ML060860155
| ML060860155 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/01/2006 |
| From: | Jane Spence Dominion |
| To: | Johnson G Operations Branch I |
| Conte R | |
| References | |
| Download: ML060860155 (77) | |
Text
t NRC Written Examination Answer Sheet tone Unit Two Reactor Onerator . .
J J
J Page 1 of 2
NRC Written Examination Answer Sheet tone Unit Two Reactor Operator Examlnatlon J
J All work done on this examination is my own; I have neither given, nor received aid.
Examinee's Signature Page2of 2
The plant was being shutdown per AOP-2575, Rapid Downpower, when the US directed a Manual Trip.
The Crew has since transitioned to EOP 2526, Reactor Trip Recovery and is performing applicable actions.
[Take the conditions stated in each of the following choices SEPERATELY. Assume there are NO other problems with ANY other components or operational guidelines.]
Which of the following plant conditions would REQUIRE the use of an AOP, in conjunction with the use of EOP-2526?
A The plant shutdown was required by AOP-2560, Storms, High Winds and High Tides, due to an 2:
approaching hurricane.
n The trip was required by AOP-2575 due to the Secondary Operator allowing RCS Tavg to get 15 I__
"F off of program band.
c The plant shutdown was required by AOP-2556, CEA Malfunctions, due to a dropped CEA that j 1 could not be recovered.
D While performing EOP-2526, the CONVEX must deenergize the RSST due to an internal fault in jj the transformer.
Justification A - CORRECT: AOP-2560 has further guidance for preparing the plant for an approaching storm that is not covered in any other Operations procedure. Shutting down the plant is just one of the many preparatory actions that must be accomplished before the storm actually arrives.
B -WRONG: If Tavg deviates from program band by more than 10 O F , AOP-2575 requires a plant trip (assumes the operators have lost control of the plant). However, once the plant is tripped, RCS temperature is controlled by the Steam Dump System, which, with no stated problems, would make this point irrelevant.
C -WRONG: Once the plant is tripped, all CEAs are fully inserted. Therefore AOP-2556would no longer be of use.
D -WRONG: If the RSST is lost while performing EOP-2526the crew must transition to EOP-2528, Loss Of Forced Circulation. The A O P dealing electrical problems does not cover the required actions for Natural Circulation.
Reference NRC KIA SystemIEIA NRC KIA Generic System E02 Reactor Trip Recovery Number EK3.2 Knowledge of the reasons for normal, abnormal and emergency operating procedures associated with Reactor Trip Recovery as they apply to the Reactor Trip Recovery.
Importance ROlSRO 2.8 3.5 IOCFR Link (CFR: 41.5 I41 .IO,45.6,45.13)
Page 1 of 75
Question ID 6054015 0"g'n Mod, , ] r-Memory? (check=Yes)'
The plant has tripped from 100% power due to a Main Turbine trip The following conditions now exist
- A PORV is stuck open on the trip and its block valve is deenergized.
- RCS pressure has stabilized at -1200 psia.
- RVLMS has just dropped to 29%.
- Pressurizer level is 60% and rising.
- Both Steam Generators are at 900 psia and stable.
- The "C" HPSl pump has tripped on overload.
Which one of the following actions is required and sufficient to ensure adequate core cooling?
A The open PORV Block Valve must be re-energized and closed. 1 I... 1
,~
8 Start the "B" HPSl pump, aligned to Fac. 2, to restore HPSl capacity. l i C All available Charging Pumps must be running with Letdown secured. !4 1I D RCS pressure must be reduced by increasing RCS heat removal. p:
Justification D - CORRECT: This event is a small break LOCA. A cooldown will result in a decrease in RCS pressure allowing HPSl to inject.
A -WRONG: While this will stop the LOCA (loss of RCS inventory), it will likely take some time to re-energize the bus and close the block valve. The time it takes to re-energize the block valve may be long enough to cause core uncovery.
B -WRONG: a second centrifugal pump running in PARALLEL with the existing one will NOT improve HPSl capability. The problem is RCS pressure.
C -WRONG: Maximizing Charging with no letdown flow will NOT be adequate to restore inventory control. The stem states that RVLMS has just dropped to 29% with the 3 charging pumps running. The open PORV equates to the "worse case" size SB-LOCA for MP2, based on the continual loss of inventory being greater than charging pump capacity but lower than core heat removal requirements.
Reference EOP 2532, Lesson Plan, R1, C1, pg. 14 of 23:
D. Perform a Controlled Cooldown [EO# - RO-gB]
- 1. The cooldown should be initiated within 1hr of the event to conserve condensate inventory and comply with the long term cooling analysis. A minimum cooldown rate of 40"F/hr should be applied to ensure that at the 8 -
10hr point of the event boron precipitation conditions can be met. The starting point for the cooldown is the point at which the RCS stabilizes initially after the LOCA. Tc is used for monitoring RCS cooldown. However, if reflux boiling is occuring (occurs with a partially voided hot leg where steam leaves the core region, travels to the steam generator tubes, condenses, and the condensate flows back to the core via the hot leg) CETs should be utilized.
- 2. In the case of a SBLOCA that is inhibiting Safety Injection, cooldown initiation is extremely important.
Delaying the cooldown will prolong the duration of the LOCA with insufficient injection flow and thereby raise the potential for core damage due to voiding and inadequate core heat removal. In this case the operators must take actions to initiate a cooldown at the maximum rate as expeditiously as possible.
NRC WA System/E/A NRC WA Generic System 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)
Number AK1.02 Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: Change in leak rate with change in pressure I in portance ROlSRO 3.1 3.7 IOCFR Link (CFR41.8/41.10/45.3)
Page 2 of 75
- A small break LOCA is in progress.
- All ESAS equipment is operating as designed.
- Moments ago, the Crew commenced steaming both Steam Generators.
- Reactor vessel level has just dropped to 0%.
Both Cold Leg temperatures are 563 OF and stable.
Both Hot Leg temperatures are 564 "F and rising slowly.
RCS Pressure is 1170 psia and steady.
Which one of the following describe what these plant parameters indicate?
A The core is now uncovered with inadequate core cooling from any source. ii B The core is covered and being cooled by single phase natural circulation and SI flow. L c The core is covered and being cooled by two-phase natural circulation and SI flow. I D The core is covered and being cooled by reflux boiling and Safety Injection.
Justification D - CORRECT: The RCS is at saturation, indicating the core is still covered (Le.; not superheated). Although vessel level indicates 0%. this indication is still above the core.
A - WRONG: Safety injection flow and heater removal from the steam generators by way of reflux boiling is indicative of adequate inventory.
B -WRONG: Single phase natural circulation is NOT possible with 0% vessel level. The Steam Generator tube are NOT in contact with any water in the hot or cold legs.
C -WRONG: Two phase natural circulation is NOT possible with 0% vessel level. The Steam Generator tube are NOT in contact with any water in the hot or cold legs.
Reference Steam Tables and Generic Fundamentals NRC WA System/E/A NRC K/A Generic System 009 Small Break LOCA Number EK1.02 Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Use of steam tables Importance ROlSRO 3.5 4.2 IOCFR Link (CFR41.8/41.10/45.3)
Page 3 of 75
- , 4.- I Unit 2 was operating at 100% power when a large break LOCA occurred. Given the following events and conditions:
- RSST is unavailable, both EDGs started and loaded on the LNP.
- "A" Service Water Pump will not restart post-trip.
- "B"Service Water (SW) Pump remains secured.
- 24E is aligned to 24D.
- Unit 3 is in the process of checking availability of supplying power to 24E.
- "8" RBCCW Pump is being restored to service by Maintenance (PM's) and is due back within the hour.
- ALL other equipment is operating as designed.
Then, the "C" RBCCW Pump trips on overload. Maintenance is dispatched to determine the cause of the trip.
Which one of the following statements describes how the above equipment malfunctions will affect the mitigating strategy of this event?
A No effect; both trains of CTMT Spray are available to handle the initial energy release and one - j j ~.~
train of RBCCW and SW remain for long term cooling.
B The initial containment pressure spike will exceed design pressure due to the unavailability of the J CAR fans but long term heat removal will be adequate.
C Bus 24E must be energized from Unit 3 or, the "C" RBCCW Pump must be restored by the time 'Z SRAS occurs to ensure continued core cooling.
D To ensure continued core cooling, steps must be taken outside the EOPs, including the alignment of Facility 2 Service Water to Facility 1 RBCCW.
Justification C - CORRECT: The given conditions will result in a loss of all RBCCW. When SRAS occurs, RBCCW is required to provide flow to at least one SDC Heat Exchanger. If Bus 24E is energized from Unit 3, the "B" RBCCW Pump may be started to supply Facility 2. If the "C" RBCCW Pump is started, then Facility 2 is functional again.
A - WRONG: NO RBCCW is in operation for Heat Removal B -WRONG: CTMT Spray will prevent CTMT pressure from exceeding the design pressure. Additionally, long term cooling may not be available due to the loss of RBCCW needed for SRAS.
D - WRONG. It would be acceptable to crosstie facilities in order to restore RBCCW. However, since the "A' SW pump was lost following the LNP and the "8" SW pump is aligned to Facility 2 (24E is aligned to 24D), there was no cooling water available for the "A' EDG and it had to be emergency tripped. Therefore, there is no power for any Facility 1 equipment, including the RBCCW pumps.
Reference 1. ECC-01-C rev 3 page 19
- 2. ECCOI figure 6 (M2105-11-98)
- 3. CSS-00-C rev 4 chg 1 pages 6-8, 37
- 4. FSAR Figure 25203-3001 Main Single Line Diagram 1-10-92
- 5. RBC-00-C rev 5 pages 24-25
- 6. EOP 2532, Rev. 24, Step 38 NRC KIA SystemlElA NRC K/A Generic System 01 1 Large Break LOCA Number EA2.03 Ability to determine or interpret the following as they apply to a Large Break LOCA:
Consequences of managing LOCA with loss of ccw Importance ROlSRO 3.7 4.2 IOCFR Link (CFR 43.5 / 45.13)
Page 4 of 75
- 5 . 1 Question ID 6000001 I ]Memory7 (Ch&k=Yes)
The plant is operating normally at 100% power when high vibration alarms are received on the "B" RCP Based on the vibration level and procedure guidance, the US orders the following:
- Trip the Reactor
- Trip the Main Turbine
- Trip the "B" RCP "In that order" Which one of the following describes why the specific ORDER of the directed actions is so important?
A Tripping the Main Turbine before the RCP ensures the Main Turbine does NOT trip on overspeed due to loss of load when the RCP trips.
B 4
Tripping the reactor before the RCP avoids challenging core integrity due to e i $ b e + b s & e low p RCS flow with the reactor at 100% power.
c Tripping the Main Turbine between the reactor and the RCP ensures 25B has a chance to transfer ;'
to the RSST before the sudden power surge from the RCP loss takes out 25B on overload.
D Tripping the RCP after the reactor prevents formation of major flow oscillations in the core, d resulting in uneven changes in nucleate boiling and subsequently unacceptable radial flux distributions.
Jirstification B - CORRECT: for the reason stated A -WRONG: The Main Turbine will NOT trip on overspeed due to a loss of one RCP C -WRONG: The loss of one RCP will Not cause a large enough sudden power surge to cause the bus supply to trip on overload, locking out the 6.9 bus and the available condensate pump. The electrical system can handle the power surge caused by one RCP tripping.
D -WRONG: The loss of one RCP will NOT have a significant impact on core flow or cause unacceptable disturbances in the core.
Reference ARP 259OC-054, RCP HIGH VIBRATION, Rev. 000-04 (ARP ref. Gives actual sequence of required actions)
MANUAL Trip before AUTO Trip OP-2260; RPS ANTICIPATORY MANUAL TRIPS Approaching RPS Auto Setpoints or Various Turbine trips & efforts to regain control is unsuccessful.
NRC KIA System/E/A NRC K/A Generic System 015 Reactor Coolant Pump Malfunctions Number AK3.03 Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow) : Sequence of events for manually tripping reactor and RCP as a result of an RCP malfunction Importance ROlSRO 3.7 4.0 IOCFR Link (CFR 41.5,41.10 145.6 /45.13)
Page 5 of 75
The plant has been shutdown for refueling for approximately eight (8) days, with Fac. 1 Shutdown Cooling presently in service and all temperatures are stable. The "B" RBCCW header is secured, isolated and about to be drained to do a weld repair on the -25' level of the Aux. Building.
Approximately five (5) minutes after the Control Room is notified that personnel have begun draining the "B" RBCCW header, the crew notices that the "RCS return from SDC" temperature is beginning to rise with NO operator action.
Which of the following describes other indications that could be used to diagnose the probable CAUSE of t h e SDC temperature rise?
A RBCCW Surge Tank makeup valve opened in automatic; "A" header level 5% and dropping, "B" Li header level 50% and stable.
B The "A" RBCCW pump amps are beginning to fluctuate and the "A" header flow and pressure are .q, fluctuating and dropping.
C A PEO touring the "A" Safeguards Room reports hearing an unusually high noise level coming I-,
from "A' SDC Heat Exchanger.
D 2-SI-657, "SDC HX Flow Control Valve", is slowly opening and 2-SI-306, "SDC Total Flow Control ~
Valve", is slowly closing.
Justification B - CORRECT: Cavitating RBCCW pump conditions and loss of header flow and pressure are sure indications that the wrong RBCCW header is being drained, resulting in degraded performance of the in-service RBCCW pump and a loss of SDCIRHR heat sink.
A - WRONG: If a RBCCW header had been prepared for draining, the Surge Tank level must be lowered to
-33%, which is the level of the weir that separates the two facilities, not the normal 50% level. This would require the makeup valve be placed in manual/closed as it is preset to maintain 50% level in the surge tank.
VALID DISTRACTOR: because the applicant may expect the RBCCW Surge tank to indicate auto-makeup if the "A" header is accidentally drained, with the Makeup valve maintaining the level normal on the "E' header side based on the presence of the weir.
C -WRONG: This would be indicative of excessive SDC flow or pump noise being reflected through the system to the heat exchanger (largest load).
VALID DISTRACTOR: because the applicant may conclude the abnormal noise is caused by SDC flow automatically adjusting for the loss of heat sink or abnormal RBCCW pump noise being reflected into the largest heat exchanger on the system.
D -WRONG: SDC valve, as presently configured, will not auto-adjust for rising temperatures due to the degrading heat sink.
VALID DISTRACTOR: because the applicant may consider 2-SI-306 auto-response to changing SDC total flow and assume both SDC control valves will respond to changing demands on the SDC system. 241-657 can NOT be operated in automatic temperature control.
Reference Generic Fundamentals: indication of a pump losing NPSH (covered in Loss Of SDCIRHR Training Materials and AOP for the SDC Pump).
NRC KIA SystemlElA NRC KIA Generic System 025 Loss of Residual Heat Removal System (RHRS)
Number AA1.04 Ability to operate and / or monitor the following as they apply to the Loss of Residual Heat Removal System: Closed cooling water pumps Importance ROlSRO 2.8 2.6 IOCFR Link (CFR 41.7 I 4 5 . 5 145.6)
Page 6 of 75
Question ID. 650001 5 P ," ~ ..~,,
, ^I The reactor is shutdown, maintaining Hot Standby conditions.
The following indications for "A'RCP are noted by the PPO:
- Bleedoff temperature - 197 O F and rising
- Lower Seal temperature - 150 OF and stable
- Motor Stator Temperature - 210°F and slowly rising Given these conditions, identify the correct response related to conditions of "A" RCP.
A Trip pump required because Bleedoff Temperature exceeds the limit !/-
5 Trip pump NOT required because no RCP limits are exceeded.
C Trip pump because Lower Seal Temperature exceeds the limit.
D Trip pump because Motor Stator Temperature exceeds the limit L-:
Justification A CORRECT: ARP 25908-070, step 4, requires the pump to be tripped when the annunciator alarms. Also, AOP-2564, Loss Of RBCCW, step 3.3, bullet #6, Page 7 of 46, gives parameters to be monitored, and associated contingency actions required, if a parameter (temperature) is exceeded based on the loss of cooling water.
B -WRONG: Bleedoff temperature exceeded the limit for a pump trip.
C -WRONG: The Lower Seal Temperature did NOT exceed the required trip limit D -WRONG: The Stator Temperature did NOT exceed the required trip temperature.
Reference 1. ARP-2590B-070, RCP A BLEED-OFF TEMP HI, Rev. 000, Alarm setpoint is 180°F. Procedure requires a pump trip at 195°F
- 2. ARP-2590B-066, "RCP A STRTEMP HI", Rev. 000, Alarm setpoint is 260°F. Procedure requires a pump trip above 260°F
- 3. ARP-2590B-079, "RCP A LOWER SEAL TEMP HII", Rev. 000, Alarm setpoint is 170°F. Procedure reqiiiers a pump trip at 170°F rising 4 . AOP-2564, Loss Of RBCCW, step 3.3, bullet #6, Page 7 of 46, parameters to be monitored, and associated contingency actions required, on Bleedoff Flow high temp due to RBCCW loss.
NRC KIA System/E/A NRC K/A Generic System 026 Loss of Component Cooling Water (CCW)
Number AA2.04 Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The normal values and upper limits for the temperatures of the components cooled by CCW Importance ROlSRO 2.5 2.9 10CFR Link (CFR: 43.5 I45.13)
Page 7 of 75
The plant is at 83% power and stable, awaiting the repair and return of the "B" Condensate Pump.
Per OP 2204, the following other plant conditions exist:
- two (2) charging pumps are operating in manual (Mode Switch in LVL-1).
- Pressurizer backup heaters have been manually energized to force sprays
- PZR pressure has been stabilized at the normal operating pressure.
Then, unknown to the Crew, the in-service Pressurizer (PZR) pressure control transmitter fails AS-IS (output can NOT change).
Which one of the following describes the system response if Tcold were to NOW rise two (2) degrees due to a primary-secondary power mismatch?
A The running backup charging pump would stop. LI B PZR pressure and Spray flow would rise slightly. !f C The proportional heaters would go to minimum. 'LJ D Pressurizer level setpoint would rise to maximum. ii Jiistification B - CORRECT: PZR pressure would rise only slightly due to an insurge because while forcing PZR spray flow there is no "dead-band" for the PZR pressure controller. As soon as pressure started to rise, the pressure controller would immediately open the spray valves further and stop the rise at some value based on the controllers proportional band.
A -WRONG: Pressurizer level would Not rise enough to cause the backup Charging Pump to stop (3.6%
required).
C - WRONG: The Proportional Heaters are already at minimum due to forcing sprays D -WRONG: If the controller fails AS-IS, then the level controller will NOT see any change in actual level; therefore, the setpoint (or output) will NOT change.
Reference MSS-00-C, Main Steam System (Training Mat.), Rev. 6, Page 47 of 69.
PLPCS, Pressurizer Level and Pressure Control System (Training Mat.), Rev. 3, Page 41 of 61.
ARP 25908, A-39, Rev. 000-02, PRESSURIZER CH Y LEVEL HVLO.
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NRC KIA System/E/A NRC KIA Generic System 027 Pressurizer Pressure Control System (PZR PCS) Malfunction Number AK1.02 Knowledge of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions: Expansion of liquids as temperature increases Iin portance ROlSRO 2.8 3.1 I 10CFR Link (CFR41.8/41.10/45.3)
Page 8 of 75
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Question ID. 6000003
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The plant is operating at 100% power when a major earthquake causes a main feedwater pump to trip.
The US has ordered a plant trip on low Steam Generator level, but numerous malfunctions have prevented the crew from deenergizing the CEDS busses by any means available in the control room. All CEAs remain fully withdrawn.
Which of the following describes the required procedure action, or automatic system action, that would now occur to prevent core damage7 A The control operators will manually initiate boron injection immediately to shut down the reactor ' I.
B High RCS pressure will cause Aux. Feedwater to maintain a heat sink for the at-power reactor. IIj c An automatic main turbine trip on SG level will trigger an alternate means of tripping the reactor. ;A D The control operators will immediately initiate SlAS to start boron injection and a reactor shutdown. L!
Justification A - CORRECT: EOP 2525 requires the operator to manually initiate Emergency Boration on a plant trip where two or more CEAs do NOT fully insert.
B - WRONG: Auxiliary Feed will be automatically actuated by this condition; however, Auxiliary Feedwater flow is NOT adequate to remove heat from the RCS at greater than approximately 12% power. Failure of all CEAs to insert will result in power remaining much greater than 12%.
C - WRONG: A Reactor trip signal will be generated by the Turbine trip, but that signal is processed the same way as the Low SG Level trip and the manual trip, which has been stated in the stem to have failed; therefore, the Reactor will still remain at power.
D -WRONG: Initiation of SlAS will start Boron injection, but this is NOT a procedurally directed action. Initiating SIAS will immediately isolate letdown flow. This will compound the problem because there is not enough room in the PZR steam space at NOPlNOT to inject enough boric acid to adequately shutdown the reactor. Additionally, starting HPSl Pumps will NOT provide any additional Boric Acid flow until RCS pressure is reduced below HPSl shutoff head. This method of Boric Acid injection will NOT shut down the Reactor any faster than the manual initiation of Emergency Boration.
Reference EOP-2525; Immediate Actions, Rev. 20, Contingency for ensuring reactor is tripped EOP-2540A; Functional Revcovery, Reactivity Safety Function guidance for loss.
NRC KIA SystemlEIA NRC K/A Generic System 029 Anticipated Transient Without Scram (ATWS)
Number EK1.03 Knowledge of the operational implications of the following concepts as they apply to the ATWS:
Effects of boron on reactivity Importance ROlSRO 3.6 3.8 10CFR Link (CFR41.8/41.10/45.3)
Page 9 of 75
- IO, 1 P Question ID: 610
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The plant has tripped due to a loss of the electrical grid. On the trip, both facilities of Vital AC were repowered by their respective EDG. All other plant systems and components responded as designed.
Five (5) minutes into the performance of EOP-2525, a Steam Generator Tube Rupture (SGTR) occurred on #I SG. The Crew has now completed EOP-2525 and the US is about to direct the Board Operators to commence a cooldown per EOP-2534, SGTR.
Then, VA-IO deenergizes when it's main breaker fails open. PEOs are now being sent out in the plant to perform various tasks due to the loss of VA-10.
Which of the following local PEO tasks would require direct HP coverage due to ALARA concerns from the SGTR?
A Verification of the continued operation of the EDGs as designed. ._ I B Local-Manual operation of the "A'ADV during the plant cooldown. 2 c Local-Manual operation of the "A'Aux. Feedwater Reg. Valve. I .1 D Swapping the "B" charging pump from facility 1 to facility 2 power. 'J Justification B - CORRECT: Local-Manual operation of the " A ADV will result in above normal exposure to the PEO due to the radioactive steam of the # I SGTR.
A -WRONG: Although checking of the EDGs requires passing through the Aux. Building control point, the SGTR has no effect on background radiation in any of the areas or pathways utilized for check both EDGs.
C -WRONG: It is unlikely any contaminated steam made it into the Turbine Building with the plant tripping on a LOOP five minutes before the SGTR occurred. The SPO should have closed MSlVs before the SGTR due to the LNP. However, even if some amount of contaminated steam passed into the Turbine Building, the location of the
- I AFRV is not near any component that would now be an increased source of exposure.
D - WRONG: The "B" charging pump Kirk-Key interlock is located in the same area as the charging pumps, however, a SGTR does not effect the rad. Levels in this area like a LOCA or any type of event involving fuel damage would.
Reference OP-2316A (Main Steam System), Rev. 032-02, Page 10 of 106 E34-01-C (EOP-2534, SGTR Training Material), Rev. 2, Ch. 2; Concerns for SGTR NRC WA System/E/A NRC K/A Generic System 038 Steam Generator Tube Rupture Radiation Control (SGTR)
Number GA 2.3.2 SEE GENERIC K/A Knowledge of facility ALARA program linportance ROlSRO 2.5 2.9 IOCFR Link (CFR: 41.12/43.4. 45.9/45.10)
Page 10 of 75
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1 ' _Memory7 (Check=Yx The plant is at 100% power when it trips due to a loss of Battery Bus 201 B. On the trip, a large rupture occurs on the "6" Main Steam Header inside Containment.
Which of the following describes the plant system design, if any, and a response that will automatically mitigate this event?
A Aux. Feed Water requires power to actuate, so the loss of power to Facility 2 AFW will 0 automatically prevent AFW from feeding the ESD.
B The Turbine Battery will power VA-20 to ensure the #2 Main Feed Reg. Valve closes on a MSI 2 signal, preventing main feed flow to the ESD.
c All ESAS cabinets have dual power supplies to ensure both ESAS facilities are available to automatically actuate all necessary components.
u D This condition is beyond the design basis of the plant and there are NO system or plant i L1 components that will automatically mitigate the event.
Justification B - CORRECT The Turbine Battery is the back up power to VA-20 through Inverter 6. Maintaining VA-20 energized will allow the #2 Main Feed regulating Valve to automatically close on an MSI.
A -WRONG: The loss of Facility 2 DC power will cause the #2 Aux Feed Regulating Valve to fail open, and admit the maximum feed flow to the #2 Steam Generator.
C -WRONG: This statement is NOT true; however, The sensor cabinets do have dual power supplies; however, t h e actuation cabinets have one power supply. Additionally, this has nothing to do with mitigation of an ESD.
D -WRONG: This condition is NOT beyond design basis.
Reference CLP 125VDC1120VAC (TDC-00-C) Rev. 2, Chg. 1; Page 9 of 81 NRC KIA System/E/A NRC K/A Generic System E05 Excess Steam Demand Number EK2.1 Knowledge of the interrelations between the Excess Steam Demand and components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features.
Iin po rtance ROlSRO 3.3 3.6 10CFR Link (CFR: 41.7 / 45.7)
Page '1 1 of 75
- 12 1 Question ID ~ 0 0 0 10 ~
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The plant is at 100% power with the "B" Auxiliary Feedwater (AFW) pump out of service.
Then, the following plant events occur:
- The main feeder breaker on Vital DC distribution panel trips open on a bus fault, de-energizing DV-10.
- The plant trips on high Pressurizer pressure due to the closure of both MSIVs.
- On the plant trip, the "A" Main Steam Header ruptures in CTMT.
Which of the following actions are required to establish at least one Steam Generator as a controllable RCS heat sink?
A Ensure the "A" AFW pump is running, close 2-FW-44, close 2-FW-44, take manual control of #2 rj AFW Reg. Valve and feed only #2 SG.
B Ensure the "A'AFW pump is running and feed only #2 SG with it's AFW Reg. Valve in Local-Manual in the Turbine Building.
c Start the Turbine Driven AFW pump, place the # I AFW Reg. Valve in override and feed only #2 7 SG using the #2 AFW Reg. Valve D Start the Turbine Driven AFW pump, close 2-FW-44, take manual control of the #2 AFW Reg d' Valve and feed only the #2 SG.
Justification D - CORRECT; The TDAFP is the only AFW pump available once 2-FW-44 is closed. FW-44 isolates the discharge header to # I SG AFW Reg. Valve, which cannot be controlled from the control room and has failed open due to the loss of DV-IO.
A -WRONG; 150 gpm per SG is within the pump's capacity. but DV-10 supplies control power to the pump breaker.
B -WRONG; 150 gpm per SG is within the pump's capacity, but DV-10 supplies control power to the pump breaker, also this addresses the wrong side of R/v-44.
C -WRONG; Only in the fact that TDAFP is on the wrong side of FW-44 Reference AFW-00-C, Rev. 5, Ch. 3, Pages 11, 18 & 19 NRC KIA System/E/A NRC KIA Generic System 054 Loss of Main Feedwater (MFW)
Number AA1.02 Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW):
Manual startup of electric and steam-driven AFW pumps Importance RO/SRO 4.4 4.4 IOCFR Link (CFR 41.7 / 45.5 I45.6)
Page 12 of 75
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~ ~
The plant has tripped and suffered a total loss of ALL Vital AC. Vital AC is expected to be lost for at least three hours. Based on this expectation, the crew is working through the required actions of EOP-2530.
Which one of the following describes a required action, and the reason for the action's time restriction, per EOP 2530, Station Blackout?
A If Vital AC is not restored within 60 minutes, open PPC UPS breakers, preventing loss of PPC due c1 to battery backup power degradation.
B Within 60 minutes of commencing a cooldown, and every 50 O F thereafter, sample the RCS for ,~.;
Boron to ensure Shutdown Margin is met.
c If Vital AC is not restored within 30 minutes, align supplemental cooling to all vital areas due to :v:'
loss of ventilation and ventilation cooling.
D If Vital AC is not restored within 30 minutes, ensure backup air is aligned to applicable valves to 0 preclude loss of valve position control.
Justification C - CORRECT: The requirement to align supplemental cooling is to limit the heat loading in the Vital DC Switchgear Rooms. This requirement is one of the most important time-critical operator actions in 2530 as it could result in the loss of ALL remaining Safety Channels of instrumentation.
A -WRONG: When the PPC is lost due to slow voltage degradation, the first indication is "BAD" or unjustified indications displayed by parameters monitored by the PPC. The danger here is operators may respond to these indications, not knowing the PPC is slowly going "brain dead". Also, the PPC does NOT "die" well when it is lost due to a slow power supply degradation. It will take MUCH longer to restore when lost in this fashion.
B - WRONG: There is a requirement in 2530 to verify RCS boron concentration meets shutdown margin every 50 "F on a cooldown, but the concentration MUST be verified BEFORE the cooldown is started and for the NEXT 50 "F temperature.
D -WRONG: The backup air available to certain valves is designed to assist in mitigating a plant accident (i.e.;
LOOF) combined with a loss of Instrument Air. The Station Blackout event assumes no other plant events require mitigation, except for those directly attributed to the lack of Vital AC.
Reference CLP Station Blackout (E30-01-C) Rev. 3, Page 15 of 22 NRC KIA System/E/A NRC K/A Generic System 055 Loss of Offsite and Onsite Power (Station Blackout)
Number EK3.02 Knowledge of the reasons for the following responses as the apply to the Station Blackout:
Actions contained in EOP for loss of offsite and onsite power Importance R O E RO 4.3 4.6 10CFR Link (CFR41.5/41.10/45.6/45.13)
Page 13 of 75
- 14, 1 The plant is at 55% power in preparation for restoring the "A" Main Feedwater Pump to service The following additional plant conditions exist:
- "C" charging pump is operating as the "running" pump.
- " A ' charging pump is running in manual (Mode Switch in LVL-1).
- "B" charging Pump is aligned to Facility Two.
- All Pressurizer backup heaters have been manually energized to force Pressurizer sprays
- Channel "X" PZR Level and Pressure Control are in service.
- All plant parameters are stable.
Then, the plant trips when Vital DC bus 201 6 deenergizes due to a faulty supply breaker.
The Crew is performing steps in EOP-2525, but has NOT taken any actions OUTSIDE of the Control Room.
Which one of the following describes an expected plant status, and why, as a result of the loss of power and subsequent plant trip?
A Letdown will be isolated (auto) and at least two charging pumps will be running, causing a c continuous rise in Pressurizer level and pressure.
B VR-21 will also be deenergized preventing the operation of any charging pumps. Pressurizer level j:
will be dropping continuously due to this loss.
C Ten minutes post-trip, RCS pressure will be slowly rising due to only one bank of proportional v heaters and one charging pump being available.
D VA-20 will also deenergize causing Ch. "Y" Pressurizer level to fail low and trip all Pressurizer 1 --I L
heaters. This will cause RCS pressure to drop.
Justification C - CORRECT: The loss of 201 B causes a complete loss of Fac. 2 AC power, including 25B, 248, 24D, 22F and VR-21. When VR-21 deenergizes. the High PZR Pressure interlock for the backup heaters fails in the "triggered" mode, tripping ALL backup heater breakers and preventing them from being used. Even though two groups of backup heaters have energized busses per the given scenario, they are still deenergized by a "failure" in there control circuit and can't be energized as long as VR-21 is dead. With only one Facility of Vital AC available (due to the described loss of DC), this leaves only one banklgroup of PZR proportional heaters and available to maintain pressure. Ten minutes post-trip, the PZR will still be "recovering" with only one group of proportional heaters and one charging pump.
A - WRONG: Letdown should not isolate automatically because VR-11 was NOT deenergized and "A" charging pump never stopped (it never lost power and its control circuit also remained energized. However, only one charging pump is available until "B" charging pump power supply is LOCALLY aligned to Fac. One.
B -WRONG: If VR-11 were deenergized all available (powered) charging pumps would be operating due to a power loss to part of their control circuit. If Channel "Y" LEVEL control were in service, the loss of power would deenergize it requiring manual control of the available charging pump.
D -WRONG: VA-20 will swap to its alternate power source, INV-5, which is powered by the Turbine Battery and unaffected. This power supply swap is a "make-before-break and, therefore, there will NOT be a momentary loss of power to Ch. "Y" PZR level transmitter. This will prevent the heaters from tripping on a "failed" LO-LO LEVEL trigger.
Reference AOP-2504B, Rev. 003-04, Loss of VR-21; Page 4 of 52 - "Discussion" section.
NRC KlA SystemlElA NRC KIA Generic System 056 Loss of Offsite Power Number AA2.73 Ability to determine and interpret the following as they apply to the Loss of Offsite Power: PZR heater on/off Importance ROlSRO 3.5 3.6 IOCFR Link (CFR: 43.5 145.13)
Page 14 of 75
- I5 1 With the plant in Mode 5 during a refueling outage with the following conditions:
VA-10 has been placed on it's ALTERNATE source for electrical PM's on Inverter 1 (INV-I). The PMs have been completed and Inverter 1 and its Static Switch are presently being restored to operation with the following conditions existing:
- The "AUTO/MAN" switch inside the Inverter 1 (INV-1) cabinet is in the MANUAL position.
- The "SYNC" switch on the Static Switch is in the ON position.
- The Inverter is energized with all input and output breakers closed.
- Voltage of the two sources are approximately equal.
- Both sources in synch.
Then, INV-5 trips off line due to an internal fault.
Which one of the following describes the status of, or the required actions for, VA-IO?
A When INV-5 deenergized, static switch # I "auto" transferred VA-10 to its normal power source.
8 INV-1 must be completely deenergized (shutdown) and restarted in order to reenergize VA-10. 7
~~
c The "SYNC" switch on # I static switch must be placed in the "OFF" position to recover VA-10.
7 D The "AUTO/MAN" switch inside INV-1 must be placed in the "AUTO" position to recover VA-IO. p
~ ~~ ~~ ~~ ~ ~~~~ ~~~ ~ ~ ~ ~ ~ . ~.~~ ~ ~~ ~ ~-
Justification D - CORRECT: The nomenclature of the "AUTO/MAN" switch inside INV-1 is similar to the "AUTOIMAN" switch OUTSIDE the INV-1 cabinet on the Static Switch. However, their functions are quite different. With the INV-1 switch specified in the stem in the applicable position, the Static Switch will NOT transfer to the Alternate power supply and VA-10 will be deenergized.
A -WRONG: The transfer is blocked from happening until the above mentioned switch is repositioned B - WRONG: Shutting down the inverter completely is NOT required, but if done correctly, would correct the switch misalignment and reenergize VA-IO.
C -WRONG: Placing the "SYNC" switch to ON is one of the actions that triggered the actual event that occurred at MP2.
NOTE: This question is similar to an actual scenario that occurred at Millstone Unit 2 and resulted in a loss of RHR and an RCS temperature rise of >lO°F (Classifiable event).
Reference CLP 125VDC/120VAC (LVD-00-C) Rev. 5, Chg. 1; Page 26 of 81 NRC KIA SystemlEIA NRC KIA Generic System 057 Loss of Vital AC Electrical Instrument Bus Number AA1.01 Ability to operate and / or monitor the following as they apply to the Loss of Vital AC Instrument Bus:
Manual inverter swapping Importance ROlSRO 3.7 3.7 10CFR Link (CFR 41.7 145.5 145.6)
Page 15 of 75
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- ., 16, I , Memory? (Ch&k=Yes)
The plant has tripped from 100% power with the following conditions'
- ALL electrical components and busses were aligned for normal operation.
- Loss of off-site power (cause of plant trip).
- On the trip, the "B" Battery Bus tie to 201B trips open due to a breaker failure
- All other plant components and systems are operating as designed.
Which of the following describes indications that would result from the above loss of electrical busses?
A 201A indicates: 134 Battery Volts, -5 Battery Amps and +I 5 Bus Amps.
201 B indicates: 0 Battery Volts, 0 Battery Amps and 0 Bus Amps.
8 201A indicates: 126 Battery Volts, +20 Battery Amps and + I 5 Bus Amps. . .-
I 3
201 B indicates: 126 Battery Volts, +20 Battery Amps and +20 Bus Amps.
c 201A indicates: 134 Battery Volts, -5 Battery Amps and + I 5 Bus Amps.
201B indicates: 134 Battery Volts, 0 Battery Amps and 0 Bus Amps.
D 201A indicates: 134 Battery Volts, -5 Battery Amps and + I 5 Bus Amps.
201B indicates: 126 Battery Volts, -5 Battery Amps and + I 5 Bus Amps.
Justification C - CORRECT:
201A indicates: 134volts indicates a battery on float charge with an indicated 5 amp float and 15 amps worth of load on the bus (normal state);
2018 indicates: 134volts indicates a battery on float charge with an indicated 0 amp current flow (battery is isolated) and 0 amps of load on the bus (Bus is isolated from any power source).
A -WRONG:
201A indicates: 134volts indicates a battery on float charge with an indicated 5 amp float and 15 amps worth of load on the bus (normal state);
201B indicates: 0 battery volts would indicate the battery bus "fuses" have blown in the disconnect B -WRONG: 126 volts is the low voltage alarm setpoint and would indicate a battery in being drained due to a lack of an energized battery charger, NOT disconnected from the load bus. This would be correct if it was believed the battery chargers have been lost due to the breaker failure or LOOPILNP.
D -WRONG: This has the expected indication for ZOIA, but indicates 201B is still connected and being discharged at a normal rate.
Reference CLP 125VDC/IZOVAC (LVD-00-C) Rev. 5, Chg. 1; Page 14 of 81 NRC K/A System/E/A NRC K/A Generic System 058 Loss of DC Power Number AKI .01 Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation Iin portance ROlSRO 2.8 3.1 IOCFR Link (CFR 41.8 /41.10/45.3)
Page 16 of 75
I
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I
' -'Memory? (Check=Yes)
- 17,,
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The plant is operating in MODE 4 shutting down for refueling. The following plant conditions exist:
- "A" & "B" RCPs are operating.
- SDC is presently operating in "Warm-up Recirc" with "SDC to RCS temperature, T351Y" at 125°F and rising.
- RCS pressure is presently 255 psia and stable..
Then, the following control board indications are received:
- RBCCW Surge Tank level is 55% and rising.
- RBCCW Radiation Monitor indication is stable (NOT in alarm).
Which of the following conditions would result in the above parameters?
A Heat input from " A ' Shutdown Cooling (SDC) Heat Exchanger B A leak from the Spent Fuel Pool (SFP) Cooling Heat Exchanger II! 1 c A leak from the "A' Reactor Coolant Pump (RCP) Seal Cooler I
_-I i
D Mussel clogging of the Facility One ( I ) RBCCW heat exchanger. 4 Justification D: CORRECT: Mussel clogging would cause a loss of Service Water flow through the HX (loss of heat sink). In the closed RBCCW System the resulting temperature rise would the surge tank level to go up.
A: WRONG: This would cause the above conditions if the SDC HX was aligned for RBCCW flow. During 'Warm-up Recirc" RBCCW is not aligned.
B: WRONG: The SFP cooling system pressure is too low to leak into RBCCW system.
C: WRONG: A leak from a RCP seal cooler would result in RBCCW radiation monitor rise/alarm.
Reference ARP-2590E-048, Rev. 000, "RBCCW HX TEMP HI", Step 6; references high temp. caused by HX fouling NRC KIA System/E/A NRC K/A Generic System 062 Loss of Nuclear Service Water Number A A I .05 Ability to operate and / or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): The CCWS surge tank, including level control and level alarms, and radiation alarm Im porta nce ROlSRO 3.1 3.1 IOCFR Link (CFR 41.7 I45.5 I45.6)
Page 17 of 75
- , 1,8 1 - -
Question ID 6000009 _. . - (Check=Yes)
'gMemory?
The plant has tripped from 100% power due to an air leak in the Turbine Building. The crew has transitioned to EOP-2526 and are referencing AOP-2563, Loss Of Instrument Air, for guidance on system restoration. All plant conditions are stable.
Analysis of the leak location using the P&IDs reveals that the Unit 3 air system cross tie could supply the temporarily isolated systems and allow for return of remote operation from the Control Room. However, the ILoss of Instrument Air AOP does NOT provide steps for restoration of air once the leak is isolated.
Which of the following describes an administrative requirement that must be met in order to allow the restoration of the Instrument Air System.
A The "on-watch" Unit 2 SM or US must declare entry into 10-CFR-50.54X in order to deviate from ,J the applicable AOP.
B Only Technical Support or Plant Management can authorize this temporary change as it applies to ,I an AOP and not an NOP.
c Senior Licensed personnel from Unit 2 and Unit 3 must approve the temporary procedure change [J to the Loss of Instrument Air AOP.
D An "on-watch" Unit 2 Senior License must review/approve the temporary procedure change to the 2 Loss of Instrument Air AOP.
Justification D - CORRECT: This is considered a "Temporary Procedure Change", therefore all aspects of the change require approval by the "on-watch SM or US in the applicable unit.
A -WRONG: 50.54X can only be implemented if emergency conditions exist, and in order to protect the general public.
B -WRONG: These requirements are only applicable if restoring the plant in a post accident condition C -WRONG: SRO review/approval from both units not applicable even though cross-tie affects both units.
Reference Minor Procedure Revision; MP-05-MMM, Page 22
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NRC KIA System/E/A NRC KIA Generic System 065 Loss of Instrument Air Equipment Control Number GA 2 2.11 SEE GENERIC WA Knowledge of the process for controlling tern porary changes.
Importance RO/SRO 2.5 3.4 IOCFR Link (CFR: 41.10 / 43.3 / 45.13)
Page 18 of 75
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I iMemory7
=_ (Check=Yes)-
- I9 The plant IS in normal operation at 100% power, when the Ch "Y" Pressurizer (PZR) Level Controller, LIC-11OY, began to operate erratically The PPO has selected Ch "X" as the controlling channel of PZR level per the applicable ARP and the PZR IS now stable at NORMAL, 100% power parameters.
I&C IS now working in the applicable Spec-200 cabinet, troubleshooting the LIC-11OY circuit and comparing it's output values to that of LIC-11OX circuit Then, while I&C is troubleshooting PZR Level Control, the PPO notices the following plant indications
- PZR Level Ch. "X" PPC indications is slowly rising.
- PZR Level Ch. "Y" PPC indications is slowly lowering.
- PZR Level Controller LIC-11OX "red" needles (vertical scale) is slowly rising.
- PZR Level Controller LIC-11OY "red" needles (vertical scale) is slowly lowering.
- PZR Level Controller LIC-11OX "black" needle (horizontal scale) is rising.
- PZR Level Controller LIC-11OY "black" needle (horizontal scale) is stable (NOT changing).
- Letdown flow is slowly rising.
- ALL Safety and Control channel indications of PZR Pressure are slowly lowering.
- RCS temperature is stable at 100% power values.
Which of the following describes the malfunction and some of the required actions to be taken, per the applicable ARP, for PZR Level Control?
A Ch. "X" PZR Level TRANSMITTER is failing; the bias output on the letdown flow controller must be lowered to stabilize Pressurizer parameters.
B Ch. "X" PZR Level TRANSMITTER is failing; LIC-11OX must be placed in MANUAL and the output :V lowered to stabilize Pressurizer parameters.
c Ch. "X" PZR Level CONTROLLER is failing, swap level control to LIC-1 IOY, in AUTOMATIC, and 1
'1 ensure PZR parameters stabilize at normal.
D Ch. "X" PZR Level CONTROLLER is failing, swap level control to LIC-1 IOY, in MANUAL and lower '7 the output to stabilize Pressurizer parameters.
Jiistification B -CORRECT: Pressurizer level is lowering because the Ch. "X" PZR Level TRANSMITTER is failing high.
Therefore, LIC-11OX must be placed in MANUAL and the output lowered to stabilize Pressurizer parameters.
A -WRONG: Lowering the bias will not restore PZR. Level as the Chg. Pumps are controlled from the level controller.
C -WRONG: The transmitter is failing not the controller. Also, LIC-I 1OY is not responding to the rising input and therefore must be considered inop due to I&C troubleshooting.
D -WRONG: Per the stem, LIC-11OY is not responding and therefore considered unavailable Reference PLC-01-C, Rev. 3, Pg. 31 of 65 NRC KIA SystemlElA NRC KIA Generic System 028 Pressurizer (PZR) Level Control Malfunction Number AA2.10 Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: Whether the automatic mode for PZR level control is functioning improperly, necessity of shift to manual modes Importance ROlSRO 3.3 3.4 10CFR Link (CFR: 43.5 I45.13)
Page 19 of 75
Question ID. 6000020 1- 4 Memory?
--- - - --- ~-
(Check=Yes)
Then, the high voltage power supply to the Channel "A" Wide Range Nuclear Instruments fails, such that the channel is now reading eight (8) decades lower.
Which of the following describes the change in planffcomponent conditions due to this power supply fai Iu re?
A The channel indication on C04 will swap from "% Power" to "CPS". F B The Power Trip Test Interlock (PTTI) for the channel is now armed. u c The Dropped CEA annunciator will alarm on C04 and this channel i D The Level 1/2 Bistables are bypassing the applicable interlocks. I1 Justification A - CORRECT: The Wide Range Nl's auto-swap .indication from "%-Power" to "CPS" when the output reaches 1000 counts. If the detector is now 10E8 less sensitive, it should be reading in the source range, or CPS.
B -WRONG: The PTTl interlock would indeed be armed for this channel, IF this were the Linear Channel detector power supply that failed.
C -WRONG: The "Dropped C E A alarm would activate on this failure, IF it were the Safety Nl's not the Wide Range Detectors.
D - WRONG: The Level 1/2 bistables will activate on this channel, but ALL four channels must activate for the applicable interlocks to be affected.
Reference NIS-014, Rev. 3, Ch. 2, Pg. 15 of 57, Pg 31 of 57 ARP 259OC-092, Rev. 000 NRC KIA System/E/A NRC WA Generic System 032 Loss of Source Range Nuclear Instrumentation Number AK2.01 Knowledge of the interrelations between the Loss of Source Range Nuclear Instrumentation and the following: Power supplies, including proper switch positions Im po rtance ROlSRO 2.7 3.1 10CFR Link (CFR 41.7 /45.7)
Page 20 of 75
Question ID. 6100044 I-flMemory7 (Check=Yes) '
A reactor startup IS being performed with all systems and components operating as designed Then, with Group 2 Regulating CEAs at 70 steps withdrawn, 'A' wide range NI channel fails HIGH Which one of the following describes the required actions and why?
A The startup must be suspended because any further CEA withdrawal would require bypassing and making inoperable the CMI interlock.
B The Startup must be suspended because continued CEA withdrawal is not allowed by Tech. LI Specs. with only three (3) wide range detectors.
C The startup may continue once the remaining wide range detectors are verified operable by -I
-J performing the applicable Tech. Spec. surveillance.
D The startup may continue without pause if the remaining three (3) wide range detectors have been 3 monitored as trending normally.
~ ~ ~ ~
Justification A - CORRECT: The maximum CEA insertion limit is automatically bypassed when all four wide range NI channels sense reactor power to be below 1E-4%. This allows CEAs to be withdrawn for a reactor startup without the need to bypass the CMI circuit, which is required by Tech. Specs. To be operable before entering Mode 2.
However, when the NI channel failed high, it caused the PDlL interlock associated with CEAPDS to become "live". Because a startup is in progress, the CEAs are not yet above the minimum PDIL (Group 4 above 72 steps) and would, therefore, immediately trigger a PDIL - CEA Motion Inhibit (CMI). The only way to continue moving CEAs in any direction at that point would be to bypass the CMI.
B -WRONG: Tech. Specs allows the reactor to be started up with only two wide range NI channels operable, but the Reactor Startup procedure requires all four NI channels be operable.
C -WRONG: Performing a Tech. Spec. surveillance on the alternate facilities to prove their operability is a requirement of some Tech. Specs. To allow continued operation. but not the one for the Wide Range Channels D - WRONG: This is allowed by Tech. Specs., but not by the Reactor Startup procedure, OP-2202.
Reference CLP CED-01-C [Control Element Drive System]; page 32 of 67.
NRC KIA SystemIElA NRC KIA Generic System 033 Loss of Intermediate Range Nuclear Instrumentation Number AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation:
Termination of startup following loss of intermediate- range instrumentation Importance ROlSRO 3.2 3.6 IOCFR Link (CFR 41.5,41.10/45.6/45.13)
Page 21 of 75
Memory? (Check=Yes)
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A PEO has just reported the start of a Clean Liquid Radwaste (CLR) discharge to Long Island Sound Five (5) minutes later, the SPO notices the Unit Two stack radiation monitor is going up Which one of the following system or component conditions would cause this rise in the stack radiation detector?
A The CLR Monitor Tank Pump recirculation Valve has been left open while discharging the tank. 7 B The CLR Receiver Tank, NOT in service, is left aligned to the Monitor Tank that is discharging. 2 c The CLR discharge filter is plugged causing the Monitor Tank Pump discharge relief valve to lift. !_J D The CLR Monitor Tank Pump seal is leaking several gallons per minute onto the tank room floor. 2; Justification D - CORRECT: As the CLR system is "sealed" from the atmosphere due to entrained RCS gas, Pump seals leaking on the CLR Monitor Tank pump would "vent" contaminated gasses to the Aux. Bldg.
A -WRONG: The CLR Monitor Tank CANNOT be aligned for discharge, due to system interlocks, if the "recirc."
valve is opened.
B -WRONG: The CLR Monitor Tank CANNOT be aligned for discharge, due to system interlocks, if a CLR Receiver Tank is still aligned to the Monitor Tank.
C -WRONG: The CLR discharge relief just recircs. back to the Monitor Tank, therefore should not be a source of gas release.
Reference ALR-04-C, Rev. 3, Figure 1 (Rev 4)
NRC KIA System/E/A NRC KIA Generic System 059 Accidental Liquid Radwaste Release Number AK2.02 Knowledge of the interrelations between the Accidental Liquid Radwaste Release and the following: Radioactive-gas monitors Importance ROlSRO 2.7 2.7 10CFR Link (CFR 41.7 / 45.7)
Page 22 of 75
-I _-
- ??,, , I Y Question ID. 6000007 I .o.p;-lw;;.
, ~-
4Memory7 (Check=Yes)'
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The unit is at 100% power, when an Engineering Analysis determines both Containment High Range Radiation Monitors are inoperable. The applicable Technical Specification LCO has been entered.
Due to the loss of the CTMT radiation monitors, the Unit Supervisor has directed the PPO to monitor the Main Steam Line Radiation Monitor as an alternate indication, in the event a major accident occurs causing high radiation in containment.
Which one of the following describes a limitation of using this alternate means of monitoring CTMT radiation level?
A Alternate indication will only respond if Steam Generator Tube Leakage/Rupture is present. -_
B Alternate indication will be several decades below the real containment radiation level. 'Z c Alternate indication will only respond if the unit is at power and N-16 is present in the RCS. El D Alternate indication will be unusable in the event of a Main Steam Isolation (MSI) actuation. ir Justification B - CORRECT: Although the MS line Rad. Monitors are designed to monitor N-16, they would be sensitive to any form of radiation because they are simply a very sensitive radiation detector positioned in direct contact with the main steam line. However, although they are more sensitive, due to their position OUTSIDE CTMT the detectors would "see" the equivalent radiation seen by the CTMT High Range detectors as a much lower than actual radiation field. The CTMT wall would attenuate much of the radiation preventing the MS line detectors from getting a true reading.
A - WRONG: This is what the steam line detectors were DESIGNED to detect, but, based on their design, it is not the only thing they CAN detect.
C -WRONG: The MS line detectors were designed to detect N-16, but due to their design would pick up radiation from any source that strikes them..
D -WRONG: This implies the detector sensitivity is the only factor to be considered in using these detectors for the purpose stated in the stem.
Reference OP-23838, Rev. 008-04, Note on Page 6 MP2 Radiation Monitor Manual (Monman2.doc), Page 34 of 98, "Purpose of Main Steam Line radiation monitors (last sentence).
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NRC K/A System/E/A NRC WA Generic System 061 Area Radiation Monitoring (ARM)
System Alarms Number AKI 01 Knowledge of the operational implications of the following concepts as they apply to Area Radiation Monitoring (ARM) System Alarms: Detector limitations Importance ROlSRO 2.5 2.9 IOCFR Link CFR41.8/41.10/45.3)
Page 23 of 75
- 2A I The plant is at 100% power when a fire breaks out in the 25' Cable Vault
. Smoke is beginning to fill the Control Room and the crew has just noticed that the Plant Process Computer has STOPPED updating any data.
The SM then immediately directs the following crew actions per AOP-2579A, Appendix "R" Fire:
I)Trip the reactor and verify it is fully tripped
- 2) Then, all personnel immediately evacuate the Control Room Considering the plant conditions, why are these the ONLY crew actions directed to be taken before evacuating the Control Room?
A If two (2) or more CEAs do not fully insert, Emergency Boration MUST first be initiated from the Control Room.
B There is NO indication that all the CEAs fully inserted outside of the Control Room, when the E reactor was tripped.
C If the reactor is tripped from outside the Control Room, the Main Turbine would then trip BEFORE 1 the reactor.
D There is NO way to start a backup charging pump for Emergency Boration if the running pump Lj trips on a hot short.
Justification B - CORRECT: The ONLY way to verify the CEAs are fully inserted, without the PPC, is by the Core Mimic or CEAPDS. Both can be monitored only in the Control Room. Although it may be possible to "see" the indication through the rear door window of the Control Room, if the room were filled with smoke this would not be possible.
A -WRONG: App. "R" procedures have directions for establishing Emergency Boration from outside the Control Room.
C -WRONG. When the reactor is tripped manually, from ANY location, the Turbine is tripped automatically by the loss of power to the CEA buss undervoltage relays. App. "R" does have steps to locally trip the Main Turbine, and this could be directed before the reactor is tripped, but it would be a violation of the procedure sequence and Operational Guidelines.
D -WRONG: It is still possible to take "local" control of the Fac. 2 charging pump@) at C-I 0 (Fire S/D Panel) and run it from there even if it tripped on a "hot short". There are, however, NO controls available outside the control room to start any "backup" charging pumps.
Reference AOP-2579A, Rev. 9-04, Pg 6 of 64, Step 5a
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NRC KIA SystemlEiA NRC KIA Generic System 067 Plant fire on site Number AK3.02 Knowledge of the reasons for the following responses as they apply to the Plant Fire on Site:
Steps called out in the site fire protection plan, FPS manual, and fire zone manual Importance ROlSRO 2.5 3.3 IOCFR Link (CFR 41.5,41.10/45.6 /45.13)
Page 24 of 75
- 25 I The plant has tripped and EOP 2525 is being performed due to a large SGTR on #I SG. On the trip, the RSST was lost due to grid instabilities.
All other systems responded normally with the following actions having been taken thus far:
- All initial and subsequent actions of EOP-2525 have been completed.
- The PPO has verified Natural Circulation establishing and SIAS/CIAS/EBFAS manually actuated on low RCS pressure.
- The SPO has closed both MSIVs, broke condenser vacuum and stabilized RCS temperature using both Atmospheric Dump Valves (ADVs).
- The PPO and SPO are now awaiting diagnostic queries from the US.
Then, a fuel pin suddenly ruptures due to a manufacturing defect.
Which one of the following radiation monitors would indicate the fuel failure under present plant conditions?
A Containment Refueling Bridge Area and the Containment High Range radiation monitors. -J B Steam Jet Air Ejector and the Steam Generator Blowdown radiation monitors. '13 c The Main Steam Line radiation monitors for the " A ' main steam header and "A" ADV. 3 r- -
D The Main Steam Line radiation monitors for both main steam headers and both ADVs I -!
Justification C - CORRECT: Because the "A" ADV taps off the steam header very close to CTMT, it has it's own Rad. Monitor (unlike the "B" ADV). Ordinarily, a fuel failure would show up on several Rad. Monitors, but not with a SGTR event affecting it and a SIASICIAS actuation.
A -WRONG: Location of RMs inside CTMT would require significant clad failure for alarm to come in B -WRONG: SGBD sampling flow path was isolated on the ClAS and the SJAE has no steam supply D -WRONG: The "B" MSL Rad. Monitor only sees the "B" main steam header, there is no "shine" from CTMT with a SGTR and the rad monitors are UPSTREAM of where the two headers mix. Also, there is no rad. Monitor for the "E"ADV, only the "A'ADV, due to the "A"sproximity to the CTMT wall.
Reference ARP-2590A-117, Rev. 000, Pg. 1 of 1 NRC KIA System/E/A NRC KIA Generic System 076 High Reactor Coolant Activity Number AK2.01 Knowledge of the interrelations between the High Reactor Coolant Activity and the following:
Process radiation monitors Importance ROlSRO 2.6 3.0 IOCFR Link (CFR 41.7 / 45.7)
Page 25 of 75
,Oyg,In,.z; I ' 'JMemGy7 (Check=Yesr
-~
The plant tripped from 100%. On the trip both Bus 25A and B u s 25B failed to transfer to the RSST The difference between Thot and CET temperatures is found to be 15°F Which one of the following describes the operator actions that must be taken to mitigate this condition in accordance with EOP 2528, Loss of Offsite Power/Loss of Forced Circulation?
A Stop any cooldown and ensure both steam generator saturation temperatures are within 10°F. 13 B Energize all available Pressurizer heaters and ensure RCS subcooled per highest Thot.
I I
c Stop any RCS depressurization and begin venting the reactor head using the installed head vents.
D Ensure steam generator (SG) pressure is less than saturation pressure for existing Tcold by 2 steaming the SGs.
~-
Justification D - CORRECT This IS the guidance given in the procedure, as it indicates not enough heat IS being withdrawn from the RCS by the SGs This will soon result in stalled NC flow, indicative of the larger than expected delta-T between CETs and Thot A - WRONG: This is the course of action when the two RCS loops become "uncoupled" due to uneven steam generator heat removal.
B -WRONG: This implies the abnormal delta-T is due to void formation, but the void would have to reach the top of the core and effect both detectors.
C -WRONG: This implies noncondensible gasses have come out of solution and stagnated RCS flow, resulting in the abnormal delta-T.
Reference EOP-2528, Rev. 15,Pg. 8 of 36, Step 11, Contingency NRC KIA SystemlEIA NRC K/A Generic System A I 3 Natural Circulation Operations Number AK1.3 Knowledge of the operational implications of annunciators and conditions indicating signals and remedial actions associated with the Natural Circulation Operations as they apply to the Natural Circulation Operations.
Importance ROlSRO 3.1 3.4 10CFR Link (CFR: 41.8/41.10/45.3)
Page 26 of 75
- 27, 1 L _ '<Memory? (Check=Yes)
._.____pi A plant startup IS in progress at End-Of-Life following an uncomplicated plant trip Power is being held at 1% while operators prepare to start the first Main Feedwater Pump.
Then, a major plant accident occurs with the following conditions being noted
- "A" Steam Generator level is 200" and dropping rapidly.
- "B" Steam Generator level is 15% and dropping slowly.
- "A" Steam Generator pressure is 350 psia and dropping rapidly.
- "6" Steam Generator pressure is 740 psia and dropping slowly.
- Pressurizer pressure is 1380 psia and dropping.
- Pressurizer level is off scale low.
- RCS temperature is 460°F and dropping.
- CTMT pressure is 28 psig and rising.
- No rad monitors are rising or in alarm.
Which one of the following describe administrative limits that are in place to prevent this accident from exceeding design bases?
A The Moderator Temperature Coefficient has a maximum, allowable negative limit when critical. 3 5 The steam generators have a maximum allowable water level when operating in Mode 2. El c The combination of ASI, Tcold, Q-power and R C S pressure are within limits above the POAH. Ji D The maximum amount of CEA insertion allowable for the given power level of the reactor.
Justification A - CORRECT: Limiting the maximum negative value of MTC ensures the safety analysis remains valid for limiting the return to power following an ESD event and the subsequent cooldown.
B - WRONG: Although the amount of mass in the SG will certainly have an impact on the magnitude of the ESD event, and the maximum allowable SG water level is administratively limited by plant procedures, this limit is based on protecting plant systems and equipment.
C -WRONG: This basically describes the TMlLP limit function, which, although certainly affected by an ESD, is not designed to protect the plantlcore in such an event.
D -WRONG: This defines the PDIL, which will be affected as plant power changes with the ESD. Also, the PDlL is one of the foundation assumptions used in the design of various LSSS. However, the bases for the PDIL does not encompass any plant transients bounded by an ESD event.
Reference Tech. Spec. Bases 3/4, Page 1-1; Shutdown Margin (ITC)
Tech. Spec. Bases 3/4, Page I - l a ; MTC CLP E36-01-C; Page 9 of 33; Subsection 2.j CLP E36-01-C; Figure 4 - Safety Analysis Assumptions CLP E36-01-C; Figure 7 - FSAR Worse Case Analysis NRC WA System/E/A NRC KIA Generic System A l l RCS Overcooling Number AA2.2 Ability to determine and interpret adherence to appropriate procedures and operation within the limitations in the facility's license and amendments as they apply to the RCS Overcooling.
Importance ROlSRO 3.0 3.4 10CFR Link (CFR: 43.5 / 45.13)
Page 27 of 75
New Selected for Exam u Past NRC Exam? 6000110 A plant heatup is in progress in accordance OP 2201, "Plant Heatup". SDC is in operation and the crew is ready to rack up the breakers for "A" and "B" RCPs to commence concurrent RCP/SDC operation.
The current RBCCW valve alignment is:
2-RB-30.1A, RBCCW CTMT ISOL HDR A SPLY - CLOSED 2-RB-37.2A, RBCCW CTMT ISOL HDR A RTN - OPEN 2-RB-30.1 B, RBCCW CTMT ISOL HDR B SPLY - CLOSED 2-RB-37.2B, RBCCW CTMT ISOL HDR B RTN - OPEN Which of following describes the system response with this valve alignment?
A Annunciator "RCP B CLG WTR FLOW LO", would clear when the "B" RCP breaker is racked- 0 up and the "B" RCP can be started.
B Annunciator "RCP A CLG WTR FLOW LO", would NOT clear when the "A" RCP breaker is B racked-up and the " A RCP can NOT be started.
C Annunciator "RCP A CLG WTR FLOW LO", would clear when the "A' RCP breaker is racked- W up and the "A" RCP can NOT be started.
D Annunciator "RCP B CLG WTR FLOW LO", would NOT clear when the "B" RCP breaker is 0 racked-up, but the "B" RCP can be started.
B - CORRECT Since the RBCCW supply to " A is isolated because 2-RB-30 1A is closed, the low flow annunciator will not clear when the breaker is racked-up and the RCP will not start because one of the start permissive for the RCP is adequate RBCCW flow A - WRONG The low flow annunciator will not clear because there is no RBCCW flow to "B" RCP and the RCP cannot be started because the start permissive ISnot met C -WRONG: The low flow annunciator will not clear because there is no flow to "A' RCP D - WRONG: The "6" RCP can not be started because the RBCCW flow permissive is not meet
$$ Per GAJ, corrected 'typo' in choice "A" & " D by changing "'A' RCP breaker" to "'B' RCP breaker". Also changed typo in Justification - RLC $$
NRC KIA System/E/A System 003 Reactor Coolant Pump System (RCPS)
Number K6.04 RO 2.8 SRO 3.1 CFR Link (CFR: 41.7 14%)
Knowledge of the effect of a loss or malfunction on the following will have on the RCPS: Containment isolation valves affecting RCP operation NRC KIA Generic System Number RO SRO CFR Link Page 54 of 127
- ,29 J Memory? (Check=Yesr
~
The plant is at 100% power, steady state A ground exists somewhere in the Letdown Flow Control circuit causing continuous letdown flow oscillations Letdown Flow Control Valve 2-CH-I IOQ is being placed in service and 2-CH-11 OP is being swap out in hopes of isolating the source of the ground The control operator has NOT had a chance to fully restore CVCS controls to their normal, at power, alignment Then, immediately after operators complete unisolating 2-CH-1 IOQ, the now in-service CH-1 10Q fails OPEN, resulting in the following conditions:
- Letdown flow is 128 gpm.
- Letdown Temperature through the Ion Exchangers (1.X.k) is 150 O F .
- Letdown Pressure upstream of the Pressure Control Valves is oscillating between 300 - 350 psi Which of the following describes the plant or system response, and the required actions based on that response, to CH-110Q failing open at this time.
A Letdown Flow is exceeding the design limit of the Ion Exchangers; they must be verified bypassed -3 before the resin is channeled.
B Letdown Pressure is oscillating above the letdown design limit; the backpressure controller must 1111 be stabilized in manual to close the relief valve.
C Letdown flow is exceeding the capacity of the Letdown Heat Exchanger; ensure the 1.X 's are 2 bypassed to prevent I.X. damage on temperature.
D Letdown flow is exceeding the capacity of the running charging pumps; the plant must be tripped due to inability to maintain PZR level.
~~
Justification C - CORRECT: LID flow may be exceeding the capacity of the HX. The auto system function should bypass the IX on high LID temp., preventing IX damage on temperature, IF the "bypass" switch has been returned to "auto".
Based on the conditions given in the stem, the control operator must not have yet returned the switch to auto and the I.X.'s are NOT being bypassed. Therefore, they are in immediate danger of damage on high temperature.
A - WRONG: The Letdown Limiter circuit is designed to ensure letdown flow does not exceed the IX capacity by limiting the amount of "open" signal the normal control system can send to the Letdown Flow Control valves.
However, even if the flow control valves FAIL open, the letdown valves are "mechanically blocked" to prevent letdown flow from exceeding 128 gpm.
B -WRONG: Pressure may INITIALLY lift the relief, but a backpressure of -35Opsi is within the design limit of the letdown piping because the relief does not lift to over 500 psi. The back-pressure control valves are operating to limit the magnitude of the pressure rise due to the high letdown flow. No operator actions are required for this condition.
D -WRONG: Even with the LID control valve failed full open, LID flow is not designed to exceed the capacity of all three charging pumps with normal RCS pressure. The backup charging pumps will automatically start on PZR level and maintain it within limits without the need to trip the plant.
Reference CVS-00-C, (CVCS Training Material) Rev. 8, Ch. 3, Pages 28 & 29 NRC KIA SystemIElA NRC KIA Generic System 004 Chemical and Volume Control System Number A221 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Excessive letdown flow, pressure, and temperatures on ion exchange resins (also causes)
I tn po rtance R O E RO 2.7 2.7 Page 29 of 75
10CFR Link (CFR: 41.51 4315 14513 14515)
,]
- _ _ _ ,,?,O,, 1 P Question =l16 , Memory?
__ (Check=Yes)
The following plant conditions exist:
20 gpm RCS leak into CTMT has recently appeared.
Control room operators are performing a Rapid Downpower at 30%/hr.
Plant power is presently 93% and dropping at the intended rate.
2 charging pumps are running, Letdown is at approximately 60 gpm.
Pressurizer level is 65% and stable.
VCT level is 75% and stable.
Boric Acid addition to the Charging pump suction at the required rate.
Additional 20 gpm blended makeup to the Charging pump suction.
A plant Chemist is regenerating a CPF Ion Exchanger resin bed.
Then, during the load reduction, the standby PMW pump trips off line on overload, resulting in a PMW Low Pressure alarm.
Based on the above plant conditions and the design limits of the CVCS and it's support systems, how will the plant respond to the PMW pump trip?
A The PMW and Boric Acid Flow Controllers, both operating in automatic, will maintain BOTH flow 11 rates at setpoint.
B Pressurizer level will begin to drop BELOW Level Setpoint, with all RCS temperatures being (J maintained on program.
c The RATE of the plant downpower will accelerate without any operator action on ANY reactivity 9 control systems.
)J Boric Acid and PMW flow rates will BOTH drop, maintaining the preset blend concentration to the 7 charging pumps.
Justification C - CORRECT: With PMW pump tripped and chemistry performing a regen., PMW pressure will drop enough to prevent PMW flow to the CVCS A -WRONG: The flow controller cannot make up for the loss of PMW pressure.
B -WRONG: RCS temperatures will not be maintained D -WRONG: ONLY the PMW flow rate will drop, not BOTH Reference ARP-2590B-020 Alarm D-5, NOTE 2 NRC KIA SystemlEIA NRC WA Generic System 004 Chemical and Volume Control System Number A1.10 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: Reactor power Importance ROlSRO 3.7 3.9 IOCFR Link (CFR: 41.5145.5)
Page 30 of 75
When entering 'Reduced Inventory Region' (RIR), SDC flow is reduced to between 1400 and 1600 gpm What is the reason for the MINIMUM, REQUIRED flow of 1400 gpm?
A This is the minimum flow required to prevent thermal stratification in the RCS hot and cold legs. 1 B Lower flow rates have the potential to cause RCS heatup due to inadequate decay heat removal. IJ c This reduced flow rate is consistent with the minimum required for adequate boron mixing in the '2:
smaller RCS volume.
D This is the minimum required flow for the LPSI (SDC) Pump, lower flow rates would potentially 1 overheat the pump.
Justification C - CORRECT: Due to the reduced volume in the reactor, a lower flow rate is adequate A - WRONG: Thermal stratification will not occur with the RCS in this configuration B -WRONG: Flow rates below 1000 gallons/minute would be required before inadequate decay heat removal became an issue.
D -WRONG: This is not the minimum flow for LPSl pump.
Reference OP-2310, Rev. 23-00, page 8 of 107, step 3.9 NRC KIA SystemlEIA NRC KIA Generic System 005 Residual Heat Removal System (RHRS)
Number K5.09 Knowledge of the operational implications of the following concepts as they apply the RHRS:
Dilution and boration considerations Importance RO/SRO 3.2 3.4 IOCFR Link (CFR: 41.5145.7)
Page 31 of 75
Question"1D. 6071905 I_1x__"',, ,"",
'"'"'1,-
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The plant has experienced a Loss of Coolant Accident and the RWST level is approaching the SRAS
'qMemory7
' .---. ~-
(Check=Yes) setpoint Which of the following valves REQUIRE operator action to ensure they go closed when SRAS fully actuates?
A RWST header isolation valves (2-CS-13.1A & 2-CS-13.1B)
B Both CTMT Sump Outlet isolation valves (2-CS-16.1A & CS-16.1B). .-
c 2-SI-659 & 2-SI-660, Safety Injection Pump Recirc. Valves. I&
D RBCCW outlet isolation valves from the SDC heat exchangers. 1-l Jirstification C - CORRECT: Components 2-SI-659 & 2-SI-660 require the switches on C-01 be placed in the A -WRONG: The header isolations due not get a closed signal as any remaining pressure in CTMT, and head of water in the CTMT suction line, will seat the check valves in this line and make it unnecessary.
B -WRONG: CTMT Sump outlet isolations get an OPEN signal, and they get it without operator action.
D -WRONG: These valves get an OPEN signal on a SRAS and should not be closed. However, in all other ESAS situations these valves would be closed (a common misconception).
Reference EOP-2532, Rev. 24, Step 21 & 48.
NRC KiA System/E/A NRC KIA Generic System 006 Emergency Core Cooling System (ECCS)
Number A4.02 Ability to manually operate andlor monitor in the control room: Valves Importance ROlSRO 4.0 3.8 10CFR Link (CFR: 41.7 / 45.5 to 45.8)
Page 32 of 75
Question ID
'" ,- ', ~
6071028
' _'.' ] or,?-
1 I 'q Mernory7<Check=Yes)
~ ______,
When drawing a bubble in the Pressurizer, which of the following sets of conditions indicates that a bubble has been formed?
, . ?I A Pressurizer level lowering with constant or rising pressure. 3 2 Quench tank level, temperature and pressure rising with no water being added to the RCS.
3 ,
8 Pressurizer level lowering with constant or rising pressure. e Pressurizer temperature corresponds to saturation temperature for present RCS pressure.
c Pressurizer pressure rising with rising water space temperature. u
~,
Quench tank level, temperature and pressure rising with no water being added to the RCS.
D Pressurizer pressure rising while maintaining constant charging flow. !__i Pressurizer steam and water space temperatures rising above hot leg temperature.
Justification B - CORRECT: OP 2301 D states, "Monitor for spontaneous bubble formation by the following:
- Pressurizer level lowering with constant or rising pressure.
- Pressurizer temperature corresponds to saturation temperature for present RCS pressure.
A -WRONG: PORVs are closed as soon as the PZR indicates filled and vented by way of flooding over into the Quench Tank. Gasses that come out of solution upon heating will be vented out the PZR steam space sample line, which remains open. Leaving the PORVs open would challenge the Quench Tank rupture disk.
C - WRONG: This is indicative of a solid RCS, but does occur to some extent when drawing a bubble. However, it can NOT be used as confirmation that a bubble EXISTS in the PZR. Also, the PORVs are closed as soon as the PZR indicates filled and vented, so QT parameters should NOT change.
D -WRONG: This will occur as soon as the process for drawing a bubble in the PZR begins and the PORVs are closed. Also, PZR water space temperature must be at SATURATION for the RCS pressure, NOT just above Thot.
Reference OP-2301 D, Rev. 27-02, Pg. 44 of 83, Step 4.4.27 NRC KIA System/E/A NRC KIA Generic System 007 Pressurizer Relief Tank/Quench Tank System (PRTS)
Number K5.02 Knowledge of the operational implications of the following concepts as the apply to PRTS: Method of forming a steam bubble in the PZR Importance ROlSRO 3.1 3.4 IOCFR Link (CFR:41.5/45.7)
Page 33 of 75
- 34 I ]w N
=
- e '2Memory? (Check=Yes)
~ ___
Which of the following will cause the "RBCCW SURGE TK AUTO MAKEUP" annunciator to alarm on C-06/7?
A Makeup valve open for 3 minutes and makeup valve in "AUTO" Y4 B
9 3 As soon as the makeup valve opens, regardless of PMW header pressure. .I c Only when the setpoint for "RBCCW SURGE TK Level HIILO" setpoint is reached D When makeup valve opens, if PMW header pressure is insufficient to supply makeup. iIli Jtistification A - CORRECT: The annunciator setpoint is "valve open for 3 minutes".
B -WRONG; There is no alarm function from the valve position.
C -WRONG; The surge HllLO alarm is not tied to the SURGE TK excessive makeup alarm.
D -WRONG; Valve position and PMW pressure do not feed into this alarm.
Reference ARP-2590E, Rev. 009-01, B-8, 046 NRC KIA System/E/A NRC KIA Generic System 008 Component Cooling Water System (CCWS)
Number K4.02 Knowledge of CCWS design feature(s) andlor interiock(s) which provide for the following:
Operation of the surge tank, including the associated valves and controls tin portance ROlSRO 2.9 2.7 IOCFR Link (CFR: 41.7)
Page 34 of 75
Origin: - RO id SRO Student Handout? ; :1 Lower Order? Question ID:
- 35 L_i Mod E Selected for Exam u Past NRC Exam? 6069977 The plant is in Mode 5 on Shutdown Cooling. The crew is presently making preparations to perform PORV Block Valve stroke testing. The test involves cycling each Block Valve fully closed then fully open.
Which of the following conditions meet the system requirements that would allow stroke testing of the PORV Block Valves?
RCS Pressure RCS Tc LTOP Switch Position SDC Status
- I)250 psia 150 O F Low In Service
- 2) 250 psia 150 "F High In Service
- 3) 300 psia 175 OF High In Service
- 4) 500 psia 175 O F Low Secured A #I 0 B #2 c #3 D #4 u B - CORRECT Conditions are within the SDC System limits, below LTOP limits (LTOP ACTUATION > 410 PSlA & c 275 DEG F) and allow the PORV Block Valve to be closed (LTOP Switch MUST be in "High")
A -WRONG The LTOP switch in "Low" prevents closing of the PORV Block Valve before opening the PORV, which would blow the PRT rupture disk C -WRONG 300 PSlA is above the design pressure for the SDC System D -WRONG 500 PSlA and 175 "F is above the LTOP limit and would open the PORV as soon as the LTOP switch was selected to "Low" References I..."' '
RCS-00-C, Rev 7, Ch 1, Pages 34 of 116 RCS-00-C, Figure 14 (Rev 1)
{Reworded stem to test more complex concepts solicited from Block Valve testing and reworded Choice "B"to be correct for conditions - RLC}
NRC KIA SystemlEIA System 010 Pressurizer Pressure Control System (PZR PCS)
Number A3.01 RO 3.0 SRO 3.2 CFR Link (CFR: 41.7 / 45.5)
Ability to monitor automatic operation of the PZR PCS, including: PRT temperature and pressure during PORV testing NRC K/A Generic System Number RO SRO CFR Link Page 66 of 127
Y Question IO 6054501
"' ,,,~ I ; ' Origin,," MvO,d ] Memory7
.- (Check=YesL The Main Turbine has just tripped from 100% power due to a failure in the Electro-Hydraulic Control cabinet. The turbine trip results in RCS pressure spiking above 2410 psia before the reactor is shutdowdtripped automatically. RCS pressure then immediately dropped to a normal "post-trip" value Which of the following is an indication that RPS failed to trip the reactor, but that it tripped due to actuation of the Diverse Scram System (DSS)?
A The MG sets are still running but their output contactors are tripped open. @
B Both MG set 480 volt supply breakers are open with amber indicating lights. ! -!
c 3 minutes and 20 seconds after the trip, both channels of AFAS will actuate. .-I D All four (4) of the CEDS Bus Undervoltage Relays have been deenergized. _I Jtistification -
A -CORRECT: With the RPS High Pressure trip set 140 psi above normal operating pressure, the plant should have tripped on High Pressure. If it tripped on activation of the DSS, then the MG set output contactors would be open. This does NOT trip the MG sets, just deenergizes the CEA buses.
B - WRONG: The DSS does NOT trip MG set input power, just the output power.
C -WRONG: The DSS will acttiate the AFAS if the high pressure signal stays in for the ENTIRE time (3 minutes, 20 seconds).
D -WRONG. The CEDS Undervoltage relays will be deenergized on the loss of power to the CEA buses, but this happens regardless of HOW the plant was tripped. Therefore it can NOT be used as an indication of the plant tripping on DSS actuation.
Reference CED-00-C, Control Rod Drive System (Training Material), Rev. 3, Ch. 4, Page 9 of 67, page 36 of 67 AFW-00-C, Aux. Feedwater (Traininq Mat.), Rev. 5. Ch. 3, Paqes 11 & 12 of 57 NRC K/A System/E/A NRC KIA Generic System 012 Reactor Protection System Number K3.01 Knowledge of the effect that a loss or malfunction of the RPS will have on the following: CRDS Importance ROlSRO 3.9 4.0 IOCFR Link (CFR: 41.7 I45.6)
Page 36 of 75
- 37 1 '-
Question ID. 6 0 5 4 0 d
~ I " "
z,~,,~ M-od, ~
I-IMeinory? (Check=Yes)
The following plant conditions exist:
- Plant is at 100% power, steady state.
- Vital 120 VAC electrical bus WAC-3 (VA-30) is aligned to its alternate power supply.
- While in this alignment a complete Loss of Normal Power occurs CONCURRENT with a loss of Inverter 4 (INV-4 is off-line).
During the performance of EOP-2525, the STA reports the following:
- RPS channels "C" and "D" momentarily deenergized until the Diesel Generators loaded.
.- All parameters monitored by safety channels "C" and "D", EXCEPT for RWST level indications, momentarily failed until the Diesel Generators loaded.
- Both facilities of the ESAS fully actuated, with the EXCEPTION of AEAS and SRAS.
Which of the following describes why a SRAS actuation did NOT occur with the rest of the ESAS actuations?
A The ESAS SRAS actuation only triggers if a High CTMT pressure SlAS actuation is also sensed $3 by the ESAS Sensor Cabinets.
B The RWST level channels have a backup power supply directly from non-vital 125 VDC, which i?
auto-swaps on loss of the WAC.
c The RWST level channels are powered from the ESAS Sensor Cabinets and each Sensor Cabinet 7-has a backup power supply D When power is restored to ESAS, the "crow bar" circuit blows the Actuation Cabinet fuses before i ESAS can actuate a SRAS.
Justification C - CORRECT: All ESAS sensor channels have backup power from the opposite facility. Channel 'A' from VA-IO/
40, Channel 'B' from VA-20130, Channel 'C' from VA-30/20 and Channel ' D from VA-40/10. Because the RWST level loops get their power from the ESAS sensor cabinets, which never totally lost power, SRAS does not fire.
A -WRONG:
B - WRONG: When the backup power to VA-30 & VA-40 (VR-11 & 21) momentarily deenergized on the LNP, ESAS saw two channels of safety parameters fail to their accident condition. This meets the 214 requirement and will result in an actuation of ALL systems except for SRAS and AEAS. Channel "Y" of PZR Pressure Control has the aforementioned backup DC power supply, not the RWST indication.
D -WRONG: The "crow bar" circuit will blow the Actuation Cabinet fuses IF the Actuation Cabinets lost power and then were reenergized. However ESAS Actuation Cabinets are powered from VIAC-10 and -20, which did NOT deenergize in the question scenario.
Reference ESA-01-C, Rev. 3, Pg. 8 & 39 of 73 NRC KIA SystemIElA NRC KIA Generic System 013 Engineered Safety Features Actuation System (ESFAS)
Number K2.01 Knowledge of bus power supplies to the following: ESFASkafeguards equipment control Im p o r t ance ROlSRO 3.6 3.8 IOCFR Link (CFR: 41.7)
Page 37 of 75
Orlg,l""-;,,k$ik _ ", 3 I 1 -Memory?
(Check=Yes)
The plant has tripped from 100% power due to a Large Break - Loss of Coolant Accident.
Which of the following describes the concern if the Containment Spray System is NOT operated per EOP-2532. "Loss Of Coolant Accident"?
A CTMT equipment is not designed for operation in a harsh environment without the CTMT Spray El System.
8 Containment pressure could exceed the design limit if the CTMT Spray system is not utilized as './J designed.
c Containment Sump Recirculation will NOT be effective as designed in providing a long term heat sink D Containment hydrogen control could jeopardize the CTMT integrity design criteria without spray l.-a flow.
Justification B - CORRECT: 2532 requires that CTMT spray not be secured UNTIL pressure drops to less than 7 psig. It is SECURED to reduce the effects of moisture/water on non-qualified equipment. However, if secured prior to the procedure requirement, CTMT pressure should exceed the design limit.
A -WRONG: All required safety equipment is certified for operation in harsh CTMT environment.
C -WRONG: In a large break LOCA, the long term heat sink will be maintained by SRAS actuation, which aligns RBCCW as the heat sink for HPSl pumplsump recirc.
D -WRONG: Spray flow has minimal effect on CTMT hydrogen concentration or hydrogen burn capability.
Reference CSS-00-C, CTMT Spray (Training Mat.), Rev. 4, Ch. 1, Page 5 of 51 System 022 Containment Cooling System (CCS)
Number A1.02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Containment pressure Importance ROlSRO 3.6 3.8 IOCFR Link (CFR: 41.5 / 45.5)
Page 38 of 75
=,ni--;- 1 _ _ _ ~ _ _ _ _ _ _-_
Memory? (Check=Yes)
'dd During operation in an Event Specific EOP, the Crew encounters a step that states: "IF condition 'X' exists, THEN Go To EOP 2540, Functional Recovery" Which of the following describes the procedure usage requirements if is determined that condition 'X' does exist?
A Immediately exit the EOP in use, enter EOP 2540, and do not exit until its exit conditions are met. -2 B Exit the EOP in use, enter EOP 2540 and perform the required actions, then return to the initial EOP c Continue in the event based EOP while using EOP 2540 in-parallel to address condition ' X A D Complete the section of the EOP in use, then exit that EOP and enter EOP 2540 3 Justification A - CORRECT: OP 2260 states: "The words 'Go To' are used to accomplish branching." Branching is used to direct the user to exit the procedure in use and not return unless otherwise directed.
B -WRONG: 'Go To' does not allow for automatic return to the branching procedure. This must be so directed in the applicable procedure.
C -WRONG: EOP 2540 cannot be used in parallel with any other event based EOP.
D -WRONG: 'Go To' action must be immediately carried out, therefore, the original procedure must be immediately exited.
Reference 013-2260, (EOP Users Guide), Rev. 009-001, Page 15
- ~ ~ -~ _ _ ~- ~- ~
NRC KIA SystemIElA NRC KIA Generic System 022 Containment Cooling System (CCS) 2.4 Emergency Procedures /Plan Number GS 2.4.5 SEE GENERIC K/A "Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions."
Importance ROlSRO 2.9 3.6 10CFR Link (CFR: 41.10/43.5/45.13)
Page 39 of 75
- , 40 1 ,Origi,n Bank, 1 ~ idMemory7 (Ch&k=Yes)i
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What effect does a SRAS actuation have on the Containment Spray System7 A Containment Spray pumps are tripped and the RBCCW valves to SDC heat exchangers are -
closed B Containment spray pumps are aligned to the containment sump and the RBCCW valves to SDC ]:I heat exchangers are closed.
c Containment spray pumps are aligned to the containment sump and the RBCCW valves to SDC p; heat exchangers are opened.
D Containment spray pumps are tripped and the RBCCW valves to the SDC heat exchangers are 1-1 opened.
Justification A. WRONG: CTMT spray are not tripped on a SRAS (the LSPl pumps trip NOT CTMT Spray Pumps) and the RBCCW is supplied to SDC HXs for cooling of recirculated water.
B.WRONG: The first part of the statement is true about the CTMT spray aligned to sump but second part is wrong (same reason as Aabove.
C. CORRECT: Per EOP 2532 step 48, LPSl pumps trip and RBCCW outlet valves from SDC HX open D. WRONG: CTMT sprays do NOT trip on SRAS.
Reference CSS-00-C, (Training Mat.), Rev. 4, Chg. 1, Page 7 of 51 NRC KIA System/E/A NRC KIA Generic System 026 Containment Spray System (CSS)
Number K1.O1 Knowledge of the physical connections and/or cause-effect relationships between the CSS and the following systems: ECCS Importance ROlSRO 4.2 4.2 10CFR Link (CFR: 41.2 to 41.9 / 45.7 to 45.8)
Page 40 of 75
, O ~ I , ~ ~
,I W ' ' ~ M ~ ~ ~ ~ ~ e c' k = Y e s )
The plant has tripped from 100% power due to a large break LOCA The crew is carrying out the actions of EOP-2532, LOCA, and all ESAS equipment has actuated, including SRAS.
Then, the crew notices that HPSl pump performance is degrading due to CTMT sump clogging. The crew wants to stop both CTMT Spray pumps to mitigate the degradation of Safety Injection flow.
In order to secure CTMT Spray during degraded HPSl performance, CTMT pressures must be maintained below which of the following values?
B 7 PSig '1
-2 c 39 psig. iJ D 54 psig. 'LI' Justification D - CORRECT: Some clogging requires securing of CTMT Spray as soon as possible to limit the degradation of HPSl injection flow.
A -WRONG: This is the TECH SPEC limit for CTMT pressure during normal plant operations.
B -WRONG: This is the normal limit for securing CTMT Spray if sump clogging is not a factor.
C -WRONG: This is the CTMT pressure limit, assuming 54 PSlG is actually 54 PSIA.
Reference EOP-2532. Rev. 024, Pg. 42 of 96, Step (contingency) 50.2.a
~ ~ ~~ - ~- ~- ~~
NRC K/A SystemlEIA NRC KIA Generic System 026 Containment Spray System (CSS)
Number K3.02 Knowledge of the effect that a loss or malfunction of the CSS will have on the following:
Recirculation spray system importance ROlSRO 4.2 4.3 IOCFR Link (CFR: 41.7 145.6)
Page 41 of 75
1 -
iJ- Memory?_ _(Check=Yes)
_ _ _ __ _ _ -> I The plant is operating at 90% power for Control Valve Testing, when a High Pressure turbine 6th stage Extraction Steam supply valve fails closed Which of the following is a plant change that would result from this valve failing closed, with NO operator actions?
A Reactor power as seen by the Nuclear Instruments will begin to rise. I&
B The Main Turbine will soon trip due to loss of heater drain pump flow. U c Actual Main Generator electrical megawatt output will begin to lower. Id D Water hammer in the High Pressure Feed Water Heaters will begin. CI Justification A - CORRECT. The extraction steam valve closing will cause feedwater temperature to drop, resulting in positive reactivity addition.
B -WRONG: Heated drain pump flow will be lower, but not enough to trip the main feed pumps on low suction pressure.
C -WRONG: Generator output will rise due to the extraction steam now going through the high pressure turbine D -WRONG: Feed water heater water hammer requires loss of all extraction steam.
Reference FHD-00-C, (Extraction Steam & Feedwater Heaters Training Mat.), Rev. 4, Page 29 & 30 NRC K/A SystemlEIA NRC K/A Generic System 039 Main and Reheat Steam System (M RSS)
Number K5.08 Knowledge of the operational implications of the following concepts as the apply to the MRSS:
Effect of steam removal on reactivity Importance ROlSRO 3.6 3.6 10CFR Link (CFR: 441.5 / 45.7)
Page 42 of 75
- 43,. 1 A plant start up is in progress with power currently at 70%.
The following equipment is in service:
o 'A' and 'C' Condensate pumps o both Heater Drain Tank pumps o both Main Feed Pumps.
Then, the flow transmitter for the Condensate Pump Minimum Flow controller fails low Without any operator action, what effect does this failure have on the Condensate/Main Feed System?
I A Condensate Surge Tank level will RISE. 1-1 B SGFP suction pressure will LOWER ,d C Condenser hotwell level will LOWER D S/G level will begin to RISE Justification E3 - CORRECT: The transmitter failing low causes the minimum flow control valve to go to its full open position.
At least 3200 gpm of condensate pump discharge flow will be diverted to the condenser. This action lowers the suction pressure to the Main Feed pumps.
A -WRONG: This implies confusion between the Condensate Minimum Flow Valve and the Hot Well Reject Valve, which discharges to the Surge Tank. The Condensate Reject Valve taps off of the same pipe extension as the Minimum Flow Valve and is physically located right next to it.
C -WRONG: Condenser level would rise due to the water being diverted back to it, vice going to S/G D -WRONG: S/G level would lower due to less flow to it.
Reference CON-00-C, Condensate System (Training Mat.), Rev. 7, Ch. 8, Pages 23 & 37 NRC KIA System/E/A NRC K/A Generic System 056 Condensate System Number K1.03 Knowledge of the physical connections and/or cause-effect relationships between the Condensate system and the following systems:
MFW Importance ROlSRO 2.6 2.6 IOCFR Link (CFR: 41.2 to 41.9145.7 to 45.8)
Page 43 of 75
- 44, 1 Question ID. I O O O O ~
I , , .. I r; __-.
Memory' __(Check=Yes)'
-- - p i The following plant conditions exist
- 80% power increasing at 3% per hour.
- 3 condensate pumps, 5 CPF demins., 2 heater drains pumps, 2 Steam Generator Feed Pumps (SGFP).
- 'A' & 'B' SGFP are in Auto speed control at 4130 and 4050 rpm, respectively Then, one of the ' A SGFP suction pressure detector fails low (0 psi).
Which of the following is a response of the MFW system to this instrument failure, and what action, (if any), will be required to stabilize the plant?
A No immediate response occurs in the MFW system since SGFPs are in speed control. If a second :7J detector were to fail, the 'A' SGFP would trip after five seconds, resulting in a necessary plant trip.
B Both SGFPs shift to Minimum Discharge Pressure control when the detector fails low. The 6th 1 CPF demin. must be placed in service due to higher flowldemin. DP when the 'A' SGFP speeds up.
C 'A' SGFP shifts to Manual and goes to 3400 rpm due to loss of the suction pressure signal. 'A' Main Feed Reg. Valve must be shifted to Manual and adjust to maintain required feed flow condition.
D "A" SGFP will speed up attempting to maintain Feed Reg. Valve DP. However, when the speed 'J exceeds 4200 rpm with the low suction pressure signal, the SGFP will trip requiring a plant trip.
Justification A - CORRCT: Suction pressure is a critical monitored parameter for the SGFP speed control system. If one is lost, only an alarm is received. However, if a second detector fails, the applicable pump will immediately trip. At 80% power, one SGFP is insufficient to maintain adequate main feed flow necessary to maintain SG level.
B -WRONG: The pumps are in speed control, but suction pressure still has an input to the SGFP control circuit and must be dealt with. Placing a demin in service will have no impact on feed flow as the pump has not changed speed.
C -WRONG: The pump does not shift to manual, however, 3400 rpm is the speed at which auto control takes over.
D - WRONG: SGFPs do not control on main feed suction pressure, however, if 213 pressure were to go low, they do have a control input from feed pump speed and would cause the applicable pump to speed up, as stated in the distracter.
Reference MFW-00-C, Rev. 8, various locations.
FWL-00-C, Rev. 2, various locations.
NRC K/A System/E/A NRC K/A Generic System 059 Main Feedwater (MFW) System Number A1.07 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW controls including: Feed Pump speed, including normal control speed for ICs Importance ROlSRO 2.5 2.6 10CFR Link (CFR: 41.5/45.5)
Page 44 of 75
- 45, I A plant startup is in progress with current power being held at 1YOwhile preparations are being made to start the first Main Feedwater pump.
Then, VA-10 deenergizes.
Which one of the following describes an effect on the plant at this time, due to this loss of power?
.~~~
A ONLY # I SG level will begin to rise without operator action. Y!
B BOTH SG levels will begin to rise without operator action. '-3 c #I SG AFW Reg. Valve is locked up in it's "throttled" position. !J D NEITHER Facility of AFAS can actuate, regardless of SG level Justification A - CORRECT; Both channels of Aux. Feed. Require power to actuate. As Fac. 1 is powered from VA-IO, loss of this bus will inop. It. VA-10 also powers the Aux. Fd. Reg. Valve, which will now fail full open, overfeeding the # I SG.
B -WRONG; Although the loss of VA-10 disables Fac. 1 AFAS, and Fac. 1 AFAS can start the Fac. 2 pump, this facility has no control over the Fac. 2 AFW Reg. Valve specifically. Therefore, there is no change to #2 SG level.
C -WRONG; The #I Main Feed Reg. Valve locks up on loss of VA-10, not the AFW Reg. Valve. The AFW valve will fail full open on loss of VA-10.
D -WRONG; AFAS Fac. 1 is disabled and Ch. "A" to both facilities is disabled and cannot send an actuation signal. However, Fac. 2 of AFAS can still actuate if two of the remaining three channels ('B', 'C' & ID') drop below their actuation setpoint.
Reference AFW-00-C, Rev. 5, Ch. 3, Pages 11, 18 & 19 NRC KIA SystemlEIA NRC WA Generic System 061 Auxiliary / Emergency Feedwater (AFW) System Number K3.02 Knowledge of the effect that a loss or malfunction of the AFW will have on the following: S/G Importance ROlSRO 4.2 4.4 10CFR Link (CFR: 41.7 145.6)
Page 45 of 75
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Question ID. 0062980 Memory? (Check=Yes)
- , . 4 6 ,,,,,, i__r--,"-, , ,',, ' '
- - -~
The plant was operating at power when a low Steam Generator water level trip occurred due to a loss of BOTH Main Feedwater Pumps In the performance of EOP-2525, the SPO has started both motor-driven AFW pumps and aligned them to feed the SGs. However, upon verifying the AFW flow rate, the SPO notes there is zero (0) feed flow to the either SG.
A Plant Equipment Operator (PEO) is sent to the AFW pumps to monitor, and reports the following:
- Both the "A"and "B" AFW pumps are operating
- Both pumps are extremely noisy
- The pump discharge lines are extremely hot
- Local temperature indicators are reading 265 OF
- NO system valves were found out of position Which one of the following actions should the Crew initiate?
A With both AFW pumps running, shut the pump manual discharge valves and then slowly open them.
B Override the AFAS, secure the AFW pumps, and attempt to restart each pump on recirculation '1 flow.
C Secure both AFW pumps, shut the AFW pump manual discharge valves and vent the pump iz casings.
D Shut down either the "A'or "B" AFW pump to provide higher suction pressure to the running AFW i_
pump.
Justification C - CORRECT; The local indications are that the AFW pumps are steam bound. Per OP-2322, the pumps must be shutdown and vented.
A -WRONG; Local throttling of the AFW pump discharge to more "finely tune" discharge pressure will NOT collapse the steam void inside the pumps.
B - WRONG; Continued operation, in any form, prior to venting has the potential to cause severe damage to the pumps.
D -WRONG; Although the problem is lack of NPSH, this solution will NOT collapse the steam void in the pump casing that is causing the problem.
Reference OP-2322, Rev. 026-00, Pg. 50 - 54 of 58 NRC KIA SystemlEIA NRC K/A Generic System 061 Auxiliary / Emergency Feedwater (AFW) System Number K6.02 Knowledge of the effect of a loss or malfunction of the following will have on the AFW components:
Pumps Importance ROlSRO 2.6 2.7 10CFR Link (CFR: 41.7 / 45.7)
Page 46 of 75
-Memory?
___ - (Check=Yes)-
The plant is stable at 98% power following the loss of VA-10 (deenergized). Plant personnel are trouble-shooting the deenergized bus.
Then, the following occurs:
- "A" main steam header breaks in containment and the plant is tripped.
- On the trip, 24C fails to transfer to the RSST due to a failure of the RSST-to-24C Feeder breaker
- 24C is deenergized.
+ ALL other equipment operates as appropriate for the above conditions.
- 2-FW-44 (Aux. Feed Header X-tie) has been closed.
- Both electric AFW Pumps are secured.
- #I Steam Generator (SG) is at 550 psia and dropping.
Then, the STA reports that Main Feed Flow is beginning to show flow to the #I SG. The US directs the SPO to IMMEDIATELY secure all feed flow to the #I SG.
Which one of the following actions should the SPO take to immediately secure Main Feed flow to the #I SG?
A Ensure all Facility 1 SlAS & ClAS actuations on C05. Ll B Reenergize 4160 VAC buss 24C via 24E and Unit 3. L.1 c Manually isolate #I Aux. Feed Regulating Valve. Li D Ensure all running Condensate Pumps are secured. .-
I%/
Justification 0 - CORRECT; The loss of VA-10 locks up the #I Main Feed Reg. Valve (FRV) in the 100% power position. On the trip, the loss of the RSST to 24C will deenergize 24A and therefore 22A&22C. These 480 load centers power the motor operated valves that are normally used to isolate the locked up #I FRV. As power to these valves as lost BEFORE any signal could close them, they are all fully open. As soon as the #ISG depressurizes to </= 500 psia, any running condensate will start feeding it and accelerate the depressurization.
A -WRONG; As power is still lost to these valves, no ESAS or control board signal can effect them. If power WAS available, SlAS & ClAS actuation would already have isolate feed flow.
B -WRONG; The depressurization (and subsequent feeding) will jeopardize CTMT long before any lost bus can be reenergized.
C - WRONG; With the electric Aux feed pumps secured and RN-44 closed, Aux feed is effectively isolated from the #I SG.
Reference AOP-2404C, Rev. 003-06, Pg. 3 of 24. (Main Feed Reg. Valve locks u p on loss of VA-10)
NRC KIA SystemIElA NRC WA Generic System 062 A.C. Electrical Distribution Number A1.03 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: Effect on instrumentation and controls of switching power supplies Importance ROlSRO 2.5 2.8 10CFR Link (CFR: 41.5 145.5)
Page 47 of 75
Question ID. 6000019
- " x , , I , " I ,
'Oris_l", ,, ,, 1 [ ?--Memory7 (Check=Yes) 1 The plant IS operating at 100% power with the "B" Emergency Diesel Generator (EDG) running for a Tech Spec surveillance The EDG is in parallel to grid and loaded per the applicable Tech Spec surveillance procedure All EDG parameters are normal and stable.
Then, Unit 3 trips due to a fault on one of their vital AC busses.
Which of the following describes the effect of the Unit 3 trip on the "B" EDG operation?
A The EDG indications must be checked locally to ensure the throttle position and governor are still I .J within required limits.
B The EDG indications must be checked locally to ensure the jacket and oil temperatures are still J within required limits.
c The EDG indications on C08 must be checked to ensure the Megawatt loading on the EDG is still 1 within required limits.
D The EDG indications on C08 must be checked to ensure the VAR loading on the EDG is still within q' required limits.
Justification D - CORRECT: When Unit 3 trips, it will "dump" all of the VARs is carrying onto the "local" grid, of which the EDG is now attached. The EDG will pick up some of these VARs and have to be adjusted accordingly from C08.
A -WRONG: When the EDG is run for a T.S. surv., it is operated in "Unit Parallel" mode. This means the governor is in a "load set" type of control mode and will maintain EDG MWe loading at a constant setting regardless of how the grid it is tied to fluctuates.
B -WRONG: As the EDG can be assumed to be started and operated per procedure, once the VAR loading is corrected, the EDG should be running fine with all of it's auto systems performing as designed.
C -WRONG: MW loading should not change, regardless of grid instability, with the EDG operating in "Unit Parallel" mode. However, VARS can and will change with the grid fluctuations.
Reference SP-2613L, (EDG Tech. Spec. Surveilance), Rev. 00-01, Pg. 13, Note & Caution NRC KIA System/E/A NRC K/A Generic System 062 A.C. Electrical Distribution Number A1.O1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: Significance of D/G load limits I in portance R O E RO 3.4 3.8 IOCFR Link (CFR: 41.5 I 4 5 . 5 )
Page 48 of 75
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- A9 "r191.1 v-Mernory7
- -~ - __
(Check=Yes)
During the performance of EOP 2525, the Secondary Plant Operator noticed that D C control power was lost to the "A" Emergency Diesel Generator (EDG)
All local actions were taken in accordance with the procedure to secure the EDG When DC control power is restored, which of the following sets of actions must be performed to completely restore the EDG to an operable status?
A Reopen starting air isolation valves and reset the shutdown relay B Reset the fuel racks and reset the shutdown relay. _.
c Reset the fuel racks and reopen the starting air isolation valves. I -A D Reset the fuel racks, reopen the starting air isolation valves and reset the shutdown relay. ./'
Justification D - When DC was lost, the EDG fuel racks had to be tripped to stop the engine. A micro switch is activated which sends a signal to energize the SDR when DC is restored. EOP 2525 Contingencies require the DG be tripped locally and the air start valves be closed. In order to restart, the fuel racks have to be reset and the air start valves have to be reopened. Then the SDR is reset by pressing the "alarm reset" button on the skid.
A -WRONG; The fuel racks had to be manually tripped locally on the loss of DC control power in order to shutdown the EDG. Therefore, in order to "restore" the EDG the fuel racks must be manually reset locally. The governor cannot do this automatically.
B - WRONG; The Air start valves must be closed when securing the EDG because the air start solenoids are deenergized. This would cause the EDG to continue to roll in an attempt to start. Therefore, in order to "restore" the EDG, these valves must be reopened.
C -WRONG; The SDR has triggered and must be reset before the EDG governor will allow the EDG to start Reference EDG-00-C, Emergency Diesel Generators (Training Mat.), Rev. 7, Pages 137 8138 of 143 ARP 2591A-018, Rev. 001 NRC WA System/E/A NRC K/A Generic System 063 D.C. Electrical Distribution Number K2.01 Knowledge of bus power supplies to the following:
Major DC loads Importance ROISRO 2.9 3.1 IOCFR Link (CFR: 41.7)
Page 49 of 75
Question ID. 6000021 c, .' ' ' ' , '..' ' . Origin" New, , ,, _I 4 Memory? (Check=Yes) ~
The plant is in Mode 5 for a refueling outage. DC bus 201A is cross-tied to 201 B for replacement of the "B" Battery ALL other electrical systems are in a normal alignment for a shutdown plant.
Then, the in-service battery charger trips off line on an internal fault.
How long is the battery DESIGNED to carry all loads under the present conditions?
A Approximately one (1) hour.
B Approximately four (4) hours.
c Approximately eight (8) hours D Approximately sixteen (16) hours. I u'
Justification A - CORRECT: Cross-tying the buses will result in a substantial decrease in battery life. The MODE 5 loads are NOT substantially different than in any other mode. Through engineering calculations it has been determined that the battery will NOT degrade linearly, but will deplete exponentially.
B - WRONG: This is the original, expected time for the battery while carrying two busses, based on a linear decay of capacity.
C -WRONG: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is the normal battery life when carrying only one DC bus, but two are now drawing on the battery.
D -WRONG: This would be the estimated capacity if the required logic in answering the question were reversed.
Reference 125 VDC/120 VAC, LVD-00-C, Rev 5, Change 1, Page 12 of 77.
~
NRC KIA System/E/A NRC KIA Generic System 063 DC Electrical Distribution System Number A1.O1 Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including: Battery capacity as it is affected by discharge rate Importance RO/SRO 2.5 3.3 IOCFR Link (CFR: 41.5 I 4 5 . 5 )
Page 50 of 75
z;Mo,d,;;;,. ,I q -Memory7
-~
(Check=Yes)
The "A" EDG IS being unloaded and shutdown from the control room following a normal Tech Spec surveillance run However, the EDG is unloaded too quickly and the output breaker trips on reverse power, causing the following annunciator to alarm on Panel C-08 "DIESEL GEN 12U BKR CLOSING CKT BLOCKED" No other annunciators are in alarm for the "A" EDG.
Which of the following describes actions that MUST be taken to reset any protective functions for the "A" EDG that this condition may have triggered?
A Reset the "A' EDG output breaker from C08 and verify the associated alarm clears. k 5 Reset the "A" EDG output breaker from C08, reset the EDG shutdown relay and verify all alarms clear c Reset the "A" EDG output breaker trip relays locally at A312 and verify the associated alarm clears. 7 D Reset the "A'EDG output breaker trip relays locally at A312, reset the EDG shutdown relay and I?
verify all alarms clear Justification A - CORRECT: The EDG breaker need only be reset at C08 per the applicable ARP, no further action is necessary to restore the EDG to standby.
B -WRONG: The EDG electronic governor shutdown relay is not actuated by this alarm condition, even though the EDG is rendered inoperable by the triggering of the applicable trip circuit.
C -WRONG. This particular breaker trip will trigger local relays, but they do not have any "interlock" function They can be reset at any time and do not inop. The EDG in and of themselves.
D -WRONG. This would be the requirement if the reverse power caused a generator fault and actuated the EDG differential-lockout relay, which also disables the EDG.
Reference Per ARP 2590F, Rev. 007-07, Page 202 Of 203, CORRECTIVE ACTIONS #2, this the correct method to reset the alarm. Also, classroom training material titled, Emergency Diesel Generator System (EDG-00-C), Revision 7, chg.1 Pages 138 & 139 discusses the closing logic for the diesel output breaker.
NRC KIA SystemlEIA NRC KIA Generic System 064 Emergency Diesel Generators (ED/G)
Number A2.05 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loading the ED/G I m portance ROlSRO 3.1 3.2 10CFR Link (CFR: 41.5/43.5/45.3/45.13)
Page 51 of 75
- 52 I ,0,1gi,n I ,,M,od , 1 I Memory? (Check=Yes)~
With the plant operating normally at 100% power, a PROCESS RAD MONITOR HI HVFAIL alarm is received on C-06/7 . The Radiation Monitor indications on the PPC indicate as follows:
- RM8434B, Aux Building Gaseous Radiation Monitor in alarm and rising.
- RM8999, Radwaste Exhaust Particulate has an Unexplained Rise.
- RM8132B, U2 Stack Gaseous Radiation Monitor has an Unexplained Rise.
The response of these radiation monitors was caused by which of the following?
A The Volume Control Tank vent line is slowly leaking to the Waste Gas header.
B A packing leak on the discharge valve of the running Waste Gas Compressor. dl c Evaporation of water in the SFP which lowers level by eighteen inches.
D The inadvertent opening of a relief valve on the running Charging Pump
~~ ~ ~~ ~~~ ~~~~ ~ ~ ~~~ ~~~~~ ~ ~ ~~~ ~~~~ ~ ~ _ _ ~
Justification B - CORRECT: Due to the configuration of the Aux Building ventilation system, a packing leak on the discharge valve of the running Waste Gas Compressor would be seen by RM8434B, Aux Building Gaseous Radiation Monitor. The other two Rad Monitors are in the ventilation flow path to the Unit 2 stack, where most Aux Building ventilation is discharged.
A -WRONG: A pinhole leak in the VCT vent to waste gas would result in an alarm on RM8998, Radwaste Exhaust Particulate Rad Monitor, NOT on RM8434B.
C -WRONG: Any activity from the Spent Fuel Pool area would result in an alarm in RM8145A andlor 6 , Spent Fuel Pool Exhaust Particulate andlor Gaseous Rad Monitor, NOT on RM8434B.
D -WRONG. A radioactive release in the Charging Pump area would result in an alarm on RM8997, Radwaste Exhaust Particulate Rad Monitor, NOT on RM8434B.
Reference PPC display printout of Unit Two Rad Monitors.
(Also shown on P&ID 25203-26029, sheet 2, and sheet 3 - not included)
NRC KIA System/E/A NRC KIA Generic System 073 Process Radiation Monitoring Conduct of Operations (PRM) System Number GS 2.1.7 SEE GENERIC K/A Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.
Importance ROlSRO 3.7 4.4 IOCFR Link (CFR: 43.5 145.12 145.13)
Page 52 of 75
Question ID 6000108 Memory7
. -~
(Check=Yes)
___I^__x___" " " ' -'
Which of the following is an accurate list of the Service Water components that receive a SlAS signal and the type of signal received (Le., start, stop, open closed)?
8 --;
A 1, ONLY the two in-service Service Water Pumps receive a start signal. I.-_'
- 2. Both Diesel Generator Service Water Bypass Valves, 2-SW-231A and B receive an open signal.
B 1. ONLY the two in-service Service Water Pumps receive a start signal. -.J
- 2. Both Diesel Generator Service Water Supply Valves, 2-SW-89A and 8 receive an open signal.
c 1. All three Service Water Pumps receive a start signal. , ~~
- 2. Vital Chiller Crosstie Valves, 2-SW-102 and 104, receive a close signal.
D 1. All three Service Water Pumps receive a start signal. 'V
.~ ~~.
- 2. TBCCW Service Water Supply Valves, 2-SW-3.2A and 3.26, receive a close signal.
Justification D - CORRECT: All three SW Pumps receive a start signal to ensure all required emergency heat loads are supplied, even though only the two in-service piimps actually start. The spare pump is prevented from starting by maintaining the control switch in Pull-To-Lock. The Service Water Isolation Valves to TBCCW will close to ensure an adequate supply of Service Water to all components requiring Service water cooling during an accident. All three RBCCW HX TCVs will open to ensure RBCCW has adequate cooling during an accident.
A -WRONG: All three Service Water Pumps receive a start signal, even though only two pumps actually start.
Also, the DG Service Water Bypass Valves close on the DG start, NOT from a safety injection actuation signal.
B - WRONG: All three Service Water Pumps receive a start signal, even though only two pumps actually start.
Also, the DG Service Water Supply Valves open on the DG start, NOT from a safety injection actuation signal.
Additionally, all three RBCCW TCVs get an open signal. The spare HX is outlet isolation is throttle or closed limit or prevent flow through it.
C -WRONG: Vital Chiller Cross tie Valves were removed from the system several years ago. Also, all three RBCCW TCVs get an open signal. The spare HX is outlet isolation is throttle or closed limit or prevent flow through it.
Reference SWS-O4-C, Service Water System, Rev. 5, Figure 1 (Rev. 1) and Pg. 8, 13, 14 & 15 of 59 NRC K/A System/E/A NRC K/A Generic System 076 Service Water System (SWS)
Number A3.02 Ability to monitor automatic operation of t h e SWS, including: Emergency heat loads I in portance ROISRO 3.7 3.7 IOCFR Link (CFR: 41.71455)
Page 53 of 75
, # ,,54 1 The Instrument Air supply to the Containment Air Receiver must be isolated to repair an air leak on the s~ipplypiping. In order to maintain air to Containment for valve operation, the Station Air System must be aligned to the Containment Air Receiver. A dedicated operator has been selected and briefed.
Cross-tying Station Air to supply the Containment Instrument Air Receiver A must be minimized to prevent moisture build up in the Containment Instrument Air System which Z may impact the operation of air operated valves in Containment B must be limited to one hour to ensure the CONTAINMENT INTEGRITY Technical Specification iI action statement is NOT exceeded C will result in a lower supply pressure to air operated valves in Containment which may cause the 1; valves to operate more slowly during an event D will require the Containment Instrument Air System to be purged prior to use for breathing air due ' .'
to the oil vapor contained in the Station Air System Justification A - CORRECT: Caution prior to step 4.4.1 in OP 2332A, Station Air, states, "Station Air System has NO air dryer, cross-tying Station Air to supply the Containment Instrument Air Receiver should be minimized to prevent moisture build up in the Containment portion of the Instrument Air System."
B -WRONG: With a dedicated operator stationed, there is NO limit on the CONTAINMENT INTEGRITY Tech Spec LCO.
C - WRONG The Station Air System operates at roughly the same pressure as the Instrument Air System.
Additionally, air pressure is regulated at the individual components at a much lower pressure than the system D -WRONG: There is NO requirement to purge the Instrument Air System after cross-tying to the Station Air System.
Reference OP-2332A, Station Air, Rev. 011-00, Pg. 9 or 13, Caution prior to Step 4.4.1 NRC K/A System/E/A NRC K/A Generic System 078 Instrument Air System (IAS)
Number K1.03 Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: Containment air Importance ROlSRO 3.3 3.4 IOCFR Link (CFR: 41.2 to 41.9145.7 to 45.8)
Page 54 of 75
- ,s5, ] Question ID" 600ij107 1 -- "
Origin
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The plant is operating normally at 100% power. A small, unisolable Instrument Air leak in Containment Memory7 (Check=Yes)?
has caused Containment air pressure to slowly rise over the past several days. (The Instrument Air leak is NOT large enough to adversely impact the Instrument Air System.)
Which of the following provides the maximum pressure allowed by Technical Specifications and, also, provides the action that must be taken to prevent exceeding this limit?
A 27.7 inches of water. After obtaining Containment samples, vent Containment through the i4 Hydrogen Purge flow path and EBFS to Millstone Stack.
27.7 inches of water. After obtaining Containment samples, initiate Containment Purge using I
B EBFS and the Containment Cleanup flow path.
C 20 inches of water. After obtaining Containment samples, vent Containment through the .I Hydrogen Purge flow path and EBFS to Millstone Stack.
D 20 inches of water. After obtaining Containment samples, initiate Containment Purge using EBFS T.,
and the Containment Cleanup flow path.
Justification A - CORRECT: The Technical Specification limit for Containment pressure is 1.0 psig which equates to 27.7 inches of water. The only approved method of depressurizing Containment in MODE 1 is to use the Hydrogen Purge valves, through EBFAS, out to the Millstone Stack.
B -WRONG: The Containment Cleanup flow path can ONLY be used in MODES 5 or 6 C -WRONG 20 inches of water is the maximum pressure referenced in OP 2314B, Containment and Enclosure Building Purge; however, Technical Specifications allow up to 27.7 inches of water (1 .O psig).
D -WRONG: 20 inches of water is the maximum pressure referenced in OP 23148, Containment and Enclosure Building Purge; however, Technical Specifications allow up to 27.7 inches of water (1 .O psig). Additionally, the Containment Cleanup flow path can ONLY be used in MODES 5 or 6.
Reference OP 23148, Containment and Enclosure Building Purge, Rev. 020-05, Section 4.14, Pg. 39 of 62 Millstone 2 Technical Specification, 3.6.1.4 NRC KIA System/E/A NRC KIA Generic System 103 Containment System Number A1.O1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including: Contain ment pressure, temperature, and humidity Importance ROlSRO 3.7 4.1 10CFR Link (CFR: 41.5 / 45.5)
Page 55 of 75
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The crew just performed a plant shutdown from 100% power to approximately 3% power in six (6) hours.
You have been directed to maintain power between 2% and 4%. After approximately 30 minutes without any operator action, you notice that delta-T power is reading 2.5% and nuclear power is reading 5 E-5%.
Remembering your last direction, you withdraw CEAs 10 steps.
How does RCS temperature INITIALLY respond to the change in CEA position and why?
A RCS temperature will rise. Adding positive reactivity at this time will result in a temperature rise 'l:.:
because the Reactor is still critical.
B RCS temperature will NOT change. A positive reactivity addition significantly below the Point of -4 Adding Heat will NOT cause a temperature rise.
C RCS temperature will rise. With Delta-T power above the Point of Adding Heat, a positive reactivity addition will cause temperature to rise.
D RCS temperature will NOT change. Plant systems will NOT allow RCS temperature to rise until nuclear power is greater than 5%.
Justification B - CORRECT: The Reactor is below the point of adding heat, but is still likely critical. Withdrawing CEAs will add positive reactivity, but until the Reactor is above the point of adding heat, RCS temperature will NOT increase A -WRONG: Even if the Reactor is critical, RCS temperature will NOT be affected by a positive reactivity addition until power is above the point of adding heat which is around 1% (nuclear) power.
C -WRONG: Nuclear power must be above the point of adding heat for a positive reactivity insertion to cause RCS temperature to rise; NOT Delta-T power.
D -WRONG: RCS temperature will NOT rise until nuclear power is above the point of adding heat (approximately 1%). Plant systems will limit the rise in temperature, but temperature must rise for the systems to respond.
Reference Generic Fundamentals (applicable pages supplied to Lead Examiner)
NRC KiA System/E/A NRC K/A Generic System 001 Control Rod Drive System Number K5.95 Knowledge of the following operational implications as they apply to the CRDS: Effect of reactor power changes on RCS temperature importance ROlSRO 3.4 3.7 IOCFR Link (CFR: 41.5/45.7)
Page 56 of 75
- 57 1 Question ID 6000105 Memory7 (Check=Yes)
___ - _ ~ _
The plant is in normal operation at 100% power with TWO (2) charging pumps in service in manual (Mode Switch in LVL-1).
Then, the Loop 1Thot input to the Reactor Regulating System suddenly fails low (minimum)
Assuming NO operator action, how will actual Pressurizer level respond and why?
-7 A RISES; Letdown Flow goes to MINIMUM with NO change in Charging Pumps.
B RISES; Letdown Flow goes to MINIMUM and ALL Charging Pumps are running. if c LOWERS; Letdown Flow goes to MAXIMUM with TWO (2) Charging Pump running. L D LOWERS; Letdown Flow goes to MAXIMUM with ONE (1) Charging Pump running. '2 Justification When the Thot input to RRS fails low, the RRS Tavg calculator will immediately drop the calculated Tavg based on the "new" RCS temperature input. This will cause the PZR setpoint calculator in RRS to then drop the PZR auto-setpoint output based on the "bad" Tavg. The auto-setpoint is fed directly to the PZR Level Controllers and used to control PZR level in normal operation. When the auto-setpoint drops due to the instrument failure, it will cause the in-service PZR Level Controller to "think" actual PZR level is much higher than it should be for the given RCS temperature. The applicable controller will then RAISE Letdown Flow to MAXIMUM (-128 gpm) and STOP any "backup" Charging Pumps that are presently running (40 gpm charging flow). Based on this system response, actual PZR level will LOWER.
D - CORRECT: requires analysis of the failure's effect on the RRS and the effect that the RRS output change would have on the PZR Level Control System.
A - WRONG. implies the malfunction directly effects the Letdown Flow Controller, instead of the PZR Level Controller.
B -WRONG: implies the reverse direction of the input failure, which is a malfunction also covered in Training.
C -WRONG: implies the Charging Pumps are unaffected by the PZR Level Controller when running in LVL-1 mode. The pumps would only be unaffected if they were running for surveillance testing (start on an ESAS signal)
Reference RRS-01-C, Reactor Regulating System, Rev. 3, Ch. 3, Page 18 NRC KIA System/E/A NRC K/A Generic System 01 1 Pressurizer Level Control System (PZR LCS)
Number K6.04 Knowledge of the effect of a loss or malfunction on the following will have on the PZR LCS:
Operation of PZR level controllers Importance ROlS RO 3.1 3.1 IOCFR Link (CFR: 41.7 I 4 5 . 7 )
Page 57 of 75
-~_-
' J Memory7 (Check=Yes)]
Which one of the following CEDS interlocks or design features will be affected by the loss of the Plant Process Computer (PPC)?
A Group Deviation Backup I. !
B Upper Electrical Limit id C Sequential Permissive d CEA Motion Inhibit ._I Justification C - CORRECT: The Sequential Permissive is generated by the PPC to allow withdrawal of the next group of CEAs when its preceding group withdraws above its Upper Sequential Permissive. Without the PPC, this signal is unavailable and CEA groups must be withdrawn in individual mode only.
A -WRONG: Although the PPC generates two different Group Deviation alarms, the CEDS interlock on Group Deviation is generated by CEAPDS (RPIS). Therefore, if the PPC is lost, the alarms are affected but the INTERLOCK is not.
B -WRONG: The Upper Electrical Limit is driven by a separate set of reed switches that also feed the core mimic, NOT the PPC. The Upper Core Stop (sometimes confused with the UEL) is the CEDS interlock that is controlled by the PPC.
D -WRONG: The CMI is a function of CEAPDS Reference CED-01-C, Rev. 4, Control Element Drive System; Table 5 - Interlocks and Alarms, [or Table 6 - Steps vs Functions table (end of the document)l.
NRC K/A SystemlEIA NRC K/A Generic System 014 Rod Position Indication System (RPIS)
Number K4.06 Knowledge of RPIS design feature(s) and/or interlock(s) which provide for the following:
Individual and group misalignment Itn portance ROlSRO 3.4 3.7 IOCFR Link (CFR: 41.5 /45.7)
Page 58 of 75
' -Memory7
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The plant is operating normally at 100% power when 120 Volt Vital AC Bus, VA-20, suddenly deenergizes Which of the following describes the affect of a loss of this bus on the Nuclear Instrumentation System?
A Only Channel "B" Wide Range and Channel "B" Power Range Safety Channel are lost. Channel 3 "Y" Power Range Control Channel and lncore Detectors are NOT affected by this loss of power.
B Channel "B" Wide Range, Channel "B" Power Range Safety Channel, and Channel "Y" Power G Range Control Channel are lost. lncore Detectors is NOT affected by this loss of power.
c Only Channel "B" Power Range Safety Channel is lost. Channel "B" Wide Range, Channel "Y" Power Range Control Channel, and lncore Detectors are NOT affected by this loss of power.
D Channel "B" Wide Range, Channel "B" Power Range Safety Channel, Channel "Y" Power Range j:
Control Channel, and all the Facility 2 lncore Detectors are lost.
Justification A - CORRECT: VA-20 supplies Channel "B" Wide Range and Power Range Safety Channel. VR-11 Supplies Channel " Y Control Channel. The lncore Detectors are self powered.
B - WRONG: Channel "Y" Power Range Control Channel is power from VR-11, C -WRONG: Channel "B" Wide Range is powered from VA-20 D -WRONG: Channel "Y" Power Range Control Channel is power front VR-11 and there are NO Facility 2 lncore Detectors.
Reference NIS-00-C, Rev. 3, Ch. 2, Pg. 8 of 54 AOP-2504D, Loss of 120 VAC Vital Instrument Panel, VA-20, Rev. 003-05, Pg. 21 of 23, Attachment 2.
NRC K/A System/E/A NRC WA Generic System 015 Nuclear Instrumentation System Number K2.01 Knowledge of bus power supplies to the following:
NIS channels, components, and interconnections Importance ROlSRO 3.3 3.7 10CFR Link (CFR: 41.7)
Page 59 of 75
- , 65! I OrbIn,", N-5,. ,, 1 -~
' q Memory? (Check=Yes)
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A Large Break LOCA concurrent with a loss of the RSST has occurred. EOP 2525 is complete and the crew is performing steps of EOP 2532, Loss of Coolant Accident. Peak Containment pressure was 34 psig, crushing the Containment Air Recirculation Fan ductwork Several hours after the event, a Hydrogen concentration of 4.2% is detected in Containment. The RSST is still deenergized.
Which of the following describes the flow path of air to the Hydrogen Recombiners at this point?
A The Containment Auxiliary Recirculation Fans take a suction from the Containment dome li discharging to the Containment Air Recirculation Fan suctions. The Containment Air Recirculation Fan discharge ductwork, crushed by the excess Containment pressure, causes the fans to discharge directly to the Hydrogen Recombiners.
E3 The Post Incident Recirculation Fans are taking a suction on the blowout panels at the discharge of the Containment Air Recirculation Fans. The Post Incident Recirculation Fans discharge directly to the Containment Auxiliary Recirculation Fans which discharge directly to the suction of the Hydrogen Recombiners.
C The Containment Air Recirculation Fans discharge through the open blowout panels to the suction 1' 111 of the Post Incident Recirculation Fans which are discharging to the Containment lower levels to provide mixing. The Hydrogen Recombiners use convection air flow to reduce Containment Hydrogen concentration.
D The Post Incident Recirculation Fans take a suction on the Containment dome and discharge to !@
the Containment Air Recirculation Fans which discharge through the open blowout panels to the lower levels of Containment to provide mixing. The Hydrogen Recombiners use convection air flow to reduce Hydrogen concentration.
Justification D - CORRECT: The Post Incident Recirculation Fans take a suction on the Containment dome and discharge to the Containment Air Recirculation Fans which discharge through the open blowout panels to the lower levels of Containment to provide mixing. The Hydrogen Recombiners use convection air flow to reduce Hydrogen concentration.
A -WRONG: The Containment Auxiliary Recirculation Fans are deenergized due to the loss of the RSST.
B -WRONG: The Post Incident Recirculation Fans do NOT take a suction form the Containment Air Recirculation Fans, which do NOT discharge directly to the Hydrogen Recombiners.
C -WRONG: The Post Incident Recirculation Fans do NOT take a suction form the Containment Air Recirculation Fans.
Reference Containment and Containment Systems, CCS-00-C, Rev. 8, Change 4, Pages 35 & 37. Also Figure 41 (Rev. 1).
Hydrogen Control System, HCS-00-C, Rev. 3, Pages 10 & 11, and Figure 8.2 (Rev. 2)
P&ID 25203-26028, sheet1 and sheet 3 [NOT supplied]
NRC K/A System/E/A NRC K/A Generic System 028 Hydrogen Recombiner and Purge Control System (HRPS)
Number K1.O1 Knowledge of the physical connections and/or cause-effect relationships between the HRPS and the following systems: Containment annulus ventilation system (including pressure limits)
Importance R O E RO 2.5 2.5 10CFR Link (CFR: 41.2 to 41.9 145.7 to 45.8)
Page 60 of 75
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Origi,n New, , 1 3 I
g > - Memory'
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(Check=Yes)
The plant has just tripped from 100% power due to a loss of condenser vacuum, with all systems responding as designed.
The US has directed the SPO to stabilize RCS Tavg at 532 O F using the Atmospheric Dump Valves (ADV) .
Which of the following is the required automatic setpoint for the ADV controllers to maintain the directed Tavg7 A 900 psla r/
B 900 psig 1-c 880 psia I d D 920 psia -2 I
Justification A - CORRECT: Per Steam Tables and units on ADV controller, Tavg of 532 "F = 900 psia.
B -WRONG: ADV is upstream of MSIVs, therefore, units are in psia NOT psig like Condenser Dumps.
C -WRONG: Setpoint for Condenser Dumps.
D -WRONG: Normal ADV setpoint, but will maintain Tavg at - 534 "F.
Reference Steam Tables NRC K/A System/E/A NRC KIA Generic System 041 Steam Dump System (SDS) and Turbine Bypass Control Number K5.02 Knowledge of the operational implications of the following concepts as the apply to the SDS: Use of steam tables for saturation temperature and pressure I in po rtance R O E RO 2.5 2.8 10CFR Link (CFR: 41.5 I45.7)
Page 61 of 75
Question ID. 6055138 origin,..,. M,jd 1
_ _ _ " I _ _
'd Memory7 A__-__ -
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(Check=Yes)
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Shell warming of the main turbine is in progress Shell pressure ISbeing maintained between 60 and 100 Ps'g.
Which of the following is the correct status of the Turbine Valves?
A The Control Valves are throttled open with the potentiometer, the Main Stop Valves are open, and the Intermediate Stop Valves are open.
B The Control Valves are open, the Main Stop Valves are closed, and #2 Intermediate Stop Valve is :2 throttled open with the potentiometer.
C The Control Valves are open, #2 Main Stop Valve Bypass is throttled open with the potentiometer, i?
and all Intermediate Valves are closed.
D #2 Control Valve Bypass is throttled open with the potentiometer, the Main Stop Valves are closed, -3 and all Intercept Valves are closed.
~ ~~ ~ ~ ~~~ ~ ~~~ ~ ~ ~~~
Justification C - CORRECT: When the SHELL WARM button is pressed, all 4 Control Valves go fully open. As the potentiometer is rotated, the #2 Main Stop Valve internal bypass opens to admit steam to the Turbine shell. Shell pressure is maintained at 60-100 psig for the duration of Shell Warming All the Intercept and Intercept Stop Valves remain closed to allow the shell to be pressurized to 60-1 00 psig.
A - WRONG: The Control Valves are NOT throttled open by the potentiometer. Additionally, the Intermediate Stop Valves must remain closed to allow the shell to be pressurized.
B -WRONG: The Main Stop Valves cannot remain closed; otherwise the Shell could NOT be pressurized Additionally, the potentiometer cannot control any of the Intermediate Valves.
D -WRONG: The potentiometer does NOT control the #2 Control Valve Bypass (No Control Valve Bypass exists). Additionally, the Main Stop Valves cannot remain closed. The shell must be pressurized for Shell Warming.
Reference OP 2323A, Main Turbine, Rev. 022-05, Pg. 17 of 61 MTC-00-C, Rev. 5 [NOT included]
~~ ~ ~ ~- ~~
NRC KIA SystemIEIA NRC KIA Generic System 045 Main Turbine Generator (MT/G)
System Number A401 Ability to manually operate and/or monitor in the control room: Turbine valve indicators (throttle, governor, control, stop, intercept), alarms, and annunciators Importance ROlSRO 3.1 2.9 10CFR Link (CFR: 41.7 / 45.5 to 45.8)
Page 62 of 75
Question ID. 0055797 O r y ,%,nk,
, , 1 q Memory? (Check=Yes)
A The Mechanical Vacuum pumps (hoggers) and the Condenser Air Removal Fans. I (2
/
B The SJAEs and the demister. Id C The demister and the isolation valve to the Millstone Stack. LA D The demister and Fan 55 A & B. A I
Justification B - CORRECT: Per the P&ID, this is the order of components in the system A -WRONG: MV pumps and Cond. Air Removal Fans are both downstream of the RE-5099 C -WRONG: Demister and Isolation valves are downstream of the RE-5099 D -WRONG: Demister and Cond. Air Removal Fans are downstream of the RE-5099 and demister.
Reference P&ID for Condenser Air Removal, 25203-26012 CAR-00-C, (CAR Training Mat.), Rev. 3, Chg. 3, Figure 1 NRC KIA System/E/A NRC KIA Generic System 055 Condenser Air Removal System (CARS)
Number K1.06 Knowledge of the physical connections and/or cause-effect relationships between the CARS and the following systems: PRM system Importance ROlSRO 2.6 2.6 IOCFR Link (CFR: 41.2 to 41.9 / 45.7 to 45.8)
Page 63 of 75
, ~ - i n , ' " T F w,, 1 d Memory? (Check=Yes) ~ -.-
You have been directed to perform a Gaseous Rad Waste discharge of the "A" Waste Gas decay Tank with an OPERABLE Waste Gas Discharge Radiation Monitor (RM-9095).
Which of the following components in the discharge flow path are used to prevent exceeding the Dressure limit in the Discharge Radiation Monitor?
A Waste Gas Discharge Isolation Valve, 2-GR-37 2, can be throttled using the variator on Panel C- ,-
61.
B Waste Gas Discharge Pressure Controller, 2-GR-9.1 is set to a maximum of 10 psig in the '2 discharge header.
C Waster Gas Discharge Header Stop, 2-GR-71 is manually throttled to maintain 10 psig in the U discharge header.
D Radiation Monitor Outlet Isolation, 2-GR-93, is manually throttled to maintain less than 20 psig. I Justification B - CORRECT: 2-GR-9.1 is the inlet to the discharge header and is procedurally throttled to limit discharge header pressure to 10 psig; thus, limiting the pressure in the Radiation Monitor to 10 psig.
A -WRONG: Although 2-GR-37.2 can be throttled, it is after the Radiation Monitor and will only change the flow rate of the discharge. Procedurally, it is NOT throttled, but fully opened.
C -WRONG: 2-GR-71 is the bypass around the Radiation Monitor and is used to maintain the discharge flow fate below the maximum allowed by the discharge permit.
D - WRONG: 2-GR-93 is used to maintain flow through the Radiation Monitor between 2.5 and 3.5 CFM.
Reference SP-2617B, Rev. 011-10, Section 4.1, Pg. 14 of 33.
P&lD 25203-26021, Sh. 2 [NOT provided]
NRC KiA SystemiElA NRC WA Generic System 071 Waste Gas Disposal System (WGDS)
Number K4.05 Knowledge of design feature(s) and/or interlock(s) which provide for the following: Point of Release Importance ROlSRO 2.7 3.0 IOCFR Link (CFR: 41.7)
Page 64 of 75
The plant is in normal operation at 100% power, when a Fire System Trouble annunciator is received on C06/7. The ONLY annunciator lit on Fire Panel C-26 is Zone 12, "A" Diesel Room TROUBLE.
Which of the following conditions will cause ONLY this annunciator to alarm?
A The Manual Break Glass station for the "A'Diesel Generator Room has been actuated. l i
B ALL the Smoke Detectors for the "A' Diesel Generator Room have been actuated. L:
c ONLY the Heat Detectors for the "A" Diesel Generator Room have been actuated.
D "A'Diesel Room Deluge Valve Isolation Valve, 2-FIRE-22 is closed for surveillance Justification CHOICE (A) - NO WRONG: Activation of the Manual Break Glass station will cause the Deluge Valve to open, resulting in the actuation of the "A" Diesel Room ALALRM, NOT the TROUBLE annunciator.
VALID DISTRACTOR: An applicant may think that the Manual Break Glass station only results in a TROUBLE annunciator on C-26.
CHOICE (B) - NO WRONG: Actuation of any 8 smoke detectors in the "A" DG Room will trigger an ALARM annunciator on C-26.
VALID DISTRACTOR: an applicant may think that both heat and smoke detectors are required to actuate an ALARM and a TROUBLE and that only the smoke detectors will trigger a TROUBLE annunciator.
CHOICE (C) - NO WRONG: Actuation of only the Heat Detectors will trigger an ALARM annunciator on C-26.
VALID DISTRACTOR: an applicant may think that both heat and smoke detectors are required to actuate an ALARM and a TROUBLE and that only the heat detectors will trigger a TROUBLE annunciator.
CHOICE (D) -YES CORRECT: Closing the " A Diesel Room Deluge Valve Isolation Valve will actuate a TROUBLE annunciator on C-26.
Reference ARP-25901, "Alarm Response for Fire Panel, C-26" (Zone 12), Revision 02-05, (Pg 20 of 106)
NRC K/A System/E/A NRC WA Generic System 086 Fire Protection System (FPS)
Number A1.05 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Fire Protection System operating the controls including: FPS Lineups Importance ROlSRO 2.9 3.1 IOCFR Link (CFR: 41.5 145 5)
Page 65 of 75
Question ID: 3100003 I - I L l M e m o j ? (Check=Yes)
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The plant IS operating normally at 100% power. A student in training for an RO license is standing a watch under the supervision of the Secondary Plant Operator (SPO).
The Primary Plant Operator (PPO) suddenly becomes ill and has to leave. Due to vacation coverage, a third (work control) Licensed Operator is NOT available. A replacement is called in and will report to working about 30 minutes, however, the PPO is leaving NOW.
Then, five (5) minutes later, the plant trips.
Which of the following statements, as allowed by procedure, describes the Operator responsibilities for this event?
A The SM will perform the duties of the PPO while the US performs the command and control i I - .
m1 :
function.
B The US performs the duties of the PPO while the Shift Manager maintains a command and control ,g function.
c The Station Duty Officer performs the command and control function while the US performs the in duties of the PPO.
D The SPO trainee will perform the required actions on the primary plant under the direct supervision ' , I of the US.
Justification B - CORRECT: OP 2260, step 1.21.5, states, "In the event it becomes necessary to provide assistance to the board operators, this assistance should be provided by the US in all cases. The Shift Manager should maintain his overall perspective of plant conditions." In this case, a loss of DV-20 will result in a plant trip due to closure of the #2 MSIV. The US should perform the actions of the PPO until relieved.
A -WRONG: The SM should always 'maintain his overall perspective of plant conditions C -WRONG: The Station Duty Officer, by definition, is NOT licensed on Unit 2 and CANNOT perform licensed duties.
D -WRONG: The trainee can ONLY 'operate controls with permission of the SM/US and under the direct supervision of a qualified Operator.' The US cannot maintain direct supervision of the trainee during the performance of EOP 2525. Additionally, trainee operation of equipment must be immediately suspended during unanticipated or abnormal events.
Reference MP-14-OPS-GDL300 OP 2260, step 1.21.5 NRC K/A System/E/A NRC KIA Generic System 2.1 Conduct of Operations Conduct of Operations Number G 2.1.2 SEE GENERIC K/A Knowledge of operator responsibilities during all modes of plant operation.
Importance RO/SRO 3.0 4.0 10CFR Link (CFR: 41.10 145.13)
Page 66 of 75
Question ID 0056587 0r;g;n." , , , . ~ n .k ~ 1 v-Memory7 (Check=Yes)
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Shift turnover IS in progress and the on-coming Control Operator last had the watch four (4) days ago Which of the following items is required to be reviewed by the on-coming Control Operator BEFORE he can assume the watch, per MP-14-OPS-GDL200, Conduct of Operations?
A All Tech Spec. surveillances performed in the last four (4) days. /...J B The shift turnover report for the period covering the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. lq' c All Unit Two Condition Reports generated in the last four (4) days. !L D Clearances for all tags hung on control boards in the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. I i d
)
Justification B - CORRECT: MP-14-OPS-GDL200, Conduct of Operations states that the on-coming watch standers must review, as a minimum, the turnover report for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
A -WRONG: Although the Operations Schedule has Licensed Operators off for four to seven days at a time, review of all T.S. surv. Completed over the last 4 days could take several hours and is totally unnecessary.
C -WRONG: Although the Operations Schedule has Licensed Operators off for four to seven days at a time, review of all CRs entered over the last 4 days could take several hours and is totally unnecessary.
D -WRONG: The number of tags hung on the boards could number close to a hundred, too much time would be required for review and is not necessary.
Reference MP-14-OPS-GDL200, Rev. 012-01, Conduct of Operations, Pg. 15 NRC K/A System/E/A NRC KIA Generic System 2.1 Conduct of Operations Conduct of Operations Number G 2.1.3 SEE GENERIC KIA Knowledge of shift turnover practices.
Importance ROlSRO 3.0 3.4 IOCFR Link (CFR: 41.10 I45.13)
Page 67 of 75
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ro~emory7 (Check=Yeq The plant has tripped due to an Excess Steam Demand Event in CTMT.
On the trip, off-site power was lost and the "A" EDG output breaker would NOT close. The SPO then emergency tripped the ' A EDG.
All other equipment responded as designed The crew completed EOP-2525, diagnosed an ESD and entered EOP-2536.
Twenty (20) minutes after the plant trip, the following conditions now exist:
- The "A'EDG output breaker has been repaired and is cleared for operation.
- The SPO has been directed to start and load the "A' EDG and then restore Facility 1 equipment.
- CTMT pressure is 23 psig and slowly dropping.
Which of the following is correct, with respect to restoring Facility 1 equipment?
A Start the "A" EDG and verify the sequencer energizes bus 24C, starts the 'A SW pump and then the "A' RBCCW pump.
B Close the RBCCW supply to "A' SDC Heat Exchanger, 2-RB-13.1A., BEFORE allowing the "A" IJ RBCCW pump to start.
c After restoring 24C and Fac. 1 SW, start and run the Fac. 1 CAR fans for > I 5 minutes prior to rI1 starting Facility 1 RBCCW.
D Start the "A" EDG and energize 24C; restore Facility 1 SW flow, but do NOT restore Fac. 1 g RBCCW to the CAR fans.
Justification D - CORRECT: CTMT pressure is >20#. The reason for not starting the RBCCW pump is water hammer caused by the hot RBCCW in idle CAR cooler flashing on pp start.
A - WRONG: This will start the RBCCW pump before the RBCCW in the CAR fan heat exchanger can be cooled.
B -WRONG: The SDC has no RBCCW flow at this time and 2-RB-13.1 is not open. Even so, the RBCCW in this heat exchanger is not the problem.
C -WRONG: Running the CAR fan will NOT cool the RBCCW in it's heat exchanger enough if CTMT pressure is still above 20 psig.
Reference EOP-2525, Rev. 20, Pg. 5 of 26, Contingency Action NRC K/A System/E/A NRC WA Generic System 2.1 Conduct of Operations Conduct of Operations Number G 2.1.20 SEE GENERIC WA Ability to execute procedure steps.
Importance ROlSRO 4.3 4.2 10CFR Link (CFR: 41 .IO I43.5 / 45.12)
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1
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I Memory?
- (Check=Yes) I
~# 69 The PPO is about to perform SP-2601D, RPS - Calorimetric calibration.
Which of the following groups contains ONLY those parameters that are directly affected by adjustments made to the NI or Delta-T power calibration potentiometers? (Assume that only the calibration adjustments are made, with no other operator actions.)
A CEAPDS PDlL indication and TaveiTref recorder Power indication U B Power Ratio Calculator recorder AS1 setpoint and Calorimetric Power indication 0 c LPD trip setpoint and CEA Motion Inhibit on PDlL "trigger" setpoint I$
D Linear Range Power indication and Diverse Scram Aux. Feed actuation setpoint U
~~ ~~~~~~~~ ~ ~ ~~ ~~~~ ~~~~
Justification C - CORRECT: NI or Delta-T potentiometers will affect Q-power. Q-power is not an input to the Tavgnref recorder or the wide range power drawers.
A -WRONG: PDlL setpoint is fed by RPS, not the actual indication. Also, power indication on the recorders is delta-T power.
B -WRONG: The PRR AS1 setpoint is fed by the highest of the Q-Powers from all four channels, however, the Calorimetric calculation of the PPC does not receive and input from RPS.
D -WRONG: Linear power indication is directly off the Nl's, therefore, unaffected by the Q-power Cal. Pot.
Reference RPS-01-C, Reactor Protection System (Training Mat.), Rev. 6, Ch. 2, Page 52 of 80 NRC K/A System/E/A NRC K/A Generic System 2.2 Equipment Control Equipment Control Number G 2.2.12 SEE GENERIC K/A Knowledge of surveillance procedures.
Importance ROlSRO 3.0 3.4 10CFR Link (CFR: 41.10145.13)
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IgMCmory? (Check=Yes)
~ ____ -J The 'A'RBCCW pump motor is being red tagged out of service for maintenance during which the supply breaker will be completely removed from its cubicle for repair. The Work Order does NOT allow for testing at this time Which of the following is the proper method for hanging the tags for the 'A' RBCCW pump motor?
A Hang a BLUE tag on the control board switch and a RED tag at the breaker cubicle. ;7
- --I B Hang a RED tag on the control board switch and a YELLOW tag at the breaker cubicle. 2 C Hang a RED tag on the control board switch and a BLUE tag at the breaker cubicle. u D Hang a YELLOW tag on the control board switch and a RED tag at the breaker cubicle. !z Justification D - CORRECT: WC-2 requires that personnel "hang tags for removed breakers on [the] breaker cubicle." Yellow tags are used on the hand switches in the Control Room to signify a component tagged out in the plant.
A -WRONG: Ordinarily, if testing were allowed, a Blue tag would be hung on the control board to allow for this.
However, testing is not allow per the stem.
B -WRONG: If the Switch were also part of the maintenance and troubleshooting package then this would be correct. However, that has to be specifically stated.
C -WRONG: This configuration could be used if the breaker were to be locally tested by Maintenance Reference WC-2,Rev. 006-03, Step 1.6.4 NRC KIA System/E/A NRC K/A Generic System 2.2 Equipment Control Equipment Control Number G 2.2.13 SEE GENERIC WA Knowledge of tagging and clearance procedures.
Importance ROlSRO 3.6 3.8 IOCFR Link (CFR: 41.10/45.13)
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-#- ,71 ,I A Fuel Handling Accident has occurred inside the Unit #2 Containment Building. All personnel were immediately evacuated.
The containment isolation was completed effectively. Management personnel are interested in making a Containment Entry to evaluate the condition and location of the dropped fuel assembly.
Which radiation monitor(s) will be utilized in evaluating conditions inside the Containment, prior to the actual entry?
A Unit 2 Stack Radiation Monitors u B CTMT High Range Radiation Monitors 0 c Millstone Stack Radiation Monitor U D Refueling Platform Radiation Monitor LZ Justification D - CORRECT: Refuel platform is an area rad. Mon. that has a range capable of detecting CTMT rads. From fuel leaking.
A -WRONG: The stack would not see the rads. Because CTMT would not be vented through normal ventilation (Main Exhaust) while containing fission product gasses.
B -WRONG: The High Range monitors have too high a range for just fission product gasses. They are designed to see levels in the hundreds of rads.
C -WRONG: Millstone stack would see the rads., but the stack is used by both units and is substantially diluted by normal discharge.
Reference AOP-2577 (Fuel Handling Accident), Rev. 008-01, Section 2.0 NRC WA System/E/A NRC K/A Generic System 2.2 Equipment Control Equipment Control Number G 2.2.30 SEE GENERIC K/A "Knowledge of RO duties in the control room during fuel handling such as alarms from fuel handling area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation."
Importance ROlSRO 3.5 3.3 IOCFR Link (CFR: 45.12)
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'Question ID 2000021
' ,, " ,v-,, r I Memory?
(Check=Yes)
~ -
A minor fuel handling accident has just occurred in the Spent Fuel Pool and gaseous activity is bubbling up to the surface The unit supervisor directs the PPO to verify equipment actuation preventing/mitigating the discharge of activity to the environment.
Which of the following system alignments must the PPO verify to accomplish this task?
A Fan F-20 supplying outside air, discharge aligned to the EBFAS. :-I B Fan F-20 off with the outside air supply isolated, discharge aligned to EBFAS. I 2
c Fan F-20 supplying outside air, discharge aligned to the Main Exhaust System. El 0 Fan F-20 off with the outside air supply isolated, discharge aligned to the Main Exhaust System. I]
Justification B - CORRECT: 2 of 4 SFP area RMs exceeding 50 mR/hr actuates AEAS which places the SFP area under a negative pressure by securing supply air (F-20 off and outside air dampers closed) and aligning suction to EBFS so any activity will be filtered and monitored.
A - WRONG: F-20 must be secured because it could pressurize the area around the SFP and cause air leakage directly to the environment.
C -WRONG: See "A" and "D", both concepts are wrong D -WRONG: The discharge would not be filtered enough (no activated charcoal for the iodine) if sent to Main Exhaust. Also, it would be a "ground" release.
Reference AOP-2577 (Fuel Handling Accident), Rev. 008-01, Attachment 2, Pg. 10 NRC WA System/E/A NRC WA Generic System 2.3 Radiation Control Radiation Control Number G 2.3.1 1 SEE GENERIC K/A Ability to control radiation releases Importance RO/SRO 2.7 3.2 IOCFR Link (CFR: 45.9 145.10)
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1 VJMemoty? (Check=Yes)]
Preparations for a plant shutdown from 100% is in progress due to a Tech. Spec. LCO requirement. The IDENTIFIED leakage from the RCS to the Primary Drain Tank is in excess of the RCS Leakage Tech Spec The US has directed the PPO to start forcing Pressurizer sprays, start a second charging pump in manual, and notify HP (per the guidance in AOP-2575, Rapid Down Power).
Which of the following describes the REASON for the notification of the Health Physics Department, per A 0P-2575?
A The expected background radiation rise due to the rise in the RCS identified leakage. u B The expected lowering of radiation levels in the plant due to the upcoming shutdown. 0 c The expected radiation rise when balancing letdown flow for the second charging pump. F D The expected radiation rise from dumping RCS to radwaste as PZR level setpoint lowers. u
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Justification The "Parent question (# 5000023) was SRO ONLY, this Modification is not.
C - CORRECT: Starting a 2nd pump requires increasing letdown flow. The higher flow rate would cause a rise in rad. Levels due to the higher amount of N-16.
A -WRONG: Leakage to the PDT would not be seen outside CTMT until the PDT is pumped to rad waste.
B -WRONG: HP will be informed of the rad. Level changes of the shutdown when the shutdown is discussed (brief). This will be done in several minutes by the US.
D -WRONG: PZR setpoint does not lower as fast as the PZR level lowers due to the RCS cooldown on the power reduction. Therefore, there will be no radwaste generated by the shutdown.
Reference 1. RPM-1.1.2, "Radiation Protection Program and ALARA Program", Revision 3 (8/19/04) (Pg 5,8,9,10,11,16 of 33)
- 2. OP-2304E, "Charging Pumps", Revision 15 (03/09/04), Step 4.4.10 (Pg 21 of 25)
NRC KIA System/E/A NRC K/A Generic System 2.3 Radiation Control Radiation Control Number G 2.3.10 SEE GENERIC K/A Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
Importance ROlSRO 2.9 3.3 10CFR Link (CFR: 43.4 I45.10)
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- , 7 4 I Following a trip from full power due to loss of all feedwater, stopping RCPs, a major step in the mitigation strategy of EOP 2537, precedes the Contingency Action to initiate once through cooling.
Which of the following describes the basis for stopping RCPs at this point in the procedure?
A To prevent RCP operation outside of design limits during the upcoming RCS depressurization. i:3 B To minimize RCP seal damage due to the impending ClAS which will isolate seal bleedoff flow. 0 c To eliminate RCP heat input into the RCS and to conserve the remaining SG water inventory '?
D To minimize subsequent RCS voiding and mass loss if Once Through Cooling must be initiated. 7 Justification C - CORRECT: RCS heat input is the primary bases for the securing of ALL RCPs immediately after implementing EOP-2537.
A -WRONG: If depressurization were a concern, the RCPs would be secured at the same points in the depressurization as is done in a normal plant cooldown.
B -WRONG: Once ClAS has initiated and isolated the normal RCP Bleedoff flow path to the VCT, RCP seal bleedoff flow can be sent to the PDT via the CTMT bleedoff flow relief valve.
D - WRONG: The potential for accelerated mass loss of a Small Break LOCA (TMI scenario) is real, but RCPs are required to be secured long before the initiation of OTC and RCS pressure reaches the SlAS initiation setpoint.
Reference E37-01-C, Loss Of All Feedwater (Training Mat.), Rev. 1, Ch. 2, Page 13 of 28 NRC KIA System/E/A NRC K/A Generic System 2.4 Emergency Procedure /Plan Emergency Procedures /Plan Number G 2.4.23 SEE GENERIC K/A Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.
Importance ROlSRO 2.8 3.8 10CFR Link (CFR: 41.10/45.13)
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Memory? (Check=Yesc The plant IS operating normally at 100% power. I&C personnel have removed the #I 'alternate' steam flow detector for a scheduled calibration. The #I steam flow selector switch was placed in the 'MAIN' position on C-05 prior to the start of the calibration.
Approximately 15 minutes after the start of the calibration, the SPO reports the following:
- SG LEVEL SET POINT DEVIATION HI/LO annunciator
- #I and #2 SG levels are lowering.
- #I SG main steam flow indication on the SGTR screen is lowering and appears to be caused by a failing detector.
Which of the following actions must the SPO take to stabilize the plant?
A Take manual control of #I FRV. Place BOTH SGFP speed controls in 'MANUAL' and maintain I/
FRV DIP to between 20 and 100 psid while manually restoring #I SG level.
, ~-
B Manually control #I SG FRV and raise the setpoint of #2 FRV, to restore SG levels to normal I-i level. Ensure the SGFPs automatically maintain the required delta-P.
c Take manual control of the #I FRV and the "A" SGFP. Restore #I SG level to 70% while I
I-!
operating the "A' SGFP in manual to maintain the required delta-P.
D Take manual control of BOTH SG FRVs and position them to restore SG levels. Open the CPF 0 bypass valve to raise SGFP suction pressure and compensate for the greater demand in feed flow.
Justification A - CORRECT: The steam flow detector is an input to automatic control of SGFP speed as well as automatic FRV position. ARP 25900 requires manual control of SGFP speeds and manual control of BOTH FRVs, until SG levels are stabilized, on a failure of the steam flow detector.
B -WRONG: There is nothing directly wrong with the #2 SG FRV control system. However, the # 1 FRV changing position due to the failure, along with the SGFPs changing speed, requires placing the #2 FRV in manual a necessary burden to the recovery of the plant. NOT placing both SGFPs in manual will severely jeopardize the mitigation of this event.
C -WRONG: BOTH SGFPs need to be manually controlled to mitigate this failure. Controlling only one will allow the other to shut down, and one SGFP cannot supply sufficient feed flow above 75% power.
D -WRONG: Controlling #2 SG FRV in manual is counter-productive. Opening the CPF bypass will raise SGFP suction pressure, but the control system will then LOWER SGFP speed even faster in an attempt to maintain a perceived lower need of delta-P.
Reference ARP-2590D (C05 Alarm Response), Rev. 002-12, D-16, Pg. 064 OP-2385 (Feedwater Control System Operation), Rev. 010-00, Page NRC WA System/E/A NRC K/A Generic System 2.4 Emergency Procedure /Plan Emergency Procedures /Plan Number G 2.4.50 SEE GENERIC K/A Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
Importance ROlSRO 3.3 3.3 IOCFR Link (CFR: 45.3)
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