ML14017A036
| ML14017A036 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/15/2013 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | |
| References | |
| Download: ML14017A036 (209) | |
Text
GGNS LOT 2013 NRC INITIAL LICENSED OPERATOR WRITTEN EXAMINATION RO EXAM ANSWER KEY 1 A 26 D 51 B 2 D 27 B 52 A 3 B 28 C 53 C 4 B 29 B 54 C 5 B 30 D 55 A 6 C 31 D 56 B 7 B 32 B 57 A 8 B 33 D 58 C 9 C 34 A 59 D 10 C 35 D 60 A 11 A 36 C 61 C 12 C 37 D 62 C 13 B 38 B 63 B 14 A 39 C 64 A 15 A 40 C 65 B 16 B 41 A 66 B 17 D 42 D 67 B 18 C 43 A 68 A 19 D 44 C 69 D 20 D 45 C 70 A 21 B 46 A 71 D 22 A 47 C 72 B 23 C 48 A 73 B 24 A 49 D 74 D 25 D 50 B 75 D
Examination Outline Cross-Reference Level RO 295001 Partial or Complete Loss of Forced Core Flow Circulation K/A # 295001 AK3. Knowledge of the reasons for the following responses Rating 3.4 as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : (CFR: 41.5 / 45.6)
Rev / Date 0 AK3.01 Reactor water level response Question 1 The plant is operating at rated power.
Reactor Recirc pump A trips.
Reactor water level rises to +45 Narrow range and is now slowly returning to normal.
Which of the following describes the reason for the level rise?
A. The response of the DFCS results in over feeding the vessel.
B. Jet pump reverse flow forces water back into the downcomer.
C. A rise in vessel temperature causes water in the downcomer to thermally expand.
D. A RPV pressure reduction causes level transmitters to indicate a higher level.
Answer: A Explanation:
Per FSAR Table 15.3-1 and Figure 15.3-1, when a single Recirc pump trips from rated power, the feed water flow will rise slightly for approximately 5 seconds before turning and reducing feed to the reactor. The feed water system is controlled digitally by a three element DFCS (Digital Feedwater Control System). The feed water system uses inputs from level, steam flow, and feed flow instruments to adjust feed pump speed.
Initially there is a corresponding rise in turbine steam flow according to Figure 15.3-1 which will cause the DFCS to raise feed flow in anticipation of a lower reactor water level; however, level will rise and peak after about 5 seconds (the same time feed flow turns)
A is correct
B is wrong, reverse flow will conservatively begin at about 5 seconds into the transient when the level transient has reached its peak level. Since the level rise is immediate, the subsequent reverse flow condition has no effect on the level rise.
C is wrong minimal temperature change. Downcomer temperature is affected mostly by feed water temperature. It is slightly sub-cooled and not operating in saturated conditions which would make pressure a factor in temperature. The temperature in the core is affected more by pressure and thermal power.
D is wrong, minimal pressure change and any pressure change is sensed by the reference and variable legs equally and therefore will not affect level indication.
Technical
References:
FSAR 15.3.1.2 FSAR 15.3.1.3.3.1 FSAR Table 15.3-1 FSAR Figure 15.3-1 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-MCD13 Obj 2 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b) 55.43(b)(2)
Examination Outline Cross-Reference Level RO 295003 Partial or Complete Loss of AC K/A # 295003 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics / reactor behavior /
Rating 3.7 and instrument interpretation.
Rev / Date 0 Question 2 The plant is operating at rated power when the following occurs:
- Feeder breaker to BOP Transformer 12B trips (breaker internal fault), no other breakers are affected.
The Loss of AC Power ONEP has been entered and Immediate Operator Actions taken.
Preceding any reactor scram that may occur, what are at least two additional ONEPs that will have to be entered?
A. Loss of Instrument Air Turbine and Generator Trips B. Automatic Isolations Reduction in Recirculation System Flow Rate C. Automatic Isolations Loss of Condenser Vacuum D. Loss of Condenser Vacuum Reduction in Recirculation System Flow Rate Answer: D Explanation:
With BOP 12B Xfmr loss and resultant loss of bus 11HD, one Circ Pump and one Recirc pump will trip, requiring entry into Loss of Condenser Vacuum and Reduction in Recirculation System Flow Rate ONEPs. The Turbine does not trip, the reactor does not scram and no isolations are received.
A is wrong due to no loss of instrument air or turbine trip B is wrong, due to no Isolation signal is received
C is wrong due to no Isolation signal is received D is correct due to explanation above.
Technical
References:
05-1-02-III-3 05-1-02-V-8 04-1-01-B33-1, Att III 04-1-01-N71-1, Att III E0001 References to be provided to applicants during exam: E0001 Learning Objective: GLP-OPS-ONEP, OBJ. 1 Question Source: Bank # GGNS-LORQT-06385 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b) 55.43(b)(2)
Examination Outline Cross-Reference Level RO 295004 Partial or Complete Loss of D.C. Power AK2. Knowledge of the interrelations between PARTIAL OR K/A # 295004 COMPLETE LOSS OF D.C. POWER and the following: (CFR: Rating 3.0 41.7 / 45.8)
Rev / Date 0 AK2.02 Batteries Question 3 Which of the following describe the amount of time the 1B3 batteries can provide DC power to all required emergency loads if there are no battery chargers in service?
A. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Answer: B Explanation:
Per FSAR 8.3.2.1.6.2 Batteries 1A3 and 1B3 have sufficient stored energy to operate connected essential loads continuously for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
A is wrong due to this time is the design for 1C3 batteries B is correct C is wrong this time is for complete restoration of 1C3 batteries by the chargers.
D is wrong, the A and B DC system battery chargers are designed to fully charge the batteries from minimum voltage to full within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Technical
References:
FSAR 8.3.2.1.6.2 Battery Capacity Considerations
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-L1100, OBJ. 2 Question Source: Bank # GGNS-OPS-01645 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b) 55.43(b)(2)
Examination Outline Cross-Reference Level RO 295005 Main Turbine Generator Trip AA1. Ability to operate and/or monitor the following as they K/A # 295005 apply to MAIN TURBINE GENERATOR TRIP : (CFR: 41.7 / Rating 2.7 45.6)
Rev / Date 0 AA1.04 Main generator controls Question 4 The plant was at rated conditions when the Main Turbine tripped.
The Turbine and Generator Trips ONEP directs actions to protect the voltage gradient capacitors in 500KV breakers.
Which of the following describes the time requirements to take action to ensure damage does not occur?
A. Immediately but within 15 minutes B. within 30 minutes C. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D. within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Answer: B Explanation:
Per 05-1-02-I-2, Turbine Generator Trip ONEP CAUTION prior to step 3.9. Do not allow a 500 kv breaker to remain open with voltage on it for > 30 minutes. Voltage gradient capacitors in the breaker will overheat. Step 3.9 or 3.10 should be performed within 30 minutes of T/G trip.
A is wrong does not meet with the CAUTION time requirements but plausible due to the time limit is the same as an EAL call.
B is correct C is wrong, does not meet with the CAUTION time requirements but plausible due to the time limit of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is used in Tech Specs and reportable events.
D is wrong, does not meet with the CAUTION time requirements but plausible due to the time limit of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is used in Tech Specs and reportable events
Technical
References:
05-1-02-I-2, Turbine Generator Trip ONEP CAUTION prior to step 3.9 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-ONEP, OBJ. 8.0 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b) 55.43(b)(2)
Examination Outline Cross-Reference Level RO 295006 SCRAM AK1. Knowledge of the operational implications of the K/A # 295006 following concepts as they apply to SCRAM : (CFR: 41.8 to Rating 3.4 41.10)
Rev / Date 0 AK1.02 Shutdown margin Question 5 A reactor scram has occurred.
Which of the following ensures that sufficient shutdown margin is maintained and that the reactor will remain subcritical under all conditions?
A. Two peripheral control rods are at position 48.
B. One center core control rod is at position 48.
C. 50% of the control rods are at position 04 with the rest fully inserted D. >50% of the control rods are at position 02 or beyond.
Answer: B Explanation:
Per Tech Specs definitions, SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that: (c) All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
After a scram the crew can determine the reactor shutdown with only one control rod not full in. If more that one is withdrawn then a calculation must be performed.
A is wrong does not meet the definition with > than one rod not inserted.
B is correct C is wrong, same as A D is wrong, same as A Technical
References:
Tech Spec Section 1.1 Definitions References to be provided to applicants during exam: None Learning Objective: GLP-OPS-TS001, OBJ. 4.13 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b) 55.43(b)(2)
Examination Outline Cross-Reference Level RO 295016 Control Room Abandonment AA2. Ability to determine and/or interpret the following as K/A # 295016 they apply to CONTROL ROOM ABANDONMENT : (CFR: Rating 3.8 41.10 / 43.5 / 45.13)
Rev / Date 0 AA2.05 Drywell pressure Question 6 The CRS has determined to evacuate the Main Control Room due to toxic fumes.
All actions were taken prior to leaving the Control Room.
No Security threat exists.
After arriving at the Remote Shutdown Panel you observe the following:
- RHR A and B pumps are operating.
- SSW A and B pumps are operating.
- RCIC system is operating with the Gland Seal Compressor is not running.
Which of the following has occurred?
A. BUV on both 15AA and 16AB B. Reactor water level lowered to -75 inches C. Drywell pressure rose to 1.42 psig D. Total loss of Offsite power Answer: C Explanation:
The Remote Shutdown Panels do not have indication for Drywell Pressure, however, by the indications given the student can determine drywell pressure. Will both RHR pumps and both SSW pumps running this would indicate an auto start of low level or high drywell pressure. Along with the RCIC Gland Seal Comp and CRD aux oil pumps with no indication, all of these indications are compatible with a LOCA signal.
RCIC is started upon Control Room abandonment. The RCIC gland seal compressor
is locked out on an LSS LOCA (1.39 psig in the DW or -150.3 RPV water level)
A is wrong RCIC Gland Seal comp. would be running and CRD aux oil pump would have Green indication.
B is wrong, RCIC Gland Seal comp. would be running and CRD aux oil pump would be running.
C is correct.
D is wrong, RCIC Gland Seal comp. would be running and CRD aux oil pump would have Green indication.
Technical
References:
04-1-01-R21-1 Table 1 GLP-OPS-R2100, Page 19 of 40 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-R2100, OBJ. 11, 14, 16 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b) 55.43(b)(2)
Examination Outline Cross-Reference Level RO 295018 Partial or Complete Loss of Component Cooling Water K/A # 295018 AA2.02 AA2.02 - Ability to determine and/or interpret the following as they Rating 3.1 apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Cooling water temperature Revision 0 Question 7 The plant is operating at rated conditions.
CCW Temperature Control Valve (P44-F501) has failed closed.
Which of the following will auto isolate due to the effects of this condition?
A. Reactor Recirculation Pumps B. Reactor Water Cleanup System C. Fuel Pool Cleaning and Cleanup System D. Control Rod Drive Pumps Answer: B Explanation:
With the Temp control valve failing close the CCW temp will rise causing RWCU temp to rise and cause the G33-F004 to auto close at 140 degrees Non regen outlet temp. Therefore, B is the correct answer.
Technical
References:
04-1-01-G33-1 Step 3.1 05-1-02-V-1, Loss of CCW ONEP Step 3.2.3 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-G3336, Objective 8.2 Question Source: Bank # GGNS-OPS-09630 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No
Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 295019 Partial or Complete Loss of Instrument Air AK2. Knowledge of the interrelations between PARTIAL OR K/A # 295019 COMPLETE LOSS OF INSTRUMENT AIR and the following: Rating 2.8 (CFR: 41.7 / 45.8)
Rev / Date 0 AK2.08 Plant Ventilation Question 8 A total loss of Instrument Air has occurred.
Which of the following indicates the available Drywell cooling?
A. A Drywell Coolers A Drywell Chillers B. B Drywell Coolers B Drywell Chillers C. A Drywell Coolers B Drywell Chillers D. B Drywell Coolers A Drywell Chillers Answer: B Explanation:
Per SOI 04-1-01-M51-1, P&L step 3.3, The outlet dampers for Drywell Cooler Fans 1A-6A fail closed on loss of power or air. 1B-6B fail open on loss of power or air.
Per SOI 04-1-01-P72-1, P&L step3.9, Temp Control valves on the 1A and 2A Drywell chillers fail closed on loss of power or air, 1B and 2B TCVs fail open on loss of air.
A, C, and D is wrong - see explanation above.
B is correct wrong - Bus 16AB is de-energized Technical
References:
04-1-01-M51-1, step 3.3
04-1-01-P72-1. step 3.9 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-ONEP, Objective GLP-OPS-M5100, OBJ. 9.3 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 295021 Loss of Shutdown Cooling AK3. Knowledge of the reasons for the following responses K/A # 295021 as they apply to LOSS OF SHUTDOWN COOLING : (CFR: 41.5 Rating 3.3
/ 45.6)
Rev / Date 0 AK3.01 Raising reactor water level Question 9 The plant is in Mode 4.
All RHR and ADHR heat removal capability is lost.
Which of the following describes the reason for raising reactor water level when RCS temperature cannot be maintained below 200°F when core cooling is lost?
A. Ensure Natural Circulation inside the core B. Ensure enough core coverage to verify adequate core cooling C. Provide a flowpath through the SRVs into the Suppression Pool D. Increase inlet subcooling Answer: C Explanation:
A is wrong - This is correct if there is no forced circulation (no recirc pumps running) because reactor coolant temperature indication is no longer accurate. Raising level in this case allows operators to gain accurate level indication. See step 3.4.2 of Inadequate Decay Heat Removal ONEP.
B is wrong - Adequate core cooling is assured as long as reactor water level is above TAF (-167).
C is correct - Step 3.4.3.f(3) of Inadequate Decay Heat Removal ONEP directs operators to raise reactor water level to between +101 and +125 to establish flow through open SRVs back to the Suppression Pool if adequate cooling cannot be reestablished prior to reaching 200F.
D is wrong - this is not a reason given in the Inadequate Decay Heat Removal ONEP; however, it is plausible because increasing sub-cooling suggests gaining margin to fuel damage.
Technical
References:
05-1-02-III-1, Inadequate Decay Heat Removal References to be provided to applicants during exam: None Learning Objective: GLP-OPS-ONEP, Objective 11.0 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 295023 Refueling Accidents AA1. Ability to operate and/or monitor the following as they K/A # 295023 apply to REFUELING ACCIDENTS : (CFR: 41.7 / 45.6) Rating 3.4 AA1.04 Radiation monitoring equipment Rev / Date 0 Question 10 Core Alterations are in progress when a dropped fuel bundle results in the following P601 alarms:
Where can control room operators check the status of the associated radiation monitors?
A. Only Trip Units on two Upper Control Room panels and a 4-channel recorder on a Main Control Room backpanel B. Only Trip Units on a Main Control Room backpanel and a 4-channel recorder on an Upper Control Room panel C. Trip Units on two Upper Control Room panels, Trip Units on two Main Control Room backpanels, and a 4-channel recorder on a Main Control Room backpanel D. Trip Units on one Upper Control Room panel, Trip Units on one Main Control Room backpanel, and 4-channel recorder on an Upper Control Room panel Answer: C Explanation:
See ARIs P601-18A-D5 (for Divs 1 and 4) and D6 (for Divs 2 and 3).
Per step 3.2 of the D5 ARI, operators can check an indicating Trip Unit for rad monitor D17K609A (Div 1) at Upper Control Room panel P669, and an indicating Trip Unit for rad monitor K609D (Div 4) at Main Control Room backpanel P672.
Per step 3.2 of the D6 ARI, operators can check an indicating Trip Unit for rad monitor D17K609B (Div 2) at Main Control Room backpanel P670, and an indicating Trip Unit for rad monitor K609C (Div 3) at Upper Control Room panel P671.
Technical
References:
ARI P601-18A-D5 and D6 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-D1721 Obj 13 Question Source: Bank # NRC Exam Bank X 15 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 295024 High Drywell Pressure EK1. Knowledge of the operational implications of the K/A # 295024 following concepts as they apply to HIGH DRYWELL Rating 3.9 PRESSURE : (CFR: 41.8 to 41.10)
Rev / Date 0 EK1.02 Containment building integrity: Mark-III Question 11 Which of the following is a potential consequence of exceeding an internal drywell pressure of 30 psig during a LOCA?
A. A loss of the pressure suppression function of Primary Containment B. Lowered efficiency of the charcoal filter train which may be used to vent the containment via 6 vent lines.
C. Failure of Primary Containment due to exceeding the Primary Containment maximum external-to-internal d/p limit D. Re-pressurization of the RPV from decay heat due to the inability of SRVs to remain open Answer: A Explanation:
Exceeding 30 psig drywell internal pressure can result in a loss of drywell integrity, allowing the LOCA blowdown to bypass the horizontal vents and hence the suppression pool (which otherwise provides the pressure suppression function); i.e.,
drywell discharges would be directly to the Primary Containment air atmosphere.
B is wrong. The 6 vent valves are no longer available with drywell pressure above 1.23 psig. Additionally, pressure is not considered a factor in plant procedures for the efficiency of the charcoal filter train.
C is wrong. The Primary Containment maximum external-to-internal d/p limit of 3 psid is anything but challenged when the drywell breaches due to high internal pressure.
The drywell breach having bypassed the suppression pool will raise Primary Containment internal pressure, rather than lower it. Plausible to the Applicant who has not grasped the relationships between drywell versus Primary Containment internal pressure.
D is wrong. The SRVs have no limitation on the maximum drywell pressure against which they are able to remain open. Plausible to the weak Applicant, generally.
Technical
References:
UFSAR, Section 6.2, Containment Systems References to be provided to applicants during exam: None Learning Objective: GLP-OPS-M4101 Obj 4.5 Question Source: Bank # NRC Exam Bank X 110 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam 2012 Yes Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 295025 High Reactor Pressure 2.4.50 Ability to verify system alarm setpoints and operate K/A # 295025 controls identified in the alarm response manual. (CFR: 41.10 / Rating 4.2 43.5 / 45.3)
Rev / Date 0 Question 12 The plant is operating at rated conditions.
A transient occurs causing the following alarms.
- RX PRESS HI P680-5A-B2
- RX PRESS HI DR LINE TEMP LO P680-3A-A2
- RX PRESS HI P680-2A-E4 (1) Which of the following describes the peak pressure that was achieved and (2) what are the immediate actions required?
A. (1) 1064.7 psig (2) Verify pressure control system is operating properly B. (1) 1095 psig (2) Verify pressure control system is operating properly C. (1) 1126 psig (2) Place Reactor mode switch to SHUTDOWN D. (1) 1325 psig (2) Notify NRC that a Safety limit has been exceeded Answer: C Explanation:
Setpoints are as follows
- RX PRESS HI P680-5A-B2 1064.7 psig
- RX PRESS HI DR LINE TEMP LO P680-3A-A2 1125 psig
- RX PRESS HI P680-2A-E4 1095 psig The peak pressure achieved was at least 1126 psig, the ARI states to Carry out action of Scram ONEP which is Place Mode Switch to SHUTDOWN.
A is wrong peak pressure reached 1126 and action is wrong B is wrong same as A C is correct D is wrong this is a safety limit and there is no pressure alarm to show this pressure.
Technical
References:
- RX PRESS HI P680-5A-B2
- RX PRESS HI DR LINE TEMP LO P680-3A-A2
- RX PRESS HI P680-2A-E4 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-C71 Obj 4.5, C11A Obj, ONEP Obj.
Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 295026 Suppression Pool High Water Temperature EK3. Knowledge of the reasons for the following responses K/A # 295026 as they apply to SUPPRESSION POOL HIGH WATER Rating 3.9 TEMPERATURE: (CFR: 41.5 / 45.6)
Rev / Date 0 EK3.05 Reactor SCRAM Question 13 A SRV has stuck open at rated power.
If operators are unable to close the SRV the reactor will be scrammed.
What is the purpose for this action?
A. Protect Suppression Pool level instrumentation.
B. Tech Spec required action to reduce heat input into the suppression pool C. Required action of EP-3 when the Tech Spec entry condition is met D. Minimize the rise in Containment Temperature.
Answer: B Explanation:
Per EP-3, a suppression pool temp of 110°F would require placing the mode switch to shutdown. Per Tech Specs this is done to reduce the rate of energy production and thus the heat input to the suppression pool.
A is wrong, suppression pool level instrumentation is not the reason.
B is correct.
C is wrong, 95°F is only a EP 3 entry not a plant shutdown.
D is wrong, suppression pool level instrumentation is not the reason.
Technical
References:
02-S-01-40 Att VI page 7 of 34 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-G3336, Objective 8.2 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 295027 High Containment Temperature (Mark III Containment Only)
K/A # 295027 EK1. Knowledge of the operational implications of the Rating 3.8 following concepts as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY) : (CFR:
Rev / Date 0 41.8 to 41.10)
EK1.03 Containment integrity: Mark-III Question 14 Per 02-S-01-40, EP Technical Bases, which of the following is a basis for Emergency Depressurizing the RPV before CTMT temperature reaches 185°F?
A. Ensure the RPV is depressurized before CTMT temperature gets high enough to damage the CTMT.
B. Ensure the RPV is depressurized while the SRVs are still functional.
C. Limit the release of energy into CTMT in order to keep from exceeding the CTMT temperature LCO limit.
D. Limit the release of energy into CTMT in order to preserve the pressure suppression capacity of the suppression pool.
Answer: A Explanation:
See EP Tech Bases, Attachment VI, page 17 of 34, bottom-most paragraph.
B is wrong. It suggests the basis for ED before exceeding a DW temperature of 330°F (see Attachment VI, page 12 of 34).
C is wrong. The Tech Spec LCO limit for CTMT temperature is 95°F (see Attachment VI, page 13 of 34).
D is wrong. There is no need to preserve the pressure suppression capacity of the suppression pool post ED (See HCTL curve/bases). Additionally, the release of energy to the containment has nothing to do with rather or not the pressure suppression capacity is maintained because this is a function of reactor pressure, suppression pool level, and suppression pool temperature (See HCTL curve).
Technical
References:
02-S-01-40, EP Technical Bases References to be provided to applicants during exam: None Learning Objective: GLP-OPS-EP3, Objective 7 Question Source: Bank # 2010 NRC bank X
- 425 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 295028 High Drywell Temperature EK2. Knowledge of the interrelations between HIGH K/A # 295028 DRYWELL TEMPERATURE and the following: (CFR: 41.7 / Rating 3.2 45.8)
Rev / Date 0 EK2.02 Components internal to the drywell Question 15 EP-3 (Containment Control) directs operators to emergency depressurize if drywell temperature cannot be maintained below 330°F.
This specific temperature is the drywell design temperature; however, a temperature of 340°F would challenge the ability of certain components within the drywell to operate as designed.
Per the EP Technical Bases, what are those components?
A. SRVs B. MSIVs C. Drywell Purge Supply/Initial Vacuum Relief Valves D. Post-LOCA Vacuum Valves Answer: A Explanation:
See EP Tech Bases, Attachment VI, page 12 of 34, for EP-3, Step DWT-5.
All Distracters are plausible because each of these components is either in, or interface with the drywell.
Technical
References:
02-S-01-40, EP Technical Bases References to be provided to applicants during exam: None Learning Objective: GLP-OPS-EP3 Obj 7
Question Source: Bank # 2012 NRC Exam X NRC Exam Bank 100 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam Yes Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 295030 Low Suppression Pool Water Level EA2. Ability to determine and/or interpret the following as K/A # 295030 they apply to LOW SUPPRESSION POOL WATER LEVEL : Rating 3.7 (CFR: 41.10 / 43.5 / 45.13)
Rev / Date 0 EA2.03 Reactor pressure Question 16 Step SPL-9 of the Suppression Pool Level leg of EP-3 (Containment Control) directs operators to perform an Emergency Depressurization (E.D.) if suppression pool level cannot be maintained above 14.5 feet.
(1) What is the operational concern associated with this particular step?
(2) What action is required if suppression pool level is 12.0 feet before this step is actually implemented?
A. (1) Suppression Pool coverage of the horizontal vents.
(2) Emergency Depressurization using SRVs is prohibited; operators must use an alternate method to lower reactor pressure.
B. (1) Suppression Pool coverage of the horizontal vents.
(2) Emergency Depressurize using 8 ADS/SRVs to lower reactor pressure.
C. (1) Suppression Pool level below the SRV Tail Pipe Level Limit.
(2) Emergency Depressurization using SRVs is prohibited; operators must use an alternate method to lower reactor pressure.
D. (1) Suppression Pool level below the SRV Tail Pipe Level Limit.
(2) Emergency Depressurize using 8 ADS/SRVs to lower reactor pressure.
Answer: B
Explanation:
EP bases for SPL-9 states that the concern for 14.5 is horizontal vent coverage. Per EP-2, ED is allowed as long as Suppression Pool level remains above 10.5. Below 10.5 alternate depressurization means must be used per table 3 of EP-2. The STPLL (SRV Tail Pipe Level Limit) is based on potential damage to the SRV tailpipes if an SRV is opened with SP level too high.
Technical
References:
02-S-01-40 EP Bases EP-2 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-EP3, Objective 7 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 295031 Reactor Low Water Level 2.1.31 Ability to locate control room switches, controls, and K/A # 295031 indications, and to determine that they correctly reflect the Rating 4.6 desired plant lineup. (CFR: 41.10 / 45.12)
Rev / Date 0 Question 17 The reactor was operating at rated power when the reactor scrammed due to instrument failures.
Which of the following indications confirm that reactor water level is below +11.4 inches and above -41.6 inches?
A. On P680 panel - Reactor Recirc pumps A and B tripped to OFF On P601 panel - ADS Confirmatory Level annunciator.
B. On P680 panel - Reactor Recirc pumps A and B tripped to OFF On P601 panel - ADS Timers have initiated.
C. On P680 panel - Reactor Recirc pumps A and B running in slow speed.
On P601 panel - ADS Timers have initiated.
D. On P680 panel - Reactor Recirc pumps A and B running in slow speed.
On P601 panel - ADS Confirmatory Level annunciator.
Answer: D Explanation:
On P601 panel the ADS confirmatory level is <+11.4 reactor water level. Recirc pumps will also shift to slow speed at this level on cavitation interlocks. At -41.6, the Recirc pumps will trip off due to ATWS/ARI initiation. This means that in order to confirm level between11.4 and -41.6 Recirc pumps must be in slow speed and the ADS Confirmatory Level annunciator must be in alarm.
ADS timers will initiate when only when an ECCS/LSS LOCA signal is received (1.39 psig in DW or -150.3 reactor level 1).
Technical
References:
04-1-02-1H13-P680-2A-C15
04-1-02-1H13-P601-18A-A2 04-1-02-1H13-P601-18A-C2 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-E2202, Objective 15.0 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 295038 High Off-Site Release Rate EK3. Knowledge of the reasons for the following responses K/A # 295038 as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.5 Rating 3.9
/ 45.6)
Rev / Date 0 EK3.02 System isolations Question 18 Which of the following is a reason for the automatic isolation feature of normal Secondary Containment Ventilation?
A. Ensure Auxiliary Building d/p is maintained negative.
B. Ensure Auxiliary Building d/p is maintained positive.
C. Prevent untreated airborne radioactivity from being released to the outside.
D. Prevent the release of airborne radioactivity beyond ODCM limits from within the Turbine Building.
Answer: C Explanation:
Normal Secondary Containment ventilation is actually the combined Aux Bldg Ventilation (T41) and Fuel Handling Area Ventilation (T42) systems. These auto-isolate (and SGTS auto-initiates) on the associated vent exhaust high-high radiation conditions. In doing so, this isolation feature prevents untreated airborne radioactivity from being released to the outside environment. This implies that although T41/T42 releases are untreated, SGTS is treated before its release.
Answers A & B are wrong mainly because they do not speak directly to the fundamental reason for the automatic isolation described in the stem. They provide sufficient plausibility based on the fact that both the normal ventilation systems and SGTS also serve to maintain a negative building d/p.
Answer D is wrong because the Turbine Building is not part of the Secondary CTMT boundary.
Technical
References:
GLP-OPS-T4100 GLP-OPS-T4200 GLP-OPS-T4800 EP-2; EP-4 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-T4200, OBJ. 4A Question Source: Bank # NRC Exam Bank X 413 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 600000 Plant Fire On Site AK3.04 - Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE: Actions contained in the K/A # 600000 abnormal procedure for plant fire on site Rating 2.8 Revision 0 Question 19 Heavy smoke and flames is being reported coming from load center 15B51.
Which of the following describes the required response of the Control Room Crew?
A. Dispatch the 5 Man Fire Brigade, only B. Dispatch the 3 Man Fire Brigade with Claiborne County Fire Department Supplement, only C. Dispatch the Claiborne County Fire Department, only D. Dispatch the 5 Man Fire Brigade with Claiborne County Fire Department for backup support, only Answer: D Explanation:
Per 10-S-03-2, Response to fires, 6.2.3.f, The Operations 5 man Fire Brigade is the Primary Responder for all fires in: U1 and U2 Power Block, Diesel Generator Bays, SSW A and B, Independent Fuel Storage, ESF Transformers (11, 21, 12), Fire Water Pump House, and all other areas of the plant containing structures, systems or components important to safety (manholes, tanks, etc.). Claiborne County will provide backup support. .
A is wrong. Procedure states Claiborne County will provide backup support B is wrong. See explanation above C is wrong. See explanation above D is correct. See explanation above Technical
References:
10-S-03-2, 6.2.3.f CR-GGN- 2013-3589
Standing Order 13-0008 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-PROC, Objective 58.2, 58.3 Question Source: Bank # 2011 Audit Exam (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 700000 Generator Voltage and Electric Grid Disturbances AK2. Knowledge of the interrelations between GENERATOR K/A # 700000 VOLTAGE AND ELECTRIC GRID DISTURBANCES and the Rating 3.0 following: (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)
Rev / Date 0 AK2.03 Sensors, detectors, indicators Question 20 The plant is operating at rated power.
Grid voltage lowers from 500KV to 490KV.
(1) P680 Gen Volt Meter (N41-R605) will (2) P680 Gen MVAR Meter (N41-JR-R608) will A. (1) read the same (2) read lower B. (1) read the same (2) read higher C. (1) read lower (2) read lower D. (1) read lower (2) read higher Answer: D Explanation:
According to the SOI, you raise VAR by raising the VR output (that is raising generator no-load voltage relative to grid voltage). When tied to grid, generator voltage and frequency are set by the grid. This is the reason there is no concern in the SOI for generator voltage output when tied to the grid (it is only concerned with generator VARs). Thus, when grid voltage lowers this is equivalent to raising on the TVR regulator and therefore generator VARs increase. Since grid voltage lowered, the generator output voltage will also lower in kind.
This is also validated using the simulator and generic fundamentals for operating AC sources in parallel.
Technical
References:
04-1-01-N40-1 section 4.4 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-N4151, Objective 8, 13.1, 15, 16 Question Source: Bank # NRC Exam Bank 2012 Audit Exam 19 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 295009 Low Reactor Water Level AK1. Knowledge of the operational implications of the K/A # 295009 following concepts as they apply to LOW REACTOR WATER Rating 3.3 LEVEL : (CFR: 41.8 to 41.10)
Rev / Date 0 AK1.05 Natural circulation Question 21 The plant is scrammed due to a trip of both reactor recirculation pumps.
Reactor water level is being maintained at 85" on Shutdown Range.
Which of the following will be the effect on natural circulation flow rate if reactor water level is lowered below the steam separators?
A. Flow rate will decrease initially and then increase to a new thermal equilibrium value slightly less than the original flow rate.
B. Flow rate will significantly decrease due to the loss of communication between the core and the annulus.
C. Flow rate will increase to a new stable value as the temperature of the water in the core increases to a new stable value.
D. Flow rate will not be significantly affected because the thermal driving head is primarily dependent on the differential temperature between the core and the annulus.
Answer: B Explanation:
Per FSAR 4.4.3.6, the natural circulation achieved a lower vessel levels "are minimums, it should be noted that the flow rates would be the lowest flow achieved."
Therefore, as water level is lowered the natural circulation flow rate will also lower and not return to the original value. This is especially true once reactor water level is below the steam separators at a level of 82". This is also the reason for actions in the SCRAM ONEP to raise reactor water level above 82" to allow for maximum natural circulation.
This question is a modification of NRC Generic Fundamentals question B891.
Technical
References:
FSAR 4.4.3.6 05-1-02-I-1 Reactor Scram ONEP References to be provided to applicants during exam: None Learning Objective: GLP-OPS-MCD01 Obj 3.2 Question Source: Bank # GGNS-OPS-04003 X (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(5) 55.43(b)
Examination Outline Cross-Reference Level RO 295012 High Drywell Temperature 2.2.44 Ability to interpret control room indications to verify K/A # 295012 the status and l operation of a system, and understand how Rating 4.2 operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12)
Rev / Date 0 Question 22 A LOCA is in progress Drywell Temperature is 150°F and rising.
CRS directs to maximize Drywell cooling per EPs.
The Operator notices the following indications:
- Drywell Chilled water pump A running, pump B has no indication.
- ALL Drywell Coolers have no indication.
Which of the following indicates the cause for the given indication?
A. Drywell pressure 1.39 psig B. Reactor water level <-41.6 inches C. Loss of power on 16AB D. Loss of Instrument Air Answer: A Explanation:
To achieve the given indication a LOCA signal of >1.39psig or < -150.3 has to occur.
the B Drywell Chilled Water system is powered from 16AB but during a LOCA sequence this load is locked out and not powered. The A skid is not running due to no chilled water flow from the CTMT isolation valves being closed on hi drywell pressure. The Drywell coolers are powered from 15B42 and 16B42 MCC which are shed and not sequenced back, but, can be manually re-energized. Therefore the only signal that could all of the indications is a LOCA signal of -150.3 RPV level or
>1.39 psig Drywell pressure.
A is correct.
B is wrong. Level 2 would cause the CTMT isolation valves to close and cause the indications on the A skid chillers but not the loss of power to the other components.
C is wrong. This would cause the indication on the B skid with no power, but, would not cause a loss of chilled water flow, the A skid will still be running normal.
D is wrong. This would cause the chilled water cooling water control valve to fail open on one skid and close on the other but would not affect the indication of any component.
Technical
References:
04-1-01-P72-1, step 3.9 05-1-02-III-5, page 12 GLP-OPS-M5100, pages 14 & 15 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-M51, Objective 9.2 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 295013 High Suppression Pool Temperature AA1. Ability to operate and/or monitor the following as they K/A # 295013 apply to HIGH SUPPRESSION POOL TEMPERATURE : (CFR: Rating 3.9 41.7 / 45.6)
Rev / Date 0 AA1.02 Systems that add heat to the suppression pool.
Question 23 The plant is at rated conditions.
RCIC is operating CST to CST for surveillance.
Which of the following indicates the maximum Suppression pool temperature for continued RCIC operation?
A. 90°F B. 95°F C. 105°F D. 110°F Answer: C Explanation:
Per Tech Specs 3.6.2.1 Suppression pool average temperature shall be
<95°F at >1% power and no testing
<105°F at >1% power and testing that adds heat to the suppression pool is being performed.
<110°F at <1% power A is wrong, This is the alarm setpoint that tells the operator to start suppression pool cooling.
B is wrong, This is an EP entry condition that requires all supp pool cooling to be started C is correct.
D is wrong. This is a required Reactor Scram
Technical
References:
06-OP-1M24-V-0001 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-TS001, Objective 39 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 295015 Incomplete SCRAM AK2. Knowledge of the interrelations between INCOMPLETE K/A # 295015 SCRAM and the following: Rod control and information system: Rating 3.2 Plant-Specific Rev / Date Question 24 The plant is operating at rated power.
A Turbine Trip occurs and all control rods did not insert due to hydraulic block.
Reactor power is 12%.
The crew is maintaining level -70 to -130 Wide Range in the Level Control leg of EP-2A.
ECCS systems have been terminated and prevented.
Power and Pressure legs of EP-2A have not been executed.
Which of the following describes actions necessary to attempt further control rod motion using RC&IS?
A. Restart CRD Perform EP Attachments 18, 19 and 20 B. Restart CRD Perform EP Attachment 18 only C. Perform EP Attachments 18, 19 and 20 only D. Perform EP Attachment 20 only Answer: A Explanation:
With completion of the level control leg the maximum water level would be -70 therefore an Instrument Air to Containment and Drywell isolation would have occurred. ATWS/ARI is manually initiated and a scram condition is present.
Restoring Instrument Air and Attachments 18 and 19 are required to be able to scram rods.
B is incorrect because attachment 19 is also required
C is incorrect because the Aux building must be restored D is incorrect because attachments 18 and 19 are required to allow movement by RCIS.
Technical
References:
EOP flow charts References to be provided to applicants during exam: None Learning Objective: GLP-OPS-EP07 OBJ. 1 & 2 Question Source: Bank # Audit Initial Exam X (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41 55.43
Examination Outline Cross-Reference Level RO 295032 High Secondary Containment Area Temperature EK3. Knowledge of the reasons for the following responses K/A # 295032 as they apply to HIGH SECONDARY CONTAINMENT AREA Rating 3.6 TEMPERATURE : (CFR: 41.5 / 45.6)
Rev / Date 0 EK3.02 Reactor SCRAM Question 25 The reactor is operating at rated power.
- There is a steam leak in the RCIC room.
- E51-F063, RCIC steam supply inboard isolation valve, failed to isolate.
- E51-F064, RCIC steam supply outboard isolation valve, isolated as required.
- 5 minutes later RCIC room temperature has now stabilized near the max safe value and is 2°F below its peak temperature.
EP-4 steps 8 and 9 direct entry into EP-2 before any area temperature reaches its max safe value.
(1) Per 02-S-01-40, EP Technical Bases, what is the reason for entering EP-2?
(2) Is a reactor scram required for the conditions above?
A. (1) This condition is indicative of a wide-spread problem and threatens safe operation of the plant.
(2) A reactor scram is required because the steam leak may not be isolated.
B. (1) This condition is indicative of a wide-spread problem and threatens safe operation of the plant.
(2) A reactor scram is not required because the steam leak was isolated.
C. (1) EP-2 requires a reactor scram which will reduce the heat input into the RCIC room.
(2) A reactor scram is required because the steam leak may not be isolated.
D. (1) EP-2 requires a reactor scram which will reduce the heat input into the RCIC room.
(2) A reactor scram is not required because the steam leak was isolated.
Answer: D
Explanation:
EP-4 bases states the bases for entering EP-2 in steps 8 and 9 is to scram the reactor to reduce heat input, radioactivity release, and rate of the leak into the affected room. The bases for a normal reactor shutdown in step 7 is because 2 max safe values exist and is indicative of a wide-spread problem and threatens safe operation of the plant.
Step 6 of EP-4 is an override such that if the system cannot be isolated from the RPV a scram is directed before any max safe value is exceeded.
For the conditions in the stem, it is apparent that the steam leak was isolated as indicated by RCIC room temperature stabilizing and lowering. In this case, a normal reactor shutdown is directed only if 2 or more max safe values are exceeded.
The applicant must evaluate plant conditions and based on the EP-4 bases recall that a scram is not required.
Technical
References:
02-S-01-40, EP Technical Bases References to be provided to applicants during exam: None Learning Objective: GLP-OPS-EP4 obj. 7 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43
Examination Outline Cross-Reference Level RO 295034 Secondary Containment Ventilation High Radiation EK2. Knowledge of the interrelations between SECONDARY K/A # 295034 CONTAINMENT VENTILATION HIGH RADIATION and the Rating 3.8 following: (CFR: 41.7 / 45.8)
Rev / Date 0 EK2.02 Area radiation monitoring system Question 26 The plant is in Mode 5.
A fuel handling accident occurs on the Fuel Handling Platform causing Standby Gas Treatment to automatically initiate.
Which of the following describes the Area Radiation monitors that should be in alarm?
A. Dryer/Separator Storage Area B. Fuel Pool Cooling and Cleanup Area C. Containment Ventilation Filter Train Area D. Spent Fuel Handling Area Answer: D Explanation:
First the student must recognize that a accident occurred on the Fuel handling platform which is in the Aux building. This is also verified by an auto start of SBGT, therefore the ARMs that will be in alarm will be the Spent Fuel Handling Area. The setpoint for the ARMs in this area is 2.5 MR/HR, the auto start of SBGT is 3.5 mr/hr Sweep or 35 mr/hr exhaust.
Answer A is inside the containment and would not be in alarm Answer B is on elevation 185 and should not be in alarm at this time Answer C is also inside containment and would not be in alarm.
D is correct
Technical
References:
04-1-02-1H13-P844-1A-A4 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-EP04 OBJ. 1 & 2 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41 55.43
Examination Outline Cross-Reference Level RO 295036 Secondary Containment High Sump/Area Water Level EK1. Knowledge of the operational implications of the K/A # 295036 following concepts as they apply to SECONDARY Rating 2.9 CONTAINMENT HIGH SUMP/AREA WATER LEVEL : (CFR:
41.8 to 41.10) Rev / Date 0 EK1.01 Radiation releases Question 27 A radioactivity release from an Aux Building Equipment Drain Sump is detected before being processed in the Rad Waste Building by __________(1)___________.
A significant release would result in ______________(2)_______________.
A. (1) radiation detectors in the Fuel Handling Ventilation System (2) automatic isolation of all inputs to the associated Aux Building Equipment Drain Sump B. (1) radiation detectors in the Fuel Handling Ventilation System (2) a SGTS system initiation C. (1) the Process Liquid Radiation Monitoring System (2) a SGTS system initiation D. (1) the Process Liquid Radiation Monitoring System (2) automatic isolation of all inputs to the associated Aux Building Equipment Drain Sump Answer: B Explanation:
Using the Floor & Equipment Drain P&ID, M-1094A, a vent pipe is routed to the Fuel Handling Ventilation System which will alarm and initiate the SGTS (ARI P870-2A-A3) in the event of a radioactivity release.
The inputs fed to the Aux Blding Equipment Drain Sumps are gravity fed drains with no automatic isolation capability.
The Process Liquid Radiation Monitoring System monitors PSW, SSW, CCW, and Radwaste discharges.
Technical
References:
04-1-02-1H13-P870-2A-A3
M-1094A References to be provided to applicants during exam: None Learning Objective: GLP-OPS-P4500 Obj 3 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41b(8) 55.43
Examination Outline Cross-Reference Level RO 203000 RHR/LPCI: Injection Mode (Plant Specific)
K6. Knowledge of the effect that a loss or malfunction of the K/A # 203000 following will have on the RHR/LPCI: INJECTION MODE Rating 3.0 (PLANT SPECIFIC) : (CFR: 41.7 / 45.7)
Rev / Date 0 K6.10 Component cooling water systems Question 28 A LOCA is in progress.
Only RHR Pump C is available.
It is operating in LPCI mode to maintain reactor water level when SSW Pump B trips and cannot be re-started.
Consider the following:
- 1. LPCI loop C injection water temperature
- 2. RHR Pump C seal temperature
A. 1 and 2, only B. 1 and 3, only C. 2 and 3, only D. 1, 2, and 3 Answer: C Explanation:
See SSW P&IDs M-1061B and D. Unlike RHR loops A and B, (which have heat exchangers cooled by the respective SSW subsystem) RHR loop C has no heat exchanger. Therefore LPCI loop C injection water temperature is unaffected by a loss of SSW B. However, both the RHR Pump C seal cooler and the RHR Pump C Room cooler is supplied by SSW B.
Technical
References:
P&IDs M-1061B and D, SSW System References to be provided to applicants during exam: None Learning Objective: GLP-OPS-E1200 Obj 13.3 Question Source: Bank # NRC Exam Bank X 124 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam Yes Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41 55.43
Examination Outline Cross-Reference Level RO 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)
K/A # 205000 A2. Ability to (a) predict the impacts of the following on the Rating 2.9 SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) ; and (b) based on those predictions, use procedures Rev / Date 0 to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
A2.10 Valve Operation: Plant-Specific Question 29 The plant is in Mode 4.
- RHR 'A' is in Shutdown Cooling Mode.
- RHR 'B' E12-F004B Suppression Pool Suction valve is closed with it's breaker open.
- 14BE1 is out of service due to an over current trip on the feeder breaker.
Consider the E12-F428A Pressure Lock Isolation for E12-F024A Test Return to the Suppression Pool.
(1) What is the operational implication having E12-F428A open for this plant condition?
(2) What action is required if E12-F428A is discovered to be stuck in the open position?
A. (1) A potential loss of Reactor Coolant to the Suppression Pool.
(2) Place ADHR in RPV Cooling Mode.
B. (1) A potential loss of Reactor Coolant to the Suppression Pool.
(2) Place RHR 'B' in Shutdown Cooling Mode.
C. (1) RHR Heat Exchanger Inlet temperature indication will not represent actual core temperature.
(2) Place ADHR in RPV Cooling Mode.
D. (1) RHR Heat Exchanger Inlet temperature indication will not represent actual core temperature.
(2) Place RHR 'B' in Shutdown Cooling Mode.
Answer: B Explanation:
04-1-01-E12-2 P&L 3.8.12 states that when in shutdown cooling mode with RHR, E12-F428A should be closed to prevent possible loss of reactor coolant to suppression pool due to disc flexing of E12-F024A.
Both ADHR pumps are powered from 14BE1. ADHR cannot, therefore, be used for decay heat removal. ADHR is plausible for the applicant who cannot recall the power supply for ADHR pumps.
RHR 'B' shutdown cooling is not affected by the status of E12-F004B; however, this offers additional plausibility for the use of ADHR for the applicant who is unsure of how the interlocks associated with RHR suction valves work.
The Inadequate Decay Heat ONEP refers to situations that render indicated reactor core temperature inaccurate. This distracter is plausible based on the applicant not remembering what those situations are; however, for this situation the temperature indications are still representative of actual core temperature.
Technical
References:
04-1-01-E12-2 P&L 3.8.12 Shutdown Cooling and ADHR Operation References to be provided to applicants during exam: None Learning Objective: GLP-OPS-E1200 Obj 14 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43
Examination Outline Cross-Reference Level RO 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)
K/A # 205000 K2. Knowledge of electrical power supplies to the following: Rating 3.1 (CFR: 41.7)
Rev / Date 0 K2.01 Pump motors .
Question 30 The plant is in Mode 4 with RHR A in shutdown cooling.
Division 1 Diesel Generator is tagged out due to an air leak.
ST-21 transformer trips on sudden pressure.
Which of the following describes the affect on shutdown cooling?
A. A RHR pump trips, but, can be restarted after 15AA is re-energized.
B. A RHR pump trips, B RHR must be started to maintain cooling.
C. A RHR pump continues to operate, but, B RHR must be started to maintain cooling.
D. A RHR pump continues to operate, no other action required.
Answer: D Explanation:
With an ST-21 trip power is lost on 16AB bus. This will not affect the RHR A system.
For the above reasons, only choice D is correct.
A is wrong, RHR A pump will not trip 15AA does not de-energize.
B is wrong, RHR A pump will not trip C is wrong, RHR A will continue to operate normally. There is no reason to start RHR B Technical
References:
04-1-01-R21-15, step 3.3 04-1-01-R21-16, step 3.3 E0001 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-E1200, Objective 7.1 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 209001 Low Pressure Core Spray System K3. Knowledge of the effect that a loss or malfunction of the K/A # 209001 LOW PRESSURE CORE SPRAY SYSTEM will have on Rating 2.9 following: (CFR: 41.7 / 45.4)
Rev / Date 0 K3.03 Emergency generators Question 31 A loss of offsite power with a LOCA has occurred.
Div 1 Diesel Generator is carrying 15AA.
A loss of which of the following components will have the largest effect on the electrical loading of the Diesel Generator?
A. RHR pump A B. SSW pump A C. CRD pump A D. LPCS pump Answer: D Explanation:
Component listing shows LPCS pump to be Approx. 1200 kw with RHR A at 803 and SSW A at 997. CRD is lower than any listed load.
Also 05-1-01-I-4 Loss of AC power ONEP, states that during a Station Blackout and cross tie Div 3 D/G with 15 or 16 bus CAUTION prior to step 3.2.10 j states Do not attempt to start LPCS with Div 3 diesel generator cross connected to Div 1 bus. This is due to the heavy load and the minimum amount that can be used.
Technical
References:
05-1-01-I-4 3.2.10 j
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-ONEP, Objective 9-12 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 209001 Low Pressure Core Spray System A1 - Ability to predict and/or monitor changes in parameters K/A # 209001 associated with operating the LOW PRESSURE CORE SPRAY Rating 3.4 SYSTEM controls including:
Rev / Date 0 A1.01Core spray flow Question 32 The plant has experienced an ATWS.
All Low pressure ECCS systems have been overridden.
An emergency Depressurization is in progress with 8 ADS/SRVs open.
The CRS has called for injection with Feedwater at 2 mlbm/hr.
Bus 15AA has a loss of power but is restored by Div 1 Diesel Generator.
Which of the following describes the flow indication for the LPCS system?
A. Indicates 0 gpm B. Indicates > 7000 to the RPV C. Indicates < half of rated flow D. Indicates minimum flow only.
Answer: B Explanation:
LPCS overrides will be lost when 15AA is deenegized. As soon as the bus is restored the LPCS pump will auto start and inject. With the CRS just calling for feedwater flow injection at 2 mil reactor pressure has just went below 206 psig per EP-2A.
Tech Specs 3.5.1 SR 3.5.1.4 states that LPCS is required to deliver >7115 gpm at 290 psid. Therefore LPCS should be indicating >7000 gpm A is wrong because LPCS will restart and inject B is correct
C is wrong, it should indicate full flow due to the ability of the pump to deliver 7115 at 290 psid.
D is incorrect due to is will inject at full flow and not be on min flow Technical
References:
Tech Specs 3.5.1 GLP-OPS-E2100 Page 16 of 39 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-E21, Objective 9.7 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 209002 High Pressure Core Spray System (HPCS) 2.4.18 Knowledge of the specific bases for EOPs. (CFR: 41.10 K/A # 209002
/ 43.1 / 45.13) Rating 3.3 Rev / Date 0 Question 33 What is the bases for overriding HPCS injection when entering EP-2A?
A. HPCS could dilute the concentration of Boron inside the core shroud and can only be used after Hot Boron Weight has been injected by SLC.
B. HPCS is used when the preferred systems cannot maintain RPV water level before an Emergency Depressurization.
C. HPCS could dilute the concentration of Boron inside the core shroud and can only be used after Cold Boron Weight has been injected by SLC.
D. HPCS is used when the preferred systems cannot maintain RPV water level after an Emergency Depressurization.
Answer: D Explanation:
Per EP Bases, HPCS can only be used if the preferred systems cannot maintain RPV water level within the desired band and only after emergency RPV depressurization has been performed. This makes D correct and B wrong.
The bases discussion for EP-2A states that injection systems are selected and operated so as to minimize the risk of Boron dilution and cold water injection; however, no mention of weighting for Hot/Cold Shutdown Boron Weight is made. This makes answer choices A &
C plausible but wrong.
Technical
References:
02-S-01-40 EP Bases Attachment V References to be provided to applicants during exam: None
Learning Objective: GLP-OPS-EP02A Obj 7 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)
Examination Outline Cross-Reference Level RO 211000 Standby Liquid Control System K4. Knowledge of STANDBY LIQUID CONTROL SYSTEM K/A # 211000 design feature(s) and/or interlocks which provide for the following: Rating 3.8 (CFR: 41.7)
Rev / Date 0 K4.04 Indication of fault in explosive valve firing circuits Question 34 The plant has experienced an ATWS.
The CRS directs you to initiate Standby Liquid Control system.
Which of the following describes the indications that the SLC Squib Valves have fired?
A. White SQUIB VALVE READY light OFF Annunciator SLC SYS A and B OOSVC in alarm Amber status light SQUIB A and B LOSCNT OR PWRLOSS is ON B. White SQUIB VALVE READY light ON Annunciator SLC SYS A and B OOSVC is clear Amber status light SQUIB A and B LOSCNT OR PWRLOSS is ON C. White SQUIB VALVE READY light OFF Annunciator SLC SYS A and B OOSVC in alarm Amber status light SQUIB A and B LOSCNT OR PWRLOSS is OFF D. White SQUIB VALVE READY light ON Annunciator SLC SYS A and B OOSVC is clear Amber status light SQUIB A and B LOSCNT OR PWRLOSS is OFF Answer: A Explanation:
04-1-01-C41-1 Attachment VI, Verification of SLC Injection, step 1.Verfiy system initiation by Observing the following; a. F004A and F004B SQUIB VALVES FIRED:
- White SQUIB VALVE READY light OFF
- Annunciator SLC SYS A and B OOSVC
- Amber status light SQUIB A and B LOSCNT OR PWRLOSS is ON A is correct B is wrong, Ready light would be OFF and Annunciator would be on C is wrong, Status light would be ON D is wrong, This shows system in normal standby not squibs fired.
Technical
References:
04-1-01-C41-1, Attachment VI References to be provided to applicants during exam: None Learning Objective: GLP-OPS-C41, Objective 9.3, 10.4 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 212000 Reactor Protection System K1. Knowledge of the physical connections and/or cause-effect K/A # 212000 K1 relationships between REACTOR PROTECTION SYSTEM and the following: A.C. electrical distribution Rating 3.4 Rev / Date 0 Question 35 The plant is operating at rated power.
ST-11 transformer trips and a lockout occurs.
All 4160kv and 480v electrical buses have been restored per Loss of AC Power ONEP.
No other operator actions have been taken.
What is the status of the A RPS Bus Power?
A. RPS A Energized from normal source, the alternate source available B. RPS A De-energized, the alternate source available, only C. RPS A Energized from alternate source, the normal source available D. RPS A De-energized, neither source available Answer: D Explanation:
With an ST-11 lockout comes a loss BOP buses 12HE, 13AD and 15AA. RPS A normal source (M/G set) is powered from 13AD and Alternate source is from 15AA. Even though the Div 1 DG has re-powered Bus 15AA, and with it MCC 15B42 (i.e., Div 1 LSS has re-sequenced the MCC back on), that Alternate Feed light is nonetheless OFF. The reason is that the Alternate Source RPS A EPA Breakers both tripped open on Undervoltage/Underfrequency on the initial power loss. Thus, the power is still not available to the A RPS Bus until an operator is dispatched to manually reset the EPA relays in the MG Set Room. Therefore, RPS A is de-energized and neither source is available from the control room.
For the above reasons, only choice D is correct.
A, B, C are wrong for reasons described above. They are plausible based on the Applicants need to recall and comprehend the power distribution relationship with RPS and the effect that a loss of that power has on the EPA trip relay status.
Technical
References:
04-1-01-C71-1, Attachment III E-0001, Main One Line (electrical distribution)
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-C7100, Objective 11 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last 2 NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 215003 Intermediate Range Monitor (IRM) System K1. Knowledge of the physical connections and/or cause K/A # 215003 effect relationships between INTERMEDIATE RANGE Rating 3.3 MONITOR (IRM) SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Rev / Date 0 K1.05 Display control system: Plant-Specific Question 36 A plant startup is in progress with reactor power indicating on the IRMs.
IRM D is indicating 45 on Range 4 when it is inadvertently placed on Range 6.
Which of the following will result from this action?
A. Half-Scram RPS A B. Half-Scram RPS B C. Rod withdrawal block D. Down pushbutton illuminates, only.
Answer: C Explanation:
With IRM D indicating 45 on range 4 going to range 5 would indicate 4.5, but on the 0 to 40 scale which would be 11% of scale. Continuing to range 6 would indicate on the 0 to 125 scale and would be 3.6 % of scale which is <5 % of scale that is the Control Rod Withdrawal block, therefore C is the only correct answer.
A and B are incorrect due to no scram setpoint is exceeded, but plausible if the student is confused on ranging down instead of up.
C is correct D is wrong, a Rod block will occur along with the down pushbutton.
Technical
References:
Tech Spec TR 3.3.2.1
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-C5102, Objective 7.2 Question Source: Bank # GGNS-OPS-09489a (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 215003 Intermediate Range Monitor (IRM) System K6. Knowledge of the effect that a loss or malfunction of the K/A # 215003 following will have on the INTERMEDIATE RANGE MONITOR Rating 3.6 (IRM) SYSTEM : (CFR: 41.7 / 45.7)
Rev / Date 0 K6.02 24/48 volt D.C. power: Plant-Specific Question 37 The plant startup is in progress, with the following conditions:
- Reactor power 8%
- Mode switch position RUN
- IRM A is BYPASSED The High Voltage Power Supply unit for IRM Drawer B fails and output voltage drops to zero.
Which of the following describes the status of the plant?
A. Full Reactor Scram, Control Rod Withdrawal Block B. Half Reactor Scram, Control Rod Withdrawal Block C. No RPS scram signal, Control Rod Withdrawal Block D. No RPS scram signal, No Control Rod Withdrawal Block Answer: D Explanation:
A loss of the High Voltage power supply unit which the output is 20 VDC, will cause a loss of DC power to the detector. The IRM INOP trip will actuate with a loss of High voltage power. IRM B will receive an INOP signal. However, with the plant Mode Switch in RUN all trips and rod blocks are bypassed.
A and B are wrong due to no scram signal is received due to IRMs are bypassed.
C is wrong, Rod block signal is bypassed D is correct.
Technical
References:
E1171 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-C5102, Objective 6.3, 7.4, 8.1 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 215004 Source Range Monitor (SRM) System A1. Ability to predict and/or monitor changes in parameters K/A # 215004 associated with operating the SOURCE RANGE MONITOR Rating 3.6 (SRM) SYSTEM controls including: (CFR: 41.5 / 45.5)
Rev / Date 0 A1.05 SCRAM, rod block, and period alarm trip setpoints Question 38 A reactor startup is in progress with the reactor subcritical.
Control rods are being withdrawn to achieve critically.
SRM E spikes to 2 x 105 cps.
Which of the following describes the plant response?
A. SRM Upscale alarm only B. SRM Upscale alarm Control Rod Withdrawal Block C. SRM Upscale alarm Control Rod Withdrawal Block Half scram D. SRM Upscale alarm Control Rod Withdrawal Block Full scram Answer: B Explanation:
A spike of any SRM during startup (IRM < range 8) of 2 x 105 is above the setpoint of 1 x 105 Control rod block.
A is wrong, yes you would receive the upscale alarm however a rod block would also be received.
B is correct C is wrong, a half scram would not occur.
D is wrong, a full scram would not occur.
Technical
References:
Tech Spec TR 3.3.2.1 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-C5101, Objective 8.2 Question Source: Bank # GGNS-OPS-06835 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 215005 Average Power Range Monitor/Local Power Range Monitor System K/A # 215005 A3. Ability to monitor automatic operations of the AVERAGE Rating 3.2 POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM including: (CFR: 41.7 / 45.7)
Rev / Date 0 A3.04 Annunciator and alarm signals Question 39 The plant is operating at 8% power in Mode 2.
The following alarms were received:
- CONTROL ROD WITHDRAWAL BLOCK
A. Reactor Mode Switch was taken to RUN B. Reactor power indicated <5%
C. Reactor power indicated >18%
D. Only 3 valid LPRM detectors per level are feeding Channel 1 Answer: C Explanation:
At 8% power the given alarms can only be generated by an Upscale signal A is wrong, If the mode switch was taken to run no alarms would be received B is wrong, <5% power is an downscale rod block not an upscale trip.
C is correct, D is wrong, 3 LPRMs per level is the minimum required but no inop signal is generated no alarms would be received.
Technical
References:
GLP-OPS-C5104 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-C5104, Objective 7.1 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 217000 Reactor Core Isolation Cooling System (RCIC)
A2. Ability to (a) predict the impacts of the following on the K/A # 217000 REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ; Rating 3.8 and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those Rev / Date 0 abnormal conditions or operations: (CFR: 41.5 / 45.6)
A2.15 Steam line break Question 40 A reactor scram has occurred due to a 'B' main steam line break in the Aux Building Steam Tunnel.
The Control Room Operators are implementing the actions of EP-2.
The MSIV's automatically closed on a NSSSS Group 1 Isolation.
(1) What is the impact of the above conditions on RCIC and (2) What actions will allow continued RCIC operation, if any?
A. (1) RCIC will automatically isolate immediately due to low steam flow since RCIC steam flow comes from the 'B' MSL.
(2) RCIC operation is administratively prohibited, use the Automatic Isolations ONEP to ensure automatic isolations occur.
B. (1) RCIC will automatically isolate immediately due to low steam flow since RCIC steam flow comes from the 'B' MSL.
(2) Install EP Attachment 3 to defeat RCIC isolations and non-mechanical turbine trips.
C. (1) RCIC will automatically isolate 30 minutes after the MSIV's closed assuming MSL tunnel temperatures remain high.
(2) Install EP Attachment 3 to defeat RCIC isolations and non-mechanical turbine trips.
D. (1) RCIC will automatically isolate 30 minutes after the MSIV's closed assuming MSL tunnel temperatures remain high.
(2) RCIC operation is administratively prohibited, use the Automatic Isolations ONEP to ensure automatic isolations occur.
Answer: C Explanation:
RCIC taps off the 'A' MSL (P&ID M-1077A).
RCIC will isolate after a 30 minute time delay when the Aux Building main steam tunnel temperature exceeds 185F (Automatic Isolations ONEP).
EP-2 (via bases discussion and allowance for Att 3 on EP-2 Table 1) specifically allows the use of Attachment 3 to defeat RCIC isolations. There is no administrative prohibition for using RCIC in this case.
For the above reasons, only the answer C is correct. All distracters are wrong but plausible based on the applicant knowing which steam line supplies RCIC and knowing that EP-2 allows for defeating automatic isolations for the given conditions.
Technical
References:
02-S-01-40 EP Bases Attachment IV 05-1-02-III-5 Automatic Isolations ONEP P&ID M-1077A References to be provided to applicants during exam: None Learning Objective: GLP-OPS-EP-4, Objective 19 Question Source: Bank # X (note changes; attach parent) Modified Bank #
New Question History: Last 2 NRC Exams No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(11) 55.43(b)
Examination Outline Cross-Reference Level RO 218000 Automatic Depressurization System K5. Knowledge of the operational implications of the K/A # 218000 following concepts as they apply to AUTOMATIC Rating 3.8 DEPRESSURIZATION SYSTEM : (CFR: 41.5 / 45.3)
Rev / Date 0 K5.01 ADS logic operation Question 41 A LOCA is in progress.
- Drywell pressure reaches 1.39 psig and continues to rise.
- Then, the MSIVs automatically close on low reactor water level.
When will the ADS Valves automatically open?
A. 105 seconds after the MSIVs get their automatic close signal B. 9.2 minutes after the MSIVs get their automatic close signal C. 9.2 minutes after drywell pressure reached 1.39 psig D. 105 seconds after drywell pressure reached 1.39 psig Answer: A Explanation:
If an ECCS pump is running (assumed by stem conditions), ADS will auto-initiate 105 seconds after the presence of ALL 3 of the following signals: a -150.3" signal, a confirmatory +11.4" signal, and a 1.39 psig drywell pressure signal. The need for the 1.39 psig DW pressure signal is auto-bypassed if level stays below -150.3" for at least 9.2 minutes;however, ADS still requires a running ECCS pump and the 105 second time delay.
Since the stem conditions indicate DW pressure has reached 1.39 psig, the 9.2 minute bypass is irrelevant;therefore, distracters 1 & 2 are wrong. Distracter 3 is wrong because the start of the 105 second timer always needs the coincident -150.3" level.
Technical
References:
Loop locic 17-S-06-5 M1077B
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-E2202, Objective 12.2 Question Source: Bank #
(note changes; attach parent) Modified Bank # GGNS-OPS-09496 New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off K/A # 223002 A4. Ability to manually operate and/or monitor in the control Rating 3.6 room: (CFR: 41.7 / 45.5 to 45.8)
Rev / Date 0 A4.03 Reset system isolations Question 42 The plant is preparing a hot startup following a plant transient.
The MSIVs are currently closed.
The mechanical vacuum pumps are maintaining main condenser vacuum at 7 Hg Vac.
Reactor water level is +35 Narrow Range Reactor pressure is 550 psig What actions (if any) are required to OPEN the MSIVs?
A. The MSIVs cannot be opened with current plant conditions.
B. Depress the NSSS INBD ISOL RESET and NSSS OTBD ISOL RESET pushbuttons.
All eight MSIVs will automatically open C. Place all eight MSIV handswitches to CLOSE Depress the NSSS INBD ISOL RESET and NSSS OTBD ISOL RESET pushbuttons.
Place all four OTBD MSIV handswitches to OPEN D. Place the NSSS DIV 1, 2, 3, and 4 CNDSR LO VAC BYP switches in BYP, Place all eight MSIV handswitches to CLOSE Depress the NSSS INBD ISOL RESET and NSSS OTBD ISOL RESET pushbuttons.
Place all four OTBD MSIV handswitches to OPEN Answer: D Explanation:
To reopen the MSIVs after a Group 1 signal the handswitches for each MSIV (8) must be taken to the CLOSE position then NSSSS must be reset (INBD and OTBD). Only the OTBD valves can be opened due to the DP across the INBD valves is >100 psig the steam lines must be equalized. Also with Main Condenser vacuum being 7 Hg Vac a group 1 signal is still present therefore the Condenser Lo Vac bypass switches must be placed in bypass before NSSSS can be reset. D is correct.
A is wrong, The MSIVs can be reopened, plausible due to the student may think that 7Hg Vac would prevent an opening.
B is wrong, Vacuum must be bypassed and the handswitches must be placed in closed.
Also the MSIVs will not auto open. Plausible due to the MSIVs will reopen on loss of air or loss of solenoid power then restored.
C is wrong, Condenser vac must be bypassed.
Technical
References:
04-1-01-B21-1 Section 4.3 04-1-01-M71-1 Section 5.4 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-M7101, Objective 7.1 & 9.0 Question Source: Bank # GGNS-OPS-08812a (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 239002 Relief/Safety Valves K2. Knowledge of electrical power supplies to the following:
K/A # 239002 (CFR: 41.7) Rating 2.8 K2.01 SRV solenoids Rev / Date 0 Question 43 What are the power supplies to A and B solenoids of the Safety Relief Valves (SRVs)?
A. 11DA & 11DB B. 11DD & 11DE C. 11DE & 11DK D. 11DA & 11DC Answer: A Explanation:
SRV solenoids are powered from ESF DC buses 11DA and 11DB.
A is correct B is wrong, plausible due to these are 125 vdc busses also but used for BOP, C is wrong, plausible due to these supply power to the ATWS/ARI solenoids.
D is wrong, 11DA is correct but 11DC is wrong, plausible due to 11DC is also an ESF DC bus.
Technical
References:
04-1-01-B21-1 E1161 SH 4 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-E2202, Objective 9.1 Question Source: Bank # GGNS-OPS-09948
(note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 259002 Reactor Water Level Control System A3. Ability to monitor automatic operations of the REACTOR K/A # 259002 WATER LEVEL CONTROL SYSTEM including: (CFR: 41.7 / Rating 3.0 45.7)
Rev / Date 0 A3.01 Runout flow control: Plant-Specific Question 44 Only the A Reactor Feed Pump is operating.
Considering only the Feedwater System Operating Instruction P&Ls, which of the following describes the limitations on the operating feed water pump?
Speed is A. administratively restricted only.
B. clamped at 5850 rpm in FW Auto only.
C. clamped at 5850 rpm in Speed Auto and FW Auto only.
D. clamped at 5850 rpm in Emergency Manual, Speed Auto and FW Auto.
Answer: C Explanation:
Per 04-1-01-N21-1 step 3.15. In SPEED AUTO AND FW AUTO, RFPT speed is clamped at 5800 rpm for dual feed pump operation and 5850 for single feed pump operation. Step 3.3 states that there is no speed restriction in emergency manual operation.
Technical
References:
04-1-01-N21-1, P&Ls 3.3 and 3.15 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-N2100, Objective 19.0, 29.0 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X
Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 261000 Standby Gas Treatment System K1. Knowledge of the physical connections and/or cause K/A # 261000 effect relationships between STANDBY GAS TREATMENT Rating 2.8 SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
Rev / Date 0 K1.08 Process radiation monitoring system .
Question 45 Which of the following will cause a Standby Gas Treatment system B initiation?
A. Fuel Handling Area exhaust rad monitor A indicating 4.2 mrem/hr Fuel handling area exhaust rad monitor C mode switch in STANDBY B. Fuel Pool Sweep exhaust rad monitor B indicating 7.9 mrem/hr Fuel Pool Sweep exhaust rad monitor C indicating 8.1 mrem/hr C. Fuel Handling Area exhaust rad monitor B indicating 5.3 mrem/hr Fuel Handling Area exhaust rad monitor C INOP trip D. Fuel Pool Sweep exhaust rad monitor B indicating 31 mrem/hr Fuel Pool Sweep exhaust rad monitor D INOP trip Answer: C Explanation:
For SBGT to initiate on process rad monitor system the A systems must see A and D monitors and B must see B and C monitors, for Fuel handling area > 3.6 mrem/hr or INOP OR Fuel pool sweep >30 mrem/hr or INOP.
A is wrong, due to A and C monitors will not initiate either system B is wrong, due to B and C monitors did not reach their setpoint of 30 mrem/hr.
C is correct, see explanation above.
D is wrong, due to B and D monitors will not initiate either system Technical
References:
04-1-01-T48-1 E1257 SH01 E1177 SH 32 & 33 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-T4801, Objective 8.7 Question Source: Bank # GGNS-OPS-00141a (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 262001 A.C. Electrical Distribution K4. Knowledge of A.C. ELECTRICAL DISTRIBUTION design K/A # 262001 feature(s) and/or interlocks which provide for the following: Rating 2.8 (CFR: 41.7)
Rev / Date 0 K4.04 Protective relaying Question 46 A loss of ST-11 has occurred (1) Which of the following describes the BOP buses that were de-energized?
(2) What is required to re-energize the lost busses and restart major loads after the bus is re-energized from an alternate source?
A. (1) 12HE and 13AD (2) Locally reset Bus Undervoltage relays.
B. (1) 11HD and 12AE (2) Locally reset Bus Undervoltage relays.
C. (1) 12HE and 13AD (2) Bus Undervoltage relays will auto reset D. (1) 11HD and 12AE (2) Bus Undervoltage relays will auto reset Answer: A Explanation:
Per 05-1-02-I-4 Technical
References:
05-1-02-I-4, step 2.0
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-ONEP, Objective 14.0 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 262001 A.C. Electrical Distribution A4. Ability to manually operate and/or monitor in the control K/A # 262001 room: (CFR: 41.7 / 45.5 to 45.8) Rating 3.6 A4.04 Synchronizing and paralleling of different A.C. supplies Rev / Date 0 Question 47 The division 3 Diesel Generator is running in parallel with the grid.
All control switches are in the normal handswitch lineup.
Which of the following describes the effect of placing the DIV 3 Diesel Generator Voltage Regulator switch to RAISE?
A. Division 3 Diesel Generator Real load or Watts will rise.
B. Division 3 Diesel Generator output voltage will rise.
C. Division 3 Diesel Generator Reactive load or VARS will rise.
D. Division 3 Diesel Generator output frequency will rise.
Answer: C Explanation:
With the Diesel Gen in parallel raising the Voltage reg switch will cause reactive load or VARS to rise, therefore C is correct.
Technical
References:
04-1-01-P81-1 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-P8100, Objective 17 Question Source: Bank # GGNS-OPS-02697 (note changes; attach parent) Modified Bank #
New Question History: Last 2 NRC Exams No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 262002 Uninterruptable Power Supply (A.C./D.C.)
A4 - Ability to manually operate and/or monitor in the control K/A # 262002 room: Rating 2.8 A4.01 Transfer from alternative source to preferred source Rev / Date 0 Question 48 The following annunciator was received and sealed in:
STATIC INVRTR 1Y97 TROUBLE A local operator reports that 1Y97 indicates Alternate Source Supplying Load.
Which of the following describes when this annunciator should clear?
After Normal source is restored and .
A. static transfer switch auto transfers to Normal source.
B. manual transfer switch is transferred to Normal source to load.
C. manually depress the Inverter to load pushbutton.
D. local alarm acknowledge pushbutton is depressed.
Answer: A Explanation:
1Y97 is a BOP inverter therefore when loss of normal supply the alarm is received.
After normal supply is restored the BOP inverters will auto transfer back to the normal supply which will then clear the alarm.
A is correct Technical
References:
04-1-01-L62-1
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-L6200, Objective 17 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last 2 NRC Exams No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 263000 D.C. Electrical Distribution 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 /
K/A # 263000 43.5 / 45.13) Rating 3.7 Rev / Date 0 Question 49 A LOCA has occurred with a Station Blackout.
DC electrical bus 11DA has tripped.
Which of the following procedures can be used to operate the RCIC system for EOP mitigation?
A. RCIC System Operating Instructions (SOI)
B. 05-1-02-I-4, Loss of AC Power ONEP C. 05-1-02-II-1, Shutdown From the Remote Shutdown Panel ONEP D. 05-S-01-STRATEGY, Alternate Strategy Answer: D Explanation:
Alternate Strategy procedure provides pre-established strategies for dealing with significant events well outside design bases and current procedures. This guidance is focused on operation of starting RCIC from outside the control room.
A is wrong, the system SOI does not go outside the design B is wrong, the ONEP does not cover loss of DC.
C is wrong, Remote shutdown panel does not cover loss of DC D is correct, Technical
References:
05-S-01-STRATEGY References to be provided to applicants during exam: None Learning Objective: GLP-OPS-B5B00, OBJ. 2 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last 2 NRC Exams No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 263000 DC Electrical Distribution K3.02 - Knowledge of the effect that a loss or malfunction of the K/A # 262001 D.C. ELECTRICAL DISTRIBUTION will have on the following: Rating 3.3/3.6 Components using D.C. control power (i.e. breakers)
Rev / Date 0 Question 50 Bus 15AA is energized.
CRD pump A is operating.
Battery bus 11DA is subsequently de-energized.
Circuit breaker 152-1505 for CRD pump A:
A. will not be capable of any remote or local operation until bus 11DA is restored.
B. can be opened, closed and reopened once, manually, at the breaker.
C. can be opened remotely but must be closed manually, at the breaker.
D. will only trip automatically on a protective relay actuation.
Answer: B Explanation:
A is wrong; the breaker will not operate remotely without DC power B is correct: Any 4160 breaker that is already closed when the loss of DC occurs can be opened due to the opening springs are compressed by the closing of the breaker, the closing springs are also compressed when the breaker is closed C is wrong; the breaker will not operate remotely without DC power D is wrong; trip coil is DC powered Technical
References:
04-1-01-L11-1
References to be provided to applicants during exam: NONE Learning Objective: GLP-OPS-L1100. OBJ 10.1 Question Source: Bank # GGNS-OPS-02739a (note changes; attach parent) Modified Bank #
New Question History: Last 2 NRC Exams No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b) 55.43(b)(2)
Examination Outline Cross-Reference Level RO 264000 Emergency Generators (Diesel/Jet)
K5. Knowledge of the operational implications of the K/A # 264000 following concepts as they apply to EMERGENCY Rating 3.4 GENERATORS (DIESEL/JET) : (CFR: 41.5 / 45.3)
Rev / Date 0 K5.06 Load sequencing Question 51 A Station Blackout has occurred.
Reactor water level is -172 inches and slowly lowering.
Drywell pressure is 2.5 psig and slowly rising.
Operators are restoring 16AB from Div 3 Diesel Generator.
Which of the following describes the order components are restored after 16AB is energized?
A. RHR B relay logic is restored and allow LSS to perform a LOCA sequence.
B. SSW B pump is manually started then RHR B logic is restored C. RHR B relay logic is restored causing SSW B to auto start.
D. RHR B is manually started causing SSW B to auto start.
Answer: B Explanation:
Per 05-1-02-I-4, Loss of AC Power, 3.2.10 , LSS is shutdown prior to energizing the 16AB bus and remains shutdown to protect the Div 3 D/G from overloading and prevent surge loading from simultaneous start of large pumps.
After the power is restored in step 3.2.10 I Starts SSW pump B manually prior to energizing the RHR logic power.
A is wrong; SSW is manually started prior to restoring relay logic and LSS is shutdown and will not perform its function.
B is correct:
C is wrong; SSW is manually started prior to restoring relay logic to prevent a simultaneous start of large components D is wrong; Same as C Technical
References:
05-1-02-I-4 References to be provided to applicants during exam: NONE Learning Objective: GLP-OPS-ONEP. OBJ 13.0 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last 2 NRC Exams No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b) 55.43(b)(2)
Examination Outline Cross-Reference Level RO 300000 Instrument Air System (IAS)
K6 Knowledge of the effect that a loss or malfunction of the K/A # 300000 following will have on the INSTRUMENT AIR SYSTEM: (CFR: Rating 2.7 41.7 / 45.7)
Rev / Date 0 K6.03 Temperature indicators Question 52 Which of the following will cause a Plant Air Compressor trip?
A. Compressor Cabinet Temperature failing HIGH B. Compressor Cooling Water Temperature failing HIGH C. Compressor Oil Filter DP failing LOW D. Compressor Oil Temperature failing LOW Answer: A Explanation:
Per 04-S-02-SH13-P854-1A-A4, A is Correct B is wrong: The compressors do not have a cooling water temp trip C is wrong; Low oil filter D/P does nothing, but high filter D/P would cause a trouble alarm.
D is wrong; Oil temp high or pressure low will trip compressor Technical
References:
04-S-02-SH13-P854-1A-A4 References to be provided to applicants during exam: NONE Learning Objective: GLP-OPS-P5100. OBJ 15.0 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last 2 NRC Exams No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b) 55.43(b)(2)
Examination Outline Cross-Reference Level RO 400000 Component Cooling Water System (CCWS)
A4. Ability to manually operate and / or monitor in the control K/A # 400000 room: (CFR: 41.7 / 45.5 to 45.8) Rating 3.1 A4.01 CCW indications and control Rev / Date 0 Question 53 A LOCA has occurred.
Reactor water level dropped to -50 Wide Range and is currently -30 and rising.
Drywell pressure is 1.29 psig and slowly rising.
You are verifying initiations and isolations per EP-2.
Which of the following describes the status of the SSW systems?
A. SSW A is running with flow going to all components SSW B is in Standby SSW C is supplying all of its associated components B. SSW A is in Standby SSW B is running supplying flow to all components SSW C is in Standby C. SSW A is running supplying flow to the RCIC room cooler SSW B is in Standby SSW C is supplying all of its associated components D. SSW A is in Standby SSW B is running supplying flow to the RCIC room cooler SSW C is supplying all of its associated components
Answer: C Explanation:
With level reaching <-41.6 a RCIC auto initiation occurred, SSW A starts when the E51-F045 is not full closed but will only supply flow to the ECCS room coolers only. Also HPCS auto initiated and SSW C will supply all its associated loads.
A is wrong, due to only the RCIC room cooler (along with all ECCS room coolers) is receiving flow from SSW A.
B is wrong, due to SSW B does not provide flow for RCIC room cooler and SSW C will be running due to HPCS initiation C is correct, see explanation above.
D is wrong, due to due to SSW B does not provide flow for RCIC room cooler.
Technical
References:
04-1-01-E51-1 04-1-01-P41-1 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-P41, Objective 10.1 & 10.2 Question Source: Bank # GGNS-OPS-00778 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 201003 Control Rod and Drive Mechanism K1. Knowledge of the physical connections and/or cause K/A # 201003 effect relationships between CONTROL ROD AND DRIVE Rating 3.1 MECHANISM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Rev / Date 0 K1.03 RPIS Question 54 Which of the following is the reason for CONT ROD OVERTRAVEL annunciator?
A. The control rod has inserted too far as detected by an independent reed switch closing past the 00 indication.
B. The control rod has inserted too far as detected by the reed switch for 00 indication no longer being closed.
C. The control rod has become uncoupled as detected an independent reed switch closing past the 48 indication.
D. The control rod has become uncoupled as detected by the reed switch for 48 indication no longer being closed.
Answer: C Explanation:
If the coupling spud becomes uncoupled from the control rod the drive piston will withdraw a small additional distance to reach its lover mechanical end stop. In this overtravel position, the attached ring magnet actuates the overtravel reed switch in the indicator probe, thus providing an indication of control rod and drive separation.
A and B are wrong, due to this is inserted which cant be overtravel or uncoupled.
C is correct, see explanation above.
D is wrong, due to an independent reed switch exist called the Overtravel reed switch past the 48 position.
Technical
References:
GLP-OPS-C111B, Page 14 of 51 References to be provided to applicants during exam: None
Learning Objective: GLP-OPS-C111B, Objective 5.2, 7.2 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 202001 Recirculation System 2.4.9 Knowledge of low power/shutdown implications in K/A # 202001 accident (e.g., loss of coolant accident or loss of residual Rating 3.8 heat removal) mitigation strategies. (CFR: 41.10 / 43.5 / 45.13)
Rev / Date 0 Question 55 A plant startup is in progress with all IRMs on Range 8.
Which of the following must be met in order to satisfy assumptions for providing core cooling in the first few seconds of a DBA LOCA in a Recirc Loop?
A. Recirc loop flows are matched to within 10% of rated core flow.
B. Recirc loop flows are matched to within 5% of rated core flow.
C. ESF Diesel Generators come up to rated speed and voltage within 10 seconds.
D. HPCS or LPCS inject to the core at > 7000 gpm within 15 seconds.
Answer: A Explanation:
FSAR section 5.4.1.7.8 states that jet pump flows are closely matched for a DBA LOCA to ensure the design bases peak cladding temperatures are not exceeded during the first few seconds of the event. (This is also explained in the TS bases for TS 3.4.1)
TS 3.4.1 (applicable in Modes 1 & 2) requires loop jet pump flows be matched within 10%
when operating at < 70% rated core flow. In reactor Mode 2, Recirc Pumps are in slow speed and core flow is below 70% rated. If core flow were above 70% rated, TS 3.4.1 SR 3.4.1.1 would require loop flow to be within 5%. The plant conditions in the stem make the answer correct and distracter 1 wrong.
Distracter 2 is plausible because ESF Diesel Generators are required to be at rated speed and voltage within 10 seconds per SR 3.8.1.2.
Distracter 3 is plausible because one method of adequate core cooling is for HPCS or LPCS to inject > 7000 gpm; however, for the DBA it is assumed that 3 ECCS systems are operable in order to re-flood the core.
Technical
References:
FSAR section 5.4.1.7.8 TS 3.4.1 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-B3300 Obj 2 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)
Examination Outline Cross-Reference Level RO 204000 Reactor Water Cleanup System K4. Knowledge of REACTOR WATER CLEANUP SYSTEM K/A # 204000 design feature(s) and/or interlocks which provide for the Rating 3.3 following: (CFR: 41.7)
Rev / Date 0 K4.08 Reducing reactor pressure upstream of low pressure piping: LP-RWCU Question 56 The plant is in Mode 4.
Reactor water level is being maintained by RWCU blowdown.
G33-F033 is 15% open.
The Operator at the controls notices reactor water level rising due to the G33-F033 automatically closing.
Which of the following describes the reason for the G33-F033 automatic action?
A. Reactor Water level reached +53.5 Narrow Range B. Line pressure downsteam of the G33-F033 is >140 psig.
C. Filter Demin Inlet Temperature >140°F D. Any Group 8 RWCU isolation signal is received Answer: B Explanation:
The G33-F033 will auto close upon receipt of either low pressure upsteam or high pressure downstream of the F033. IF F033 is > 5% open.
A is wrong. No auto actions for RWCU at this level B is correct, C is wrong, This will cause the F004 to auto close and cause a pump trip.
D is wrong, Group 8 isolation does not affect the F033.
Technical
References:
GLP-OPS-G3336, page 26 of 43 04-1-02-1H13-P680-11A-C5 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-G3336, Objective 8.6 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 216000 Nuclear Boiler Instrumentation K5. Knowledge of the operational implications of the K/A # 216000 following concepts as they apply to NUCLEAR BOILER Rating 3.5 INSTRUMENTATION : (CFR: 41.5 / 45.3)
Rev / Date 0 K5.13 Reference leg flashing: Design-Specific Question 57 A significant leak has occurred from the isolation valve on the Reference leg of the transmitter that supplies C Narrow Range reactor water level detector.
Which of the following describes the expected response of the C Narrow Range level indication?
The C Narrow Range indicated level will A. be higher than either A or B indicated levels.
B. be lower than either A or B indicated levels.
C. be the same as A or B indicated levels.
D. oscillate higher and lower than A or B indicated levels.
Answer: A Explanation:
With a leak on the reference leg of a level transmitter the DP will reduce causing indicated level to rise.
A is correct.
B is wrong, due to the reason A is correct the indicated level will rise C is wrong, due to the reason A is correct the indicated level will rise.
D is wrong, due to the reason A is correct the indicated level will rise.
Technical
References:
04-1-01-B21-1 GLPL-OPS-COM07
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-B2101, Objective 12.2 GLP-OPS-COM07, Objective 10a Question Source: Bank # GGNS-OPS-00372 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 223001 Primary Containment System and Auxiliaries K4. Knowledge of PRIMARY CONTAINMENT SYSTEM AND K/A # 223001 AUXILIARIES design feature(s) and/or interlocks which Rating 3.1 provide for the following: (CFR: 41.7)
Rev / Date 0 K4.06 Maintains proper containment/secondary containment to drywell differential pressure Question 58 Which of the following conditions will the Normal Drywell Vacuum Relief valves (E61-F007 & E61-F020) OPEN?
A. Upon receipt of a LOCA B. 30 seconds after the receipt of a LOCA C. When Drywell pressure is 0.186 psig less than Containment pressure and no LOCA signal present.
D. When a LOCA signal is present and Drywell pressure is 0.88 psig greater than Containment pressure.
Answer: C Explanation:
The Normal Drywell Vacuum Relief valves will auto open when NO LOCA signal is present and Drywell pressure is 0.186 psig less than Containment pressure.
A is wrong, The E61 CTMT and Drywell H2 Analyzers will auto upon receipt of a LOCA B is wrong, the 30 second time delay is in the E61 Drywell purge compressor and Post LOCA Vacuum Relief valves (E61 F003A & B) start/open logic C is correct.
D is wrong, These signals is associated with the Post LOCA Drywell Vacuum relief valve.
Technical
References:
04-1-01-E61-1 GLP-OPS-E6100
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-B2101, Objective 12.2 GLP-OPS-COM07, Objective 10a Question Source: Bank # GGNS-OPS-00339 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 233000 Fuel Pool Cooling and Clean-up A2. Ability to (a) predict the impacts of the following on the FUEL K/A # 233000 POOL COOLING AND CLEAN-UP ; and (b) based on those Rating 2.7 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: Rev / Date 0 41.5 / 45.6)
A2.09 - A.C. electrical power failures Question 59 The plant is operating at rated conditions.
Fuel Pool Cooling and Cleanup is operating with both pumps and both filters in service.
Both Filter Bypass Valves are closed.
Incoming feeder breaker for 15AA bus trips.
15AA power is restored by its Diesel Generator.
Which of the following describes:
(1) the status of the Fuel Pool Cooling and Cleanup system?
(2) the procedure being used to mitigate the transient?
A. (1) Both pumps running (2) Loss of AC Power ONEP B. (1) A pump running, B pump tripped (2) Loss of AC Power and Inadequate Decay heat Removal ONEPs C. (1) B pump running, A pump tripped (2) Loss of AC Power and Inadequate Decay heat Removal ONEPs D. (1) Both pumps are tripped (2) Loss of AC Power and Inadequate Decay heat Removal ONEPs
Answer: D Explanation:
A loss of either 15AA or 16AB will cause one pump to trip due to loss of power, but, the Air operated Filter Demin inlet valves F019 and F045 are powered from the 15 and 16 bus also and are in series. Therefore on a loss of either bus one pump will trip and cause the closure of one filter inlet valve causing a trip of the other pump on low flow.
A is wrong, Both pumps will trip and Decay heat removal ONEP will be entered B is wrong, Both pumps will trip C is wrong, Both pumps will trip D is correct, Technical
References:
04-1-01-G41-1, Att III page 1 of 3 05-1-02-III-1, 4.1 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-G4146, Objective 7.6 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 239001 Main and Reheat Steam System K6. Knowledge of the effect that a loss or malfunction of the K/A # 239001 following will have on the MAIN AND REHEAT STEAM Rating 3.3 SYSTEM: (CFR: 41.7 / 45.7)
Rev / Date 0 K6.08 Main condenser vacuum Question 60 The plant is experiencing a loss of condenser vacuum.
Which of the following identifies the order in which automatic actions occur in response to the degrading vacuum?
A. 1) Main Turbine Trip
- 2) Feed Pump Turbine Trip
- 3) Turbine Bypass Valve closure
- 4) MSIV closure B. 1) Main Turbine Trip
- 2) Turbine Bypass Valve closure
- 3) Feed Pump Turbine Trip
- 4) MSIV closure C. 1) Feed Pump Turbine Trip
- 2) Main Turbine Trip
- 3) Turbine Bypass Valve closure
- 2) Turbine Bypass Valve closure
- 3) Feed Pump Turbine Trip
- 4) Main Turbine Trip Answer: A Explanation:
With degrading vacuum the auto trip setpoints are as follows:
21 Main Turbine Trip 16 Feedpump Turbine Trip
12 Bypass valve closure 9 MSIV closure A is correct, B is wrong, the feedpump will trip before bypass valve closure C is wrong, Main turbine will trip before the feedpump turbine D is wrong, These are in the opposite order.
Technical
References:
04-1-01-N62-1 05-1-02-V-8, section 5.0 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-N6200, Objective 14 GLP-OPS-ONEP, Objective 39.0 Question Source: Bank # GGNS-OPS-01990 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 245000 Main Turbine Generator and Auxiliary Systems K3. Knowledge of the effect that a loss or malfunction of the K/A # 245000 MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS will Rating 3.7 have on following: (CFR: 41.7 / 45.4)
Rev / Date 0 K3.08 Reactor/turbine pressure control system: Plant-Specific Question 61 The plant is operating at rated power.
At P680, the CF TK LVL LO alarm is received and seals-in.
Local operators report that EHC tank level is lowering and theyve found a large EHC leak.
Control room operators perform the following:
- Place the Mode Switch in SHUTDOWN
- Trip Main Turbine
- Trip the EHC pumps.
An ATWS occurs and EP-2A is entered.
Operators will use ______(1)_______ for pressure control and maintain reactor pressure within a band of ______(2)______.
A. (1) Bypass Control Valves (2) 800 psig to 1060 psig B. (1) Bypass Control Valves (2) 450 psig to 600 psig C. (1) SRVs / MSL Drains (2) 800 psig to 1060 psig D. (1) SRVs / MSL Drains (2) 450 psig to 600 psig Answer: C Explanation:
Per ARI P680-10A-B4, If a large leak is found, remove the turbine generator from service and shutdown the control fluid pumps to repair leakage. With the EHC
pumps shutdown the bypass valves would fail closed, therefore pressure control would be from SRVs with a band of 800 psig to 1060 psig.
A is wrong. The bypass control valves will fail closed due to low EHC pressure.
B is wrong. The bypass control valves will fail closed due to low EHC pressure. The lower band is incorrect due to MSIVs are still open and Feedwater is unaffected.
C is correct. See explanation above.
D is wrong. The lower band is incorrect due to MSIVs are still open and Feedwater is unaffected.
Technical
References:
04-1-02-1P680-10A-B4.
04-1-01-N32-1, 3.10 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-N3200, Objective 18.1 Question Source: Bank # 240 (note changes; attach parent) Modified Bank #
New Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 259001 Reactor Feedwater System K5. Knowledge of the operational implications of the K/A # 259001 following concepts as they apply to REACTOR FEEDWATER Rating 2.8 SYSTEM : (CFR: 41.5 / 45.3)
Rev / Date 0 K5.03 Turbine operation: TDRFP's-Only Question 62 The plant is operating at rated power when the Main Turbine trips.
The operator at the controls aligns for startup level control using Reactor Feed Pump A.
10 minutes after the Scram the transient is complete; all plant parameters are stable again.
Which of the following describes the Governor Control Valve position for RFPT A?
A. Slightly higher than original indication due to more load on the feed pump turbine B. Slightly lower than original indication due to the Scram.
C. More than double the original indication due to a swap to its high pressure steam source.
D. One-half the original indication due to the Scram.
Answer: C Explanation:
Rated position for the governor valve is 20% after a scram the governor will slowly move to high pressure steam and governor valve will be 60%.
A is wrong. See explanation above B is wrong. See explanation above C is correct. See explanation above D is wrong. See explanation above
Technical
References:
04-1-01-N21-1, Feedwater SOI References to be provided to applicants during exam: None Learning Objective: GLP-OPS-N2100, Objective 17, 18 Question Source: Bank # 2012 Audit 1 X (note changes; attach parent) Modified Bank #
New Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 271000 Offgas System A4. Ability to manually operate and/or monitor in the control K/A # 271000 room: (CFR: 41.7 / 45.5 to 45.8) Rating 2.8 A4.01 Reset system isolations Rev / Date 0 Question 63 The plant is operating at 25% power.
The N64-F060, Offgas System Discharge to Radwaste Ventilation Valve, has automatically closed on Post Treatment Hi Hi Hi radiation.
Which of the following describes how Offgas flow is re-established?
Verify Post Treatment radiation indication is below the trip setpoint and:
A. Depress the NSSSS Reset Pushbuttons B. Verify N64-F060 auto reopens C. Place handswitch for N64-F060 to CLOSE and back to AUTO/OPEN D. Place Brass Key lock switch for N64-F045, Adsorber Train Bypass Valve, to BYPASS and back to TREAT Answer: B Explanation:
Once the F060 closes on a Post Treat rad signal it will auto reset once the signal is cleared. 04-1-02-1H13-P601-19A-C8 step 3.7, If the radiation levels drop below the Hi Hi Hi trip, ensure N64-F060 opens and indicates open on panel P845.
A is wrong. This is required for and NSSSS isolation valve, N64-F060 is not part of NSSSS.
B is correct. See explanation above C is wrong. This is required for opening an MSIV after auto closure.
D is wrong. This will only open the F045 and close the F060 when taken to BYPASS Technical
References:
04-1-02-1H13-P601-19A-C8 step 3.7 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-N6465, Objective 11.9, 13.0 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 286000 Fire Protection System A3 - Ability to monitor automatic operations of the FIRE K/A # 286000 PROTECTION SYSTEM including: Rating 3.4 A3.01 Fire water pump start Rev / Date 0 Question 64 The deluge system for ESF Transformer 12 inadvertently initiates.
15 seconds later Fire Water system header pressure drops to 102 psig and rising.
Operators isolate the deluge and fire water header pressure rises to 150 psig.
Which of the following describe the status of the Fire Water Pumps 10 minutes after the deluge was isolated?
A. A and B Diesel Driven Fire Water pumps running Motor Driven Fire Water pump secured Jockey pump secured B. A and B Diesel Driven Fire Water pumps running Motor Driven Fire Water pump running Jockey pump secured C. A and B Diesel Driven Fire Water pumps secured Motor Driven Fire Water pump running Jockey pump secured D. A and B Diesel Driven Fire Water pumps secured Motor Driven Fire Water pump secured Jockey pump running
Answer: A Explanation:
Jockey pump will auto start on 135 psig and auto stop at 147 psig. Motor driven pump will auto start at 129 psig and auto stop at 141 psig, after 7 minutes. Diesel driven pump A auto starts at 123 psig + 5 seconds and B auto starts at 117 psig + 5 seconds. With pressure at 102 psig for at least 15 seconds all pumps should have initially started. 10 minutes after pressure has stabilized at 150 psig the jockey pump and motor driven pumps should have auto stopped. The diesel driven do not have an auto stop feature they must be manually shutdown.
A is correct. This is required for and NSSSS isolation valve, N64-F060 is not part of NSSSS.
B is wrong. Motor driven would not be running C is wrong. Diesel Driven would be running.
D is wrong. Diesel Driven would be running Technical
References:
GLP-OPS-P6400, page 13 and 15 of 73 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-P6400, Objective 6.1, 6.2, 6.3 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 290003 Control Room HVAC K4. Knowledge of CONTROL ROOM HVAC design feature(s)
K/A # 290003 and/or interlocks which provide for the following: (CFR: 41.7) Rating 3.1 K4.01 System initiations/reconfiguration: Plant-Specific Rev / Date 0 Question 65 The plant is operating at 100% power when the following occurs:
At time - 0 minutes All 4 Control Room Vent Radiation Monitors are indicating 6 mRem/hr.
At time - 5 minutes A LOCA occurs.
At time - 10 minutes Reactor water level reaches -51 Wide Range, HPCS and RCIC auto initiate and restore level to normal band.
Which of the following describes the operation of the Fresh Air Inlet valves, Z51-F007 and Z51-F016, for the Control Room Standby Fresh Air Ventilation system?
A. Receives isolation signal at time 0 minutes Interlocked closed until all signals are manually reset.
B. Receives isolation signal at time 0 minutes Can be manually opened at time 10 minutes.
C. Receives isolation signal at time 10 minutes Interlocked closed until all signals are manually reset.
D. Receives isolation signal at time 10 minutes Can be manually opened at time 20 minutes.
Answer: B Explanation:
On the receipt of a control room isolation signal of -41.6 RPV level or 1.23 psig D/W pressure or >5 mrem/hr on Control Room vent rad monitor, the SBFA unit will auto start in the Recirc mode. Fresh air inlet valves F007 and F016 will auto close and be interlocked closed for 10 minutes. Any subsequent initiation signal during the 10 min time delay has
no effect.
A is wrong, The valves can be re-opened after 10 min time delay.
B is correct C is wrong, the Valves will auto close at time 0 and can be re-opened after 10 min time delay.
D is wrong, the Valves will auto close at time 0 and can be re-opened after 10 min time delay.
Technical
References:
04-1-01-Z51-1 E-0131 SH 3 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-Z5100, Objective 9, 11 Question Source: Bank # GGNS-OPS-09533 (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 2.1.21 Ability to verify the controlled procedure copy.
(CFR: 41.10 / 45.10 / 45.13) K/A # 2.1.21 Rating 3.5 Rev / Date 0 Question 66 When performing SOI actions to operate a plant system in the Control Room, operators must A. verify the copy used is the current revision using the Electronic Data Management System (EDMS) except in an emergency.
B. use a Control Room controlled copy or verify the copy used is the current revision using the Electronic Data Management System (EDMS).
C. use only a Control Room controlled copy, other copies are not allowed to be used in the Control Room.
D. verify the copy used is the current revision using the Electronic Data Management System (EDMS) in all cases Answer: B Explanation:
EN-HU-106 requires all personel to verify prior to use the correct revision of the procedure and refers to EN-AD-103 for verification practices. EN-AD-103 states that revisions are verified in EDMS only, but that Control Room controlled copies may be used without verifying the current revision in EDMS (it is assumed that they are always the most recent revision).
A is wrong, procedure allows use of controlled copies.
B is correct C is wrong, procedure copies are allowed as long as current rev is checked.
D is wrong, procedure allows use of controlled copies
Technical
References:
EN-AD-103 EN-HU-106 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-PROC Obj. 12.3 Question Source: Bank # 552 X (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 2.1.20 Ability to interpret and execute procedure steps. (CFR:
41.10 / 43.5 / 45.12)
K/A # 2.1.20 Rating 4.6 Rev / Date 0 Question 67 Operators are in EP-2A and have just begun to re-inject to the RPV at 4000 gpm with RHR A.
Per 02-S-01-27, Operations Philosophy, the rate of injection is to be raised (if necessary) in increments of A. 500 gpm B. 1000 gpm C. 2000 gpm D. 2500 gpm Answer: B Explanation:
02-S-01-27, Section 6.2.11 states, If reactor level does not begin to indicate an upward trend, then raise injection rate in approximately 0.5 Mlbm/hr (1000 gpm) increments B is correct Technical
References:
02-S-01-27 step 6.2.11 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-PROC Obj. 57.8 Question Source: Bank # LORQT-06501a X (note changes; attach parent) Modified Bank #
New
Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.
(CFR: 41.10 / 43.5 / 45.12)
K/A # 2.1.5 Rating 2.9 Rev / Date 0 Question 68 Which of the following is a Tech Spec requirement for MINIMUM shift manning?
A. One RO and one SRO present in the control room during MODE 1, 2, or 3.
B. One RO and one SRO present in the control room when fuel is in the reactor.
C. Two non-licensed operators present on the site when fuel is in reactor.
D. Two ROs and one SRO in the control room during MODE 1, 2, or 3.
Answer: A Explanation:
The several sources are: Tech Spec Administrative Controls Section 5.2.1, Technical Requirements Manual (TRM) Section 7.0 (including Table 7.2.2-1), and Conduct of Operations procedure, EN-TQ-115, Attachment 9.3, Section 1.
Per Tech Spec Administrative Controls Section 5.2.1and the Conduct of Operation Att. 9.3 combination of step 5 and 5a states that MODE 1, 2, or 3, there need be only one RO and one SRO in the control room.
B is wrong. Fuel is in the reactor (prospectively) even in MODE 5. The Conduct of Operations Att. 9.3 step 5a (allowing just an RO to be in the control room in, for example, MODE 5) is consistent with The TRM Table 7.2.2-1 (allowing just an RO to be present in the control room in MODE 5). Therefore, this choices claim that the MINIMUM is having one RO and one SRO in the control room when fuel is in the reactor is wrong.
C is wrong. The Conduct of Operations Att. 9.3 reference speaks to the requirements for non-licensed operators on site when fuel is in the reactor. There the MINIMUM only one non-licensed (not two). This is consistent with the TRM Table 7.2.2-1 (allowing just one non-licensed operator on site in MODE 4, or 5)
C is wrong. TRM Table 7.2.2-1 requires there to be at least 2 ROs and one SRO for minimum crew manning and does not specify the location (1 RO may rove).
Technical
References:
Tech Spec Administrative Controls Section 5.2.1 Technical Requirements Manual (TRM) Section 7.0 (including Table 7.2.2-1)
Conduct of Operations procedure, EN-OP-115, Attachment 9.3, Section 1.
References to be provided to applicants during exam: None Learning Objective: GLP-OPS-PROC Obj. 1.12, GLP-OPS-CFR01 Obj. 17 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) 55.43(b)
Examination Outline Cross-Reference Level RO 2.1.40 Knowledge of refueling administrative requirements. (CFR:
41.10 / 43.5 / 45.13) K/A # 2.1.40 Rating 2.8 Rev / Date 0 Question 69 Which of the following describes when continuous communication must be maintained between the Control Room and Refueling Platform?
A. Moving SRM detector from full out position to full in position.
B. Moving Control Rod 32-33 from full in to full out position with an empty cell.
C. Removing the CRD Mechanism from Control Rod 48-33.
D. Moving a Control Rod to perform One-Rod-Out interlock checks during Core Verification.
Answer: D Explanation:
A is wrong. Per Tech Spec Definition movement of source range monitors are not considered to be core alterations.
B is wrong. Per Tech Spec Definition, Control Rod movement provided there are no fuel assemblies in the associated core cell.
C is wrong. Per 03-1-01-5, IOI, 2.15, Continuous communication is required with the Control room and Refuel Floor when changing CRD mechanisms.
D is correct. Per Tech Spec Definition, Movement of any fuel, sources, or reactivity control components within the reactor vessel with the head removed.
Technical
References:
Tech Specs, Definitions, Core Alteration 03-1-01-5, 2.6 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-IOI05, Objective 2.3
Question Source: Bank # 252 X (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 2.2.41 Ability to obtain and interpret station electrical and mechanical drawings. (CFR: 41.10 / 45.12 / K/A # 2.2.41 45.13) Rating 3.5 Rev / Date 0 Question 70 Use your provided references to answer this question.
Consider the control schematic for TBCW Pump A.
One of the 35 amp control power fuses is burned out.
How is electrical control of the pump impacted by this fuse failure?
A. Cannot electrically stop the pump from any location, and automatic tripping is disabled.
B. Can electrically start and stop the pump at the breaker, but not from the control room.
C. Cannot electrically start the pump from any location, but all electrical stopping and automatic tripping are still functional.
D. Can electrically stop the pump at the breaker or from the control room, but automatic tripping is disabled.
Answer: A Explanation:
Refer to electrical drawing E-1227-001. The two 35 amp fuses power all electrical trippinglocal manual stopping, remote manual (control room) stopping, and protective tripping. Thus, having either one of these two fuses blown interrupts continuity of power to all of that.
All distracters are all plausible to the Applicant who has not become proficient at being able to read/understand the control scheme print for a typical 4.16 KV breaker.
NOTE - Concerning the KA match (Tier 3) for this item. Weve chosen the control drawing for a TBCW pump only because it represents the typical control scheme for 4.16 KV pumps, generically. For this reason, we have chosen not to swap the original
KA; i.e., this Exam Author feels that this item as presented meets the intent of ES-401, Section D.2.a (1st paragraph) concerning plant-wide.
A is correct.
Technical
References:
E-1227-001, TBCW Pump A electrical control References to be provided to applicants during exam:
E-1227-001, TBCW Pump A electrical control Learning Objective: GLP-OPS-R2700 Objective 26 Question Source: Bank # 134 X (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 2.2.43 Knowledge of the process used to track inoperable alarms.
K/A # 2.2.43 Rating 3.0/3.3 Revision 0 Question 71 Leak detection computer point E31N16A1 for B21-F022A, Stem Leakoff Temperature, has failed and is causing a nuisance alarm.
This signal to LDS TROUBLE annunciator has been bypassed.
How should the associated annunciator be identified?
A. No identification is required for the alarm window.
B. Place a length of red tape diagonally across the alarm window C. Place two lengths of red tape diagonally across the alarm window to form an X D. Place two vertical lengths of red tape on the alarm window.
Answer: D Explanation:
First thing the student must recognize is that the leak detection stem leakoff temp feed into a multiple point alarm.
If an alarm has multiple inputs then Per 02-S-01-25, 6.3.5, If an annunciator input is bypassed, the Annunciator window is marked as follows:
Two vertical lengths of red tape are applied to distinguish an annunciator with a bypassed input.
A is wrong. See explanation above B is wrong. See explanation above C is wrong. See explanation above D is correct. See explanation above
Technical
References:
02-S-01-25, 6.3.5 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-PROC, Objective 33.4 Question Source: Bank # 612 X (note changes; attach parent) Modified Bank #
New Question History: Last 2 NRC Exams No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) & (10) 55.43(b)
Examination Outline Cross-Reference Level RO 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.
K/A # 2.3.4 Rating 3.2 Rev / Date 0 Question 72 Which of the following is required for a worker to receive 20 Rem TEDE during an emergency?
- 1. Authorization by the Emergency Director or Emergency Plant Manger
- 2. Life Saving
- 3. Protecting Valuable Property
- 4. Only Voluntary A. 1, 2, 3, and 4 B. 1 and 2 only C. 1, 2, and 3 only D. 1 and 3 only Answer: B Explanation:
A is wrong. Only >25 Rem is voluntary B is correct. 10-S-01-17, 6.1 Table.
C is wrong. Protecting property is limited to 10 Rem.
D is wrong. Protecting property is limited to 10 Rem.
Technical
References:
10CFR20 10-S-01-17, 6.1 Table References to be provided to applicants during exam: None
Learning Objective: ENS Generic RadWorker Training, Objective 4.1 Question Source: Bank # 613 X (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(12) 55.43(b)
Examination Outline Cross-Reference Level RO 2.3.13 Knowledge of radiological safety procedures pertaining to licensedl operator duties, such as response to radiation monitor alarms, containment entry requirements, K/A # 2.3.13 fuel handling responsibilities, access to locked high- Rating 3.4 radiation areas, aligning filters, etc.l (CFR: 41.12 / 43.4 / 45.9 /
45.10) Rev / Date 0 Question 73 Per EN-RP-105, Radiological Work Permits (RWP), a radiation worker must participate in an RWP Pre-Job Brief prior to entering A. the RHR C Pump Room.
B. the Aux Building Steam Tunnel.
C. the LPCS Pump Room.
D. the El. 119 Piping Penetration Room.
Answer: B Explanation:
See EN-RP-105, section 5.3[8], 4th bulletVHRA or LHRA entry requires the RWP Pre-Job Brief.
Of the 4 areas among the answer choices, only the Aux Bldg Steam Tunnel (choice B) is a Locked HRA. The other 3 areas are only Radiation Areas, for which the RWPs require no pre-job briefs.
Technical
References:
EN-RP-105, Radiological Work Permits References to be provided to applicants during exam: None Learning Objective: GLP-OPS-PROC, Objective 50.3 Question Source: Bank # 79 X (note changes; attach parent) Modified Bank #
New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X
Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(12) 55.43(b)
Examination Outline Cross-Reference Level RO 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: K/A # 2.4.45 41.10 / 43.5 / 45.3 / 45.12) Rating 4.1 Rev / Date 0 Question 74 A Reactor Scram has occurred.
The CRS has implemented transient alarm response.
Which of the following valid alarms should be immediately reported to the CRS?
A. P680-11A-D7, CRD WTR DIS. O 2 HI B. P870-9A-A4, CTMT-DRWL ISOL DIV 2 OPER C. P807-3A-G3, STATIC INVRTR 1Y98 TROUBLE D. P870-10A-G2, RHR C PMP RM FLOODED Answer: D Explanation:
Per EN-OP-115-08, Annunciator Response, section 9, The announcement of transient alarms during Abnormal/ONEP and EOP is not required. In such cases, the operators are expected to announce those alarms that are of significance to the implementation of the applicable Abnormal/ONEP and EOP.
Per 04-1-01-C82-1, 4.2.2 a (1), Glowing cerise - a color bordering the annunciators encompassing those alarms associated with EP-4 entry conditions.
A is wrong - This alarm is not of significance to the implementation of ONEPs or EPs.
B is wrong - This alarm will come in when an isolation is complete, but, not required to be announced.
C is wrong - This alarm is not of significance to the implementation of ONEPs or EPs.
D is correct - This alarm indicates an entry into EP-4 and should be announced to the CRS.
Technical
References:
EN-OP-115-08 04-1-01-C82-1 References to be provided to applicants during exam: None Learning Objective: GSMS-RO-IN002, Objective B 10 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(12) 55.43(b)
Examination Outline Cross-Reference Level RO 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate K/A # 2.4.47 control room reference material. (CFR: 41.10 / 43.5 / Rating 4.2 45.12)
Rev / Date 0 Question 75 Refer to the next sheet that shows part of Table 10 of EP-4, Auxiliary Building Control.
Operators have entered EP-4.
There is an unisolable steam leak in the RCIC room.
Which of the following situations requires that an Emergency Depressurization be performed?
A. P680-8A1-C4 annuciator, RCIC RM SMP LVL HI-HI with RCIC EQUIP AREA TEMP at 250°F B. RCIC ROOM at 9 x 104 mr/hr with SBGT Filter Train at 80 mr/hr.
C. P680-8A1-C4 annuciator, RCIC RM SMP LVL HI-HI with P870-2A-A1 annuciator, RCIC PMP RM FLOODED D. RCIC Room at 8.5 x 104 mr/hr with Main Steam Line Rad Monitor readings of 1 x 105 mr/hr
Answer: D Explanation:
Per EP-4, step 10, an ED is required only when 2 or max safe values are reached for a
single given parameter (i.e., 2 temps, or 2 water levels, or 2 rad levelsnot combinations among these parameters).
D is correct because both of these rad levels are above their max safe values.
B is wrong RCIC rad levels are above the Max safe, however the SBGT rad levels are well below the Max safe limit by a factor of 10.
A and C are wrong because RCIC RM SMP LVL HI-HI is not a max safe value Technical
References:
EP-4, Aux Building Control References to be provided to applicants during exam:
Table 10 of EP-4, Auxiliary Building Control.
Learning Objective: GLP-OPS-EP4, Objective 3 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(12) 55.43(b)
GGNS LOT 2013 NRC INITIAL LICENSED OPERATOR WRITTEN EXAMINATION SRO EXAM ANSWER KEY 76 A 77 A 78 A 79 C 80 C 81 D 82 B 83 D 84 C 85 B 86 A 87 C 88 B 89 B 90 C 91 A 92 D 93 A 94 D 95 C 96 B 97 B 98 D 99 A 100 A
Examination Outline Cross-Reference Level SRO 295003 Partial or Complete Loss of A.C. Power AA2. Ability to determine and/or interpret the following as K/A # 295003 they apply to PARTIAL OR COMPLETE LOSS OF A.C. Rating 3.5 POWER : (CFR: 41.10 / 43.5 / 45.13)
Rev / Date 0 AA2.03 Battery status: Plant-Specific Question 76 Use your provided references to answer this question.
Bus 17AC was de-energized for two days due to a bus fault.
The fault has been corrected and bus 17AC is now energized.
Battery Charger 1C4 has NOT been re-energized yet.
Division 3 battery 1C3 pilot cell voltage is now 2.11 VDC.
Which of the following describes the OPERABILITY of Division 3 battery 1C3?
A. Battery 1C3 must be declared INOPERABLE.
B. Battery 1C3 may be considered OPERABLE for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> only if average voltage for each connected cell is a minimum of 2.07 VDC.
C. Battery 1C3 may be considered OPERABLE for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> only if average voltage for each connected cell is a minimum of 2.13 VDC.
D. Battery 1C3 may be considered OPERABLE for 31 days for stated conditions if charger 1C5 is in service.
Answer: A Explanation:
With bus 17B01 de-energized, Div 3 battery charger 1C4 is inoperable, requiring entry into TS 3.8.4 A. Charger 1C4 is the only charger that can be credited for TS 3.8.4 for Div 3. This TS action requires verification of category A battery cell parameters to be within limits or TS 3.8.4 B must be entered, which requires declaring the associated
battery inoperable. Cell voltage is given as 2.11 V, which does not meet the minimum category A limit listed in TS table 3.8.6-1; therefore, TS 3.8.4 is not met, and Div 3 battery is inoperable, as reflected by answer A. All other answers are wrong since they state Div 3 battery may be considered operable given certain other conditions, but with the one .
Answers B and C are plausible if the student only considers the requirements of TS 3.8.6 for battery parameters.
Answer D is plausible if the student assumes charger 1C5 can meet operability requirements.
Technical
References:
04-1-01-L11-1 TS 3.8.4 and Bases TS 3.8.6 References to be provided to applicants during exam: TS 3.8.4, TS 3.8.6 Learning Objective: GLP-OPS-TS001, OBJ. 40 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 295006 SCRAM 2.2.39 Knowledge of less than or equal to one hour K/A # 295006 Technical Specification action statements for systems. Rating 4.5 (CFR: 41.7 / 41.10 / 43.2 / 45.13)
Rev / Date 0 Question 77 Tech Spec 3.1.5 Actions require inserting a manual scram in Mode 1 within 20 minutes if charging water header pressure is < 1520 psig and two control rod scram accumulators associated with withdrawn control rods are inoperable.
What is a Tech Spec basis for the allowed completion time of 20 minutes for this action?
A. Allow time to start a CRD pump.
B. Allow time to recharge the inoperable scram accumulators.
C. Allow time to drain water from the inoperable scram accumulator instrument blocks.
D. Allow time to fully insert and disarm the control rods associated with the inoperable scram accumulators.
Answer: A Explanation:
TS 3.1.5 Action B.1 assumes no CRD pumps are running if charging water header pressure is <1520 psig. The completion time for Action B starts when 2 or more scram accumulators for withdrawn control rods are declared inoperable concurrently with charging water header pressure < 1520 psig. If the condition cannot be corrected within 20 minutes, Action B.1 is not met and Action D must be entered, which requires placing the Reactor Mode Switch in SHUTDOWN immediately. The TS Bases states the 20 minute completion time provided for Action B.1 should be adequate for starting a CRD pump, as reflected by Answer A. Answers B and C are plausible because accumulator low pressure and water detection in the accumulator instrument block both cause the CRD accumulator trouble alarm, required operable by TR 3.1.5. But
both answers are wrong. Moisture in the accumulator instrument block does not directly require declaring the accumulator inoperable. Low pressure does require individual accumulators to be declared inoperable, but this is addressed by TS 3.1.5 Actions B.2.1 and B.2.2. Answer D is plausible because inserting and disarming a control rod would be necessary if the rod was declared inoperable, which is an alternative to declaring a control rod slow when its accumulator is inoperable.
However, answer D is wrong because TS 3.1.5 Action B does not address urgency forinserting/disarming a control rod, only for restoring a CRD pump to operation.
Technical
References:
TS 3.1.5 Bases TS 3.1.5 Action B.1 TS 3.1.3 TR 3.1.5 05-1-02-IV-1 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-TS001, OBJ. 39 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7),(10) 55.43(b)(2)
Examination Outline Cross-Reference Level SRO 295018 Partial or Complete Loss of Component Cooling Water K/A # 295018 AA2. Ability to determine and/or interpret the following as Rating 3.4 they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER :
Rev / Date 0 (CFR: 41.10 / 43.5 / 45.13)
AA2.01 Component temperatures Question 78 The plant is at rated power.
CCW Temperature Control Valve P44-F501 has drifted to 10% open due to a failure in the valve positioner.
RWCU Filter Demin inlet temperature is 139°F.
One point, RECIRC PUMP A SEAL COOLING WATER DISCHARGE, on Recirc Temperature Recorder 1B33R601 has just reached its alarm setpoint.
No actions have been taken by the crew.
The CRS should NEXT A. Enter 05-1-02-V-1, Loss of CCW, and direct tripping both RWCU pumps.
B. Enter 05-1-02-V-1, Loss of CCW, and direct placing both RWCU filter Demins in HOLD.
C. Enter 05-1-02-V-11, Loss of PSW, and direct reducing core flow to 70 mlbm/hr.
D. Enter 05-1-02-V-11, Loss of PSW, and direct placing the Reactor Mode Switch to SHUTDOWN.
Answer: A Explanation:
Conditions stated in the stem reflect a partial loss of CCW, as described in ONEP 05-1-02-V-1. All answers are plausible because they reflect actions for varying degrees
of effects of degraded CCW capability. Loss of CCW ONEP is the appropriate procedure for this condition, not Loss of PSW ONEP, since overall PSW capability is not impaired, only cooling to CCW. An RWCU inlet temperature above 130°F necessitates tripping both RWCU pumps, without first placing filters in hold, per ONEP step 3.2.3b. Therefore, answer A is correct and answer B is wrong. Answer C is wrong because the stem asks for the NEXT action that should be taken, and for a partial loss of CCW, power reduction is required only if isolation of major heat loads on CCW, FPCCU and/or RWCU, does not result in stabilizing Recirc pump/motor temperatures. Answer D is wrong because the point in alarm on recorder 1B33R601 is a cooling water temperature, not a pump/motor metal or seal cavity temperature.
Per the note at ONEP section 2.0, a Recirc pump/motor metal or seal cavity temperature alarm would require declaration of a total loss of CCW, which would necessitate a manual scram.
Technical
References:
05-1-02-V-1 04-1-02-1H13-P680-3A-A8 04-1-02-1H13-P680-11A-D6 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-ONEP, OBJ. 3, 34 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 295026 Suppression Pool High Water Temperature 2.2.25 K/A # 295026 / 2.2.25 Knowledge of the bases in Technical Specifications for Rating 4.2 limiting conditions for operations and safety limits.
(CFR: 41.5 / 41.7 / 43.2)
Rev / Date 0 Question 79 One basis for the Tech Spec ACTION to immediately suspend RCIC flow testing based on high Suppression Pool temperature is.
A. Elevated RCIC turbine lube oil temperature that could damage RCIC turbine.
B. RCIC turbine exhaust check valve chatter that can damage RCIC piping.
C. Ensure the Containment temperature limit is not exceeded.
D. Preclude having to enter EP-3, Containment Control.
Answer: C Explanation: In the stem, RCIC flow testing implies power is above 1%. TS 3.6.2.1 Action C.1 requires immediately suspending testing that adds heat to the Supp Pool when Supp Pool average temperature exceeds 105°F. The implication in the stem is Supp Pool temperature has just exceeded 105°F. TS bases 3.6.2.1 states the Supp Pool temperature limit was developed to address technical concerns that include limiting containment average air temperature to < 185°F during the DBA; therefore, answer C is correct. Answer A is plausible because it describes undesired effects of elevated Supp Pool temperature on RCIC. It is each wrong because 04-1-01-E51-1 step 3.3 states RCIC suction temperature is limited to 140°F for non-emergency operation to prevent bearing/turbine seals overheating, and during emergencies, EP caution 4 states RCIC equipment damage could occur if Supp Pool temperature exceeds 225°F, well above the 105°F condition implied in the stem and because RCIC operability is not a concern in the DBA analysis and is not mentioned in the bases for TS 3.6.2.1. Answer B is plausible because exhaust check valve operation could theoretically be affected mechanistically by Supp Pool parameters, but it is wrong because 04-1-01-E51-1 step 3.2 states this is a concern for operation of RCIC
at speeds below 2000 rpm and because RCIC operability is not a concern in the DBA analysis and is not mentioned in the bases for TS 3.6.2.1. Answer D is plausible because entry into an Emergency Procedure is undesirable and implies elevated risk and challenge to plant safety, but it is wrong because TS 3.6.2.1 Action C that requires suspending testing that is adding heat to the Supp Pool at 105°F, so the Supp Pool Temperature EP-3 entry condition of 95°F would have already been met, so EP-3 entry cannot be avoided.
Technical
References:
TR 3.6.2.1 and bases EP-3 EP Caution 4 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-TS001, OBJ. 39 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(5,7) 55.43(b)(2)
Examination Outline Cross-Reference Level SRO 295028 High Drywell Temperature EA2. Ability to determine and/or interpret the following as K/A # 295028 they apply to HIGH DRYWELL TEMPERATURE : Rating 4.1 (CFR: 41.10 / 43.5 / 45.13)
Rev / Date 0 EA2.03 Reactor water level Question 80 A steam leak in the drywell occurred at rated power.
All but one control rod fully inserted on the resulting scram.
Due to instrument failures, the only Reactor Water Level indicator that is functioning is Upset range, which is indicating +55 inches, slowly trending down.
Drywell temperature is 196°F, slowly trending up.
Containment pressure is 4 psig, slowly trending up.
The CRS should NEXT A. direct overriding Low Pressure ECCS systems.
B. direct terminating injection from all sources outside of primary containment.
D. Exit EP-2A and enter EP-5A.
Answer: C Explanation: EP Caution 1 states Upset range water level instrumentation may not be used if drywell temperature at elev. 166 is above 195°F and indicated Upset level is <159 inches. The stem states all other water level indication is unavailable, so the stem stating drywell temperature is above 195°F indicates Upset level indication cannot be used, and now all level indication is unavailable. Stating drywell temperature, in general, is above 195°F implies temperature at elev. 166 is above
195°F, since upper elevations will always be the hottest. Conditions in the stem, that a scram has resulted due to a leak in the drywell, imply that either EP-2 or EP-2A would have been entered. Since all but one control rod are fully inserted, EP-2 must have been entered. EP-2 step 3 requires exiting EP-2 and entering EP-5; therefore, only answer C is correct. Answer A is plausible because the indicated level given is high, above the top of the allowed control band in EP-2, 53.5 inches, and 02-S-01-27, Operations Philosophy, requires terminating injection to get back into band if level is actually high, and EP-2 step ED-2 actually requires preventing low pressure ECCS injection if not desired, but here, level has to be assumed to be unknown. Answer B is plausible for the same reasons as answer A, plus EP-2 step L-2 has action that states terminate injection from outside Containment, though it is not applicable here. Answer D is plausible if the student believes an operation is in EP-2A due to one control rod being stuck out, but the reactor is considered shutdown based on shutdown margin.
Technical
References:
EP-2 EP-5 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-EP02, OBJ. 3,4,7; GLP-OPS-EP5, OBJ. 5 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown K/A # 295037 EA2. Rating 4.1 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR Rev / Date 0 POWER ABOVE APRM DOWNSCALE OR UNKNOWN :
(CFR: 41.10 / 43.5 / 45.13)
EA2.04 Suppression pool temperature Question 81 Use the figure on the next page to answer this question.
An ATWS is in progress.
- Reactor power is 7%.
- Reactor Water Level is -130 inches on wide range.
- Reactor Pressure is 910 psig.
- Suppression Pool temperature is 160°F.
- Suppression Pool level is 18.5 feet..
The CRS should NEXT A. Direct terminating injection to a new RPV level band of -191 inches to -167 inches.
B. Direct lowering reactor pressure to 450 psig to 600 psig.
C. Direct fully opening Main Bypass Valves in anticipation of Emergency Depressurization.
D. Enter the Emergency Depressurization leg.
Answer: D Explanation: With suppression pool temperature 160°F, suppression pool level 18.5 ft, and RPV pressure 910 psig, operation is in the unsafe zone of the HCTL curve. EP-3 step SPT-5 requires emergency depressurization, which is the overriding priority, as stated in EP-2A step P-1. Therefore, answer D is correct, and other answers are wrong since EP-2A step P-1 requires exiting level and pressure legs at that point since they would delay emergency depressurization.
Answer A is plausible because with suppression pool temperature above 110°F
and power above 5%, EP-2A step L-5 requires performance of step L-8 to lower level to reduce power. Answer B is plausible because one strategy to avoid the unsafe zone of HCTL is to reduce reactor pressure, as stated in EP-2A step P-1. But EP-3 step SPT-5 does not accommodate restoring operation to the safe zone of the curve, as does the PSP curve. Once in the unsafe zoneof HCTL, it is too late for that strategy. Answer C is plausible because for non-ATWS situations where emergency depressurization is inevitable, the strategy is to dump the maximum energy possible to the main condenser. But this is only stated in EP-2 step P-1.
Technical
References:
EP-2A EP-3 EP-2 References to be provided to applicants during exam: HCTL curve Learning Objective: GLP-OPS-EP02A, OBJ. 22, GLP-OPS-EP3 Obj. 22 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 700000 Generator Voltage and Electric Grid Disturbances 2.1.20 Ability to interpret and execute procedure steps.
K/A # 700000 (CFR: 41.10 / 43.5 / 45.12) Rating 4.6 Rev / Date 0 Question 82 Refer to the surveillance data sheet on the next page.
The plant is at 40% power during start up.
05-1-02-I-4, Loss of AC Power has been entered for grid instability due to grid voltage fluctuations.
06-OP-1R20-W-0001, Plant AC and DC Electrical Power Distribution Weekly Lineup, has been performed in accordance with the ONEP.
Based on this data, the CRS should A. Initiate a Potential LCO, only, for the offsite supply from Baxter Wilson and Franklin lines being inoperable for TS 3.8.1 Condition A.
B. Initiate a Potential LCO, only, for the offsite supply from the 115KV Port Gibson line being inoperable for TS 3.8.1 Condition A.
C. Initiate an actual TS LCO for the offsite supply from Baxter Wilson and Franklin lines being inoperable for TS 3.8.1 Condition A. Perform TS 3.8.1 Action A.1 as specified and restore the sources operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
D. Initiate an actual TS LCO for the offsite supply from the 115KV Port Gibson line being inoperable for TS 3.8.1 Condition A. Perform TS 3.8.1 Action A.1 as specified and restore the sources operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
GRAND GULF NUCLEAR STATION SURVEILLANCE PROCEDURE 06-OP-1R20-W-0001 Revision 109 Attachment I Page 2 of 12 Page XRef DATA SHEET I PLANT AC AND DC ELECTRICAL POWER DISTRIBUTION WEEKLY LINEUP SAFETY RELATED (Step 5.1.2) Div 1, 2 & 3 Offs ite Feeders OFFSITE ENERGIZED VOLTAGE RECORDED FREQUENCY RECORDED INITIALS FEEDER INDICATOR VOLTAGE INDICATOR FREQUENCY YES/NO (LOCATION) (ACCEPTANCE (LOCATION) (ACCEPTANCE CRITERIA) CRITERIA)
BAXTER 500 kV FREQ
$ WILSON
- YES JACKSON 494 kV SR27-SR-R600 60.5 Hz DISPATCHER (H13-P807)
$ FRANKLIN
- YES (496-525kV)*** or Pine Bluff (58.5-61.8Hz)
Dispatcher
- 4.05 x 27.64 152-1511 = 111.94 kV 115kV LINE 152-1611 (120.75-
$ PORT
- YES 152-1704 112.13) kV GIBSON
- To determine status of offsite feeders, CONTACT load dispatcher. ENSURE that the feeders are independently energized from the grid, such that the loss of one feeder would NOT result in the loss of another
- Allowable Value of minimum voltage is >491 kV for operability of Offsite Feeders. This value is based on analysis of the Class 1E ESF buses AND includes an allowance for instrument uncertainty associated with the voltage measurement in the switchyard. Extended operation beyond the normal continuous operating limits Should be evaluated AND caution Should be taken when starting large loads under these conditions.
Answer: B
Explanation: The value listed for the 115KV Port Gibson line, 111.94KV, is less than the minimum required for operability, 112.13KV; therefore it is inoperable. The value given for the 500KV Baxter and Franklin lines, 494KV is less than the nominal minimum, 496KV, but it is greater than the operability limit, 491KV, listed in note***. Therefore, the 500KV lines meet operability requirements.
This alone makes answer A inicorrect. TS 3.8.1 requires two offsite supplies operable, and since the 500KV feeds are operable, the LCO is met. This makes answers C and D wrong since they imply the LCO is not met. Since the 115KV source is inoperable, only a Potential LCO should be initiated as defined in 02-S-01-17 step 5.1.2b; therefore, answer B is correct. Answer A and C are plausible if the student does not know meeting the allowable value is sufficient for operability. Answer D is plausible if the student thinks inoperability of one offsite source that can in part satisfy LCO 3.8.1 constitutes failure to meet the LCO.
Technical
References:
TS 3.8.1 06-OP-1R20-W-0001 Att I ENS-DC-199 02-S-01-17 References to be provided to applicants during exam: partially completed 06-OP-1R20-W-0001 Att I page 2 (above); TS 3.8.1 first 2 pages Learning Objective: GLP-OPS-TS001, OBJ. 40, GLP-OPS-TSLCO Obj. 3, 7 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 295020 Inadvertent Containment Isolation AA2.
K/A # 295020 Ability to determine and/or interpret the following as Rating 3.4 they apply to INADVERTENT CONTAINMENT ISOLATION :
(CFR: 41.10 / 43.5 / 45.13) Rev / Date 0 AA2.02 Drywell/containment temperature.
Question 83 Use your provided references to answer this question.
The coil for relay 1M71-R65 fails, resulting in inadvertent closure of Division 2 Drywell Chilled Water containment isolation valves.
Power has been reduced to 70%.
Average drywell temperature has risen to 145°F.
CRD Cavity temperature is 155°F.
All temperatures at other drywell elevations are below 149°F.
What is the limiting action statement at this point?
A. Declare equipment within the drywell inoperable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
B. Restore drywell temperature to within limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or then declare equipment within the drywell inoperable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
C. Restore drywell temperature to within limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or immediately initiate action to provide a record of the cumulative time the limit was exceeded and an analysis to demonstrate continued operability of equipment within the drywell.
D. Restore drywell temperature to within limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Answer: D
Explanation: All answers are plausible, because both TS 3.6.5.5 and TR 6.7.3 impose limits on either average drywell temperature or local drywell temperatures. TS 3.6.5.5 requires drywell average temperature to be maintained 135°F. TR 6.7.3 additionally limits CRD Cavity temperature to 185°F and other local drywell temperatures to 150°F. Only TS 3.6.5.5 limit has been exceeded. Answer D reflects TS 3.6.5.5 Actions; therefore it is correct. Answers A, B, and C reflect TR 6.7.3 actions, and since LCO TR 6.7.3 is still met, no TR 6.7.3 actions are required, these answers are wrong.
Technical
References:
TS 3.6.5.5 TR 6.7.3 06-OP-1000-W-0001 Att. I, data sheet II References to be provided to applicants during exam: TS 3.6.5.5 and TR 6.7.3 Learning Objective: GLP-OPS-TS001, OBJ. 40 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 295035 Secondary Containment High Differential Pressure EA2.
K/A # 295035 Ability to determine and/or interpret the following as they Rating 3.9 apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:
Rev / Date 0 (CFR: 41.8 to 41.10)
EA2.01 Secondary containment pressure: Plant-Specific Question 84 Use your provided references to answer this question.
The plant is at rated power.
SGTS A is operating for performance of 06-OP-1T48-M-0001, Standby Gas Treatment System A Operability.
Five hours into the SGTS A surveillance, personnel using door 1A401, elevation 166 ft Turbine Bldg. to Auxiliary Bldg. equipment door, report the door has jammed in the open position.
Auxiliary Bldg D/P rises to -0.05 w.c.
What is the Tech Spec completion time for having to be in Mode 3 if this condition continues?
A. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B. 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
C. 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.
D. 7 days, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Answer: C Explanation:
Secondary Containment Integrity surveillance requirements SR 3.6.4.1.3 and SR 3.6.4.1.4 require that Secondary Containment be leak tight enough that one SGTS can draw down Secondary Containment pressure to -0.25 w.c. within 180 seconds and maintain -0.266 w.c. for 1 continuous hour. The listed SGTS surveillance tests both proper operation of SGTS and leak tightness of Secondary Containment. For the
stated conditions, SGTS is unable to maintain the required DP, due to excessive Aux.
Bldg. leakage, not due to a failure of SGTS. Therefore, the applicable TS action is TS 3.6.4.1, Action A.1, restore Secondary Containment operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or enter TS Condition B and be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; so Mode 3 would have to be attained within 16 total hours if the problem is not resolved. This is reflected in correct answer C.
Answer A is plausible if the student incorrectly believes both SGTS are rendered inoperable, incorrectly enters TS 3.6.4.3 Action D, and then misapplies TS 3.0.3, allowing only 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to be in Mode 3 versus the 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> allowed by TS 3.0.3.
Answer B is plausible if the student believes both SGTS are rendered inoperable, incorrectly enters TS 3.6.4.3 Action D, and then applies TS 3.0.3, which allows 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> to be in Mode 3.
Answer D is plausible if the student incorrectly believes only SGTS A is rendered inoperable, since it is the subsystem in operation, and then applies TS 3.6.4.3 Condition A, projecting Condition B.
Technical
References:
TS 3.6.4.1, Secondary Containment TS 3.6.4.3, SGTS References to be provided to applicants during exam: TS 3.6.4.1, TS 3.6.4.3 (including surveillance requirements)
Learning Objective: GLP-OPS-TS001, OBJ. 40 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 295036 Secondary Containment High Sump/Area Water Level 2.2.42 K/A # 295036 / 2.2.42 Ability to recognize system parameters that are entry-level Rating 4.6 conditions for Technical Specifications.
(CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3) Rev / Date 0 Question 85 The plant is in Mode 1.
RHR C room sump pumps are tagged out of service for discharge check valve replacement.
A PSW leak from ADHR heat exchangers results in flooding RHR C pump room with two feet of water.
RHR C Jockey Pump remains operating.
For the initial operability screening of the associated Condition Report, Engineerings judgment based on industry operating experience is that RHR C Jockey Pump motor winding insulation may fail at any time due to submersion.
For the immediate operability determination, the SRO should A. Screen the CR as Non-Functional, and immediately enter TS 3.5.1.
B. Screen the CR as Inoperable, and immediately enter TS 3.5.1.
C. Screen the CR as Functional, and enter TS 3.5.1 after 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or upon failure of RHR C Jockey Pump, whichever come first.
D. Screen the CR as Operable, and enter TS 3.5.1 after 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or upon failure of RHR C Jockey Pump, whichever come first.
Answer: B Explanation: There is no direct LCO entry based on Secondary Containment water levels. However, the effect of high water level in a Secondary Containment area can pose operability issues, requiring an immediate operability determination per EN-OP-104, Operability Determination. RHR C Jockey Pump is required for LPCI C OPERABILITY. In this case, the student must recognize Engineering Judgment does not provide a Reasonable Expectation of Operability. EN-OP-104 defines Reasonable Expectation as a high standard and states the supporting basis should provide a high degree of confidence that the SSC remains OPERABLE. Definition 3.0[24] states when system capability is degraded to a point where it cannot perform with Reasonable Expectation or reliability, the system should be judged INOPERABLE. If RHR C may be expected to fail at any time, RHR C may not be considered operable, since ECCS acceptance criterion 5 for long-term cooling cannot be guaranteed. Per EN-OP-104 definitions, the appropriate CR operability coding for equipment listed in Tech Specs is OPERABLE/INOPERABLE versus FUNCTIONAL/NON-FUNCTIONAL, which applies to equipment only required by TRM, E-Plan, Security System, etc. With RHR C inoperable, TS 3.5.1 Condition A must be entered immediately. Therefore, answer B is correct.
Answer A is plausible if the student does not know the required distinction for TS equipment versus other required equipment defined in EN-OP-104, but it is wrong since RHR C Jockey Pump is required by TS.
Answer C is plausible if the student does not consider equipment qualification with respect to operability and only considers that RHR C Jockey Pump continues to run.
But it is wrong because the code FUNCTIONAL cannot be applied to TS equipment and because there is no Reasonable Expectation of Operability as defined in EN-OP-104.
Answer D is plausible if the student does not consider equipment qualification with respect to operability and only considers that RHR C Jockey Pump continues to run.
Technical
References:
EN-OP-104 TS 3.5.1 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-PROC Obj. 42.1, 42.3; GLP-OPS-TS001, OBJ. 40 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7,10) 55.43(b)(2,3)
Examination Outline Cross-Reference Level SRO 203000 RHR/LPCI: Injection Mode (Plant Specific) 2.1.19 K/A # 203000 / 2.1.19 Ability to use plant computers to evaluate system or Rating 3.8 component status.
(CFR: 41.10 / 45.12)
Rev / Date 0 Question 86 Refer to the attached figure.
EP-2 and EP-3 are being executed.
RHR A and B are available.
All other injection systems are unavailable.
Operation is in the Unsafe Zone of the PSP curve.
Operation is in the Unsafe Zone of the HCTL curve.
Emergency Depressurization is in progress.
How should RHR A and B be aligned for the plant conditions depicted on the attached SPDS display?
A. Align both RHR A and B for LPCI injection via 1E12F042A and B.
B. Align both RHR A and B for Containment Spray.
C. Align both RHR A and B for Suppression Pool Cooling.
D. Align RHR A for injection via 1E12F053A and align RHR B for Containment Spray.
Answer: A Explanation: Adequate core cooling is the priority objective of the EPs. For the conditions represented on SPDS, adequate core cooling does not exist. Level is below the Minimum Zero Injection RPV Water Level, -204 and Minimum Steam Cooling Water Level, -191. EP-2 step L-14 directs maximizing RPV injection with all available systems due to RPV level <-191, depicted by Fuel Zone (FZ) level on SPDS. Here, only RHR A and B are available. Therefore, answer A is correct.
Answer B is plausible because the PSP curve has been exceeded. It is wrong because EP-3 step PCP-5 states only initiate Containment Spray with systems not required for adequate core cooling. Since only RHR A and B are available for injection, adequate core cooling cannot be assured by other means, so RHR A and B should be used for injection, not Containment Spray. Therefore, answer B is wrong.
Answer C is plausible because the HCTL curve has been exceeded. It is wrong because EP-3 step SPT-5 states only initiate Suppression Pool Cooling with systems not required for adequate core cooling. Since only RHR A and B are available for injection, adequate core cooling cannot be assured by other means, so RHR A and B should be used for injection, not Suppression Pool Cooling. Therefore, answer C is wrong.
Answer D is plausible because the PSP curve has been exceeded. It is wrong because EP-3 step PCP-5 states only initiate Containment Spray with systems not required for adequate core cooling. Since only RHR A and B are available for injection, adequate core cooling cannot be assured by other means, so RHR A and B should both be used for injection until adequate core cooling is achieved. Therefore, answer B is wrong Technical
References:
EP-2, EP-3 02-S-01-27, Operations Philosophy References to be provided to applicants during exam: none Learning Objective: GLP-OPS-EP02, OBJ. 8; GLP-OPS-EP3, OBJ. 8, Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X
Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10)
Examination Outline Cross-Reference Level SRO 211000 Standby Liquid Control System A2.
K/A # 211000 Ability to (a) predict the impacts of the following on the Rating 3.2 STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or Rev / Date 0 mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.07 Valve closures Question 87 The plant is at rated power.
06-OP-1C41-Q-0001, Standby Liquid Control Functional Test, is being performed with SLC A pump running, circulating the SLC Test Tank.
A Loss of Offsite Power occurs.
An ATWS due to multiple stuck control rods exists with power 10%.
Immediate Containment evacuation is directed, and SLC pump A continues to circulate the SLC Test Tank.
(1) What effect will these conditions have on SLC system B injection to the RPV?
(2) What direction should be given by the CRS concerning Power Control for EP-2A?
A. (1) SLC B will inject SLC Boron Tank contents, only.
(2) Trip SLC pump B when SLC tank level drops to 2000 gallons.
B. (1) SLC B will inject SLC Boron Tank contents, only.
(2) Trip SLC pump B when SLC tank level drops to 0 gallons.
C. (1) SLC B will inject SLC Test Tank contents, only.
(2) Direct installation of Attachment 28, Alternate SLC Injection.
D. (1) SLC B will inject SLC Boron Tank contents diluted with SLC Test Tank contents.
(2) Direct installation of Attachment 28, Alternate SLC Injection.
Answer: C
Explanation: During the listed surveillance, SLC pump suction from SLC Test Tank valve 1C41F031 is open. Limit Switches on F031 provide a open permissive to SLC pumps boron tank suction valves 1C41F001A and B and provide a separate start permissive to SLC pumps. With 1C41F031, neither 1C41F001A or B will open when the respective SLC pump is initiated from control room panel 1H13P601, although the respective SLC pump will start since it sees a suction path available. In the situation given, SLC pump B will immediately start with suction from the SLC Test Tank only, and neither suction from the SLC boron tank will open. SLC B injection valve will fire when SLC B is initiated; therefore, SLC B will start and pump SLC Test Tank to the reactor. Since SLC will not inject boron under these conditions, EP-2A step Q-4 requires Att. 28 for alternate boron injection. For these reasons, answer C is correct.
All other answers are plausible since they either pertain to SLC B being functional and include actions that would follow successful SLC injection, as in answers A and B, or they pertain to a disfunctional SLC B and include the associated contingency of Att.
28, as in answer D. Answers A, B and D are fundamentally wrong since they imply SLC boron tank contents would be injected, purely or diluted, as if 1C41F001A/B would open, but F001A/B would not open if F031 is open.
Technical
References:
EP-2A EP Att. 28 E1169-01, 02, 12, 14 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-EP02A, OBJ. 8; GLP-OPS-C4100 OBJ. 10 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(5)
Examination Outline Cross-Reference Level SRO 215005 Average Power Range Monitor/Local Power Range Monitor System K/A # 215005 A2. Rating 3.7 Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER Rev / Date 0 RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.02 Upscale or downscale trips Question 88 The plant is at rated power.
APRM 4 fails to 120%.
The crew places APRM 4 in BYPASS on 1H13P680.
(1) What is the impact of APRM 4 failing to 120%?
(2) In addition to placing APRM 4 in BYPASS on 1H13P680, which of the following are other actions required per plant procedures?
A. (1) Half-trip signal to voters, only. No half-scram.
(2) Initiate a Technical Specification (TS) LCO tracking report in ESOMS for APRM4.
B. (1) Half-trip signal to voters, only. No half-scram.
(2) Initiate a Potential Tech Spec (PTS) LCO tracking report in ESOMS for APRM4.
C. (1) Division 2 half-scram.
(2) Reset the half scram, and initiate a Technical Specification (TS) LCO tracking report in ESOMS for APRM4.
D. (1) Division 2 half-scram.
(2) Reset the half scram, and initiate a Potential Tech Spec (PTS) LCO tracking report in ESOMS for APRM4.
Answer: B
Explanation: This relates to the new PRNM system installed during refueling outage 18 in 2012. Eight APRM channels feeding 2 trip systems were replaced by 4 digital APRM channels feeding a single, VOTER logic trip system. For a single APRM trip, no half scram is produced, but only one vote. Any 2 APRM channels will now produce a full scram, not a half-scram. No single APRM channel will produce a half-scram. This is different than all other RPS logic, which requires a trip in each of 2 trip systems to produce a full scram and where any channel tripping produces a half-scram. TS 3.3.1.1 and TR 3.3.2.1 require only 3 of 4 APRM channels operable in Mode 1. No action statement is entered for only one APRM inoperable. For this situation, 02-S-01-17, Control of Limiting Conditions of Operation requires a Potential LCOTR to be initiated to track the inoperability of equipment potentially required operable so that if another APRM goes inoperable, it would be recognized that 2 APRMS are inoperable and the appropriate LCO conditions entered. Since no half-scram is produced and a PTSLCO is proper, answer B is correct. Other answers are plausible because they each include variations of the concepts discussed above.
Answer A is wrong because it implies actual entry into a TS condition statement.
Answers C and D are wrong because they imply a half-scram was produced by the APRM failure. Additionally, answer C is wrong because it implies actual entry into a TS condition statement.
Technical
References:
TS 3.3.1.1 TR 3.3.2.1 02-S-01-17 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-C5100 Obj. 3.5, 7.2, 14; GLP-OPS-TS001, OBJ. 40; GLP-OPS-TSLCO OBJ. 7 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(5)
Examination Outline Cross-Reference Level SRO 259002 Reactor Water Level Control System 2.4.34 K/A # 259002 / 2.4.34 Knowledge of RO tasks performed outside the main control Rating 4.1 room during an emergency and the resultant operational effects.
Rev / Date 0 (CFR: 41.10 / 43.5 / 45.13)
Question 89 The plant is at rated power.
An event occurs that requires Control Room evacuation.
The CRS implementing 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, directs the RO to open Reactor Protection System breakers:
- CB2A, CB5A, CB7A, CB8A
- CB2B, CB5B, CB7B, CB8B (1) For what situation does the CRS direct this action?
(2) What effect will this action have on reactor level control using Feedwater?
A. (1) Control room fire.
(2) Feedwater will NOT be affected by this RO action and will automatically maintain level in the normal band unless disabled by the fire.
B. (1) Control room fire.
(2) Feedwater will become unavailable due to this RO action.
C. (1) Security event.
(2) Feedwater will NOT be affected by this RO action and will automatically maintain level in the normal band unless disabled by the fire.
D. (1) Security event.
(2) Feedwater will become unavailable due to this RO action.
Answer: B Explanation: For evacuation of the Control Room due to a fire, 05-1-02-II-1 step 3.1.4 specifically requires an RO to perform Att. XXII. The first section of this
procedure requires opening RPS breakers CB-2A,5A,7A,and 8A on panel 1C71P001 in RPS A MG set room located on 189 elevation, Control Bldg and CB-2B,5B,7B,and 8B on panel 1C71P002 in RPS B MG set room located on 148 elevation, Control Bldg. The CB-2A(B) and 8A(B) result in deenergization of RPS A scram solenoids.
The CB-5A(B) and 7(A) deenergize Division 1 and 2 MSIV solenoids. This does not directly affect Feedwater Level Control, but it does result in MSIV closure, which isolates steam to the reactor feed pump turbines, thus disabling Feedwater system.
Since the breakers are in the control building and Feedwater is disabled, answer B is correct and all other answers are wrong. Answer A is plausible because the breakers do not directly affect Feedwater level control. Answers C and D are plausible because there are many actions taken outside of the main control room associated with certain security events, but not these specific actions.
Technical
References:
05-1-02-II-1 step 3.1.4 and Att. XXII References to be provided to applicants during exam: none Learning Objective: GLP-OPS-N2100 Obj. 36; GLP-OPS-C7100 Obj. 4.1, 23; GLP-OPS-C3400, OBJ. 22 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 400000 Component Cooling Water System (CCWS) 2.4.2 K/A # 400000 / 2.4.2 Knowledge of system set points, interlocks and automatic Rating 4.6 actions associated with EOP entry conditions.
(CFR: 41.7 / 45.7 / 45.8) Rev / Date 0 Question 90 Use your provided references to answer this question.
(1) Which one of the following signals, including its setpoint, is common to both RPS instrumentation and Containment Isolation instrumentation for Drywell Chilled Water System valves and is an Emergency Procedure entry condition?
(2) What is the completion time for placing the channel in trip if only one channel of this instrumentation is inoperable?.
A. (1) Reactor Water Level Low.
(2) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B. (1) Reactor Water Level Low.
(2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C. (1) Drywell Pressure High.
(2) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D. (1) Drywell Pressure High.
(2) 24hours.
Answer: C Explanation: Drywell Chilled Water System isolation valves 1P72F121,F122,F123,F124,F125,F126 are part of Group 6. Drywell Pressure High from trip units 1C71N650A,B,C,D is common to both RPS and Group 6 containment isolation instrumentation. With a nominal setpoint of 1.23 psig, it is the basis for EP-2 and EP-3 entry conditions. With one channel of this instrumentation inoperable, TS 3.3.1.1 Action A.1 applies and has a completion time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Also, TS 3.3.6.1 Action A.1 applies, and since this represents TS 3.3.6.1 function 2b, also has a completion time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, as it is for all containment isolation signals that are common to RPS. Therefore, answer C is correct. Answer A and B are plausible because Reactor Level Low (level 3) is a signal common to TS 3.3.1.1, TS 3.3.6.1, and an EP-2 entry condition. But it is wrong in this case because Level 3 is not
specifically common to Group 6, but only Groups 2 and 3. Answer D is plausible because the completion time one channel inoperable of TS 3.3.6.1 instrumentation that is not common to RPS is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per TS 3.3.6.1 Action A.1. Though the associated TS instrumentation tables are not to be provided for this question, the SRO student should recognize the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> completion listed for TS 3.3.6.1 Action A is for instrumentation common to RPS and the TS 3.3.6.1 Action A completion time is and is designed to be consistent with the TS 3.3.1.1 Action A.1 completion time.
Technical
References:
TS 3.3.1.1 TS 3.3.6.1 EP-2 EP-3 17-S-06-5 References to be provided to applicants during exam: TS 3.3.1.1 first page (Condition A), TS 3.3.6.1 first page (Condition A)
Learning Objective: GLP-OPS-TS001 Obj. 40 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)
Examination Outline Cross-Reference Level SRO 201003 Control Rod and Drive Mechanism A2.
K/A # 201003 Ability to (a) predict the impacts of the following on the Rating 3.1 CONTROL ROD AND DRIVE MECHANISM ; and (b) based on those predictions, use procedures to correct, control, or Rev / Date 0 mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.06 Loss of CRD cooling water flow Question 91 The plant is at rated power.
Control Rod 32-22 is required to be isolated for maintenance.
04-1-01-C11-1, Control Rod Drive Hydraulic System, section 5.7, Total Isolation of an HCU, is performed for the protective tagging boundary.
(1) Which of the following is a possible consequence of this performing this section of the SOI at rated power?
(2) What is a mitigating action required for this consequence?
A. (1) CRD Mechanism drive seal degradation due to seal temperature above 450°F.
(2) Assign an action to engineering in PCRS to evaluate scram time.
B. (1) CRD Mechanism drive seal degradation due to seal temperature above 450°F.
(2) Assign an action to engineering in PCRS to evaluate HCU Accumulator pre-charge pressure.
C. (1) CRD Mechanism drive seal degradation due to constant application of charging header pressure while the tagout is hanging.
(2) Assign an action to engineering in PCRS to evaluate scram time.
D. (1) CRD Mechanism drive seal degradation due to constant application of charging header pressure while the tagout is hanging.
(2) Assign an action to engineering in PCRS to evaluate HCU Accumulator pre-charge pressure.
Answer: A Explanation: 04-1-01-C11-1, Control Rod Drive Hydraulic System, section 5.7, Total Isolation of an HCU, results in isolation of CRD cooling water to the CRDM.
Precaution 3.9 and 3.11 of 04-1-01-C11-1 warn against this because high CRDM temperature will result if RCS temperature is above 250°F. Alarm 1H13-P680-4A2-A4, CRD HYD TEMP HI would be expected for reduced cooling water flow to a CRDM at rated conditions. ARI 04-1-02-1H13-P680-4A2-A4 step 4.4 states if CRD temperature reaches 400°F, a CR should be initiated for tracking the condition by Engineering. ARI step 3.3 also states Engineering should be contacted to evaluate adding time to the previous scram time if the CRDM temperature reaches 450°F. The administrative mechanism for Engineering evaluations related to operability is the CR process in PCRS. Therefore, answer A is correct. Answer B is plausible because the listed consequence is correct and because HCU Accumulator pressure can mechanistically affect scram time. However, it is wrong because there is no process or requirement to vary HCU Accumulator pressure to quicken or reduce scram time.
Answers C and D are plausible because seal damage is a concern if insert riser valve 101XX is reopened before withdraw riser valve 102XX per 04-1-01-C11-1 step 3.7.
Seal damage is also a concern per 04-S-03-C11-4 if a control rod is scrammed from position 10 or less. Similarly for single rod scram, seal damage might concern the student who does not know the piping configuration for an HCU or that HCU isolation includes isolation of insert riser valve 101XX and withdraw riser valve 102XX, such that the CRDM is isolated from CRD charging water header. The second parts of answers C and D relate to seal health and are, therefore, plausible.
Technical
References:
04-1-01-C11-1 04-1-02-1H13-P680-4A2-A4 04-S-03-C11-4 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-C111B Obj. 4.4,8.1,8.2,9 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X
Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(5)
Examination Outline Cross-Reference Level SRO 219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode A2. Ability to (a) predict the impacts of the following on the K/A # 219000 RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE; Rating 3.2 and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those Rev / Date 1 abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.03 Valve closures Question 92 The plant is at rated power.
- RHR pump A is tagged out of service for maintenance.
- Safety Relief Valve B21-F041F fails open.
- RHR B is placed into Suppression Pool Cooling mode.
- A spurious Division 2 ECCS initiation signal is received.
- The RHR B/C Initiation signal can NOT be reset.
- Suppression Pool temperature is 94°F.
(1) What is the status of RHR B for Suppression Pool Cooling mode?
(2) What action should be taken NEXT by the CRS for current conditions?
A. (1) Interlocks prevent opening 1E12F024B, RHR B TEST RTN TO SUPP POOL.
(2) Direct placing SRV hand switches to OFF for B21-F041F.
B. (1) Interlocks prevent opening 1E12F024B, RHR B TEST RTN TO SUPP POOL.
(2) Direct placing the Reactor Mode Switch to SHUTDOWN.
C. (1) 1E12F024B, RHR B TEST RTN TO SUPP POOL can be opened from 1H13P601.
(2) Direct placing the Reactor Mode Switch to SHUTDOWN.
D. (1) 1E12F024B, RHR B TEST RTN TO SUPP POOL can be opened from 1H13P601.
(2) Direct placing SRV hand switches to OFF for B21-F041F.
Answer: D Explanation: A division 2 ECCS signal will cause RHR B Return to Supp Pool valve E12-F024B to close, interrupting RHR B Supp Pool Cooling flow. The automatic valve closure can be manually overridden by placing the 1H13P601 hand switch for E12-F024B to Open. Therefore, RHR B Supp Pool Cooling can be aligned.
Ops Philosophy allows ROs to take actions without direction for a limited number of conditions. In the case above ROs and SROs should have the system knowledge required to know placing the SRV hand switches in off will (or should) cause the SRV to de-energize and shut. However, the SRO must make an operational decision in this case because this action will require entry into TS 3.5.1 Condition H which is a LCO 3.0.3 shutdown statement. Although the plant will begin preps for shutdown, this will allow the plant to perform a controlled shutdown vice scram the plant.
EP-3 step SPT 3 and 4 requires entry into EP-2 to effect a manual scram BEFORE, i.e.no later than when, Supp Pool temperature reaches 110°F. At the current suppression pool temperature, the SRO is expected to take the actions required to prevent an unnecessary plant transient (scram); however, EP-3 will be entered at 95F and the applicant may mistake that Entry condition for a need to scram the plant.
Technical
References:
EP-3 EP-2 E1181-68, 69 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-E1200 Obj. 8.10, 20; GLP-OPS-EP3 Obj. 3,7,8 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(5)
Examination Outline Cross-Reference Level SRO 256000 Reactor Condensate System 2.4.1 Knowledge of EOP entry conditions and immediate K/A # 256000 / 2.4.1 action steps. Rating 4.8 (CFR: 41.10 / 43.5 / 45.13)
Rev / Date 0 Question 93 The plant is at rated power.
A spurious IP Condenser Hotwell Level Low signal, <1-11 wc, occurs due to a relay failure, indicated by amber alarm 1H13-P680-2A-E9, CONDSR HTWL LVL LO.
As a result, the highest priority for the CRS is to implement A. EP-2, RPV Control.
B. 05-1-02-III-3, Reduction in Recirculation System Flow Rate.
C. 05-1-02-V-7, Feedwater Systems Malfunctions.
D. ARI 04-1-02-1H13-P680-2A-E9, CONDSR HTWL LVL LO.
Answer: A Explanation: Low IP hotwell level is sensed by N19 level switches N105, N106, and N107, which actuate auxiliary relay N19 63X1/N105. This relay is a single point failure component that provides a common trip signal to all three Condensate Pumps.
If the relay contacts fail to the energized state, i.e. closed, all running Condensate Pumps trip, resulting in trip of all Condensate Booster Pumps and Reactor Feed Pumps. Plant data shows for this transient, reactor water level rapidly falls to around -
80 inches wide range, due to loss of all feed flow and due to shrink following the scram at reactor Level 3. EP-2 entry condition <11.4 reactor water level will be met within seconds of the initiating event. There would be no time for the RO or SRO to refer to the ARI before EP-2 execution is required. Additionally, no actions of the ARI would provide immediate, rapid recovery of Condensate in order to avert an automatic
scram or HPCS/RCIC initiations. Therefore, any time before an automatic scram on low water level should be spent immediately effecting a manual scram per EN-OP-115 to protect the reactor when it is in jeopardy. EP-2 entry condition on low water level will be met for either an automatic or a manual scram, due to shrink. Emergency Procedure execution is higher priority than lower tier procedures. The lower tier procedures should only delay EP execution when they are needed to accomplish specific EP steps. Answer A is correct. All other answers are plausible because entry requirements for the listed procedures are met. However, for answers B and C, the reactor operators are responsible for performing immediate ONEP actions without requiring direction from the CRS, and though those actions may seem important to the initial license candidate, EP-2 is more important in the onset of an event to ensure the overall plant is stabilized. Answers B and C are wrong because ONEP subsequent action execution in this case is of lower priority than restoring and stabilizing reactor water level using the more comprehensive strategies given in EP-2.
Answer D is plausible because the alarm is the initial manifestation of the problem and because the ARI provides instructions for defeating the failed relay. However, EP-2 begins with particularly arranged steps necessary to ensure reactor and personnel safety under all conditions that must be addressed before focus is dedicated to individual systems or components. No actions of the ARI would provide immediate, rapid recovery of Condensate in order to avert an automatic scram or HPCS/RCIC initiations. Recovery of Condensate will be prioritized according EP-2 itself, and then the ARI would be executed to compliment/facilitate EP-2 steps. Therefore, answer D is wrong.
Technical
References:
04-1-02-1H13-P680-2A-E9 EP-2 05-1-02-III-3 05-1-02-V-7 E1148-15 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-N1900 Obj. 13; GLP-OPS-EP2 Obj. 1,2,3,5,8 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
K/A # 2.1.23 (CFR: 41.10 / 43.5 / 45.2 / 45.6) Rating 4.4 Rev / Date 0 Question 94 Following a LOCA with limited fuel damage, the Shift Technical Advisor reports containment hydrogen concentration has reached 3%.
Based on this; A. EP-2, RPV Control, is exited and EP-3, Containment Control is continued.
B. EP-2, RPV Control, is exited, EP-5, RPV Flooding is entered, and EP-3, Containment Control, is continued.
C. EP-2, RPV Control, is continued, EP-3, Containment Control is exited, and the SAPs are entered.
D. All EPs are exited and all SAPs are entered.
Answer: D Explanation: This is incorrect because the EPs state if containment or drywell hydrogen concentration exceeds 2.9%, all the EPs are exited and all the SAGs are entered.
B. This is incorrect because the EPs state if containment or drywell hydrogen concentration exceeds 2.9%, all the EPs are exited and all the SAGs are entered C. This is incorrect because the EPs state if containment or drywell hydrogen concentration exceeds 2.9%, all the EPs are exited and all the SAGs are entered D. This is correct based on the discussion above..
Technical
References:
EP-3, Containment Control, step H-1.
References to be provided to applicants during exam: none Learning Objective: GLP-OPS-EP3 Obj. 8 Question Source: Bank # 2009 NRC Q#95 (note changes; attach parent) Modified Bank #
New Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.
K/A # 2.1.37 (CFR: 41.1 / 43.6 / 45.6) Rating 4.6 Rev / Date 0 Question 95 Power is to be reduced from 100% to 65% for a control rod sequence exchange.
Who is responsible for reviewing the Reactivity Maneuver Plan and ensuring that the control rod pull sheets are highlighted to emphasize areas of concern?
A. Reactor Engineering B. Field Support Supervisor C. Reactivity Management SRO D. Operations Manager Answer: C Explanation: 02-S-01-27, Operations Philosophy, classifies the described power change as a Type 3 power maneuver per step 6.8.1a(3), which requires staffing an additional SRO, the Reactivity Management SRO (RMSRO). Step 6.8.1b(2) states the RMSRO is responsible for reviewing the Reactivity Maneuver Plan and ensuring that the control rod pull sheets are highlighted to emphasize areas of concern. Step 6.8.2d states the RMSRO is responsible for instructing ROs on the execution of the specific pull sheets. Therefore, answer C is correct. Answer A is plausible since Reactor Engineers develop the control rod pull sheets and are involved in during the sequence exchange. Answer B is plausible since this is an on shift SRO who does not normally have the control room command function and might be considered capable of assuming dedicated reactivity management duties. Answer D is plausible because Ops Management is present in the control room for management oversight during power reductions for sequence exchanges.
Technical
References:
02-S-01-27, Operations Philosophy References to be provided to applicants during exam: none Learning Objective: GLP-OPS-PROC Obj. 4.10 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(1) 55.43(b)(6)
Examination Outline Cross-Reference Level SRO 2.2.35 Ability to determine Technical Specification Mode of Operation.
K/A # 2.2.35 (CFR: 41.7 / 41.10 / 43.2 / 45.13) Rating 4.5 Rev / Date 0 Question 96 Use your provided references to answer this question.
A refueling outage is in progress.
The following conditions exist:
- RPV head bolt de-tensioning is in progress.
- Primary and Secondary containment will be relaxed on the next shift.
- A complete loss of Shutdown Cooling occurred.
- The refuel floor supervisor has directed that all head bolts be re-tensioned
- Recirc loop temperature reached 201°F at 0630.
- It is now 0743 with:
- Recirc loop temperature reached 202°F
- Two RPV head bolts remain de-tensioned (1) What EAL entry is required?
(2) What is the current Tech Spec MODE of Operation?
A. (1) CA3 - Alert (2) Hot Shutdown B. (1) CA3 - Alert (2) Refueling C. (1) CU3 - Unusual Event (2) Hot Shutdown D. (1) CU3 - Unusual Event (2) Refueling
Answer: B Explanation:
The given stem conditions places the plant in EAL CA3 because even though it suggests maintaining Cold Shutdown (Mode 4), it is applicable in Modes 4 & 5 (10-S-01-1 Att. 2 page 30 of 113, Rev 122).
The EAL is met because the RCS temperature has been above 200°F for 60 minutes.
CU3 is plausible based on misinterpreting the EAL conditions for CA3.
Hot Shutdown is plausible based on temperature being above 200°F; however, Mode 5 has no temperature requirements.
Technical
References:
10-S-01-1 flowchart TS table 1.1-1 References to be provided to applicants during exam: EAL flowchart page 2 of 2 Learning Objective: GLP-OPS-TS001 Obj. 5; GLP-EP-EPTS6, Obj. 1 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7,10) 55.43(b)(2)
Examination Outline Cross-Reference Level SRO 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting K/A # 2.2.36 conditions for operations. Rating 4.2 (CFR: 41.10 / 43.2 / 45.13)
Rev / Date 0 Question 97 The plant is at rated power.
Corrective Maintenance is scheduled to replace an ECCS jockey pump.
Which of the following is required to be performed before the tagout lineup to drain the ECCS jockey pump piping necessary for this work?
A. Quarterly functional surveillance for the associated ECCS system valves B. Closed loop leakage test for jockey pump containment boundary valves C. Monthly lineup and fill and vent surveillance for the associated ECCS D. Quarterly functional surveillance for the associated jockey pump Answer: B Explanation: per 02-S-01-17 section 6.12, a closed loop leakage test is required before breaching the closed system outside containment to ensure the overall containment leak rate limit, L a , is not exceeded. Otherwise, an LCO for Containment Integrity would have to be entered. Therefore, answer B is correct. Other answers are plausible because they are all necessary retests possibly required for this type of maintenance, but they are wrong since they are required post-maintenance, not pre-maintenance.
Technical
References:
TS 3.8.7 and bases TS 3.8.1 References to be provided to applicants during exam: none Learning Objective: GLP-OPS-TSLCO Obj. 14
Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(2)
Examination Outline Cross-Reference Level SRO 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or K/A # 2.3.14 activities. Rating 3.8 (CFR: 41.12 / 43.4 / 45.10)
Rev / Date 0 Question 98 Operators are in EP-3 (CTMT Control) attempting to control a rising CTMT pressure.
The CRS has determined it is now necessary to vent the CTMT in order to keep CTMT pressure below 22.4 psig.
The EOF is fully operational.
Just before opening the 20 vents, the CRS/Shift Manager is required to directly notify the ________(1)________ that ________(2)_________ release of radioactivity will occur.
A. (1) Radiological Assessment Coordinator (2) a Filtered B. (1) Radiological Assessment Coordinator (2) an Un-filtered C. (1) Emergency Director (2) a Filtered D. (1) Emergency Director (2) an Un-filtered Answer: D Explanation: See EP-1, Attachment 13 (page 3 of 10), step 2.7, where direction is given to notify the Emergency Director that an Un-filtered release will begin, not the Radiological Assessment Coordinator and not filtered. Therefore, answer D is correct Radiological Assessment Coordinator is a plausible distracter because that position is
responsible for overall dose assessment.
Filtered release is a plausible distracter because the smaller 6 containment vent path (Low Volume Purge) for EP Att. 14 vents through the Ctmt exhaust charcoal filter train.
Technical
References:
EP-1, Attachment 13 - Containment Venting/Defeating Containment Vent Path Isolation Interlocks References to be provided to applicants during exam: none Learning Objective: GLP-EP-EPTS26 Obj. 40 Question Source: Bank #
(note changes; attach parent) Modified Bank # 408 X New Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(12) 55.43(b)(4)
Examination Outline Cross-Reference Level SRO 2.4.11 Knowledge of abnormal condition procedures.
(CFR: 41.10 / 43.5 / 45.13)
K/A # 2.4.11 Rating 4.2 Rev / Date 0 Question 99 The plant is operating at rated power.
- Feedwater and Condensate conductivity is trending up but do not exceed limits.
- Reactor Water Conductivity is 1.5 umho/cm.
The CRS will A. (1) Enter Condensate/Reactor Water High Conductivity ONEP and TRM 6.4.1 (2) Restore reactor chemistry to within limits in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. (1) Enter Condensate/Reactor Water High Conductivity ONEP and TRM 6.4.1 (2) Manually scram the reactor C. (1) Enter TRM 6.4.1, Only (2) Restore reactor chemistry to within limits in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> D. (1) Enter Condensate/Reactor Water High Conductivity ONEP, Only (2) Manually scram the reactor Answer: A Explanation:
TRM 6.4.1 is entered in Mode 1 with Conductivity 1.0 umho/cm. The Condensate/Reactor Water High Conductivity ONEP directs actions with as little as 0.3 umho/cm so it also requires entry.
The TRM action in this case is to restore within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The ONEP requires a scram when conductivity exceeds 5 umho/cm.
Technical
References:
TRM 6.4.1, Chemistry
05-1-02-V-12, Condensate/Reactor Water High Conductivity ONEP References to be provided to applicants during exam: none Learning Objective: GLP-OPS-TS001 obj 4 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b) 55.43(b)(5)
Examination Outline Cross-Reference Level SRO 2.4.25 Knowledge of fire protection procedures.
(CFR: 41.10 / 43.5 / 45.13)
K/A # 2.4.25 Rating 3.7 Rev / Date 0 Question 100 Use your provided references to answer this question.
Painters are covering one smoke detector in the Control Building per an approved work order.
All fire rated assemblies in the area are OPERABLE.
All other smoke detectors and all heat detectors in the Control Building are OPERABLE.
What is the minimum action (i.e. least manpower and material required) required by the TRM for this activity?
A. Establish an hourly fire watch patrol, only, for the affected area within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B. Establish an hourly fire watch patrol with backup fire suppression for the affected area within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
C. Establish a continuous fire watch, only, for the affected area within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
D. Establish a continuous fire watch with backup fire suppression for the affected area within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Answer: A Explanation: Per 10-S-03-1, Fire Protection System Impairment, step 2.4, the Shift Manager is responsible for ensuring LCO requirements are met. A smoke detector is a Function A device per TR 6.2.1, and covering one causes that detector to be inoperable. TR 6.2.8 Action A.1 requires a hourly fire watch patrol; therefore, answer
A is correct. Answer C is plausible because a continuous firewatch may be required per TR 6.2.8 based on fire rated assembly operability in association with fire detection instrumentation operability. But here, fire rated assemblies are operable, so only an hourly FW is required. And although a continuous FW would satisfy the purpose of an hourly FW patrol, the stem asks for the minimum action, which would be an hourly patrol rather than an individual dedicated to one specific area. Answers B and D are plausible because inoperability of some fire detection instrumentation also inops fire suppression systems, such as CO2, sprinklers, and halon, which would require backup fire suppression, if affected, per TR 6.2.3, TR 6.2.4, TR 6.2.5. However, only Function B detectors initiate these systems and smokle detectors are Function A; therefore, answers B and D are wrong.
Technical
References:
10-S-03-1, Fire Protection System Impairment TR 6.2.1 TR 6.2.8, TR 6.2.3, TR 6.2.4, TR 6.2.5 References to be provided to applicants during exam: TR 6.2.1 Learning Objective: GLP-OPS-FIRELCO Obj. 1,19,34 Question Source: Bank #
(note changes; attach parent) Modified Bank #
New X Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10) 55.43(b)(5)