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{{#Wiki_filter:QUESTION 1 Given the following conditions:
* The plant is operating at 1 00%
* A transient caused by a short in the reactor recirculation control circuitry occurs Immediately following the transient, the plant stabilizes with the following parameters:
* ReactorPower50%
* "A" Recirc pump Drive Motor breaker tripped * "B" Recirc pump flow is _22 Kgpm
* Loop "A" jet pump flow is 5 Mlbmlhr
* Loop "B" jet pump flow is 36 Mlbmlhr
* Total indicated core flow 31 Mlbm/hr What is actual core flow? A 26 Mlbm/hr.
B. 31 Mlbm/hr.
C. 36 Mlbm/hr.
D. 41 Mlbm/hr.
K&A Statement:
Justification:
Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:
AA2.03 Actual Core Flow A. Incorrect but plausible:
Jet pump flows must be added. B. Incorrect but plausible:
Jet pump flows must be added. C. Incorrect but plausible:
Jet pump flows must be added. D. Correct:
Below 48% running recirc loop speed, Loop flow will be less than 23 Kgpm. Jet pump loop flows are both positive and added together.
Setpoints must be adjusted to single loop values within 4 hours.
==References:==
HC. OP-AB. RPV. 0003, HC. OP-ST. BB-0007, HC.OP-DL.ZZ-0026 Attach 3V TS 2.2.1 and 3.4.1 Student Ref: NONE Learning Objective:
L!RIB/ Core flow determination following RR pump trip Question Source: HC Bank #137 Question History:
NRC 2002 Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 2 Given the following conditions:
* The plant is in Operational Condition 5 with the Electrical Distribution System aligned in the Normal lineup.
* An internal short on Transformer 1 BX-501 causes a sudden pressure fault on the transformer.
Which ONE of the following describes the resulting availability of Power for the Safe Shutdown Systems?
A. Power to both 4.16KV switchgear 1 OA402 and 1 OA404 fast transfers to Transformer 1AX501. 13 KV breaker BS 1-2 trips open. B and D diesel generators START but their output breakers DO NOT CLOSE. B. Power to both 4.16KV switchgear 1 OA402 and 1 OA404 fast transfers to Transformer 1AX501. 13 KV Breaker BS 1-2 trips open. Band D diesel generators DO NOT START. C. Power is lost permanently to both 4.16KV switchgear 1 OA401 and 1 OA403. 13 KV breaker BS 1-2 stays closed. B and D Diesel Generators start but their output breakers DO NOT CLOSE. D. Power is lost momentarily to both 4.16KV switchgear 1 OA402 and 1 OA404. 13 KV breaker BS 1-2 trips open. Power is restored when the B and D Diesel generators output breakers close.
Question 2 K&A Rating: 295003 Partial or Complete Loss of AC Power AA2.05 (3.9) K&A Statement:
Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of AC Power: Whether a partial or complete loss of A. C. Power has occurred.
Justification:
A. Incorrect but plausible:
The B & D Diesel Generators DO NOT START B. Correct:
13 Kv Breakers BS 2-3 and BS 1-2 trip open. Bus section 2 is energized, Bus section 1 remains energized.
The bus infeed breaker swap to the AX501 feed. The loads remain energized.
Because one infeed is always available, the Diesels do NOT start. C. Incorrect but plausible:
Power is NOT permanently lost to both 4.16KV switchgears.
Power is restored when the bus infeed breaker swaps to the AX501 feed. D. Incorrect but plausible:
Power is NOT restored from the B & D Diesel Generators
==References:==
Drawing E-0001 and 066-01: Class 1 E AC Power Distribution NOH01 EAC00 CLASS 1 E AC POWER DISTRIBUTION, page 32 of 93 Student Ref: NONE Learning Objective:
Question Source: Hope Creek Question Q76871 -Modified Question History:
2005 HC Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 3 A loss of the Channel "C" Class 1 E 125 VDC System will result in the loss of .... A. the ability to control RCIC components.
B. Reactor Feedwater Pump speed control.
C. "C" 4.16 KV Class 1 E electrical breaker control.
D. automatic or remote manual operation of HPCI MOVs.
K&A Rating: 295004AK2.03 (3.3/3.3)
K&A Statement:
Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following:
AK2.03 D.C. Bus Loads Justification:
A. Incorrect but plausible:
is powered from Channel B 125 VDC. B. Incorrect but plausible:
RFP controls receive control power from non-1 E 125 VDC. C. Correct:
'C' 1 E 125 VDC supplies 4KV breaker control power. Without control power, remote breaker operation and automatic breaker trips are disabled.
D. Incorrect but plausible:
HPCI MOVs receive control power from "A" bus of 125 VDC.
==References:==
N/A Learning Objective:
OAB 150E003 Question Source: HC Bank Q146 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
10CFR: CFR 41.7/45.8 Comments:
Student Ref: NONE QUESTION 4 Following a main generator load rejection at full power, a reactor recirculation pump trip is initiated in anticipation of a ... A. Reactor Pressure Increase B. Reactor Pressure Decrease C. Reactor Water Level Increase D. Reactor Water Level Decrease K&A Rating: K&A Statement:
Justification:
295005AK1.01 (4.0/4.1)
Knowledge of the operational implications of the following concepts as the apply to MAIN TURBINE GENERATOR TRIP: AK1.01 Pressure effects on reactor power A. Correct:
When a main generator load rejection signal is received the turbine is tripped.
and a SCRAM signal will be generated.
The SCRAM is initiated to reduce reactor power in anticipation of the reactor pressure increase and subsequent power increase due to void collapse.
Tripping of a recirculation pump will also serve to reduce reactor power and minimize reactor pressure increase.
B. Incorrect but plausible:
reactor pressure will increase, not decrease due to the turbine trip and control valve closure.
C. Incorrect but plausible:
reactor level will decrease not increase due to the rising reactor pressure from the turbine trip and control valve closure.
D. Incorrect but plausible:
reactor level will decrease due to the turbine trip and closure of the control valves, however the SCRAM and recirc pump trip is in anticipation of the pressure increase which will be more limiting than the level drop.
==References:==
Main Turbine Construction and Components Lesson Plan Learning Objective:
R4 Question Source: Bank Question History:
Taken from HC Bank-Question#
214 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 5 Due to a spurious MSIV isolation, the Unit has tripped from 100% power. Given the following conditions:
* All control rods are fully inserted
* HPCI is running in pressure control mode
* RCIC is running in level control
* Torus temperature is 135°F and rising at about 1 degree/minute Caution #4 of HC.OP-EO.ZZ-0101, RPV Control, warns against operating HPCI and RCIC with high suction temperatures due to ... A tripping the turbine(s) due to low suction pressure B. cavitation of the jockey fill pump C. inadequate condensation in the barometric condenser D. inadequate cooling to the lube oil cooler K&A Rating: 295006 G2.4.20 (3.8/4.3)
K&A Statement:
Knowledge of the operational implications of the EOP warnings,
: cautions, and notes Justification:
A. Incorrect but plausible:
if applicant believes that the purpose of the caution is to prevent elevated HPCI!RCIC suction temperatures causing cavitation of the HPCI!RCIC pumps and the accompanying pressure surges could lead to a low suction pressure trip B. Incorrect but plausible:
if applicant believes that the purpose of the caution is to prevent elevated torus temperatures from causing cavitation of the jockey fill pump C. Incorrect but plausible:
if the applicant believes that the purpose of the caution is to alert operators to the potential of inadequate condensation in the barometric condenser due to lower heat transfer caused by the elevated torus temperatures D. Correct:
HC.OP-EO.ZZ-0101, caution #4 states, "Operation of HPCI OR RCIC turbines with suction temperatures above 170°F may result in equipment damage".
lAW with the bases document, it states, "Caution
#4 warns against operating HPCI and RCIC with high suction temperatures due to higher than normal oil temperatures".
This is caused due to the lube oil coolers being cooled by the suction source, resulting in less cooling to the lube oil.
==References:==
HC.OP-EO.ZZ-01 01, Rev. 11 Student Ref: NONE Learning Objective:
E01 01 PE006 Question Source: HC bank 109059 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
1 OCFR55: CFR 41.10/43.5/45.13 Comments:
QUESTION 6 Given the following conditions:
* Fire and smoke in the control room has led to a control room evacuation.
* The Control Room is being evacuated per HC.OP-AB.HVAC-0002.
* HPCI and RCIC both automatically initiated and are injecting.
* Control has been established at the Remote Shutdown Panel (RSP).
* The RSP operator trips RCIC when reactor water level reaches +54 inches.
* No field actions of AB.HVAC-0002 have been completed.
Reactor water level will ... A. lower until RCIC automatically re-initiates at -38 inches. B. lower until HPCI automatically re-initiates at -38 inches. C. lower until both HPCI and RCIC automatically reinitiate.
D. continue to rise due to HPCI injection.
K&A Rating: 295016 AK2.02 (4.0/4.1)
K&A Statement:
Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:
Local control stations:
Plant-Specific Justification:
A. Incorrect but plausible:
Plausible if the applicant does not recall that RCIC auto start is disabled at RSP when the transfer switches are placed in the EMER. B. Correct:
HPCI will continue to cycle between -38 inches and +54 inches. C. Incorrect but plausible:
Plausible if the applicant does not recall that RCIC auto start is disabled at RSP when the transfer switches are placed in the EMER. D. Incorrect but plausible:
Plausible if the applicant does not recall that HPCI auto trip at +54 inches is still in effect.
==References:==
HC.OP-IO.ZZ-0008(Q),
Rev. 31 HC.OP-AB.HVAC-0002(Q),
Rev. 7 Learning Objective:
RCICOOE004 Question Source: New Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Student Ref: NONE Comprehensive/Analysis:
X 10CFR: CFR 41.7/45.8 Comments:
QUESTION 7 Given the following conditions:
* The plant is operating at 100% rated thermal power
* A service water pipe on the 77' level of the Reactor Building develops a large flange leak
* The RACS ROOM FLOODED alarm is received on all channels
* All expected automatic actions occur
* The EO reports 2" of water covering the entire RACS Pump and HX room floor, with the rate of rise slowing Which ONE of the following actions is required lAW HC.OP-AB.COOL-003
'Reactor Auxiliary Cooling'?
A. Begin a normal reactor shutdown at 15% per hour. B. Manually close valve 1-EA-V453 SSW to RACS Outboard Isolation valve. C. Remove the floor drain plugs and pump out the room via the SW Dewatering pump. D. Reduce Reactor Recirc Pump Speed to minimum and lock the mode switch in SHUTDOWN.
K&A Rating: K&A Statement:
Justification:
2950 18AK2. 02 (3.4) Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER and the following:
AK2.02 Plant Operations A. Incorrect but plausible:
A total loss of RAGS has occurred.
The amount of water on the floor prevents recovery of SSW to RAGS. Action for flooding in rooms from EOP-103 Table 2. B. Incorrect but plausible:
Manual closing of 1-EA-V453 directed if HV-2346 fails to close. Not true with given stem conditions.
C. Incorrect but plausible:
Floor drains and SSW dewatering tank are not interconnected.
D. Correct:
Since all automatic actions occurred, SSW to RAGS is lost and a loss of RAGS has occurred.
Retainment override of AB-COOL-0003 reduce RRPs to minimum and lock Mode Switch in S/D.
==References:==
AB-COOL-0003 Student Ref: NONE Learning Objective:
ABCOL3E001 Question Source: Bank Question History:
HC Bank #308 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 8 Given the following conditions:
Hope Creek is starting up from a Refueling outage. The plant is currently in Operational Condition 3 with RCS temperature at 240°F. Instrument Air pressure is 105 psig. The Instrument/Service Air Compressors are aligned as follows:
Compressor Control Mode OOK107 LEAD 10K107 LAG 10K100 AUTO Subsequently, 7.2KV Bus 1 OA 120 loses power. Status Running OFF OFF Assuming NO operator
: actions, which ONE of the following describes the response of the Instrument/Service Air systems?
A Instrument Air header pressure remains at 105 psig. B. Instrument Air header pressure drops to 92 psig, when Service Air Compressor 1 OK1 07 starts and returns pressure to -105 psig. C. Service Air receiver pressure drops to 92 psig, when Service Air Compressor 1 OK1 07 starts and returns pressure to -105 psig. D. Service Air receiver pressure drops to 85 psig when Emergency Air Compressor 1 OK1 00 starts and returns pressure to -95 psig.
Question 8 K&A Rating: 295019 Partial or Complete Loss of lnst. Air AA 1.03 (3.0) K&A Statement:
Ability to operate and/or monitor the following as they apply to Partial or Complete Loss of lnst. Air: Instrument Air Compressor power supplies.
Justification:
A. Incorrect but plausible:
Power to SAC 10K107 is from 7.2 KV bus 10A110, NOT 10A120 B. Incorrect but plausible:
SAC 1 OK1 07 will NOT start at 92 psig in the instrument air header. It starts when the service air receiver is 92 psig. I. C. Correct:
SAC 1 OK1 07 will start when the service air receiver pressure is 92 psig .. D. Incorrect but plausible:
Loss of Power to 1 OA 120 causes a loss of Power to SAC OOK107, Instrument Air header will not drop to 85 psig, because 10K107 will prevent the pressure getting that low. Plausible because EIAC 1 OK1 00 would start at 85 psig.
==References:==
NOH01SERAIR-01, SERVICE AIR SYSTEM, p.47-48 NOH011NSAIR-01, INSTRUMENT AIR SYSTEM, p15, 42 Student Ref: NONE Learning Objective:
Question Source: New Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR41.7/45.6 Comments:
QUESTION 9 Given the following conditions:
* The reactor has been shutdown for 90 hours following 1000 EFPD of operation.
* The plant is in Cold Shutdown with RPV metal and RCS temperature of 140°F.
* A total loss of Shutdown Cooling occurred at 1200 hours.
* All efforts to restore heat removal from the RPV have failed.
* Both Recirculation pumps have been secured.
Assuming NO additional operator action, when will the plant reach OPCON 3? A. 1245 B. 1307 C. 1330 D. 1352 (Note: See attached Figure 1) 0 ... ..... LL. OJ a>-c .VI ... 0 :::l 0 0 N.c 0 +' Q) E 1-FIGURE 1 Total Loss of Heat Removal from Rx Vessel 1000 EFPD of Operations 5.000
-'!"'_,_..,...,...,..
___
I 100"F IY 4.000 ...,.. __ ...,__.....,.._,_""'1
_,.-!-.,.11 IL r'2o*F / v 3.ooo ..,.. __
... I L ,4"FI r-I 0.000 ...._ __ ....... _ .............
1_...,1 .... 1....._1
.....,. .... 1 ___ .....__ ... 1_.._ ... ,_.. ..............
10 100 Time from Rx Shutdown (hours) ** Represents initial starting temperature as listed in the Figure. Each line represents a different initial starting temperature 1000 Question 9 K&A Rating: 295021 Loss of Shutdown Cooling AA2.01 (3.5) K&A Statement:
Ability to determine and/or interpret the following as they apply to Loss of Shutdown Cooling:
Reactor water heatup/cooldown rate. Justification:
A. Incorrect but plausible:
Value obtained by using the 160°F curve. B. Correct::
Operational Condition 3 is achieved when the Reactor temperature reaches 200°F. The 140°F curve of Figure 1 intersects the 90-hour line between the 1.000 and 1.250 hour lines. 1307 is the only option that is between 1 hour and 1 hour and fifteen minutes following the loss of SOC. C. Incorrect but plausible:
Value obtained by using the 120°F curve. D. Incorrect but plausible:
Value obtained by using the 1 00°F curve.
==References:==
HC.OP-AB.RPV-0009, Figure 1and Technical Specification Table 1.2 Student Ref: Figure 1 of HC.OP-AB.RPV-0009 Learning Objective:
Question Source: HC.OP-AB.RPV-0009, Figure 1and Technical Specification Table 1.2 Question History:
2005 HC NRC Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 10 The plant is in OPCON 1 with irradiated fuel being shuffled in the Spent Fuel Pool when the following VALID alarms are received in the Control Room:
* RADIATION MONITORING ALARM/TRBL
* NEW FUEL CRITICALITY RAD HI
* REFUEL FLR EXH RAD ALARM/TRBL
* RB EXH RADIATION ALARM/TRBL Which ONE of the following is the appropriate action lAW HC.OP-AB.CONT-0005, IRRADIATED FUEL DAMAGE'?
A. Verify Reactor Building Ventilation isolates and fans trip. B. Verify the Drywell Integrity Airlock surveillance test is current.
C. Verify "A" and "C" SACS pumps start if not already running.
D. Verify Emergency Instrument Air Compressor starts if not already running.
K&A Rating: K&A Statement:
Justification:
295023AK3.01 (3.3/3.6)
Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS:
AK3.03 Ventilation Isolation A. Correct:
Hi alarms on Reactor building and RF floor indicate an isolation signal and supply and exhaust fans trip. B. Incorrect but plausible:
The concern is secondary not primary containment.
C. Incorrect but plausible:
The "A" and "B" SACS pump will start if not running, not the "A" and "C" SACS pumps. D. Incorrect but plausible:
The EAIC would trip on the High Radiation Signal.
==References:==
HC.OP-AB.CONT-0005, Rev 4 Student Ref: NONE Learning Objective:
ABCNT5E004 Question Source: Modified from HC Bank (Question
#405) Question History:
Modified from bank question used on 2010 HC Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 11 A transient has occurred.
The current containment conditions are as follows:
* Drywell pressure:
4.5 psig rising
* Drywell temperature:
150 F rising
* Torus pressure:
4.0 psig rising
* Torus water temperature:
82 F steady
* Torus airspace temperature:
145 F rising Which of the following events would explain the current plant conditions?
A. A safety relief valve (SRV) has lifted and is discharging through a T-Quencher.
B. The containment is functioning as designed following a feedwater line break inside containment.
C. A main steam line break has occurred inside containment with a Torus to drywell vacuum breaker open. D. The containment is functioning as designed following a total loss of drywell cooling.
QUESTION 11 K&A Rating: K&A Statement:
Justification:
295024 High Drywell Pressure EA2.05 (3.6) Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:
EA2.05 Suppression chamber air-space temperature:
Plant Specific.
A Incorrect but plausible:
Since the Torus airspace temperature would not be rising with the discharge of the SRV into the water portion of the torus. B. Incorrect but plausible:
Since the Torus airspace temperature would not be rising if the containment was responding normally.
C. Correct:
Drywell and torus air space trending together is indication that there is a bypass of the containment, which in this case is the torus to drywell vacuum breaker being open. D. Incorrect but plausible:
Since the Torus airspace temperature would not be rising with the loss of drywell cooling.
==References:==
NOHOIPRICONC, "Primary Containment Structure" Student Ref: NONE Learning Objective:
PRICONE008 Question Source: Mod INPO Bank PB2 Question Question History:
2007 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 12 Which ONE of the following describes the reason for sustained (vice intermittent)
SRV openings during performance of HC.OP-EO.ZZ-01 01, 'RPV CONTROL' if the pneumatic supply is lost to the SRVs? Sustained SRV openings
... A. conserve SRV pneumatic supply for sufficient cooldown to less than the shutdown cooling high pressure interlocks.
B. prevent exceeding the 100°F/hr cooldown limit during depressurization while conserving the SRV pneumatic supply. C. conserve SRV accumulator pneumatic supply for later use if an Emergency Depressurization is required.
D. allow the operator to depressurize without regard to the Technical Specification cooldown limits before the pneumatic supply is depleted.
K&A Rating: 295025 2.1.28 (4. 1/4.1) K&A Statement:
High Reactor Pressure 2.1.28: Knowledge of the purpose and function of major system components and controls.
Justification:
A. Incorrect but plausible:
Sustained opening of the SRVs is specified in step RC/P-6 to conserve accumulator pressure for a subsequent Emergency Depressurization, NOT to allow for further cooldown.
B. Incorrect but plausible:
Although there is a 1 OOF/hr cooldown rate limit applied in step RC/P-6, sustained opening of the SRV vice intermittent opening is not used to control cooldown rate. C. Correct:
lAW HC.OP-EO.ZZ-0101 bases step RC/P-6, sustained SRV opening conserves accumulator pressure when the source of pressure to the SRV pneumatic supply is isolated or otherwise out of service.
Such action to reduce the number of cycles on the SRVs prolongs SRV availability should more degraded conditions later required SRVs be opened for rapid depressurization of the RPV. D. Incorrect but plausible:
HC.OP-EO.ZZ-01 01 step RC/P-6 limits cooldown rate to 1 OOF/hr.
==References:==
HC.OP-EO.ZZ-0101 Bases Document Learning Objective:
E01 01 PE005 Question Source: HC Bank #131 Question History:
Not used on previous 2 NRC exams Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
10CFR: CFR 41.7 Comments:
Student Ref: NONE QUESTION 13 The plant is at rated power with the following conditions:
* HPCI testing is in progress lAW HC.OP-IS.BJ-0001 "HPCI Main and Booster Pump Set-OP204 and OP217 -lnservice Test"
* Torus level is 75.5 inches
* Div 1 average suppression pool temperature indicates 96°F and is slowly trending higher
* Div 2 average suppression pool temperature indicates 98°F and is slowly trending higher WHICH ONE of the following describes the required action(s)?
A. Enter HC.OP-EO.ZZ-01 02 "Primary Containment Control" AND immediately suspend testing.
B. Place the Mode Switch in Shutdown and place Suppression Pool cooling in service.
C. Enter HC.OP-EO.ZZ-0102 "Primary Containment Control".
Suspend testing before average suppression pool temperature reaches 1 05°F. D. Do NOT enter HC.OP-EO.ZZ-01 02 "Primary Containment Control".
Suspend testing before average suppression pool temperature reaches 1 05°F.
K&A Rating: K&A Statement:
Justification:
295026 EA 1.03 (3.9/3.9)
Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:
EA1.03 Temperature Monitoring A. Incorrect but plausible:
If applicant believes that EOP entry is required.
EOP entry is not required until 105°F while performing testing lAW HC.OP-IS.BJ-0001.
Additionally, testing is not required to be suspended until 1 05°F suppression chamber average water temperature lAW TS 3.6.2.1 B. Incorrect but plausible:
If applicant believes that the Mode Switch is required to be placed in Shutdown due to one temperature above 11 0°F. This is only required if the Suppression Pool average water temperature exceeds 11 ooF C. Incorrect but plausible:
If applicant believes that EOP entry is required.
EOP entry is not required until 1 05°F while performing testing lAW HC.OP-IS.BJ-0001.
D. Correct:
EOP entry is not required until 105°F while performing testing lAW IS. BJ-0001.
Additionally, testing is not required to be suspended until1 05°F suppression chamber average water temperature lAW TS 3.6.2.1
==References:==
HC.OP-IS.BJ-0001, Rev. 60 TS 3.6.2.1, Amendment 110 Learning Objective:
Question source: Modified Limerick 2008 Question History:
None Applicant Ref: NONE Cognitive level: Memory/Fundamental knowledge:
Comprehensive/Analysis:
X 1 OCFR Part 55: 41.7/45.6 Comments:
QUESTION 14 SELECT the reactor pressure, drywell temperature and level instrument that would indicate valid reactor water level of ( -1 00) inches. A. RPV pressure is 300 psig, Drywell temperature is 400 degrees F on SPDS point A2266, Narrow Range A. B. RPV pressure is 200 psig, Drywell temperature is 350 degrees F on SPDS point A2277, Wide Range B. C. RPV pressure is 100 psig, Drywell temperature is 350 degrees F on SPDS point A2281, Upset Range. D. RPV pressure is 100 psig, Drywell temperature is 250 degrees F on SPDS point A2280, Fuel Zone A. RPV Saturation Temperature 600 G:' 550 L. c: :I 0:: 500 i: Ill E 450 .s "' .s 400 lij QJ z 350 ::1 300 Ill a. E Ill 250 r-.....---_.,..,.,..
./ / ., /"' I ! ' I 200 0 100 200 300 400 500 600 700 800 900 1 000 11 00 RPV Pressure (psi g) -Results K&A Rating: 295028 EK1.01 (3.5/3.7)
K&A Statement:
Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE:
Reactor water level measurement Justification:
A. Incorrect but plausible:
RPV pressure-300 psig, Drywell temperature-400 degrees F on SPDS point A2266, Narrow Range A-Within the curve, but 100 inches is above the top of the narrow range band (0-60 inches).
B. Correct:
pressure-200 psig, Drywell temperature-350 degrees F on SPDS point A2277, Wide Range B-Within the curve and A2277 is the nearest SPDS point to Wide Range B. C. Incorrect but plausible:
RPV pressure-100 psig, Drywell temperature-350 degrees F on SPDS point A2281, Upset Range-Outside of the curve. D. Incorrect but plausible:
RPV pressure-100 psig, Drywell temperature-250 degrees F on SPDS point A2280, Fuel Zone A-Within the curve, outside of the range of the Fuel Zone ( -111 to -311 inches).
==References:==
EOP Caution 1 HC.OP-EO.ZZ-LIMITS-CONV, Rev. 5 Learning Objective:
E01 01 LE007 Question Source: Bank# 33436 Question History:
Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR41.8, 41.9, 41.10 Comments:
Student Ref: NONE QUESTION 15 Given the following:
* A leak has developed in the Torus requiring emergency makeup
* Torus level is 71 inches and lowering slowly
* The Condensate transfer lines to Core Spray Suctions are NOT available Emergency Makeup to the Suppression Pool via RCIC is accomplished by which ONE of the following?
A. Running RCIC with the full flow test flowpath open. B. Running RCIC with the min flow discharge flowpath open. C. Overriding and opening both RCIC suction MOVs simultaneously.
D. Running the RCIC Jockey Pump with the Suppression Pool suction MOV open.
K&A Rating: 295030 2.4.11 (4.0) K&A Statement:
Low Suppression Pool Water Level 2.4.11: Knowledge of abnormal condition procedures.
Justification:
A. Incorrect but plausible:
Running RCIC with the full flow test flowpath open would only recirculate CST water. B. Correct:
Running RCIC with the min flow discharge flowpath open, lAW EOP-313 will provide emergency makeup water to the suppression pool. C. Incorrect but plausible:
Overriding and opening both RCIC suction MOVs simultaneously will not provide emergency makeup water to the suppression pool. There is a check valve in the suppression pool suction path that prevents gravity draining the CST into the suppression pool. D. Incorrect but plausible:
Running the RCIC jockey pump with the suppression pool suction MOV open will not provide water to the suppression pool. There is a check valve between the suppression pool and the jockey pump suction line, and the flow direction of the jockey pump is incorrect.
==References:==
EOP-313 Student Ref: NONE Learning Objective:
EOP300E004, E0101AE006 Question Source: HC Bank # 184 Question History:
Used on 1999 Audit Exam Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
1 OCFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 16 Given the following conditions:
* The plant has experienced a LOCA with a loss of ALL injection.
* All Control Rods have inserted.
* RPV pressure is being controlled with SRVs.
* RPV water level has lowered to -196 inches (Corrected Fuel Zone). What is the status of core cooling?
Adequate core cooling exists ... A. only if injection is established at this water level. B. only when the SRVs are closed. C. at this RPV water level. D. only if RPV water level is raised 10 inches.
K&A Rating: K&A Statement:
Justification:
295031 EA2.04 (4.6/4.8)
Ability to determine and/or interpret the following as they apply to Reactor Low Water Level EA2.04 Adequate Core Cooling A. Incorrect but plausible:
Injection at this level would reduce steam generation needed to assure adequate core cooling.
B. Incorrect but plausible:
Adequate steam generation is present whether SRVs are open or closed. C. Correct:
At levels above -200 inches with no injection there is sufficient steam flow to provide adequate core cooling.
D. Incorrect but plausible:
Because compensated level is above -200, adequate core cooling exists at this level.
==References:==
Student Ref: NONE Learning Objective:
E01 01 LE006 Question Source: INPO Bank Q101 HC Bank Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
1 OCFR: CFR 41.10/43.3/45.13 Comments:
QUESTION 17 Unit 1 was at 100% power when the following occurred:
* A loss of the 1 OA 1 02 bus resulted in the loss of two RFPTs.
* The remaining RFPT was unable to maintain RPV water level.
* The reactor scrammed on RPV L VL 3.
* Numerous control rods remained withdrawn with power at 5%.
* RPV water level lowered to -110".
* HPCI and RCIC initiated, and RPV water level then started to rise. Current conditions are:
* Reactor power is 5% and rising 1% every two minutes due to RPV water injection.
* RPV water level is -11 0" and rising at 15"/min.
* RPV pressure is 920 psig and steady.
* NO operator actions have been taken. Which ONE of the following describes the associated RRCS function status four minutes later? A. A feedwater runback has NOT initiated, RPT breakers are closed. B. ARI valves are open, SLC pumps did NOT initiate.
C. RPT breakers are tripped, SLC pumps have initiated.
D. A feedwater runback has initiated, ARI valves are closed.
K&A Rating: 295037EK2.02 (4.0/4.2)
K&A Statement:
Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following:
RRCS Justification:
A Incorrect but plausible:
The -38" RPT breaker trip signal is a seal in signal. 9 seconds after water level dropped below -38", the RPT breakers will trip B. Incorrect but plausible:
The -38" level to the SLC pump initiation circuit does NOT seal in. However, starting from -11 0", rising at 15" per minute, RPV water level will be -50" in four minutes.
Since RPV water level will be below -38" when the 230 second timer times out, SLC pumps will initiate C. Correct:
The -38" RPT breaker trip signal is a seal in signal. 9 seconds after water level went below -38", the RPT breakers will trip. The -38" level to the SLC pump initiation circuit does NOT seal in. However, starting from -11 0", rising at 15" per minute, RPV water level will be -50" in 240 seconds.
Since RPV water level will be below -38" when the 230 second timer times out, SLC pumps will initiate D. Incorrect but plausible:
A feedwater runback is initiated only by reactor pressure reaching 1071 psig. There is NO reason for reactor pressure to have reached 1071 psig during the lowering level transient.
Additionally, ARI valves opened immediately at -38" and do NOT automatically reset
==References:==
NOH04RRCS Student Ref: NONE Learning Objective:
RRCSOOE007 Question Source: HC Bank (Question ID#65237)
Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7/45.8 Comments:
QUESTION 18 HC.OP-EO.ZZ-0103/4, Reactor Building
& Rad Release Control, step RR-5, directs isolation of all primary systems discharging into areas outside Primary Containment or Reactor Building, except those systems required to assure adequate core cooling and/or shutdown the reactor.
In accordance with the EOP Bases document, HC.OP-EO.ZZ-1 03/4. Reactor Building
& Rad Release Control, these systems are specifically exempted from isolation, because ... A. additional radiological consequences from them are unlikely.
B. they are required to support alternate reactor depressurization methods.
C. isolation of a EOP support system requires an upgrade of the Emergency Classification.
D. systems operated for RPV control are given a higher priority than stopping a rad release.
Question 18 K&A Rating: 295038 High Off-site Release Rate EK3.02 (3.9) K&A Statement:
Knowledge of the reasons for the following responses as they apply to High Off-Site Release Rate: System Isolations Justification:
A. Incorrect but plausible:
NOT in accordance with bases document B. Incorrect but plausible:
NOT in accordance with bases document C. Incorrect but plausible:
NOT in accordance with bases document D. Correct:
Per EOP Bases document 103/104:
The objectives of RPV Control, Primary Containment
: Control, and the EPG contingencies are given higher priority than the steps of the objectives of Radioactivity Release Control.
Systems that must be operated to perform other EPGs are therefore NOT isolated in this step.
==References:==
BWROG, EPGs/SAGs Appendix B, section 9 Radioactivity Release control HC.OP-EO.ZZ-103/4.
Reactor Building
& Rad Release Control Bases Document-
: p. 13 & 14 Student Ref: NONE Learning Objective:
EOP1 03E006 Question Source: HC NRC Exam 2005 Question History:
HC NRC Exam 2005 Cognitive Level: 10CFR: Comments:
Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
CFR 41.10/43.5/45.13 QUESTION 19 Given:
* The plant Is operating at 100% power
* A loss of MCC OOB590 occurs.
* Field operator confirms NO fire onslte. Based on the above, it will be necessary to ... A. secure the motor driven fire pump to prevent pumping down the fire water storage tank. B. manually start the electric fire pump due to a loss of the diesel driven fire pump battery chargers.
C. manually start the diesel fire pump due to a loss of the motor driven fire pump power supply. D. secure the diesel driven fire pump to prevent pumping down the fire water storage tank.
K&A Rating: 600000 AK3.04 (2.8/3.4)
K&A Statement:
Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE: Actions contained in the abnormal procedure for plant fire on site Justification:
A. Incorrect but plausible:
Plausible if the applicant does not recall that the motor driven fire pump has no power, it is not running.
B. Incorrect but plausible:
Plausible if the applicant does not recall that the battery chargers are powered from 1AJ483. C. Incorrect but plausible:
Plausible if the applicant does not recall that the diesel driven fire pump auto starts on a loss of power to the motor driven fire pump. D. Correct:
The diesel driven fire pump auto starts on a loss of power to the motor driven fire pump. Since there is no fire, the fire water storage tank is being pumped down.
==References:==
HC.OP-AR.QK-0002 Student Ref: NONE Learning Objective:
FIRPROE008 Question Source: HC Bank Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.7/45.8 Comments:
QUESTION 20 The plant was operating at full power when a grid disturbance results in the actuation of a Main Generator Differential Overcurrent Lockout condition.
The Main Turbine trips and the Alterex exciter field breaker trips. Which ONE of the following describes the plant response on the below listed equipment?
A. lsophase Bus Duct Cooling Fans are unaffected.
Stator Water Cooling Pumps TRIP. Generator output breakers BS6-5 & BS2-6 remain closed. B. lsophase Bus Duct Cooling Fans are unaffected.
Stator Water Cooling Pumps TRIP. Generator output breakers BS6-5 & BS2-6 TRIP. C. lsophase Bus Duct Cooling Fans TRIP. Stator Water Cooling Pumps are unaffected.
Generator output breakers BS6-5 & BS2-6 TRIP. D. lsophase Bus Duct Cooling Fans TRIP. Stator Water Cooling Pumps are unaffected.
Generator output breakers BS6-5 & BS2-6 remain closed.
K&A Rating: K&A Statement:
Justification:
700000 AK2.02 (3.1/3.3)
Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following:
AK2.02 Breakers, Relays A. Incorrect but plausible:
Generator output breakers BS6-5 & BS2-6 TRIP. B. Correct:
The plant was operating at full power when a grid disturbance results in the actuation of a Main generator differential overcurrent lockout condition.
The Main generator trips and the Alterex field breaker trips. The following automatic actions are initiated when the Main Generator regular lockout relay is actuated due to a generator differential overcurrent condition:
Trips and blocks closing of gen output breakers BS6-5 & BS2-6. C. Incorrect but plausible:
lsophase bus duct cooling fans are unaffected, stator water cooling pumps trip. D. Incorrect but plausible:
lsophase Bus Duct cooling fans are unaffected, Stator Water Cooling Pumps TRIP, generator output breakers BS6-5 & BS2-6 TRIP.
==References:==
HC.OP-AR.ZZ-0015 Student Ref: NONE Learning Objective:
MNGENOE011 Question Source: HC Bank ID 120299 Question History:
Used on 2010 NRC Audit Exam Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
10CFR55:
CFR 41.4/41.5/41.7/41.10/45.8 Comments:
QUESTION 21 Given the following:
* The reactor has scrammed on high Drywell pressure.
* Drywell Floor Drain Sump pumps (CP267/DP267) have stopped running.
* Drywell pressure continues to rise.
* No operator actions have been taken. Under these conditions, which ONE of the following caused the Drywell Floor Drain Sump pumps to stop running?
A. The Reactor Recirc Seal Staging flow is isolated.
B. The HB-HV-5258 DRYWELL FLOOR DRN PMPS DSCH VLV has failed closed. C. The Drywell Leak Detection (DLD) Sump Monitoring System has failed. D. The power source to the pumps is load shed.
K&A Rating: K&A Statement:
Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE:
AA1.02 Drywell floor and equipment drain sumps Justification:
A. Incorrect but plausible:
Reactor Recirc Seal Staging flow is directed to the Drywell Equipment Drain Sump. Loss of this flow input would have no effect on the DWFDS. B. Incorrect but plausible:
The HV-5258 is interlocked to open on a start of either CP267 or DP267; however, it is NOT a permissive for the pumps to run. If it failed shut it would prevent the pumps from pumping, but it would NOT prevent them from running.
(The pumps DO receive a permissive to run from the drywell containment isolation valves F003/F004).
C. Incorrect but plausible:
The OLD SMS does not control operation of the drywell sump pumps. D. Correct:
The power supply to the CP267 and DP267 are 252064 and 262064, respectively.
These MCCs are load shed on High Drywell Pressure.
==References:==
E-0023-1 Sheet 1 Student Ref: NONE HC.OP-SO.SM-0001 Table SM-20 HC.OP-AR.ZZ-0014 Attachment D3-C3 and 02363. Learning Objective:
L/R/A/ Drywell Floor Drain Sump pumps trip Question Source: HC Bank #58 Question History:
NRC EXAM 2002 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.7/45.6 Comments:
QUESTION 22 Given the following conditions:
* A plant startup is in progress following a forced outage.
* The plant has been operating with a known fuel leak.
* The plant scrammed 40 hours ago.
* The "A" Mechanical Vacuum Pump (MVP) is placed in service with the suction valve throttled.
* The Main Condenser Vacuum Breakers are closed. lAW HC.OP-SO.CG-0001,
'Condenser Air Removal Operation',
which ONE of the following actions is permitted if the South Plant Vent (SPV) RMS Effluent monitor reaches the ALERT alarm set point and condenser vacuum cannot be maintained?
A. Stop the MVP. B. Open the Main Condenser Vacuum Breakers.
C. Throttle the MVP Suction valve further closed. D. Increase MVP flow to the High Alarm setpoint.
K&A Rating: 295017AK1.02 (3.8) K&A Statement:
Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE: AK1.02 Protection of the general public Justification:
A. Incorrect but plausible:
Stopping of the MVP would only be required if the HIGH SPV RMS Effluent monitor set point were reached.
B. Incorrect but plausible:
Not required per HC.OP-SO.CG-0001.
Additionally, opened the Main Condenser Vacuum Breakers would increase effluent flow. C. Incorrect but plausible:
lAW HC.OP-SO.CG-0001, Caution 5.8.13, the MVP suction is throttled when the HIGH alarm setpoint is reached.
D. Correct:
lAW HC.OP-SO.CG-0001, Caution 5.8.13, if condenser vacuum cannot be maintained with the flow reduced such that the SPV RMS Effluent ALERT setpoint is not exceeded, flow may be increased but not to exceed the HIGH alarm setpoint (which ensures that ODCM release rate limits are not exceeded).
==References:==
HC.OP-SO.CG-0001, Rev 45, Section 5.8.13 Student Ref: NONE Learning Objective:
ABBOP6E001 Question Source: Modified from HC Bank (Question
#261) Question History:
Original question used on 2003 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 23 The plant was operating at 100% power when an inadvertent PCIG Isolation occurred due to multiple instrument failures.
lAW EOP-101 "RPV Control",
and assuming NO SCRAM ACTIONS or OTHER OPERATOR ACTIONS are taken for 40 minutes, which ONE of the following, if any, is available to initiate a controlled RPV depressurization?
A Manual operation of SRVs only. B. Manual operation of Bypass Valves only. C. Both SRVs and Bypass Valves. D. Neither SRVs nor Bypass Valves.
K&A Rating: 295020 AK2.01 (3.6/3.7)
K&A Statement:
Knowledge of the interrelations between INADVERTENT CONTAINMENT ISOLATION and the following:
Main steam system Justification:
A. Correct:
Inboard MSIVs will close due to their accumulators bleeding down(-30 min), rendering the Bypass Valves unable to support pressure control.
: However, the (ADS) SRVs can be operated manually due to the pneumatic supply accumulators.
B. Incorrect but plausible:
Plausible if candidate does not know that Inboard MSIVs will close due to loss of Instrument Nitrogen.
Inboard MSIVs will close due to their accumulators bleeding down(-30 min), rendering the Bypass Valves unable to support pressure control.
C. Incorrect but plausible:
Plausible if candidate does not know that Inboard MSIVs will close due to loss of Instrument Nitrogen.
Inboard MSIVs will close due to their accumulators bleeding down(-30 min), rendering the Bypass Valves unable to support pressure control.
D. Incorrect but plausible:
Plausible if candidate does not know that Inboard MSIVs will close due to loss of Instrument Nitrogen.
Inboard MSIVs will close due to their accumulators bleeding down (-30 min), rendering the Bypass Valves unable to support pressure control.
==References:==
NOH01 MSTEAMC-08 NOH01 PCIGOOC-06 NOH04NSSSSOC-04 Learning Objective:
MSTEAME003, PCIGOOE008, Question Source: New Question History:
None Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
10CFR: CFR 41.7/45.8 Comments:
QUESTION 24 Plant conditions are as follows:
* Reactor Power is 1 00%
* Torus water level is 81 inches and rising slowly The CRS directs lowering Torus water level using B RHR to Radwaste.
Which ONE of the following describes the reason for this action? A. Torus to Drywell Vacuum Breakers will be covered B. HPCI Turbine exhaust back pressure limits could be exceeded C. RCIC Turbine exhaust back pressure limits could be exceeded D. Primary Containment pressure could be exceeded during a subsequent LOCA K&A Rating: K&A Statement:
295029 EK3.02 (3.6/4.0)
Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL WATER LEVEL: EK3.02 Lowering suppression pool water level e-orred-vacuum breakers occurs at 124 inches. B. Incorrect but plausible:
While plausible due to the increased backpressure felt by HPCI due to increased Torus water level and possibly primary containment
: pressure, there is no limit associated with HPCI exhaust backpressure.
C. Incorrect but plausible:
While plausible due to the increased backpressure felt by RCIC due to increased Torus water level and possibly primary containment
: pressure, this could only result in a trip of the RCIC turbine on high exhaust pressure.
There is no limit associated with RCIC exhaust backpressure D. Correct:
Per HC.OP-EO.ZZ-01 02, Primary Containment
: Control, step SP/L-24, "SRV operation with suppression pool water level above the SRV Tail Pipe Level Limit (STPLL) could damage the SRV discharge lines. This, in turn, could lead to containment failure from direct pressurization and damage to equipment inside the containment (ECCS piping, RPV water level instrument runs, wetwell-to-dry well vacuum breakers, etc.) from pipe-whip and jet-impingement loads. The RPV is therefore not permitted to remain at pressure if suppression pool water level and RPV pressure cannot be restored and maintained below the STPLL." With the given plant conditions, STPLL will be violated at approximately 11 0 inches.
==References:==
HC.OP-EO.ZZ-0102, Rev. 12 Student Ref: NONE Learning Objective:
EOP1 02E009 Question Source: Bank, LGS Bank 575226 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
10CFR55:
CFR 41.5/45.6 Comments:
QUESTION 25 Hope Creek is at 100% power when ONLY the following overhead alarms are received:
* 81-85 HPCI OUT OF SERVICE
* 03-A 1 HPCIIRHR A LEAK TEMP HI When the operator checks the HPCI panel, HPIC inboard and outboard steam line isolation valves are stroking closed.
* The HPCI turbine was NOT running at the time.
* HPCI Steam line pressure is 900 psig and lowering.
Which ONE of the following would cause this isolation?
A. HPCI Pipe Chase High Temperature.
B. HPCI Pump Room High Temperature.
C. HPCI Steam Line Low Supply Pressure.
D. HPCI Steam Line High Differential Pressure.
K&A Rating: K&A Statement:
Justification:
295032EA2.01 (3.8/3.8)
Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE:EA2.01 Area Temperature A. Incorrect but plausible:
Isolates after a 15min time delay. Would also cause annunciator D3-B1 HPCI STM LK ISLN TIMER INITIATED.
B. Correct:
Isolates immediately on room high temperature.
High HPCI room temperature causes D3-A 1 annunciator.
This creates a HPCI turbine trip, which causes B 1-B5 annunciator.
B1-A4 HPCI TURBINE TRIP annunciator does not alarm because the F001 is closed (HPCI not in service).
C. Incorrect but plausible:
Isolates at 100 psig Reactor pressure.
Pressure is at least 900 psig in the Reactor D. Incorrect but plausible:
Would cause B1-A4 annunciator, HPCI STEAM LINE DIFF PRESSURE HI.
==References:==
HC.OP-AR.ZZ-0014 Student Ref: NONE Learning Objective:
EOP1 03E006 Question Source: HC Bank Q3552 Question History:
NRC 2003 Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 26 Given the following conditions:
* Reactor Building Ventilation is in a normal lineup.
* Reactor Building differential pressure is negative at 0.55 inches water gauge. Which ONE of the following will cause a degradation of Reactor Building differential pressure?
A. Level 3 Low RPV water level. B. Inadvertent actuation of the "A" RHR Room Blowout Panel. C. Inadvertent closure of GUD-925 (Rx building supply manual damper).
D. Refuel Floor Vent Exhaust effluent monitor reading 1.5 x 1 o-3 K&A Rating: K&A Statement:
Justification:
295035 2.2.44 ( 4.2) Secondary Containment High Differential Pressure:
2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
A. Incorrect but plausible:
RBVS automatically isolates at Level 2 low Reactor Water level (-38"), not Level 3. B. Incorrect but plausible:
The ECCS blowout panels relieve pressure from the ECCS spaces to the torus space, where increased pressure can be manually vented out of the steam vent. An inadvertent opening of the 'A' RHR blowout panel will not change the overall pressure inside the secondary containment.
C. Correct:
Inadvertent closure of GUD-925 will trip the RBVS Supply fans causing degradation in secondary containment 0/P. This is an actual event documented in LER-00-009 and specifically covered in training material (NOH01 SECCONC-04
). D. Incorrect but plausible:
The Refuel Floor Vent Exhaust monitor automatically isolates RBVS at 2.0 x 1 o-3 11Cilcc.
The Reactor Building Vent Exhaust monitor automatically isolates RBVS at 1.0 x 1 o-3
==References:==
HC.OP-SO.GR-0001 Student Ref: NONE LER 00-009 Learning Objective:
SECCONE004-008 Question Source: New Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 27 The crew needs to sample containment atmosphere using the H2/02 Analyzers to monitor for high containment hydrogen concentrations post LOCA. Which of the following describes:
(1) ALL sample locations monitored by the Hydrogen/Oxygen Analyzers, and (2) which containment isolation signals to the associated containment isolation valves can be overridden to ensure availability for sampling?
A. (1) The upper drywell and the torus, only (2) Only isolation signals to the upper drywell sample valves can be over ridden B. (1) The upper drywell and the torus, only (2) Isolation signals to ALL the sample valves can overridden C. (1) The upper drywell, lower drywell, and the torus (2) Only isolation signals to the upper drywell sample valves can be overridden D. (1) The upper drywell, lower drywell, and the torus (2) Isolation signals to ALL the sample valves can be overridden K&A Rating: 500000 EK1.01 (3.3/3.9)
K&A Statement:
Knowledge of the operational implications of the following concepts as they apply to HIGH CONTAINMENT HYDROGEN CONCENTRATION:
EK1.01 Containment Integrity Justification:
A. Incorrect but plausible:
the lower drywell is also sampled B. Incorrect but plausible:
the lower drywell is also sampled and all containment isolation signals to the isolation can be manually overridden C. Incorrect but plausible:
all containment isolation signals to the isolation can be manually overridden D. Correct:
per LP NOH01 H202AN Each package (H202 analyzer) takes samples from three different locations; High -Drywell head region, Low -Drywell cylindrical region, Suppression Chamber Air Space. CIVs can be individually opened after the associated isolation override P.B. is depressed at (1 OC650E)
==References:==
LP NOH01 H202AN-03 Section III.A.2.a.2)
& III.B.1.e.3)
Learning Objective:
NOH01 H202AN-03 Obj. 2 & 3 Question Source: Question History:
Cognitive Level: 10CFR55:
Comments:
NRC 2010 None Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
CFR 41.8-41.7 Student Ref: NONE 
QUESTION 28 Given the following conditions:
* The plant is operating at 100 percent power when a loss of off-site power occurs
* All EDGs have begun their start sequence Then, a LOCA occurs while the EDGs are starting
* A valid LPCI initiation signal is present from all 4 divisions of LPCI logic Which ONE of the following describes when the RHR pumps will start? A. RHR pumps will start 5 seconds after the Core Spray Pumps start. B. RHR pumps will start immediately when the EDG output breakers close. C. RHR pumps will start 5 seconds after receiving the LPCI initiation signal. D. RHR pumps will start immediately when RPV pressure lowers to 450psig.
K&A Rating: K&A Statement:
Justification:
203000A4.01 (4.3/4.1)
Ability to manually operate and/or monitor in the control room:A4.01 Pumps A. Incorrect but plausible:
With off-site power unavailable, the RHR pumps will start immediately when their associated EDG output breaker closes, with an automatic initiation signal present.
Core Spray Pumps start after the RHR Pump starts. B. Correct:
RHR Pump D will start immediately when the D EDG output breaker closes. C. Incorrect but plausible:
With off-site power unavailable, the RHR pumps will start immediately when their associated EDG output breaker closes, with an automatic initiation signal present D. Incorrect but plausible:
With off-site power unavailable, the RHR pumps will start immediately when their associated EDG output breaker closes, with automatic initiation signal present.
==References:==
HC.OP-SO.BC-0001 Student Ref: NONE Learning Objective:
RHRSYSE008 Question Source: HC Bank #35561 Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.7/43.5/45.8 Comments:
QUESTION 29 Given the following conditions:
* The reactor is in cold shutdown with loop "B" of RHR in shutdown cooling.
* A break results in a loss of reactor coolant inventory.
* Water level has lowered to -150 inches. Select the correct RHR loop "B" status for these conditions.
A. RHR pump (BP202) will be running.
RHR Injection Valve (HV-F017B) will be full open. RHR Suppression Pool Suction Valve (HV-F004B) will be full open. B. RHR pump (BP202) will be tripped.
RHR Injection Valve (HV-F017B) will be full closed. RHR Suppression Pool Suction Valve (HV-F004B) will be full closed. C. RHR pump (BP202) will be running RHR Injection Valve (HV-F017B) will be full closed. RHR Suppression Pool Suction Valve (HV-F004B) will be full open. D. RHR pump (BP202) will be tripped.
RHR Injection Valve (HV-F017B) will be full open. RHR Suppression Pool Suction Valve (HV-F004B) will be full closed.
K&A Rating: K&A Statement Justification:
205000A2.05 (3.5/3.7)
Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on those predictions, use procedures to correct,
: control, or mitigate the consequences of those abnormal conditions or operations A2.05 System Isolation A. Incorrect but plausible:
Suction valves being closed will prevent the pump starting.
Suppression pool suction will not receive an open signal. B. Incorrect but plausible:
LPCI signal will open the Injection Valve. C. Incorrect but plausible:
LPCI signal will open the Injection Valve. Suction valves being closed will prevent the pump starting.
Suppression pool suction will not receive an open signal. D. Correct:
Closing the SOC suction valves on the isolation signal will trip the RHR pump. The LPCI signal will open the RHR Injection Valve. The Suppression Pool Suction valve does not receive an open signal on a LPCI signal.
==References:==
HC.OP-SO.BC-0001 Student Ref: NONE Learning Objective:
Question Source: HC Bank #209 Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.5/45.6 Comments:
QUESTION 30 HPCI is injecting to the Reactor Vessel with the Flow Controller in MANUAL at 5600gpm.
Reactor Pressure rises from 850 psig to 950 psig. Which ONE of the following describes the effect on HPCI, steady-state to steady-state?
A. RPM will RISE, pump flow will RISE. B. RPM will RISE, pump flow will REMAIN CONSTANT.
C. RPM will REMAIN CONSTANT, pump flow will LOWER. D. RPM will REMAIN CONSTANT, pump flow will REMAIN CONSTANT.
K&A Rating: K&A Statement:
Justification:
206000K5.01 (3.3) Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM: K5.01 Turbine Operation A. Incorrect but plausible:
With FIC-R600 HPCI controller in MANUAL, pump speed is maintained at a constant speed without operator action. B. Incorrect but plausible:
With FIC-R600 HPCI controller in MANUAL, pump speed is maintained at a constant speed without operator action. C. Correct:
With FIC-R600 HPCI controller in MANUAL, pump speed is maintained at a constant speed without operator action. If Reactor Pressure rises, pump speed remains constant and pump flow will lower due to increased backpressure (more resistance).
D. Incorrect but plausible:
With FIC-R600 HPCI controller in MANUAL, pump speed is maintained at a constant speed without operator action. However, if Reactor Pressure rises, pump flow will lower due to increased backpressure, not remain constant.
==References:==
NOH01 HPCI1 00-10 Student Ref: NONE Learning Objective:
Obj 6a,b,c,d Question Source: New Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 31 While responding to a Loss of Coolant Accident, the Plant Operator (PO) notices that the "A" Core Spray Loop fails to initiate automatically (the 'Initiation Logic Seal-In' lights are NOT illuminated).
The PO manually starts both "A" Loop pumps, and then attempts to open inboard isolation valve HV-F005A.
Which ONE of the following describes the response of HV-F005A?
A. HV-F005A will open if outboard isolation valve HV-F004A is closed. B. HV-F005A will open provided the PO also depresses AUTO OPEN OVRD. C. HV-F005A will open immediately upon both 'A' loop pumps starting.
D. HV-F005A will open when RPV pressure is less than 461 psig.
K&A Rating: K&A Statement:
Justification:
209001 K4.01 (3.2/3.4)
Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following:
K4.01 Prevention of over pressurization of core spray piping A. Correct:
Without an auto initiation signal, HV-F005A will only open if outboard isolation valve HV-F004A is closed to prevent over pressurization of the core spray piping. B. Incorrect but plausible:
The AUTO OPEN OVRD, is only active with an initiation signal present and allows the operator to close the valve, there is no initiation signal as evident by 'Initiation Logic Seal-In' lights are not illuminated.
C. Incorrect but plausible:
The auto open signal requires an initiation signal and RPV pressure<
461 psig, there is no initiation signal as evident by 'Initiation Logic Seal-In' lights are not illuminated.
D. Incorrect but plausible
: The auto open signal requires an initiation signal, with 'Initiation Logic Seal-In' lights are not illuminated, the valve will not open.
==References:==
NA Learning Objective:
Question Source: Question History:
Cognitive Level: 10CFR: Comments:
HC Bank Q#72 NA Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
CFR41.07 Student Ref: NONE QUESTION 32 Given the following conditions:
* The Reactor scrammed from 100% power on turbine stop valve closure
* All control rods have NOT fully inserted
* APRMs are reading 12%
* The RO manually starts both SLC pumps at the direction of the CRS
* Initial SLC tank level is 4850 gallons 45 minutes later, and assuming NO further operator
: actions, the SLC pumps are ... A. tripped on SLC tank low level. B. running with the SLC tank empty. C. tripped with the reactor shutdown.
D. running and injecting to the vessel.
K&A Rating: 211000 At01 (3.6/3.7)
K&A Statement:
Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM: A1.01 Tank Level Justification:
A. Incorrect but plausible:
The SLC pumps would still be running because level would be >325 gal after 45 minutes B. Incorrect but plausible:
The SLC tank will be above the low level trip. The Low level trip would not be bypassed when started from the Control Room C. Incorrect but plausible:
The reactor would be shutdown because <11 00 gal remain in the tank but the SLC pumps would still be running D. Correct:
Pumps auto trip on low level at 325 gals remaining provided the pumps were not running in test. Min required capacity (2 pumps)= 84.2 gpm (OP-IS-BH-0001/0002),
(4850-325) I 84.2 = 53.2 minutes.
Max acceptable flow rate= 48.6 gpm X 2 = 97.2 gpm. 4850 -325/ 97.2 = 46.6 minutes.
Therefore the pumps should still be running at the 45 minute mark
==References:==
NOH01SLCSYSC, Rev. 4 HC.OP-AR.ZZ-0008 Att. E1, Rev. 43 Learning Objective:
SLCSYSEOO 18/E007 Student Ref: NONE Question Source: Modified from PB Bank (Question
#1) Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/45.5 Comments:
QUESTION 33 Given the following conditions:
* The Reactor scrammed on low Reactor water level
* All rods failed to insert
* SLC was initiated
* RPV level was lowered lAW HC.OP-EO.ZZ-0101A, ATWS-RPV Control
* RPV level is being maintained between -150" and -185"
* The "A" RPS Bus is de-energized and cannot be restored
* The implementation of HC.OP-EO.ZZ-0301, "Bypassing MSIV Isolation Interlocks" and reopening of the Main Steam Isolation Valves (MSIVs) has been directed Following completion of HC.OP-EO.ZZ-0301, which ONE of the following describes the ability to operate MSIVs from the control room? A. None of the eight MSIVs can be reopened.
B. Only the four outboard MSIVs can be reopened.
C. All eight of the MSIVs can be reopened.
D. Only the four inboard MSIVs can be reopened.
K&A Rating: 212000K1.14 (3.6/3.7)
K&A Statement:
Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following:
K1.14: Main Steam system. Justification:
A. Correct:
Both A and B RPS buses must be energized to allow NSSSS logic to be reset. E0-301 only bypasses the Lo-Lo-Lo level isolation.
B. Incorrect but plausible:
Because both A and B busses need to be energized.
Plausible if the candidate thinks that one RPS is enough power to open the outboard valves. C. Incorrect but plausible:
Because both A and B busses need to be energized.
Plausible if the candidate thinks that bypassing the LO-LO-LO logic is all that is required to reopen the MSIV's. D. Incorrect but plausible:
Because both A and B busses need to be energized.
Plausible if the candidate thinks that one RPS is enough power to open the inboard valves.
==References:==
HC.OP-EO.ZZ-0301, Section 2.6 NONE Learning Objective:
NA Question Source: Bank (Question
#255) Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.2 to 9/43.7/45.8 Comments:
Student Ref:
QUESTION 34 Given the following conditions:
* The Mode Switch is in START UP
* A half SCRAM has been inserted on RPS "B" due to a Reactor Vessel Water Level instrumentation malfunction.
* APRM "C" is BYPASSED IRM "E" fails such that annunciator IRM A/B/E/F UPSCALE INOP/TRIP is received.
Which ONE of the following describes the plant condition after the IRM "E" malfunction?
A. A rod block is generated by the Rod Block circuitry; A full SCRAM is received.
B. A rod block is generated by the Rod Block circuitry; A full SCRAM is NOT received.
C. A rod block is NOT generated by the Rod Block circuitry; A full SCRAM is received.
D. A rod block is NOT generated by the Rod Block circuitry; A full SCRAM is NOT received.
K&A Rating: 215003K1.02 (3.6) K&A Statement:
Knowledge of the physical connections and/or cause effect relationship between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following:
K1.02 Reactor manual control Justification:
A. Correct:
RPS "B" has a half SCRAM as stated in the stem. With the mode switch in startup, the IRM "E" malfunction and IRM NB/E/F UPSCALE INOP/TRIP alarm is an indication of a half SCRAM on RPS "A" AND a rod block will be generated by the Rod Block circuitry.
A half SCRAM on both channels of RPS satisfies the conditions for a full reactor SCRAM. B. Incorrect but plausible:
A rod block IS generated, but a full SCRAM is also generated, not a half SCRAM. C. Incorrect but plausible:
A rod block IS generated as evidenced by the IRM "E" malfunction and IRM NB/E/F UPSCALE INOP/TRIP alarm stated in the stem. D. Incorrect but plausible:
A rod block IS generated as evidenced by the IRM "E" malfunction and IRM NB/E/F UPSCALE INOP/TRIP alarm stated in the stem.
==References:==
NOH01 RPSOOC-09 HC.OP-AR.ZZ-0009(Q)
Learning Objective:
RPSOOOE007 Question Source: New Question History:
N/A Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/
Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 35 Given the following conditions:
* A Reactor Startup is in progress
* The Reactor Protection System shorting links are removed
* Reactor power is increasing with a stable positive period of 150 seconds
* The SRM Channel 'A' detector is stuck and will NOT withdraw
* The SRM Channel 'A' indication increases to 2x1 05 cps Assume NO operator actions are performed.
Indicated Reactor power will ... A continue to increase and Reactor period will remain positive.
B. decrease and Reactor period will be negative.
C. continue to increase and Reactor period will become shorter.
D. decrease and Reactor period will initially go to infinity.
K&A Rating: 215004K6.04 (2.9/2.9)
K&A Statement:
Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRH) SYSTEM: K6.04 Detectors Justification:
A. Incorrect but plausible:
The Reactor SCRAM will result in a lowering count rate and negative period. B. Correct:
The Reactor will SCRAM, causing a lowering count rate and a negative period. C. Incorrect but plausible:
The Reactor SCRAM will result in a lowering count rate and negative period. D. Incorrect but plausible:
The Reactor SCRAM will result in a negative period.
==References:==
HC.OP-SO.SB-0001 Student Ref: NONE HC.OP-SO.SE-0001 Learning Objective:
SRMSYSSE01 0 Question Source: Bank Question History:
HC Bank# 164 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR41.7/45.7 Comments:
QUESTION 36 Given the following conditions:
* Reactor Power is 55%
* Control rod 30-31 is selected Subsequently, Recirculation Flow Unit "A" fails downscale.
Which ONE of the following describes ALL the Scram (if any) and withdraw Rod Block signals generated from this failure?
A. No scram signals are generated.
Withdraw rod block signals are generated by APRM Channels A, C, and E; RBM Channel B; and the A and B Recirculation Flow Units. B. No scram signals are generated.
Withdraw rod block signals are generated by APRM Channels A, C, and E; RBM Channel A; and the A and B Recirculation Flow Units. C. Scram signals are generated by APRM Channel A, C, and E. Withdraw rod block signals are generated by APRM Channel A, C, and E; and Recirculation Flow Unit A, ONLY. D. Scram signals are generated by APRM Channels A, C, and E. Withdraw rod block signals are generated by APRM Channels A, C, and E; RBM Channel A and 8; and the A and B Recirculation Flow Units.
Question 36 K&A Rating: K&A Statement:
Justification:
215005 K4.01 (3.7/3.7)
Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following:
K4.01 Rod withdrawal blocks A. Incorrect but plausible:
RBM channel B is not affected by flow converter A failure.
Plausible if applicant does not know which flow converters input into which RBM. B. Correct:
The APRM flow biased rod block setpoint at zero flow is 53%. The reactor power is 2112/3840 which is 55% therefore Control Rod withdrawal block signals are generated from APRM Channels A, C, and E. RBM Channel A zero flow setpoint on the low range is 49%. A and B Recirculation Flow Units with 10% difference between flow comparators will generate a rod block. C. Incorrect but plausible:
The APRM flow biased scram setpoint at zero flow is 58%. The reactor power is 2112/3840 which is 55% and therefore no scram signal will be generated.
D. Incorrect but plausible:
The APRM flow biased scram setpoint at zero flow is 58%. The reactor power is 2112/3840 which is 55% and therefore no scram signal will be generated.
==References:==
HC.OP-SO.SF-0002-RBM Student Ref: NONE Learning Objective:
NA Question Source: HC BANK #64976 Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7 Comments:
QUESTION 37 The plant was operating at 80% power when a transient occurred requiring a manual initiation of RCIC. Approximately 15 minutes later the PO notes that HV-F031 RCIC PUMP SUCTION FROM THE SUPPRESSION POOL has drifted to a 50% OPEN position.
Which ONE of the following describes the system response?
A. HV-F010 PUMP SUCTION VALVE FROM CST will immediately CLOSE. B. HV-F022 RCIC PUMP DISCHARGE TEST RETURN TO CST valve will receive an OPEN signal. C. HV-F010 PUMP SUCTION VALVE FROM CST will remain OPEN until CST Level reaches -22,500 gallons.
D. HV-F010 PUMP SUCTION VALVE FROM CST will remain OPEN until CST Level reaches -70,000 gallons.
K&A Rating: K&A Statement:
Justification:
217000K2.02 (2.6) Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC): K6.04 Condensate storage and transfer system A. Incorrect but plausible:
HV-F01 0 will immediately close if HV-F031 goes FULL OPEN. Since HV-F031 is only partially open, HV-F010 will not immediately close. B. Incorrect but plausible:
HV-F022 receives a CLOSE signal, not an OPEN signal when HV-F031 is not full CLOSE. C. Incorrect but plausible:
HV-F01 0 will automatically close when CST level reaches -70,000 gallons.
The setpoint was changed from -22,500 gallons due to pump vortexing concerns.
D. Correct:
HV-F01 0 will automatically close when CST level reaches -70,000 gallons
==References:==
NOH04RCIC00-1 0 Student Ref: NONE Learning Objective:
RCICOOEE05, RCICOOEE09 Question Source: New Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 38 OP-EO.ZZ-1 03, Reactor Building
: Control, has been entered due to an unisolable primary coolant leak in the RWCU Non-Regenerative Heat Exchanger Room. The following conditions exist:
* RPV pressure is 725 psig and slowly lowering
* RPV water level is -140 inches
* Drywell Pressure is 0.5 psig and steady
* ADS LOGIC CH B MANUAL INHIBIT Switch is in "INHIB"
* ADS LOGIC CH D MANUAL INHIBIT Switch has failed in the "NORM" position
* All low pressure ECCS Pumps are running with normal indications 395 seconds after reaching RPV Level 1, the LOGIC D/H I NIT RESET Pushbutton is depressed and released.
Which ONE of the following identifies the effect on the ADS valves? A. Valves will close and re-open 300 seconds later B. Valves will close and re-open 405 seconds later C. Valves will remain closed; valves will auto open 405 seconds later D. Valves will remain closed; valves will NOT auto open on any time delay K&A Rating: K&A Statement:
218000 A3.09 (4.2/4.3)
Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including:
Reactor vessel water level Justification:
A. Incorrect but plausible:
Plausible if the applicant does not understand ADS Initiation Logic and System response with respect to (1) operation of the 300 second ADS High Drywell Bypass Timer and the 105 second Logic D/H (CH D) ADS Initiation Timer, and (2) manipulation of the Manual Inhibit Switches and the Logic Reset Pushbuttons.
See Answer 'C' discussion.
B. Incorrect but plausible:
Plausible if the applicant does not understand ADS Initiation Logic and System response with respect to (1) operation of the 300 second ADS High Drywell Bypass Timer and the 105 second Logic D/H (CH D) ADS Initiation Timer, and (2) manipulation of the Manual Inhibit Switches and the Logic Reset Pushbuttons.
See Answer 'C' discussion.
C. Correct:
The LOGIC D/H I NIT RESET Pushbutton is depressed 10 seconds prior to ADS initiation (auto initiation occurs at 405 seconds).
The 300 second ADS High Drywell Bypass Timer will be reset. The CH DADS Initiation Timer will also be reset. The ADS valves never opened because the auto initiation at 305 seconds (105 seconds after the 300 second ADS High Drywell Bypass Timer times out) was interrupted when the LOGIC D/H !NIT RESET Pushbutton was depressed at time 395 seconds.
The ADS valves will open 405 seconds later when both the ADS High Drywell Bypass Timer (300 seconds) and the Logic D/H (CH D) ADS Initiation Timer ( 105 seconds) successfully time out (300 + 1 05 = 405 seconds) after having been reset. D. Incorrect but plausible:
Plausible if the applicant does not understand ADS Initiation Logic and System response with respect to (1) operation of the 300 second ADS High Drywell Bypass Timer and the 105 second Logic D/H (CH D) ADS Initiation Timer, and (2) manipulation of the Manual Inhibit Switches and the Logic Reset Pushbuttons.
See Answer 'C' discussion.
==References:==
NOH04ADSSYSC-08 Student Ref: NONE Learning Objective:
Obj. 2 of NOH04ADSSYSC-08 Question Source: New Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR41.7 Comments:
QUESTION 39 A reactor startup is in progress with the following conditions:
* Reactor pressure 750 psig
* The Reactor Mode Switch is in "Startup/Hot Standby"
* The main turbine is tripped Subsequently, a valid Group 1 MSIV isolation occurs
* No conditions to cause a reactor scram have been generated
* All systems operated as designed Which ONE of the following plant parameters generated the Group 1 MSIV isolation signal? A. Reactor water level B. Main steam line pressure C. Main steam line radiation D. Main steam line tunnel temperature K&A Rating: K&A Statement:
Justification:
223002K1.01 (3.8/3.9)
Knowledge of the physical connections and/or cause-effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the following:
K1.01 Main Steam System. A. Incorrect but plausible:
Since a low level scram did not occur, reactor watr level is >12.5 inches and will not cause an isolation.
B. Incorrect but plausible:
Because the mode switch would have to be in run to enable the low pressure isolation.
C. Incorrect but plausible:
Main Steam Line Radiation does not cause a group 1 isolation.
Plausible because it does cause a group 2 isolation.
D. Correct:
High Main Steam Line temperature will cause an isolation in these plant conditions.
==References:==
HC.OP-SO.SM-0001 B21-1 090-0062 Learning Objective:
Question Source: Bank #82 Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
Student Ref: NONE Comprehensive/Analysis:
X 10CFR: CFR 41.2 to 9/45.7 to 8 Comments:
QUESTION 40 Given the following:
* The ADS Manual Initiation Channel D and H pushbuttons (S6D and S6H) have been armed and depressed
* There is NO Safety Relief Valve response Which ONE of the following bus losses would cause this system response?
A. 125 VDC bus 1BD417 B. 125 VDC bus 1DD417 C. 120 VAC bus 1BJ481 D. 120 VAC bus 1 DJ481 K&A Rating: 239002 K2.01 (2.8/3.2)
K&A Statement:
Knowledge of electrical power supplies to the following:
K2.01 SRV Solenoids Justification:
A. Incorrect but plausible:
If applicant does not recall power supply to ADS logic D&H; this is the power supply to ADS logic B&F B. Correct:
This is the power supply to ADS logic D&H C. Incorrect but plausible:
If applicant does not understand effect of this loss. This will result in a loss of automatic initiation of ADS channel B, but will not prevent a manual initiation D. Incorrect but plausible:
If applicant does not understand effect of this loss. This will result in a loss of automatic initiation of ADS channel D, but will not prevent a manual initiation
==References:==
NOH04ADSSYSC Rev. 8 HC.OP-AB.ZZ-0136, Rev. 19 Learning Objective:
ADSSYSE006 Question Source: New Question History:
None Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
10CFR55:
CFR41.7 Comments:
QUESTION 41 Given the following conditions:
* The plant is at 1 00% power
* The 'C' Steam Flow detector for the Feedwater Level Control System fails low (its output indicates 0 lbm/hr steam flow) Which ONE of the following describes the automatic plant response with NO operator action? ACTUAL reactor water level will ... A. remain the same. The Feedwater Level Control System will shift to single element control.
B. decrease and stabilize at a lower than normal value. The Feedwater Level Control System will shift to single element control.
C. decrease and stabilize at a lower than normal value. The Feedwater Level Control System will remain in three element control.
D. increase and stabilize at a higher than normal value. The Feedwater Level Control System will remain in three element control.
K&A Rating: 259002A 1. 01 (3. 8/3. 8) K&A Statement:
Ability to predict and/or monitor changes in parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including:
A1.01 Reactor Water Level Justification:
A. Correct:
The system transfers to single element control.
With stable conditions, single and three element control should control level at the same setpoint.
B. Incorrect but plausible:
Level should not decrease.
C. Incorrect but plausible:
Level should not decrease.
The system will shift to single element control.
D. Incorrect but plausible:
Level should not increase.
The system will shift to three element control.
==References:==
DWG H-1-AE-ECS-0128-0 Student Ref: NONE Learning Objective:
FWCONTE012 Question Source: HC Bank# 233 Question History:
Used on 2003 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/45.5 Comments:
QUESTION 42 Given the following conditions:
* The plant is at 50% power
* The FRVS Recirculation Fans are in AUTO Subsequently, a Loss of Offsite Power (LOP) occurs:
* HPCI and RCIC are manually initiated
* The minimum water level reached was -32 inches Assuming NO other operator
: actions, the total FRVS recirculation flow 3 minutes after the LOP will be ... A. 0 cfm. B. 60,000 cfm. C. 120,000 cfm. D. 180,000 cfm.
K&A Rating: 261000A2.10 (3.1/3.2)
K&A Statement:
Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct,
: control, or mitigate the consequences of those abnormal conditions or operations:
K2.1 0 Low reactor water level: Plant-Specific Justification:
A. Correct:
There are no automatic isolation signals present; No FRVS fans are running.
B. Incorrect but plausible:
All six FRVS recirculation fans start at RPV level 2 (-38 inches).
Plausible if the applicant believes that some combination of FRVS fans would be running given the stem conditions.
C. Incorrect but plausible:
All six FRVS recirculation fans start at RPV level 2 (-38 inches).
Plausible if the applicant believes that some combination of FRVS fans would be running given the stem conditions.
D. Incorrect but plausible:
All six FRVS recirculation fans start at RPV level 2 (-38 inches).
Plausible if the applicant believes that some combination of FRVS fans would be running given the stem conditions.
==References:==
HC.OP-AB.CONT-0003 Student Ref: NONE HC.OP-SO.GR-0001 Learning Objective:
RBVENTE004 Question Source: Bank Question History:
HC Bank # 130 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/45.6 Comments:
Question meets part (a) of KIA.
QUESTION 43 Opening the toggle switch associated with the 500KV breaker Emergency Trip handle ... A. transfers all control functions to the blockhouse.
B. prevents the Emergency Trip function.
C. removes control power for the breaker.
D. enables removal of kirk key at the breaker.
K&A Rating: K&A Statement Justification:
262001 K5.02 (2.6) Knowledge of the operational implications of the following concepts as they apply to A. C. ELECTRICAL DISTRIBUTION.
K5.02 Breaker Control A. Incorrect but plausible:
No local control at the breaker except for test. B. Incorrect but plausible:
Opening the toggle switch does not prevent emergency operation of the breaker.
C. Correct:
Interrupts breaker control circuit to prevent electrical operation.
D. Incorrect but plausible:
Kirk Key can be removed with toggle in either position.
==References:==
NOH01 MNPWR0-04, MAIN POWER SYSTEM, p21 Student Ref: NONE Learning Objective:
MNPWROE014 Question Source: Hope Creek Question-Q68164 Question History:
2005 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
1 OCFR: CFR 41.5/45.3 Comments:
QUESTION 44 A UPS Manual Bypass Switch has been repositioned from "BYPASS TO ALTERNATE" to "ISOLA TED AFTER PREFERRED" for testing on a faulty Static Switch. Which ONE of the following describes the system response if a loss of offsite power occurs, and the MCCs for the Preferred and Backup power are NOT 1 E switchgear supplied?
The input to the Regulator/Static Switch Cabinet will be supplied by _(1)_ and power to system loads will _(2)_ . A. (1) 125 VDC power (2) be maintained.
B. (1) 125 VDC power (3) be lost. C. (1) normal480 VAC power (2) be lost. D. (1) normal 480 VAC power (2) be maintained.
(See attached UPS Power Control Circuit Diagram)
UPS POWER CONTROL (TYPICAL)
BACKUP POWER 4a0 VAC CBJ01 ) (BACK UP mn AC POWER SUPPLY REGULATOR CABINET SWITCH POSITION NORMAL BYPASS TO PREFERRED ISOLATE (AFTER PREF.) BYPASS TO ALTERNATE ISOLATE (AFTER ALT.) CB21 1 X X X BLOCKING DIODES 2 3 X X X X Ai..TERHA.TE POWER 125 voc I) CB" (; LT DC INPUT At; POWER SUPPL 'f RECTIFIER CABINET LOAD 4 5 X X X X K&A Rating: 262002A4.01 (2.8/3.1)
K&A Statement:
Ability to manually operate and/or monitor in the control room: A4.01 Transfer from alternative source to preferred source. Justification:
A Correct:
Taking the Manual Bypass Switch to "ISOLATED TO PREFERRED" places the load directly on the preferred (inverter output) power source and all power is removed from the static switcrl The preferred power source consists of the DC output of the rectifier cabinet which comes from the auctioneered high input. The inputs to the rectifier cabinet are normal480 VAC or 125 VDC (alternate preferred source).
In the case of a LOP on a non 1 E switchgear, power will be supplied to the regulator cabinet via the preferred source powered by 125 VDC. B. Incorrect but plausible:
Supply to the system loads will not be lost, the static switch is de-energized to the preferred source position, which will be powered by 125 VDC. C. Incorrect but plausible:
The output of the rectified cabinet will take auctioneered high power source -> 125VDC since 480 VAC normal will be lost. D. Correct:
Backup 480 VAC power will only be supplied to the regulator/static switch cabinet if the static switch was in the alternate power position.
==References:==
NOH01 EAC00-04 Student Ref: NONE Learning Objective:
NON1 EOE003 Question Source: New Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7/45.5 to 45.8 Comments:
QUESTION 45 Given the following condition:
* 125 VDC Bus B (1 BD417) tripped off due to a bus fault. Which ONE of the following describes the location(s),
if any, from which SRV 'F' can be opened? A. NOT available from any location.
B. Available from the Main Control Room ONLY. C. Available from the Remote Shutdown Panel ONLY. D. Available from the Main Control Room AND the Remote Shutdown Panel.
Question 45 K&A Rating: K&A Statement:
Justification:
263000 K2.01 (3.2/3.3)
Knowledge of electrical power supplies to the following:
K2.01: Major D.C. loads A. Correct : 1 BD417 feeds all of the SRV "A" solenoids.
SRV F is a non-ADS SRV, so it only has one solenoid and it is an 'A' solenoid.
SRV F has one control station in the MCR and one at the RSP; however, the valve cannot be opened from either location since both locations use the same 'A' solenoid and there is no power available.
B. Incorrect but plausible:
Will not open due to no power. C. Incorrect but plausible:
Will not open due to no power. D. Incorrect but plausible
: Will not open due to no power
==References:==
NA Student Ref: NONE Learning Objective:
loss of 1 BD417 Question Source: Bank HC #191 Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
10CFR55:
CFR41.7 Comments:
QUESTION 46 One second after paralleling the 'A' EDG to the 10A401 bus, a Loss of Offsite Power occurs and the 'A' EDG output breaker trips (its indication is flashing).
After inspection, the SM directs reclosing the 'A' EDG output breaker.
Which ONE of the following operator actions will energize the 1 OA401 4.16 KV bus? A. Depress the TRIP Pbs on the Normal and Alternate supply breakers, then allow the EDG output breaker to auto close. B. Locally reset the EDG output breaker, then using the EDG output breaker CLOSE Pb, reenergize the 10A401 bus. C. Reset the Test Lockout relay, then using the EDG output breaker CLOSE Pb, reenergize the 10A401 bus. D. Acknowledge the flashing TRIP Pb on the EDG output breaker, then allow the EDG output breaker to auto close.
K&A Rating: 264000 G2.1.28 (4.1/4.1)
K&A Statement:
2 1.28 Knowledge of the purpose and function of major system components and controls.
Justification:
A. Incorrect but plausible:
Depress the TRIP Pbs on the Normal and Alternate supply breakers, then allow the EDG output breaker to auto close.Will prevent the EDG from loading AB-135 Note 4.2. B. Incorrect but plausible:
Locally reset the EDG output breaker, then using the EDG output breaker CLOSE Pb, reenergize the 10A401 bus. Cannot be reset locally.
C. Incorrect but plausible:
Reset the Test Lockout relay, then using the EDG output breaker CLOSE Pb, reenergize the 1 OA401 bus. Test LO would not be tripped.
D. Correct:
Acknowledge the flashing TRIP Pb on the EDG output breaker, then allow the EDG output breaker to auto close. AB-135 5.16-The Anti-pump circuitry on the DIG output breaker could cause the output breaker to fail open, if a LOP were to occur within 2 seconds of closing the output breaker when testing and loading the DIG to the grid. To load the D/G under this condition the operator must wait a minimum of 2 seconds from the time the breaker was originally closed, then depress the TRIP push-button (even though the breaker is already tripped) to reset the logic. When the TRIP push-button is released, then the breaker will close and the D/G will load.
==References:==
NOH04EDGOOOC-03, ABZZ-0135 Student Ref: NONE Learning Objective:
EDGOOOE020, OAB135E004 Question Source: Bank# 33519 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR41.7 Comments:
QUESTION 47 The 1 OK1 07 Service Air Compressor is in service in the LEAD mode of o Subsequently, the following conditions are noted:
* 2ND INTERCOOLER HIGH LEVEL alarm is in on the OK107 local control panel
* AFTERCOOLER AIR OUTLET HIGH TEMP SHU OWN alarm is in on the 1 OK1 07 local control panel
* Lube Oil pressure is 25 psig Which ONE of the following now describes th status of the 1 OK1 07 Service Air Compressor?
The compressor
... A B. C. D. on aftercooler air outlet temperature.
K&A Rating: 300000K5.01 (2.5/2.5)
K&A Statement:
Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: K5.01 Air Compressors Justification:
A. Incorrect but plausible:
The low oil pressure trip occurs at 1 Opsig. B. Incorrect but plausible:
While there is an alarm for high demister level, there is no associated compressor trip. C. Correct:
None of the conditions listed in the question stem will result in a Service Air Compressor trip. D. Incorrect but plausible:
While there is an alarm for high aftercooler air outlet temperature, there is no associated compressor trip. An aftercooler air outlet high temperature of 115F NO LONGER will cause its associated compressor to trip.
==References:==
NOH01 SERAIR-05 Student Ref: NONE Learning Objective:
Obj 3 Question Source: New Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
1 OCFR55: CFR 41 .5/45.3 Comments:
QUESTION 48 Given the following conditions:
* Reactor power is 83%
* Normal electrical lineup
* TACS is being supplied from the 'A' SACS loop Which ONE of the following describes the response of the SACS to depressing the TRIP pushbutton for breaker 40108 ( 1 OA401 Normal Feeder Breaker)?
A. The 'A' SACS Pump AP21 0 will trip, TACS swaps to supply system loads. B. The 'A' SACS Pump AP21 0 will trip, loop 'A' will remain online supplying both SACS and TACS loads at a reduced flow. C. The 'C' SACS Pump CP21 0 will trip, loop 'A' will remain online supplying both SACS and TACS loads at a reduced flow. D. The 'A' SACS Pumps AP21 0 and the 'C' SACS Pump CP21 0 will both instantly trip, the TACS swaps to supply system loads.
K&A Rating: K&A Statement:
Justification:
400000 A3.01 (3.0/3.0)
Ability to monitor automatic operations of the CCWS including:
A3.01 Setpoints on instrument signal levels for normal operations,
: warnings, and trips that are applicable to the CCWS A. Correct:
Pump AP21 0 will trip on undervoltage, which closes the associated SACS to TACS isolation, causing a low flow condition in the 'A' loop. The low flow will start the AUTO pump in the 'B' loop, and open the 'B' loop isolation valves B. Incorrect but plausible:
Pump AP21 0 will trip on undervoltage, which will close the SACS to T ACS isolation C. Incorrect but plausible:
If applicant does not recall correct pump power supplies.
'C' SACS pump is supplied from the 1 OA403 bus D. Incorrect but plausible:
If applicant does not recall correct pump power supplies.
'C' SACS pump is supplied from the 1 OA403 bus
==References:==
NOH04STACSOC, Rev. 8 Student Ref: NONE Learning Objective:
ST ACSOE004 Question Source: HC Bank 33814 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7/45.7 Comments:
QUESTION 49 Given the following (assume no operator actions):
* HPCI has initiated from a valid Hi Drywell Pressure signal
* RPV Water level is 35"
* Reactor pressure is 980 psig Then, the operator observes that the HPCI turbine control valve (FV-4879) is in the throttled position AND that turbine speed is decreasing.
Additionally, the operator observes the following valves going closed:
* FV-4880, HPCI turbine stop valve
* HV-F006, HPCI pump discharge to Core Spray
* HV-F012, HPCI minimum flow Finally, the operator later notices that HPCI turbine speed is again increasing and HPCI realigns for injection (without operator actions).
Which ONE of the following caused this HPCI response?
A. Incorrectly set HPCI flow controller.
B. Mechanical overspeed HPCI trip. C. Automatic reset of the HPCI isolation logic. D. Automatic swap of the HPCI pump suction source.
K&A Rating: 206000 A3.01 (3.6/3.5)
K&A Statement:
Ability to monitor automatic operations of the HIGH PRESSURE COOLANT INJECTION SYSTEM including:
A3.01 Turbine Speed Justification:
A. Incorrect but plausible:
Represents a valid, incorrect condition for the applicant to evaluate.
B. Correct:
The HPCI turbine will respond to a mechanical overspeed trip (approximately 5200 rpm) by removing hydraulic oil from the turbine stop valve actuator and then reapplying it after the turbine speed has slowed sufficiently.
The turbine control valve will throttle to attempt to control the speed. C. Incorrect but plausible:
Represents a valid, incorrect condition for the applicant to evaluate.
D. Incorrect but plausible:
Represents a valid, incorrect condition for the applicant to evaluate.
==References:==
HC.OP-SO.BJ-0001 NOH0*1 HPCI00-1 0 Learning Objective:
HPCIOOE004 Question Source: Bank# 14/33365 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: Comments:
Student Ref: None QUESTION 50 Hope Creek is shutdown with refueling in progress.
The shorting links ARE installed.
Which ONE of the following will occur if the "A" Source Range Monitor (SRM) drawer mode switch is taken out of the OPERATE position?
Alarm Status Block/RPS Status A. A SRM downscale alarm Rod Block and Reactor Scram will occur B. A SRM downscale alarm Rod Block occurs, Reactor Scram will NOT occur C. A SRM Upscale/lnop alarm Rod Block occurs, Reactor Scram will NOT occur D. A SRM Upscale/lnop alarm Rod Block and Reactor Scram will occur Question 50 K&A Rating: K&A Statement:
Justification:
295004 K3.01 (3.4) Knowledge of the effect that a loss or malfunction of the source range monitor system (SRM) will have on the following:
K3.01 RPS A Incorrect but plausible:
Plausible if the applicant determines that placing the drawer mode switch out of Operate creates a SRM downscale trip, and determines that full scram will occur based on SRM downscale condition.
B. Incorrect but plausible:
Plausible if the applicant determines that placing the drawer mode switch out of Operate creates a SRM downscale trip, and determines that rod block will occur. C. Correct:
Placing the drawer mode switch out of Operate creates a SRM Inoperative trip causing Upscale/lnop alarm, and due to the upscale/inop alarm a rod block will also occur. D. Incorrect but plausible:
Plausible due to partially correct that Placing the drawer mode switch out of Operate creates a SRM Inoperative trip causing Upscale/lnop alarm, however SRM channels generate a scram signal on an INOP condition for loss of power only condition.
==References:==
Learning Objective:
Question Source: Question History:
Cognitive Level: 10CFR55:
Comments:
New NA Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
CFR 41.7/43.5/45.4 Student Ref: NONE QUESTION 51 Given the following conditions:
* The plant is at 85% power
* All three Reactor Feed Pumps are in AUTO
* RPV Narrow Range Level instruments indicate:
o N004A=34inches o N0048=35inches o N004C=35.5 inches Which ONE of the following describes the response of ACTUAL Reactor water level if a slow leak developed through the N004B detector equalizing valve, eventually causing a gross failure of N004B? ACTUAL Reactor water level would: A. LOWER 0.5 inch, then RISE 1.5 inches B. RISE 1.0 inch, then LOWER 0.5 inches C. LOWER 1.0 inch, then RISE 0.5 inches D. RISE 0.5 inch, then LOWER 1.5 inches K&A Rating: 259002K3.02 (3.8/3.8)
K&A Statement:
Knowledge of the effect that a loss or malfunction of the REACTOR WATER LEVEL CONTROL SYSTEM will have on the following:
K3.01 Reactor Water Level Justification:
A. Correct:
Initially, N004B is selected since DFCS selects the MEDIAN RPV level signal when 3 good signals are available.
With a leak through the N004B equalizing valve, N004B indicated level would begin to rise, resulting in a lowering of actual RPV level. As soon as N004B exceeded 35.5 inches indicated, N004C would become the median RPV level signal. Actual RPV level would have lowered 0.5 inch during this transition.
When N004B gross fails, N004A (the lowest of the two remaining signals) will become the controlling level signal. RPV water level will then rise since indicated level on N004A is 34 inches. This is a 1.5 inch rise from the previous level. B. Incorrect but plausible:
Initially lowers 0.5 inches, then rises 1.5 inches. C. Incorrect but plausible:
Initially lowers 0.5 inches, then rises 1.5 inches. D. Incorrect but plausible:
Initially lowers 0.5 inches.
==References:==
DWG H-1-AE-ECS-0128-0 Student Ref: NONE Learning Objective:
FWCONTEOI 0 Question Source: HC Bank# 95 Question History:
Used on 2009 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7/45.4 Comments:
QUESTION 52 125VDC bus 1 CD417 is de-energized when an Emergency Diesel Generator start signal is received.
Diesel Generator 1 CG400 will ... A. NOT automatically start B. automatically start but the output breaker can only be shut manually C. automatically start but in the DROOP mode D. automatically start but all trips will be disabled K&A Rating: K&A Statement:
Justification:
263000 K3.01 (3.4/3.8)
Knowledge of the effect that a loss or malfunction of the D. C. ELECTRICAL DISTRIBUTION System will have on the following:
K3.01 Emergency Generators:
Plant-Specific A. Correct:
The diesel generator will NOT automatically start; power is lost to the air start solenoids B. Incorrect but plausible:
The diesel generator will NOT automatically start C. Incorrect but plausible:
The diesel generator will NOT automatically start D. Incorrect but plausible:
The diesel generator will NOT automatically start
==References:==
NO HOI DCELEC, Rev. 3 Student Ref: NONE Learning Objective:
DCELECE003 Question Source: HC Bank 33239 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X IOCFR55:
CFR 41.7/45.4 Comments:
QUESTION 53 Which ONE of the following conditions (signals) will cause a trip of the Emergency Instrument Air Compressor?
A. Instrument Air Header pressure of 70 psig. B. RACS cooling water return from compressor temperature of 137" F. C. RPV level of -30". D. Emergency Instrument Air Receiver pressure of 110 psi g.
K&A Rating: K&A Statement Justification:
300000 K1.04 (2.8/2.9)
Knowledge of the connections and I or cause effect relationships between INSTRUMENT AIR SYSTEM and the following:
Cooling water to compressor A Incorrect but plausible:
Low Instrument Air Header pressure 70 psig. Isolates the service air header. B. Correct:
RACS cooling water return from compressor high temperature of 135&deg; F will cause auto trip of emergency instrument air compressor.
C. Incorrect but plausible:
Low RPV level of -129" indicated.
A LOCA will trip open the 1 E feeder breaker.
There is no override.
D. Incorrect but plausible:
High Pressure in the Emergency Instrument Air Receiver 70 psig. Actual trip occurs at 120 psig.
==References:==
HC.OP-SO.KB-0001 Student Ref: NONE Learning Objective:
INSAIRE003 Question Source: Modified Bank# 34761 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
10CFR: CFR 41.2-41.9 Comments:
QUESTION 54 Which ONE of the following is an indication that a control rod has become uncoupled and stuck? A. A red full-out indication on the full core display.
B. NO response on nuclear instrumentation when the control rod is being withdrawn from fully inserted.
C. A"--" indication on the Four Rod Display for that control rod when a withdraw signal is applied at position
: 48. D. Control rod position indication does NOT change when a withdraw signal is applied to a control rod at position
: 48.
K&A Rating: 201003 K4.02 (3.8) K&A Statement:
Knowledge of CONTROL ROD AND DRIVE MECHANISM design feature(s) and/or interlocks which provide for the following K4.02 Detection of an uncoupled rod Justification:
A. Incorrect but plausible:
A red full out indication is indication that the rod is overtraveled and uncoupled but does not mean that the rod is stuck. B. Correct:
No response on Nl's during rod withdrawal is an indication the rod is not moving. The rod position indication (Reed Switches) are in the mechanism which is moving, however the control rod is uncoupled and stuck. C. Incorrect but plausible:
A coupled rod during coupling checks will indicate-and is indication that the rod is coupled.
D. Incorrect but plausible:
A coupled rod at 48 will not move beyond 48 and is indication that the rod is coupled.
==References:==
NA Learning Objective:
000005E006 Question Source: HC Bank 56330 Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
10CFR55:
CFR41.74 Comments:
Student Ref: NONE QUESTION 55 The Reactor is operating normally at 90% power when the CP501 Circ Water Pump trips.
* AP501, BP501, and DP501 Circ Water Pumps remain in service
* Condenser pressure was steady at 5.0" HgA before the CP501 pump tripped
* RECIRC RUNBK BYPASS switch is in the IN (ARMED) position.
Which ONE of the following describes the response of the Reactor Recirculation System? A. The MG Set Drive Motor Breaker will trip. B. A full recirc pump runback to 30% speed will occur. C. An intermediate recirc pump runback to 45% speed will occur. D. The Recirculation System will maintain recirc pumps at their current speed.
K&A Rating: 202002 2.1. 7 ( 4.4/4. 7) K&A Statement:
Recirculation Flow Control 2.1. 7 Ability to evaluate plant pertormance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Justification:
A. Incorrect but plausible:
Plausible if the examinee mistakes the MG Set Drive Motor Breaker trip conditions with the Recirc Runback conditions.
B. Incorrect but plausible:
Plausible if the examinee believes that a full Recirc Pump Runback would occur on the loss of 1 Circ Water Pump with 4 running.
(A full runback to 30% occurs only on the loss of 1 Circ Water Pump with 3 running AND Condenser pressure
>5.8" HgA.) C. Correct:
An intermediate Recirc Pump runback to 45% speed will occur on the loss of 1 Circ Water Pump with 4 Circ Water Pumps running AND Condenser pressure
> 4.5" HgA. D. Incorrect but plausible:
Plausible if the examinee believes that the plant can take the loss of 1 Circ Water pump with 4 running without a Recirc Runback.
Condenser pressure being given as 5.0" HgA in the stem may lead the examinee to believe that the loss of 1 Circ water Pump intermediate runback will not be active.
==References:==
HC.OP-SO.BB-0002 NOH01 RECCON SO-DA-0001 step 5.14.2 Learning Objective:
R11, R15 Question Source: New Question History:
N/ A Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/43.5/45.12/45.13 Comments:
QUESTION 56 Given the following:
* OPCON 1 at 100% power.
* All control rods are fully withdrawn.
* An "A" side half SCRAM is in due to performance of HC.OP-ST.SF-003, "RPS Manual SCRAM Test-Weekly".
Subsequently, the PILOT SCRAM VALVE TRIP ACTUATOR LOGIC Group 4B SOLENOIDS LOGIC B NORMAL light extinguishes at 1 OC651 C. This condition was NOT caused by the weekly test. Which ONE of the following describes the indications on the full core display based ONLY on the INITIAL RPS response?
A. Control Rod Groups 1, 2 & 3 RED lamps are lit, and Control Rod Group 4 GREEN, BLUE, & AMBER lamps are lit. B. All Control Rods have RED lamps lit C. All Control Rods have GREEN, BLUE, & AMBER lamps lit. D. Control Rod Group 1 & 4 GREEN, BLUE, & AMBER lamps are lit, and Control Rod Groups 2 & 3 have RED lamps lit.
K&A Rating: K&A Statement:
Justification:
214000 A4.02 (4.0/4.1)
Ability to manually operate and/or monitor in the control room: Control rod position A. Correct:
The Operator's action causes an "A" side half scram and all "A" side scram pilot solenoids de-energize.
No rod motion occurs. Now, the Group 48 solenoid logic status light extinguishes indicating a loss of power to "8' side scram pilot solenoids for all control rods in Group 4 only. So, the "A" and "8" scram pilot solenoid valves are both energized on all control rods in Group 4, and this vents air from all scram valves for control rods in Group 4 ONLY. The Full Core Display will show RED (full out indication) for rods in Groups 1 ,2, & 3, and Group 4 control rods will display GREEN (full in), BLUE (scram valves open), and AMBER (rod drift). B. Incorrect but plausible:
Plausible if the applicant determines that sufficient logic did not complete to cause any of the rods to scram. RED indicates full out position.
Group 4 control rods went full in, so this answer is incorrect.
C. Incorrect but plausible:
Plausible if the applicant determines that sufficient logic is made up to cause full core scram. D. Incorrect but plausible:
Plausible if the applicant determines that the control rods associated with Groups 1 & 4 scrammed (A 1 manual pushbutton was depressed and the Group 48 logic power failure, i.e. RPS A & B de-energized on Groups 1 & 4).
==References:==
HC.OP-STSF-0003, "RPS Manual Scram Test-Weekly" Student Ref: NONE NOH01 RPSOOC-09 NOH04MANCONC-07 Learning Objective:
MANCONE002 Question Source: NRC HC 2007 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR: CFR 41.7/45.7 Comments:
QUESTION 57 Given the following conditions:
* The plant is at full power
* A TIP system trace is in progress
* An instrument technician error causes actuation of the NSSSS Channel A manual isolation switch
* The inserted TIP detector becomes stuck in the core Which ONE of the following describes the TIP system response to this condition?
A. The TIP Shear Valve automatically fires to cut the detector cable and seal the guide tube. B. The TIP Guide Tube Ball Valve automatically closes, cutting the detector cable and sealing the guide tube. C. The inserted TIP detector will receive a signal to automatically withdraw to the "in-shield" position.
The TIP Guide Tube Ball Valve automatically closes. D. The inserted TIP detector will receive a signal to automatically withdraw to the "in-shield" position.
The TIP Guide Tube Ball Valve will NOT automatically close.
K&A Rating: 215001A 1.02 (2.5/2.4)
K&A Statement:
Ability to predict and/or monitor changes in parameters associated with operating the TRAVERSING IN-CORE PROBE controls including A1.02 Detector Position Justification:
A. Incorrect but plausible:
The Shear Valves must be manually initiated.
B. Incorrect but plausible:
The Ball Valve will not close with the cable inside the valve. C. Incorrect but plausible:
the TIP Guide Tube Ball Valve will not automatically close due to the detector being stuck and not at the "in-shield" limit, the ball valve will not close with the cable inside the valve. D. Correct:
The TIP detector not at its "in-shield" position will receive a signal to automatically withdraw to the "in-shield"
: position, and the TIP Guide Tube Ball Valve will not automatically close due to the detector being stuck and not at the "in-shield" limit, the ball valve will not close with the cable inside the valve.
==References:==
NOHOTIPS00-02 HC.OP-SO.SM-0001 HC.RE-SO.SE-0001 Learning Objective:
TIPSOOE006 Question Source: Bank Question History:
HC # 328 Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/45.5 Comments:
QUESTION 58 The crew is responding to a Station Blackout.
The following conditions exist:
* A&C EDGs failed to energize their respective buses
* Lowest RPV level was+ 5", and is currently
+1 0" and steady using "B" Loop of Core Spray
* RPV pressure is 200 to 400 psig and being controlled by SRV's * "B" RHR has been placed in Suppression Pool cooling Subsequently, Drywell pressure reaches 1.72 psig due to the loss of containment cooling.
What actions will place "B" RHR in Suppression Pool Cooling and Spray? A Press AUTO OPEN OVRDs for BC-HV-F024B and BC-HV-F027B, and open the valves. B. Press AUTO OPEN OVRD for BC-HV-F017B, and close it; then press AUTO CL OVRDs for BC-HV-F024B and BC-HV-F027B, and open the valves. C. Press AUTO OPEN OVRDs for BC-HV-F0248 and BC-HV-F027B, and verify they stroke open. D. Press AUTO OPEN OVRD for BC-HV-F017B, and close it; then press AUTO OPEN OVRDs for BC-HV-F024B and BC-HV-F027B, and verify they stroke open.
K&A Rating: 219000 K3.01 (3.9/4.1)
K&A Statement:
Knowledge of the effect that a loss or malfunction of the RHRILPCI:
TORUS/SUPPRESSION POOL COOLING MODE: K3.01: Suppression pool temperature control Justification:
A. Incorrect but plausible:
AUTO CLOSE OVRD must be depressed to allow re-opening of BC-HV-F024B and BC-HV-F027B.
The BC-HV-F017B must also be closed. B. Correct:
lAW HC.OP-SO.BC-0001-Step 3.3.7 The following overrides are provided during a LPCI Initiation AND the associated Logic Initiation Reset must be pressed to restore to Auto position:
* RHR PUMP A(B,C,D)P202 (STOP PB should be pressed).
* BC-HV-F017A(B,C,O),
RHR LOOP A(B,C,D)
LPCIINJ MOV (AUTO OP OVRD PB should be pressed).
* BC-HV-F024A(B),
RHR LOOP A(B) TEST RET MOV (AUTO CL OVRD PB should be pressed).
* BC-HV-F027A(B),
RHR LOOP A(B) SUPP CHAMBER SPRAY HDR ISLN MOV (AUTO CL OVRD PB should be pressed)
C. Incorrect but plausible:
AUTO CLOSE OVRD must be depressed to allow re-opening of BC-HV-F024B and BC-HV-F027B.
They will not auto open. Also, BC-HV-F017B must be closed. D. Incorrect but plausible:
AUTO CLOSE OVRD must be depressed to allow re-opening of BC-HV-F024B and BC-HV-F027B.
==References:==
HC.OP-SO.BC-0001-Step 3.3.7 Student Ref: NONE Learning Objective:
Question Source: New Question History:
New Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7/45.4 Comments:
QUESTION 59 A function of the Suppression Chamber-to-Drywell Vacuum Breakers is to ... A. prevent low Suppression Chamber pressure from damaging the drywell after a LOCA. B. allow hot liquid to enter the Suppression Chamber during the blowdown phase of a LOCA. C. prevent a negative pressure from occurring in the Suppression Chamber following a LOCA. D. allow non-condensable gasses to reenter the drywell to prevent exceeding the drywell external pressure limit.
K&A Rating:
293001 K5.01 (3.1/3.3)
K&A Statement:
Knowledge of the operational implications of the following concepts as they apply to PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES:
K5.01 Vacuum breaker/relief operation Justification:
A. Incorrect but plausible:
Suppression chamber pressure is relieved back to the drywell after steam condensation to protect the drywell from implosion.
B. Incorrect but plausible:
Hot liquid from the blowdown will gravity drain from the drywell to the suppression chamber through the downcomers.
C. Incorrect but plausible:
The reactor building to suppression pool vacuum breakers prevent a negative pressure from occurring in the suppression chamber following a LOCA not the suppression chamber to drywell vacuum breakers.
D. Correct:
The suppression chamber to drywell vacuum breakers purpose is to prevent exceeding design external pressures of the drywell.
==References:==
NOH01 PRICONC Student Ref: NONE Learning Objective:
PRICONE003 Question Source: Bank Question History:
HC Bank# 79 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
1 OCFR55: CFR 41.5/45.3 Comments:
QUESTION 60 The plant was operating at 85% power when a high OW pressure condition occured.
* "B" RHR and "B" Core Spray pumps failed to start automatically.
Both pumps have been placed in service manually.
Ten minutes later, conditions are as follows:
* "A" RHR has been placed in Suppression pool cooling and spray
* Torus pressure is 9.5 psig and continues to slowly rise
* The CRS has directed "B" loop of RHR be placed in Drywell Spray
* The "B" RHR Loop Inboard Spray Valve (HV-F021 B) will open but the Outboard Drywell Spray Valve (HV-F016B) remains closed (electrical power is available)
Which ONE of the following conditions is most likely preventing OW Spray Valve HV-F016B from opening?
A The HX Bypass valve (HV-F048A) is not fully closed. B. The "B" channel Drywell pressure instrument is failed low. C. The RHR Full Flow test Valve (HV-F024B) is not fully closed. D. The "B" channel RPV water level indicates above -129 inches.
K&A Rating: 226001 K3.02 (2.9/2.9)
K&A Statement:
Knowledge of RHR/LPCI:
CONTAINMENT SPRAY SYSTEM MODE design feature(s) and/or interlocks which provide for the following:
K4.12 Prevention of Inadvertent Containment Spray Activation Justification:
A. Incorrect but plausible:
There is no interlock between these valves. B. Correct:
A high drywell pressure signal is required to open F0168. (ALL of the following conditions must be satisfied to allow both F0168 RHR LOOP 8 OUTS CONT SPRAY ISLN MOV AND F021 8 RHR LOOP 8 INBD CONT SPRAY MOVs to open simultaneously:
LPCI initiation signal present, High drywell pressure signal present, and F-178 RHR LOOP 8 LPCIINJ MOV CLOSED.)
C. Incorrect but plausible:
There is no interlock between these valves. D. Incorrect but plausible:
Manual initiation will seal in the LPCI initiation signal.
==References:==
HC.OP-SO.BC-0001 Student Ref: NONE RHRSYSE008 Learning Objective:
RHRSYSE008 Question Source: Bank Question History:
HC Bank# 329 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7 Comments:
QUESTION 61 Fuel handling activities are being conducted in the Spent Fuel Pool in preparation for the upcoming refuel outage. The following conditions occur:
* A Spent fuel bundle was physically damaged during a moving evolution to a new pool location during the lift
* The damaged fuel bundle is currently over an area of empty fuel storage racks
* The Refuel Floor Exhaust Radiation Monitor is reading 2.3 x 10-31JCilcc
* The Refuel Floor Area Radiation Monitor is reading 14 mR!Hr lAW HC.OP-AB.CONT-0005(0)
"Irradiated Fuel Damage",
which ONE of the following immediate actions is required?
A. Place spent fuel bundle in its new designated storage location.
B. Place spent fuel bundle in the nearest storage location.
C. Start Filtration, Recirculation and Ventilation System (FRVS) fans. D. Suspend handling of the damaged fuel bundle.
K&A Rating: K&A Statement:
Justification:
234000 A 1.02 (3.3/3.8)
Ability to predict and/or monitor changes in parameters associated with operating the FUEL HANDLING EQUIPMENT controls including:
Refuel floor radiation levels/ airborne levels A. Incorrect but plausible:
Plausible if the applicant determines that placing spent fuel bundle in its new designated storage location is an immediate safe action. B. Incorrect but plausible:
Plausible if the applicant determines that placing spent fuel bundle in the nearest storage location is the safest immediate action based on stem condition indicating fuel bundle over empty storage location.
C. Incorrect but plausible:
Plausible due to Refuel Floor Exhaust Radiation Monitor reading > 2.0 x 10-3 !JCi/cc,
: however, Filtration, Recirculation and Ventilation System (FRVS) fans start automatically.
D. Correct:
HC.OP-AB.CONT-0005(0)
"Irradiated Fuel Damage" directs suspending of the handling of Irradiated Fuel/Components based on entry conditions due to refuel floor radiation alarm.
==References:==
HC.OP-AB.CONT-0005(0)
"Irradiated Fuel Damage" Student Ref: NONE Learning Objective:
Question Source: Question History:
Cognitive Level: 10CFR: Comments:
New None Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
CFR 41 .5/45.5 Changed from A4.01 to A1.02, due to Q 56 change to balance sample plan.
QUESTION 62 While operating at 35% power, the following indications are received:
* Alarm D3-C5 TURBINE GENERATOR VIB HI
* GRIDS point A2529 MAIN TURBINE BEARING 11 VIB indicates 9 mils and steady lAW HC.OP-AB.BOP-0002
'MAIN TURBINE',
which ONE of the following is required?
A. With system operator concurrence, adjust MVAR loading.
B. Lock the Mode Switch in SHUTDOWN and then immediately trip the Main Turbine.
C. Monitor GRIDS or Main Control Room System 1 Computer for further bearing degradation.
D. Reduce Recirc Pump speed to minimum, then lock the Mode Switch in SHUTDOWN and immediately trip the Main Turbine.
K&A Rating: 241 OOOA 1.23 (2.8/2.8)
K&A Statement:
Ability to predict and/or monitor changes in parameters associated with operating the REACTOR/TURBINE PRESSURE REGULATING SYSTEM controls including:
A 1.23 Main Turbine Vibration Justification:
A. Incorrect but plausible:
Plausible if the operator does not recall that the bearing vibration magnitude does not meet the immediate action criteria (Bearing 1-10 must be >= 12 mils for tripping the reactor and turbine).
Section B.1 of HC.OP-AB.BOP-0002 directs the operator to adjust MVAR loading.
B. Correct:
lAW HC.OP-AB.BOP-002 Immediate
: Actions, if bearing 11-12 Vibration>=
8 mils@ 1800rpm AND reactor power is>+ 24%, then lock the mode switch in shutdown and immediately trip the main turbine.
C. Incorrect but plausible:
Plausible if the operator does not recall that the bearing vibration magnitude does not meet the immediate action criteria (Bearing 1-10 must be >= 12 mils for tripping the reactor and turbine).
D. Incorrect but plausible:
Plausible if the operator does not recall that the bearing vibration magnitude does not meet the immediate action criteria.
The HC.OP-AB.BOP-002 Retainment Override states that if bearing 11-12 Vibration
>= 7 mils @ 1800rpm AND reactor power is>+ 24%, then reduce recirc pump speed to minimum and then lock the mode switch in shutdown and immediately trip the main turbine.
The immediate action criteria
(>=8mils, no recirc pump speed reduction) overrides the retainment override.
==References:==
HC.OP-AB.
BOP-0002 HC.OP-AR.ZZ-0014 Learning Objective:
MNTURBE019 Question Source: New Question History:
N/A Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/45.5 Comments:
QUESTION 63 Given the following conditions:
* The Reactor failed to SCRAM when the Mode Switch was placed in Shutdown
* Reactor power is 14% and Reactor pressure is 11 05 psig
* The 'A' RFP speed has slowed to approximately 2500 rpm and remains steady
* The RFP TURBINE AUTO XFR TO MANUAL (B3-F3) annunciator is in alarm Which ONE of the following describes the reason for the 'A' RFP speed reduction?
The 'A' RFP is responding to ... A. a control signal failure.
B. a gross failure of a Main Steam Flow transmitter.
C. a Redundant Reactivity Control System runback.
D. the Setpoint Setdown feature of Digital Feedwater Control.
K&A Rating: 25900*1A3.11 (3.2/3.7)
K&A Statement:
Ability to monitor automatic operations of the REACTOR FEEDWATER SYSTEM including:
A3.11 Reactor feed pump runbacks:
Plant-Specific Justification:
A Incorrect but plausible:
Speed will remain essentially constant with a Control Signal failure.
B. Incorrect but plausible:
A single steam flow transmitter failure changes control to single element.
C. Correct:
With RPV pressure at 11 05psig, RRCS will swap the RFP to manual, initiate a RFP Runback signal and lower speed to 2500rom (RRCS Setpoint:
>1071psig w/ .4% pwr for 25 seconds)
D. Incorrect but plausible:
Calling for a lower level RFP speed would not stabilize pump speed at 2500 rpm and a transfer to manual control would not occur ..
==References:==
HC.OP-SO.SA-0001 Student Ref: NONE Learning Objective:
FWCONTE011 Question Source: Bank Question History:
HC Bank #318 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7/45.7 Comments:
QUESTION 64 A plant startup is in progress when the 'A' RPS Motor-Generator Voltage Regulator fails causing generator output voltage to decrease to approximately 1 OOVAC. Which ONE of the following describes the effect of this condition on the Main Steam Line (MSL) Radiation Monitors?
A. The reduced voltage causes a DOWNSCALE trip of MSL Radiation Monitors RE-N006A and RE-N0068.
B. The reduced voltage causes a DOWNSCALE trip of MSL Radiation Monitors RE-N006A and RE-N006C.
C. Power is lost to MSL Radiation Monitors RE-N006A and RE-N006C, resulting in an INOP trip. D. Power is lost to MSL Radiation Monitors RE-N006A and RE-N0068, resulting in an INOP trip.
K&A Rating: 272000K6.01 (3.0/4.2)
K&A Statement:
Knowledge of the effect that a loss or malfunction of the following will have on the RADIATION MONITORING SYSTEM K6.01 Reactor protection system Justification:
A. Incorrect but plausible:
Any voltage reduction would be momentary due to the UV trip of the EPA breakers; an INOP trip occurs on a loss of power to the MSL Rad Monitors.
This would result in an INOP trip of MSL Rad Monitors RE-N006A
& C since they are powered from RPS Bus 'A' B. Incorrect but plausible:
Any voltage reduction would be momentary due to the UV trip of the EPA breakers; C. Correct:
When the 'A' RPS MG output is less than 108 VAC, the EPA breakers on the MG output to the 'A' RPS Bus trip on undervoltage, causing a loss of the 'A' RPS bus. This results in an INOP trip of the MSL Rad monitors RE-N006A
& C since they are powered from RPS Bus 'A'. D. Incorrect but plausible:
RE-N006A
& C trip, not A & B. See 'C' for additional explanation.
==References:==
HC.OP-SO.SB-0001 Student Ref: NONE HC.OP-SO.SP-0001 Learning Objective:
RMSYSOE002 Question Source: Bank Question History:
HC Bank #52 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
1 OCFR55: CFR 41.7/45.7 Comments:
QUESTION 65 Given the following conditions:
* The Diesel Generator Room Carbon Dioxide Fire protection system is aligned for automatic operation.
* A valid EDG room high temperature condition has just occurred.
Which ONE of the following describes how the Diesel Generator Room Carbon Dioxide Fire protection system responds?
A A discharge alarm occurs. C02 with a wintergreen scent is discharged into the room immediately.
B. A pre-discharge alarm is activated.
After a time delay, C02 with a wintergreen scent is discharged into the room. C. A pre-discharge alarm is activated.
No C02 is discharged into the room until a valid smoke detector alarm is received.
D. A pre-discharge alarm is activated and a wintergreen scent is discharged into the room. After a time delay, C02 is discharged into the room.
K&A Rating: 286000A4.01 (3.3/3 2) K&A Statement:
Ability to manually operate and/or monitor from the control room A4.01 System alarms and indicating lights. Justification:
A Incorrect but plausible:
A pre-discharge alarm is activated.
After a time delay, C02 with a wintergreen scent is discharged into the room, not immediately.
B. Correct:
lAW HC.OP-SO.KC-0002, Section 3.2.3 and 5.2, a pre-discharge alarm is activated.
After a time delay, C02 with a wintergreen scent is discharged into the room. (The wintergreen scent comes from the odorizer bottle which is activated by the C02 flow.) C. Incorrect but plausible:
A pre-discharge alarm is activated.
After a time delay, C02 with a wintergreen scent is discharged into the room. The EDG room alarm actuates on Hi Temp, not a smoke detector alarm. D. Incorrect but plausible:
A pre-discharge alarm is activated.
After a time delay, C02 with a wintergreen scent is discharged into the room. The wintergreen scent comes in with the C02, not before.
==References:==
HC.OP-SO.KC-0002 Student Ref: NONE Learning Objective:
FIRPROE015 Question Source: Bank Question History:
HC Bank# 229 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7/45.5 to 45.8 Comments:
QUESTION 66 Which ONE of the following situations complies with OP-AA-103-1 01, CONTROL ROOM ACCESS CONTROL, and OP-AA-101-111, ROLES AND RESPONSIBILITIES OF ON-SHIFT PERSONNEL?
A. During a Reactor Startup the NCO "At the Controls" has ultimate responsibility to limit access to the control room to only Licensed Personnel.
B. In OPCON 1, 2, or 3 a minimum of 2 NCOs will be "At the Controls".
It may be lowered to 1 NCO if properly relieved by the CRS. C. In OPCON 1 with only 2 NCOs assigned to the control room, when the PO is in the computer room retrieving a P-1, the "At the Controls" NCO is allowed to initiate normal drywell makeup on the 1 OC650 panel. D. In OPCON 1 with only 2 NCOs in the control room, the "At the Controls" NCO is allowed to check a reading on the SRV temperature recorder on the 1 OC650C panel.
K&A Rating: 2.1.1 (3.8/4.2)
K&A Statement:
2.1.1 Knowledge of conduct of operations requirements.
Justification:
A. Incorrect but plausible:
OP-AA-1 03-101 lists personnel who are allowed in the control room for a startup including QA and the NRC. B. Incorrect but plausible:
OP-AA-103-101 states that the on duty SM and CRS shall remain in the Protected Area unless properly relieved.
C. Incorrect but plausible:
lAW OP-AA-1 01-111, section 4.9, it is expected that three duty Licensed Operators will be in the Control Room boundary at all times in OPERATIONAL CONDITIONS 1, 2 and 3, except that one NCO may go to the Computer Room to retrieve the printouts from the PPC (Plant Process Computer) or CMS (Core Monitoring System) as long as the other NCO is "at the controls" in the Inner Horseshoe and is within view of the control panels. By initiating a normal drywell Mkeup, this NCO is NOT within view of the panels and therefore this is NOT an allowed activity.
D. Correct:
OP-AA-1 01-111, Section 4.9 states that "At the Controls" NCO may briefly enter the "Outer Horseshoe" for activities such as checking an instrument reading or responding to an alarm. (the SRV temperature recorder on the 1 OC650C panel is in the outer horseshoe).
E.
==References:==
OP-AA-1 01-111, Section 4.9 Student Ref: NONE Learning Objective:
ADMPR05CE008 Question Source: Bank Question History:
HC Bank# 123 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
1 OCFR55: CFR 41.10/45.13 Comments:
QUESTION 67 While operating with 300 Mvars overexcited (Lagging),
a main generator hydrogen leak occurs.
* Hydrogen pressure has stabilized at 65 psig. Which ONE of the following is the main generator power output limit, based on these conditions?
A. 1000 Mw B. 1150 Mw C. 1250 Mw D. 1325 Mw (Note: See attached graph)
HC.OP-SO.MA-0001 (Q) ATTACHMENT 1 POWER CHANGES DURING OPERATION GENERATOR REACTIVE CAPABILITY AND MINIMUM EXCITATION LIMIT CURVE GENERATOR REACTIVE CAPABILJTY CURVE AT8 d POLE !Jil!OO KVA lBCG RPH VOLTS 0.94 PF U sen 75 PSTG H\DnOGEN PRESSURf 530 VOLTS 1500 120(1 lOGO aoo 400 "' <{ _J (,'1 .20() 0: > 0 _J -I) 0 GO pr "/b
. ' 1 o 7o 60 PSLG )------. v O.DO PDI;; v vu.f]5 O,SC* ----I 1--Jo r-psr [?:S;, / ;< /< //""--...
>-. / '-1'-' / ">< '""
/I L / ... ,"' _1_ r----
ex
' V::'/ ------K ,_ .
I [,;' 20*8 0 6 C* 8(lQ 1\JC ;J 11200 14()1) 150 I I IV 1000
...... r--Ml11lmum Ewr::tt.:.tlott Limit curve
' ""' *-----.___
t9-I ,_ \ ', _,; "--.. .....-
e: k-?5: -----, c 9'5 -:::-......-:
=---::::::,.
C.<JO ' "* ' G.E5 l \\ 0. tliJ 0. EG
-oorJ c r,.c;]qo,rF-<
-800 -1 DOfJ Page43 of47 Rev. 56 K&A Rating: G 2.1.25 (3.9/4.2)
K&A Statement:
Ability to interpret reference materials, such as graphs, cuNes, tables, etc. Justification:
A. Incorrect but plausible:
Based on 400 Mvar line and 45 PSIG pressure cuNe. B. Incorrect but plausible:
Based on 300 Mvar line and 45 PSIG pressure cuNe. C. Correct : based on 300 Mvar line intersecting the 65 PSIG cuNe. D. Incorrect but plausible:
based on 300 Mvar line and 75 PSIG pressure cuNe
==References:==
HC.OP-SO.MA-0001 attachment
#1 Student Ref: Att #1 Learning Objective:
MNGENOE01 0 Question Source: HC Bank Q#214 Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/
Analysis:
X 10CFR: CFR 41.10/43.5/45.12 Comments:
QUESTION 68 lAW TS 2.1.1, Thermal Power shall not exceed 24% of rated thermal power with the Reactor Vessel Steam Dome Pressure less than _(1)_ psig OR Core Flow less than _(2)_% of Rated Flow. w m A. 756 15 B. 756 10 C. 785 15 D. 785 10 K&A Rating: 2.2.22 (4.0/4.7)
K&A Statement:
2.2.22 Knowledge of limiting conditions for operations and safety limits. Justification:
A. Incorrect but plausible:
Pressure too low and flow too high. B. Incorrect but plausible:
Pressure too low. C. Incorrect but plausible:
Flow too high. D. Correct:
lAW Tech Spec 2.1.1, thermal power shall not exceed 24% of rated thermal power with the reactor vessel steam dome pressure less than 785psig or core flow less than 10% or rated flow.
==References:==
TS 2.1.1 Student Ref: NONE Learning Objective:
TECSPECE001 Question Source: Bank Question History:
HC Bank #11 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
1 OCFR55: CFR 41.5/43.2/45.2 Comments:
QUESTION 69 Given the following conditions:
* A Reactor Startup is in progress.
* The Mode Switch in Startup/Hot Standby.
Subsequently, SLCS TANK TROUBLE (C1-E1) annunciator alarms
* The EO isolates a severely leaking SLC storage tank drain line.
* 4450 gallons remains in the storage tank.
* A Chemistry sample shows current boron concentration is 13.7%. Based on this information, the SLC system storage tank sodium pentaborate solution volume/concentration Technical Specification requirement
... A. is being met. B. needs to have only water added to bring it into specification.
C. needs to have only boron added to bring it into specification.
D. needs to have both boron AND water added to bring it into specification.
K&A Rating: 2.2.42 (3.9/4.6)
K&A Statement:
2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
Justification:
A. Incorrect but plausible:
Total volume is below limits. The question stem indicates that the plant is in Mode 2 (mode switch in Startup/Hot Standby),
therefore the applicability statement applies (3.1.5 applies in Modes 1 and 2). Plausible if the applicant believes that the Tech Spec does not apply given that a Reactor startup is in progress.
B. Incorrect but plausible:
Adding water only will dilute the concentration below 13.6%. C. Incorrect but plausible:
Adding boron only will not bring the total volume above limits. D. Correct:
TS 3.1.5 requires a minimum volume of -4746.5 gallons at a concentration of 13.7%. (4782-4640=142 gallons between 13.6% and 14.0%. Subtract%
of this to estimate minimum requirement at 13.7% due to linear slope of line.) Water must be added to bring level up to the minimum.
(-296 gallons.)
Starting at 13.7%, this would dilute the concentration down to -12.8%, less than the minimum required
( 4450/4 7 46x13. 7%). Therefore, boron must also be added to bring the concentration volume within acceptable limits.
==References:==
TS 3.1.5 Student Ref: Tech Spec 3.15 Learning Objective:
SLCSYSE027 Question Source: Modified Question History:
Modified from HC Bank #115 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
41.7 I 41.10 I 43.2 I 43.3 I 45.3 Comments:
QUESTION 70 A vent valve is to be worked on, and it is the only vent valve capable of removing stored energy in the associated system. lAW OP-AA-1 09-115, "Safety Tagging Operations"
... A. a White Caution Tag (WCT) is applied to the vent valve and the work is conducted on the valve with the WCT attached.
B. the Clearing Agent and the Tagging Authority agree that the valve can be VER, and the work is conducted on the valve. C. a Worker's Blocking Tag (WBT) is applied to the vent valve and the work is conducted on the valve with the WBT attached D. an Information Tag (IN F) is applied to the vent valve and the work is conducted on the valve with the INF attached.
Question 70 K&A Rating: 2.2.13 (4.1) K&A Statement:
Knowledge of tagging and clearance procedures.
Justification:
A. Incorrect but plausible:
Not in accordance with the procedure.
B. Correct:
Per component and tagging rules this is the correct process to control the valve. C. Incorrect but plausible:
Not in accordance with the procedure.
D. Incorrect but plausible:
Not in accordance with the procedure
==References:==
0 P-AA-1 09-115 Attachment 2 Component Tagging Rules Learning Objective:
NA0015E004 Question Source: HC Bank Q#62 Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
1 OCFR55: CFR 41.10/45.13 Comments:
Student Ref: NONE QUESTION 71 Given:
* The Refueling SRO reports he is ready to commence core alterations.
* Radiation Protection has been notified and are evacuating the upper regions of the Drywell.
lAW RP-AA-403,
'Administration of the Radiation Work Permit Program',
which ONE of the following actions/controls is a requirement if rad levels in the upper regions of the Drywell are determined to be 1100 mrem/hr?
A. The area is roped off, conspicuously posted, and a flashing light provided as a warning device. B. Qualified personnel will provide direct or remote (e.g., closed circuit TV) roving surveillance and the area is conspicuously posted. C. The area is conspicuously posted and qualified personnel provide direct or remote (e.g., closed circuit TV) continuous surveillance.
D. Qualified personnel will provide direct or remote (e.g., closed circuit TV) continuous surveillance ONLY.
K&A Rating: 2.3.12 (3.2/3.7)
K&A Statement:
2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning
: filters, etc. Justification:
A. Correct:
Correct lAW RP-AA-403 B. Incorrect but plausible:
Requires continuous surveillance C. Incorrect but plausible:
Combination of incorrect statements.
D. Incorrect but plausible:
This is used in lieu of stay time.
==References:==
RP-AA-403 Student Ref: NONE RP-AA-406 Learning Objective:
NOH04ADM024E Question Source: Bank Question History:
HC Bank #26 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
10CFR55:
CFR 41.12,45.9,45.10 Comments:
Learning objective NOH04ADM024-E001 From Memory Describe what the worker is acknowledging when signing a RWP prior to use. lAW RP-AA-403, Administration of the Radiation Work Permit Program QUESTION 72 Given the following conditions:
* The Reactor core has been operating with one or more known fuel pin leaks
* A Reactor SCRAM occurred from 1 00% power
* Both SCRAM Discharge Volume Drain Valves did NOT go full close
* The Reactor Building HVAC Exhaust Rad monitor is reading 4E-5 1-JCi/mL and steady Which ONE of the following rooms would become the most significant radiological hazard AND what action would reduce the radiological hazard? A Reactor Building Equipment Sump BT266; Enter EOP 103/4. B. Reactor Building Equipment Sump AT266; Enter EOP 103/4. C. Reactor Building Equipment Sump BT266; Reset the Reactor SCRAM. D. Reactor Building Equipment Sump AT266; Reset the Reactor SCRAM.
K&A Rating: 2.2.11 (3.8/4.3)
K&A Statement:
Ability to control radiation releases.
Justification:
A. Incorrect but plausible:
The South Sump Room will experience rising rad levels as described in answer choice 'C'. However, the entry condition for EOP 103/4 is Reactor Building HVAC exhaust rad levels above 5E-5 !JCilml.
The stem indicated that current Reactor Building HVAC exhaust levels are currently reading 4E-5 !JCi/mL.
Therefore, EOP 103/4 entry is NOT required.
B. Incorrect but plausible:
The North Sump Room will NOT experience rising rad levels as described in answer choice 'C'. The entry condition for EOP 103/4 is Reactor Building HVAC exhaust rad levels above 5E-5 !JCi/ml.
The stem indicated that current Reactor Building HVAC exhaust levels are currently reading 4E-5 !JCi/mL.
Therefore, EOP 103/4 entry is NOT required.
C. Correct:
The North and South SCRAM discharge volumes drain through a common line to the Reactor Building Equipment Drain Sump (1 BT266) located in the South Reactor Building Sump Room (54' elevation).
If the drain valves did not close (as stated in the stem), a LOCA would exist discharging into the South Sump Room. The leaking fuel would severely raise rad levels in that room. With drain valves open, resetting the SCRAM will reclose the valves and terminate the LOCA. D. Incorrect but plausible:
With drain valves open, resetting the SCRAM will reclose the valves and terminate the LOCA. However, the North Sump Room will NOT experience risif)g rad levels as described in answer choice 'C'.
==References:==
EOP 103/4 Student Ref: M-61-1 Sht 1 & 2 Learning Objective:
CRDHYDE01 0 Question Source: Bank Question History:
HC Bank #205 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.11/43.4/45.10 Comments:
This question is RO level. The first part of the two part answer choice can be answered solely by system knowledge.
Direct EOP entry conditions are considered RO level knowledge.
Understanding that resetting the Reactor SCRAM will reduce radiation levels can be answered solely by system knowledge.
QUESTION 73 Concerning the EOPs, which ONE of the following is the correct name for this EOP symbol? A. Decision Step B. Entry Condition C. Contingent Step D. Execute Concurrently Step K&A Rating: 2.4.17 (3.9/4.3)
K&A Statement:
Knowledge of EOP terms and definitions.
Justification:
A. Incorrect but plausible:
Contains a question the answer to which determines the next step. All decision steps require a choice between two opposite possibilities, for example "YES/NO" or "HIGH/LOW".
B. Incorrect but plausible:
Entry conditions or symptoms are generally brief noun phrases that initiate the EOP actions.
The entry conditions are located at the top of each flowchart within entry condition symbols.
C. Correct:
A contingent action step indicates that until the expected condition contained within the step is met, the corresponding action cannot be performed.
Once operators reach a contingent action step, they wait until the expected condition is met before proceeding.
D. Incorrect but plausible:
Requires the operator to enter the designated procedure and perform the stated actions while continuing in the existing flow path. The execute concurrently symbol is used when an operator is to enter another procedure while at the same time continuing in the current flow path. The symbol contains the number of the concurrent procedure
==References:==
EOP Bases Student Ref: NONE Learning Objective:
INTEOPE001 Question Source: Bank Question History:
HC Bank #78 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
1 OCFR55: CFR 41.10/45.13 Comments:
QUESTION 74 Gross mechanical damage has occurred at the top of the Reactor core. Attempts are being made to determine the full extent of the mechanical damage. Which ONE of the following systems can be used to determine the full extent (axial AND radial location) of the mechanical damage? A. TIP B. SRM C. LPRM D. IRM K&A Rating: K&A Statement:
2.4.3 Ability to identify post-accident instrumentation.
Justification:
A. Correct:
The driving range of the TIP detectors can be used to determine the axial and radial distribution of damage. TIP detector position is read from the TIP Drive Control Units. Axial position above BAF (in inches) can be determined by subtracting the core bottom limit from the detector reading.
Radial location can be determined from Drive Control Unit and Indexer position assignment to LPRM strings.
B. Incorrect but plausible:
SRMs and IRMs axial position can be determined as a distance from the detector full-in or full-out position.
Since there is no definite method of determining that these detectors have stopped moving, except at the full-in and full-out positions, their use is only to determine that full traverse movement is possible.
Therefore, SRMs and IRMs are not useful at defining the extent of (axial) of all damage. They only traverse from below the active fuel to 15" above active fuel centerline.
C. Incorrect but plausible:
LPRMs detectors are not moveable, and therefore are of limited use in determining the extent of damage. D. Incorrect but plausible:
SRMs and IRMs axial position can be determined as a distance from the detector full-in or full-out position.
Since there is no definite method of determining that these detectors have stopped moving, except at the full-in and full-out positions, their use is only to determine that full traverse movement is possible.
Therefore, SRMs and IRMs are not useful at defining the extent of (axial) of all damage. They only traverse from below the active fuel to 15" above active fuel centerline.
==References:==
NOH01 MCD000-00 Student Ref: NONE Learning Objective:
NMRESPE003 Question Source: Bank Question History:
HC Bank #84 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
1 OCFR55: CFR 41.6/45.4 Comments:
QUESTION 75 Given the following:
* The plant is operating at 20% Reactor Power
* The standby EHC pump is C/T out of service
* EHC pressure is 1600 psig Subsequently, the in-service EHC pump trips.
* EHC pressure is lowering at 100 psi/second 5 seconds later, which ONE of the following operator action(s) is required lAW AB.OP-AB.BOP-003
'Turbine Hydraulic Pressure'?
A. Ensure the Bypass Valves control Reactor pressure.
B. Ensure the Control Valves fast close when EHC pressure reaches 530 psig. C. Reduce Recirc speed to minimum and then lock the Mode Switch in SHUTDOWN.
D. Lock the Mode Switch in SHUTDOWN.
Recirc speed is NOT required to be reduced to minimum prior to this action.
K&A Rating: 2.4.11 (4.0/4.2)
K&A Statement:
Knowledge of abnormal condition procedures.
Justification:
A. Incorrect but plausible:
BPVs fail shut with no EHC pressure after-1 minute B. Incorrect but plausible:
The control valves fast close when the fast acting solenoids are energized, which is only on a power load unbalance.
C. Incorrect but plausible:
Recirc pump speed is NOT required to be reduced to minimum prior to this action. D. Correct:
The retainment override in HC.OP-AB.BOP-003 directs the mode switch placed in SHUTDOWN if EHC pressure is <1200psig and lowering.
With both pumps tripped, it will be there immediately.
==References:==
HC.OP-AB.BOP-003 Student Ref: NONE Learning Objective:
ABBOP3E003 Question Source: Bank Question History:
HC Bank #33 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Ana lysis: 1 OCFRSS: CFR 41.10/41.5/43.5/45.13 Comments:
QUESTION 76 Given the following conditions:
* The plant is in Operational Condition 3 with the Electrical Distribution System aligned in the Normal plant lineup
* The E3-B4 STA SERVICE TRANSFORMER 1BX501 TRBL alarm is received in the control room
* A Field Operator subsequently reports that the F-1 FAULT PRESSURE alarm is lit on the 1 BX501 local alarm panel.
* No other actions have been taken Which ONE of the following is the status of power for the vital buses and what is required by Technical Specifications (TS)? The diesel generators for the affected ESF buses ... A. will be supplying bus power. Verify correct breaker alignments and indicated power availability for all channels of the 1 E distribution system. B. will NOT be supplying bus power. Verify correct breaker alignments and indicated power availability for all channels of the 1 E distribution system. C. will be supplying bus power. Verify correct breaker alignments and indicated power availability for ONLY the B and D channels of the 1 E distribution systems.
D. will NOT be supplying bus power. Verify correct breaker alignments and indicated power availability for ONLY the B and D channels of the 1 E distribution systems.
K&A Rating: 295003 2.4.31 (4.2/4 1) K&A Statement:
Partial or complete loss of AC: 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.
Justification:
A. Incorrect but plausible:
Power from 1AX501 is available.
13kV bkrs BS 2-3 and BS 1-2 trip open and Bus section 2 is de-eneergized.
The bus infeed bkr swaps to the 1AX501 feed and the loads remain energized.
The EDGs do NOT start because one infeed is always available.
B. Correct:
Power from 1AX501 is still available, only the 1 BX501 xfmr is OOS as indicated by two alarms given in stem. 13kV bkrs BS 2-3 and BS 1-2 trip open and Bus section 2 is de-eneergized.
The bus infeed bkr swaps to the 1AX501 feed and the loads remain energized.
The EDGs do not start because one infeed is always available.
lAW TS 3.8.1.1 (applicable in Modes 1, 2, and 3) with one offsite circuit unavailable (1 BX501 xfmr OOS), demonstrate operability of the remaining AC sources by performing SR 4.8.1.1.
1.a within 1 hr and at least once per 8 hours thereafter.
SR 4.8.1.1.1.a requires that each of the required independent circuits shall be determined operable by verifying correct breaker alignments and indicated power availability.
C. Incorrect but plausible:
Power from 1AX501 is available.
The bus infeed bkr swaps to the 1AX501 feed and the loads remain energized and the EDGs do NOT start. However, TS 3.8.1. 1 requires that SR 4.8.1.1.1.a be performed within 1 hr. SR 4.8.1.1.1
.a states that each of the required independent circuits be determined operable by verifying correct breaker alignments and indicated power availability.
The SR checks all 4 channels, not just B and D. D. Incorrect but plausible:
TS 3.8.1.1 requires that SR 4.8.1.1.1.a be performed within 1 hr. SR 4.8.1.1.1.a states that each of the required independent circuits be determined operable by verifying correct breaker alignments and indicated power availability.
The SR checks all 4 channels, not just B and D. SRO Only Justification:
This question is SRO only as it requires assessing facility operating limitations in the Tech Specs and their bases, cannot be answered solely by knowing <= 1 hr TS actions or above the line information, cannot be answered by knowing TS safety limits, and involves application of TS required actions and surveillance requirements.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
TS 3.8.1.1.a DWG E-0001 Learning Objective:
1 EACOOE015 Question Source: Bank Student Ref: None Question History:
HC Bank #446 (stem modified
: slightly, but still counts as Bank) Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/45.3 Comments:
QUESTION 77 Hope Creek is at 20% power following a startup from a refueling outage when the plant scrams. You have entered HC.OP-AB.ZZ-0000, Reactor Scram and note the following plant conditions:
* RPV Level-(+33" stable)
* RPV Pressure
-1 000 psig stable
* Mode Switch-Locked in Shutdown position
* All Control Rods fully inserted You have reached step S-8: IF Conditions permit THEN RESET the Scram AND INSERT a Half Scram (if required)
Which of the following conditions would REQUIRE you to INSERT a Half Scram? I. APRM channels "A" and "C" INOP II. IRM channels "E" and "F" INOP Ill 1 Reactor Vessel Steam Dome Pressure High Transmitter INOP A. I only B. II only C. I, II only D. I, II, and II K&A Rating: K&A Statement:
Justification:
29500682.4.47 (4.2/4.2)
SCRAM 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
A. Correct:
Per TS 3.3.1.a.
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least 1 trip system in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.
For the APRM's in OPCON 3-Minimum OPERABLE Channels per Trip System is 2, therefore if 2 APRM were INOP you would only have 1 in that Trip System OPERABLE and would need to insert a Half-scram B. Incorrect but plausible:
Per TS 3.3.1-1 in OPCON 3 you are only required to have 2 IRM's OPERABLE per trip system, since you have 3 available having 1 INOP still leaves 2 that are OPERABLE and so you would NOT have to insert a Half-Scram C. Incorrect but plausible:
see answer choice 'B' above D. Incorrect but plausible:
see answer choice 'B' above, also Reactor Steam Dome Pressure High transmitter is only required in OPCON 1 or 2, since the stem indicates that the plant is in OPCON 3 this would be N/A SRO Only Justification:
This question is SRO only as it requires assessing facility operating limitations in the Tech Specs and their bases, cannot be answered solely by knowing <= 1 hr TS actions or above the line information, cannot be answered by knowing TS safety limits, and involves application of TS required actions and surveillance requirements.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
TS 3.3.1.a Student Ref: NONE Learning Objective:
Question Source: Bank Question History:
HC Bank #479 (2005 NRC Exam, 2009 NRC Audit) Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.5/45.12 Comments:
Used on 2005 NRC exam and 2009 NRC audit exam with no reference given.
QUESTION 78 Given the following conditions:
* Reactor power at 100% * "A" SACS Loop is declared inoperable at 0930 today (9/13)
* Delaware river temperature is 84F and steady
* No other systems or components are OOS/Inoperable Which ONE of the following correctly identifies the latest times/dates for completion of the FIRST Tech Spec required action? A Immediately B. 1130 9/13 C. 0930 9/16 D. 0930 10/16 K&A Rating: K&A Statement:
Justification:
Partial or Total Loss of CCW 2.2.40 Ability to apply Technical Specifications for a system. A. Incorrect but plausible:
Plausible if the applicant believes that 3.7.1.3 is the most restrictive applicable action -> Immediately commence a down power to hot shutdown.
B. Correct:
lAW TS 3.7.1. 1 Action 2, with one SACS subsystem inoperable (as given in the stem-> one SACS loop OOS), and if continued plant operation is permitted by LCO 3.7.1.3 (even with river temperature above 3.7. 1.3 limit of 85F, continued operation is allowable due to the fact that both emergency discharge valves are open and emergency discharge pathways are available->
no other systems OOS in stem), realign at least one affected EDG to the OPERABLE SACS subsystem within 2 hours ->1130 on 9/13. C. Incorrect but plausible:
Plausible if the applicant believes that 3. 7.1. 1 Action a.1.b is the most restrictive applicable action -> the HX must be restored within 72hrs. D. Incorrect but plausible:
Plausible if the applicant believes that 3. 7. 1.1 Action a.1.a is the most restrictive applicable action -> the SACS pump must be restored within 30 days. SRO Only Justification:
This question is SRO only as it requires assessing facility operating limitations in the Tech Spe.cs and their bases, cannot be answered solely by knowing <= 1 hr TS actions or above the line information, cannot be answered by knowing TS safety limits, and involves application of TS required actions and surveillance requirements.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
Learning Objective:
0302-*000.00H-00011 0-10 Obj 8 Question Source: Modified Question History:
Modified from HC Bank #325 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.5/45.11 Comments:
QUESTION 79 A malfunction in the recirculation pump trip circuitry causes a trip of the 'A' Recirculation pump resulting in the following plant conditions:
* Core Flow is 45 percent
* Reactor power is 50 percent Subsequently, the 'B' Recirculation pump trips. All appropriate operator actions are taken for the given plant conditions.
Based on the above, the ... A. NRC must be notified no later than one hour. B. NRC must be notified no later than four hours. C. NRC must be notified no later than eight hours. D. State of New Jersey must be notified no later than 15 minutes.
Question 79 K&A Rating: K&A Statement:
Justification:
295001 G2.4.30 (4.1) 295001 Partial or Complete Loss of Forced Core Flow Circulation G 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external
: agencies, such as the State, the NRC, or the transmission system operator.
A. Incorrect but plausible:
No EAL or one hour notification thresholds reached.
B. Correct:
NRC must be notified in four hours.-RAL 11.3 for manual RPS actuation.
C. Incorrect but plausible:
Eight hour reports are for actuations other than RPS. D. Incorrect but plausible:
No EAL thresholds reached therefore no 15 minute notification required.
SRO Only Justification:
This question is SRO only as it requires determining off site notifications which is a SRO ONLY function.
==References:==
RAL 11.3. Student Ref: EALs and RALs Learning Objective:
NA Question Source: HC Bank 112333 Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.5/45.11 Comments:
QUESTION 80 The plant is operating at 1 00% power with the following:
* I&C has just discovered that all suppression pool temperature indicating channels have a calibration error
* Average suppression pool temperature was indicating 82&deg;F
* Actual average suppression pool temperature has just been confirmed as 112&deg;F
* HPCI is running as part of the quarterly HC.OP-IS.BJ-0001 HPCI MAIN AND BOOSTER PUMP SET 1ST
* No other actions have been taken Which ONE of the following identifies the most limiting TS required action? A. Restore suppression pool average water temperature to within limits within 1 hr or be in at least HOT SHUTDOWN within the next 24hrs and in COLD SHUTDOWN within the following 24hrs. B. Stop all testing which adds heat to the suppression pool and restore the average temperature to less than 95&deg;F within 24hrs or be in at least HOT SHUTDOWN with the next 12hrs and in COLD SHUTDOWN within the following 24hrs. C. Place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in the suppression pool cooling mode. D. Depressurize the reactor pressure vessel to less than 200 psig within 12hrs.
K&A Rating: 295026EA2.01 (4.1/4.2)
K&A Statement:
Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:
EA2.01 Suppression Pool Temperature Justification:
A. Incorrect but plausible:
With suppression pool water temperature
>95F (112F) TS 3.6.2.1.b is applicable.
The 1 hr action reference is for 3.6.2.1.a suppression pool water level outside limits, not temperature.
B. Incorrect but plausible:
The given TS action is required if suppression pool average temperature is greater than 1 05F but less than 11 OF. C. Correct:
Per TS 3.6.2.1, with suppression pool average water temperature greater than 11 OF, place the reactor mode switch in the Shutdown position and operate at least one RHR loop in the suppression pool cooling mode. D. Incorrect but plausible:
The given TS action is required if suppression pool average temperature is greater than 120F (Given as 112F in the stem). SRO Only Justification:
This question is SRO only as it involves application of required TS actions, cannot be answered by solely knowing <= 1 hr TS actions, cannot be answered by solely knowing above the line LCO information, and cannot be answered solely by knowing TS safety limits. The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
TS 3.6.2.1 Student Ref: NONE Learning Objective:
Question Source: New Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.5/45.13 Comments:
QUESTION 81 The plant was at 100% power when a loss of off-site power occurred.
* The reactor has scrammed (all control rods at 00)
* HPCI and RCIC have tripped
* All low pressure ECCS has started on the Level 1 RPV water level signal
* ADS is inhibited Current Plant conditions are:
* RPV level:
* RPV pressure:
* SP level: -140 inches and dropping slowly 900 psig and slowly rising 76inches
* SP temperature:
113&deg; F lAW HC.OP-EO.ZZ-01 01, which ONE of the following does the CRS direct? A. Use the main turbine bypass valves to lower reactor pressure to 400 psig to allow the ECCS systems to inject B. Immediately emergency depressurize the reactor using alternate depressurization systems C. Ensure that one or more low pressure ECCS systems are available and before -185 inches emergency depressurize the reactor D. Use the Safety Relief Valves to lower reactor pressure to 600 psig and restore level with the secondary condensate pumps K&A Rating: 295031 EA2.04 (4.6/4.8)
K&A Statement Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: EA2.04 Adequate Core Cooling Justification:
A. Incorrect but plausible:
Guidance lAW AL/C-1 0 of HC.OP-EO.ZZ-01 01, guidance is to perform an ED before -185 inches. With a loss of offsite power, there is no powerfor EHC pumps, and MSIV are closed B. Incorrect but plausible:
ED criteria is not currently met, ED not permitted C. Correct:
lAW AL/C-10 of HC.OP-EO.ZZ-0101, with one or more low pressure ECCS pumps running and lined up for injection, ED is required before -185 inches which will allow for low pressure ECCS restoration of RPV level. D. Incorrect but plausible:
Loss of offsite power, no power to secondary condensate pumps SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
HC.EO-OP.ZZ-01 01, Rev. 11 HC.EO-OP.ZZ-01 01 Bases, Rev. 4 Learning Objective:
E01 01 E006 Question Source: HC Bank 34404 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
10CFR55:
CFR 41.10/43.5/45 13 Comments:
Student Ref: NONE X QUESTION 82 Given the following conditions:
* The plant is operating at 100% power
* LAC Police report a marsh fire directly beneath the New Freedom 500kV line (5023) lAW HC.OP-AB.BOP-0004, "Grid Disturbances",
the CRS will implement procedures that will ... A. scram the reactor.
B. remove the 5023 line from service.
C. result in a Reactor power reduction.
D. declare all off-site power sources inoperable per TS 3.8.1.1 or 3.8.1.2.
K&A Rating: 70000AA2.02 (3.2/3.8)
K&A Statement:
Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:
AA2.05 Operational status of offsite circuit Justification:
A. Incorrect but plausible:
A Reactor SCRAM is not required per HC.OP-AB.BOP-004.
: However, a Reactor SCRAM would be required if the applicant mistakenly believes that a fire off site would result in entry into HC.OP-AB.ZZ-0135 SBO/LOOP/EDG Malfunctions.
B. Correct:
lAW HC.OP-AB.BOP-0004, Grid Disturbances, part D, if a fire is reported that threatens or has the poses a threat to an off-site distribution line, measures shouls be taken to remove the line from service expeditiously
. C. Incorrect but plausible:
A power reduction is not required per HC.OP-AB.BOP-0004 in the event of a gr D. Incorrect but plausible:
Entry into TS would only be required per HC.OP-AB.BOP-0004 if the 500kV switchyard voltage falls below 493kV. SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator actions.
AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
HC.OP-AB.BOP-0004 Student Ref: NONE HC.OP-AB.ZZ-0135 Learning Objective:
Question Source: New Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/43.5/45.5/45.7/45/8 Comments:
QUESTION 83 Given the following conditions:
* A LOCA has occurred
* All rods are full in
* RHR pumps ''A" and "8" are NOT available
* HPCI and RCIC are not available
* Reactor water level is -100 inches and steady
* Reactor Pressure is 700 psig
* Drywell pressure is 25 psig and rising at 5 psig per minute
* Suppression Chamber Pressure is 23 psig and rising at 5 psig per minute
* Suppression Pool level is 1 06 inches and rising at 2 inches per minute
* Secondary Condensate Pumps total flow is 12,000 gpm Which ONE of the following is the correct EOP mitigation strategy for this event? A. Emergency Depressurize and then inject with sources internal to the containment.
B. Emergency Depressurize and then inject with sources external to the containment.
C. Inhibit ADS and remain at pressure to conserve inventory; inject with sources internal to the containment.
D. Inhibit ADS and remain at pressure to conserve inventory; inject with sources external to the containment.
Question 83 K&A Rating:
295009 G 2.4.6 (4.7) K&A Statement:
Knowledge of EOP Mitigation Strategies Justification:
A. Correct:
Since PSP can not be maintained ED is required.
If adequate core cooling is assured then terminate injection from outside conatainment.
B. Incorrect but plausible:
SP is high and going higher, do not want to continue with injection external to containment.
. C. Incorrect but plausible:
Need to ED because PSP can not be maintained.
D. Incorrect but plausible:
SP is high and going higher, do not want to continue with injection external to containment.
. SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
EOP 101 and 102 Student Ref: EOP 102 graphs Learning Objective:
NA Question Source: HC Bank 34414 Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.5/45.13 Comments:
QUESTION 84 Given the following conditions:
* Unit 1 is at 60% power for waterbox cleaning and control rod patter
* Control rods 18-43, 18-19, 42-43, and 42-19 have just been ins
* A CRD pump is currently tagged
* At 0127, B CRD pump trips, maintenance is investigating
* At 0128, accumulator trouble alarms are received for H 30-31 and 42-43
* At 0130, charging header pressure
<940 psig
* At 0131, Field Operator reports HCU pressure for. CU 30-31 and 42-43 are 925 psig
* At 0133, accumulator trouble alarms are recei d for HCU 18-15 and 22-31, also confirmed with the Field Operator at 930 ps* In accordance with HC.OP-AB.IC-0001, Cant I Rod: ( 1) Which of the following actions is re ired? (2) What is the basis for the require ent of that action? A. (1) Place the mode switch* SHUTDOWN immediately (2) Allow reactor shutdo n based upon average control rod scram times B. (1) Place the mod (2) Allow chargi witch in SHUTDOWN immediately header pressure alone to fully insert control rods C mode switch in SHUTDOWN by 0153 (2) Allow dequate time to restore a CRD pump to service D. ace the mode switch in SHUTDOWN by 0153 Allow charging header pressure alone to fully insert control rods K&A Rating: 295022 AA2.01 (3.5/3.6)
K&A Statement:
Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS: AA2.01 Accumulator Pressure Justification:
A. Incorrect but plausible:
If applicant believes that immediate action is required.
This would be required with RPV pressure
<900 psig. While control rod scram times are affected by lower accumulator pressures, this is not the reason for the required action B. Incorrect but plausible:
If applicant believes that immediate action is required.
This would be required with RPV pressure
<900 psig. Accumulator pressure is the first to start control rod motion for a scram, which is followed up by reactor pressure completing the control rod scram; this is not the reason for the required action C. Correct:
lAW HC.OP-AB.IC-0001 and TS 3.1.3.5, if reactor pressure is >900 psig the mode switch must be placed in SHUTDOWN within 20 minutes if: two or more scram accumulators are inoperable coincident with charging header pressure
<940 psig. The basis for this action is that it allows for adequate time to return a CRD pump to service to restore charging header pressure.
D. Incorrect but plausible:
If applicant believes that 20 minutes is allowed.
Accumulator pressure is the first to start control rod motion for a scram, which is followed up by reactor pressure completing the control rod scram; this is not the reason for the required action
==References:==
HC.OP-AB.IC-0001, Rev. 14 Student Ref: NONE TS 3.1.3.5 Bases, Amendment 183 Learning Objective:
NOH01 CRMECHC:
13 Question Source: New Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.5/45.13 Comments:
QUESTION 85 The plant is at 100% power with all systems operational and in a normal lineup. A LOCA occurs, and all systems respond as designed and initiate injection into the vessel. Current plant conditions are as follows:
* RPV Pressure is 60 psig and lowering slowly
* RPV water level is -66 inches and rising slowly
* Torus level is 126 inct1es and rising slowly
* Drywell Sprays are in service
* Drywell pressure is 7 psig, lowering slowly lAW HC.OP-EO.ZZ-01 02, "Primary Containment Control",
which ONE of the following systems is to be secured?
A. Drywell Sprays B. HPCI C. Core Spray D. RCIC K&A Rating: 295029 EA2.01 (3.9/3.9)
K&A Statement Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: EA2.01 Suppression Pool Water Level Justification:
A Correct:
Per SP/L-19 and SP/L 20 of HC.OP-EO.ZZ-0102, Primary Containment
: Control, termination of drywell sprays is required if Torus Level cannot be maintained below 124 inches. This is to ensure operability of the suppression chamber to drywell vacuum breakers, to relieve noncondensibles into the drywell and equalize drywell and suppression chamber pressures.
B. Incorrect but plausible:
If applicant does not recall that HPCI suction has transferred to the Torus with Torus level at 126 inches. Additionally, if applicant does not recall that HPCI isolates on low steam pressure at 100 psig, may also choose this answer. C. Incorrect but plausible:
If applicant does not recall that termination of systems injecting into the RPV is only required for those systems with suction outside of primary containment.
D. Incorrect but plausible:
If applicant does not recall either (1) that RCIC suction does not realign to Torus on high Torus water level (2) RCIC isolates on low steam pressure at 64.5 psig. SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
HC. OP-EO.ZZ-01 02, Primary Containment Control Student Ref: NONE HC.OP-EO.ZZ-0102 Bases Learning Objective:
EOP1 02E009 Question Source: Modified HC Bank ID 35663 Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/
Analysis:
X 10CFR55:
CFR 41.10/43.5/45.13 Comments:
QUESTION 86 Plant conditions are as follows:
* Reactor shutdown and all control rods are fully inserted
* No High Pressure Injection is available
* Reactor water level-130 inches and dropping slowly
* Reactor pressure 330 psig and dropping slowly
* Drywell pressure 19 psig and rising slowly HV-F017B, LPCIInjection Valve, starts to auto open and, with all ECCS Pumps available and reactor level rising, the PO manually overrides HV-F017B.
One minute later, a 1 OA401 Bus Lockout causes a trip of 'A' RHR and 'A' Core Spray pumps resulting in reactor level dropping to -170 inches. The PO depresses the HV-F017B OPEN pushbutton.
As the valve begins to open the following alarms and indications are observed:
* "RHR DIV 2 OUT OF SERVICE" computer alarm
* Amber status light next to HV-F017B is flashing
* HV-F017B Valve shows dual indication
* 'B' RHR Pump Flow is 1,500 gpm and steady Which ONE of the following identifies the impact on 'B' RHR and the action(s) required per HC.OP-EO.ZZ-0101,
'RPV Control'?
Impact on 'B' RHR A Pump continues to run with Min Flow Valve closed B Pump continues to run with Min Flow Valve open C Pump continues to run with Min Flow Valve closed D Pump continues to run with Min Flow Valve open Action Required lAW HC.OP-EO.ZZ-101 Steam Cooling PRIOR to RPV Emergency Depressurization RPV Emergency Depressurization ONLY RPV Emergency Depressurization ONLY Steam Cooling PRIOR to RPV Emergency Depressurization K&A Rating: 203000 A2.11 (3.4/3.6)
K&A Statement:
Ability to (a) predict the impacts of the following on the RHR/LPCI:
INJECTION MODE (PLANT SPECIFIC);
and (b) based on those predictions, use procedures to correct,
: control, or mitigate the consequences of those abnormal conditions or operations:
A2.11 Motor operated valve failures Justification:
A. Incorrect but plausible:
With RHR pump flow above 1270 gpm, the min flow valve will be closed. The flashing indicator next to F017 is indicative of tripping of valve thermal overloads.
Steam cooling is NOT required since injection sources are available.
B. Incorrect but plausible:
With RHR pump flow above 1270 gpm, the min flow valve will not be open. The flashing indicator next to F017 is indicative of tripping of valve thermal overloads.
RPV ED will be required per EOP-101 since injection systems are available.
C. Correct:
With RHR pump flow above 1270 gpm, the min flow valve will be closed. The flashing amber indicator next to the F017 valve indicates that the valve has stopped traveling due to thermal overload actuation.
RPV ED will be required per EOP-101 since injection systems are available.
D. Incorrect but plausible:
With RHR pump flow above 1270 gpm. the min flow valve will not be open. The flashing indicator next to F017 is indicative of tripping of valve thermal overloads.
Steam cooling is NOT required since injection sources are available.
SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
NOH01 RHRSYSC, Rev. 10 HC.OP-EO.ZZ-01 01, Rev. 11 HC.OP-AR.ZZ-0005, Rev. 22 EOP 101 Learning Objective:
RHRSYSE007 Question Source: New Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Student Ref: NONE Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/45.6 QUESTION 87 Given the following conditions:
* The plant is operating at 90% power
* The HPCI Pump is being operated in the Full Flow Test Lineup
* HPCI discharge pressure is 1150 psig While attempting to adjust pump flow, the flow controller setpoint remains stationary at 5300 gpm in AUTO. Subsequently, the PO shifts the HPCI flow controller to MANUAL and reports that the pump develops rated flow. lAW Hope Creek Technical Specifications, HPCI ... A. is Operable B. is Inoperable C is Operable But Degraded D. is Operable But Non-Conforming K&A Rating: K&A Statement:
Justification:
206000A2.
i4 (3.3/3.4)
Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use procedures to correct,
: control, or mitigate the consequences of those abnormal conditions or operations:
A2.14 Flow controller failure A. Incorrect but plausible:
lAW the TS Bases, HPCI must be in AUTO with a setpoint of 5700gpm and be capable of rated flow and discharge pressure to be declared Operable.
B. Correct:
HPCI is inoperable because it is NOT capable of meeting all surveillance requirements listed in the TS. lAW the TS 3.5. i Bases, HPCI must be in AUTO with a setpoint of 5600gpm and be capable of rated flow and discharge pressure to be declared operable.
C. Incorrect but plausible:
HPCI could not be considered Operable But Degraded unless the setpoint was stuck at 5600gpm in AUTO. D. Incorrect but plausible:
Operable But Non-Conforming is not applicable to the conditions given in the stem. SRO Only Justification:
This question is SRO only as it requires knowledge of information in the TS Bases. No reference was provided (similar to example IV of the NRC "Clarification Guidance for SRO-Only Questions").
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
HC.OP-SO.BJ-0001 SH.OP-AP.ZZ-Oi 08 TS 3.5.1 Learning Objective:
HPCIOOE013 Question Source: Bank Student Ref: NONE Question History:
HC Bank #48 (Used on 2003 NRC Exam) Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/45.6 Comments:
QUESTION 88 Given the following conditions:
* Refueling is in progress.
* The Reactor Mode Switch is locked in REFUEL.
* Source Range Monitors A, C, and D are operable
* SRM B is inoperable.
* Shutdown margin has been verified.
* All control rods are at position
: 00. As a fuel assembly is taken to the fuel pool through the transfer canal, I&C determines that the 'C' SRM has failed its weekly surveillance test. lAW Hope Creek Technical Specifications, core alterations
... A. can continue with no restrictions.
B. must be immediately suspended.
C. can continue if no fuel movement occurs. D. can continue only in the quadrants monitored by SRMs A and D.
Question 88 K&A Rating: 215004 G 2.2.25 (4.2) K&A Statement:
Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. Ju.stification:
A. Incorrect but plausible:
Must restrict movement in quadrants that have no SRM's operable.
B. Incorrect but plausible:
Suspending all moves is overly restrictive.
Moves can be performed if there is SRMs operable in the quadrant and an adjacent quadrant.
C. Incorrect but plausible:
Core alterations other than fuel moves are also effected by this TS. D. Correct:
One operable SRM must be in the quadrant where the core alteration is taking place and one in an adjacent quadrant..
SRO Only Justification:
This question is SRO only as it requires knowledge of the Bases in the technical specifications which is a SRO function.
==References:==
TS 3.9.2 Bases Student Ref: NONE Learning Objective:
NA Question Source: HC Bank #34795 Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 43.2 Comments:
QUESTION 89 lAW HC.OP-EO.ZZ-0202,
'EOP Bases', which ONE of the following states the MINIMUM number of Safety Relief Valves (SRVs) that must be opened during an Emergency Depressurization and the basis for that minimum number? A. 4 SRVs; To remove all decay heat from the core at a pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow. B. 4 SRVs; To provide the minimum depressurization rate required to ensure the low pressure ECCS systems inject soon enough to minimize the amount of time water level is below the top of active fuel. C. 5 SRVs; To remove all decay heat from the core at a pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow. D. 5 SRVs; To provide the minimum depressurization rate required to ensure the low pressure ECCS systems inject soon enough to minimize the amount of time water level is below the top of active fuel.
K&A Rating: 218000G2.4.18 (3.3/4.0)
K&A Statement:
ADS 2.4.18 Knowledge of the specific bases for EOPs. Justification:
A. Incorrect but plausible:
See answer choice 'C' explanation.
Plausible if the applicant does not recall that 5 SRVs are required.
B. Incorrect but plausible:
See answer choice 'C' explanation.
Plausible if the applicant does not recall that 5 SRVs are required and/or the bases for requiring 5 SRVs. C. Correct:
lAW HC.OP-EO.ZZ-0202, 5 ADS valves are required for RPV-ED. The Minimum Number of SRVs Required for Emergency Depressurization (MNSRED) is the least number of SRVs which corresponds to a Minimum Steam Cooling Pressure (MSCP) sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding MSCP. The MNSRED is utilized to ensure the RPV will depressurize and remain depressurized when emergency depressurization is required.
D. Incorrect but plausible:
See answer choice 'C' explanation.
Plausible if the applicant does not recall the bases for requiring 5 SRVs. SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
HC.OP-EO.ZZ-0202 Student Ref: NONE Learning Objective:
EOP202E003 Question Source: Bank Question History:
HC Bank #138 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/
Analysis:
1 OCFR55: CFR 41.10/43.1/45.13 Comments:
QUESTION 90 Given the following conditions:
* The plant is in OPCON 4
* RCS Temperature is 182 oF with a 10&deg;F/hr heat up rate
* RPV Level is 88 inches and steady
* RHR is in the Shutdown Cooling mode of operation with the '8' Loop in service
* The 'A' and 'C' EDGs are tagged out for planned maintenance A Loss of Offsite Power occurs. Following the Emergency start of the 'B' and 'D' EDGs, a generator differential overcurrent condition occurs on the '8' EDG. Assuming no additional action, which ONE of the following identifies the required actions for this event lAW HC.OP-AB.RPV-0009,
'Shutdown Cooling'?
A. CLOSE the MSIVs B. Restore the '8' RHR pump C. Place RWCU in service and fully OPEN ED-V035 D. Cross tie 'D' RHR to '8' RHR loop K&A Rating: 2640002.4.9 (3.8/4.2)
K&A Statement:
EDGs 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.
Justification:
A. Incorrect but plausible:
lAW HC.OP-AB.RPV-0009, the MSIVs should only be closed it forced circulation canot be established using preferred RHR loops or reactor recirculation IF RPV level cannot be maintained
<90 inches. B. Incorrect but plausible:
HC.OP-AB.RPV-0009 does direct the restoration of the tripped RHR pump, however, due to the trip of the '8' EDG, the '8' RHR pump cannot be returned to service.
C. Incorrect but plausible:
HC.OP-AB.RPV-0009 does direct that RWCU be placed in service as an alternate means of decay heat removal,
: however, RWCU is not available due to the LOOP (power supplies:
108254 and 108264. D. Correct:
Upon LOOP, 10A401 and 10A403 are denergized due to the A and C EDGs being tagged out. The generator differential on the 8 EDG will trip the 8 EDG rendering the 1 OA402 vital bus denergized.
The only bus with power will be the 1 OA404 bus being supplied by the D EDG. Loop 8 RHR SID Cooling will deenergize (8 RHR Pump). lAW HC.OP-AB.RPV-0009, Step E.5, operators should attempt to cross tie the D RHR pump into the 8 Loop of S/0 cooling using Attachment 3 to restore core cooling.
SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
NOH01RHRSYSC-i0 Student Ref: NONE HC.OP-A8.RPV-0009 Learning Objective:
RHRSYSE001 Question Source: New Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.5/45.13 Comments:
QUESTION 91 While operating at 100% power, the following indications are noted:
* The Reactor Building Steam Vent Blowout Panel 1 AS224 indicates open on the RM-11
* Visual observation confirms that 1AS224 is NOT fully closed
* Reactor Building d/p is holding steady at -.28" WG Which ONE of the following describes the operational impact and actions (if any) required?
A. Loss of secondary containment integrity.
Take actions lAW HC.OP-AB.CONT-0003, Reactor Building.
B. Loss of secondary containment integrity.
Take actions lAW EOP-1 03/4, Reactor Building and Rad Release Control.
C. Potential loss of secondary containment.
Start an additional RBVS exhaust fan JAW HC.OP-SO.GR-0001, Reactor Building Ventilation System Operation.
D. Potential loss of secondary containment.
NO additional actions are required unless Reactor Building d/p is greater than -.30" WG.
Question 91 K&A Rating: 290001 A2.02 (3.7) K&A Statement:
Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT; and (b) based on those predictions, use procedures to correct,
: control, or mitigate the consequences of those abnormal conditions or operations:
A2.02 Excessive outleakage Justification:
A. Correct:
By TS definition, TS 3.6.5.1, secondary containment integrity is lost because the blowout panel is not closed and sealed. AB.CONT-0003, Reactor Building discusses mitigation actions.
B. Incorrect but plausible:
No entry condition exists for EOP 103/4. C. Incorrect but plausible:
secondary containment integrity is lost because the blowout panel is not closed and sealed. D. Incorrect but plausible:
secondary containment integrity is lost because the blowout panel is not closed and sealed. SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
TS 3.6.5.1 AB.CONT-0003 Student Ref: NONE Learning Objective:
ABCNT3E007 Question Source: HC Bank 80648 Question History:
HC NRC Exam 2010 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 43.5 Comments:
QUESTION 92 Given the following conditions:
* Main Generator output is 915 MWe
* The 'B' Instrument Gas Compressor is tagged for routine maintenance Subsequently, the 'A' Instrument Gas Compressor trips.
* Instrument Gas pressure is 55 psig and slowly lowering
* All attempts to restore Instrument Gas pressure have failed Instrument Gas pressure is now 49 psig. lAW HC.OP-AB.COMP-0002
'Primary Containment Instrument Gas', the CRS must direct ( 1) in order to prevent (2) A. (1) closing the Inboard MSIVs, THEN locking the Mode Switch in Shutdown (2) the pressure and power spike from the Inboard MSIVs drifting closed. B. (1) reducing Recirc Pump speed to Minimum, THEN locking the Mode Switch in Shutdown (2) the pressure and power spike from the Outboard MSIVs drifting closed. C. (1) reducing Recirc Pump speed to Minimum, THEN locking the Mode Switch in Shutdown (2) the pressure and power spike from the Inboard MSIVs drifting closed. D. (1) closing the Outboar*d MSIVs, THEN locking the Mode Switch in Shutdown (2) the pressure and power spike from the Outboard MSIVs drifting closed.
K&A Rating: K&A Statement:
Justification:
239001 2.4.11 (4.0/4.2)
Main and Reheat Steam: 2.4.11 Knowledge of abnormal condition procedures.
A. Incorrect but plausible:
MSIV closure is an RPS actuation B. Incorrect but plausible:
PCIG has no effect on outboard MSIVs, RRPs are at minimum due to the PCP trip. C. Correct:
lAW HC.OP-AB.COMP-0002 Retainment Overide (not an Immediate Action),
if lnstriument Gas Pressure is<= 50 psig, reduce recirc pump speed to minimum, lock the mode switch in shutdown, place RCIC and/or HPIC in service, close the Inboard MSIVs, and Close HV-5124A and HV-51248.
D. Incorrect but plausible:
MSIV closure is an RPS actuation.
Outboard MSIVs are not affected by a loss of instrument gas. SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
HC.OP-AB.COMP-0002 Student Ref: NONE Learning Objective:
ABCMP2E001 Question Source: Bank Question History:
HC Bank #226 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.5/45.13 Comments:
QUESTION 93 The plant is operating at 100% power with the following conditions:
* Pressure transmitter SM-PT-N076A has failed its surveillance calibration check
* The failure is non-conservative Which ONE of the following describes the actions(s) required by Hope Creek Technical Specifications?
A. Trip the inoperable channel in one trip system within 12 hours. B. Trip the inoperable channel in one trip system within 24 hours OR be in at least STARTUP within 6 hours. C. Trip the inoperable channel in one trip system within one hour AND trip one channel in the other trip system within 24 hours. D. Commence a normal shutdown within one hour AND be in at least STARTUP within 6 hours, HOT SHUTDOWN within 6 hours, and COLD SHUTDOWN within the following 24 hours.
Question 93 K&A Rating: 216000 G 2.2.22 (4.0) K&A Statement:
Knowledge of limiting conditions for operations and safety limits Justification:
A. Incorrect but plausible:
This more limiting action is for instruments common to RPS. B. Correct:
Actions for number of OPERABLE channels less than required for ONE trip system. N076A is NOT common to NSSSS and RPS. Tech Specs 3.3.1 and 3.3.2.b.1.b C. Incorrect but plausible:
This action is for OPERABLE channels less than required for BOTH trip systems.
D. Incorrect but plausible:
These actions are lAW 3.0.3 when there is no applicable condition.
This is not the case. SRO Only Justification:
This question is SRO only as it requires assessing operability and technical specifications which is an SRO function.
==References:==
Tech Spec sections 3.3.1 through 3.3.4; P&ID M-42-1 Sheet 2 Student Ref: Tech Spec sections 3.3.1 through 3.3.4; P&ID M-42-1 Sheet 2 Learning Objective:
NSSSSOE009 Question Source: HC Bank #35703 Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.5/43.5 Comments:
QUESTION 94 Given the following conditions:
* The plant is in Operational Condition 4
* The Reactor Head detensioning machine is being lowered into position to detension the reactor head lAW HC.OP-IO.ZZ-0005, Cold Shutdown to Refueling, who may authorize detensioning the first RPV Head Stud? A. Control Room Supervisor B. Reactor Engineer C Refueling Floor SRO D. Refueling Outage Manager K&A Rating: 2.1 .41 ( 3. 7) K&A Statement:
Knowledge of the refueling process Justification:
A. Correct:
lAW HC.OP-IO.ZZ-0005, Step 5.4.22, this direction comes from the SM/CRS B. Incorrect but plausible:
Non-licensed personnel are not authorized to direct mode changes C. Incorrect but plausible:
Even though the refueling floor SRO is a licensed
: operator, this direction must come from the licensed SRO in the MCR D. Incorrect but plausible:
The refueling outage manager may recommend mode changes, but direction for making mode changes comes from the licensed SRO in the MCR SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
HC.OP-IO.ZZ-0005, Rev. 35 Student Ref: NONE Learning Objective:
Question Source: Question History:
Cognitive Level: 1 OCFR55: Comments:
HC Bank #35519 HC NRC Exam 2003 Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
CFR 41.2/41.10/43.6/45.13 QUESTION 95 It is 0200 during normal full power operation.
The CRS needs to leave the site due to a personal emergency.
* At 0205 the CRS departs as directed by the Shift Manager (SM).
* At 0210 the SM calls the Operations Manager to inform him of the reduction in crew composition.
* At 0220 the SM reaches a relief watch stander for the CRS and directs the reflief to come to work.
* At 0415 the CRS relief arrives and completes a turnover with the SM. lAW Technical Specification 6.2 and OP-AA-1 01-111, 'Roles And Responsibilities Of On-Shift Personnel',
which ONE of the following is correct?
A. The operating crew has complied fully with shift manning requirements.
B. The CRS position should have been manned by a relief by 0405. C. The CRS should not have left until the Operation Manager gave him permission.
D. The CRS should not have left until his relief had arrived and turnover was completed.
K&A Rating: K&A Statement:
Justification:
G2.1.4 (3 3/3.8) Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 1 OCFR55, etc. A. Incorrect but plausible:
Plausible if the applicant recalls that a reduction in shift crew composition is allowable per Tech Specs as long as immediate action is taken to restore compliance, but does not recall that minimum shift crew composition must be restored within two hours. B. Correct:
Correct response.
Tech Spec 6.2.2 allows shift crew composition to be less than the minimum requirements Table 6.2.2-1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore shift crew composition.
C. Incorrect but plausible:
The Operation Manager needs to only be notified.
Operations manager permission does not need to be obtained.
D. Incorrect but plausible:
Tech Specs allow for a two hour window to restore minimum shift crew composition.
SRO Only Justification:
This question is SRO only as it requires knowledge of facility operating limitations in the Technical Specifications and their bases (minimum shift crew staffing requirements).
==References:==
Tech Spec 6.2 Student Ref: NONE Learning Objective:
Question Source: Modified Question History:
Modified from 2005 VY LOI Exam Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.2 Comments:
QUESTION 96 Surveillance HC.OP-IS.BC-0004(0),
'DP202, D RESIDUAL HEAT REMOVAL PUMP SERVICE TEST', is scheduled to be performed seven days from now. The system engineer wants to perform a Temporary Change related to minimum flow valve, F0007D. This change will involve cycling HV-F007D, D RHR PUMP MIN FLOW ISLN VLVwith HV-F01 OB D RHR Pump Test Return Vlv OPEN, and taking readings on discharge pressure and flow. lAW AD-AA-1 01, 'PROCESSING OF PROCEDURES AND T&RMS' ... A. the test may NOT be conducted as a Temporary Change. The test also may NOT be conducted as an On The Spot Change. B. the test may be conducted as a Temporary Change. Obtain an approval signature from a station qualified reviewer (SQR) and an SRO. C. the test may NOT be conducted as a Temporary Change. The procedure changes should be processed as an On The Spot Change. D. the test may be conducted as a Temporary Change. Perform a full review, approval and authorization within 14 days of implementing the Temporary Change. K&A Rating: G.2.2.5 (3.2) \
K&A Statement:
Knowledge of the process for making design or operating changes to the facility.
Justification:
A. Correct:
The proposed test represents a change of intent to the procedure.
lAW AD-AA-101 4.2.3, Temporary Changes are only allowed if they DO NOT change the intent of a procedure.
Since the proposed test changes plant configuration, it should be processed as a Interim Change per section 4.2.2 of AD-AA-101 B. Incorrect but plausible:
Plausible if the applicant does not recognize that the proposed test represents a change of intent to the procedure.
C. Incorrect but plausible:
Plausible if the applicant recognizes that the test should not be processed as a Temporary Change, but does not recognize that an Interim Change should be used instead of an On The Spot Change. D. Incorrect:
Plausible if the applicant does recognize that the proposed test represents a change of intent to the procedure.
SRO Only Justification:
This question is SRO only as it requires This question, from the Generic portion of the written exam, is SRO only as it requires application of 1 OCFR55.43(b
)(3), facility licensee procedures required to obtain authority for design and operating changes to the facility.
==References:==
AD-AA-1 01, "Processing of Procedures and T&RMs. Student Ref: NONE Learning Objective:
NA Question Source: NEW Question History:
NA Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
1 OCFR: CFR 41.10/43.5/45.13 Comments:
QUESTION 97 The plant is operating at rated conditions when the N01 OA, N011 A, N012A, and N013A Steam Tunnel Temperature switches fail low. These four switches input temperature to Main Steam Line 'A'. Which ONE of the following describes the required Technical Specification actions?
A Place the inoperable channel(s) in the tripped condition within 1 hr ONLY. B. Place the inoperable channel(s) in the tripped condition within 24hrs ONLY. C. Place the inoperable channel(s) in one trip system in the tripped condition within one hour, AND place the inoperable channel(s) in the remaining trip system in the tripped condition within 1 hr.* D. Place the inoperable channe!(s) in one trip system in the tripped cbndition within one hour, AND place the inoperable channel(s) in the remaining trip system in the tripped condition within 24hrs.
K&A Rating: K&A Statement:
Justification:
2.2.37 (3.6/4.6)
Ability to determine operability and/or availability of safety related equipment.
A. Incorrect but plausible:
Plausible if the applicant misinterprets the TSs and believes they are in TS 3.3.2.b.1.a.
B. Incorrect but plausible:
Plausible if the applicant misinterprets the TSs and believes they are in TS 3.3.2.b.1.c.
C. Correct:
All 4 steam tunnel temperature switches on the 'A' Main Steam Line are inoperable as given in the stem (N01 OA, 11 A, 12A & 13A). TS Table 3.3.2-1 (3.f) states that a minimum of 2 operable channels per trip system are required in OPCONS 1,2 and 3. The inoperable temp switches result in 0 Operable channels on the 'A' Steam Line for BOTH the A and B trip systems.
Therefore, TS 3.3.2 action c.1 and 2.1 are applicable.
D. Incorrect but plausible:
Plausible if the applicant misinterprets the TSs and believes they are in TS 3.3.2.c.2.3.
SRO Only Justification:
This question is SRO only as it involves application of required TS actions, cannot be answered by solely knowing <= 1 hr TS actions, cannot be answered by solely knowing above the line LCO information, and cannot be answered solely by knowing TS safety limits. The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
TS 3.3.2 Student Ref: TS Learning Objective:
Question Source: New Question History:
N/A Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.7/43.5/45.12 Comments:
QUESTION 98 Given the following conditions:
* You are filling out a dose extension lAW RP-AA-203 for an operator to enter the Drywell
* The operator's current annual dose is 3500 mrem TEDE
* The operator is expected to receive 400-450 mrem on this job. lAW RP-AA-203, Exposure Control and Authorization, which of the following personnel must approve this extension?
: 1. Work Group Supervisor
: 2. RP Manager 3. Station/Plant Manager 4. Site Vice President A 1 and 3 only. B. 2 and 4 only. C. 1, 2, and 3 only. D. 1, 2, 3, and 4.
K&A Rating: 2.3.4 (3.2/3.7)
K&A Statement:
Justification:
2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.
A. Incorrect but plausible:
See response C. B. Incorrect but plausible:
See response C. C. Correct:
lAW RP-AA-203, to raise the ADCL to 3000 mrem, 1 & 2 are required.
To raise the ADCL to 4000mrem, 1, 2, and 3 are required.
To raise the ADCL above 4000mrem, 1, 2, 3 and 4 are required.
The stem indicates that a maximum ADCL of 3950 will be required for the job (3500mrem
+ 450mrem),
therefore 1, 2, and 3 are required.
D. Incorrect but plausible:
See response C. SRO Only Justification:
This question is SRO only as it requires knowledge of radiation hazards that may arise including maintenance activities and various contamination conditions.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
RP-AA-203 Student Ref: NONE Learning Objective:
NOH04ADM024E-004 Question Source: Bank Question History:
HC Bank# 213 Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.12/43.4/45.10 Comments:
QUESTION 99 During a fire in the Turbine Building, the Shift Manager is required to assign a Fire Department Liaison.
lAW OP-HC-112-1 01-1001 "Shift Turnover Responsibilities",
: 1) who, by title, can be assigned the Liaison role, and 2) what is their duty? A. 1) ST A AND the Auxiliary Building EO; 2) Communicates to the MCR on status of the event. B. 1) WCC Supervisor AND the Radwaste EO; 2) Communicates to the MCR on status of the event. C. 1) ST A AND the CRS; 2) Advises the Shift Manager on what equipment needs to be removed from service.
D. 1) WCC Supervisor AND the CRS; 2) Advises the Shift Manager on what equipment needs to be removed from service.
Question 99 K&A Rating: G.2.4.27 (3.4/3.9)
K&A Statement:
Knowledge of "fire in the plant" procedures.
Justification:
A. Incorrect but plausible:
Cannot fill dual function of STA and Fire Liason, see OP-AA-1 01-111 Attachment 1 Hope Creek Shift Complement note 3 B. Correct:
OP-AA-101-111 Attachment 1 Hope Creek Shift Complement note 3 states: the SM should designate an appropriate EO, RWEO, RO or SRO (cannot be concurrently assigned to fill the position of SM, CRS, STA, NC0(2), CM1, CM2 or OSCC in Attachment 2 table) to function as the station fire brigade liaison.
They should also function as liaison for other emergencies (ex. chemical spill, toxic gas, environmental, etc.). The purpose of the liaison is to provide real time communication to the control room regarding the status of the event. A. Incorrect but plausible:
Cannot fill dual function of STA and Fire Liason see OP-AA-1 01-111 Attachment 1 Hope Creek Shift Complement note 3 C. Incorrect but plausible:
Cannot fill dual function of CRS and Fire Liason, see OP-AA-1 01-111 Attachment 1 Hope Creek Shift Complement note 3 SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
OP-HC-112-101-1001 Learning Objective:
ADMPR05CE001 Question Source: Bank Question History:
HC Bank #221 Cognitive Level: Memory/Fundamental Knowledge:
X Comprehensive/Analysis:
1 OCFR55: CFR 41.3/43.5/45/13 Comments:
Student Ref: None QUESTION 100 The plant is at 100% power. Which ONE of the following situations constitutes sufficient information to declare the associated SSC INOPERABLE?
A. Receipt of OHA B1-C4 HPCI TURBINE TROUBLE concurrent with CRIDS point 05430 HPCI VACUUM TANK LEVEL HI; and indication the Vacuum tank Condensate Pump is NOT running.
B. Receipt of OHA E4-A2 DIESEL GEN PNL A/B/C/D C422 concurrent with CRIDS points D4581 DIESEL GENERATOR A RCP 1AC422 TRBL AND D3776 DG A REGULAR LOCKOUT RELAY TRIPPED; and the Aux Bid EO reports the REGULAR LOCKOUT RELAY for A EDG is tripped.
C. Receipt of OHA C6-D4 CRD ACCUM TROUBLE concurrent with CRIDS point D5268 CRD ACCUM AND an amber ACCUM light for 02-19 on the Full Core Display; and the RB EO reports accumulator pressure for 02-19 is 1150psig.
D. Receipt of OHA C1-E1 SLC TANK TROUBLE concurrent with CRIDS point D2382 SLCS TANK LEVEL HILO and an indicated SLC tank level of 4640 gallons.
K&A Rating: 2.4.45 (4.3) K&A Statement:
Ability to prioritize and interpret the significance of each annunciator or alarm Justification:
A. Incorrect but plausible:
Pump down of the HPCI vacuum tank is NOT required for HPCI to perform its safety function.
Although the alarm condition is valid, it does NOT inherently render HPCI inoperable B. Correct:
A trip of the Regular Lockout Relay will prevent the EDG from starting under ANY conditions.
lAW HC.OP-AR.ZZ-0017, Lockout Relays Activated is one of the inputs into E4-A2. lAW HC.OP-SO.KJ-0001, an EDG trips upon receipt of an 86R lockout.
This will prevent it from fulfilling its safety function.
The CRIDS points identify the 'A' EDG as the affected EDG C. Incorrect but plausible:
lAW HC.OP-AR.ZZ-0011, this alarm may be due to EITHER high water level OR low gas pressure.
Only low gas pressure would be indicative of an inoperable condition.
Additionally, the allowable alarm setpoint could be above the pressure that requires declaring the HCU inoperable.
Local investigation is required to provide sufficient information erode confidence in the assumption of operability.
(Note that under different conditions, such as a current gas leakage problem with the associated HCU and knowledge that it alarms very close to the point at which it becomes inoperable, the assessment might be different)
D. Incorrect but plausible:
SLC operability is a combination of tank level and solution concentration.
The tank low level alarm is within the allowable operating region for the SLC tank lAW T/S 3.1 .5. Comparison of the SLC tank level with the solution concentration is required to assess operability SRO Only Justification:
This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.
The question cannot be answered by solely knowing systems knowledge, immediate operator
: actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.
==References:==
HC.OP-AR.ZZ-01 07, Rev. 6 HC.OP-AR.ZZ-0011, Rev. 58 HC.OP-AR.ZZ-0008, Rev. 43 TS 3.5.1 Learning Objective:
Question Source: HC Bank 654 79 Student Ref: NONE Question History:
None Cognitive Level: Memory/Fundamental Knowledge:
Comprehensive/Analysis:
X 10CFR55:
CFR 41.10/43.5/45.3/45.12 Comments:}}

Revision as of 02:36, 4 July 2018

Hope Creek - Final Written Examination with Answer Key (401-5 Format) (Folder 3)
ML13267A389
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 08/27/2013
From: Fish T H
Operations Branch I
To:
Public Service Enterprise Group
Jackson D E
Shared Package
ML13077A118 List:
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TAC U01872
Download: ML13267A389 (207)
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Text

QUESTION 1 Given the following conditions:

  • The plant is operating at 1 00%
  • A transient caused by a short in the reactor recirculation control circuitry occurs Immediately following the transient, the plant stabilizes with the following parameters:
  • ReactorPower50%
  • "A" Recirc pump Drive Motor breaker tripped * "B" Recirc pump flow is _22 Kgpm
  • Loop "A" jet pump flow is 5 Mlbmlhr
  • Loop "B" jet pump flow is 36 Mlbmlhr
  • Total indicated core flow 31 Mlbm/hr What is actual core flow? A 26 Mlbm/hr.

B. 31 Mlbm/hr.

C. 36 Mlbm/hr.

D. 41 Mlbm/hr.

K&A Statement:

Justification:

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:

AA2.03 Actual Core Flow A. Incorrect but plausible:

Jet pump flows must be added. B. Incorrect but plausible:

Jet pump flows must be added. C. Incorrect but plausible:

Jet pump flows must be added. D. Correct:

Below 48% running recirc loop speed, Loop flow will be less than 23 Kgpm. Jet pump loop flows are both positive and added together.

Setpoints must be adjusted to single loop values within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

References:

HC. OP-AB. RPV. 0003, HC. OP-ST. BB-0007, HC.OP-DL.ZZ-0026 Attach 3V TS 2.2.1 and 3.4.1 Student Ref: NONE Learning Objective:

L!RIB/ Core flow determination following RR pump trip Question Source: HC Bank #137 Question History:

NRC 2002 Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 2 Given the following conditions:

  • The plant is in Operational Condition 5 with the Electrical Distribution System aligned in the Normal lineup.
  • An internal short on Transformer 1 BX-501 causes a sudden pressure fault on the transformer.

Which ONE of the following describes the resulting availability of Power for the Safe Shutdown Systems?

A. Power to both 4.16KV switchgear 1 OA402 and 1 OA404 fast transfers to Transformer 1AX501. 13 KV breaker BS 1-2 trips open. B and D diesel generators START but their output breakers DO NOT CLOSE. B. Power to both 4.16KV switchgear 1 OA402 and 1 OA404 fast transfers to Transformer 1AX501. 13 KV Breaker BS 1-2 trips open. Band D diesel generators DO NOT START. C. Power is lost permanently to both 4.16KV switchgear 1 OA401 and 1 OA403. 13 KV breaker BS 1-2 stays closed. B and D Diesel Generators start but their output breakers DO NOT CLOSE. D. Power is lost momentarily to both 4.16KV switchgear 1 OA402 and 1 OA404. 13 KV breaker BS 1-2 trips open. Power is restored when the B and D Diesel generators output breakers close.

Question 2 K&A Rating: 295003 Partial or Complete Loss of AC Power AA2.05 (3.9) K&A Statement:

Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of AC Power: Whether a partial or complete loss of A. C. Power has occurred.

Justification:

A. Incorrect but plausible:

The B & D Diesel Generators DO NOT START B. Correct:

13 Kv Breakers BS 2-3 and BS 1-2 trip open. Bus section 2 is energized, Bus section 1 remains energized.

The bus infeed breaker swap to the AX501 feed. The loads remain energized.

Because one infeed is always available, the Diesels do NOT start. C. Incorrect but plausible:

Power is NOT permanently lost to both 4.16KV switchgears.

Power is restored when the bus infeed breaker swaps to the AX501 feed. D. Incorrect but plausible:

Power is NOT restored from the B & D Diesel Generators

References:

Drawing E-0001 and 066-01: Class 1 E AC Power Distribution NOH01 EAC00 CLASS 1 E AC POWER DISTRIBUTION, page 32 of 93 Student Ref: NONE Learning Objective:

Question Source: Hope Creek Question Q76871 -Modified Question History:

2005 HC Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 3 A loss of the Channel "C" Class 1 E 125 VDC System will result in the loss of .... A. the ability to control RCIC components.

B. Reactor Feedwater Pump speed control.

C. "C" 4.16 KV Class 1 E electrical breaker control.

D. automatic or remote manual operation of HPCI MOVs.

K&A Rating: 295004AK2.03 (3.3/3.3)

K&A Statement:

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following:

AK2.03 D.C. Bus Loads Justification:

A. Incorrect but plausible:

is powered from Channel B 125 VDC. B. Incorrect but plausible:

RFP controls receive control power from non-1 E 125 VDC. C. Correct:

'C' 1 E 125 VDC supplies 4KV breaker control power. Without control power, remote breaker operation and automatic breaker trips are disabled.

D. Incorrect but plausible:

HPCI MOVs receive control power from "A" bus of 125 VDC.

References:

N/A Learning Objective:

OAB 150E003 Question Source: HC Bank Q146 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

10CFR: CFR 41.7/45.8 Comments:

Student Ref: NONE QUESTION 4 Following a main generator load rejection at full power, a reactor recirculation pump trip is initiated in anticipation of a ... A. Reactor Pressure Increase B. Reactor Pressure Decrease C. Reactor Water Level Increase D. Reactor Water Level Decrease K&A Rating: K&A Statement:

Justification:

295005AK1.01 (4.0/4.1)

Knowledge of the operational implications of the following concepts as the apply to MAIN TURBINE GENERATOR TRIP: AK1.01 Pressure effects on reactor power A. Correct:

When a main generator load rejection signal is received the turbine is tripped.

and a SCRAM signal will be generated.

The SCRAM is initiated to reduce reactor power in anticipation of the reactor pressure increase and subsequent power increase due to void collapse.

Tripping of a recirculation pump will also serve to reduce reactor power and minimize reactor pressure increase.

B. Incorrect but plausible:

reactor pressure will increase, not decrease due to the turbine trip and control valve closure.

C. Incorrect but plausible:

reactor level will decrease not increase due to the rising reactor pressure from the turbine trip and control valve closure.

D. Incorrect but plausible:

reactor level will decrease due to the turbine trip and closure of the control valves, however the SCRAM and recirc pump trip is in anticipation of the pressure increase which will be more limiting than the level drop.

References:

Main Turbine Construction and Components Lesson Plan Learning Objective:

R4 Question Source: Bank Question History:

Taken from HC Bank-Question#

214 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 5 Due to a spurious MSIV isolation, the Unit has tripped from 100% power. Given the following conditions:

  • HPCI is running in pressure control mode
  • RCIC is running in level control
  • Torus temperature is 135°F and rising at about 1 degree/minute Caution #4 of HC.OP-EO.ZZ-0101, RPV Control, warns against operating HPCI and RCIC with high suction temperatures due to ... A tripping the turbine(s) due to low suction pressure B. cavitation of the jockey fill pump C. inadequate condensation in the barometric condenser D. inadequate cooling to the lube oil cooler K&A Rating: 295006 G2.4.20 (3.8/4.3)

K&A Statement:

Knowledge of the operational implications of the EOP warnings,

cautions, and notes Justification:

A. Incorrect but plausible:

if applicant believes that the purpose of the caution is to prevent elevated HPCI!RCIC suction temperatures causing cavitation of the HPCI!RCIC pumps and the accompanying pressure surges could lead to a low suction pressure trip B. Incorrect but plausible:

if applicant believes that the purpose of the caution is to prevent elevated torus temperatures from causing cavitation of the jockey fill pump C. Incorrect but plausible:

if the applicant believes that the purpose of the caution is to alert operators to the potential of inadequate condensation in the barometric condenser due to lower heat transfer caused by the elevated torus temperatures D. Correct:

HC.OP-EO.ZZ-0101, caution #4 states, "Operation of HPCI OR RCIC turbines with suction temperatures above 170°F may result in equipment damage".

lAW with the bases document, it states, "Caution

  1. 4 warns against operating HPCI and RCIC with high suction temperatures due to higher than normal oil temperatures".

This is caused due to the lube oil coolers being cooled by the suction source, resulting in less cooling to the lube oil.

References:

HC.OP-EO.ZZ-01 01, Rev. 11 Student Ref: NONE Learning Objective:

E01 01 PE006 Question Source: HC bank 109059 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

1 OCFR55: CFR 41.10/43.5/45.13 Comments:

QUESTION 6 Given the following conditions:

  • Fire and smoke in the control room has led to a control room evacuation.
  • The Control Room is being evacuated per HC.OP-AB.HVAC-0002.
  • HPCI and RCIC both automatically initiated and are injecting.
  • The RSP operator trips RCIC when reactor water level reaches +54 inches.
  • No field actions of AB.HVAC-0002 have been completed.

Reactor water level will ... A. lower until RCIC automatically re-initiates at -38 inches. B. lower until HPCI automatically re-initiates at -38 inches. C. lower until both HPCI and RCIC automatically reinitiate.

D. continue to rise due to HPCI injection.

K&A Rating: 295016 AK2.02 (4.0/4.1)

K&A Statement:

Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:

Local control stations:

Plant-Specific Justification:

A. Incorrect but plausible:

Plausible if the applicant does not recall that RCIC auto start is disabled at RSP when the transfer switches are placed in the EMER. B. Correct:

HPCI will continue to cycle between -38 inches and +54 inches. C. Incorrect but plausible:

Plausible if the applicant does not recall that RCIC auto start is disabled at RSP when the transfer switches are placed in the EMER. D. Incorrect but plausible:

Plausible if the applicant does not recall that HPCI auto trip at +54 inches is still in effect.

References:

HC.OP-IO.ZZ-0008(Q),

Rev. 31 HC.OP-AB.HVAC-0002(Q),

Rev. 7 Learning Objective:

RCICOOE004 Question Source: New Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Student Ref: NONE Comprehensive/Analysis:

X 10CFR: CFR 41.7/45.8 Comments:

QUESTION 7 Given the following conditions:

  • The plant is operating at 100% rated thermal power
  • The RACS ROOM FLOODED alarm is received on all channels
  • All expected automatic actions occur
  • The EO reports 2" of water covering the entire RACS Pump and HX room floor, with the rate of rise slowing Which ONE of the following actions is required lAW HC.OP-AB.COOL-003

'Reactor Auxiliary Cooling'?

A. Begin a normal reactor shutdown at 15% per hour. B. Manually close valve 1-EA-V453 SSW to RACS Outboard Isolation valve. C. Remove the floor drain plugs and pump out the room via the SW Dewatering pump. D. Reduce Reactor Recirc Pump Speed to minimum and lock the mode switch in SHUTDOWN.

K&A Rating: K&A Statement:

Justification:

2950 18AK2. 02 (3.4) Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER and the following:

AK2.02 Plant Operations A. Incorrect but plausible:

A total loss of RAGS has occurred.

The amount of water on the floor prevents recovery of SSW to RAGS. Action for flooding in rooms from EOP-103 Table 2. B. Incorrect but plausible:

Manual closing of 1-EA-V453 directed if HV-2346 fails to close. Not true with given stem conditions.

C. Incorrect but plausible:

Floor drains and SSW dewatering tank are not interconnected.

D. Correct:

Since all automatic actions occurred, SSW to RAGS is lost and a loss of RAGS has occurred.

Retainment override of AB-COOL-0003 reduce RRPs to minimum and lock Mode Switch in S/D.

References:

AB-COOL-0003 Student Ref: NONE Learning Objective:

ABCOL3E001 Question Source: Bank Question History:

HC Bank #308 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 8 Given the following conditions:

Hope Creek is starting up from a Refueling outage. The plant is currently in Operational Condition 3 with RCS temperature at 240°F. Instrument Air pressure is 105 psig. The Instrument/Service Air Compressors are aligned as follows:

Compressor Control Mode OOK107 LEAD 10K107 LAG 10K100 AUTO Subsequently, 7.2KV Bus 1 OA 120 loses power. Status Running OFF OFF Assuming NO operator

actions, which ONE of the following describes the response of the Instrument/Service Air systems?

A Instrument Air header pressure remains at 105 psig. B. Instrument Air header pressure drops to 92 psig, when Service Air Compressor 1 OK1 07 starts and returns pressure to -105 psig. C. Service Air receiver pressure drops to 92 psig, when Service Air Compressor 1 OK1 07 starts and returns pressure to -105 psig. D. Service Air receiver pressure drops to 85 psig when Emergency Air Compressor 1 OK1 00 starts and returns pressure to -95 psig.

Question 8 K&A Rating: 295019 Partial or Complete Loss of lnst. Air AA 1.03 (3.0) K&A Statement:

Ability to operate and/or monitor the following as they apply to Partial or Complete Loss of lnst. Air: Instrument Air Compressor power supplies.

Justification:

A. Incorrect but plausible:

Power to SAC 10K107 is from 7.2 KV bus 10A110, NOT 10A120 B. Incorrect but plausible:

SAC 1 OK1 07 will NOT start at 92 psig in the instrument air header. It starts when the service air receiver is 92 psig. I. C. Correct:

SAC 1 OK1 07 will start when the service air receiver pressure is 92 psig .. D. Incorrect but plausible:

Loss of Power to 1 OA 120 causes a loss of Power to SAC OOK107, Instrument Air header will not drop to 85 psig, because 10K107 will prevent the pressure getting that low. Plausible because EIAC 1 OK1 00 would start at 85 psig.

References:

NOH01SERAIR-01, SERVICE AIR SYSTEM, p.47-48 NOH011NSAIR-01, INSTRUMENT AIR SYSTEM, p15, 42 Student Ref: NONE Learning Objective:

Question Source: New Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR41.7/45.6 Comments:

QUESTION 9 Given the following conditions:

  • The reactor has been shutdown for 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> following 1000 EFPD of operation.
  • The plant is in Cold Shutdown with RPV metal and RCS temperature of 140°F.
  • A total loss of Shutdown Cooling occurred at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />.
  • All efforts to restore heat removal from the RPV have failed.
  • Both Recirculation pumps have been secured.

Assuming NO additional operator action, when will the plant reach OPCON 3? A. 1245 B. 1307 C. 1330 D. 1352 (Note: See attached Figure 1) 0 ... ..... LL. OJ a>-c .VI ... 0 :::l 0 0 N.c 0 +' Q) E 1-FIGURE 1 Total Loss of Heat Removal from Rx Vessel 1000 EFPD of Operations 5.000

-'!"'_,_..,...,...,..

___

I 100"F IY 4.000 ...,.. __ ...,__.....,.._,_""'1

_,.-!-.,.11 IL r'2o*F / v 3.ooo ..,.. __

... I L ,4"FI r-I 0.000 ...._ __ ....... _ .............

1_...,1 .... 1....._1

.....,. .... 1 ___ .....__ ... 1_.._ ... ,_.. ..............

10 100 Time from Rx Shutdown (hours) ** Represents initial starting temperature as listed in the Figure. Each line represents a different initial starting temperature 1000 Question 9 K&A Rating: 295021 Loss of Shutdown Cooling AA2.01 (3.5) K&A Statement:

Ability to determine and/or interpret the following as they apply to Loss of Shutdown Cooling:

Reactor water heatup/cooldown rate. Justification:

A. Incorrect but plausible:

Value obtained by using the 160°F curve. B. Correct::

Operational Condition 3 is achieved when the Reactor temperature reaches 200°F. The 140°F curve of Figure 1 intersects the 90-hour line between the 1.000 and 1.250 hour0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> lines. 1307 is the only option that is between 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and fifteen minutes following the loss of SOC. C. Incorrect but plausible:

Value obtained by using the 120°F curve. D. Incorrect but plausible:

Value obtained by using the 1 00°F curve.

References:

HC.OP-AB.RPV-0009, Figure 1and Technical Specification Table 1.2 Student Ref: Figure 1 of HC.OP-AB.RPV-0009 Learning Objective:

Question Source: HC.OP-AB.RPV-0009, Figure 1and Technical Specification Table 1.2 Question History:

2005 HC NRC Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 10 The plant is in OPCON 1 with irradiated fuel being shuffled in the Spent Fuel Pool when the following VALID alarms are received in the Control Room:

  • RADIATION MONITORING ALARM/TRBL
  • NEW FUEL CRITICALITY RAD HI
  • RB EXH RADIATION ALARM/TRBL Which ONE of the following is the appropriate action lAW HC.OP-AB.CONT-0005, IRRADIATED FUEL DAMAGE'?

A. Verify Reactor Building Ventilation isolates and fans trip. B. Verify the Drywell Integrity Airlock surveillance test is current.

C. Verify "A" and "C" SACS pumps start if not already running.

D. Verify Emergency Instrument Air Compressor starts if not already running.

K&A Rating: K&A Statement:

Justification:

295023AK3.01 (3.3/3.6)

Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS:

AK3.03 Ventilation Isolation A. Correct:

Hi alarms on Reactor building and RF floor indicate an isolation signal and supply and exhaust fans trip. B. Incorrect but plausible:

The concern is secondary not primary containment.

C. Incorrect but plausible:

The "A" and "B" SACS pump will start if not running, not the "A" and "C" SACS pumps. D. Incorrect but plausible:

The EAIC would trip on the High Radiation Signal.

References:

HC.OP-AB.CONT-0005, Rev 4 Student Ref: NONE Learning Objective:

ABCNT5E004 Question Source: Modified from HC Bank (Question

  1. 405) Question History:

Modified from bank question used on 2010 HC Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 11 A transient has occurred.

The current containment conditions are as follows:

  • Drywell pressure:

4.5 psig rising

  • Drywell temperature:

150 F rising

  • Torus pressure:

4.0 psig rising

  • Torus water temperature:

82 F steady

  • Torus airspace temperature:

145 F rising Which of the following events would explain the current plant conditions?

A. A safety relief valve (SRV) has lifted and is discharging through a T-Quencher.

B. The containment is functioning as designed following a feedwater line break inside containment.

C. A main steam line break has occurred inside containment with a Torus to drywell vacuum breaker open. D. The containment is functioning as designed following a total loss of drywell cooling.

QUESTION 11 K&A Rating: K&A Statement:

Justification:

295024 High Drywell Pressure EA2.05 (3.6) Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

EA2.05 Suppression chamber air-space temperature:

Plant Specific.

A Incorrect but plausible:

Since the Torus airspace temperature would not be rising with the discharge of the SRV into the water portion of the torus. B. Incorrect but plausible:

Since the Torus airspace temperature would not be rising if the containment was responding normally.

C. Correct:

Drywell and torus air space trending together is indication that there is a bypass of the containment, which in this case is the torus to drywell vacuum breaker being open. D. Incorrect but plausible:

Since the Torus airspace temperature would not be rising with the loss of drywell cooling.

References:

NOHOIPRICONC, "Primary Containment Structure" Student Ref: NONE Learning Objective:

PRICONE008 Question Source: Mod INPO Bank PB2 Question Question History:

2007 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 12 Which ONE of the following describes the reason for sustained (vice intermittent)

SRV openings during performance of HC.OP-EO.ZZ-01 01, 'RPV CONTROL' if the pneumatic supply is lost to the SRVs? Sustained SRV openings

... A. conserve SRV pneumatic supply for sufficient cooldown to less than the shutdown cooling high pressure interlocks.

B. prevent exceeding the 100°F/hr cooldown limit during depressurization while conserving the SRV pneumatic supply. C. conserve SRV accumulator pneumatic supply for later use if an Emergency Depressurization is required.

D. allow the operator to depressurize without regard to the Technical Specification cooldown limits before the pneumatic supply is depleted.

K&A Rating: 295025 2.1.28 (4. 1/4.1) K&A Statement:

High Reactor Pressure 2.1.28: Knowledge of the purpose and function of major system components and controls.

Justification:

A. Incorrect but plausible:

Sustained opening of the SRVs is specified in step RC/P-6 to conserve accumulator pressure for a subsequent Emergency Depressurization, NOT to allow for further cooldown.

B. Incorrect but plausible:

Although there is a 1 OOF/hr cooldown rate limit applied in step RC/P-6, sustained opening of the SRV vice intermittent opening is not used to control cooldown rate. C. Correct:

lAW HC.OP-EO.ZZ-0101 bases step RC/P-6, sustained SRV opening conserves accumulator pressure when the source of pressure to the SRV pneumatic supply is isolated or otherwise out of service.

Such action to reduce the number of cycles on the SRVs prolongs SRV availability should more degraded conditions later required SRVs be opened for rapid depressurization of the RPV. D. Incorrect but plausible:

HC.OP-EO.ZZ-01 01 step RC/P-6 limits cooldown rate to 1 OOF/hr.

References:

HC.OP-EO.ZZ-0101 Bases Document Learning Objective:

E01 01 PE005 Question Source: HC Bank #131 Question History:

Not used on previous 2 NRC exams Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

10CFR: CFR 41.7 Comments:

Student Ref: NONE QUESTION 13 The plant is at rated power with the following conditions:

  • HPCI testing is in progress lAW HC.OP-IS.BJ-0001 "HPCI Main and Booster Pump Set-OP204 and OP217 -lnservice Test"
  • Torus level is 75.5 inches
  • Div 1 average suppression pool temperature indicates 96°F and is slowly trending higher
  • Div 2 average suppression pool temperature indicates 98°F and is slowly trending higher WHICH ONE of the following describes the required action(s)?

A. Enter HC.OP-EO.ZZ-01 02 "Primary Containment Control" AND immediately suspend testing.

B. Place the Mode Switch in Shutdown and place Suppression Pool cooling in service.

C. Enter HC.OP-EO.ZZ-0102 "Primary Containment Control".

Suspend testing before average suppression pool temperature reaches 1 05°F. D. Do NOT enter HC.OP-EO.ZZ-01 02 "Primary Containment Control".

Suspend testing before average suppression pool temperature reaches 1 05°F.

K&A Rating: K&A Statement:

Justification:

295026 EA 1.03 (3.9/3.9)

Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

EA1.03 Temperature Monitoring A. Incorrect but plausible:

If applicant believes that EOP entry is required.

EOP entry is not required until 105°F while performing testing lAW HC.OP-IS.BJ-0001.

Additionally, testing is not required to be suspended until 1 05°F suppression chamber average water temperature lAW TS 3.6.2.1 B. Incorrect but plausible:

If applicant believes that the Mode Switch is required to be placed in Shutdown due to one temperature above 11 0°F. This is only required if the Suppression Pool average water temperature exceeds 11 ooF C. Incorrect but plausible:

If applicant believes that EOP entry is required.

EOP entry is not required until 1 05°F while performing testing lAW HC.OP-IS.BJ-0001.

D. Correct:

EOP entry is not required until 105°F while performing testing lAW IS. BJ-0001.

Additionally, testing is not required to be suspended until1 05°F suppression chamber average water temperature lAW TS 3.6.2.1

References:

HC.OP-IS.BJ-0001, Rev. 60 TS 3.6.2.1, Amendment 110 Learning Objective:

Question source: Modified Limerick 2008 Question History:

None Applicant Ref: NONE Cognitive level: Memory/Fundamental knowledge:

Comprehensive/Analysis:

X 1 OCFR Part 55: 41.7/45.6 Comments:

QUESTION 14 SELECT the reactor pressure, drywell temperature and level instrument that would indicate valid reactor water level of ( -1 00) inches. A. RPV pressure is 300 psig, Drywell temperature is 400 degrees F on SPDS point A2266, Narrow Range A. B. RPV pressure is 200 psig, Drywell temperature is 350 degrees F on SPDS point A2277, Wide Range B. C. RPV pressure is 100 psig, Drywell temperature is 350 degrees F on SPDS point A2281, Upset Range. D. RPV pressure is 100 psig, Drywell temperature is 250 degrees F on SPDS point A2280, Fuel Zone A. RPV Saturation Temperature 600 G:' 550 L. c: :I 0:: 500 i: Ill E 450 .s "' .s 400 lij QJ z 350 ::1 300 Ill a. E Ill 250 r-.....---_.,..,.,..

./ / ., /"' I ! ' I 200 0 100 200 300 400 500 600 700 800 900 1 000 11 00 RPV Pressure (psi g) -Results K&A Rating: 295028 EK1.01 (3.5/3.7)

K&A Statement:

Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE:

Reactor water level measurement Justification:

A. Incorrect but plausible:

RPV pressure-300 psig, Drywell temperature-400 degrees F on SPDS point A2266, Narrow Range A-Within the curve, but 100 inches is above the top of the narrow range band (0-60 inches).

B. Correct:

pressure-200 psig, Drywell temperature-350 degrees F on SPDS point A2277, Wide Range B-Within the curve and A2277 is the nearest SPDS point to Wide Range B. C. Incorrect but plausible:

RPV pressure-100 psig, Drywell temperature-350 degrees F on SPDS point A2281, Upset Range-Outside of the curve. D. Incorrect but plausible:

RPV pressure-100 psig, Drywell temperature-250 degrees F on SPDS point A2280, Fuel Zone A-Within the curve, outside of the range of the Fuel Zone ( -111 to -311 inches).

References:

EOP Caution 1 HC.OP-EO.ZZ-LIMITS-CONV, Rev. 5 Learning Objective:

E01 01 LE007 Question Source: Bank# 33436 Question History:

Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR41.8, 41.9, 41.10 Comments:

Student Ref: NONE QUESTION 15 Given the following:

  • A leak has developed in the Torus requiring emergency makeup
  • Torus level is 71 inches and lowering slowly
  • The Condensate transfer lines to Core Spray Suctions are NOT available Emergency Makeup to the Suppression Pool via RCIC is accomplished by which ONE of the following?

A. Running RCIC with the full flow test flowpath open. B. Running RCIC with the min flow discharge flowpath open. C. Overriding and opening both RCIC suction MOVs simultaneously.

D. Running the RCIC Jockey Pump with the Suppression Pool suction MOV open.

K&A Rating: 295030 2.4.11 (4.0) K&A Statement:

Low Suppression Pool Water Level 2.4.11: Knowledge of abnormal condition procedures.

Justification:

A. Incorrect but plausible:

Running RCIC with the full flow test flowpath open would only recirculate CST water. B. Correct:

Running RCIC with the min flow discharge flowpath open, lAW EOP-313 will provide emergency makeup water to the suppression pool. C. Incorrect but plausible:

Overriding and opening both RCIC suction MOVs simultaneously will not provide emergency makeup water to the suppression pool. There is a check valve in the suppression pool suction path that prevents gravity draining the CST into the suppression pool. D. Incorrect but plausible:

Running the RCIC jockey pump with the suppression pool suction MOV open will not provide water to the suppression pool. There is a check valve between the suppression pool and the jockey pump suction line, and the flow direction of the jockey pump is incorrect.

References:

EOP-313 Student Ref: NONE Learning Objective:

EOP300E004, E0101AE006 Question Source: HC Bank # 184 Question History:

Used on 1999 Audit Exam Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

1 OCFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 16 Given the following conditions:

  • The plant has experienced a LOCA with a loss of ALL injection.
  • RPV pressure is being controlled with SRVs.
  • RPV water level has lowered to -196 inches (Corrected Fuel Zone). What is the status of core cooling?

Adequate core cooling exists ... A. only if injection is established at this water level. B. only when the SRVs are closed. C. at this RPV water level. D. only if RPV water level is raised 10 inches.

K&A Rating: K&A Statement:

Justification:

295031 EA2.04 (4.6/4.8)

Ability to determine and/or interpret the following as they apply to Reactor Low Water Level EA2.04 Adequate Core Cooling A. Incorrect but plausible:

Injection at this level would reduce steam generation needed to assure adequate core cooling.

B. Incorrect but plausible:

Adequate steam generation is present whether SRVs are open or closed. C. Correct:

At levels above -200 inches with no injection there is sufficient steam flow to provide adequate core cooling.

D. Incorrect but plausible:

Because compensated level is above -200, adequate core cooling exists at this level.

References:

Student Ref: NONE Learning Objective:

E01 01 LE006 Question Source: INPO Bank Q101 HC Bank Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

1 OCFR: CFR 41.10/43.3/45.13 Comments:

QUESTION 17 Unit 1 was at 100% power when the following occurred:

  • A loss of the 1 OA 1 02 bus resulted in the loss of two RFPTs.
  • The remaining RFPT was unable to maintain RPV water level.
  • The reactor scrammed on RPV L VL 3.
  • RPV water level lowered to -110".
  • HPCI and RCIC initiated, and RPV water level then started to rise. Current conditions are:
  • Reactor power is 5% and rising 1% every two minutes due to RPV water injection.
  • RPV water level is -11 0" and rising at 15"/min.
  • RPV pressure is 920 psig and steady.
  • NO operator actions have been taken. Which ONE of the following describes the associated RRCS function status four minutes later? A. A feedwater runback has NOT initiated, RPT breakers are closed. B. ARI valves are open, SLC pumps did NOT initiate.

C. RPT breakers are tripped, SLC pumps have initiated.

D. A feedwater runback has initiated, ARI valves are closed.

K&A Rating: 295037EK2.02 (4.0/4.2)

K&A Statement:

Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following:

RRCS Justification:

A Incorrect but plausible:

The -38" RPT breaker trip signal is a seal in signal. 9 seconds after water level dropped below -38", the RPT breakers will trip B. Incorrect but plausible:

The -38" level to the SLC pump initiation circuit does NOT seal in. However, starting from -11 0", rising at 15" per minute, RPV water level will be -50" in four minutes.

Since RPV water level will be below -38" when the 230 second timer times out, SLC pumps will initiate C. Correct:

The -38" RPT breaker trip signal is a seal in signal. 9 seconds after water level went below -38", the RPT breakers will trip. The -38" level to the SLC pump initiation circuit does NOT seal in. However, starting from -11 0", rising at 15" per minute, RPV water level will be -50" in 240 seconds.

Since RPV water level will be below -38" when the 230 second timer times out, SLC pumps will initiate D. Incorrect but plausible:

A feedwater runback is initiated only by reactor pressure reaching 1071 psig. There is NO reason for reactor pressure to have reached 1071 psig during the lowering level transient.

Additionally, ARI valves opened immediately at -38" and do NOT automatically reset

References:

NOH04RRCS Student Ref: NONE Learning Objective:

RRCSOOE007 Question Source: HC Bank (Question ID#65237)

Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7/45.8 Comments:

QUESTION 18 HC.OP-EO.ZZ-0103/4, Reactor Building

& Rad Release Control, step RR-5, directs isolation of all primary systems discharging into areas outside Primary Containment or Reactor Building, except those systems required to assure adequate core cooling and/or shutdown the reactor.

In accordance with the EOP Bases document, HC.OP-EO.ZZ-1 03/4. Reactor Building

& Rad Release Control, these systems are specifically exempted from isolation, because ... A. additional radiological consequences from them are unlikely.

B. they are required to support alternate reactor depressurization methods.

C. isolation of a EOP support system requires an upgrade of the Emergency Classification.

D. systems operated for RPV control are given a higher priority than stopping a rad release.

Question 18 K&A Rating: 295038 High Off-site Release Rate EK3.02 (3.9) K&A Statement:

Knowledge of the reasons for the following responses as they apply to High Off-Site Release Rate: System Isolations Justification:

A. Incorrect but plausible:

NOT in accordance with bases document B. Incorrect but plausible:

NOT in accordance with bases document C. Incorrect but plausible:

NOT in accordance with bases document D. Correct:

Per EOP Bases document 103/104:

The objectives of RPV Control, Primary Containment

Control, and the EPG contingencies are given higher priority than the steps of the objectives of Radioactivity Release Control.

Systems that must be operated to perform other EPGs are therefore NOT isolated in this step.

References:

BWROG, EPGs/SAGs Appendix B, section 9 Radioactivity Release control HC.OP-EO.ZZ-103/4.

Reactor Building

& Rad Release Control Bases Document-

p. 13 & 14 Student Ref: NONE Learning Objective:

EOP1 03E006 Question Source: HC NRC Exam 2005 Question History:

HC NRC Exam 2005 Cognitive Level: 10CFR: Comments:

Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

CFR 41.10/43.5/45.13 QUESTION 19 Given:

  • The plant Is operating at 100% power
  • A loss of MCC OOB590 occurs.
  • Field operator confirms NO fire onslte. Based on the above, it will be necessary to ... A. secure the motor driven fire pump to prevent pumping down the fire water storage tank. B. manually start the electric fire pump due to a loss of the diesel driven fire pump battery chargers.

C. manually start the diesel fire pump due to a loss of the motor driven fire pump power supply. D. secure the diesel driven fire pump to prevent pumping down the fire water storage tank.

K&A Rating: 600000 AK3.04 (2.8/3.4)

K&A Statement:

Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE: Actions contained in the abnormal procedure for plant fire on site Justification:

A. Incorrect but plausible:

Plausible if the applicant does not recall that the motor driven fire pump has no power, it is not running.

B. Incorrect but plausible:

Plausible if the applicant does not recall that the battery chargers are powered from 1AJ483. C. Incorrect but plausible:

Plausible if the applicant does not recall that the diesel driven fire pump auto starts on a loss of power to the motor driven fire pump. D. Correct:

The diesel driven fire pump auto starts on a loss of power to the motor driven fire pump. Since there is no fire, the fire water storage tank is being pumped down.

References:

HC.OP-AR.QK-0002 Student Ref: NONE Learning Objective:

FIRPROE008 Question Source: HC Bank Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.7/45.8 Comments:

QUESTION 20 The plant was operating at full power when a grid disturbance results in the actuation of a Main Generator Differential Overcurrent Lockout condition.

The Main Turbine trips and the Alterex exciter field breaker trips. Which ONE of the following describes the plant response on the below listed equipment?

A. lsophase Bus Duct Cooling Fans are unaffected.

Stator Water Cooling Pumps TRIP. Generator output breakers BS6-5 & BS2-6 remain closed. B. lsophase Bus Duct Cooling Fans are unaffected.

Stator Water Cooling Pumps TRIP. Generator output breakers BS6-5 & BS2-6 TRIP. C. lsophase Bus Duct Cooling Fans TRIP. Stator Water Cooling Pumps are unaffected.

Generator output breakers BS6-5 & BS2-6 TRIP. D. lsophase Bus Duct Cooling Fans TRIP. Stator Water Cooling Pumps are unaffected.

Generator output breakers BS6-5 & BS2-6 remain closed.

K&A Rating: K&A Statement:

Justification:

700000 AK2.02 (3.1/3.3)

Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following:

AK2.02 Breakers, Relays A. Incorrect but plausible:

Generator output breakers BS6-5 & BS2-6 TRIP. B. Correct:

The plant was operating at full power when a grid disturbance results in the actuation of a Main generator differential overcurrent lockout condition.

The Main generator trips and the Alterex field breaker trips. The following automatic actions are initiated when the Main Generator regular lockout relay is actuated due to a generator differential overcurrent condition:

Trips and blocks closing of gen output breakers BS6-5 & BS2-6. C. Incorrect but plausible:

lsophase bus duct cooling fans are unaffected, stator water cooling pumps trip. D. Incorrect but plausible:

lsophase Bus Duct cooling fans are unaffected, Stator Water Cooling Pumps TRIP, generator output breakers BS6-5 & BS2-6 TRIP.

References:

HC.OP-AR.ZZ-0015 Student Ref: NONE Learning Objective:

MNGENOE011 Question Source: HC Bank ID 120299 Question History:

Used on 2010 NRC Audit Exam Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

10CFR55:

CFR 41.4/41.5/41.7/41.10/45.8 Comments:

QUESTION 21 Given the following:

  • The reactor has scrammed on high Drywell pressure.
  • Drywell Floor Drain Sump pumps (CP267/DP267) have stopped running.
  • Drywell pressure continues to rise.
  • No operator actions have been taken. Under these conditions, which ONE of the following caused the Drywell Floor Drain Sump pumps to stop running?

A. The Reactor Recirc Seal Staging flow is isolated.

B. The HB-HV-5258 DRYWELL FLOOR DRN PMPS DSCH VLV has failed closed. C. The Drywell Leak Detection (DLD) Sump Monitoring System has failed. D. The power source to the pumps is load shed.

K&A Rating: K&A Statement:

Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE:

AA1.02 Drywell floor and equipment drain sumps Justification:

A. Incorrect but plausible:

Reactor Recirc Seal Staging flow is directed to the Drywell Equipment Drain Sump. Loss of this flow input would have no effect on the DWFDS. B. Incorrect but plausible:

The HV-5258 is interlocked to open on a start of either CP267 or DP267; however, it is NOT a permissive for the pumps to run. If it failed shut it would prevent the pumps from pumping, but it would NOT prevent them from running.

(The pumps DO receive a permissive to run from the drywell containment isolation valves F003/F004).

C. Incorrect but plausible:

The OLD SMS does not control operation of the drywell sump pumps. D. Correct:

The power supply to the CP267 and DP267 are 252064 and 262064, respectively.

These MCCs are load shed on High Drywell Pressure.

References:

E-0023-1 Sheet 1 Student Ref: NONE HC.OP-SO.SM-0001 Table SM-20 HC.OP-AR.ZZ-0014 Attachment D3-C3 and 02363. Learning Objective:

L/R/A/ Drywell Floor Drain Sump pumps trip Question Source: HC Bank #58 Question History:

NRC EXAM 2002 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.7/45.6 Comments:

QUESTION 22 Given the following conditions:

  • A plant startup is in progress following a forced outage.
  • The plant has been operating with a known fuel leak.
  • The plant scrammed 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> ago.
  • The "A" Mechanical Vacuum Pump (MVP) is placed in service with the suction valve throttled.

'Condenser Air Removal Operation',

which ONE of the following actions is permitted if the South Plant Vent (SPV) RMS Effluent monitor reaches the ALERT alarm set point and condenser vacuum cannot be maintained?

A. Stop the MVP. B. Open the Main Condenser Vacuum Breakers.

C. Throttle the MVP Suction valve further closed. D. Increase MVP flow to the High Alarm setpoint.

K&A Rating: 295017AK1.02 (3.8) K&A Statement:

Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE: AK1.02 Protection of the general public Justification:

A. Incorrect but plausible:

Stopping of the MVP would only be required if the HIGH SPV RMS Effluent monitor set point were reached.

B. Incorrect but plausible:

Not required per HC.OP-SO.CG-0001.

Additionally, opened the Main Condenser Vacuum Breakers would increase effluent flow. C. Incorrect but plausible:

lAW HC.OP-SO.CG-0001, Caution 5.8.13, the MVP suction is throttled when the HIGH alarm setpoint is reached.

D. Correct:

lAW HC.OP-SO.CG-0001, Caution 5.8.13, if condenser vacuum cannot be maintained with the flow reduced such that the SPV RMS Effluent ALERT setpoint is not exceeded, flow may be increased but not to exceed the HIGH alarm setpoint (which ensures that ODCM release rate limits are not exceeded).

References:

HC.OP-SO.CG-0001, Rev 45, Section 5.8.13 Student Ref: NONE Learning Objective:

ABBOP6E001 Question Source: Modified from HC Bank (Question

  1. 261) Question History:

Original question used on 2003 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 23 The plant was operating at 100% power when an inadvertent PCIG Isolation occurred due to multiple instrument failures.

lAW EOP-101 "RPV Control",

and assuming NO SCRAM ACTIONS or OTHER OPERATOR ACTIONS are taken for 40 minutes, which ONE of the following, if any, is available to initiate a controlled RPV depressurization?

A Manual operation of SRVs only. B. Manual operation of Bypass Valves only. C. Both SRVs and Bypass Valves. D. Neither SRVs nor Bypass Valves.

K&A Rating: 295020 AK2.01 (3.6/3.7)

K&A Statement:

Knowledge of the interrelations between INADVERTENT CONTAINMENT ISOLATION and the following:

Main steam system Justification:

A. Correct:

Inboard MSIVs will close due to their accumulators bleeding down(-30 min), rendering the Bypass Valves unable to support pressure control.

However, the (ADS) SRVs can be operated manually due to the pneumatic supply accumulators.

B. Incorrect but plausible:

Plausible if candidate does not know that Inboard MSIVs will close due to loss of Instrument Nitrogen.

Inboard MSIVs will close due to their accumulators bleeding down(-30 min), rendering the Bypass Valves unable to support pressure control.

C. Incorrect but plausible:

Plausible if candidate does not know that Inboard MSIVs will close due to loss of Instrument Nitrogen.

Inboard MSIVs will close due to their accumulators bleeding down(-30 min), rendering the Bypass Valves unable to support pressure control.

D. Incorrect but plausible:

Plausible if candidate does not know that Inboard MSIVs will close due to loss of Instrument Nitrogen.

Inboard MSIVs will close due to their accumulators bleeding down (-30 min), rendering the Bypass Valves unable to support pressure control.

References:

NOH01 MSTEAMC-08 NOH01 PCIGOOC-06 NOH04NSSSSOC-04 Learning Objective:

MSTEAME003, PCIGOOE008, Question Source: New Question History:

None Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

10CFR: CFR 41.7/45.8 Comments:

QUESTION 24 Plant conditions are as follows:

  • Reactor Power is 1 00%
  • Torus water level is 81 inches and rising slowly The CRS directs lowering Torus water level using B RHR to Radwaste.

Which ONE of the following describes the reason for this action? A. Torus to Drywell Vacuum Breakers will be covered B. HPCI Turbine exhaust back pressure limits could be exceeded C. RCIC Turbine exhaust back pressure limits could be exceeded D. Primary Containment pressure could be exceeded during a subsequent LOCA K&A Rating: K&A Statement:

295029 EK3.02 (3.6/4.0)

Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL WATER LEVEL: EK3.02 Lowering suppression pool water level e-orred-vacuum breakers occurs at 124 inches. B. Incorrect but plausible:

While plausible due to the increased backpressure felt by HPCI due to increased Torus water level and possibly primary containment

pressure, there is no limit associated with HPCI exhaust backpressure.

C. Incorrect but plausible:

While plausible due to the increased backpressure felt by RCIC due to increased Torus water level and possibly primary containment

pressure, this could only result in a trip of the RCIC turbine on high exhaust pressure.

There is no limit associated with RCIC exhaust backpressure D. Correct:

Per HC.OP-EO.ZZ-01 02, Primary Containment

Control, step SP/L-24, "SRV operation with suppression pool water level above the SRV Tail Pipe Level Limit (STPLL) could damage the SRV discharge lines. This, in turn, could lead to containment failure from direct pressurization and damage to equipment inside the containment (ECCS piping, RPV water level instrument runs, wetwell-to-dry well vacuum breakers, etc.) from pipe-whip and jet-impingement loads. The RPV is therefore not permitted to remain at pressure if suppression pool water level and RPV pressure cannot be restored and maintained below the STPLL." With the given plant conditions, STPLL will be violated at approximately 11 0 inches.

References:

HC.OP-EO.ZZ-0102, Rev. 12 Student Ref: NONE Learning Objective:

EOP1 02E009 Question Source: Bank, LGS Bank 575226 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

10CFR55:

CFR 41.5/45.6 Comments:

QUESTION 25 Hope Creek is at 100% power when ONLY the following overhead alarms are received:

  • 81-85 HPCI OUT OF SERVICE
  • 03-A 1 HPCIIRHR A LEAK TEMP HI When the operator checks the HPCI panel, HPIC inboard and outboard steam line isolation valves are stroking closed.
  • The HPCI turbine was NOT running at the time.
  • HPCI Steam line pressure is 900 psig and lowering.

Which ONE of the following would cause this isolation?

A. HPCI Pipe Chase High Temperature.

B. HPCI Pump Room High Temperature.

C. HPCI Steam Line Low Supply Pressure.

D. HPCI Steam Line High Differential Pressure.

K&A Rating: K&A Statement:

Justification:

295032EA2.01 (3.8/3.8)

Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE:EA2.01 Area Temperature A. Incorrect but plausible:

Isolates after a 15min time delay. Would also cause annunciator D3-B1 HPCI STM LK ISLN TIMER INITIATED.

B. Correct:

Isolates immediately on room high temperature.

High HPCI room temperature causes D3-A 1 annunciator.

This creates a HPCI turbine trip, which causes B 1-B5 annunciator.

B1-A4 HPCI TURBINE TRIP annunciator does not alarm because the F001 is closed (HPCI not in service).

C. Incorrect but plausible:

Isolates at 100 psig Reactor pressure.

Pressure is at least 900 psig in the Reactor D. Incorrect but plausible:

Would cause B1-A4 annunciator, HPCI STEAM LINE DIFF PRESSURE HI.

References:

HC.OP-AR.ZZ-0014 Student Ref: NONE Learning Objective:

EOP1 03E006 Question Source: HC Bank Q3552 Question History:

NRC 2003 Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 26 Given the following conditions:

  • Reactor Building differential pressure is negative at 0.55 inches water gauge. Which ONE of the following will cause a degradation of Reactor Building differential pressure?

A. Level 3 Low RPV water level. B. Inadvertent actuation of the "A" RHR Room Blowout Panel. C. Inadvertent closure of GUD-925 (Rx building supply manual damper).

D. Refuel Floor Vent Exhaust effluent monitor reading 1.5 x 1 o-3 K&A Rating: K&A Statement:

Justification:

295035 2.2.44 ( 4.2) Secondary Containment High Differential Pressure:

2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

A. Incorrect but plausible:

RBVS automatically isolates at Level 2 low Reactor Water level (-38"), not Level 3. B. Incorrect but plausible:

The ECCS blowout panels relieve pressure from the ECCS spaces to the torus space, where increased pressure can be manually vented out of the steam vent. An inadvertent opening of the 'A' RHR blowout panel will not change the overall pressure inside the secondary containment.

C. Correct:

Inadvertent closure of GUD-925 will trip the RBVS Supply fans causing degradation in secondary containment 0/P. This is an actual event documented in LER-00-009 and specifically covered in training material (NOH01 SECCONC-04

). D. Incorrect but plausible:

The Refuel Floor Vent Exhaust monitor automatically isolates RBVS at 2.0 x 1 o-3 11Cilcc.

The Reactor Building Vent Exhaust monitor automatically isolates RBVS at 1.0 x 1 o-3

References:

HC.OP-SO.GR-0001 Student Ref: NONE LER 00-009 Learning Objective:

SECCONE004-008 Question Source: New Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 27 The crew needs to sample containment atmosphere using the H2/02 Analyzers to monitor for high containment hydrogen concentrations post LOCA. Which of the following describes:

(1) ALL sample locations monitored by the Hydrogen/Oxygen Analyzers, and (2) which containment isolation signals to the associated containment isolation valves can be overridden to ensure availability for sampling?

A. (1) The upper drywell and the torus, only (2) Only isolation signals to the upper drywell sample valves can be over ridden B. (1) The upper drywell and the torus, only (2) Isolation signals to ALL the sample valves can overridden C. (1) The upper drywell, lower drywell, and the torus (2) Only isolation signals to the upper drywell sample valves can be overridden D. (1) The upper drywell, lower drywell, and the torus (2) Isolation signals to ALL the sample valves can be overridden K&A Rating: 500000 EK1.01 (3.3/3.9)

K&A Statement:

Knowledge of the operational implications of the following concepts as they apply to HIGH CONTAINMENT HYDROGEN CONCENTRATION:

EK1.01 Containment Integrity Justification:

A. Incorrect but plausible:

the lower drywell is also sampled B. Incorrect but plausible:

the lower drywell is also sampled and all containment isolation signals to the isolation can be manually overridden C. Incorrect but plausible:

all containment isolation signals to the isolation can be manually overridden D. Correct:

per LP NOH01 H202AN Each package (H202 analyzer) takes samples from three different locations; High -Drywell head region, Low -Drywell cylindrical region, Suppression Chamber Air Space. CIVs can be individually opened after the associated isolation override P.B. is depressed at (1 OC650E)

References:

LP NOH01 H202AN-03 Section III.A.2.a.2)

& III.B.1.e.3)

Learning Objective:

NOH01 H202AN-03 Obj. 2 & 3 Question Source: Question History:

Cognitive Level: 10CFR55:

Comments:

NRC 2010 None Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

CFR 41.8-41.7 Student Ref: NONE

QUESTION 28 Given the following conditions:

  • The plant is operating at 100 percent power when a loss of off-site power occurs
  • All EDGs have begun their start sequence Then, a LOCA occurs while the EDGs are starting
  • A valid LPCI initiation signal is present from all 4 divisions of LPCI logic Which ONE of the following describes when the RHR pumps will start? A. RHR pumps will start 5 seconds after the Core Spray Pumps start. B. RHR pumps will start immediately when the EDG output breakers close. C. RHR pumps will start 5 seconds after receiving the LPCI initiation signal. D. RHR pumps will start immediately when RPV pressure lowers to 450psig.

K&A Rating: K&A Statement:

Justification:

203000A4.01 (4.3/4.1)

Ability to manually operate and/or monitor in the control room:A4.01 Pumps A. Incorrect but plausible:

With off-site power unavailable, the RHR pumps will start immediately when their associated EDG output breaker closes, with an automatic initiation signal present.

Core Spray Pumps start after the RHR Pump starts. B. Correct:

RHR Pump D will start immediately when the D EDG output breaker closes. C. Incorrect but plausible:

With off-site power unavailable, the RHR pumps will start immediately when their associated EDG output breaker closes, with an automatic initiation signal present D. Incorrect but plausible:

With off-site power unavailable, the RHR pumps will start immediately when their associated EDG output breaker closes, with automatic initiation signal present.

References:

HC.OP-SO.BC-0001 Student Ref: NONE Learning Objective:

RHRSYSE008 Question Source: HC Bank #35561 Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.7/43.5/45.8 Comments:

QUESTION 29 Given the following conditions:

  • Water level has lowered to -150 inches. Select the correct RHR loop "B" status for these conditions.

A. RHR pump (BP202) will be running.

RHR Injection Valve (HV-F017B) will be full open. RHR Suppression Pool Suction Valve (HV-F004B) will be full open. B. RHR pump (BP202) will be tripped.

RHR Injection Valve (HV-F017B) will be full closed. RHR Suppression Pool Suction Valve (HV-F004B) will be full closed. C. RHR pump (BP202) will be running RHR Injection Valve (HV-F017B) will be full closed. RHR Suppression Pool Suction Valve (HV-F004B) will be full open. D. RHR pump (BP202) will be tripped.

RHR Injection Valve (HV-F017B) will be full open. RHR Suppression Pool Suction Valve (HV-F004B) will be full closed.

K&A Rating: K&A Statement Justification:

205000A2.05 (3.5/3.7)

Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on those predictions, use procedures to correct,

control, or mitigate the consequences of those abnormal conditions or operations A2.05 System Isolation A. Incorrect but plausible:

Suction valves being closed will prevent the pump starting.

Suppression pool suction will not receive an open signal. B. Incorrect but plausible:

LPCI signal will open the Injection Valve. C. Incorrect but plausible:

LPCI signal will open the Injection Valve. Suction valves being closed will prevent the pump starting.

Suppression pool suction will not receive an open signal. D. Correct:

Closing the SOC suction valves on the isolation signal will trip the RHR pump. The LPCI signal will open the RHR Injection Valve. The Suppression Pool Suction valve does not receive an open signal on a LPCI signal.

References:

HC.OP-SO.BC-0001 Student Ref: NONE Learning Objective:

Question Source: HC Bank #209 Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.5/45.6 Comments:

QUESTION 30 HPCI is injecting to the Reactor Vessel with the Flow Controller in MANUAL at 5600gpm.

Reactor Pressure rises from 850 psig to 950 psig. Which ONE of the following describes the effect on HPCI, steady-state to steady-state?

A. RPM will RISE, pump flow will RISE. B. RPM will RISE, pump flow will REMAIN CONSTANT.

C. RPM will REMAIN CONSTANT, pump flow will LOWER. D. RPM will REMAIN CONSTANT, pump flow will REMAIN CONSTANT.

K&A Rating: K&A Statement:

Justification:

206000K5.01 (3.3) Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM: K5.01 Turbine Operation A. Incorrect but plausible:

With FIC-R600 HPCI controller in MANUAL, pump speed is maintained at a constant speed without operator action. B. Incorrect but plausible:

With FIC-R600 HPCI controller in MANUAL, pump speed is maintained at a constant speed without operator action. C. Correct:

With FIC-R600 HPCI controller in MANUAL, pump speed is maintained at a constant speed without operator action. If Reactor Pressure rises, pump speed remains constant and pump flow will lower due to increased backpressure (more resistance).

D. Incorrect but plausible:

With FIC-R600 HPCI controller in MANUAL, pump speed is maintained at a constant speed without operator action. However, if Reactor Pressure rises, pump flow will lower due to increased backpressure, not remain constant.

References:

NOH01 HPCI1 00-10 Student Ref: NONE Learning Objective:

Obj 6a,b,c,d Question Source: New Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 31 While responding to a Loss of Coolant Accident, the Plant Operator (PO) notices that the "A" Core Spray Loop fails to initiate automatically (the 'Initiation Logic Seal-In' lights are NOT illuminated).

The PO manually starts both "A" Loop pumps, and then attempts to open inboard isolation valve HV-F005A.

Which ONE of the following describes the response of HV-F005A?

A. HV-F005A will open if outboard isolation valve HV-F004A is closed. B. HV-F005A will open provided the PO also depresses AUTO OPEN OVRD. C. HV-F005A will open immediately upon both 'A' loop pumps starting.

D. HV-F005A will open when RPV pressure is less than 461 psig.

K&A Rating: K&A Statement:

Justification:

209001 K4.01 (3.2/3.4)

Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following:

K4.01 Prevention of over pressurization of core spray piping A. Correct:

Without an auto initiation signal, HV-F005A will only open if outboard isolation valve HV-F004A is closed to prevent over pressurization of the core spray piping. B. Incorrect but plausible:

The AUTO OPEN OVRD, is only active with an initiation signal present and allows the operator to close the valve, there is no initiation signal as evident by 'Initiation Logic Seal-In' lights are not illuminated.

C. Incorrect but plausible:

The auto open signal requires an initiation signal and RPV pressure<

461 psig, there is no initiation signal as evident by 'Initiation Logic Seal-In' lights are not illuminated.

D. Incorrect but plausible

The auto open signal requires an initiation signal, with 'Initiation Logic Seal-In' lights are not illuminated, the valve will not open.

References:

NA Learning Objective:

Question Source: Question History:

Cognitive Level: 10CFR: Comments:

HC Bank Q#72 NA Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

CFR41.07 Student Ref: NONE QUESTION 32 Given the following conditions:

  • The Reactor scrammed from 100% power on turbine stop valve closure
  • The RO manually starts both SLC pumps at the direction of the CRS
  • Initial SLC tank level is 4850 gallons 45 minutes later, and assuming NO further operator
actions, the SLC pumps are ... A. tripped on SLC tank low level. B. running with the SLC tank empty. C. tripped with the reactor shutdown.

D. running and injecting to the vessel.

K&A Rating: 211000 At01 (3.6/3.7)

K&A Statement:

Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM: A1.01 Tank Level Justification:

A. Incorrect but plausible:

The SLC pumps would still be running because level would be >325 gal after 45 minutes B. Incorrect but plausible:

The SLC tank will be above the low level trip. The Low level trip would not be bypassed when started from the Control Room C. Incorrect but plausible:

The reactor would be shutdown because <11 00 gal remain in the tank but the SLC pumps would still be running D. Correct:

Pumps auto trip on low level at 325 gals remaining provided the pumps were not running in test. Min required capacity (2 pumps)= 84.2 gpm (OP-IS-BH-0001/0002),

(4850-325) I 84.2 = 53.2 minutes.

Max acceptable flow rate= 48.6 gpm X 2 = 97.2 gpm. 4850 -325/ 97.2 = 46.6 minutes.

Therefore the pumps should still be running at the 45 minute mark

References:

NOH01SLCSYSC, Rev. 4 HC.OP-AR.ZZ-0008 Att. E1, Rev. 43 Learning Objective:

SLCSYSEOO 18/E007 Student Ref: NONE Question Source: Modified from PB Bank (Question

  1. 1) Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/45.5 Comments:

QUESTION 33 Given the following conditions:

  • The Reactor scrammed on low Reactor water level
  • All rods failed to insert
  • SLC was initiated
  • RPV level was lowered lAW HC.OP-EO.ZZ-0101A, ATWS-RPV Control
  • RPV level is being maintained between -150" and -185"
  • The "A" RPS Bus is de-energized and cannot be restored
  • The implementation of HC.OP-EO.ZZ-0301, "Bypassing MSIV Isolation Interlocks" and reopening of the Main Steam Isolation Valves (MSIVs) has been directed Following completion of HC.OP-EO.ZZ-0301, which ONE of the following describes the ability to operate MSIVs from the control room? A. None of the eight MSIVs can be reopened.

B. Only the four outboard MSIVs can be reopened.

C. All eight of the MSIVs can be reopened.

D. Only the four inboard MSIVs can be reopened.

K&A Rating: 212000K1.14 (3.6/3.7)

K&A Statement:

Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following:

K1.14: Main Steam system. Justification:

A. Correct:

Both A and B RPS buses must be energized to allow NSSSS logic to be reset. E0-301 only bypasses the Lo-Lo-Lo level isolation.

B. Incorrect but plausible:

Because both A and B busses need to be energized.

Plausible if the candidate thinks that one RPS is enough power to open the outboard valves. C. Incorrect but plausible:

Because both A and B busses need to be energized.

Plausible if the candidate thinks that bypassing the LO-LO-LO logic is all that is required to reopen the MSIV's. D. Incorrect but plausible:

Because both A and B busses need to be energized.

Plausible if the candidate thinks that one RPS is enough power to open the inboard valves.

References:

HC.OP-EO.ZZ-0301, Section 2.6 NONE Learning Objective:

NA Question Source: Bank (Question

  1. 255) Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.2 to 9/43.7/45.8 Comments:

Student Ref:

QUESTION 34 Given the following conditions:

  • The Mode Switch is in START UP

Which ONE of the following describes the plant condition after the IRM "E" malfunction?

A. A rod block is generated by the Rod Block circuitry; A full SCRAM is received.

B. A rod block is generated by the Rod Block circuitry; A full SCRAM is NOT received.

C. A rod block is NOT generated by the Rod Block circuitry; A full SCRAM is received.

D. A rod block is NOT generated by the Rod Block circuitry; A full SCRAM is NOT received.

K&A Rating: 215003K1.02 (3.6) K&A Statement:

Knowledge of the physical connections and/or cause effect relationship between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following:

K1.02 Reactor manual control Justification:

A. Correct:

RPS "B" has a half SCRAM as stated in the stem. With the mode switch in startup, the IRM "E" malfunction and IRM NB/E/F UPSCALE INOP/TRIP alarm is an indication of a half SCRAM on RPS "A" AND a rod block will be generated by the Rod Block circuitry.

A half SCRAM on both channels of RPS satisfies the conditions for a full reactor SCRAM. B. Incorrect but plausible:

A rod block IS generated, but a full SCRAM is also generated, not a half SCRAM. C. Incorrect but plausible:

A rod block IS generated as evidenced by the IRM "E" malfunction and IRM NB/E/F UPSCALE INOP/TRIP alarm stated in the stem. D. Incorrect but plausible:

A rod block IS generated as evidenced by the IRM "E" malfunction and IRM NB/E/F UPSCALE INOP/TRIP alarm stated in the stem.

References:

NOH01 RPSOOC-09 HC.OP-AR.ZZ-0009(Q)

Learning Objective:

RPSOOOE007 Question Source: New Question History:

N/A Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/

Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 35 Given the following conditions:

  • A Reactor Startup is in progress
  • Reactor power is increasing with a stable positive period of 150 seconds
  • The SRM Channel 'A' detector is stuck and will NOT withdraw
  • The SRM Channel 'A' indication increases to 2x1 05 cps Assume NO operator actions are performed.

Indicated Reactor power will ... A continue to increase and Reactor period will remain positive.

B. decrease and Reactor period will be negative.

C. continue to increase and Reactor period will become shorter.

D. decrease and Reactor period will initially go to infinity.

K&A Rating: 215004K6.04 (2.9/2.9)

K&A Statement:

Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRH) SYSTEM: K6.04 Detectors Justification:

A. Incorrect but plausible:

The Reactor SCRAM will result in a lowering count rate and negative period. B. Correct:

The Reactor will SCRAM, causing a lowering count rate and a negative period. C. Incorrect but plausible:

The Reactor SCRAM will result in a lowering count rate and negative period. D. Incorrect but plausible:

The Reactor SCRAM will result in a negative period.

References:

HC.OP-SO.SB-0001 Student Ref: NONE HC.OP-SO.SE-0001 Learning Objective:

SRMSYSSE01 0 Question Source: Bank Question History:

HC Bank# 164 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR41.7/45.7 Comments:

QUESTION 36 Given the following conditions:

  • Reactor Power is 55%
  • Control rod 30-31 is selected Subsequently, Recirculation Flow Unit "A" fails downscale.

Which ONE of the following describes ALL the Scram (if any) and withdraw Rod Block signals generated from this failure?

A. No scram signals are generated.

Withdraw rod block signals are generated by APRM Channels A, C, and E; RBM Channel B; and the A and B Recirculation Flow Units. B. No scram signals are generated.

Withdraw rod block signals are generated by APRM Channels A, C, and E; RBM Channel A; and the A and B Recirculation Flow Units. C. Scram signals are generated by APRM Channel A, C, and E. Withdraw rod block signals are generated by APRM Channel A, C, and E; and Recirculation Flow Unit A, ONLY. D. Scram signals are generated by APRM Channels A, C, and E. Withdraw rod block signals are generated by APRM Channels A, C, and E; RBM Channel A and 8; and the A and B Recirculation Flow Units.

Question 36 K&A Rating: K&A Statement:

Justification:

215005 K4.01 (3.7/3.7)

Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following:

K4.01 Rod withdrawal blocks A. Incorrect but plausible:

RBM channel B is not affected by flow converter A failure.

Plausible if applicant does not know which flow converters input into which RBM. B. Correct:

The APRM flow biased rod block setpoint at zero flow is 53%. The reactor power is 2112/3840 which is 55% therefore Control Rod withdrawal block signals are generated from APRM Channels A, C, and E. RBM Channel A zero flow setpoint on the low range is 49%. A and B Recirculation Flow Units with 10% difference between flow comparators will generate a rod block. C. Incorrect but plausible:

The APRM flow biased scram setpoint at zero flow is 58%. The reactor power is 2112/3840 which is 55% and therefore no scram signal will be generated.

D. Incorrect but plausible:

The APRM flow biased scram setpoint at zero flow is 58%. The reactor power is 2112/3840 which is 55% and therefore no scram signal will be generated.

References:

HC.OP-SO.SF-0002-RBM Student Ref: NONE Learning Objective:

NA Question Source: HC BANK #64976 Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7 Comments:

QUESTION 37 The plant was operating at 80% power when a transient occurred requiring a manual initiation of RCIC. Approximately 15 minutes later the PO notes that HV-F031 RCIC PUMP SUCTION FROM THE SUPPRESSION POOL has drifted to a 50% OPEN position.

Which ONE of the following describes the system response?

A. HV-F010 PUMP SUCTION VALVE FROM CST will immediately CLOSE. B. HV-F022 RCIC PUMP DISCHARGE TEST RETURN TO CST valve will receive an OPEN signal. C. HV-F010 PUMP SUCTION VALVE FROM CST will remain OPEN until CST Level reaches -22,500 gallons.

D. HV-F010 PUMP SUCTION VALVE FROM CST will remain OPEN until CST Level reaches -70,000 gallons.

K&A Rating: K&A Statement:

Justification:

217000K2.02 (2.6) Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC): K6.04 Condensate storage and transfer system A. Incorrect but plausible:

HV-F01 0 will immediately close if HV-F031 goes FULL OPEN. Since HV-F031 is only partially open, HV-F010 will not immediately close. B. Incorrect but plausible:

HV-F022 receives a CLOSE signal, not an OPEN signal when HV-F031 is not full CLOSE. C. Incorrect but plausible:

HV-F01 0 will automatically close when CST level reaches -70,000 gallons.

The setpoint was changed from -22,500 gallons due to pump vortexing concerns.

D. Correct:

HV-F01 0 will automatically close when CST level reaches -70,000 gallons

References:

NOH04RCIC00-1 0 Student Ref: NONE Learning Objective:

RCICOOEE05, RCICOOEE09 Question Source: New Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 38 OP-EO.ZZ-1 03, Reactor Building

Control, has been entered due to an unisolable primary coolant leak in the RWCU Non-Regenerative Heat Exchanger Room. The following conditions exist:
  • RPV pressure is 725 psig and slowly lowering
  • RPV water level is -140 inches
  • Drywell Pressure is 0.5 psig and steady
  • ADS LOGIC CH B MANUAL INHIBIT Switch is in "INHIB"
  • ADS LOGIC CH D MANUAL INHIBIT Switch has failed in the "NORM" position
  • All low pressure ECCS Pumps are running with normal indications 395 seconds after reaching RPV Level 1, the LOGIC D/H I NIT RESET Pushbutton is depressed and released.

Which ONE of the following identifies the effect on the ADS valves? A. Valves will close and re-open 300 seconds later B. Valves will close and re-open 405 seconds later C. Valves will remain closed; valves will auto open 405 seconds later D. Valves will remain closed; valves will NOT auto open on any time delay K&A Rating: K&A Statement:

218000 A3.09 (4.2/4.3)

Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including:

Reactor vessel water level Justification:

A. Incorrect but plausible:

Plausible if the applicant does not understand ADS Initiation Logic and System response with respect to (1) operation of the 300 second ADS High Drywell Bypass Timer and the 105 second Logic D/H (CH D) ADS Initiation Timer, and (2) manipulation of the Manual Inhibit Switches and the Logic Reset Pushbuttons.

See Answer 'C' discussion.

B. Incorrect but plausible:

Plausible if the applicant does not understand ADS Initiation Logic and System response with respect to (1) operation of the 300 second ADS High Drywell Bypass Timer and the 105 second Logic D/H (CH D) ADS Initiation Timer, and (2) manipulation of the Manual Inhibit Switches and the Logic Reset Pushbuttons.

See Answer 'C' discussion.

C. Correct:

The LOGIC D/H I NIT RESET Pushbutton is depressed 10 seconds prior to ADS initiation (auto initiation occurs at 405 seconds).

The 300 second ADS High Drywell Bypass Timer will be reset. The CH DADS Initiation Timer will also be reset. The ADS valves never opened because the auto initiation at 305 seconds (105 seconds after the 300 second ADS High Drywell Bypass Timer times out) was interrupted when the LOGIC D/H !NIT RESET Pushbutton was depressed at time 395 seconds.

The ADS valves will open 405 seconds later when both the ADS High Drywell Bypass Timer (300 seconds) and the Logic D/H (CH D) ADS Initiation Timer ( 105 seconds) successfully time out (300 + 1 05 = 405 seconds) after having been reset. D. Incorrect but plausible:

Plausible if the applicant does not understand ADS Initiation Logic and System response with respect to (1) operation of the 300 second ADS High Drywell Bypass Timer and the 105 second Logic D/H (CH D) ADS Initiation Timer, and (2) manipulation of the Manual Inhibit Switches and the Logic Reset Pushbuttons.

See Answer 'C' discussion.

References:

NOH04ADSSYSC-08 Student Ref: NONE Learning Objective:

Obj. 2 of NOH04ADSSYSC-08 Question Source: New Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR41.7 Comments:

QUESTION 39 A reactor startup is in progress with the following conditions:

  • Reactor pressure 750 psig
  • The Reactor Mode Switch is in "Startup/Hot Standby"
  • No conditions to cause a reactor scram have been generated

Justification:

223002K1.01 (3.8/3.9)

Knowledge of the physical connections and/or cause-effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the following:

K1.01 Main Steam System. A. Incorrect but plausible:

Since a low level scram did not occur, reactor watr level is >12.5 inches and will not cause an isolation.

B. Incorrect but plausible:

Because the mode switch would have to be in run to enable the low pressure isolation.

C. Incorrect but plausible:

Main Steam Line Radiation does not cause a group 1 isolation.

Plausible because it does cause a group 2 isolation.

D. Correct:

High Main Steam Line temperature will cause an isolation in these plant conditions.

References:

HC.OP-SO.SM-0001 B21-1 090-0062 Learning Objective:

Question Source: Bank #82 Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

Student Ref: NONE Comprehensive/Analysis:

X 10CFR: CFR 41.2 to 9/45.7 to 8 Comments:

QUESTION 40 Given the following:

  • The ADS Manual Initiation Channel D and H pushbuttons (S6D and S6H) have been armed and depressed
  • There is NO Safety Relief Valve response Which ONE of the following bus losses would cause this system response?

A. 125 VDC bus 1BD417 B. 125 VDC bus 1DD417 C. 120 VAC bus 1BJ481 D. 120 VAC bus 1 DJ481 K&A Rating: 239002 K2.01 (2.8/3.2)

K&A Statement:

Knowledge of electrical power supplies to the following:

K2.01 SRV Solenoids Justification:

A. Incorrect but plausible:

If applicant does not recall power supply to ADS logic D&H; this is the power supply to ADS logic B&F B. Correct:

This is the power supply to ADS logic D&H C. Incorrect but plausible:

If applicant does not understand effect of this loss. This will result in a loss of automatic initiation of ADS channel B, but will not prevent a manual initiation D. Incorrect but plausible:

If applicant does not understand effect of this loss. This will result in a loss of automatic initiation of ADS channel D, but will not prevent a manual initiation

References:

NOH04ADSSYSC Rev. 8 HC.OP-AB.ZZ-0136, Rev. 19 Learning Objective:

ADSSYSE006 Question Source: New Question History:

None Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

10CFR55:

CFR41.7 Comments:

QUESTION 41 Given the following conditions:

  • The plant is at 1 00% power
  • The 'C' Steam Flow detector for the Feedwater Level Control System fails low (its output indicates 0 lbm/hr steam flow) Which ONE of the following describes the automatic plant response with NO operator action? ACTUAL reactor water level will ... A. remain the same. The Feedwater Level Control System will shift to single element control.

B. decrease and stabilize at a lower than normal value. The Feedwater Level Control System will shift to single element control.

C. decrease and stabilize at a lower than normal value. The Feedwater Level Control System will remain in three element control.

D. increase and stabilize at a higher than normal value. The Feedwater Level Control System will remain in three element control.

K&A Rating: 259002A 1. 01 (3. 8/3. 8) K&A Statement:

Ability to predict and/or monitor changes in parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including:

A1.01 Reactor Water Level Justification:

A. Correct:

The system transfers to single element control.

With stable conditions, single and three element control should control level at the same setpoint.

B. Incorrect but plausible:

Level should not decrease.

C. Incorrect but plausible:

Level should not decrease.

The system will shift to single element control.

D. Incorrect but plausible:

Level should not increase.

The system will shift to three element control.

References:

DWG H-1-AE-ECS-0128-0 Student Ref: NONE Learning Objective:

FWCONTE012 Question Source: HC Bank# 233 Question History:

Used on 2003 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/45.5 Comments:

QUESTION 42 Given the following conditions:

  • The plant is at 50% power
  • The FRVS Recirculation Fans are in AUTO Subsequently, a Loss of Offsite Power (LOP) occurs:
  • The minimum water level reached was -32 inches Assuming NO other operator
actions, the total FRVS recirculation flow 3 minutes after the LOP will be ... A. 0 cfm. B. 60,000 cfm. C. 120,000 cfm. D. 180,000 cfm.

K&A Rating: 261000A2.10 (3.1/3.2)

K&A Statement:

Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct,

control, or mitigate the consequences of those abnormal conditions or operations:

K2.1 0 Low reactor water level: Plant-Specific Justification:

A. Correct:

There are no automatic isolation signals present; No FRVS fans are running.

B. Incorrect but plausible:

All six FRVS recirculation fans start at RPV level 2 (-38 inches).

Plausible if the applicant believes that some combination of FRVS fans would be running given the stem conditions.

C. Incorrect but plausible:

All six FRVS recirculation fans start at RPV level 2 (-38 inches).

Plausible if the applicant believes that some combination of FRVS fans would be running given the stem conditions.

D. Incorrect but plausible:

All six FRVS recirculation fans start at RPV level 2 (-38 inches).

Plausible if the applicant believes that some combination of FRVS fans would be running given the stem conditions.

References:

HC.OP-AB.CONT-0003 Student Ref: NONE HC.OP-SO.GR-0001 Learning Objective:

RBVENTE004 Question Source: Bank Question History:

HC Bank # 130 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/45.6 Comments:

Question meets part (a) of KIA.

QUESTION 43 Opening the toggle switch associated with the 500KV breaker Emergency Trip handle ... A. transfers all control functions to the blockhouse.

B. prevents the Emergency Trip function.

C. removes control power for the breaker.

D. enables removal of kirk key at the breaker.

K&A Rating: K&A Statement Justification:

262001 K5.02 (2.6) Knowledge of the operational implications of the following concepts as they apply to A. C. ELECTRICAL DISTRIBUTION.

K5.02 Breaker Control A. Incorrect but plausible:

No local control at the breaker except for test. B. Incorrect but plausible:

Opening the toggle switch does not prevent emergency operation of the breaker.

C. Correct:

Interrupts breaker control circuit to prevent electrical operation.

D. Incorrect but plausible:

Kirk Key can be removed with toggle in either position.

References:

NOH01 MNPWR0-04, MAIN POWER SYSTEM, p21 Student Ref: NONE Learning Objective:

MNPWROE014 Question Source: Hope Creek Question-Q68164 Question History:

2005 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

1 OCFR: CFR 41.5/45.3 Comments:

QUESTION 44 A UPS Manual Bypass Switch has been repositioned from "BYPASS TO ALTERNATE" to "ISOLA TED AFTER PREFERRED" for testing on a faulty Static Switch. Which ONE of the following describes the system response if a loss of offsite power occurs, and the MCCs for the Preferred and Backup power are NOT 1 E switchgear supplied?

The input to the Regulator/Static Switch Cabinet will be supplied by _(1)_ and power to system loads will _(2)_ . A. (1) 125 VDC power (2) be maintained.

B. (1) 125 VDC power (3) be lost. C. (1) normal480 VAC power (2) be lost. D. (1) normal 480 VAC power (2) be maintained.

(See attached UPS Power Control Circuit Diagram)

UPS POWER CONTROL (TYPICAL)

BACKUP POWER 4a0 VAC CBJ01 ) (BACK UP mn AC POWER SUPPLY REGULATOR CABINET SWITCH POSITION NORMAL BYPASS TO PREFERRED ISOLATE (AFTER PREF.) BYPASS TO ALTERNATE ISOLATE (AFTER ALT.) CB21 1 X X X BLOCKING DIODES 2 3 X X X X Ai..TERHA.TE POWER 125 voc I) CB" (; LT DC INPUT At; POWER SUPPL 'f RECTIFIER CABINET LOAD 4 5 X X X X K&A Rating: 262002A4.01 (2.8/3.1)

K&A Statement:

Ability to manually operate and/or monitor in the control room: A4.01 Transfer from alternative source to preferred source. Justification:

A Correct:

Taking the Manual Bypass Switch to "ISOLATED TO PREFERRED" places the load directly on the preferred (inverter output) power source and all power is removed from the static switcrl The preferred power source consists of the DC output of the rectifier cabinet which comes from the auctioneered high input. The inputs to the rectifier cabinet are normal480 VAC or 125 VDC (alternate preferred source).

In the case of a LOP on a non 1 E switchgear, power will be supplied to the regulator cabinet via the preferred source powered by 125 VDC. B. Incorrect but plausible:

Supply to the system loads will not be lost, the static switch is de-energized to the preferred source position, which will be powered by 125 VDC. C. Incorrect but plausible:

The output of the rectified cabinet will take auctioneered high power source -> 125VDC since 480 VAC normal will be lost. D. Correct:

Backup 480 VAC power will only be supplied to the regulator/static switch cabinet if the static switch was in the alternate power position.

References:

NOH01 EAC00-04 Student Ref: NONE Learning Objective:

NON1 EOE003 Question Source: New Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7/45.5 to 45.8 Comments:

QUESTION 45 Given the following condition:

  • 125 VDC Bus B (1 BD417) tripped off due to a bus fault. Which ONE of the following describes the location(s),

if any, from which SRV 'F' can be opened? A. NOT available from any location.

B. Available from the Main Control Room ONLY. C. Available from the Remote Shutdown Panel ONLY. D. Available from the Main Control Room AND the Remote Shutdown Panel.

Question 45 K&A Rating: K&A Statement:

Justification:

263000 K2.01 (3.2/3.3)

Knowledge of electrical power supplies to the following:

K2.01: Major D.C. loads A. Correct : 1 BD417 feeds all of the SRV "A" solenoids.

SRV F is a non-ADS SRV, so it only has one solenoid and it is an 'A' solenoid.

SRV F has one control station in the MCR and one at the RSP; however, the valve cannot be opened from either location since both locations use the same 'A' solenoid and there is no power available.

B. Incorrect but plausible:

Will not open due to no power. C. Incorrect but plausible:

Will not open due to no power. D. Incorrect but plausible

Will not open due to no power

References:

NA Student Ref: NONE Learning Objective:

loss of 1 BD417 Question Source: Bank HC #191 Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

10CFR55:

CFR41.7 Comments:

QUESTION 46 One second after paralleling the 'A' EDG to the 10A401 bus, a Loss of Offsite Power occurs and the 'A' EDG output breaker trips (its indication is flashing).

After inspection, the SM directs reclosing the 'A' EDG output breaker.

Which ONE of the following operator actions will energize the 1 OA401 4.16 KV bus? A. Depress the TRIP Pbs on the Normal and Alternate supply breakers, then allow the EDG output breaker to auto close. B. Locally reset the EDG output breaker, then using the EDG output breaker CLOSE Pb, reenergize the 10A401 bus. C. Reset the Test Lockout relay, then using the EDG output breaker CLOSE Pb, reenergize the 10A401 bus. D. Acknowledge the flashing TRIP Pb on the EDG output breaker, then allow the EDG output breaker to auto close.

K&A Rating: 264000 G2.1.28 (4.1/4.1)

K&A Statement:

2 1.28 Knowledge of the purpose and function of major system components and controls.

Justification:

A. Incorrect but plausible:

Depress the TRIP Pbs on the Normal and Alternate supply breakers, then allow the EDG output breaker to auto close.Will prevent the EDG from loading AB-135 Note 4.2. B. Incorrect but plausible:

Locally reset the EDG output breaker, then using the EDG output breaker CLOSE Pb, reenergize the 10A401 bus. Cannot be reset locally.

C. Incorrect but plausible:

Reset the Test Lockout relay, then using the EDG output breaker CLOSE Pb, reenergize the 1 OA401 bus. Test LO would not be tripped.

D. Correct:

Acknowledge the flashing TRIP Pb on the EDG output breaker, then allow the EDG output breaker to auto close. AB-135 5.16-The Anti-pump circuitry on the DIG output breaker could cause the output breaker to fail open, if a LOP were to occur within 2 seconds of closing the output breaker when testing and loading the DIG to the grid. To load the D/G under this condition the operator must wait a minimum of 2 seconds from the time the breaker was originally closed, then depress the TRIP push-button (even though the breaker is already tripped) to reset the logic. When the TRIP push-button is released, then the breaker will close and the D/G will load.

References:

NOH04EDGOOOC-03, ABZZ-0135 Student Ref: NONE Learning Objective:

EDGOOOE020, OAB135E004 Question Source: Bank# 33519 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR41.7 Comments:

QUESTION 47 The 1 OK1 07 Service Air Compressor is in service in the LEAD mode of o Subsequently, the following conditions are noted:

  • 2ND INTERCOOLER HIGH LEVEL alarm is in on the OK107 local control panel
  • AFTERCOOLER AIR OUTLET HIGH TEMP SHU OWN alarm is in on the 1 OK1 07 local control panel
  • Lube Oil pressure is 25 psig Which ONE of the following now describes th status of the 1 OK1 07 Service Air Compressor?

The compressor

... A B. C. D. on aftercooler air outlet temperature.

K&A Rating: 300000K5.01 (2.5/2.5)

K&A Statement:

Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: K5.01 Air Compressors Justification:

A. Incorrect but plausible:

The low oil pressure trip occurs at 1 Opsig. B. Incorrect but plausible:

While there is an alarm for high demister level, there is no associated compressor trip. C. Correct:

None of the conditions listed in the question stem will result in a Service Air Compressor trip. D. Incorrect but plausible:

While there is an alarm for high aftercooler air outlet temperature, there is no associated compressor trip. An aftercooler air outlet high temperature of 115F NO LONGER will cause its associated compressor to trip.

References:

NOH01 SERAIR-05 Student Ref: NONE Learning Objective:

Obj 3 Question Source: New Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

1 OCFR55: CFR 41 .5/45.3 Comments:

QUESTION 48 Given the following conditions:

  • Reactor power is 83%
  • Normal electrical lineup
  • TACS is being supplied from the 'A' SACS loop Which ONE of the following describes the response of the SACS to depressing the TRIP pushbutton for breaker 40108 ( 1 OA401 Normal Feeder Breaker)?

A. The 'A' SACS Pump AP21 0 will trip, TACS swaps to supply system loads. B. The 'A' SACS Pump AP21 0 will trip, loop 'A' will remain online supplying both SACS and TACS loads at a reduced flow. C. The 'C' SACS Pump CP21 0 will trip, loop 'A' will remain online supplying both SACS and TACS loads at a reduced flow. D. The 'A' SACS Pumps AP21 0 and the 'C' SACS Pump CP21 0 will both instantly trip, the TACS swaps to supply system loads.

K&A Rating: K&A Statement:

Justification:

400000 A3.01 (3.0/3.0)

Ability to monitor automatic operations of the CCWS including:

A3.01 Setpoints on instrument signal levels for normal operations,

warnings, and trips that are applicable to the CCWS A. Correct:

Pump AP21 0 will trip on undervoltage, which closes the associated SACS to TACS isolation, causing a low flow condition in the 'A' loop. The low flow will start the AUTO pump in the 'B' loop, and open the 'B' loop isolation valves B. Incorrect but plausible:

Pump AP21 0 will trip on undervoltage, which will close the SACS to T ACS isolation C. Incorrect but plausible:

If applicant does not recall correct pump power supplies.

'C' SACS pump is supplied from the 1 OA403 bus D. Incorrect but plausible:

If applicant does not recall correct pump power supplies.

'C' SACS pump is supplied from the 1 OA403 bus

References:

NOH04STACSOC, Rev. 8 Student Ref: NONE Learning Objective:

ST ACSOE004 Question Source: HC Bank 33814 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7/45.7 Comments:

QUESTION 49 Given the following (assume no operator actions):

  • HPCI has initiated from a valid Hi Drywell Pressure signal
  • RPV Water level is 35"
  • Reactor pressure is 980 psig Then, the operator observes that the HPCI turbine control valve (FV-4879) is in the throttled position AND that turbine speed is decreasing.

Additionally, the operator observes the following valves going closed:

  • HV-F012, HPCI minimum flow Finally, the operator later notices that HPCI turbine speed is again increasing and HPCI realigns for injection (without operator actions).

Which ONE of the following caused this HPCI response?

A. Incorrectly set HPCI flow controller.

B. Mechanical overspeed HPCI trip. C. Automatic reset of the HPCI isolation logic. D. Automatic swap of the HPCI pump suction source.

K&A Rating: 206000 A3.01 (3.6/3.5)

K&A Statement:

Ability to monitor automatic operations of the HIGH PRESSURE COOLANT INJECTION SYSTEM including:

A3.01 Turbine Speed Justification:

A. Incorrect but plausible:

Represents a valid, incorrect condition for the applicant to evaluate.

B. Correct:

The HPCI turbine will respond to a mechanical overspeed trip (approximately 5200 rpm) by removing hydraulic oil from the turbine stop valve actuator and then reapplying it after the turbine speed has slowed sufficiently.

The turbine control valve will throttle to attempt to control the speed. C. Incorrect but plausible:

Represents a valid, incorrect condition for the applicant to evaluate.

D. Incorrect but plausible:

Represents a valid, incorrect condition for the applicant to evaluate.

References:

HC.OP-SO.BJ-0001 NOH0*1 HPCI00-1 0 Learning Objective:

HPCIOOE004 Question Source: Bank# 14/33365 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: Comments:

Student Ref: None QUESTION 50 Hope Creek is shutdown with refueling in progress.

The shorting links ARE installed.

Which ONE of the following will occur if the "A" Source Range Monitor (SRM) drawer mode switch is taken out of the OPERATE position?

Alarm Status Block/RPS Status A. A SRM downscale alarm Rod Block and Reactor Scram will occur B. A SRM downscale alarm Rod Block occurs, Reactor Scram will NOT occur C. A SRM Upscale/lnop alarm Rod Block occurs, Reactor Scram will NOT occur D. A SRM Upscale/lnop alarm Rod Block and Reactor Scram will occur Question 50 K&A Rating: K&A Statement:

Justification:

295004 K3.01 (3.4) Knowledge of the effect that a loss or malfunction of the source range monitor system (SRM) will have on the following:

K3.01 RPS A Incorrect but plausible:

Plausible if the applicant determines that placing the drawer mode switch out of Operate creates a SRM downscale trip, and determines that full scram will occur based on SRM downscale condition.

B. Incorrect but plausible:

Plausible if the applicant determines that placing the drawer mode switch out of Operate creates a SRM downscale trip, and determines that rod block will occur. C. Correct:

Placing the drawer mode switch out of Operate creates a SRM Inoperative trip causing Upscale/lnop alarm, and due to the upscale/inop alarm a rod block will also occur. D. Incorrect but plausible:

Plausible due to partially correct that Placing the drawer mode switch out of Operate creates a SRM Inoperative trip causing Upscale/lnop alarm, however SRM channels generate a scram signal on an INOP condition for loss of power only condition.

References:

Learning Objective:

Question Source: Question History:

Cognitive Level: 10CFR55:

Comments:

New NA Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

CFR 41.7/43.5/45.4 Student Ref: NONE QUESTION 51 Given the following conditions:

  • The plant is at 85% power
  • All three Reactor Feed Pumps are in AUTO
  • RPV Narrow Range Level instruments indicate:

o N004A=34inches o N0048=35inches o N004C=35.5 inches Which ONE of the following describes the response of ACTUAL Reactor water level if a slow leak developed through the N004B detector equalizing valve, eventually causing a gross failure of N004B? ACTUAL Reactor water level would: A. LOWER 0.5 inch, then RISE 1.5 inches B. RISE 1.0 inch, then LOWER 0.5 inches C. LOWER 1.0 inch, then RISE 0.5 inches D. RISE 0.5 inch, then LOWER 1.5 inches K&A Rating: 259002K3.02 (3.8/3.8)

K&A Statement:

Knowledge of the effect that a loss or malfunction of the REACTOR WATER LEVEL CONTROL SYSTEM will have on the following:

K3.01 Reactor Water Level Justification:

A. Correct:

Initially, N004B is selected since DFCS selects the MEDIAN RPV level signal when 3 good signals are available.

With a leak through the N004B equalizing valve, N004B indicated level would begin to rise, resulting in a lowering of actual RPV level. As soon as N004B exceeded 35.5 inches indicated, N004C would become the median RPV level signal. Actual RPV level would have lowered 0.5 inch during this transition.

When N004B gross fails, N004A (the lowest of the two remaining signals) will become the controlling level signal. RPV water level will then rise since indicated level on N004A is 34 inches. This is a 1.5 inch rise from the previous level. B. Incorrect but plausible:

Initially lowers 0.5 inches, then rises 1.5 inches. C. Incorrect but plausible:

Initially lowers 0.5 inches, then rises 1.5 inches. D. Incorrect but plausible:

Initially lowers 0.5 inches.

References:

DWG H-1-AE-ECS-0128-0 Student Ref: NONE Learning Objective:

FWCONTEOI 0 Question Source: HC Bank# 95 Question History:

Used on 2009 NRC Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7/45.4 Comments:

QUESTION 52 125VDC bus 1 CD417 is de-energized when an Emergency Diesel Generator start signal is received.

Diesel Generator 1 CG400 will ... A. NOT automatically start B. automatically start but the output breaker can only be shut manually C. automatically start but in the DROOP mode D. automatically start but all trips will be disabled K&A Rating: K&A Statement:

Justification:

263000 K3.01 (3.4/3.8)

Knowledge of the effect that a loss or malfunction of the D. C. ELECTRICAL DISTRIBUTION System will have on the following:

K3.01 Emergency Generators:

Plant-Specific A. Correct:

The diesel generator will NOT automatically start; power is lost to the air start solenoids B. Incorrect but plausible:

The diesel generator will NOT automatically start C. Incorrect but plausible:

The diesel generator will NOT automatically start D. Incorrect but plausible:

The diesel generator will NOT automatically start

References:

NO HOI DCELEC, Rev. 3 Student Ref: NONE Learning Objective:

DCELECE003 Question Source: HC Bank 33239 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X IOCFR55:

CFR 41.7/45.4 Comments:

QUESTION 53 Which ONE of the following conditions (signals) will cause a trip of the Emergency Instrument Air Compressor?

A. Instrument Air Header pressure of 70 psig. B. RACS cooling water return from compressor temperature of 137" F. C. RPV level of -30". D. Emergency Instrument Air Receiver pressure of 110 psi g.

K&A Rating: K&A Statement Justification:

300000 K1.04 (2.8/2.9)

Knowledge of the connections and I or cause effect relationships between INSTRUMENT AIR SYSTEM and the following:

Cooling water to compressor A Incorrect but plausible:

Low Instrument Air Header pressure 70 psig. Isolates the service air header. B. Correct:

RACS cooling water return from compressor high temperature of 135° F will cause auto trip of emergency instrument air compressor.

C. Incorrect but plausible:

Low RPV level of -129" indicated.

A LOCA will trip open the 1 E feeder breaker.

There is no override.

D. Incorrect but plausible:

High Pressure in the Emergency Instrument Air Receiver 70 psig. Actual trip occurs at 120 psig.

References:

HC.OP-SO.KB-0001 Student Ref: NONE Learning Objective:

INSAIRE003 Question Source: Modified Bank# 34761 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

10CFR: CFR 41.2-41.9 Comments:

QUESTION 54 Which ONE of the following is an indication that a control rod has become uncoupled and stuck? A. A red full-out indication on the full core display.

B. NO response on nuclear instrumentation when the control rod is being withdrawn from fully inserted.

C. A"--" indication on the Four Rod Display for that control rod when a withdraw signal is applied at position

48. D. Control rod position indication does NOT change when a withdraw signal is applied to a control rod at position
48.

K&A Rating: 201003 K4.02 (3.8) K&A Statement:

Knowledge of CONTROL ROD AND DRIVE MECHANISM design feature(s) and/or interlocks which provide for the following K4.02 Detection of an uncoupled rod Justification:

A. Incorrect but plausible:

A red full out indication is indication that the rod is overtraveled and uncoupled but does not mean that the rod is stuck. B. Correct:

No response on Nl's during rod withdrawal is an indication the rod is not moving. The rod position indication (Reed Switches) are in the mechanism which is moving, however the control rod is uncoupled and stuck. C. Incorrect but plausible:

A coupled rod during coupling checks will indicate-and is indication that the rod is coupled.

D. Incorrect but plausible:

A coupled rod at 48 will not move beyond 48 and is indication that the rod is coupled.

References:

NA Learning Objective:

000005E006 Question Source: HC Bank 56330 Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

10CFR55:

CFR41.74 Comments:

Student Ref: NONE QUESTION 55 The Reactor is operating normally at 90% power when the CP501 Circ Water Pump trips.

  • AP501, BP501, and DP501 Circ Water Pumps remain in service
  • Condenser pressure was steady at 5.0" HgA before the CP501 pump tripped
  • RECIRC RUNBK BYPASS switch is in the IN (ARMED) position.

Which ONE of the following describes the response of the Reactor Recirculation System? A. The MG Set Drive Motor Breaker will trip. B. A full recirc pump runback to 30% speed will occur. C. An intermediate recirc pump runback to 45% speed will occur. D. The Recirculation System will maintain recirc pumps at their current speed.

K&A Rating: 202002 2.1. 7 ( 4.4/4. 7) K&A Statement:

Recirculation Flow Control 2.1. 7 Ability to evaluate plant pertormance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Justification:

A. Incorrect but plausible:

Plausible if the examinee mistakes the MG Set Drive Motor Breaker trip conditions with the Recirc Runback conditions.

B. Incorrect but plausible:

Plausible if the examinee believes that a full Recirc Pump Runback would occur on the loss of 1 Circ Water Pump with 4 running.

(A full runback to 30% occurs only on the loss of 1 Circ Water Pump with 3 running AND Condenser pressure

>5.8" HgA.) C. Correct:

An intermediate Recirc Pump runback to 45% speed will occur on the loss of 1 Circ Water Pump with 4 Circ Water Pumps running AND Condenser pressure

> 4.5" HgA. D. Incorrect but plausible:

Plausible if the examinee believes that the plant can take the loss of 1 Circ Water pump with 4 running without a Recirc Runback.

Condenser pressure being given as 5.0" HgA in the stem may lead the examinee to believe that the loss of 1 Circ water Pump intermediate runback will not be active.

References:

HC.OP-SO.BB-0002 NOH01 RECCON SO-DA-0001 step 5.14.2 Learning Objective:

R11, R15 Question Source: New Question History:

N/ A Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/43.5/45.12/45.13 Comments:

QUESTION 56 Given the following:

  • OPCON 1 at 100% power.

Subsequently, the PILOT SCRAM VALVE TRIP ACTUATOR LOGIC Group 4B SOLENOIDS LOGIC B NORMAL light extinguishes at 1 OC651 C. This condition was NOT caused by the weekly test. Which ONE of the following describes the indications on the full core display based ONLY on the INITIAL RPS response?

A. Control Rod Groups 1, 2 & 3 RED lamps are lit, and Control Rod Group 4 GREEN, BLUE, & AMBER lamps are lit. B. All Control Rods have RED lamps lit C. All Control Rods have GREEN, BLUE, & AMBER lamps lit. D. Control Rod Group 1 & 4 GREEN, BLUE, & AMBER lamps are lit, and Control Rod Groups 2 & 3 have RED lamps lit.

K&A Rating: K&A Statement:

Justification:

214000 A4.02 (4.0/4.1)

Ability to manually operate and/or monitor in the control room: Control rod position A. Correct:

The Operator's action causes an "A" side half scram and all "A" side scram pilot solenoids de-energize.

No rod motion occurs. Now, the Group 48 solenoid logic status light extinguishes indicating a loss of power to "8' side scram pilot solenoids for all control rods in Group 4 only. So, the "A" and "8" scram pilot solenoid valves are both energized on all control rods in Group 4, and this vents air from all scram valves for control rods in Group 4 ONLY. The Full Core Display will show RED (full out indication) for rods in Groups 1 ,2, & 3, and Group 4 control rods will display GREEN (full in), BLUE (scram valves open), and AMBER (rod drift). B. Incorrect but plausible:

Plausible if the applicant determines that sufficient logic did not complete to cause any of the rods to scram. RED indicates full out position.

Group 4 control rods went full in, so this answer is incorrect.

C. Incorrect but plausible:

Plausible if the applicant determines that sufficient logic is made up to cause full core scram. D. Incorrect but plausible:

Plausible if the applicant determines that the control rods associated with Groups 1 & 4 scrammed (A 1 manual pushbutton was depressed and the Group 48 logic power failure, i.e. RPS A & B de-energized on Groups 1 & 4).

References:

HC.OP-STSF-0003, "RPS Manual Scram Test-Weekly" Student Ref: NONE NOH01 RPSOOC-09 NOH04MANCONC-07 Learning Objective:

MANCONE002 Question Source: NRC HC 2007 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR: CFR 41.7/45.7 Comments:

QUESTION 57 Given the following conditions:

  • The plant is at full power
  • A TIP system trace is in progress
  • An instrument technician error causes actuation of the NSSSS Channel A manual isolation switch
  • The inserted TIP detector becomes stuck in the core Which ONE of the following describes the TIP system response to this condition?

A. The TIP Shear Valve automatically fires to cut the detector cable and seal the guide tube. B. The TIP Guide Tube Ball Valve automatically closes, cutting the detector cable and sealing the guide tube. C. The inserted TIP detector will receive a signal to automatically withdraw to the "in-shield" position.

The TIP Guide Tube Ball Valve automatically closes. D. The inserted TIP detector will receive a signal to automatically withdraw to the "in-shield" position.

The TIP Guide Tube Ball Valve will NOT automatically close.

K&A Rating: 215001A 1.02 (2.5/2.4)

K&A Statement:

Ability to predict and/or monitor changes in parameters associated with operating the TRAVERSING IN-CORE PROBE controls including A1.02 Detector Position Justification:

A. Incorrect but plausible:

The Shear Valves must be manually initiated.

B. Incorrect but plausible:

The Ball Valve will not close with the cable inside the valve. C. Incorrect but plausible:

the TIP Guide Tube Ball Valve will not automatically close due to the detector being stuck and not at the "in-shield" limit, the ball valve will not close with the cable inside the valve. D. Correct:

The TIP detector not at its "in-shield" position will receive a signal to automatically withdraw to the "in-shield"

position, and the TIP Guide Tube Ball Valve will not automatically close due to the detector being stuck and not at the "in-shield" limit, the ball valve will not close with the cable inside the valve.

References:

NOHOTIPS00-02 HC.OP-SO.SM-0001 HC.RE-SO.SE-0001 Learning Objective:

TIPSOOE006 Question Source: Bank Question History:

HC # 328 Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/45.5 Comments:

QUESTION 58 The crew is responding to a Station Blackout.

The following conditions exist:

  • A&C EDGs failed to energize their respective buses
  • Lowest RPV level was+ 5", and is currently

+1 0" and steady using "B" Loop of Core Spray

  • RPV pressure is 200 to 400 psig and being controlled by SRV's * "B" RHR has been placed in Suppression Pool cooling Subsequently, Drywell pressure reaches 1.72 psig due to the loss of containment cooling.

What actions will place "B" RHR in Suppression Pool Cooling and Spray? A Press AUTO OPEN OVRDs for BC-HV-F024B and BC-HV-F027B, and open the valves. B. Press AUTO OPEN OVRD for BC-HV-F017B, and close it; then press AUTO CL OVRDs for BC-HV-F024B and BC-HV-F027B, and open the valves. C. Press AUTO OPEN OVRDs for BC-HV-F0248 and BC-HV-F027B, and verify they stroke open. D. Press AUTO OPEN OVRD for BC-HV-F017B, and close it; then press AUTO OPEN OVRDs for BC-HV-F024B and BC-HV-F027B, and verify they stroke open.

K&A Rating: 219000 K3.01 (3.9/4.1)

K&A Statement:

Knowledge of the effect that a loss or malfunction of the RHRILPCI:

TORUS/SUPPRESSION POOL COOLING MODE: K3.01: Suppression pool temperature control Justification:

A. Incorrect but plausible:

AUTO CLOSE OVRD must be depressed to allow re-opening of BC-HV-F024B and BC-HV-F027B.

The BC-HV-F017B must also be closed. B. Correct:

lAW HC.OP-SO.BC-0001-Step 3.3.7 The following overrides are provided during a LPCI Initiation AND the associated Logic Initiation Reset must be pressed to restore to Auto position:

  • RHR PUMP A(B,C,D)P202 (STOP PB should be pressed).
  • BC-HV-F017A(B,C,O),

RHR LOOP A(B,C,D)

LPCIINJ MOV (AUTO OP OVRD PB should be pressed).

  • BC-HV-F024A(B),

RHR LOOP A(B) TEST RET MOV (AUTO CL OVRD PB should be pressed).

  • BC-HV-F027A(B),

RHR LOOP A(B) SUPP CHAMBER SPRAY HDR ISLN MOV (AUTO CL OVRD PB should be pressed)

C. Incorrect but plausible:

AUTO CLOSE OVRD must be depressed to allow re-opening of BC-HV-F024B and BC-HV-F027B.

They will not auto open. Also, BC-HV-F017B must be closed. D. Incorrect but plausible:

AUTO CLOSE OVRD must be depressed to allow re-opening of BC-HV-F024B and BC-HV-F027B.

References:

HC.OP-SO.BC-0001-Step 3.3.7 Student Ref: NONE Learning Objective:

Question Source: New Question History:

New Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7/45.4 Comments:

QUESTION 59 A function of the Suppression Chamber-to-Drywell Vacuum Breakers is to ... A. prevent low Suppression Chamber pressure from damaging the drywell after a LOCA. B. allow hot liquid to enter the Suppression Chamber during the blowdown phase of a LOCA. C. prevent a negative pressure from occurring in the Suppression Chamber following a LOCA. D. allow non-condensable gasses to reenter the drywell to prevent exceeding the drywell external pressure limit.

K&A Rating:

293001 K5.01 (3.1/3.3)

K&A Statement:

Knowledge of the operational implications of the following concepts as they apply to PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES:

K5.01 Vacuum breaker/relief operation Justification:

A. Incorrect but plausible:

Suppression chamber pressure is relieved back to the drywell after steam condensation to protect the drywell from implosion.

B. Incorrect but plausible:

Hot liquid from the blowdown will gravity drain from the drywell to the suppression chamber through the downcomers.

C. Incorrect but plausible:

The reactor building to suppression pool vacuum breakers prevent a negative pressure from occurring in the suppression chamber following a LOCA not the suppression chamber to drywell vacuum breakers.

D. Correct:

The suppression chamber to drywell vacuum breakers purpose is to prevent exceeding design external pressures of the drywell.

References:

NOH01 PRICONC Student Ref: NONE Learning Objective:

PRICONE003 Question Source: Bank Question History:

HC Bank# 79 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

1 OCFR55: CFR 41.5/45.3 Comments:

QUESTION 60 The plant was operating at 85% power when a high OW pressure condition occured.

Both pumps have been placed in service manually.

Ten minutes later, conditions are as follows:

  • "A" RHR has been placed in Suppression pool cooling and spray
  • Torus pressure is 9.5 psig and continues to slowly rise
  • The CRS has directed "B" loop of RHR be placed in Drywell Spray
  • The "B" RHR Loop Inboard Spray Valve (HV-F021 B) will open but the Outboard Drywell Spray Valve (HV-F016B) remains closed (electrical power is available)

Which ONE of the following conditions is most likely preventing OW Spray Valve HV-F016B from opening?

A The HX Bypass valve (HV-F048A) is not fully closed. B. The "B" channel Drywell pressure instrument is failed low. C. The RHR Full Flow test Valve (HV-F024B) is not fully closed. D. The "B" channel RPV water level indicates above -129 inches.

K&A Rating: 226001 K3.02 (2.9/2.9)

K&A Statement:

Knowledge of RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE design feature(s) and/or interlocks which provide for the following:

K4.12 Prevention of Inadvertent Containment Spray Activation Justification:

A. Incorrect but plausible:

There is no interlock between these valves. B. Correct:

A high drywell pressure signal is required to open F0168. (ALL of the following conditions must be satisfied to allow both F0168 RHR LOOP 8 OUTS CONT SPRAY ISLN MOV AND F021 8 RHR LOOP 8 INBD CONT SPRAY MOVs to open simultaneously:

LPCI initiation signal present, High drywell pressure signal present, and F-178 RHR LOOP 8 LPCIINJ MOV CLOSED.)

C. Incorrect but plausible:

There is no interlock between these valves. D. Incorrect but plausible:

Manual initiation will seal in the LPCI initiation signal.

References:

HC.OP-SO.BC-0001 Student Ref: NONE RHRSYSE008 Learning Objective:

RHRSYSE008 Question Source: Bank Question History:

HC Bank# 329 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7 Comments:

QUESTION 61 Fuel handling activities are being conducted in the Spent Fuel Pool in preparation for the upcoming refuel outage. The following conditions occur:

  • A Spent fuel bundle was physically damaged during a moving evolution to a new pool location during the lift
  • The damaged fuel bundle is currently over an area of empty fuel storage racks
  • The Refuel Floor Exhaust Radiation Monitor is reading 2.3 x 10-31JCilcc
  • The Refuel Floor Area Radiation Monitor is reading 14 mR!Hr lAW HC.OP-AB.CONT-0005(0)

"Irradiated Fuel Damage",

which ONE of the following immediate actions is required?

A. Place spent fuel bundle in its new designated storage location.

B. Place spent fuel bundle in the nearest storage location.

C. Start Filtration, Recirculation and Ventilation System (FRVS) fans. D. Suspend handling of the damaged fuel bundle.

K&A Rating: K&A Statement:

Justification:

234000 A 1.02 (3.3/3.8)

Ability to predict and/or monitor changes in parameters associated with operating the FUEL HANDLING EQUIPMENT controls including:

Refuel floor radiation levels/ airborne levels A. Incorrect but plausible:

Plausible if the applicant determines that placing spent fuel bundle in its new designated storage location is an immediate safe action. B. Incorrect but plausible:

Plausible if the applicant determines that placing spent fuel bundle in the nearest storage location is the safest immediate action based on stem condition indicating fuel bundle over empty storage location.

C. Incorrect but plausible:

Plausible due to Refuel Floor Exhaust Radiation Monitor reading > 2.0 x 10-3 !JCi/cc,

however, Filtration, Recirculation and Ventilation System (FRVS) fans start automatically.

D. Correct:

HC.OP-AB.CONT-0005(0)

"Irradiated Fuel Damage" directs suspending of the handling of Irradiated Fuel/Components based on entry conditions due to refuel floor radiation alarm.

References:

HC.OP-AB.CONT-0005(0)

"Irradiated Fuel Damage" Student Ref: NONE Learning Objective:

Question Source: Question History:

Cognitive Level: 10CFR: Comments:

New None Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

CFR 41 .5/45.5 Changed from A4.01 to A1.02, due to Q 56 change to balance sample plan.

QUESTION 62 While operating at 35% power, the following indications are received:

  • Alarm D3-C5 TURBINE GENERATOR VIB HI
  • GRIDS point A2529 MAIN TURBINE BEARING 11 VIB indicates 9 mils and steady lAW HC.OP-AB.BOP-0002

'MAIN TURBINE',

which ONE of the following is required?

A. With system operator concurrence, adjust MVAR loading.

B. Lock the Mode Switch in SHUTDOWN and then immediately trip the Main Turbine.

C. Monitor GRIDS or Main Control Room System 1 Computer for further bearing degradation.

D. Reduce Recirc Pump speed to minimum, then lock the Mode Switch in SHUTDOWN and immediately trip the Main Turbine.

K&A Rating: 241 OOOA 1.23 (2.8/2.8)

K&A Statement:

Ability to predict and/or monitor changes in parameters associated with operating the REACTOR/TURBINE PRESSURE REGULATING SYSTEM controls including:

A 1.23 Main Turbine Vibration Justification:

A. Incorrect but plausible:

Plausible if the operator does not recall that the bearing vibration magnitude does not meet the immediate action criteria (Bearing 1-10 must be >= 12 mils for tripping the reactor and turbine).

Section B.1 of HC.OP-AB.BOP-0002 directs the operator to adjust MVAR loading.

B. Correct:

lAW HC.OP-AB.BOP-002 Immediate

Actions, if bearing 11-12 Vibration>=

8 mils@ 1800rpm AND reactor power is>+ 24%, then lock the mode switch in shutdown and immediately trip the main turbine.

C. Incorrect but plausible:

Plausible if the operator does not recall that the bearing vibration magnitude does not meet the immediate action criteria (Bearing 1-10 must be >= 12 mils for tripping the reactor and turbine).

D. Incorrect but plausible:

Plausible if the operator does not recall that the bearing vibration magnitude does not meet the immediate action criteria.

The HC.OP-AB.BOP-002 Retainment Override states that if bearing 11-12 Vibration

>= 7 mils @ 1800rpm AND reactor power is>+ 24%, then reduce recirc pump speed to minimum and then lock the mode switch in shutdown and immediately trip the main turbine.

The immediate action criteria

(>=8mils, no recirc pump speed reduction) overrides the retainment override.

References:

HC.OP-AB.

BOP-0002 HC.OP-AR.ZZ-0014 Learning Objective:

MNTURBE019 Question Source: New Question History:

N/A Student Ref: NONE Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/45.5 Comments:

QUESTION 63 Given the following conditions:

  • The Reactor failed to SCRAM when the Mode Switch was placed in Shutdown
  • Reactor power is 14% and Reactor pressure is 11 05 psig
  • The 'A' RFP speed has slowed to approximately 2500 rpm and remains steady
  • The RFP TURBINE AUTO XFR TO MANUAL (B3-F3) annunciator is in alarm Which ONE of the following describes the reason for the 'A' RFP speed reduction?

The 'A' RFP is responding to ... A. a control signal failure.

B. a gross failure of a Main Steam Flow transmitter.

C. a Redundant Reactivity Control System runback.

D. the Setpoint Setdown feature of Digital Feedwater Control.

K&A Rating: 25900*1A3.11 (3.2/3.7)

K&A Statement:

Ability to monitor automatic operations of the REACTOR FEEDWATER SYSTEM including:

A3.11 Reactor feed pump runbacks:

Plant-Specific Justification:

A Incorrect but plausible:

Speed will remain essentially constant with a Control Signal failure.

B. Incorrect but plausible:

A single steam flow transmitter failure changes control to single element.

C. Correct:

With RPV pressure at 11 05psig, RRCS will swap the RFP to manual, initiate a RFP Runback signal and lower speed to 2500rom (RRCS Setpoint:

>1071psig w/ .4% pwr for 25 seconds)

D. Incorrect but plausible:

Calling for a lower level RFP speed would not stabilize pump speed at 2500 rpm and a transfer to manual control would not occur ..

References:

HC.OP-SO.SA-0001 Student Ref: NONE Learning Objective:

FWCONTE011 Question Source: Bank Question History:

HC Bank #318 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7/45.7 Comments:

QUESTION 64 A plant startup is in progress when the 'A' RPS Motor-Generator Voltage Regulator fails causing generator output voltage to decrease to approximately 1 OOVAC. Which ONE of the following describes the effect of this condition on the Main Steam Line (MSL) Radiation Monitors?

A. The reduced voltage causes a DOWNSCALE trip of MSL Radiation Monitors RE-N006A and RE-N0068.

B. The reduced voltage causes a DOWNSCALE trip of MSL Radiation Monitors RE-N006A and RE-N006C.

C. Power is lost to MSL Radiation Monitors RE-N006A and RE-N006C, resulting in an INOP trip. D. Power is lost to MSL Radiation Monitors RE-N006A and RE-N0068, resulting in an INOP trip.

K&A Rating: 272000K6.01 (3.0/4.2)

K&A Statement:

Knowledge of the effect that a loss or malfunction of the following will have on the RADIATION MONITORING SYSTEM K6.01 Reactor protection system Justification:

A. Incorrect but plausible:

Any voltage reduction would be momentary due to the UV trip of the EPA breakers; an INOP trip occurs on a loss of power to the MSL Rad Monitors.

This would result in an INOP trip of MSL Rad Monitors RE-N006A

& C since they are powered from RPS Bus 'A' B. Incorrect but plausible:

Any voltage reduction would be momentary due to the UV trip of the EPA breakers; C. Correct:

When the 'A' RPS MG output is less than 108 VAC, the EPA breakers on the MG output to the 'A' RPS Bus trip on undervoltage, causing a loss of the 'A' RPS bus. This results in an INOP trip of the MSL Rad monitors RE-N006A

& C since they are powered from RPS Bus 'A'. D. Incorrect but plausible:

RE-N006A

& C trip, not A & B. See 'C' for additional explanation.

References:

HC.OP-SO.SB-0001 Student Ref: NONE HC.OP-SO.SP-0001 Learning Objective:

RMSYSOE002 Question Source: Bank Question History:

HC Bank #52 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

1 OCFR55: CFR 41.7/45.7 Comments:

QUESTION 65 Given the following conditions:

  • The Diesel Generator Room Carbon Dioxide Fire protection system is aligned for automatic operation.
  • A valid EDG room high temperature condition has just occurred.

Which ONE of the following describes how the Diesel Generator Room Carbon Dioxide Fire protection system responds?

A A discharge alarm occurs. C02 with a wintergreen scent is discharged into the room immediately.

B. A pre-discharge alarm is activated.

After a time delay, C02 with a wintergreen scent is discharged into the room. C. A pre-discharge alarm is activated.

No C02 is discharged into the room until a valid smoke detector alarm is received.

D. A pre-discharge alarm is activated and a wintergreen scent is discharged into the room. After a time delay, C02 is discharged into the room.

K&A Rating: 286000A4.01 (3.3/3 2) K&A Statement:

Ability to manually operate and/or monitor from the control room A4.01 System alarms and indicating lights. Justification:

A Incorrect but plausible:

A pre-discharge alarm is activated.

After a time delay, C02 with a wintergreen scent is discharged into the room, not immediately.

B. Correct:

lAW HC.OP-SO.KC-0002, Section 3.2.3 and 5.2, a pre-discharge alarm is activated.

After a time delay, C02 with a wintergreen scent is discharged into the room. (The wintergreen scent comes from the odorizer bottle which is activated by the C02 flow.) C. Incorrect but plausible:

A pre-discharge alarm is activated.

After a time delay, C02 with a wintergreen scent is discharged into the room. The EDG room alarm actuates on Hi Temp, not a smoke detector alarm. D. Incorrect but plausible:

A pre-discharge alarm is activated.

After a time delay, C02 with a wintergreen scent is discharged into the room. The wintergreen scent comes in with the C02, not before.

References:

HC.OP-SO.KC-0002 Student Ref: NONE Learning Objective:

FIRPROE015 Question Source: Bank Question History:

HC Bank# 229 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7/45.5 to 45.8 Comments:

QUESTION 66 Which ONE of the following situations complies with OP-AA-103-1 01, CONTROL ROOM ACCESS CONTROL, and OP-AA-101-111, ROLES AND RESPONSIBILITIES OF ON-SHIFT PERSONNEL?

A. During a Reactor Startup the NCO "At the Controls" has ultimate responsibility to limit access to the control room to only Licensed Personnel.

B. In OPCON 1, 2, or 3 a minimum of 2 NCOs will be "At the Controls".

It may be lowered to 1 NCO if properly relieved by the CRS. C. In OPCON 1 with only 2 NCOs assigned to the control room, when the PO is in the computer room retrieving a P-1, the "At the Controls" NCO is allowed to initiate normal drywell makeup on the 1 OC650 panel. D. In OPCON 1 with only 2 NCOs in the control room, the "At the Controls" NCO is allowed to check a reading on the SRV temperature recorder on the 1 OC650C panel.

K&A Rating: 2.1.1 (3.8/4.2)

K&A Statement:

2.1.1 Knowledge of conduct of operations requirements.

Justification:

A. Incorrect but plausible:

OP-AA-1 03-101 lists personnel who are allowed in the control room for a startup including QA and the NRC. B. Incorrect but plausible:

OP-AA-103-101 states that the on duty SM and CRS shall remain in the Protected Area unless properly relieved.

C. Incorrect but plausible:

lAW OP-AA-1 01-111, section 4.9, it is expected that three duty Licensed Operators will be in the Control Room boundary at all times in OPERATIONAL CONDITIONS 1, 2 and 3, except that one NCO may go to the Computer Room to retrieve the printouts from the PPC (Plant Process Computer) or CMS (Core Monitoring System) as long as the other NCO is "at the controls" in the Inner Horseshoe and is within view of the control panels. By initiating a normal drywell Mkeup, this NCO is NOT within view of the panels and therefore this is NOT an allowed activity.

D. Correct:

OP-AA-1 01-111, Section 4.9 states that "At the Controls" NCO may briefly enter the "Outer Horseshoe" for activities such as checking an instrument reading or responding to an alarm. (the SRV temperature recorder on the 1 OC650C panel is in the outer horseshoe).

E.

References:

OP-AA-1 01-111, Section 4.9 Student Ref: NONE Learning Objective:

ADMPR05CE008 Question Source: Bank Question History:

HC Bank# 123 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

1 OCFR55: CFR 41.10/45.13 Comments:

QUESTION 67 While operating with 300 Mvars overexcited (Lagging),

a main generator hydrogen leak occurs.

  • Hydrogen pressure has stabilized at 65 psig. Which ONE of the following is the main generator power output limit, based on these conditions?

A. 1000 Mw B. 1150 Mw C. 1250 Mw D. 1325 Mw (Note: See attached graph)

HC.OP-SO.MA-0001 (Q) ATTACHMENT 1 POWER CHANGES DURING OPERATION GENERATOR REACTIVE CAPABILITY AND MINIMUM EXCITATION LIMIT CURVE GENERATOR REACTIVE CAPABILJTY CURVE AT8 d POLE !Jil!OO KVA lBCG RPH VOLTS 0.94 PF U sen 75 PSTG H\DnOGEN PRESSURf 530 VOLTS 1500 120(1 lOGO aoo 400 "' <{ _J (,'1 .20() 0: > 0 _J -I) 0 GO pr "/b

. ' 1 o 7o 60 PSLG )------. v O.DO PDI;; v vu.f]5 O,SC* ----I 1--Jo r-psr [?:S;, / ;< /< //""--...

>-. / '-1'-' / ">< '""

/I L / ... ,"' _1_ r----

ex

' V::'/ ------K ,_ .

I [,;' 20*8 0 6 C* 8(lQ 1\JC ;J 11200 14()1) 150 I I IV 1000

...... r--Ml11lmum Ewr::tt.:.tlott Limit curve

' ""' *-----.___

t9-I ,_ \ ', _,; "--.. .....-

e: k-?5: -----, c 9'5 -:::-......-:

=---::::::,.

C.<JO ' "* ' G.E5 l \\ 0. tliJ 0. EG

-oorJ c r,.c;]qo,rF-<

-800 -1 DOfJ Page43 of47 Rev. 56 K&A Rating: G 2.1.25 (3.9/4.2)

K&A Statement:

Ability to interpret reference materials, such as graphs, cuNes, tables, etc. Justification:

A. Incorrect but plausible:

Based on 400 Mvar line and 45 PSIG pressure cuNe. B. Incorrect but plausible:

Based on 300 Mvar line and 45 PSIG pressure cuNe. C. Correct : based on 300 Mvar line intersecting the 65 PSIG cuNe. D. Incorrect but plausible:

based on 300 Mvar line and 75 PSIG pressure cuNe

References:

HC.OP-SO.MA-0001 attachment

  1. 1 Student Ref: Att #1 Learning Objective:

MNGENOE01 0 Question Source: HC Bank Q#214 Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/

Analysis:

X 10CFR: CFR 41.10/43.5/45.12 Comments:

QUESTION 68 lAW TS 2.1.1, Thermal Power shall not exceed 24% of rated thermal power with the Reactor Vessel Steam Dome Pressure less than _(1)_ psig OR Core Flow less than _(2)_% of Rated Flow. w m A. 756 15 B. 756 10 C. 785 15 D. 785 10 K&A Rating: 2.2.22 (4.0/4.7)

K&A Statement:

2.2.22 Knowledge of limiting conditions for operations and safety limits. Justification:

A. Incorrect but plausible:

Pressure too low and flow too high. B. Incorrect but plausible:

Pressure too low. C. Incorrect but plausible:

Flow too high. D. Correct:

lAW Tech Spec 2.1.1, thermal power shall not exceed 24% of rated thermal power with the reactor vessel steam dome pressure less than 785psig or core flow less than 10% or rated flow.

References:

TS 2.1.1 Student Ref: NONE Learning Objective:

TECSPECE001 Question Source: Bank Question History:

HC Bank #11 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

1 OCFR55: CFR 41.5/43.2/45.2 Comments:

QUESTION 69 Given the following conditions:

  • A Reactor Startup is in progress.
  • The Mode Switch in Startup/Hot Standby.

Subsequently, SLCS TANK TROUBLE (C1-E1) annunciator alarms

  • The EO isolates a severely leaking SLC storage tank drain line.
  • 4450 gallons remains in the storage tank.
  • A Chemistry sample shows current boron concentration is 13.7%. Based on this information, the SLC system storage tank sodium pentaborate solution volume/concentration Technical Specification requirement

... A. is being met. B. needs to have only water added to bring it into specification.

C. needs to have only boron added to bring it into specification.

D. needs to have both boron AND water added to bring it into specification.

K&A Rating: 2.2.42 (3.9/4.6)

K&A Statement:

2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Justification:

A. Incorrect but plausible:

Total volume is below limits. The question stem indicates that the plant is in Mode 2 (mode switch in Startup/Hot Standby),

therefore the applicability statement applies (3.1.5 applies in Modes 1 and 2). Plausible if the applicant believes that the Tech Spec does not apply given that a Reactor startup is in progress.

B. Incorrect but plausible:

Adding water only will dilute the concentration below 13.6%. C. Incorrect but plausible:

Adding boron only will not bring the total volume above limits. D. Correct:

TS 3.1.5 requires a minimum volume of -4746.5 gallons at a concentration of 13.7%. (4782-4640=142 gallons between 13.6% and 14.0%. Subtract%

of this to estimate minimum requirement at 13.7% due to linear slope of line.) Water must be added to bring level up to the minimum.

(-296 gallons.)

Starting at 13.7%, this would dilute the concentration down to -12.8%, less than the minimum required

( 4450/4 7 46x13. 7%). Therefore, boron must also be added to bring the concentration volume within acceptable limits.

References:

TS 3.1.5 Student Ref: Tech Spec 3.15 Learning Objective:

SLCSYSE027 Question Source: Modified Question History:

Modified from HC Bank #115 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

41.7 I 41.10 I 43.2 I 43.3 I 45.3 Comments:

QUESTION 70 A vent valve is to be worked on, and it is the only vent valve capable of removing stored energy in the associated system. lAW OP-AA-1 09-115, "Safety Tagging Operations"

... A. a White Caution Tag (WCT) is applied to the vent valve and the work is conducted on the valve with the WCT attached.

B. the Clearing Agent and the Tagging Authority agree that the valve can be VER, and the work is conducted on the valve. C. a Worker's Blocking Tag (WBT) is applied to the vent valve and the work is conducted on the valve with the WBT attached D. an Information Tag (IN F) is applied to the vent valve and the work is conducted on the valve with the INF attached.

Question 70 K&A Rating: 2.2.13 (4.1) K&A Statement:

Knowledge of tagging and clearance procedures.

Justification:

A. Incorrect but plausible:

Not in accordance with the procedure.

B. Correct:

Per component and tagging rules this is the correct process to control the valve. C. Incorrect but plausible:

Not in accordance with the procedure.

D. Incorrect but plausible:

Not in accordance with the procedure

References:

0 P-AA-1 09-115 Attachment 2 Component Tagging Rules Learning Objective:

NA0015E004 Question Source: HC Bank Q#62 Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

1 OCFR55: CFR 41.10/45.13 Comments:

Student Ref: NONE QUESTION 71 Given:

  • The Refueling SRO reports he is ready to commence core alterations.
  • Radiation Protection has been notified and are evacuating the upper regions of the Drywell.

lAW RP-AA-403,

'Administration of the Radiation Work Permit Program',

which ONE of the following actions/controls is a requirement if rad levels in the upper regions of the Drywell are determined to be 1100 mrem/hr?

A. The area is roped off, conspicuously posted, and a flashing light provided as a warning device. B. Qualified personnel will provide direct or remote (e.g., closed circuit TV) roving surveillance and the area is conspicuously posted. C. The area is conspicuously posted and qualified personnel provide direct or remote (e.g., closed circuit TV) continuous surveillance.

D. Qualified personnel will provide direct or remote (e.g., closed circuit TV) continuous surveillance ONLY.

K&A Rating: 2.3.12 (3.2/3.7)

K&A Statement:

2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning

filters, etc. Justification:

A. Correct:

Correct lAW RP-AA-403 B. Incorrect but plausible:

Requires continuous surveillance C. Incorrect but plausible:

Combination of incorrect statements.

D. Incorrect but plausible:

This is used in lieu of stay time.

References:

RP-AA-403 Student Ref: NONE RP-AA-406 Learning Objective:

NOH04ADM024E Question Source: Bank Question History:

HC Bank #26 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

10CFR55:

CFR 41.12,45.9,45.10 Comments:

Learning objective NOH04ADM024-E001 From Memory Describe what the worker is acknowledging when signing a RWP prior to use. lAW RP-AA-403, Administration of the Radiation Work Permit Program QUESTION 72 Given the following conditions:

  • The Reactor core has been operating with one or more known fuel pin leaks
  • A Reactor SCRAM occurred from 1 00% power
  • The Reactor Building HVAC Exhaust Rad monitor is reading 4E-5 1-JCi/mL and steady Which ONE of the following rooms would become the most significant radiological hazard AND what action would reduce the radiological hazard? A Reactor Building Equipment Sump BT266; Enter EOP 103/4. B. Reactor Building Equipment Sump AT266; Enter EOP 103/4. C. Reactor Building Equipment Sump BT266; Reset the Reactor SCRAM. D. Reactor Building Equipment Sump AT266; Reset the Reactor SCRAM.

K&A Rating: 2.2.11 (3.8/4.3)

K&A Statement:

Ability to control radiation releases.

Justification:

A. Incorrect but plausible:

The South Sump Room will experience rising rad levels as described in answer choice 'C'. However, the entry condition for EOP 103/4 is Reactor Building HVAC exhaust rad levels above 5E-5 !JCilml.

The stem indicated that current Reactor Building HVAC exhaust levels are currently reading 4E-5 !JCi/mL.

Therefore, EOP 103/4 entry is NOT required.

B. Incorrect but plausible:

The North Sump Room will NOT experience rising rad levels as described in answer choice 'C'. The entry condition for EOP 103/4 is Reactor Building HVAC exhaust rad levels above 5E-5 !JCi/ml.

The stem indicated that current Reactor Building HVAC exhaust levels are currently reading 4E-5 !JCi/mL.

Therefore, EOP 103/4 entry is NOT required.

C. Correct:

The North and South SCRAM discharge volumes drain through a common line to the Reactor Building Equipment Drain Sump (1 BT266) located in the South Reactor Building Sump Room (54' elevation).

If the drain valves did not close (as stated in the stem), a LOCA would exist discharging into the South Sump Room. The leaking fuel would severely raise rad levels in that room. With drain valves open, resetting the SCRAM will reclose the valves and terminate the LOCA. D. Incorrect but plausible:

With drain valves open, resetting the SCRAM will reclose the valves and terminate the LOCA. However, the North Sump Room will NOT experience risif)g rad levels as described in answer choice 'C'.

References:

EOP 103/4 Student Ref: M-61-1 Sht 1 & 2 Learning Objective:

CRDHYDE01 0 Question Source: Bank Question History:

HC Bank #205 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.11/43.4/45.10 Comments:

This question is RO level. The first part of the two part answer choice can be answered solely by system knowledge.

Direct EOP entry conditions are considered RO level knowledge.

Understanding that resetting the Reactor SCRAM will reduce radiation levels can be answered solely by system knowledge.

QUESTION 73 Concerning the EOPs, which ONE of the following is the correct name for this EOP symbol? A. Decision Step B. Entry Condition C. Contingent Step D. Execute Concurrently Step K&A Rating: 2.4.17 (3.9/4.3)

K&A Statement:

Knowledge of EOP terms and definitions.

Justification:

A. Incorrect but plausible:

Contains a question the answer to which determines the next step. All decision steps require a choice between two opposite possibilities, for example "YES/NO" or "HIGH/LOW".

B. Incorrect but plausible:

Entry conditions or symptoms are generally brief noun phrases that initiate the EOP actions.

The entry conditions are located at the top of each flowchart within entry condition symbols.

C. Correct:

A contingent action step indicates that until the expected condition contained within the step is met, the corresponding action cannot be performed.

Once operators reach a contingent action step, they wait until the expected condition is met before proceeding.

D. Incorrect but plausible:

Requires the operator to enter the designated procedure and perform the stated actions while continuing in the existing flow path. The execute concurrently symbol is used when an operator is to enter another procedure while at the same time continuing in the current flow path. The symbol contains the number of the concurrent procedure

References:

EOP Bases Student Ref: NONE Learning Objective:

INTEOPE001 Question Source: Bank Question History:

HC Bank #78 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

1 OCFR55: CFR 41.10/45.13 Comments:

QUESTION 74 Gross mechanical damage has occurred at the top of the Reactor core. Attempts are being made to determine the full extent of the mechanical damage. Which ONE of the following systems can be used to determine the full extent (axial AND radial location) of the mechanical damage? A. TIP B. SRM C. LPRM D. IRM K&A Rating: K&A Statement:

2.4.3 Ability to identify post-accident instrumentation.

Justification:

A. Correct:

The driving range of the TIP detectors can be used to determine the axial and radial distribution of damage. TIP detector position is read from the TIP Drive Control Units. Axial position above BAF (in inches) can be determined by subtracting the core bottom limit from the detector reading.

Radial location can be determined from Drive Control Unit and Indexer position assignment to LPRM strings.

B. Incorrect but plausible:

SRMs and IRMs axial position can be determined as a distance from the detector full-in or full-out position.

Since there is no definite method of determining that these detectors have stopped moving, except at the full-in and full-out positions, their use is only to determine that full traverse movement is possible.

Therefore, SRMs and IRMs are not useful at defining the extent of (axial) of all damage. They only traverse from below the active fuel to 15" above active fuel centerline.

C. Incorrect but plausible:

LPRMs detectors are not moveable, and therefore are of limited use in determining the extent of damage. D. Incorrect but plausible:

SRMs and IRMs axial position can be determined as a distance from the detector full-in or full-out position.

Since there is no definite method of determining that these detectors have stopped moving, except at the full-in and full-out positions, their use is only to determine that full traverse movement is possible.

Therefore, SRMs and IRMs are not useful at defining the extent of (axial) of all damage. They only traverse from below the active fuel to 15" above active fuel centerline.

References:

NOH01 MCD000-00 Student Ref: NONE Learning Objective:

NMRESPE003 Question Source: Bank Question History:

HC Bank #84 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

1 OCFR55: CFR 41.6/45.4 Comments:

QUESTION 75 Given the following:

  • The plant is operating at 20% Reactor Power
  • The standby EHC pump is C/T out of service
  • EHC pressure is 1600 psig Subsequently, the in-service EHC pump trips.
  • EHC pressure is lowering at 100 psi/second 5 seconds later, which ONE of the following operator action(s) is required lAW AB.OP-AB.BOP-003

'Turbine Hydraulic Pressure'?

A. Ensure the Bypass Valves control Reactor pressure.

B. Ensure the Control Valves fast close when EHC pressure reaches 530 psig. C. Reduce Recirc speed to minimum and then lock the Mode Switch in SHUTDOWN.

D. Lock the Mode Switch in SHUTDOWN.

Recirc speed is NOT required to be reduced to minimum prior to this action.

K&A Rating: 2.4.11 (4.0/4.2)

K&A Statement:

Knowledge of abnormal condition procedures.

Justification:

A. Incorrect but plausible:

BPVs fail shut with no EHC pressure after-1 minute B. Incorrect but plausible:

The control valves fast close when the fast acting solenoids are energized, which is only on a power load unbalance.

C. Incorrect but plausible:

Recirc pump speed is NOT required to be reduced to minimum prior to this action. D. Correct:

The retainment override in HC.OP-AB.BOP-003 directs the mode switch placed in SHUTDOWN if EHC pressure is <1200psig and lowering.

With both pumps tripped, it will be there immediately.

References:

HC.OP-AB.BOP-003 Student Ref: NONE Learning Objective:

ABBOP3E003 Question Source: Bank Question History:

HC Bank #33 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Ana lysis: 1 OCFRSS: CFR 41.10/41.5/43.5/45.13 Comments:

QUESTION 76 Given the following conditions:

  • The plant is in Operational Condition 3 with the Electrical Distribution System aligned in the Normal plant lineup
  • The E3-B4 STA SERVICE TRANSFORMER 1BX501 TRBL alarm is received in the control room
  • A Field Operator subsequently reports that the F-1 FAULT PRESSURE alarm is lit on the 1 BX501 local alarm panel.
  • No other actions have been taken Which ONE of the following is the status of power for the vital buses and what is required by Technical Specifications (TS)? The diesel generators for the affected ESF buses ... A. will be supplying bus power. Verify correct breaker alignments and indicated power availability for all channels of the 1 E distribution system. B. will NOT be supplying bus power. Verify correct breaker alignments and indicated power availability for all channels of the 1 E distribution system. C. will be supplying bus power. Verify correct breaker alignments and indicated power availability for ONLY the B and D channels of the 1 E distribution systems.

D. will NOT be supplying bus power. Verify correct breaker alignments and indicated power availability for ONLY the B and D channels of the 1 E distribution systems.

K&A Rating: 295003 2.4.31 (4.2/4 1) K&A Statement:

Partial or complete loss of AC: 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

Justification:

A. Incorrect but plausible:

Power from 1AX501 is available.

13kV bkrs BS 2-3 and BS 1-2 trip open and Bus section 2 is de-eneergized.

The bus infeed bkr swaps to the 1AX501 feed and the loads remain energized.

The EDGs do NOT start because one infeed is always available.

B. Correct:

Power from 1AX501 is still available, only the 1 BX501 xfmr is OOS as indicated by two alarms given in stem. 13kV bkrs BS 2-3 and BS 1-2 trip open and Bus section 2 is de-eneergized.

The bus infeed bkr swaps to the 1AX501 feed and the loads remain energized.

The EDGs do not start because one infeed is always available.

lAW TS 3.8.1.1 (applicable in Modes 1, 2, and 3) with one offsite circuit unavailable (1 BX501 xfmr OOS), demonstrate operability of the remaining AC sources by performing SR 4.8.1.1.

1.a within 1 hr and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

SR 4.8.1.1.1.a requires that each of the required independent circuits shall be determined operable by verifying correct breaker alignments and indicated power availability.

C. Incorrect but plausible:

Power from 1AX501 is available.

The bus infeed bkr swaps to the 1AX501 feed and the loads remain energized and the EDGs do NOT start. However, TS 3.8.1. 1 requires that SR 4.8.1.1.1.a be performed within 1 hr. SR 4.8.1.1.1

.a states that each of the required independent circuits be determined operable by verifying correct breaker alignments and indicated power availability.

The SR checks all 4 channels, not just B and D. D. Incorrect but plausible:

TS 3.8.1.1 requires that SR 4.8.1.1.1.a be performed within 1 hr. SR 4.8.1.1.1.a states that each of the required independent circuits be determined operable by verifying correct breaker alignments and indicated power availability.

The SR checks all 4 channels, not just B and D. SRO Only Justification:

This question is SRO only as it requires assessing facility operating limitations in the Tech Specs and their bases, cannot be answered solely by knowing <= 1 hr TS actions or above the line information, cannot be answered by knowing TS safety limits, and involves application of TS required actions and surveillance requirements.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

TS 3.8.1.1.a DWG E-0001 Learning Objective:

1 EACOOE015 Question Source: Bank Student Ref: None Question History:

HC Bank #446 (stem modified

slightly, but still counts as Bank) Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/45.3 Comments:

QUESTION 77 Hope Creek is at 20% power following a startup from a refueling outage when the plant scrams. You have entered HC.OP-AB.ZZ-0000, Reactor Scram and note the following plant conditions:

  • RPV Level-(+33" stable)

-1 000 psig stable

  • Mode Switch-Locked in Shutdown position
  • All Control Rods fully inserted You have reached step S-8: IF Conditions permit THEN RESET the Scram AND INSERT a Half Scram (if required)

Which of the following conditions would REQUIRE you to INSERT a Half Scram? I. APRM channels "A" and "C" INOP II. IRM channels "E" and "F" INOP Ill 1 Reactor Vessel Steam Dome Pressure High Transmitter INOP A. I only B. II only C. I, II only D. I, II, and II K&A Rating: K&A Statement:

Justification:

29500682.4.47 (4.2/4.2)

SCRAM 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

A. Correct:

Per TS 3.3.1.a.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least 1 trip system in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.

For the APRM's in OPCON 3-Minimum OPERABLE Channels per Trip System is 2, therefore if 2 APRM were INOP you would only have 1 in that Trip System OPERABLE and would need to insert a Half-scram B. Incorrect but plausible:

Per TS 3.3.1-1 in OPCON 3 you are only required to have 2 IRM's OPERABLE per trip system, since you have 3 available having 1 INOP still leaves 2 that are OPERABLE and so you would NOT have to insert a Half-Scram C. Incorrect but plausible:

see answer choice 'B' above D. Incorrect but plausible:

see answer choice 'B' above, also Reactor Steam Dome Pressure High transmitter is only required in OPCON 1 or 2, since the stem indicates that the plant is in OPCON 3 this would be N/A SRO Only Justification:

This question is SRO only as it requires assessing facility operating limitations in the Tech Specs and their bases, cannot be answered solely by knowing <= 1 hr TS actions or above the line information, cannot be answered by knowing TS safety limits, and involves application of TS required actions and surveillance requirements.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

TS 3.3.1.a Student Ref: NONE Learning Objective:

Question Source: Bank Question History:

HC Bank #479 (2005 NRC Exam, 2009 NRC Audit) Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.5/45.12 Comments:

Used on 2005 NRC exam and 2009 NRC audit exam with no reference given.

QUESTION 78 Given the following conditions:

  • Reactor power at 100% * "A" SACS Loop is declared inoperable at 0930 today (9/13)
  • Delaware river temperature is 84F and steady
  • No other systems or components are OOS/Inoperable Which ONE of the following correctly identifies the latest times/dates for completion of the FIRST Tech Spec required action? A Immediately B. 1130 9/13 C. 0930 9/16 D. 0930 10/16 K&A Rating: K&A Statement:

Justification:

Partial or Total Loss of CCW 2.2.40 Ability to apply Technical Specifications for a system. A. Incorrect but plausible:

Plausible if the applicant believes that 3.7.1.3 is the most restrictive applicable action -> Immediately commence a down power to hot shutdown.

B. Correct:

lAW TS 3.7.1. 1 Action 2, with one SACS subsystem inoperable (as given in the stem-> one SACS loop OOS), and if continued plant operation is permitted by LCO 3.7.1.3 (even with river temperature above 3.7. 1.3 limit of 85F, continued operation is allowable due to the fact that both emergency discharge valves are open and emergency discharge pathways are available->

no other systems OOS in stem), realign at least one affected EDG to the OPERABLE SACS subsystem within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ->1130 on 9/13. C. Incorrect but plausible:

Plausible if the applicant believes that 3. 7.1. 1 Action a.1.b is the most restrictive applicable action -> the HX must be restored within 72hrs. D. Incorrect but plausible:

Plausible if the applicant believes that 3. 7. 1.1 Action a.1.a is the most restrictive applicable action -> the SACS pump must be restored within 30 days. SRO Only Justification:

This question is SRO only as it requires assessing facility operating limitations in the Tech Spe.cs and their bases, cannot be answered solely by knowing <= 1 hr TS actions or above the line information, cannot be answered by knowing TS safety limits, and involves application of TS required actions and surveillance requirements.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

Learning Objective:

0302-*000.00H-00011 0-10 Obj 8 Question Source: Modified Question History:

Modified from HC Bank #325 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.5/45.11 Comments:

QUESTION 79 A malfunction in the recirculation pump trip circuitry causes a trip of the 'A' Recirculation pump resulting in the following plant conditions:

  • Core Flow is 45 percent
  • Reactor power is 50 percent Subsequently, the 'B' Recirculation pump trips. All appropriate operator actions are taken for the given plant conditions.

Based on the above, the ... A. NRC must be notified no later than one hour. B. NRC must be notified no later than four hours. C. NRC must be notified no later than eight hours. D. State of New Jersey must be notified no later than 15 minutes.

Question 79 K&A Rating: K&A Statement:

Justification:

295001 G2.4.30 (4.1) 295001 Partial or Complete Loss of Forced Core Flow Circulation G 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external

agencies, such as the State, the NRC, or the transmission system operator.

A. Incorrect but plausible:

No EAL or one hour notification thresholds reached.

B. Correct:

NRC must be notified in four hours.-RAL 11.3 for manual RPS actuation.

C. Incorrect but plausible:

Eight hour reports are for actuations other than RPS. D. Incorrect but plausible:

No EAL thresholds reached therefore no 15 minute notification required.

SRO Only Justification:

This question is SRO only as it requires determining off site notifications which is a SRO ONLY function.

References:

RAL 11.3. Student Ref: EALs and RALs Learning Objective:

NA Question Source: HC Bank 112333 Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.5/45.11 Comments:

QUESTION 80 The plant is operating at 1 00% power with the following:

  • I&C has just discovered that all suppression pool temperature indicating channels have a calibration error
  • Average suppression pool temperature was indicating 82°F
  • Actual average suppression pool temperature has just been confirmed as 112°F
  • HPCI is running as part of the quarterly HC.OP-IS.BJ-0001 HPCI MAIN AND BOOSTER PUMP SET 1ST
  • No other actions have been taken Which ONE of the following identifies the most limiting TS required action? A. Restore suppression pool average water temperature to within limits within 1 hr or be in at least HOT SHUTDOWN within the next 24hrs and in COLD SHUTDOWN within the following 24hrs. B. Stop all testing which adds heat to the suppression pool and restore the average temperature to less than 95°F within 24hrs or be in at least HOT SHUTDOWN with the next 12hrs and in COLD SHUTDOWN within the following 24hrs. C. Place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in the suppression pool cooling mode. D. Depressurize the reactor pressure vessel to less than 200 psig within 12hrs.

K&A Rating: 295026EA2.01 (4.1/4.2)

K&A Statement:

Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

EA2.01 Suppression Pool Temperature Justification:

A. Incorrect but plausible:

With suppression pool water temperature

>95F (112F) TS 3.6.2.1.b is applicable.

The 1 hr action reference is for 3.6.2.1.a suppression pool water level outside limits, not temperature.

B. Incorrect but plausible:

The given TS action is required if suppression pool average temperature is greater than 1 05F but less than 11 OF. C. Correct:

Per TS 3.6.2.1, with suppression pool average water temperature greater than 11 OF, place the reactor mode switch in the Shutdown position and operate at least one RHR loop in the suppression pool cooling mode. D. Incorrect but plausible:

The given TS action is required if suppression pool average temperature is greater than 120F (Given as 112F in the stem). SRO Only Justification:

This question is SRO only as it involves application of required TS actions, cannot be answered by solely knowing <= 1 hr TS actions, cannot be answered by solely knowing above the line LCO information, and cannot be answered solely by knowing TS safety limits. The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

TS 3.6.2.1 Student Ref: NONE Learning Objective:

Question Source: New Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.5/45.13 Comments:

QUESTION 81 The plant was at 100% power when a loss of off-site power occurred.

  • All low pressure ECCS has started on the Level 1 RPV water level signal
  • ADS is inhibited Current Plant conditions are:
  • SP level: -140 inches and dropping slowly 900 psig and slowly rising 76inches
  • SP temperature:

113° F lAW HC.OP-EO.ZZ-01 01, which ONE of the following does the CRS direct? A. Use the main turbine bypass valves to lower reactor pressure to 400 psig to allow the ECCS systems to inject B. Immediately emergency depressurize the reactor using alternate depressurization systems C. Ensure that one or more low pressure ECCS systems are available and before -185 inches emergency depressurize the reactor D. Use the Safety Relief Valves to lower reactor pressure to 600 psig and restore level with the secondary condensate pumps K&A Rating: 295031 EA2.04 (4.6/4.8)

K&A Statement Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: EA2.04 Adequate Core Cooling Justification:

A. Incorrect but plausible:

Guidance lAW AL/C-1 0 of HC.OP-EO.ZZ-01 01, guidance is to perform an ED before -185 inches. With a loss of offsite power, there is no powerfor EHC pumps, and MSIV are closed B. Incorrect but plausible:

ED criteria is not currently met, ED not permitted C. Correct:

lAW AL/C-10 of HC.OP-EO.ZZ-0101, with one or more low pressure ECCS pumps running and lined up for injection, ED is required before -185 inches which will allow for low pressure ECCS restoration of RPV level. D. Incorrect but plausible:

Loss of offsite power, no power to secondary condensate pumps SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

HC.EO-OP.ZZ-01 01, Rev. 11 HC.EO-OP.ZZ-01 01 Bases, Rev. 4 Learning Objective:

E01 01 E006 Question Source: HC Bank 34404 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

10CFR55:

CFR 41.10/43.5/45 13 Comments:

Student Ref: NONE X QUESTION 82 Given the following conditions:

  • The plant is operating at 100% power
  • LAC Police report a marsh fire directly beneath the New Freedom 500kV line (5023) lAW HC.OP-AB.BOP-0004, "Grid Disturbances",

the CRS will implement procedures that will ... A. scram the reactor.

B. remove the 5023 line from service.

C. result in a Reactor power reduction.

D. declare all off-site power sources inoperable per TS 3.8.1.1 or 3.8.1.2.

K&A Rating: 70000AA2.02 (3.2/3.8)

K&A Statement:

Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

AA2.05 Operational status of offsite circuit Justification:

A. Incorrect but plausible:

A Reactor SCRAM is not required per HC.OP-AB.BOP-004.

However, a Reactor SCRAM would be required if the applicant mistakenly believes that a fire off site would result in entry into HC.OP-AB.ZZ-0135 SBO/LOOP/EDG Malfunctions.

B. Correct:

lAW HC.OP-AB.BOP-0004, Grid Disturbances, part D, if a fire is reported that threatens or has the poses a threat to an off-site distribution line, measures shouls be taken to remove the line from service expeditiously

. C. Incorrect but plausible:

A power reduction is not required per HC.OP-AB.BOP-0004 in the event of a gr D. Incorrect but plausible:

Entry into TS would only be required per HC.OP-AB.BOP-0004 if the 500kV switchyard voltage falls below 493kV. SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator actions.

AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

HC.OP-AB.BOP-0004 Student Ref: NONE HC.OP-AB.ZZ-0135 Learning Objective:

Question Source: New Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/43.5/45.5/45.7/45/8 Comments:

QUESTION 83 Given the following conditions:

  • All rods are full in
  • RHR pumps A" and "8" are NOT available
  • Reactor water level is -100 inches and steady
  • Reactor Pressure is 700 psig
  • Drywell pressure is 25 psig and rising at 5 psig per minute
  • Suppression Chamber Pressure is 23 psig and rising at 5 psig per minute
  • Suppression Pool level is 1 06 inches and rising at 2 inches per minute
  • Secondary Condensate Pumps total flow is 12,000 gpm Which ONE of the following is the correct EOP mitigation strategy for this event? A. Emergency Depressurize and then inject with sources internal to the containment.

B. Emergency Depressurize and then inject with sources external to the containment.

C. Inhibit ADS and remain at pressure to conserve inventory; inject with sources internal to the containment.

D. Inhibit ADS and remain at pressure to conserve inventory; inject with sources external to the containment.

Question 83 K&A Rating:

295009 G 2.4.6 (4.7) K&A Statement:

Knowledge of EOP Mitigation Strategies Justification:

A. Correct:

Since PSP can not be maintained ED is required.

If adequate core cooling is assured then terminate injection from outside conatainment.

B. Incorrect but plausible:

SP is high and going higher, do not want to continue with injection external to containment.

. C. Incorrect but plausible:

Need to ED because PSP can not be maintained.

D. Incorrect but plausible:

SP is high and going higher, do not want to continue with injection external to containment.

. SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

EOP 101 and 102 Student Ref: EOP 102 graphs Learning Objective:

NA Question Source: HC Bank 34414 Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.5/45.13 Comments:

QUESTION 84 Given the following conditions:

  • Unit 1 is at 60% power for waterbox cleaning and control rod patter
  • A CRD pump is currently tagged
  • At 0127, B CRD pump trips, maintenance is investigating
  • At 0128, accumulator trouble alarms are received for H 30-31 and 42-43
  • At 0130, charging header pressure

<940 psig

  • At 0131, Field Operator reports HCU pressure for. CU 30-31 and 42-43 are 925 psig
  • At 0133, accumulator trouble alarms are recei d for HCU 18-15 and 22-31, also confirmed with the Field Operator at 930 ps* In accordance with HC.OP-AB.IC-0001, Cant I Rod: ( 1) Which of the following actions is re ired? (2) What is the basis for the require ent of that action? A. (1) Place the mode switch* SHUTDOWN immediately (2) Allow reactor shutdo n based upon average control rod scram times B. (1) Place the mod (2) Allow chargi witch in SHUTDOWN immediately header pressure alone to fully insert control rods C mode switch in SHUTDOWN by 0153 (2) Allow dequate time to restore a CRD pump to service D. ace the mode switch in SHUTDOWN by 0153 Allow charging header pressure alone to fully insert control rods K&A Rating: 295022 AA2.01 (3.5/3.6)

K&A Statement:

Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS: AA2.01 Accumulator Pressure Justification:

A. Incorrect but plausible:

If applicant believes that immediate action is required.

This would be required with RPV pressure

<900 psig. While control rod scram times are affected by lower accumulator pressures, this is not the reason for the required action B. Incorrect but plausible:

If applicant believes that immediate action is required.

This would be required with RPV pressure

<900 psig. Accumulator pressure is the first to start control rod motion for a scram, which is followed up by reactor pressure completing the control rod scram; this is not the reason for the required action C. Correct:

lAW HC.OP-AB.IC-0001 and TS 3.1.3.5, if reactor pressure is >900 psig the mode switch must be placed in SHUTDOWN within 20 minutes if: two or more scram accumulators are inoperable coincident with charging header pressure

<940 psig. The basis for this action is that it allows for adequate time to return a CRD pump to service to restore charging header pressure.

D. Incorrect but plausible:

If applicant believes that 20 minutes is allowed.

Accumulator pressure is the first to start control rod motion for a scram, which is followed up by reactor pressure completing the control rod scram; this is not the reason for the required action

References:

HC.OP-AB.IC-0001, Rev. 14 Student Ref: NONE TS 3.1.3.5 Bases, Amendment 183 Learning Objective:

NOH01 CRMECHC:

13 Question Source: New Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.5/45.13 Comments:

QUESTION 85 The plant is at 100% power with all systems operational and in a normal lineup. A LOCA occurs, and all systems respond as designed and initiate injection into the vessel. Current plant conditions are as follows:

  • RPV Pressure is 60 psig and lowering slowly
  • RPV water level is -66 inches and rising slowly
  • Torus level is 126 inct1es and rising slowly
  • Drywell Sprays are in service
  • Drywell pressure is 7 psig, lowering slowly lAW HC.OP-EO.ZZ-01 02, "Primary Containment Control",

which ONE of the following systems is to be secured?

A. Drywell Sprays B. HPCI C. Core Spray D. RCIC K&A Rating: 295029 EA2.01 (3.9/3.9)

K&A Statement Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: EA2.01 Suppression Pool Water Level Justification:

A Correct:

Per SP/L-19 and SP/L 20 of HC.OP-EO.ZZ-0102, Primary Containment

Control, termination of drywell sprays is required if Torus Level cannot be maintained below 124 inches. This is to ensure operability of the suppression chamber to drywell vacuum breakers, to relieve noncondensibles into the drywell and equalize drywell and suppression chamber pressures.

B. Incorrect but plausible:

If applicant does not recall that HPCI suction has transferred to the Torus with Torus level at 126 inches. Additionally, if applicant does not recall that HPCI isolates on low steam pressure at 100 psig, may also choose this answer. C. Incorrect but plausible:

If applicant does not recall that termination of systems injecting into the RPV is only required for those systems with suction outside of primary containment.

D. Incorrect but plausible:

If applicant does not recall either (1) that RCIC suction does not realign to Torus on high Torus water level (2) RCIC isolates on low steam pressure at 64.5 psig. SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

HC. OP-EO.ZZ-01 02, Primary Containment Control Student Ref: NONE HC.OP-EO.ZZ-0102 Bases Learning Objective:

EOP1 02E009 Question Source: Modified HC Bank ID 35663 Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/

Analysis:

X 10CFR55:

CFR 41.10/43.5/45.13 Comments:

QUESTION 86 Plant conditions are as follows:

  • No High Pressure Injection is available
  • Reactor water level-130 inches and dropping slowly
  • Reactor pressure 330 psig and dropping slowly
  • Drywell pressure 19 psig and rising slowly HV-F017B, LPCIInjection Valve, starts to auto open and, with all ECCS Pumps available and reactor level rising, the PO manually overrides HV-F017B.

One minute later, a 1 OA401 Bus Lockout causes a trip of 'A' RHR and 'A' Core Spray pumps resulting in reactor level dropping to -170 inches. The PO depresses the HV-F017B OPEN pushbutton.

As the valve begins to open the following alarms and indications are observed:

  • "RHR DIV 2 OUT OF SERVICE" computer alarm
  • Amber status light next to HV-F017B is flashing
  • HV-F017B Valve shows dual indication
  • 'B' RHR Pump Flow is 1,500 gpm and steady Which ONE of the following identifies the impact on 'B' RHR and the action(s) required per HC.OP-EO.ZZ-0101,

'RPV Control'?

Impact on 'B' RHR A Pump continues to run with Min Flow Valve closed B Pump continues to run with Min Flow Valve open C Pump continues to run with Min Flow Valve closed D Pump continues to run with Min Flow Valve open Action Required lAW HC.OP-EO.ZZ-101 Steam Cooling PRIOR to RPV Emergency Depressurization RPV Emergency Depressurization ONLY RPV Emergency Depressurization ONLY Steam Cooling PRIOR to RPV Emergency Depressurization K&A Rating: 203000 A2.11 (3.4/3.6)

K&A Statement:

Ability to (a) predict the impacts of the following on the RHR/LPCI:

INJECTION MODE (PLANT SPECIFIC);

and (b) based on those predictions, use procedures to correct,

control, or mitigate the consequences of those abnormal conditions or operations:

A2.11 Motor operated valve failures Justification:

A. Incorrect but plausible:

With RHR pump flow above 1270 gpm, the min flow valve will be closed. The flashing indicator next to F017 is indicative of tripping of valve thermal overloads.

Steam cooling is NOT required since injection sources are available.

B. Incorrect but plausible:

With RHR pump flow above 1270 gpm, the min flow valve will not be open. The flashing indicator next to F017 is indicative of tripping of valve thermal overloads.

RPV ED will be required per EOP-101 since injection systems are available.

C. Correct:

With RHR pump flow above 1270 gpm, the min flow valve will be closed. The flashing amber indicator next to the F017 valve indicates that the valve has stopped traveling due to thermal overload actuation.

RPV ED will be required per EOP-101 since injection systems are available.

D. Incorrect but plausible:

With RHR pump flow above 1270 gpm. the min flow valve will not be open. The flashing indicator next to F017 is indicative of tripping of valve thermal overloads.

Steam cooling is NOT required since injection sources are available.

SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

NOH01 RHRSYSC, Rev. 10 HC.OP-EO.ZZ-01 01, Rev. 11 HC.OP-AR.ZZ-0005, Rev. 22 EOP 101 Learning Objective:

RHRSYSE007 Question Source: New Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Student Ref: NONE Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/45.6 QUESTION 87 Given the following conditions:

  • The plant is operating at 90% power
  • The HPCI Pump is being operated in the Full Flow Test Lineup
  • HPCI discharge pressure is 1150 psig While attempting to adjust pump flow, the flow controller setpoint remains stationary at 5300 gpm in AUTO. Subsequently, the PO shifts the HPCI flow controller to MANUAL and reports that the pump develops rated flow. lAW Hope Creek Technical Specifications, HPCI ... A. is Operable B. is Inoperable C is Operable But Degraded D. is Operable But Non-Conforming K&A Rating: K&A Statement:

Justification:

206000A2.

i4 (3.3/3.4)

Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use procedures to correct,

control, or mitigate the consequences of those abnormal conditions or operations:

A2.14 Flow controller failure A. Incorrect but plausible:

lAW the TS Bases, HPCI must be in AUTO with a setpoint of 5700gpm and be capable of rated flow and discharge pressure to be declared Operable.

B. Correct:

HPCI is inoperable because it is NOT capable of meeting all surveillance requirements listed in the TS. lAW the TS 3.5. i Bases, HPCI must be in AUTO with a setpoint of 5600gpm and be capable of rated flow and discharge pressure to be declared operable.

C. Incorrect but plausible:

HPCI could not be considered Operable But Degraded unless the setpoint was stuck at 5600gpm in AUTO. D. Incorrect but plausible:

Operable But Non-Conforming is not applicable to the conditions given in the stem. SRO Only Justification:

This question is SRO only as it requires knowledge of information in the TS Bases. No reference was provided (similar to example IV of the NRC "Clarification Guidance for SRO-Only Questions").

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

HC.OP-SO.BJ-0001 SH.OP-AP.ZZ-Oi 08 TS 3.5.1 Learning Objective:

HPCIOOE013 Question Source: Bank Student Ref: NONE Question History:

HC Bank #48 (Used on 2003 NRC Exam) Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/45.6 Comments:

QUESTION 88 Given the following conditions:

  • Refueling is in progress.
  • The Reactor Mode Switch is locked in REFUEL.
  • Source Range Monitors A, C, and D are operable
00. As a fuel assembly is taken to the fuel pool through the transfer canal, I&C determines that the 'C' SRM has failed its weekly surveillance test. lAW Hope Creek Technical Specifications, core alterations

... A. can continue with no restrictions.

B. must be immediately suspended.

C. can continue if no fuel movement occurs. D. can continue only in the quadrants monitored by SRMs A and D.

Question 88 K&A Rating: 215004 G 2.2.25 (4.2) K&A Statement:

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

A. Incorrect but plausible:

Must restrict movement in quadrants that have no SRM's operable.

B. Incorrect but plausible:

Suspending all moves is overly restrictive.

Moves can be performed if there is SRMs operable in the quadrant and an adjacent quadrant.

C. Incorrect but plausible:

Core alterations other than fuel moves are also effected by this TS. D. Correct:

One operable SRM must be in the quadrant where the core alteration is taking place and one in an adjacent quadrant..

SRO Only Justification:

This question is SRO only as it requires knowledge of the Bases in the technical specifications which is a SRO function.

References:

TS 3.9.2 Bases Student Ref: NONE Learning Objective:

NA Question Source: HC Bank #34795 Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 43.2 Comments:

QUESTION 89 lAW HC.OP-EO.ZZ-0202,

'EOP Bases', which ONE of the following states the MINIMUM number of Safety Relief Valves (SRVs) that must be opened during an Emergency Depressurization and the basis for that minimum number? A. 4 SRVs; To remove all decay heat from the core at a pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow. B. 4 SRVs; To provide the minimum depressurization rate required to ensure the low pressure ECCS systems inject soon enough to minimize the amount of time water level is below the top of active fuel. C. 5 SRVs; To remove all decay heat from the core at a pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow. D. 5 SRVs; To provide the minimum depressurization rate required to ensure the low pressure ECCS systems inject soon enough to minimize the amount of time water level is below the top of active fuel.

K&A Rating: 218000G2.4.18 (3.3/4.0)

K&A Statement:

ADS 2.4.18 Knowledge of the specific bases for EOPs. Justification:

A. Incorrect but plausible:

See answer choice 'C' explanation.

Plausible if the applicant does not recall that 5 SRVs are required.

B. Incorrect but plausible:

See answer choice 'C' explanation.

Plausible if the applicant does not recall that 5 SRVs are required and/or the bases for requiring 5 SRVs. C. Correct:

lAW HC.OP-EO.ZZ-0202, 5 ADS valves are required for RPV-ED. The Minimum Number of SRVs Required for Emergency Depressurization (MNSRED) is the least number of SRVs which corresponds to a Minimum Steam Cooling Pressure (MSCP) sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding MSCP. The MNSRED is utilized to ensure the RPV will depressurize and remain depressurized when emergency depressurization is required.

D. Incorrect but plausible:

See answer choice 'C' explanation.

Plausible if the applicant does not recall the bases for requiring 5 SRVs. SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

HC.OP-EO.ZZ-0202 Student Ref: NONE Learning Objective:

EOP202E003 Question Source: Bank Question History:

HC Bank #138 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/

Analysis:

1 OCFR55: CFR 41.10/43.1/45.13 Comments:

QUESTION 90 Given the following conditions:

  • The plant is in OPCON 4
  • RCS Temperature is 182 oF with a 10°F/hr heat up rate
  • RPV Level is 88 inches and steady
  • The 'A' and 'C' EDGs are tagged out for planned maintenance A Loss of Offsite Power occurs. Following the Emergency start of the 'B' and 'D' EDGs, a generator differential overcurrent condition occurs on the '8' EDG. Assuming no additional action, which ONE of the following identifies the required actions for this event lAW HC.OP-AB.RPV-0009,

'Shutdown Cooling'?

A. CLOSE the MSIVs B. Restore the '8' RHR pump C. Place RWCU in service and fully OPEN ED-V035 D. Cross tie 'D' RHR to '8' RHR loop K&A Rating: 2640002.4.9 (3.8/4.2)

K&A Statement:

EDGs 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Justification:

A. Incorrect but plausible:

lAW HC.OP-AB.RPV-0009, the MSIVs should only be closed it forced circulation canot be established using preferred RHR loops or reactor recirculation IF RPV level cannot be maintained

<90 inches. B. Incorrect but plausible:

HC.OP-AB.RPV-0009 does direct the restoration of the tripped RHR pump, however, due to the trip of the '8' EDG, the '8' RHR pump cannot be returned to service.

C. Incorrect but plausible:

HC.OP-AB.RPV-0009 does direct that RWCU be placed in service as an alternate means of decay heat removal,

however, RWCU is not available due to the LOOP (power supplies:

108254 and 108264. D. Correct:

Upon LOOP, 10A401 and 10A403 are denergized due to the A and C EDGs being tagged out. The generator differential on the 8 EDG will trip the 8 EDG rendering the 1 OA402 vital bus denergized.

The only bus with power will be the 1 OA404 bus being supplied by the D EDG. Loop 8 RHR SID Cooling will deenergize (8 RHR Pump). lAW HC.OP-AB.RPV-0009, Step E.5, operators should attempt to cross tie the D RHR pump into the 8 Loop of S/0 cooling using Attachment 3 to restore core cooling.

SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

NOH01RHRSYSC-i0 Student Ref: NONE HC.OP-A8.RPV-0009 Learning Objective:

RHRSYSE001 Question Source: New Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.5/45.13 Comments:

QUESTION 91 While operating at 100% power, the following indications are noted:

  • The Reactor Building Steam Vent Blowout Panel 1 AS224 indicates open on the RM-11
  • Visual observation confirms that 1AS224 is NOT fully closed
  • Reactor Building d/p is holding steady at -.28" WG Which ONE of the following describes the operational impact and actions (if any) required?

A. Loss of secondary containment integrity.

Take actions lAW HC.OP-AB.CONT-0003, Reactor Building.

B. Loss of secondary containment integrity.

Take actions lAW EOP-1 03/4, Reactor Building and Rad Release Control.

C. Potential loss of secondary containment.

Start an additional RBVS exhaust fan JAW HC.OP-SO.GR-0001, Reactor Building Ventilation System Operation.

D. Potential loss of secondary containment.

NO additional actions are required unless Reactor Building d/p is greater than -.30" WG.

Question 91 K&A Rating: 290001 A2.02 (3.7) K&A Statement:

Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT; and (b) based on those predictions, use procedures to correct,

control, or mitigate the consequences of those abnormal conditions or operations:

A2.02 Excessive outleakage Justification:

A. Correct:

By TS definition, TS 3.6.5.1, secondary containment integrity is lost because the blowout panel is not closed and sealed. AB.CONT-0003, Reactor Building discusses mitigation actions.

B. Incorrect but plausible:

No entry condition exists for EOP 103/4. C. Incorrect but plausible:

secondary containment integrity is lost because the blowout panel is not closed and sealed. D. Incorrect but plausible:

secondary containment integrity is lost because the blowout panel is not closed and sealed. SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

TS 3.6.5.1 AB.CONT-0003 Student Ref: NONE Learning Objective:

ABCNT3E007 Question Source: HC Bank 80648 Question History:

HC NRC Exam 2010 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 43.5 Comments:

QUESTION 92 Given the following conditions:

  • Main Generator output is 915 MWe
  • The 'B' Instrument Gas Compressor is tagged for routine maintenance Subsequently, the 'A' Instrument Gas Compressor trips.
  • Instrument Gas pressure is 55 psig and slowly lowering
  • All attempts to restore Instrument Gas pressure have failed Instrument Gas pressure is now 49 psig. lAW HC.OP-AB.COMP-0002

'Primary Containment Instrument Gas', the CRS must direct ( 1) in order to prevent (2) A. (1) closing the Inboard MSIVs, THEN locking the Mode Switch in Shutdown (2) the pressure and power spike from the Inboard MSIVs drifting closed. B. (1) reducing Recirc Pump speed to Minimum, THEN locking the Mode Switch in Shutdown (2) the pressure and power spike from the Outboard MSIVs drifting closed. C. (1) reducing Recirc Pump speed to Minimum, THEN locking the Mode Switch in Shutdown (2) the pressure and power spike from the Inboard MSIVs drifting closed. D. (1) closing the Outboar*d MSIVs, THEN locking the Mode Switch in Shutdown (2) the pressure and power spike from the Outboard MSIVs drifting closed.

K&A Rating: K&A Statement:

Justification:

239001 2.4.11 (4.0/4.2)

Main and Reheat Steam: 2.4.11 Knowledge of abnormal condition procedures.

A. Incorrect but plausible:

MSIV closure is an RPS actuation B. Incorrect but plausible:

PCIG has no effect on outboard MSIVs, RRPs are at minimum due to the PCP trip. C. Correct:

lAW HC.OP-AB.COMP-0002 Retainment Overide (not an Immediate Action),

if lnstriument Gas Pressure is<= 50 psig, reduce recirc pump speed to minimum, lock the mode switch in shutdown, place RCIC and/or HPIC in service, close the Inboard MSIVs, and Close HV-5124A and HV-51248.

D. Incorrect but plausible:

MSIV closure is an RPS actuation.

Outboard MSIVs are not affected by a loss of instrument gas. SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

HC.OP-AB.COMP-0002 Student Ref: NONE Learning Objective:

ABCMP2E001 Question Source: Bank Question History:

HC Bank #226 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.5/45.13 Comments:

QUESTION 93 The plant is operating at 100% power with the following conditions:

  • Pressure transmitter SM-PT-N076A has failed its surveillance calibration check
  • The failure is non-conservative Which ONE of the following describes the actions(s) required by Hope Creek Technical Specifications?

A. Trip the inoperable channel in one trip system within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. B. Trip the inoperable channel in one trip system within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OR be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. C. Trip the inoperable channel in one trip system within one hour AND trip one channel in the other trip system within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. D. Commence a normal shutdown within one hour AND be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Question 93 K&A Rating: 216000 G 2.2.22 (4.0) K&A Statement:

Knowledge of limiting conditions for operations and safety limits Justification:

A. Incorrect but plausible:

This more limiting action is for instruments common to RPS. B. Correct:

Actions for number of OPERABLE channels less than required for ONE trip system. N076A is NOT common to NSSSS and RPS. Tech Specs 3.3.1 and 3.3.2.b.1.b C. Incorrect but plausible:

This action is for OPERABLE channels less than required for BOTH trip systems.

D. Incorrect but plausible:

These actions are lAW 3.0.3 when there is no applicable condition.

This is not the case. SRO Only Justification:

This question is SRO only as it requires assessing operability and technical specifications which is an SRO function.

References:

Tech Spec sections 3.3.1 through 3.3.4; P&ID M-42-1 Sheet 2 Student Ref: Tech Spec sections 3.3.1 through 3.3.4; P&ID M-42-1 Sheet 2 Learning Objective:

NSSSSOE009 Question Source: HC Bank #35703 Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.5/43.5 Comments:

QUESTION 94 Given the following conditions:

  • The plant is in Operational Condition 4
  • The Reactor Head detensioning machine is being lowered into position to detension the reactor head lAW HC.OP-IO.ZZ-0005, Cold Shutdown to Refueling, who may authorize detensioning the first RPV Head Stud? A. Control Room Supervisor B. Reactor Engineer C Refueling Floor SRO D. Refueling Outage Manager K&A Rating: 2.1 .41 ( 3. 7) K&A Statement:

Knowledge of the refueling process Justification:

A. Correct:

lAW HC.OP-IO.ZZ-0005, Step 5.4.22, this direction comes from the SM/CRS B. Incorrect but plausible:

Non-licensed personnel are not authorized to direct mode changes C. Incorrect but plausible:

Even though the refueling floor SRO is a licensed

operator, this direction must come from the licensed SRO in the MCR D. Incorrect but plausible:

The refueling outage manager may recommend mode changes, but direction for making mode changes comes from the licensed SRO in the MCR SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

HC.OP-IO.ZZ-0005, Rev. 35 Student Ref: NONE Learning Objective:

Question Source: Question History:

Cognitive Level: 1 OCFR55: Comments:

HC Bank #35519 HC NRC Exam 2003 Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

CFR 41.2/41.10/43.6/45.13 QUESTION 95 It is 0200 during normal full power operation.

The CRS needs to leave the site due to a personal emergency.

  • At 0205 the CRS departs as directed by the Shift Manager (SM).
  • At 0210 the SM calls the Operations Manager to inform him of the reduction in crew composition.
  • At 0220 the SM reaches a relief watch stander for the CRS and directs the reflief to come to work.

which ONE of the following is correct?

A. The operating crew has complied fully with shift manning requirements.

B. The CRS position should have been manned by a relief by 0405. C. The CRS should not have left until the Operation Manager gave him permission.

D. The CRS should not have left until his relief had arrived and turnover was completed.

K&A Rating: K&A Statement:

Justification:

G2.1.4 (3 3/3.8) Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 1 OCFR55, etc. A. Incorrect but plausible:

Plausible if the applicant recalls that a reduction in shift crew composition is allowable per Tech Specs as long as immediate action is taken to restore compliance, but does not recall that minimum shift crew composition must be restored within two hours. B. Correct:

Correct response.

Tech Spec 6.2.2 allows shift crew composition to be less than the minimum requirements Table 6.2.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore shift crew composition.

C. Incorrect but plausible:

The Operation Manager needs to only be notified.

Operations manager permission does not need to be obtained.

D. Incorrect but plausible:

Tech Specs allow for a two hour window to restore minimum shift crew composition.

SRO Only Justification:

This question is SRO only as it requires knowledge of facility operating limitations in the Technical Specifications and their bases (minimum shift crew staffing requirements).

References:

Tech Spec 6.2 Student Ref: NONE Learning Objective:

Question Source: Modified Question History:

Modified from 2005 VY LOI Exam Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.2 Comments:

QUESTION 96 Surveillance HC.OP-IS.BC-0004(0),

'DP202, D RESIDUAL HEAT REMOVAL PUMP SERVICE TEST', is scheduled to be performed seven days from now. The system engineer wants to perform a Temporary Change related to minimum flow valve, F0007D. This change will involve cycling HV-F007D, D RHR PUMP MIN FLOW ISLN VLVwith HV-F01 OB D RHR Pump Test Return Vlv OPEN, and taking readings on discharge pressure and flow. lAW AD-AA-1 01, 'PROCESSING OF PROCEDURES AND T&RMS' ... A. the test may NOT be conducted as a Temporary Change. The test also may NOT be conducted as an On The Spot Change. B. the test may be conducted as a Temporary Change. Obtain an approval signature from a station qualified reviewer (SQR) and an SRO. C. the test may NOT be conducted as a Temporary Change. The procedure changes should be processed as an On The Spot Change. D. the test may be conducted as a Temporary Change. Perform a full review, approval and authorization within 14 days of implementing the Temporary Change. K&A Rating: G.2.2.5 (3.2) \

K&A Statement:

Knowledge of the process for making design or operating changes to the facility.

Justification:

A. Correct:

The proposed test represents a change of intent to the procedure.

lAW AD-AA-101 4.2.3, Temporary Changes are only allowed if they DO NOT change the intent of a procedure.

Since the proposed test changes plant configuration, it should be processed as a Interim Change per section 4.2.2 of AD-AA-101 B. Incorrect but plausible:

Plausible if the applicant does not recognize that the proposed test represents a change of intent to the procedure.

C. Incorrect but plausible:

Plausible if the applicant recognizes that the test should not be processed as a Temporary Change, but does not recognize that an Interim Change should be used instead of an On The Spot Change. D. Incorrect:

Plausible if the applicant does recognize that the proposed test represents a change of intent to the procedure.

SRO Only Justification:

This question is SRO only as it requires This question, from the Generic portion of the written exam, is SRO only as it requires application of 1 OCFR55.43(b

)(3), facility licensee procedures required to obtain authority for design and operating changes to the facility.

References:

AD-AA-1 01, "Processing of Procedures and T&RMs. Student Ref: NONE Learning Objective:

NA Question Source: NEW Question History:

NA Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

1 OCFR: CFR 41.10/43.5/45.13 Comments:

QUESTION 97 The plant is operating at rated conditions when the N01 OA, N011 A, N012A, and N013A Steam Tunnel Temperature switches fail low. These four switches input temperature to Main Steam Line 'A'. Which ONE of the following describes the required Technical Specification actions?

A Place the inoperable channel(s) in the tripped condition within 1 hr ONLY. B. Place the inoperable channel(s) in the tripped condition within 24hrs ONLY. C. Place the inoperable channel(s) in one trip system in the tripped condition within one hour, AND place the inoperable channel(s) in the remaining trip system in the tripped condition within 1 hr.* D. Place the inoperable channe!(s) in one trip system in the tripped cbndition within one hour, AND place the inoperable channel(s) in the remaining trip system in the tripped condition within 24hrs.

K&A Rating: K&A Statement:

Justification:

2.2.37 (3.6/4.6)

Ability to determine operability and/or availability of safety related equipment.

A. Incorrect but plausible:

Plausible if the applicant misinterprets the TSs and believes they are in TS 3.3.2.b.1.a.

B. Incorrect but plausible:

Plausible if the applicant misinterprets the TSs and believes they are in TS 3.3.2.b.1.c.

C. Correct:

All 4 steam tunnel temperature switches on the 'A' Main Steam Line are inoperable as given in the stem (N01 OA, 11 A, 12A & 13A). TS Table 3.3.2-1 (3.f) states that a minimum of 2 operable channels per trip system are required in OPCONS 1,2 and 3. The inoperable temp switches result in 0 Operable channels on the 'A' Steam Line for BOTH the A and B trip systems.

Therefore, TS 3.3.2 action c.1 and 2.1 are applicable.

D. Incorrect but plausible:

Plausible if the applicant misinterprets the TSs and believes they are in TS 3.3.2.c.2.3.

SRO Only Justification:

This question is SRO only as it involves application of required TS actions, cannot be answered by solely knowing <= 1 hr TS actions, cannot be answered by solely knowing above the line LCO information, and cannot be answered solely by knowing TS safety limits. The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

TS 3.3.2 Student Ref: TS Learning Objective:

Question Source: New Question History:

N/A Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.7/43.5/45.12 Comments:

QUESTION 98 Given the following conditions:

  • You are filling out a dose extension lAW RP-AA-203 for an operator to enter the Drywell
  • The operator's current annual dose is 3500 mrem TEDE
  • The operator is expected to receive 400-450 mrem on this job. lAW RP-AA-203, Exposure Control and Authorization, which of the following personnel must approve this extension?
1. Work Group Supervisor
2. RP Manager 3. Station/Plant Manager 4. Site Vice President A 1 and 3 only. B. 2 and 4 only. C. 1, 2, and 3 only. D. 1, 2, 3, and 4.

K&A Rating: 2.3.4 (3.2/3.7)

K&A Statement:

Justification:

2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

A. Incorrect but plausible:

See response C. B. Incorrect but plausible:

See response C. C. Correct:

lAW RP-AA-203, to raise the ADCL to 3000 mrem, 1 & 2 are required.

To raise the ADCL to 4000mrem, 1, 2, and 3 are required.

To raise the ADCL above 4000mrem, 1, 2, 3 and 4 are required.

The stem indicates that a maximum ADCL of 3950 will be required for the job (3500mrem

+ 450mrem),

therefore 1, 2, and 3 are required.

D. Incorrect but plausible:

See response C. SRO Only Justification:

This question is SRO only as it requires knowledge of radiation hazards that may arise including maintenance activities and various contamination conditions.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

RP-AA-203 Student Ref: NONE Learning Objective:

NOH04ADM024E-004 Question Source: Bank Question History:

HC Bank# 213 Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.12/43.4/45.10 Comments:

QUESTION 99 During a fire in the Turbine Building, the Shift Manager is required to assign a Fire Department Liaison.

lAW OP-HC-112-1 01-1001 "Shift Turnover Responsibilities",

1) who, by title, can be assigned the Liaison role, and 2) what is their duty? A. 1) ST A AND the Auxiliary Building EO; 2) Communicates to the MCR on status of the event. B. 1) WCC Supervisor AND the Radwaste EO; 2) Communicates to the MCR on status of the event. C. 1) ST A AND the CRS; 2) Advises the Shift Manager on what equipment needs to be removed from service.

D. 1) WCC Supervisor AND the CRS; 2) Advises the Shift Manager on what equipment needs to be removed from service.

Question 99 K&A Rating: G.2.4.27 (3.4/3.9)

K&A Statement:

Knowledge of "fire in the plant" procedures.

Justification:

A. Incorrect but plausible:

Cannot fill dual function of STA and Fire Liason, see OP-AA-1 01-111 Attachment 1 Hope Creek Shift Complement note 3 B. Correct:

OP-AA-101-111 Attachment 1 Hope Creek Shift Complement note 3 states: the SM should designate an appropriate EO, RWEO, RO or SRO (cannot be concurrently assigned to fill the position of SM, CRS, STA, NC0(2), CM1, CM2 or OSCC in Attachment 2 table) to function as the station fire brigade liaison.

They should also function as liaison for other emergencies (ex. chemical spill, toxic gas, environmental, etc.). The purpose of the liaison is to provide real time communication to the control room regarding the status of the event. A. Incorrect but plausible:

Cannot fill dual function of STA and Fire Liason see OP-AA-1 01-111 Attachment 1 Hope Creek Shift Complement note 3 C. Incorrect but plausible:

Cannot fill dual function of CRS and Fire Liason, see OP-AA-1 01-111 Attachment 1 Hope Creek Shift Complement note 3 SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

OP-HC-112-101-1001 Learning Objective:

ADMPR05CE001 Question Source: Bank Question History:

HC Bank #221 Cognitive Level: Memory/Fundamental Knowledge:

X Comprehensive/Analysis:

1 OCFR55: CFR 41.3/43.5/45/13 Comments:

Student Ref: None QUESTION 100 The plant is at 100% power. Which ONE of the following situations constitutes sufficient information to declare the associated SSC INOPERABLE?

A. Receipt of OHA B1-C4 HPCI TURBINE TROUBLE concurrent with CRIDS point 05430 HPCI VACUUM TANK LEVEL HI; and indication the Vacuum tank Condensate Pump is NOT running.

B. Receipt of OHA E4-A2 DIESEL GEN PNL A/B/C/D C422 concurrent with CRIDS points D4581 DIESEL GENERATOR A RCP 1AC422 TRBL AND D3776 DG A REGULAR LOCKOUT RELAY TRIPPED; and the Aux Bid EO reports the REGULAR LOCKOUT RELAY for A EDG is tripped.

C. Receipt of OHA C6-D4 CRD ACCUM TROUBLE concurrent with CRIDS point D5268 CRD ACCUM AND an amber ACCUM light for 02-19 on the Full Core Display; and the RB EO reports accumulator pressure for 02-19 is 1150psig.

D. Receipt of OHA C1-E1 SLC TANK TROUBLE concurrent with CRIDS point D2382 SLCS TANK LEVEL HILO and an indicated SLC tank level of 4640 gallons.

K&A Rating: 2.4.45 (4.3) K&A Statement:

Ability to prioritize and interpret the significance of each annunciator or alarm Justification:

A. Incorrect but plausible:

Pump down of the HPCI vacuum tank is NOT required for HPCI to perform its safety function.

Although the alarm condition is valid, it does NOT inherently render HPCI inoperable B. Correct:

A trip of the Regular Lockout Relay will prevent the EDG from starting under ANY conditions.

lAW HC.OP-AR.ZZ-0017, Lockout Relays Activated is one of the inputs into E4-A2. lAW HC.OP-SO.KJ-0001, an EDG trips upon receipt of an 86R lockout.

This will prevent it from fulfilling its safety function.

The CRIDS points identify the 'A' EDG as the affected EDG C. Incorrect but plausible:

lAW HC.OP-AR.ZZ-0011, this alarm may be due to EITHER high water level OR low gas pressure.

Only low gas pressure would be indicative of an inoperable condition.

Additionally, the allowable alarm setpoint could be above the pressure that requires declaring the HCU inoperable.

Local investigation is required to provide sufficient information erode confidence in the assumption of operability.

(Note that under different conditions, such as a current gas leakage problem with the associated HCU and knowledge that it alarms very close to the point at which it becomes inoperable, the assessment might be different)

D. Incorrect but plausible:

SLC operability is a combination of tank level and solution concentration.

The tank low level alarm is within the allowable operating region for the SLC tank lAW T/S 3.1 .5. Comparison of the SLC tank level with the solution concentration is required to assess operability SRO Only Justification:

This question is SRO only as it requires assessing plant conditions and then selecting a procedure or section of a procedure to mitigate, recover or with which to proceed.

The question cannot be answered by solely knowing systems knowledge, immediate operator

actions, AOP or EOP entry conditions, or the purpose, overall sequence of events, or overall mitigative strategy of a procedure.

References:

HC.OP-AR.ZZ-01 07, Rev. 6 HC.OP-AR.ZZ-0011, Rev. 58 HC.OP-AR.ZZ-0008, Rev. 43 TS 3.5.1 Learning Objective:

Question Source: HC Bank 654 79 Student Ref: NONE Question History:

None Cognitive Level: Memory/Fundamental Knowledge:

Comprehensive/Analysis:

X 10CFR55:

CFR 41.10/43.5/45.3/45.12 Comments:

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