ML083190115
| ML083190115 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 09/20/2008 |
| From: | Caruso J Operations Branch I |
| To: | Hunter J Exelon Generation Co |
| Hansell S | |
| Shared Package | |
| ML081060533 | List: |
| References | |
| TAC U01745 | |
| Download: ML083190115 (200) | |
Text
1 B
2 B
3 C
4 C
5 A
6 B
7 A
8 B
9 B
10 c
11 A
12 B
13 D
14 D
15 C
16 D
17 D
18 D
19 A
20 c
21 D
22 c
23 A
24 C
25 B
26 D
27 D
28 D
29 D
30 C
31 B
32 A
33 D
34 D
K I*: u Limcriclt 2008 NRCl Written Esani 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 B
A D
C D
A A
D D
A C
A C
c D
C C
C A
B C
D A
D A
B B
B C
C C
B D
C 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 B
B C
D D
B D
D B
B D
B A
B B
A C
B A
B A
B A
C B
A A
B A
C C
D Page 1 of 1
QUESTION 1
Unit 1 plant conditions are as follows:
0 50% power 0
65% total core flow on recorder XR-042-1 R613 1 B Reactor Recirc pump trips.
New plant conditions are:
0 39% total core flow on recorder XR-042-1 R613 0
1A Reactor Recirc speed at 57%
0 Core Plate dp 1.5 psid on recorder XR-042-1 R613 Assume no operator actions are taken.
WHICH ONE of the following identifies the core power level and recirculation flow configuration one minute later?
A.
40% power; idle loop flow is in the reverse direction B.
40% power; idle loop flow is in the forward direction C.
37% power; idle loop flow is in the reverse direction D.
37% power; idle loop flow is in the forward direction
K&A #
295001 Partial or Complete Loss of Core flow Importance Rating 3.3 QUESTION 1 K&A Statement:
K1.02 - Knowledge of the operational implications of power to flow distribution as it applies to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION Justification:
A.
Incorrect but plausible. Idle loop flow is always subtracted, but for recirc pump speed less than 6O%, actual flow is forward through the jet pumps. Since actual flow is forward, to determine actual total core flow, the flow through the idle loop should not be subtracted. The correct reactor power is therefore determined based on core dp measurement. This choice provides the correct power level; however, the basis for the answer is incorrect.
B.
Correct - Total core power is determined based on core dp and rodline. Actual idle loop flow is forward through the jet pumps when the operating recirc pump is less than 60% speed.
C.
Incorrect but plausible if the applicant does not understand that indicated flow is less than actual flow when in single loop operalion with the recirc pump speed less than 60%. Using indicated flow instead of core dp results in underpredicting actual core power.
D.
Incorrect but plausible if applicant does not understand indicated flow is less than actual flow, but understands flow is forward in the idle loop.
References:
OT-I 12 pg.4, Attachment 1, Student Ref: required
- Yes, LLOT1 540 power to flow map WIO labeling Learning Objective:
N/A Question source:
New Quest ion His tory :
None Cognitive level:
Memory/ F u n d a m e nta I know ledge :
ComprehensivelAnalysis:
X IOCFR 55.41(5) X
QUESTION 2
Unit 2 plant conditions are as follows:
0 A Loss of offsite power has occurred 0
ALL diesel generators FAILED to start.
WHICH ONE of the following identifies instruments that can be used to determine Reactor Water Level and Reactor Pressure?
A.
Div 1 PAMS Level Narrow Range Pressure on 20C603 B.
Wide Range Level on 20C603 HPCI Steam Line Pressure on 20C647 C.
Wide Range Level on 206603 Narrow Range Pressure on 20C603 D.
Div 1 PAMS Level HPCl Steam Line Pressure on 20C647
K&A #
295001 Partial or Complete Loss of AC Importance Rating 4.2 QUESTION 2 K&A Statement:
A2.02-Ability to deterrnine and/or interpret Reactor power/
pressure/ and level as it applies to PARTIAL OR TOTAL LOSS OF AC Justification:
A.
Incorrect but plausible if applicant does not recall that the PAMS is safety related but not DC powered. Narrow range pressure not available.
- 5.
Correct - Wide range level instrumentation on 10C603 is available during a station blackout with loss of all diesels. HPCl steam line pressure indication is available during a station blackout with loss of all diesels.
C.
Incorrect but plausible if applicant does recall that narrow range instrumentation is not a va i la b le.
D.
Incorrect but plausible if applicant does not recall that the PAMS is safety related but not DC powered.
References:
LGSOPS0042, E-I, rev.33 Att. 1 Student Ref: required N
Learning Objective:
LGSOPS0042 I11 0 Question source:
Bank (Limerick)
Question History:
None Cognitive level:
Memo ryl F u nd a m e n t a I know I ed g e:
X Com pre hen sive/Ana I ys is:
1 OCFR 55.41(7) X
QUESTION 3
Plant conditions are as follows:
0 Reactor power is 100%
0 An electrical fault has resulted in a loss of 125/250 VDC Non-Safeguard BOP power 0
Electrical Maintenance is investigating A grid disturbance results in a Unit 1 Main Generator lockout.
WHICH ONE of the following identifies the effect on operation of the 11 Unit Aux Bus supply breakers?
A.
The 11 Bus Breaker will trip 10-1 1 Bus Breaker will close B.
The 11 Bus Breaker will trip The 10-1 1 Bus Breaker will NOT close C.
The 11 Bus Breaker will NOT trip The 10-1 1 Bus Breaker will NOT close D.
The 11 Bus Breaker will NOT trip The 10-1 1 Bus Breaker will close WHEN the 11 Bus Breaker is tripped at the cubicle
K&A #
295004 Partial or Complete Loss of DC Power Importance Rating 3.2 QUESTION 03 K&A Statement:
Justification:
A.
A2.04 - Ability to determine and/or interpret system lineups as it applies to PARTIAL OR TOTAL LOSS of DC POWER.
Incorrect. Plausible if applicant believes 11 Unit Aux Bus control power is supplied from a safeguards source.
B.
Incorrect. Plausible if applicant believes 11 Unit Aux Bus and the 10-1 1 breaker control power is supplied from different sources.
C.
Correct. Both breakers are supplied from the 125/250 VDC non-safeguards source.
D.
Incorrect. Plausible if applicant believes the 11 Bus Breaker receives non-safeguards power and the 10-1 1 Bkr gets its control power from the 220 yard.
References:
LLOT0670 Rev 12 Student Ref.
LGSOPS0092A Rev 00 Required: No Learning Objective:
113 (LGSOPSOO92A)
Obj 10, (LLOT0670)
Quest ion source :
New Question History:
Cog nit ive level:
Memory/Fundamental knowledge:
ComprehensivelAnalysis:
X IOCFR 55.41 X
QUESTION 4
A Unit 2 GP-3 plant shutdown is in progress in preparation for a refueling outage with the following conditions:
0 Reactor power reduced from 48% to 18% over the last 30 minutes 0
"2A" and "28" Reactor Feed Pumps are in service 0
"2C" Reactor Feed Pump is in STANDBY 0
The Main Turbine is about to be removed from service per step 3.1.33 of GP-3 Immediately after the Main Turbine is tripped, the "28" Reactor Feed Pump trips.
WHICH ONE of the following describes the expected plant response over the next 15 minutes and the basis for this response?
Reactor Power Response Basis for Response A.
Lowers to a new stable value Change in xenon concentration B.
Lowers to a new stable value Change in recirc pump speed C.
Rises to a new stable value Change in feedwater temperature D.
Rises to a new stable value Change in RPV pressure
K&A #
295005 Main Turbine Generator Trip Importance Rating 3.8 QUESTION 4 K&A Statement:
Justification:
A.
K2.01 - Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and Feedwater temperature Incorrect but plausible. Xenon would initially build in due to the power decrease resulting in further reducing reactor power. Incorrect because xenon is not decaying at this point. Plausible distractor because xenon is a poison and will cause power to change depending on if it is building in or burning out. The applicant needs to understand the characteristics of xenon during power changes.
B.
Incorrect but plausible since recirc runbacks do occur on feed pump trips. However at 18% power the recirc pumps would already be at min speed.
C.
Correct - Following a turbine trip, a loss of extraction steam to the feedwater heaters would lower feedwater temperature. FW temperature lowering would add positive reactivity to the core.
D.
Incorrect but plausible. Expect RPV pressure to change slightly when removing the turbine from service. Several minutes after turbine is tripped pressure should return to normal based on BPVs taking steam loads.
References:
GP-3 Normal Plant Shutdown, Student Ref: required N
LLOT0540 pg. 26; LGSOPSOOOIA pg.
51 Learning Objective:
LGSOPSOOOI A IL8f Question source:
Bank (Susquehanna)
Quest ion History:
None Cognitive level:
Memory/Fundamental knowledge:
ComprehensivelAnalysis:
X 1 OCFR 55.41(5) X
QUESTION 5
Unit 1 conditions are as follows:
Reactor Power is 100%
0 Reactor level is 35 0
3 out of 4 Steam Flow Transmitters fail resulting in a FWLCS TROUBLE alarm 0
Alarm IXX-FW301.ISFE, Steam flow SMS Err is noted on the FWLC Alarm List A Main Turbine trip results in the following:
e RPS actuates and scram valves open 0
150 control rods fail to insert due to hydraulic lock Reactor power remains at 50%
WHICH ONE of the following describes the FWLC response, and the impact on Reactor water level after 30 seconds?
A.
SCRAM Profile will activate, reactor water level will go down
- 6.
SCRAM Profile will NOT activate on incomplete SCRAM, reactor water level will remain constant C.
SCRAM Profile will activate, reactor water level will go up D.
SCRAM Profile will NOT activate on incomplete SCRAM, reactor water level will go down
K&A #
295006 SCRAM Importance Rating 3.9 QUESTION 5 K&A Statement:
Justification:
A.
A I.02-Ability to operate and/or monitor Reactor water level control system as it applies to SCRAM Correct. SCRAM profile activates by C71 -K14 RPS relays located in panels 609 and 61 1. SCRAM profile is not dependent on rod position. SCRAM profile will reduce feedwater flow 6%/sec after the initial 10 seconds. Feedwater flow reduction will result in RPV level reduction.
B.
Incorrect but plausible if applicant does not understand the conditions under which the SCRAM profile activates.
C.
Incorrect but plausible if applicant does not understand the system response based on the SCRAM profile.
D.
Incorrect but plausible if applicant does not understand the conditions under which the SCRAM profile activates.
References:
LGSOPS0550 pg. 31, 32, 33 Student Ref: required N
Learning Objective:
N/A Question source:
Bank (Limerick)
Question Hi story:
None Cognitive level:
Memory/ Funda men tal knowledge:
X ComprehensivelAnalysis:
IOCFR 55.41(3) X
QUESTION 6
WHICH ONE of the following is the reason why the reactor is scrammed prior to evacuating the main control room in accordance with SE-1, Plant Shutdown from the Remote Shutdown Panel?
A.
Ensures that inventory makeup requirements will be within HPCl capability.
- 8.
Ensures that inventory makeup requirements will be within RClC capability.
C.
Scramming from outside of the control Room would require access to plant areas that may be inaccessible due to post-accident high radiation levels.
D.
Scramming from outside of the control room would require RPS bus power to be tripped causing concurrent isolations of NSSSS valve groups.
K&A #
295016 Control Room Abandonment Importance Rating 4.1 QUESTION 6 K&A Statement:
Justification:
A.
A3.01-Knowledge of the reasons for Reactor SCRAM as it applies to CONTROL ROOM ABANDONMENT Incorrect - HPCl is only used with SE-10 and is not applicable for this condition.
B.
Correct - RClC is used for inventory makeup per SE-1. Scramming the reactor reduces the inventory loss by reducing the heat load.
C.
Incorrect - Accidents are not within the scope of SE-1. Plausible if applicant thinks that accident conditions should be considered as part of remote shutdown.
D.
Incorrect - MSlVs are manually closed prior to evacuation of the control room and all Group isolations are expected during SE-1. Plausible if applicant is considering the actions to take for Alternate Remote Shutdown, SE-6, which directs opening of breakers to initiate the SCRAM.
References:
SE-1, SE-6, Student Ref: required N
Learning Objective:
N/A Question source:
Modified Bank (Peach Bottom)
Question History:
None Cognitive level:
Memory/ F u n d a menta 1 know I ed g e :
X Com pre hensive/Anal ysis :
1 OCFR 55.41(7) X
QUESTION 7
Unit 1 plant conditions are as follows:
e Reactor Power is 35%
e e
e e
e IA Stator Water Cooling pump is tripped & in PULL-TO-LOCK 1 B Stator Water Cooling pump is running Stator Cooling outlet temperature is 76 C Stator Cooling water inlet pressure is 23 psig Stator Cooling Storage Tank level is 4 below normal level WHICH ONE of the following describes the status of the Main Turbine due to the conditions above?
A.
Will trip if generator stator current is 8,000 amps after 3.5 minutes B.
Will trip if generator stator current is 22,000 amps after 2 minutes C.
Will trip immediately
- 0.
Will remain on-line indefinitely
K&A #
295018 CCW Importarice Rating 3.4 QUESTION 7 K&A Statement:
Justification:
A.
K2.02 - Knowledge of the interrelations between PARTIAL OR TOTAL LOSS OF CCVJ and Plant operations Correct - Plant conditions indicate a loss of cooling to the generator based on low stator coolant flow supply pressure, 43 psig. If the current carried by the main generator is not reduced to 7468 amps in 3 1/2 minutes a turbine trip will be generated.
B.
Incorrect but plausible if the applicant does NOT remember the settings for the first checkpoint. A trip signal would be generated at two minutes if load was not reduced to 26,173 amps.
C.
Incorrect but plausible if the applicant believes that loss of stator cooling causes an immediate turbine trip.
D.
Incorrect but plausible if the applicant believes that stator cooling runback set point has not been reached.
References:
LLOTO630 pages 14, 15, 16 Student Ref: required No Learning Objective:
N/A Question source:
Bank (Limerick)
Question History:
None Cognitive level:
Me rn o r y/ F u n d a menta I know I edge :
Comprehensive/Analysis:
X IOCFR 55.41(4) X
QUESTION 8
Unit 1 plant conditions are as follows:
0 Instrument Air and Service Air are in a normal configuration 0 B-2, "IA" INSTRUMENT AIR LOW PRESSURE alarm is received 0 C-2, "1 B" INSTRUMENT AIR LOW PRESSURE alarm is received 0 1A Instrument Air header pressure is reading 65 psig 0 1B Instrument Air header pressure is reading 68 psig WHICH ONE of the following describes the status of the Service Air Compressor?
A.
The Service Air Compressor is supplying the Service Air Header ONLY.
B.
The Service Air Compressor is supplying the 1A Instrument Air Header ONLY.
C.
The Service Air Compressor is supplying the 1A and 1 B Instrument Air Headers ONLY.
D.
The Service Air Compressor is supplying the Service Air Header and the 1A and 1 B Instrument Air Headers.
K&A #
29501 9 Partial or Total Loss of Inst. Air Importance Rating 3.6 QUESTION 8 K&A Statement:
A2.02 - Ability to determine and/or interpret Status of safety-related instrument air system loads as it applies to PARTIAL OR TOTAL LOSS OF INSTRUMENT AIR Incorrect but plausible if the applicant does NOT remember that when both Instrument Air headers are less than 70 psig, Service Air header isolates from Service Air compressor or does not think that the set point has been reached.
J us t i fica t ion:
A.
- 6.
Correct - Service Air will automatically backup IA when IAheader pressure drops below the Service Air header pressure. When both IA headers drop to 70 psig, PV-01 5-*67 closes to isolate service air header from the service air compressor. Service Air feeds the 1A instrument Air header through a manually positioned valve.
C.
Incorrect. Plausible if the applicant does NOT remember that Service Air compressor feeds only the 1A Instrument Air header through a manually positioned valve.
D.
Incorrect but plausible if the applicant does NOT remember that when both IA headers are less than 70 psig, Service Air header isolates from Service Air compressor.
References:
LLOT0730 pages 4, ON-I 19 Student Ref: required No Learning Objective:
LLOT0730 #2 Question source:
Bank (Limerick)
Question History:
Cognitive level:
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
1 OCFR 55.41(7) X
QUESTION 9
Unit 1 Plant conditions are as follows:
o OPCON4 o "IA" RHR is in Shutdown Cooling o "1A" and "1 6" Reactor Recirc pumps are secured The "1A" RHR Pump trips.
WHICH ONE of the following RPV level indications provide for crediting natural circulation as an alternative method of reactor coolant circulation?
A.
Upset level indication at 66 inches B.
Shutdown level indication at 64 inches C.
Wide Range level indication at 58 inches D.
Narrow Range level indication at 56 inches
K&A #
295021 Loss of Shutdown Cooling Importance Rating 3.9 QUESTION 9 K&A Statement:
Justification:
A.
K1.03 - Knowledge of the operational implications of Adequate core cooling as it applies to LOSS OF SHUTDOWN COOLING Incorrect but plausible. Vessel level must be maintained above 78 inches on the Upset level indication.
- 6.
Correct - Vessel level must be maintained above 60 inches on the Shutdown level indica tion C.
Incorrect but plausible if the applicant does NOT remember wide range indication could be off scale. Must use either Upset level or Shutdown level indication.
D.
Incorrect but plausible if the applicant does NOT remember narrow range indication would be off scale.
References:
S51.8.8, Rev.66 pg.2; Drawing 042-01 Student Ref: required No Learning Objective:
LLOT Question source:
Bank (Limerick)
Question History:
Cognitive level:
Memory! Fu n d a m en ta I know ledge :
Compre hensivelAnalysis:
X IOCFR 55.41(7) X
QUESTION 10 Unit 1 plant conditions are as follows:
OPCON5 Fuel bundle 43-08 is being lowered into the core SRM C count rate increases from 70 cps to 300 cps and has stabilized Remaining SRMs continue to indicate 70 to 80 cps Fuel Floor has just reported that fuel bundle 43-08 is approaching its seated position in the correct location WHICH ONE of the following describes the required action (if any) and the basis for this action?
A.
No action required unless SRM countrate exceeds 560 cps. Shutdown margin is not significantly impacted at 300 cps.
B.
Fully seat the fuel bundle and request Reactor Engineering to determine if count rate is within expected range. This action verifies higher count rate is consistent with a fuel bundle positioned adjacent to an SRM.
C.
Notify fuel handling director to stop lowering the bundle and enter ON-1 20, Fuel Handling Problems. This action addresses an unexpected count rate increase.
D.
Evacuate the Fuel Floor. This action ensures personnel are safe from the effects of an inadvertent criticality.
K&A #
295023 Refueling Accident I m porta nce Rating 3.2 QUESTION 10 K&A Statement:
Justification:
A.
K1.02 - Knowledge of the operation implications of shutdown margin as it applies to Refueling Accident.
Incorrect. Plausible if applicant believes the ON entry threshold is at three doublings.
- 3.
Incorrect but plausible since ON-I 20 states that a doubling of count rate should be expected for loading of fuel assemblies adjacent to SRM detectors.
C.
Correct. FH-105 Attachment identifies required immediate RO actions.
D.
Incorrect but plausible since this would be the appropriate action if an increasing trend were observed. However, stable counts indicate a subcritical condition and evacuation would not be required.
References:
LLOT0760, Rev 14 Student Ref.
Yes LLOTI 550, Rev. 13 FH-105, ON-120, Rev. 017.
Core Map Learning Objective:
Obj 2, (LLOTI 550)
Question source:
Modified Bank (Limerick)
Changed conditions and distractors Changes resulted in new answer.
Quest ion History :
NRC-05 (LGS), OYS CERT-04 Cognitive level:
Me rn o r y/Fu n da m en t a I know ledge :
Corn prehensive/Analysis:
X 1 OCFR 55.41 X
QUESTION 11 A loss of drywell cooling has occurred. Containment temperature and pressure are rising and current values are as follows:
0 Drywell Pressure 1.3 psig 0
Drywell Temperature 135°F 0
Suppression Pool Air Temperature 132°F 0
Suppression Pool Level 23'6" 0
Suppression Pool Temperature 91 "F 0
Containment Leak Detector 50 cpm and steady Suppression pool pressure 0.3 psig WHICH ONE of the following correctly describes the ability to vent the drywell?
A.
May NOT be vented based on current conditions
- 6.
May be vented until Drywell Temperature increases to 140°F C.
May be vented until Drywell Pressure drops to 0.9 psig D.
May be vented until Drywell Temperature increases to 137.5"F OR Drywell Pressure drops to 1.I psig
K&A #
295024 High Drywell Pressure Importance Rating 3.9 QUESTION 11 K&A Statement:
Justification:
A.
G2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation Correct. SP N2 mass is determined from OT-101 Attachment 1 Page 1 as 7710, based on rounding SP temperature up to '140°F and pressure down to 0.2 psig. Use of this value yields a limit defined by the N2 mass line of 7500 pounds on Page 1. Plotting this mass with DW pressure of 1.3 psig and DW temperature of 135°F determines conditions are unsafe to vent the DW.
- 6.
Incorrect. Venting is not allowed. Plausible if applicant determines the next higher nitrogen mass on Attachment 1 by not rounding values in the correct direction as noted in Justification A or incorrectly plots values/rounding values on Attachment 2.
C.
Incorrect. Venting is not allowed. Plausible if applicant determines the next higher nitrogen mass on Attachment 1 by not rounding values in the correct direction as noted in Justification A or incorrectly plots values/rounding values on Attachment 2.
D.
Incorrect. Venting is not allowed. Plausible if applicant determines the next higher nitrogen mass on Attachment 1 by not rounding values in the correct direction as noted in Justification A or incorrectly plots values/rounding values on Attachment 2.
References:
OT-I 01, High Drywell Pressure Student Ref: OT-1 01, Yes Pages 9, 10 (Step 3.10), Attachment 1 & 2 Learning Objective:
Question source:
Modified Bank (Limerick)
Quest ion His tory:
Cognitive level:
Memory/Fundamental knowledge:
Com pre he nsive/Ana lysis:
X IOCFR 55.41(7) X
QUESTION 12 Unit 1 is at 100% powe when th following occurs:
0 An EHC malfunction results in a Reactor scram from rated power.
All Control Rods are at position 00.
SRVs are cycling to control pressure.
Turbine bypass valves are approximately 3% open.
Given the above conditions and using the logic diagram below, which one of the following is consistent with the above conditions?
A.
B.
C.
D.
Load Limit output has failed to 3%.
Maximum Combined Flow Limiter has failed to 3%.
Pressure Set has failed to 980 psi.
A Steam Throttle Pressure input has failed to 980 psi.
Load Rejvct Runtmck o,
~
SYNC S W P ~
not Soledo0
-LOSE of Stator Water Cooling Runbeck 2 %,
Ruribilch 25 A Canbind Prebrura Set Smilfl Close mils Bypas& Jach
K&A #
295025 High Reactor Pressure Importance Rating 3.5 QUESTION 12 K&A Statement:
K3.08 - Knowledge of the reasons for Reactor/Turbine pressure regulating system operation as it applies to High Reactor Pressure.
Justification:
A.
Incorrect. Load limit at 3% will result in control valves closing and, ultimately, a high pressure reactor scram and turbine trip. Plausible if applicant thinks the limit will prevent BPV operation > 3%.
- 5.
Correct. Maximum Combined Flow Limiter will limit bypass valve demand signal to 3%.
C.
Incorrect. Pressure set failing to 980 will result in RPV pressure increase on control and bypass valve closure but will not limit BPV operation to 3%.
D.
Incorrect. Controlling throttle pressure input failed to 980 will result in backup regulator taking control at a slightly higher reactor pressure.
References:
LGSOPS0031 B Rev000 Student Ref. required No Learning Objective:
IL 2, IL 5 Question source:
New Question History :
Cognitive level:
10CFR MemorylFundamentaI knowledge:
Com p re hens ivelAna lysis:
X 55.41 X
QUESTION 13 Unit 2 plant conditions are as follows:
0 Unit 2 is at rated power.
0 Quarterly HPCI flow testing is in progress IAW ST-6-055-230-2 HPCI Pump, Valve and Flow Test.
0 Suppression pool level is 22 IO.
Div 1 SPOTMOS indicates 92OF and slowly trending higher.
0 Div 2 SPOTMOS indicates 90°F and slowly trending higher.
Highest individual suppression pool temperature sensor indicates 97OF.
WHICH ONE of the following describes the required action(s)?
A.
Enter T-I 02 Primary Containment Control AND immediately suspend testing.
- 6.
Suspend testing when average Suppression Pool temperature reaches 95OF.
C.
Enter T-I02 Primary Containment Control AND suspend testing before highest individual suppression pool temperature reaches 105°F.
D.
T-I02 Primary Containment Control entry is NOT required at this time.
Suspend testing before average Suppression Pool temperature reaches 105OF.
K&A #
295026 Suppression Pool High Water Temperature Importance Rating 3.9 QUESTION 13 K&A Statement:
Al.03 - Ability to operate and/or monitor temperature monitoring as it applies to Suppression Pool High Water Temperature.
Justification:
A.
Incorrect but plausible if applicant doesnt recognize that entry into T-I 02 is based on the average suppression pool temperature reaching 95OF versus the highest individual temperature indication AND is unaware of exception during testing which allows heat to be added to the Suppression Pool until the maximum average Suppression Pool temperature approaches 105OF.
- 6.
Incorrect but plausible if the applicant is unaware of exception during testing which allows heat to be added to the Suppression Pool until the maximum average Suppression Pool temperature approaches 105°F.
C.
Incorrect but plausible if applicant doesnt recognize that entry into T-102 is based on the average suppression pool temperature reaching 95OF versus the highest individual temperature indication AND is unaware of the exception during testing which allows heat to be added to the Suppression Pool until the maximum average Suppression Pool temperature approaches 105°F.
D.
Correct - Average Suppression Pool Temperature is below 95OF, therefore entry into T-102 is not required yet. An approved Tech Spec exception allows maximum Suppression Pool Temperature to be increased from 95°F to 105°F during testing which adds heat to the Suppression Chamber.
References:
LLOTOI 30 Rev01 6 Student Ref. required No Learning Objective:
Obj. 4 Question source:
New Question History:
Cognitive level:
Memory/Fundamental knowledge:
Com pre hensivelAna lysis:
X 1 OCFR 55.41 X
QUESTION 14 Given the following initial conditions:
0 The plant is stable at 85% power 0
RPV dome pressure is 1032 psig 0
Drywell temperature is 128°F 0
Drywell pressure is 0.3 psig During the subsequent power increase to 1 OO%, a loss of drywell cooling occurs.
Current plant conditions are as follows:
0 The plant is at 100% power.
0 Power, pressure and indicated level are stable.
0 RPV dome pressure is 1045 psig 0
Drywell temperature is 140°F and rising slowly Drywell pressure is 0.5 psig and rising slowly WHICH ONE of the following correctly completes the statement to describe Narrow Range Level instrumentation indication relative to actual RPV level?
Indicated level is actual level and indicated level will trend actual level as drywell temperature rises to 145°F.
A.
less than, toward B.
less than, away from C.
greater than, toward D.
greater than, away from
K&A #
295028 High Drywell Temperature Importance Rating 3.6 QUESTION 14 K&A Statement:
Justification:
A.
K2.03 - Knowledge of interrelations between HIGH DRYWELL TEMPERATURE and Reactor water level indication Incorrect but plausible. NR level is non-density compensated and is calibrated for 1045 psig PRV pressure and 135 F drywell temperature. Plausible that indicated is less than actual if applicant assumes the increased variable leg density at 1045 psig RPV pressure has de-calibrated the instrument, since lower density water column on the variable leg will lower the indicated level.
- 6.
Incorrect but plausible. NR level is non-density compensated and is calibrated for 1045 psig PRV pressure and 135 F drywell temperature. Plausible, and correct, that indicated level will trend away from actual level if drywell temperature continues to rise.
C.
Incorrect but plausible. NR level is no-density compensated and is calibrated for 1045 psig RPV pressure and 135 F drywell temperature. Plausible that indicated level is greater than actual level is applicant assumes instrument was at calibrated conditions prior to the drywell temperature increase to 135 F.
D.
Correct. NR level is non-density compensated and is calibrated for 1045 psig PRV pressure and 135 F drywell temperature. NT indicated level will be greater than and trend away from actual level as drywell temperature continues to increase beyond the instrument calibration value of 135 F because density of the reference leg will decrease as it heats up, reducing the reference leg pressure on the d/p sensor.
References:
LGSOPS0042 pg. 37, 39 IL 7G, Dwg.
Student Ref: required N
042-01, 006-04 Learning Objective:
LGSOPS0042, #7 Question source:
New Question History:
Cognitive level:
1 OCFR Memory/Fundamental knowledge:
Co m pre he n s i ve/An a I ys is :
X 55.41(5) X
QUESTION 15 Unit 2 plant conditions are as follows:
0 An Emergency Blowdown in progress.
0 Suppression Pool level is 14 and lowering.
2A and 2B RHR are in Suppression Pool Cooling with suction temperature at 1 15OF.
0 2A and 2B Core Spray loops are injecting.
0 Division 1 SPOTMOS indicates 131OF 0
Division 2 SPOTMOS is de-energized.
Which one of the following is the actual value of Suppression Pool temperature and the status of RHR and Core Spray NPSH limits?
Suppression Pool RHR and Core Spray Temperature NPSH Limits A.
B.
C.
D.
131OF Met 131OF Not Met 11 5°F Met 115OF Not Met
K&A #
295030 Low Suppression Pool Water Level Importance Rating 3.6 QUE.STION 15 K&A Statement:
AI.01 - Ability to operate and/or monitor ECCS systems (NPSH considerations) as it applies to Low Suppression Pool Water Level.
Justification:
A.
Incorrect but plausible if the applicant doesnt recognize that the Suppression Pool temperature thermocouples are uncovered below 17.8 ft.
B.
Incorrect but plausible if the applicant doesnt recognize that the Suppression Pool temperature thermocouples are uncovered below 17.8 ft and doesnt recognize current Suppression Pool level is above the NPSH limit of 13.5 ft.
C.
Correct - Note 2 associated with the step SP/L-4 of T-102 Primary Containment Control informs the procedure user that below 17.8 ft the RHR suction temperature should be used as a valid indicator of Suppression Pool temperature. Note 3 associated with step SP/L-5 informs the procedure user that 13.5 ft is the minimum level for NPSH and Vortex limits.
D.
Incorrect but plausible if the applicant doesnt recognize that current Suppression Pool level is above the NPSH limit of 13.5 ft.
References:
LLOTI 560 Student Ref: required No T-I 02 Bases Primary Containment Control - Bases, Rev. 022, pp.79 &
80 Learning Objective:
IL3 Question source:
Bank (Limerick)
Question History:
Cognitive level:
Memo ry/Fund a m en ta I know ledge:
Co m pre hens ive/Ana lysis :
X 1 OCFR 55.41 X
QUESTION 16 Unit 1 startup is in progress e
e Reactor level is 35 e
e Reactor pressure is 350 psig Reactor Feed Pumps Bypass (LIC-006-120) is in AUTO, and is 60% open Reactor Feed Pump Bypass (LIC-006-138) is in AUTO and is closed Due to a leak in Containment the following occurs:
e Reactor level drops to 15 WHICH ONE of the following describes the automatic DFWLC response?
A.
LIC-006-120 will remain open, LIC-006-138 will open
- 6.
LIC-006-120 will close, LIC-006-138 will remain closed C.
LIC-006-120 will close, LIC-006-138 will open D.
LIC-006-120 will remain open, LIC-O06-138 will remain closed
K&A #
295031 Reactor Low Water Level.
Importance Rating 4.3 QUESTION 16 K&A Statement:
Justification:
A.
A I.I 3-Ability to operate and/or monitor the Reactor Water Level Control as it applies to Reactor Low Water Level.
Incorrect. Plausible if applicant believes the 138 will open on lowering level at this RPV pressure.
B.
Incorrect. Plausible if applicant believes requirements for 120 closure is met and the 138 open requirements are not met.
C.
Incorrect. Plausible if applicant believes the valve swap occurs on low level at this level.
- 0.
Correct. The 120 to 138 valve swap occurs based on 120 valve position and reactor pressure. While the position interlock may be met as level drops, the pressure interlock is not met. Therefore the 138 valve remains closed.
References:
LLOT0550, Rev. 01 8 Student Ref. required No Learning Objective:
Obj 03 Question source:
New Question History:
Cognitive level:
Me m or y/Fu n d a menta I k now I ed g e :
Comprehensive/Analysis:
x 1 OCFR 55.41 X
QUESTION 17 Unit 1 is at 100% power with a half-scram in on "B" RPS due to I&C testing.
Subsequently, power is lost to 1AY160.
Current plant conditions are as follows:
0 Unit 1 is 100% power e
RPV pressure is 1038 psig 0
RPV level is 32" WHICH ONE of the following describes the appropriate procedure to perform and the basis for performing it?
PROCEDURE A.
ON-I 17 LOSS of TECW Loss of 1 AY160 causes loss of TECW
- 6.
ON-I04 Control Rod Problems CRD Flow Controller has lost power C.
OT-I 00 Reactor Low Level Failure of Feed Pump Speed Control D.
T-214 Manual Initiation of ARI Unit 1 has experienced an ATWS
K&A #
295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown Importance Rating 4.1 QUESTION 17 K&A Statement:
K2.03-Knowledge of the interrelations between SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown and ARI/RPT/ATWS Justification:
A.
Incorrect The I&C surveillance has caused a half scram on Channel B. Per E-1AY160 there has been a loss of 1A RPS UPS power. There should have been a FULL SCRAM. Plausible if applicant does not associate response to loss of 1AY160 with SCRAM on A side and/or does not recognize that ATWS takes precedence over TECW.
B.
Incorrect - Loss of 1AY160 will not affect CRD Flow Controllers (E-IAYIGO Section 1.O). Plausible if some other power supply is assumed lost or if associates with RECW/RWCU isolations.
C.
Incorrect but plausible if the applicant does not know that Feedwater has its own, separate UPS.
D.
Correct The I&C surveillance has caused a half scram on Channel B. Per E-1AY160 there has been a loss of 1A RPS UPS power. There should have been a FULL SCRAM. With the reactor at 100% power and with the operator completing applicable portions of OT-I 17, the Mode Switch should be in SHUTDOWN and actions of T-101 in progress. With NO scram functions working, the next appropriate action (directed in T-101) is to manually initiate ARI.
References:
E-1AY16O page 1, OT-I 17 page 3 Student Ref: required No LGSOPS0071 pages 40 and 41 Learning Objective:
N/A Question source:
Bank (Limerick)
Question History:
None Cognitive level:
Memory/Fundamental knowledge:
ComprehensivelAnalysis:
X I OCFR 55.41(5) X
QUESTION 18 WHICH ONE of the following describes the bases for the Technical Specification Limit for the Specific Activity of Primary Coolant?
The Technical Specification primary coolant activity limit ensures that whole body dose limits at the site boundary are not exceeded in the event of a main steam line rupture (1) primary containment and are based on 10 CFR (2) limits.
A.
B.
C.
D.
inside 50 inside 100 outside 50 outside 100
K&A #
295038 High Off-site Release Rate Importance Rating 4.2 K1.02 - Knowledge of Ihe operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE: Protection of the general public Incorrect. Plausible if applicant assumes the TS is based on 10CFR50 limits.
QUESTION 18 K&A Statement:
Justification:
A.
B.
Incorrect. Plausible if applicant assumes an unisolable break inside containment is the higher risk for extended offsite dose due to the potential for core damage and/or breach of containment.
C.
Incorrect. Plausible if applicant assumes the TS is based on 10CFR50 limits.
D.
Correct. Tech Spec 3/4 4.5 limit is based an ensuring the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses resulting from a main steam line break outside containment will not exceed the dose guidelines on 10 CFR Part 100.
References:
Tech spec 3/4 4.5 Student Ref: required N
Learning Objective:
Question source:
Modified Bank (Pilgrim)
Question History:
None Cognitive level:
Memory/Fundamental knowledge:
X Com p re hensive/Ana lysis:
1 OCFR 55.41(12 X
)
QUESTION 19 Unit 1 is at 100% reactor power when the following alarms occur:
e e
MOTOR DRIVEN FIRE PUMP RUNNING REAC I EL 177 PB NE STAIR on FIRE panel WHICH ONE of the following identifies the immediate operator action?
A.
Dispatch the fire brigade 6.
C.
D.
Dispatch the fire brigade leader ONLY Start Diesel Driven Fire Pump Evacuate Unit 1 Reactor Enclosure
K&A #
600000 Plant Fire Onsite Importance Rating 3.8 QUESTION 19 K&A Statement:
Justification:
A.
G2.4.11-Knowledge of the abnormal condition procedures as it relates to PLANT FIRE ON SITE Correct. Per SE-8 immediate operator actions, indications of a fire alarm and fire pump starting require dispatching fire brigage.
B.
Incorrect. Plausible as FB Leader would be dispatched without the brigade if fire alarm without indications of a fire pump running.
C.
Incorrect. Plausible as SE-8 directs starting the diesel driven fire pump if the motor-driven fire pump does not auto start.
D.
Incorrect. Plausible does direct evacuation of the immediate area, but not the entire reactor enclosure.
References:
SE-8, Fire Student Ref: required N
Learning Objective:
LGSOPS-2000 Objective 2 Question source:
Bank (Limerick)
Question History:
LOT 7109 Cognitive level:
Me m or y! F u n d a menta I know I edge :
X Com pre hen sive/Ana lysis :
1 OCFR 55.41(10 X
)
QUESTION 20 Unit 1 is at 100% power with the following plant conditions:
0 HPCl is operating in the full flow test mode on the condensate storage tank for post maintenance testing 0
RHR Loop A is operating in the Suppression Pool Cooling Mode to support the HPCl test Average Suppression Pool temperature is 100°F and steady 0
D11 Diesel Generator has been started for ST-6-092-1 11-1, D11 Diesel Generator 24 Hour Endurance Test 0
The PRO is adjusting D11 Diesel Generator speed so that the sync scope indicator will rotate slowly in the fast direction A grid disturbance causes a ground overcurrent relay to trip on 101 Safeguard Transformer and the following alarm is received:
0 120 F1 101 SAFEGUARD XFMR DlFF GRD LOCKOUT WHICH ONE of the choices below completes the following statement?
Five minutes after receiving the 101 transformer alarm, Suppression pool temperature will be (1) and the D11 Bus will be energized from (2).
0 (2)
A.
B.
C.
D.
Stable 20 1 Safeguards Transformer Stable D11 Diesel Generator I ncreas i ng D11 Diesel Generator Increasing 201 Safeguards Transformer
K&A #
700000 Generator Voltage and Electric Grid Disturbances Importance Rating 3.9 QUESTION 20 K&A Statement:
A I.05 - Ability to operate and/or monitor Engineered Safety Features as it applies to GENERATOR AND ELECTRIC GRID DISTURBANCES Justification:
A.
Incorrect. RHR Pump 1A will not restart on restoration of bus voltage. Pool temperature will increase due to loss of cooling flow. Plausible if applicant thinks diesel will re-energize the bus before loads are shed on undervoltage.
B.
Incorrect. RHR Pump 1A will not restart on restoration of bus voltage. Pool temperature will increase due to loss of cooling flow. Plausible if applicant thinks diesel will re-energize the bus before loads are shed on undervoltage.
C.
Correct. The 101 Safeguards Transformer ground overcurrent fault will trip the Dlx-101 breakers, de-energizing D11 thru D34 Buses. D11 Diesel Generator is ready to load so its output breaker will automatically close onto the D11 Bus after bus voltage has decayed to < 40%. Dll-201 will not close because D11 Bus voltage will be restored before the 1 second transfer closure interlock time delay elapses. RHR Pump 1A will shed on the initial bus undervoltage and will not auto start upon bus voltage restoration.
D.
Incorrect. D11-201 has a 1 second time delay following bus undervoltage before it will close. D11 Diesel Generator output breaker has a shorter 0.5 second time delay. The diesel will re-energize the bus before the closure interlock is satisfied on D11-201 Breaker.
References:
LLOT0370 pages 12&25, S51.8.A, Student Ref: required No LGSOPS0092A Learning Objective:
LLOT0370 # I 4, LGSOPS0092A IL4 Question source:
New Question History:
Cognitive level:
Memory/Fundamental knowledge:
ComprehensivelAnalysis:
X 1 OCFR 55.41(7) X
QUESTION 21 Plant conditions are as follows:
0 85% power 0
C Inboard MSIV has inadvertently closed and RPV pressure peaked at 1060 psig.
0 OT-102, High Reactor Pressure has been entered.
Which one of the following is the basis for reducing reactor power to less than 75% for the failed closed MSIV?
A.
Ensure reactor power does not increase to greater than 100%.
B.
Restore the balanced steam admission characteristics on the HP turbine.
- c.
Ensure reactor pressure transient will not result in SRV actuation.
D.
Restore the margin between actual steam flow and the Group 1 isolation trip point.
K&A #
295015 High Reactor Pressure Importance Rating 3.8 QUESTION 21 K&A Statement:
Justification:
A.
K2.02 - Knowledge of the interrelations between High Reactor Pressure and reactor power.
Incorrect but plausible since the void collapse would result in a power increase.
B.
Incorrect but plausible since this is the basis for power reduction in the event of a TCV or Stop Valve closure.
C.
Incorrect but plausible if candidate is unsure of pressure setpoint where SRVs actuate.
D.
Correct. The basis discussion for step 3.4 of OT-I 02 High Reactor Pressure states this concern as the reason for reducing power to 75%.
References:
LLOTl540, Rev. 09 Student Ref. required No OT-I 02 Bases, Reactor High Pressure - Bases, pg. 4, Rev. 17 Learning Objective:
IL5 Question source:
Bank (Limerick)
Modified distractor c Question History:
NRC-05, Oyster Crk. Cert - 04 Cognitive level:
Memo r y/ F u n d a me n t a I knowledge :
Com p re hensivebna lysis :
X 1 OCFR 55.41 X
QUESTION 22 Unit 2 is at 50% power when the following alarm is received in the MCR:
207 D-4, FWLCS FAILURE Investigation reveals both AFI 00 buses have failed.
WHICH ONE of the following identifies the available Feedwater Control and the response of the Reactor Recirculation pumps?
A.
- 6.
C.
D.
Available Feedwater Control MSC MIA Manual MSC M/A Manual Reactor Recirculation Pump Response 42% Runback 42% Runback 28% Runback 28% Runback
K&A#
295009 Importance Rating 3.9 QUESTION 22 K&A Statement:
Justification:
A I.01 - Ability to operate and/or monitor Reactor Feedwater as it applies to LOW REACTOR WATER LEVEL.
A.
B.
C.
D.
Incorrect but plausible since on a loss of the AFIOO buses the FWLCS will transfer to the Manual Speed Controller. However, rather than RRP runback to 42% speed, a RRP runback to 28% speed should have occurred due to the loss of the AFlOO buses.
Incorrect but plausible if the applicant doesnt know that control has transferred to the MSC on failure of the AFIOO buses. Also the applicant may believe that only the conditions for a runback to 42% RRP speed is required.
Correct - As the result of loss of both AFlOO buses RFP control transfers to the Manual Speed Controller (MSC). Also, as a result of a loss of both AFIOO buses, the RRP should have runback to 28% speed.
Incorrect but plausible if the applicant doesnt know that control is transferred to the MSC on failure of the AFIOO buses.
References:
OT-I 00, Low Reactor Level LLOT0550, Rev. 017 LLOT0540, Rev.024 S06.1.HU/2, Responding to Alarms and Selected Events at the Feedwater Level Control System Operator Station, Rev. 4 Learning Objective:
Obj. 10 (LLOT0550)
Student Ref. required No Question source:
New Q ues t io n H is t o r y :
Cognitive level:
Memory/Fundamental knowledge:
X ComprehensivelAnalysis:
1 OCFR 55.41 X
QUESTION 23 Unit 2 plant conditions are as follows:
Feedwater level control failure has occurred.
0 Reactor level is +75 and steady.
0 Reactor pressure is 500 psig.
WHICH ONE of the following identifies the method for controlling reactor pressure per OT-I I O, Reactor High Level?
A.
Bypass valves B.
HPCI in pressure control mode C.
A, C, N SRVs ONLY D.
B, C, J SRVs ONLY
QUESTION 23 K&A Statement:
K&A #
295008 High Reactor Water Level Importance Rating 3.6 AA1.01 - Ability to operate or monitor [reactor level control as it applies] to HIGH REACTOR WATER LEVEL Note: RPV pressure control under high level conditions provides level control by removing inventory.
Justification:
A.
Correct. Level is high but below the steam lines. Nothing in the stem indicates the condenser is not available. Therefore bypass valves are available and acceptable for use.
- 6.
Incorrect. HPCI trips at +54 high level. OT-1 IO does not allow or direct bypassing the high level trip. Plausible because HPCI can be used under other conditions to control RPV pressure and level.
C.
Incorrect. Any SRV may used under these conditions. Plausible if applicant confuses the OT-I 10 SRVs with the RSP SRVs.
D.
Incorrect. Any SRV may used under these conditions. Plausible because B, C, and J are used in OT-1 10 when RPV level is > 118 (flooding MSLs).
References:
OT-1 I O, Reactor High Level Student Ref. required No Learning Objective:
LLOTI 540, Obj. #5 Question source:
New Question History:
Cognitive level:
Memory/ F u n d a menta I know I e d g e :
Com pre hens ive/Ana lysis:
X 1 OCFR 55.41 X
QUESTION 24 Given the following conditions:
0 Reactor power is 18%
An ATWS is in progress on Unit 1 RPV level has been intentionally lowered to -75 in accordance with T-I 17, Level / Power Control WHICH ONE of the following explains the reason for this RPV level reduction?
A.
Concentrate boron in the core B.
Increase void fraction inside the shroud C.
Increase preheating of the incoming feedwater D.
Reduce natural circulation driving head through the core
K&A #
29501 5 Incomplete SCRAM Importance Rating 3.8 QUESTION 24 K&A Statement:
Justrfication:
A.
Incorrect. Plausible because reducing feed will limit dilution.
G2.4.9 - Knowledge of low power/shutdown implications in accident mitigation strategies, as it relates to Incomplete SCRAM.
B.
Incorrect. Plausible because void fraction could increase due to lowered head of column of water.
C.
Correct. Feedwater spargers are uncovered by lowering level, which increases feedwater preheating. Water is therefore less cold and dense which will lower reactor power.
D.
Incorrect. Plausible in that driving head will decrease with lower level. However, this is not the basis.
References:
LLOTI 560, Rev. 12 Student Ref. required No T-I17 Bases, LeveVPower Control -
Bases, pg. 7, Rev. 12 Learning Objective:
IL6 Question source:
Bank (Limerick)
Question History:
NRC-05, Oyster Crk. Cert - 04 Cognitive level:
Memory/Fundamental knowledge:
Com pre hens ive/Ana I ys is:
X 1 OCFR 55.41 X
QUESTION 25 Unit 2 plant conditions are as follows:
0 2B Containment Hz Recombiner is in operation for post-maintenance testing.
RPV Wide Range Level Transmitter LT-042-2N081 D fails to an indicated level of -42.
Subsequently, RPV Wide Range Level Transmitter LT-042-2N081 C fails off-scale low.
WHICH ONE of the following identifies the current status of these valves associated with the Drywell Unit Coolers and Hydrogen Recombiners?
A.
A Drywell Chilled Water Return - OPEN B Drywell Chilled Water Supply - OPEN Drywell Supply to B Recombiner - OPEN B.
A Drywell Chilled Water Return - OPEN B Drywell Chilled Water Supply - OPEN Drywell Supply to B Recombiner - CLOSED C.
A Drywell Chilled Water Return - CLOSED B Drywell Chilled Water Return - CLOSED Drywell Supply to B Recombiner - OPEN D.
A Drywell Chilled Water Supply - CLOSED B Drywell Chilled Water Supply - CLOSED Drywell Supply to B Recombiner - CLOSED
K&A #
295020 Inadvertent Containment Isolation Importance Rating 3.1 QUESTION 25 K&A Statement:
K2.03 - Knowledge of the interrelations between Inadvertent Containment Isolation and Drywell/containment ventilationlcooling.
Justification:
A.
Incorrect but plausible if the applicant believes that both the Hydrogen Recombiners and cooling water to the Drywell Cooler Units isolate on a Level 1 (-129).
B.
Correct - Cooling to the Drywell Cooler Units isolates on a Level 1 (-129) signal on actuation of either trip channels A & B or C & D. Level 1 has only been reached on trip channel C (LT-042-2N081 C). Drywell Supply to Hydrogen Recombiner B isolates on Level 2 (-38) signal on trip channel D.
C.
Incorrect but plausible if the applicant believes cooling to the Drywell Cooler Units isolate on Level 2 and that Drywell Supply to Hydrogen Recombiner B receives a close signal either when Level 1 is reached or a Level 2 is reached on trip channel C.
D.
Incorrect but plausible if the applicant believes that both systems isolate on Level 2 signal.
References:
- LLOT0180, Rev. 015 Student Ref. required No
- SSN Drawing ## 058-01, Rev. 2
- LGSOPS0042, Rev. 000
- GP-8.1, Automatic Actuations by Isolation Signals, Rev.14
- S58.1A, Placing Containment Hydrogen Recombiners in Ready Mode Rev,. 008 Learning Objective:
Question source:
Question History:
Cognitive level:
1 OCFR Obj 2, (LLOTO180)
IL-9 (LGSOPS0042)
Modified Bank (Limerick)
(Modified) Changed trip signal and s ys tem s affected NRC-05, Oyster Crk. Cert - 04 Memory/Fundamental knowledge:
ComprehensivelAnalysis:
X 55.41 X
QUESTION 26 Unit 1 was at 100% when the seismic event occurred Unit 1 plant conditions are as follows:
0 Reactor is scrammed 0
Div 1 125 VDC is de-energized 0
Div 2 125 VDC is de-energized A steam leak in the HPCl room has been confirmed 0
HV-55-1 FOO2 HPCI Steam Line Inboard Isolation Valve failed to isolate.
Conditions in the HPCl Equipment room are as follows:
Room temperature is 230°F and rising.
Radiation levels are 0.8 mr/hr and rising.
WHICH ONE of the following describes the status of gaseous activity levels in the Unit 1 HPCl Pump Room and the plant response?
A.
B.
C.
D.
Status of Gaseous Activity Level Quickly stabilize in the HPCl Room.
Continue to rise in the HPCl room and a filtered release to the South Stack exists.
Continue to rise in the HPCl Room and is confined to the room.
Continue to rise in the HPCl room and a filtered release to the North Stack exists.
Plant Response The release was terminated by the automatic isolation of the HPCl Steam Supply Line on high room temperature.
HPCl Steam Flooding Dampers closed on high temperature in the HPCl Room.
HPCl Steam Flooding Dampers closed on loss of power.
The RE HVAC Supply and Exhaust dampers closed on a loss of power.
K&A ##
295032 High Secondary Containment Temperature Importance Rating 3.6 QUESTION 26 K&A Statement:
K1.02 - Knowledge of the operational implications of Radiation releases as it applies to HIGH SECONDARY CONTAINMENT TE M PERATU RE.
Justification:
A.
Incorrect but plausible since normally HV-55-FO03 HPCI Steam Line Outboard Isolation Valve would close on high room temperature. However, the signal for closing HV-55-FO03 is powered from Div 2.
- 6.
Incorrect but plausible since the Steam Flooding Isolation Dampers normally close on a high d/P signal. However, with the loss of Div 1 and Div 2 the solenoids to the actuating arm will be de-energized. These solenoids need to energize to release the actuating arm for the dampers.
C.
Incorrect but plausible since most all of the dampers associated with the RE HVAC system close on a loss of DC power. However, the solenoids associated with the Steam Flooding Isolation Dampers need to energize to release the actuating arm for the dampers.
D.
Correct - Conditions within the HPCl equipment room will continue to worsen. HV-155-FO03 receives its isolation signal from Div 2, which means with HV-55-1 F002 failed open, an unisolable leak path exists from the ruptured HPCl steam supply line into the HPCl Equipment Room. The RE HVAC system dampers are supplied by 125V DC from Div 1 and Div 2 and on a loss of power the exhaust and supply dampers would close. This would confine the steam and radiation within the RE HVAC and eventually would reach the North Stack through the filtered flowpath provided by the SGTS.
References:
LLOT0200, Rev. 01 8 LLOT0340, Rev. 25 Learning Objective:
Obj 7, (LLOT0200)
Obj. 14 (LLOT0340)
Question source:
New Question History:
Cognitive level:
Student Ref. required No Me m or y/Fu n d a menta I know I ed g e :
Comprehensive/Analysis:
X IOCFR 55.41 X
QUESTION 27 Plant conditions are as follows:
0 Unit 1 is in a refueling outage performing RPV pressure test.
0 Secondary containment is established in Unit 1 Reactor Enclosure and the Refuel Floor with Zones I and Ill cross-tied.
0 Unit 2 is at 100% power.
0 Secondary containment is established in Unit 2 Reactor Enclosure.
The following annunciators are received in the MCR:
109 B-4, REACTOR ENCL AREA HI RADIATION 109 E-I, 1 REAC ENCL REFUEL FLR VENT EXHAUST RAD MON A/B HI - HI / DOWNSCALE Rising radiation levels are noted on Panel 1OC605 WHICH ONE of the following describes the status of Unit 1 and Unit 2 Reactor Enclosure HVAC 1 minute later?
Unit 1 RE HVAC Unit 2 RE HVAC A.
Isolated Isolated B.
Not Isolated Not Isolated C.
Not Isolated Isolated D.
Isolated Not Isolated
K&A #
295034 Secondary Contain men t Ventilation High Radiation Importance Rating 3.9 QUESTION 27 K&A Statement:
K2.04-Knowledge of the interrelations between SECONDARY CONTAINMENT VENTILATION HIGH RADIATION and Secondary Containment ventilation Justification:
A.
Incorrect. Plausible if applicant thinks both units HVAC would respond to the HI-HI rad condition on a Unit 1 instrument. The refuel floor HVAC is common and an isolation signal on one unit would also trip the same division on the other units logic.
Possible that applicant would believe that a similar response would occur for reactor enclosure HVAC.
B.
Incorrect. It only takes one HI-HI signal to isolate reactor enclosure HVAC, so Unit 1 reactor enclosure HVAC is isolated. Plausible if the applicant thinks that two (2) Hi-HI inputs are required for an isolation on Unit 1.
C.
Incorrect. Plausible if applicant thinks that Unit 1 A/B monitors are downscale based on the alarm, which would not cause an isolation on Unit 1. Applicant may believe that Unit 2 would be affected due to venting condition.
D.
Correct. It only takes one HI-HI signal to isolate reactor enclosure HVAC. Reactor enclosure isolation is divided into two divisions. Each division controls one set of dampers. An isolation signal on either division will result in a full isolation of the reactor enclosure ventilation. Unit 2 reactor enclosure ventilation is separate from Unit 1. The isolation signal present is on Unit 1; therefore, Unit 2 ventilation is not affected.
References:
LLOT0200 pg. 6, 7, 28, 29; ARC Student Ref: required No 10800 E-I, 1 OC800B-4 Learning Objective:
LLOT0200 Obj. 3 Question source:
Modified Bank (Susquehanna)
Question History:
None Cognitive leve I :
Memory/Fundamental knowledge:
X Com pre hensive/Ana lysis:
1 OCFR 55.41(9) X
QUESTION 28 The following describes the initial Unit 2 plant conditions:
0 Reactor is at 100% power.
0 RHR Loop A is running in Suppression Pool Cooling mode 0
RHR A discharge pressure is 215 psig.
Subsequently, a reactor coolant leak occurs on Unit 2 and the following conditions exist:
0 Reactor pressure is 500 psig and dropping 0
Reactor level is -100 and dropping Drywell pressure is 15.3 psig and going up WHICH ONE of the following describes the current position of the following RHR Loop A valves:
0 0
0 0
HV-C-51-2F048A, 2A RHR Htx Shell Side Bypass Valve HV-51-2F024A, 2A RHR Pp Full Flow Test Return Valve HV-51-2F007A, 2A RHR Pp Min Flow Valve HV-51-2F017A, 2A RHR LPCl Injection PClV A.
F048A - Open F024A - Closed F007A - Open F017A - Closed B.
F048A - Closed F024A - Closed F007A - Open F017A - Closed C.
F048A - Open F024A - Closed F007A - Closed F017A - Open D.
F048A - Closed F024A - Open F007A - Closed F017A - Closed
K&A #
203000 RHR/LPCI:
Injection Mode Importance Rating 4.2 QUESTION 28 K&A Statement:
K4.01 - Knowledge of RHWLPCI: INJECTION MODE (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for Automatic system initiation/ injection.
Incorrect but plausible since this would be the status of these valves if just high drywell pressure was an initiation signal.
Justification:
A.
B.
Incorrect but plausible if the applicant believes high drywell pressure was an initiation signal and that the signal only repositioned valves that rob flow from LPCl flowpath. Following this premise, the heat exchanger bypass and LPCl injection valve would not reposition until conditions were met for LPCl injection valve opening (i.e., 74 psid across injection valve).
C.
Incorrect but plausible if the applicant believes that high drywell pressure was an initiation signal and that the LPCl Injection valve opens immediately on an initiation signal.
D.
Correct - LPCl initiation needs either reactor level less than -1 29 OR Drywell pressure greater than 1.68 psig along with reactor pressure less than 455 psig and less than 74 psid across the 17 valve. At the current conditions, none of the valves would have repositioned yet.
References:
LLOT0370, Rev. 01 7 Student Ref. required No Learning Objective:
Obj. 6, Obj. 8 Question source:
New Question History:
Cognitive level:
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
1 OCFR 55.41 X
QUESTION 29 Given the following Unit 2 plant conditions:
0 RHR loop 6 is operating in the Shutdown Cooling (SDC) Mode 0
Reactor coolant temperature is at 290 F and slowly rising 0
RPV pressure is 45 psig and slowly rising The following alarms are received:
0 222 D22 A2 201 D22 Bus breaker trip 222 D22 A I D22 bus diff/overcurrent lockout 0
222 D22 61 D22 safeguard bus undervoltage 222 D22 63 D224 load center xfmr breaker trip 0
222 D22 C4 D22 Diesel Running WHICH ONE of the following describes the automatic response of the Shutdown Cooling Suction Isolation Inboard and Outboard Valves (HV51-2FOO9 and HV51-2F008) when reactor pressure exceeds 75 psig.
A.
HV-51-2FOO9 will remain in its current position. HV-51-2FOO8 will close.
- 6.
HV-51-2FOO9 and HV-51-2FOO8 will remain in their current positions.
C.
HV-51-2FOO9 and HV-51-2FOO8 will BOTH close.
D.
HV-51-2FOO8 will remain in its current position. HV-51-2FOO9 will close.
K&A #
205000 Shutdown Cooling Importance Rating 3.3 QUESTION 29 K&A Statement:
Justification:
A.
K6.05 - Knowledge of the effect that a loss or malfunction of AC electrical power will have on the SHUTDOWN COOLING system Incorrect but plausible if the applicant does not associate the loss of Bus 22 to a failure of F008; rather than FOO9.
- 6.
incorrect but plausible if applicant thinks both valves are powered from D22.
C.
Incorrect but plausible if applicant thinks neither valve is powered from D22 or if applicant thinks the EDG has powered the bus.
D.
Correct. Alarms indicate lockout conditions on Bus D22 with the bus de-energized.
2F008 is powered from D22 bus and will not operate. RPV pressure will increase from decay heat input with loss of Loop B RHR cooling. 2F009 will isolate on pressure interlock at 75 psig RPV pressure.
References:
LLOT0370 pages 25 Student Ref: required No Learning Objective:
LLOT0370 #I 4
e S51.I
.7.B, Defeating The Rhr Shutdown Cooling Auto Isolation, Rev 007, Steps 4.2.3, 4.2.4 (power supplies)
Question source:
Question History:
Cognitive level:
1 OCFR Bank (Perry)
Chgd pwr supply to fit Limerick Memory/Fundamental knowledge:
Comprehensive/Analysis:
X 55.41(7) X
QUESTION 30 Unit 2 plant conditions are as follows:
- r. A feedwater line break on the "B" header inside containment results in a reactor scram on low level
- r. HPCl and RClC both running
- r. RPV Level is + I O " and going up A DC distribution panel low voltage alarm is received (2PPB1/2PPB3 125 VDC DlST PANELS UNDERVOLTAGE -Window G-4 on AR-222).
The crew determines that Div 2 DC is deenergized.
WHICH ONE of the following identifies the expected plant conditions?
A.
HPCl turbine is running, RPV level is going down B.
HPCl turbine is running, RPV level is going up.
C.
HPCl turbine is NOT running, RPV level is going down.
D.
HPCl turbine is NOT running, RPV level is going up.
K&A #
206000 HPCl Importance Rating 2.8 QUESTION 30 K&A Statement:
K2.03 - Knowledge of the electrical power supplies to initiation logic KA match. HPCl initiation logic loses power. HPCl not able to trip on high level or start on low level. Applicant must realize loss of initiation logic will not trip the turbine under the given circumstances.
Justification :
A.
Incorrect but plausible. HPCl does not trip on loss of Div II DC. Plausible a failure low of the flow controller would decrease speed to -750 rpm. However, a loss of DC will close the governor valve, reducing speed to zero.
B.
Incorrect but plausible. Level would not increase because neither HPCl nor RClC is injecting. Plausible if applicant thinks HPCl flow controller not affected by loss of Div II DC.
C.
Correct - Div II provides HPCl initiation logic power. HPCl will have started post trip as expected on level decrease below Level 2 (-38.5 inches) based on a low level scram from full power at Level 3 (+12.5 inches) with loss of both reactor feedwater pumps. The subsequent loss of DC power to 2BD102, Division II distribution Panel will disable automatic HPCl initiation on low level, render HPCl turbine trips inoperable and remove control power from the HPCl flow controller, resulting in closure of the HPCl governor valve. RClC would be discharging through the break and not contributing to vessel inventory.
D.
Incorrect but plausible. HPCl does not trip on loss of Div 11 DC. Plausible if applicant thinks RClC is injecting.
References:
ARC-MCR-222 G4; LLOT0340 rev. 25; Student Ref: required No LLOT0690, rev. 12 Learning Objective:
LLOT0690 Obj. 6c; LLOT0340 obj. 14a Question source:
Modified Bank (Limerick)
Question History:
Cognitive leve I:
Me m o r y/ F u n d a m e n t a I know led g e :
Com p re hen sive/Ana lysis:
X 1 OCFR 55.41(7) X
QUESTION 31 Unit 1 is at 100% power hl len the following occurs:
0 0
RPV level is -50".
Reactor scrams on low reactor level.
The following HPCl indications are observed:
HPCl flow controller in AUTO with a setpoint of 5600 gpm.
0 HPCl Initiation Signal White SEAL IN light is lit.
0 FV-56-112, HPCI Turbine Stop Valve is OPEN.
e FV-56-111, HPCl Turbine Control Valve is CLOSED.
Which of the following explains the HPCI system response?
A.
Speed Feedback to Turbine Speed Governor Controller fails low.
B.
Ramp Generator output signal fails low C.
Auxiliary Oil Pump failed to start.
D.
The Turbine Trip Solenoid Valve is energized.
I-
- - - I r -
A
K&A #
206000 High Pressure Coolant Injection 1 m portance Rating 3.4 QUESTION 31 K&A Statement:
K4.11 - Knowledge of HIGH PRESSURE COOLANT INJECTION design feature(s) andlor interlocks which provide for Turbine Speed Control.
Justification:
A.
Incorrect but plausible since if the applicant did not understand how a failure of this would effect the Turbine Speed Governor Controller. However, a low failure of the speed feedback signal would result in generating in an open signal to the HPCl Turbine Control Valve (TCV) and speed and flow to increase.
- 6.
Correct - The ramp generator overrides the flow controller during turbine startup to allow a controlled rate of acceleration. Once the ramp generator output exceeds the signal from the flow controller, the controller takes over and maintains control until the ramp generator is reset. The ramp generator is reset whenever the turbine stop valve is fully closed. With the ramp generator signal failed low, its output will never exceed the flow controller and will not reset.
C.
Incorrect but plausible since failure of the Auxiliary Oil pump would lead to no control oil getting to the TCV and the Turbine Stop Valve (TSV). However, since the initial conditions listed in the question state that the Turbine Stop Valve is open then this couldnt be the cause of the problem.
D.
Incorrect but plausible since the turbine trip solenoid valve energized would close both the turbine stop and turbine control valves.
References:
LLOTO340, Rev. 025 Student Ref. required No Learning Objective:
Obj. 6b, Obj 10 Question source:
New Question History:
Cognitive level :
1 OCFR MemorylFundamental knowledge:
Com pre hen sive/Ana lysis :
x 55.41 X
QUESTION 32 Unit 1 Plant conditions are as follows:
0 Reactor Power is 100%
0 Core spray pumps 1 B and 1 D are operating in the full flow test mode Core spray suction valve HV-52-1 Fool D position indication limit switch fails, producing a signal corresponding to an intermediate valve position.
WHICH of the following describes the impact on the 1 D Core Spray Pump, based on the above conditions?
A.
B.
C.
D.
Pump will continuing running Automatic Start is Available Manual Start is Available Pump will continuing running Automatic Start is NOT Available Manual Start is Available Pump will Trip CORE SPRAY OUT OF SERVICE Annunciator will actuate Automatic Start is NOT Available Manual Start is Available Pump will Trip CORE SPRAY OUT OF SERVICE Annunciator will actuate Automatic Start is NOT Available Manual Start is NOT Available
K&A #
209001 LPCS Importance Rating 2.5 QUESTION 32 K&A Statement:
A I.08 -Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including system lineup.
Justification:
A.
Correct. The pump will continue running. Automatic and manual pump starts are not affected based on suction valve position.
B.
Incorrect but plausible. The pump will conttnue to run. Plausible if the applicant believes that an interlock exits to protect the pump if the suction valve is not fully open.
C, Incorrect - The pump continue to run. The annunciator does not alarm. Plausible if the applicant believes that the pump would trip based on valve position to protect the pump if the suction valve is not fully open and that an interlock exits on the automatic pump start.
D.
Incorrect - The pump continues to run. The annunciator does not alarm. Automatic and manual starts are not affected. Plausible if the applicant believes that the pump trips based on suction valve position and that interlocks exist on pump start.
References:
LLOT0350 pg. 17, 18, S52.5.8, rev. 26 Student Ref: required N
Core Spray Header Flush and Vessel Floodup, Dwg. 052-01 rev.2; Dwg.
8031-M-52 rev. 48 Learning Objective:
LLOT0350 Obj. #6, 7, 8 Quest ion source :
Modified Bank (Hope Creek)
Question History:
None Cognitive level:
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
1 OCFR 55.41(7) X
QUESTION 33 Unit 1 plant conditions are as follows:
0 D13 bus is locked out 0
Reactor level is -140 0
Reactor pressure is 370 psig Which one of the following describes the status of the 1 B and 1 D Core Spray Pumps?
A.
Injecting approximately 3,000 gpm.
B.
Injecting approximately 6,000 gpm.
C.
Not injecting, only one Core Spray pump is running on min flow.
D.
Not injecting, both Core Spray pumps are running on min flow.
K&A ##
209001 Low Pressure Core Spray Importance Rating 3.7 QUESTION 33 K&A Statement:
Justification:
A.
A3.04 - Ability to monitor automatic operations of the Low Pressure Core Spray controls including system flow.
Incorrect but plausible since this would be the approximate flow if either of the 1 B Loop pumps received power from the D13 bus and if applicant confuses the setpoint for the injection valves opening (455 psig) with the pump shutoff head (-330 psig),
then design flow could be reached at reactor pressure equal to 370 psig.
B.
Incorrect but plausible since this would be the approximate flow if both the 1 B Loop pumps were injecting into the vessel and if applicant confuses the setpoint for the injection valves opening (455 psig) with the shutoff head (-330 psig) then design flow could be reached at reactor pressure equal to 370 psig.
C.
Incorrect but plausible since this would be the status of the pumps if either of the 1 6 Loop pumps received power from the D13 bus.
D.
Correct-Neither the 1 B or 1 D Core Spray pumps would be effected by the loss of 013 bus. However, the current reactor pressure exceeds the shutoff head (-330 psig) of the Core Spray pumps.
References:
LLOT0350, Rev. 01 5 Student Ref. required No Learning Objective:
Obj 5, Obj 9 Question source:
Modified Bank (Limerick)
(Modified) Changed affected Core Spray loop (1 B), changed reactor pressure from 230 psig to 370 psig, and removed any reference to a specific pumps within distractors.
Question History:
NRC-05, Oyster Crk. Cert - 04 Cognitive level:
Memory/Fundamental knowledge:
ComprehensivelAnalysis:
X 1 OCFR 55.41 X
QUESTION 34 Unit 1 conditions are as follows:
0 A Group I Isolation has occurred.
0 Reactor power is 15%.
0 SRVs are cycling AUTOMATICALLY to control reactor pressure.
All three SLC pumps have been manually started from the Control Room with the following system indications:
0 SLC Tank level is 3,800 gallons 0
SLC Pump discharge pressure are:
0 SQUIB READY status lights indicate as follows:
A = 1 100 psig A = OFF 6 = OFF C = OFF B = 1 195 psig C = 1405 psig Five minutes following manual initiation of SLC, annunciator MCR 108 1-2, STANDBY LIQUID TANK HVLO LEVEL, is received. SLC Tank level is 3,585 gallons.
Given these conditions, which one of the following describes the SLC pumps that are injecting boron solution into the RPV?
Pump A Pump B Pump C A.
Injecting Injecting Not Injecting B.
Injecting Injecting Injecting C.
Not injecting Not injecting Injecting D.
Not injecting Injecting Not injecting
K&A #
21 1000 Standby Liquid Control Importance Rating 3.8 QUESTION 34 K&A Statement:
K4.04 - Knowledge of STANDBY LIQUID CONTROL system design features and/or interlocks which provide for indication of fault in explosive valve firing circuits.
Justification:
A.
- 6.
C.
D.
NOTE: For this event, A SLC pumps squib valve opened but its discharge pressure is too low to inject info the reacfor vessel and C squib valve has lost continuity but the squib valve did not fire.
Incorrect but plausible since the SQUIB READY status lights for squib valves A and B are out which is one possible indication that boron is injecting. However, A SLC pump discharge pressure is below reactor pressure, so it is not injecting. For C SLC pump, although its continuity light is out, the high discharge pressure of 1405 psi would indicate that flow is not getting through C SLC injection line.
Incorrect. Plausible if applicant only used SQUIB READY status lights OFF as confirmation that all SLC pumps are injecting into the core. Also if the applicant did not recognize that with reactor pressure being controlled by SRVs that at a minimum SLC pressure must exceed the SRV lift setpoints (1 170 to 1 I90 psig).
Incorrect. Plausible since it is accurate for A SLC pump given that reactor pressure is greater than A SLC pump discharge pressure. Even though continuity light is out, A SLC pump would not be injecting. Also. applicant may confuse where reactor pressure is being controlled by SRVs. May believe that B SLC pump would not be injecting into the core as well. With only one pump injecting, this would be consistent with the level change observed on the SLC tank (See discussion below.).
Correct. Only B SLC pump is injecting into the core. Anormal pressures on A and C pumps. Volume change provides confirmation that only one pump is injecting.
The volume change over the five minute injection period is 215 gallons (3800 -
3585). Each pump is designed for 43 gpm. This indicates that only one SLC pump can be injecting into the vessel (i.e. 43 gpm x 5 min = 215 gallons.)
References:
LGSOPS0048, Rev. 000 Student Ref. required No Learning Objective:
E04, IL2, IL6 Question source:
New Question History:
Cog n it ive level :
Me m o r y/ F u n d a m en t a I know I ed g e :
Comprehensive/Anal ysis:
X 1 OCFR 55.41 X
QUESTION 35 Unit 1 plant conditions are as follows:
0 Reactor power is at 97% power coasting down for a refueling outage 0
RPS SCRAM Functional testing is in progress 0
Group AI and A4 white lights are extinguished on panel 10C603 0
Group A2 and A3 white lights are lit on panel 10C603 Group BI RPS solenoid power fuse blows.
WHICH ONE of the following identifies the status of Control Rods ten (1 0) seconds after the fuse blows, assuming no operator action?
A.
No rods inserted B.
45 rods inserted C.
93 rods inserted D.
185 rods inserted
K&A #
212000 RPS Importance Rating 3.2 QUESTION 35 K&A Statement:
A4.04 - Knowledge of the effect that a loss or malfunction of REACTOR PROTECTION SYSTEM will have on RPS logic channels.
Justification:
A.
Incorrect RPS K14 relay is de-energized for Group AI and A4 control rods based on white light indication. Blown fuse on Group BI circuit results in venting of air from scram pilot valves for Group 1 control rods (45 rods SCRAM). Plausible if applicant does not recall the RPS logic or impact of blown fuse.
- 6.
Correct - In a coast-down all rods would be expected to be full out (position 48).
RPS K14 relay is de-energized for Group AI and A4 control rods based on white lights being extinguished. A side scram pilot valves have repositioned for control rod Groups 1 and 4. Blown fuse on Bl results in de-energizing Blre1ay. This repositions the second set of scram pilot valves on Group I control rods and vents air from Group 1 scram valves. Venting air from the scram valves results in scramming of Group 1 control rods. There are 45 rods in Group 1.
C.
Incorrect - RPS K14 relay is de-energized for Group AI and A4 and has repositioned A scram pilot valves for Group 1 control rods. Blow fuse on BI de-energizes BI relay only and results in repositioning of BI scram pilot valves.
Plausible if applicant does not understand that the blown fuse does not affect the B4 control rod circuitry (which would result in scram of 48 rods).
D.
Incorrect - RPS K14 relay is de-energized for Group AI and A4 control rods and has repositioned A side scram pilot valves Blow fuse.on BI de-energizes 81 relay only (45 rods insert). Plausible if applicant thinks both Group 1 and 4 SCRAM.
References:
LGSOPS0071, Dwg. 071-01a, 071-03a Student Ref: required N
Learning Objective:
LGSOP0071 E04 Question source:
Modified Bank (Limerick)
Question History:
None Cognitive I eve I:
Memo r y/Fu n d a me n t a I know I ed g e :
Com pre hens ive/Ana lysis:
X IOCFR 55.41(7) X
QUESTION 36 A Unit 2 turbine trip from full power caused a reactor scram.
RPV water level lowered to -20 during the initial transient.
Level has been restored to +35.
The scram has NOT been reset.
WHICH ONE of the following describes the status of the RPS Backup Scram valves?
A.
B.
C.
D.
Energized and aligned to vent the scram air header De-energized and aligned to vent the scram air header Energized and aligned to supply air to the scram air header De-energized and aligned to supply air to the scram air header
K&A #
212000 RPS Importance Rating 3.4 QUESTION 36 K&A Statement:
A I.08 -Ability to predict and/or monitor changes in parameters associated with operating the REACTOR PROTECTION SYSTEM including valve position.
Justification:
A.
Correct. Backup SCRAM valves are normally deenergized. Backup SCRAM valves energize to operate and reposition to vent the SCRAM air header. Since the SCRAM has not been reset, even though the SCRAM condition cleared, the valves would still be energized and venting.
B.
Incorrect. Back up SCRAM valves energize to close. Plausible if the applicant believes that the backup SCRAM valves are de-energize to reposition, similar to the SCRAM valves.
C.
Incorrect. Backup SCRAM valves are energize to operate. The backup SCRAM valves receive an initiation signal on the SCRAM and would reposition. Since the SCRAM has not been reset, even though the SCRAM condition cleared, the valves would still be energized and venting. Plausible if the applicant believes that once the scram condition has cleared the valves would return to their de-energized state.
D.
Incorrect. The backup SCRAM valves have received an initiation signal and would have energized and closed. Plausible if applicant believes that the valves are de-energize to operate.
References:
LGSOPS 0071 pg. 22 Student Ref: required N
Learning Objective:
LGSOPS E04, IL7 Question source:
Modified Bank (Duane Arnold)
Question History:
None Cog nit ive level :
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
1 OCFR 55.41(7) X
QUESTION 37 A startup is in progress on Unit 2:
IRM Channel D failed upscale earlier in the shift and has been bypassed.
B IRM is reading 30 on Range 4 The Reactor Operator selects Range 3 on the B IRM Range Switch.
Which of the following describes the expected plant response (if any)?
A.
B.
NO rod blocks or scram signals generated.
Control Rod block and half-scram generated.
C.
Control Rod block and full-scram generated.
D.
Control Rod block generated and NO scram signal.
K&A #
21 5003 IRM Importance Rating 3.6 QUESTION 37 K&A Statement:
K3.02 - Knowledge of the effect that a loss or malfunction of the INTERMEDIATE RANGE MONITOR system will have on Reactor Manual Control.
Justification:
A.
Incorrect but plausible since this is true. However, it is not the only response to the event. Placing IRM B to Range 3 would raise the value by a factor of 3.16 (3.16 x 30 = 94). The new value of 94 would exceed the upscale setpoint of 85/125 for IRM B and a control rod block would be generated by RMCS.
- 8.
Incorrect but plausible since a rod block would occur and if the applicant believes with one IRM inop (IRM D) and that IRM B exceeded the scram setpoint (1201125) then a half-scram would be generated on RPS trip system B. However, with IRM D bypassed it is removed from the trip circuit and the new value of 94 would not reach the scram setpoint (1201125) for IRM 6. In addition, the two IRMs are on the same trip system.
C.
Incorrect but plausible since a rod block would occur and if the applicant believes with one IRM inop (IRM D) and that IRM B exceeded the scram setpoint (120/125) then a full -scram would be generated with a tripped detector one each RPS trip system the same. However, with IRM D bypassed it is removed from the trip circuit and the new value of 94 would not reach the scram setpoint (120/125) for IRM 6.
In addition, the two IRMs are on the same trip system.
D.
Correct - IRMD is bypassed. Therefore no trips or blocks will be generated from IRM D. IRM B will indicate 94 of 125 scale, which is greater than the 85/125 scale for a rod block signal generated by RMCS but less than the 120/125 scram signal generated by RPS.
References:
LLOT0250, Rev. 01 2 Student Ref. required No Learning Objective:
Obj 10 Question source:
New Question History:
Cog nit ive level :
Memo ry/F u n d a m e n t a I know I ed g e:
Comprehensive/Analysis:
X 1 OCFR 55.41 X
QUESTION 38 Unit 2 is in a refueling outage with the following conditions established:
0 Mode switch is in REFUEL 0
Control rod 30-31 is at position 24; all others are at 00 0
Fuel is being moved in the spent fuel pool 0
Shorting links are installed An I&C Technician, troubleshooting a problem on the A Source Range Monitor (SRM), moves the drawer Mode switch out of OPERATE.
WHICH ONE of the following describes the effects on the SRMs and Control Rod 30-31?
Effect on SRMs Effect on Control Rod 30-31 A.
SRM Downscale Alarm No effect B.
SRM Downscale Alarm Cannot be withdrawn C.
SRM Upscale/lnop Alarm Cannot be withdrawn D.
SRM Upscale/lnop Alarm Rapidly inserts on SCRAM signal
K&A #
215004 SRM Importance Rating 3.3 QUESTION 38 K&A Statement:
A3.02-Ability to monitor automatic operations of the.Source Range Monitor system including annunciator and alarm conditions.
Justification:
A.
Incorrect but plausible if the applicant does not understand the SRM out of OPERATE will result in an Upscale alarm.
B.
Incorrect but plausible if the applicant does not understand the SRM out of OPRATE will result in an Upscale alarm. Downscale alarm does not cause a Rod Block.
C.
Correct - With the shorting links installed, one SRM upscale will cause a Rod Block.
D.
Incorrect but plausible if the applicant does not understand that the control logic requires 2 out of 4 SRM channels upscale to initiate a SCRAM
References:
LLOT0240 pages 16 Student Ref: required No Learning Objective:
N/A Question source:
Bank (Limerick)
Question History:
None Cognitive level:
Memo ry/F u n d a m e n t a I know I ed g e:
X Com pre hensive/Ana lysis :
I OCFR 55.41(7) X
QUESTION 39 A unit 2 plant startup is in progress per GP-2 with the following conditions:
0 Reactor Mode Switch in STARTUP All lRMs on Range 10 e
APRM #I 15.2%
0 APRM #2 12.8%
APRM #3 1 I
.5%
0 APRM #4 1 I
.8%
WHICH ONE of the following describes the status of RPS and RDCS for the conditions above?
Status of RPS Status of RDCS A.
Scram No rod block B.
Scram Withdraw rod block C.
No scram No rod block D.
No scram Withdraw rod block
K&A #
21 5005 APRM/LPRM Importarice Rating 3.6 QUESTION 39 K&A Statement:
A4.06 - Ability to manually operate and/or monitor in the control room: Verification of proper functioningloperability.
Justification:
A.
Incorrect. Plausible if applicant thinks scram logic made up but rod block logic is not.
B.
Incorrect. Plausible if applicant thinks both logic circuits each require just a single high input for actuation.
C.
Incorrect. Plausible if applicant does not think the given values exceed the logic trip values.
D.
Correct. Withdrawal Rod block only generated at 12%-13% Rx power with mode switch not in Run for at least 1 APRM in trip. Rx scram at 15%-20% with mode switch not in Run for at least 2 APRM's in trip.
References:
LLOT0275, Rev. 004, TS 2.2.1, GP-2 Student Ref. required No Learning Objective:
Obj 10, Obj 17 Quest ion source :
Bank (Limerick)
Quest ion His tory:
Cognitive level :
Memo r yl F u n d a m en t a I know led g e :
Corn p re hensive/Ana lysis:
X 1 OCFR 55.41 X
QUESTION 40 Unit 1 plant conditions are as follows:
0 Reactor power is 85%
0 Recirc Loop "A" flow is 39,200 gpm on FR-043-1 R614 on Panel 1 OC602 0
Recirc Loop "B" flow is 38,600 gpm on FR-043-1 R614 on Panel 10C602 Subsequently, Reactor Recirc pump 'A' trips. The following Recirc Loop flows are observed after conditions stabilize:
0 Recirc Loop "A" flow is 0 gpm on FR-043-1 R614 on Panel 1 OC602 Recirc Loop "B" flow is 39,100 gpm FR-043-1 R614 on Panel 1 OC602 I&C has performed applicable ST-2 procedures for single loop operations Assuming 10Oo/0Total Recirc Drive Flow equals 88,000 gpm, WHICH ONE of the following identifies the APRM flow-biased scram setpoints being enforced before AND after the Recirc pump trip?
(Assume setpoint values have been rounded to nearest whole number).
BEFORE AFTER A.
- 6.
C.
D.
117%
87%
1 2 1 O/O 8 7 O/o 11 7%
9 2 O/O 121%
92%
K&A #
21 5005 APRMILPRM Importance Rating 3.7 QUESTION 40 K&A Statement:
K4.07 - Knowledge of Average Power Range Monitor/Local Power Range Monitor system design feature(s) and/or interlocks which provide for: Flow-biased trip setpoints.
Correct - Before: APRM Flow-biased setpoint is clamped at 116.6% (-1 17%). After:
For single-loop operation, the setpoint uses the following algorithm Justification:
A.
0.66 [(39,100/88,000)100 - 7.61 + 62.8 = -87%
B.
Incorrect but plausible since using the correct algorithms for two-loop and single-loop operation will yield these results.
Incorrect but plausible if the applicant uses the incorrect AW calculate the setpoint for single loop operation.
Incorrect but plausible if the applicant uses the incorrect AW calculate the setpoint for single loop operation.
C.
D.
0.0% -VS-7.6%) to 0.0% -VS-7.6%) to
References:
LLOT0275, Rev. 004 Student Ref. required No Learning Objective:
Obj 14 Question source:
Bank (Limerick)
Modified lesson plan question Question History:
Cognitive level:
Memory/ F u n d a m en t a I know 1 edge :
Com pre hens ive/Ana lysis:
X 1 OCFR 55.41 X
QUESTION 41 WHICH ONE of the following describes the interlock associated with the RClC pump suction valves?
A.
The suction source will automatically swap from the CST to the Suppression Pool upon a low level in the CST.
B.
The suction source will automatically swap from the CST to the Suppression Pool upon a high level in the Suppression Pool.
C.
The suction source will automatically swap from the Suppression Pool to the CST upon a low level in the Suppression Pool.
D.
The suction source will automatically swap from the Suppression Pool to the CST upon a high level in the Suppression Pool.
K&A #
217000 RClC Importance Rating 3.6 QUESTION 41 K&A Statement:
K1.03 - Knowledge of the physical connections and/or cause-effect relationship between REACTOR CORE ISOLATION COOLING and Suppression pool Justification :
A.
Correct - The Suppression Pool suction valves automatically open on a CST low level after a 20 second time delay. The CST suction valve receives a close signal when the suppression pool suction valves are open.
- 5.
Incorrect - The Suppression Pool suction valves automatically open on a CST low level. Plausible if the applicant recalls that the suppression pool suction valves are manual open and close but does not recall the automatic function.
C.
Incorrect - There is not an automatic swap on suppression pool low level. Plausible if applicant thinks that on low suppression pool level the correct action would be to swap to the CST. If the pool is empty, then no suction source is aligned. The automatic swap from the CST to the Suppression Pool can be overridden by manual action.
D.
Incorrect - There is not an automatic swap on suppression pool low level. Plausible if applicant thinks that on high suppression pool level the correct action would be to swap to the CST.
References:
LLOT0380 pg. 1 1, 13 Student Ref: required N
Learning Objective:
LLOT0380 Obj. 7a, 7b Question source:
Modified Bank (Perry)
Question History:
Cognitive level:
Memo ry/Fu nd a m e n t a I know I ed g e :
X Co m pre he nsive/Ana I ys is :
QUESTION 42 Unit 2 plant conditions are as follows:
0 Reactor is at 100% power 0
RClC monthly pump surveillance is in progress A loss of 125 VDC Div 3 power occurs.
Subsequently, the following alarms are received in the MCR:
216 A-I, RClC OUT OF SERVICE 216 C-4, RClC TURB EXH LO VACUUM DIAPHRAGM RUPTURED 219 A-2, A REAC ENCL HVAC PNL 2AC208 TROUBLE 006, L-I -L REAC II EL 177 PB RClC PUMP FIRE WHICH ONE of the following predicts the impact on RClC AND what operator action is required?
A.
RClC steam line is isolated.
Enter Fire Safe Shutdown Guide for the RClC room.
B.
RClC steam line is NOT isolated.
Depress the RClC Manual isolation Pushbutton.
C.
RClC steam line is NOT isolated.
Perform a Rapid Plant Shutdown.
D.
RClC steam line is isolated.
Close the RClC Steam Line Inboard Isolation (F007) valve.
K&A #
217000 RClC Importance Rating 3.3 QUESTION 42 K&A Statement:
A2.14 - Ability to (a) predict the impacts of Rupture disc failure:
Exhaust Diaphram on the REACTOR CORE ISOLATION COOLING SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations.
Justification:
A.
Incorrect Entry into the Fire Safe Shutdown Guides is only after a fire is confirmed by the Fire Brigade. Plausible if applicant does NOT associate the rupture disk with the fire alarm.
B.
Incorrect but plausible since a turbine trip will occur from Div 1 However there is no direction to reset the trip at this point.
C.
Incorrect but plausible. Entry into T-103 required due to initiation of steam leak detection (Div I),
however there is NO primary system discharging into secondary containment since the steam leak has been isolated by the closure of F008 and F076 (as well as Turbine Trip valve).
D.
Correct - On high equipment room temperature, Div 1 and Div 3 isolation signals are generated. The RCIC OUT OF SERVICE alarm should indicate an isolation has initiated. Div 1 closes F008 and F076; Div 3 closes F007. Since Div 3 power is lost, F007 failed to close automatically. Appropriate operator action, per Roles and Responsibilities is to complete the RClC system isolation by manually closing F007 from the control board. F007 is an AC powered motor operated valve.
References:
LLOT0380 pg. 11, 12, 13, 15, 16, 17 Student Ref: required N
OP-AA-101-111, Roles And Responsibilities Of On-Shift Personnel, Section 4.6.2.5 Learning Objective:
NIA Question source:
Bank (Limerick)
Question History:
None Cognitive level:
MemorylFundamental knowledge:
Com pre hensive/Ana lysis:
X 1 OCFR 55.41(7) X
QUESTION 43 Unit 2 was operating at 100% power:
0 Div 3 125 VDC power is lost.
A reactor coolant leak occurs.
T=O RPV water level has dropped to -129 and continues lowering:
0 0
0 0
0 0
0 Drywell pressure is 16.7 psig A Core Spray pump discharge pressure is 140 psig B Core Spray pump discharge pressure is 285 psig D Core Spray pump discharge pressure is 300 psig A RHR Pump discharge pressure is 11 5 psig No other ECCS pumps are running ADS NORM-INHIBIT switches are in the NORM position Which one of the following describes the response of ADS logic?
A.
The ADS valves will open in 105 seconds.
B.
The ADS valves will open in 420 seconds.
C.
The ADS valves will open in 525 seconds.
D.
The ADS valves will NOT open.
K&A #
21 8000 Automatic Depressurization System Importance Rating 3.9 QUESTION 43 K&A Statement:
K6.01 - Knowledge of the effect that a loss or malfunction of RHR/LPCI system pressure will have on the Automatic Depressurization System.
Justification:
A.
Incorrect. Level setpoint is satisfied but pump logic is not. Plausible if applicant thinks pump logic is satisfied.
B.
Incorrect. Bypass timer in combination with pump pressure switches cannot satisfy ADS requirements. Plausible if applicant thinks logic is satisfied by the drywell bypass timer and pump combination.
C.
Incorrect. The pump logic is not made up. Plausible if thinks logic is satisfied by 105 and 420 second timers and the pump combination.
D.
Correct - Div 3 initiation logic is inoperable with the loss of Div 3 125 VDC.
Therefore, only the inputs to the Div 1 initiation logic will result in automatic initiation of ADS. While the initiation logic inputs for High Drywell Pressure (+I.68 psig) Low Reactor Water Level (-129) and 105 second timer will be met; the inputs for low pressure pump availability will not. Div 1 initiation logic looks at the following combination of low pressure ECCS pumps:
RHR Pumps A or C running Core Spray Pump A running A Loop Core Spray, which consists of A and C Core Spray pumps, only indicates 140 psig. A Loop RHR/LPCI, which consists of A and C RHR pumps only indicates 15 psig. With these low pressures on Div 1 logic and the loss of Div 3 logic, the ADS valves will not receive an open signal.
Channel A Channel E Settinq 125 psig 145 psig RHR Pumps A or C running Core Spray Pump C running
References:
LLOT0330, Rev. 009 Student Ref. required No Learning Objective:
Obj 8 Question source:
New Question History:
Cognitive level:
Memory/Fundamental knowledge:
Comprehensive/Analysis:
X 1 OCFR 55.41 X
QUESTION 44 Unit 1 was in OPCON 4 with IB Loop RHR in Shutdown Cooling when an RPV level transient results in the following conditions:
e e
e Indicated reactor vessel level is + I 5 inches and slowly rising MCR Annunciator REACTOR 107 H-I, REACTOR WATER LEVEL BELOW LEVEL 3 TRIP is lit.
MCR Annunciator REACTOR 107 H-2, REACTOR HVLO LEVEL is lit.
HV-51-1 FOO9, RHR Shutdown Cooling Suction Inboard is OPEN HV-51-1 F008, RHR Shutdown Cooling Suction Outboard is CLOSED HV-51-1 FOI 56, RHR Shutdown Cooling Return Outboard is CLOSED HV-51-151 B, RHR Shutdown Cooling Test Check Equalizing Line Inboard is OPEN HV-51-1 F0506, RHR Shutdown Cooling Testable Check is CLOSED WHICH ONE of the following describes the status of the associated NSSSS logic relays?
Channel A Channel B Channel C Channel D A.
E ne rg ized Energized Deenergized Deenergized
- 8.
Energized Deenergized Energized Deenergized C.
Deenergized Deenergized En erg ized Energized D.
Deenergized Energized Deenergized Energized
K&A #
223002 PClS / Nuclear Steam Supply Shutoff System Importance Rating 2.6 QUESTION 44 K&A Statement:
A I.04 - Ability to predict and/or monitor changes in parameters associated with operating the PCIS/ Nuclear Steam Supply Shutoff System including Individual system relay status.
Justification:
A.
Correct - A Group IIA isolation has occurred that only affected the outboard isolation valves. This means that the B trip relay channel tripped which was due to trip conditions on Channels C and D. The NSSSS logic uses de-energized state to achieve a trip condition.
B.
Incorrect but plausible if the applicant believes the B and D channels are required to trip to affect the outboard isolation valves.
C.
Incorrect but plausible if the applicant believes the A and B channels are required to trip to affect the outboard isolation valves.
D.
Incorrect but plausible if the applicant believes the A and C channels are required to trip to affect the outboard isolation valves.
References:
LLOTOI 80, Rev. 01 5 Student Ref. required No Learning Objective:
Obj 2 Question source:
New Question History:
Cognitive level:
Memory/ F u n d a m e n ta I know I ed g e :
Com pre hens ive/Ana lysis:
X 1 OCFR 55.41 X
QUESTION 45 Div 1 125 VDC has been lost.
WHICH ONE of the following describes the impact on MANUAL operation of the non-ADS SRVS from the Main Control Room (MCR) and the Remote Shutdown Panel (RSP)?
A.
- 6.
C.
D.
MCR Available RSP Available Unavailable Available Unavailable Available Unavailable Unavailable
K&A #
239002 Safety Relief Valves Importance Rating 2.8 K2.01 - Knowledge of electrical power supplies to SRV solenoids.
QUESTION 45 K&A Statement:
Justification :
A.
Incorrect but plausible if the applicant believed that the controls switches at both locations were powered from Div 3 125 VDC.
B.
Incorrect but plausible if the applicant believed Div 1 125 VDC powered SRV control switches in the MCR and Div 3 125 VDC powered SRV control switches in the RSP.
C.
Correct - The manual switches for the non-ADS SRVs in the MCR and RSP are both supplied from Div 1 125 DC.
D.
Incorrect but plausible if the applicant believed Div 3 125 VDC powered SRV control switches in the MCR and Div 1 125 VDC powered SRV control switches in the RSP.
References:
LLOTOI 20, Rev. 01 4 Student Ref. required No Learning Objective:
Obj 3 Question source:
New Question History:
Cognitive level:
Me m o r y/ F u n d am e n t a I know led g e :
X Com pre hensive/Ana lysis :
1 OCFR 55.41 X
QUESTION 46 Unit 1 plant conditions are as follows:
e 100% power e
1 D Narrow Range Level Transmitter has failed upscale Ten (1 0) minutes later, 1 C Feedwater Narrow Range Level Transmitter fails upscale.
WHICH ONE of the following identifies the status of the RFP Turbines and the Main Turbine?
RFP Turbines Main Turbine A.
operating opera ti ng B.
tripped tripped C.
tripped ope rating D.
operating tripped
K&A #
259002 Rx Water Level Control Importance Rating 3.1 QUESTION 46 K&A Statement:
K5.03 - Knowledge of operational implications of Water level measurement as it applies to REACTOR WATER LEVEL CONTROL SYSTEM Justification:
A.
Correct - There are four narrow range level instruments (A, 6, C &D). All four signals are used to determine reactor level for control and for +54 trips. The system takes the four reactor water level inputs and checks for signal validity and then averages the valid ones. (I.E. The output is the average of the valid signals.) The system only provides a good output when there are 2 or more valid input signals. If there is only one valid input signal, a LEVEL SIGNAL FAILURE condition exists. For each level transmitter, a failure is defined as: a deviation of greater than 4 inches from the SMS output for 3 second OR a hardware failure, eg. Outside 4-20mA, on any sensing element used in the calculation. A failure results in an automatic bumpless disconnection of the signal from the soft majority selector. The detected error is annunciated as a trouble alarm in the MCR. More detailed alarm information is available at the operator work station. Bumpless reconnection of the repaired signal is performed automatically. If 3 out of 4 reactor level signals are in error, or if two errors occur simultaneously, a FWLC failure will occur.
B.
Incorrect but plausible if the applicant does NOT remember the required level instruments to initiate a high level trip.
C.
Incorrect but plausible if the applicant does NOT remember the required instruments to initiate a high level trip.
D.
Incorrect but plausible if the applicant does NOT remember the required instruments to initiate a high level trip.e
References:
LLOT0550 pages 14, 15 Student Ref: required No Learning Objective:
N/A Question source:
Bank (Limerick)
Question History:
Cognitive level:
MemorylFundamental knowledge:
x Corn pre hensive/Ana lysis:
IOCFR 55.41(7) X
QUESTION 47 Given the following conditions:
A valid initiation signal for the Standby Gas Treatment System is received SGTS Fan A is in AUTO and running SGTS Fan B is in STANDBY WHICH ONE of the following will initiate the B fan of the Standby Gas Treatment System?
A.
Refueling Floor Exhaust radiation level of 2 mR/hr B.
Reactor Enclosure Exhaust radiation level of 1 OmR/hr C.
Low flow condition on Standby Gas Treatment System for 20 seconds D.
Low flow condition on Reactor Enclosure Recirc System for 20 seconds
K&A #
261 000 SBGTS Importance Rating 3.2 QUESTION 47 K&A Statement:
Justification:
A.
A3.01-Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including System flow Incorrect - Since the B fan is in STBY, it will only start following a low flow condition on the system. Plausible if the applicant believes that in STBY the fan will start on a second initiation signal. Refuel Floor radiation of 2mR/hr is an initiation signal.
B.
Incorrect - Since the B fan is in STBY, it will only start following a low flow condition on the system. Plausible if the applicant believes that in STBY the fan will start on a second in it ia t ion signa I.
C.
Correct - Since the 73 fan is in STBY, it will only start following a low flow condition on the system.
D.
Incorrect - Since the B fan is in STBY, it will only start following a low flow condition on the SBGT system. Plausible if the applicant believes that the fan will start on a low flow condition of RERS.
References:
LLOT0200 pg. 21, 28, 29 Student Ref: required No Learning Objective:
LLOT0200 3, 1 Ob Question source:
Modified Bank (Dresden)
Question History:
Cognitive level:
Memo ry/F u n d a m e n t a I k n o wI ed g e :
X Comprehensive/Analysis:
1 OCFR 55.41(7) X
QUESTION 48 Unit 1 plant conditions are as follows:
0 100°/~ power Normal electrical lineup The 101 Safeguard Transformer Feeder Breaker trips.
WHICH ONE of the following identifies the response of the D114 Load Center feeder breaker to the above condition?
A.
Trips and recloses 3 seconds after 201-Dl 1 breaker closes B.
Trips and recloses 3 seconds after D11 EDG output breaker closes C.
Remains closed and supplies power to D l 14 Load Center when 201 -D11 breaker closes D.
Remains closed and supplies power to D114 Load Center when D11 EDG output breaker closes
K&A #
262001 AC Electrical Distribution Importance Rating 3.8 QUESTION 48 K&A Statement:
K3.01 - Knowledge of the effect that a loss or malfunction of the AC ELECTRICAL DISTRIBUTION system will have on Emergency Generators.
Justification :
A.
Incorrect but plausible since on a LOCA coridition Load Center Transformer Breakers trip and then re-close after a three second time delay B.
Incorrect but plausible since the on a LOCA condition Load Center Transformer Breakers trip and then re-close after a three second time delay C.
Correct - D114 Load Center Transformer Breaker (4.16kV) is closed. Power is supplied by D*l bus via a 4.16 KV/480 V transformer. Load Center Transformer Breakers remain closed on undervoltage. All Load Center MCC Feeder Breakers (480 V) remain closed except D*14-G-D and D*24-G-D NON-Safeguard MCC feeder breakers.
D.
Incorrect but plausible EDG does not supply D114 Load Center.
References:
LLOT0650 pages 13, LGSOPS 092A, Student Ref: required No Learning Objective:
N/A Question source:
Bank (Limerick)
Quest ion H isto ry :
Cognitive level:
Memory/Fundamental knowledge:
Corn p re hens ive/Ana lysis:
X 1 OCFR 55.41(7) X
QUESTION 49 During Unit 1 startup at 20% power the following alarms on MCR PNL122 annunciate:
0 1B RPS & UPS STATIC INVERTER TROUBLE IDB-1 250 VDC MCC UNDERVOLTAGE 1 DB-2 250 VDC MCC UNDERVOLTAGE All RPS white status lights are lit.
WHICH ONE of the following identifies the source currently providing power to 1 B RPS?
A.
480V Non-Safeguard bypassing the RPWUPS cabinet B.
480V Non-Safeguard through the RPS/UPS cabinet.
C.
TSC UPS inverter bypassing the RPS/UPS cabinet.
D.
TSC UPS inverter through the RPS/UPS cabinet.
K&A#
262002 Importance Rating 2.7 QUESTION 49 K&A Statement:
K6.03-Knowledge of the effect that a loss or malfunction of Static Inverter will have on the UNINTERRUPTABLE POWER SUPPLY System Justification:
A.
Incorrect. Plausible if applicant thinks the TSC inverter is not available based on receipt of the static switch trouble alarrn. However, the bus is still powered and bypass operation is a manual operation through a maintenance bypass switch.
- 8.
Incorrect. Plausible if applicant believes non-safeguards power is aligned as bus backup through the static switch. This alignment is only used when the TSC inverter is out of service.
C.
Incorrect. Plausible if applicant thinks the TSC inverter normally feeds the bus as a backup source through the bypass switch rather than the static switch.
D.
Correct. TSC inverter is normally aligned as the backup source to the RPS bus and is routed through the RPS/UPS cabinet to the static switch and then out to the RPS bus.
References:
ARC 122 D-12, F-4, & A-5; LLOT 0650 Student Ref: required NONE pg 9; S94.2.B pgs 2 and 3, S94.9.A Learning Objective:
N/A Question source:
New Question History:
None Cognitive level :
Memo ry/F u nd a m e n t a I k n owl E: d g e :
ComprehensivelAnal ysis:
X 1 OCFR 55.41(7) X
QUESTION 50 A metal scaffold inadvertently falls across the positive and negative output terminals of Battery Bank 1 B1 and short circuits the bank.
Which one of the following describes the immediate impact this event will have on the Div 2 DC Distribution system?
A.
A complete loss of 125 VDC loads and undervoltage on 250 VDC loads.
B.
A complete loss of 125 VDC loads and normal voltage on 250 VDC loads.
C.
A partial loss of 125 VDC loads and undervoltage on 250 VDC loads.
D.
A partial loss of 125 VDC loads and normal voltage on 250 VDC loads.
K&A #
263000 DC Electrical Distribution Importance Rating 3.2 QUESTION 50 K&A Statement:
K1.02 - Knowledge of the physical connections and/or cause-effect relationships between DC ELECTRICAL DISTRIBUTION SYSTEM and Battery charger and battery.
Justification:
A.
Incorrect but plausible if the applicant did not understand how 125 VDC is generated on Div 2 using two battery banks. The second half of the answer is correct for the effect on 250 VDC.
B.
Incorrect but plausible if the applicant did not understand how the loss of a battery bank would effect the generation of 125 and 250 VDC on Div 2.
C.
Correct - The Div 2 DC system uses two battery banks / battery chargers to generate the 125 VDC and 250 VDC power supply. Output from one battery chargerlbattery bank feeds one terminal within the Div 2 DC Fuse Box. Output from the other battery charger/ battery bank feeds the other terminal of the fuse box.
Each battery bank/ charger generates 125 VDC. 250 VDC power is generated by tapping off each of these positive and negative 125 VDC strips, while 125 VDC power is generated by tapping off one or the other of the strips within the fuse box to a neutral leg. Loss of a battery bank and its associated charger will result in a loss of power to one of the strips. Those 125 volt DC loads tapping off the terminal fed by the faulted battery will lose power and all the 250 VDC loads would be supplied at only 125 volts.
D.
Incorrect but plausible since the first half of the answer is correct concerning partial loss of 125 VDC loads and if the applicant did not understand how the 250 VDC signal is generated.
References:
LLOT0690, Rev. 01 2 Student Ref. required No Learning Objective:
Obj. 2 Quest ion source :
New Question History:
Cog nit ive level :
Memory/Fundamental knowledge:
X Com p re hen s ive/Ana lysis:
IOCFR 55.41 X
QUESTION 51 Unit 2 is at 100°/~ power when the following concerns developed:
While performing ST-6-092-364-0, D24 Diesel Generator Operability Verification, an internal Jacket Water Cooling leak is observed on EDG D24 and the diesel is declared inoperable.
ST-6-092-366-0, Inoperable Unit 2 Safeguard Power Supply Actions for Both Units is being performed.
Visual inspection of remaining EDGs reveals leaks at the same locations on EDG D21 and D22.
Results of inspection of Jacket Water Cooling on EDG D11, D12, D13, D14 and D23 were satisfactory.
Based on the results of the inspection, operations management has declared EDG 021 and D22 inoperable.
Which one of the following describes actions that are required to be taken within one hour?
A.
Determine power available to the 101 and 201.Safeguard Transformers ONLY.
B.
Verify the operability of at least two LPCl subsystems ONLY.
C.
Determine power available to the 101 and 201 Safeguard Transformers AND verify that D23 can start and accelerate to synchronous speed.
D.
Verify that D23 can start and accelerate to synchronous speed AND verify the operability of at least two LPCl subsystems.
K&A #
264000 Emergency Diesel Generators Importance Rating 3.9 QUESTION 51 K&A Statement:
G.2.2.39 - Knowledge of less than or equal to one hour Technical Specification action statements for systems as they relate to EDGs.
Justification :
A.
Incorrect but plausible since this surveillance (SR 4.8.1-.1.I
.a,) would be required anytime a diesel is lost. However, it is not the only one-hour action that would be required.
B.
Incorrect but plausible since the operability of LPCl is required anytime two or more diesels are declared inoperable (action e), but it has a two hour completion time rather than a one hour completion time..
C.
Correct - Condition C of LCO 314.8.1 AC Sources Operating states that SR 4.8.1.I
. l a should be performed for the two required offsite circuits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and that SR 4.8.1.I
.2.a for verification of operability for the last diesel should be performed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
D.
Incorrect but plausible since the operability of LPCl is required anytime two or more diesels are declared inoperable (action e), but it has a two hour completion time rather than a one hour completion time. Verification of operability for the last diesel within one hour is correct.
References:
LLOT0670, Rev. 01 2 Learning Objective:
Obj. 13 Student Ref. required Yes TS 3.8.1.1 (Pages 3/4 8-1 thru 3/4 8-7a)
Question source:
New Question History:
Cognitive level:
Memory/Fundamental knowledge:
X ComprehensivelAnalysis:
1 OCFR 55.41 X
QUESTION 52 Unit 2 plant conditions are as follows:
0 Reactor Power is 100%
0 Bus D21 is de-energized while Electrical Maintenance replaces an u nd e rvo I tag e re lay A bus differential overcurrent condition de-energizes D22 Bus.
WHICH ONE of the following lists procedures that will be required to be performed within 15 minutes?
A.
LOSS of TECW (ON-I 17) ONLY LOSS of RECW (ON-I 13) ONLY B.
C.
D.
Loss of TECW (ON-I 17) AND Loss of Instrument Air (ON-I 19)
Loss of RECW (ON-I 13) AND Recirculation Pump Trip (OT-I 12)
K&A #
300000 Instrument Air lmportance Rating 2.8 QUESTION 52 K&A Statement:
Kl.04 - Knowledge of the physical connections and/or cause-effect relationships between Instrument Air System and Cooling water to the compressors.
Justification:
A.
Incorrect. Loss of TECW (ON-I 17) ONLY is not correct. Loss of IA procedure (ON-119) is also required on the loss of compressors.
B.
Incorrect. Loss of RECW (ON-I 13) ONLY is not correct. RECW will not be lost under the given conditions. Plausible if applicant thinks D21 and D22 loss will impact the running RECW pump.
C.
Correct. Loss of TECW due to de-energized motor will result in trip of running Instrument Air Compressor.
D.
Incorrect. May enter OT-I 12 on sustained loss of Instrument Air but will not need to enter ON-I 13 as RECW is not lost.
References:
LLOT0730, Rev. 014 Student Ref. required No Learning Objective:
Obj. 15b (LLOT0730)
Question source:
New Question History:
Cognitive level:
1 OCFR Memory/Fundamental knowledge:
Comprehensive/Analysis:
X 55.41 X
QUESTION 53 Plant conditions are as follows:
0 0
A Loss of offsite power has occurred E-10120, Loss Of Offsite Power is being performed OB AND OD ESW Pumps are TRIPPED and cannot be restarted The Shift Manager has directed available ESW to be used to cool RECW AND TECW Heat Exchangers.
WHICH ONE of the following describes the Heat Exchangers that can be cooled per E-l0/20, given the conditions above?
Unit 1 Unit 2 A.
TECW RECW B.
TECW TECW C.
RECW TECW D.
RECW RECW
K&A #
400000 Component Cooling Water Importance Rating 4.6 QUESTION 53 K&A Statement:
Justification:
A.
G2.1.20 - Ability to interpret and execute procedure steps as it relates to COMPONENT COOLING WATER Correct. A Loop of ESW can be backed up to Unit 1 TECW and Unit 2 RECW. B Loop is unavailable under given conditions and B Loop can be aligned to cool Unit 1 RECW and Unit 2 TECW. Important because of critical loops that must be cooled for safe shutdown of Unit.
B.
Incorrect. Plausible if applicant believes this is the correct alignment. Applicant must understand E-I 0/20 capabilities.
C.
Incorrect. Plausible if applicant believes this is the correct alignment. Applicant must understand E-I 0/20 capabilities.
D.
Incorrect. Plausible if applicant believes this is the correct alignment. Applicant must understand E-I 0120 capabilities.
References:
LGOPS0013, LLOT 1570, LLOT 01 10 Student Ref: required none Learning Objective:
LLOT 1570 Obj. 1 le, LLOT 01 10 Obj. 3 Question source:
New Question History:
none Cognitive level:
Memory/Fundamental knowledge:
X Com pre hensive/Ana lysis :
1 OCFR 55.41(7) X
QUESTION 54 Unit 1 is operating normally at rated power when the instrument air line breaks at the actuator of the in-service CRD Flow Control Valve, FVC-46-1 F002A.
WHICH ONE of the following describes the impact of the air line break on the CRD system?
A.
The flow control valve opens.
CRD pump will trip on high flow.
B.
The flow control valve closes.
CRD mechanism temperatures will increase.
C.
The flow control valve opens.
CRD mechanism will move more than one notch in response to a R MCS wit hd rawa I command.
D.
The flow control valve closes.
Drive water d/p drops to zero.
K&A #
201003 Control Rod Drive and Mechanism Importance Rating 3.2 QUESTION 54 K&A Statement:
K1.01 - Knowledge of the physical connections and/or cause-effect relationships between Control Rod Drive and Mechanism System and Control Rod Drive Hydraulics,System.
Justification:
A.
Incorrect but plausible if the applicant doesnt know which direction the flow control valve fails. Also, failure of the valve will result in a low flow / high pressure condition on the discharge of the CRD pumps.
B.
Correct - Flow control valve will close (-20% open with mechanical gag) on loss of air. With the Drive Water Pressure Control Valve, F003 throttled to maintain Drive Water pressure higher than reactor pressure, the already limited flow (0.2 - 0.3 gpm) providing cooling flow to each CRD mechanism will become more restricted and CRD mechanism temperatures will increase.
C.
lncorrect but piausible if the applicant doesnt know which direction the flow control valve fails.
D.
Incorrect. Drive header pressure is maintained because the flow control valve cannot fully close (-20% open). If applicant thinks valve fully closes, then the applicant will logically conclude drive water dlp goes to zero.
References:
LGSOPS0046 Rev001 Student Ref. required No Learning Objective:
IL 9 Question source:
Modified Bank (Susquehanna)
Changed stem from loss of power to controller to loss of air to flow control valve.
Question History:
Cognitive leve I :
Me m o r y/ F u n d a m en t a I know I edge :
Com pre he nsive/Ana lysis :
X 1 OCFR 55.41 X
QUESTION 55 Unit 1 is initially at steady state 100% power The hold down bolt on #I4 Jet Pump failed allowing the Rams Head for the Jet pump to become disengaged from the Jet pump inlet.
WHICH ONE of the following identifies the plant response to the failure?
A.
Indicated total core flow increases Reactor power increases B.
Loop A drive flow increases Reactor power increases C.
Indicated dp on Jet Pump #I 3 decreases Reactor power decreases D.
Loop 6 drive flow increases Reactor power decreases
K&A #
202001 Recirculation Importance Rating 3.5 QUESTION 55 K&A Statement:
Justification:
A.
K6.01 - Knowledge of the effect that a loss or malfunction of Jet Pumps will have on the Recirculation System Incorrect. Core power decreases on a jet pump failure. Indicated core flow will rise due to the addition of reverse flow through the failed jet pump. Plausible if applicant uses the fact that indicated flow increases and an increase in total core flow would cause power to increase.
B.
Incorrect. Core power decreases on a jet pump failure. The recirc drive flow in the loop containing the failed jet pump will increase due to the decrease of flow resistance; however, total core flow decreases causing a decrease in core power.
C.
Correct - Indicated dp on the jet pump sharing the riser with the failed jet pump decreases. The dp will drop due to the preferential flow of drive water out the failed jet pump. Actual total core flow decreases, causing a decrease in reactor power.
D.
Incorrect. Loop B drive flow will not change. Actual power increases. Plausible if applicant does not remember that loop A discharges through jet pump 14 or believes that drive flow will increase in the unaffected loop.
References:
ON-1 00; LGSOPS0043A, LLOTl550 Student Ref: required No Learning Objective:
LLOTI 550 Obj. 1 Question source:
New Question History:
Cognitive level:
Memory/ F u nd a m e n ta I know I ed g e:
X Comprehensive/Analysis:
IOCFR 55.41(2) X
QUESTION 56 The 1 A Recirc Motor Generator (MG) Set is being started IAW S43.1.A, Startup of Recirculation System The RO places the IA MG Set Drive Motor Control (MOTOR) to START at 1 OC602.
WHICH ONE of the following lists the values for Speed Demand and MG Set Generator Voltage 30 seconds later?
A.
B.
C.
D.
Speed Demand MG Set Generator Voltage (nominal) 40%
161 0 volts 40%
805 volts 2 0 O/O 1610 volts 20%
805 volts
K&A #
202002 Recirculation Flow Control Importance Rating 3.3 QUESTION 56 K&A Statement:
Justification:
A.
A4.01 - Ability to manually operate and/or monitor the control room: MG sets Incorrect. After 30 seconds the speed controller has reduced speed demand to approximately 20% and voltage would be approximately 805. Voltage increase does occur so answer is plausible if the applicant does not recall the settle speed of the MG set start sequence.
B.
Incorrect. After 30 seconds the speed controller has reduced speed demand to approximately 20%. Plausible since voltage is correct for the 20% settle speed.
C.
Incorrect. After 30 seconds the speed controller has reduced speed demand to approximately 20% and voltage would be approximately 805. Voltage increase does occur so answer is plausible if the applicant does not recall the voltage (805) associated with the settle speed.
D.
Correct - The speed demand is initially from the startup signal generator which positions the scoop tube between 35%-45?/0 speed to provide sufficient torque for breakaway MG set startup. After the Generator Field breaker closes, speed control transfers to the speed controller, which reduces speed demand to approximately 20%. Generator voltage increases when the field breaker closes and is controlled by the MG Set voltage regulator to maintain a 70V/Hz relationship. Six pole sync generator speed varies from 230 rpm to 1150 rpm depending on scoop tube controlled drive coupling. Voltage = (70V/Hz)*(N*P)/120, where N=gen spd in rpm, P=number of poles. Voltage approx equal to 805V at 20% min speed of 230 rpm.
References:
LGSOPS0043B pg. 6, 7, 10-12; Student Ref: required No S43.1.A, rev.59 pg.13, 14, S43.9A LGSOPS0043A pg. 13, 14 Learning Objective:
N/A Question source:
Bank (Limerick)
Question History:
Cognitive level:
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
1 OCFR 55.41(7) X
QUESTION 57 Unit 2 plant conditions are as follows:
0 0
Reactor Power is at 75%
Traversing In-core Probe (TIP) scans are in progress The B TIP is stuck at the indexer A main turbine trip occurred, resulting in the following:
0 RClC and HPCl automatically start on a valid signal WHICH ONE of the following describes the position of the TIP Shear and Ball Valves for the B TIP two (2) minutes later with no operator action?
Shear Valve A.
Open Ball Valve Open B.
Open Closed C.
Closed Open D.
Closed Closed
K&A#
215001 Traversing In-Core Probe Importance Rating 3.1 QUESTION 57 K&A Statement:
K6.04 - Knowledge of the effect that a loss or malfunction of Primary Containment Isolation System will have on the TRAVERSING IN-CORE PROBE SYSTEM Justification:
A.
B.
C.
D.
Correct - A containment isolation signal is present, so TIP should automatically shift into reverse, withdraw detectors and close associated ball valves. 8 detector is stuck at the indexer, B TIP is not withdrawn into its shield and the ball valve will not automatically close. The shear valve requires operator action to close. Therefore for the given conditions, both valves remain open.
Note: The WA is matched since the TIP rnalfunction causes a PClS function to NOT be satisfied. That is, the ball valve cannot close with the TIP partially inserted.
The result is that the shear valve is the only action left. The author chose to NOT include manual action (to fire the shear valve) in the stem to avoid potentially conflicting answers.
Incorrect - The ball valve will not automatically close on the isolation signal if the TIP is not withdrawn into its shield during a containment isolation. Plausible if the applicant does not recall the interlock between the automatic closure of the ball valve and the TIP detector location or it the applicant thinks that the TIP detector being at the indexer will allow the ball valve to close.
Incorrect - The shear valve requires operator action to close. Plausible if the applicant thinks that under an isolation condition, the shear valve would receive an automatic signal.
Incorrect - The ball valve will not automatically close on an isolation signal if the TIP detector is not withdrawn into its shield. The shear valve requires operator action.
Plausible if the applicant thinks that the shear valve would receive an automatic signal to allow the ball valve to close on a containment isolation signal.
References:
LLOT0290 pg. 8, 9, 12, 17, 18 Student Ref: required No Learning Objective:
N/A Question source:
Modified Bank (Limerick)
Question History:
Cognitive level:
Memory/Fundamental knowledge:
Comprehensive/Analysis:
X IOCFR 55.41(9) X
QUESTION 58 Plant conditions are as follows:
0 Reactor power is at 31%
0 Control Rod 22-27 is selected Subsequently, the 'C' level detector on the LPRM string located at 24-33 fails downscale.
WHICH ONE of the following describes the effect this will have on the Rod Block Monitoring (RBM) channels "A" and "B"?
A.
B.
C.
D.
RBM "A" and "B" will be unaffected since level 'C' LPRM detectors are not part of the RBM circuit.
RBM "A" will register a change in the "Average Flux" value since the control rod has been selected.
RBM "B" will be unaffected since the level 'C' LPRM detectors are not part of the RBM 'B' circuit RBM "B" will register a change in the "Average Flux" value since the control rod has been selected.
RBM "A" will be unaffected since level 'C' LPRM detectors are not part of the RBM 'A' circuit RBM "A" and "B" will register a change in the "Average Flux" value since level 'C' LPRM detectors are used by both channels to generate an "Average Flux" value.
K8A #
21 5002 Rod Block Monitor Importance Rating 2.8 QUESTION 58 K8A Statement:
Justification:
A.
K6.05 - Knowledge of the effect the a loss or malfunction LPRM detectors will have on the Rod Block Monitoring System.
Incorrect but plausible since A level detectors are excluded from the averaging circuit.
B.
Incorrect but plausible since RBM A would be effected and RBM B would be unaffected if the failure was on a D level detector.
C.
Incorrect but plausible since RBM B would be effected and RBM A would be unaffected if the failure was on a B level detector.
D.
Correct - The C level detectors provide inputs to both RBM averaging circuits.
References:
LGSOPS00746 Rev001 Learning Objective:
E007 Student Ref. required Yes Core Map Question source:
New Quest ion His tory:
Cognitive level:
Memory/Fundamental knowledge:
X Com pre hensive/Analysis:
1 OCFR 55.41 X
QUESTION 59 A seismic event has resulted in the following Unit 2 plant conditions:
D21 and D22 Safeguard Buses are de-energized A DBA LOCA has occurred in the Reactor Recirculation System piping Both suppression pool-to-drywell vacuum relief valves on one downcomer are stuck open Suppression Pool level is 24 ft. and rising very slowly WHICH ONE of the following describes the response of the primary containment and the status of containment integrity?
A.
Drywell pressure will equalize with Suppression Pool pressure; containment integrity is threatened.
B.
Drywell pressure will equalize with Suppression Pool pressure; containment integrity is NOT threatened.
C.
Drywell pressure will rise to about 5 psig greater than Suppression Pool pressure; containment integrity is threatened.
D.
Drywell pressure will rise to about 5 psig greater than Suppression Pool pressure; containment integrity is NOT threatened.
K&A #
223001 Primary CTMT and Aux.
Importance Rating 2.8 QUESTION 59 K&A Statement:
K5.03 - Knowledge of the operational implications of Down comer operation as it applies to PRiMARY CONTAINMENT SYSTEM AND AUXILIARIES Justification:
A.
Correct - A broken downcomer would bypass suppression function and equalize drywell pressure with suppression pool pressure. A reduction in suppression capability could challenge primary containment integrity on a DBA.
B.
Incorrect - Plausible because pressure will equalize and incorrect because containment integrity IS threatened.
C.
Incorrect - Plausible because integrity IS threatened. Incorrect because pressure will equalize between drywell and SP.
D.
Incorrect - Plausible if applicant thinks suppression capability is maintained and incorrect because pressure will equalize.
References:
LLOT0130 Obj 2 Student Ref: required No Learning Objective:
N/A Question source:
Bank (Limerick)
Question History:
Cognitive level:
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
1 OCFR 55.41(9) X
QUESTION 60 I I NOTE T1 A LOCA has occurred on Unit 2. Operators are performing Step 4.6.13 of T-225, Startup and Shutdown of Suppression Pool and Drywell Spray Operation to spray the drywell with 2A RHRSW (step shown below).
Step 4.6.13 will require coordination between an Operator at OOC667 AND a second
~
Operator at 20C681
~
4.6.1 3 Simultaneously PERFORM the following to maintain RHR Service Water discharge I
pressure 75 to 120 psig as indicated an PI-12-001A-1, Pump A/C Disch (Px), at I
I OOC667 Main Control Room):
CAUTION Slowlv throttling open Outboard Drywell Spray valve will prevent rapid pressure drop.
0 Throttle Fully CLOSED HV-51-2F068A, 2A RHR Htx SW Outlet Vlv (2A) at 006667 (Main Control Room).
0 Throttle Fully OPEN HV-51-2F016A, 2A RHR Cntmt Spray Line Outboard PCIV (OUTBOARD) to initiate spray AND MAXIMIZE flowrate as indicated on FI-51-2R603A, FL.
While opening the HV-51-2F016A and closing HV-51-2F068A, RHRSW discharge pressure rises to 1 15 psig with RPV pressure steady at 180 psig.
WHICH ONE of the following identifies:
0 RHRSW Alignment 2A RHRSW Pump Status?
RHRSW Alignment A.
Aligned to Drywell ONLY
- 9.
Aligned to Drywell AND RPV C.
Aligned to Drywell ONLY D.
Aligned to Drywell AND RPV 2A RHRSW Pump Status Running Running Tripped Tripped
K&A #
226001 RHR/LPCI:
Containment Spray Importance Rating 4.3 QUESTION 60 K&A Statement:
G.2.1.23 - Ability to perform specific system and integrated plant procedures during all modes of plant operation, as it relates to RHWLPCI: CONTAINMENT SPRAY.
Justification:
A.
Incorrect but plausible if applicant does not realize that 74 psid permissive is met to align RHRSW to RPV.
B.
Correct - With reactor pressure at 180 psiy allowing the pressure to exceed 105 psig will result in LPCl injection valves getting an auto open signal due to RHR pressure to reactor pressure being less than 74 psid. RHRSW pump remains running because trip setpoint of > 120 psig is not met and, additionally, the RHRSW Hi Rad (disch press) Bypass is in BYPASS earlier in T-225.
C.
Incorrect but plausible if applicant believes hi pressure pump trip setpoint is reached and does not realize that 74 psid permissive is met to align RHRSW to RPV.
D.
Incorrect but plausible if applicant believes hi pressure pump trip setpoint is reached
References:
LLOTI 561, Rev.007 T-225, Rev. 20 Learning Objective:
IL2, IL3 Student Ref. required No Question source:
New Question History:
Cognitive level :
Memory/Fundamental knowledge:
Co m pre he n s ive/Ana I ys is :
X 1 OCFR 55.41 X
QUESTION 61 Unit 2 Refueling operations are in progress with the following conditions:
0 Core shuffle part 2 is in progress 0
"B" and "C" SRMs are INOPERABLE 0
All other SRMs are OPERABLE A Core Component Transfer Authorization Sheet (CCTAS) is planned with the following Steps:
Step #I Move fuel assembly from Spent Fuel Pool to Core Location 27-50 Step #2:
Move fuel assembly from Spent Fuel Pool to Core Location 13-30.
Step #3:
Move fuel assembly from Spent Fuel Pool to Core Location 47-1 4.
Step #4:
Move Fuel Assembly from Core Location 21-38 to Spent Fuel Pool.
WHICH ONE of the following describes the steps that may be performed?
Core Map provided to locate positions of planned fuel moves.
A.
Immediately stop all core alterations. Do not perform Step #I.
B.
Fuel moves can continue up through Step #2. However, core alterations must be terminated prior to Step #3.
C.
Fuel moves can continue up through Step #3. However, core alterations must be terminated prior to Step #4.
D.
Fuel moves can continue up through Step #4.
K&A ##
234000 Fuel Handling Equipment Importance Rating G.2.2.39 QUESTION 61 K&A Statement:
G.2.2.39 - Knowledge of less than or equal to one hour Technical Specification action statements for systems as they relate to Fuel Handling Equipment.
Justification:
A.
Incorrect SRMs A and D are operable. Fuel assembly 27-50 is located in the same quadrant as SRM A and is adjacent to SRM D.
B.
Correct -T.S. 3.9.2 requires two SRMs be operable; the SRM in the quadrant that the assembly is located and the adjacent SRM. Fuel assembly location 13-30 is in the same quadrant as SRM D and adjacent to SRM A.; therefore, this assembly can be lowered into the core. Fuel assembly location 47-14 is located in the same quadrant as SRM C; therefore this assembly cannot be lowered into the core and all core alterations must be immediately suspended prior to step #3.
C.
Incorrect - T.S. 3.9.2 requires two SRMs be operable for core alterations. All core alterations must be immediately suspended if...One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant is not operable. In this instance, fuel position 47-14 is in the same quadrant SRM C and would require suspension of core alterations prior to performing this fuel movement, since SRM C is inoperable.
D.
Incorrect but plausible if the applicant believed that as long as two SRMs were operable adequate coverage was provided. It would also seem reasonable, since removing fuel from the core would increase shutdown margin decreasing the likelihood of inadvertent criticality.
References:
LLOT0760, Rev. 01 4 Student Ref. required No Learning Objective:
Obj 12 (LLOT0760), Obj. 3 (LLOT0240)
LLOT0240, Rev. 009 Question source:
New Question History:
Cognitive level :
Memory/ F u n d a menta I know ledge :
Com pre hensive/Ana I ys is :
X IOCFR 55.41 X
QUESTION 62 Unit 2 is at 33% power when a 4 inch line upstream of the Turbine Stop Valve on the B MSL is inadvertently isolated.
WHICH ONE of the following describes the impact on the plant of isolating the steam load(s) off this 4 inch line? (Assume NO operator action.)
A.
RFP turbines will swap to low pressure steam supply.
B.
Turbine will trip on low condenser vacuum.
C.
Main Condenser vacuum will lower but stabilizes above 23 D.
Reactor will scram on low reactor water level.
K&A #
239001 Main Steam and Reheat System I m portarice Rating 2.9 QUESTION 62 K&A Statement:
K1.08 - Knowledge of the physical connections and/or cause-effect relationships between MAIN AND REHEAT STEAM SYSTEM and Condenser Air Removal.
Justification:
A.
Incorrect. On Unit 2, the reactor feed pump turbines are supplied by the C main steam line. Plausible since on Unit 1, the reactor feed pump turbines are supplied by the 8 main steam line.
B.
Correct. On Unit 2, the condenser air removal system is supplied from the Main Steam system off of B MSL which would affect the Steam Jet Air Ejectors (,SJAE) and Steam Seal Evaporator (SSE). Loss of steam flow to the SJAEs would result in non-condensable gases building up in the condenser. The Turbine will trip when condenser vacuum degrades to 21 Hg Vac.
C.
Incorrect. Loss of steam flow to the SJAEs would result in non-condensable gases building up in the condenser. Loss of steam flow to the SJAEs would result in non-condensable gases building up in the condenser.
D.
Incorrect but plausible since on Unit 1 steam to the RFPTs is supplied from B MSL.
Loss of high pressure steam to the RFPTs could result in a tow reactor level condition because of inadequate low pressure steam supply pressure to the feed pump turbines at low main turbine power level. However, Unit 2 high pressure steam supply to the RFPTs is from C MSL.
References:
LGSOPS0007, Rev. 000 Student Ref. required No LGSOPS0069, Rev. 000 LLOTl870, Rev.OO1 Learning Objective:
IL3 (LGSOPS0007), IL3 (LGSOPS0069), Obj 1 (LLOTI 870)
Question source:
New Question History:
Cognitive level:
Memory/Fundamental knowledge:
X Com p re hensive/Ana I ysis :
1 OCFR 55.41 X
QUESTION 63 Unit '2 plant conditions are as follows:
0 Reactor power is 100%
0 Both running Reactor Enclosure (RE) Exhaust Fans have tripped The Standby RE Exhaust Fan fails to start.
WHICH ONE of the following describe RE Supply Fans and Standby Gas Treatment Fan status thirty (30) minutes later with NO operator action?
A.
B.
C.
D.
RE Supply Fans Standby Gas Treatment Fans Tripped Running Running Off Tripped Off Running Running
K-A #
2 9000 Plant Ventilation Importance Rating 3.5 QUESTION 63 K&A Statement:
Justification:
A.
A4.01 - Ability to manually operate and/or monitor in the control room: Start and stop fans Incorrect. Plausible if applicant thinks SGTS fans would start within 30 minutes.
B.
Incorrect. Plausible if applicant believes RE supply fans remain running.
C.
Correct. RE supply fans will trip on a loss of supply fans. SGTS fans will start on a low d/p signal, but on a 50 minute time delay.
D.
Incorrect. Plausible if applicant believes RE supply fans remain running and that SGTS fans would start within 30 minutes.
Learning Objective:
LLOT-0200 Objective 4 Question source:
Bank (Limerick)
Question History:
Cognitive level:
Me m o r yl Fu n d a m e n t a I know I ed g e :
X Com p re hensive/Ana I ysi s :
Student Ref No Required:
1 OCFR 55.41(7) X
QUESTION 64 Unit 2 is at 100% power.
The following alarm is received at MCR 227 8-1 :
2 UNIT RECOMBINER OUTLET HI TEMP Recombiner Outlet Temperature is reading 905°F on TE-69-235 A, B, and C WHICH ONE of the following describes the expected plant response AND the required procedural action?
A.
B.
C.
D.
Plant Response First Stage Air Valve closes.
First Stage Steam Valve closes.
First Stage Air Valve closes.
First Stage Steam Valve closes.
Procedural Action Swap Air Ejectors per S07.6.A, Placing Alternate Steam Jet Air Ejector in Service.
Swap Air Ejectors per S07.6.A, Placing Alternate Steam Jet Air Ejector in Service.
Ensure Hydrogen Water Chemistry has tripped per ON-103, Control Of Sustained Combustion In The Offgas System.
Ensure Hydrogen Water Chemistry has tripped per ON-103, Control Of Sustained Combustion In The Offgas System.
K&A#
271000 Offgas Importance Rating 2.7 QUESTION 64 K&A Statement:
A2.12 - Ability to (a) predict the impacts of Recombiner high temperature on the OFFGAS SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations Incorrect. Plausible if applicant believes they have a condition which isolated SJAE and restoration is needed. Operator action to swap to standby SJAE would be appropriate for a failed SJAE. However, due to combustion in the off-gas, it is not appropriate.
Justification:
A.
B.
Incorrect but plausible if applicant believes that the steam valve closes instead of the air valve due to the SJAE isolation. Operator action to swap to standby SJAE would be appropriate for a failed SJAE. However, due to combustion in the off-gas, it is not appropriate.
C.
Correct - ON-I 03 is for Combustion in Offgas. Also, while closing the steam valve would normally be part of isolating the SJAE (plausible), it's desired for the steam valve to remain open under these conditions in order to help purge the offgas system.
D.
Incorrect but plausible if applicant believes that the steam valve closes instead of the air valve due to the SJAE isolation. However, entry into ON-I 03 is appropriate.
References:
LGSOPS0069, ARC 127-B-1, ON-I 03 Student Ref: required No Learning Objective:
LGSOPS0069 IL2, IL4 Question source:
New Quest ion History:
Cognitive level:
Memory/Fundamental knowledge:
X Com p re he nsive/Ana I ys is :
1 OCFR 55.41(4) X
QUESTION 65 Unit 1 plant conditions are as follows:
0 Reactor power is 80%
0 Control Rod 22-47 is at position 10 Control Rod 22-47 is required to be withdrawn to position 22 for a rod pattern adjustment.
WHICH ONE of the following describes the procedural restrictions for positioning Rod 22-47 in accordance with S73.1.A, Normal Operation of the Reactor Manual Control System.
A.
22-47 can be continuously withdrawn to position 22.
B.
22-47 can be continuously withdrawn 3 notches to position 16, allowed to settle, and then a continuous withdraw to position 22 can be performed.
C.
22-47 can be continuously withdrawn to position 18, then single notch withdraw to position 22 must be performed.
D.
22-47 must be withdrawn one notch at a time until it reaches position
- 22.
K&A #
201002 Reactor Manual Control System Importance Rating 3.8 QUESTION 65 K&A Statement:
Justification:
A.
2.1.32 Ability to explain and apply system limits and precautions.
In-Correct but plausible if applicant believes that continuous withdraw can be used up to the target position. This would be acceptable if target position was 48.
B.
In-Correct but plausible if applicant knows that continuous must be stopped two notches before target but believes that continuous can be used again to achieve target position.
C.
Correct - In accordance with S73. I.A, Section 4.3, for continuous withdraws, the withdraw signal is removed two notches before its target position. Single notch withdraws are then performed to the target position.
D.
In-Correct but plausible since this would be the correct method if the difference between current and target positions is three notches or less.
References:
S73.1.A Student Ref. required No Learning Objective:
Obj. 1 (LLOT0085) Trainee shall perform the following from memory...: Recognize responsibilities of licensed operator with respect to reactivity maneuvers Question source:
New Question History:
Cognitive level:
Memo r yl Fu nda menta I know I edge :
X Comprehensive/Analysis:
1 OCFR 55.41(10) X
QUESTION 66 Plant conditions are as follows:
Unit 1 is at 100% power 0
Unit 2 is in OPCON 5 0
Unit 2 Core Shuffle is in progress 0
A fuel bundle is grappled but remains seated in the core There are no other activities in progress on the Fuel Floor ARM Channel 32 - NEW FUEL STORAGE VAULT alarm is received WHICH ONE of the following identifies the required action per ON-I 20, Fuel Handling Problems?
A.
Raise grappled fuel assembly.
B.
Evacuate affected area(s) of the fuel floor.
C.
Verify SBGT system has initiated.
D.
Notify Reactor Engineering to verify shutdown margin.
K&A #
Plant Generic - Conduct of Operations Importance Rating 3.9 QUESTION 66 K&A Statement:
G2.1.44 - Knowledge of RO duties in the control room during fuel handling such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.
Incorrect but plausible if the applicant does slot distinguish between area radiation monitor alarm and observing SRM doublings as indicating inadvertent criticality (ON-120 step 2.1) Raising grappled assembly is response to indicated criticality.
Justification:
A.
B.
Correct - If any Fuel Floor Radiation Monitor alarms, and is not due to object handling near water surface which is immediately re-submersible, THEN Evacuate affected area(s) of the Fuel Floor. (ON-120- step 2.2.1)
C.
Incorrect but plausible if the applicant does not know that SBGT initiation is not an action dedicated to an ARM alarming.
D.
Incorrect but plausible if the applicant does not recognize that ARM alarming is not indicative of inadvertent criticality.
References:
ON-120, LLOT760 pg. 37, ARC 109 C-Student Ref: required No 5, D-5 Learning Objective:
N/A Question source:
Bank (Limerick)
Question His tory:
None Cognitive level :
Memory/Fundamental knowledge:
X Com p re hensive/Ana I ysi s :
IOCFR 55.41(10 X
)
QUESTION 67 Unit '2 is at 100% power when the following occurs:
0 Main Turbine trips.
0 RPS fails to de-energize.
0 Four (4) SRVs open automatically to control reactor pressure.
Two minutes later, the following conditions exist:
0 MCR Annunciator 107 C-I, SCRAM DISCHARGE VOLUME HIGH LEVEL TRIP is lit.
0 All but five (5) control rods are fully inserted.
WHICH ONE of the procedures can be used to insert the control rods that did not fully insert?
A.
T-214, Manual initiation of ARI.
- 6.
T-215, De-energization of Scram Solenoids.
C.
T-216, Manual Isolation and Vent of Scram Air Header D.
T-217, RPS/ARI Reset And Backup Method Of Draining Scram Discharge Volume.
K&A #
Plant Generic Importance Rating 4.3 QUESTION 67 K&A Statement:
Justification:
A.
G.2.1.23 - Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Incorrect. ARI has already initiated. This procedure does not provide actions needed to insert remaining rods. Plausible because this T-200 procedure is called out in any ATWS.
B.
Incorrect. Scram air header is already depressurized as a result of ARI. Plausible because this procedure is called out for an electrical ATWS.
C.
Incorrect. Scram air header is already depressurized as a result of ARI. Plausible because this procedure is called out for an electrical ATWS.
D.
Correct. Scram valves are open due to automatic initiation of ARI. This procedure is appropriate given the conditions.
References:
LGSOPS0046 Rev001 LLOTI 560, Rev 007 Student Ref. required Yes T-I 01, w/out entry conditions or notes Learning Objective:
IL 5(LLOT1560)
Question source:
New Question History:
LGS NHC-05, OYS CERT-04 Cognitive level:
Memory/Fundamental knowledge:
ComprehensivelAnalysis:
X 1 OCFR 55.41 X
QUESTION 68 Unit 2 plant conditions are as follows:
0 Unit 2 is at 100% power 0
2A CRD pump tripped 10 minutes ago Annunciator CRD HYDRAULIC HI TEMP (108 G-5) is lit 0
Plant operator reports high temperatures greater than 250 degrees on 30 CRDs A plant operator is placing CRD Pump 2B in service using procedure S46.1.A, Control Rod Drive Hydraulic System Startup.
2A PUMP ONLY Slowly THROTTLE 046-201 4A, A CRD Pump Min-Flow Stop Valve, approximately 5 - 5 3i4 turns in close direction from full open position.
2B PUMP ONLY Slowly THROTTLE 046-1014B, B CRD Pump Min-Flow Stop Valve, approximately 5 114 - 5 112 turns in close direction from flu11 open position.
4.18 4.1 9 WHICH ONE of the following describes the proper actions based on the conditions above?
A.
CRS initiates a Procedure Problem Identification System issue (PPIs),
then plant operator repositions 046-201 4B.
B.
Plant operator repositions 046-2014B to correctly lineup the system to immediately restore cooling flow to reactor recirc pumps and CRDs.
C.
CRS processes a Temporary Change (TC), then plant operator repositions 046-201 4B.
D.
Plant operator repositions valve 046-201 4B to correctly lineup the system to immediately restore cooling flow to reactor recirc pumps and CRDs, then CRS processes a Temporary Change (TC) to document system alignment.
K&A @
Equipment Control Importance Rating 3.0 QUESTION 68 K&A Statement:
Justification:
A.
G.2.2.6 - Knowledge of the process for making changes to procedures.
Incorrect. Errors in equipment numbers shall be considered TC's. The change is beyond a PPI.
- 9.
Incorrect. Procedure must be corrected prior to manipulating the valve.
C.
Correct. Per HU-AA-104, the procedure must be addressed first. Errors in equipment numbers shall be considered TC's.
D.
Incorrect. Per HU-AA-104, the procedure change must be addressed before the valve can be manipulated.
References:
HU-AA-104 Student Ref: required No Learning Objective:
None Question source:
Bank (Limerick)
Question History:
Cognitive level:
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
1 OCFR 55.41(10 X
)
QUESTION 69 WHICH ONE of the following describes a condition that will violate a Unit 1 Technical Specification Safety Limit?
A
- 6.
C.
D.
Drywell pressure rises to 60 psig Reactor level drops to -1 70 inches Reactor pressure rises to 1280 psig Minimum Critical Power Ratio (MCPR) lowers to 1.I 0
K&A #
Equipment Control Importance Rating 4.0 QUESTION 69 K&A Statement:
Justification:
A.
2.2.22-Knowledge of limiting conditions for operations and safety limits Incorrect-Drywell pressure is not a safety limit but plausible since it exceeds the maximum design pressure for the drywell.
B.
Correct - Although the limit does not apply in OPCON 1, the plant will scram at
+12.5 inches (decreasing) and be in an applicable OPCON at that point.
C.
Incorrect - Reactor pressure rising to 1280 psig does not exceed a Safety Limit but plausible since it is above the Safety valve setpoint.
D.
Incorrect Plausible since MCPR must be greater than I
.09 with one recirc loop in operation or 1.07 with 2 two loops in operation.
References:
Tech Spec. 2.1, LGSOPSI 800 Student Ref: required No Learning Objective:
N/A Chg: init pwr level, MCPR distractor.
Question source:
Modified (Limerick)
Quest ion His tory:
Cognitive level:
Memory/Fundamental knowledge:
X ComprehensivelAnaIysts:
QUESTION 70 Unit 2 plant conditions are as follows:
0 HPCl has started on low reactor level Level is 20 inches and rising.
0 Trip Unit B21-2N693B fails upscale (HPCI level 8 trip).
Which ONE of the following describes the HPCl system response to the failed trip unit?
A.
HPCl trips.
Automatic low reactor level start capability is lost.
B.
HPCl remains running.
Automatic low reactor level start capability is maintained C.
HPCl remains running.
Automatic low reactor level start capability is lost.
D.
HPCl trips.
Automatic low reactor level start capability is maintained
K&A #
Equipment Control Importance Rating 3.6 QUESTION 70 K&A Statement:
Just if ica t ion:
A.
G2.2.37-Ability to determine operability andlor availability of safety related equipment Incorrect. There are four level 8 trip units. The logic is one-out-of-two taken twice, required to initiate a HPCl turbine shutdown. Plausible if applicant believes that the HPCl turbine will trip on a single Level 8 channel upscale and that auto start is also impacted.
B.
Correct. The high level trip logic is not met. Automatic low level start is not affected.
C.
Incorrect. The high level trip logic is not met and auto start is not affected.
Plausible if applicant thinks the high level failure impacts the low level start logic.
D.
Incorrect. There are four level 8 trip units. The logic is one-out-of-two taken twice, required to initiate a HPCl turbine shutdown. Plausible if applicant believes that the HPCl turbine will trip on a single Level 8 channel upscale and that auto start is not impacted.
References:
LLOT0340 pg. 32 Student Ref: required No Learning Objective:
N/A Question source:
New Question History:
None Cognitive level:
Memory/Fundamental knowledge:
X Comprehensive/Analysis IOCFR 55.41(7) X
QUESTION 71 Repairs on a piece of equipment are planned in a high radiation area.
Temporary shielding will be installed. Job estimates are as follows:
e e
e Job duration - 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (NOT including shielding install I remove time)
Shielding installation / removal duration - 15 minutes each Dose rate at the piece of equipment without shielding - 1.2 R/hr Dose rate with the shielding installed - 125 mR/hr Your current exposure is 1200 mrem. You are expected to install and remove the shielding as part of completing this job.
Which one of the following contains the correct dose that would be received AND correctly identifies whether this is within or exceeds the annual dose control limit (ADCL)?
A.
Total dose for this job will be 675 mrem AND you will exceed the ADCL.
B.
Total dose for this job will be 675 mrem AND you will NOT exceed the ADCL.
C.
Total dose for this job will be 975 mrem AND you will exceed the ADCL.
D.
Total dose for this job will be 975 mrem AND you will NOT exceed the ADCL.
K&A #
Plant-wide Generic I m portance Rating 3.2 QUESTION 71 K&A Statement:
Justification:
A.
G2.3.4 - Knowledge of radiation exposure limits under normal or emergency conditions.
Incorrect. Plausible if applicant does not account for both installation and removal of shielding and, additionally, does not know the ADCL limit of 2000 mRem.
Total dose = (I x 0.25 x 1200) + (3 x 125) + 1200 mrem = 675 mrem + 1200
= 1,875 mrem
- 6.
Incorrect. Plausible if applicant does not account for both installation and removal of shielding.
Total dose = (I x 0.25 x 1200) + (3 x 125) + 1200 mrem = 675 mrem + 1200
= 1,875 mrem C.
Correct.
Total dose = (2 x 0.25 x1200) + (3 x 125) + 1200 mrem = 975 mrem + 1200
= 2.125 mrem D.
Incorrect. Plausible if applicant does not know the ADCL limit of 2000mRem.
Total dose = (2 x 0.25 x1200) + (3 x 125) + 1200 mrem = 975 mrem + 1200
= 2,125 mrem
References:
LLOTI 760, Rev. 01 0 Student Ref: required No Learning Objective:
Obj 5 Question source:
Modified Bank (Pilgrim)
Changed from an ALARA question to a dose rate question.
Question History:
Pilgrim Cognitive level:
Memory/ F u n d a m e n t a I know I ed g e:
Co m pre he nsive/Ana I ys i s :
X I
OCFR 55.43(5) x
QUESTION 72 Unit 1 plant conditions are as follows:
o A startup is in progress.
o There are indications of leakage on Condensing Pot XY 1 D004A.
The Shift Manager directs you to use P&ID M-42 to determine condensing pot elevation in the drywell to support drywell entry.
Assume the condensing pot is located at the same level as its penetration.
Vessel zero elevation is 2663 above sea level.
WHICH ONE of the following identifies:
0 The maximum allowed reactor power level?
0 IF a low-dose zone for workers during performance of radiological surveys is required per RP-LG-460-105, Drywell Entries at Power?
A.
B.
C.
D.
Maximum Allowed Reactor Power 3% or less 7% or less 3% or less Low-Dose Zone For Workers During Performance Of Rad iolon ical S urvevs Required YES YES NO 7% or less NO
K&A #
Plant-wide Generics Importance Rating 3.2 QUESTION 72 K&A Statement:
G.2.3.12 - Knowledge of radiological safety practices pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
Justification:
A.
B.
C.
D.
Incorrect but plausible if the applicant determines the instrument to be located at the 296 Elevation. Step 4.1.2.1 of RP-LG-460-105 states that Reactor Power Levels shall be maintained 3% or less for access to 296elevation. Also, step 4.9.4 states If access is required on 296 elevation or within 6 feet of bioshield penetrations on 277 (N1, N2, or N8), then USE the foliowing special precautions and radiological survey techniques:. One of the special precautions states IDENTIFY low dose area and ALLOW worker to stand in low dose area (Le., stairwell or outside of drywell) while radiation surveys are performed.
Incorrect but plausible if the applicant determines the instrument to be located at the 277 Elevation. Step 4.1.2.2 of RP-LG-460-105 states that Reactor Power Levels shall be maintained 7% or less for access to 277, 286, 303 or 31 3 elevations.
Also, step 4.9.4 states lf access is required on 296 elevation or within 6 feet of bioshield penetrations on 277(NI, N2, or N8), then USE the following special precautions and radiological survey techniques:. One of the special precautions states IDENTIFY low dose area and ALLOW worker to stand in low dose area (/.e.,
stairwell or outside of drywell) while radiation surveys are performed..
Incorrect but plausible if the applicant determines the instrument to be located at the 296 Elevation and doesnt read the other restriction concerning the establishment of a low dose area. See explanation in choice A.
Correct - The applicant must first determine the elevation Condensing Pot XY 1 D004A is located. Using M-42 Sheet 1, it can be determined that Condensing Pot XY 1D004A comes off RPV penetration N12 and that Vessel Zero is 266 3. Using Figure I:
Elevation Correlation Chart on M-42 Sheet 2, it can be determined that N12 is 599.0 above Vessel Zero, which would put the Condensing Pot on 313 Elev.
(266 3 + 599.0 = 266 3 + 49 11 = 316 2 or - 313).
Step 4.1.2.2 of RP-LG-460-105 states that Reactor Power Levels shall be maintained 7% or less for access to 277, 286, 303 or 313 elevations.
References:
LLOTI 760, Rev. 01 0 Student Ref. required Yes ON-122, Rev.017 RP-LG-460-105, Drywell Entries at Power.
M-42, Sheets 18.2 Learning Objective:
Obj. 11 (LLOTI 760)
Question source:
New Question History:
Cognitive level:
Memory/ F u nd a menta 1 know ledge :
Co m pre hensiveiAnal ysis :
X IOCFR 55.41 X
QUESTION 73 Unit 1 Plant conditions are as follows:
0 Division 2 DC is de-energized GP-4, Rapid Plant Shutdown To Hot Shutdown, is performed Following the scram:
0 0
RPV level drops and RClC automatically starts RPV level is restored to normal within one minute WHICH ONE of the following describes the status of the 1A and 1 B Reactor Recirculation Pumps?
A.
B.
C.
D.
1 A Reactor Recirculation Pump Running Running Tripped Tripped 1 B Reactor Recirculation Pump Running Tripped Running Tripped
K&A #
Equipment Control I m port an ce Rating 4.0 QUESTION 73 K&A Statement:
G2.4.21-Knowledge of parameters and logic used to assess the status of safety functions such as reactivity control, core cooling, reactor coolant system integrity, containment conditions, radioactivity release control, etc Justification:
A.
Incorrect. "IN' Recirc Pump: Running; "1 B" Recirc Pump: Running, is not correct because neither Recirc Pump would be running due to RPT breakers for both pumps being tripped.
B.
Incorrect. lvlN1 Recirc Pump: Running; "1 6" Recirc Pump: Tripped, is not correct because the "IA" Recirc Pump would be tripped (not running).
C.
Incorrect. "IA" Recirc Pump: Tripped; "1B" Recirc Pump: Running, is not correct because the "1 B" Recirc Pump would be tripped (not running).
D.
Correct. RPT breaker logic will normally trip both RPTs when RPV level drops to -
38 inches. The RCIC System automatic start indicates that RPV level has reached -
38 inches. The loss of Division 2 DC will prevent one RPT breaker per Recirc Pump from tripping, but the other RPT breaker will trip. As a result, both Recirc Pumps will be tripped. Therefore, for the given plant conditions, "IA" Recirc Pump: Tripped; "1 B" Recirc Pump: Tripped, is the correct answer.
References:
E-I FB, ARC MCR 11 1 C-5, ARC MCR Student Ref: required No 112 D-3, GP-18, Step 1.O Learning Objective:
N/A Question source:
Bank (Limerick)
Question History:
NRC Exam 2006 Cognitive level:
Memory/Fundamental knowledge:
Com prehensive/Analysis:
X 1 OCFR 55.41(7) X
QUESTION 74 Unit 2 initial plant conditions are as follows:
0 Reactor power is 40%.
0 RPV level is 32.
The following MCR alarm windows annunciate on Panel 21 2 0
0 D-2, 2B RECIRC PUMP MOTOR TRIP E-3,2B RECIRC M-G GENERATOR LOCKOUT TRIP Reactor Recirc flows are:
0 e
RRP A Drive flow = 44,000 gpm RRP B Drive flow = 0 gpm WHICH ONE of the following describes the required immediate operator actions in accordance with OT-112, Recirculation Pump Trip?
A.
Immediately manually scram the reactor.
Manually control RPV level as necessary until RPV level is normal.
B.
Manually control RPV level as necessary until RPV level is normal.
If a scram condition occurs then enter T-I00 Scram / Scram Recovery.
C.
Manually adjust RRP drive flow to ensure plant is outside the OPRM Trips Enabled region.
If a scram condition occurs then enter T-100 Scram / Scram Recovery D.
Manually adjust RRP A flow to ensure plant is outside the OPRM Trips Enabled region.
Manually control RPV level as necessary until RPV level is normal.
K&A #
Plant-wide Generic Importance Rating 4.6 QUESTION 74 K&A Statement:
G2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
Justification:
A.
Incorrect but plausible since these would be the actions if both recirculation pumps were running.
- 6.
Correct. With only one RRP trip, the first "applicable" immediate actions are controlling RPV level and scramming if a scram condition occurs.
C.
Incorrect but plausible since the tripping of an RRP can cause the plant to enter the area where OPRM trips are enables (Le., THI possible). The second action (Scram if a scram condition occurs) is a correct imrriediate action.
D.
Incorrect but plausible since the tripping of an RRP can cause the plant to enter the area where OPRM trips are enables (i.e., THI possible). The second action (RPV level control) is a correct immediate action.
References:
LLOTI 540, Rev. 009 Learning Objective:
OT-I 12 Obj 2, (LLOTI 540)
Question source:
New Question History:
Cognitive level:
10CFR Student Ref: required No Me m or y/ F u n d a m e n t a I know I edge :
X Comprehensive/Analysis:
55.41 X
QUESTION 75 Unit 2 is operating at 100% power. The following alarm windows annunciate on MCR Annunciator Panel 21 1 :
A-I, A RECIRC PUMP SEAL STAGE HI/LO FLOW A-2,2A RECIRC PUMP SEAL LEAKAGE HI FLOW The 2A RRP seal indications are as follows:
0 Seal#l pressure 380 psig 0
Seal #2 pressure 350 psig Which of the following resulted in the above alarms and indications on the 2A Recirculation Pump?
A.
Only Seal # I has failed.
B.
Breakdown orifice between the seals has plugged.
C.
Only Seal #2 has failed.
D.
Both Seal # I and Seal #2 have failed.
K&A #
Plant-wide Generic Importance Rating 4.2 QUESTION 75 K&A Statement:
Justification:
A.
G2.4.31 - Knowledge of annunciator alarms, indications, or response procedures.
Incorrect but plausible since the pressures would be about equal in both seal #I and seal #2. However, the pressure would be much higher (-1,000 psig) if seal # I failed.
Also only alarm A-I would be received if seal #I failed.
B.
Incorrect but plausible since both of the listed alarms (A-I and A-2) would be received if the breakdown orifice plugged. However, pressure on #2 seal would decrease while pressure on #2 seal would remain at approximately 1,000 psig.
C.
Incorrect but plausible since both of the listed alarms (A-I and A-2) would be received if #2 seal failed. However, pressure on #I seal would remain approximately 1,000 psig and #2 seal pressure would drop below 500 psig.
D.
Correct. Normal pressure on the # I and #2 seals would be 1,000 psig and 500 psig, respectively. Approximately equally low pressure on both seals is an indication of failure of both pump seals.
References:
LGSOPS0043A, Rev. 001 Student Ref: required No Learning Objective:
IL 6 Question source:
Modified Bank (Brunswick)
Modified conditions to change answer from plugging #I seal breakdown orifice to failure of both seals.
Question History:
BRWK NRC-07 Exam Cognitive level:
Memory/Fundamental knowledge:
X Com pre hensive/Ana lysis:
IOCFR 55.41 X
QUESTION 76 Unit 1 plant conditions are as follows:
0 1 6 Recirc Pump tripped IA Recirc Pump speed is 52%
Core Plate delta P is 1.O psid 0
Reactor Power is 57%
0 One OPRM is out of service for testing, all other equipment is in service The following alarms are observed:
16 RECIRC M-G DRIVE MOTOR TRIP 0
OPRM TRIPS ENABLED WHICH ONE of the following actions is required?
A.
Manually SCRAM the reactor B.
Restart the 1 6 Recirc Pump C.
Lower IA Recirc Pump speed D.
Insert control rods per RMSl
K&A #
295001 Partial or Complete Loss of Core flow Importance Rating 3.5 QUESTION 76 K&A Statement:
A2.01 - Ability to determine and/or interpret Powerlflow map as it applies to PARTIAL OK COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION.
Justification:
A.
Incorrect but plausible if applicant does not differentiate between the actions to take with OPRMs operable. OT-I 12 directs operator to SCRAM if OPRMs are inoperable.
B.
Incorrect but plausible if the applicant know that restarting the tripped recirc pump is not allowed prior to exiting the restricted region.
C.
Incorrect but plausible if applicant does not understand power to flow map.
D.
Correct - OT-112 directs either increasing core flow or inserting control rods to exit the restricted region.
References:
OT-I 12 pg.4, Attachment 1, LLOT2.75 Student Ref: required
- Yes, power to flow map without labeling Learning Objective:
NIA Quest ion source :
Bank (Limerick)
Question History:
None Cognitive level :
Memo ry/F u n d a m e n t a I know ledge:
Co m pre he n s ivelAna I ys is :
X 1 OCFR 55.41(7) X
QUESTION 77 Unit 1 plant conditions are as follows:
0 Unit startup is in progress per GP-2 0
Reactor Power is 24%
0 Turbine first stage pressure is 186 psig The Main Turbine, Main Shaft Oil Pump Discharge pressure drops to 99 psig.
WHICH ONE of the following describes the expected plant response to the above AND the procedure to respond to the event?
A.
- 6.
C.
D.
Main Turbine Remains on Line Shutdown the Main Turbine per S01.2.A Main Turbine Trips Reactor Power increases Enter OT-104, Unexpected/Unexplained Positive or Negative Reactivity Insertion Main Turbine Trips Reactor Scrams Enter T-I 00, Scram / Scram Recovery Main Turbine Trips Reactor Recirculation Pumps Trip Reactor Scrams Enter T-100, Scram / Scram Recovery
K&A #
295005 Main Turbine Generator Trip 3.9 I m po rt a n ce Rating QUESTION 77 K&A Statement:
Justification:
A.
A2.05 - Ability to determine and/or interpret Reactor power as it applies to MAIN TURBINE GENERATOR TRIP.
In-Correct. Plausible if applicant believes Main Turbine, Main Shaft Oil Pump Discharge pressure dropping to 99 psig will not result in a turbine trip.
- 6.
Correct. A turbine trip occurs on low EHC pressure and results in a reactor scram and Reactor Recirculation Pump (RRP) trip if first stage pressure 2 190 psig power which correlates to a power level of 25%. Under the current conditions (i.e first stage pressure < 190 psig) the turbine will still trip. But instead of a reactor scram, reactor power will increase due to increased feedwater subcooling since feedwater heating is lost when extraction steam is isolated by the turbine trip.
C.
Incorrect. Plausible since a reactor scram is the expected response when power is above 25%. However, the 25% reactor power setting is based on a corresponding turbine first stage pressure of 190 psig. At a turbine first stage pressure of 186 psig, a reactor scram will not occur.
D.
Incorrect. Plausible since a RRP trip and reactor scram are expected responses when power is above 25%. However, the power level setting is based on a corresponding turbine first stage pressure of 190 psig. At a turbine first stage pressure of 186 psig, neither a reactor scram nor a RRP trip will occur.
References:
LGSOPSOOOIA, Rev. 000 Student Ref: required No Learning Objective:
IL3, IL4 Question source:
New Quest ion His tory:
Cognitive level :
Memo rylFundamen ta I knowledge:
Comprehensive/Analysis:
X 1 OCFR 55.43 (5) x
QUESTION 78 Unit 2 Plant conditions are as follows:
0 Reactor power is 80%
0 0
0 EO reports smoke is coming from the Remote Shutdown Panel Smoke and noxious odor is coming from the back of Unit 2 main control room panels Smoke and odor quickly engulf the control room Shift Supervisor orders IMMEDIATE evacuation of the control room WHICH ONE of the following describes:
0 0
An Immediate Operator Action required to be completed prior to leaving the control room Proper procedure for shifting control of equipment outside the Main Control Room A.
Place the Mode Switch in shutdown.
Enter SE-1-2.
- 0.
Place the Mode Switch in shutdown.
Enter S E-6.
C.
Initiate RCIC.
Enter SE-1-2.
D.
K&A #
295016 Control Room Abandonment Importance Rating 4.2 QUESTION 78 K&A Statement:
Justification:
A.
2.4.1 1 - Knowledge of Abnormal Condition procedures as it applies to CONTROL ROOM ABANDONMENT Incorrect - Entry into SE-1-2 is required on a Loss of Offsite Power. Entry into SE-6 is required based on plant conditions. Plausible since placing mode switch in shutdown is an immediate action.
B.
Correct - SE-1 immediate operator actions is listed. Entry into SE-6, Alternate Remote Shutdown, is required because the Remote Shutdown Panel is not accessible due to fire in the area.
C.
Incorrect - Initiating RClC is not an immediate operator action per SE-1, Remote Shutdown. Entry into SE-1-2 is required on a Loss of Offsite Power. Plausible since initiating RClC is a follow up action in SE-1.
D.
Incorrect - Initiating RClC is not an immediate operator action per SE-1, Remote Shutdown. Plausible since initiating RClC is a follow-up action in SE-1. Entry into Se-6 is required.
References:
SE-1, Remote Shutdown, LLOT0735 Student Ref: required No Learning Objective:
None Question source:
Modified Bank (LaSalle)
Question History:
none Cognitive level:
Memory/Fundamental knowledge:
X Co m pre hensive/Ana lysis:
1 OCFR 55.43(5) x
QUESTION 79 Unit 1 is shutdown with the "A" RHR loop in Shutdown Cooling with the following plant conditions :
0 Reactor pressure 25 psig 0
0 MSlVs are closed 0
Main Turbine is tripped Reactor Level is +60 inches 1A RPS & UPS DlST PNL TROUBLE alarm annunciates.
WHICH ONE of the following identifies the procedure that should be implemented first?
A.
S44.0.A Operating RWCU to Meet Plant Conditions.
B.
ON-I 13 LOSS Of RECW.
C.
D.
OT-I 12 Recirculation Pump Trip ON-I21 Loss of Shutdown Cooling.
K&A #
295021 Loss of Shutdown Coo I i ng lrnportance Rating 3.9 QUESTION 79 K&A Statement:
Justification:
A.
G2.1.20- Ability to interpret and execute procedure steps as it relates to LOSS OF SHUTDOWN COOLING Incorrect. This is not a high priority with the plant shutdown. Plausible since a loss of RWCU will occur and recovery will be needed to assist with RCS level control.
- 5.
Incorrect. This is not a high priority with the plant shutdown. Plausible since a loss of RECW will occur.
- c.
Incorrect. This is not a high priority with the plant shutdown. Plausible since both reactor recirculation pumps will receive a tripped signal (turbine is tripped).
D.
Correct. The stem conditions provide that Shutdown Cooling suction isolation valves have isolated (Loss of 1A RPS UPS POWER). Since there is no other viable decay heat removal method available then the first thing to address is the loss of Shutdown Cooling.
References:
On-I 13, OT-I 12, OT-121, S44.O.A, Student Ref: required No ARC -1 20 F5, LLOT0370, LLOTI 550 Learning Objective:
LLOTl550 0bj.l Question source:
Bank Quest ion History:
Cognitive level:
Memo ryl Fu nda men ta I know ledge:
ComprehensivelAnalysis:
X IOCFR 55.43(5) x
QUESTION 80 A steam leak in the Drywell forced the crew to perform a rapid plant shutdown on Unit 1.
The scram was NOT successful and subsequently, the Main Turbine tripped.
Unit 1 plant conditions are as follows:
Reactor Power is 18%
RPV Level is -170" Reactor pressure is 990 psig Suppression Pool level is 22' and stable Drywell pressure is 19 psig and rising slowly Drywell temperature is 225°F and rising slowly Suppression Pool pressure is 14 psig and rising slowly Suppression Pool temperature is 150°F and rising slowly WHICH ONE of the following actions is required due to the conditions above?
A.
Perform Emergency Blowdown.
B.
Reduce RPV pressure to less than 900 psig.
C.
Spray the Drywell unless required for Core cooling.
D.
Raise Suppression Pool level with condensate transfer.
K&A#
295025 I m portance Rating 4.1 QUESTION 80 K&A Statement:
Justification:
A.
A2.03 - Ability to determine and/or interpret Suppression Pool Temperature as it applies to HIGH REACTOR PRESSURE.
Incorrect. EB is not directed by either PSP or HCTL and lowering pressure will allow you to maintain safe on the HCTL. Blowdown during an ATWS is a last-resort strategy. Plausible if applicant believes carmot maintain safe side of PSP or HCTL curves.
B.
Correct. Reducing RPV pressure to 900 psig maintains operation on the safe side of HCTL curve.
C.
Incorrect. Current conditions do not allow drywell spray per drywell spray initiation limit curve. Plausible because drywell temperature and pressure are high.
D.
Incorrect. T-102 does not direct the use of T-235 at this SP level. Plausible because at low limit for suppression pool level.
References:
LLOTI 560, Rev. 01 2 T-I02 Bases, Rev 022 Student Ref: required Yes T-I 01 T-102 Learning Objective:
IL 6 Question source:
New Question History:
Cog nit ive level :
Memory/Fundamental knowledge:
Comprehensive/Analysis:
X 1 OCFR 55.43 (5) x
QUESTION 81 Unit 2 is operating at 100% power when a major seismic event occurs:
Current conditions are as follows:
0 Reactor has scrammed.
0 MSlVs are closed.
RPV Pressure is 625 psig.
0 Suppression Pool parameters:
0 Level = 4 0 0
H2 conc = 4.5%
0 O2 conc = 6.4%
0 H2 conc = 4.2%
0 O2 conc = 6.2%
0 Drywell parameters:
Based on the above conditions, which one of the following describes the course of action required AND the basis for the action?
A.
B.
C.
D.
Depressurize RPV using Turbine Bypass Valves.
Basis: To prevent exceeding the pressure capability of primary containment.
Align nitrogen inerting system to inert/purge the Drywell per T-228 using max flow rate regardless of offsite release rate.
Basis: To displace hydrogen and oxygen from the Drywell while maintaining the Drywell atmosphere inert.
Depressurize RPV using all ADS Valves.
Basis: To use available Suppression Pool Heat Capacity and ensure primary containment integrity.
Align the nitrogen inerting system to inert/purge the Suppression Pool per T-228 using max flow rate regardless of offsite release rate.
Basis: To displace hydrogen and oxygen from the Drywell while maintaining the Drywell atmosphere.
K&A#
295030 lrnportance Rating
- 4. I QUESTION 81 K&A Statement:
Justification:
G.2.4.18 - Knowledge of the specific bases for EOPs as it relates to LOW SUPPRESSION POOL LEVEL.
A.
B.
C.
D.
Correct - With SP level below 4.2 (4 2 %), decision step EB-9 directs the user to step EB-16 of T-112, Emergency Blowdown, directs using one of several flowpaths to depressurize the RPV. The Turbine Bypass Valves are one of the listed flowpaths. Below a depth of 4.2, the SRV tailpipes will become uncovered which will negate the quenching function of the SP and result in the direct pressurization of primary contain men t.
Incorrect but plausible. H2 / O2 concentrations meet the criteria for performance of T-102 DW/G-1 leg. Conditions have been met for inerting the Drywell with N2.
However, the actions to purge/inert the Drywell regardless of offsite release rate are associated with performance of T-102 DWIG-3 leg and H2 concentration isnt high enough (i.e., 5.99%) to warrant entry into DW/G-3. The basis is correct for the action to push the H2 / 02out of the Drywell space while maintaining an inert atmosphere.
Incorrect but plausible. If SP level is above 4.2 (4 2 %I),
step EB-9 directs the user to step to perform step EB-10 thru EB-12 to open ADS valves and quench the reactor discharge within the volume of the SP. However, the SP level is just below the limit where ADS valves are used to depressurize the reactor.
Incorrect but plausible since the H2 / O2 concentrations meet the criteria for performance of leg SP/G-1 of T-102. All conditions have been met for inerting the Suppression Pool with nitrogen. However, the actions to purgehnert the SP regardless of offsite release rate are associated with the performance of leg SPIG-3 of T-102 and H2 concentration isnt high enough (i.e., 5.99%) to warrant entry into the SP/G-3. The basis for the action to displace the H2 / 02from the SP vapor space with the inert nitrogen gas is correct.
References:
LLOT1560, Rev. 01 2 Student Ref: required Yes T-102 Sheets 1 &
2, T-112 Learning Objective:
Question source:
Question History:
Cognitive level:
1 OCFR IL5, IL6 New Memory/Fundamental knowledge:
Comprehensive/Analysis:
X 55.43 (5) x
QUESTION 82 The following annunciators are received in the MCR:
NORTH STACK HI-HI RADIATION (003 E l )
NORTH STACK HI RADIATION (003 E2)
RMMS shows rising radiation levels on the North Stack.
WHICH ONE of the following identifies the source of the radiation and the required action to mitigate the release?
A.
Standby Gas Treatment Exhaust Enter and execute T-I 04 Radiological Release.
- 6.
Standby Gas Treatment Exhaust Perform ST-6-1 04-880-0, Gaseous Effluent Dose Rate Determination.
C.
Reactor Enclosure Equipment Compartment Exhaust Evacuate all unnecessary personnel per SE-24, Plant Evacuations.
D.
Reactor Enclosure Equipment Compartment Exhaust Enter and execute T-I 03, Secondary Containment Control.
K&A ##
295038 High Off-site Release Rate Importance Rating 4.7 QUESTION 82 K&A Statement:
Justification:
A.
G.2.4.6 - Knowledge of EOP mitigation strategies, Incorrect. Conditions for entering T-I 04 have not been established. Plausible if applicant believes hi-hi rad alarm with rising rad is sufficient to enter T-104.
B.
Correct. The alarm response requires ST performance to determine levels.
C.
Incorrect. REECE does not discharge to North stack. Plausible because high rad in reactor enclosure would require SE-24.
D.
Incorrect. REECE does not discharge to North stack. Plausible because high rad in reactor enclosure would require entry into T-103.
References:
T-I 04 Bases, Rev. 01 3 Student Ref. required No Learning Objective:
LLOT-1560 Obj 112 Quest ion source :
Modified Bank (Limerick)
Question History:
Cognitive level:
Memory/Fundamental knowledge:
X Corn pre hensive/Ana lysis:
1 OCFR 55.43 (4) x
QUESTION 83 Unit 1 experienced a LOCA/LOOP 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ago.
Plant conditions are as follows:
Reactor pressure is 450 psig IRMs are on Range 7 Three control rods are NOT full in Suppression Pool pressure is 10 psig Suppression Pool temperature is 125°F Suppression pool level is 23 feet Drywell pressure is 15 psig Drywell temperature is 280°F Available RPV level indications are as follows:
RPV Narrow Level Indication is upscale RPV Wide Range Level Indication is at -140 and rising RPV Fuel Zone Level Indication is at -200 and dropping All other RPV level indications are downscale.
WHICH ONE of the following describes the mitigation strategy to ensure the integrity of the fuel cladding is maintained?
A.
Establish injection with 5 ADS valves open and maintain RPV pressure stable at 60 psig.
- 9.
Establish injection with 5 ADS valves open and maintain RPV pressure greater than 230 psig.
C.
Establish injection to maintain wide range level between +I 2.5 and +54.
D.
Establish injection to maintain fuel zone level between -1 61 and -186.
K&A #
295009 Low Reactor Water Level Importance Rating 4.0 QUESTION 83 K&A Statement:
Justification:
A.
G.2.4.6 - Knowledge of EOP mitigation strategies.
Incorrect. This action is taken ONLY if there is no ATWS. With three rods not full in and power level Range 7 on IRMs, ATWS conditions apply. Plausible if applicant does not follow path of ATWS leg of T-I 16.
- 8.
Correct. This is the action required since the containment conditions and RPV level indication are indicative of a loss of level indication. Level is unknown because level instrumentation is providing opposing readings. Entry into T-I 16 is required and the ATWS path is required. RPV flooding strategy is based on worst-case conditions that RPV level is low. This is the basis for the KA match of this question.
C.
Incorrect. This action requires knowledge of RPV level, which is not satisfied by stem conditions. Plausible if applicant does not identify ATWS condition and T-116 entry requirements.
D.
Incorrect. This action requires knowledge of RPV level, which is not satisfied by stem conditions. Plausible since this is an action in T-I 16 if an ATWS condition is present (which it is).
References:
T-I 01, T-102, T-291, T-116, Step RF-Student Ref: required:
T-1 01,
15 Yes T I 02, T I 16, T117 Learning Objective:
N/A Question source:
Bank (Limerick)
Question History:
None Cognitive level:
Memory/Fundamental knowledge:
Com p re hensive/Ana lysis:
X 1 OCFR 55.41(7) X
QUESTION 84 Unit 2 is operating at 100% power.
An inadvertent Group VlllA LOW LOW LOW LEVEL isolation occurs, causing HV-13-106, Recirc Pump Cooling Water Inboard Isolation Valve, to close.
The isolation signal will not clear and 6 minutes have elapsed since the valve closed.
WHICH ONE of the following describes the capability to bypass the isolation signal and operate HV-13-106?
A.
The isolation signal may be bypassed and the valve re-opened for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> because another isolation valve in series is operable.
B.
The isolation signal may be bypassed and the valve re-opened because less than 10 minutes'have elapsed since the isolation.
C.
The isolation signal may NOT be bypassed and the valve may NOT be re-opened because the administrative requirements controls for bypassing the isolation signal are not adequate.
D.
The isolation signal may NOT be bypassed and the valve may NOT be re-opened because the isolation signal has not cleared.
K&A #
295020 Inadvertent Cont.
Isolation Importance Rating 4.6 QUESTION 84 K&A Statement:
Justification:
A.
G2.1.20 - Ability to interpret and execute procedure steps, as it relates to INADVERTENT CONTAINMENT ISOLATION Correct. GP8.4 allows a containment isolation signal to be bypassed as governed by TS 3.6.3. The TS requires maintaining an operable isolation in the line and isolation of the line within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the inoperable valve cannot be restored. Once the signal is bypassed, the valve may be opened for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from time of the inadvertent isolation.
B.
Incorrect. The valve may be opened but not because of the short time duration.
Plausible if applicant thinks the 10 minute maximum isolation time for trip of the recirc pumps and the reactor is the reason for why the valves may be re-opened.
- c.
Incorrect. The isolation signal can be bypassed since HV-I 3-108 remains operable.
HV-13-108 is an isolation valve in series with HV-13-106. Tech Spec 3.6.3 allows HV-13-106 to be opened provided a second isolation valve in series remains operable. Plausible because TS 3.6.3 basis describes specific administrative controls for re-opening a valve that has been closed to meet TS 3.6.3. These controls are to station a person locally at the isolation valve that is used to isolate the line for the TS LCO Action Statement. The individual must be in constant communication with the control room. These administrative controls do not apply to HV-13-106 at this time because the valve is not yet being used to isolate the penetration to comply with action requirements.
D.
Incorrect. Under the specified conditions, the isolation signal can be bypassed to allow the valve to be opened under restrictions of TS 3.6.3. Plausible if applicant thinks signal bypass procedure requires signal reset.
References:
GP8, GP8.1, GP8.4, M-13, TS 3/4.6.3, Student Ref:
No TS Bases Learning Objective:
N/A Question source:
New Question History:
None Cognitive level:
Memory/Fundamental knowledge:
Comprehensive/Analysis:
X 1 OCFR 55.43 (2) x
QUESTION 85 Unit 1 is at 100°/o power when the following alarms are recieved:
0 0
0 REACTOR ENCL FLOOR DRAIN SUMP PUMP HIGH HIGH WATER LEVEL on MCR 127 1A - I C RHR PUMP ROOM FLOOD on MCR 113 1 B - 1 D RHR PUMP ROOM FLOOD on MCR 114 The EO reports the following:
0 0
0 0
A leak from the Suppression Pool side of the "1 A/1 C" RHR Pump room 26" of water on the floor in the "IA/IC" RHR Pump room and the level is slowly rising Water has backed up into the "1B /ID" RHR Room through the floor drain system 21" of water on the floor in the "1 B/1 D" RHR Pump room and the level is slowly rising Suppression Pool level is 22.5' and slowly lowering.
WHICH ONE of the following describes the required action(s)?
A.
Raise Suppression Pool level by performing T-234, CST to Core Spray.
B.
Transfer house loads, runback recirc to minimum, manually Scram at 60% core flow.
C.
Perform a GP-3 Normal Plant Shutdown.
D.
Enter T-I 12 and perform an Emergency Blowdown.
K&A #
295036 Secondary Containment High Sump/Area water level Importance Rating 3.2 QUESTION 85 K&A Statement:
Justification:
A.
EA2.03 Cause of the high water level Incorrect. This is an action directed by T-102. However, T-234 is only performed when directed by T-I02 and conditions provided do not justify entry into T-102.
Plausible because SP level is lower than normal and decreasing.
B.
Incorrect. Plausible answer per T-103 if the leak was considered a "primary system discharging".
C.
Correct. T-I03 drives this action per Step SCC/L-13 D.
Incorrect. Plausible if primary system discharging and 2 areas > MSO
References:
T-103, T-I 02 Student Ref: required T-102, 103 Learning Objective:
N/A Question source:
Modified Bank (Limerick)
Question History:
None Cognitive level:
Memory/ F u n d a menta I know I edge :
Com pre hensive/Ana I ys is :
X 1 OCFR 55.43(5) x
QUESTION 86 A seismic event resulted in a RPV leak in the Drywell and a rupture of the CST Pliant conditions are currently as follows:
0 0
0 0
MSlVs are closed 0
0 0
An ATWs is on Progress Reactor level is being maintained between -1 61 and -1 86 HPCl is injecting with T-251 complete RClC tripped, but is available Reactor Pressure is 540 psig and stable CST level is off-scale low Suppression Pool temperature is 173°F and slowly increasing 0
Suppression Pool level is 22 1 WHICH ONE of the following describes an immediate concern AND the mitigating action required to be taken?
A. I Heat Capacity of the Suppression Pool has been exceeded, Initiate Emergency Blowdown per 1-1 12, Emergency Blowdown.
HPCI Pump damage is possible.
Secure HPCl per T-117, Level / Power Control.
C.
Heat Capacity of the Suppression Pool has been exceeded.
Initiate RPV depressurization per T-I 01 RPV Control.
D.
HPCl Pump damage is possible.
Establish RPV level control using RClC Per S49.1.C, Recovery From RClC Turbine Trip, then secure HPCI.
QUESTION 86 K&A Statement:
KBA #
206000 High Pressure Coolant Injection Importance Rating 4.2 A2.08 - Ability to (a) predict the impacts of High Suppression Pool Temperature on the HIGH PRESSURE COOLANT INJECTION System; and based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations.
Justification :
A.
Incorrect. Plausible since for higher RPV pressures (i.e., > 700 psig), Suppression Pool heat capacity is exceeded. However, Suppression Pool temperature is just below the Heat Capacity Temperature Limit (-1 80°F) at current SP temperature.
B.
Correct. The bases for step RC/L-4 "...that operating the HPCl and RCIC Systems with high suction temperatures (above 170°F) could result in equipment damage.
With Suppression Pool temperature above 1 70"F, HPCI pump damage is a concern HPCl must be stopped per T-I17 LQ-17.
C Incorrect but plausible since for higher RPV pressures (i.e., > 700 psig), the heat capacity of the Suppression Pool is exceeded. However, at the current RPV pressure, the Suppression Pool temperature is just below the Heat Capacity Temperature Limit (-1 80°F).
D.
Incorrect but plausible since Suppression Pool temperatures above 170°F could result in damage to the HPCl pump. However, the same caution also applies to the RClC pump. High suction temperature to RClC can result in pump damage.
References:
LLOTl560, Rev. 01 2 T-102 Bases, Rev. 022 T-101 Bases, Rev. 019 T-117 Learning Objective:
IL4 Student Ref: required Yes T-101 T-I02 T-117 Question source:
New Question History:
Cognitive leve I :
Memo ry/F u n d a menta I know I ed g e :
Comprehensive/Analysis:
X 1 OCFR 55.43 (5) x
QUESTION 87 Unit 1 is at 100% power with all systems normal when the following alarm is received at 120 D11: 1A RPS & UPS DISTR PNL TROUBLE Plant conditions are as follows:
0 0
0 The PRO states that all Confirming Indications in E-lAYl60, Loss of 1A RPS UPS Power, are present The Reactor Operator reports RWCU isolation has occurred As control room supervisor, you observe various process Rad Monitor Trips and Alarms WHICH ONE of the following describes the appropriate procedure to implement and the basis for use?
Procedure Basis A.
ON-I 13 LOSS of RECW Loss of 1 AYI 60 caused RECW Isolation B.
GP-4 Rapid Plant Shutdown Loss of 1 AY160 caused loss of CRD Flow Controller power C.
OT-I 12 Recirculation Pump Trip Loss of 1AY160 caused Recirc Pump trip D.
ON-I09 Total Loss of SRM, IRM or APRM Systems Loss of 1AY160 caused loss of power to all IRMs
K&A #
21 2000 RPS Importance Rating 4.2 QUESTION 87 K&A Statement:
Justification:
A.
G2.4.11 Knowledge of Abnormal Condition Procedures as they relate to Reactor Protection System Correct. Per stem conditions indicate there has been a loss of 1A RPS UPS power.
Per E-1AY16O this will result in RECW isolation. Step 2.1 (Initial Actions) of E-l AY160 specifies Enter ON-l 13, Loss of RECW.
- 8.
Incorrect. Loss of 1 AY 160 will not affect CRD Flow Controllers (E-I AY 160 Section
-I
.O). Plausible if some other power supply is assumed lost or if associates with RWCU isolation.
C.
Incorrect but plausible. Loss of 1AY160 will not on its own cause a recir pump trip. If Recirc Pumps trip, E-1AY16O directs entering OT-112.
D.
Incorrect but plausibie since half of IRMs (A, C, E, and G) will lose power.
References:
E-1 AY 160 page 1 Student Ref: required No OT-I17 page 3 LGSOPS0071 pages 40 and 41 Learning Objective:
N/A Question source:
Bank (Limerick)
Question History:
None Cognitive level:
Me m or y/ F u n d a m e n t a I know t ed g e :
X Com pre hensive/Ana I ys is :
1 OCFR 55.43(5) x
QUESTION 88 The following plant conditions exist on Unit 1 :
The Unit is at 100% power.
The following MCR alarm windows annunciate on panel 108:
0 A-4, OPRM TRIPS ENABLED 0
B-3, APRM UPSCALE TRIP/INOP 0
B-4, APRM UPSCALE 0
E-3, APRM/RBM FLOW REF OFF NORMAL 0
0 0
Recirc Loop A Drive Flow on FI-043-1 R613 at 42,000 gpm Recirc Loop B Drive Flow on FI-043-1 R617 at 44,200 gpm Total Recirc Loop Drive Flow on Recorder FR-043-1 R614 at 86,200 gpm The following information is displayed on the ODA for the APRM channels:
WHICH ONE of the following identifies the failure AND the appropriate operator action?
A.
An upscale LPRM failure associated with APRM 2 has occurred resulting in fewer than the required number of LPRM inputs.
Perform RT-6-074-360-1 and RT-6-074-361-1, Verification Of APRM Operability.
B.
Flow transmitter FT-043-1 N024B for B RRP has failed to zero for APRM 2.
Ensure compliance with Tech Specs and then bypass APRM 2 per ARC 108 E-3, APRM/RBM FLOW REF OFF NORMAL.
C.
A downscale LPRM failure associated with APRM 2 has occurred resulting in fewer than the required number of LPRM inputs. Perform RT-6-074-360-1, Verification Of APRM Operability.
D.
Flow transmitter FT-043-1 N024A for A RRP has failed to zero for APRM 2.
Ensure compliance with Tech Specs and then bypass APRM 2 per ARC 108 E-3, APRM/RBM FLOW REF OFF NORMAL.
K&A #
21 5005 APRM/LPRM Importance Rating 3.4 QUESTION 88 K&A Statement:
A2.07 Ability to predict (a) the impacts of Recirculation flow channels flow mismatch on the APRM/LPRM System; and based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations.
Justification:
A.
B.
C.
D.
Incorrect but plausible since an upscale failure of an LPRM could result in Simulated Thermal Power increasing and would cause an APRM upscale/inop alarm. The LPRM failure would effect APRM, RBM, and OPRM indications. However, the condition would result in an APRM Upscale/lnop alarm only. Also, an upscale failure of an LPRM would not have any effect on triggering the OPRM enabled alarm. In addition, the LPRM circuit does not count the number of LPRMs that have invalid readings, but tracks how many detectors associated with an APRM are bypassed.
The LPRM failure would also have some effect APRM, RBM, and OPRM indications.
The condition would not affect total recirc drive flow.
Correct. Failure of flow transmitter FT-043-2N024B would result in I ) indicated flow for APRM channel 2 falling below the point (<6O%) where OPRM 2 would be enabled (ARC 108 A-4), 2) APRM flow-biased alarm and trip setpoints would fall to 86.7% and 94.3%, respectively (ARC 108 B-4), and 3) flow transmitter input signal would be less than I mA (ARC 108 E-3). Flow indications for loop and total drive flow would remain unchanged since they receive their flow signals from APRM Channels 1 and $4.
Incorrect but plausible since a downscale failure of an LPRM could result in Simulated Thermal Power decreasing and could cause an APRM downscale alarm.
The LPRM failure would also have some effect APRM, RBM, and OPRM indications.
The condition would not affect total recirc drive flow.
Incorrect but plausible because flow transmitter FT-043-2N024A performs the same function as 2N024B and would have similar result but the effect would be on a different APRM channel.
Student Ref. required Yes P&IDs M-43 ShtS 1 & 2
References:
LLOT0275, Rev. 012 Learning Objective:
Obj. 18 Question source:
New Question History:
Cognitive level:
Memo r y/F u nd a menta I know I ed g e :
Comprehensive/Analysis:
X 1 OCFR 55.43 (5) x
QUESTION 89 The following plant conditions exist on Unit 2:
0 Reactor power is 100%
0 RClC is running in Full Flow Test per S49.1.D The following alarms are received in the MCR:
216 A-I, RClC OUT OF SERVICE 221 G-2,2PPC1/2PPC2 125 VDC DlST PANELS UNDERVOLTAGE WHICH ONE of the following describes the effect this will have on RClC and the appropriate procedure to be performed?
A.
B.
C.
D.
RClC continues running.
HV-49-2F007, RClC Steam Line Inboard isolation valve will fail to respond to a valid isolation signal.
Shutdown the RClC Turbine per S49.1.D, RClC System Full Flow Functional Test.
RClC Isolates.
HV-49-2F007, RClC Steam Line Inboard isolation valve closes. Perform S49.1.B, Recovery from RClC Steam Line Isolation and Resultant Turbine Trip.
RClC continues running.
HV-49-2F008, RClC Steam Line Outboard isolation valve will fail to respond to a valid isolation signal.
Shutdown the RClC Turbine per S49.1.D, RClC System Full Flow Functional Test.
RClC Isolates.
HV-49-2F008, RClC Steam Line Outboard isolation valve closes.
Perform S49.1.B, Recovery from RClC Steam Line Isolation and Resultant Turbine Trip.
K&A #
223002 PCIS/Nuclear Steam Supply Shutoff System Importance Rating 3.2 A2.02 Ability to predict (a) the impacts of D.C. electrical distribution failures on the PClS / NUCLEAR STEAM SUPPLY SHUTOFF SYSTEM; and based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations.
QUESTION 89 K&A Statement:
Justification :
A.
B.
C.
D.
Correct -Div 3 125 VDC distribution panel 2PPCI supplies power to the isolation logic for the inboard RClC isolation valve. Loss of Div 3 125VDC power to the RClC isolation logic will prevent the isolation valve from closing when a valid Grp VA isolation signal is'received. Turbine shutdown is appropriate because not needed for plant shutdown or core cooling.
Incorrect but plausible since most NSSSS isolation groups will isolate on a loss of power to the initiation logic. However, the initiation logic for most NSSSS isolation groups are powered from AC not DC.
Incorrect but plausible since loss of power to Div 1 125 VDC distribution panels would effect the ability of the RClC outboard isolation valve to close on a valid Grp 5A isolation signal. Turbine shutdown is appropriate because not needed for plant shutdown or core cooling.
Incorrect but plausible since most NSSSS isolation groups will isolate on a loss of power to the initiation logic. However, the initiation logic for most NSSSS isolation groups are powered from AC not DC.
References:
LLOT0380, Rev. 024 Student Ref. required No Learning Objective:
111, IL5 Question source:
Modified Bank (Limerick)
Added different conditions and added identification of the correct procedure to study guide (LLOT0380) question Question History:
Cognitive leve I:
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
1 OCFR 55.43 (5) x
QUESTION 90 Plant Conditions are as follows:
Unit 1 is in OPCON 5.
Reactor Enclosure Secondary Containment Integrity is established.
Maintenance is replacing the drive mechanism on Control Rod 14-55.
LPRM String 40-33 is being removed for replacement.
RWCU is providing decay heat removal in accordance with S44.7.8, Using Reactor Water Cleanup as an Alternate Method of Decay Heat Removal to support testing of the SDC isolation signals.
Subsequently, laboratory test results are received, indicating that SBGT charcoal adsorber samples for both SBGT subsystems have failed the methyl iodide penetration test.
WHICH ONE of the following describes the action to take to comply with T.S.
3.6.5.3, Standby Gas Treatment System - Common System?
A.
Suspend testing on the SDC isolation signals. Re-align RHR to SDC AND Suspend alternate decay heat removal using RWCU.
B.
Suspend activities associated with replacing the drive mechanism ONLY.
Suspend activities associated with the removal of LPRM ONLY.
C.
D.
Suspend activities associated with replacing the drive mechanism AND Suspend activities associated with the removal of LPRM String.
KBA #
261000 Standby Gas Treatment System 4.7 I m I:, o rt a n ce Rating QUESTION 90 K&A Statement:
G.2.2.39 - Knowledge of less than or equal to one hour Technical Specification action statements for systems, as it relates to STANDBY GAS TREATMENT SYSTEM.
Justification:
A.
Incorrect but plausible since RWCU operations other than beyond normal makeup or letdown are specifically mentioned in GP-6.2 as operations to be considered as having a potential to drain the reactor vessel. However, only if the systems isolation capability on either Reactor Low Water Level or RWCU High Differential Flow is not maintained.
B.
Correct - T.S. 3.6.5.3, action b. states With both standby gas treatment subsystems inoperable, if in progress, suspend handling of irradiated fuel in the secondary containment, CORE ALTERA TIONS or operations with a potential for draining the reactor vessel.. Section 3.3. of GP-6.2, Shutdown Operations -
Shutdown Condition Tech Spec Actions addresses Suspending Operations with a potential for Draining the Reactor Vessel. Specifically, step 3.3.2.3 identifies CRD maintenance (Le. drive removal and drive replacement) except for HCU hydraulic line maintenance as a specific activity that falls in this category.
C.
Incorrect. Plausible since action b of TS 3.6.5.3 requires suspension of core alterations. However, the movement of LPRM detectors are specifically excluded in definition of core alterations.
D.
Incorrect. First part is correct. Action b of TS 3.6.5.3 requires suspension of core alterations. However, the movement of LPRM detectors are specifically excluded in definition of core alterations. Plausible if applicant thinks both activities must be stopped.
References:
T.S. 3.6.5.3 Student Ref. Required Yes TS3.6.5.3 (pgs 314 6-52 thru 3/4 6-54)
(SGTS)
GP-6.2, Rev.042 LGSOPSOO69 Learning Objective:
lLl0 (LGSOPSOO69)
Question source:
New Quest ion History:
Cognitive level:
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
X
QUESTION 91 Unit 1 is at 100% power, when the following occurs:
I & C determines that the following Transmitters are failed upscale:
o LT-42-1 N095A o LT-42-1 N097A WHICH ONE of the following identifies the status of the Instrumentation LCOs listed below?
TS LCO 3.3.3 A.
Not Met TS LCO 3.3.5 Not Met B.
Met Not Met C.
Not Met Met D.
Met Met
K&A #
Importance 4.7 Rating 21 6000 Nuclear Boiler Instrumentation QUESTION 91 K&A Statement:
G.2.2.22 - Knowledge of limiting conditions for operations and safety limits, as it relates to NUCLEAR BOILER I NSTRU M E NTATlO N.
Justification:
A.
Correct - From M-42 sheet 1, the affected instruments would be:
e From Sheet 2, Table I Water Level Instrumentation Utilization the function and level trip point for each level transmitter can be found in the last two columns of the table. Based on the information, LCO 3.3.3 is not met since the ADS Level 3 Permissive requires a minimum of 1 operable channel (l/trip system; two trip systems with 1 channel per trip system). LCO 3.3.5 is not met since RClC Level 2 and Level 8 trips require a minimum of 4 operable channels (4 per trip function).
LT-1 N095A (ECCS - ADS Level 3 Permissive)
LT-1N097A (RCIC Level 2, 8)
B.
Incorrect but plausible if the applicant doesnt know how many level channels are actually available for the ADS circuit.
C.
Incorrect but plausible if the applicant doesnt know how many level channels are actually available for the RClC circuits, or believes High level trip function is met due to transmitter being in a tripped condition.
D.
Incorrect but plausible since only two channels (one / subsystem) provide signals for the ADS Level 3 permissive. If the applicant doesnt notice the (***) note at the bottom of the page concerning that the Minimum Operable Channels per Trip Function is per subsystem then this choice appears to be a correct answer, and doesnt know how many level channels are actually available for the RClC circuits, or believes High level trip function is met due to transmitter being in a tripped condition.
References:
LGSOPS0042, Rev. 000 Student Ref.
Yes M-42 Nuclear Boiler required Instrumentation T.S Section 3.3 Instrumentation T.S. Sec 3.3, lnst,
M-42, Sheets 1 & 2 Learning Objective:
IL11 Question source:
New Question History:
Cognitive level:
Memory/Fundamental knowledge:
C o m p re h e n s i ve/A n a I y s i s :
X IOCFR 55.43 (2)
X
QUESTION 92 A drywell leak is in progress on Unit I.
The reactor has scrammed. The following conditions currently exist:
0 Suppression Pool Temperature 90°F 0
Suppression Pool Pressure 12 psig 0
Suppression Pool Level 24 feet Drywell Temperature 248°F 0
Drywell Pressure 15 psig 0
Reactor Level + I O inches 0
Reactor Pressure 750 psig WHICH ONE of the following describes required actions?
A. Operate all available Drywell cooling.
Spray the Drywell.
Perform Emergency Blowdown.
B. Operate all available Drywell cooling.
Spray the Suppression Pool.
Perform Emergency Blowdown.
C. Operate all available Drywell cooling.
Spray the Suppression Pool.
Begin a Normal Depressurization.
D. Spray the Suppression Pool.
Spray the Drywell.
Begin Normal Depressurization.
K&A #
230000 RHR/LPCI:
TourslPool Spray Mode Importance Rating 3.7 QUESTION 92 K&A Statement:
Justification:
A.
G.2.4.6 - Knowledge of EOP mitigation strategies.
Incorrect. Plausible if the applicant believes emergency blowdown is required and drywell spray conditions are met.
B Incorrect. Plausible if the applicant believes emergency blowdown is required.
C Correct. Suppression Pool spray is required by T-I 02. Normal depressurization, not emergency blowdown, is appropriate per T-101. The available drywell cooling would be required by T-I 02.
D.
Incorrect. Plausible if the applicant believes drywell spray conditons are met.
References:
T-I 02 (PUP)
Student Ref. required Yes LP LOR-9403C T-I 01 T-IO2 Learning Objective:
LOR-9403C, Obj #6 Question source:
Modified Bank (Limerick)
Changed to require SP Spray vs SP Cooling.
Question History:
LIMERICK LOR 0871 Cognitive level:
Memory/Fu nda me n ta I know ledge:
Co m pre hen sive/Ana lysis:
X 1 OCFR 55.43(5) x
QUESTION 93 Unit 1 plant conditions are as follows:
0 Reactor Power is 20% power.
0 RT-6-031-321-1, Mechanical Trip Valve and Master Trip Solenoid Valves Operability Test is in progress.
Unit Aux Buses 11 and 12 are being powered from the 10 and 20 Startup Buses respectively.
0 When the Oil Trip Pushbutton is depressed, all main turbine stop valves CLOSE.
Two bypass valves fail to open and all other bypass valves function properly.
Which one of the following identifies the expected plant response, AND the procedure to be entered?
A.
Reactor Recirculation Pumps will NOT trip. The Bypass Valves will NOT control pressure below scram setpoint. Enter T-101, RPV Control.
B.
Reactor Recirculation Pumps will NOT trip. The Bypass Valves will control pressure below scram setpoint. Enter OT-I 02, High Reactor Pressure.
C.
Reactor Recirculation Pumps will trip. The Bypass Valves will control pressure below scram setpoint. Enter OT-I 02, High Reactor Pressure.
D.
Reactor Recirculation Pumps will trip. The Bypass Valves will NOT control pressure below scram setpoint. Enter T-I 01, RPV Control.
K&A #
241 000 Reactor/Turbine Pressure Regulator Importance Rating 3.9 QUESTION 93 K&A Statement:
A2.05 - Ability to (a) predict the impacts of Failed openlclosed main stop valve(s) on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations.
Justification:
A.
Incorrect but plausible if the applicant believes the steam flow would exceed the flow capacity of the available bypass valves, then the reactor would scram on high pressure.
B.
Correct - Bypass capacity with all nine BPV can handle up to 30% of full power steam flow, with each bypass valve capable of handling 3.33% of the flow. Two inoperable BPVs would reduce the systems bypass capacity to 23.7% (30% -
2(3.33%)). Since reactor power is less than the bypass capacity, the operable BPVs can control reactor pressure. One entry condition for OT-I02 is the Unexpected/unexplained opening of a Main Turbine Bypass Valve.
C.
Incorrect but plausible if the applicant wisunderstands what will cause the RRPs to trip. While the RRPs will trip when pressure is greater than 1049 psig, the trip is not enabled unless power is above 25%.
D.
Incorrect but plausible if the applicant misunderstands what will cause the RRPs to trip and miscalculates the flow capacity of the available bypass valves. The RRPs will trip when pressure is greater than 1049 psig, but the trip is not enabled unless power is above 25%. The available bypass capacity (23.7%) is greater than the current power level (22%).
References:
LGSOPS0031 B, Rev. 000 Student Ref. required No Learning Objective:
IL2 OT-I 02, Rev.017 Question source:
Modified Bank (Clinton)
Changed power level, word choice.
Added inoperable BPVs.
Quest ion H is tory:
Clinton Bank Cognitive level:
Memo ry/Fu nda m en ta I knowledge :
X Corn p re hens ive/Ana I ys is :
1 OCFR 55.43 (5) x
QUESTION 94 Unit 1 power ascension is in progress following a refueling outage. Reactor power is 60% in two-loop operation.
At 11 05: The Outage Control Center reports that the IA Reactor Recirculation Pump A RPT Breaker has a faulty EOC-RPT trip coil which was not replaced during the outage period.
At 1140: P-I Edit shows Minimum Critical Power Ratio (MCPR) is 1.423.
At 121 5: Reactor engineering reports the new adjusted Minimum Critical Power Ratio (MCPR) limit is 1.473.
W HlCH ONE of the following identifies the required Technical Specification action?
A.
By 141 5, restore MCPR to within the limit or be less than 25% by 181 5.
B.
By 131 5, take action to place Unit 1 in STARTUP by21 15 C.
By 2305, place the affected EOC-RPT trip system in the tripped condition by EOC-RPT channel in the tripped condition.
D.
Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from 1105, restore the EOC-RPT breaker to operable status OR take action to place Unit 1 in STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
K&A #
Plant-wide Generics Importance Rating 4.6 QUESTION 94 K&A Statement:
Justification:
A.
G.2.1.37 - Knowledge of procedures, quidelines, or limitations associated with reactibity management.
Correct. MCPR is below the limit and one RPT trip system is INOPERABLE.
Therefore, TS 3.2.3 Action b applies.
- 8.
Incorrect. Plausible if applicant evaluates MCPR 00s and believes 3.3.2 Action a is not met. This incorrect assessment would lead to 3.0.3 entry.
C.
Incorrect. Plausible if applicant thinks the EOC-RPT trip is a channel failure requiring action under TS 3.3.4.2 Action b.
D.
Incorrect. Plausible if applicant thinks the EOC-RPT trip system inoperability requires TS 3.3.4.2 Action d before they apply TS 3.2.3 Action a.
References:
TS 3.2.3 and 3.3.4.2 Student Ref. required Yes TS 3.2.3 and 3.3.4.2 (Pg 314 2-8, 314 2-9, 314 3-46)
Learning 0 bj e c t i ve :
Question source:
Question History:
Cognitive level :
1 OCFR New Memory/Fundamental knowledge:
Com pre hen sive/Ana I ys is:
X 55.43 (5) x
QUESTION 95 Unit 1 plant conditions are as follows:
0 0
0 GP-2 startup is in progess Reactor Mode switch is in STARTUP/HOT STANDBY Reactor pressure is 210 psig WHICH ONE of the following choices completes the statement below?
All of the listed instruments are required to be OPERABLE per Tech Specs EXCEPT:
A.
Reactor Vessel Pressure High for ARI
- 9.
Reactor Vessel Level 1 for ADS C.
Reactor Vessel Pressure for High Pressure Scram D.
Reactor Vessel Level 2 RWCU System Isolation
K&A #
Equipment Control Importance Rating 4.7 QUESTION 95 K&A Statement:
Justification:
A.
G2.2.22 - Knowledge of limiting conditions for operations and safety limits Correct - ARI is not required in per Tech Specs.
B.
Incorrect - Per Tech spec 3/4 3.3.3-1, pg. 3/4 3-34 instrumentation is required in OPCON 2 with reactor pressure at value noted.
C.
Incorrect - Per Tech spec 3/4 3.3.3-1, pg. 314 3-2 instrumentation is required in OPCON 2.
D.
Incorrect-Per Tech spec 3/4 3.3.3-1, pg. 3/4 3-1 2 instrumentation is required in OPCON 2.
References:
Tech. Spec. 3/4 3.3.3-1 Student Ref: required No Learning Objective:
LLOT Obj. 3, 6a, 7 Question source:
Bank (Fermi)
Question History:
None Cognitive level:
MemorylFundamental knowledge:
X Com prehensive/Ana lysis:
1 OCFR 55.41(10 X
)
QUESTION 96 Unit 1 Reactor power is 100% when the following occurs:
At 0800:
The operator completing ST-6-1 07-590-1, DAILY SURVEILLANCE LOG/OPCON 1,2,3 reports the following for HPCl Drywell Pressure instrumentation:
o Two trip unit channels are reading 0.5 psig o Two trip unit channels are reading 0.1 psig o All channels were reading 0.4 psig on the previous day I&C confirms the readings obtained by the operators.
At 0900:
I&C Technicians completing the ADS Surveillance Procedure report:
o The ADS timer as left value cannot be adjusted below 120 seconds o Manual actuation as well as ADS Timer Override are unaffected by the as left condition Considering a Start time of 0900, Is there an LCO action statement in effect with an action of less When, if ever, will a change in 0 P E RAT IO N AL C 0 N D I T IO N than 7 days?
be required?
A.
- 8.
C.
D.
Yes 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Yes 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br /> No 7 days No OPCON change not required
K&A #
Equipment Control Importance Rating 4.6 G2.2.23 - Ability to track Tech Spec limiting conditions for operation QUESTION 96 K&A Statement:
Justification:
A.
Incorrect - Plausible if applicant declares ADS INOP immediately.
- 6.
Correct - With two HPCl initiation channels inop, HPCl must be declared INOP (3.3.3). With ADS timer setpoint out of spec, 3.3.3 Action 31 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before declaring ADS ECCS subsystem INOP. With both HPCl and ADS inop, TS 3.5.1 Actions do not address situation. Therefore, TS 3.0.3 applies. Mode change action required in 24 plus 1 plus 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br /> total).
C.
Incorrect - Plausible if applicant zeros in on the otherwise statement in 3.3.3.
D.
Incorrect-Plausible if applicant thinks ACTION 30 (a) applies. That is, the only action necessary is to place affected channel in Tripped. This is, typically, what happens with 3.3.3 but is NOT applicable in this instance since ADS timer is also INOP and there is only one channel provided.
References:
ON-120, LLOT760 pg. 37 Student Ref:
YES ST-6-1 07-590-1 TS pgs 3/4 3-32 thru 3-36a, 3-37 w/stpts removed except ADS timer, Pages 5-1 thru 5-3 whotes removed from bottom of 5-1 Learning Objective:
NIA Question source:
New Question History:
None Cognitive leve I:
Memory! F u nd a menta I know ledge :
Com p re hens ive/Ana lysis:
X 1 OCFR 55.43(2)
X
QUESTION 97 Unit 2 plant conditions are as follows:
0 Reactor power is 100%
There is a steam leak from the 2B RWCU pump Dose rates in the 2B RWCU pump room are 2 R/hour Entry is required into the 28 RWCU pump room.
WHICH ONE of the following describes the type of Locked High Radiation Area and required authorization for release of the key?
Type of Locked High Radiation Area for 28 RWCU Pump Required Authorization For Release of Kev A.
Level 1 RP Supervision
- 6.
Level 1 Operations Shift Manager C.
Level 2 RP Supervision D.
Level 2 Operations Shift Manager
K&A #
Radiation Control Importance Rating 4:7 QUESTION 97 K&A Statement:
G2.3.13 - Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
Justification:
A.
Correct - Based on dose rates (2000 mRlhr), the 2B RWCU pump room is a Level I
Locked High Radiation Area (less then 15,000 MWhr). Per RP-AA-460-001 entry into a locked High Rad area requires RP Supervision authorization except in emergencies.
B.
Incorrect - Plausible if applicant does not remember that entry requires RP supervision approval except in emergencies.
C.
Incorrect - Plausible if applicant does not know values for level 2 is greater than 15,000. mR/hr D.
Incorrect - Plausible if applicant does not remember that entry requires RP supervision approval except in emergencies, and does not know values for level 2 is greater than 15,000. mR/hr
References:
RP-LG-460-1010, RP-LG-460-1016, Student Ref: required No LLOTI 760 Learning Objective:
LLOT1760 Obj. 1 Question source:
Bank (Limerick)
Chgs: location, dose Question History:
ILT NRC Exam 2005 Cognitive level:
Memory/Fundamental knowledge:
X Comprehensive/Analysis:
X
QUESTION 98
- ATTACHMENT Partial GP-5 provided****
Unit 1 Reactor power has been 100% for the last 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />.
An Offgas release rate calculation is in progress per GP-5, Steady State Operations.
The following data has been obtained for A and B channels.
0 Previous hourly Off-gas release rate was 99.80 pCi/sec.
Radiation levels as read from RR-026-1 R601, point 1 and 2, respectively:
o Channel A = 37.7 Mrem/hr Channel B = 42.4 Mrem/hr 0
System Flow as read from FR-069-115, SJAE Discharge Flow from Recombiner is 53 scfm.
0 The U1 Off gas placard contains the following information:
I Sum of six values I139 Using the provided portion of GP-5:
WHICH ONE of the following, describes the status of the calculated offgas release rate AND the action to take concerning the calculated offgas release rate?
A.
B.
C.
D.
The calculated offgas release rate is invalid.
Recalculate using flow readings from FIT-070-1 50, HEPA Filter I OF371 Dsh Flow, Local Indication, Low Range.
The calculated offgas release rate has risen.
Request Chemistry to perform ST-5-041 -885-1, Dose Equivalent 1-1 31 De t e r m i n at i o n.
The calculated offgas release rate has risen.
Request Chemistry to perform ST-5-070-885-1, Isotopic Off-gas Analysis.
The calculated offgas release rate has risen.
Request Chemistry to perform ST-5-041 -885-1, Dose Equivalent 1-1 31 Determination AND ST-5-070-885-1, Isotopic Off-gas Analysis.
K&A #
Plant-wide Generics Importance Rating 4.3 QUESTION 98 K&A Statement:
Justification:
A.
G.2.3.11 - Ability to control radiation releases.
Incorrect but plausible since a flow rate below 50 scfm would invalidatethe calculation and require use of FIT-070-150 B.
Incorrect but plausible since an increase of more than 15% from the last hourly calculation in the offgas release rate requires performance of ST-5-041-885-1, Dose Equivalent 1-1 31 Determination. However, this is only applicable if offgas release rates are above 75,00OpCi/sec.
C.
Correct - The applicant must first determine the average offgas release rate using the following formula for each channel.
RR = rad level x offgas flow x K constant Ave RR = (RRchA + RRCh B) / 2 = 171.81 pCi/sec This results in an increase of 72% increase over the last hourly increase.
Performance of ST-5-070-885-1, Isotopic Off-gas Analysis is required if the offgas release rate increases by more than 50% from the last hourly calculation regardless of the offgas release rate level.
D.
Incorrect but plausible since an increase of more than 15% in the offgas release rate calculation requires performance of ST-5-041-885-1, Dose Equivalent 1-1 31 Determination. In addition, an increase of more than 50% in the offgas release rate calculation requires performance of ST-5-070-885-1, Isotopic Off-gas Analysis.
However, ST-5-041-885-1, Dose Equivalent 1-1 31 Determination would only be required if offgas release rates are above 75,00OpCi/sec.
References:
LLOTI 790, Rev. 006 Student Ref. required Yes GP-5, Rev. 135 GP-5, pp 15-16 Spec. step 3. I. 18 Learning Objective:
Question source:
New Question History:
Cognitive level:
Memory/Fundamental knowledge:
Com pre hens ive/Ana I ys is :
X 1 OCFR 55.43 (5) x
QUESTION 99 The following conditions exist on Unit 2.
0 A reactor scram from full power has occurred due to a loss of offsite power.
0 Shortly after the scram, lit alarm windows on MCR annunciator panels 201 through 212, 215 through 217,220 and 222 go out.
0 Many MCR instruments are indicating downscale.
0 The crew enters ON-122, Loss of Main Control Room Annunciators.
0 Twenty minutes later the screens on the Plant Monitoring System go blank.
Which one of the following identifies the required action for the annunciator failure per ON-I 22 AND the Emergency Action Level per EP-AA-1008 for the given conditions?
A.
B.
C.
D.
Required Action per ON-I 22 Dispatch operator to investigate conditions on non-safeguard electrical DC panels.
Dispatch operator to investigate conditions on safeguard electrical distribution panels.
Dispatch operator to investigate conditions on non-safeguard electrical DC panels.
Dispatch operator to investigate conditions on safeguard electrical distribution DC panels.
Emergency Action Level Alert, MA6 Alert, MA6 Site Area Emergency, MS6 Site Area Emergency, MS6
K&A #
Plant-wide Generics Importance Rating 4.0 G.2.4.32 - Knowledge of operator response to loss of all annunciators.
QUESTION 99 K&A Statement:
Justification:
A.
Incorrect but plausible since MCR Annunciator power is provided by Non-safeguard 125 VDC. However, the conditions exceed the Alert Emergency Action Level.
- 0.
Incorrect but plausible if the applicant believes that the MCR annunciators are supplied from the Safeguard 125V DC system and that the conditions meet the threshold values for an Alert classification.
C.
Correcl - MCR Annunciator power is provided by Non-safeguard 125 VDC. The non-safeguard electrical distribution panels are those that supply annunciators that have lost power. A Site Area Emergency is declared since a majority of the annunciators have been lost for more than 15 minutes, a plant transient is in progress (reactor scram) and Compensatory Non-alarming indications are unavailable (PMS).
D.
Incorrect but plausible if the applicant believes that the MCR annunciators are supplied from the Safeguard 125V DC system.
References:
LLOT0690, Rev. 12 Student Ref. required Yes ON-122, Rev.017 EP-AA-1008 EAL HOT Matrix Learning Objective:
Obj. 4 (LLOT0690)
Question source:
Modified Bank (Limerick)
Changed conditions 1) correct ans now Site Area Emergency and 2) requiring examinee to distinguish non-safeguard and safeguard DC buses.
Question History:
Limerick Cognitive level:
Memory/Fundamental knowledge:
Co m pre hens ive/Ana I ys is :
X 1 OCFR 55.43 (5) x
QUESTION 100 SENSITIVE INFORMATION - EXEMPT FROM PUBLIC DISCLOSURE Security-related question. Can not be released to general public.
A.
B.
C.
D.
SENSITIVE INFORMATION - EXEMPT FROM PUBLIC DISCLOSURE