ML040120739
| ML040120739 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 12/08/2003 |
| From: | Christman R Entergy Nuclear Northeast |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-286/03-303 | |
| Download: ML040120739 (117) | |
Text
Examination Outline Cross-reference:
1 1
/%,/
Tier ##
WA #
00007EK1.06 Importance Rating 4.1 O3 Proposed Question:
Given the following conditions:
Indian Point 3 was operating at 100% power for 300 consecutive days when a loss of offsite power occurred. The crew has stabilized the plant in accordance with station EOPs, and have entered POP 3.1- "Plant Shutdown From 45% Power". Total AFW flow to the Steam Generators has lowered from 300 gpm to 250 gpm over the last hour to properly maintain NR levels.
WHICH ONE of the following describes the reason for this change?
A. A SG tube leak is developing.
B. Auxiliary steam loads are being secured.
C. Decay Heat Is Lowering.
D. CST level is being conserved.
Proposed Answer:
C. Decay Heat Is Lowering Explanation (Optional):
A. There is no indication given of a SG Tube Leak, in normal POPS, not in ONOPs B. S/G Levels are not affected by Auxi1i.at-y Steam Loads C. A W Flow is removing decay heat and reactor coolant pump heat, decay heat is lowering post trip, therefore A W demand is lower D. AFW will be used as necessary to remove decay heat Technical Reference@):
None (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
Bank #
INPO-1281 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam North Anna 1 1/96 Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 X
Comments:
55.43 EK1 Knowledge of the operational implications of the following concepts as they apply to the reactor trip:
(CFR 41.8 / 41.10 / 45.3)
EKl.06 Relationship of emergency feedwater flow to S/G and decay heat removal following reactor trip
Examination Outline Cross-reference:
1 Group #
1 K/A #
0008AK2.02 Importance Rating 2.7 Proposed Question:
Given the following conditions:
The plant was operating at 100% power.
A pressurizer vapor space rupture has occurred and has caused a Reactor Trip and Safety Injection.
The crew has transitioned to E-I I "Loss of Reactor or Secondary Coolant."
0 ADVERSE CONTAINMENT conditions exist.
WHICH ONE of the following parameters checked in E-I does NOT have alternate values specified for use during ADVERSE CONTAINMENT conditions?
A. Containment pressure B. Pressurizer Level C. Pressurizer Pressure D. S/G Narrow Range Level Proposed Answer:
A. Containment Pressure Explanation (Optional):
In E-I, Pressurizer Level, Pressurizer Pressure, and S/G Narrow Range Level all have adverse containment values. Containment Pressure does not.
Technical Reference@):
E-I "Loss of Reactor or Secondarv Coolant Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-EOP-3 1 0 bi I 5
(As available)
Bank #
INPO-I0740 Modified Bank #
New Last NRC Exam IP3 7/96 Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 (Note changes or attach parent)
X AK2-Knowledge of the interrelations between Pzr Vapor Space Accident (CFR 41.7,45.7)
K2.02 Sensors and Detectors
Examination Outline Cross-reference:
1 Group #
1 WA #
00008K3.04 Importance Rating 4.6 Proposed Question:
Indian Point 3 has experienced a Reactor Trip and a Safety Injection. The following parameters are observed:
All Control Rods Have Inserted.
Pressurizer Pressure is 1100 psig and is slowly lowering.
Pressurizer Level is 100%.
RCS Subcooling is 30 degrees F as read on qualified CETs.
Containment Pressure is 3 psig and is slowly rising.
Containment Particulate and Gaseous Radiation levels are rising slowly.
HHSl is operating as required.
All 4 Reactor Coolant Pumps are running.
What is the required Reactor Coolant Pump configuration for this condition?
A. Run all 4 RCPs To Allow Proper Mixing of Borated Safety Injection Water.
B. Trip 3 of 4 RCPs and leave 31 RCP running for pressurizer spray flow.
C. Trip all 4 RCPs as required due to low RCS Subcooling.
D. Trip all 4 RCPs since a Phase B actuation should have occurred.
Proposed Answer:
C. Trip all 4 RCPs as required due to low RCS Subcooling.
Explanation (Optional):
A. Running all 4 RCPs is not allowed per E-0 B. Must trip all 4 RCPs per E-0 C. Answer is correct, HHSl is running and Subcooling is 40F per E-0 D. Phase B has not occurred Technical Reference(s):
E-0 SteD 13 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
Comments:
55.43 AK3-Knowledge of reasons for the following as they apply to Pzr Vapor Space Accidents (CFR 4 1.5,4 1.1,45.6,45.13)
K3.04 RCP Tripping Requirements
Examination Outline Cross-reference:
Tier #
1 Group #
WA #
00009EA2.01 Importance Rating 4.8 Proposed Question:
Safety Injection flow has been terminated for a Small Break LOCA using ES-1.2 Post LOCA Cooldown and Depressurization. Subsequently the Reactor Operator notes the following:
RCS subcooling is 10 degrees F as indicated on qualified Core Exit Thermocouples SG levels indicate 28% on all S/Gs Total AFW flow is 450 gpm RCS pressure is 1750 psig and slowly decreasing Pressurizer level is 10% and stable Normal Charging is aligned and charging flow is indicated ONE HHSl pump has been stopped e
What is the NEXT action required based on the above data?
A. Re Enter E-I, Loss of Reactor Or Secondary Coolant.
B. Manually start and align HHSl pumps.
C. Depressurize the RCS to refill the PZR.
D. Continue with next step in ES-I.2. No contingency actions are required.
Proposed Answer:
B. Manually start and alinn HHSl DumDs.
Explanation (Optional):
A. Not an appropriate transition-Situation requires SI pump restart-Transitions recommended by TSC B. Action called out by step 18 in ES 1.2 Post LOCA Cooldown and Depressurization Technical Reference(s):
ES 1.2 Post LOCA Cooldown and DeDressurization-Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-EOP-33 Obi V None Question Source:
Bank #
INPO 084 Braidwood 6/99 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam 6/99 Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
Memory or Fundamental Knowledge Comprehension or Analysis 55.41 55.43 x
EA2 Ability to determine or interpret the following as they apply to a small break LOCA:
(CFR43.5 /45.13)
EA2.01 Actions to be taken, based on RCS temperature and pressure, saturated and superheated X
Examination Outline Cross-reference:
1 Group #
1 KIA #
00001 1 EA1.07 Importance Rating 4.4 Proposed Question:
Indian Point 3 was operating at 100% when a Design Basis Large Break Loss of Coolant Accident occurred ten minutes ago. The following parameters currently exist:
All Control Rods are inserted Pressurizer Pressure is I00 psig Pressurizer Level is 0%
RVLIS indicates that the core is covered Safety Injection equipment has functioned normally RCS Subcooling is OF Containment radiation levels are in alarm and slowly rising Containment Pressure is 17 psig and slowly lowering.
Which of the following is correct concerning Containment Isolation?
A. Phase A has actuated on high containment pressure signal, and Phase B has actuated on high containment pressure signal.
B. Phase A has actuated on the Safety Injection Signal and Phase B has actuated on high containment pressure signal.
C. Phase A has not actuated and Phase B has not actuated since RCPs are required for this condition.
D. Phase A has actuated on the Safety Injection Signal and Phase B has not actuated.
Proposed Answer:
B. Phase A has actuated on the Safety Injection Signal and Phase B has actuated on high containment pressure signal.
Explanation (Optional):
A. Phase A does not actuate on a reactor trip signal B. Phase A would have actuated on SI signal, Phase B actuates on containment pressure C. Both Phase A and B have actuated and RCPs need to be secured per E-0 D. Phase A Statement is correct, however Phase B should have actuated. Large Break LOCA of indicated magnitude causes containment pressures much higher than the trip setpoint of 22 psig.
Containment pressure of 17 psig is meant to show that containment cooling and containment spray are having an effect.
Technical Reference( s) :
Proposed references to be provided to applicants during examination:
None Learning Objective: LIC-ESS-08 Obi 8 Question Source:
Bank #
Modified Bank #
New X
Question History:
Last NRC Exam (Note changes or attach parent)
Question Cognitive Level:
Memory or Fundamental Knowledge
Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
Comments:
55.43 EA1 Ability to operate and monitor the following as they apply to a Large Break LOCA:
(CFR 41.7 / 45.5 / 45.6)
EAl.07 Containment isolation system
Examination Outline Cross-reference:
I Group #
1 WA #
0001 5/17AA1.05 Importance Rating 3.8 Proposed Question:
Indian Point Unit 3 is operating at 100% power when the shaft on the No. 32 Reactor Coolant Pump (RCP) seizes, causing RCS flow to lower. Which of the following occurs as a result of the pump failure?
A. Reactor trip caused by RCP No. 32 breaker over current, followed by a safety injection caused by low RCS flow.
B. Reactor trip caused by low level in No. 32 steam generator, followed by a safety injection caused by high steam-flow coincident with low steam pressure in No. 32 steam generator.
C.OT-delta run back in the affected loop, followed by a reactor trip caused by low RCS flow.
D. Reactor trip caused by the opening of RCP No. 32 breaker, followed by a safety injection caused by steamline differential pressure.
Proposed Answer:
D. Reactor trip caused by the opening of RCP No. 32 breaker, followed by a safety injection caused by steamline differential pressure.
Explanation (Optional):
A. Low RCS Flow does not cause a Safety Injection
- 6. Stopping an RCP in a loop will not cause a High Steam Flow Condition or a Low Steam Pressure Condition C. OT-Delta T runback does not occur on a loss of a running RCP D. Reactor Trip signal above P-8 driven off of breaker open contacts, a steam line DIP condition will develop Technical Reference(s):
(Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective:
(As available)
None Question Source:
Bank #
INPO 10531 IP3 4/96-Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam 4/96 Question Cognitive Level:
Memory or Fundamental Knowledge -
Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
Comments:
APE : 015/017 Ability to operate and/or monitor the following as they apply to Reactor Coolant Pump (RCP)
Malfunctions (CFR 41.7 45.5,45.6)
Examination Outline Cross-reference:
1 Group #
1 WA #
000022AA1.09 Importance Rating 3.3 Proposed Question:
Indian Point 3 is operating at 100% power and has just received a "RCP Seal Inj. Filter High D/P" alarm on panel SFF. In addition, "Thermal Barrier CCW Header Low Flow" on Panel SGF is in alarm.
The Reactor Operator determines that Seal Injection Flow has been lost to all Reactor Coolant Pumps. The following RCP indications exist:
RCP Seal Diff Pressure Frame Vibration Shaft Vibration Seal Outlet Temp 31 400 psid 1.3 mils 9.3 mils 204 degrees F 32 400 psid 1.9 mils 10.1 mils 210 degrees F 33 400 psid 1.9 mils 12.2 mils 241 degrees F 34 400 psid 1.4 mils 8.5 mils 215 degrees F Which statement describes the required actions, if any, for these conditions?
A. RCP operation may continue, there are no parameters requiring an immediate RCP Shutdown.
B. At Least ONE RCP must be immediately shutdown due to RCP Seal Outlet Temperatures exceeding operating limits.
C. At Least ONE RCP must be immediately shutdown due to RCP Shaft Vibration levels exceeding operating limits.
D. All RCPs must be immediately shutdown due to simultanous Loss of RCP Seal Injection Flow and "Thermal Barrier CCW Header Low Flow"" alarm.
Proposed Answer:
B.. At Least ONE RCP must be immediately shutdown due to RCP Seal Outlet Temperatures exceeding operating limits.
Explanation (Optional):
A. RCP Seal Outlet Temperature Requires an immediate RCP shutdown above 235F B. RCP Seal Outlet Temperature Requires an immediate RCP shutdown above 235F C. RCP Vibration Levels are within limits D. ONOP RCS-5 Requires Attachment 1 actions be taken without a direct step to trip any RCPs Technical Reference@):
ARP - 009. ONOP-RCP-5 Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
(As available)
Bank #
Modified Bank #
(Note changes or attach parent)
New X
Last NRC Exam N/A Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 X
55.43 Comments:
AA1. Ability to operate and I or monitor the following as they apply to the Loss of Reactor Coolant Pump Makeup:
(CFR 41.7 145.5 / 45.6)
AAl.09 RCP seal flows, temperatures, pressures, and vibrations
Examination Outline Cross-reference:
1 1
Group #
WA #
00025AK3.03 Importance Rating 4.1 Proposed Question:
While operating in Reduced Inventory/Midloop conditions with the RCS temperature 130F. RCS level indication is noted to be 61' 8 " and lowering. Which of the following actions would be correct to implement for this condition ?
I.
Raise RCS Level using HHSl pump(s), to raise pump Net Positive Suction Head.
- 2. Stop all running RHR pumps, to minimize the possibility of Gas Binding the pumps.
- 3. Raise RCS Pressure to greater than 150 psig, to minimize the possibility of Gas Binding the RHR pumps.
- 4. Establish level in at least 2 S/G's >15% narrow range, to establish a back up heat sink.
A. Actions 1 & 2 B. Actions I&
3 C. Actions 1, 2 & 3 D. Actions 2, 3 & 4 Proposed Answer:
A. Actions 1 & 2 Explanation (Optional):
Items 1 and 2 are direct procedure steps in ONOP RHR-2 " Loss of RHR With the RCS Drained or At Midloop" Technical Reference(s):
ONOP RHR-2 Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
Bank #
INPO 577 Beaver Valley 3197-Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Beaver Vallev 3/97 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
55.43 Comments:
AK3 Knowledge of the reasons for the following responses as they apply to a loss of RHR: (CFR 41.5,41.145.6,45.13)
AK3.03 Immediate Actions contained in the EOPs for Loss of RHR
Examination Outline Cross-reference:
1 WA ##
000026AA2.02 Importance Rating 3.6 Group #
1 Proposed Question:
Indian Point 3 is at 100% power.
a 31 and 32 CCW surge tank High-Low alarm conditions are present.
31 and 32 CCW surge tank levels are LOWERING.
Actions have not yet been taken to split out CCW headers The CCW Surge Tank makeup valves are OPEN.
Both CCW Surge Tank levels are 4%
and operators have stopped all CCW Pumps per a
a a
a ONOP CC-1"Loss of Component Cooling" Which ONE of the following is the cause of the loss of CCW?
A Tube Leak In.....
A. The Non-Regenerative Heat Exchanger B. An RCP Thermal Barrier Heat Exchanger C. The Seal Water Heat Exchanger D. The Excess Letdown Heat Exchanger Proposed Answer:
C. The Seal Water Heat Exchanger Explanation (Optional):
The only system listed that operates at a sufficiently low pressure to cause leakage of CCW into the system is the Seal Water Heat Exchanger.
Technical Reference@):
ONOP CC-I Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
(As available)
Bank #
INPO-5108 Farlev 10/95 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam 10/95 Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 X
55.43 AA2. Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water:
(CFR: 43.5 / 45.13)
AA2.02 The cause of possible CCW loss
Examination Outline Cross-reference:
1 Group #
1 WA #
2.4.41 Importance Rating
- 4. I Proposed Question:
Indian Point 3 has just experienced a Manual Reactor Trip and Safety Injection from 100% power.
A Pressurizer PORV has failed open. Actions in ONOP RCS-2 "Malfunction of Pressurizer Pressure Control System" have been taken, and the associated PORV block MOV failed to close. Actions of E-0 "Reactor Trip or Safety Injection" are being taken.
Other plant conditions are as follows:
All Control Rods are Inserted Total HHSl flow is 200gpm RCS Pressure is 11OOpsig Pressurizer Level is 100%
R-I 1 and R-12 read 100 mWhr PRT Level is offscale high PRT Pressure is lpsig Containment Pressure is 1 psig What is the appropriate EAL Classification?
A. Notification of Unusual Event B. Alert C. Site Area Emergency D. General Emergency Proposed Answer:
B. Alert Explanation (Optional):
Indications of a LOCA-1 Barrier Failed-Alert Technical Reference@):
E Plan-EALs Proposed references to be provided to applicants during examination:
E Plan-EALs Learning Objective:
LIC-ERT-13 Obi-5 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
Comments:
(CFR:
43.5 / 45.1 1) 55.41 55.43 x
2.4.4 1 Knowledge of the emergency action level thresholds and classifications.
Examination Outline Cross-reference:
1 WA #
00029EA1.08 Group #
1 Importance Rating 4.5 Proposed Question:
Given the following plant conditions:
Reactor power was 65% when the turbine tripped and an ATWS occurred The reactor tripped 20 seconds later when Train A reactor trip breaker was opened via the reactor trip pushbutton trip pushbutton Train B reactor trip breaker is failed closed, and will not operate locally or from the reactor No controls other than control rods and boration controls have been operated Which one of the following correctly describes the operation of the steam dumps for these conditions?
Steam Dumps will open.....
A. immediately following the turbine trip and modulate to stabilize Tavg 5 degrees above its no-load value.
B. when the reactor trip breaker is opened and modulate to stabilize Tavg at its no-load value.
C. immediately following the turbine trip and modulate to stabilize Tavg at its no-load value.
D. when the reactor trip breaker is opened and will be blocked closed when Tavg falls below its low-low value.
Proposed Answer:
A. immediately following the turbine trip and modulate to stabilize Tavg 5 degrees above its no-load value.
Explanation (Optional):
A. The Steam Dumps are armed immediately by the load rejection, and the reactor trip controller is not used since the B Reactor Trip Breaker stays shut.
B. Both concepts are wrong-The steam dumps open before the reactor trip breaker opens, and reactor trip controller is not active so Tavg stays 5 degrees above no-load value C. Reactor trip controller is not active so Tavg stays 5 degrees above no-load value D. The steam dumps open before the reactor trip breaker opens Technical Reference@):
(Attach if not previously provided)
Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-SPC-02 Obi 3 Question Source:
Bank #
INPO 15170- Salem 1-2/99 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam 2/99 Salem I Question Cognitive Level:
10 CFR Part 55 Content:
Memory or Fundamental Knowledge 55.41 X
55.43 Comprehension or Analysis X
Comments:
EA1 Ability to operate and monitor the following as they apply to a reactor trip:
(CFR 41.7 / 45.5 / 45.6)
EAl.08 Rx Trip Switch Pushbuttons.
Examination Outline Cross-reference:
1 Group #
1 KIA #
00029EK1.03 Importance Rating 3.8 Proposed Question:
Which ONE of the following is the preferred method of injecting highly borated water into the RCS to insert the desired reactivity change during an ATWS per EOP FR-S.l "Response To Nuclear Power Generation-ATWS?
A. Emergency Boration using the RWST B. Emergency Boration by failing air to FCV-11 OA C. Emergency Boration through CH-MOV-333 D. Emergency Boration using Normal Boration via the Blender Proposed Answer:
C. Emergency Boration through CH-MOV-333 Explanation (Optional):
Four paths are mentioned, CH-MOV-333 is the Normal Response Path, the others are in the response not obtained column.
Technical Reference(s):
EOP FR-S. 1 "ResDonse To Nuclear Power Generation-ATWS Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-EOP-37 Obi1 0 Bank #
INPO-1157 Callawav-2/97 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam 2/97 Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X
EK1 Knowledge of the operational implications of the following concepts as they apply to the ATWS:
(CFR 41.8 / 41.IO / 45.3)
EK1.03 Effects of boron on reactivity
Examination Outline Cross-reference:
1 Group #
1 WA #
00038EA2.12 Importance Rating 4.2 Proposed Question:
Indian Point 3 has just experienced a SGTR in 32 S/G. E-0
'I Reactor Trip or Safety Injection" actions have led to a transition to E-3 "Steam Generator Tube Rupture". The operator has attempted to isolate steam supplies from the ruptured S/G by closing 32 MSIV. Overhead Annunciator "Main Steam Isolation Valve Closed" on Panel FAF "Turbine First Out Annunciator has not alarmed. What is the next action that must be taken to limit the release of radioactivity from 32 S/G?
A. Stop 32 RCP to limit heat transfer from 32 S/G and commence a cooldown with Steam Generators via Normal Steam Dumps.
B. Continue attempts to close the 32 MSIV and continue actions to cool down with all Steam Generators via Normal Steam Dumps.
C. Isolate the remaining Steam Generators by closing their MSlVs and cool down using intact Steam Generator Atmospheric Valves.
D. Close down on intact Steam Generator AtmosphericValves, continue attempts to close 32 MSIV.
Proposed Answer:
C. Isolate the remaining Steam Generators by closing their MSlVs and cool down using intact Steam Generator Atmospheric Valves.
Explanation (Optional):
The strategy spelled out in E-3 is to isolate the intact steam generators and cooldown in an expeditious manner to limit release from the ruptured S/G. Normal Steam Dumps become unavailable as a result of the Safety Injection.
Technical Reference(s):
E-3 "Steam Generator Tube RuDture"-
Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-35 Obi1 4 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis I O CFR Part 55 Content:
55.41 55.43 x
X Comments:
EA2 Ability to determine or interpret the following as they apply to a SGTR:
(CFR 43.5 / 45.13)
EA2.12 Status of MSIV activating system
Examination Outline Cross-reference:
I KIA #
000040AK2.02 Importance Rating 2.6 Group #
1 Proposed Question:
Given the following conditions:
e The plant is at a steady-state power level of 90%.
Pressurizer pressure and level have suddenly started lowering.
Pressurizer pressure and level control systems are responding properly in e
e AUTO.
WHICH ONE of the following parameters ALONE can be used, PRIOR to a plant trip to determine that the pressurizer changes are the result of a Steam Line Rupture?
A. Pressurizer Level B. Loop Differential Temperature C. Containment Humidity D. Containment Particulate Radiation RMS Proposed Answer:
B. Loop Differential Temperature Explanation (Optional):
A. Pressurizer Level will lower for both events prior to a plant trip.
B. Since loss of coolant accident does not have an immediate effect on RCS temperature and a Steam Line Rupture does, this is the determining sensor.
C. Containment Humidity rise for both events.
D. Response time is too fast for the Containment Particulate Detector to see the event.
Technical Reference(s):
Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-35 Obi2 (As available)
Question Source:
Bank #
lNPOl0633 IP3 7/96 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam IP3 7/96 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 X
Comments:
AK2. Knowledge of the interrelations between the Steam Line Rupture and the following: (CFR 41.7 / 45.7)
AK2.02 Sensors and detectors
Examination Outline Cross-reference:
000054AA2.03 4.1 Tier #
Group #
Importance Rating Proposed Question:
Indian Point Unit 3 is operating at 100% power when offsite electrical power is lost and a simultaneous SI signal is generated. The control room crew immmediately enters E-0. At 1 min. and 20 sec. after losing offsite power, the crew notes that the turbine-driven A B M pump is running but that the motor driven ABFW pumps are not running.
Should the motor-driven ABFW pumps be expected to be running by this point? Why?
A. No. The SI sequencer does not start the motor-driven pumps on a loss of offsite power: only the turbine-driven ABFW pump is available.
B. No. The non-SI sequencer does not start the motor-driven ABFW pumps until 90 seconds have elapsed.
C. Yes. The non-SI sequencer should have started the motor-driven ABFW pumps by now.
D. Yes. The SI sequencer should have started the motor-driven ABFW pumps by now.
Proposed Answer:
D. Yes. The SI sequencer should have started the motor-driven ABFW pumps by now.
Explanation (Optional):
There are two concepts being examined. First, the SI Sequencer should be in use for the conditions noted. Second, The motor driven ABFW pumps should have started on the loss of all main feedwater as appropriately timed on the SI Sequencer.
Technical Reference@):
None (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-SPC-009 Obi 5 (As available)
Bank #
INPO-10498 IP3 4/96-Modified Bank #
New (Note changes or attach parent)
Last NRC Exam lP3 4/96 Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 55.43 x
Comments:
AA2. Ability to determine and interpret the following as they apply to the Loss of Main Feedwater 0:
(CFR: 43.5 / 45.13)
AA2.03 Conditions and Reasons for AFW Pp Startup
Examination Outline Cross-reference:
1 WA #
00055EA2.03 Importance Rating 3.9 1
Group #
Proposed Question:
After a Station Blackout, WHICH ONE of the following 48OVAC bus combinations must be reenergized to restore operability to the safeguards valves required to regain availability of both trains of safety injection?
A. Busses 2A and 3A B. Busses 3A and 5A C. Busses 5A and 6A D. Busses 3A and 5A Proposed Answer:
C. Busses 5A and 6A Explanation (Optional):
ECA- 0.0 "Loss of All AC Power" Provides bus restoration guidance. LIC-ESS-01 Lesson Guide States that two trains of ECCS equipment are needed, energizing buses 5A and 6A gives 2 full trains, none of the other combinations do.
Technical Reference( s) :
ECA- 0.0 "Loss of All AC Power" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source :
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-EOP-36 Obi1 1 (As available)
Bank #
INPO-I0605 IP3-7/96-Modified Bank #
New (Note changes or attach parent)
Last NRC Exam IP3 7/96 Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X
EA2 Ability to determine or interpret the following as they appiy to a Station Blackout: (CFR 43.5 / 45.13)
EA2.03 Action Necessary To Restore Power
Examination Outline Cross-reference:
1 Tier #
Group #
Importance Rating 3.7 WA #
00056 2.4.8 Proposed Question:
A Station Blackout occurred 2 hrs. ago. After the first hour, Emergency Diesel Generators were started and continue to run. A significant RCP Seal Leak developed during the Station Blackout. The control room crew has been implementing FR-C.l, "Response to Inadequate Core Cooling," when the RWST Low Level alarm energizes and an additional Red Path develops on Secondary Heat Sink.
Which one of the following actions should be taken by the operating crew?
A. Remain in FR-C.l until T-hot and the CETs show a lowering trend, and then transition to ES-1.3, I' Transfer to Cold Leg Recirculation."
- 8. Immediately transition to ES-1.3, "Transfer to Cold Leg Recirculation."
C. Immediately transition to FR-H.1, "Response to Loss of Secondary Heat Sink".
D. Remain in FR-C.1 until a Core Cooling Purple Path is established, then transition to FR-H.l, "Response to Loss of Secondary Heat Sink."
Proposed Answer:
B. Immediately transition to ES-1.3, "Transfer to Cold Leg Recirculation."
Explanation (Optional):
Westinghouse EOP Rules of usage prioritize swap to Cold Led Recirculation on Low RWST Level as taking precedence over all FRP's in use at the time that swapover is required.
Technical Reference@):
Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-31 Obi 12-(As available)
Question Source:
Bank #
INPO-5587 Salem 1/96 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Salem 1/96 Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
Memory or Fundamental Knowledge 55.41 55.43 X b(5)
Comprehension or Analysis X
2.4.8 Knowledge of how the event-based emergency/abnormal operating procedures are used in conjunction with the symptom-based EOPs.
(CFR 41.10/43.5 /45.13)
Examination Outline Cross-reference:
1 Group #
1 KIA #
00057 2.2.2 Importance Rating 3.5 Proposed Question:
Indian Point Unit 3 is in Mode 1 at 12% reactor power with a plant start up in progress. Power is being raised to roll the main turbine and place it on the grid. 33 Instrument Bus Channel IV (Yellow) is lost and can not be restored for approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. What is the preferred mode of heat removal that is available to the operator?
A. Steam Dumps to the Main Condenser in Automatic, Tave Control Mode B. Steam Dumps to the Main Condenser in Automatic, Main Steam Pressure Control Mode C. Steam Dumps to the Main Condenser In Manual Operation D. Steam Dumps to the Main Condenser are not available, Steam Generator Atmospherics must be uti ked.
Proposed Answer:
C. Steam Dumps to the Main Condenser In Manual Operation Explanation (Optional):
ONOP-EL3 "Loss of an Instrument Bus" states that for a loss of 33 Instrument Bus, that Steam Dumps are only available in Manual Mode.
Technical Reference(@:
ONOP-EL3 "Loss of an Instrument Bus"-
Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-ONP-02 Obi4 Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X
2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.
(CFR 45.2)
Examination Outline Cross-reference:
1 WA #
000058AA1.03 Importance Rating 3.3 Group #
I Proposed Question:
Indian Point 3 has experienced a Loss of DC Bus 31. The crew has entered ONOP-EL6 "Loss of a DC Bus". Maintenance has determined that there is a fault on Battery 31. What is the proper sequence of actions required in accordance with ONOP-EL-5 "Loss of a DC Bus" to allow maintenance on Battery 31?
A. Isolate Battery 31 by opening the incoming breaker, Remove Battery Output Fuses, Place DC Bus 31 on its Battery Charger.
B. Place DC Bus 31 on its Battery Charger, Isolate Battery 31 by opening the incoming breaker, Remove Battery Output Fuses.
C. Remove Battery Output Fuses, Isolate Battery 31 by opening the incoming breaker, Place DC Bus 31 on its Battery Charger.
D. Isolate Battery 31 by opening the incoming breaker, Place DC Bus 31 on its Battery Charger, Remove Battery Output Fuses.
Proposed Answer:
A. Isolate 31 Battery by opening the incoming breaker, Remove Battery Output Fuses, Place DC Bus 31 on its Battery Charger.
Explanation (Optional):
Procedurally, answer A provides the only correct sequence Technical Reference(@:
ONOP-EL-5 "Loss of a DC Bus Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-EDS-7 Obi 7 Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X
AA1. Ability to operate and lor monitor the following as they apply to the Loss of DC Power:
(CFR 41.7 / 45.5 / 45.6)
AA 1.03 Vital and battery bus components
Examination Outline Cross-reference:
1 Group #
I WA #
000062AA2.02 Importance Rating 3.6 Proposed Question:
Indian Point Unit 3 is operating at 100% power, with the 4/5/6 service water header as the essential header and the Nos. 31 and 35 pumps in operation. The following alarms are received in the control room:
8 "480V SWGR MOTOR TRIP (COMMON)"
"SERVICE WTR HDR (34.35.36) HIGH LOW PRESS" 0
8 "SERVICE WATER STRAINER TROUBLE" What action should the operating crew take in response to the alarms, and why?
A. Stop the No. 38 service water pump after it auto starts, since the the strainer is clogged on 38 service water pump.
B. Start the No. 38 service water pump, since the 35 service water pump breaker tripped.
C. Start the No. 34 service water pump, since the 35 service water pump breaker tripped.
D. Stop the No. 34 service water pump after it auto starts, since the strainer is clogged on 35 service water pump.
Proposed Answer:
C. Start the No. 34 service water pump, since the 35 service water pump breaker tripped.
Explanation (Optional):
A. 38 Service Water pump does not auto start, and can not be used as Tech Spec operable B. 38 Service Water Pump is a backup pump and would only be started if normal header pumps could not be started.
C. 34 Service Water Pump is one of two pumps that should be started, indications are that 35 Pump tripped.
D. 34 Service Water Pump does not Auto Start on low header pressure Technical Reference( s) :
ONOP-RW-1 Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-ONP-37 Obi 3 Question Source:
Bank #
Modified Bank ##
INPO-10517 IP3 4/96 (Changed Stem and Answer/Distractors to answer cause of loss of SWS as well as actions)
New Question History:
Last NRC Exam IP3 4/96 Question Cognitive Level:
Memory or Fundamental Knowledge 10 CFR Part 55 Content:
55.41 55.43 X b.5 Comments:
AM. Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:
(CFR:
43.5 / 45.13)
AA2.02 The cause of possible SWS loss X
Comprehension or Analysis
Examination Outline Cross-reference:
I Group #
1 WA #
000065 2.1.22 Importance Rating 3.3 Proposed Question:
Indian Point 3 is at 100% power when a Loss of Instrument Air occurs. ONOP-IA-1" Loss of Instrument Air" has been entered. Instrument Air header pressure is at 60 psig and lowering due to an unisolable leak. Over the next 5 minutes, what operating MODE and condition will the plant be in?
A. Mode 1 at 100% power, normal operation B. Mode lat 50% power, due to decreased Boiler Feed C. Mode 3 at 0% power, due to a reactor trip only D. Mode 3 at 0% power, due to a reactor trip and safety injection Proposed Answer:
B. C. Mode 3 at 0% power, due to a reactor trip only Explanation (Optional):
ONOP-IA-1 has 6Opsig as the threshold for initial operator actions. A manual react-r trip is required.
Technical Reference@):
ONOP-IA-1 Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-ONP-27 Obi 3 Bank #
INPO-10516 IP3 4/96 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam IP3 4/96 Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X
Comments:
2.1.22 Ability to determine Mode of Operation.
(CFR 43.5 / 45.13)
Examination Outline Cross-reference:
Tier #
Group #
I WA #
W/E04 EK3.4 Importance Rating 3.8 Proposed Question:
Given the following conditions:
0 A LOCA outside containment has occurred The Reactor Operator has manually actuated a Safety Injection.
The crew has completed ECA-1.2, "LOCA Outside Containment," and transitioned to 0
ECA-1.1, "Loss of Emergency Coolant Recirculation" Why is subcooling minimized once cooldown has been started?
A. This lowers the RCS pressure reducing the amount of RCS inventory loss.
B. It allows the operator to stop all ECCS pumps.
C. It allows RHR to be placed in service in cooldown mode earlier.
D. To limit concerns with causing Pressurized Thermal Shock of RCS Components.
Proposed Answer:
A. This lowers the RCS pressure reducing the amount of RCS inventory loss.
Explanation (Optional):
Technical Reference(s):
ECA-1.2. "LOCA Outside Containment,"/ECA-I. 1."Loss of Emeraency Coolant Recirculation" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
??&??
Bank #
INPO-19486 Kewaunee 12/2000 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam Kewaunee I U2000 Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 X
55.43 Comments:
EK3. Knowledge of the reasons for the following responses as they apply to the @OCA Outside Containment)
(CFR 41.5 /41.10,45.6,45.13)
EK3.4 RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.
Examination Outline Cross-reference:
1 Tier #
Group #
WA #
W/E04 EA2.2 Importance Rating 4.2 Proposed Question:
Given the following conditions on Indian Point 3:
0 A LOCA has occurred outside of the Containment ECA-I.I "Loss Of Emergency Recirculation" is the procedure in effect Containment Pressure is 23 psig due to a steam line break on 32 Steam Generator.
e 0
Which one of the following correctly describes the reasons for the operator's actions associated with the Containment Spray System?
The Containment Spray System is operated as directed in...
A. ECA-1. I "Loss Of Emergency Recirculation" because it establishes minimum required containment spray flow and conserves RWST inventory.
B. ECA-1.1 "Loss Of Emergency Recirculation" since FRPs are not implemented during the performance of ECA-I.I.
C. FR-Z. 1 "Response To High Containment Pressure" because the actions concerning Containment Spray operation are more restrictive.
D. FR-Z. 1 "Response To High Containment Pressure" since restoration of the critical safety function takes precedence.
Proposed Answer:
A. ECA-1.1 "Loss Of Emergency Recirculation" because it establishes minimum required containment spray flow and conserves RWST inventory.
Explanation (Optional):
A. Correct-ECA 1. I establishs desired relationship between heat removal and containment spray strategy with a loss of recirc capability B. FRPs are implemented during ECA 1.1 C. FR-Z.l actions are less restrictive, and procedure defers to ECA I.I if in effect D. FR-Z.l does take precedence, however the Containment Spray System operation is defered to ECA 1.1 Technical Reference@):
Proposed references to be provided to applicants during examination:
ECA-1.I "Loss Of Emeraencv Recirculation". FR-Z. 1 "ResDonse To High Containment Pressure" None Learning Objective:
LIC-EOP-41 Obi 11 Question Source:
Bank #
INPO-15557 Salem 2/99 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Salem 2/99 Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 55.43 x EA2. Ability to determine and interpret the following as they apply to
the (LOCA Outside Containment)
(CFR 43.5 / 45.13)
EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.
Examination Outline Cross-reference:
1 WA #
W/E 11 EK2.2 Importance Rating 4.3 Tier #
Group #
Proposed Question:
The plant has experienced a Loss of Coolant Accident. Both trains of Emergency Coolant Recirculation have become unavailable. RWST level has lowered to less than 1.5 feet and all suction sources from the RWST have been stopped. Step 27 of ECA-1.1, "Loss of Emergency Coolant Recirculation" directs that all S/Gs be depressurized to less than 700 psig. What is the desired outcome of this action?
A. It allows maximum AFW flow to the S/Gs, so that levels can rapidly be restored.
B. It sets up conditions for controlled injection to the RCS from the accumulators.
C. It ensures that there is adequate subcooling for restart of the RCPs.
D. It decreases RCS temperature and pressure which reduces break flow from the LOCA.
Proposed Answer:
B. It sets up conditions for controlled injection to the RCS from the accumulators.
Explanation (Optional):
The Westinghouse EOP Basis document describes the reason that Steam Generators are depressurized in the instance is to inject Accumulators in a controlled fashion to add RCS inventory.
Technical Reference(s):
Westinahouse Owners Group Guidance For ECA-1.I, "Loss of Emercrencv Coolant Recirculation" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
????
Bank #
Modified Bank #
New INPO-3414 9/98 Braidwood (Note changes or attach parent)
Last NRC Exam 8/98 Braidwood Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 X
55.43 EK2. Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following:
(CFR 41.7 / 45.7)
EK2.2 Facility*s heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
Examination Outline Cross-reference:
Level Tier #
Group #
WA #
I I W/E05 EK2.2 4.2 Proposed Question:
A Reactor Trip with SI occurs. The operators perform the immediate action steps, verify ECCS flow, and check Auxiliary Feedwater flow. Minimum Auxiliary Feedwater flow cannot be established, operators enter FR-H.1, "Response to Loss of Secondary Heat Sink". An operator checks RCS pressure; it is less than all S/G pressures. The operators are directed by FR-H.l to E-I, "Loss of Reactor or Secondary Coolant".
Based on this information, select the ONE statement that correctly summarizes plant conditions:
A. A Large Break LOCA is in progress; a secondary heat sink is required.
B. A Large Break LOCA is in progress; a secondary heat sink is not required.
C. A Small Break LOCA is in progress; a secondary heat sink is required.
D. A Small Break LOCA is in progress; a secondary heat sink is not required.
Proposed Answer:
B. A Large Break LOCA is in progress; a secondary heat sink is not required.
Explanation (Optional):
Per WOG Guidance, RCS Pressure being less than S/G pressure in FR-H.l indicates a sufficiently large LOCA in progress to remove heat via break flow, and ECCS injection.
A. Incorrect-Secondary Heat Sink not required B. Correct-Large LOCA and Secondary Heat Sink Is Not Required C. Incorrect-By definition LOCA is "large" D. Incorrect-By definition LOCA is "large" Technical Reference@):
FR-H.1, "ResDonse to Loss of Secondarv Heat Sink" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-EOP-39 Obi-J Bank #
INPO-4196 Braidwood 4/96 Modified Bank #
New Last NRC Exam Braidwood 4/96 Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 (Note changes or attach parent)
X Comments:
EK2. Knowledge of the interrelations between the (Inadequate Heat Transfer) and the following:
(CFR: 41.7 / 45.7)
EK2.2 Facility*s heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
Examination Outline Cross-reference:
I Group #
2 WA #
000003AK3.05 Importance Rating 4.1 Proposed Question:
A dropped rod has occurred and the Operating Crew is responding in accordance with plant procedures. A step in the procedure, which references Technical Specifications, states, "If Axial Flux Difference (AFD) cannot be restored within the required band, then reduce thermal power to less than or equal to 50% within the next 30 minutes."
The basis for this action includes all EXCEPT which one of the following choices?
A. To ensure adequate shutdown margin is maintained during recovery of the dropped rod.
B. To limit power distribution skewing so core peaking factors are consistent with assumptions used in the safety analyses.
C. To prevent invalidating the conclusions of the transient and accident analyses with regard to fuel clad integrity.
D. To ensure Heat Flux Hot Channel Factor is NOT exceeded during normal operation or in the event of xenon redistribution following power changes.
Proposed Answer:
A. To ensure adequate shutdown margin is maintained during recovery of the dropped rod.
Explanation (Optional):
Answers B,C,D are directly extracted from the 1P3 Tech Spec Basis Document. Answer A is the exception since Shutdown Margin is not affected by the position of a dropped rod at power.
Technical Reference(s):
IP3 Tech SDecs and Basis Document Proposed references to be provided to applicants during examination:
None Learning Objective:
?R?
Question Source:
Bank #
INPO - 20664 Point Beach 2/02 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Point Beach 2/02 Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 X
55.43 Comments:
AK3. Knowledge of the reasons for the following responses as they apply to the Dropped Control Rod:(CFR 41.5,41.10 145.6 145.13)
AK3.05 Tech Spec Limit s For Load Reduction to 50% power if flux cannot be brought back within the specified target band.
Examination Outline Cross-reference:
1 2
KIA #
000028AA2.02 Importance Rating 3.8 Group #
Proposed Question:
The following conditions exist:
0 Reactor and turbine power are at 40%.
Rod control is in manual All other systems are normally aligned in AUTOMATIC 0
0 If a T-hot RTD were to fail HIGH, assuming NO operator action, the pressurizer level control system will attempt to:
A. lower pressurizer level to 46.6%
B. maintain pressurizer level at its present value C. lower pressurizer level to 23.1 %
D. raise pressurizer level to 51.3%
Proposed Answer:
D. raise pressurizer level to 51.3%
Explanation (Optional):
Thot input failing high drives its channels Tavg high. Tave for Pzr level control is on an auctioneered high Tave input. Tave no load is 547F which corresponds to 23.1% Pzr Level. Tave Failed High is clamped at 547F which corresponds to Pzr Level of 51.3%. Tave at 100% Power at IP3 is 567F (Not Westinghouse Standard) which corresponds to a Pzr Level of 46.6%. A,B,C Distractors fit various possible failure modes, but for this question, D is the correct response Technical Reference@):
ONOP-RPC-1 Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-IXC-10 Obi 5 Bank #
INPO-5330 Salem 1/96 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam Salem 1/96 Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 55.43 x
AA2. Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions:
(CFR: 43.5 / 45.13)
AA2.02 PZR level as a hction of power level or T-ave. including interpretation of malhction
Examination Outline Cross-reference:
1 KIA #
000051 AA2.02 Importance Rating 4.1 Group #
2 Proposed Question:
Given the following conditions:
0 Plant power is at 88% and being reduced because decreasing condenser vacuum has been The hogging jets have been placed in service.
noted on recorder PR-1151.
0 WHICH ONE of the following combinations of condenser vacuum conditions requires an immediate turbine trip?
Condenser 311 Condenser 32/ Condenser 33 (Turbine Exhaust Hood Temperature)
A. 25.6 in. Hg/ 27.5 in. Hg/ 26.5 in. Hg (Turbine Exhaust Hood Temperature-200F)
B. 26.0 in. Hg/ 26.5 in. Hg/ 29.7 in. Hg (Turbine Exhaust Hood Temperature-180F)
C. 27.2 in. Hg/ 29.5 in. Hg/ 29.8 in. Hg (Turbine Exhaust Hood Temperature-245F)
D. 26.2 in. Hg/ 25.7 in. Hg/ 28.2 in. Hg (Turbine Exhaust Hood Temperature-225F)
Proposed Answer:
B. 26.0 in. Hg/ 26.5 in. Hg/ 29.7 in. Hg (Turbine Exhaust Hood Temperature-180F)
Explanation (Optional):
There are 3 ONOP referenced Rx Trip criteria >P-g:Condenser Vacuum ~25.5in Hg, Shell Differential of >3 in Hg, Turbine Exhaust Hood Temp > 250F for 15 Min. Mathmatically B is the only correct answer. The candidate must apply all 3 criteria.
Technical Reference@):
ON OP-C-1 Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
I O CFR Part 55 Content:
Comments:
LIC-ONP-I 0 Obi-3 Bank #
Modified Bank #
New INPO-10747 IP3 7/96 (Added Turbine Exhaust Hood Criteria To Stem and Distractors)
Last NRC Exam IP3 7/96 Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 55.43 X b.5 AA2. Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum:
(CFR 43.5 / 45.13)
AA2.02 Conditions requiring reactor andor turbine trip
Examination Outline Cross-reference:
2 Tier #
Group #
KIA #
000061 AA2.02 Importance Rating 3.2 Proposed Question:
WHICH ONE of the following Area Radiation Monitor (ARM) readings will cause an automatic action to occur?
A. TSC HVAC Duct ARM (R-44A) reading 50 mR/hr
- 6. Control Room ARM (R-I) reading 0.4 mWhr C. VC 80' ARM (R-2) reading 60 mR/hr D. CVCS Tank 31 ARM (R-MA) reading 120 mR/hr Proposed Answer:
A. TSC HVAC Duct ARM (R-44A) reading 50 mR/hr Explanation (Optional):
A. Alarm and Actuation Setpoint-10 mR/hr-Correct Answer
- 6. Alarm and Actuation Setpoint-1 mWhr C. No Auto Action-Alarm Setpoint-50mR/hr (at power), 20mR/hr (S/D)
D. No Auto Action-Alarm Setpoint-lOOmR/hr Technical Reference@):
LIC-RDM-03 Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-RDM-03 Obi. 2 Bank #
INPO-10747 IP3 7/96 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam IP3 7/96 Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 X
55.43 Comments:
AA2. Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms:
(CFR: 43.5 / 45.13)
AA2.02 Normal radiation intensity for each ARM system channel
Examination Outline Cross-reference:
Tier #
Group #
2 WA #
W/E 14 AK1.O1 Importance Rating 3.1 Proposed Question:
The plant has experienced a Large Break LOCA. All reactor protective systems and engineered safety features systems functioned normally. Containment pressure peaked at 45 psig and has been lowering in a linear fashion. At 22 psig decreasing, the containment pressure rapidly reduced to 3 psig in approximately 5 minutes. What caused the change in the rate of containment depressurization?
At 22 psig......
A. Containment spray becomes more effective at removing energy from the containment atmosphere.
B. Containment coolers become more effective at removing energy from the containment atmosphere.
C. Containment leak rate experienced a prompt increase due to one or more failed containment penetration seals.
D. The containment was vented in accordance with FR-Z.l "Response To High Containment Pressure".
Proposed Answer:
C. At 22 psig, containment leak rate experienced a prompt increase due to one or more failed containment penetration seals.
Explanation (Optional):
A. As pressure and containment temperature lowers the rate of energy removal lowers.
B. As pressure and containment temperature lowers the rate of energy removal lowers.
C. Correct Answer-Used as a basis in EAL Guidelines to call containment safety barrier failed.
D. FR-Z. 1 "Response To High Containment Pressure" does not have venting actions.
Technical Reference( s) :
FR-Z. 1 "ResDonse To Hish Containment Pressure" Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-41 Obi 3 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
55.43 Comments:
AK1. Knowledge of the operational implications of the following concepts as they apply to Loss of Containment Integrity:
(CFR 41.8 / 41.10 / 45.3)
AKl.01 Effect of pressure on leak rate
Examination Outline Cross-reference:
Tier #
Group #
2 KIA #
WIE 14 EK2.2 Importance Rating 3.8 Proposed Question:
Following a small break loss of coolant accident inside containment concurrent with a faulted SG inside containment, the following conditions are noted:
b RCS subcooling is indicating 1OF b
RCS pressure is I380 psig All ECCS pumps are operating HHSl pumps are injecting to the RCS Containment spray has automatically actuated, and containment pressure is 25 psig The crew is implementing 2-EOP-FRCE-1, Response to High Containment Pressure b
b b
0 Which one of the following correctly describes why the reactor coolant pumps are tripped under these conditions?
A. To prevent RCP motor winding damage due to exposure to spray water.
B. To prevent subsequent core damage due to pumping a two-phase mixture.
C. To prevent RCP seal damage due to lack of cooling.
D. To prevent RCP bearing damage due to lack of cooling.
Proposed Answer:
D. To prevent RCP bearing damage due to lack of cooling.
Explanation (Optional):
A. Incorrect-RCPs are designed to operate with spray, as long as cooling to pump to present B. Incorrect-RCS Pressure is not below RCP Trip Criteria C. Incorrect-Seal Injection is not lost-Seals will remain intact D. Correct-Phase B Isolates RCP Cooling for bearings Technical Reference@):
FR-Z. 1 'I Resoonse To Hiah Containment Pressure" Proposed references to be provided to applicants during examination:
None Learning Objective:
????
Question Source:
Bank #
INPO-17212 (Salem 1 1/98)
Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Salem 1 1/98 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis I O CFR Part 55 Content:
55.41 X
55.43 Comments:
Em.
Knowledge of the interrelations between the(High Containment Pressure) and the following: (CFR: 41.7/45.7)
EK2.2 Facility *s heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
Examination Outline Cross-reference:
1 Group #
2 KIA #
W/E6 EK1.3 Importance Rating 3.6 Proposed Question:
The following conditions exist at Indian Point 3 after a Reactor Trip and Safety Injection:
0 All Control Rods Are Inserted a
Cooldown is in Progress on Recirculation Cooling 0
RCS Subcooling is OF 0
RCPs are NOT running 0
Highest Qualified CET Temperature is 685F RVLIS Full Range Level is 50%
The following alarms exist on Panel SAF- "High Tave", "Pressurizer Low Pressure", and 0
0 "Pressurizer Low Level" What is the current status of the Core Cooling Critical Safety Function (CSF), and the core cooling strategy based on the Functional Recovery Procedure that may be used?
A. Core Cooling CSF is Green, continue with cooldown on Recirculation Cooling.
B. Core Cooling CSF is Yellow, FR-C.3 "Response To Saturated Core Cooling" may be implemented
, and actions are taken to raise subcooling margin.
C. Core Cooling CSF is Orange, FR-C.2 I' Response To Degraded Core Cooling" is implemented, and take actions to raise water level in the reactor vessel.
D. Core Cooling CSF is Red, FR-C.l 'I Response To Inadequate Core Cooling" is implemented, and reactor coolant pumps are operated to improve core cooling.
Proposed Answer:
B. Core Cooling CSF is Yellow, FR-C.3 "Response To Saturated Core Cooling" may be implemented
, and actions are taken to raise subcooling margin.
Explanation (Optional):
Conditions indicate that a Core Cooling Yellow Path exists, this makes A,C,D wrong.
Technical Reference@):
FR-C.1,FR-C.2,FR-C.3 Proposed references to be provided to applicants during examination:
CFSTs Learning Objective:
LIC-EOP 38 Obi2.3.4 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam NJA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
Comments:
EK1 Knowledge of Operational Implications of the following concepts as they apply to Saturated Core Cooling (CFR 41.8,41.1,45.3)
EKl.3 Annunciators and conditions indicating signals and remedial actions 55.43
Examination Outline Cross-reference:
1 2
WA #
W/E6 EA2.2 Importance Rating
- 4. I Group #
Proposed Question:
When performing actions in accordance with FR-C.2 'I Response To Degraded Core Cooling", the operator is directed to verify SI valve alignment per attachment "Injection Phase SI Valve Line Up" with the SI system in the Cold Leg Injection Mode. Which of the following valves should be CLOSED per this system alignment?
A. BIT Outlet Valve (1835 A,B)
B. CCW to RHR Heat Exchanger (822 A,B)
C. RWST To Safety Injection Pumps (1 81 0)
D. RHR Discharge To Containment Building Spray Header (889 A,B)
Proposed Answer:
D. RHR Discharge To Containment Building Spray Header (889 A,B)
Explanation (Optional):
A. Cold Leg Injection Injects Through The BIT B. CCW is Aligned to the RHR Heat Exchanger C. RWST is aligned as the suction source to the SI Pumps D. Should be closed during injection phase, but may be opened as a part of swapping to Cold Leg Recirc.
Technical Reference@):
FR-C.2 " ResDonse To Decrraded Core Coolincl" Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-38 Obi 15 Question Source:
Bank #
INPO 579 Beaver Vallev 3/97 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Beaver Vallev 3/97 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 x
Comments:
EA2 Ability to determine and interpret the following as they apply to degraded core cooling (CFR 43.5,45.13)
EA 2.2 Adherence to appropriate procedures and operation within limitations in the facilities license and amendments
Examination Outline Cross-reference:
1 Group #
2 WA #
000076AK3.05 Importance Rating 3.8 Proposed Question:
Indian Point 3 experienced a reactor trip from 100% power due to a main transformer protective relaying fault. Operators have transitioned from E-0 "Reactor Trip or Safety Injection" to ES-0.1 "Reactor Trip Response". Valid alarms are received on R63A and B (GFF Detectors) and annunciator "R63NB GFFDI is in alarm. Charging pump area radiation monitors show a rising trend, but have yet to alarm. The CRS implements ONOP RCS-4 I' High Radioactivity In Reactor Coolant System". Which statement describes the correct actions to take with the CVCS system and the basis for those actions?
A. Letdown is diverted to the CVCS HUT to limit radiation levels in the charging pump area.
B. Charging and Letdown are maximized through CVCS ion exchangers to maximize clean up.
C. Charging and Letdown are isolated to contain the high radioactivity within the containment building.
D. Excess Letdown is placed in service through CVCS ion exchangers to maximize clean up.
Proposed Answer:
B. Charging and Letdown are maximized through CVCS ion exchangers to maximize clean up.
Explanation (Optional):
Per ONOP-RCS-4 Charging and Letdown are maximized to clean up RCS. Excess letdown goes though seal water heat exchanger directly to VCT, and can not be directed through ion exchangers.
Technical Reference(s):
ONOP-RCS-4 "Hiah Radioactivitv In Reactor Coolant Svstem" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-ONP-30 Obi 2 Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam NiA Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X
AK3. Knowledge of the reasons for the following responses as they apply to theHigh Reactor Coolant Activity : (CFR 41.5,41.10 /45.6/45.13)
AK3.05 Corrective actions as a result of high fission product radioactivity level in the RCS
Examination Outline Cross-reference:
1 Group #
2 K/A #
W/EOI EA1.3 Importance Rating 3.8 Proposed Question:
The operating crew is responding to a Main Steam Line Break on 31 Steam Generator inside containment. The following conditions presently exist:
The operating crew has transitioned from E-2, "FAULTED STEAM GENERATOR ISOLATION", to E-I, "LOSS OF REACTOR OR SECONDARY COOLANT".
0 31 S/G is dry with pressure 4 0 0 psig.
32 S/G narrow range level is 20% and rising, with pressure stable at 950 psig.
33 S/G narrow range level is 5% and rising, with pressure stable at 950 psig.
Pressurizer Level is 15% and rising.
Containment Pressure is 18 psia and lowering.
Containment Temperature is 240 degrees F and lowering.
RCS Subcooling indicates stable at 80 degrees F.
RCS Pressure is 2150 psig and rising slowly.
0 0
34 S/G narrow range level is 10% and rising, with pressure stable at 950 psig.
0 0
0 0
0 Which ONE of the following conditions will complete the transition criteria from E-I, "LOSS OF REACTOR OR SECONDARY COOLANT to ES-I.I, "SI TERMINATION"?
A. 33 S/G level rises to 14%.
B. Pressurizer level rises to 32%.
C. RCS Subcooling rises to 112 degrees F D. RCS pressure rises to 2215 psig.
Proposed Answer:
B. Pressurizer level rises to 32%.
Explanation (Optional):
A. S/G level criteria has already been met by 32 S/G >14%
B. Pressurizer level must rise to >32% for adverse containment conditions C. RCS subcooling criteria has already been met, it is greater than 63 degrees F.
D. RCS Pressure criteria has already been met, it is greater than 2000 psig.
Technical Reference(s):
E-I. "LOSS OF REACTOR OR SECONDARY COOLANT", ES1.l "SI Termination" Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-EOP-33 Obi V Question Source:
Bank #
INPO-15387 Surrv 1 4/8/99 None Modified Bank #
New Memory or Fundamental Knowledge (Note changes or attach parent)
Question History:
Last NRC Exam Surrv 1 4/8/99 Question Cognitive Level:
Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
Comments:
EA1. Ability to operate and / or monitor the following as they apply to the (Reactor Trip or Safety Inj ectiodRediagnosis)
(CFR 41.7 /45.5,45.6)
EAI.3 Desired operating results during abnormal and emergency situations.
55.43
Examination Outline Cross-reference:
2 Tier #
Group #
Importance Rating 3.3 KIA #
W/E 01 2.3.10 Proposed Question:
Indian Point 3 has experienced a Reactor Trip and Safety Injection due to a LOCA. Conditions have been met for SI Termination in accordance with ES 1.I,
'I SI Termination". An operator must enter the RHR pump room to prepare for Shutdown Cooling. Radiation levels in the RHR pump room are 2 Rem/hr due to fuel failures that were realized during the event. The operator is predicted to exceed radiation exposure levels specified in plant procedures. Which one of the following identifies the individual who must give final authorization for radiation exposures greater than N+l Lifetime total Effective Dose Equivalent (TEDE)? (where N = the individual's age)?
A. The Individual's Supervisor B. The Radiation and Environmental Services Manager C. The Vice President Nuclear Operations D. The Plant Manager Proposed Answer:
D. The Plant Manager Explanation (Optional):
The answer is appropriate per Lesson Plan LIC-RAD-5 Technical Reference(@:
AP-7 Radiation Protection Plan Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-RAD-5 Obi 5.1.5 Question Source:
Bank #
INPO-10560 IP-3 7/96 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam IP-3 7/96 Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis I O CFR Part 55 Content:
55.41 55.43 x
Comments:
2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
(CFR: 43.4 / 45.10)
Examination Outline Cross-reference:
Tier #
Group #
2 K/A #
W/E 15 2.4.21 Importance Rating 4.3 Proposed Question:
FR-Z.2 "Response to Containment Flooding" Step #I states: "Try to identify unexpected source of water to sump" This is based on a water level greater than the design basis flood level as provided by water from the RWST, RCS and what other sources?
A. Condensate Storage Tank and Safety Injection Accumulators B. Component Cooling Water and Safety Injection Accumulators C. Primary Water Storage Tank and Component Cooling Water D. Primary Water Storage and Service Water Proposed Answer:
A. Condensate Storage Tank and Safety Injection Accumulators Explanation (Optional):
There are 4 credited sources of water to enter the containment in the design analysis. They are:
Reactor Coolant System, Refueling Water Storage Tank, Condensate Storage Tank and Safety Injection Accumulators. Other distractors are chosen from sources that are checked in FR-Z.2. Only answer A8 is correct.
Technical Reference(s):
FR-Z.2 "ResDonse to Containment Floodinu" Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-41 Question Source:
Bank #
INPO-19553 Cook 1 5/2001 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Cook I 5/2001 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 x
Comments:
2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions including:
- 1. Reactivity control
- 2. Core cooling and heat removal
- 3. Reactor coolant system integrity
- 4. Containment conditions
- 5. Radioactivity release control.
(CFR: 43.5 / 45.12)
Examination Outline Cross-reference:
2 Tier #
Group #
KIA #
W/E 16 EK 1.2 Importance Rating 3.2 Proposed Question:
Given the following conditions for Indian Point 3:
A LOCA has occurred.
and Depressurization".
The crew is performing cooldown as directed by ES-1.2 'I Post LOCA Cooldown HHSl Pumps are still operating in injection phase.
The Containment Iodine Filter Fans are not running.
Two Containment FCUs are running.
Containment pressure is stable at 2.2 psig.
The CRS transitions to FR-Z.3 "Response to High Containment Radiation Level" in response to a YELLOW path condition.
In addition to running Containment Iodine Filter Fans, what action does the CRS direct in FR-Z.3 "Response to High Containment Radiation Level" in order to help reduce containment radiation levels?
A. Containment Post Accident Ventilation is initiated B. An RHR Pump is aligned to supply the associated Containment Spray Header C. Initiate a Phase B Containment Isolation Signal D. The Idle Containment FCUs are started Proposed Answer:
D. The Idle Containment FCUs are started Explanation (Optional):
There are only a few actions in the FR-Z.3 strategy: Verify Containment Ventilation Isolation, Run Containment Iodine Filter Fans, Start All Containment FCUs Tec hn ica I Reference( s) :
Proposed references to be provided to applicants during examination:
None None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-EOP-41 Obi 12 Bank #
INPO-19494 Kewaunee 12/2000 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam Kewaunee 12/2000 Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 X
55.43
EK1. Knowledge of the operational implications of the following concepts as they apply to the Wigh Containment Radiation)
(CFR 41.8 /41.10,45.3)
EKl.2 Normal, abnormal and emergency operating procedures associated with (High containment Radiation).
Examination Outline Cross-reference:
1 Group #
2 WA #
W/E09 E A 1 2 Importance Rating 3.9 Proposed Question:
An early step in ES-0.2, "Natural Circulation Cooldown" requires the crew to verify that all CRDM fans are running. WHICH ONE of the following is the basis for this step?
A. To prevent overheating damage to the RPI coils and CRDM motor windings.
B.To remove heat from the reactor vessel head to prevent voiding in the vessel during depressurization.
C. To provide cooling for the core exit thermocouple extension leads to ensure reliable temperature indication.
D. To eliminate PTS concerns in the reactor vessel head during the RCS cooldown.
Proposed Answer:
B.To remove heat from the reactor vessel head to prevent voiding in the vessel during depressurization.
Explanation (Optional):
Answer A-CRDM Windings are not energized, RPI coils no hotter than normal, and cooling down Answer B-Correct per procedure Answer C-CETs do not require cooling for temperature compensation Answer D-Head is not limiting component for PTS Technical Reference( s) :
ES-0.2. "Natural Circulation Cooldown" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
???
Bank #
INPO-10652 IP3 7/96 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam IP3 7/96 Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X
EA1. Ability to operate and I or monitor the following as they apply to the (Natural Circulation Cooldown) (CFR:
41.7 145.5 145.6)
EA 1.2 Operating behavior characteristics of the facility.
Examination Outline Cross-reference:
2 Group #
1 WA #
003 A3.03 Importance Rating 2.6 Proposed Question:
The plant is in Mode 1 at 100% power with the folllowing indications and alarms:
"RCP No.1 Seal Low D/P" annunciator in ALARM "RCP No.1 Seal Return High/Low Flow" annunciator in ALARM 32 RCP No. I Seal Differential Pressure..... l50psid 32 RCP Seal Inlet Temperature..... l50F 32 RCP Seal Outlet Temperature..... l80F 32 RCP Seal Injection Flow..... 6 gpm 32 RCP Standpipe Level Off Normal Light..... Not Lit 32 RCP Seal Return Recorder Indication..... 6 gpm What is the status of the 32 RCP Seal?
A. No.1 and No.2 RCP Seals have failed and full pressure drop is across No.3 RCP Seal B. No.2 Seal has failed and full pressure drop is across No.1 and No.3 RCP Seals C. No.1 Seal has failed and full pressure drop is across No.2 and No.3 RCP Seals D. No.3 Seal has failed and full pressure drop is across No.1 and No.2 RCP Seals Proposed Answer:
C. No.? Seal has failed and full pressure drop is across No.2 and No.3 RCP Seals Explanation (Optional):
A. Incorrect-Would require No. 1 Seal D/P to be >150psid and Standpipe in alarm B. Incorrect-Would require No I Seal D/P to be >150psid C. Correct-No. 1 Seal D/P Low and Leakoff flow high indicates a No.1 Seal Failure D. Incorrect-Would require Standpipe level low.
Technical Reference(s):
ONOP-RCS-5. SOP-RCS-1 Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-NSS-3 Obi 2 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content:
55.41 X
55.43 X
Comments:
A3 Ability to monitor automatic operation of the RCPS, including: (CFR 41.7 / 45.5)
A3.03 Seal DR
Examination Outline Cross-reference:
2 WA #
004 K2.02 Importance Rating 3.1 Group ##
1 Proposed Question:
What are the power supplies for 31 and 32 Boric Acid Transfer Pumps?
A. 31-MCC 36B, 32-MCC 36A B. 31-MCC 36C, 32-MCC 36D C. 31-MCC 36A, 32-MCC 36B D. 31-MCC 36D, 32-MCC 36C Proposed Answer:
C. 31-MCC 36A, 32-MCC 36B Explanation (Optional):
Answers A,B,D include wrong MCCs for the listed loads, All MCCs listed are in control building Technical Reference@):
480V Distribution System Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-PSA-01 Obi 4 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 X
55.43 Comments:
K2 Knowledge of bus power supplies to the following:
(CFR: 41.7)
K2.02 Makeup pumps
Examination Outline Cross-reference:
2 I K/A #
004 K3.04 Importance Rating 3.9 Group #
Proposed Question:
Indian Point 3 has just experienced a loss of all CVCS charging capability, while at power. The operating crew isolated CVCS letdown flow and has entered the appropriate Technical Specification action statements. What, if any, is the affect on the operation of Reactor Coolant Pumps (RCPs)?
A. The RCPs must be immediately tripped to prevent damage to the RCP seal packages from uncontrolled heatup.
B. RCP operation is allowed while restoring RCP Seal Injection, since Thermal Barrier Cooling is in service.
C. The RCPs must be immediately tripped to prevent damage to the RCP seal packages via contaminants from the RCS backflowing into the RCP Seals.
D. RCP operation is allowed to continue while restoring RCP Seal Injection even if Thermal Barrier Cooling is NOT in service.
Proposed Answer:
B. RCP operation is allowed while restoring RCP Seal Injection, since Thermal Barrier Cooling is in service.
Explanation (Optional):
Answer A is wrong, the RCPs do not immediately have to be tripped.
Answer B is correct per ONOP-RCS-5 Answer C is wrong, the RCPs do not immediately have to be tripped, however the reason is mentioned in the lesson plan as a negative outcome Answer D is wrong, RCPs must be immediately tripped in this condition Technical Reference@):
ONOP-RCS -5 "RCP Malfunctions" Proposed references to be provided to applicants during examination:
None Lea rni ng 0 bject ive:
LIC-ONP-31 Obi 2 Question Source:
Bank #
Modified Bank #
New X
Question History:
Last NRC Exam (Note changes or attach parent)
X Question Cognitive Level:
10 CFR Part 55 Content:
55.41 X
Comments:
K3 Knowledge of the effect that a loss or malfunction of the CVCS will have on the following:
(CFR: 41.7/45/6)
K3.04 RCPS Memory or Fundamental Knowledge Comprehension or Analysis 55.43
Examination Outline Cross-reference:
2 Group ##
1 WA #
005A4.0 I Importance Rating 3.4 Proposed Question:
Given the following:
0 The Unit has tripped from 100% power due to a LOCA with a loss of offsite power.
The Safety Injection (SI) Signal has been reset IAW E-I "Loss of Reactor or Secondary RCS pressure is 225 psig.
Emergency Diesel Generator 31 has started and successfully loaded Emergency Diesel Generator 32 has failed to start.
No flow is indicated for RHR pump 31 0
Coolant
0 0
RWST Level is 15ft.
e 0
0 WHICH ONE (I) of the following is the reason that RHR pump 31 has no flow indication?
A. RCS pressure is above RHR pump shutoff head.
B. 480 V bus 3A has no power C. SI has been reset and SI termination requirements have been met, and actions taken to reduce SI flow D. RWST Suction Valve (MOV-882) has been closed per ES-1.3 "Transfer To Cold Leg Recirculation" Proposed Answer:
A. RCS pressure is above RHR pump shutoff head Explanation (Optional):
A. Correct-Shutoff Head 160psid B. Incorrect-31 RHR Pump is powered from 31 EDG via 480V bus 3A C. Incorrect-SI Termination Requirements are not met-RCS Pressure4 65Opsig D. Incorrect-Requirements for Cold Leg Recirc not met-RWST Level> 11.5 feet Technical Reference@):
ES-1.3 "Transfer To Cold Lea Recirculation", E-I "Loss of Reactor or Secondarv Coolant
Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-PSA-3 Obi 2 None Question Source:
Bank #
Modified Bank #
INPO-3033 2/96 Watts Bar(Changed Distractor 'ID'. and SI Injection Signal and Electrical Conditions in Stem))
New Question History:
Last NRC Exam 2/96 Watts Bar Question Cognitive Level:
10 CFR Part 55 Content:
55.41 X
Comments:
A4 Ability to manually operate and/or monitor in the control room: (CFR 41.7 / 45.5 to 45.8)
A4.01 Controls and Indication for RHR Pumps Memory or Fundamental Knowledge Comprehension or Analysis X
55.43
Examination Outline Cross-reference:
1 Tier #
Group #
KIA #
005 2.2.5 Importance Rating 2.7 Proposed Question:
It has been determined that due to a problem with a temperature probe on the outlet of 31 Residual Heat Removal Heat Exchanger, a Temporary Modification is required to ensure that RHR HX Temperature Recorder TR-636 receives a valid input from 31 RHR Heat Exchanger outlet temperature.
As you review the package that is to be installed on your shift, you note the following:
0 The plant is at 100% power, 60 days before a refueling outage.
A strap on thermocouple will be taking the place of the installed RTD.
The instrumentation will be powered from a supply in the vicinity of 31 RHR Heat Exchanger that The cart that is holding the equipment has been seimically restrained.
The Modification package has a removal date of 120 days A Temp Alt Tag has been prepared for hanging on the required equipment.
The electrical connection at the recorder will be accomplished by lifting and taping the leads 0
0 is safety related and from that train.
s s
s 0
from the RTD and landing similar leads from the Thermocouple output.
Which statement is correct concerning this plant modification?
A. This modification meets the requirements as a Temporary Alteration and should be approved and installed.
B. The Temporary Alteration procedure is NOT required in this instance and this package should not be approved for installation.
C. The proposed installation period is too long, the package should not be approved since it will be installed for 120 days or more.
D. The proposed Temporary Alteration should not be installed because taping of lifted leads is not authorized in the RHR Heat Exchanger room.
Proposed Answer:
C. The proposed installation period is too long, the package should not be approved since it will be installed for 120 days or more.
Explanation (Optional):
Technical Reference(@:
Proposed references to be provided to applicants during examination:
Learning Objective:
???????????
Question Source:
Bank #
ENN-DC-I 36 "TemDorarv Alterations" None Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
Comments:
55.41 55.43 x
2.2.5 Knowledge of the process for making changes in the facility as described in the safety analysis report.
(CFR: 43.3 / 45.13)
Examination Outline Cross-reference:
2 Group #
1 WA #
006 A3.01 Importance Rating 4.3 Proposed Question:
The following plant conditions exist:
Plant is in Mode 4 a
RCS Press-375 psig RCS Temp-21 0 degrees F Maintenance is being performed on the containment pressure detectors when an inadvertent SI signal is received by the ECCS circuitry. Which of the following describes the response of the ECCS Accumulators to the inadvertent SI under the given plant conditions?
The accumulators will:
A. discharge into the RCS because the accumulator outlet MOVs (valves 894 A, B. C, & D) will open on the SI signal.
B. discharge into the RCS because the accumulator outlet MOVs (valves 894 A. B, C. & D) are open with their power supply locked out.
C. not discharge into the RCS because the accumulator outlet MOVs (valves 894 A. B. C, & D) are closed with their power supply locked out.
D. not discharge into the RCS because accumulator pressure is lower than RCS pressure.
Proposed Answer:
C. not discharge into the RCS because the accumulator outlet MOVs (valves 894 A. B. C, & D) are closed with their power supply locked out.
Explanation (Optional):
A. Incorrect-No Discharge Into RCS B. Incorrect-No Discharge Into RCS C. Correct-In Mode 4 Accumulators are locked out D. Incorrect-Wrong reason, and accumulators are not required to be operable.
Technical Reference(s):
POP 3.3 I' Plant Cooldown-Hot to Cold Shutdown" Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-POP 4 Obi 2 Question Source:
Bank #
INPO-10479 IP3 4/96 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam IP3 4/96 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
55.43 Comments:
A3 Ability to monitor automatic operation of the ECCS, including: (CFR 41.7 / 45.5)
A3.01 Accumulators
Examination Outline Cross-reference:
2 Group #
1 K/A #
007 K1.03 Importance Rating 3.2 Proposed Question:
Given the following conditions:
Pressurizer Relief Tank (PRT) level is 68%.
PRT pressure is 3 psig.
RCS pressure is 2255 psig.
One pressurizer PORV is leaking by the seat.
a WHICH ONE of the following describes the expected downstream tailpipe condition?
A. Saturated steam-water mixture at 21 3F.
B. Superheated steam at 313F.
C. Superheated steam at 635F.
D. Saturated steam-water mixture at 222 deg F.
Proposed Answer:
D. Saturated steam-water mixture at 222 deg F Explanation (Optional):
A. Incorrect-Wrong place on Mollier Diagram B. Incorrect-Not Superheated C. Incorrect-Not Superheated D. Correct-Approx Enthalpy I 1 15 BTU/lbm, constant enthalpy process Technical Reference(s):
Mollier Diaaram Proposed references to be provided to applicants during examination:
Mollier Diaaram Learning Objective:
LIC-ONP-51 Obi 1 Question Source:
Bank #
INPO-10712 and 20583 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam IP3 7/96 and Point Beach 1 2/02 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
I O CFR Part 55 Content:
55.41 X
55.43 Comments:
K1 Knowledge of the physical connections and/or cause effect relationships between the PRTS and the following systems:
(CFR 41.2 to 41.9 / 45.7 to 45.8)
K1.03 RCS
Examination Outline Cross-reference:
Tier #
Group #
1 Importance Rating 3.3 WA #
008K4.01 Proposed Question:
The following plant conditions exist:
The plant is in Mode I at 80% power 31 and 33 CCW pumps are in service 32 CCW pump is in standby A leak has developed on the discharge of 31 CCW pump What is the highest CCW header pressure for which the standby CCW pump will automatically start?
A. 11 0 psig B. 100 psig C. 90 psig D. 80 psig Proposed Answer:
B. 100 psig Explanation (Optional):
The standby CCW pump starts at 100 psig lowering Technical Reference(s):
SOP-CC-001 B Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-PSA-2 Obi 4 Bank ##
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X
K4 Knowledge of CCWS design feature@) andfor interlock(s) which provide for the following:
(CFR 4 1.7)
K4.0 1 Automatic Start of Standby Pump
Examination Outline Cross-reference:
2 1
KIA #
010K2.02 Importance Rating 2.7 Group ##
Proposed Question:
The following plant conditions exist:
0 The plant is in Mode 1 at 75% power A Loss of 32 Instrument Bus (Channel I
, Red) has just occurred ONOP-EL3 "Loss of an Instrument Bus" has been implemented The Pressurizer Master Pressure Controller was taken to Manual per the Initial Operator Actions 0
0 0
When can the Pressurizer Master Pressure Controller be returned to Automatic?
A. When 32 Instrument Bus is restored, since power was lost to the Pressurizer Master Pressure Controller.
B. As a part of contingency actions of the procedure, since Instrument power was not lost to the Pressurizer Master Pressure Controller.
C. When 32 Instrument Bus is restored, since power was lost to the Pressurizer Spray Valve Controllers.
D. When 32 Instrument Bus is restored, since power was lost to the Pressurizer Heater Controller.
Proposed Answer:
B. As a part of contingency actions of the procedure, since Instrument power was not lost to the Pressurizer Master Pressure Controller.
Explanation (Optional):
Power is not lost to any Pressurizer controls for the loss of 32 Instrument Bus Technical Reference(@:
ONOP-EL3 "Loss of an Instrument Bus Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
I O CFR Part 55 Content:
LIC-ONP-2 Obi 3 Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 Comments:
K2 Knowledge of bus power supplies to the following:
(CFR 41.7)
K2.02 Controller for PZR spray valve X
Examination Outline Cross-reference:
2 KIA #
0 I OA3.02 Importance Rating 3.5 Group #
1 Proposed Question:
Given the following conditions:
0 Reactor is at 100% power.
Pressurizer Pressure Channel Ill is the controlling channel.
Pressurizer Pressure controls are in AUTO.
All Pressurizer Pressure channels are operable.
0 0
0 All Pressurizer control components are operable.
WHICH ONE of the following describes the final result if Channel 111 pressurizer pressure detector (PT-457) fails LOW under these conditions, and NO operator action is taken?
A. PORV PCV-456 will maintain RCS pressure 90 to 100 psig above normal.
B. No effect on RCS pressure because the Master Pressure Controllerwill select the non-failed channel.
C. Backup heaters will maintain RCS pressure 35 to 50 psig below normal.
D. A High Pressurizer Pressure Reactor Trip will occur, since pressure rises and PORVs will not open.
Proposed Answer:
A. PORV PCV-456 will maintain RCS pressure 90 to 100 psig above normal.
Explanation (Optional):
A. Correct-PCV-456 is independent from the Master Pressure Controller and will maintain pressure 2335-2325 psig.
B. Incorrect-Sray Valve actuation is controlled by the Pressurizer Pressure Controller which thinks pressure is low due to the failure C. Incorrect-Backup heaters will be on, however actual plant pressure will continue to rise above the control band D. Incorrect-PCV-456 will operate to prevent a High Pressure Reactor Trip Technical Reference(s):
LIC-IXC-11 Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-IXC-I1 Obi 5 Question Source:
Bank #
INPO 10645 IP3 7/96 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam IP3 7/96 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
55.43 Comments:
A3 Ability to monitor automatic operation of the PZR PCS, including: (CFR: 41.7 / 45.5)
A3.02 Pzr Pressure
Examination Outline Cross-reference:
2 WA #
01 2K1. O l Group #
1 Importance Rating 3.7 Proposed Question:
During a startup, with the reactor critical at 1 E-8 amps. the breaker from the No. 31 static inverter supply to instrument bus No. 31 trips open. The transfer to the back-up power supply fails, and power is lost to the 120 VAC instrument bus No. 31.
The result of the above will be:
A. A reactor trip due to the de-energization of Intermediate Range channel N35.
B. A source range high flux reactor trip due to the de-energization of permissive P-6.
C. A reactor trip due to the de-energization of Intermediate Range channel N36.
D. The de-energization of several nuclear instrument channels, with no effect on nuclear power.
Proposed Answer:
C. A reactor trip due to the de-energization of Intermediate Range channel N36.
Explanation (Optional):
A. Incorrect-N-35 does not de-energize B. Incorrect-P-6 does not de-energize C. Correct-in 1 of 2 logic while in IR, loss of power causes trip RPS channel D. Incorrect-Trip does occur Technical Reference@):
ONOP-EL3 "Loss of an Instrument Bus Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-IXC-I 3 Obi 4 Bank #
INPO-10499 IP3 4/96 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam IP3 4/96 Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 X
55.43 Comments:
K1 Knowledge of the physical connections and/or cause effect relationships between the RPS and the following systems:
(CFR 41.2 to41.9/45.7 to45.8)
K1.O1 120V vital/hstrument power system
Examination Outline Cross-reference:
2 KIA #
012 K4.07 Importance Rating 3.2 Group #
1 Proposed Question:
The plant is operating at 70% power when the breaker supplying 32 RCP trips. Which statement below correctly describes the condition of the "Reactor Trip First Out Alarm Panel" Panel FDF?
A. "Loss of Flow Single Loop" annunciator will be flashing, and will go solid when acknowledged.
B. I' Reactor Trip Breaker Open" annunciator will be flashing, and will go solid when acknowledged.
C. "Loss of Flow Single Loop" annunciator will be solid, and will begin flashing when acknowledged.
D. I' Reactor Trip Breaker Open" annunciator will be solid, and will begin flashing when acknowledged.
Proposed Answer:
A. "Loss of Flow Single Loop" annunciator will be flashing, and will go solid when acknowledged.
Explanation (Optional):
A. Correct-This condition will be sensed first-First Out flashes until acknowledged, then goes solid B. Incorrect-This condition will not be the "first out" C. Incorrect-First Out flashes until acknowledged, then goes solid D. Incorrect-This is not "first out", and would be shown as solid and stays solid Technical Reference@):
ARP-002 "Panel FDF-Reactor Trip First Out" Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-IXC-13 Obi 1 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
55.43 Comments:
K4 Knowledge of RPS design feature@) and/or interlock@)
which provide for the following:
(CFR 4 1.7)
K4.07 First-out indication
Examination Outline Cross-reference:
2 I
WA #
01 3K5.01 Importance Rating 3.2 Group #
Proposed Question:
Given the following:
e An automatic safety injection occurred following a steam break outside containment.
The steam break has been isolated.
RCS pressure has been restored to 2200 psig.
Main steamline pressures have been restored to 1000 psig.
ESF Busses are all energized.
Reactor Trip Breaker "Ag opened by SI actuation.
Reactor Trip Breaker " B I is closed, never opened.
Both trains of Safety Injection have been RESET.
SI flow has been terminated.
e e
e e
e e
e e
WHICH ONE (1) of the following describes the response of the SI and RHR pumps when RCS pressure rapidly decreases to 1500 psig? (ASSUME no operator action is taken.)
A. ECCS pumps will NOT start in either Train "A" or Train I ' B I.
B. ECCS pumps will start in both Train "A" and Train "B".
C. Train Af ECCS pumps will start; Train "B" ECCS pumps will NOT start.
D. Train " B ECCS pumps will start; Train " A ECCS pumps will NOT start.
Proposed Answer:
B. ECCS pumps will start in both Train "A" and Train "B.
Explanation (Optional):
Safety Injection will automatically reinitiate, because the "B'l RTB is still shut, arming " B Safety Injection initiation. "B" Automatic SI Initiates and picks up both trains of SI.
Technical Reference@):
Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-ESS-1 Obi 1 Question Source:
Bank #
INPO-5365 Salem 1 7/96 Westinclhouse Technoloqv Systems Manual Modified Bank #
New Question History:
Last NRC Exam Salem 1 7/96 Question Cognitive Level:
Memory or Fundamental Knowledge I O CFR Part 55 Content:
55.41 55.43 x
Comments:
(Note changes or attach parent)
Comprehension or Analysis X
K5 Knowledge of the operational implications of the following concepts as they apply to the ESFAS: (CFR: 41.5 /
K5.0 1 Defmitions of safety train and ESF channel 45.7)
Examination Outline Cross-reference:
2 1
WA #
022 K2.01 Importance Rating 3.1 Group #
Proposed Question:
The plant has experienced a Design Basis Loss of Coolant Accident. Containment Cooling has initiated, and both Containment Spray Pumps and all Fan Cooler Units have started and are running.
An electrical fault causes the loss of 480V Bus 5A. How does this impact the design cooling capability of the Containment Cooling Systems?
Containment Cooling......
A. IS NOT adequate since 1 Containment Spray Pump and 3 Fan Cooling Units are in operation.
B. IS NOT adequate since 1 Containment Spray Pump and 4 Fan Cooling Units are in operation.
C. IS adequate since BOTH Containment Spray Pumps and 4 Fan Cooling Units are in operation.
D. IS adequate since 1 Containment Spray Pump and 3 Fan Cooling Units are in operation.
Proposed Answer:
D. IS adequate since I Containment Spray Pump and 3 Fan Cooling Units are in operation.
Explanation (Optional):
A. Incorrect-This configuration is adequate for containment cooling by design B. Incorrect-Only 3 FCUs are operating C. Incorrect-Only 1 CS pump is operating D. Correct-With bus 5A 00s only 1 CS Pump and 3 FCUs are operating, and this is adequate by design Technical Reference(s):
Westinahouse Technoloav Course, and LIC-ESS-3 and 4 Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-ESS-4 Obi 3 Question Source:
Bank #
None Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge 10 CFR Part 55 Content:
55.41 X
Comments:
K2 Knowledge of power supplies to the following:
(CFR: 41.7)
K2.01 Containment cooling fans (Note changes or attach parent)
Comprehension or Analysis X
55.43
Examination Outline Cross-reference:
1 Tier #
Group #
WA #
026K5.01 Importance Rating 2.9 Proposed Question:
Given the following plant conditions:
The plant was at 100% power 0
A design basis LOCA occurred The operator notes, several minutes after Containment Spray initiation that Spray additive tank 0
flow indicates "O", and that Containment Spray Additive Tank Level is not lowering.
Which ONE (1) of the following describes the effect of this failure?
A. Radioactive Iodine will NOT be as effectively removed and held in solution since there will be a lack of elemental Sodium (Na) in the Containment Spray Solution.
B. Radioactive Cesium will be effectively removed and the lower pH from the lack of NaOH in the Containment Spray Solution has little effect.
C. Radioactive Iodine will NOT be as effectively removed and held in solution since spray solution pH will be too low.
D. Radioactive Iodine will be effectively removed and the lower pH from lack of NaOH in the Containment Spray Solution has no effect.
Proposed Answer:
C. Radioactive iodine will NOT be as effectively removed and held in solution since spray solution pH will be too low.
Explanation (Optional):
A. Incorrect-Na has no effect on removal of Iodine B. Incorrect-Containment Spray is not credited for Cesium removal C. Correct-Removal of Iodine is ph dependent D. Incorrect-Iodine removal is enhanced by presence of NaOH in solution to lower pH.
Technical Reference@):
Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-ESS-3 Obi 3 Question Source:
Bank ##
Westinclhouse Course Manual and LIC-ESS- 03 None Modified Bank #
INPO-5185 Robinson-8/96 (Changed Stem for New Memory or Fundamental Knowledge spray add delay, and changed all distractors Question History:
Last NRC Exam Robinson 8/96 Question Cognitive Level:
10 CFR Part 55 Content:
55.41 X
Comments:
Comprehension or Analysis X
55.43 K4 Knowledge of CSS Design Features and/or Interlocks which provide for the following:
(CFR 41.7) K4.06 Iodine Scavenging via the CS System
Examination Outline Cross-reference:
1 Tier #
Group ##
WA #
026A2.08 Importance Rating 3.7 Proposed Question:
Given the following:
The crew has just entered EOP E-I, "Loss of Reactor or Secondary Coolant." Containment Spray is currently reducing containment pressure. Containment readings are as follows:
e Containment Temperature-230 degrees F and lowering Containment Pressure Chart Recorder-Pressure peaked at 24 psig. Currently pressure is 19 psig and lowering 0
Containment Spray Pumps-Both Operating Normally Fan Cooling Units-3 In Service 0
WHICH ONE (1) of the following is the HIGHEST pressure below which containment spray may be secured per EOP E-I, "Loss of Reactor or Secondary Coolant"?
A. Conditions curently met B. 16.0 psig C. 11.5 psig.
D. 4.0 psig Proposed Answer:
B. 16.0 psig Explanation (Optional):
A. Incorrect-Pressure needs to be 46psig or 5 FCUs running B. Correct-Meets above conditions C. Incorrect-1 I
.5 pounds is setpoint to initiate CS D. Incorrect-4.0 psig is Phase A setpoint on many Westinghouse plants Technical Reference(s):
E-1 I "Loss of Reactor or Secondanr Coolant Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-EOP-33 Obi V None Question Source:
Bank #
Modified Bank #
INPO-6198 Diablo Canvon 4/98 (Changed stem to use containment atmosphere parameters, Changed all distractors and answer New Memory or Fundamental Knowledge Question History:
Last NRC Exam Diablo Canvon 4/98 Question Cognitive Level:
Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
55.43 Comments:
A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
(CFR: 41.5/43.5/45.3 /45.13)
A2.08 Safe securing of containment spray when it can be done.
Examination Outline Cross-reference:
I Tier #
Group #
WA #
039 2.2.2 Importance Rating 3.5 Proposed Question:
The plant is conducting a start up per POP 1.3 "Plant Startup From Zero to 45% Power". SOP-MS-001 "Main and Reheat Steam System" is being utilized to warm up the Main Steam system. Which of the following is the correct sequence of events when warming up the Main Steam System during plant st a rt u p?
A. Open MSlV Bypass Valves, Achieve 100 psid Across MSIVs, Close MSlV Bypass Valves, Open MSIVs.
B. Open MSlV Bypass Valves, Achieve 50 psid Across MSIVs, Close MSlV Bypass Valves, Open MSIVs.
C. Open MSlV Bypass Valves, Achieve 100 psid Across MSIVs, Open MSIVs, Close MSlV Bypass Valves.
D. Open MSlV Bypass Valves, Achieve 50 psid Across MSIVs, Open MSIVs, Close MSlV Bypass Valves.
Proposed Answer:
D. Open MSlV Bypass Valves, Achieve 50 psid Across MSIVs, Open MSIVs, Close MSlV Bypass Valves.
Explanation (Optional):
Correct DIP across MSlVs per SOP-MS-001 is 50psid, and Bypasses are opened, MSlVs are then opened, and Bypasses are shut.
Technical Reference@):
SOP-MS-001 "Main and Reheat Steam Svstem" Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-POP-11 Obi 3, LIC-SPC-1 Obi 7 Question Source:
Bank #
Modified Bank #
New X
Question History:
Last NRC Exam N/A (Note changes or attach parent)
Question Cognitive Level:
10 CFR Part 55 Content:
55.41 X
Memory or Fundamental Knowledge Comprehension or Analysis X
55.43 Comments:
2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. (CFR 45.2)
Examination Outline Cross-reference:
Group #
WA #
056 2.4.47 Importance Rating 3.7 Proposed Question:
The following conditions and alarms exist with the plant at 100% power:
0 Steam Generator Levels In All 4 S/Gs have shown a brief lowering trend, but have recovered.
All Feedwater Regulating Valves have opened beyond normal steady state positions and have returned to near normal.
All 4 S/Gs showed a brief Steam Flow/ Feed Flow Mismatch, and have recovered.
All 4 S/G Feedwater Flows lowered slightly and have recovered MBFP Suction Pressure lowered to
, and has since raised to "6900 V Motor Trip (Common)" Annunciator is in alarm on Panel SHF CD-AOV-521 "Polisher Vessels and Post Filters Bypass" has opened 0
0 0
0 What is the diagnosed problem and required recovery action for the stated conditions?
A. A Condensate Pump has tripped and an immediate power reduction is necessary per ONOP-FW-001 "Loss of Feedwater".
B. A Condensate Pump has tripped and a power reduction is not currently required per ONOP-FW-001 "Loss of Feedwater".
C. A Condensate Booster Pump has tripped and an immediate power reduction is necessary per ONOP-FW-001 "Loss of Feedwater".
D. A Condensate Booster Pump has tripped and a power reduction is not currently required per ONOP-FW-001 "Loss of Feedwater".
Proposed Answer:
D. A Condensate Booster Pump has tripped and a power reduction is not required per ONOP-FW-001 "Loss of Feedwater".
Explanation (Optional):
Indications are indicative of a condensate booster pump trip since levels are not degrading. No actions are required to lower power per the ONOP.
Technical Reference(s):
ONOP-FW-001 "Loss of Feedwater" Proposed references to be provided to applicants during examination:
ARP-011 Panel SHF-
"Electrical" Daqe 2 Learning Objective:
LIC-SPC-4 Obi 6 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
I O CFR Part 55 Content:
55.41 55.43 x Comments:
2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
(CFR 41.10,43.5 /45.12)
Examination Out I ine Cross-reference:
1 Tier #
Group #
WA #
059 A3.04 Importance Rating 2.6 Proposed Question:
Given the following plant conditions:
0 The plant is operating at 80% power.
Both main boiler feed pumps (MBFPs) are in AUTO.
The "MBFP No. 31 & 32 SPEED CONTROL TROUBLE" annunciator has just alarmed.
The 'I% Feedwater" display is blank.
The "Failure Hold" light is illuminated.
0 0
0 0
The "% Hold" display indicates 77%.
Under these conditions, what action is CAUTIONED against by operating procedure ONOP-FW-1. "Loss of Feedwater"?.
A. Placing the MBFP Foxboro speed controller in manual to control speed.
B. Depressing "Reset Track and Hold" to transfer MBFP speed control to the Foxboro speed controller.
C. Allowing the track and hold circuitry to maintain control of MBFP speed.
D. Controlling MBFP speed locally at the Lovejoy Control Panel.
Proposed Answer:
B. Depressing "Reset Track and Hold" to transfer MBFP speed control to the Foxboro speed controller.
Explanation (Optional):
Actions A,C, and D are not cautioned against in ONOP-RN-0001, Action B is cautioned against in ONOP-FW-001 Technical Reference@):
ONOP-FW-001 "Loss of Feedwater Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-ONP-4 Obi 5 None Question Source:
Bank #
INPO-10538 IP3 4/96 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam IP3 4/96 Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 55.43 10 CFR Part 55 Content:
55.41 X
Comments:
A3 Ability to monitor automatic operation of the MFW, including:
(CFR 41.7 /45.5)
A3.04 Turbine driven feed pump
Examination Outline Cross-reference:
1 Tier #
Group #
K/A ##
061 K3.01 Importance Rating 4.6 Proposed Question:
The plant has experienced a Loss of Main and Auxiliary Feedwater. FR-H.l I Response To Loss of Secondary Heat Sink procedure has been implemented. FR-H.l contains a caution statement that ensures that the bleed and feed process is promptly initiated when the appropriate criteria are met.
What is the basis for starting bleed and feed without delay when the criteria are met?
A. Steam generator dryout cannot be observed, therefore action must be taken at an observable SG level.
B. This ensures some water left in the S/Gs so that thermal stress is reduced on a later initiation of main or auxiliary feedwater.
C. The ability to cool the core by all available methods will become unavailable if bleed and feed is delayed.
D. Core uncovery will begin a few minutes after bleed and feed initiation criteria occurs.
Proposed Answer:
C. The ability to cool the core by all available methods will become unavailable if bleed and feed is delayed.
Explanation (Optional):
Answer C is the basis for the bleed and feed process per WCAP.
Technical Reference(s):
FR-H.l I Response To Loss of Secondary Heat Sink Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-EOP-39 Obi K Bank #
INPO-6091 DC Cook 1/96 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam DC Cook 1/96 Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 X
55.43 Comments:
K3 Knowledge of the effect that a loss or malfunction of the AFW will have on the following:
(CFR: 41.7/45.6)
K3.01 RCS
Examination Outline Cross-reference:
Group #
WA #
061A1.01 Importance Rating 4.2 Proposed Question:
The plant is operating at 100% power at steady state when the following conditions are encountered:
a 480 V Bus 6A is lost due to an electrical fault 34 S/G Feedwater Regulating Valve positions to 50% open due to an internal control problem What will be the impact on the Auxiliary Feedwater System as a result of these conditions (Assume no operator action)?
A. 31 ABFP starts on Low Low Steam Generator Level at 12% in 34 S/G, Aux Feed is supplied to 33 and 34 S/Gs.
B. 31 ABFP starts on Low Low Steam Generator Level at 8% in 34 S/G, Aux Feed is supplied to 31 and 32 SIGs.
C. 33 ABFP starts on Low Low Steam Generator Level at 8%in 34 S/G, Aux Feed is supplied to 31 and 32 SIGs.
D. 33 ABFP starts on Low Low Steam Generator Level at 12% in 34S/G, Aux Feed is supplied to 33 and 34 SIGs.
Proposed Answer:
B. 31 ABFP starts on Low Low Steam Generator Level at 8% in 34 S/G, Aux Feed is supplied to 31 and 32 S/Gs.
Explanation (Optional):
Electric Driven ABFPs are 31 and 33.31 is powered from 3A 480V Bus, 33is powered from 6A 480V Bus. 31 injects to 31 and 32 S/Gs, 33 injects to 33 and 34 S/Gs. Electric pumps start on 2 of 3 detectors on I of 4 S/Gs at 8% S/G level. Only B Answer is correct.
Technical Reference@):
LIC-SPC-9 Proposed references to be provided to applicants during examination:
Learning Objective:
None LIC-SPC-9 Obi 1 and 4 Question Source:
Bank #
Modified Bank #
New X
Question History:
Last NRC Exam Question Cognitive Level:
Memory or Fundamental Knowledge 10 CFR Part 55 Content:
55.41 X
Comments:
A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW controls including: (CFR 41.5 / 45.5)
A1.O1 S/G level (Note changes or attach parent)
Comprehension or Analysis X
55.43
Examination Outline Cross-reference:
1 Tier #
Group #
KIA #
062K1.02 Importance Rating 4.4 Proposed Question:
Which of the following conditions would result in the starting and loading of the 33 Emergency Diesel Generator onto the 5A 480V Bus?
A. A Loss of Coolant Accident has caused a Safety Injection Actuation.
B. The 5A 480V Bus has it's voltage drop to 85% of Normal Voltage for 40 seconds.
C. The 5A 480V Bus has a rapid drop in bus voltage to 40% of Normal voltage.
D. The 5A 480V Bus has its Normal Feeder Breaker open on overcurrent.
Proposed Answer:
C. The 5A 480V Bus has a rapid drop in bus voltage to 40% of Normal voltage.
Explanation (Optional):
A. SI Signal only starts diesel, no loading B. Degraded Bus Voltage relay picks up at 80% of normal bus voltage for 40 seconds C. The instantaneous voltage relay picks up at 46% of normal bus voltage D. Overcurrent condition will lock the bus out. Diesel will not load.
Technical Reference(s):
LIC-EDS-I I Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-EDS-11 Obi 4 Bank #
Modified Bank #
New X
Last NRC Exam N/A Memory or Fundamental Knowledge (Note changes or attach parent)
Comprehension or Analysis X
55.41 X
55.43 Comments:
K1 Knowledge of the physical connections and/or cause effect relationships between the ac distribution system and the following systems:
(CFR 4 1.2 to 4 1.9)
K1.02 ED/G
Examination Outline Cross-reference:
1 Tier #
Group #
Importance Rating 3.3 WA #
062A2.05 Proposed Question:
Given the following plant conditions:
0 RCS Temperature-450 degrees F An RCS Cooldown is in progress A Loss of All Offsite Power has just occurred.
No 480V AC Buses are Energized No Emergency Diesel Generators Are Running RCS Pressure-1400 psig Based on these conditions, what procedure must be entered and what actions must be taken?
A. Enter ECA- 0.0, "Loss of All AC Power", Place listed Engineered Safeguards equipment control switches in Trip/Pullout/Off to prevent an uncontrolled start of large loads on Safeguards AC Buses.
B. Enter ECA- 0.0, Loss of All AC Power", Place listed Engineered Safeguards equipment control switches in Trip/Pullout/Off, to prevent an uncontrolled cooldown of the Reactor Coolant System.
C. Enter ONOP-EL-4, "Loss of Offsite Power", Place listed Engineered Safeguards equipment control switches in Trip/Pullout/Off to prevent an uncontrolled start of large loads on Safeguards AC Buses.
D. Enter ONOP-EL-4, "Loss of Offsite Power", Place listed Engineered Safeguards equipment control switches in Trip/Pullout/Off, to prevent an uncontrolled cooldown of the Reactor Coolant System.
Proposed Answer:
A. A. Enter ECA- 0.0, "Loss of All AC Power", Place listed Engineered Safeguards equipment control switches in Trip/Pullout/Off to prevent an uncontrolled start of large loads on Safeguards AC Buses.
Explanation (Optional):
Answer A is correct per WOG Basis Document for ECA 0.0, Mode 3 Conditions require use of EOPs Technical Reference@):
ECA-00 'I Loss of All AC Power Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
LIC-?????
Bank #
INPO-2814 Point Beach 8/99 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam Point Beach 8/99 Memory or Fundamental Knowledge
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 x Comments:
A2 Ability to (a) predict the impacts of the following malfunctions or op X
ration on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
(CFR 41.5 /43.5 /45.3 /45.13)
A2.05 Methods for energizing a dead bus
Examination Outline Cross-reference:
1 Tier #
Group #
WA #
063K2.0 1 Importance Rating 4.4 -
Proposed Question:
125 Volt DC Bus 32 has been lost due to a fault on the bus. Channel A Pressurizer Level Bistables are in the tripped condition for an I/C Surveillance. All other equipment has functioned as expected. what is the affect on continued plant operation?
A. A Reactor Trip occurs due to an shunt trip condition on the "A" reactor trip breaker.
B. A Reactor Trip occurs due to an undervoltage condition on the "A" reactor trip breaker.
C. A Reactor Trip occurs due to 2 of 4 logic met for High Pressurizer Level Trip.
D. No Immediate Reactor Trip occurs, however a 2 Hour Shutdown T.S LCO is entered.
Proposed Answer:
B. A Reactor Trip occurs due to an undervoltage condition on the "A" reactor trip breaker.
Explanation (Optional):
RTB and BYPA are powered from DC Bus 31, RTA and BYPB are powered from DC Bus 32, The associated vital AC bus is not lost as the inverter completes a rapid transfer to alternate AC source.
Technical Reference@):
None Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-EDS-7 Obi 5 Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 X
55.43 Comments:
K2 Knowledge of bus power supplies to the following:
(CFR 4 1.7)
K2.01 Major DC loads
Examination Outline Cross-reference:
1 Tier #
Group #
KIA #
063A1.01 Importance Rating 3.3 Proposed Question:
Given the following plant conditions:
A Station Blackout has occurred 31,32,33,and 34 125 Volt DC Electrical Buses are discharging at their design rate Restoration of AC Electrical Power is not expected for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Required Emergency Operating Procedure actions have been completed 0
By design, how long are the Station Batteries expected to last?
A. 1 Hour B. 2 Hours C. 3 Hours D. 4 Hours Proposed Answer:
B. 2 Hours Explanation (Optional):
Answer of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is given in the IP3 FSAR Chapter 8 Table 8.2-2 Technical Reference(s):
IP3 FSAR ChaDter 8 Table 8.2-2 Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-EDS-7 Obi 4 Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 X
55.43 Comments:
A1 Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including: (CFR 41.5 / 45.5)
A1.O1 Battery capacity as it is affected by discharge rate
Examination Outline Cross-reference:
Group #
KIA #
064K1.01 Importance Rating 4.4 Proposed Question:
Given the following plant conditions:
a The 2A 480V Vital bus experienced a loss of bus voltage The 31 EDG energized the 2A 480V bus Loads were appropriately sequenced onto the bus.
The normal source to the bus is now available 0
0 0
Which one of the following correctly completes the description of the method for restoration of the normal power supply to the 2A 480V Vital Bus in accordance with SOP-EL-001, "DIESEL GENERATOR OPERATION"?
The EDG is...
A. unloaded in Unit Mode, placed in parallel with the normal feeder breaker closed and then removed from the bus.
B. transferred to Parallel Mode, using the "Unit/ Parallel" Switch, placed in parallel with the normal feeder breaker closed and then removed from the bus.
C. unloaded in Unit Mode and removed from the bus before the normal feeder breaker is closed.
D. transferred to Parallel Mode automatically when the sequencer is reset, unloaded and removed from the bus before the normal feeder breaker is closed.
Proposed Answer:
B. transferred to Parallel Mode, using the "Unit/ Parallel" Switch, placed in parallel with the normal feeder breaker closed and then removed from the bus.
Explanation (Optional):
A. Incorrect-Shifted to parallel B. Correct per procedure C. Incorrect-Shifted to parallel D. Does not auto shift to parallel mode Technical Reference(s):
SOP-EL-001, "DIESEL GENERATOR OPERATION" Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
Bank #
INPO-I5179 Salem I 2/99 None LIC-EDS-I 1 Obi 6 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Salem 1 2/99 Question Cognitive Level:
Memory or Fundamental Knowledge -
Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
Comments:
55.43
K1 Knowledge of the physical connections and/or cause-effect relationships between the EDG system and the following systems:
K1.O1 AC Distribution System
Examination Outline Cross-reference:
1 Tier #
Group #
KIA #
064A2.02 Importance Rating 2.9 Proposed Question:
The 32 Emergency Diesel Generator has been paralleled to the 6A 480 Volt Bus for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to complete surveillance testing. The following indications exist:
Frequency-60.3 Hertz AC Kilowatts-2050 KW a
KVARS-1200 KVARs (Lagging)
Per "SOP-EL-001, "Diesel Generator Operation", what actions, if any, must be taken to ensure the continued operation of 32 EDG?
A. EDG WARS limit is being exceeded and must be corrected by raising 32 EDG terminal voltage.
B. EDG KVARs limit is being exceeded and must be corrected by lowering 32 EDG terminal voltage.
C. EDG AC Kilowatts limit is being exceeded, and must be corrected by raising 32 EDG engine speed.
D. EDG AC Kilowatts limit is being exceeded, and must be corrected by lowering 32 EDG engine speed.
Proposed Answer:
D. EDG AC Kilowatts limit is being exceeded, and must be corrected by lowering 32 EDG engine speed.
Explanation (Optional):
A. Incorrect-KVAR Lag limit is 1300 lagging B. Incorrect-KVAR Lag limit is 1300 lagging C. Incorrect-Must lower engine speed to unload "real" power from the diesel generator D. Correct-KW limit is 1950 KW for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 1750 steady state, must lower engine speed to unload
real power from the diesel generator Technical Reference@):
SOP-EL-001, "DIESEL GENERATOR OPERATION" Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
Bank #
None LIC-EDS-1 I Obi 5 Modified Bank #
New X
Memory or Fundamental Knowledge (Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
55.43
Comments:
A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based OD those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
(CFR 41.5143.5145.3 145.13)
A2.02 Load, VARS, pressure on air compressor, speed droop, kequency, voltage, fuel oil level, temperatures
Examination Outline Cross-reference:
Group #
7 KIA #
073K3.01 Importance Rating 4.2 Proposed Question:
A liquid waste release is in progress from the Monitor Tank. Process Radiation Monitor R-18 subsequently loses electrical power. What is the status of the liquid waste release, and what actions are required?
A. The release continues, and a second sample of effluent must be taken within 15 minutes.
B. The release is automatically terminated, no further action is required.
C. The release must be manually terminated, and can be resume once sampled.
D. The release continues, and a second sample must be immediately taken.
Proposed Answer:
B. The release is automatically terminated, no further action is required.
Explanation (Optional):
A. Incorrect-The release terminates, the 15 minute sample is required if the release is started with R-18 inoperable.
B. Correct-Loss of power isolates RCV-018 C. Incorrect-Isolates in Auto D. The release terminates Technical Reference(@:
Proposed references to be provided to applicants during examination:
SOP-WDS-014 "Liquid Waste Releases". ONOP-RM-001 "Failures of Radiation Monitors" None Learning Objective:
LIC-RDM-2 Obi 5 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam NIA Question Cognitive Level:
I O CFR Part 55 Content:
55.41 X
Comments:
Memory or Fundamental Knowledge Comprehension or Analysis X
55.43 K3 Knowledge of the effect that a loss or malfunction of the PRM system will have on the following:
(CFR: 41.7 / 45.6)
K3.01 Radioactive effluent releases
Examination Outline Cross-reference:
1 Tier #
Group #
KIA #
076A1.02 Importance Rating 2.6 Proposed Question:
The temperature of the Hudson River has risen to 82 degrees F. Which of the following Alarms would be indicative of the need to raise service water flow to the Component Cooling Water Heat Exchanger?
A. "lnst Air Comp 31 or 32 Trouble/ Auto Trip" B. 32 Emergency Diesel Generator "High Water Temperature" C. "SGBDHX-4 Blowdown Out. Temp. High" D. "Spent Fuel Pit High Temperature" Proposed Answer:
D. "Spent Fuel Pit High Temperature" Explanation (Optional):
A. Incorrect-SW Direct Supply B. Incorrect-SW Direct Supply C. Incorrect-SW Direct Supply D. Correct-Supplied By CCW Technical Reference@):
ARP-0IQ"Diesel Generators". ARP-012" Coolina Water and Air",
ARP-013 "Bearina Monitor". ARP-006 "Condensate and Feedwater" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-SAU-003 Obi 1 Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X -
A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including:
(CFR 41.5 /45.5)
A 1.02 Reactor and turbine building closed cooling water temperatures.
Examination Outline Cross-reference:
1 Tier #
Group #
WA #
078A3.01 Importance Rating 3.2 Proposed Question:
A large leak (approximately 400scfm) has occurred in the Nuclear Loop of the Instrument Air System.
ONOP-IA-I, "Loss of Instrument Air has been entered. Which statement below is correct concerning actions that are taken for this leak?
A. No mitigating actions are required since the cross connect with the Conventional Loop of Instrument Air automatically provides adequate volume of air to Nuclear Loop components.
B. Verify that PCV-I 142, "Station Air Backup To Instrument Air" valve automatically opens at 9Opsig to provide adequate volume of air to Nuclear Loop components.
C. Verify that PCV-1142, "Station Air Backup To Instrument Air" valve automatically opens at 95psig to provide adequate volume of air to Nuclear Loop components.
D. Manually open the inlet to PCV-1142, "Station Air Backup To Instrument Air" valve to provide adequate volume of air to Nuclear Loop components.
Proposed Answer:
B. Verify that PCV-1142, "Station Air Backup To Instrument Air" valve automatically opens at 9Opsig to provide adequate volume of air to Nuclear Loop components.
Explanation (Optional):
A. Incorrect-Conventional Loop can only provide 225 SCFM due o an orifice.
B. Correct-Correct setpoint for auto valve opening and volume of Station Air is 9OOSCFM C. Incorrect-Wrong setpoint D. Incorrect-Auto Only Technical Reference@):
Proposed references to be provided to applicants during examination:
SOP-IA-001 "Instrument Air System O~eration",ONOP-IA-l, "Loss of Instrument Air" None Learning Objective:
LIC-PSS-02 Obi 7 Question Source:
Bank #
Modified Bank #
New X
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge 10 CFR Part 55 Content:
55.41 X
Comments:
A2 Ability to (a) predict the impacts of the following malfunctions or operations on the IAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
(CFR 41.5 143.5 /45.3 145.13)
A3.01 Air pressure (Note changes or attach parent)
Comprehension or Analysis X
55.43
Examination Outline Cross-reference:
I Tier #
Group #
WA #
078 2.1.2 Importance Rating 4.0 Proposed Question:
"INST. AIR COMPR. 31 OR 32 TROUBLE/AUTO TRIP" alarm annunciator is received on Panel SJF.
Instrument air system parameters are normal. In accordance with AP-21, "Conduct of Operations" which of the following is the correct response to the alarm?
A. Depress the Alarm Acknowledge Pushbutton, Announce the alarm to the other licensed personnel, Verify acknowledgement from the CRS.
B. Announce the alarm to other licensed personnel, Verify acknowledgement from the CRS, Depress the Alarm Acknowledge Pushbutton.
C. Depress the Alarm Acknowledge Pushbutton, Announce the alarm to the other licensed personnel, Acknowledgement from the CRS is not required.
D. Announce the alarm to the other licensed personnel, Depress the Alarm Acknowledge Pushbutton, Acknowledgement from the CRS is not required.
Proposed Answer:
A. Depress the Alarm Acknowledge Pushbutton, Announce the alarm to the other licensed personnel, Verify acknowledgement from the CRS.
Explanation (Optional):
A. Correct-Sequence correct per AP-21, "Conduct of Operations" B. Incorrect-Incorrect sequence C. Incorrect-Must get acknowledgement from CRS D. Incorrect-Must get acknowledgement from CRS Technical Reference@):
AP-21, "Conduct of ODerations" Proposed references to be provided to applicants during examination:
None Learning Objective:
???????
Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 X
Comments:
55.43 2.1.2 Knowledge of operator responsibilities during all modes of plant operation.
(CFR 41.10/45.13)
Examination Outline Cross-reference:
1 Tier #
Group #
WA #
2.4.46 Importance Rating 3.6 Proposed Question:
The plant is at 100% power when the crew receives the "High Containment Pressure (SI) Channel Trip" alarm on panel SBF2. The following are the readings for Containment Pressure:
PC-948A-3.3 psig PC-948B-3.0 psig PC-948C-2.9 psig PC-948D-2.9 psig PC-948E-2.5 psig PC-948F-2.8 psig Which of the following statements and associated procedural action is correct about the "High Containment Pressure (SI) Channel Trip" alarm on panel SBF2?
A. The alarm is valid, 2 of the associated Containment Pressure Channels are tripped, enter E-0 "Reactor Trip or Safety Injection".
B. The alarm is valid, 1 of the associated Containment Pressure Channels is tripped, enter E-0 "Reactor Trip or Safety Injection".
C. The alarm is not valid, 1 of the associated Containment Pressure Channels is tripped, enter the issue into the corrective action system.
D. The alarm is not valid, none of the associated Containment Pressure Channels is tripped, enter the issue into the corrective action system.
Proposed Answer:
D. The alarm is not valid, none of the associated Containment Pressure Channels is tripped, enter the issue into the corrective action system.
Explanation (Optional):
A. Incorrect-A,B, and C Channels are not associated with this alarm.
B. Incorrect-A,B, and C Channels are not associated with this alarm.
C. Incorrect-DIE, and F Channels are below the 3 psig setpoint.
D. Correct-D,E, and F Channels are below the 3 psig setpoint.
Technical Reference@):
ARP-005, "SBF2-Safeauards" Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-ESS-1 Obi I Question Source:
Bank #
None Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam NIA Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 x
Comments:
2.4.46 Ability to verify that the alarms are consistent with the plant conditions.
(CFR: 43.5 / 45.3 / 45.12)
Examination Outline Cross-reference:
2 Tier #
Group #
WA #
001K5.88 Importance Rating 3.4 Proposed Question:
Given the following conditions, choose the correct statement concerning reactivity effects:
0 The plant is at 100% power, equilibrium conditions Rod Control is in Automatic, with the control bank inserted 40 steps.
A Single Shutdown Bank Control Rod has just ratcheted inward to 65 steps on IRPl No reactor trip has occurred Current RCS Boron Concentration is 235 ppm Tave is 567.1 degrees F 0
0 0
RCS Temperature........
A. change is negligiable since plant is early in core life, rod motion does not occur.
B. initially lowers, then automatic rod withdrawal restores Tave to 567.1 degrees F.
C. initially lowers, then automatic rod withdrawal restores Tave to 566. I degrees F.
D. change is negligiable since plant is late in core life, rod motion does not occur.
Proposed Answer:
Chitially lowers, then automatic rod withdrawal restores Tave to 566.1 degrees F.
Explanation (Optional):
A. Incorrect-Late in life, rod motion occurs B. Incorrect-Restores temperature to within I degree F due to "lock up" C. Correct-Restores temperature to within I degree F due to "lock up" D. Incorrect-Late in life should get rod motion Technical Reference(s):
ONOP-RC-1 "Dropped or Misalianed Control Rod Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-IXC-7 Obi 3 Question Source:
Bank #
Modified Bank #
New X
Memory or Fundamental Knowledge (Note changes or attach parent)
Question History:
Last NRC Exam NIA Question Cognitive Level:
10 CFR Part 55 Content:
55.41 X
Comments:
Comprehension or Analysis X
55.43 K5 Knowledge of the following OPS Implications as applied to CRDs K5.88 Effects of boron on temperature coefficient
Examination Outline Cross-reference:
2 Tier #
Group #
WA #
002A4.0 1 Importance Rating 3.8 Proposed Question:
The plant is at 100% power. Indications of a small RCS leak have been noticed by the operating crew.
The crew enters ONOP-RCS-7, "Excessive RCS Leakage". Step 3 "Monitor RCS Leakage" is implemented. The following computer trends are monitored:
Time VCT Level Pressurizer Level 0800 28%
48%
081 5 29%
46%
What is the calculation of RCS leakage per ONOP-RCS-7, "Excessive RCS Leakage"?
A. 11.3 gpm B. 9.5 gpm C. 8.7 gpm D. 7.1 gpm Proposed Answer:
C. 8.7 gpm Explanation (Optional):
A. Incorrect-Added VCT Level contribution B. Incorrect-Outside Answer Band C. Correct-19.3 gal/% for VCT Change, 75 gal/% for Pressurizer Level Change D. Incorrect-Outside Answer Band Technical Reference(s):
ONOP-RCS-7, "Excessive RCS Leakaae" Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-ONP-51 Obi 3 Question Source:
Bank ##
None Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge -
Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
Comments:
55.43 A4 Ability to manually operate and/or monitor in the control room: (CFR 41.7 / 45.5 to 45.8)
A4.01 RCS leakage calculation program using the computer
Examination Outline Cross-reference:
2 Tier #
Group #
KIA #
014A2.02 Importance Rating 3.6 -
Proposed Question:
A Loss of Offsite Power has occurred, and 31 EDG has failed to start. 32 and 33 EDGs have been loaded. What actions, if any, are required associated with Control Rod Position Indication?
A. Commence Emergency Boration per the response not obtained column of E-0, "Reactor Trip or Safety Injection" for Rod Bottom Lights NOT Being Lit.
B. Take all Rod Position verification actions per E-0, "Reactor Trip or Safety Injection" and ES-0.1 I' Reactor Trip Response" Rod Position Indication functions normally.
C. Commence Emergency Boration per response not obtained column of ES-0.1 I' Reactor Trip Response".
D. Emergency Boration is NOT required in either E-0, "Reactor Trip or Safety Injection" and ES-0.1 'I Reactor Trip Response", since Reactor Power is <5%, and IR Start Up Rate is negative.
Proposed Answer:
C. Commence Emergency Boration per response not obtained column of ES-0.1 " Reactor Trip Response".
Explanation (Optional):
A. Incorrect-Actions in E-0 inapropriate, Rx is Shutdown
- 6. Incorrect-Power is lost to ARPls C. Correct-With no ARPl indication-this action is correct D. Incorrect-Must verify rod position or borate in ES-0.1 Technical Reference(s):
Proposed references to be provided to applicants during examination:
Learning Objective:
LIC-EOP-32 Obi 12 Question Source:
Bank #
ES-0. I
'I Reactor TriD ResDonse" None Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge -
Comprehension or Analysis X
55.41 55.43 x 10 CFR Part 55 Content:
Comments:
A2 Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on those on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
(CFR 41.5 143.5 145.3 145.13)
A2.02 Loss of power to the RPIS
Examination Outline Cross-reference:
2 Tier #
Group #
WA #
017A2.01 Importance Rating 3.5 Proposed Question:
The plant has experienced a Loss of Coolant Accident, in which several vital ECCS systems have failed.
RVLIS full range indication is 50%. While monitoring Core Exit Thermocouple (CET) temperatures it is noticed that several indications have failed both high and low indicating a combination of shorted and open CET circuits. According to the Critical Safety Function Status Trees, which one of the following correctly states how incore thermocouples would be used to determine the onset of inadequate core cooling, and transition to FR-C.l I' Response To Inadequate Core Cooling" ?
A. When any 5 core exit TCs are greater than 715 degrees F.
B. When any non-failed core exit TC is greater than 1200 degrees F.
C. When any 5 core exit TCs are greater than 1200 degrees F.
D. When any non-failed core exit TC is greater than 715 degrees F.
Proposed Answer:
C. When any 5 core exit TCs are greater than 1200 degrees F.
Explanation (Optional):
A. Incorrect-With RVLIS>33% not an entry condition B. Incorrect-Decision not based on 1 CET reading-5 chosen to account for failures C. Correct-Per Safety Function Status Trees D. Incorrect-Decision not based on 1 CET reading-5 chosen to account for failures Technical Reference(s):
CSFTs Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-38 Obi 11 Question Source:
Bank #
INPO 931 1 Ginna 10/98 Modified Bank #
New Question History:
Last NRC Exam Ginna 10/98 Question Cognitive Level:
Memory or Fundamental Knowledge 10 CFR Part 55 Content:
55.41 X
Comments:
(Note changes or attach parent)
Comprehension or Analysis X
55.43 A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ITM system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:
(CFR: 41.5 143.5 145.3 145.5)
A2.0 1 Thermocouple open and short circuits
Examination Outline Cross-reference:
2 Group #
2 WA #
027K5.0 1 Importance Rating 3.4 Proposed Question:
Given the following plant conditions:
e The plant was operating at 100 % power, for 100 consecutive days A Design Basis LOCA has occurred The Containment Atmosphere has been sampled and Iodine is present e
e Which ONE (1) of the following describes a performance and design attribute of the HEPA and Charcoal filters found in the Containment Fan Cooler Units?
A. The HEPA filter removes particulates while the charcoal filter removes elemental and organic iodines, to meet the requirements of 10CFR50 App. A, GDC 60, "Control of Releases of Radioactive Materials To The Environment".
B. The HEPA filter removes particulates while the charcoal filter removes elemental and organic iodines, to meet the requirements of 10CFR50 App. A, GDC 41" Containment Atmosphere Cleanup" C. The HEPA filter removes elemental and organic iodines while the charcoal filter removes particulate, to meet the requirements of 1 OCFRSO App. A, GDC 41" Containment Atmosphere Cleanup" D. The HEPA filter removes elemental and organic iodines while the charcoal filter removes particulate, to meet the requirements of 10CFR50 App. A, GDC 60, "Control of Releases of Radioactive Materials To The Environment".
Proposed Answer:
B. The HEPA filter removes particulates while the charcoal filter removes elemental and organic iodines, to meet the requirements of 10CFR50 APP. A, GDC 41" Containment Atmosphere Cleanup" Explanation (Optional):
A. Incorred-GDC 41" Containment Atmosphere Cleanup" is correct per TS basis B. Correct-HEPA removes particulate, lodines adsorbed in Charcoal, and GDC 41" Containment Atmosphere Cleanup" per TS basis C. Incorrect-HEPA removes particulate D. Incorrect-HEPA removes particulate Technical Reference@):
Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
Bank #
INPO-4941 Robinson 2/96 LIC-ESS-4 and FSAR None LIC-ESS-4 Obi 2 Modified Bank #
New Memory or Fundamental Knowledge (Note changes or attach parent)
Question History:
Last NRC Exam Robinson 2/96 Question Cognitive Level:
10 CFR Part 55 Content:
55.41 X
Comprehension or Analysis X
55.43
Comments:
K5 Knowledge of the operational implications of the following concepts as they apply to the CIRS:
(CFR 41.7 / 45.7)
K5.01 Purpose of charcoal filters
Examination Outline Cross-reference:
2 Tier #
Group #
WA #
029A1.03 Importance Rating 3.3 Proposed Question:
The plant is at 100% power late in core life. The following vapor containment parameters are observed:
Vapor Containment Temperature: 1 15 degrees F Vapor Containment Pressure: -2.5 psig Vapor Containment Humidity: 70%
what, if any, vapor containment parameter is out of specification, and what action is required to return the parameter to specification?
A. Vapor Containment Temperature is out of specification, and running additional Fan Cooler Units is required.
B. Vapor Containment Humidity is out of specification, and running additional Fan Cooler Units is required.
C. Vapor Containment Pressure is out of specification, and performing a Containment Pressure Relief is required.
D. All parameters are within specification for the Vapor Containment, no actions are required.
Proposed Answer:
C. Vapor Containment Pressure is out of specification, and performing a Containment Pressure Relief is required.
Explanation (Optional):
A. Incorrect-TS Limit 120 degrees F B. Incorrect-No limit for humidity C. Correct-Pressure is out of spec Low, Operations of Purge system for vacuum/pressure relief is required D. Incorrect-Pressure is out of spec Technical Reference(s):
Oueration Proposed references to be provided to applicants during examination:
Question Source:
Bank #
SOP-CB-003 "containment Pressure Relief and Purge System None Learning Objective:
?????????????
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 55.43 10 CFR Part 55 Content:
55.41 X
Comments:
A1 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the Containment Purge System controls including:
(CFR: 41.5 /45.5)
Al.03 Containment pressure, temperature, and humidity
Examination Outline Cross-reference:
2 Tier #
Group #
KIA #
035A2.02 Importance Rating 4.4 Proposed Question:
A reactor trip has occurred due to failed UN coil in "A" reactor trip breaker. All systems functioned normally during the reactor trip. The following conditions exist:
All Control Rods inserted All 480V buses are powered by their normal sources Secondary System parameters are normal for post trip conditions E-0, "Reactor Trip or Safety Injection" was entered, and at step 4 a transition was made to ES-0.1, "Reactor Trip Response" Condenser Steam Dumps and S/G Atmospheric Relief Valves are closed Tave is 539 degree F and lowering slowly What is the status of Feedwater Flow to the Steam Generators, and what actions are required to be taken for the given plant conditions?
A. Main Feedwater is supplying flow via Feedwater Regulating Valve Bypass Valves at 75% open to all four Steam Generators. Take manual control of Feedwater Regulating Bypass Valves and lower flow to 1% of total feed flow.
B. Main Feedwater is supplying flow via Feedwater Regulating Valves at 75% open to all four Steam Generators. Take manual control of Feedwater Regulating Valves and lower flow to 1% of total feed flow.
C. Auxiliary Feedwater is supplying flow via 31 and 33 Auxiliary Feedwater Pumps to all four Steam Generators. Throttle AFW flow to slightly greater than 365 gpm until 1 S/G level is greater than 9%.
D. Auxiliary Feedwater is supplying flow via 32 Auxiliary Feedwater Pump to all four Steam Generators.
Throttle AFW flow to slightly greater than 365 gpm until 1 S/G level is greater than 9%.
Proposed Answer:
C. Auxiliary Feedwater is supplying flow via 31 and 33 Auxiliary Feedwater Pumps to all four Steam Generators. Throttle AFW flow to slightly greater than 365 gpm until I SIG level is greater than 9%.
Explanation (Optional):
A. Incorrect-Main Feedwter isolates post trip, close to an outdated mod.
B. Incorrect-Main Feedwater isolates post trip C. Correct-Auto start and feed of 31 and 33 AFW pumps D. Incorrect-32 Pump starts and runs in standby Technical Reference@):
LIC-SPC-009 Proposed references to be provided to applicants during examination:
None Learning 0 bject ive:
LIC-SPC-005 Obi 5. LIC-SPC- 009 Obi 2 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
10 CFR Part 55 Content:
55.41 X
Memory or Fundamental Knowledge Comprehension or Analysis 55.43 X
Comments:
A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SIG; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
(CFR 41.5 143.5 145.3 / 45.5)
A2.02 Reactor triplturbine trip
Examination Outline Cross-reference:
2 KIA #
01 5K6.01 Importance Rating 3.2 Tier #
Group #
Proposed Question:
During a normal plant shutdown, intermediate range detector channel N36 fails high with reactor power at 6%. Which ONE of the following statements describes how this failure affects the reactor and subsequent operation of the Nuclear Instrumentation system?
The reactor will:
A. trip on high IR flux and source range Nls will have to be manually re-energized.
B. NOT trip and source range Nls will have to be manually reenergized.
C. NOT trip, and source range Nls will re-energize when N35 reaches the proper setpoint.
D. trip on high IR flux, and source range Nls will re-energize when N35 reaches the proper setpoint.
Proposed Answer:
A. trip on high IR flux and source range Nls will have to be manually re-energized.
Explanation (Optional):
A. Correct-Power is Ranges to energize source range Nls B. Incorrect-Trip occurs C. Incorrect-Trip occurs D. Incorrect-Must manually energize Source Range Instruments P-10 so Intermediate Range trip is enabled. Takes 2/2 lowering Intermediate Technical Reference@):
LIC-IXC-5 Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-IXC-5 Obi 5 Question Source:
Bank #
INPO-11042 Kewaunee 0211 997 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Kewaunee 02/1997 X
Question Cognitive Level:
10 CFR Part 55 Content:
55.41 X
Comments:
Memory or Fundamental Knowledge Comprehension or Analysis 55.43 K6 Knowledge of the effect of a loss or malfunction on the following will have on the NIS (CFR41.7,45.7)
K6.01 Sensors, Detectors, and Indicators
Examination Outline Cross-reference:
2 Tier #
WA #
2.2.17 Importance Rating 3.5 Group #
Proposed Question:
The 32 Steam Jet Air Ejector(SJAE) is experiencing backfiring, and must be swapped from service for troubleshooting and repair during plant operations. There will be no breach of the SJAE, Maintenance plans to ultrasonically test the "air-lane" to ensure that there is no foreign material in it. How is appropriate plant configuration maintained during this maintenance evolution?
A. The Maintenance Work Package must include steps to realign valves to place 32 SJAE back into service, after maintenance since there are no other means to control the system status.
B. The 32 SJAE must be swapped to a standby SJAE and the 32 SJAE must be tagged out for the purpose of controlling configuration.
C. The 32 SJAE is swapped to a standby SJAE utilizing a plant operating procedure, and as such is controlled such that configuration is maintained.
D. The 32 SJAE is swapped to a standby SJAE, and configuration control must be controlled with a "Check Off List" in accordance with OD-9, "System Status Control".
Proposed Answer:
C. The 32 SJAE is swapped to a standby SJAE utilizing a plant operating procedure, and as such is controlled such that configuration is maintained.
Explanation (Optional):
A. Incorrect-Not required if removed and returned via an approved procedure B. Incorrect-Not required if removed and returned via an approved procedure, and the maintenance does not require tagging for personnel or plant safety reasons C. Correct-This is an acceptible exclusion to the requirements of OD-9, a checkoff list is not required D. Incorrect-Controlled by an approved procedure, this is an exclusion to OD-9 Technical Reference@):
OD-9, "System Status Control" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
???????????
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X
I O CFR Part 55 Content:
55.41 55.43 x
Comments:
2.2.17 Knowledge of the process for managing maintenance activities during power operations.
(CFR 43.5 / 45.13)
Examination Outline Cross-reference:
2 KIA #
071 K4.06 Importance Rating 3.5 Tier #
Group #
Proposed Question:
Which of the following combinations of Process Radiation Monitors are interlocked such that individually they will cause an automatic closure of RCV-014 'Waste Gas Discharge Flow Control Va1ve"during a Waste Gas Decay Tank release, terminating the release?
A. R-14 "Plant Vent Gas PRM" AND R-12 "VC Gas Activity PRM" B. R-20 Waste Gas Activity PRM" AND R27" Wide Range Plant Vent Gaseous Activity PRM" C. R-12 "VC Gas Activity PRM" AND R-20 'Waste Gas Activity PRM" D. R-14 "Plant Vent Gas PRM" AND R-27" Wide Range Plant Vent Gaseous Activity PRM" Proposed Answer:
D. R-I4 "Plant Vent Gas PRM" AND R-27" Wide Range Plant Vent Gaseous Activity PRM" Explanation (Optional):
A. Incorrect-R-12, although mentioned in WGDT release procedure does not auto isolate release B. Incorrect-R-20, although mentioned in WGDT release procedure does not auto isolate release C. Incorrect-Neither R-20 or R-12 terminate a release D. Correct-Either detector terminates a release.
Technical Reference@):
SOP-WDS-013 "Gaseous Waste Releases" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
LIC-PSA-6 Obi 5 Bank ##
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 X
55.43 K4 Knowledge of design feature@) and/or interlock@) which provide for the following: (CFR 41.7)
K4.06 Sampling and monitoring of waste gas release tanks
Examination Outline Cross-reference:
2 KIA #
072K4.02 Importance Rating 3.4 Tier #
Group #
Proposed Question:
The plant is in Mode 6 with fuel reconstitution activities underway in the Fuel Storage Building. Area Radiation Monitor R-5 "Fuel Storage Building ARM" comes into alarm. Which of the following describes automatic actions, if any, as a result of this alarm?
A. FSB Supply Fan Stops, FSB Exhaust Fan Starts, FSB Sliding Door Shuts, Charcoal Filter Inlet and Outlet Face Dampers Open.
B. FSB Supply Fan Starts, FSB Exhaust Fan Starts, FSB Sliding Door Shuts, Charcoal Filter Inlet and Outlet Face Dampers Open.
C. FSB Supply Fan Starts, FSB Exhaust Fan Starts, FSB Sliding Door Opens, Charcoal Filter Inlet and Outlet Face Dampers Isolate.
D. Area Radiation Monitor R-5 "Fuel Storage Building ARM" has only alarm function, there are no automatic actions.
Proposed Answer:
A. FSB Supply Fan Stops, FSB Exhaust Fan Starts, FSB Sliding Door Shuts, Charcoal Filter Inlet and Outlet Face Dampers Open.
Explanation (Optional):
" A has the only correct sequence of events Technical Reference(s):
SOP-RM-1 "Area Radiation Monitors" Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-RDM-3 Obi 5 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 55.43 10 CFR Part 55 Content:
55.41 X
Comments:
K4 Knowledge of ARM system design feature@) and/or interlock(
s) which provide for the following:
(CFR: 41.7)
K4.02 Fuel building isolation
Examination Outline Cross-reference:
2 Tier #
Group #
WA #
086K6.04 Importance Rating 2.9 Proposed Question:
A fire has started in oil soaked lagging on 32 Main Boiler Feed Pump. The Heat Activated Detector (HAD) near the fire has failed. Assuming the fire is of sufficient size to actuate the HAD if it were operable, which statement below is correct?
A. The system is a "wet pipe" fire suppression system, and will provide fire suppression when local "fusible links" melt.
B. The system is a "pre-action" fire suppression system, and will NOT supply fire suppression since the Heat Activated Detector (HAD) has failed.
C. The system is a "deluge" fire suppression system, and will NOT supply fire suppression since the Heat Activated Detector (HAD) has failed.
D. The system provides a local and control room alarm ONLY, and manual fire suppression action is necessary via portible equipment.
Proposed Answer:
C. The system is a "deluge" fire suppression system, and will NOT supply fire suppression since the Heat Activated Detector (HAD) has failed.
Explanation (Optional):
Answer "C" is the only correct answer, the MBFP has a deluge system and will not actuate since the HAD has failed.
Technical Reference@):
LIC-PSS-5 Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-PSS-5 Obi 2 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 55.43 10 CFR Part 55 Content:
55.41 X
Comments:
K6 Knowledge of the effect of a loss or malfunction on the Fire Protection System following will have on the : (CFR 41.7 / 45.7)
K6.04 Fire, smoke, and heat detectors
Examination Outline Cross-reference:
Tier #
Group #
KIA #
2.1.11 Importance Rating 3.8 Proposed Question:
Given the following plant conditions:
e The Plant is in MODE I.
A PORV block valve was closed, and would not reopen.
Which ONE of the following must be performed within I hour per Technical Specifications?
A. Verify the other PORV block valve is OPERABLE.
B. Place the affected PORV block valve in MANUAL.
C. Remove electrical power from the affected PORV block valve.
D. Initiate action to place the unit in a MODE in which the specification does not apply.
Proposed Answer:
B. Place the affected PORV block valve in MANUAL.
Explanation (Optional):
A. Incorrect-Listed in other Tech Specs, but not in IP3 B. Correct-Per ITS C. Incorrect-Action for failed PORV D. Incorrect-Can operate with block closed for 7 days Technical Reference(s):
Proposed references to be provided to applicants during examination:
ImDroved Technical Specifications-IP3 None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
I O CFR Part 55 Content:
Bank #
INPO-5113 Farlev 10/95 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam Farlev 10/95 Memory or Fundamental Knowledge x
Comprehension or Analysis 55.41 55.43 x
Comments:
2.1.1 1 Knowledge of less than one hour technical specification action statements for systems.
(CFR: 43.2 / 45.13)
Examination Outline Cross-reference:
Tier #
Group #
WA #
2.1.14 Importance Rating 3.3 Proposed Question:
Which of the following conditions would require notification of UNIT 2 Operations Department personnel "as time permits" per AP-21, "Conduct of Operations"?
A. Entry into a I Hour Technical Specification Limiting Condition of Operation Action Statement.
B. An Injury to plant personnel that does not require an ambulance.
C. A small fire that was extinquished in 5 minutes in the Operations Management office area.
D. A planned WGDT release per SOP-WDS-013, "Gaseous Waste Releases".
Proposed Answer:
C. A small fire that was extinquished in 5 minutes in the Operations Management office area.
Explanation (Optional):
A,B, and D are incorrect as they are not listed in AP-21 pgs 23 and 24 C. Correct-Plant Fires is a listed item Technical Reference@):
AP-21. "Conduct of Ooerations" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 x
X Comments:
2.1.14 Knowledge of system status criteria which require the notification of plant personnel.
(CFR: 43.5 / 45.12)
Examination Outline Cross-reference:
Tier #
Group #
WA #
2.1.31 Importance Rating 3.9 Proposed Question:
The Operations Manager has required that a Valve Lineup Verification be completed on the Chemical and Volume Control System as a corrective action for some identified mis-positioning events coming out of the recent plant outage. One of the valves is in a "very high radiation area" and is a locked open component. Which of the methods listed below is acceptable to verify the locked valve's position per OD-35 "Component Verification and Control"?
A. A verification of Radiation Protection records is completed to ensure that no one entered the area since the last time the valve was positioned.
B. Flow may be initiated through the system from the Control Room and a verification of flow can be verified on Control Room Panels.
C. A paperwork review of the previous stop tag release paperwork and previous check off list is completed such that the adminstrative position is understood.
D. It is not possible to enter the area that the valve is in, so the verification paperwork is marked "N/A" for this valve.
Proposed Answer:
B. Flow may be initiated through the system from the Control Room and a verification of flow can be verified on Control Room Panels.
Explanation (Optional):
A. Incorrect-Not listed as acceptable in procedure B. Correct-Per alternate methods of verification in OD-35 C. Incorrect-paperwork review is not a method in OD-35 D. Incorrect- "N/A is not a method in OD-35 Technical Reference@):
Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
Bank #
OD-35 "ComDonent Verification and Control"?
None Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
Comments:
55.43 2.1.29 Knowledge of how to conduct and verify valve lineups (CFR 41.10/45.1/45.12)
Examination Outline Cross-reference:
Tier #
Group #
WA #
2.1.24 Importance Rating 3.1 Proposed Question:
The valve body print designation shown below is indicative of which of the following type valve?
A. Gate Valve B. Diaphragm Valve C. Needle Valve D. Butterfly Valve Proposed Answer:
C. Needle Valve Explanation (Optional):
Per Dwg # 9321 -D-20163 Technical Reference@):
Dwcl# 9321-D-20163 "Flow Diaaram Symbols Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
?????????
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge Comprehension or Analysis 55.41 X
55.43 X
2.1.24 Ability to obtain and interpret station electrical and mechanical drawings.
(CFR 45.12 /45.13)
Examination Outline Cross-reference:
Tier #
Group #
KIA #
2.2.29 Importance Rating 2.2.29 Proposed Question:
Which ONE (1) of the following is a responsibility of the Refueling SRO?
A. Authorize changes in the fuel load pattern, as recommended by Reactor Engineering B. Provide Command and Control for Loss of RHR Cooling during Refueling C. Provide Command and Control for Loss of Refueling Pool Level during Refueling D. Provide Command and Control for Loss of Offsite Power during Refueling Proposed Answer:
C. Provide Command and Control for Loss of Refueling Pool Level during Refueling Explanation (Optional):
A. Incorrect-Authorized by Reactor Engineering B. Incorrect-Control Room CRS runs this casualty C. Correct-Refuel SRO directs putting fuel in safe place and evacuation D. Incorrect-Control Room CRS runs this casualty Technical Reference@):
Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Bank #
INPO-10924 Kewaunee 12/97 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam Kewaunee 12/97 Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 55.43 x
Comments:
2.2.29 Knowledge of SRO fuel handling responsibilities
Examination Outline Cross-reference:
Tier #
Group #
WA ##
2.2.21 Importance Rating 3.5 Proposed Question:
Which of the following individuals is responsible for 'I declaring a system inoperable" when a system is taken out of service for a planned LCO action statement entry for maintenance and for "declaring a system operable" after all Post Work and Operability testing is completed?
A. Field Support Supervisor B. Control Room Supervisor C. Shift Manager D. WCC Supervisor Proposed Answer:
C. Shift Manager Explanation (Optional):
C is the only correct answer per AP-53," Control of Maintenance Activities Under LCO Action Statements" Step 4.8.7 and 4.7.4 Technical Reference(s):
AP-53." Control of Maintenance Activities Under LCO Action Statements" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
????????????
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge Comprehension or Analysis 55.41 55.43 x
X Comments:
2.2.21 Knowledge of pre-and post-maintenance operability requirements. (CFR 43.2)
Examination Outline Cross-reference:
Tier #
Group #
KIA #
2.2.6 Importance Rating 3.3 Proposed Question:
Which one of the following correctly states the MAXIMUM time allowable between the implementation date, and expiration date of a Temporary Procedure Change (TPC)NOT associated with Out of Service Equipment?
Proposed Answer:
A. 90 days with no extension B. 90 days with a possible extension of up to 60 days C. 14 days with no extension D. 14 days with a possible extension of up to 6 months Explanation (Optional):
Technical Reference( s) :
AP-3 Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
I O CFR Part 55 Content:
???????????
Bank #
Modified Bank #
New IP 2 Exam Bank (Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 X
55.43 Comments:
2.2.6 Knowledge of the process for making changes in procedures as described in the S A R
Examination Outline Cross-reference:
Tier #
Group ##
KIA ##
2.2.25 Importance Rating 3.7 Proposed Question:
What is safety limit basis for the over-power delta-temperature reactor trip setpoint?
A. Fuel maximum kwlft.
B. Fuel peak clad temperature.
C. Fuel surface DNBR.
D. RCS over-pressurization.
Proposed Answer:
A. Fuel maximum kw/ft.
Explanation (Optional):
Technical Reference(s):
Improved Tech SDecs Basis Document Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Bank #
INPO-1051 1 IP3 4/96 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam IP-3 4/96 Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 55.43 X
Comments:
2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (CFR: 43.2)
Examination Outline Cross-reference:
Tier #
Group #
KIA #
2.3.6 Importance Rating 3.4 Proposed Question:
While reviewing a release permit for a waste monitor tank, it is determined that R-18 I' Waste Disposal Liquid" Monitor has failed its source check. Per SOP-WDS-14, "Liquid Waste Releases" what are the required actions, if any, of the Control Room Supervisor ?
A. Do not approve the liquid waste discharge, secure the lineup, liquid waste discharge is not permitted until R-18 is repaired.
B. Approve the liquid waste discharge, no other actions required, R-I 8 has a backup monitor.
C. Approve the liquid waste discharge, as long a a second sample is drawn and calculations are second verified prior to the release.
D. Approve the liquid waste discharge, and ensure that continuous effluent sampling is conducted throughout the liquid waste discharge.
Proposed Answer:
C. Approve the liquid waste discharge, as long a a second sample is drawn and calculations are second verified prior to the release.
Explanation (Optional):
Per SOP-WDS-14, "Liquid Waste Releases", with R-18 00s the release can continue as long as a second sample is taken and the calculations are second checked.
Technical Reference(@:
SOP-WDS-14. "Liauid Waste Releases" Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-PSA-5 Obi 7 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam NfA Question Cognitive Level:
10 CFR Part 55 Content:
Memory or Fundamental Knowledge Comprehension or Analysis 55.41 55.43 x
x Comments:
2.3.6 Knowledge of the requirements for reviewing and approving release permits. (CFR 43.4 / 45.10)
Examination Outline Cross-reference:
Tier #
Group #
WA #
2.3.4 Importance Rating 3.1 Proposed Question:
An ALERT was declared at Indian Point Unit 3 No emergency response facilities are staffed. An operator must go into a room requiring radiation exposure in excess of the limits prescribed in 1 OCFRZO.
Assuming that no authorities have been delegated, who must authorize the operator to proceed?
A. Field Supervisor B. Senior Radiation Protection Supervisor present on site C. Shift Manager D. Control Room Supervisor Proposed Answer:
C. Shift Manager Explanation (Optional):
Technical Reference@):
Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Bank #
INPO-19575 DC Cook 5/01 Modified Bank #
New (Note changes or attach parent)
Last NRC Exam DC Cook 5/01 Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 55.43 x
Comments:
2.3.4 Knowledge of Radiation Exposure Limits and Contamination Control, including permissible levels in excess of those authorized (CFR: 43.4/45.10)
Examination Outline Cross-reference:
Tier #
Group #
WA #
2.3. I 1 Importance Rating 3.2 Proposed Question:
A Steam Generator Tube Rupture has just occurred on 32 Steam Generator. E-3, "Steam Generator Tube Rupture" has been implemented. Which of the following actions is performed to directly limit the potential radiation release to the public?
A. Maintain feed flow to the ruptured S/G until level in that S/G is greater than 9% (14%).
B. Adjust ruptured S/G atmospheric relief valve controller to 1040 psig.
C. Entering E-2, "Faulted S/G Isolation" for a S/G depressurizing in an uncontrolled manner.
D. Verifying that the ruptured S/G pressure is greater than 400 psig.
Proposed Answer:
B. Adjust ruptured S/G atmospheric relief valve controller to 1040 psig.
Explanation (Optional):
A. Incorrect-Basis to prevent S/G depressurization prematurely B. Correct-Preferentially keeps ruptured S/G atmospheric valve closed, other S/Gs remove decay heat C. Incorrect-Stops feeding a faulted S/G with action steps in E-3 D. Incorrect-Determines appropriate cooldown strategy Technical Reference(s):
EOP Technical Basis Document Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-35 Obi14 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 X
Comments:
55.43 2.3.1 1 Ability to control radiation releases.
(CFR 45.9 / 45.10)
Examination Outline Cross-reference:
Tier #
Group #
WA #
2.3.1 Importance Rating 3.0 Proposed Question:
Which ONE (I) of the following is the required radiological posting for an area in which a person could receive a radiation dose of 0.075 rem in an hour?
A. No Posting Is Required B. Radiation Area C. High Radiation Area D. Very High Radiation Area Proposed Answer:
C. High Radiation Area Explanation (Optional):
Technical Reference@):
Proposed references to be provided to applicants during examination:
None Learning Objective:
????????
Question Source:
Bank #
INPO-17831 Seauovah 6/98 Modified Bank #
New (Note changes or attach parent)
Question History:
Last NRC Exam Seauovah 6/98 Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 X
55.43 Comments:
2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements.
(CFR 41.12 /43.4.45.9/45.10)
Examination Outline Cross-reference:
Tier #
Group ##
WA #
2.3.10 Importance Rating 3.3 Proposed Question:
ONOP-RM-002, "High Activity-Radiation Monitoring System" has been entered due to a valid alarm on the R-33 "Control Room Gas" Radiation Monitor. What actions are required to protect against personnel exposure?
A. Immediately Evacuate the Main Control Room B. Place Control Room Ventilation In "100% Recirculation" Mode C. Secure the Control Room Ventilation System Until Operators Don EABs D. Place Control Room Ventilation In "10% Incident" Mode Proposed Answer:
D. Place Control Room Ventilation In "10% Incident Mode Explanation (Optional):
This is the correct action per ONOP-RM-2 Att 7 Technical Reference(s):
ONOP-RM-002. "Hiah Activitv-Radiation Monitorina Svstem" Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
LIC-ONP-33 Obi 4 Bank #
Modified Bank #
New x
(Note changes or attach parent)
Last NRC Exam N/A Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 X
55.43 Comments:
2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
(CFR 43.4 / 45.10)
Examination Outline Cross-reference:
Tier #
Group #
WA #
2.4.38 Importance Rating 4.0 Proposed Question:
A Loss of Offsite Power has occurred at Indian Point 3, you are filling in as the Shift Manager, and have assumed the duties of the Emergency Director during the early stages of this event. What are the time limits to complete the Emergency Classification and complete the Immediate Actions in the flow chart of the Emergency Plan Implementing Procedure IP-2001?
A. Emergency Classification-within 15 minutes of the initiating conditions, Immediate Actions-within 15 minutes of the "declaration" B. Emergency Classification-within 15 minutes of the initiating conditions, Immediate Actions-within 30 minutes of the "declaration" C. Emergency Classification-within 30 minutes of the initiating conditions, Immediate Actions-within 15 minutes of the "declaration" D. Emergency Classification-within 30 minutes of the initiating conditions, Immediate Actions-within 30 minutes of the "declaration" Proposed Answer:
A. Emergency Classification-within 15 minutes of the initiating conditions, Immediate Actions-within 15 minutes of the "declaration" Explanation (Optional):
Per IP-2001 steps 4.2 A and B, it should take no longer than 15 minutes to classify the event and no longer than 15 minutes to complete Immediate Actions once declaration is complete Technical Reference@):
Per IP-2001 steps 4.2 A and B Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-ERT-12 Obi 6 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 55.41 55.43 X
10 CFR Part 55 Content:
Comments:
2.4.38 Ability to take actions called for in the facility emergency plan, including (if required) supporting or acting as emergency coordinator.
(CFR: 43.5 / 45.1 1)
Examination Outline Cross-reference:
Tier #
Group #
WA #
2.4.13 Importance Rating 3.9 Proposed Question:
The plant has experienced a Reactor Trip, and the crew has transitioned to ES-0.1 "Reactor Trip Response". Step 1 of this Emergency Operating Procedure has an Asterisk (*) next to the step. What is the meaning of the Asterisk(*) in terms of how the step is carried out by the operating crew?
A. The CRS reads that the step is a "continuous action step" and if at any time during the EOP in effect conditions require the "response not obtained" actions to be completed, then the crew returns to that step and completes the actions.
B. The CRS reads that the step is a "continuous action step" and if at any time in the EOP Network conditions require the "response not obtained" actions to be completed, then the crew returns to that step and completes the actions.
C. The CRS reads that the step is an "Immediate Action Step" and that completion of the step needs to proceed without delay. No crew informational briefs will be conducted during the completion of the step and external communications will be kept to a minimum.
D. The CRS reads that there is a "caution or note" associated with the step in effect. Then the "caution or note" is read and the CRS receives confirmation that the "caution or note" is understood before continuing with the procedure.
Proposed Answer:
A. The CRS reads that the step is a "continuous action step" and if at any time during the EOP in affect conditions require the "response not obtained" actions to be completed, then the crew returns to that step and completes the actions.
Explanation (Optional):
A. Correct-Per station EOP rules of usage the
- denote a continuous action step, and that step is only continuous for the time while in that procedure.
B. Incorrect-Per station EOP rules of usage the
- denote a continuous action step, and that step is only continuous for the time while in that procedure.
C. Incorrect-Per station EOP rules of usage the
- denote a continuous action step D. Incorrect-Per station EOP rules of usage the
- denote a continuous action step Technical Reference@):
WOG EOP Guidance Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-31 Obi 5 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A
Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 10 CFR Part 55 Content:
55.41 X
55.43 Comments:
2.4.13 Knowledge of crew roles and responsibilities during EOP flowchart use. (CFR: 41.10 / 45.12)
Examination Outline Cross-reference:
Tier #
Group #
KIA #
2.4.1 1 Importance Rating 3.6 Proposed Question:
Indian Point 3 is conducting a plant start up after a short outage. Reactor Power is 30%. An automatic turbine trip occurs, and the crew enters ONOP-TG-4, "Turbine Trip Below P-8'. Which of the following is an initial operator action for this condition?
A. Place Steam Dumps in Manual control.
B. Place Feed Water Regulating Valves in Manual control.
C. Place the Rod Control System in Manual control.
D. Place Main Boiler Feed Pump Speed Control in Manual control.
Proposed Answer:
C. Place the Rod Control System in Manual control.
Explanation (Optional):
A. Incorrect-Prefer Steam Dumps in Auto to respond to the transient B. Incorrect-Prefer Feedwater Regulating Valves in Auto to respond to the transient C. Correct-Prevent rods from driving in and allows plant to stabilize. Large Tavernref error develops due to the turbine trip.
D. Incorrect-Prefer Main Boiler Feed Pump Speed Control in Auto to respond to the transient Technical Reference(s):
ONOP-TG-4, "Turbine TriD Below P-8" Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-ONP-52 Obi 3 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
10 CFR Part 55 Content:
55.41 X
Memory or Fundamental Knowledge Comprehension or Analysis X
55.43 Comments:
2.4.1 1 Knowledge of abnormal condition procedures. (CFR 41.10 / 43.5 / 45.13)
Examination Outline Cross-reference:
Tier #
Group #
WA #
2.4.16 Importance Rating 4.0 Proposed Question:
A Reactor Trip has occurred due to a loss of Instrument Air. The crew is taking actions per ES-0.1,"
Reactor Trip Response". The RO has recommended to enter ONOP-IA-I, "Loss of Instrument Air".
Which of the following statements is correct regarding this recommendation?
A. While in the EOP Network, use of lower tier procedures is prohibited, therefore ONOP-IA-1, "Loss of Instrument Air" can not be implemented.
B. ONOP-IA-1, "Loss of Instrument Air" can be completed concurrently with the EOP Network with the agreement of the CRS.
C. ONOP-IA-1, "Loss of Instrument Air" takes immediate priority over ES-0.1 'I Reactor Trip Response" actions since Instrument Air is assumed to be available in the EOP network.
D. There is no need to implement ONOP-IA-1, "Loss of Instrument Air", as there are direct actions in ES-0.1 I' Reactor Trip Response" to restore Instrument Air.
Proposed Answer:
B. ONOP-IA-1, "Loss of Instrument Air" can be completed concurrently with the EOP Network with the agreement of the CRS.
Explanation (Optional):
A. Incorrect-Concurrent actions are allowed and in some cases necessary in the EOPs B. Correct-Per station EOP usage guidelines C. Incorrect-Can be worked in parallel, but do not take priority over EOPs D. Incorrect-EOP ES-0.1 does not restore instrument air.
Technical Reference(s):
WOG EOPUsaae Document Proposed references to be provided to applicants during examination:
None Learning Objective:
LIC-EOP-31 Obi 19 Question Source:
Bank #
Modified Bank #
New X
(Note changes or attach parent)
Question History:
Last NRC Exam N/A Question Cognitive Level:
Memory or Fundamental Knowledge X
Comprehension or Analysis 55.43 I O CFR Part 55 Content:
55.41 X
Comments:
2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures.
(CFR41.10/43.5/45.13)
Group #
WA #
Importance Rating Examination Outline Cross-reference:
Proposed Question:
Proposed Answer:
Explanation (Optional):
Technical Reference@):
Proposed references to be provided to applicants during examination:
Learning Objective:
Question Source:
Question History:
Question Cognitive Level:
I O CFR Part 55 Content:
Comments:
Bank #
Modified Bank #
(Note changes or attach parent)
New Last NRC Exam Memory or Fundamental Knowledge -
Comprehension or Analysis 55.41 55.43