RIA Obj R2 GEN2.4 2.4.4 - GENERIC - Emergency Procedures / Plan Emergency Procedures / Plan Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) 401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 128 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS013 K2.01 - Engineered Safety Features Actuation System (ESFAS) 65 65 B Knowledge of bus power supplies to the following: (CFR: 41.7)
ESFAS/safeguards equipment control ...............................
Given the following Unit 1 conditions:
Initial conditions:
Reactor power = 100%
1KVIA Panelboard de-energized Current conditions:
MSLB inside the Reactor Building occurs Lowest RCS pressure = 1137 psig Reactor Building Pressure peaked at 32 psig Which ONE of the following describes ALL ES Actuation Logic Channels that have actuated?
A. 1, 3, 5, 7 B. 2, 4, 6, 8 C. 1, 5, 7 ONLY D. 2, 6, 8 ONLY Friday, October 04, 2013 Page 129 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 65 65 B General Discussion Answer A Discussion Incorrect. KVIA provides power to the odd digital channels. With KVIA de-energized, the Odd channels cannot actuate.
Answer B Discussion Correct. KVIA provides power to the odd digital channels. With KVIA de-energized, the Odd channels cannot actuate. Since RB pressure has exceeded 10 psig, all channels 8 channels would receive an actuation signal however only the odd channels have power and therefore they are all that can actuate.
Answer C Discussion Incorrect. Plausible since RCS pressure has reached the Low RCS pressure setpoint for HPI injection but has not reached the LPI injection setpoint of 550 psig. The misconception that HPI and LPI only actuate from RCS pressure rather than from either RCS pressure OR RB pressure would lead to believing that channels 2 and 4 had not yet received an actuation signal. Additionally, the power supplys to the Actuation Logic channels is split based on odd and even channels. Channel 1&2 RCS pressure setpoint has already been reached therefore under the misconception that ES channels 3 and 4 (LPI) only actuate on low RCS pressure this is plausible.
Answer D Discussion Incorrect. Plausible since RCS pressure has reached the Low RCS pressure setpoint for HPI injection but has not reached the LPI injection setpoint of 550 psig. The misconception that HPI and LPI only actuate from RCS pressure rather than from either RCS pressure OR RB pressure would lead to believing that channels 2 and 4 had not yet received an actuation signal. Additionally, the power supplys to the Actuation Logic channels is split based on odd and even channels. Channel 1&2 RCS pressure setpoint has already been reached therefore under the misconception that ES channels 3 and 4 (LPI) only actuate on low RCS pressure this is plausible.
Basis for meeting the KA Requires knowledge of ES powers supplies for the equipment that controls actuation of the components.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory MODIFIED ILT40 Q40 Development References Student References Provided Obj IC-ES R2, R26 IC-ES SYS013 K2.01 - Engineered Safety Features Actuation System (ESFAS)
Knowledge of bus power supplies to the following: (CFR: 41.7)
ESFAS/safeguards equipment control ...............................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 130 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS022 A4.04 - Containment Cooling System (CCS) 66 66 B Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
Valves in the CCS ................................................
Given the following Unit 1 conditions:
Initial conditions:
Time = 1200 Reactor Power = 100%
1A MSLB inside the Reactor Building Current conditions:
Time = 1201 Reactor Building Pressure = 3 psig increasing Which ONE of the following describes the operation of 1LPSW-18?
A. It is NORMALLY fully open however it will receive a signal to open from ES-5 at 1201 B. It is NORMALLY throttled and will go fully open when it receives a signal to open from ES-5 at 1201 C. It is NORMALLY fully open however it will receive a signal to open from ES-5 at 1204 D. It is NORMALLY throttled and will go fully open when it receives a signal to open from ES-5 at 1204 Friday, October 04, 2013 Page 131 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 66 66 B General Discussion Answer A Discussion Incorrect. The valve being fully open at 1200 is plausible since the associated RBCU inlet valve (1LPSW-16) normal position is fully open. ES-5 does send an open signal to 1LPSW-18 at 1201.
Answer B Discussion Correct. The RBCU Cooler outlet valves are throttled during normal operation and go fully open when an ES signal is received. Since ES 5&6 actuate at 3 psig RB pressure, 1LPSW-18 will receive its open signal at 1201.
Answer C Discussion Incorrect. The valve being fully open at 1200 is plausible since the associated RBCU inlet valve (1LPSW-16) normal position is fully open. Not receiving an open signal until 1204 is plausible since the start signal to the RBCU's is delayed for 3 minutes following ES 5&6 to ensure adequate bus voltages. Since the RBCU does not receive a start signal until 1204 it is plausible to believe that the associated LPSW outlet valve does not receive a signal to open until the RBCU has received a signal to start.
Answer D Discussion Incorrect. The valve is throttled at 1200. Not receiving an open signal until 1204 is plausible since the start signal to the RBCU's is delayed for 3 minutes following ES 5&6 to ensure adequate bus voltages. Since the RBCU does not receive a start signal until 1204 it is plausible to believe that the associated LPSW outlet valve does not receive a signal to open until the RBCU has received a signal to start.
Basis for meeting the KA Requires the ability to monitor Containment Cooling System valves (LPSW cooling water to RBCU's) for proper operation following an ES signal.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided PNS-RBC Obj R1, R6 PNS-RBC SYS022 A4.04 - Containment Cooling System (CCS)
Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
Valves in the CCS ................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 132 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS022 K2.01 - Containment Cooling System (CCS) 67 67 B Knowledge of power supplies to the following: (CFR: 41.7)
Containment cooling fans .........................................
Given the following Unit 1 conditions:
Initial conditions:
Reactor power = 50%
Current conditions:
LBLOCA occurs 1TD de-energized 1B RBCU switch in OFF Which ONE of the following describes the status of the below listed Reactor Building Cooling Units five (5) minutes after ES actuates?
ASSUME NO OPERATOR ACTIONS 1B RBCU 1C RBCU A. LOW LOW B. LOW OFF C. OFF LOW D. OFF OFF Friday, October 04, 2013 Page 133 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 67 67 B General Discussion Answer A Discussion Incorrect: Plausible as the RBCU power supplies are not sequenced such that the letter designator follows the power supply arrangement. If 1A RBCU fan is applied to TD bus this choice would be plausible.
Answer B Discussion Correct. The 1B RBCU will be in Low even thought its control board switch is in OFF. 1C RBCU is powered from 1TD power string therefore it would not have power available.
Answer C Discussion Incorrect. Plausible to believe that 1TD is power supply to 1B RBCU which is the misconception that makes this answer plausible.
Answer D Discussion Incorrect. Plausible under the assumption that the 1B will not start since its switch is in OFF. 1TD would be OFF since it is de-energized.
Basis for meeting the KA Requires knowledge of power supplies to Reactor Building Cooling Units (RBCUs)
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED 2009 Q38 Development References Student References Provided PNS-RBC, Obj R1 PNS RBC ES Power Supply table SYS022 K2.01 - Containment Cooling System (CCS)
Knowledge of power supplies to the following: (CFR: 41.7)
Containment cooling fans .........................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 134 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS026 A3.02 - Containment Spray System (CSS) 68 68 B Ability to monitor automatic operation of the CSS, including: (CFR: 41.7 / 45.5)
Verification that cooling water is supplied to the containment spray heat exchanger .............................................
Given the following plant conditions:
Unit 2 Reactor Power = 100%
SBLOCA has occurred on Unit 1 Reactor Building Pressure = 11.2 psig slowly decreasing Which ONE of the following describes the actions directed (if any) by Enclosure 5.1 (ES Actuation) to ensure the required LPSW flow exists in the 1A LPI cooler?
A. None B. Place 1LPSW-251 in Failed Open AND fully open 1LPSW-4 C. Place 1LPSW-251 in Failed Open AND Throttle LPSW flow to approximately 3000 gpm using 1LPSW-4 D. Place 1LPSW-251 in Failed Open AND Throttle LPSW flow to approximately 5200 gpm using 1LPSW-4 Friday, October 04, 2013 Page 135 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 68 68 B General Discussion Answer A Discussion Incorrect. Plausible since there are no LPSW failures and LPSW-251 is designed to automatically control LPSW flow, Additionally plausible since the valve is maintained in Auto with a setpoint of 3000 gpm.
Answer B Discussion Correct. With all LPSW pumps operating, Encl. 5.1 directs placing LPSW-251 and 252 in "Failed Open" and the fully opening 1LPSW-4 & 5.
Answer C Discussion Incorrect. Plausible since these actions are taken at other times based on component failures following ES actuation. Additional plausibility based on the fact that 3000 gpm is the setpoint that is normally maintained on LPSW-251 and 252.
Answer D Discussion Incorrect. Placing LPSW-251 in failed open and throttling with LPSW-4 is plausible becasuse those actions are directed as a result of component failures following ES actuation. The flow rate is plausible since it is where LPSW-251 will auto control LPSW flow following a condition where flow exceeds 5900 gpm.
Basis for meeting the KA Chief Examiner said that using LPSW flow to LPI coolers would be sufficient to match KA since ONS does not have coolers specific to the RBS system. This question d Demonstrates the ability to verify cooling water to the LPI coolers by displaying knowledge of the proper actions required to ensure appropriate LPSW flow under specific plant conditions that require RBS flow.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-LOSCM Obj R28 EAP-LOSCM, SSS-LPW EOP Encl 5.1 & 5.12 SYS026 A3.02 - Containment Spray System (CSS)
Ability to monitor automatic operation of the CSS, including: (CFR: 41.7 / 45.5)
Verification that cooling water is supplied to the containment spray heat exchanger .............................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 136 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS039 A3.02 - Main and Reheat Steam System (MRSS) 69 69 D Ability to monitor automatic operation of the MRSS, including : (CFR: 41.5 / 45.5)
Isolation of the MRSS ............................................
Given the following Unit 1 conditions:
Initial conditions:
Reactor power = 100%
Loss of offsite power occurs Current conditions:
Main Feeder Buses remain de-energized
- 1) The position of 1MS-112 (SSRH Control) is __(1)__.
- 2) 1MS-77 (MS to MSRH) __(2)__ be operated from the control room switch.
Which ONE of the following completes the statements above?
A. 1. open
- 2. can B. 1. closed
- 2. can C. 1. open
- 2. can NOT D. 1. closed
- 2. can NOT Friday, October 04, 2013 Page 137of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 69 69 D General Discussion Answer A Discussion Incorrect: Plausible since 1MS-112 is normally open at 100% power and it would be logical to assume that the valve would not operate with no AC power. Second part is plausible because other electric valves can be operated from the control room with the MFBs de-energized (Ex..CCW-8).
Answer B Discussion Incorrect: Plausible since 1MS-112 is normally open at 100% power and it would be logical to assume that the valve would not operate with no AC power. Second part is correct.
Answer C Discussion Incorrect: First part is correct. Second part is plausible because other valves can be operated from the control room with the MFBs de-energized (Ex. CCW-8.).
Answer D Discussion Correct: 1MS-112 will close on a loss of power due to IA porting off. 1MS-77 is an electric valve which cannot be operated from its control room switch.
Basis for meeting the KA Question requires knowledge of automatic actions (Isolation) of the MSRs following a LOOP.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2009B Q8 Development References Student References Provided STG-MSR Obj R18 STG-MSR SYS039 A3.02 - Main and Reheat Steam System (MRSS)
Ability to monitor automatic operation of the MRSS, including : (CFR: 41.5 / 45.5)
Isolation of the MRSS ............................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 138 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS039 K5.08 - Main and Reheat Steam System (MRSS) 70 70 A Knowledge of the operational implications of the following concepts as the apply to the MRSS: (CFR: 441.5 / 45.7)
Effect of steam removal on reactivity ...............................
Given the following Unit 1 conditions:
Reactor Power = 50%
1A Turbine Bypass Valve fails OPEN Which ONE of the following describes the plant response?
ASSUME NO OPERATOR ACTIONS Reactor power will...
A. Increase then return to pre-transient level.
B. Increase and stabilize at a higher power level.
C. Decrease then return to pre-transient level.
D. Decrease and stabilize at a lower power level.
Friday, October 04, 2013 Page 139 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 70 70 A General Discussion Answer A Discussion Correct. With ICS in automatic the failed open TBV will add positive reactivity due to the initial cooldown. Since ICS is maintaining Core Thermal Power at setpoint it will see the increase in CTP and reduce FDW and Reactor to bring CTP back to setpoint. The end result would be CTP returning to setpoint and the steam being lost out of the TBV would result in lower MW production.
Answer B Discussion Incorrect. Plausible since it would be correct if ICS were maintaining Megawatts instead of CTP. It is plausible to believe that is the case since prior to our ICS upgrade that is the way ICS worked.
Answer C Discussion Incorrect. Plausible since it would be correct if moderator temperature coefficient were positive. It is possible to believe that the MTC is positive since reactor power is not at 100% and prior to our 24 month cores we could have a positive MTC during the early stages of our initial startups following a refueling outage.
Answer D Discussion Incorrect. Plausible since it would be correct if ICS were tracking Megawatts which it does under different conditons. Also plausible to believe power decreases since it would be correct for a positive MTC.
Basis for meeting the KA Requires knowledge of the operational implications of the effect of increased steam removal on reactivity. This knowledge is required to be able to predict the response of the ICS system to an increase in steam flow.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided SAE-L226 Obj R9 SAE-L226 SYS039 K5.08 - Main and Reheat Steam System (MRSS)
Knowledge of the operational implications of the following concepts as the apply to the MRSS: (CFR: 441.5 / 45.7)
Effect of steam removal on reactivity ...............................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 140 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS059 A1.03 - Main Feedwater (MFW) System 71 71 B Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW controls including: (CFR: 41.5 / 45.5)
Power level restrictions for operation of MFW pumps and valves. .......
Given the following Unit 1 conditions:
Reactor Power = 80% stable ICS in Manual 1B Main Feedwater Pump trips Which ONE of the following is the MAXIMUM power level allowed in accordance with AP/1 (Plant Runback).
A. 74%
B. 65%
C. 60%
D. 55%
Friday, October 04, 2013 Page 141of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 71 71 B General Discussion Answer A Discussion Incorrect. Plausible since this would be correct for a RCP trip Answer B Discussion Correct, per AP/1 initiate a runback to < or = 65%
Answer C Discussion AP/1.Incorrect. Plausible since this is the power level of a CRD out inhibit put in place following an asymmetric rod runback by ICS as well as it is the allowable thermal power during a dropped rod with 4 RCP's and therefore is discussed as a limit several times in AP/1.
Answer D Discussion Incorrect. Plausible since this would be correct for a dropped control rod.
Basis for meeting the KA Requires the ability to predict the maximum power level allowed following a loss of one of the two operating Main Feedwater Pumps. With ICS in Manual the restrictions on power level would be associated with manual operations of the MFW controls.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EAP-APG Obj R9 AP/1 SYS059 A1.03 - Main Feedwater (MFW) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW controls including: (CFR: 41.5 / 45.5)
Power level restrictions for operation of MFW pumps and valves. .......
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 142 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS059 K1.04 - Main Feedwater (MFW) System 72 72 D Knowledge of the physical connections and/or cause-effect relationships between the MFW and the following systems: (CFR: 41.2 to 41.9 /
45.7 to 45.8)
S/GS water level control system ....................................
Given the following Unit 3 conditions:
Initial conditions:
Reactor tripped from 35% power due to 1TA lockout 3A Main FDW pump operating 3FDW-35 & 3FDW-44 (3A and 3B Startup FDW Control) in MANUAL 3A and 3B SG levels = 38" SU and stable Current conditions:
3FDW-35 & 44 are placed in Automatic Which ONE of the following describes the response of 3FDW-35 & 44?
A. Travel open to increase SG levels to 240" XSUR.
B. Travel open to increase SG levels to 50% on Operating level.
C. Travel closed to decrease SG level to 30" on XSUR.
D. Travel closed to decrease SG level to 25" on SU level.
Friday, October 04, 2013 Page 143 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 72 72 D General Discussion Answer A Discussion Incorrect. Plausible since this would be correct if on EFDW and RCP's were off. 1TA lockout makes the 240" level plausible since it results in loss of 2 RCP's.
Answer B Discussion Incorrect. Plausible since this would be correct if RCP's were off. 1TA lockout makes the level plausible since it results in loss of 2 RCP's.
Answer C Discussion Incorrect. Plausible since this would be correct if on EFDW.
Answer D Discussion Correct. On Main FDW with RCP's operating, SG levels would be controlled at 25" on startup indication following a Rx trip.
Basis for meeting the KA Requires knowledge of the relationship between SG level control system and Main FDW.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED ILT40 Q28 Development References Student References Provided CF-FDW Obj R28 CF-FDW SYS059 K1.04 - Main Feedwater (MFW) System Knowledge of the physical connections and/or cause-effect relationships between the MFW and the following systems: (CFR: 41.2 to 41.9 /
45.7 to 45.8)
S/GS water level control system ....................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 144 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS061 K5.05 - Auxiliary / Emergency Feedwater (AFW) System 73 73 A Knowledge of the operational implications of the following concepts as the apply to the AFW: (CFR: 41.5 / 45.7)
Feed line voiding and water hammer .................................
Which ONE of the following describes the:
- 2) method used to determine if Main Feedwater backleakage into the EFDW discharge piping is occurring?
A. 1. Vapor binding of the EFDW pumps
- 2. locally monitoring EFDW pump discharge piping for increasing temperature B. 1. Vapor binding of the EFDW pumps
- 2. Monitoring EFDW temperature OAC points for increasing temperature C. 1. Overpressurizing the EFDW system piping
- 2. locally monitoring EFDW pump discharge piping for increasing temperature D. 1. Overpressurizing the EFDW system piping
- 2. Monitoring EFDW temperature OAC points for increasing temperature Friday, October 04, 2013 Page 145 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 73 73 A General Discussion Answer A Discussion Correct. Back leakage from the MFDW system can result in vapor binding of the EFDWP's, this phenomenon has occurred here at ONS. At Oconee, the ONLY means of detecting the back leakage is by locally monitoring the EFDWP discharge piping by touch or with a pyrometer.
Answer B Discussion Incorrect. First part is correct. Second part is plausible since EFDW temperatures are a significant concern here at ONS and EFDW suction side temperatures are carefully monitored on the OAC to ensure heat removal capacity credited to EFDW in the FSAR therefore it would be plausible to believe the same process of monitoring EFDW temps would be available for the discharge side of the EFDW pumps.
Answer C Discussion Incorrect. First part is plausible since there are systems where we are concerned with leakage through check valves resulting in over pressurizing system piping (specifically the LPI system). In fact it is such a concern with LPI that there is an Inter system LOCA test done just to verify leakage is within limits. Given the focus on intersystem leakage it would be plausible to believe that over pressurizing piping would be a concern. Even if the candidate did not believe it could over pressurize the EFDW discharge piping it would be still be plausible to believe that back leakage through the EFDW pump could over pressurize the suction side piping since it is not rated for SG pressure and thereby making this choice plausible. Second part is correct.
Answer D Discussion Incorrect. First part is plausible since there are systems where we are concerned with leakage through check valves resulting in over pressurizing system piping (specifically the LPI system). In fact it is such a concern with LPI that there is an Inter system LOCA test done just to verify leakage is within limits. Given the focus on intersystem leakage it would be plausible to believe that over pressurizing piping would be a concern. Even if the candidate did not believe it could over pressurize the suction side piping since it is not rated for SG pressure and thereby making this choice plausible.. Second part is plausible since EFDW temperatures are a significant concern here at ONS and EFDW suction side temperatures are carefully monitored on the OAC to ensure heat removal capacity credited to EFDW in the FSAR therefore it would be plausible to believe the same process of monitoring EFDW temps would be available for the discharge side of the EFDW pumps.
Basis for meeting the KA This KA requires knowledge of the operational implications of voiding in the EFDW lines as a result of back leakage from the Main Feedwater system. It would also be an operational implication to understand how to detect the issue of back leakage which is what leads to the voiding.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided CF-EF R51, R52 CF-EF SYS061 K5.05 - Auxiliary / Emergency Feedwater (AFW) System Knowledge of the operational implications of the following concepts as the apply to the AFW: (CFR: 41.5 / 45.7)
Feed line voiding and water hammer .................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 146 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS062 K1.02 - AC Electrical Distribution System 74 74 A Knowledge of the physical connections and/or cause-effect relationships between the ac distribution sys- tem and the following systems : (CFR:
41.2 to 41.9)
ED/G ..........................................................
Given the following plant conditions:
No Keowee Units are operating ACB-3 closed
- 1) KHU 1X switchgear is being powered from __ (1) __.
- 2) Keowee control power will be available for a MINIMUM of approximately __ (2) __
hour(s) following a loss of ALL AC power.
Which ONE of the following completes the statements above?
A. 1. 1TC
- 2. one B. 1. 1TC
- 2. four C. 1. the 230 KV switchyard
- 2. one D. 1. the 230 KV switchyard
- 2. four Friday, October 04, 2013 Page 147 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 74 74 A General Discussion Keowee Unit 2 is the underground unit which is determined by ACB-4 being closed. If Keowee Unit 1 is operating to the grid and receives an Emergency Start signal, it will separate from the grid by opening ACB-1 and then operate in standby until needed or manually shut down.
Answer A Discussion Correct.
With KHU-1 aligned to the underground its auxiliaries are supplied from the CX transformer which gets its power from 1TC 4160V switchgear.
The Keowee batteries will last about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Answer B Discussion Incorrect.
First part is correct.
Second part is plausible because 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is a common TS completion time and since the Keowee batteries are required by TS, it may be confused with how long the battery will last. Additionally, 4 hrs is the length of time N2 backup supply to various components will last following a loss of IA and since both N2 and Keowee batteries serve as backup on loss of normal energy supplies, it would be plausible to confuse the two.
Answer C Discussion Incorrect.
First part is plausible since it would be correct if ACB-4 were closed.
Second part is incorrect and plausible. The student may assume that the yellow bus is not automatically isolated from the grid when a switchyard isolation occurs.
Answer D Discussion Incorrect.
First part is plausible since it would be correct if ACB-4 were closed.
Second part is plausible because 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is a common TS completion time and since the Keowee batteries are required by TS, it may be confused with how long the battery will last. Additionally, 4 hrs is the length of time N2 backup supply to various components will last following a loss of IA and since both N2 and Keowee batteries serve as backup on loss of normal energy supplies, it would be plausible to confuse the two.
Basis for meeting the KA Since Oconee uses Hydro units for emergency power, this question matches the KA intent by requiring knowledge of the ONS AC distribution systems connection with the KHU electrical systems.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED ILT42 Q13 Development References Student References Provided EL-KHG Obj R22 EL-KHG SYS062 K1.02 - AC Electrical Distribution System Knowledge of the physical connections and/or cause-effect relationships between the ac distribution sys- tem and the following systems : (CFR:
41.2 to 41.9)
ED/G ..........................................................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 148 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION SYS063 K4.01 - DC Electrical Distribution System 75 75 B Knowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
Manual/automatic transfers of control ...............................
Given the following plant conditions:
3CA Battery Charger fails - output voltage = 0 VDC 3CA Battery voltage = 120 VDC 3DCB Bus voltage = 123 VDC Unit 1 DCA/DCB Bus voltage = 125 VDC Unit 2 DCA/DCB Bus voltage = 127 VDC Which ONE of the following will automatically supply power to 3DIA panelboard?
A. 3CA Battery B. Unit 1 DC Bus C. 3DCB Bus D. Unit 2 DC Bus Friday, October 04, 2013 Page 149 of 150
FOR REVIEW ONLY - DO NOT DISTRIBUTE ILT44 ONS RO NRC Examination QUESTION 75 75 B General Discussion Answer A Discussion Incorrect. Plausible because the 3CA battery will supply power to the bus if its voltage is higher than the backup source. In this case it is not.
Unit 1's voltage is higher.
Answer B Discussion Correct, The voltage from Unit 1 is higher than the 3CA battery voltage since Unit 1 is being supplied from the charger, so Unit 1 will supply power.
Answer C Discussion Incorrect. For the Vital DC system, the 3DCB bus is not aligned to the 3DCA bus. Plausible because 3DCB Bus is aligned to backup the essential inverters Answer D Discussion Incorrect. Unit 2's DC Bus is not connected to Unit 3. Plausible because Unit 2 is next to Unit 3.
Basis for meeting the KA Requires knowledge of a design feature which provides for automatic transfer of control power to components powered from the 3DIA panelboard.
Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ILT41 Q47 Development References Student References Provided EL-DCD Obj R4 EL-DCD SYS063 K4.01 - DC Electrical Distribution System Knowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
Manual/automatic transfers of control ...............................
401-9 Comments: Remarks/Status Friday, October 04, 2013 Page 150 of 150
Enclosure 5.1 AP/1/A/1700/034 Generator Capability Page 1 of 3 Curve
QUESTION 62 Room 61 LPI & RBS Pumps Survey # M-021506-17 Date/Time Today 0412 ROOM 61 LPI AND RB PUMPS N
AHU 1-6 *50 LEWA
+30 112 1A 2 35 RBS Significant Dose Contributor STAIRS SUMP PUMPS 88 1 58
+1245
+975 3 1C LPI 1A LPI 95100 76 50 Summary of Highest Readings Comments: CONTACT RP REGARDING ANY ATTEMPTS TO CLEAN LPI ROOM SUMP Smears Air Samples & Wipes
- 1) 554 DPM/100 cm2 /
- 2) 485 DPM/100 cm2 /
- 3) 1453 DPM/100 cm2 /
Symbol Legend (for example only) Type: Job Coverage Dose Rate
- 150 Contact Reading HS-50 Hot Spot RWP: 5036
+75 30 cm Reading RCA Posting Reactor Power = 100%
20 General Area Drip Bag 15 Smear 15 Air Sample 15 Wipe Unless otherwise noted, dose rates in mrem/hr.
Surveyor: W. Walters Approved by: N. Wriston, Date: Today