ML081330625
| ML081330625 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/25/2008 |
| From: | Ludlam G AmerGen Energy Co |
| To: | Caruso J Operations Branch I |
| Hansell S | |
| Shared Package | |
| ML072851077 | List: |
| References | |
| U01688 05000219/2008-301 | |
| Download: ML081330625 (285) | |
Text
ES-401 Site-Specific k0 Written Examination Form ES-401-7 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:
Date: Facility I Unit: Oyster Creek Region: Reactor Type: W 0CE 0BW 0GE Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.
Applicant Certification All work on this examination is my own. I have neither given nor received aid.
Examination Value Points I1 Applicant's Score Points I Applicant's Grade Percent ES-401, Page 30 of 33
OC ILT 07-1 RO NRC Written Exam Required References 1 Question 1 Reference To Be Provided 1 1 Attachment 202.1-5 13 Large HCTL, PCPL, CSIL, WIT, PSP and TLL curves (this represents all curves from PCC EOP) 31 I EMG-SP4 1 33 I Drawing 148F723 (WITH Note 10 DELETED) 68 Attachment 202.1-2 73 Attachment 203-2
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Name: Date:
oc LT 07-1 RO NRC Exam KEY The plant was at 80% power. Recirculation Pump A has just been shutdown and the following valves are closed:
0 PUMP SUCTION 0 DISCHARGE DlSCH BYPASS IAW procedure 202.1 , Power Operation, which of the following limits is reduced due to the new operating loop configuration?
A. MCPR, as required by the fuel vendor.
B. FLLLP, as required by the USAR safety analysis.
C. MAPLHGR, as required by Technical Specifications.
D. MLHGR, as required by the Core Operating Limits Report.
Page 2 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question ##
C Question Developer InitialdDate: NTP 11/13/07 Answer Knowledge and Ability Reference Information 295001 AK3.05 Importance Knowledge of the reasons for the following responses Rating as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Reduced loop operating requirements Level RO Tier # 1 Group# 1 References 202.1 TS 3.3.F2.a. 1 I
The question stem shows the plant at > 25% power, with the primary containment inerted, and with one recirculation pump isolated. IAW procedure 202.1, in this configuration, only MAPLHGR must be reduced from the normal 5-loop operating configuration to a 4-lOOp Explanation:
configuration, with power > 25% and the primary containment interted. A reduction in MAPLHGR is required by Technical Specifications 3.3.F.2.a.l. Answer C is correct. All other distracters are plausible and incorrect.
Objective Given a set of system indications or data, evaluate and interpret then to determine limits, trends and system status.
Question Source 1 Bank 1 1 1 New X Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
Time to Complete: 1-2 minutes Page 3 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant had just been shutdown when a complete loss of offsite power occurred due to a fault on both 34.5 KV power lines.
Which of the following is correct regarding AC power supplies? (Assume NO operator actions unless stated)
A. Combustion Turbine #1 can be manually aligned directly to Bus 1A to provide power to Feedwater Pump 1A and Condensate Pump 1A.
- 6. Emergency Diesel Generator #1 will automatically start and load onto Bus 1C to provide power to Bus 1A2 to provide power to CRD Pump 1A.
C. When the fault has been cleared on Bank 5, breaker S1B will automatically close to provide power to Feedwater Pump 1B and Condensate Pump 18.
D. Emergency Diesel Generator #2 will automatically start and load onto Bus 1D to provide power to Bus 1B2 to provide power to a Condensate Transfer Pump 1-2.
Page 4 of 211
OC ILT 07-1 RO NRC Exam KEY Question #
B Question Developer InitialdDate: NTP 11/14/07 Answer Knowledge and Ability Reference Information RO SRO 295003 AK3.01 Importance 3.3 3.5 Knowledge of the reasons for the following responses Rating as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Manual and auto bus transfer Level RO Tier # 1 Group# 1 Refere nces 341 USAR 8.3.1.1.1
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Explanation:
Page 5 of 21 1
OC ILT 07-1 RO NRC Exam KEY powered from 1832 - not 1B2. Bus 1B32 remains de-energized under the conditions presented in the question.
Answer D is incorrect.
components and expected system response including power loss or failed components.
Question Source [Bank I I Modified Bank I 1 I New-Question Cognitive Memory or X Comprehension Level: Fundamental 1:B or Analysis Know ledge
~~
10 CFR Part 55 55.41 5 55.43 Content:
Time to Complete: 1-2 minutes Page 6 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when an event occurred. Indications and investigations revealed the following:
0 Battery Charger MG Set A Breaker has opened 0 Battery A Main Breaker has opened Which of the following states the proper function of a DC Distribution System Automatic Transfer Switch under the given conditions?
The power to 125 VDC Bus (1) has automatically transferred to 125VDC Bus (2) .
(1) (2)
A. DC-F DC-C B. DC-1 DC-C C. DC-2 DC-B D. DC-E DC-B Page 7 of 21 1
OC ILT 07-1 HO NRC Exam KEY I Question ## 1 I Question Developer InitiaWDate: NTP 11/14/07 I Knowledge and Ability Reference Information RO I SRO 295004 2.1.28 (Partial/complete loss of DC power)
Knowledge of the purpose and function of major system components and controls.
Level RO Tier # 1 Group#
Importance Rating 1
l-r--References 0-3033 RAP-9XF4e ABN-54 BR 3028 The question stem describes a loss of power to 125 VDC Bus DC-A (both the battery charger and battery become disconnected from the Bus). When this bus de-energizes, then automatic transfer switch DC-E swaps from DC-A as the source of input power to 125 VDC Bus DC-B. Answer D is correct. Bus DC-F normally receives power from Explanation:
Bus DC-C, which is not affected by the loss of DC-A. Answer A is incorrect. Bus DC-1 normally receives power from Bus DC-B, which is not affected by the loss of DC-A. Answer B is incorrect. Bus DC-2 normally receives power from Bus DC-C, which is not affected by the loss of DC-A. Answer D is incorrect.
References to Learning 2621.828.0.001 2 01121 Objective State potential consequences on plant operation, plant equipment, and environment due to failure of DC electrical system.
I Question Source 1 Bank I 1 Modified Bank 1 1 1 New Question Cognitive Memory or Comprehension X Level: Fundamental or Ana Iysis 3:SPK Knowledge I 10 CFR Part 55 Content:
I 55.41 17 1 I55.43 I Time to Complete: 1-2 minutes Page 8 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was starting up after an outage. The Operator had just completed synchronizing the main generator to the grid and generator output indicated 130 MWe, when the following annunciator alarmed:
GENERATOR - FIELD LOST Which of the following actions is required?
A. Manually scram the reactor and enter ABN-1 , Reactor Scram.
B. Manually trip the turbine and enter ABN-I 0, Turbine Generator Trip.
C. Verify RPV pressure stabilized below 1045 psig with the Turbine Bypass Valves.
D. Control RPV water level 138 - 175 with Condensate/Feedwater and CRD.
Page 9 of 21 1
OC ILT 07-1 RO NRC Exam KEY I Answer I Knowledge and Ability Reference Information RO SRO 295005 AA2.04 Importance 3.7 3.8 Ability to determine and/or interpret the following as Rating they apply to MAIN TURBINE GENERATOR TRIP:
Reactor Pressure Level RO Tier # 1 Group #
References 1 RAP-R3a 1 ABN-10 The question stem describes a plant startup and completion of synchronizing the main generator to the grid with electrical output at 130 MWe (which approximates 20% electrical output). With grid synchronization complete, all turbine bypass valves are closed which means that reactor power is also approximately 20%. The loss of field alarm is a main generator trip that also provides a turbine trip. At 20%
reactor power, the turbine trip reactor scram is bypassed. A subsequent action in ABN-I 0 directs that if reactor power was e 30%,
Explanation: then stabilize reactor pressure < 1045 with turbine bypass valves.
Answer C is correct. IAW ABN-10, if reactor power was >30% on the turbine-generator trip, then a manual scram IAW ABN-1 is required.
Answer A is incorrect. Since the turbine did automatically trip, there is no requirement to manually trip the turbine. Answer B is incorrect.
Direction for RPV water level control in ABN-IO directs 155 - 165, not 138 - 175 as in RPV Control - No ATWS EOP. Also, there are no entry conditions into this EOP. Answer D is incorrect.
Learning 2621.828.0.0025 248-10445 Objective Given a set of system indications or data, evaluate and interpret then to determine limits, trends, and system status Pageloof211
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3: PEO Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
ITime to Complete: 1-2 minutes Page 11 of 21 1
OC ILT 07-1 RO NRC Exam KEY Following an automatic scram from rated power, the Operator placed the REACTOR MODE SELECTOR switch in SHUTDOWN.
Which of the following indications, ALONE, allows the Reactor Operator to confirm that the reactor will remain shutdown under all conditions without boron?
A. All APRMs indicate < 2% power.
B. All control rods indicate position 04.
C. All LPRM amber lights on Panel 4F are LIT.
D. All control rods full-in EXCEPT 2 control rods at position 30.
Page12of211
OC ILT 07-1 RO NRC Exam KEY Question #
Question Developer InitiaWDate: NTP 11/15/07 Answer Knowledge and Ability Reference Information RO SRO 295006 AK2.06 Importance 4.2 4.3 Knowledge of the interrelations between SCRAM and Rating the following: Reactor power Level RO Tier# 1 Group# 1 2000-BAS-References 3200.02 (EOP Users Guide)
IAW the EOP Users Guide, position 04 is the Maximum Subcritical Banked Withdrawal Position (MSBWP). If all control rods are inserted to at least position 04, the reactor will remain shutdown under all RPV water temperature, xenon, and boron conditions. Answer B is correct.
If all APRMs indicate 1.5% (which is less than 2%), then the reactor is not shutdown presently. Answer A is incorrect. LPRM amber lights come on when the LPRMs are less than 2%. Similar to answer A, the reactor cannot be confirmed shutdown by these indications alone.
Explanation: Answer C is incorrect. IAW the reference, all control rods to 04 or beyond or all control rods at position 00 with any single control rod at any position, will ensure the reactor will remain shutdown under all conditions. With all rods full-in except 2 control rods at position 30, these conditions are not satisfied and the control room operator can not by himself guarantee the reactor will remain shutdown under all conditions without boron. The core engineer can be used to make this determination.
Learning 2621.845.0.0040 3053 Objective Explain the basis for each step of the RPV Control - No ATWS entry conditions.
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OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis l:B Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
Page 14 of 211 I
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power for 1 week following a refuel outage when an event occurred that required a manual scram. One hour later, the plant was cooling down with the Turbine Bypass Valves. Current plant conditions are as follows:
RPV pressure is 600 psig and lowering RPV water level is in the normal band Primary Containment parameters are normal An event then occurred which required a Control Room Evacuation (not due to a fire).
The Operators were able to accomplish ALL Control Room Subsequent Operator Actions IAW ABN-30, Control Room Evacuation, PRIOR to leaving the Control Room. All required Shutdown Panels have been activated.
Which of the following indications at the Remote Shutdown Panel is correct?
A. RPV pressure indication will be rising since the MSIVS are closed.
B. RPV pressure indication will be lowering since Isolation Condenser B is in service.
C. RPV water level indication will be rising since the Feedwater/Condensate System is injecting.
D. Isolation Condenser shell water level indication will be lowering since Isolation Condenser A is in service.
Page 15 of 211
OC ILT 07-1 RO NRC Exam KEY Knowledge and Ability Reference Information 295016 AA2.03 Importance Ability to determine and/or interpret the following as Rating they apply to CONTROL ROOM ABANDONMENT:
Reactor pressure Level I RO I Tier# 11 1 Group # 1 References I346 I
1 ABN-30 I I
The question describes the plant at rated power for 1 week after a refuel outage, when a manual scram was inserted. With this operating history, decay heat will be small. IAW ABN-30, the actions performed prior to evacuating the control room include the following:
scram the reactor, trip all recirculation pumps, close the MSIVs, trip all feedwater pumps, trip the turbine, trip all condensate pumps, initiate isolation condenser B, and defeat the automatic initiation of isolation condenser A.
Explanation: Because there is very little decay heat, and since one isolation condenser can carry approximately 3% power at rated pressure, it is expected that RPV pressure will be lowering as indicated at the RSD from the operation of IC-6. Answer B is correct.
Since RPV pressure will be lowering, answer A is incorrect. Answer B is incorrect since all feedwater and condensate pumps were manually tripped in the control prior to evacuating the control room. Answer D is incorrect since its auto start has been defeated by actions in the control room prior to evacuating.
References to provided durii 1
Learning 2621.828.0.0064 10446 Objective Identify and explain system operating controls/indications under all plant operating conditions.
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Question Source Bank 1 I Modified Bank I Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3: PEO Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
Page 17 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was shutdown and cooled down. An outage activity requires that all main condenser waterboxes be drained and opened.
Which of the following states the required action PRIOR to waterbox draining and opening?
A. Lineup Fire Protection to cool the Station Air Compressors.
B. Align the Service Water System to the TBCCW heat exchangers.
C. Place an additional TBCCW heat exchanger in service IAW procedure.
D. Align the Emergency Service Water System to the RBCCW heat exchangers.
Page18of211
OC ILT 07-1 RO NRC Exam KEY Question #
B Question Developer InitiaWDate: NTP 11/17/07 Answer Knowledge and Ability Reference Information RO SRO 295018 AAI .01 Importance 3.3 3.4 Ability to operate and/or monitor the following as Rating they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Backup systems Level 1 RO I Tier# 11 1 Level 1 RO References 323 322 Explanation:
Learning 2621.828.0.0048 274-10450 Objective Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation in accordance with plant procedures.
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OC ILT 07-1 RO NRC Exam KEY Question Source
~
Bank 1 Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 2:DR Knowledge 10 CFR Part 55 55.41 7 55.43 Content :
Page 20 of 211
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when the following annunciators alarmed:
TBCCW - DISCH PRESS LO TBCCW - SURGE TANK LVL HVLO TURB BLDG SUMP LVLS - LUBE OIL BAY 1-1 LEVEL HI TURB BLDG SUMP LVLS -CONDENSATE BAY 1-2 LEVEL HI The Operator reports TBCCW HX OUTLET PRESS indicates 15 psig and lowering and that the TBCCW Surge Tank indicates low and CANNOT be raised.
Which of the following states the required Subsequent Operator Actions IAW ABN-20, TBCCW Failure Response?
First Action Second Action A. Manually scram the reactor Stop all recirculation pumps B. Manually scram the reactor Manually trip the turbine C. Secure TBCCW Pumps Manually scram the reactor D. Perform a rapid power reduction Manually trip the turbine Page 21 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
A Question Developer InitialdDate: NTP 11/16/07 Answer Knowledge and Ability Reference Information RO SRO 295018 2.1.23 (PartiallComplete loss of component Importance 3.9 4.0 cooling water) Rating Ability to perform specific system and integrated plant procedures during different modes of plant operation.
Level RO Tier # 1 Group# 1 References ABN-20 RAP-M4a Explanation:
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OC ILT 07-1 RO NRC Exam KEY in TBCCW and is a plausible distractor.
Answer D is incorrect.
Learning 2621.828.0.0048 274-10450 Objective Describe and interpret procedure sections and steps for plant emergency and off-normal conditions that involve this system including personnel allocation and equipment operation IAW plant procedures.
Modified Question Source Bank New X Bank I I Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 2:DR Knowledge 10 CFR Part 55 55.41 2 55.43 Content:
Page 23 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when a total loss of station air pressure occurred.
Which of the following valves has an established backup air system that can be manually connected?
A. CRD Flow Control Valves.
B. Main Feed Regulating Valves.
C. Drywell Vent and Purge Valves.
D. Isolation Condenser Makeup Valves.
Page 24 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
Question Developer InitiaWDate: NTP 11/16/07 Answer Knowledge and Ability Reference Information RO SRO 295019 AA1.01 Importance 3.5 3.3 Ability to operate and/or monitor the following as they Rating apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Backup air supply Level RO Tier# 1 Group# 1 I I I References
~~ 307 The question describes the plant at rated power when all station air pressure is lost. IAW procedure 307, Isolation Condenser System, the Isolation Condenser makeup valves have an established method to supply an alternate/backup air supply. Answer D is correct.
All other listed valves are air operated, which makes these distractors plausible, but do not have an alternate/backup air supply. Answers B-D are incorrect. CRD can be used for RPV injection and insertion of Explanation: control rods. Feedwater can supply high pressure injection into the RPV. The Drywell vent and purge valves can be used to control Primary Containment parameters in the Primary Containment Control EOP. All systems directly provide major support to the RPV or primary containment. The KA is directly matched in that the question provides a loss of air and the question asks about a backup air supply.
References to provided durii Learning Objective 02029 2621.828.0.0023 Describe the relationships between the Isolation Condenser System and the following: Instrument & Service Air System Page 25 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank I I Question Cognitive Memory or X Comprehension Level: Fundamental 1:1 or Analysis Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
Page 26 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is shutdown and a cooldown is in-progress with the Shutdown Cooling System. Current plant conditions are as follows:
0 Shutdown Cooling Pump B and C are in service Shutdown Cooling Pump A is tagged out of service 0 RPV water level is 182" RPV water temperature is 275 O F and lowering slowly All Recirculation Pumps are OFF The following annunciators then alarmed:
1B2 MN BRKR TRIP 182 MN BRKR OL TRIP Which of the following actions is required?
A. To control RPV pressure, open the EMRVs and use the Condensate System for makeup.
B. To control RPV pressure, open the EMRVs and use the Control Rod Drive System for makeup.
C. To control the RPV cooldown, raise Reactor Water Cleanup System flow in letdown mode and use the Core Spray System for makeup.'
D. To control the RPV cooldown, raise Reactor Water Cleanup System flow in letdown mode and use the Condensate System for makeup.
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OC ILT 07-1 RO NRC Exam KEY Knowledge and Ability Reference Information RO SRO 295021 AK3.04 Importance 3.3 3.4 Knowledge of the reasons for the following responses Rating as they apply to LOSS OF SHUTDOWN COOLING:
Maximizing reactor water cleanup flow I RO I
Level Tier # 1 Group# 1 ABN-48 References ABN-3 303 RAP-QIc The question stem describes an overload condition and loss of 480 VAC Bus 1B2, which causes a loss of power to both operating SDC pumps. IAW ABN-3, an acceptable method under the present plant conditions to maintain the cooldown is to initiate RWCU in the letdown mode with makeup through the condensate system. To maximize the cooldown rate through the cleanup system, then Explanation: cleanup flow rate would need to me raised. Answer D is correct.
The use of isolation condensers is not allowed due high RPV water level. Answer A is incorrect. The use of EMRVs is allowed but RPV makeup is by the core spray system - not CRD. Answer A and B are incorrect. Using cleanup system letdown can be done with either condensate or CRD - not core spray. Answer C is incorrect.
Learning 2621.828.0.0045 205-10450 Objective Describe and interpret procedure sections and steps for plant emergency and off-normal conditions that involve this system including personnel allocation and equipment operation IAW ABN, EOP & EOP Support Procedures, and EPIPs.
Page 28 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Source I Bank 1
I Modified Bank 1 I 1 I I New I
~~
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 1 X 3:SPK 10 CFR Part 55 Content:
55.41 5 I 55.43 I Page 29 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is shutdown for a refuel outage with fuel moves in progress on the refuel floor.
The refuel floor SRO has just notified the Control Room that a fuel bundle has dropped onto the top of the reactor core. The Control Room Operator reports the following radiation monitor readings:
Radiation Monitor B9 indicates 75 mr/hr Radiation Monitor C10 indicates 80 mr/hr 0 Reactor Building Ventilation Exhaust Radiation Monitor 1 indicates 20 mr/hr Which of the following states the status of the RB Ventilation System AND the reason for this system status?
RB Ventilation Status Reason Am Trips and isolates BUT is manually To reduce refuel floor radiation levels restarted as quickly as possible B' Trips and isolates BUT is manually To ensure the greatest amount of air restarted dilution prior to discharge C. Trips and isolates AND remains The system is not designed for high isolated temperature air D.
Trips and isolates AND remains Ensure air is discharged through a isolated filtration system Page 30 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
11 d Question Developer InitialdDate: NTP 11/17/07 295023 AK3.03 Importance Knowledge of the reasons for the following responses Rating as they apply to REFUELING ACCIDENTS:
Ventilation isolation Level RO Tier # 1 Group# 1 References I RAP-1OFlf 1 USAR 6.5.1.1 I The question describes a refuel accident during refueling. The indications provide the following information: radiation monitor B9 is above its setpoint (50 mr/hr) and starts a 2-minute delay until the normal RB vent system isolates and SGTS starts; the RB vent radiation monitor is above its setpoint (9 mr/hr) to immediately isolate the normal RB vent system and start SGTS. Therefore, the normal RB vent system is isolated and SGTS has started to ensure the radioactive atmosphere is discharged through a filtration system.
Answer D is correct.
IAW the station procedures, if ONLY the refuel area radiation monitors B9 or C9 have isolated the normal RB vent system and SGTS initiated, then the EOP directs placing the normal RB vent system back in service. This makes distractors A and B plausible, but not correct and the correct answer less obvious. There is no Explana ion:
procedural allowance to override the vent systems when the RB vent monitors cause a valid isolation.
Because the radioactivity in the discharged air will be decaying, this decay results in a temperature increase and distractor C is plausible.
The KA match is direct in that it matches a refuel accident with the reason for ventilation isolation.
An override on the Secondary Containment Control EOP talks about operating RB vent under conditions similar, but not identical, to those in the question. Because the SRO will receive this EOP as a reference and the RO does not, this override will be deleted from the SRO reference. This has been annotated on the SRO Reference Summary.
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OC ILT 07-1 RO NRC Exam KEY Learning 2621.828.0.0042 261-1 0446 Objective Identify and explain system operating controls/indications under all plant operating conditions.
Modified Question Source Bank New Bank Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis 1:F Knowledge
~ ~~~ ~ ~ ~~
10 CFR Part 55 55.41 5 55.43 Content:
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OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when a LOCA occurred. The following conditions currently exist:
Containment Spray Pump 51A is operating in the DW SPRAY mode 0 Containment Spray Pump 51C is operating in the TORUS CLG mode Drywell pressure is 13 psig and lowering The following annunciators then alarmed:
S l A BRKR TRIP BUS 1A U N Which of the following states the response of the Containment Spray Pumps 51A and 51C?
Containment Spray Pump 51A Containment Spray Pump 51C A. Trips AND can be re-started Trips AND can be re-started immediately after AC power is immediately after AC power is restored restored B. Trips AND will automatically restart Remains running after a time delay after AC power is restored C. Remains running Trips AND will automatically restart after AC power is restored D. Trips AND can be re-started after a Remains running time delay after AC power is restored Page 33 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question # 12 D Question Developer InitialdDate: NTP 11/17/07
! Answer I
Knowledge and Ability Reference Information RO SRO 295024 EA1.17 Importance 3.9 3.9 Ability to operate and/or monitor the following as Rating they apply to HIGH DRYWELL PRESSURE:
Containment spray: Plant-Specific Level I RO ITier# 11 I Group# 11 BR 3000 References 237E901 sh 1 11688328 sh. 11a RAP-S1f
~~
The question shows that containment spray pump 51A (powered from USS Bus 1A2, which is powered from 4160 VAC Bus 1C) is spraying the drywell, and that containment spray pump 51C (powered from USS Bus 162, which is powered from 4160 VAC Bus 1D) is cooling the torus. The alarm given describes a loss of the startup transformer (SA) to Bus 1A and onto Bus 1C (which powers bus 1A2). When this occurs, containment spray pump 51A will trip, and EDG1 will start and load onto bus l C , which will automatically re-energize bus 1A2.
But, there is a 200 second time delay after the EDG has loaded onto the bus to allow for sequenced loading. There is no auto start of the Explanation: pumps, even if they were previously running when the startup power was lost. Therefore, containment spray pump 51A will trip, and can be manually re-started after a time delay after the bus power is restored.
The loss of the startup transformer SA does not impact the running containment spray pump 51C, since it is still powered from the second startup transformer, SB. Therefore, it will remain running.
Answer D is correct.
All other answers, although plausible, are incorrect.
Learning 2621.828.0.0009 Objective State the pressure and temperature limits of the Primary Containment Page 34 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
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OC ILT 07-1 RO NRC Exam KEY Which of the following conditions would opening of the EMRVs for RPV depressurization be UNABLE to prevent Primary Containment failure?
A. RPV pressure of 1050 psig Torus temperature of 168 O F Torus water level of 154 B. RPV pressure of 700 psig Torus temperature of 171 O F Torus water level of 144 C. Drywell temperature of 350 O F Drywell pressure of 3 psig D. Torus pressure of 30 psig Primary Containment water level of 3 6 0 Page 36 of 21 1
OC ILT 07-1 RO NRC Exam KEY I I I I Question #
13 A Question Developer InitialdDate: NTP 11/19/07 Answer Knowledge and Ability Reference Information RO SRO 295025 EA2.04 Importance 3.9 3.9 Ability to determine and/or interpret the following as Rating they apply to HIGH REACTOR PRESSURE:
Suppression pool level I I Level RO Tier # 1 Group# 1 BAS-3200.02 References (EOP Users Guide)
IAW the reference, the heat capacity temperature limit (HCTL) is a function of RPV pressure, torus water temperature, and torus water level. It is the maximum torus water temperature at a given torus water level at which initiation of RPV depressurization will not result in exceeding the torus temperature at which the primary containment pressure limit is reached before the rate of energy transfer from the RPV is within the capacity of either a single 12 (or larger) containment vent or the hardened vent. In other words, the capacity of the containment to either absorb or pass the decay heat of the reactor without failing is measured in terms of torus temperature, RPV pressure, and torus level. The indications in answer A show that the HCTL is violated, and is the correct answer. Under the same torus water temperature, torus water level, but a lower RPV pressure, Explanation: margin is gained to the HCTL curve. Under given conditions of torus level and temperature, lowering RPV pressure maintains/improves the margin to HCTL and thus to the potential loss of the primary containment failure.
HCTL is not violated in answer B, which is an incorrect answer. The responses in answer C show conditions are on the bad side of the containment spray initiation curve, which is not related to RPV depressurization. Answer C is incorrect. The responses in answer D place the plant on the good side of the primary containment pressure limit (PCPL). PCPL is based only on the structural considerations which impact the integrity of the primary containment. RPV depressurization is not one of those considerations. Answer D is incorrect.
Page 37 of 211
OC ILT 07-1 RO NRC Exam KEY curves from PCC EOP) complied with, being exceeded, or about to be exceeded based upon existing conditions.
Question Source I I
Bank I I
I I
Modified Bank New 7 Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
Time to Complete: 2-3 minutes Page 38 of 211
OC ILT 07-1 RO NRC Exam KEY The plant is at 3% power on a startup after a refuel outage. A pre-job brief is being conducted in preparation for performing Procedure 602.4.003, Electromatic Relief Valve Operability Test. Torus water temperature is currently 88 O F and Torus Cooling is not in service.
An open EMRV raises Torus water temperature by 2 "F/minute. Which of the following states how long the EMRVs can remain open during the surveillance test UNTIL a Technical Specification reactor scram requirement is FIRST met? (assume a constant heatup rate)
A. 3 Y2 minutes B. 8 % minutes C. 11 minutes D. 13 minutes Page 39 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaMDate: NTP 11/19/07 Knowledge and Ability Reference Information RO SRO 295026 2.2.22 (Suppression Pool High Water Importance 3.4 4.1 Temperature) Rating Knowledge of limiting conditions for operations and safety limits.
Level RO Tier# 1 Group# 1 References TS 3.5 TS 3.5.A.1 .c(l) states that during normal power operation, the pool temperature limit is 95 O F .
IAW TS 3.5.A.1 .c(2): During testing which adds heat to the suppression pool, the water temperature shall not exceed 10 OF above the normal POWER OPERATION limit specified in (1) above, which is 95 O F . Thus the maximum allowed Torus water temperature during this test is 105 O F .
TS 3.5.A.l.c(3) states that the reactor shall be scrammed from any power condition if the pool temperature reaches 110 O F . At 2 "F/minute for 11 minutes (= 22 OF), the pool temperature will reach 88 Explanation:
+ 22 = 110 O F . Thus, after 11 minutes the scram requirement will be first met. Answer C is correct.
Answer A places the torus water temperature at the normal power operation TS temperature limit of 95 O F . Answer B places the torus water temperature at the TS testing temperature limit of 105 O F .
Answer D places the torus water temperature above the TS scram temperature limit of 110 OF but is past the time when the scram is first required as in answer B. All distractors are plausible since the values are TS values in some fashion.
Learning 2621.828.0.0030 01032 Objective Analyze Technical Specification requirements when given applicable sections.
Page 40 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Source Bank 1 X Question Cognitive Memory or Comprehension X Level: Fundamental or Ana I ysis 3:SPR Knowledge 10 CFR Part 55 Content:
Time to Complete: 1-2 minutes Page 41 of 21 1
OC ILT 07-1 RO NRC Exam KEY A step in the Primary Containment Control EOP requires entry into the RPV Control -
No ATWS EOP, prior to reaching 281 O F in the Drywell. What is the basis for this step?
A. This ensures that Drywell Sprays will be effective.
B. This will prevent RPV water level instrument inaccuracies.
C. This ensures the environmental qualification of the EMRVs is not exceeded.
D. This reduces the rate at which heat is transferred from the RPV to the Drywell.
Page 42 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
Question Developer InitiaWDate: NTP 11/19/07 Answer Knowledge and Ability Reference Information RO SRO 295028 EA2.01 Importance 4.0 4.1 Ability to determine and/or interpret the following as Rating they apply to HIGH DRYWELL TEMPERATURE:
Drywell temperature 3
Level 1 RO I Tier# 11 1 Group# 1 BAS-3200.02 References (EOP Users Guide)
IAW the reference, entry into RPV Control - No ATWS allows the Operator to reduce reactor pressure via normal means should Drywell sprays prove unsuccessful in terminating the drywell temperature increase. A reduction in RPV pressure lowers the Explanation:
saturation temperature in the RPV and reduces the rate at which heat is transferred from the reactor to the Drywell. Answer D is correct.
Answers A, B, and C are incorrect.
References 1 provided dui Learning 1621.845.0.0042 3000 0bjective k i n g procedure EMG-3200.02, evaluate the technical basis for each
,tep in the procedure, and apply this evaluation to determine correct
- ourses of action under emergency conditions.
Page 43 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Source I Bank 1 I
X I
I Modified Bank Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis l:B Knowledge 10 CFR Part 55 Content:
~
55.41 10 1 1 55.43 Page 44 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question 16 I- . . i
. > )
The EOP Users Guide lists several adverse effects of a lowering Torus water level with the reactor at power.
Which of the following is correct if Torus water level were 125 while at power?
A. Added stress to the EMRV downcomers when the EMRVs are opened for emergency depressurization.
B. The Torus will not be able to be vented due to the loss of the ability of the Torus vent valves to function.
C. The use of the EMRVs during an emergency depressurization will result in a direct pressurization of the Torus air space.
D. The Torus water temperature will heat up faster during an Emergency Depressurization and results in a lower heat capacity.
Page 45 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
16 D Question Developer InitiaWDate: NTP 11/19/07 Answer Knowledge and Ability Reference Information RO SRO 295030 EKl.01 Importance 3.8 4.1 Knowledge of the operational implications of the Rating following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Steam condensation Level 1 RO ( T i e r # 11 IGroup# 11 6s-3200.02 References (EOP Users Guide)
IAW the reference, there are several adverse effects of a lowering torus water level: 1) loss of core spray NPSH; 2) Vortex formation; 3) reduced capacity for condensing steam that is discharged in the torus; and, 4) uncovery of the drywell vent header downcomer (1 10) and EMRV discharge (90). At a lower torus water level, there is less water volume to absorb steam energy during an ED. For a given amount of steam energy deposited in the Torus at a lower level, the water temperature will rise faster and the heat capacity goes down (heat capacity = Q/AT). Answer D is correct.
E: planation: When torus water level is <go, then the EMRV quenchers are uncovered and the use of the EMRVs is prohibited. The use of the EMRVs at this torus water level will result in a direct pressurization of the torus air space. Answer C is incorrect. The inability to vent the torus can occur when torus water level is high - not low. Answer B is incorrect. The EOP Users Guide makes no mention of additional stresses to the EMRV quenchers on low torus water level. When torus water level is high, then there are added stresses to the EMRV downcomers when the EMRVs are opened. Answer A is incorrect.
Learning 2621.845.0.042 3000 Objective Using procedure EMG-3200.02, evaluate the technical basis for each step in the procedure, and apply this evaluation to determine correct Page 46 of 21 1
I OC ILT 07-1 RO NRC Exam KEY I courses of action under emergency conditions.
Question Source Bank I
1 I
Modified Bank 1 1 I I I New Question Cognitive Memory or X Comprehension Level: Fundamental 1:1 or Analysis Knowledge 10 CFR Part 55 Content:
1 55.41 18 I 55.43 1 Time to Complete: 1-2 minutes Page 47 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when an ATWS occurred.
The Unit Supervisor has directed that RPV injection be terminated and prevented, IAW Support Procedure 17, Termination and Prevention of Injection, and that RPV water level be lowered to 3 0 .
Which of the following states the Core Spray Main Pump and Feedwater/Condensate Pump configuration AFTER implementation of Support Procedure 17, and the basis for lowering RPV water level?
Pump Configuration Basis for 30" RPV Water Level A. All Core Spray Main Pumps in PTL The lowered water level will lower and ALL Feedwater/Condensate reactor power from increased voids Pumps OFF B. All Core Spray Main Pumps in PTL The lowered water level will reduce and ALL Feedwater/Condensate subcooling to minimize power Pumps OFF EXCEPT one oscillations Condensate Pump ON C. ALL Core Spray Main Pumps in PTL; The lowered water level will lower All Feedwater/Condensate Pumps reactor power from increased voids ON with MFRVs CLOSED D. All Core Spray Main Pumps ON with The lowered water level will reduce the Parallel Isolation Valves CLOSED; subcooling to minimize power All Feedwater/Condensate Pumps oscillations OFF Page 48 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
17 B Question Developer InitialdDate: NTP 11/19/07 Answer Knowledge and Ability Reference Information RO SRO 295031 EKl.03 Importance 3.7 4.1 Knowledge of the operational implications of the Rating following concepts as they apply to REACTOR LOW WATER LEVEL: Water level effects on reactor power 1 RO 1 Tier# 1 1 I I Level ~ )Group# 1 I
BAS-3200.02 (EOP References EMG-SP17 Users Guide)
The question stem describes a high power ATWS (power > 2%) and the requirement to terminate/prevent injection and to lower RPV water level to 30IAW the ATWS EOP. IAW SP17, all core spray main pumps are placed in pull-to-lock (PTL), and all feedwaterkondensate pumps except one are secured. This one Explanation: pump is required to supply the SJAE condensers to maintain main condenser vacuum. IAW the EOP Users Guide, RPV water level is lowered to minimize feedwater subcooling to prevent large power oscillations that could cause fuel damage. Answer B is correct. The other answers list the incorrect pump configuration or incorrect basis.
Learning 2621.845.0.0041 3055 Objective Given a copy of RPV Control, describe in detail each step or conditional statement, including technical basis, and how to perform each step as required.
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OC ILT 07-1 RO NRC Exam KEY Question Source
~~~ ~~
Bank I ~
1 I
Modified Bank
~
New X Question Cognitive Memory or X Comprehension Level: Fundamental 1:P or Analysis Knowledge 10 CFR Part 55 Content:
55.41 8 -T I 55.43 Page 50 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when an event occurred. Present plant conditions are as follows:
The MSlVs indicate green lights ON Reactor power is 60%
RPS GROUP SCRAM SOLENOIDS white lights are ON Both CRD Pumps are running Which of the following states a method to shutdown the reactor IAW Support Procedure 21, Alternate Insertion of Control Rods?
A. Bypass the automatic reactor scram signals and then reset the scram and manually scram the reactor.
B. Confirm closed the CRD Charging Header Supply valve, V-15-52, and raise CRD cooling water differential pressure.
C. Confirm the REACTOR MODE SELECTOR switch is in SHUTDOWN and manually drive rods with the CRD System.
D. Confirm closed the CRD Cooling Water PCV, V-15-24, and then vent the Control Rod Drive under piston volume.
Page 51 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
18 Question Developer InitiaWDate: NTP 11/20/07 Answer Knowledge and Ability Reference Information RO SRO 295037 EK2.05 Importance 4.0 4.1 Knowledge of the interrelations between SCRAM Rating CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following: CRD hydraulic system Level 1 RO ) T i e r # 11 IGroup# (1 References EMG-SP21 The reactor was at rated power when an event occurred. With the MSlVs indicating closed, the reactor must have scrammed. But with power at 60%, there must also be an ATWS. Since the RPS scram group solenoids are lit (these are lit when there is no scram condition present), then this is an electric ATWS. One method to insert control rods during an electric ATWS is to close the CRD charging header supply valve and to raise CRD cooling water AP. Answer B is correct.
Explanation: Bypassing the scram signals, resetting the scram and scramming again are actions for a hydraulic ATWS. Answer A is incorrect.
Manually driving control rods is always an option, but the reactor mode selector switch is in refuel - not shutdown. Answer C is incorrect. Venting the CRD over piston volume is an option for an electric ATWS, but answer D calls for venting the under piston volume - not the over piston volume. Answer D is incorrect.
Learning !621.828.0.0011 10450 Objective Iescribe and interpret procedure sections and steps for plant
!mergency or off-normal conditions that involve this system including iersonnel allocation and equipment operation IAW applicable ABNs, IOP & EOP Support procedures and EPIPs.
Page 52 of 21 1
oc LT 07-1 RO NRC Exam KEY Question Source Bank
~
1 Modified Bank lNew I Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis T r 3:SPK Knowledge
~
10 CFR Part 55 55.41 7 55.43 Content:
~~
Time to Complete: 1-2 minutes Page 53 of 211 I
OC ILT 07-1 RO NRC Exam KEY An event has occurred which caused entry into EMG-3200.12, Radioactivity Release Control. This procedure includes the following Conditional Statement:
IF the release is from the Turbine Building, THEN operate available Turbine Building ventilation per Support Procedure 51 Which of the following states the basis for this Conditional Statement?
A. To reduce the amount of radioactivity released.
- 6. To ensure a greater dilution factor during release.
C. To ensure the release is ONLY through an elevated release.
D. To ensure BOTH ground and elevated releases are monitored.
Page 54 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
Question Developer Initials/Date: NTP 11/20/07 Answer Knowledge and Ability Reference Information RO SRO 295038 EK2.03 Importance 3.6 3.8 Knowledge of the interrelations between HIGH OFF- Rating SITE RELEASE RATE and the following: Plant ventilation systems Level RO Tier # 1 Group# 1 EMG-3200.03 References (EOP Users EMG-3200.12 ENG-SP51 Guide)
~ _ _ _
The EOP for radioactivity release control is entered when an alert emergency classification from offsite release rate has been declared.
From the EOP User's Guide: "This Conditional Statement directs the operator to maintain the Turbine Building Ventilation System in service to preserve Turbine Building accessibility, and ensure that any radioactivity is discharged through a monitored release point, either the Main Stack for an elevated release, or via the Turbine Explanation: Building Stack, which is considered a ground level release. When required, Support Procedure - 51 provides the necessary directions for restarting the Turbine Building Ventilation System." Some of the TB vent systems started discharge to the main stack (elevated release; ie., Exhaust Fan EF 1-7) and some to the TB stack (ground release; ie., exhaust fan EF 1-1). Therefore, answer D is correct. All other distracters are plausible but incorrect.
Learning 2621.845.0.0012 2483 Objective Using procedure EMG-3200.12, evaluate the technical basis for each step and apply this evaluation to determine the correct course of action under emergency conditions.
Page 55 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New Bank Question Cognitive Memory or X Comprehension Level: Fundamental l:B or Analysis Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
Page 56 of 211
OC ILT 07-1 RO NRC Exam KEY An electrical fire started inside the 4160 Volt Switchgear C and D Vault. Which of the following states the fire suppression system and initiation method designed to suppress this fire?
Suppression System Initiation Method A. Halon 1301 Manual B. Dry pipe sprinkler Automatic C. Low pressure C02 Manual D. High pressure C02 Automatic Page 57 of 21 1
OC ILT 07-1 RO NRC Exam KEY Knowledge and Ability Reference Information RO SRO 600000 AA1.08 Importance 2.6 2.9 Ability to operate and / or monitor the following as Rating they apply to PLANT FIRE ON SITE: Fire fighting equipment used on each class of fire Level I RO I Tier # I 1 IGroup# 11 I References I 333.1 1 ABN-29 Learning 2621.828.0.001 9 286-10446 Objective identify and explain system operating controls/indications under all plant operating conditions.
Question Source I Bank 1 1 Modified Bank lNew 1 X Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis Knowledge l:F 10 CFR Part 55 Content:
1 1 55.41 io 1 55.43 1 Time to Complete: 1-2 minutes Page 58 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is starting up after an outage. The following conditions currently exist:
All IRMs are mid-range on Range 8 0 Turbine warming is in-progress with ALL Turbine Stop Valves open The Operator reports the following:
Annunciator COND VAC LO 25 INCHES alarms CONDENSER VACUUM 1A, 1B and 1C indicate 2 4 HG and are lowering at a rate of ?hHG/minute RPV pressure is 590 psig and is rising at a rate of 4 psig/minute Assume the rates above remain constant and the ONLY Operator action is to range the IRMs, if required.
The reactor will scram from which of the following scram signals?
A. The turbine trip in 3 minutes.
B. The turbine trip in 4 minutes.
C. Low condenser vacuum in 3 minutes.
D. Low condenser vacuum in 4 minutes.
Page 59 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
2, D Question Developer InitiaWDate: NTP 11/21/07 Answer Knowledge and Ability Reference Information 295002 AA2.02 Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM: Reactor power it Importance Rating RO SRO Level RO Tier # 1 Group # 2 I
References RAP-J1b RAP-Qlc ABN-10 I I The question stem shows that the reactor is starting up (with the mode switch in STARTUP) with RPV pressure e 600 psig. At this low pressure, the main condenser low vacuum scram signal and the turbine stop valve closure scram signal are bypassed. The low vacuum scram setpoint is 22 hg. In 3 minutes, condenser vacuum be still be above 22, but RPV pressure will be 602 psig and the low Explanation: vacuum scram is no longer bypassed. In 4 minutes, condenser vacuum drops to 22 hg, the scram and turbine trip setpoint. In this same 4 minutes, RPV pressure has risen to 606 psig. At this pressure, the low vacuum signal is no longer bypassed and the reactor will scram from a low vacuum scram signal. Answer D is correct.
Learning Objective I 2621.828.0.0051 249-10445 Given a set of system indications or data, evaluate and interpret them to determine limits, trends and system status.
Page 60 of 21 1
OC ILT 07-1 RO NRC Exam KEY I Question Source Question Cognitive Level:
10 CFR Part 55 Content:
Bank I Memory or Fundamental Knowledge 55.41 10 55.43 Comprehension or Analysis 7 2: DR Time to Complete: 1-2 minutes I
Page 61 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power, with the following abnormal lineup:
0 Isolation Condenser A DC Condensate Return Valve V-14-34 switch is in the CLOSE position A turbine trip then occurred. The following indication was noted for 3 seconds:
0 EMRV 108A indicated in the VALVE OPEN REGION Which of the following states the expected indications of the Isolation Condenser System and the reason for this indication?
Expected IC Indications Reason A. ISOL CONDENSER A LEVEL AND BOTH Isolation Condensers have ISOL COND B LEVEL begin to lower. automatically initiated.
B. ISOL COND A STEAM INLET NEITHER Isolation Condenser has temperature AND ISOL COND B automatically initiated.
STEAM INLET temperature remain at their initial value.
C. ISOL CONDENSER A PRESS BOTH Isolation Condensers have remains at its initial value and ONLY automatically initiated, BUT Isolation S O L CONDENSER B PRESS begins Condenser A has automatically to lower. isolated.
D. ISOL COND A STEAM INLET Isolation Condenser B has temperature remains at its initial value automatically initiated AND Isolation and ONLY ISOL COND B STEAM Condenser A initiation is defeated.
INLET temperature begin to rise.
Page 62 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question#
22 D Question Developer InitiaWDate: NTP 11/21/07 Answer Knowledge and Ability Reference Information RO SRO 295007 AK3.01 Importance 4.0 4.2 Knowledge of the reasons for the following responses Rating as they apply to HIGH REACTOR PRESSURE:
Isolation condenser operation Level 1 RO 1 Tier# 1 1 Group# I 307 References 420 RAP-C5a RAP-C1a The question shows the plant at rated power with the isolation condenser A DC condensate return valve switch in the closed position. This valve is normally closed with the switch in the auto position. With this switch in the closed position, IC A will not auto initiate. With this lineup, an event occurs which results in EMRV NRl08A opening for 3 seconds (due to a high RPV pressure). The pressure setpoint is above the setpoint for auto IC initiation for longer than the IC time delay (3 seconds EMRV and 1.5 second TD for IC Explanation:
initiation). Therefore, a condition existed where the ICs should have auto initiated. But since IC A is defeated, only IC B will initiate. When only the IC B initiates, the steam inlet temperatures rise but remain the same for IC A. Answer D is correct. Answer A is incorrect since this is the expected response if both ICs initiated. Since IC B auto initiated, answer B is incorrect. Regardless of which IC initiated, (but not isolated), the IC pressure will follow RPV pressure as it lowers.
Answer C is incorrect.
Learning 2621.828.0.0023 02030 Objective Describe the isolation condenser features and/or interlocks which provide the following: 1) automatic system initiation; 2) automatic system isolation.
Page 63 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 2: DR Knowledge I I I 10 CFR Part 55 55.41 5 55.43 Content:
Time to Complete: 1-2 minutes Page 64 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when an event occurred. Present plant conditions are as follows:
All control rods indicate green backlight EXCEPT 6 control rods which indicates position 04 Drywell pressure indicates 3.3 psig RPV water level is 144 and rising The following annunciators are in alarm:
o S1A BRKR OL TRIP/BRKR PERM OPN o S1B BRKR OL TRIP/BRKR PERM OPN Reactor Building AP indicates -0.1 Which of the following states the Abnormal Procedures (ABNs) and EOPs that should be implemented under the given conditions? (Other than ABN-1, Reactor Scram)
A. RPV Control - No ATWS EOP ABN-37, Station Blackout Secondary Containment Control EOP
- 6. RPV Control -With ATWS EOP ABN-36, Loss of Offsite Power Primary Containment Control EOP C. RPV Control - No ATWS EOP ABN-36, Loss of Offsite Power Primary Containment Control EOP D. RPV Control - No ATWS EOP ABN-37, Station Blackout Primary Containment Control EOP Page 65 of 21 1
OC ILT 07-1 RO NRC Exam KEY I Question #
Answer 1 23 Question Developer InitialdDate: NTP 11/21/07 I
Knowledge and Ability Reference Information RO SRO 295010 2.4.4 (High drywell pressure) Importance 4.0 4.3 Ability to recognize abnormal indications for system Rating operating parameters which are entry-level conditions for emergency and abnormal operating procedures:
Level RO Tier # 1 Group# 2 References EMG-3200.01A EMG-3200.02 ABN-36 The question stem shows that the reactor is shutdown by control rods, and that drywell pressure is above the entry condition for Primary Containment Control EOP and RPV Control - No ATWS. It also shows that the startup transformers have been lost (by Explanation:
annunciation). With the loss of only the startup transformers (with the turbine off-line), ABN-36 entry and implementation is required. The RB AP is below the entry into the Secondary Containment Control EOP.Answer C is correct.
None Learning 2621.845.0.0042 3025 Objective Given key plant parameters, determine if entry conditions for the EOPs have been met.
2621.845.0.0040 3052 State plant conditions requiring entry into RPV Control - No ATWS.
[ Question Source 1 I
Bank I I I 1 Modified Bank I
I 1
I New 1 I
Question Cognitive Memory or X Comprehension Level: Fundamental 1:P or Analysis Knowledge I 10 CFR Part 55 Content:
1 1 55.41 i o 1 55.43 1 I Time to Complete: 1-2 minutes Page 66 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when the Offgas Radiation Monitor readings rose slightly, and a small increase in the Stack RAGEMS was also observed. Chemistry has confirmed that a very small, but greater than normal, amount of fission products are present in the reactor coolant. Both Offgas Radiation Monitors and Stack RAGEMS readings are now currently trending down very slowly.
An explosion in the Offgas System then occurred. The following conditions are noted:
OFFGAS ISOL ACT I AND OFFGAS S O L ACT II annunciators in alarm CONDENSER OFF GAS AIR EXTRACTION VALVES V-7-1 through V-7-6 indicate closed If another similarly sized fuel failure were to occur now, under the plant conditions above, which of the following states the expected response of the Offgas Radiation Monitors and Stack RAGEMS? (NEGLECT any impact from the Steam Seal Exhauster Blower)
Offqas Radiation Monitors Stack RAGEMS A. Rise Continue to Lower
- 6. Continue to Lower Rise C. Rise Rise D. Continue to Lower Continue to Lower Page 67 of 211
OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP 11/24/07 Knowledge and Ability Reference Information RO SRO 295017 A A I .07 Importance 3.4 3.6 Ability to operate and/or monitor the following as Rating they apply to HIGH OFF-SITE RELEASE RATE:
Process radiation monitoring system I I Level RO Tier # 1 Group# 2 I I I References The question stem shows that slight fuel failures have resulted in an increase in the offgas radiation monitors and the stack radiation monitoring system (RAGEMS). The radiation monitors are currently lowering. Offgas discharges to the stack, after being processed in the offgas system. The alarms provided in the stem describe an event which resulted in either high pressure or high temperature in the offgas system (in both sensor channels I and II), and this has resulted Explanation: in an offgas system isolation from the main condenser, and offgas system flow goes to 0. With the offgas system isolated from the condenser, the offgas radiation monitors will not detect further fuel failures and the offgas radiation monitors will continue to lower.
Likewise, since there is no further offgas flow to the stack, the stack RAGEMS will also continue to lower. Answer D is correct.
Learning 2621.828.0.033A 273-10435 Objective Given plant operating conditions, describe or explain the purpose/function of the system and its components.
Page 68 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Source 1 Bank I 1I Modified Bank 7
I Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
I Time to Complete: 1-2 minutes Page 69 of 211
OC ILT 07-1 RO NRC Exam KEY Consider the following two plant conditions:
- 1. Condition 1: The plant is at 800 psig during a startup
- 2. Condition 2: The plant is at rated power Which of the following is correct regarding a reactor scram from the SUSTAlNED loss of both CRD Pumps?
Condition Action A reactor scram is required IMMEDIATELY due to the ...
A. 1 LOWER RPV pressure.
2 I HIGHER RPV pressure. I 1 1 REDUCED control rod worth. I D. 2 FASTER CRD seal degradation.
Page 70 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitialdDate: NTP 11/24/07 295022 AKI .02 Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS: Reactivity control Level References RO Tier #
RAP-HI c h Group#
Importance 2
Rating I-RO With the sustained loss of both CRD pumps, charging water has also SRO been lost to the CRD accumulators. On a scram, the accumulators provide the initial force to insert the control rods. As the accumulator pressure decays away, the HCU check valve will re-position to allow RPV pressure to provide the force to complete the insertion of the control rods. When the charging supply is lost at lower RPV pressures, the driving force from the RPV has been reduced, and this Explanation: can impact control rod scram times. The RAP requires a scram if charging water cannot be immediately restored, with RPV pressure e 850 psig. Answer A is correct. At rated power (or any pressure > 850 psig), the scram can be delayed until 2 accumulator alarms are received. Answer B is incorrect. Rod worth and seal failures are not the correct reason. Answers C and D are incorrect.
References to provided durii Learning 2621.828.0.001 1 10444 Objective State the function and interpretation of system alarms, alone and in combination, as applicable, in accordance with the system RAPS.
Page 71 of 211
oc LT 07-1 RO NRC Exam KEY Question Source Bank 1 I Modified Bank 1 1 New Question Cognitive Memory or Comprehension x Level: Fundamental or Analysis 2:DR Knowledge 10 CFR Part 55 55.41 8 55.43 Content:
Page 72 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is shutdown for a refuel outage in mid-July. The following conditions currently exist:
Fuel shuffling is underway SGTS Fan 2 is out of service An event occurred on the refuel floor that resulted in the following radiation monitoring annunciators alarming:
0 AREAMONHI 0 NORTH WALL C10 HIGH OPER FLOOR B9 HI VENT TRIP RX BLDG -VENT HI Which of the following states the plant impact of these conditions?
A. Turbine Building AP will immediately begin to lower.
B. Refuel floor radiation levels will begin to lower in two minutes.
C. Reactor Building general area temperatures will begin to rise.
D. Total air flow through the Main Stack has dropped to 2600 SCFM.
Page 73 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
26 C Question Developer Initials/Date: NTP 11/24/07 Answer Knowledge and Ability Reference Information RO SRO 295034 EKI .02 Importance 4.1 4.4 Knowledge of the operational implications of the Rating following concepts as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION:
Radiation releases Level RO Tier # 1 Group# 2 References 1 RAP-1OFlf I BR 2011 I The question stem describes an event on the refuel floor that has initiated an immediate signal to start SGTS and trip RB normal ventilation, and a signal to perform these same functions after a 2-minute time delay. Thus, S G T S has immediately started and the normal RB ventilation system has immediately secured and isolated.
Normal RB vent flow is 76,700 SCFM. The flow through one SGTS fan is 2600 SCFM. Cooling for the RB is from an air wash system on the inlet air supply to the normal RB ventilation fans. Since the normal fans have tripped, there is no longer any air flow through the air wash system. The SGTS provides no cooling to the RB atmosphere. Thus, with the air cooling function removed, RB general air temperatures will begin to rise. Answer C is correct.
The events described in the question stem only impact the RB Explanation: atmosphere. Turbine Building AP compares pressures in the TB to the outside pressure, not against the RB inside pressure. The event in the question stem does not impact either the TB inside pressure (since TB vent is not impacted) nor the outside pressure. Answer A is incorrect. As stated, one train of SGTS flows 2600 SCFM and it does discharge to the main stack. But the TB vent system, which was not affected by the event, is still operating and discharging to the main stack. Thus main stack flow is 2600 + TB vent flow, not just 2600 SCFM from SGTS. Answer D is incorrect. The radiation levels on the refuel floor are due the source of the radiation. It may be incorrectly concluded that SGTS will start after the 2-minute time delay and that this will impact the radiation levels on the floor. Since SGTS is already running and the radiation source is not impacted, radiation levels will not change in 2 minutes. Answer B is incorrect.
Page 74 of 21 1
OC ILT 07-1 RO NRC Exam KEY Learning 2621.828.0.0042 261 -1 0445 Objective Given a set of system indications or data, evaluate and interpret them to determine limits, trends, and system status.
I Modified New Question Source Bank X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 8 55.43 Content:
Time to Complete: 1-2 minutes Page 75 of 21 1
oc LT 07-1 RO NRC Exam KEY
. t ' <
Question
... . -27
' . .. . . I : .>.
Which of the following states why BOTH hydrogen and oxygen levels in the Primary Containment are monitored during accident conditions?
A. Inadequate core cooling during a LOCA could result in combustible levels of both hydrogen and oxygen.
B. Exceeding LHGR prior to a LOCA could result in combustible levels of both hydrogen and oxygen from radiolysis.
C. Both hydrogen and oxygen concentrations need to be known to accurately determine amount of fuel damage.
D. The uranium-water reaction generates large amounts of both hydrogen and oxygen that could lead to Primary Containment failure.
Page 76 of 211
OC ILT 07-1 RO NRC Exam KEY I Question ## 1 27 A 1 Question Developer InitialdDate: NTP 11/26/07 I Knowledge and Ability Reference Information 1 RO I SRO 500000 EK2.02 Importance I 3.1 I 3.5 Knowledge of the interrelations between HIGH CONTAINMENT HYDROGEN CONCENTRATIONS the following: Containment oxygen monitoring systems Level RO Tier # 1 Group # 2 References I BAS-3200.02 (EOP Users Guide) I ,
IAW the reference, a LOCA resulting in inadequate core cooling could result in the generation of large amounts of hydrogen.
Combined with oxygen, a combustible mixture could result. Answer A is correct.
IAW the reference, radiolysis does produce hydrogen at rates that are generally not of concern. Answer B is incorrect.
Explanation: Other methods exist to determine the amount of fuel damage other than by knowing H a 0 2 levels. Ifthese levels are not known, fuel damage can still be estimated. Answer C is incorrect.
IAW the reference, a combustible mixture could fail the Primary Containment, but hydrogen generation is from the metal-water (clad-water) reaction, not a uranium-water reaction. Answer D is incorrect.
Learning 2621.845.0.0042 3000 Objective Using procedure EMG-3200.02, evaluate the technical basis for each step in the procedure, and apply this evaluation to determine the correct courses of action under emergency conditions.
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oc LT 07-1 RO NRC Exam KEY Question Source I Modified Bank I 1 1 New Question Cognitive Level:
Memory or Fundamental Knowledge
,:F 1 X Comprehension or Analysis 10 CFR Part 55 7 I 55.43 I Content: I I
55m41II I I I Time to Complete: 1-2 minutes Page 78 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is shutdown and a cooldown is in-progress with the Shutdown Cooling system. Present plant conditions are as follows:
0 RPV coolant temperature is 325 O F and lowering 0 RPV water level is 155 0 Shutdown Cooling Pumps A, B, and C are in-service Which of the following annunciators would indicate a condition in which ALL Shutdown Cooling Pumps are automatically tripped?
A. TORUS/DRWELL - DW PRESS HVLO B. SHUT DN CLG - SD HX PUMP RM TEMP HI C. REACTOR LEVEL - RX LVL LO I AND RX LVL LO II D. CORE SPRAY - SYSTEM 1 AUTOSTART AND SYSTEM 2 AUTOSTART Page 79 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
28 Question Developer InitialdDate: NTP 11/26/07 Answer Knowledge and Ability Reference Information RO SRO 205000 K5.02 Importance 2.8 2.9 Knowledge of the operational implications of the Rating following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): Valve operation Level RO Tier # 2 Group# 1 EMG-SP1 RAP-H5e RAP-Ble References RAP-C8d RAP-B3g RAP-C2d The question stem shows the reactor shutdown and cooling down with SDC. All pumps trip when SDC inlet isolation valve V-17-19 closes. SDC receives an isolation signal at RPV water level 10-10 (90) or high drywell pressure (>3.0 psig). This will close the SDC inlet/outlet isolation valves V-17-19 and V-17-54. The closure of the SDC inlet isolation valve will trip all SDC pumps. Core spray will initiate from either RPV water level 10-10 OR drywell high pressure.
The receipt of Core Spray System 1 Autostart (or 2) annunciator will start core spray and the EDGs. These same parameters will close Explanation: SDC inlet isolation valve, which will trip all SDC pumps. Answer D is correct.
The drywell hillo pressure alarm alarms at 1.4 psig, which is not high enough to cause a SDC isolation of 3 psig. Answer A is incorrect.
There is no automatic action if a SDC leak is detected in the SDC pump room. Answer B is incorrect.
The RPV low water level annunciator alarms at about 138, which is above the 10-10 level setpoint to isolate SDC. Answer C is incorrect.
Learning 2621.828.0.0045 205-10440 Objective Given the system logic/electrical drawings, describe the system nauto isolation signals, setpoints, and expected system response including Page 80 of 21 1
OC ILT 07-1 RO NRC Exam KEY I power loss or failed components.
Modified Question Source Bank New X Bank Comprehension X Level: or Analysis 2: DR Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
Page 81 of 211
. , .. . t Question 29
. I r The plant was at rated power when an event occurred. Present plant conditions are as follows:
0 RPV pressure is 700 psig and lowering 0 RPV water level is 80 0 Core Spray has auto started The MSlVs are closed 0 Drywell pressure is 2 psig and rising slowly 0 ABN-1, Reactor Scram, Immediate Operator Actions have been performed Which of the following will ensure that the Isolation Condensers maintain the MAXIMUM heat transfer capability?
A. Verifying the Isolation Condenser vent valves remain open to prevent air binding in the condensers.
- 6. Verifying the Isolation Condenser vent valves remain open to prevent shell pressurization from non-condensable gases.
C. Ensuring the Isolation Condenser shell water levels are properly maintained as required, from the Condensate Transfer System.
D. Ensuring the Isolation Condenser shell water levels are properly maintained as required, from the Demineralized Water System.
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OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP 11/26/07 Knowledge and Ability Reference Information RO 1 SRO 207000 K5.03 Importance 2.7 3.0 Knowledge of the operational implications of the Rating following concepts as they apply to ISOLATION (EMERGENCY) CONDENSER: Heat transfer I I Level RO Tier # 2 Group# 1 References I
307 r I n
I The question describes an event where RPV water level is below the 10-10 setpoint: with drywell pressure at 2 psig and core spray running, then RPV water level must be below the lo-lo setpoint. With water level at this level and lowering (and core spray is not injecting due to RPV pressure), then the isolation condensers have also auto initiated (RPV lo-lo level or high RPV pressure). When the ICs auto initiate, the IC vent valves close. To maximize heat transf.er of the ICs, the Explana ion:
tube bundles' must remain covered with water. Makeup to the ICs, when they are in service, is from the condensate transfer system.
Answer C is correct. Makeup from demineralized water is used when the ICs are in a standby condition. Answer D is incorrect. Since the IC vent valves have auto closed, answers A and B are incorrect. The vent valves do not close on a automatic system isolation.
References ta provided duril Learning 2621.828.0.002310457 0bjective Describe the sequence of operation of the isolation Condenser System when an initiation signal is received.
2621.828.0.002302029 Describe the relationship between the Isolation Condenser System and the following: Condensate transfer System Page 83 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank X New Bank
~~~
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 1 X 2:DR 10 CFR Part 55 55.41 5 55.43 Content:
Page 84 of 21 1
. je ,! I . . :
The plant was at rated power when a manual scram was inserted. A fire has disabled USS 1B3 and the plant is now shutdown and is cooling down with Isolation Condenser B in service. The following annunciator then alarmed:
ISOL COND -SHELL B LVL HVLO The Operator reports the S O L CONDENSER B LEVEL indicates 7 2 and that ISOL COND B SHELL temperature indicates 104 O F .
Which of the following states the impact on the Isolation Condensers from this alarm and the action IAW the RAP?
Isolation Condenser B Required Action Impact A. The shell level is high Verify ISOLATION CONDENSER MAKEUP valves closed; perform those actions to drain the IC B shell IAW Procedure 307, Isolation Condenser System.
B. The shell level is high Verify ISOLATION CONDENSER MAKEUP valves closed; request Chemistry to sample the IC B shell water, and isolate Isolation Condenser B if evidence of a tube leak exists.
C. The shell level is low Add makeup to the Isolation Condenser B using the preferred water source; use of the associated ISOLATION CONDENSER MAKEUP valve is NOT required.
D. The shell level is low Add makeup to the Isolation Condenser B using the alternate water source (NOT local hoses);
use of the associated ISOLATION CONDENSER MAKEUP valve is required.
Page 85 of 21 1
OC ILT 07-1 RO NRC Exam KEY Knowledge and Ability Reference Information RO SRO 207000 A4.06 Importance 3.8 4.0 Ability to manually operate and/or monitor in the Rating control room: Shell side makeup valves
' I Level RO Tier # 2 Group# 1 References I
I307 1 RAP-C6b I I I
The question stem describes the plant cooling down with the isolation condenser B, with USS 1B3 inoperable due to a fire. The indications provided show a low level in the IC shell, not high. A high level could occur from a tube leak. Because the IC B is in-service, a higher shell temperature is expected and normal. Because the isolation condenser is in service, makeup to the shell is normally provided by condensate transfer (preferred source) or fire water (alternate source)
(through the normal path or through the use of fire hoses). Because USS 183 is de-energized, both condensate transfer pumps are lost and thus the normal preferred makeup method utilizing condensate Explanation: transfer is lost. Using fire protection (the alternate water source) (not local hoses) is available, and the lineup includes both in-plant manipulations and control room manipulations. When the isolation condenser is in standby, makeup requires no control room manipulations. Answer D is correct.
Answers A and B are incorrect since the shell level is low - not high.
On a high level, the answers provided are plausible and IAW procedure. Answer C is incorrect since condensate transfer is not available.
Learning 2621.828.0.0023 02029 Describe the relationship between the Isolation Condenser System and the following: Condensate transfer System, Demineralized Water System.
Page 86 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
Page 87 of 211
OC ILT 07-1 RO NRC Exam KEY The plant was operating at rated power. A reactor recirculation leak has caused an automatic scram. The following conditions currently exist:
0 Core Spray Pumps NZOlNNZ03A are injecting into the RPV at 4000 GPM 0 Core Spray Pumps NZO1B/NZ03B are injecting into the RPV at 4200 GPM Torus water temperature is 150" F 0 Torus water level is 120" 0 RPV water level is 9 8 TAF and rising (Torus Pressure - STRAINER) equals -2.9 psig Which of the following, if any, is required regarding the Core Spray System?
A. Secure Core Spray Booster Pump NZ03A.
B. Secure Core Spray Booster Pump NZ03B.
C. No actions regarding Core Spray are required.
D. Secure BOTH Core Spray Booster Pumps NZ03A and NZ03B.
Page 88 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer Initials/Date: NTP 11/26/07 Knowledge and Ability Reference Information RO SRO 209001 A2.09 Importance 3.1 3.3 Ability to (a) predict the impacts of the following on Rating the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low suppression pool level Level 1 RO 1 Tier # 12 I Group# 1 I
~~~ ~
References EMG-SP4 From the reference provided, it can be ascertained that core spray system B is on the bad side of the core spray NPSH limit curve, and will be secured. Answer B is correct.
Learning 2621.845.0.0040 10445 Objective Given a set of system indications or data, evaluate and interpret then to determine limits, trends, and system status.
Modified Question Source Bank X New Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
I I I I I Time to Complete: 1-2 minutes Page 89 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when an event occurred. Present plant conditions are as follows:
0 Drywell pressure is 3.6 psig and rising 0 RPV water level is 1 2 0 and rising 0 FEED PUMPS DISCHARGE PRESSURE indicates 800 psig The Operator notes the following Core Spray System indications:
0 MAIN PUMP AMPS NZOlA indicates 50 AC AMPERES 0 MAIN PUMP AMPS NZO1D indicates 0 AC AMPERES 0 SYS 1 FLOW indicates approximately 100 GPM 0 SYS 2 PUMP DISCH PRESS BOOSTERS indicates approximately 330 psig Which of the following is correct regarding the observed Core Spray indications?
A. Core Spray Pump NZOlD has tripped.
B. Core Spray Pump NZOlA is currently running on minimum flow.
C. Core Spray System 2 is NOT currently indicating the expected discharge head.
D. Core Spray System 1 CANNOT provide core cooling when the RPV depressurizes.
Page 90 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
32 B Question Developer InitialdDate: NTP 11/28/07 Answer Knowledge and Ability Reference Information RO SRO 209001 A4.11 Importance 3.7 3.6 Ability to manually operate and/or monitor in the Rating control room: System flow Level RO Tier # 2 Group# 1 RAP-61E!
References I341 USAR 6.3.1.3.3 RAP-62e The question stem describes the plant at power when an event resulted in a low RPV water condition and a high drywell pressure condition. Under the given conditions, core spray 1 (main pump A and booster pump a) and core spray 2 (main pump B and booster pump B) will start. With feedwater discharge pressure at 800 psig, then RPV pressure is close to this value. With core spray running at an RPV pressure > 305 psig, the core spray parallel isolation valves are closed and core spray is running on minimum flow back to the torus. This flow is approximately 100 gpm. Therefore, core spray A is running on minimum flow. Answer B is correct.
As stated, core spray A and B start on their signals. Core spray C and Explanation: D will still be in standby (off), unless a preferred core spray system fails. Since there is no indication of this in the question stem, then core spray D will be off and no amps is the expected condition - not tripped. Answer A is incorrect.
With core spray system B running on minimum flow, the discharge pressure is approximately as listed in answer C. Answer C is incorrect.
Answer D is incorrect since the provided indications are the expected indications, and core spray A will provide core cooling, as designed, when RPV pressure drops < 305 psig. Answer D is incorrect.
Learning 1 2621.828.0.001 0 209-10444 Page 91 of 21 1
OC ILT 07-1 RO NRC Exam KEY Objective Describe the interlock signals and setpoints for the affected system components and expected system response including power loss or failed components.
Modified Question Source Bank New X Bank I orComprehension 1
~~
Question Cognitive Level:
Memory or Fundamental Knowledge Analysis I X 3:SPK 10 CFR Part 55 Content:
1 55.41 17 1 55.43 1 Time to Complete: 1-2 minutes Page 92 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is at rated power.
If Standby Liquid Control System (SLC) manual valve V-19-25 (SLC System Discharge Header Isolation Valve) is in the closed position when the Operator placed the STANDBY LIQUID CONTROL switch to the FIRE SYS 2 position, which of the following states the expected plant response?
Reactor Water Cleanup SLC System 2 SQUIBS SLC PUMP DISCH System Liqht PRESS A. NOT isolated Off < RPV pressure B. NOT isolated On > RPV pressure C. Isolated Off 0 psig D. Isolated On > RPV pressure Page 93 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitialdDate: NTP 11/28/07 Knowledge and Ability Reference Information RO SRO 21 1000 Al.09 Importance 4.0 4.1 Ability to predict and/or monitor changes in Rating parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including:
SBLC system lineup Level 1 RO I Tier# 12 1 Group#
t 1
References 1 I
GE 148F723 I I
RAP-Glb I
NOTE: DELETE Note 10 on P&iD With the SLC discharge header isolation valve closed when SLC is activated, there will be no SLC flow to the RPV. Also, RWCU does not isolate as expected since the flow switch necessary to isolate RWCU is downstream of the header isolation valve and it does not see any SLC flow. When the SLC system is activated, the normally OFF squibs light will energize. With the SLC flow path isolated, and SLC pumps are positive displacement pumps, the pump pressure will Explanation: be greater than RPV pressure. The pressure transmitter is upstream of the isolation valve and will indicate SLC system pressure. Answer B is correct.
Answer A is incorrect since the squibs light is on. Answers C and D are incorrect since RWCU will not isolate in this condition when SLC is initiated.
Learning 2621.828.0.0046 21 1-10438 Objective Using the system P&ID, locate each of the system components and explain its operation and limitations within the system.
Page 94 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank l x I Bank New Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR T r Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
I I I Time to Complete: 1-2 minutes Page 95 of 211
OC ILT 07-1 RO NRC Exam KEY Which of the following pairs of instruments, when their trip setpoint is EXCEEDED, will result in de-energizing ALL RPS SCRAM SOLENOID GROUPS?
0 RPV Pressure RE03A, B, C, D Drywell Pressure RE04A, 9,C, D A. RE03A AND RE03C B. RE03B AND RE04D C. RE03D AND RE04D D. RE03C AND RE04B RE03A 7k RE03D RE04A -fL REO4D 1 K51 1 K52 2K51 2K52 Page 96 of 211
OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP 11/29/07 Knowledge and Ability Reference Information RO SRO 212000 K5.02 Importance 3.3 3.4 Knowledge of the operational implications of the Rating following concepts as they apply to REACTOR PROTECTION SYSTEM: specific logic arrangements Level I RO (Tier# ( 2 (Group# 11 References I 237E566 sh 1, 3,537 RPS scram logic is 1 out of 2 taken twice logic. RPV pressure instruments RE03A and RE03C input into RPSl, as do drywell pressure instruments RE04A and RE04C. A trip of any of these instruments will generate a '/2 scram, regardless of how many have tripped. RPV pressure instruments RE03B and RE03D input into RPS2, as do drywell pressure instruments RE04B and RE04D. A trip Explanation: of any of these instruments will generate a '/2 scram, regardless of how many have tripped. A reactor scram requires a trip in RPSl and in RPS2. Whan a scram signal is generated, all RPS SCRAM SOLENOID GROUPS will be de-energized (lights out). The only combination that trips both RPSl and RPS2 is answer D. All others generate a % scram only.
Learning 2621.828.0.0037 212-10445 Objective Given a set of system indications or data, evaluate and interpret them to determine limits, trends, and system status.
Page 97 of 21 1
I OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank I I Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 Content:
I 55.41 15 I I 55.43 Time to Complete: 1-2 minutes Page 98 of 21 1
.. 1
.. 1 . !
The plant is starting up after an outage. All IRMs are approximately mid-scale on Range 6 and steady.
The following annunciator then alarmed:
NEUTRON MONITORS - IRM DNSCL The Operator reports that the IRM ALL IN light is EXTINGUISHED and IRM 11 indicates downscale on Range 6.
Which of the following states the cause of these indications and the plant response?
Cause Plant Response A. IRM 11 has been driven out of the Control rod withdrawal block ONLY core B. IRM 11 has been driven out of the Control rod withdrawal block AND a core % scram C. IRM 11 has failed downscale Control rod withdrawal block ONLY D. IRM 11 has failed downscale Control rod withdrawal block AND a
% scram Page 99 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
35 A Question Developer InitiaWDate: NTP 11/29/07 Answer Knowledge and Ability Reference Information 215003 K6.03 Knowledge of the effect that a loss or malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM: Detector drive motor Level References RO Tier #
RAP-H7a 2 Group #
RAP-G4e Importance Rating 1
~
The question describes a plant startup (with power ascension halted) with the reactor mode switch in startup and power being monitored by the IRMs. Two indications are then provided: an IRM is downscale,
~~
SRO along with the IRM all in light OUT. The all in light is expected to be lit under a normal startup. This would show that all IRMs are fully inserted. With the all in light out, then at least one IRM is not fully Explanation: inderted. Therefore, one IRM has been driven out of the core. As the IRM is driven out of the core, the count rate would go down, thus the IRM downscale annunciator. A downscale IRM will generate a control rod withdraw block. Answer A is correct.
A failed downscale IRM would give the annunciator in the stem but it would not explain the all in light. There is no M scram from a downscale IRM. Other answers are plausible but incorrect.
Learning 2621.828.0.0029 215-10444 Objective Describe the interlock signals and setpoints for the affected system components and expected system response including power loss or failed components.
Page 100 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Source Bank I I Modified Bank 1 lNew I Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3: PEO Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
Page 101 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is starting up after an outage. The Operator is withdrawing control rods with the reactor subcritical. Current plant conditions are as follows:
SRM 21 indicates 100 CPS SRM 22 indicates 120 CPS 0 SRM 23 indicates 110 CPS SRM 24 indicates 115 CPS All IRMs indicate downscale The following annunciators then alarm:
NEUTRON MONITORS - SRM DNSCL NEUTRON MONITORS - SRM HI-HI The SRMs indications are as follows:
SRM 21 indicates 0 CPS SRM 22 indicates 120 CPS SRM 23 indicates 110 CPS SRM 24 indicates 1E6 CPS Which of the following states the impact on RPS and/or RMCS from this event?
Impact from SRM 21 Impact from SRM 24 A. Rodblock ONLY Rodblock ONLY B. Rodblock ONLY Rodblock AND ?h scram C. Rodblock AND ?hscram Rodblock AND ?h scram D. None Rodblock ONLY Page 102 of 21 1
OC ILT 07-1 RO NRC Exam KEY I I I I Question ##
36 D Question Developer InitiaWDate: NTP 11/29/07 Answer Knowledge and Ability Reference Information RO SRO 215004 K4.01 Importance 3.7 3.7 Knowledge of SOURCE RANGE MONITOR (SRM) Rating SYSTEM design feature(s) and/or interlocks which provide for the following: Rod withdrawal blocks I Level 1 RO I Tier# 12 1 Group# 1 I References I RAP-G5d I RAP-H7a ~ _ _ _
The question stem describes a plant startup when 2 SRMs fail: one fails downscale and one fails upscale. Plant impacts from SRM failures are not bypassed since IRMs are downscale. A downscale Explanation: SRM has no impact (it does provide a rod block if not fully inserted; all SRMs are fully inserted at this point in the startup) and an upscale SRM provides a rod block only. Answer D is correct. Neither type of SRM failure produces a scram signal.
components and expected system response including power loss or failed components.
Modified Question Source Bank New X Bank Question Cognitive Memory or X Comprehension Level: Fundamental 1:F or Analysis Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
ITime to Complete: 1-2 minutes Page 103 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power, with APRM 1 bypassed, when the following annunciator alarmed:
NEUTRON MONITORS - LPRM HI The Operator reports that LPRM 28-33A (input into APRM 1) indicates upscale.
Which of the following states the impact of this failure to APRM 1 indicated reactor power on Panel 4F AND to reactor power as calculated by heat balance?
Impact on APRM 1 Impact on Heat Balance Indicator/Recorder on Panel 4F A. Indicates higher Indicates higher B. Indicates higher No impact C. No impact Indicates higher D. No impact No impact Page 104 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question#
37 B Question Developer InitiaWDate: NTP 12/2/07 Answer Knowledge and Ability Reference Information I RO I SRO 215005 K1.07 Importance 2.6 2.9 Knowledge of the physical connections and/or Rating cause/effect relationships between AVERAGE POWER RANGE MONITOWLOCAL POWER RANGE MONITOR SYSTEM and the following:
Process computer, performance monitoring system I
Level RO Tier # 2 Group# 1 References RAP-G6f NF-AB-770 Explanation:
References to provided durii Learning Objective 2621.828.0.0029 215-10445 Given a set of system indications or date, evaluate and interpret then to determine limits, trends or system status.
Page 105 of 21 1
~ __ _ _ ~ ~
Modified Question Source 1 Bank New X Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis 1:1 Knowledge 10 CFR Part 55 Content:
I I Time to Complete: 1-2 minutes Page 106 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power. Current plant conditions are as follows:
0 RECIRC TOTAL FLOW on Panel 4F indicates 150.0 X 1O3 GPM Reactor Recirculation Pump C is in local manual control, and is locked at its present speed All recirculation pump flows are matched Which of the following would result in a flow comparator alarm and ROD BLOCK?
A. Division 1 total reactor recirculation flow output fails to 140 x lo3 GPM.
B. Division 2 total reactor recirculation flow output fails to 160 x 1O3 GPM.
C. Reactor Recirculation Pump B flow transmitter FT-IA6OB fails to 0 GPM.
D. Reactor recirculation flow is lowered 10% by the MASTER RECIRC SPEED CONTROLLER.
Page 107 of 21 1 i
OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP 12/1/07 Knowledge and Ability Reference Information 1 RO I SRO 215005 A3.05 Ability to monitor automatic operations of the AVERAGE POWER RANGE MONITOWLOCAL POWER RANGE MONITOR SYSTEM including:
Flow converterkomparator alarms Level I RO 1 Tier # 12 1 Group# 1 References USAR 7.5.1.8.7 RAP-H7a RAP-G5f 3 The question stem describes the plant at rated power with one recirculation pump locked in local manual control. The speed of this pump is not affected by the master recirc speed controller. The comparator mismatch alarm is received at 10% or 16,000 GPM difference between the division 1 total reactor recirculation flow and the division 2 total recirculation flow. The original total flow is 150,000 GPM, and with all pumps matched at 30,000 GPM. Each recirculation loop has 2 flow transmitters. One set of 5 transmitters input into division 1 total recirculation flow and the second set of 5 transmitters input into division 2 total recirculation flow. When one individual lopp flow transmitter fails in only 1 division, that division will drop by Explanation:
30,000 gpm for a total loop flow of 120,000 and a drop of 30,000. Div 2 still sees 150,000 gpm. This difference is greater than that required for a flow comparator trip. Answer C is correct.
Answer A total flow drops by only 10,000 and answer 6 only rises by 10,000. Both are too small and there is no flow comparator trip.
Answers A and B are incorrect.
In answer D, recirculation flows are changed in 4 recirculation pumps.
This would result in the flows being unbalanced among the pumps, but division 1 flow would still equal division 2 flow and no comparator alarmhodblock is present. Answer D is incorrect.
Learning I 2621.828.0.0029 215-10445 Page 108 of 21 1
oc LT 07-1 RO NRC Exam KEY Objective Given a set of system indications or date, evaluate and interpret them to determine limits, trends, and system status.
Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 Content:
I I
55.41 17 I
1 I
55.43 1 I I Time to Complete: 1-2 minutes Page 109 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when a failure resulted in the following annunciator:
ADS SV/EMRV - EMRV POWER LOST/DISABLED Investigation has revealed that control power from 125 VDC Panel F to the EMRVs has been lost. Which of the following states the impact of this event if an Emergency Depressurization (ED) became necessary?
A. NO EMRVs will function. ED can be accomplished with the Turbine Bypass Valves and/or the Isolation Condensers.
B. ALL EMRVs will function. There is no need to supplement ED with the Turbine Bypass Valves and/or the Isolation Condensers.
C. EMRVs NR108A and NR1086 ONLY will function. ED can be supplemented with the Turbine Bypass Valves and/or the Isolation Condensers.
D. EMRVs NR108C, NR108D and NR108E ONLY will function. ED can be supplemented with the Turbine Bypass Valves and/or the Isolation Condensers.
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OC ILT 07-1 RO NRC Exam KEY Knowledge and Ability Reference Information 218000 A2.05 Importance I Ro 3.4 SRO 3.6 Ability to (a) predict the impacts of the following on Rating the AUTOMATIC DEPRESSURIZATION SYSTEM ;
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of A.C.
or D.C. power to ADS valves I
Level RO Tier # 2 Group # 1 References 729E182 RAP-B5g I
Explanation:
References to provided durii Learning Objective 2621.828.0.0005 00369 State how the following systems interrelate with ADS: Vessel and Primary Containment instrumentation; Core Spray; NSSS; Vital AC Power; and 125 VDC Power.
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I I Modified Question Source Bank I Bank New X Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis 1:1 Knowledge 10 CFR Part 55 55.41 5 Content:
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Time to Complete: 1-2 minutes Page 112 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when the turbine tripped and all Turbine Bypass Valves failed open. With NO Operator action, which of the following states the ultimate status of the RPS MSlV relays 1K73, 1K74, 2K73, and 2K74? (see attached drawing)
A. All relays will be energized.
B. All relays will be de-energized.
C. ONLY 1K73 and 2K73 relays will be de-energized.
D. ONLY 1K74 and 2K74 relays will be de-energized.
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OC ILT 07-1 RO NRC Exam KEY a 7 I I 1 lKlD IWF-t 1
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OC ILT 07-1 RO NRC Exam KEY Page 115 of 21 1
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Question #
4o B Question Developer InitiaWDate: NTP 12/3/07 Answer Knowledge and Ability Reference Information RO SRO 223002 Al.04 Importance 2.6 2.8 Ability to predict and/or monitor changes in Rating parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR 1
STEAM SUPPLY SHUT-OFF controls including:
Individual system relay status Level 1 RO ,I Tier# 12 Group # 1 References 237E566 sh 2, 3, 6, 7, 12, 13 1 When the turbine trips from rated power, the turbine bypass valves (TBVs) usually act to control RPV pressure. When the TBVs fail open, and no operator action, the RPV will depressurize. Since the Reactor Mode switch is still in RUN, the RPV will depressurize to less than 825 psig, at which time the MSlVs will automatically close to Explanation: stop the rapid cooldown. RPS relays 1K117, 1K118,2K117, and 2K118 will de-energize which cause relays 1K73, 1K74, 2K73, and 2K74 to de-energize. The relays must be energized to open the MSIVs, and de-energized to close. With these relays de-energized, the MSlVs will automatically close. Answer B is correct.
Learning 2621.828.0.0030 02456 Objective Describe RPS isolation logic trip signals and function, including the following: purpose/design basis; setpoints; conditions that allow bypassing isolation signals; how bypassing isolation signals is accomplished.
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I Question Source Bank Modified Bank New X Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
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OC ILT 07-1 RO NRC Exam KEY The plant was at 80% power and shutting down, with the following abnormal switch configuration:
AUTO DEPRESS VALVE NR108A switch is in the OFF position The NORMAUDISABLE switch for EMRV NR108B is in the DISABLE position Which of the following states those EMRVs which can function in the Pressure Relief Mode to control RPV pressure and/or in the ADS Mode during a LOCA?
Pressure Relief Mode ADS Function A. EMRVs B, C, D and E ONLY All EMRVs
- 6. EMRVs A, C, D and E ONLY All EMRVs C. EMRVs C, D and E ONLY EMRVs A, C, D and E ONLY D. EMRVs C, D and E ONLY EMRVs B, C, D and E ONLY Page 118 of 211
OC ILT 07-1 RO NRC Exam KEY Question#
4, C Question Developer InitiaWDate: NTP 12/3/07 Answer Knowledge and Ability Reference Information RO SRO 239002 K3.02 Importance 4.2 4.4 Knowledge of the effect that a loss or malfunction of Rating the RELIEFSAFETY VALVES will have on following:
Reactor over pressurization Level RO Tier # 2 Group# 1 The plant is at power an abnormal switch configuration. For an EMRV to open, its solenoid must energize. With the front control panel switch in OFF, the affected EMRV will not function in the pressure relief mode, but will function in the ADS mode. With the interior panel switch in DISABLE, the solenoid will not energize at all: the affected and NRl086, will function in the pressure relief mode. Answer C is correct. This question is a bank question that was used on the last Page 119 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Source Bank Memory or I x Bank Modified 1 INew I Comprehension X Question Cognitive Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 Content: I 55m41 I I 55.43 I
Time to Complete: 1-2 minutes Page 120 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when an event occurred. Current conditions are as follows:
RPV pressure has risen to 1078 psig 0 RPV water level has lowered to 70 Drywell pressure has risen to 17 psig Which of the following is correct regarding the actuation of the EMRVs and the effect of RPV pressure?
A. No EMRVS are open since their setpoint has not yet been reached, and the ADS signal is bypassed. Reactor pressure continues to rise.
B. 5 EMRVs indicated in the VALVE OPEN REGION and remained open until manually bypassed, resulting in a lowering RPV pressure until closed.
C. 2 EMRVs indicated in the VALVE OPEN REGION resulting in a lowering RPV pressure. The EMRVs close automatically when RPV pressure lowers.
D. 3 EMRVs indicated in the VALVE OPEN REGION resulting in a lowering RPV pressure. The EMRVs close automatically when RPV pressure lowers.
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OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP 12/3/07 RO SRO 239002 A3.06 Importance 4.1 4.1 Ability to monitor automatic operations of the Rating RELIEF/SAFETY VALVES including: Reactor pressure Level RO Tier # 2 Group# 1 References
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I420 1 RAP-B3g 1 RAP-Blg The question stem describes the plant at rated when some event occurred that resulted in a peal RPV pressure of 1075 psig and RPV water level lowering to 60. Either of these conditions should have scrammed. AN RPV pressure of 1065 psig (from one reference) or 1074 psig (from another reference) will result in the opening of 2 EMRVs in the pressure relief mode. When pressure drops to their Explanation: close setpoint, the valves will auto close. Answer C is correct.
An ADS signal of 64.6 AND high drywell pressure of 2.9 psig will result in all 5 EMRVs opening, which can be closed by manual bypass. Since the RPV level setpoint has not been reached, then ADS will not automatically actuate.
References ta provided duri, I
Learning 2621.828.0.0005 00368 Objective Describe the EMRV initiation logic for both over-pressure operation and operation in the ADS mode. Include the following: initiation signals and setpoints, timer and setpoints, control switches, and panel indications.
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OC ILT 07-1 RO NRC Exam KEY I Modified Question Source I Bank I Bank New X I
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Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 1 X 3:PEO 10 CFR Part 55 Content:
I 55.41 17 55.43 Page 123 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is at rated power when the main steam flow transmitter FT-ID0033A, which inputs into the Digital Feedwater Control System (DFCS), fails to a 0 output. Which of the following states the impact on the DFCS?
This will cause....
A. a bump-less transfer to the alternate DCC.
B. the in-service DCC to control using the last good value of steam flow.
C. the in-service DCC to double the steam flow input from FT-ID0033B.
D. both DCCs to halt and transfer feedwater level control to the Moore Stations.
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OC ILT 07-1 RO NRC Exam KEY Question #
Question Developer InitiaMDate: NTP 12/5/07 Answer I Knowledge and Ability Reference Information RO SRO 259002 K4.10 Importance 3.4 3.4 Knowledge of REACTOR WATER LEVEL CONTROL Rating SYSTEM design feature(s) and/or interlocks which provide for the following: Three element control (main steam flow, reactor feedwater flow and reactor water level provide input)
Level RO ITier# ( 2 1 References I
MDD-OC-625, page 20 of 70 I 1 The plant is at power when 1 of 2 steam flow transmitters fails (1 per steam line). When this occurs, the DFCS will ignore the bad signal and double the good signal from the second steam flow transmitter, Explanation: and will continue to operate with the in-service digital control computer. Answer C is correct. Other answers are plausible but incorrect.
Learning 2621.828.0.001 8 259-10444 Objective Describe the interlock signals and setpoints for the affected system components and expected system response including power loss and failed components.
Modified Question Source Bank New X Bank Question Cognitive, Memory or X Comprehension Level: Fundamental or Analysis 1:1 Knowledge 10 CFR Part 55 Content:
1 55.41 17 1 55.43 Time to Complete: 1-2 minutes Page 125 of 211
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when the following annunciators alarmed:
MAIN STEAM - FLOW HI/MN STM LINE AREA TEMP HI-HI 1 MAIN STEAM - FLOW HVMN STM LINE AREA TEMP HI-HI II Current plant conditions are as follows:
RPV water level is 1 8 8 and rising slowly RPV pressure is 950 psig and rising slowly Primary Containment parameters are normal ABN-1, Reactor Scram, has been entered. Which of the following states an action to control either RPV pressure or RPV water level, given the conditions above?
A. Trip all operating Feedwater Pumps to prevent injecting into the RPV.
B. Augment RPV pressure control with use of the Isolation Condensers Vents.
C. Stabilize RPV pressure below 1045 psig using the Main Turbine Bypass Valves.
D. Trip all operating Feedwater Pumps AND Condensate Pumps to prevent injecting into the RPV.
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OC ILT 07-1 RO NRC Exam KEY Knowledge and Ability Reference Information RO SRO 259002 2.4.6 (Reactor water level control) Importance 3.1 4.0 Knowledge symptom based EOP mitigation Rating strategies.
Level RO Tier # 2 Group # 1 References I
ABN-1 I
RAP-J3a 1
The plant was at rated power when the hi-hi temperature alarms result in the closure of the MSIVs, which generated a scram signal.
ABN-1, reactor Scram, is then entered. When RPV water level cannot be restored and maintained below 170, all operating feedwater pumps are tripped. Answer B is correct.
The use of isolation condensers vents is not allowed since the IC should be manually isolated due to RPV high water level. Since the Explanation:
IC cannot be used, the vents also become unavailable. Answer B is incorrect. The use of the turbine bypass valves is not allowed since the MSlVs have closed. Answer C is incorrect. Tripping of feedwater pumps and condensate pumps is only done on high RPV water level and RPV pressure is low (<350psig; the condensate pump shutoff head). Answer D is incorrect.
Learning 2621.845.0.00 200-10445 Objective Given a set of system indications or data, evaluate and interpret them tp determine limits, trends, and system status.
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1 OC ILT 07-1 RO NRC Exam KEY Question Source Bank 1 I
I I
Modified Bank New Question Cognitive Memory or Comprehension Level: Fundamental or Ana Iysis 3:SPK Knowledge 10 CFR Part 55 55.41 10 55.43 Content:
I Time to Complete: 1-2 minutes Page 128 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power with the STANDBY GAS SELECT switch in SYS 2, when the following radiation monitoring annunciator alarmed:
0 AREANENT DNSCL Investigation revealed that REACTOR BUILDING VENT MANIFOLD NO. 1 radiation monitor indicates downscale. Which of the following states the impact on the Standby Gas treatment System (SGTS)?
A. Both SGTS Fans are in standby and BOTH can auto start.
B. Both SGTS Fans have auto started and will remain running.
C. ONLY SGTS Fan 2 has auto started and will remain running.
D. BOTH SGTS Fans have auto started and SYS 1 fan will shutdown after a time delay.
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OC ILT 07-1 RO NRC Exam KEY I Question #
Answer 1 45 Question Developer InitialdDate: NTP 12/5/07 RO SRO 261000 K6.04 Importance Knowledge of the effect that a loss or malfunction of Rating the following will have on the STANDBY GAS TREATMENT SYSTEM: Process radiation monitoring Level RO Tier # Group #
References I RAP-1OFlf I RAP-1OF4g I651.4.001 The question stem describes a downscale indication of the #1 RB vent manifold radiation monitor (of which there are 2). The logic for SGTS auto initiation is for either vent manifold radiation monitor to exceed the upscale trip point. When this occurs, both SGTS fans start. When it has been assured that the selected fan is functioning properly, the secondary fan will auto secure after a time delay. The impact of a single vent manifold radiation monitor downscale failure is Explanation: there is none. The SGTS remains in standby and will auto initiate as designed when the operable radiation monitor detects an upscale trip. Answer A is correct.
Answer B, C, D and are incorrect since no SGTS fans have started.
The logic for SGTS auto start is independent of which radiation monitor senses an upscale trip to start both SGTS fans - radiation monitor #I (2) is not dedicated to the auto start of SGTS fan #I (#2).
Learning 2621.828.0.0042, 02456 0bjective Describe the RPS indication logic trip signals and functions, including the following: purpose/design basis; setpoints; conditions that allow bypassing isolation signals; how bypassing isolation signals is accomplished .
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OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank 1 1 Comprehension 1
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Question Cognitive Memory or X Level: Fundamental or Ana Iys is Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
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OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when the following event occurred:
An electrical fault on Bus 1D caused the Main Breaker 1D to automatically open Annunciator MN BRKR 1D 86 LKOUT TRIP alarmed Which of the following states the response of EDG 2?
EDG 2 will .....
A. idle start.
B. NOT start.
C. fast start and load onto Bus 1D.
D. fast start but NOT load onto Bus 1D.
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I OC ILT 07-1 RO NRC Exam KEY I Question I 46 B I Question Developer InitialdDate: NTP 12/5/07 I Knowledge and Ability Reference Information I Ro SRO 4.3 262001 K1.O1 Knowledge of the physical connections and/or cause/effect relationships between A.C.
ELECTRICAL DISTRIBUTION and the following:
Emergency generators
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Level
~~
1 RO 1 Tier# 12 I Group# 1 I
~~~
References [RAP-&
The plant is at power when a fault occurs on 4160 VAC Bus 1D. EDG 2 can idle start from a LOCA signal (the EDG accelerates to 400 RPM) and does not load onto the bus. A loss of voltage to the bus will Explanation: cause EDG 2 to fast start (accelerate to 900 RPM) and load onto bus 1D (normally). But since there is a fault (lockout) on bus 1D, the EDG is prevented from fast starting and loading onto the bus. Answer B is correct. Other answers are plausible but incorrect.
References to provided durii I
Learning 2621.828.0.0013 264-10445 Objective Given a set of system indications or data, evaluate and interpret them to determine limits, trends, and system status.
Modified Question Source Bank New X Bank Question Cognitive Memory or X Comprehension Level: Fundamental 1:1 or Analysis Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
Time to Complete: 1-2 minutes Page 133 of 211
OC ILT 07-1 RQ NRC Exam KEY The plant is at 4% power during a startup. Current plant conditions are as follows:
0 Feedwater Pump IC is in-service 0 Condensate Pumps 19 and 1C are in-service 0 Reactor Recirculation Pump B is in IDLE 0 Power ascension is in-progress The following annunciators then alarm:
STARTUP XFMRS - LKOUT RELAY 86/S1B TRIP STARTUP XFMRS - S1B VOLTS LO Which of the following IMMEDIATE OPERATOR ACTIONS is required?
A. Manually insert CRAM rods due to reduced core flow.
B. Manually scram the reactor due to reduced core flow.
C. Manually scram the reactor due to reduced feedwater flow.
D. Manually reduce recirculation flow due to reduced feedwater flow.
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OC ILT 07-1 RO NRC Exam KEY Question #
47 C Question Developer InitiaWDate: NTP 12/6/07 Answer Knowledge and Ability Reference Information RO SRO 262001 2.4.49 (AC Electrical Distribution) Importance 4.0 4.0 Ability to perform without reference to procedures Rating those actions that require immediate operation of system components and controls.
Level RO Tier # Group #
References
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202.1 Power Ops. Curve
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ABN-2
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I ABN-17
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The question stem describes the plant starting up with a single feedwater pump and 2 condensate pumps running (all powered from 4160 VAC bus 1B). Bus 1B also powers 2 recirculation pumps: pump B (which is currently off) and D. Under the given conditions, power to the station is provided by the startup transformers SA and SB. SB powers bus 1B. Thus the alarms provided show a loss of startup transformer SB and the loss of bus 1B. The running feedwater pump, both running condensate pumps and recirculation pump D will trip.
IAW ABN-17, the trip of multiple condensate pumps requires an Explanation: immediate operator action to scram the plant. Answer C is correct.
Inserting cram rods is not the correct immediate action for a single recirculation pump trip (Cram rods could be required if the exclusion zone was entered. Burt power is below the exclusion zone and would not be entered upon the recirculation pump trip). Answer A is incorrect. IAW ABN-2, the trip of multiple recirculation pumps would require an immediate scram. Answer B is incorrect. If a single feedwater pump or single condensate pump trips, then lowering power with recirculation flow is required IAW ABN-17. This is an incorrect action for multiple pump trips. Answer D is incorrect.
Learning 2621.828.0.0017 256-10450 Objective describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation in accordance with Page 135 of 21 1
OC ILT 07-1 RO NRC Exam KEY I ABN, EOP & EOP Support Procedures and EPIPs.
Modified Question Source Bank X New Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 2:RI Knowledge 10 CFR Part 55 55.41 10 55.43 Content:
I I I I I Time to Complete: 1-2 minutes Page 136 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power. An over-voltage condition resulted in the loss of a power supply. The following annunciators alarmed:
RPS - SCRAM CONTACTOR OPEN RPS - RPS MG SET 1 TRIP PWR LOST - PROT SYS PNL 1 PWR LOST STATION BAT/CHG - BAT CHG C1 TROUBLE The Operator notes that RBCCW 1-1 is still running.
Which of the following states the plant impact and the required action?
Plant Impact Required Action A. RPS MG Set 1 has tripped Place RPS MG Set 1 on the alternate power supply B. VMCC 1A2 has tripped Verify that the Rotary Inverter has transferred to the alternate power supply C. USS 1A2 has tripped Secure the Reactor Building Ventilation system and start the Standby Gas Treatment System D. VMCC 1A2 has tripped Verify Vital Lighting Distribution Panel VLDP-1 has transferred to the alternate power supply Page 137 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP 12/6/07 Knowledge and Ability Reference Information RO SRO 262002 A2.02 I Importance 2.5 2.7 Rating I I Level RO Tier # 2 Group# 1 References I3013 1 ABN-50 I The plant is at power when an electrical power loss occurs. USS 1A2 supplies power to VMCC 1A2, and VMCC 1A2 supplies power to RPS MG Set 1 and to battery chargers C1 and C2 (only 1 charger is in-service at a time). The first three alarms provided point to the loss of: USS 1A2, or VMCC 1A2, or RPS MG Set 1. The last alarm, charger C1 trouble, alarms when AC power is lost to the in-service charger, or DC output is low. Since the charger is powered from VMCC 1A2 (which is powered from USS 1A2), the possible plant impact is either the loss of USS 1A2 or VMCC 1A2. The question stem also says that the RBCCW pump 1-1 is still running (which is powered from USS 1A2), and therefore, the plant impact must be the loss of VMCC 1A2. When VMCC 1A2 is lost, VLDP-1 (which is Explanation: normally powered from VMCC 1A2), then the input power automatically transfers to VMCC 182. ABN-50 requires verification of power transfer for VDLP-1. Answer D is correct.
It is correct in Answer A, that RPS MG Set 1 has tripped. But, the loads on the RPS MG are manually transferred to another power supply - not the power to the MG itself. Answer A is incorrect.
Answer B is incorrect since power to the rotary inverter (which is normally powered from VMCC 1B2) is not affected and no transfer takes place. Answer B is incorrect. Since USS 1A2 has not been lost, then answer C is incorrect. The action is correct for the loss of USS 1A2.
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OC ILT 07-1 RO NRC Exam KEY Learning 2621.828.0.0056 212-10445 Objective Given a set of system indications or data, evaluate and interpret them to determine limits, trends, and system status.
Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
Time to Complete: 1-2 minutes Page 139 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when the following annunciator alarmed:
DC PWR LOST - BUS C UV Bus DC-C was confirmed to have 0 volts.
Which of the following states the impact on the plant?
A. 125 VDC Power Panel E (DC-E) will de-energize.
B. 4160 VAC Bus 1A will NOT transfer to transformer SA on a scram.
C. 41 60 VAC Bus 1B will NOT transfer to transformer SB on a scram.
D. 125 VDC Power Panel F (DC-F) will transfer to an alternate power supply.
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OC ILT 07-1 RO NRC Exam KEY Knowledge and Ability Reference Information I RO I SRO 263000 K3.03 Knowledge of the effect that a loss or malfunction of the D.C. ELECTRICAL DISTRIBUTION will have on following: Systems with D.C. components (Le. valves, motors, solenoids, etc.)
I Level RO Tier # 2 Group # 1 References ABN-55 3033 13028 sh. 1 The plant is at rated power when indications of the loss of 125 VDC Bus C occurs. This bus normally provides DC power to the breakers on 4160 VAC Bus 1A. When the reactor scrams, breaker 1A (aux.
transformer supply to bus 1A) will automatically open and breaker S l A (startup transformer SA supply to bus 1A) will automatically close. With the loss of breaker power, this transfer does not occur.
Answer B is correct.
DC-E is normally supplied from DC-A and is unaffected by the DC Explanation:
loss. Answer A is incorrect.
Bus 1B breaker power supply is normally from bus DC-B and is unaffected by the DC loss. Answer C is incorrect.
DC-F is powered from DC-C and is affected by the DC loss, but unlike many DC panels, there is no alternate DC power supply and DC-F is de-energized. Answer D is incorrect.
Learning 2621.828.0.0012 01 121 Objective Given a set of system indications or data, evaluate and interpret them to determine limits, trends, and system status.
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I Question Source I I
Bank I I
X I
I Modified Bank I 1 I I I New I
Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis 1:1 Knowledge I 10 CFR Part 55 Content:
1 1
55.41 17 I
1 1 I
55.43 I I Time to Complete: 1-2 minutes Page 142 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when a total Loss of Offsite Power (LOOP) occurred.
Three minutes later, EDG 1 experienced an overspeed condition.
Which of the following states the Core Spray System Pumps and Containment Spray System Pumps that are able to perform their function in this condition?
Core Sprav Svstem Pumps Containment Sprav Svstem Pumps A. A and B A and B B. D and A B and C C. B and C C and D D. C and D D and A Page 143 of 21 1
1 OC ILT 07-1 RO NRC Exam KEY Question #
C Question Developer InitiaMDate: NTP 12/7/07 Answer Knowledge and Ability Reference Information RO SRO 264000 K3.03 Importance 4.1 4.2 Knowledge of the effect that a loss or malfunction of Rating the EMERGENCY GENERATORS (DIESEUJET) will have on following: Major loads powered from electrical buses fed by the emergency generator(s)
Level RO Tier # 2 Group # 1 3001C References RAP-T4b 13000 1 3002 sh. 2 The plant was at rated power when all offsite power was lost. Both EDG 1 and EDG 2 will start and load onto their respective busses (1C and 1D), Loads on bus 1C (EDG 1) are: core spray loops A & D (main pumps); and USS 1A2, which powers core spray loops A & D (booster pumps) and containment spray loops A & B. Loads on bus 1D (EDG 2) are: core spray loops C & B (main pumps); and USS 1B2, which powers core spray loops C & B (booster pumps) and Explanation:
containment spray loops C & D. Many trips for the EDGs are bypassed if they are started due to loss of bus voltage (fast start).
EDG overspeed is not bypassed and does result in an EDG trip.
Therefore, EDG 1 trips and EDG 2 only, continues to run and power the loads: core spray loops C & B; and containment spray loops C &
D. Answer C is correct. All others are the incorrect arrangement of pumps.
Learning 2621.828.0.0013 262-10445 Objective Given a set of system indications or data, evaluate and interpret them to determine limits, trends, and system status.
Page 144 of 21 1
I OC ILT 07-1 RO NRC Exam KEY 1 1 I I I Question Source Bank Modified Bank 1New 1 X Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis 10 CFR Part 55 Content:
1 Knowledge 55.41 17 Time to Complete: 1-2 minutes r T1:F 55.43
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Page 145 of 211
OC ILT 07-1 RO NRC Exam KEY The plant is at rated power with the following air system lineup:
COMPRESSOR 1 is in LEAD and indicates red light ON (Panel 7F)
COMPRESSOR 2 is in LAG and indicates green light ON (Panel 7F)
COMPRESSOR 3 is in standby The following annunciator then alarms:
SERVICE AIR - COMPR 1 TRIP Which of the following describes the impact on the Instrument Air Compressors?
A. When air pressure drops to 95 psig, Air Compressor #2 will automatically swap to the LEAD Compressor.
B. When air pressure drops to 90 psig, Air Compressor #3 will automatically swap to the LAG Compressor.
C. Air Compressor #2 automatically swapped to the LEAD Compressor when Air Compressor #1 tripped.
D. Air Compressor 2 will remain as the LAG Compressor until manually transferred to the LEAD Compressor.
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OC ILT 07-1 RO NRC Exam KEY Question #
51 D Question Developer InitialdDate: NTP 12/8/07 Answer Knowledge and Ability Reference Information RO SRO 300000 K4.01 Importance 2.8 2.9 Knowledge of (INSTRUMENT AIR SYSTEM) design Rating feature(s) and or interlocks which provide for the following: ManuaVautomatic transfers of control Level RO Tier # 2 Group# 1 I I References 334 The question shows that air compressor #1 is the lead compressor, and controls air pressure 105-120 psig. Air compressor #2 is the lag compressor, and is designed to maintain pressure 95-110 psig.
Establishing Lead (air pressure controlling band) and Lag (a different air pressure controlling band) compressors is performed locally.
Under normal air usage conditions, one air compressor is lead, one is lag, and one is in standby. When air pressure gets low enough, the lag compressor will start. When the lead compressor trips, air pressure will fall until the lead compressor gets signaled to start. The lag compressor will remain in lag until locally swapped to lead. When air compressor 1 trips, the lag compressor will start and control air Explanation: pressure at the lag pressure band. Air compressor 3 will remain in standby and will start and load at a lower pressure. Answer D is correct.
It is plausible that following the air compressor trip, the lag compressor will become the lead at a higher pressure and that compressor #3 become the new lag compressor at a lower pressure.
But since placing a compressor in lead/lag is a manual manipulation, all answers which specify transfer to a different mode automatically is incorrect. Air compressor #3, stated in the stem to be in standby, is a recognized and understood mode of operation: it is not currently operating, but is powered and ready to function if required.
Learning I 621.828.0.0043, Service, Instrument and Breathing Air Objective 10441, Given the system logic/electrical drawings, describe the system trip signals & setpoints and expected system response including Page 147 of 21 1
OC ILT 07-1 RO NRC Exam KEY I I power loss or failed components.
Question Source I Bank I I Modified Bank 1 1 I New Question Cognitive Memory or X Comprehension Level: Fundamental 1:1 or Analysis Knowledge 10 CFR Part 55 Content:
1 55.41 17 1 55.43 I Time to Complete: 1-2 minutes Page 148 of 211
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when Motor Control Center 1621A was de-energized.
Which of the following systems has valve operators that are directly affected by this power loss?
A. ESW B. TBCCW C. RBCCW D. Circulating Water Page 149 of 21 1
RO SRO 400000 K2.02 Importance Knowledge of electrical power supplies to the Rating following: CCW valves I Level I RO
~~
) T i e r # 12 )Group# ) I I References 1 3004 sh. 3 ~
The plant was at power when MCC 1B21A was de-energized. Of the Explanation: systems listed, only RBCCW has motor operators powered from this bus. Answer C is correct.
Learning 2621.828.0.0035 0005 Objective State how service water, shutdown cooling, RWCU, primary containment, AC electrical distribution and chemical treatment systems interrelate with the RBCCW system.
I Question Source 1 Bank 1 Modified
/Bank 1 /New 1 X Question Cognitive Memory or X Comprehension Level: Fundamental 1:F or Analysis Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
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OC ILT 07-1 RQ NRC Exam KEY The plant was at rated power when an event occurred, resulting in the following annunciators alarming:
REACTOR LEVEL - RX LVL LO-LO I AND RX LVL LO-LO II DW PRESS - DW PRESS HI-HI I AND DW PRESS HI-HI I I STARTUP XFMRS - LOCKOUT RELAY 86/SlA TRIP STARTUP XFMRS - LOCKOUT RELAY 86/S1 B TRIP Which of the following states the response of the RBCCW Pumps?
Both RBCCW Pumps are OFF and.....
A. will automatically start after a 60 second time delay.
B. will automatically start after a 166 second time delay.
C. receive an automatic trip signal but can be bypassed and manually started.
D. receive an automatic trip signal but can be manually started WITHOUT being bypassed.
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OC ILT 07-1 RO NRC Exam KEY Question #
Question Developer InitialdDate: NTP 12/8/07 Answer Knowledge and Ability Reference Information 400000 A3.01 Ability to monitor automatic operations of the CCWS including: Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the CCWS Level References I RO ITier# 12 IGroup# 11 Importance Rating I 223R0173 sh. l a I 11688328 sh. 13 I 341 tRO The plant is at power when a LOCA occurs (Drywell pressure > 3 SRO psig and RPV water level e 86) and a loss of offsite power (startup transformers SA and SB are locked-out). This will start and load both EDGs. Under LOOP conditions only, both RBCCW pumps will automatically start after a 166 second time delay following EDG start/load. But with the LOCA as well, the pumps do not receive a start signal, but do receive a trip signal. The pumps can however be Explanation: manually started from the control room, but only after a switch located at the respective pump breakers is placed in the bypass position (which bypasses the trip signal). This will allow manual pump start from the control room. Answer C is correct. Other answers are plausible but incorrect. The CRD Pumps receive a start signal with a 60 second time delay during LOOP LOCA conditions.
Learning 2621.828.0.0035 0006 Objective Explain the logic for the RBCCW pump breakers for a loss of power with and without a LOCA.
Page 152 of 211
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
Page 153 of 21 1
OC ILT 07-1 RQ NRC Exam KEY With NO control rod manipulations in-progress, the in-service CRD FCV (NC30A) indicated red light ON and green light OFF.
Which of the following states the system impact and the required action IAW Procedure 235, Determination and Correction of Control Rod Drive System Problems?
Svstem Impact Required Action A. Indicated DRV WTWREACTOR AP Place CRD DRIVE WATER PRESS rises CONTROL switch to CLOSE B. Indicated CLG WTWREACTQR AP Place CRD COOLING WATER lowers PRESS CONTROL switch to OPEN C. Stabilizer Valve flow will lower Place the alternate CRD FCV (NC30B) in service D. Indicated CHARGING WATER Place the alternate CRD FCV PRESS lowers (NC30B) in service Page 154 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitialdDate: NTP 12/8/07 1
RO SRO 201001 A2.11 Importance Ability to (a) predict the impacts of the following on Rating the CONTROL ROD DRIVE HYDRAULIC SYSTEM ;
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve openings Level RO Tier # 2 Group# 2 References 235 237E487 The plant is at rated power when the in-service CRD FCV shows indications of full open. Normally the valve displays intermediate position (red and green). With this valve full open, charging pressure will be reduced (since charging comes off upstream of the valve), and cooling waterheactor AP and drive water/reactor AP have risen (since they come off downstream of the valve). IAW procedure, the correct Explanation: action to correct a reduction in charging pressure due to a failed FCV is to place the alternate FCV in service. Answer D is correct.
Since drive waterheactor AP rises, placing the CRD drive switch to close will result in a greater AP. Answer A is incorrect. Since cooling waterheactor AP rises, answer B is incorrect. The Stabilizer Valves come off downstream of the CRD FCV, so that they too would see a higher pressure and could result in greater flow - not less.
References to provided durii I
Learning 2621.828.0.001 1 10450 Objective Describe and interpret procedure sections and steps for plant emergency of off-normal conditions that involve this system including personnel allocation and equipment operation IAW applicable ABN, EOP & EOP Support Procedures and EPIPs.
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OC ILT 07-1 RO NRC Exam KEY Question Source I I
Bank 1 I
1I Modified Bank New X Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 Content:
I I
55.41 15 I
r T I
55.43 I I Time to Complete: 1-2 minutes Page 156 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when the following annunciator alarmed:
ROD CNTRL - ROD DRIFT The Operator attempts to insert the control rod back to its original position, but it withdraws to the full out position. The Operator then scrams the control rod IAW procedure by placing the appropriate toggle switch to the open (scram) position (up position).
Which of the following states the control rod position indication (1) after the control rod has completed drifting, and, (2) after the control rod is scrammed BUT before the toggle switch is taken to the closed position (down position).
0 (2)
Indication after Drift Indication after Scram A. blank-blank 00 with green backlight B. Red backlight ONLY 00 with green backlight C. 48 with red backlight Green backlight ONLY D. Red backlight ONLY Green backlight ONLY Page 157 of 211
OC ILT 07-1 RO NRC Exam KEY I Question #
Answer 1 55 Question Developer InitialdDate: NTP 12/8/07 Knowledge and Ability Reference Information RO SRO 201003 A3.01 Importance 3.7 3.6 Ability to monitor automatic operations of the Rating CONTROL ROD AND DRIVE MECHANISM including: Control rod position I I Level RO Tier # 2 Group# 2 I I I References I ABN-6 1302.2 I The plant was at rated when a control rod drifted to the full out position. Indication at this position is 48 with a red backlight. When the scram signal is still applied and the control rod scrams, the control rod will past full in and will show only a green backlight. Answer C is correct.
Explanation: A rod that is at any non-fully inserted position (except 48) will show the number with no red/green backlight. A control rod that is uncoupled and past position 48 will display only the red backlight. A control rod at 00 (but not in a scram state) will show 00 with green backlight. Answer A is incorrect since the 00 would not be displayed.
Answer B is incorrect since the control rod is not uncoupled. Answer D is incorrect since 48 will have a red backlight.
Learning 2621.828.0.001 1 00079 Objective Describe the backlighting scheme on the full core display for the following: overtravel in; full in; full out; and, overtravel out.
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OC ILT 07-1 RO NRC Exam KEY I I I I Question Source I 1 Modified Bank I INew I Question Cognitive Level:
Bank Memory or Fundamental I
Comprehension or Analysis 1 3:SPK Knowledge 10 CFR Part 55 Content:
1 I
55.41 17 I
I 1 I
55.43 I I Time to Complete: 1-2 minutes Page 159 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is starting up after the third forced outage this calendar year. Current plant conditions are as follows:
RPV water temperature is 120 O F and steady 0 11 control rods have been withdrawn to their withdraw limit The RWM fails and is declared inoperable. The Reactor Engineer states that he recalls a similar RWM failure during the last startup at the exact same step in the control rod sequence and that the RWM was repaired after exceeding 10% power.
Which of the following is correct regarding the reactor startup?
A. The startup CAN continue as long as a second Licensed Operator verifies the Operator at the controls is following the control rod sequence.
B. The startup CAN NOT continue because a startup with the RWM failure with less than 12 control rods withdrawn has been performed this year.
C. The startup CAN NOT continue until a temporary change to procedure 409, Operation of the Rod Worth Minimizer, is processed by the SM or US.
D. The startup CAN continue as long as a second Licensed Operator AND A Reactor Engineer verifies the Operator at the controls is following the control rod sequence.
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OC ILT 07-1 RO NRC Exam KEY Question #
56 B Question Developer InitialdDate: NTP 12/8/07 Answer Knowledge and Ability Reference Information 201006 2.1.32 (RWM) Importance Ability to explain and apply system limits and Rating precautions.
I I Level RO Tier # 2 Group ## 2 References 409 TS 3.2 The plant is starting up with 11 control rods withdrawn when the RWM is declared inoperable. A similar RWM failure occurred this year in which the startup continued with the RWM inoperable with <
12 control rods withdrawn. Because that startup took place this year, a second startup with an inoperable RWM and c 12 control rods withdrawn is not allowed IAW procedure 409 (and IAW TS 3.2).
Answer B is correct.
Explanation:
If more than 12 control rods had been withdrawn when the RWM failed, then answer A states the correct compensatory action. Answer A is incorrect. If the RWM failed with < 12 control rods withdrawn for the first time this year, then answer D states the correct compensatory action. Answer B Is incorrect. Because the procedural precaution reflects a TS requirement, it cannot be changed with a procedural temporary change. Answer C is incorrect.
References to provided durii I
Learning 2621.828.0.0041 217-10447 Objective Given the normal operating procedures and documents for the system, describe or interpret the procedural steps.
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OC ILT 07-1 RO NRC Exam KEY Question Source I I
Bank 1 I
I Modified Bank X New Question Cognitive Memory or X Level: Fundamental 1:P 10 CFR Part 55 Content:
~~
Knowledge 55.41 10
~
Time to Complete: 1-2 minutes r T1 55.43 Page 162 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is at 80% power with Reactor Recirculation Pump B in local manual control.
Which of the following states the impact if the Operator moves the scoop tube operating lever in the RAISE direction that scoops LESS oil.
A. The MG Set generator output current, voltage, and power will rise.
B. The MG Set generator output current, voltage, and power will lower.
C. The MG Set AC drive motor input current, voltage, and power will rise.
D. The MG Set AC drive motor input current, voltage, and power will lower.
Page 163 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
57 A Question Developer InitialdDate: 12/9/07 Answer Knowledge and Ability Reference Information RO SRO 202002 Al.03 Importance 2.5 2.4 Ability to predict and/or monitor changes in Rating parameters associated with operating the RECIRCULATION FLOW CONTROL SYSTEM controls including: MG set generator current, power, voltage Level 1 RO 1 Tier# 12 1 Group# 2 References I Simulator I The plant is at power with a recirculation pump in local manual control, when the operator moves the scoop tube operating lever in the direction that scoops less oil. This will enhance the coupling between the input AC drive motor and the generator and will result in generator speed, current, voltage and power rising, and the recirculation pump will pump more water. (Recall that power =
voltage x current) Answer A is correct.
Explanation: If the candidate thinks that scooping less oil will result in the recirculation pump pumping less water, then answer B would be correct. If recirculation pump flow is increased, the MG input AC motor will have an increase in current and power, but voltage will remain constant. If recirculation pump flow is decreased, the MG input AC motor will have a decrease in current and power, but voltage will remain constant. Answers C and D are incorrect.
Learning 2621.828.0.0040 00158 Objective Describe the following components associated with the recirculation flow control system, including location, purpose, construction, operation and power supply: fluid coupler.
Page 164 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank X New Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 2: DR Knowledge
~
10 CFR Part 55 55.41 5 55.43 Content:
Page 165 of 21 1
OC ILT 07-1 RO NRC Exam KEY Which of the following aids in providing cooling to the Containment Spray Pump motors when operating in the Torus Cooling Mode?
A. A recirculation fan is designed to always be running in the Containment Spray Pump corner rooms.
B. A recirculation fan will auto start in the Containment Spray Pump corner rooms when a Containment Spray Pump starts.
C. Containment Spray Pump corner room inlet and outlet dampers auto open on pump start to increase air flow into/out of the corner rooms.
D. A recirculation fan is automatically started in the Containment Spray Pump corner rooms whenever room temperature reaches the auto start temperature setpoint.
Page 166 of 211
OC ILT 07-1 RO NRC Exam KEY Question #
Question Developer InitiaWDate: NTP 12/9/07 Answer Knowledge and Ability Reference Information RO SRO 219000 K4.06 Importance 2.7 2.7 Knowledge of RHWLPC I: TO RUS/SUPPRESSI ON Rating POOL COOLING MODE design feature(s) and/or interlocks which provide for the following: Pump motor cooling Level I RO 1 Tier # 12 1 Group# 2 References I310 I
I I I When a containment spray pump is started, as would be for initiating torus cooling, a recirculation fan (RF-1-10 or RF-1-11; one fan for Explanation: each pump set) is set up to auto start. Answer B is correct. Other answers are plausible but incorrect.
Learning 2621-828-0-0009 226-10444 Objective Describe the interlock signals and setpoints for the affected system components and expected system response including power loss or failed components.
Question Source I Bank Bank New X I
Question Cognitive Memory or X Comprehension Level: Fundamental 1:F or Analysis Knowledge
~
10 CFR Part 55 55.41 7 55.43 Content:
I I Time to Complete: 1-2 minutes Page 167 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is at rated power with the following conditions:
0 Drywell average temperature is 120 OF 0 Drywell Recirculation fans are in the NORMAL configuration IAW Procedure 312.9, Primary Containment Control A fault on Bus 1B23 then occurred. Which of the following states the status of the Drywell Recirculation Fans?
A. Drywell Recirculation Fans 1-1, 1-2, and 1-3 have tripped.
Drywell Recirculation Fans 1-4 and 1-5 remain running.
B. Drywell Recirculation Fans 1-1 and 1-2 have tripped, and Drywell Recirculation Fan 1-3 is unavailable.
Drywell Recirculation Fans 1-4 and 1-5 remain running.
C. Drywell Recirculation Fans 1-4, and 1-5 have tripped.
Drywell Recirculation Fans 1-1,l-2, and 1-3 remain running.
D. Drywell Recirculation Fans 1-4 and 1-5 have tripped.
Drywell Recirculation Fans 1-1 and 1-2 remain running and Drywell Recirculation Fan 1-3 is available.
Page 168 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitialdDate: NTP 12/9/07 Knowledge and Ability Reference Information RO SRO 223001 K2.09 Importance 2.7 2.9 Knowledge of electrical power supplies to the Rating following: Drywell cooling fans I I Level RO Tier# 2 Group # 2 References 312.9 I I At rated power, average drywell temperature is around 120 O F with the normal recirculation fan lineup of fans 1, 2, 4, and 5.
The plant is at power when bus 1B23 is lost. Normally, fans 1, 2, 4, Explanation: and 5 are operating, with fan 3 in standby. Fans 1, 2, and 3 are powered from Bus 1A23; fans 4 and 5 are powered from Bus 1923.
Therefore, when Bus 1B23 is lost, fans 4 and 5 trip. Fans 1 and 2 remain running and fan 3 is available (but not energized). Answer D is correct.
References to provided durii Objective Given a set of plant conditions, interpret control room and/or local primary containment system indications and evaluate then in terms of limits and trends using available data.
Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
Page 169 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when a LOCA occurred. Drywell Sprays have been initiated. IAW the EOP Users Guide, which of the following states the basis for the Conditional Statement below?
IF Drywell sprays have been initiated, AND Torus or Drywell pressure drops below 1 psig, THEN confirm termination of Drywell Sprays.
It provides operating margin to.. ...
A. the Torus suction header vortex limit.
B. operation of the Drywell-to-Torus vacuum breakers.
C. the Containment Spray Pump trip on low Drywell pressure.
D. operation of the Reactor Building-to-Torus vacuum breakers.
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OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP 12/9/07 Knowledge and Ability Reference Information 1 RO 1 SRO 226001 K5.06 Importance 12.6 I 2.8 Knowledge of the operational implications of the Rating following concepts as they apply to RHWLPCI:
CONTAINMENT SPRAY SYSTEM MODE: Vacuum breaker operation I I 1 I I Level RO Tier# 12 Group# 2 I
~~
EOP Users References TS 5.2 Guide The plant was at rated power when a LOCA occurred and drywell sprays have been manually initiated. There are 2 reasons IAW the users guide: 1) it provides margin to operation of the RB-Torus vacuum breakers which could add oxygen; 2) it maintains a positive margin to the design negative pressures of the drywell and Torus (from TS 5.2.A, these valves are: Drywell: -2 psid; and Torus: -1 Explanation: psig). Answer D is correct.
Answer A is plausible but is more related to low Torus water level instead of Torus pressure. Answer B is incorrect since it refers to the wrong vacuum breaker. It is true that a running containment spray pump will trip at low drywell pressures (0.6 psig), and this is to also prevent going negative in the drywell. But it is not the basis for the conditional statement. Answer C is incorrect.
Learning 2621.845.0.0042 3000 Objective Using procedure EMG-3200.02, evaluate the technical basis for each step in the procedure and apply this evaluation to determine correct courses of action under emergency conditions.
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I 1
I Question Source I Bank I 1 Modified Bank lNew X Question Cognitive Memory or X Comprehension Level: Fundamental l:B or Analysis Knowledge 10 CFR Part 55 Content:
~~
55.41 5 1 55.43 Page 172 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant is shutdown and fuel shuffling is taking place. The following annunciator is then received in the Control Room:
ROD CNTRL - ROD BLOCK Which of the following states the cause of this alarm?
A. The Main Fuel Hoist was just loaded with a fuel bundle over the core.
- 6. The Monorail Auxiliary Hoist was just loaded with a control rod blade over the core.
C. The Main Fuel Hoist positioned on a fuel bundle when the grapple ENGAGED light went ON.
D. The Main Fuel Hoist was loaded with fuel in the Spent Fuel Pool when a control rod was withdrawn to position 02.
Page 173 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
61 A Question Developer InitiaWDate: NTP 12/9/07 Answer I
Knowledge and Ability Reference Information 234000 A3.02 Ability to monitor automatic operations of the FUEL HANDLING EQUIPMENT including: tlnterlock operation Level References RO 1
I Tier #
USAR Table 7.7-1 1
I Group #
656.4.001 Importance Rating I
I360 tRO SRO The plant is shutdown and fuel shuffling is underway. When the hoist loaded light comes on, this means that the hoist is loaded with fuel (as sensed by the load cell). When the hoist is loaded with fuel over the core, a control rod block is installed. Answer A is correct.
Even with the bridge over the core, a loaded Monorail Auxiliary Hoist Explanation: does not install a control rod block. Answer B is incorrect. In answer C, there is not yet any load on the fuel hoist, even though it is positioned over the core. The grapple engaged light verifies that the grapple is closed. It does not input into the rodblock circuit. Answer C is incorrect. A loaded hoist in the SFP does not create a control rod block nor does a single control rod withdrawn to position 2. Answer D is incorrect.
2621.812.0.0003 02391 Demonstrate understanding of the interlocks and rod blocks associated with the following refueling platform components, including their purpose and applicable technical specifications: bridge and trolley, main hoist, aux. hoist.
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OC ILT 07-1 RO NRC Exam KEY I Question Source Bank I 1 Modified Bank 1 1 1 New Question Cognitive Memory or X Comprehension Level: Fundamental 1:1 or Analysis Knowledge I 10 CFR Part 55 Content:
55.41 7 55.43 Page 175 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when a leak in the steam pressure sensing line to the EPR has developed, causing sensed pressure to slowly decay.
Which of the following states the impact on the Turbine Generator System?
A. Generator load (MWe) will rise.
- 6. Generator load (MWe) will lower.
C. Reactor Pressure Vessel pressure will lower.
D. Generator terminal voltage will require manual adjustment.
Page 176 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP la10/07 Knowledge and Ability Reference Information RO SRO 241000 K1.24 Importance 2.7 2.8 Knowledge of the physical connections and/or Rating cause/effect relationships between REACTOFUTURBINE PRESSURE REGULATING SYSTEM and the following: Main generator I 1 Level RO Tier # 2 Group# 2 I I I Lesson Plan References 262 1.828.0.005 1
~ ~ ~~
The plant is at power when the pressure sensing line to the electronic pressure regulator (EPR) develops a leak and sensed pressure starts to slowly decay. The EPR is designed to maintain the pressure setpoint. Normally, when reactor power is reduced, RPV pressure lowers and the EPR will close down on the turbine control valves (TCVs) to maintain pressure. This will result in a lower generator output (load). The same response on the generator is expected when the sensing line results in a lower sensed pressure to the EPR - it will act to close down on the TCVs to try to maintain the pressure. Thus, Explanation:
generator load will lower. Answer B is correct. Answer A is incorrect since generator load goes down - not up. Because a change in reactor power is not the cause for the sensed pressure reduction, and with the turbine control valves closing down, RPV pressure will start to rise. Answer C is incorrect. Because generator terminal voltage is automatically controlled, there will be no requirement to manually manipulate the generator terminal voltage (even if it does change).
Answer D is incorrect.
Learning 2621.828.0.0051 249-10445 Objective Given a set of system indications or date, evaluate and interpret them to determine limits, trends and status.
Page 177 of 21 1
1 OC ILT 07-1 RO NRC Exam KEY Question Source Bank 1 Modified Bank I
X New Question Cognitive Comprehension X Level: or Analysis 3: PEO L
10 CFR Part 55 55.41 5 55.43 Content:
Page 178 of 21 1
OC ILT 07-1 RO NRC Exam KEY Which of the following conditions would result in a HIGHER GROUND-LEVEL radioactivite release rate during the DBA refuel accident?
A. Reactor Building Differential Pressure LOWEST INDICATED changes from
-0.10 to +0.15 inches of water.
B. The Reactor Building outer Railroad Airlock door is found to be stuck in the open position and it cannot be closed and sealed.
C. The electric heating coil in the running Standby Gas Treatment System loop is de-energized and the air stream humidity rises to 100%.
D. The Standby Gas Treatment System 1 Fan trips immediately upon startup, with the STANDBY GAS SELECT switch in the SYS 1 position.
Page 179 of 21 1
OC ILT 07-1 RO NRC Exam KEY I Question# I I Question Developer InitiaMDate: NTP 12/10/07 I
~ ~~ ~
Knowledge and Ability Reference Information RO SRO 290001 K3.01 Importance 4.0 4.4 Knowledge of the effect that a loss or malfunction of Rating the SECONDARY CONTAINMENT will have on Following: Offsite radioactive release rate Level RO Tier # 2 Group# 2 Lesson Plan References TS 3.5 basis USAR 6.5.1 262 1.828.0.0042
~~~
With a positive pressure in the reactor building during the DBA refuel accident, the leakage from the RB to the outside is greater than when a negative pressure is maintained, as is the norm. Answer A is correct.
Even though the RB outer Railroad Airlock door is open, the RB is still closed as long as the inner door is closed. Nothing in the question implies that these other doors are inoperable. Answer B is Explanation: incorrect. If the heating coil in the running SGT loop failed, the efficiency of the charcoal filters is reduced and this will result in a greater elevated release - not ground level. Answer C is incorrect.
When SGT auto starts, both loops initiate. If the selected fan does not provide adequate flow, the non-preferred fan will remain running. This fan trip will not affect either the elevated or ground release rates.
Answer D is incorrect.
Learning 261-10435 2621.828.0.0042 Objective Given plant operating conditions, describe or explain the purpose/function of the system and its components.
Page 180 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Source jsank-1 Modified X
Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 2: DR Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
Page 181 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when an event occurred. It has been determined that the .Standby Liquid Control System injection line has completely sheared just inside the Drywell penetration. Which of the following states the Control Room indicator that is unreliable because of this pipe break?
A. Core AP
- 6. REACTOR LEVEL FUEL ZONE A ONLY C. STANDBY LIQUID CONTROL PUMP DISCH PRESS D. REACTOR LEVEL NARROW RANGE GEMAC A ONLY Page 182 of 211
OC ILT 07-1 RO NRC Exam KEY Knowledge and Ability Reference Information 290002 K6.05 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR VESSEL INTERNALS: SBLC Level References I RO I Tier # 12 USAR 7.6.1.1.6 I Group#
USAR 9.3.5.2 Importance Rating 2
148F712 IRO SRO Explanation:
Learning 2621.828.0.0046 21 1-10453 Objective Explain or describe how this system is interrelated with other systems.
Page 183 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or X Comprehension Level: Fundamental 1:1 or Analysis Knowledge 10 CFR Part 55 55.41 7 55.43 Content:
Page 184 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when the following annunciator alarmed:
CIRC & SERVICE WATER - NRW CHLORINE LEAK The Intake Operator reports an audible alarm and sees a greenish-yellow vapor cloud.
Which of the following states the required status or mode of operation of the Control Room Heating and Ventilation System for these conditions IAW ABN-33, Toxic or Flammable Gas Release?
Status or Mode A. Purge B. FulI Recirculation C. Partial Recirculation D. Manually tripped & isolated Page 185 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP la11/07 Knowledge and Ability Reference Information RO SRO 290003 A4.04 Importance 2.8 3.0 Ability to manually operate and/or monitor in the Rating control room: Environmental conditions Level RO Tier# 2 Group# 2 The plant is at rated power when the control room receives indications of a chlorine gas leak to the environment outside of the control room. IAW the references, the control room HVAC shall be operated in the full recirculation mode. This mode of operation is Explanation: provided to minimize the intrusion of toxic gases into the control room during a release using no outside air. Answer B is correct. All other answers are plausible vent system modes.
2621.828.0.0054 02324 modes of control room ventilation damper alignment and the effects of the damper alignment modes on control room habitability.
Modified Question Source Bank X New Bank Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis l:P Knowledge I 10 CFR Part 55 Content:
I 55.41 17 1 55.43 1 I Time to Complete: 1-2 minutes Page 186 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when an event occurred which allowed the use of Transient Alarm Response. Which of the following states the expectation for alarm announcement by this response AND when Transient Alarm Response is exited?
Transient Alarm Response Alarm Transient Alarm Response Exited Announcement A. ONLY those alarms associated with When announced by the Unit EOP entry conditions should be Supervisor announced
- 6. ONLY those alarms associated with When all EOPs have been exited EOP entry conditions should be announced C. ONLY critical alarms should be When announced by the Unit announced Supervisor D. ONLY critical alarms should be When all EOPs have been exited announced Page 187 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
66 C Question Developer InitiaWDate: NTP 12/12/07 Answer Knowledge and Ability Reference Information RO SRO 2.1.1 Importance 3.7 3.8 Knowledge of conduct of operations requirements Rating Level RO Tier# 3 Group #
I 1 I I I OP-OC-101-References
~111-10Q1 IAW procedure OP-OC-101-111-1001, Strategies for Successful Transient Mitigation, when transient alarm response is allowed, only critical alarms and results should be announced to the US (Unit Supervisor). The US shall appraise the transient and as conditions Explanation: permit, exit transient alarm response by announcing to the crew that transient alarm response is being exited. Answer C is correct.
Other distractors are plausible in that some are related to normal alarm response or are mis-interpretations of the transient alarm response guideline in the procedure.
References to provided durii I
Learning Objective Question Source Bank I
1 Modified Bank I
I I
I New 1 I
Question Cognitive Level: Fundamental 1:P or Analysis Knowledge
- ~ -~ ~~
7 10 CFR Part 55 55.41 10 55.43 Content:
Page 188 of 211
OC ILT 07-1 RO NRC Exam KEY The plant is shut down and is cooling down. The Shutdown Cooling System was placed into service with SDC Pumps A, B and C. The cooldown rate plot over the first hour from procedure 203, Plant Shutdown, is provided (see next page).
Which of the following correctly describes the cooldown rate and what action can be taken to change the cooldown rate? (The starting temperature is as shown on the plot.)
Cooldown Rate Action A. The cooldown rate should be reduced Throttle closed RBCCW INTO the SDC Heat Exchangers B. The cooldown rate should be reduced Throttle closed RBCCW OUT OF the SDC Heat Exchangers C. The cooldown rate should be raised Throttle open RBCCW INTO the SDC Heat Exchangers D. The cooldown rate should be raised Throttle open RBCCW OUT OF the SDC Heat Exchangers Page 189 of 21 1
OC ILT 07-1 RO NRC Exam KEY 630 G:
e CI 3 54Q Eai 450 F
w c:
ItJ II O
s 1 2 3 4 5 6 7 8 Hours Page 190 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
67 B Question Developer InitiaWDate: NTP 12/12/07 Answer Knowledge and Ability Reference Information RO SRO 2.1.7 Importance 3.7 4.4 Ability to evaluate plant performance and make Rating operational judgments based on operating characteristics / reactor behavior / and instrument interpretation.
Level RO Tier # 3 Group #
References BR 2006 sh. 1 305 203 The question stem describes a reactor cooldown with SDC,with the cooldown rate plot provided. It can be seen from the plot that the rate is approximately 95 "F/hr (305 - 210 in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />). The procedurally allowed cooldown rate limit is 15 "F/lO minutes = 90 "F/hr. Therefore, the cooldown rate is greater than procedurally allowed and must be reduced. Cooldown rate is reduced by throttling closed V-5-106, SD CLG WTR OUTLET, which is the RBCCW out of the SDC heat exchangers. Answer B is correct and answer A is incorrect. Answers C and D state the incorrect cooldown rate relationship.
Learning 2621.828.0.0045 205-10445 Objective Given a set of system indications or data, evaluate and interpret then to determine limits, trends and system status.
Question Source 1 Bank 1 I New X Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
I I I I I Time to Complete: 1-2 minutes Page 191 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at 60% power. A malfunction occurred in the master recirculation controller which caused recirculation flow and reactor power to lower. The Reactor Operator has taken all recirculation speed controllers to MANUAL and the flow/power reduction has ceased. The following conditions exist:
Reactor power is 45% and steady 0 Reactor recirculation flow is 6.5 x lo4 GPM Which of the following actions are required?
A. Manually scram the reactor.
- 6. Raise reactor power to 60% with control rods.
C. Lower reactor power to 30% with control rods.
D. Raise reactor recirculation flow to 7.0 x 1O4 GPM.
Page 192 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question #
D Question Developer InitiaWDate: NTP la12/07 Answer Knowledge and Ability Reference Information 1 RO 1 SRO 2.1.25 Importance 2.8 3.1 Ability to obtain and interpret station reference Rating materials such as graphs / monographs/ and tables which contain performance data.
Level I RO 1 Tier# 13 1 Group#
~~
I I References I 202.1 I
1301.2 The question describes an event in which power and recirculation flow place the plant in the Exclusion Zone on the Power Operations Curve. IAW procedure 202.1, Power Operation, the operator is to exit the zone using rods or flow. The recirculation flow in answer D places the plant outside of the zone. Answer D is correct.
Explanation: Scramming the reactor would place the plant outside of the zone but this is not the intent of the procedural step. Answer A is incorrect.
Raising reactor power to 60% would move the plant out of the zone, but it would also pass the scram setpoint. Answer B is incorrect.
Lowering reactor power to 30% would not place the plant outside of the zone. Answer C is incorrect.
Learning 2621.828.0.0040 00224 Objective Identify and interpret normal operating procedures for the recirculation flow control system.
Modified Question Source Bank X New Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 Content:
55.41 10 55.43 Time to Complete: 1-2minutes
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Page 193 of 21 1
OC ILT 07-1 RO NRC Exam KEY Which of the following would require the use of a grounding device for a clearance IAW Procedure OP-MA-109-101, Clearance and Tagging? The work will require replacing the motor in each case.
A. SDCPump B. ESWPump C. Core Spray Booster Pump D. Containment Spray Pump Page 194 of 21 1
OC ILT 07-1 RO NRC Exam KEY Knowledge and Ability Reference Information RO SRO 2.2.13 Importance 3.6 3.8 Knowledge of tagging and clearance procedures. Rating Level RO I Tier # .F1 Group #
OP-MA-I09-References I101 IAW the reference, proper safety grounding shall be applied prior to working on high voltage equipment when contact with exposed conductors is planned or possible. The reference also defines high voltage as an energy source 600 volts or above. In the work activities listed in the question stem, all will require removal of the motor and the potential for exposed conductors exists. Of the equipment listed, Explanation only the ESW Pump is powered from a bus greater than 600 VAC (Bus 1C or 1D). Answer B is correct.
SDC pumps are powered from 480 VAC USS 1A2/1B2. Answer A is incorrect. Core Spray Booster pumps are powered from 480 VAC USS 1A2/1B2. Answer C is incorrect. Containment Spray pumps are powered from 480 VAC USS 1A2/1B2. Answer D is incorrect.
Learning 0bjective Page 195 of 21 1
OC ILT 07-1 RO NRC Exam KEY I Question Source Question Cognitive Bank Memory or Modified Bank New Comprehension X
X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 10 55.43 Content:
I I I Time to Complete: 1-2 minutes Page 196 of 211
OC ILT 07-1 RO NRC Exam KEY The plant is shutdown for a refuel outage. Fuel shuffling is in-progress. The fuel cell shown below must be emptied for inspection. Which of the following steps, IN ORDER, are performed to allow the control rod to be withdrawn?
1st Step 2"'Step Step A. Remove bundles 1 & 3 Insert blade guide Remove bundles 2 & 4 B. Remove bundles 2 & 3 Remove bundles 1 & 4 Insert blade guide C. Remove bundles 1 & 4 Insert blade guide Remove bundles 2 & 3 D. Remove bundles 2 & 4 Remove bundles 1 & 3 Insert blade guide Page 197 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP 12/14/07 Knowledge and Ability Reference Information RO 1 SRO 2.2.27 Importance Knowledge of the refueling process. Rating Level RO Tier # 3 Group #
References 205 I
The fuel cell must have all fuel removed prior to withdrawing the control rod. Diagonal bundles are removed first, then a blade guide Explanation: installed to support the control rod, then the other diagonal bundles are removed. Answer C is correct.
Learning 2621.812.0.0003 07442 Objective Describe in general, refueling and fuel handling procedures to include precautions and limitations per procedure 205.
Modified Question Source Bank New X Bank Question Cognitive Memory or X Comprehension Level: Fundamental 1:P or Analysis Knowledge 10 CFR Part 55 Content:
I 55.41 15 1 55.43 1 Time to Complete: 1-2 minutes Page 198 of 21 1
OC ILT 07-1 RO NRC Exam KEY A reactor startup is in-progress with the REACTOR MODE SELECTOR switch in STARTUP. The Drywell is being inerted IAW 312.1 1, Nitrogen System and Containment Atmosphere Control.
Which of the following states the valves used to allow nitrogen to flow into the Drywell AND the discharge path for air leaving the Drywell IAW procedure 312.1 l ?
Nitroaen Flow In Air Discharge Out A. Drywell Purge Valves Air is exhausted to the Standby Gas V-23-13 and V-23-14 Treatment System B. Drywell Purge Valves Air is exhausted to the RB Ventilation V-23-13 and V-23-14 System C. Torus Purge Valves Air is exhausted to the Standby Gas V-23-15 and V-23-16 Treatment System D. Torus Purge Valves Air is exhausted to the RB Ventilation V-23-15 and V-23-16 System Page 199 of 21 1
OC ILT 07-1 RO NRC Exam KEY IQuestion #
I Question Developer InitialdDate: NTP 12/14/07 I Knowledge and Ability Reference Information RO SRO 2.3.9 Importance 2.5 3.4 Knowledge of the process for performing a Rating containment purge.
Level RO Tier # 3 Group #
References 312.11 13432.19-1 BR 201 1 sh. 2 Explanation:
References to provided durir 7
Learning 2621.828.0.0032 00446 Objective Identify and interpret normal, abnormal and Emergency Operating Procedures for the Primary Containment System.
Page 200 of 21 1
OC ILT 07-1 RO NRC Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or X Comprehension Level: Fundamental 1:P or Ana Iysis Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
Page 201 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when the following annunciator alarmed:
RADIATION MONITORS PROCESS OFF GAS - OFF GAS HI The Operator reports that the Offgas Radiation Monitors are at the alarm setpoint.
Under the given conditions, which of the following states the required action IAW ABN-26, High Main Steam/Offgas/Stack Effluent Activity?
A. Confirm isolation of the Off Gas System.
- 6. Reduce reactor power to clear the alarm.
C. Scram the reactor IAW ABN-1, Reactor Scram.
D. Commence a plant shutdown IAW 203, Plant Shutdown.
Page 202 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question#
72 B Question Developer Initials/Date: NTP la15/07 Answer Knowledge and Ability Reference Information RO SRO 2.3.1 1 Importance 2.7 3.2 Ability to control radiation releases. Rating Level RO Tier # 3 Group #
References ABN-26 RAP-IOF~C
~~
The plant is at power when the offgas system radiation monitor alarms. The alarm response refers to ABN-26. IAW ABN-26, the correct action is to reduce power in an attempt to clear the alarm.
Answer B is correct.
Explanation: A second alarm occurs at a higher offgas radiation level (offgas radiation hi-hi). When this alarm occurs, the offgas system will isolate after a 15-minute time delay. Answer A is incorrect. Answer C is an appropriate action if the hi-hi alarm is in and cannot be cleared in 15 minutes. Answer C is incorrect. For hi-hi offgas radiation, it is appropriate to initiate a plant shutdown. Answer D is incorrect.
Learning 2621.828.0.0004 00200 Objective Interpret given Augmented Off Gas System alarms, and describe the required operator actions IAW the applicable alarm response Question Source procedure.
Question Cognitive Bank Memory or Modified Bank X
New Comprehension r x Level: Fundamental 1:P or Analysis Knowledge
~
10 CFR Part 55 Content:
55.41 5 55.43 Page 203 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when a TOTAL loss of feedwater occurred. The Unit Supervisor has entered the ONLY required EOP: RPV Control - No ATWS. Current plant conditions are as follows:
0 RPV coolant temperature is 436 O F Immediate Operator Actions IAW ABN-1, Reactor Scram, have been performed 0 An RPV isolation has occurred Which of the following states the additional plant response?
A. The ADS timers are timing-down.
B. Core Spray has automatically started and is injecting.
C. Core Spray has automatically started and the EDGs have idle started.
D. Core Spray has automatically started and the EDGs have fast started.
Page 204 of 21 1
OC ILT 07-1 RO NRC Exam KEY I Question #
Answer I 73 Question Developer InitialsIDate: NTP la15/07 Knowledge and Ability Reference Information RO SRO 2.4.2 Importance 4.3 4.6 Knowledge of system set points 1 interlocks and Rating automatic actions associated with EOP entry conditions.
Level RO Tier # 3 Group #
I I Refere nces EMG-SP1 341 I
I ABN-1 203 Explanation:
Learning 2621.828.0.001 0 209-10444 Objective Describe the interlock signals and setpoints for the affected system components and expected system response including power loss or Page 205 of 21 1
I failed components.
2621.828.0.001 3 264-10444 Describe the interlock signals and setpoints for the affected system components and expected system response including power loss or failed components.
Question Source Question Cognitive Level:
Bank Memory or Fundamental Modified Bank Comprehension or Analysis New 1
X X
3:SPK Knowledge 10 CFR Part 55 55.41 10 55.43 Content:
Page 206 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power when a break in the Service Water 30discharge header occurred during excavation at the intake area. The Intake Operator reports a large visible Service Water geyser. The Control Room Operator reports that SERVICE WATER HEADER PRESS indicates 58 psig and steady.
Which of the following actions is required IAW ABN-18, Service Water Failure Response?
A. Reduce Reactor Recirculation flow to 8.5 x 1O4 gpm.
B. Scram the reactor and trip all reactor recirculation pumps.
C. Swap RBCCW Heat Exchanger cooling to ESW System I.
D. Stop all Service Water Pumps and initiate a reactor shutdown.
Page 207 of 21 1
OC ILT 07-1 RO NRC Exam KEY 1 Question # 1 1 Question Developer InitiaWDate: NTP la15/07 I Knowledge and Ability Reference Information RO SRO 2.4.24 Importance 3.3 3.7 Knowledge of loss of cooling water procedures. Rating Level 1 RO I Tier # 13 1 Group#
References ABN-18 I
The plant is at power when a major leak occurred in the common SW discharge line and cannot be isolated. This event occurred at Oyster Creek during the current operating cycle and required entry into ABN-
- 18. Answer D, stopping service water pumps, was required by the then-current ABN-18 revision.
With a non-isolable leak that effects both SW pumps, the required actions are to enter ABN-19, RBCCW Failure Response and swap Explanation: RBCCW HX cooling to ESW I.
Answer A is incorrect since a reactor scram is only required for a total loss or imminent total loss of SW. With SW discharge pressure steady at 58 psig, the loss is not imminent (normal pressure is approximately 70 psig). Answer A is incorrect. Answer B and D are also actions for an actual or imminent loss and are incorrect.
operations of the service water system.
Page 208 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Source 1 Bank 1 Question Cognitive Memory or Comprehension Level: Fundamental or Analysis Knowledge 10 CFR Part 55 Content:
1 1 I
55.41 I
io I I I Time to Complete: 1-2 minutes Page 209 of 21 1
OC ILT 07-1 RO NRC Exam KEY The plant was at rated power, when the Shift Manager has declared an Unusual Event.
A Reactor Operator can fill which of the following Emergency Plan positions, which notifies the New Jersey State Police Office of Emergency Management of the INITIAL event from the Control Room?
A. Incident Assessor B. Shift Communicator C. ENS Communicator D. Operations Communicator - Control Room Page 210 of 21 1
OC ILT 07-1 RO NRC Exam KEY Question Developer InitiaWDate: NTP 12/17/07 Knowledge and Ability Reference Information RO SRO 2.4.39 Importance 3.3 3.1 Knowledge of the RO's responsibilities in emergency Rating plan implementation.
1 Tier# 13 1 Group#
References I EP-AA-1010 1 EP-AA-112-1OO-F-o3 The Control Room Operators become qualified as the Shift Explanation: Communicator. Answer B is correct. All other answers are emergency positions but are incorrect.
Learning Objective Modified Question Source Bank New X Bank Question Cognitive Memory or X Comprehension Level: Fundamental or Analysis 1:P Knowledge 10 CFR Part 55 55.41 5 55.43 Content:
~~ ~
Time to Complete: 1-2 minutes Page 21 1 of 21 1
ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with a 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.
Applicant Certification All work on this examination is my own. I have neither given nor received aid.
ResuIts ROISRO-OnlyPTotal Examination Values 1- Points licants Scores ES-401, Page 31 of 33
OC ILT 07-1 NRC Exam Required References - SRO Written Exam Question Reference To Be Provided 1 TS 3.3 (No Basis) 2 10CFR50.72.
EP-AA-1010 Hot Matrix in COLOR 4 EP-AA-1010 Cold Matrix in COLOR 5 EMG-SP38 EMG-3200.02 (PCC EOP) 6 EMG-3200.01B (RPVC - with ATWS EOP)
DELETE SP-21 actions in Power Leg 7 EMG-3200.01A (RPVC - NOATWS EOP) 8 TS 3.2 (No Basis) 10 EMG-3200.11 (SCC EOP)
NOTE: DELETE the major override that refers to SP-49 & -50.
11 TS 3.5 (No Basis) 12 EMG-3200.01A (RPVC - NOATWS EOP) 13 TS Figures 3.2.1 and 3.2.2 (SLC graphs)
TS Figure 4.5.1 TS 3.5 15 EMG-3200.01B (RPVC - with ATWS EOP)
DELETE SP-21 actions in Power Leg 16 TS 5.3 (No Basis) 17 EMG-3200.01A (RPVC - NOATWS EOP) 18 10CFR50.72 20 TS 3.3.E (No Basis)
TS 3.6 24 Calculator Page 1 of 1
- 20. @@
- 21. qj@@
- 24. @e
- 25. @@a .
Name: Date:
OC ILT 07-1 NRC SRO Exam KEY A plant startup was in-progress. The following conditions currently exist:
All IRM Range switches are on Range 10 The REACTOR MODE SELECTOR switch is in STARTUP An event occurred which resulted in reduced recirculation pump flow, and NO operator actions have occurred. Total core flow is 30.2 x lo6 Ib/hr.
Which of the following states the Technical Specification requirements due to this plant condition, AND the basis for this requirement?
Tech Spec Requirement . Basis for TS Requirement A. The plant shall be placed in COLD To prevent transition boiling during a SHUTDOWN within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. transient.
B. The plant shall be placed in COLD To prevent fuel cladding temperature SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. exceeding 1500 O F during a LOCA.
C. The plant shall be placed in the To prevent exceeding 1% plastic SHUTDOWN CONDITION within 30 strain on the cladding during a hours. transient.
D. The plant shall be placed in the To prevent fuel cladding failure during SHUTDOWN CONDITION within 24 a LOCA.
hours.
OC ILT 07-1 NRC SRO Exam KEY Page 2 of 70
OC ILT 07-1NRC SRO Exam KEY Question Developer InitiaWDate: NTP 10/19/07 Knowledge and Ability Reference Information I RO 1 SRO 295001 2.2.22(PartiaKomplete loss OF forced core Importance 3.4 4.1 flow) Knowledge of limiting conditions for operations Rating and safety limits.
Level SRO Tier # 1 Group# 1 References I TS 3.3.H I TS 3.0 I TS 3.3.H says that a minimum flow of 39.65 x lo6 Ib/hr is required while in Range 10 of the IRMs and the Reactor Mode switch in STARTUP. This is done to ensure the transient MCPR limits are not violated. Maintaining within the MCPR limits will preventheduce the amount of transition boiling during a transient. Because this TS does Explanation: not provide any actions if exceeded, then TS 3.0.A applies, which requires the plant be placed in Cold Shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Answer A lists the correct TS requirement and the correct basis. All other answers are related to other thermal limits and/or provide an incorrect thermal limit basis.
Learning 2621.828.0.004000221 Objective Given plant conditions and applicable sections of Technical Specifications, describe which specification applies and what action is required.
OC ILT 07-1 NRC SRO Exam KEY Page 3 of 70
OC ILT 07-1 NRC SRO Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 55.43 2 Content:
OC ILT 07-1 NRC SRO Exam KEY Page 4 of 70
OC ILT 07-1 NRC SRO Exam KEY The plant is at rated power.
Which of the following would require an IMMEDIATE (1 Hour) notification to the NRC IAW 10CFR50.72?
A. RPV water level lowered to 76 during a transient.
B. 34.5 KV Bank 5 and Bank 6 were de-energized for 30 minutes.
C. Both EDGs were declared inoperable due to diesel fuel contamination.
D. A vehicle fire located in the main parking lot that was extinguished in 20 minutes.
OC ILT 07-1 NRC SRO Exam KEY Page 5 of 70
OC ILT 07-1 NRC SRO Exam KEY Question Developer InitiaWDate: NTP 10/23/07 I
Knowledge and Ability Reference Information RO SRO 295003 2.4.30 (PartiaKomplete loss of AC power) Importance 2.2 3.6 Knowledge of which events related to system Rating operations/status should be reported to outside agencies.
I I Level SRO Tier# 1 Group# 1 EP-AA-1010 References LS-AA-I400 10CFR50.72 Hot Matrix IAW EP-AA-1010, the loss of startup transformers SA and SB for >
15 minutes would constitute an unusual event emergency classification (this is represented by Bank 5 and Bank 6). IAW 10CFR50.72, this would require an immediate NRC notification.
Answer B is correct.
A 4-hour report to the NRC due to RPS actuation or ECCS discharge into the reactor coolant system. Answer A is incorrect. A tech. spec.
required shutdown would be required if both EDGs were declared Explanation: inoperable. A 4-hour report would then be required from the initiation of this shutdown. Answer C is incorrect. A fire, not extinguished within 15 minutes could result in an emergency classification, depending on the location. A vehicle fire in the main parking lot does not have the potential to damage safety systems in any Table H2 areas (from EP-AA-1010, UE classification for firelexplosion, HU6).
Thus, no immediate NRC notification is required. Answer D is incorrect.
)e EP-AA-1010 3 exam: 10CFR50.72 Hot Matrix in COLOR Learning Objective OC ILT 07-1 NRC SRO Exam KEY Page 6 of 70
OC ILT 07-1 NRC SRO Exam KEY Question Source 1 Modified Bank 1 1 1 I I New Question Cognitive Level:
Memory or Fundamental Comprehension or Analysis 1 X r T -I 3:SPR Knowledge 10 CFR Part 55 Content:
I 55.41 I I 55m43II I Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 7 of 70
OC ILT 07-1 NRC SRO Exam KEY A plant startup was in-progress. The following conditions currently exist:
RPV pressure is 990 psig 0 5 Turbine Bypass Valves are open 0 The REACTOR MODE SELECTOR switch is in RUN 0 Preparations are being made to place the generator on line Which of the following states an expected reactor scram signal and the Technical Specifications basis for the scram signal, if ALL Turbine Bypass Valves SIMULTANEOUSLY failed opened? (Assume NO operator action)
Scram Signal TS Basis A. Turbine Stop Valve closure Anticipatory scram for pressure and flux transients B. MSlV closure Anticipatory scram for pressure and flux transients C. Turbine Stop Valve closure To maintain margin to the reactor coolant system pressure safety limit MSlV closure To maintain margin to the reactor coolant system pressure safety limit OC ILT 07-1 NRC SRO Exam KEY Page 8 of 70
OC ILT 07-1 NRC SRO Exam KEY Question Developer InitialdDate: NTP 10/23/07 I Knowledge and Ability Reference Information SRO 295006 AA2.06 3.8 Ability to determine and/or interpret the following as Rating they apply to SCRAM: Cause of reactor SCRAM Level RO Tier# 1 Group # 1 I References I TS 2.3 Basis I I The question stems describes a plant startup, in preparation for placing the main generator on-line. In this condition the reactor Mode Selector switch is in RUN. Steam to the turbine at this point is minimal, with the majority going through the 5 open turbine bypass valves. When the other 4 turbine bypass valves fail open, reactor pressure will drop below 825 psig and the MSlVs will auto close (the reactor mode switch is still in RUN). MSlV closure scram is to anticipate flux and pressure transients from MSlV closure events and is the correct answer (B). The reactor scram signal from turbine stop valve closure is indeed to anticipate flux and pressure transients. But at this point in the startup, this scram is bypassed due to low turbine load. Answer A and C are incorrect. Answer D lists the incorrect basis (but related to reactor pressure as is the question) and is incorrect.
References to provided durii Learning 2621.850.0.0090 01658 Objective State the reactor coolant and fuel clad integrity safety limits and briefly describe the bases.
OC ILT 07-1 NRC SRO Exam KEY Page 9 of 70
OC ILT 07-1 NRC SRO Exam KEY Modified Question Source Bank New X Bank
~ ~
Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 55.43 2 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 10 of 70
OC ILT 07-1 NRC, SRO Exam KEY The plant was in a refueling outage with refueling in-progress, when an event occurred.
Which of the following would require an emergency classification and entry into the Site Emergency Plan? (Emergency classifications from Emergency Director Judgment are NOT to be considered.)
A. RPV water level suddenly dropped by 10.
B. RPV water level indication is lost for 20 minutes.
C. RB ARM C-1 verified to be reading at the MAX SAFE level.
D. Stack RAGEMS indicates 7.93 E3 cps LRM for 30 minutes.
OC ILT 07-1 NRC SRO Exam KEY Page 11 of 70
OC ILT 07-1 NRC SRO Exam KEY Question #
C Question Developer InitiaWDate: NTP 10/23/07 Answer Knowledge and Ability Reference Information RO SRO 295023 AA2.05 Importance 3.2 4.6 Ability to determine and/or interpret the following as Rating they apply to REFUELING ACCIDENTS: Entry conditions of emergency plan Level SRO Tier # 1 Group# 1 References EP-AA-1010 I EMG-3200.11 (SCC E-p)
Answer C states that a RB ARM (Area Radiation Monitor) indicates at the MAX SAFE value, which is 1000 mr/hr (the upscale reading). A UE for Abnormal Rad Levels (RU2) is appropriate if an ARM indicates a valid upscale reading. Answer C is correct.
A UE classification would be appropriate if there is an uncontrolled drop in SFP water level or reactor cavity and a valid rise in the refuel floor radiation monitors. Since there are no indications provided that show an increase in these radiation monitors, the UE does not apply (see RU2 for abnormal rad levels). Answer A is incorrect.
Explanation:
A UE classification would be appropriate if all RCS temperature and RPV water level indications are lost for > 15 minutes (MU5 for Decay Heat), but since only water level indication is lost, this classification does not apply. Answer B is not correct.
A UE classification would be appropriate if stack RAGEMS rose above 7.92 E3 cps for 2 60 minutes (RUl for radiological effluent).
But since the given rad levels are above this setpoint for only 30 minutes, the classification does not apply. Answer D is incorrect.
,e EP-AA-1010 Cold Matrix in aexam: COLOR Learning Objective OC ILT 07-1 NRC SRO Exam KEY Page 12 of 70
OC ILT 07-1 NRC SRO Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 55.43 5 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 13 of 70
OC ILT 07-1 NRC SRO Exam KEY The plant was at rated power when a LOCA occurred. The following conditions currently exist:
0 The RPV has been emergency depressurized 0 Primary Containment hydrogen indicates 0.5%
0 Containment High Range Radiation Monitors indicate 22,000 Whr 0 Torus pressure is 32 psig (Torus - Drywell) AP has been calculated as 8 psid 0 Wide range Torus level recorder indicates upscale Which of the following is the correct Primary Containment vent strategy?
Vent From Vent Through A. Torus Torus Vent Valves V-23-21 and V-23-22 B. Drywell Drywell Purge Valves V-27-3 and V-27-4 C. Torus Torus Vent Valves V-28-18 and V-28-47 D. Drywell N2 Purge Valves V-23-13 and V-23-14 OC ILT 07-1 NRC SRO Exam KEY Page 14 of 70
OC ILT 07-1 NRC SRO Exam KEY I Question #
I Question Developer InitiaWDate: NTP 10/23/07
.r Answer Knowledge and Ability Reference Information RO SRO 295024 EA2.01 Importance Ability to determine and/or interpret the following as Rating they apply to HIGH DRYWELL PRESSURE: Drywell pressure Level SRO Tier # 1 Group# 1 References EMG-3200.02, PCC EOP I Support Procedure 38 The given conditions put the Candidate deep in the pressure leg of the Primary Containment Control EOP. The vent path depends on the Primary Containment Water Level (Torus water level). Information is provided, along with the handout to look on a graph and determine this value (which is 550). Given this value and the other information, Explanation: the only correct path is to vent the Drywell through the Drywell Purge Valves, V-27-3 and V-27-4. Answer B is correct. All other options are possible vent paths (in the pressure leg and the combustible gas leg),
but are incorrect with the given conditions.
EMG-3200.02, PCC EOP 2621.845.0.0008 03000 step in the procedure and apply this evaluation to determine correct courses of action under emergency conditions.
OC ILT 07-1 NRC SRO Exam KEY Page 15 of 70
OC ILT 07-1 NRC SRO Exam KEY Modified Question Source Bank New Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 Content:
1 55.41 1 I 55.43 15 I Time to Complete: 2-3 minutes OC ILT 07-1 NRC SRO Exam KEY Page 16 of 70
OC ILT 07-1 NRC SRO Exam KEY The plant was at rated power when a feedwater level control event occurred. The following conditions currently exist 1 minute after the Operator placed the REACTOR MODE SELECTOR switch in SHUTDOWN:
RPV water level is 1 8 4 Isolation Condensers have automatically initiated EMRV NRl08A is cycling Control rod position indication has been lost Some LPRM downscale lights are cycling ON and OFF Annunciators FLOW HVMN STM LINE AREA TEMP HI-HI I AND II have alarmed Which of the following states the INITIAL RPV pressure control strategy?
A. Open EMRVs to maintain RPV pressure 920 psig to below 1045 psig.
B. Depressurize the RPV with EMRVs to maintain the cooldown rate below 100 "F/hr.
C. Manually open the Turbine Bypass Valves to maintain RPV pressure below 1045 psig.
D. Continue to use Isolation Condensers to maintain RPV pressure 920 psig to below 1045 psig.
OC ILT 07-1 NRC SRO Exam KEY Page 17 of 70 1
OC ILT 07-1 NRC SRO Exam KEY Question Developer InitiaWDate: NTP 10/23/07 Knowledge and Ability Reference Information I RO 1 SRO 295025 2.4.6 (High reactor Pressure) Importance 3.1 4.0 Knowledge symptom based EOP mitigation Rating strategies.
Level I SRO I Tier# 1 1 1 Group# 1 References 1 EMG-3200.01B I The question stem describes a failure to scram event (EMRV cycling and LPRM downscale lights cycling on/off), and therefore entry into RPV Control -With ATWS is required. Establishing a normal cooldown is not appropriate, and thus answer B is incorrect. Manually controlling the EMRV to reduce RPV pressure to 920 psig and then maintain < 1040 is correct. Answer A is correct. Using the turbine Explanation: bypass valve to control pressure is not appropriate since there are indications that the MSlVs have automatically gone closed. Answer C is incorrect. The use of isolation condensers is not allowed in the ATWS EOP when RPV water level reaches 180, and the isolation condenser DC valves should be closed. Answer D is incorrect.
Learning 2621.845.0.0004 03080 Objective Given a copy of EMG-3200.01B, explain the actions to be taken to control level, pressure, and power, and the consequences of failing to control these parameters.
OC ILT 07-1 NRC SRO Exam KEY Page 18 of 70
OC ILT 07-1 NRC SRO Exam KEY Modified Question Source Bank New X Bank I I Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 55.43 5 Content:
OC ILT 07-1 NRC SRO Exam KEY Page 19 of 70 I
OC ILT 07-1 NRC SRO Exam KEY The reactor was at rated power when an event occurred. Current plant conditions are as follows:
0 All control rods indicate full-in 0 RPV water level lowered to 130 and has recovered to 1 8 2 RPV pressure is 900 psig 0 Drywell temperature is 225 O F and steady Drywell pressure is 2 psig and steady 0 Torus water level is 120 and steady 0 Torus water temperature is 158 OF and rising slowly Which of the following actions is required?
A. Line-up and spray the Drywell.
B. Emergency Depressurize the RPV.
C. Lower RPV pressure with EMRVs.
D. Lower RPV pressure with Isolation Condensers.
OC ILT 07-1 NRC SRO Exam KEY Page 20 of 70
OC ILT 07-1 NRC SRO Exam KEY Question Developer InitialdDate: NTP 10/23/07 Knowledge and Ability Reference Information RO SRO 295030 EA2.03 Importance 3.7 3.9 Ability to determine and/or interpret the following as Rating they apply to LOW SUPPRESSION POOL WATER LEVEL: Reactor pressure I I Level SRO Tier # 1 Group# 1 References EMG-3200.01A (RPVC-NA) 1 EOP Users Guide A loss of Drywell cooling, with controlled Drywell venting can result in the Drywell conditions listed. A Torus leak combined with EMRV leakage, with no Torus cooling can result in the Torus indications listed.
From the conditions in the question stem, it is given that the reactor has scrammed. The plant has entered RPV Control - No ATWS EOP (RPVC-NA EOP) on low RPV water level, and Primary Containment Control EOP (PCC EOP) due to low Torus water level and high DW temperature.
There are no parameters that require an emergency depressurization. Currently, the Heat Capacity Temperature Limit Curve is not violated but will be violated if RPV pressure is maintained constant and Torus temperature continues to rise. If Explanation:
Torus temperature and RPV water level cannot be maintained below HCTL, ED will be required. IAW the RPV Control - No ATWS EOP, if Torus temperature cannot be maintained below HCTL, then maintain RPV pressure below HCTL. This action will prevent the need to ED.
Because RPV water level is > 180, the Isolation Condensers cannot be used to reduce RPV pressure. The EMRVS can be used to lower RPV pressure. Answer C is correct.
Spraying the Drywell is not appropriate since Drywell is e 12 psig, and since Drywell parameters are on the bad side of the Containment Spray Initiation Limit Curve. Answer A is incorrect.
Because HCTL is not currently violated and lowering RPV pressure can prevent the need to ED, answer B is incorrect.
Answer D is incorrect since RPV water level precludes the use of the OC ILT 07-1 NRC SRO Exam KEY Page 21 of 70
OC ILT 07-1 NRC SRO Exam KEY Isolation Condensers.
Learning 2621.845.0.0042 3000 Objective Using EMG-3200.02, evaluate the technical basis for each step in the procedure and apply this evaluation to determine correct courses of action under emergency conditions.
I Modified Question Source Bank X New Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 55.43 5 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 22 of 70
oc LT 07-1 NRC SRO Exam KEY The plant was starting up after a refuel outage. Current plant conditions are as follows:
Control rod withdrawals have begun.
0 The point of adding heat has NOT yet been reached.
0 The neutron flux is increasing with a stable, positive period without additional control rod movement.
0 The last control rod movement was AFTER the -1% dk sequence step in the ECP.
0 The shutdown margin has NOT yet been demonstrated.
Which of the following states the required Technical Specifications action and the associated TS Basis?
Required TS Action TS Basis A. Immediately fully insert the control A 1% reactivity limit is considered rods in the reverse order until the unsafe since an insertion of this reactor is subcritical. reactivity into the core would lead to transients exceeding design conditions.
B. Fully insert all insertable control rods A control rod drop accident combined within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and verify operability of with a failure to meet SDM can lead to the Standby Liquid Control System. fuel damage and a loss of equipment important to safety.
C. Meet the SDM requirement within 6 If the SDM cannot be restored, hours or be in the SHUTDOWN shutdown is required to minimize the CONDITION within the following 12 potential of a malfunction of hours. equipment important to safety.
D. Meet the SDM requirement within 6 A 1% reactivity limit is considered safe hours or be in the SHUTDOWN since an insertion of this reactivity into CONDITION within the following 12 the core would not lead to transients hours. exceeding design conditions.
OC ILT 07-1 NRC SRO Exam KEY Page 23 of 70 t
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Knowledge and Ability Reference Information RO SRO 295014 AA2.01 Importance 4.1 4.2 Ability to determine and/or interpret the following as Rating they apply to INADVERTENT REACTIVITY ADDITION: Reactor power I I I 1
~
Level SRO Tier # 1 Group# 2 References 1 TS 3.2 I201 The question stem describes a reactor startup, with the indications that the reactor is critical. The 1% reactivity anomaly has been demonstrated but the reactor is critical before being able to ensure the shutdown margin has been met. IAW TS 3.2.A.3, when in the startup or run modes, the SDM must be met within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in Explanation: the shutdown condition within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Answer C is correct.
The other answers list other TS requirements (loss of SDM in the shutdown condition or reactivity anomalies, or actions required by the Startup Procedure, 201) and/or the basis is incorrect.
References to provided durii 7
Learning 2621.828.0.001 1 10451 Objective Given Technical Specifications, identify and explain associated actions for each section of the Tech Specs relating to this system including personnel allocation and equipment operation.
OC ILT 07-1 NRC SRO Exam KEY Page 24 of 70
OC ILT 07-1 NRC SRO Exam KEY Question Source Bank I Modified Bank New X Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 Content:
I 55.41 I
I I
55m43 I I I Time to Complete: 2-3 minutes OC ILT 07-1 NRC SRO Exam KEY Page 25 of 70
OC ILT 07-1 NRC SRO Exam KEY Which of the following states (1) a condition requiring entry into an EOP, AND (2) the associated Technical Specification basis for the limit of the associated EOP entry condition?
A. RPV water level 11 5 above TAF Provides margin to the fuel integrity safety limit B. Drywell pressure 3.6 psig Provides margin to the Primary Containment integrity safety limit C. Torus water level 83,000 ft3 Ensures operability of ECCS equipment in the corner rooms D. Torus water temperature of 96 O F Prevents Torus over-loading during an RPV blowdown at high pressures OC ILT 07-1 NRC SRO Exam KEY Page 26 of 70
OC ILT 07-1 NRC SRO Exam KEY Question #
A Question Developer InitiaWDate: NTP 10/24/07 Answer Knowledge and Ability Reference Information RO SRO 295009 2.4.4 (Low reactor water level) Importance 4.0 4.3 Ability to recognize abnormal indications for system Rating operating parameters which are entry-level conditions for emergency and abnormal operating procedures.
Level SRO Tier # 1 Group# 2 References I TS 2.3 Bases I I I I I
IAW TS 2.3 Bases, the LSSS for low RPV water level is set 2 115 to assure maintaining the fuel integrity safety limit. This water level equates to 137, which is an EOP entry. Answer A is correct. TS Table 3.1.1 requires a reactor scram at a Drywell pressure 53.5 psig, which would also require an EOP entry. There is no Drywell integrity safety limit. Answer B is incorrect. TS 3.5 requires a minimum Torus Explanation:
volume of 82,000 ft3, and IAW the EOP Bases, this equates to a water level of 143. This would not require an EOP entry (e 143), but the basis is correct. Answer C is incorrect. TS 3.5 also requires a maximum Torus water temperature of 95 O F , and is also an EOP entry condition (>95 O F ) , but the bases is incorrect. Answer D is incorrect.
Learning 2621.845.0.0003 03053 Objective Explain the basis for each of the EMG-3200.01 entry conditions.
I Modified Question Source Bank I Bank New X Question Cognitive Memory or X Comprehension Level: Fundamental l:B or Analysis Knowledge 10 CFR Part 55 Content:
55.41 I 55.43 12 I
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 27 of 70
OC ILT 07-1 NRC SRO Exam KEY The plant was at rated power when the following annunciators alarmed:
0 RWCU HELB I AND RWCU HELB II CU ROOM TEMP HI AREAMONHI IAW the RAP, the Operator attempted to manually isolate the Cleanup System, but valve V-16-1, Cleanup System Isolation Valve, and valve V-16-14, Cleanup SYSTEM INLET Valve, both indicate red light ON and green light OFF.
The following conditions are also noted by the Operator:
Area Radiation Monitor C-1 indicates 45 mr/hr and rising slowly Temperature indicator IB06-15 indicates 21 1 O F and rising slowly Which of the following states the correct action?
A. Shutdown the reactor when area radiation monitor C-4 reaches 50 mr/hr.
B. Immediately scram the reactor and place the Mode Switch in SHUTDOWN.
C. Shutdown the reactor when the NW Corner Room water level reaches 16.
D. Emergency depressurize the reactor when temperature indicator IB06-17 reaches 212 O F .
OC ILT 07-1 NRC SRO Exam KEY Page 28 of 70
OC ILT 07-1 NRC SRO Exam KEY Question #
Question Developer InitiaMDate: NTP 10/24/07 Answer Knowledge and Ability Reference Information RO SRO 295032 EA2.03 Importance 3.8 4.0 Ability to determine and/or interpret the following as Rating they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Cause of high area temperature Level 1 SRO ( T i e r # 11 )Group# 12 EMG-3200.12 References (SCC EOP)
The question stem presents indications of a cleanup system leak with the cleanup system failing to isolate. One temperature indicator is above the max safe value. IAW the Secondary Containment Control EOP, with a primary system discharging into the secondary containment, before any area temperature reaches the max safe value, a reactor scram and entry into the RPV Control - No ARWS EOP is required. Since this value is already exceeded, a manual reactor scram is required now. Answer B is correct.
IAW the Secondary Containment Control EOP, with a primary system discharging into secondary containment, a scram is required before any area radiation level reaches the max safe value. A shutdown is required if area radiation levels in 2 or more areas are above the max Explanation: safe. Answer A is incorrect.
IAW the Secondary Containment Control EOP, a reactor shutdown is required when water levels in 2 areas are above the max safe. No water level is above this value. A scram is required, with a primary system discharge before any water level reaches the max safe value.
The term "shutdown the reactor" and "scram the reactor" are 2 different actions. Answer D is incorrect.
When a primary system is discharging, then ED is required when area temperatures exceed the max safe value in 2 2 areas. The temperature indicators, which are above max safe, are both in the same area. Answer D is incorrect.
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OC ILT 07-1 NRC SRO Exam KEY Page 29 of 70
OC ILT 07-1 NRC SRO Exam KEY Learning 2621.845.0.001 1 03082 Objective Using procedure 3200.1 1, evaluate the technical basis for each step and apply this evaluation to determine the correct courses of action under emergency conditions.
Question Source I Bank 1 I 1 New X Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 55.43 5 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 30 of 70
OC ILT 07-1 NRC SRO Exam KEY A plant shutdown was in progress in preparation for a refuel outage. Current plant conditions are as follows:
Shutdown Cooling Pump A is in service Shutdown Cooling Pump B is in service 0 Shutdown Cooling Pump C is in standby 0 RPV coolant temperature is 325 O F and lowering The following annunciator just alarmed:
DC-1 PWRLOST The Operator reports that position indication to V-17-1 and V-17-2 have been lost (SDC Loop A suction valve and SDC Loop B suction valve).
Which of the following states the impact on the Shutdown Cooling System and the action required related to the Shutdown Cooling System ONLY?
Impact on Shutdown Cooling SDC Required Action A. The Shutdown Cooling System shall Isolate the Shutdown Cooling System be declared inoperable WITHIN 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. Shutdown Cooling Loops A and B Remove Shutdown Cooling Loops A ONLY shall be declared inoperable and B from service WITHIN 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. Declare impacted Shutdown Cooling Restore Shutdown Cooling Primary System Primary Containment Isolation Containment Isolation Valves to Valves inoperable operable BY THE TIME the REACTOR MODE SELECTOR switch is placed in RUN on plant startup D. Declare impacted Shutdown Cooling Restore Shutdown Cooling Primary System Primary Containment Isolation Containment Isolation Valves to Valves inoperable operable PRIOR TO declaring the reactor critical on plant startup OC ILT 07-1 NRC SRO Exam KEY Page 31 of 70
OC ILT 07-1 NRC SRO Exam KEY Question Developer InitiaWDate: NTP 10/25/07 Knowledge and Ability Reference Information RO SRO 205000 A2.04 Importance 2.5 2.6 Ability to (a) predict the impacts of the following on Rating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: D.C. failure I I Level SRO Tier # 2 Group# 1 TS 3.5.A.3 Procedure 305 References USAR Table 6.2-12 TS 3.5.A.3.a.3 RAP-9XF4d The question stem shows that Shutdown Cooling (SDC) is in service with Loops A and B. IAW Procedure 305 and the USAR reference, the SDC Loop suction and discharge valves are considered primary containment isolation valves. All of these 6 valves (suction &
discharge for each of 3 loops) are powered from 125 VDC MCC DC1.
Therefore, 4 of the 6 inoperable valves are open with RPV coolant temperature above 212 OF. These valves shall be declared inoperable. TS 3.5.3.A.3.a.3 allows inoperable SDC containment isolation valves with RPV coolant temperature c 350 O F . The same Tech Spec requires that the inoperable valves be made operable prior to placing the reactor in the condition where Primary Containment is required (as when the plant is started-up).
Explanation:
From TS 3.5.A.3, primary containment shall be maintained when the reactor is critical or RPV temperature is above 212 OF. Therefore, there is no requirement to alter the current SDC configuration, although the valves are inoperable. But, the valves must be made operable prior to either declaring the reactor critical, or exceeding cold shutdown temperatures (ie, > 212 OF) [since either of these conditions require primary containment integrity] Answer D is correct.
Answer A and B are incorrect since there is no requirement to remove SDC from service. Because the reactor is past initial criticality and RPV coolant temperature is in excess of 500 O F when the reactor mode switch is placed in RUN (ie, this is past the 2 conditions that require primary containment to be established), verifying containment OC ILT 07-1 NRC SRO Exam KEY Page 32 of 70
OC ILT 07-1 NRC SRO Exam KEY isolation valve operability at this point would be too late. Answer C is incorrect.
2621.828.0.0045 205-10451 for each section of the Technical Specifications relating to this system including personnel allocation and equipment operation.
Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 55.43 2 Content:
Time to Complete: 2-3 minutes OC ILT 07-1 NRC SRO Exam KEY Page 33 of 70
oc LT 07-1 NRC SRO Exam KEY A plant startup was in-progress. Current conditions are as follows:
0 RPV pressure is 700 psig and rising slowly 0 RPV water level is in the normal band 0 Control rods are being withdrawn 0 Feedwater Pump A is in service The following events then occurred:
0 RPS MG SET 1 TRIP annunciator alarmed 0 RPV water level swelled to 181 Which of the following states the strategy to lower RPV pressure as directed by the SRO?
A. Use the EMRVs.
B. Adjust the MPR setpoint.
C. Use the Isolation Condensers Vents.
D. Use the BYPASS VALVE OPENING JACK.
OC ILT 07-1 NRC SRO Exam KEY Page 34 of 70
OC ILT 07-1 NRC SRO Exam KEY Question #
Question Developer Initials/Date: NTP 10/26/07 Answer T
RO SRO 212000 A2.11 Importance Ability to (a) predict the impacts of the following on Rating the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Main steamline isolation valve closure 1 SRO
~ ~
Level ) T i e r # 12 )Group# 11 EMG-3200.01A 237E566, sheets 37 RAp-Qc, RAP-J1d References (RPVC - NO 7 408.12 ATW S) EMG-SP15 I I I Under the conditions in the stem, with RPV pressure less that 825 psig (TS value), a single RPS Bus loss will result in a full reactor scram and closure of the MSIVs. With the closure of the MSIVs, changing the MPR setpoint or the Bypass Valve Opening Jack will have no impact on RPV pressure. Answers B and D are incorrect.
Also, the Fw Pump tripped on high RPV water level (which occurs at 181). At this RPV water level, the EOPs (RPV Control - No ATWS) direct that the isolation condensers DC steam IVs closed. EMG-Explanation: SP15, Alternate Pressure Control Systems IC Tube Side Vents, requires that in order to use the IC vents, that the ICs are not required to be isolated. Since the EOP requires the DC steam valve closed, then use of the tube side vents is not allowed. Therefore answer C is incorrect.
The EMRVs can be still used to control RPV pressure. Even though the use of EMRVs require Torus water level above go, the event started at normal level of approximately 150 and can be assumed to the same. Therefore, answer A is correct.
Learning 2621.828.0.0037 212-10445 Objective Given a set of system indications or date, evaluate and interpret them OC ILT 07-1 NRC SRO Exam KEY Page 35 of 70
OC ILT 07-1 NRC SRO Exam KEY to determine limits, trends, and system status.
2621.845.0.0040 3054 Given a copy of RPV Control, describe in detail each step or conditioial statement, including technical basis, and how to perform each step as required.
Modified Question Source Bank X New Bank I I Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPR Knowledge 10 CFR Part 55 55.41 55.43 5 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 36 of 70
oc LT 07-1 NRC SRO Exam KEY The plant is at rated power. Which of the following events would require the SRO to make a notification IAW OP-AA-106-101,Significant Event Reporting?
A. The Trunion Room door was opened to allow entry of RP Technicians to perform surveys.
B. The SGTS Fan 2, whose motor failed yesterday, can not be repaired or replaced until 10 days from now.
- c. Surveillance showed that at 2600 scfm, the pressure drop across the SGTS HEPA filter measured at 2.2 inches of water.
D. Analysis of the Standby Liquid Control System Poison Tank showed a tank volume of 1400 gallons, 18 weight percent of Sodium Pentaborate Solution, at a temperature of 95 OF.
OC ILT 07-1 NRC SRO Exam KEY Page 37 of 70
OC ILT 07-1 NRC SRO Exam KEY Question #
Answer
, Question Developer InitiaWDate: NTP 10/26/07 Knowledge and Ability Reference Information RO SRO 261000 2.1.14 (SGTS) Importance 2.5 3.3 Knowledge of system status criteria which require the Rating notification of plant personnel.
~ ~ ~ ~
Level SRO Tier # 2 Group# 1 TS 3.5.b.5 References OP-OC-108- OP-AA-106-101 TS Figure 4.5.1 104-1001 IAW OP-OC-108010401001, Guidance for Limiting and Administrative Conditions for Operations, if it is determined that condition cannot be rectified prior to expiration of the LCO clock, then consider the LCO clock expired and commence a controlled shutdown. IAW OP-AA-106-101, Significant Event Reporting, notification is required if an LCO action that will not be met within the allowable time requirement. TS 3.5. b.5, Secondary Containment, allows 1 SGT System to be inoperable for 7 days. A failed motor would render 1 SGTS fan inoperable. Tech Specs allows one SGTS train to be inoperable for 7 days. Knowledge that the SGTS motor will not be fixed until 10 days from now, it will exceed the TS allowed time of 7 days. Therefore, notification IAW OP-AA-106-101 is required.
The SRO will notify the Duty Station Manager (DSM), who is a plant Explanation: management employee. Answer B is correct.
In the current plant condition, Secondary Containment integrity is required. But IAW TS definition 1.14.A, momentary opening of the Trunion Room door is allowed. Therefore, this is a non-event and no DSM notification is required. Answer A is incorrect.
IAW TS 4.5.H.1 .b.l, the pressure drop across the SGTS HEPA filter must be less than provided in Figure 4.5.1. The data provided places it less than the maximum allowed and no violation has occurred and no notifications are required. Answer C is incorrect.
The indications of the Standby Liquid Control poison tank fall within the Tech Specs limits, and therefore no notification is required.
Answer D is incorrect.
The KA is directly matched in that the question requires the candidate OC ILT 07-1 NRC SRO Exam KEY Page 38 of 70
OC ILT 07-1 NRC SRO Exam KEY to choose what events requires the notification of plant personnel.
Learning 2621.828.0.0042 261-10451 Objective Given Technical Specifications, identify and explain associated actions for each section of the Technical Specifications relating to this system including personnel allocation and equipment operation.
I I
Question Source Bank I Modified Bank I lNew 1 Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge I I I I
10 CFR Part 55 55.41 55.43 5 Content:
Time to Complete: 2-3minutes OC ILT 07-1 NRC SRO Exam KEY Page 39 of 70
OC ILT 07-1 NRC SRO Exam KEY Which of the following states the bases for Technical Specifications 3.7.D.1 .a?
Tech Spec 3.7:
D. Station Batteries and Associated Battery Chargers I. With one required station battery B or C charger inoperable:
- a. Restore associated station battery terminal voltage to greater than or equal to the minimum established float voltage within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, This provides good assurance that ....
A. the affected battery will be restored to its fully charged state within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B. the affected battery will be restored to its fully charged state within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C. the design current-carrying capacity of the affected battery will be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D. the battery charger current carrying capacity will not be exceeded when the charger is returned to service after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
OC ILT 07-1 NRC SRO Exam KEY Page 40 of 70
OC ILT 07-1 NRC: SRO Exam KEY Question# ,4 A Question Developer InitiaWDate: NTP 10/26/07 Answer I I Knowledge and Ability Reference Information RO SRO 263000 2.2.25 (DC Electrical Distribution) Importance 2.5 3.7 Knowledge of bases in technical specifications for Rating limiting conditions for operations and safety limits.
Level SRO Tier # 2 Group# 1 I I I TS 3.7 Bases I I
References I The TS Bases for 3.7.D.1 .a states that restoring the battery terminal voltage to greater than or equal to the minimum established float voltage provides good assurance that, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the associated battery will be restored to its fully charged condition (as Explanation: verified by Action 3.7.D.1 .b) from any discharge that might have occurred due to the charger inoperability. Answer A is correct and all other answers are incorrect.
References to provided durii I
Learning 2621.828.0.001 2 263-10451 Objective Given Technical Specifications, identify and explain associated actions for each section of the Technical Specifications relating to this system, including personnel allocation and equipment operation.
Question Source I Bank 1 I Modified Bank I I 1 New X Question Cognitive Memory or X Comprehension Level: Fundamental l:B or Analysis Knowledge 10 CFR Part 55 55.41 55.43 2 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 41 of 70
OC ILT 07-1 NRC SRO Exam KEY The plant was at rated power when the following annunciator alarmed:
ROD CONTROL - CONTROL AIR PRESS LO The TB Operator reports that the in-service drying tower has isolated and the standby drying tower cannot be placed into operation. The SRO ordered a manual reactor scram when INSTR AIR SUPPLY PRESS indicated e 60 psig and lowering. With the REACTOR MODE SELECTOR switch in SHUTDOWN, the current plant conditions are as follows:
0 ALL of the LPRM amber lights on the full core display are LIT 0 RPV water level is 120 and rising 0 The MASTER RECIRC SPEED CONTROLLER indicates 35 hertz 0 8 control rods indicate position 22 Assuming that a drying tower CANNOT be restored and indicated air pressure has decayed to 0 psig, which of the following states the plant impact and the required action directed by the SRO?
Plant Impact Required Action A. Main steam flow to the turbine and/or Stabilize RPV pressure below 1045 condenser is isolated psig with the Isolation Condensers
- 6. The Recirculation MG fluid couplers Place the Recirculation Pumps in local have locked up manual control and reduce to minimum C. The CRD DRIVE WATER Pressure Place the bypass Pressure Control Control valve has failed closed valve in-service and manually insert control rods D. The Feedwater MFRVs have locked Terminate and prevent Feedwater by UP closing the Heater Bank Outlet valves OC ILT 07-1 NRC SRO Exam KEY Page 42 of 70
OC ILT 07-1 NRC SRO Exam KEY Question #
15 A Question Developer InitialdDate: NTP 10/26/07 Answer Knowledge and Ability Reference Information 1 RO 1 SRO 300000 A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Air dryer and filter malfunctions Level I SRO I Tier# 12 I Group# 1 References EMG-3200.01B (RPVC -With ATWS)
ABN-35 I R: 2013, sheets 1, The question describes a loss of air event and a failure of the reactor to scram, with reactor power < 2% (since all LPRM amber lights are lit). With air pressure at 0 psig, the outside MSlVs have closed and thus steam flow to the turbine or condenser is isolated, and IAW the ATW EOP, pressure control should be stabilized < 1045 psig.
Pressure control with the Isolation Condensers is allowed (as long as RPV water level is < 160, which it is). Answer A is correct.
It is true that with a loss of instrument air, the Recirculation MG fluid couplers (scoop tubes) lock up in their current position. The question stem shows that the recirculation pumps are currently at 35 hertz, which is way above the minimum. IAW the ATWS EOP, flowing back recirculation flow to minimum is required when the main generator is Explanation:
on-line. The stem does not provide any indications that the turbine generator did not trip, and thus it is correct to assume that it has.
Since the generator is not online, reducing recirculation flow is not required (although the step is the correct way to control recirculation flow during a loss of air event). Answer B is incorrect.
It is true that the in-service CRD FCV fails closed on loss of air, but the CRD drive water PCV is motor operated, and is unaffected by the loss of air. Since the CRD FCV has failed closed, CRD water supply is not available downstream to manually insert control rods. Answer C is incorrect.
The feedwater MFRV will lock up on loss of air (but may slowly drift open or closed). But since RPV water level is 120 and reactor power OC ILT 07-1 NRC SRO Exam KEY Page 43 of 70
OC ILT 07-1 NRC SRO Exam KEY to terminate and prevent feedwater (although the listed method is one correct method to control feedwater flow during a loss of air event). Answer D is incorrect.
Learning 2621.845.0.0041 3055 Objective Given a copy of RPV Control, describe in detail each step or conditional statement, including technical basis, and how to perform each step as required 2621.828.O. 0043 Modified Question Source Bank New X Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 55.43 5 Content:
OC ILT 07-1 NRC SRO Exam KEY Page 44 of 70
OC ILT 07-1 NRC SRO Exam KEY The plant was starting up after a refuel outage. Radwaste notified the Control Room of receipt of the following NRW annunciator:
NV-37 FUEL POOL FILTER FLOW LOW 30 minutes later, the following annunciator then alarmed in the main Control Room:
FUEL POOL - POOL LEVEUTEMP HI The Reactor Building Operator reports that Fuel Pool temperature is 116 OF and rising at 1OF/minute. If this condition CANNOT be corrected in the next 10 minutes, which of the following states the plant impact and the required Technical Specifications action?
Impact Required Action A. The margin to the amount of allowable The plant shall be placed in COLD positive reactivity in the Fuel Pool is SHUTDOWN in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> reduced
- 6. The margin to the amount of allowable Alternate Fuel Pool cooling shall be Fuel Pool corrosion is reduced established prior to exceeding 150 OF C. The margin to ensuring the structural The plant shall be placed in COLD integrity of the Fuel Pool is reduced SHUTDOWN in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> D. The margin to the assumptions used The plant shall be placed in HOT in the Reactor Building flooding SHUTDOWN in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> analysis is reduced OC ILT 07-1 NRC SRO Exam KEY Page 45 of 70
OC ILT 07-1 NRC SRO Exam KEY Question Developer InitialdDate: NTP 10/27/07 Knowledge and Ability Reference Information RO SRO 233000 A2.07 Importance 3.0 3.2 Ability to (a) predict the impacts of the following on Rating the FUEL POOL COOLING AND CLEAN-UP; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High fuel pool temperature Level I SRO I Tier # 12 I Group# 2 TS 5.3.1. D References RAP-G4a TS 3.0.A
~ ~
The question stem shows the loss of fuel pool cooling (low flow and high fuel pool temperature) during a reactor startup after a refuel outage. After 10 minutes, fuel pool temperature will reach 126 O F ,
while TS 5.3.1 .D limit is 125 OF. Since there is no direct action Explanation: statement, then TS 3.0.A applies which requires the plant to be in cold shutdown in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The basis for the pool temperature limit is based on ensuring the structural integrity of the fuel pool. Answer C is correct.
References to provided durii I
Learning 2621.828.0.0020 00488 Objective With the aid of Tech Specs, explain each Tech Spec design basis applicable to Augmented & Fuel Pool Cooling System.
OC ILT 07-1 NRC SRO Exam KEY Page 46 of 70
OC ILT 07-1 NRC SRO Exam KEY Modified Question Source Bank New X Bank Question Cognitive Memory or X Comprehension X Level: Fundamental l:B or Analysis 3:SPK Knowledge 10 CFR Part 55 55.41 55.43 2 Content:
OC ILT 07-1 NRC SRO Exam KEY Page 47 of 70
OC ILT 07-1 NRC SRO Exam KEY The reactor was at 28% power when the Reactor Operator reports the following indications:
0 TOTAL STEAM FLOW indicates 2.0 Mlb/hr 0 TOTAL FEEDWATER FLOW indicates 2.50 Mlb/hr and rising RPV Water Level indicates 173 and rising The following annunciators alarmed moments later:
0 RX LVL HI I AND RX LVL HI II 0 ROPS ACTUATE A AND ROPS ACTUATE B Which of the following predicts the expected plant impact and the required action directed by the Unit Supervisor?
Plant Impact Required Action A. Automatic reactor scram AND turbine Stabilize RPV pressure below 1045 trip on a high RPV water level signal psig with the turbine bypass valves B. Automatic reactor scram AND turbine Augment RPV pressure control with trip on a high RPV water level signal the Isolation Condensers C. Automatic turbine trip AND Feedwater Restore RPV water level 138-175 Pump trip on a high RPV water level using Support Procedure 2, Feed &
signal Condensate System Operation D. Automatic turbine trip AND Feedwater Restore RPV water level 138-175 Pump trip on a high RPV water level using Support Procedure 8, Lineup for signal Condensate System Operation OC ILT 07-1 NRC SRO Exam KEY Page 48 of 70
OC ILT 07-1 NRC SRO Exam KEY Knowledge and Ability Reference Information RO SRO 259001 A2.01 Importance 3.7 3.7 Ability to (a) predict the impacts of the following on Rating the REACTOR FEEDWATER SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump trip Level SRO Tier # 2 Group # 2 References EMG-3200.01A RAP-H5d I RAP-H6f The question describes the plant at 28% power with feed flow (rising) greater than steam flow and RPV water level above normal and continuing to rise. Moments later, the RPV Level HI alarms come in (175) and results in a turbine trip. Because turbine load is c 30%
power, the reactor does not scram. RPV water level continues to rise until the ROPS alarms comes in (181). With total feedwater flow L 2.4E6 Ib/hr and RPV water level ~_181,all feedwater pumps trip.
With the reactor still at power and no feed flow, RPV water level will lower and the reactor will scram on low water level. Since the feed pumps have tripped, ROPS is now bypassed and feedwater pumps Explanation:
can be manually restarted IAW support procedure 2. Answer C is correct.
Distractors A and B are incorrect since the reactor will not scram on high RPV water level signal. Pressure control is from the pressure leg in the No ATWS EOP. Distractor D is incorrect since support procedure 2 only operates condensate pumps, whose discharge head is much lower than a post-scram reactor (with no primary leaks).
Learning 2621.845.0.0041 3055 Objective Given a copy of RPV Control, describe in detail each step or conditional statement, including technical basis, and how to perform OC ILT 07-1 NRC SRO Exam KEY Page 49 of 70
OC ILT 07-1 NRC SRO Exam KEY I each step as required.
Question Source 1 I
Bank I I
1 I
Modified Bank New X Comprehension X Level: Fundamental or Analysis 3:SPK 10 CFR Part 55 55.41 55.43 5 Content:
Time to Complete: 2-3 minutes OC ILT 07-1 NRC SRO Exam KEY Page 50 of 70
OC ILT 07-1 NRC SRO Exam KEY The plant is at rated power.
Which of the following events requires notification to an off-site agency, and what agency must be notified?
Event Offsite Aqencv A. The plant was manually scrammed 4-hour notification to the NRC due to the loss of both CRD Pumps B. Sighting or capture of a Snapping 24-hour notification to the NRC Turtle C. The declaration of an Unusual Event 15-minute notification to Lacey Township D. The discovery in March that the 8-hour notification to the NRC monthly Core Spray surveillance due in February was not performed as required OC ILT 07-1 NRC SRO Exam KEY Page 51 of 70
OC ILT 07-1 NRC SRO Exam KEY Question Developer InitiaMDate: NTP 10/29/07 Knowledge and Ability Reference Information RO SRO 201001 2.4.30 (CRD Hydraulics) 2.2 3.6 Knowledge of which events related to system Rating operations/status should be reported to outside agencies.
I I Level SRO Tier # 2 Group# 2 LS-AA-1020 References 10CFR50.72 TS 4.2 EP-AA-1010
~~
A 4-hour report to the NRC is required for an RPS actuation. Answer A is correct (IAW 10CFR50.72.) This would also be a 4-hour report for the initiation of a plant shutdown required by Tech Specs.
The sightingkapture of a sea turtle (not snapping) requires notification of NRC and National Marine Fisheries Service w/in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (LS-AA-1020, Exelon Reportability Manual). But, the question refers to a snapping turtle, which is a different species than a sea turtly. A description of a snapping turtle is provided in the procedure.
Answer B is incorrect.
Explanation:
A UE declaration does not require Lacey Township notification w/in 15 minutes (EP-AA-114-100-F-03).A UE requires that the NJ State Police Office of Emergency Management be notified within 15 minutes. Answer C is incorrect.
A missed surveillance by itself, is not a tech spec violation. TS 4.2 allows time to complete the surveillance and NOT enter the applicable LCO (if tested sat) until after the test is completed. Answer D is incorrect.
Learning Objective OC ILT 07-1 NRC SRO Exam KEY Page 52 of 70
OC ILT 07-1 NRC SRO Exam KEY Modified Question Source Bank New Bank
~ ~
Question Cognitive Memory or X Comprehension Level: Fundamental 1:F or Analysis Knowledge 10 CFR Part 55 55.41 55.43 5 Content:
~~~ ~ ~~
Time to Complete: 2-3 minutes OC ILT 07-1 NRC SRO Exam KEY Page 53 of 70
OC ILT 07-1 NRC SRO Exam KEY Procedure 301.2, Reactor Recirculation System, states the following Precaution and Limitation in regards to an ISOLATED recirculation loop with the plant at power:
The recirculation loop shall NOT be returned to service prior to achieving cold shutdown conditions Which of the following states the Technical Specification bases for the precaution?
A. The reactor coolant between the closed valves in the recirculation loop is not available during a LOCA event.
B. The restart of an isolated loop can raise concerns about the thermal stresses on the recirculation loop reactor nozzle.
C. The restart of an isolated loop can result in thermal shock to the CRD seals and result in non-conservative scram times.
D. The restart of an isolated loop will result in a cold water transient which will result in an increase in the critical power ratio.
OC ILT 07-1 NRC SRO Exam KEY Page 54 of 70
OC ILT 07-1 NRC SRO Exam KEY Knowledge and Ability Reference Information RO SRO 2.1.32 Importance 3,4 3.8 Ability to explain and apply system limits and Rating precautions.
Level 1 SRO 1 Tier# 13 I Group# 1 TS 3.3.F.2.a.l and USAR 15,4.4 References 4.2.5 Bases IAW Tech Specs Bases to TS 3.3.F.2.a.3, an isolated loop will experience a cooling of the loop temperatures greater than 50 O F and restart of an isolated loop could result in a cold water addition transient. The previous paragraph is related to an idle loop restart.
The TS Bases states that an idle loop can be restarted since the restart of the loop will not result in a cold water addition transient causing a concern from either reactivity addition or reactor nozzle thermal stresses. Therefore, startup of an isolated loop, which can result in a cold water addition transient and raises concerns due to the reactivity addition and the thermal stresses on the reactor nozzle.
Therefore, answer B is correct.
IAW TS Bases to 3.3.F.2.a.1, the water trapped between the closed E: planation recirculation suction and discharge valves is not available during a LOCA. This is the basis as to why the MAPLHGR limit is reduced during an isolated loop event. Answer A is incorrect.
Placing cold water past the CRD seals could thermally shock the seals and result in degradation. If the seals were bad enough, the scram times could be affected. But the correct answer is as provided above. Answer C is incorrect.
As stated, the restart of an isolated recirculation loop can cause a cold water addition transient but as shown in the accident analysis (USAR chapter 15) the CPR actual goes down - not up. Answer D is incorrect.
OC ILT 07-1 NRC SRO Exam KEY Page 55 of 70
OC ILT 07-1 NRC SRO Exam KEY Learning 2621.828.0.0038 202-10451 Objective Given Technical Specifications, identify and explain associated actions for each section of the Technical Specifications relating to this system including personnel allocation and equipment operations.
Modified Question Source Bank New X Bank
~
Question Cognitive Memory or X Comprehension Level: Fundamental 1:B or Analysis Knowledge 10 CFR Part 55 55.41 55.43 2 Content:
I Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 56 of 70
OC ILT 07-1 NRC SRO Exam KEY The plant is starting up after an outage.
At 0800, Chemistry has provided the following reactor coolant and AOG analysis, which was completed at 0800 today:
Analysis at 0800 o Reactor coolant pH is 8.9 o Reactor coolant conductivity is 1.9 pS/cm o Reactor coolant chlorides is 0.09 ppm o AOG hydrogen concentration downstream of the recombiner is 4.2% by volume o Reactor coolant activity of 0.1 pCi per gram dose equivalent 1-131 o Turbine steam flow is 50,000 Ib/hr At 1200, Chemistry has provided the following reactor coolant and AOG analysis, which was completed at 1200 today:
Analysis at 1200 o Reactor coolant pH is 9.1 o Reactor coolant conductivity is 2.6 pS/cm o Reactor coolant chlorides is 0.21 ppm o AOG hydrogen concentration downstream of the recombiner is 2.3% by volume o Reactor coolant activity of 0.1 8 pCi per gram dose equivalent 1-131 o Turbine steam flow is 200,000 Ib/hr The current time is 1200. Which of the following is required IAW Technical Specifications (based upon the above data)?
A. No Technical Specifications are required.
- 6. Reduce AOG hydrogen within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
C. An orderly shutdown shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
D. Chlorides and conductivity shall be reduced within the next 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
OC ILT 07-1 NRC SRO Exam KEY Page 57 of 70
OC ILT 07-1 NRC SRO Exam KEY Knowledge and Ability Reference Information RO SRO 2.1 -34 Importance 2.3 2.9 Ability to maintain primary and secondary plant Rating chemistry within allowable limits.
Level SRO Tier # 3 Group #
I I References
~~
I TS 3.3.E I TS 3.6 I The question stem shows that reactor coolant chemistry is within the TS limits with steaming rates < 100,000 Ib/hr at 8:OO am. At noon, the water chemistry does not violate TS 3.3.E.3 with steaming rates >
100,000 Ib/hr. But TS 3.3.E.5 requires that if either conductivity or chlorides is exceeded in this paragraph (which is more limiting than the limits provided in 3.3.E.3), they may remain violated up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Answer D is correct.
Explanation: All other answers are plausible as they are TS actions or interpretations, but incorrect. Other data is provided that do not violate other TS requirements. Answer A is plausible since the data is within the chemistry limits in 3.3.E.4, but 3.3.E.5 also applies. Answer B is plausible since AOG was violated at 0800, but is no longer violated at 1200. Answer C is a requirement under coolant chemistry that does not apply under the given conditions.
Learning 2621.828.0.0039 204-10451 Objective Given Technical Specifications, identify and explain associated actions for each section of the Technical Specifications relating to this system including personnel allocation and equipment operations.
OC ILT 07-1 NRC SRO Exam KEY Page 58 of 70
OC ILT 07-1 NRC SRO Exam KEY x Modified Question Source Bank New (INPO) Bank I 1 Comprehension I Question Cognitive Level:
Memory or Fundamental Knowledge or Analysis 1 X 3:SPR 10 CFR Part 55 55.41 55.43 2 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 59 of 70
OC ILT 07-1 NRC SRO Exam KEY Which of the following proposed changes would require NRC approval prior to implementation?
A. Changing the SRM upscale rodblock to 4x1O4 cps.
B. Changing the Core Spray start signal to 84 RPV water level.
C. Changing the low condenser vacuum scram signal to 23 HG.
D. Changing the Primary Containment isolation signal to 3.1 psig.
OC ILT 07-1 NRC SRO Exam KEY Page 60 of 70 t
OC ILT 07-1 NRC SRO Exam KEY I Question #
Answer I 2, Question Developer InitiaWDate: NTP 10/29/07 Knowledge and Ability Reference Information RO SRO 2.2.5 Importance 1.6 2.7 Knowledge of the process for making changes in the Rating facility as described in the safety analysis report.
Level SRO Tier# 3 Group #
References I
LS-AA-104-1 I
TS 2.3 I
IAW the reference, activities that require a change to Technical Specifications cannot be implemented until formal NRC approval is obtained. IAW TS 2.3.K, core spray starts at 2 72 TAF (86).
Explanation: Assigning a water level below this value would first require NRC approval. All other setpoint are listed in TS, but do not violate the current TS setting. Answer B is correct.
References tc provided duri I
Learning Objective I
Question Source I
/Bank 1 I
1 I
Modified Bank X New 1 Memory
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Question Cognitive or X Comprehension Level: Fundamental or Analysis Knowledge I
10 CFR Part 55 55.41 55.43 5 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 61 of 70
OC ILT 07-1 NRC SRO Exam KEY An Operator on your Crew wishes to modify a step in procedure 315.1, Turbine Generator Startup. You have verified the technical accuracy of the change and it does not involve a change of intent.
Which of the following is correct to process this as a procedure Temporary Change?
- 1. The Temporary Change must be approved by (1)
- 2. The procedure must undergo a full review within (*) days.
(1)
A. Station Qualified Reviewer AND an 14 SRO B. Station Qualified Reviewer AND the 90 Site Functional Area Manager C. Site Functional Area Manager AND an 14 SRO D. Site Functional Area Manager ONLY 90 OC ILT 07-1 NRC SRO Exam KEY Page 62 of 70
oc LT 07-1 NRC SRO Exam KEY Question ##
Question Developer InitialdDate: NTP 10/30/07 Answer Knowledge and Ability Reference Information RO SRO 2.2.1 1 Importance 2.5 3.4 Knowledge of the process for controlling temporary Rating changes.
Level RO Tier ## 3 Group ##
References I I IAW the reference, to allow use of procedure temporary change, it must be approved by a Station Qualified Reviewer and an SRO, and must undergo a full review within 14 days. Answer A is correct. A T&RM temporary change must be approved by the Site Functional Explanation: Area Manager only and be fully reviewed within 14 days. Both types of temporary changes are valid for 90 days. All other answers are incorrect.
References ta provided duri 7
Learning Objective Question Source I Bank I
I New X Question Cognitive Memory or X Comprehension Level: Fundamental 1:P or Analysis Knowledge 10 CFR Part 55 55.41 55.43 5 Content:
I Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 63 of 70
OC ILT 07-1 NRC SRO Exam KEY Radiation Protection surveys have shown that a room in the Reactor Building now shows radiation levels at 1100 mr/hr at 30 cm in the room.
IAW Technical Specification, which of the following states how unauthorized entry into this room shall be prevented (in addition to correct postings)?
A. The room shall be roped-off, with an alarm.
B. The room shall have a locked door with an alarm.
C. The room shall have a gate (locking NOT required).
D. The room shall have a locked door (alarming NOT required).
OC ILT 07-1 NRC SRO Exam KEY Page 64 of 70
OC ILT 07-1 NRC SRO Exam KEY Question #
Question Developer InitiaWDate: NTP 10/30/07 Answer Knowledge and Ability Reference Information 1I RO II SRO 2.3.1 T Importance 1 2.6 1 3.0 Knowledge of 10 CFR: 20 and related facility Rating radiation control requirements.
Level 1 SRO Tier# 13 I Group# I References I 10CFR20.1003 I TS 6.13 I RP-AA-18 IAW TS 6.13.2, a high radiation area in excess of 1000 mrem/hr shall have a locked door, alarm is not required.
I Learning Objective Modified Question Source Bank X New Bank Question Cognitive Memory or X Comprehension Level: Fundamental 1:D or Analysis Knowledge 10 CFR Part 55 55.41 55.43 2 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 65 of 70
OC ILT 07-1 NRC SRO Exam KEY You are reviewing an Operations Department job with the A U R A Committee. There are several options on how to proceed with the job. The OPTIONS are listed below:
(assume only one Operator will be required to perform the job)
Option 1 : Perform the work WITHOUT respiratory protection and WITHOUT any additional shielding 0 The dose rate in the work area is 50 mr/hr 0 Expected work completion time: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 0 Expected internal dose: 25 mrem CEDE Option 2: Perform the work WITH respiratory protection and WITHOUT any additional shielding The dose rate in the work area is 50 mr/hr 0 Expected work completion time: 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Expected internal dose: 15 mrem CEDE Option 3: Perform the work WITHOUT respiratory protection and WITH additional shielding The dose rate in the work area is 30 mr/hr Expected work completion time: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Expected internal dose: 25 mrem CEDE 0 Additional dose to install shielding: 20 mrem Option 4: Perform the work WITH respiratory protection and WITH additional shielding The dose rate in the work area is 30 mr/hr 0 Expected work completion time: 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Expected internal dose: 15 mrem CEDE Additional dose to install shielding: 20 mrem OC ILT 07-1 NRC SRO Exam KEY Page 66 of 70
OC ILT 07-1 NRC SRO Exam KEY In keeping with the ALARA principles, which job option should be selected?
A. Option 1 B. Option 2 C. Option 3 D. Option 4 OC ILT 07-1 NRC SRO Exam KEY Page 67 of 70
OC ILT 07-1 NRC SRO Exam KEY uestion Developer InitialdDate: NTP 10/30/07 Knowledge and Ability Reference Information 2.3.2 Importance Knowledge of facility ALARA program. Rating Level SRO Tier# 3 Group #
References RP-AA- 16 ARARA principles dictate that the option with the lowest achievable dose be selected. The total dose for Option 1 is 125 mrern (50x2 +
25). The total dose for Option 2 is 140 mrem (50x25 + 15). The total Explanation: dose for Option 3 is 105 mrem (30x2 + 25 + 20). The total dose for Option 4 is 110 mrem (30x25 + 15 + 20). Option 3, answer C, represents the lowest total dose to perform the job. Answer C is correct.
Learning Objective Modified Question Source Bank X New Bank Question Cognitive Memory or Comprehension X Level: Fundamental or Analysis 3:SPK Knowledge I I I 10 CFR Part 55 55.41 55.43 5 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 68 of 70
OC ILT 07-1 NRC SRO Exam KEY The plant was at rated power when an event occurred. The Shift Manager, acting as the Shift Emergency Director, has declared an Alert.
The Shift Manager has turned over the command and control responsibilities to the Station Emergency Director in the TSC. Which of the following roles and responsibilities does the Shift Manager have in these conditions?
A. Declare emergency event classifications.
- 8. Make offsite notifications (State/County).
C. Determine/issue Protective Action Recommendations.
D. Coordinate in-plant operations with the TSC Operations Manager.
OC ILT 07-1 NRC SRO Exam KEY Page 69 of 70
w OC ILT 07-1 NRC SRO Exam KEY Question #
Question Developer InitiaWDate: NTP 10/30/07 Answer 1 I Knowledge and Ability Reference Information RO SRO 2.4.37 Importance 2.0 3.5 Knowledge of the lines of authority during an Rating emergency.
Level SRO Tier # 3 Group #
References 1 I
EP-AA-112-100 o1 I
I EP-AA-ll2-2OO-F-I I I I IAW the EP-AA-112-100, Control Room Operations, and EP-AA-112-200-F-01, Station Emergency Director Checklist, the Shift Manager assumes the role as the Shift Emergency Director and turns over to the Station Emergency Director in the TSC. Prior to the turnover, the SM has the responsibilities listed in answers A-C. Once these duties Explanation: have been transferred, the SM will no longer declare emergency events, make offsite notifications to statekounty, and determine PARS. The SM will coordinate in-plant actions with the TSC Operations Manager. Answer D is correct.
References to provided durir Learning Objective Modified Question Source Bank New X Bank Question Cognitive Memory or X Comprehension Level: Fundamental 1:F or Analysis Knowledge 10 CFR Part 55 55.41 55.43 5 Content:
Time to Complete: 1-2 minutes OC ILT 07-1 NRC SRO Exam KEY Page 70 of 70