ML14084A035

Iinitial Exam 2013-302 Draft RO Written Exam

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)

Text

B 1 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 PumpsONLY trip on Unit 1 B. Both Main Feedwater pumpsONLY trip on Unit 2 C. Loss of Offsite Power on Unit 1 D. Loss of Offsite Power on Unit 2 EPE007 EK1.06 - Reactor TripKnowledge 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 ............................

B 1General DiscussionAnswer A DiscussionIncorrect. 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 - FSAR1.~Loss of Main Feedwater Highest heat load - decay heat & RCP heat.Requires the highest initial post trip EFDW flow of all analysis, therefore constitutes the design basis transient for post trip EFDW.Answer C DiscussionIncorrect. 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 DiscussionIncorrect. 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

.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References CF-EF Obj R55 CF-EFRule 7401-9 Comments:Remarks/StatusBasis for meeting the KAThis 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 CogBasis for SRO onlyEPE007 EK1.06 - Reactor TripKnowledge 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 ............................

B 2 Given the following Unit 1 conditions: Reactor power = 100% 1RC-66 (PORV) is leaking past its seat Pressurizer temperature = 648 0 F 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 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 .......................................................

B 2General DiscussionAnswer A DiscussionIncorrect. Plausible because this will be the answer if 5 psig is not converted to psia.Answer B DiscussionCORRECT: 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 DiscussionIncorrect: 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/lbAnswer D DiscussionIncorrect: Plausible with the same misconception made at TMI which was assuming constant temperature across the valve due to throttling process.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion SourceILT40 NRC Exam Q#2Student References ProvidedDevelopment ReferencesPNS-PZR R34401-9 Comments:Remarks/StatusBasis for meeting the KA Requires knowledge of pzr vapor space accident (leaking PORV) on tailpipe temp by applying thermodynamic flow characteristics of a leaking valveBasis for Hi CogBasis for SRO onlyAPE008 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 .......................................................

C 3 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 transfe r D. Ensure a secondary side levelsufficient to minimize the consequences of a total loss of feedwater durin g boiler condenser mode heat transfer EPE009 EK2.03 - Small Break LOCAKnowledge of the interrelations between the small break LOCA and the following: (CFR 41.7 / 45.7)S/Gs ...........................................................

C 3General DiscussionAnswer A DiscussionIncorrect: 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 othe r 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 DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source2009B Q26Student References ProvidedDevelopment ReferencesEAP-LOSCM Obj R6EAP-LOSCM Att. 01401-9 Comments:Remarks/StatusBasis for meeting the KABasis for Hi CogBasis for SRO onlyEPE009 EK2.03 - Small Break LOCAKnowledge of the interrelations between the small break LOCA and the following: (CFR 41.7 / 45.7)S/Gs ...........................................................

C 4 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. 1S A-15/A5 (RC PUMP MOTOR 1B1 OIL POT LOW LEVEL) in alarm D. 1S A-6/D5 (PUMP 1B2 CAVITY PRESS HI/LOW) in alarm APE015/017 2.4.45 - Reactor Coolant Pump (RCP) MalfunctionsAPE015/017 GENERICAbility to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

C 4General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. Any oil pot alarm is an entry condition for AP/16.Answer D DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-APG Obj R9EAP-AP/16401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyAPE015/017 2.4.45 - Reactor Coolant Pump (RCP) MalfunctionsAPE015/017 GENERICAbility to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

B 5 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

1) The SRO will direct actions from the __ (1) __ tab of the EOP.

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 EPE011 EA1.14 - Large Break LOCAAbility to operate and monitor the following as they apply to a Large Break LOCA: (CFR 41.7 / 45.5 / 45.6)Subcooling margin monitors .......................................

B 5General DiscussionAnswer A DiscussionIncorrect. First part is correct. Second part is plausible because it would be correct if total LPI flow were less than 3400 gp m.Answer B DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion SourceILT41 Q3Student References ProvidedDevelopment ReferencesEOP-SAEOP Rule 2EOP LOSCM401-9 Comments:Remarks/StatusBasis for meeting the KAAt 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 CogBasis for SRO onlyEPE011 EA1.14 - Large Break LOCAAbility to operate and monitor the following as they apply to a Large Break LOCA: (CFR 41.7 / 45.5 / 45.6)Subcooling margin monitors .......................................

A 6 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 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: (CF R 41.8 / 41.10 / 45.3)Loss of RHRS during all modes of operation .........................

A 6General DiscussionAnswer A DiscussionCorrect: 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 DiscussionIncorrect: First part is incorrect but plausible because load shed reset is required for many other components (seeEL-PSL). Second part is correct.Answer C DiscussionIncorrect: 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 DiscussionIncorrect: 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion SourceILT39 Q6Student References ProvidedDevelopment ReferencesEL-PSL Obj R6EL-PSL401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of actions required to restore core decay heat removal following a failure of the LPI/DHR PumpsBasis for Hi CogBasis for SRO onlyAPE025 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 .........................

D 7 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 ali gn an alternate source of seal in jection 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 ..................

D 7General DiscussionAnswer A DiscussionIncorrect: Plausible since the entry conditions for AP/20 are met and the EOP does direct entry into AP's 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 DiscussionIncorrect: 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 AP's in other conditions (Ex. AP/11, AP/25).Answer C DiscussionIncorrect: Plausible since the first part is correct in that AP/25 is directed by IMA's. 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 DiscussionCORRECT: 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.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source 2009A Q6Student References ProvidedDevelopment ReferencesEAP-IMA Obj R6 EAP-IMA401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of reason EOP IMA's direct initiating AP/25 when RCP seal injection and CC have been lostBasis for Hi CogBasis for SRO onlyAPE026 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 ..................

B 8 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 respondin g as expected B. 1. subcooled 2. has failed C. 1. saturated 2. is respondin g as expected D. 1. saturated 2. has failed APE027 AK2.03 - Pressurizer Pressure Control System (PZR PCS) MalfunctionKnowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: (CFR 41.7 / 45.7)Controllers and positioners ........................................

B 8General DiscussionAnswer A DiscussionIncorrect: 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 DiscussionCORRECT: 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 648

ûF (saturation for 2165). Saturation for actual pzr temp of 640

ûF is about 2045 psig therefore Bank 2 should be energized.

2205 psig.Answer C DiscussionIncorrect: 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 DiscussionIncorrect: 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source 2009A Q7Student References ProvidedDevelopment ReferencesPNS-PZR Obj R5, R7, R29PNS-PZR401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of how controllers for Pzr saturation circuit function and the ability to diagnose a malfunction of circuitr y.Basis for Hi CogBasis for SRO onlyAPE027 AK2.03 - Pressurizer Pressure Control System (PZR PCS) MalfunctionKnowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: (CFR 41.7 / 45.7)Controllers and positioners ........................................

A 9 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 SCM's = 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 RCP's 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.

BWE04 EK3.1 - Inadequate Heat TransferKnowledge of the reasons for the following responses as they apply tothe (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 andoperating limitations and reasons for these operating characteristics.

A 9General DiscussionAnswer 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 4The number of operating RCPs should be reduced to one to decrease the heat being added to the RCSAnswer B DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-LOHT Att 4 Obj R3, EAP-LOHT Obj 2EAP LOHT Att 4, EAP-LOHT401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyBWE04 EK3.1 - Inadequate Heat TransferKnowledge of the reasons for the following responses as they apply tothe (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 andoperating limitations and reasons for these operating characteristics.

A 10 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 EPE038 2.2.38 - Steam Generator Tube Rupture (SGTR)

EPE038 GENERICKnowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)

A 10General DiscussionAnswer A DiscussionCorrect. The TS 3.4.13 LCO limit on SG tube leakage is 150 gpd through any one SG.Answer B DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. The TS 3.4.13 LCO limit on identified leakage is 10 gpm. Plausible based on confusing the unidentified and identified leakage limits.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References Admin-ITS Obj R8 TS 3.4.13 PNS-CF401-9 Comments:Remarks/StatusBasis for meeting the KARequired knowledge of limitations on various forms of RCS Operational Leakage established in Tech Spec 3.4.13.Basis for Hi CogBasis for SRO onlyEPE038 2.2.38 - Steam Generator Tube Rupture (SGTR)

EPE038 GENERICKnowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)

A 11 Given the following Unit 1 conditions: ALL sources of feedwater have been lost Rule 4 (Initiation of HPI Forced Cooling) is complete with outstanding IAAT's 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 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 ...........................

A 11General DiscussionAnswer A DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-LOHT Obj R28EAP-LOHT Att. 4401-9 Comments:Remarks/StatusBasis for meeting the KARequires demonstrating the ability to monitor HPI flow parameters under conditions where all FDW has occurred.Basis for Hi CogBasis for SRO onlyAPE054 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 ...........................

C 12 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 MFB's?

A. Automatically through ACB-3 B. Automaticall y throu gh SL1 and SL2 C. Manually through ACB-3 D. Manually through SL1 and SL2 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 .................................

C 12General DiscussionAnswer A DiscussionIncorrect. Plausible since Zone Overlap protection will automatically close ACB-3 under certain conditions.Answer B DiscussionIncorrect. Plausible since this would be correct if the SBB's were already energized from CT-5.Answer C DiscussionCorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References EAP-BO Obj R6EAP-BOEAP-BO Att 1401-9 Comments:Remarks/StatusBasis for meeting the KARequires determining the actions directed by Encl. 5.38 to restore power to MFB's following a blackout.Basis for Hi CogBasis for SRO onlyEPE055 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 .................................

D 13 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 APE056 AA2.43 - Loss of Offsite PowerAbility 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 .......................................

D 13General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. Incorrect. First part is plausible since it is always true if Rx power is > 70%. Second part is correct.Answer C DiscussionIncorrect. 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 DiscussionCorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesIC-RPS Obj. R3, 4, 23, SAE-L216 Obj R3 IC-RPSSAE-L216401-9 Comments:Remarks/StatusBasis for meeting the KAQuestion required the ability to determine if the turbine has tripped following a loss of offsite power (switchyard isolation).Basis for Hi CogBasis for SRO onlyAPE056 AA2.43 - Loss of Offsite PowerAbility 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 .......................................

C 14 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 controlat setpoint B. Selectin g Pressurizer Le vel 2ONLY 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 APE057 AA1.06 - Loss of Vital AC Electrical Instrument BusAbility 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 .......

C 14General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. Plausible since this would be correct if SASS were in Auto.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesPNS-PZR Obj R31, R35PNS-PZR401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyAPE057 AA1.06 - Loss of Vital AC Electrical Instrument BusAbility 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 .......

B 15 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 APE065 AA2.05 - Loss of Instrument AirAbility 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 B 15General DiscussionAnswer A DiscussionIncorrect. Plausible since there are automatic actions that happen at 70 psig IA pressure which are detailed in AP/22.Answer B DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-APG Obj R9EAP-APGSAE-L035 SSS-IA AP/22401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of when a reactor trip is required based on decreasing IA pressure.Basis for Hi CogBasis for SRO onlyAPE065 AA2.05 - Loss of Instrument AirAbility 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 C 16 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 APE077 AK2.03 - Generator Voltage and Electric Grid DisturbancesKnowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: (CFR: 41.4, 41.5, 4 1.7, 41.10 / 45.8)Sensors, detectors, indicators......................................................

C 16General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source ILT41 Q16Student References ProvidedAP/34 Gen Capacity CurveDevelopment ReferencesCP05 Obj 5, EAP-APG Obj R9 AP/34 lesson and AP CP05401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 bas ed on those indications.Basis for Hi CogBasis for SRO onlyAPE077 AK2.03 - Generator Voltage and Electric Grid DisturbancesKnowledge 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......................................................

D 17 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 BWE02 2.2.3 - Vital System Status VerificationBWE02 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)

D 17General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. First part is plausible since there are times when the EOP directs to have only 1 RCP operating. Second part is correct.Answer C DiscussionIncorrect. First part is correct. Second part is plausible since the A1 RCP is the spray pump on Unit 1.Answer D DiscussionCorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-SA Obj R39EOP-SA Unit 3 SA401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyBWE02 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)

D 18 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. CETC's are used since the resultant RCS cooldown may result in Tcold being off scale low D. CETC's are used since they are qualified instrumentsand are therefore more reliable in the hostile containment environment BWE05 EK3.2 - Excessive Heat TransferKnowledge of the reasons for the following responses as they apply tothe (Excessive Heat Transfer)(CFR: 41.5 / 41.10, 45.6, 45.13)Normal, abnormal and emergency operating procedures associated with (Excessive Heat Transfer).

D 18General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. Plausible since Tcold would be the colder temperature and PTS due to decreasing RCS temperature during a MSLB are a concern.Answer C DiscussionIncorrect. CETC's are correct. The reason is plausible since the statement is a true statement. Tcold uses narrow range Tempera ture 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 DiscussionCorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-EHT Obj R8EAP-EHT401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of the reason for procedural guidance to use CETC's to stabilize RCS temperatures following a MSLB.Basis for Hi CogBasis for SRO onlyBWE05 EK3.2 - Excessive Heat TransferKnowledge of the reasons for the following responses as they apply tothe (Excessive Heat Transfer)(CFR: 41.5 / 41.10, 45.6, 45.13)Normal, abnormal and emergency operating procedures associated with (Excessive Heat Transfer).

B 19 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 Of f-gas) increased.

B. 1RIA-16 (Main Steam Line Monitor

) and 1RI A-40 increased.

C. 1RIA-59 increased while1RIA-40 remained constant.

D. 1RIA-16 increased while 1RI A-40 remained constant.

APE076 2.3.5 - High Reactor Coolant ActivityAPE076 GENERICAbility 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)

B 19General DiscussionAnswer A DiscussionIncorrect. 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 RIA's on the way to the Main Turbine however since they are N16 monitiors, the increase in activity will not impact their readings.Answer B DiscussionCorrect: 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source2009B Q24Student References ProvidedDevelopment References RAD-RIA Obj R2 RAD-RIA401-9 Comments:Remarks/StatusBasis for meeting the KADemonstrates 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 CogBasis for SRO onlyAPE076 2.3.5 - High Reactor Coolant Activity APE076 GENERICAbility 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)

C 20 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 NI's = ~ 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% withdrawn2. 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 lon ger available C. 1. Fully insert all Control Rods2. Prevents power increases when the primary power indication available to the operator is not available.

D. 1. Fully insert all Control Rods2. 2 dpm Startup Rate Control Rod Out Inhibit is no lon ger available APE032 AK3.01 - Loss of Source Range Nuclear InstrumentationKnowledge 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 ............................

C 20General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType MODIFIEDQuestion Source2007 Q20Student References ProvidedDevelopment ReferencesIC-CRI Obj R32, ADM-TSS Obj R4 TS 3.3.9 IC-CRI401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyAPE032 AK3.01 - Loss of Source Range Nuclear InstrumentationKnowledge 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 ............................

D 21 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 betw een __ (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 heat ers 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 APE037 AA1.11 - Steam Generator (S/G) Tube LeakAbility 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 ..............................................

D 21General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. First part is plausible because this is the correct level if the reactor is tripped.. Second part is correct..Answer C DiscussionIncorrect. 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 DiscussionCorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source ILT41 Q21Student References ProvidedDevelopment ReferencesEAP-SGTR Obj R4SGTR tabEAP-SGTR401-9 Comments:Remarks/StatusBasis for meeting the KARequires the ability to monitor and maintain the correct Pzr level during a SGTR shutdown.Basis for Hi CogBasis for SRO onlyAPE037 AA1.11 - Steam Generator (S/G) Tube LeakAbility 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 ..............................................

A 22 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 P ressure = 5.5 psig B. Main Turbine speed = 1955 RPM C. Loss of both Active Turbine Speed signals D. EITHER Steam Generator Level = 93% OR BWA04 AA1.2 - Turbine TripAbility to operate and / or monitor the following as they apply tothe (Turbine Trip)(CFR: 41.7 / 45.5 / 45.6)Operating behavior characteristics of the facility.

A 22General DiscussionAnswer A DiscussionCorrect. 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 DiscussionIncorrect. Plausible since rated Turbine speed is 1800 rpm and 1955 rpm is significantly greater than rated speed.Answer C DiscussionIncorrect: 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 DiscussionIncorrect: Plausible since this level is above the high level limit setpoint of 86% ORCognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References STG-EHC Obj R24STG-EHC STG-ICS401-9 Comments:Remarks/StatusBasis for meeting the KAIncorrect. Plausible since 93% is above the high level limit setpoint ofBasis for Hi CogBasis for SRO onlyBWA04 AA1.2 - Turbine TripAbility to operate and / or monitor the following as they apply tothe (Turbine Trip)(CFR: 41.7 / 45.5 / 45.6)Operating behavior characteristics of the facility.

B 23 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 open2. immediately B. 1. automatically close 2. immediately C. 1. remain open2. 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 /> APE060 2.2.36 - Accidental Gaseous-Waste ReleaseAPE060 GENERICAbility to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions f or operations. (CFR: 41.10 / 43.2 / 45.13)

B 23General DiscussionAnswer A DiscussionIncorrect, 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 DiscussionCorrect, 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 DiscussionIncorrect, 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source2009B Q50Student References ProvidedDevelopment ReferencesRAD-RIA Obj R2, R15, ADMIN-TSS Obj R3RAD-RIA, SLC-16.11.3401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyAPE060 2.2.36 - Accidental Gaseous-Waste Release APE060 GENERICAbility 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)

B 24 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 X SUR 2. TBVs BWA03 AK1.3 - Loss of NNI-YKnowledge of the operational implications of the following concepts asthey 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)

B 24General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source2007 Q25Student References ProvidedDevelopment ReferencesSTG-ICS R33 STG-ICS Intro STG-ICS Chptr 8 STG-ICS Chptr 3401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyBWA03 AK1.3 - Loss of NNI-YKnowledge of the operational implications of the following concepts asthey 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)

D 25 Given the following Unit 1 conditions:Initial conditions: Switchyard isolation occurs Current conditions: Shutdown of KHU's 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 KHU's?

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 H ydro AtOconee

) D. 1. No 2. OP/0/A/1106/019 (Keowee H ydro AtOconee

) BWA05 AA2.1 - Emergency Diesel ActuationAbility to determine and interpret the following as they apply tothe (Emergency Diesel Actuation)(CFR: 43.5 / 45.13)Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

D 25General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-APG Obj R9. EL-PSL Obj R5EL-PSL, AP/11Encl 5.41401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyBWA05 AA2.1 - Emergency Diesel ActuationAbility to determine and interpret the following as they apply tothe (Emergency Diesel Actuation)(CFR: 43.5 / 45.13)Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

A 26 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 gpm2. LPI B. 1. Maximized to < 3100 gpm2. HPI C. 1. Maximized to < 2900 gpm2. LPI D. 1. Maximized to < 2900 gpm2. HPI BWE08 EK3.1 - LOCA CooldownKnowledge of the reasons for the following responses as they apply tothe (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 andoperating limitations and reasons for these operating characteristics.

A 26General DiscussionAnswer A DiscussionCorrect. With only one LPI pump operating in the Piggyback mode LPI flow is maximized to < 3100 gpm to protect the LPI pump fro m runout.Answer B DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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 fl ow limits on the HPI pumps to protect them from damage.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-LCD Obj R6EAP-SA Obj R17EAP-LCD401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyBWE08 EK3.1 - LOCA CooldownKnowledge of the reasons for the following responses as they apply tothe (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 andoperating limitations and reasons for these operating characteristics.

C 27 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 Feedwate r B. 240 inches using Main Feedwater C. Loss of Subcooling Margin setpoint using Emergency Feedwater D. Loss of Subcooling Margin setpointusing Main Feedwate r BWE03 EK2.1 - Inadequate Subcooling MarginKnowledge of the interrelations between the (Inadequate SubcoolingMargin) and the following: (CFR: 41.7 / 45.7)Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and aut omatic and manual features.

C 27General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References CF-EF Obj R37 CF-EFRule 7401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyBWE03 EK2.1 - Inadequate Subcooling MarginKnowledge of the interrelations between the (Inadequate SubcoolingMargin) 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.

A 28 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 limits2. 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 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 ........................

A 28General DiscussionAnswer A DiscussionCorrect. 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 DiscussionIncorrect. First part is correct. Second part is plausible since it is the expected delta T at 100% power.Answer C DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesPNS-CPM Obj R8 , TA-AM! Obj R3 PNS-CPMTA-AM1401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of the effect that RCP coastdown has on RCS flow.Basis for Hi CogBasis for SRO onlySYS003 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 ........................

C 29 The Letdown Storage Tank:

1) contains approximately __(1)__ gallons of water per inch of level.

2) level setpoint that will automatically open 1HP-24 and 1HP-25 is __(2)__ inches.

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 SYS004 A1.06 - Chemical and Volume Control SystemAbility 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 .......................................................

C 29General DiscussionAnswer A DiscussionIncorrect. First part is plausible since it would be correct for the Pressurizer. Second part is correct.Answer B DiscussionIncorrect. 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 DiscussionCorrect. The LDST is approximately 31.3 gal/in and the setpoint for the LDST level interlock is <40 inches.Answer D DiscussionIncorrect. First part is correct. Second part is plausible since it is the Lo Lo level alarm setpoint for the LDST.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesPNS-PZR Obj R2, PNS-HPI Obj R8PNS-PZR, PNS-HPI401-9 Comments:Remarks/StatusBasis for meeting the KARequires ability to predict automatic operation of the HPI system controls as a function of VCT (LDST) level.Basis for Hi CogBasis for SRO onlySYS004 A1.06 - Chemical and Volume Control SystemAbility 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 .......................................................

D 30 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 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 ..............................................

D 30General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. Plausible since this would be correct if the 2A cooler were not available.Answer D DiscussionCorrect. High Pressure mode only uses one cooler and it is the A cooler.Cognitive Level ComprehensionJob Level ROQuestionType MODIFIEDQuestion Source2009B Q31Student References ProvidedDevelopment ReferencesPNS-LPI Obj 13, 35PNS-LPI401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of the effect that a loss of one of the DHR coolers will have on available DHR alignments.Basis for Hi CogBasis for SRO onlySYS005 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 ..............................................

D 31 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 SYS006 2.4.45 - Emergency Core Cooling System (ECCS)SYS006 GENERICAbility to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

D 31General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. Rule 2 is performed due to the loss of subcooling and rule 4 is performed based on SCM and Pzr level.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-LOHT Obj R23 LOHT tab of EOP401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS006 2.4.45 - Emergency Core Cooling System (ECCS)

SYS006 GENERICAbility to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

D 32 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 SYS007 2.2.36 - Pressurizer Relief Tank/Quench Tank System (PRTS)SYS007 GENERICAbility to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions f or operations. (CFR: 41.10 / 43.2 / 45.13)

D 32General DiscussionAnswer A DiscussionIncorrect, 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 DiscussionIncorrect, 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 correctAnswer C DiscussionIncorrect. First part is correct. 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 D DiscussionCorrect. The LCO of TS 3.4.12 requires the PORV to be Operable. 2RC-66 pilot valve DC solenoid and indicating lights are powere d 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..Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesADMIN-ITS Obj R8, PNS-PZR Obj R30 TS 3.4.12, 3.4.9PNS-PZR401-9 Comments:Remarks/StatusBasis for meeting the KAChief 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 CogBasis for SRO onlySYS007 2.2.36 - Pressurizer Relief Tank/Quench Tank System (PRTS)

SYS007 GENERICAbility 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)

C 33 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 RIA's response to the above conditions?

2) valves that will automatically close anytime 1RIA-49 reaches its HIGH alarm setpoint?

A. 1. increases 2. 1LWD-1 AND 1LWD-2 B. 1. remains constant 2. 1LWD-1 AND 1LWD-2 C. 1. increases 2. 1LWD-2 ONLY D. 1. remains constant 2. 1LWD-2 ONLY 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 ....................................................

C 33General DiscussionSince 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 DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType MODIFIEDQuestion Source2009 Q32Student References ProvidedDevelopment ReferencesPNS-CS Obj R7, PNS-PZR RAD-RIA401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS007 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 ....................................................

A 34 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 LAWT2. trippin g the Reactor B. 1. will overflow to the LAWT2. initiatin g a shutdown usin g AP/1/A/1700/029 (Rapid Unit Shutdown) C. 1. will overflow to a floor drain which drains to the MWHU T2. trippin g the Reactor D. 1. will overflow to a floor drain which drains to the MWHUT 2. initiatin g a shutdown usin g AP/1/A/1700/029 (Rapid Unit Shutdown) 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 ....................................................

A 34General DiscussionAnswer A DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-APG R9 AP/2401-9 Comments:Remarks/StatusBasis for meeting the KADescribes the impact of the CC leakage which is indicated by RIA alarms and uses AP/2 to mitigate the consequences of the leaka ge.Basis for Hi CogBasis for SRO onlySYS008 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 ....................................................

A 35 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

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 ............................................

A 35General DiscussionAnswer A DiscussionCorrect: 1RC-1 (Pzr Spray) setpoint is 2205 psig and 1RC-66 (PORV) is 2450 psig when in HIGH (Mode 1)Answer B DiscussionIncorrect: 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 DiscussionIncorrect: 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 DiscussionIncorrect: 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.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source2009 Q34Student References ProvidedDevelopment ReferencesPNS-PZR R5, R9PNS-PZR401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of Pzr PCS setpoints for automatic pressure control.Basis for Hi CogBasis for SRO onlySYS010 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 ............................................

C 36 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 ONLY2. Place MANUAL TRIP Keyswitch in "TRIP".

B. 1. RCS High Outlet Temperature ONLY2. Place affected RPS Channel MANUAL BYPASS ke yswitch in "BYP".

C. 1. RCS High Outlet Temperature and RCS Variable Low Pressure 2. Place MANUAL TRIP Ke yswitch in "TRIP".

D. 1. RCS High Outlet Temperature and RCS Variable Low Pressure 2. Place affected RPS Channel MANUAL BYPASS keyswitch in "BYP".

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 .........

C 36General DiscussionAnswer A DiscussionIncorrect: 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 DiscussionIncorrect: 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 DiscussionCorrect: 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 - 4706With 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 DiscussionIncorrect. First part is correct. Second part is plausible since it would be correct if the 1D RPS channel were not already in Manual Bypass.Cognitive Level ComprehensionJob Level ROQuestionType MODIFIEDQuestion Source ILT40 Q38Student References ProvidedDevelopment ReferencesAdmin-ITS Obj R8, , Admin-PIS Obj R3 IC-RPS Obj R3,4,23 1105/014 IC-RPS401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS012 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 .........

B 37 Which ONE of the following describes howRCS 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)

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 .........................

B 37General DiscussionAnswer A DiscussionIncorrect. Plausible since this is how the Thot signal to ICS is generated.Answer B DiscussionCorrect. RPS NR channel A, B, and E RCS pressures are median selected as a control signal for ICS.Answer C DiscussionIncorrect. Plausible since this is how RPS uses NR RCS pressure to determine if a high RCS pressure trip is required.Answer D DiscussionIncorrect. Plausible since this is how RPS uses NR RCS pressure to determine if a low RCS pressure trip is required.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References IC-RCI Obj R61 Th ppt IC-RCI401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS012 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 .........................

C 38 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 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 ...................................

C 38General DiscussionAnswer A DiscussionIncorrect: Plausible since it would be correct if RB pressure had reached the ES 1-6 setpoint of 3 psig.Answer B DiscussionIncorrect. Plausible since this valve did receive an ES signal however it was a signal to open.Answer C DiscussionCORRECT: 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 DiscussionD.~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.Cognitive Level ComprehensionJob Level ROQuestionType MODIFIEDQuestion Source 2009A Q39Student References ProvidedDevelopment References IC-ES R14, R18 IC-ES401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of ES actuation setpoints, as well as what components are operated from which ES digital channels. This woul d demonstrate the ability to monitor automatic operation of actuated equipment.Basis for Hi CogBasis for SRO onlySYS013 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 ...................................

C 39 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 SYS064 K1.04 - Emergency Diesel Generator (ED/G) SystemKnowledge 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 ...........................................

C 39General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. With a LOCA/LOOP occurring, the MFB would re-energize from the underground power path which means if would energize fr om 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 DiscussionIncorrect. Plausible since this would be the correct answer if the standby bus were energized from Central or Lee prior to time

= 1201.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEL-DCD Obj 06, EL-PSL Obj R24EL-DCDEL-PSL401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS064 K1.04 - Emergency Diesel Generator (ED/G) SystemKnowledge 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 ...........................................

A 40 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 stopsthe Waste Gas Exhauster but does NOT trip the running GWD compressors 2. OP/1-2/A/1104/018 (GWD S ystem) 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 S ystem) 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 Radioactivit y) D. 1. Closes the GWD tank outlet valves, stopsthe Waste Gas Exhauster, AND trips running GWD compressors 2. AP/18 (Abnormal Release of Radioactivit y) SYS073 A1.01 - Process Radiation Monitoring (PRM) SystemAbility 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 .................................................

A 40General DiscussionAnswer A DiscussionCorrect: 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 DiscussionIncorrect: 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 DiscussionIncorrect: 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 DiscussionIncorrect: 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source ILT40 Q73Student References ProvidedDevelopment ReferencesEAP-APG Obj R9 , RAD-RIA Obj R2AP/18, RAD-RIA 1SA-8/B9 /ARG401-9 Comments:Remarks/StatusBasis for meeting the KAQuestion 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 CogBasis for SRO onlySYS073 A1.01 - Process Radiation Monitoring (PRM) SystemAbility 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 .................................................

B 41 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 SYS073 K3.01 - Process Radiation Monitoring (PRM) SystemKnowledge 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 ......................................

B 41General DiscussionAnswer A DiscussionIncorrect 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 correctAnswer 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 DiscussionIncorrect. 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 correctAnswer D DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType MODIFIEDQuestion Source ILT39 Q51Student References ProvidedDevelopment References RAD-RIA R2 OP/1102/14401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of the effect that a loss of RIA's will have on Radioactive Effluent releases that are in progress.Basis for Hi CogBasis for SRO onlySYS073 K3.01 - Process Radiation Monitoring (PRM) SystemKnowledge 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 ......................................

A 42 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 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 .............................................

A 42General DiscussionAnswer A DiscussionCorrect. LPSW cools the CC coolers which in turn cool letdown therefore degraded LPSW flow would result in increasing letdown temperatures.Answer B DiscussionIncorrect. Plausible since it would be correct for degraded RCW flows.Answer C DiscussionIncorrect. Plausible since it would be correct for degraded RCW flows.Answer D DiscussionIncorrect. Plausible since it would be correct for degraded HPSW flows.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References PNS-CC Obj 03PNS-CC, SSS-IA, SSS-HPW, SSS-RCW401-9 Comments:Remarks/StatusBasis for meeting the KATo 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 CogBasis for SRO onlySYS076 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 .............................................

C 43 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 SYS078 K2.02 - Instrument Air System (IAS)Knowledge of bus power supplies to the following: (CFR: 41.7)Emergency air compressor ........................................

C 43General DiscussionAuxiliary Instrument Air System compressors are powered from non-load shed power supplies 1,2,3XP.Answer A DiscussionIncorrect, plausible because 1XD supplies the "A" Worthington compressor (Backup IA Compressor).Answer B DiscussionIncorrect, plausible because 2XF supplies the "C" Worthington compressor (Backup IA Compressor).Answer C DiscussionCorrect, Auxiliary Instrument Air System compressors are power from non-load shed power supplies 2XPAnswer D DiscussionIncorrect, plausible because this is a ES safety related bus.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source2007 Q49Student References ProvidedDevelopment References SSS-IA401-9 Comments:Remarks/StatusBasis for meeting the KA Requires knowledge of the power supply for the AIA compressors.Basis for Hi CogBasis for SRO onlySYS078 K2.02 - Instrument Air System (IAS)Knowledge of bus power supplies to the following: (CFR: 41.7)Emergency air compressor ........................................

A 44 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 SYS103 A4.04 - Containment SystemAbility to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)Phase A and phase B resets ........................................

A 44General DiscussionAnswer A DiscussionCorrect. 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 DiscussionIncorrect. Plausible since: 1HP-3 is on ES channel 1 ES Channel 1 is taken to manualRestoring letdown is desired AND directed by the EOP however it is done by Encl 5.51HP-20 is not opened in all cases adding to the plausibility of 1HP-3 being correct.Answer C DiscussionIncorrect. Plausible since this would be correct if ES Channel 1 was unable to be placed in Manual.Answer D DiscussionIncorrect. 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 5.51HP-20 is not opened in all cases adding to the plausibility of 1HP-3 being correct.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-ESA Obj R5 EAP-ESAEOP Encl 5.1401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS103 A4.04 - Containment SystemAbility to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Phase A and phase B resets ........................................

A 45 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. manuall y increase SG level B. 1. 240" XSUR 2. no actions required C. 1. 50% OR 2. manuall y increase SG level D. 1. 50% OR 2. no actions required 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 .........

A 45General DiscussionAnswer A DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. First part is plausible since it would be correct if either Main FDW pump were still in operation. Second part is correct.Answer D Discussionincorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References CF-EF R37 CF-EF Rule 7401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS016 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 .........

A 46 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 NI's 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 NI's and formula provided in OP/1/A/1105/014 D. 1. 15 2. PT/1/A/1103/019 (Backup Incore Detector S ystem) SYS015 2.2.12 - Nuclear Instrumentation System (NIS)SYS015 GENERICKnowledge of surveillance procedures. (CFR: 41.10 / 45.13)

A 46General DiscussionAnswer A DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References ADM-PIS Obj R5 ADM-PISOP/1/A/1105/014 Encl. 4.13401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of surveillance procedures as part of performing surveillances associated with nuclear instrumentation.Basis for Hi CogBasis for SRO onlySYS015 2.2.12 - Nuclear Instrumentation System (NIS)

SYS015 GENERICKnowledge of surveillance procedures. (CFR: 41.10 / 45.13)

C 47 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 procedurall y allowed to be desi gnated D. Required to be stopped and cannot be resumed until SR 1NI-1 is returned to service because other NIs are NOT procedurall y allowed to be desi gnated 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 ...................................................

C 47General DiscussionHarder this way since the LP talks about core alterations, not defeueling or refueling.Answer A DiscussionIncorrect: Plausible because the two SR NIs remain in service therefore IAW 1502/07 this would be the correct answer for defueling.Answer B DiscussionIncorrect 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 DiscussionCorrect: Procedure requires movement to be stopped until 2 NIs used to monitor core reactivity can be designated.Answer D DiscussionIncorrect 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.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source2009B Q21Student References ProvidedDevelopment References Obj. FH-FHS R20 FH-FHS OP/1/A/1502/007401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS034 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 ...................................................

C 48 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 SYS041 K4.11 - Steam Dump System (SDS)/Turbine Bypass ControlKnowledge of SDS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)T-ave./T-ref. program .............................................

C 48General DiscussionAnswer A DiscussionIncorrect. Plausible since this is the normal THP setpoint.Answer B DiscussionIncorrect. Plausible since this is setpoint plus 50 psig and is therefore correct prior to the Rx trip.Answer C DiscussionCorrect. 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 DiscussionIncorrect. Plausible since this would be correct if THP setpoint were at its normal value of 885 psig.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesSTG-ICS R10 STG-ICS chapter 3401-9 Comments:Remarks/StatusBasis for meeting the KARequired 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 CogBasis for SRO onlySYS041 K4.11 - Steam Dump System (SDS)/Turbine Bypass ControlKnowledge of SDS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)T-ave./T-ref. program .............................................

D 49 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 SYS056 A2.04 - Condensate SystemAbility 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 / 4 5.13)Loss of condensate pumps .........................................

D 49General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. First part is plausible since there are ICS runbacks that incorporate the 15%/min runback rate. Second part is corre ct,Answer C DiscussionIncorrect. First part is correct. Second part is plausible since it would be correct for the first 90 seconds of the transient.Answer D DiscussionCorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source ILT40 Q62Student References ProvidedDevelopment ReferencesObj STG-ICS R3 EAP-SA R21, R24EAP-SA STG-ICS Intro & Chptr 2401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 onlySYS056 A2.04 - Condensate SystemAbility 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 / 4 5.13)Loss of condensate pumps .........................................

D 50 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) A sustained loss of power to 1RIA-54 will trip BOTH Turbine Building Sump Pumps __(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 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 ...............................................

D 50General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesRad-RIA Obj R2 Rad-RIA401-9 Comments:Remarks/StatusBasis for meeting the KARequired knowledge of the effect of a loss of power to RIA-54 will have on Liquid Waste Releases from the Turbine Building Sump s.Basis for Hi CogBasis for SRO onlySYS068 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 ...............................................

B 51 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 RIA's 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 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 / 4 5.6)Ventilation system ...............................................

B 51General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesRAD-RIA Obj R2, BPS-DPR Obj 01RAD-RIA, AP/18401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS071 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 ...............................................

B 52 1RIA-59 setpoints are set by __(1)__ and the MINIMUM power level at which 1RI A-59is 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 SYS072 A1.01 - Area Radiation Monitoring (ARM) SystemAbility 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 .................................................

B 52General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. First part is plausible since it would be correct for most every other RIA. Second part is correct.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesRAD-RIA Obj R2, R15 RAD-RIA SGTR401-9 Comments:Remarks/StatusBasis for meeting the KARIA-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 CogBasis for SRO onlySYS072 A1.01 - Area Radiation Monitoring (ARM) SystemAbility 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 .................................................

D 53 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 SYS075 K2.03 - Circulating Water SystemKnowledge of bus power supplies to the following: (CFR: 41.7)Emergency/essential SWS pumps ...................................

D 53General DiscussionAnswer A DiscussionIncorrect. Plausible because it can supply power to the B LPSW pump but 2TD can also.Answer B DiscussionIncorrect. Plausible because it it would be correct for the C LPSW pump.Answer C DiscussionIncorrect. Plausible because they both supply power to the LPSW pumps; A and C respectively.Answer D DiscussionCorrect. The B LPSW pump can be supplied power from 1TD or 2TD.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source ILT41 Q50Student References ProvidedDevelopment ReferencesSSS-LPW R11SSS-LPW401-9 Comments:Remarks/StatusBasis for meeting the KAQuestion requires knopwledge of the power supply for one of the LPSW pumps.Basis for Hi CogBasis for SRO onlySYS075 K2.03 - Circulating Water SystemKnowledge of bus power supplies to the following: (CFR: 41.7)Emergency/essential SWS pumps ...................................

D 54 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 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 ............................................................

D 54General DiscussionAnswer A DiscussionIncorrect: Plausible since 93 psig is the pressure at which the Backup IA compressors will start.Answer B DiscussionIncorrect: Plausible since 90 psig is the pressure at which the Diesel Air Compressors will startAnswer C DiscussionIncorrect: Plausible sine 88 psig is the pressure at which the AIA compressors will startAnswer D DiscussionCORRECT: 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.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source ILT40 Q65Student References ProvidedDevelopment References Obj SSS-IA R52401-9 Comments:Remarks/StatusBasis for meeting the KA Requires knowledge of automatic cross-connect between Service air and Instrument air systems.Basis for Hi CogBasis for SRO onlySYS079 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 ............................................................

B 55 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 GEN2.1 2.1.26 - GENERIC - Conduct of OperationsConduct of OperationsKnowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (CFR: 41.10 / 45.12)

B 55General DiscussionAnswer A DiscussionIncorrect. Plausible since CO2 is a common agent used in fire prevention/extinguishing and the addition of CO2 would decrease t heconcentration of Hydrogen.Answer B DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. Plausible since foam fire retardant is used to prevent fires during flammable liquid spills.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source2009B Q67Student References ProvidedDevelopment References SSS-AGS Obj R15 SSS-AGS401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of industrial safety procedure directed by procedure to mitigate effects of a large Hydrogen leak.Basis for Hi CogBasis for SRO onlyGEN2.1 2.1.26 - GENERIC - Conduct of Operations Conduct of OperationsKnowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (CFR: 41.10 / 45.12)

C 56 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/g m D. DEI = 1.0 µCi/gm GEN2.1 2.1.34 - GENERIC - Conduct of OperationsConduct of OperationsKnowledge of primary and secondary plant chemistry limits. (CFR: 41.10 / 43.5 / 45.12)

C 56General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect 0.25 micro Ci/gm is the threshold for entry level into AP/21.Answer D Discussionincorrect. Plausible since 1.0 is a threshold value referenced several times in AP/21.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-APG Obj R9 AP/21401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of chemistry limit that is the entry condition for AP/21 (RCS Activity).Basis for Hi CogBasis for SRO onlyGEN2.1 2.1.34 - GENERIC - Conduct of Operations Conduct of OperationsKnowledge of primary and secondary plant chemistry limits. (CFR: 41.10 / 43.5 / 45.12)

C 57 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 adjustRCS temperature GEN2.1 2.1.37 - GENERIC - Conduct of OperationsConduct of OperationsKnowledge of procedures, guidelines, or limitations associated with reactivity management. (CFR: 41.1 / 43.6 / 45.6)

C 57General DiscussionAnswer A DiscussionIncorrect: Manual rod withdrawal is not permitted.Answer B DiscussionIncorrect: Increasing FDW Flow is not permitted.Answer C DiscussionCorrect: The sequence given is permitted as there is no intent to raise FDW Flow.Answer D DiscussionIncorrect: Increase in Turbine demand is only allowed if intent is to stabilize Turbine Header Pressure not to reduce pressure or RCS temperature.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source2009 Q68Student References ProvidedDevelopment References ADM-OMP Obj R23TA-PTR Obj R1TA-PTR401-9 Comments:Remarks/StatusBasis for meeting the KADemonstrates an understanding of guidelines associated with reactivity management provided in SOMP 1-02.Basis for Hi CogBasis for SRO onlyGEN2.1 2.1.37 - GENERIC - Conduct of Operations Conduct of OperationsKnowledge of procedures, guidelines, or limitations associated with reactivity management. (CFR: 41.1 / 43.6 / 45.6)

A 58 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 GEN2.2 2.2.13 - GENERIC - Equipment ControlEquipment ControlKnowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13)

A 58General DiscussionAnswer A DiscussionCorrect. White tags are used for configuration control of components and systemsAnswer B DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType BANKQuestion Source ILT40 Q70Student References ProvidedDevelopment References Obj ADM-SD R6 NSD 500401-9 Comments:Remarks/StatusBasis for meeting the KARequires generic knowledge of the tagging process defined by NSD 500Basis for Hi CogBasis for SRO onlyGEN2.2 2.2.13 - GENERIC - Equipment ControlEquipment ControlKnowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13)

B 59 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 GEN2.2 2.2.12 - GENERIC - Equipment ControlEquipment ControlKnowledge of surveillance procedures. (CFR: 41.10 / 45.13)

B 59General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionCorrect. PT.600/01 corrects the TS required 285" for allowable instrument error and uses 260" as the threshold value.Answer C DiscussionIncorrect. Plausible since this is the analytical value provided in Tech Spec 3.4.9 for maximum level.Answer D DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References Adm-ITS Obj R8 TS 3.4.9PT/600/01401-9 Comments:Remarks/StatusBasis for meeting the KARequired knowledge of PT/600/01 surveillance requirements for Pzr level.Basis for Hi CogBasis for SRO onlyGEN2.2 2.2.12 - GENERIC - Equipment ControlEquipment ControlKnowledge of surveillance procedures. (CFR: 41.10 / 45.13)

D 60 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 ONLYonce 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 hou r from the loss of air supp ly D. Will be available for Automatic operation for a MINIMUM of 2 hours from the loss of air supply GEN2.2 2.2.37 - GENERIC - Equipment ControlEquipment ControlAbility to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12)

D 60General DiscussionAnswer A DiscussionIncorrect. Plausible since both IA and AIA have been lost however there is a N2 backup supply to the valves.Answer B DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source 2007 Audit Q 54Student References ProvidedDevelopment References CF-EF Obj R39 CF-EF401-9 Comments:Remarks/StatusBasis for meeting the KARequires the ability to determine how long auto operation of FDW-315/316 is available following a loss of IA & AIA.Basis for Hi CogBasis for SRO onlyGEN2.2 2.2.37 - GENERIC - Equipment ControlEquipment Control Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12)

C 61 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?

A. 1. 2P R-1 through 2P R-6 2. Zero B. 1. 2PR-1 through 2PR-6 2. 1342 C. 1. 2P R-2 through 2P R-5 ONLY2. Zero D. 1. 2P R-2 through 2P R-5 ONLY2. 1342 GEN2.3 2.3.11 - GENERIC - Radiation ControlRadiation ControlAbility to control radiation releases. (CFR: 41.11 / 43.4 / 45.10)

C 61General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionCorrect. 2RIA-45 HIGH alarm will close 2PR-2-5. Under normal operating circumstances, when RIA 45/46 are both in service, the R IA 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References RAD-RIA Obj R2 RAD-RIA401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlyGEN2.3 2.3.11 - GENERIC - Radiation ControlRadiation ControlAbility to control radiation releases. (CFR: 41.11 / 43.4 / 45.10)

A 62 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 /> GEN2.3 2.3.7 - GENERIC - Radiation ControlRadiation ControlAbility to comply with radiation work permit requirements during normal orabnormal conditions. (CFR: 41.12 / 45.10)

A 62General DiscussionAnswer A DiscussionCorrect. 15 minutes in a 100 mr/hr field will result in 25 mrem which is the limit provided with the RWP.Answer B DiscussionIncorrect. Plausible since this would be correct for the 1A LPI pumpAnswer C DiscussionIncorrect. Plausible if using the Dose RATE setpoint with the 1C LPI pumpAnswer D DiscussionIncorrect. Plausible if using the Dose RATE setpoint instead of the Dose alarm setpointCognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedPlan viewDevelopment ReferencesPlan View401-9 Comments:Remarks/StatusBasis for meeting the KA Requires demonstrating the ability to comply with RWP requirements.Basis for Hi CogBasis for SRO onlyGEN2.3 2.3.7 - GENERIC - Radiation ControlRadiation ControlAbility to comply with radiation work permit requirements during normal orabnormal conditions. (CFR: 41.12 / 45.10)

A 63 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 lon g as the RCS remains subcooled, adequate core coolin g is assured.

B. 1. LOSCM 2. ensures RCP's are secured before pump damage renders them unavailable C. 1. SGTR 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 GEN2.4 2.4.22 - GENERIC - Emergency Procedures / PlanEmergency Procedures / PlanKnowledge of the bases for prioritizing safety functions during abnormal/emergency operations. (CFR: 41.7 / 41.10 / 43.5 / 45.1

2)

A 63General DiscussionAnswer A DiscussionCorrect. 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 DiscussionIncorrect. First part is correct. Second part is plausible since there is much focus on ensuring RCP's are secured within 2 min utes 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-LOSCM Obj R12EAP-LOSCMEAP-SGTR401-9 Comments:Remarks/StatusBasis for meeting the KAChief 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 CogBasis for SRO onlyGEN2.4 2.4.22 - GENERIC - Emergency Procedures / PlanEmergency Procedures / PlanKnowledge of the bases for prioritizing safety functions during abnormal/emergency operations. (CFR: 41.7 / 41.10 / 43.5 / 45.1

2)

B 64 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 GEN2.4 2.4.4 - GENERIC - Emergency Procedures / PlanEmergency Procedures / PlanAbility to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abn ormal operating procedures. (CFR: 41.10 / 43.2 / 45.6)

B 64General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. Plausible since it would be correct for higher power levels.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References RAD-RIA Obj R2401-9 Comments:Remarks/StatusBasis for meeting the KA Requires the ability to recognize an abnormal system parameter that is an entry conditions for the Emergency Operating Procedur e.Basis for Hi CogBasis for SRO onlyGEN2.4 2.4.4 - GENERIC - Emergency Procedures / PlanEmergency Procedures / PlanAbility 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)

B 65 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

SYS013 K2.01 - Engineered Safety Features Actuation System (ESFAS)Knowledge of bus power supplies to the following: (CFR: 41.7)ESFAS/safeguards equipment control ...............................

B 65General DiscussionAnswer A DiscussionIncorrect. KVIA provides power to the odd digital channels. With KVIA de-energized, the Odd channels cannot actuate.Answer B DiscussionCorrect. 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 DiscussionIncorrect. 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 th e misconception that ES channels 3 and 4 (LPI) only actuate on low RCS pressure this is plausible.Answer D DiscussionIncorrect. 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 th e misconception that ES channels 3 and 4 (LPI) only actuate on low RCS pressure this is plausible.Cognitive Level MemoryJob Level ROQuestionType MODIFIEDQuestion Source ILT40 Q40Student References ProvidedDevelopment References Obj IC-ES R2, R26 IC-ES401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of ES powers supplies for the equipment that controls actuation of the components.Basis for Hi CogBasis for SRO onlySYS013 K2.01 - Engineered Safety Features Actuation System (ESFAS)Knowledge of bus power supplies to the following: (CFR: 41.7)ESFAS/safeguards equipment control ...............................

B 66 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 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 ................................................

B 66General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesPNS-RBC Obj R1, R6 PNS-RBC401-9 Comments:Remarks/StatusBasis for meeting the KARequires the ability to monitor Containment Cooling System valves (LPSW cooling water to RBCU's) for proper operation following an ES signal.Basis for Hi CogBasis for SRO onlySYS022 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 ................................................

B 67 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 SYS022 K2.01 - Containment Cooling System (CCS)Knowledge of power supplies to the following: (CFR: 41.7)Containment cooling fans .........................................

B 67General DiscussionAnswer A DiscussionIncorrect: 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 DiscussionCorrect. 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 DiscussionIncorrect. Plausible to believe that 1TD is power supply to 1B RBCU which is the misconception that makes this answer plausible

.Answer D DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType MODIFIEDQuestion Source2009 Q38Student References ProvidedDevelopment ReferencesPNS-RBC, Obj R1 PNS RBCES Power Supply table401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of power supplies to Reactor Building Cooling Units (RBCUs)Basis for Hi CogBasis for SRO onlySYS022 K2.01 - Containment Cooling System (CCS)Knowledge of power supplies to the following: (CFR: 41.7)

Containment cooling fans .........................................

B 68 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 full y open 1LPSW-4 C. Place 1LPSW-251 in "Failed Open" AND Throttle LPSW flow to approximately 3000 gpm usin g 1LPSW-4 D. Place 1LPSW-251 in "Failed Open" AND Throttle LPSW flow to approximately 5200 gpm usin g 1LPSW-4 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 .............................................

B 68General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-LOSCM Obj R28EAP-LOSCM, SSS-LPWEOP Encl 5.1 & 5.12401-9 Comments:Remarks/StatusBasis for meeting the KAChief 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 CogBasis for SRO onlySYS026 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 .............................................

D 69 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. close d 2. can NOT 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 ............................................

D 69General DiscussionAnswer A DiscussionIncorrect: 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 DiscussionIncorrect: 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 DiscussionIncorrect: 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 DiscussionCorrect: 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.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source 2009B Q8Student References ProvidedDevelopment References STG-MSR Obj R18STG-MSR401-9 Comments:Remarks/StatusBasis for meeting the KAQuestion requires knowledge of automatic actions (Isolation) of the MSRs following a LOOP.Basis for Hi CogBasis for SRO onlySYS039 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 ............................................

A 70 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.

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 ...............................

A 70General DiscussionAnswer A DiscussionCorrect. 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 res ult would be CTP returning to setpoint and the steam being lost out of the TBV would result in lower MW production.Answer B DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level ComprehensionJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References SAE-L226 Obj R9SAE-L226401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS039 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 ...............................

B 71 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% SYS059 A1.03 - Main Feedwater (MFW) SystemAbility 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. .......

B 71General DiscussionAnswer A DiscussionIncorrect. Plausible since this would be correct for a RCP tripAnswer B DiscussionCorrect, per AP/1 initiate a runback to < or = 65%Answer C DiscussionAP/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 DiscussionIncorrect. Plausible since this would be correct for a dropped control rod.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment ReferencesEAP-APG Obj R9 AP/1401-9 Comments:Remarks/StatusBasis for meeting the KARequires 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 CogBasis for SRO onlySYS059 A1.03 - Main Feedwater (MFW) SystemAbility 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. .......

D 72 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.

SYS059 K1.04 - Main Feedwater (MFW) SystemKnowledge 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 ....................................

D 72General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. Plausible since this would be correct if on EFDW.Answer D DiscussionCorrect. On Main FDW with RCP's operating, SG levels would be controlled at 25" on startup indication following a Rx trip.Cognitive Level ComprehensionJob Level ROQuestionType MODIFIEDQuestion Source ILT40 Q28Student References ProvidedDevelopment References CF-FDW Obj R28 CF-FDW401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of the relationship between SG level control system and Main FDW.Basis for Hi CogBasis for SRO onlySYS059 K1.04 - Main Feedwater (MFW) SystemKnowledge 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 ....................................

A 73 Which ONE of the following describes the:

1) primary concern at ONS regarding Main Feedwater backleakage into the EFDW discharge piping?

2) method used to determine if Main Feedwater backleakage into the EFDW discharge piping is occurring?

A. 1. Vapor binding of the EFDWpumps2. locall y monitorin g EFDW pump dischar ge pipin g for increasin g temperature B. 1. Vapor binding of the EFDW pumps 2. Monitorin g EFDW temperature OAC points fo rincreasin g temperature C. 1. Overpressurizing the EFDW system piping2. locall y monitorin g EFDW pump dischar ge pipin g for increasin g temperature D. 1. Overpressurizing the EFDW system piping2. Monitoring EFDW temperature OAC points for increasing temperature SYS061 K5.05 - Auxiliary / Emergency Feedwater (AFW) SystemKnowledge 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 .................................

A 73General DiscussionAnswer A DiscussionCorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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 DiscussionIncorrect. 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.Cognitive Level MemoryJob Level ROQuestionType NEWQuestion SourceStudent References ProvidedDevelopment References CF-EF R51, R52 CF-EF401-9 Comments:Remarks/StatusBasis for meeting the KAThis 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 CogBasis for SRO onlySYS061 K5.05 - Auxiliary / Emergency Feedwater (AFW) SystemKnowledge 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 .................................

A 74 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 switch yard 2. four SYS062 K1.02 - AC Electrical Distribution SystemKnowledge 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 ..........................................................

A 74General DiscussionKeowee 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 DiscussionIncorrect.

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 DiscussionIncorrect.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 DiscussionIncorrect.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.Cognitive Level ComprehensionJob Level ROQuestionType MODIFIEDQuestion Source ILT42 Q13Student References ProvidedDevelopment ReferencesEL-KHG Obj R22 EL-KHG401-9 Comments:Remarks/StatusBasis for meeting the KASince 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 CogBasis for SRO onlySYS062 K1.02 - AC Electrical Distribution SystemKnowledge 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 ..........................................................

B 75 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 SYS063 K4.01 - DC Electrical Distribution SystemKnowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)Manual/automatic transfers of control ...............................

B 75General DiscussionAnswer A DiscussionIncorrect. 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 DiscussionCorrect, 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 DiscussionIncorrect. 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 invertersAnswer D DiscussionIncorrect. Unit 2's DC Bus is not connected to Unit 3. Plausible because Unit 2 is next to Unit 3.Cognitive Level ComprehensionJob Level ROQuestionType BANKQuestion Source ILT41 Q47Student References ProvidedDevelopment ReferencesEL-DCD Obj R4EL-DCD401-9 Comments:Remarks/StatusBasis for meeting the KARequires knowledge of a design feature which provides for automatic transfer of control power to components powered from the 3D IA panelboard.Basis for Hi CogBasis for SRO onlySYS063 K4.01 - DC Electrical Distribution SystemKnowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)Manual/automatic transfers of control ...............................

Enclosure 5.1 AP/1/A/1700/034 Generator Capability Curve Page 1 of 3

QUESTION 62 Summary of Highest Readings Smears Air Samples

& Wipes 1) 554 DPM/100 cm 2 2) 485 DPM/100 cm 2 3) 1453 DPM/100 cm 2 Surveyor: W. Walters Unless otherwise noted , dose rates in mrem

/hr.Symbol Legend (for example only

)Dose Rate Contact Reading 30 cm Reading General Area

• 150+75 20 15 Smear 15 Air Sample 15 Wipe HS-50 Hot Spot RCA Posting Drip Bag Type: Job Coverage RWP: 5036 Reactor Power

= 100%Approved by

N. Wriston , Date: Today Comments: CONTACT RP REGARDING ANY ATTEMPTS TO CLEAN LPI ROOM SUMP Room 61 LPI & RBS Pumps Survey # M-021506-17 Date/Time Today 0412 ROOM 61 LPI AND RB PUMPS N LEWA 112 SUMP PUMPS+1245 1 C LPI 1 A LPI 1 A RBS AHU 1-6 STAIRS Significant Dose Contributor 95 76 58 35 88*50+30 1 3 2 +975 50 100