ML042190268
| ML042190268 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/19/2004 |
| From: | Wilson M Dominion Nuclear Connecticut |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-423/04-301 | |
| Download: ML042190268 (140) | |
Text
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ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Question # 1 Group # 1 1 WA# EPE.007.GEN.2.4.2 Importance Rating 3.9 4.1 Proposed Question:
The reactor has tripped, and current conditions are as follows:
RCS pressure: 1880 psia and DECREASING SG A pressure: 800 psig and STABLE SG B pressure: 760 psig and DECREASING SG C pressure: 800 psig and STABLE SG D pressure: 800 psig and STABLE CTMT pressure: 18.5 psia and INCREASING Assuming NO additional operator actions have been taken, which actuations have occurred?
A. SIS and CIA ONLY.
B. SIS, CIA, and MSI ONLY.
C. SIS, CIA, CDA, and CIB ONLY.
D. SIS, CIA, MSI, CDA, and CIB.
- - : Proposed Answer: B Explanation (Optional): SIS has actuated on low PZR pressure (I892 psia), and SIS produces a CIA. A is wrong, and B is correct, since even though Main Steamline pressure is above the MSI setpoint, CTMT HI-2 MSI has actuated with CTMT pressure above 17.7 psia. C and D are wrong, since CTMT pressure is below the CDA actuation setpoint of 23 psia.
Technical Reference(s): E-0 Entry Conditions (Attach if not previously provided)
E-0, step 10 Proposed references to be provided to applicants during examination: None Learning MC-05493 Describe the operation of the following RPS controls and interlocks.. . ESF (As Objective: Actuation Signals.. . available)
Question Source: Bank #72398 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Question # 2 Group # 1 1 KJA # APE.008.AK2.02 Importance Rating 2.7 2.7 Proposed Question:
Initial Conditions:
0 The plant is in MODE 3 at normal operating temperature and pressure.
0 Charging flow control valve 3CHS*FCV121 is in MANUAL, maintaining 28% PZR level.
The reference (upper) leg tap for Pressurizer level transmitter 3RCS*LT460 breaks off of the pressurizer, and RCS pressure starts decreasing.
What will be the responses of PZR level instruments 3RCS*LI459 and LI460 to this event?
LT-459 LT-460 PZR Level Indication PZR Level Indication A. Increasing trend Offscale high B. Decreasing trend Offscale low C. Decreasing trend Offscale high D. Increasing trend Offscale low Proposed Answer: A Explanation (Optional): A is correct, since on a vapor space break, RCS pressure will drop, causing charging flow to increase. Pressure will drop to saturation in the vessel and the hot legs, and formation of a two-phase mixture will force flow up the surge line and into the pressurizer. This will cause actual pressurizer level, sensed by LT459 to increase until the PZR is full (B and C wrong). B and C are plausible, since a LOCA is in progress. A reference leg break will decrease pressure sensed by the reference side of LT-460, which appears as a high level (B and D wrong). B and D are plausible, since level indication would decrease on a variable leg break.
Technical Reference(s): Functional Drawing Sheet 1 1 (Attach if not previously provided)
P&ID 102C Proposed references to be provided to applicants during examination: None Learning MC-05342 given a failure, partial or complete, of the pressurizer pressure and level (As available)
Objective: control system, determine the effects on the system and on interrelated systems.
Question Source: Modified Bank 64307 Parent attached Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.7 Comments:
29 of 34 NUREG-I 021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5
-2 Original Bank Question # 64307 Pressurizer level transmitter LT459 is selected for control when its reference leg develops a slow leak.
Which of the following describe anticipated instrument or plant response?
LT-459 LT-460 & 461 PZR Level Indication PZR Level Indication VCT Level A. Increasing Decreasing Increasing B. Decreasing Increasing Increasing C. Increasing Decreasing Decreasing D. Decreasing Increasing Decreasing Answer: A 29 of 34 NUREG-?021,Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO L
Tier # I 1 Question # 3 Group # 1 I WA # EPE.009.GEN.2.4.7 Importance Rating 3.1 3.8 Proposed Question:
A small break LOCA is in progress, and the following sequence of events occurs:
- 1. With RCS pressure at 1700 psia, the crew transitions to ES-1.2 Post LOCA Cooldown and Depressurization.
- 2. The crew commences the cooldown of the RCS per ES-I .2, step 6 Initiate RCS Cooldown To Cold Shutdown
- 4. The RCS has adequate subcooling, and the crew stops the A Charging Pump.
- 6. The STA also reports that Pressurizer level is 25%.
Based on these conditions, what action will the crew take per ES-1.2?
A. Restart the A charging pump and transition to E-I Loss Of Reactor Or Secondary Coolant.
B. Verify one charging or SI pump is running and trip the B RCP.
C. Immediately trip the B RCP D. Continue on with ES-I .2.
Proposed Answer: D Explanation (Optional): D is correct, since RCS pressure is expected to drop during the cooldown. A LOCA is in progress, and the RCS is deliberately depressurized to maintain Pressurizer level. The goal is to get to cold shutdown and place RHR is service n the cooldown mode. Ais wrong, but plausible, since it is a misapplication of the SI reinitiation foldout page criteria. This criterion does not apply until SI has been terminated. E% and C are wrong, but plausible, since this is a misapplication of the RCP trip criteria, which is met (RCS pressure <I 500 psia, and a Charging or SI pump running). This is an action in E-I, but a cooldown is now in progress in ES-I .2, so the criterion does not apply. Part of the ES-I .2 strategy is to keep one RCP running.
Technical Reference(s): WOG Executive Volume, RCP Trip (Attach if not previously provided)
ES-1.2, steps 6, 11, and 12.
ES-1.2 Foldout Page Proposed references to be provided to applicants during examination: None Learning Objective: MC-05529 Describe the major action categories within EOP 35 ES-I .2. (As available)
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.10 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO 4
Tier # 1 1 Question # 4 Group # 1 1 KIA #
Importance Rating 2.9 3.1 Proposed Question:
With the plant operating at 8% power, feeding the SGs with the A TDMFP, the following sequence of events occurs:
The C RCP Hi-Hi vibration annunciator is received on MB4.
The crew enters AOP 3554 RCP Trip or Stopping A RCP At Power.
0 The C SG narrow range level is fed up to 70%.
The plant does not trip on the initial transient.
After the initial transient has stabilized, in what direction will the BOP operator throttle the C Feed Reg Bypass Valve in order to maintain the C SG level at 50% narrow range?
A. Closed due to increased steam pressure in the A, B, and D SGs.
B. Open due to increased steam pressure in the c SG.
C. Closed due to C SG temperature decreasing to approximately RCS T-cold.
D. Open due to C SG temperature increasing to approximately RCS T-hot.
Proposed Answer: C Explanation (Optional): When the C RCP stops, the running pumps will reverse flow through the idle C loop. The idle loop temperature will become very close to Tcold of the steaming loops ( D wrong). The affected steam generator will stop steaming due its pressure dropping as its temperature decreases to RCS T-cold (C correct, B wrong). This results in increased steam flow from the other 3 SGs, lowering their pressures (A wrong).
Technical Reference(s): AOP 3554, step 7 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-03349 FOR GIVEN PLANT CONDITIONS, QUALITATIVELY STATE THE (As available)
Objective: EFFECT OF ... RCP TRIP.. .
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.8 and41.10 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 I Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
~U Tier # 1 1 Question # 5 Group # 1 1 KIA # APE.022.AA2.03 Importance Rating 3.1 3.6 Proposed Question:
The plant is operating at 100% power with a blended makeup in progress. The RO observing the Primary Makeup System at Main Boards 3 and 4 observes the following:
0 The TOTAL MAKEUP FLOW DEVIATION annunciator is received on Main Board 3.
The Makeup To VCT Valve 3CHSFCVlllB is already closed.
0 The Makeup to Charging Pump Suction Valve 3CHS*FCVI 10B is stroking closed.
The Makeup Controller Mode transfers to OFF.
0 The Makeup Controller Control Switch transfers to STOP.
0 No other annunciators are received.
Why has this biended makeup to the VCT terminated?
A. Volume Control Tank (VCT) level has reached 54%.
B. Boric Acid Supply to Blender Valve 3CHS*FCV 110A has failed open.
C. PGS Supply to Blender Valve 3CHSFCVI 11A has failed open.
D. VCT Makeup Isolation Valve 3CHS*FCVlllB has failed closed.
Proposed Answer: C Explanation (Optional): A is wrong; since the makeup controller would return to AUTO, not STOP. A is plausible, since VCT level of 54% will terminate an auto-makeup. B is wrong, since a BORIC ACID FLOW DEVIATION annunciator would also have been received. Bis plausible, since boric acid inputs to the total flow. C is correct, since Total flow controller adjusts FCVl1 IA, and with this valve failed open, total flow would deviate from setpoint.
D is wrong, since FCVl10B is not opened for blended flow operations. It is opened during dilution operations. D is plausible, since the RO observed this valve to be closed, and if it had been open, total flow would change.
Technical Reference(s): OP 3353.MB3B, 3-7 (Attach if not previousiy provided)
OP 3304C, section 4.3 Proposed references to be provided to applicants during examination: None Learning MC-07018 Describe the operation of the cpntrols and interlocks associated with... (As available)
Objective: Primary Makeup Flow Control Valves.. .
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.6 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5
- Examination Outline Cross-reference:
Question # 6 Level Tier #
Group #
RO 1
I SRO 1
I KIA # APE.026.AAI .02 Importance Rating 3.2 3.3 Proposed Question:
The plant is operating at 100% power when the following sequence of events occurs:
- 1. MBIC, annunciator 5-7 RPCCW PUMP AUTO TRIP/OVERCURRENT alarms.
- 2. The RO reports that the A RPCCW pump has the amber and the green lights lit on MBI .
3 . The RO confirms that RPCCW flow indicates zero for the A Train RPCCW headers.
- 4. The C RPCCW pump is not available.
Which of the below Main Board operations is NOT required for this event?
A. Cross-connect the RPCCW Containment headers at MBI .
B. Shift the RCP seal return path to the top of the VCT at MB3.
C. Simultaneously close the charging header flow control valve and the letdown orifice isolation valve at MB3.
D. Isolate Auxiliary Steam to the Auxiliary Building at MB6.
Proposed Answer: B v
Explanation (Optional): A is wrong since RPCCW containment headers must be cross-tied to maintain cooling to RCPs. B is correct, since the seal water heat exchanger is cooled from the B RPCCW train. C is wrong, since the A RPCCW train cools the letdown heat exchanger. D is wrong, since cooling has been lost to the A RPCCW non-safety header, and relief valves may lift on associated Aux Steam equipment.
Technical Reference(s): AOP 3561, Att. A (Attach if not previously provided)
P&ID 121 A, B, and C Proposed references to be provided to applicants during examination: None Learning MC-04 155 Given the following failures.. . determine the effects on the RPCCW (As available)
Objective: system and on interrelated systems: High motor temperature...
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO L' Tier # 1 1 Question # 7 Group # 1 1 WA # APE.027.GEN.2.4.49 Proposed Question: Importance Rating 4.0 4.0 Initial Conditions:
0 The Unit is operating at 100% power 0 Pressurizer pressure control is in automatic The "A" set of backup Heaters is energized in "ON" 0 Actual pressurizer pressure is 2250 psia The Master Pressurizer Pressure Controller malfunctions and the setpoint is step changed from 2250 psia to 2 175 psia, and components reposition.
After placing the Pressurizer Master Pressure Controller to MANUAL, what action will the RO take with the Master Pressurizer Pressure Controller in response to the failure?
A. Push on the INCREASE pushbutton, closing both spray valves and energizing backup heaters "B", "D", and "E".
B. Push on the DECREASE pushbutton, closing both spray valves and energizing backup heaters "B","D", and "E".
C. Push on the INCREASE pushbutton, deenergizing backup heaters "B", "D",and "E", and opening both spray valves.
D. Push on the DECREASE pushbutton, deenergizing backup heaters "B", "D", and "E", and opening both spray valves.
Proposed Answer: A Explanation (Optional): With the controller attempting to maintain 2175 psia, spray valves will have opened and heaters will have deenergized. Actual pressure will have started dropping, so the RO will need to close spray valves and energize heaters ("'2" and "D' wrong). "C" and "D" are plausible since this would be the response if the controller had failed in the other direction. To raise pressure, the RO needs to go in the INCREASE direction ("A" correct, "B" and "D' wrong). "Byand " D are plausible since INCREASE can be misinterpreted to go OPEN on the spray valves.
Technical Reference(s): Process Drawing Sheet 26 (Attach if not previously provided)
Functional Drawing Sheets 11 & 12 Proposed references to be provided to applicants during examination: None Learning MC-05341 Describe the operation of the Pressurizer Pressure and Level Control (As available)
Objective: System under Normal, Abnormal, and Emergency Operating conditions.
Question Source: Modified Bank #71040 Parent Question Attached Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.1 0 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 L'
Original Question #7 1040 Initial Conditions:
The Unit is operating at 100%power Pressurizer pressure control is in automatic The "A" set of backup Heaters is energized in "ON" Actual pressurizer pressure is 2250 psia The pressurizer Master Pressure Controller malfunctions and the setpoint is step changed from 2250 psia to 2175 psia.
What will be the initial automatic responses in the pressurizer Pressure Control System as a result of this failure?
A. PORV PCV-455A opens, and both spray valves open.
B. PORV PCV-456 opens, and both spray valves open.
C. Both spray valves close and pressurizer backup heaters "B", "D",and "E" energize.
D. Both spray valves open.
Answer: D 29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO L,
Tier # 1 1 Question # 8 Group I# 1 1 KIA # EPE.029.EA2.09 lrnportance Rating 4.4 4.5 Proposed Question:
Initial Conditions:
- 1. A Safety Injection has occurred.
- 2. The crew has been progressing through the EOP network.
Current conditions:
A valid ORANGE path is received on the Subcriticality status tree.
The crew is entering FR-S.1 Response to Nuclear Power GeneratiodATWS.
Reactor Trip and Bypass Breakers are OPEN.
DRPI shows all rods on the bottom.
Intermediate Range Startup Rate is positive.
The Source Ranges have not energized.
All 4 Turbine Stop Valves are OPEN.
The A, B , and c Turbine Control Valves are closed, and the D Control Valve is stuck OPEN.
The B, C,and D MSIVs are closed, and the A MSIV is stuck OPEN.
As defined in the immediate actions of E-0 and FR-S.1, what is the status of the reactor and the turbine?
\.-
A. Both the reactor and the turbine are tripped.
B. The reactor is tripped, but the turbine is NOT tripped.
C. The reactor is.NOT tripped, but the turbine is tripped.
D. Both the reactor and the turbine are NOT tripped.
Proposed Answer: B Explanation (Optional): The reactor is tripped, since two of the three reactor-trip criteria are met (B correct, C and D are wrong). C and D are plausible, since power is increasing. The turbine is not tripped (Bcorrect, A and C are wrong). A and C are plausible, since the stuck MSIV and control valve have different alphanumeric labels.
The turbine would still be receiving steam since the 4 main steam lines join a common header before going to the turbine.
Technical Reference(s): FR-S.l, steps 1 and 2 (Attach if not previously provided)
E-0, note prior to step 1 Proposed references to be provided to applicants during examination: None Learning Objective: MC-04625 Describe the major action categories within EOP 35 FR-S.I. (As available)
Question History: New Question Cognitive LeveI: Comprehension or Analysis 10 CFR Part 55 Content: 55.41 . I O 55.43.5 Comments:
L-29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.W Tier # 1 I Question # 9 Group # 1 1 WA # EPE.0 11.EA2.03 Importance Rating 3.7 4.2 Proposed Question:
A LOCA occurs, resulting in a CDA. Current conditions are as follows:
The crew is in ES-1.2 Post LOCA Cooldown and Depressurization.
RCS pressure is 650 psia and stable.
Containment pressure is I6 psia and lowering.
No RCPs are running.
The crew has reset CDA.
A fault in the CDA reset circuitry has prevented any RPCCW pump from starting.
All other functions associated with the CDA reset have operated properly.
Technical Support is investigating options for circuit repair and local RPCCW breaker operation.
Assuming an RPCCW pump CANNOT be started, which action would the crew still be able to perform?
A. Startup Spent Fuel Pool Cooling per ES-I .2, step 3 Check Electrical Alignment, using GA-5 Spent Fuel Pool Cooling System Startup.
B. Start the B RCP per ES-I .2, step 12 Check If RCP(s) Should Be Started, using GA-6 Starting Reactor Coolant Pump.
1-C. Place a Residual Heat Removal pump in service per ES-1.2, step 29 Check if RHR System Can Be Placed In Service.
D. Transfer to Cold Leg Recirculation per ES-I .3, step 3 Align RHR and Recirc Spray Systems For Cold Leg Recirculation.
Proposed Answer: D Explanation (Optional): D is correct, since cooling for cold leg recirculation is provided by service water, and no operations are performed on the RPCCW system during the swap over to coId leg recirculation. A is wrong, since RFCCW provides cooIing to the spent fuel pool cooling heat exchangers. B is wrong, since GA-6 requires RPCCW in order to start an RCP. C is wrong, since ES-1.2 step 29 places RHR in service in the cooldown mode, which uses RPCCW for cooling.
Technical Reference(s): ES-1.2, steps 3.f. RNO, 12.d, and 29. (Attach if not previously provided)
ES-1.3 step 3.
GA-5. steu 3 and 4.
GA-6, step 1 .
Proposed references to be provided to appiicants during examination: None Learning Objective: MC-05529 Describe the major action categories within EOP 35 ES-1.2. (As availabIe)
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.8 and 55.43.5.
Comments:
29 of 34 NUREG-1021, Revision 9
~~
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Question # 10 Group # 1 1 WA # APE.040.A2.03 Importance Rating 4.6 4.7 Proposed Question:
PARAMETER: CURRENT VALUE: TREND:
Reactor power: 58% Increasing RCS pressure: 2225 PSIA Decreasing Auctioneered high Tavg: 569°F Decreasing Turbine power: 595 MWE Decreasing Containment pressure: 15 PSIA Increasing Based on the plant conditions, what event is in progress?
A. Small break RCS LOCA B. Steamline break C. RCS dilution event D. Steam generator tube rupture
.-.' Proposed Answer: B Explanation (Optional): Reactor power is increasing, indicating positive reactivity event ("A" and "D" wrong). Electric load is decreasing, indicating loss of steam to the turbine ("B" correct, "C" wrong).
Technical Reference(s): FSAR Chapter 15. I .3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-04881 DESCRIBE the major parameter changes associated with increased heat (As available)
Objective: removal by the Secondary System.
Question Source: Bank #64268 Question History: Millstone 3 2002 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Question # 11 Group # 1 1 WA # APE.054.AA2.04 Importance Rating 4.2 4.3 Proposed Question:
Initial Conditions:
Power is steady at 11% power.
0 The crew is starting up the plant per OP 3203 Plant Startup.
0 All 4 Steam Generators are being fed by the A TDMFP.
The A Feed Regulating Bypass Valve fails closed, resulting in an automatic reactor trip.
Before the BOP operator takes any manual actions, which AFW pumps will he observe to be in service?
A. The A MDAFW Pump only, providing throttled flow to the A and J3 SGs.
B. The TDAFW Pump only, providing throttled flow to all 4 SGs.
C. Both MDAFW Pumps only, providing full flow to all 4 SGs.
D. All 3 AFW Pumps, providing full flow to all 4 SGs.
-v Proposed Answer: C Explanation (Optional): With a loss of feed to only the A SG, an A SG 10-10 level reactor trip will occur. A 10-10 level in one SG sends a start signal to both MDAFW pumps (Cycorrect). Level in only the ASG will decrease to the 10-10 level setpoint, since the AFW flow control valves are normally kept fully open (Aand B wrong), and power is initially too low for significant shrink to occur. A is plausible, since the A MDAFW pump feeds the A SG, while the Bydoes not. B is plausible, since the TDAFW pump starts on 10-10 level in 2 SGs. D is plausible, since on a reactor trip from high power levels, shrink will result in the start of all 3 AFW pumps. A and By are also pIausible since AFW flow control valves are throttled to control flow when in service at low powers when feeding with AFW.
Technical Reference(s): Functional Sheet 15 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-04635 Describe the proper operation of the following Auxiliary Feedwater (As available)
Objective: System components, controls, and interlocks.. . Motor Driven Auxiliary Feed Pumps, Turbine Driven Auxiliarv Feed Pumn.. . Auxiliarv Feed Control Valves.. .
Question Source: Modified 69476 Parent question attached.
Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Original 69476 The crew is starting up the plant per OP 3203 Plant Startup when the following sequence of events occurs:
- 1. Power was steady at 8% power.
2 . A steam break occurs on the "A" S/G steam header inside containment
- 3. Steam pressure in the "A" S/G rapidly drops to 300 psig
- 4. Containment pressure rises to 18 psia
- 5. All safeguards equipment operates as designed How many AFW pumps should be in service? (Assume no operator action)
A. No AFW Pumps.
B. The TDAFW Pump only.
C. Both MDAFW Pumps only.
D. AI1 3 AFW Pumps.
Answer: C 29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO U
Tier # 1 1 Question # 12 Group # 1 1 KIA # EPE.055.EK3.02 Importance Rating 4.3 4.6 Proposed Question:
The crew is carrying out actions in ECA-0.0 Loss of All AC Power, and an operator is dispatched to locally close the RCP Seal Water Return CTMT Outer Isolation Valve 3CHS*MV8100.
What is the basis for isolating the seal water return path?
A. Prevent thermal shock of the RCP seals.
B. Prevent the formation of steam in the WCCW System.
C. Prevent overpressurizing the PRT, potentially releasing radioactivity in Containment.
D. Prevent overfilling the VCT, potentially releasing radioactivity in the Auxiliary Building.
Proposed Answer: D Explanation (Optional): A is wrong, but plausible, since this is the basis for closing the RCP Seal Supply Isolation Valves. Byis wrong, but plausible since this is the basis for closing the RPCCW CTMT return outer isolation valves.
--.- C is wrong, since seal water returns to the VCT, not the PRT, but plausible, since the PRT would release radiation to CTMT if it is overfilled. D is correct since seal return is normally aligned to the VCT, charging pumps are no longer removing water from the VCT, and if the VCT overfills, it could release radioactivity into the Aux Bldg.
Technical Reference(s): ECA-0.0, step 8 (Attach if not previously provided)
ECA-0.0 ERG Bkgd doc for step 8 Proposed references to be provided to applicants during examination: None Learning MC-03852 Discuss the basis of major procedure steps and/or sequence of steps in (As available)
Objective: EOP 35 ECA-0.0.
Question Source: Bank #67592 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5 and41.10 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
\=,
Tier # I 1 Question # 13 Group # 1 1 WA # APE.OS7.AA1.03 Importance Rating 3.6 3.6 Proposed Question:
Initial Conditions:
0 The plant is at 80% power.
All MB5 SGWLC related inputs are selected to Channel 1.
Both TDMFPs are in AUTO.
The following sequence of events occurs:
I. VIAC 1 deenergizes.
- 2. The BOP operator quickly places a11 4 Feed Regulating Valve controllers to MANUAL.
What action, if any, will the BOP operator need to take to control TDMFP speed?
A. Take the feed pump master speed controller 3FWS-SK509A to MANUAL and depress the RAISE pushbutton.
B. Take the toggles for both TDMFP Dahl controllers 3FWS-SK46A and B to the RAISE direction.
C. Take the Manual Speed Controllers 3TFC-MIA and B for both TDMFPs to the RAISE direction.
D. No action is required for TDMFP speed control, since no inputs for TDMFP control have been affected.
Proposed Answer: A Explanation (Optional): A is correct since the master speed controller is used when both TDMFPs are in auto. Going to RAISE will increase both TDMFP speeds, restoring SG level. Bis wrong, since the DAHL controllers provide coarse-adjust, and SGs are being only slightly under-fed. Also, the DAHL controllers need to be taken to MANUAL in order for the toggle switches to hnction. C is wrong, since the lower set of the Dahl speed controller and the manual speed controller controls feed pump speed, and the manual speed changer is already at the high-speed stop to allow full range of control by the Dahl controller. D is wrong since Feed Pump Speed control receives and input from total steam flow to determine program DP.
Technical Reference(s): OP 3321 Steps 4.4.34 and 35 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-04660 Describe the operation of the following Main Feedwater & SGWLC (As available)
Objective: controls and interlocks.. . TDMFP Master Speed Controller.. .
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 I .7 Comments:
29 of 34 NUREG-I021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Question # 14 Group # 1 1 WA # APE.058.AK1.01 Importance Rating 2.8 3.1 Proposed Question:
A loss of DC Bus 3 has occurred, and the crew has entered AOP 3563 Loss of DC Bus Power.
A PEO locally attempts to reenergize DC bus 3 by placing the standby charger on the deenergized DC bus.
Which component could be damaged by the PEOs improper action?
A. Thebattery.
B. The 120V inverter.
C. The Charger Rectifier stack.
D. The Charger output breaker.
Proposed Answer: C Explanation (Optional): OP 3345C, Precaution 3.6 states that a battery charger shall not be placed in service on a deenergized bus, since possible damage to the rectifier stack may occur.
./ Technical Reference(s): OP3345C, Precautions (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-03307 Describe the major administrative or procedural precautions and (As avaiIabIe)
Objective: limitations associated with the 125 VDC Distribution System, including the basis for each, identified within the following procedures... OP 3345C Question Source: Bank #68020 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.8 and 41.10 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
..- Tier # I 1 Question # 1.5 Group # 1 I KIA # APE.062.AK3.02 Importance Rating 3.6 3.9 Proposed Question:
The plant is operating at 100% power when a spurious CDA signal is received; and the reactor does NOT trip.
What Service Water System loads have isolated, providing immediate concerns during this event?
A. RPCCW Heat Exchangers and Control Building Chillers.
B. RPCCW and TPCCW Heat Exchangers.
C. Control Building Chillers and EDG Heat Exchangers.
D. EDG Heat Exchangers and TPCCW Heat Exchangers.
Proposed Answer: B Explanation (Optional): The following isolate on a CDA: Circ Water Lube Water Supply, RPCCW heat exchangers, TPCCW heat exchangers (B correct), and Hypochlorite Dilution. The following open on a CDA: EDG Cooling Water (C and D incorrect) and CTMT Recirc Heat Exchangers. Control Building chillers are not affected (A and C incorrect), but are plausible, since Service Water cools them.
Technical Reference(s): P&ID 133a-d. (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-05714 Describe the operation of the following Service Water System components (As Objective: controls and interlocks.. . available)
A. EDG Service Water Outlet Valves (SWP*AOV39A/B)
B. F. TPCCW Heat Exchanger Service Water Supply Valves (SWP*MOV71A/B)...
G. MCC & Rod Control ACU Service Water Booster Pumps (SWP*P3A/B)
H. FWCCW Heat Exchanger Isolation Valves (SWP*MOVSOA/B)
Question Source: Modified Bank #69682 Parent question attached.
Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.4and41.8 Comments:
29 of 34 NUREG-1021, Revision 9
~~
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Original Bank #69682
+-.-
Which of the following Service Water System loads isolate during a Containment Depressurization Actuation (CDA)?
A. Containment Recirc Coolers B. RPCCW Heat Exchanger C. EDG Heat Exchanger D. Control Building AC Water Chillers Answer: B 29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier ## 1 1 Question # 16 Group # 1 1 K/A ## APE.065.Al.03 Importance Rating 2.9 3.1 Proposed Question:
With the plant at 100% power, a significant instrument air leak occurs, and the following sequence of events occurs:
- 1. Air pressure starts to drop at a moderate rate, and the crew enters AOP 3562 Loss of instrument Air,
- 2. A PEO reports that the leak location has been found, and the leak can be isolated.
- 3. Several components start to reposition, but the crew is NOT required to trip the reactor.
- 4. The leak is isolated, and instrument air pressure recovers to 1 10 psig.
In accordance with AOP 3562, what action will the operators take now that air pressure has been restored?
A. Using 3304A Charging and Letdown, restore normal charging and letdown.
B. Using OP 3321 Main Feedwater, close the Main Feed Pump Recirc VaIves.
C. Take the PZR Master Pressure Controller to MANUAL and set it for 0% demand.
D. Locally place the control switches for the traveling screens to SLOW 1 Proposed Answer: A Explanation (Optional): A is correct, since letdown isolates on a loss of instrument air, and with cooling lost to the CVCS Regenerative Heat Exchanger, the crew is directed to isolate charging as well. AOP 3562 directs the restoration of charging and letdown when air pressure is restored. B is wrong, since the crew would have been required to trip the reactor if feed control had been lost. B is pIausible, since the Feed Pump Recirc Valves fail open on loss of air pressure, and AOP 3562 directs the crew to restore Main Feedwater, if required. C is wrong, but plausible, since the spray valves fail closed on a loss of air, but AOP 3562 allows the valves to respond as needed on restoration of air pressure. D is wrong since the crew places the screens in SLOW on loss of air pressure, not during the recovery of air pressure. D is plausible, since traveling screen DP instruments are pneumatic, and may not indicate properly on a loss of air pressure.
Technical Reference(s): AOP 3562, steps 1,4, 10, 12, 13, and 15. (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-07008 The crew will demonstrate the ability to safely operate the plant during a (As available)
Objective: loss of instrument air.. .
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Level RO SRO
-- Tier # 1 1 Question # 17 ic+e,J.it;..l .-e&
Group # 1 1 OU$JQ cTAT K/A# WE04.EK2.1 b y ; , d r ;&&I?&$ Importance Rating 3.5 3.9 Proposed Que:
r e -&,vt.. fiGq.4
$ e A b e, ,
The plant has tripped due to a LOCA outside Containment, and the following sequence of events occurs:
The crew commences verifjhg proper valve alignment in the ESF Building per ECA-1.2 LOCA Outside Containment.
The RO reports that RHR Loop Suction Isolation Valve 3RHS*MV8701B is OPEN.
0 The STA reports all other ECCS valves are in their proper lineup.
The US directs the RO to close 3RHS*MV8701B.
The RO improperly places the switch for 3RHS*MV8701A to OPEN, but the valve remains closed.
Which interlock prevented RHR Loop Suction Isolation Valve 3RHS*MV8701A ffom opening?
A. RHR to Charging/SI Valve 3SIL*MV8804A is CLOSED.
B. RSS to RHR Cross-connect Valve 3RSS*MV8837A is CLOSED.
C. RHR Loop Suction Isolation Valve 3SIL*MVX701B is OPEN.
D. RHR RWST Suction Isolation Valve 3RHS*MV8812A is OPEN.
Proposed Answer: D Explanation (Optional): The interlocks required to OPEN 3RHS*MV8701A are 3RHS*MV8812A CLOSED (D correct), RHR to Charging/SI Valve 3SIL*MV8804A CLOSED (A wrong), and RSS to RHR Cross-connect Valves 3RSS*MV8837A and B CLOSED ( B wrong). C is wrong since 3RHS*MV8701B is not interlocked with 3RHS*MV8701A. A and B are plausible since these valves are interlocked with 3RHS*MV8701A. C is plausible, since the RO has reported this valve out of its expected position for this event.
Technical Reference(s): LSK 27-7D (Attach if not previously provided)
ECA-1.2, step 3 Proposed references to be provided to applicants during examination: None Learning MC-05455 Describe the operation of the following Residual Heat Removal System (As available)
Objective: equipment Controls and Interlocks.. . Loop Suction Valves.. . RHR to RCS Cold Leg Injection Valves.. . RHR to CHS/SIH Pump Supply Valves.. .
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5 and41.10 Comments:
29 of 34 NUREG-1021, Revision 9
COWDlTlD11 RESULTANT (NOTB 5)
I. REFER TO WK-O-S-SM2s(nowl)NOmWLa3.
- 2. RECORDER FOR LOOP 1 AND Y HUT LEO PRESSURE EsmCmJB I S COulMl (REFER TO LSK-27-7B W 9 5IC.Z)).
8RIISWKV8701A SRHS *MW701 A as VALIE IS KEVLOCKED CLOSW TO ASSURE PROPER SYSTEU ALIOIMENT DURINB CWITAlmEllT ISOUTION.
KEY REMDVABLE 111 CENTER POSITION 0111'1.
- 9. VALVE IS TORQUE SEATED WlEM CLOSED WITH CLOSE SWITCH.
3RCSCPTW6 AND(IPTW6A ARE NORMAL S P R l r RETUIIM WIMTAMED SPRINO RETURN FIRE TRAMSFER WITCH PANEL I S ALSO A SOURCE DEVICE
I .
W C A OUTSIDE EOP 35 ECA- 1.2 Page 5 of 7 CONTAINMENT Rev. 007 STEP ACTION/MPECTED RESPONS RESPONSE NOT OBTAINED 1 I I Consult with the ADTS prior to dispatching personnel to locally
- 3. Verify Proper Valve Alignment In CLOSE valves.
ESF Building Evalves can NOT be closed,
-a. valves verify RHR suction isolation
- CLOSED THEN Locally Close valves for minimum safety function.
3RHS*MV8701A 3RHS*MV8701B 3RHS*MV8701C 3RHS*MV8702A 3RHS*Mv8702B 3RHSMV8702C
-b. Verify RHR hot leg injection valve (3SIL*MV8840) -
CLOSED
-C. Verify SI pump hot leg injection valves -
CLOSED 3SIH*MV8802A 3SIH*MV8802B
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level Tier #
RO 1
SRO 1
Question # 18 Group # 1 1 KIA # WE05.EK2.1 Importance Rating 3.7 3.9 Proposed Question:
The plant has tripped from 100% power, and the following sequence of events has occurred:
- 1. The crew enters FR-H. 1 Loss ofSecondary Heat Sink from E-0, step 4.
- 2. The crew restores feed from the Motor Driven Main Feed Pump.
- 3. The crew feeds the Steam Generators at an excessive rate, causing a Safety Injection actuation.
What actions will the crew need to take in order to allow resetting the Feedwater Isolation signal and restoring main feed from the Main Boards?
A. Reset SIS only.
B. Reset P-4 only.
C. Reset SIS and pull the RPS universal logic cards.
D. Reset P-4 and pull the RPS universal logic cards.
Proposed Answer: C
-- r Explanation (Optional): SIS generates a FWI signaI. The SIS signal combines with the P-4 signal (B plausible) to lock in the FWI. SIS must be reset (B and D wrong) in order to reset the FWI, and either P-4 must be reset, or the universal logic cards must be pulled to clear the lock-in feature of the FWI (A wrong, D plausible, and C correct).
Technical Reference(s): Functional Drawing 13 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None
~~~~i~~ MC-072 19 Describe the operation of the following Main Feedwater Systems (As available)
Objective: Controls & Interlocks.. , Safety Injection Signal Actuation (as related to the Main Feedwater System); Feedwater Isolation Reset Switches (MB2 & MB5). ..
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 I Question # 19 Group # 2 2 WA # APE.OOSK2.01 Importance Rating 2.5 2.5 Proposed Question:
With the plant initially at 100% power, the following sequence of events occurs:
- 1. DRPI for Control Bank "D" group 1 rod M-12 drops from 228 steps to 174 steps.
- 2. A ROD POSITION DEVIATION annunciator is received on MB4.
- 3. The crew enters AOP 3552 MaZ@nction of the Rod Drive System.
- 4. The crew is currently using AOP 3552, Attachment A for a misaligned rod.
- 5. I&C reports that the stationary gripper coil fuse is blown for rod M-12.
6 . I&C reports that a replacement fuse is available.
- 7. The rod has been misaligned for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- 8. No other annunciators are lit.
What is the status of rod M- 12?
A. Rod M-12 has experienced a rod position indication malfunction.
B. Rod M-12 trippability still needs to be determined.
C. Rod M-12 is ready for blown fuse replacement, and realignment.
\.-
D. Rod M-12 is stuck, so the plant needs to be shutdown.
Proposed Answer: D Explanation (Optional): This question is related the Farley Event in October, 2002, where a rod did not hlly insert during rod drop testing. With a blown stationary gripper coil fuse, and no rod motion demand signal present, the control system is not sending a signal to hold the rod up. The rod is stuck. "D" correct, since the note prior to step 4 in AOP 3552 states that if at any time a rod is determined to be untrippable, proceed to Attachment F, which shuts down the plant. "A" is wrong, since the stationary gripper coil fuse has blown, and no other annunciators are present. "A" is pIausible, since RPI for the affected rod has acted in an unusual manner. "B" is wrong, but plausible, since this action would be appropriate for the misaligned rod if the crew did not have sufficient information available to determine that the rod was stuck. "C" is wrong, but plausible, since this would be the appropriate action if the misalignment were due to a control system failure.
Technical Reference(s): AOP 3552, NOTE prior to step 4. (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None
~~~~i~~ MC-03902 Discuss the basis of major precautions, procedure steps andor sequence (As available)
Objective: of steps within AOP 3552.
Question Source: Bank #77446 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.-~-.
Tier ## 1 I Question # 20 Group ## 2 2 WA ## APE.024.GEN.2.4.45 Importance Rating 3.3 3.6 Proposed Question:
Initial Conditions:
0 One hour ago, the plant reached normal operating temperature and pressure after a refueling outage.
0 Tave has been 557°F and steady.
0 The reactor trip breakers are open.
0 Battery Charger 3 is supplying DC Bus 3.
0 Work is being conducted near 480V load center 32R.
The following sequence of events occurs:
- 1. Load Center 32R deenergizes, and both the MCC LOSS OF CONTROL POWER and the INVERTER TROUBLE annunciators are received on MB8.
- 2. An RMS TROUBLE annunciator is received on MB2.
- 3. The extra operator begins addressing the loss of 32R, while the ROs begin reviewing which equipment was lost.
- 4. 60 seconds after the loss of 32R, the STA reports that neutron flux has been unexpectedly increasing and the SHUTDOWN MARGIN MONITOR CHANNEL I and 2 annunciators are received on MB3.
- 5. Tave starts increasing, and a TAV/TREF DEVIATION annunciator is received on MB4.
At this point, which annunciator is the first priority for the crew?
b-A. RMS TROUBLE, since power has been lost to radiation monitors required by Technical Specifications.
B. INVERTER TROUBLE, since Inverter 3 has lost its normal power supply, and DC Bus 3 has lost its Charger.
C. TAV/TREF DEVIATION, since a 15-minutesurveillance is required with this annunciator lit.
D. SHUTDOWN MARGIN MONITOR, since AOP 3566 Immediate Borution must be entered.
Proposed Answer: D Explanation (Optional): D is correct, since this ARP directs the operators to AOP 3566 Immediate Boration, and an unexplained positive reactivity event is in progress. A, B , and C are wrong, since no AOP entry conditions are met based on these annunciators. A, B, and C are plausible, since all of these annunciators are expected for this event, and the concerns listed in the distractors are valid.
Technical Reference(s): OP 3353.MB4C, 2-2. (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MC-03960 Identify plant conditions that require entry into AOP-3566... (As available)
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 I Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Iv-Tier # 1 1 Question # 2 1 Group # 2 2 KIA # APE.033.A2.03 Importance Rating 2.8 3.1 Proposed Question:
Power is steady at lo- amps in the intermediate range when the following sequence of events occurs:
I. I&C requests to make an adjustment on IRNI channel N35.
- 2. As a precaution, the crew places the level trip bypass switch for N35 in the Bypass position.
- 3. The I&C Technician inadvertently creates a short in the Channel N35 IRNI drawer, and a fuse blows.
- 4. No trip bistables are lit on the Channel N35 IRNI drawer.
Which h s e has blown, and did the reactor trip on high IRNI current?
A. An instrument power fuse has blown, and the reactor did NOT trip.
B. An instrument power fuse has blown, and the reactor did trip.
C. A control power fuse has blown, and the reactor did NOT trip.
D. A control power h s e has blown, and the reactor did trip.
Proposed Answer: D Explanation (Optional): A control power fuse has blown, since control power supplies the bistable lights, which are not lit (A and B wrong). The reactor did trip, since control power also supplies the current to RPS via the bypass switch (Dcorrect, C wrong).
Technical Reference(s): Functional Sheet 3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-05229 For the following conditions, determine the effects on the NIS system (As available)
Objective: and on interrelated systems.. . Intermediate range instrument failure below P-10, Intermediate range instrument failure above P-IO... Blown power range control power fuse, Blown power range instrument power fuse...
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 I Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 I Question # 22 Group # 2 2 K/A # 036.K3.02 Importance Rating 2.9 3.6 Proposed Question:
What is the purpose of the upload limit interlock on the new fuel elevator?
A. Prevent dropping a new fuel assembly from excessive heights should the elevator cable fail.
B. Prevent raising an irradiated fuel assembly so that the minimum required depth of water shielding is maintained.
C. Prevent raising the elevator guide rollers out of the water, since the rollers are water lubricated.
D. Prevent raising the elevator when either cable binding is taking place or an obstruction is in the carts path.
Proposed Answer: B Explanation (Optional): By is correct, since the upload limit interlock stops upward motion of the erevator when loaded with a fuel assembly. Handling equipment used to raise and lower spent fuel also has a limited maximum lift height so that the minimum required depth of water shielding is maintained. A and C are plausible, since they deal with reasons for limiting upward travel. Also, lubrication is a concern when raising the rollers above existing water level.
D is plausible, since it deals with limiting travel when load is sensed.
--. Technical Reference(s): Westinghouse reheling text, page I4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-0454 1 Describe the operation of the following Fuel Handling System (As available)
Objective: equipment, controk, and interlocks... New fuel elevator...
Question Source: Bank 74350 Question History: 2001 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5 and 41.10 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO V
Tier # 1 1 Question # 23 Group ## 2 2 WA # APE.060.K3.03 Importance Rating 3.8 4.2 Proposed Question:
With the plant at 100% power, the following sequence of events occurs:
The B Train RPCCW non-safety header isolates.
The Boron Evaporator Condenser Inlet relief valve 3BRS-RV27 lifts.
The crew enters AOP 3573 Radiation Monitor Alarm Response.
Why does AOP 3573 direct the crew to start one train of SLCRS?
A. SLCRS will place the Auxiliary, Control, and ESF Buildings under a slight negative pressure, reducing the radiation release to the environment.
B. SLCRS should have AUTO-STARTED, and the AOP directs the crew to manually initiate any automatic actions that did not occur.
C. SLCRS will redirect the release from an monitored release path to a path monitored by stack radiation monitor 3GWS-RE48.
D. SLCRS will redirect the release from an unfiltered release path to a filtered release path.
Proposed Answer: D Explanation (Optional): D is correct, since SLCRS fans draw air through the SLCRS filters, reducing the radiation release. A is wrong, since SLCRS does not take suction on the Control Building, but plausible since SLCRS does reduce pressure in the buildings it draws on, reducing the radiation release. B is wrong, since SLCRS fans do not receive an auto-start signal on high radiation, but plausible, since the AOP directs operators to take manual action for automatic actions that did not occur, but the. C is wrong since 3HVR-RE10 monitors the Turbine Building stack, but plausible since the SLCRS path is a monitored release path, monitored by 3GWS-RE48.
Technical Reference(s): AOP 3573, step 2b.RNO and Att. A, page 7 of 12. (Attach if not previously provided)
P&ID 108B Proposed references to be provided to applicants during examination: None Learning MC-03019 The crew will correctly assess abnormal plant radiation conditions and, (As available)
Objective: using AOP 3573 Radiation Monitor Alarm Response, perform required actions to minimize the effects of the abnormal radiation condition.
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5 and41.10 Comments:
29 of 34 NUREG-I 021, Revision 9
~__________
ES-401 Written Examination Question Worksheet Form ES-401-5
.._-. Examination Outline Cross-reference: Level RO SRO Tier # I I Question # 24 Group # 2 2 WA # WE06.EK3.2 Importance Rating 3.5 4.0 Proposed Question:
Current Conditions:
The plant is experiencing an inadequate core cooling condition.
The crew has entered FR-C. 1 Response to Inadequate Core Cooling.
The crew is about to depressurize the secondary plant.
What is the primary purpose of depressurizingthe steam generators under these conditions?
A. To cooldown and depressurize the RCS to enhance high head safety injection.
B. To cooldown and depressurize the RCS to facilitate accumulator and RHR injection.
C. To collapse steam voids in the Steam Generator U-tubes to reestablish natural circulation.
D. To increase the Thot-Tcold Delta T, to increase thermal driving head for natural circulation.
Proposed Answer: B
-2 Explanation (Optional): RCS pressure must be reduced in order for the SI accumuIators and low-head SI (RHR) pumps to inject. The most effective way to reduce RCS pressure is a rapid secondary depressurization. (B is correct). It is assumed that high head injection is not functioning since high head injection would clear the 1200° F condition (A wrong). C and D are wrong, since natural circulation flow is not the problem, but rather a loss of RCS inventory with RCS pressure being maintained above LPSI shutoff head by a steam bubble in the vessel.
Technical Reference(s): WOG Background Doc. far FR-C.1, step 11 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-04525 Discuss the basis of major procedure steps and/or sequence of steps in (As available)
Objective: EOP 35 FR-C.l Question Source: Bank #65023 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5 and41.10 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Question # 25 Group # 2 2 WA # WlE02.EAI .3 Importance Rating 3.8 4.0 Proposed Question:
Current Conditions:
Safety Injection has actuated.
RCS pressure is 1660 psia and increasing.
CETCs are 565°F and slowly decreasing.
Pressurizer level is 18% and slowly increasing.
SG levels are 0% narrow range.
AFW flow is 600 gpm total.
Containment Temperature is 140OF.
The crew is in E- 1 Loss of Reactor or Secondary Coolant.
The crew is assessing whether SI Termination Criteria are met.
Is SI Termination Criteria met, and if not, why not?
A. Yes, SI Termination Criteria is met.
B. No, inadequate subcooling exists.
-..- C. No, inadequate heat sink exists.
D. No, inadequate RCS pressure exists.
Proposed Answer: A Explanation (Optional): A is correct since SI Termination criteria is met. B is wrong, since actual subcooling is about 45F, and minimum required is 32. C is wrong, since, although SG NR levels are 0%, all equipment operated as expected, so adequate AFW flow exists. D is wrong, since no specific setpoint exists for RCS pressure. RCS pressure must simply be stable or increasing. D is plausible, since RCS pressure is below the SI auto-actuation setpoint of I892 psia.
Technical Reference(s): E- 1, step 6 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: MC-05523 Identify plant conditions that require entry into EOP 35 ES-1.1. (As available)
Question Source: New Question Cognitive Level: Comprehension or AnaIysis 10 CFR Part 55 Content: 55.41.7 and 41.10 Comments:
29 of 34 NUREG-I021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Question ## 26 Group # 2 2 KIA ## EPE.W E 15.EA 1.1 Importance Rating 2.9 3.O Proposed Question:
Current conditions:
0 With the plant initially at 100% power, a large break LOCA occurs in CTMT, and CDA actuates.
The crew has entered EOP 35 FR-Z.2 Response To CTMT Flooding, and is trying to identify and isolate the sources of water to the CTMT sump per FR-2.2 step 1.
0 The RO has been directed to check the Steam Generator Blowdown, Fuel Pool Cooling and Purification, and Fire Protection Water systems as potential water sources into Containment.
How will the crew check these three paths, and which of these paths, if any, should still need to be isolated from CTMT?
A. The CTMT isolation valves for all three systems can be verified closed at MB 1. All three paths should already be isolated.
B. The SFC path can only be verified locally, and the FPW CTMT outer isolation valve has a bypass valve that can only be verified locally. All three paths should already be isolated.
C. The CTMT isolation valves for all three systems can be verified closed at MBI . The Blowdown System still needs
-\_.--
to be isolated.
D. The FPW path can only be verified locally. The FPW CTMT path still needs to be isolated, requiring a local manual CTMT isolation valve to be closed at the CTMT penetration area.
Proposed Answer: B Explanation (Optional): The BDG and FPW systems automatically isolate from CTMT on a SIS/CIA, and can be verified CLOSED on MBl. The FPW system has a manual bypass valve (3FPW*V666) around the CTMT isolation valve, and the SFC system has manual isolation valves that can only be checked locally (A and C wrong). The FPW and SFC manual valves are procedurally kept closed in MODE 1 (B is correct and D wrong).
Technical Reference(s): FR-2.2, step 1. (Attach if not previously provided)
P&ID I 1 lA, 123A, and 146B Proposed references to be provided to applicants during examination: None Learning MC-05993DISCUSS the basis of major procedure steps and/or sequence of steps in (As available)
Objective: EOP 35 FR-Z.2.
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.7 Comments:
29 of 34 NUREG-I021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO e
Tier # 1 1 Question ## 27 Group # 2 2 KIA # EPE.W/EI 6.EA 1.2 Importance Rating 2.9 3.O Proposed Question:
The reactor has tripped, and the crew is progressing through the EOP network. The crew enters EOP 35 FR-2.3 Response To High Containment Radiation Level.
Per FR-Z.3 guidance, The ADTS is considering the use of the Containment Air Filtration System to lower radiation levels in CTMT.
Which of the following events would the Containment Air Filtration System be most effective in reducing CTMT radiation levels?
A. A gaseous rad release during a fire in CTMT.
B. A large break loss of coolant accident.
C. An RCS leak resulting in high CTMT iodine levels and particulate levels.
D. A faulted-ruptured SG inside CTMT, resulting in CTMT temperature being at 200°F.
Proposed Answer: C L--
Explanation (Optional): C is correct, since CAF system is most effective at removing iodine in the charcoal filters and particulates in the HEPA filter. A is wrong, since smoke will greatly reduce the effectiveness of the charcoal filter.
B is wrong, since the CAF filters are located on the CTMT basement floor, which is underwater during a large LOCA.
D is wrong, since the CAF system charcoal beds do not handle high humidity conditions, and the system requires temperature to be maintained 490°F.
Technical Reference(s): FR-2.3 (Attach if not previously provided)
OP 33 13D, precautions.
Proposed references to be provided to applicants during examination: None Learning MC-0426 1 Describe the major administrative or procedural precautions and (As available)
Objective: limitations placed on the operation of the Containment Ventilation System, and the basis for each.
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO L.
Tier # 2 2 Question # 28 Group # 1 I WA # 003.Kl .I2 Importance Rating 3.0 3.3 Proposed Question:
With the plant at 100% power, what portions of a Reactor Coolant Pump are supplied by RPCCW?
A. Thermal barrier heat exchanger, and the pump upper and lower bearing oil coolers
- 3. Pump upper and lower bearing oil coolers, and RCP seal injection C. RCP seal injection, and the motor air cooler D. Thermal barrier heat exchanger, and the motor air cooler Proposed Answer: A Explanation (Optional): Motor air cooler is supplied by CDS (C, and D wrong, but plausible). Seal injection is supplied by charging (Band C wrong, but plausible). Each RCP is supplied by RPCCW to the thermal barrier heat exchanger and the pump oil coolers (A correct).
Technical Reference(s): P&ID 121B (Attach if not previously provided)
.-4 Proposed references to be provided to applicants during examination: None Learning Objective: MC-05427 Describe the folIowing RCP fluid flow paths.. . RPCCW Flow (As available)
Question Source: Bank #60154 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.2 to 41.9 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 29 Group # 1 1 KIA # 004.GEN.2.4.50 Importance Rating 3.3 3.3 Proposed Question:
With the plant operating at 100% power, an AUTO MAKEUP BLOCKED annunciator is received on Main Board 3B.
What could be the cause of this annunciator, and how can the RO correct this condition?
A. VCT level is at 40% and the REAC CLNT MAKEUP SELECT SW is in MANUAL. The RO will place the REAC CLNT MAKEUP SELECT SW in AUTO.
B. VCT level is at 45% and the REAC CLNT MAKEUP SELECT SW is in OFF. The RO will place the REAC CLNT MAKEUP SELECT SW in AUTO.
C. VCT level is at 40% and the REAC CLNT MAKEUP SELECT SW is in AUTO. The RO will place the REAC CLNT MAKEUP SELECT SW in MANUAL.
D. VCT level is at 45% and the REAC CLNT MAKEUP SELECT SW is in OFF. The RO will place the REAC CLNT MAKEUP SELECT SW in MANUAL.
Proposed Answer: A
-- Explanation (Optional): The AUTO MAKEUP BLOCKED annunciator is received when VCT lever is below the auto makeup setpoint of 41.4% (Band D wrong) AND the makeup controller is not in AUTO ((2 wrong). The ARP directs the operators to align the makeup system for automatic operation (A correct). B and D are plausible, since with the makeup select switch in OFF, auto makeup will not occur. C is plausible, since level has dropped <41%
when an auto makeup should be occurring.
Technical Reference(s): OP 3353.MB3,5-7 (Attach if not previously provided)
LSK 26-2.5C OP 3304C, section 4.2 Proposed references to be provided to applicants during examination: None Learning MC-04202 Describe the operation of the Chemical and Volume Control System (As available)
Objective: under normal, abnormal, and emergency operating conditions.
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.6, 7, 8, and 10 Comments:
29 of 34 NUREG-I021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.-4 Tier # 2 2 Question # 30 Group # 1 1 K/A # 004.A 1.06 Importance Rating 3 .O 3.2 Proposed Question:
INITIAL CONDITIONS:
0 The plant is stable at 100% power.
0 PZR Level controller 3RCS-LK459 has failed.
0 Charging Flow Control Valve 3CHS*FCV121 is in MANUAL due to the failure.
0 VCT level is 45% and stable.
0 Pressurizer level is 58% and stable.
The RO throttles 3CHS*FCV121 in the open direction, increasing charging flow by 50 gpm to raise pressurizer level.
Assuming no further operator action, what indications will exist ten minutes after 3CHS*FCVl2 I was throttled open?
A. VCT level constant at 66% with letdown diverted to Boron Recovery B. VCT level within its normal band with increased makeup C . VCT level at 4% with charging pump suction swapped to the RWST
. __. D. Letdown isolated with the reactor tripped on high pressurizer level.
Proposed Answer: B Explanation (Optional): Because charging is now greater than letdown, VCT level will decrease (A wrong). At 41%
level, makeup will initiate, and 50 gpm is within the capacity of VCT makeup of 80 gpm in AUTO (B correct, C wrong). Pressurizer level is about 125 gallons / %, so level increases by 4% ( I 0 minutes x 50 gallondminute / 125 gallons/%), which is less than the high pressurizer level trip setpoint of 89%, so the reactor will not trip (ID wrong).
Technical Reference(s): Functional Drawing #11 (Attach if not previously provided)
P&ID 104A Proposed references to be provided to applicants during examination: None Learning MC-04 I99 Describe the operation of the Controls and Interlocks associated with the (As Objective: following Chemical and Volume Control Systems components.. . Charging Line Flow available)
Control Valve.. .
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO 1-Tier # 2 2 Question # 3 1 Group # 1 1 WA # 005.K6.03 Importance Rating 2.5 2.6 Proposed Question:
Initial conditions The plant is being cooled down in preparation for refueling after a 400 day run.
RCS temperature: 175°F RCS pressure: 305 PSIA 0 The Pressurizer is solid.
0 Charging flow control is in manual 0 RHR is in the cooldown mode, with the A Train in service.
0 An undetected severe corrosion condition exists in the ARHR Heat Exchanger.
The A RHR heat exchanger suddenly develops a 50-gpm tube leak.
Assuming no operator action is taken, what will be the result of this event?
A. RPCCW surge tank level will decrease, until the RPCCW Surge Tank Makeup Valve opens, maintaining surge tank level.
B. RPCCW surge tank level decrease, until the ARPCCW pump trips, resulting in a loss of shutdown cooling.
C. RHR Pump flow increases and Pressurizer pressure decreases. RCS temperature begins to increase.
D. RHR Pump flow decreases and Pressurizer pressure decreases. RCS temperature begins to decrease.
Proposed Answer: C Explanation (Optional): RHR pressure is higher than CCP pressure; so 50 gpm is flowing from the RHR system into the RPCCW system (A and B wrong). With Charging flow control valve 3CHS*FCV121 in manual, 50 gpm is being lost from the RCS, and pressure will decrease in the pressurizer. RHR pump flow will increase, since pressure will drop at its discharge with the tube leak, and RHS*FCV618 will throttle open to maintain 4000 gpm total flow. Now a greater percentage of the RHR flow returning to the RCS is bypassing the RHS heat exchanger, so RCS temperature begins to increase (Ccorrect, D wrong).
Technical Reference(s): P&ID 1 12A (Attach if not previously provided)
OP 3310A, section 4.5.
OP 3208, steps 4.3.8 and 4.3.9 Proposed references to be provided to applicants during examination: None
~~~~i~~ MC-05459 Given a failure, partial or complete, of the Residual Heat Removal (As available)
Objective: system determine the effects on the system and on interrelated systems.
Question Source: Modified Bank #75426 Parent Question Attached Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-I 021, Revision 9
~ ~~ ~ ~~ ~ ~~ ~ -~
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Original Bank # 75426 L---
Initial conditions:
RCS temperature: 300°F RCS pressure: 340 PSIA Charging flow control is in manual RHR is in the cooldown mode, with the "A"Train in service The 'A' RHR Heat Exchanger develops a 10 gpm tube leak.
What indication will the RO observe during this event?
A. CCP surge tank level decreases.
B. Pressurizer level decreases.
C. RCS pressure increases.
D. RHR pump amps decrease.
Answer: B 29 of 34 NUREG-1021, Revision 9
~ ~~ ~~~
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO b Tier # 2 2 Question # 32 Group # 1 1 WA # 006.JS2.02 Importance Rating 2.5 2.9 Proposed Question:
With the plant at 100% power, the following sequence of events occurs:
- 1. An earthquake occurs, resulting in both a small break LOCA AND a loss of MCC 32-2R.
- 3. The crew is preparing to isolate accumulators per ES-I .2, step 22 Check if Accumulators Should Be Isolated.
What will be the impact, if any, of the loss of bus MCC 32-2R on the performance of step 22?
A. There is no impact. The crew will be able to isolate the A and C accumulators from Main Board 2.
B. The crew will not need to isolate the A and C accumulators, since the isolation valves were not able to open on the SIS signal.
C. The crew will be unable to close the isolation valves for the A and C accumulators. The crew will mitigate this by venting nitrogen off of both of these accumuIators.
D. The crew will be unable to close the isolation valves for the A and C accumuIators, and also will not be able to vent nitrogen off of these acchulators.
Proposed Answer: C Explanation (Optional): Electrical power has been lost to the A and CAccumulator Isolation Valves. A and C accumulator isolation valves are normally open MOVs (Bis wrong), powered fiom MCC 32-2R, so they cannot be closed (Awrong). Bis plausible since the isolation valves receive an OPEN signal on an SIS. ES-1.2 directs the crew to vent the unisolable accumulators to prevent N2 injection into the RCS. Venting is possible, since the vent valves are in parallel for each accumulator, powered from opposite train 125VDC power (C correct, D wrong).
Technical Reference(s): ES-1.2, step 22. (Attach if not previously provided)
EOP 35 GA-7 Isolating Accumulators EE- 1AQ Proposed references to be provided to applicants during examination: None Learning MC-062809 Given a failure, partial or complete, of the Emergency Core Cooling (As available)
Objective: System, determine the effects on the system and on interrelated systems.
Question Source: Modified Bank #76234 Parent question attached Question History: Millstone 3 2002 NRC Exam prior to modification Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5,41.8, and 41.10 Comments:
29 of 34 NUREG-I021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Original 76234 L,,
With the plant at 100% power, the following sequence of events occurs:
- 1. An earthquake occurs, resulting in both a 34C bus differential, and a small break LOCA.
- 3. The crew is preparing to isolate accumulators per ES-1.2, step 22 Check if Accumulators Should Be Isolated.
What will be the impact of the loss of bus 34C on the performance of step 22?
A. The crew will not need to isolate the A and C accumulators, since the isolation valves did not open on the SIS signal.
B. The crew will not need to isolate the A and C accumulators, since the isolation valves failed closed.
C. The crew will be unable to close the isolation valves for the A and C accumulators. The crew will mitigate this by venting nitrogen off of both of these accumulators.
D. The crew will be unable to close the isolation valves for the A and C accumulators. The crew will also not be able to vent nitrogen off of these accumulators.
Answer: C 29 of 34 NUREG-1021. Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 33 Group # 1 1 WA # 006.GEN.2.1.32 Importance Rating 3.4 3.8 Proposed Question:
Current Conditions:
0 A LOCA has occurred.
0 Reactor Trip/Safety Injection occurred due to low pressurizer pressure.
0 The crew has transitioned fi-om E-0 Reactor Trip or Safety Injection to E-1 Loss OfReactor Or Secondary Coolant.
0 RCS pressure is stable at 600 psia.
0 RHR pumps have just been stopped per E-1, step 8.
0 Containment Temperature is 170OF.
Under which condition would E-1 require the RHR pumps to be restarted?
A. RCS subcooling based on CETCs drops to 30°F.
B. Pressurizer level drops to less than 14%.
C. RCS pressure drops to 280 psia.
D. Containment temperature increases to 185°F.
Proposed Answer: C Explanation (Optional): C is correct since, to provide adequate ECCS flow, RCS pressure should be monitored to ensure that the RHR pumps are manually restarted if pressure decreases to LESS THAN 300 psia (500 psia ADVERSE CONTAINMENT). A and B are wrong, but plausible, since ECCS reinitiation criteria (per the foldout page is subcooling <32F or pressurizer level less than 16%) do not apply until SI has been terminated. D is wrong, but plausible since the RCS pressure setpoint increases from 300 to 500 psia under adverse CTMT conditions, but RCS pressure is above the adverse CTMT setpoint of 500 psia.
Technical Reference(s): E-1, CAUTION prior to step 8. (Attach if not previously provided)
E-1 Foldout Page.
Proposed references to be provided to applicants during examination: None Learning MC-06288 Describe the major administrative or procedural precautions and (As available)
Objective: limitations placed on the operation of the Emergency Core Cooling System, and the basis for each.
Question Source: Bank #71557 Question History: 2000 Millstone 3 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41 . I 0 Comments:
c 29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO C
Tier # 2 2 Question # 34 Group # 1 1 WA # 007.A3.01 Importance Rating 2.7 2.9 Proposed Question:
With the plant operating at 100% power, the following sequence of events occurs:
- 1. A PRT HI LEVEL annunciator is received on MB4.
- 2. A PRT PRESSURE HI annunciator is received on MB4.
- 3. The RO confirms that PRT Vent Valve (3RCS-PCV469) automaticallycloses on Main Board 4.
Which valve IS a potential source of water into the PRT that CANNOT be verified CLOSED via Main Board indications?
A. RCP Sealwater Return Line relief valve (3CHS*RV8113).
B. RHR Cold Leg Injection Line relief valve (3RHR*RV8856A).
C. Reactor Vessel Head Vent Valve (3RCS*SV8095A).
D. PRT Primary Water Fill Valve (3PGS-AVX030).
Proposed Answer: A
\-
Explanation (Optional): The following sources discharge to the PRT:
- 1. PZR PORVs (Main Board 4)
- 2. PZR safety valves (Main Board 4)
- 3. Reactor vessel head vent (C wrong, since position indication is on Main Board 3)
- 4. Primary Grade Water (D wrong, since position indication is on Main Board 4) 5 . CVCS relief valves 0 RCP seal return line reliefs (A correct, valve relieves to the PRT without MB indication)
Letdown line relief valve
Technical Reference(s): ARP 3353.MB4A, 2-3 PRT Level (Attach if not previously provided)
P&ID 107A, 102A, 102F Proposed references to be provided to applicants during examination: None
~~~i~~ MC-05349 Describe the Pressurizer Relief Tank System operation.. . Restoring from ( A available)
~
Objective: a High Pressurizer Relief Tank level condition.. .
Question Source: Modified Bank #68338 Parent attached.
Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Original question #68338 Which of the following could NOT be a source of leakage into the PRT?
A. The RCP Sealwater Return Line relief valve (3CHS*RV8113).
B. The RHR Cold Leg Injection Line relief valves (3RHR*RV8856 AB).
C. The RHR/RCS Suction Line relief valves (3RHS*RV37 A/B).
D. The Reactor Vessel Head Vent.
Answer: B
\--
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 35 Group # 1 1 WA # 008.A3.08 Importance Rating 3.6 3.7 Proposed Question:
Current plant conditions:
A steam break has occurred outside Containment Automatic SI and MSI signals have occurred.
The RO is monitoring Main Board 1 to verify proper response of the RPCCW system.
Which normally open RPCCW vaIves should have automatically closed during this event?
A. The containment header cross-connect valves (3CCP*AOV179A/Band 1SOA/B).
B. The cross-connect to chilled water system valves (3CCP*MOV222 thru 229).
C . The containment supply and return header isolation valves (3CCP*MOV45A/B748 A/B and 49 AB).
D. The non-safety header isolation valves (3CCP*AOP 197AB,lO A B , 194 A/B and 19 AB).
Proposed Answer: D
.~ Explanation (Optional):
A is wrong, but plausible, since these valves receive a CLOSE signal on an SIS, but they are already closed.
B is wrong, but plausible, since the CCP to CDS cross connect valves reposition to OPEN on an SIS.
C is wrong, but plausible, since the CTMT header isolates on a CIB, not CIA.
W is correct, since the non-safety header isolates on an SIS.
Technical Reference@): Attachment A of E-0 page 2 of 3 (Attach if not previously provided)
P&ID 121 A andB Proposed references to be provided to applicants during examination: None Learning MC-04 154 Describe the operation of the Reactor Plant Component Cooling System (As available)
Objective: under the following normal, abnormal, or emergency conditions.. . Sequence Safeguards Signal actuation...
Question Source: Bank #64957 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.4 I .7 Comments:
29 of 34 NUREG-I021,Revision 9
ES-40 I Written Examination Question Worksheet Form ES-40 1-5
-- Examination Outline Cross-reference: Level Tier #
RO 2 2 SRO Question # 36 Group # 1 1 WA # 0 10.A2.03 Importance Rating , 4.1 4.2 Proposed Question:
With the plant at 100% power, the following sequence of events occurs:
RCS pressure starts to slowly decrease.
0 The PZR RELIEF VALVE DIS TEMP HI annunciator is received on MB4A.
The RO reports that both PORVs indicate fully closed on MB4.
RCS pressure has stabilized at 2235 psia.
Pressurizer level has increased to 63%.
Prior to operator action, what was the effect of the leaky PORV on the Pressurizer Pressure Control System, and how will the crew determine which PORV Block Valve to close?
A. The variable control heater group 3RCS-HI C is at maximum power, and the open PORV is determined by monitoring PORV outlet temperatures on the EEQ data logger.
B. The backup heater groups that were in AUTO are now energized, and the open PORV is determined by monitoring PORV outlet temperatures on the EEQ data logger.
C. The variable control heater group 3RCS-HlC is at maximum power, and the open PORV is determined by
.-/ monitoring PORV tailpiece temperature indications on MB4.
D. The backup heater groups that were in AUTO are now energized, and the open PORV is determined by monitoring PORV tailpiece temperature indications on MB4.
Proposed Answer: A Explanation (Optional): The C Heater Group provides a variable output, which is fully powered when RCS pressure decreases to 2235 psia, and the Backup Heaters come on if pressure continues to decrease to 2225 psia. Backup heaters also energize during an insurge that raises pressurizer level 5% above program, but level has increased by only 1.5%
(B and D wrong, but plausible). The ARP directs the operators to monitor the Data Logger to determine which PORV is leaking, since the MB4 indication is common to both PORVs (A correct, C wrong).
Technical Reference(s): OP 3353.MB4A, 3-5 (Attach if not previously provided)
Functional Drawings I 1 and 12 Proposed references to be provided to applicants during examination: None Learning MC-0534 1 Describe the operation of the Pressurizer Pressure and Level Control (As available)
Objective: System under Normal, Abnormal, and Emergency Operating conditions.
Question Source: New Question Cognitive Level: Comprehension or Analysis IO CFR Part 55 Content: 55.41.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
--- Tier # 2 2 Question ## 37 Group # 1 1 WA # 0 12.K 1.06 Importance Rating 3.1 3.1 Proposed Question:
INITIAL CONDITIONS:
A load increase is in progress after an outage that included replacement of all 3 LP turbine rotors.
The plant is currently stable at 48% while preparing to place the second TDMFP in service.
The following sequence of events occurs:
- 1. A Turbine HI-HI vibration annunciator is received.
- 2. The BOP trips the turbine from MB7.
- 3. Steam Dumps open, and rods start driving in.
- 4. The extra operator takes manual control and stabilizes feed, preventing a SGWLC related reactor trip.
5 . Rods are placed in MANUAL, and reactor power steadies out at 7% power.
At this point, what is true concerning reactor trip signals?
A. The reactor should have automatically tripped when the BOP tripped the turbine.
B. The reactor will trip when AMSAC automatically actuates within the next 25 seconds.
C. The reactor will trip if pressurizer level reaches 90%.
D. The reactor wilI trip if two reactor coolant pumps trip.
Proposed Answer: A Explanation (Optional): A is correct, since the turbine tripped with power above P-9, which is currently set at 45%.
Previously, the P-9 setpoint was 51%. C and D are wrong, since P-7, P-10, and P-13 have cleared with turbine impulse pressures dropping <lo% power, and NIS power going Iess than 10% power C and D are plausible, since these trips are related to P-7, P8, and PI 0, which have changed state on the downpower. B is wrong, but plausible, since AMSAC is initially armed >40%, and will not reset for 260 seconds after power going <45%, but AMSAC has not actuated, since the extra operator stabilized SG levels above reactor trip setpoints.
Technical Reference(s): Functional Sheets 4, 5,6,7 and 16 (Attach if not previously provided)
E-0 Entry Conditions Proposed references to be provided to applicants during examination: None
~~~h~ MC-05493 Describe the operation of the following RPS controls and interlocks. .. (As Objective: Reactor Trip Signals.. . Protective Interlocks... Control Interlocks available)
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 I Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 38 Group # 1 1 WA # 0 12.K2.01 Proposed Question: Importance Rating 3.3 3.7 Initial conditions:
The plant is at 30 % power.
The swing charger is supplying power to DC Bus 1.
Bus 32T deenergizes due to an electrical fault, and a11 equipment operates as designed.
Which of the following describes the immediate effects on VIAC l ?
A. The swing charger maintains power to VIAC I via the DC bus.
B. Battery 1 begins to discharge, maintaining power to VIAC 1.
C. VIAC 1 switches over to its alternate source.
D. VIAC 1 de-energizes.
Proposed Answer: B Explanation: B is correct, and A wrong, since 32T supplies the rectifier via 32-2T, and both chargers via 32-23.
C and D are wrong since the inverter will still provide power as long as the DC bus power is acceptable.
L Technical Reference(s): EE- 1BA (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None
~~~i~~ MC-03325 Given a failure of the 480 VAC distribution system or a portion of the (As available)
Objective: system, determine the effects on the system and on interrelated systems a). Loss of 480 volt load center or MCC on applicable loads.. .
Question Source: Modified Bank #68027 Parent Question Attached Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.7 Comments: Original Bank Question #68027 The plant is at 30 % power. All electrical systems are in their normal alignment.
Bus 32T deenergizes due to an electrical fault.
Which of the following describes the effects on the 125 VDC electrical distribution system?
A. Battery 1 begins to discharge, supplying power to all of its DC bus loads including VIAC 1 .
B. Battery 1 begins to discharge, supplying power to all of its DC bus loads except VIAC 1 .
C. Battery 1 begins to discharge, but will stop dischargingwhen VIAC I transfers to the alternate supply.
D. Battery 1 will not discharge, because the Inverter 1 static switch will select the alternate source immediately upon loss of bus 32T.
Answer: A 29 of 34 NUREG-I021,Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 39 Group # 1 1 WA # 0I 3 .K5.02 Importance Rating 2.9 3.3 Proposed Question:
INITIAL CONDITIONS:
0 The plant is operating at 100% power 0 Containment Pressure Channel 111 (PT-935) has failed high 0 The appropriate bistables have been tripped 0 I&C is about to begin troubleshooting the failed channel Which safeguards signal(s) would be generated if the I&C technician inadvertentlyde-energizes the control and instrument power for the Channel I1 Containment Pressure instrument?
A, Only an SI signal will be generated.
B. Only SI and MSI signals will be generated.
C. Only SI and CDA signals will be generated.
D. SI, MSI, and CDA signals will be generated.
. L L . Proposed Answer: B Explanation (Optional): All of the above ESF actuation signals require 2 channels to cause an actuation. The SI and MSI bistables are de energize to actuate. One channel is already tripped, and when the second channel is de energized, the signals will be actuated (Awrong). The CDA signal is energize to actuate and therefore will not be affected with the second channel loss of power (Bcorrect, C and Dwrong).
TechnicaI Reference(s): Functional Drawing #8 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-05493 Describe the operation of the following RPS controls and interlocks.. . (As available)
Objective: ESF Actuation Signals.. .
Question Source: Bank #69327 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 Comments:
29 of 34 NUREG-I 021, Revision 9
ES-40 I Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO d
Tier # 2 2 Question # 40 Group # I 1 WA # 022.GEN.2. I .23 Importance Rating 3.9 4.0 Proposed Question:
Initial conditions:
The plant has just been cooled down at the start of a refueling outage.
The crew is preparing to start the Containment Purge System in the Unfiltered mode of operation.
The plan is to open the containment access hatch shortly.
Outside air temperature is 55°F.
What will be the desired Containment Purge System lineup with the Containment Access Hatch open?
A. One supply HVU and one exhaust fan running to prevent excessively cooling down CTMT.
B. Two supply HVUs and two exhaust fans running to maximize air flow in CTMT C. One supply HVU and two exhaust fans running to keep air flow into CTMT through the access hatch.
D. Two supply H W s and one exhaust fan running to keep air flow out of CTMT through the access hatch.
Proposed Answer: C Explanation (Optional): One supply HVU and two exhaust fans are desired to be running to keep air flow into CTMT through the access hatch (C correct, B and D wrong) A is wrong, since hot water heating is modulated to the HVUs to maintain 70°F outlet temperature. LAis plausible, since there is a minimum desired CTMT temperature, and only one train is allowed in the filtered mode. B is plausible since operating with 2 trains of purge is allowed with the CTMT hatch closed.
Technical Reference(s): OP 33 13F, section 4.1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-0426 1 Describe the major administrative or procedural precautions placed on (As available)
Objective: the operation of the CTMT ventilation systems, and the basis for each.
Question Source: Bank #76282 Question History: Millstone 2002 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.4 Comments:
29 of 34 NUREG-I 021,Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO v Tier # 2 2 Question # 41 Group ## 1 1 KIA # 026.K4.07 Importance Rating 3.8 4.1 Proposed Question:
With the plant initially at 100% power, a large break LOCA occurs, and the RCS rapidly depressurizes to CTMT pressure.
When will the RSS pumps start, and why?
1 A. RSS pumps will start when CTMT pressure reaches the HI-3 setpoint, in order to minimize peak CTMT pressure and reduce pressure quickly.
B. RSS pumps will start after about 660 seconds, to allow time for CTMT sump leve1 to increase, providing proper NPSH to the RSS pumps.
C. RSS pumps will start after about 660 seconds, to allow time for the RHR pumps to trip off on low-low RWST level, preventing exceeding the heat removal capabilities of the RPCCW system.
D. RSS pumps will start after about 660 seconds as part of automatic load sequencing, preventing an overload of the EDGs coincident with an LOP.
Proposed Answer: B ir Explanation (Optional): B is correct, since RSS pumps take suction on the CTMT sump, and need adequate sump level for NPSH. A is plausibIe, since CDA actuates and QSS pumps start when CTMT pressure reaches HI-3. C is plausible since RHR will trip off when RWST level reaches 520,000 gallons, both RSS and RHR have heat exchangers, and RPCCW has temperature and flow limits. D is plausible, since the sequencer does load equipment onto the EDG sequentially during an LOP, to prevent overloading the EDGs, however, no LOP is in progress, so LOP load sequencing is not in effect.
Technical Reference(s): FSAR Section 6.2.2.3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None
~~~~i~~ MC-05 171 Describe operation of the following containment de-pressurization (As available)
Objective: system components controls and interlocks.. . Containment recirculation pumps. ..
Question Source: Bank #73206 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-I 021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
,..-- -1 Tier # 2 2 Question # 42 Group # 1 1 WA # 039 .A2.04 Proposed Question: Importance Rating 3.4 3.7 Initial Conditions:
A plant startup is in progress per OP 3203 Plant Startup.
Reactor power is at 4%.
Tave is 551.
Condenser Steam Dumps are in automatic, in the steam pressure mode.
Preparations are being made to enter MODE 1.
Main Steam Header Pressure Channel 3MSS-PT507 fails HIGH.
At what temperature will Tave stabilize, and how will operators mitigate the effect of this failure of the steam dumps?
A. 553OF due to P-12. Operators will place the steam pressure controller to MANUAL and reduce the output to minimum. Atmospheric dumps will then maintain Tave at 56I0F.
B. 553°F due to P-12. Operators will place the Steam Dump Control Mode Selector Switch to Tavg. Steam Dumps will then maintain Tave at 557OF.
C. 561°F due to atmospheric dump valves. Operators will place the steam pressure controller to MANUAL and reduce the output to minimum. Atmospheric dumps will then maintain Tave at 56 1OF.
D. 561°F due to atmospheric dump valves. Operators will place the Steam Dump Control Mode Selector Switch to Tavg. Steam Dumps will then maintain Tave at 557°F.
Proposed Answer: A Explanation (Optional): Steam pressure failing high causes dumps to open (Cy and D wrong). P-I2 will close the dumps at 553OF. C and D are plausible, since atmospheric dumps would stabilize Tave at 561°F if the condenser dumps had failed closed. AOP 3571 has the operators place the steam pressure controller to MANUAL and reduce the output to minimum, causing atmospheric dump valves open to maintain SG pressure at 56 1 (A correct, Bwrong).
Byis wrong since dumps would not be armed in the Tave mode, but plausible since the failed channel does not provide input to the steam dumps in the Tave mode.
Technical Reference(s): Functional Drawing 10 (Attach if not previously provided}
AOP 3571, Attachment J Proposed references to be provided to applicants during examination: None Learning MC-05636 Given a failure, partial or complete, of the steam dump system, determine (As avaiIable)
Objective: the effects on the system and on interrelated systems.
Question Source: New Question Cognitive Level: Comprehension or Analysis I O CFR Part 55 Content: 55.41.5 Comments:
29 of 34 NUREG-I021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question ## 43 Group # 1 1 KIA # 056.K1.03 Importance Rating 2.6 2.6 Proposed Question:
The plant is operating normally at 100% power.
Which of the following conditions could result in damage to the Main Feed Pump seals?
A. Seal water header pressure exceeds 24 psig.
B. Condenser Long Recycle Valve 3FWS-V986 is closed.
C. Temperature of the condensate at the discharge of the condensate pumps exceeds 150°F.
D. Temperature of the condensate at the discharge of the 5hpoint heaters exceeds 1 5 O O F .
Proposed Answer: C Explanation (Optional): C is correct since if condensate temperature exceeds 150 degrees, the water entering the seals will not adequately cool the seals and the seals could be damaged. A is wrong since normal seal water pressure is 25 psig. A relief lifts if pressure exceeds 27 psig. B is wrong since the long recycle valve must be kept closed with feed pumps in operation to prevent overpressurizingthe long recycle piping. D is wrong since normal temperature of the
-- condensate at the discharge of the 5th point heaters is approximately 220OF.
Technical Reference(s): OP3321, Precautions 3.5 and 3.9 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-04235 Given a failure, partial or complete, of the Main Condensate and Makeup (As available)
Objective: Control systems, determine the effects on the systems and on interrelated systems.
Question Source: Bank #69650 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.4, 5, and 14 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 I Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 44 Group # 1 1 WA # 059.A3.06 Importance Rating 3.2 3.3 Proposed Question:
Initial conditions:
The plant is operating at 20% power.
The main generator has just been placed on the grid.
Main Feed is being supplied by the B TDMFP.
The following sequence of events occurs:
- 2. A Feedwater Isolation signal is received.
- 3. The reactor has NOT tripped.
Which components do the RO and BOP operators expect to see changing state on the Main Boards?
A. The Feed Reg Bypass Valves and Feed CTVs will close only.
B. The Feed Reg Bypass Valves and Feed CTVs will close, and the Main Turbine will trip only.
C. The Feed Reg Bypass Valves, Feed CTVs, and SG Chemical Feed Valves will close; and the Main Turbine and the B TDMFP will trip only.
D. The Feed Reg Bypass Valves, Feed CTVs, SG Chemical Feed Valves, and the D7 SG Blowdown and BIowdown Sample valves will close; and the Main Turbine and the B TDMFP will trip.
Proposed Answer: C Explanation (Optional): C is correct, since on aP-14 FWI, the Feed Reg Bypass Valves, Feed CTVs, and SG Chemical Feed Valves will close; and the Main Turbine and the BTDMFP will trip. A and B are plausible, since all of these components actuate on some FWI signals. Not all FWI signals trip the turbine or TDMFP. D is plausible, since the BIowdown and blowdown sample valves are normally checked closed after a trip at the FWI step in E-0, but the valves receive their close signal from the AFW pump start signal, which has not been received.
Technical Reference(s): Functional Drawing 13 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-072 19 Describe the operation of the following Feed Water System controls and (As available)
Objective: interlocks.. . P-14 Permissive Interlock.. .
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 45 Group # 1 1 K/A ## 059.A4.03 Importance Rating 2.9 2.9 Proposed Question:
Current Plant Conditions:
Reactor power is being raised per OP 3203 Plant Startup.
Reactor power has just reached 25% power.
SG level is being maintained by the Feed Reg Bypass Valves.
The crew commences transferring feedwater control fiom the Feed Reg Bypass Valves to the Feed Reg Valves.
For the A SG, how will the BOP operator transfer feedwater control to the Main Feedwater Reg valve?
A. Open the Feed Line Isolation Valve 3FWS*MOV35A. Place the Feed Reg Valve in AUTO. Reduce SG level slightly by closing the Feed Reg Bypass Valve, and observe the Feed Reg Valve automatically opening to restore SG level to the setpoint. When the Feed Reg Bypass Valve is closed and SG level is stable at 50%, place the Feed Reg Valve in AUTO.
B. Open the Feed Line Isolation Valve 3FWS*MOV35A. With both the Feed Reg Valve and the Feed Reg Bypass Valve in MANUAL, simultaneously open the Feed Reg Valve and close the Feed Reg Bypass Valve. When the Feed Reg Bypass Valve is closed and SG level is stable at 50%, place the Feed Reg Valve in AUTO.
C. Manually crack-open the Feed Reg Valve to increase SG level slightly. Observe the Feed Reg Bypass Valve
\---
automaticaIly closing to restore SG level to the setpoint. Continue opening the Feed Reg Valve and when the Feed Reg Bypass Valve is closed and SG level is stable at 50%, place the Feed Reg Valve in AUTO.
D. Place the Feed Reg Valve in AUTO. Reduce SG level slightly by closing the Feed Reg Bypass Valve, and observe the Feed Reg Valve automatically opening to restore SG level to the setpoint. When the Feed Reg Bypass Valve is closed and SG level is stable at 50%, place the Feed Reg Valve in AUTO.
Proposed Answer: B Explanation (Optional): The Feed Line isolation valve must be opened, as it is kept closed while feeding on the bypass valve to minimize leakage through the Feed Reg Valve (C and D wrong). C and Dare plausible, since a misconception may be that the Feed Line Isolation Valve isolates both the Feed Reg Valve and Feed Reg Bypass Valves. B is correct, and A is wrong, since both valves are operated in MANUAL while switching to the Feed Reg Valves. A is plausible since this method would also work for placing feed control onto the Feed Reg Valves.
Technical Reference(s): OP 3321, section 4.3.39 (Attach if not previously provided)
P&ID 130C Proposed references to be provided to applicants during examination: None Learning MC-04663 DESCRIBE the operation of. .. Main Feedwater (during) Normal At- (As available)
Objective: Power Operations while increasing or decreasing power between 25 & 100%.
Question Source: INPO Exam Bank Question History: Indian Point 2,2003 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 and41.10 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1
~
I-- ..> - . A* ----QuestionWorksheet Form ES-40 1-5 Kb4: 0 6 / -
RO SRO 2 2 1 1 061 .K5.01 Proposed Quest: Importance Rating 3.6 3.9 PLANT CONDITIONS:
A reactor trip has occurred from 100% power 0 The operators have not operated any controls post-trip.
0 The crew has just entered ES-0.1 Reactor Trip Response.
0 PZR level is 25% and slowly decreasing.
0 Steam Generator pressures are approximately 990 psig and slowly decreasing.
0 Tave is 545OF and slowly decreasing.
0 RCS pressure is 2020 psia and slowly decreasing.
What action must be taken by the crew per ES-0.1 to address the cooldown?
A. Commence immediate boration.
B. Initiate SI and return to step 1 of E-0. ,
C. Throttle Auxiliary Feedwater flow.
D. Close the MSIVs and MSIV bypass valves.
L-Proposed Answer: C Explanation (Optional): Full AFW flow is automatically on a reactor trip due to shrink in the SGs. This flow will result in an undesired cooldown shortly after the trip if operator action is not taken to throttle AFW flow. C is correct, since ES-0.1 step 1 has the crew throttle AFW if a cooldown is in progress. Also, per OP 3272, the balance of plant operator may, at any time when not required to be performing an immediate action or sequenced steps, throttle AFW flow if minimum heat sink requirements are satisfied. This includes throttling flow to minimize RCS cooldown. A is wrong, since immediate boration is required only if the cooldown causes Tave to decrease to 5 530F. B is wrong since foldout page SI requirements (PZR level 9%, 32°F subcooling) are not met. D is wrong, since MSIVs will be closed only if efforts to control the cooldown by throttling AFW are unsuccessful. A, By,and D are plausible, since each of these actions may be required if the cooldown continues.
Technical Reference(s): OP 3272, Attachment 3 (Attach if not previously provided)
ES-0.1, step 1 Proposed references to be provided to applicants during examination: None Learning MC-04454 State the conditions which would allow the action of either the throttling (As available)
Objective: or isolation of auxiliary feed water flow to a steam generator.. .
Question Source: Bank 70200 Question History: Millstone 2000 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5 Comments:
29 of 34 NUREG-I 021, Revision 9
Attachment 3 Special Considerations (Sheet 2 of 11)
Adverse Containment For parameters that can be affected by adverse containment conditions, Adverse Containment values are provided in parenthesis following the normal containment value and should be utilized by the operator when either specified adverse containment criterion is exceeded. After attaining an adverse value, the following rules apply when either containment parameter subsequently decreases below the specified value:
When Ctmt temperature decreases to less than or equal to 180'8 Adverse Ctmt values no longer apply.
When Ctmt radiation decreases to less than or equal to 10s R/hr, Adverse Ctmt values continue to apply.
Isolatin-mrottline AFW Flow The balance of plant operator may, at any time when not required to be performing an immediate action or sequenced steps, throttle AJW flow if minimum heat sink requirements are satisfied. This includes the following:
I Isolating AFW flow to a faulted SG (unless all SGs are faulted) la Throttling flow to minimize RCS cooldown (flow to all SGs should be jl throttled evenly including any ruptured SG)
However, the SMWS should direct isolating AEW flow to a ruptured SG when the minimum WR of NR ruptured SG level specified in E-3 is satisfied.
I Isolating Ruptured SG Steam Line An operator may not isolate the steam line to a ruptured steam generator until directed by the procedure. This is an identified safe condition unless specifically directed by the SGTR procedure.
I Checkine for Faulted SGs When checking for a faulted SG, use the steam pressure recorders on MB2 for proper trending accuracy.
Level of Use OP 3272 Rev. 008-02 30 of 51
REACTOR TRIP RESPONSE -
EOP 35 ES 0.1 Page 3 of 21 Rev. 020-01
.STEP ACTIONEXPECTED RESPONS~; RESPONSE NOT OBTAINED CAUTION If SI actuation occws during this procedure, immediately Go to E-0, Reactor Trip or Safety Injection.
NOTE Foldout page must be open.
- 1. Check RCS Temperature
-a. Verify RCS cold leg WR a. Perform the applicable action:
temperature -
BETWEEN 550°FAND -
IF temperature is 560°F GREATER THAN 560"E THEN Proceed to step 1.c.
fE temperature is LESS THAN 55O0F, THEN Proceed to step 1.e.
-b. Proceed to step 2.
REACTOR TRIP RESPONSE
. A STEP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED
- 1. (continued)
-c. Dump steam to condenser c. Dump steam to atmosphere using SG atmospheric relief
- 1) Verifv the following: valves or the SG atmospheric relief bypass valves (MB or MSNS - AT LEAST locally).
ONE OPEN Annunciator CONDENSER AVAIL FOR STM DUMP C-9 (M334D 5-6) - LIT
- 2) Adjust steam pressure controller to obtain zero output inMANUAL
- 3) Bansfer condenser steam dum s to Steam Pressure Mo e 2
- 4) Place both condenser steam dump interlock selectors - ON
- 5) Adjust steam ressure B
controller to ump steam to condenser
-d. Proceed to step 2.
-e. Maintain total feed flow BETWEEN 530 and 600 gpm until NR level is GREATER THAN 8% in at least one SG
-.f. CLOSE SG atmospheric relief and relief bypass valves
-g. STOP the MD FW pump and Place control switch in PULL-TO-LOCK
I REACTOR TRIP RESPONSE EOP 35 ES-0.1 Rev. 020-01 Page 5 of 21
.STEP c
ACTIONEXPECTED RESPONS I
RESPONSE NOT OBTAINED
- 1. (continued)
-h. TRIP the TD F W pumps
-i. Check SG code safety valves 1. Consult OMOC to determine I closed if safetv valve(s1 not closed shouldbe ga&;d using Flow switches (MBS) - GA- 17.
NOT LIT Local observation of safety valves (MSVB roof) -
NO STEAM OBSERVED
-j. Verify RCS cold leg WR j. Place both condenser steam temperature - dump interlock selector STABLE OR INCREASING switches to OFF.
IF RCS cooldown continues, THEN CLOSE the MSWs and MSIV bypass valves.
-k. Verify RCS cold leg W R k. Perform the following:
temperature -
GREATER THAN 530°F 1) CLOSE the MSIVs and MSIV bypass valves.
- 2) Using AOP 3566, Immediate Boration, Immediate borate.
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 47 Group # 1 1 KIA # 062.KI -04 Importance Rating 3.7 4.2 Proposed Question:
With the plant at 100% power, a "brown-out" condition on the grid occurs. Offsite power being supplied to MP3 busses has been at 80% voltage for 5 minutes.
What will happen to the 6.9KV and 4,160V busses?
A. All 6.9KV buses to remain energized at the low voltage, and the normal 4. I6KV buses de-energize, while the emergency 4.16KV buses are powered from the emergency diesels.
B. AI1 6.9KV buses de-energize, and the normal 4.16KV buses remain energized at the low voltage, while the emergency 4.16KV buses slow transfer.
C. All 6.9KV buses to remain energized at the low voltage, and the normal 4.16KV buses remain energized at the low voltage, while the emergency 4.16KV buses are powered from the emergency diesels.
D. AI1 6.9KV buses de-energize, and the normal 4.16KV buses de-energize, while the emergency 4.16KV buses are powered from the emergency diesels.
Proposed Answer: C Explanation (Optional): Sustained UV for the 6.9 KV buses is 2/2 less than 70%. Since voltage is at SO%, they will not de energize. Undervoltage to lockout NSST for the 4160 buses is also 70% so the breakers will not open. However, after 4.5 minutes, the 4160V buses will attempt to transfer, since RSST is less than 97%, fast and slow transfer will not occur. The tiebreakers will open; the diesels will start and energize the emergency buses from the EDG (C correct).
Technical Reference(s): LSK 24-3 series (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-03337 Describe the 4kV Distribution System operation under normal, abnormal (As available)
Objective: and emergency conditions: At power operations, Main Generator trip, Loss of NSSA, Loss of RSSA, LOP sequence of operations, MB8 alarm response Question Source: Bank #68084 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.4,41.7, and41.8 Comments:
29 of 34 NUREG-1021 Revision 9 I
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
-?-_--I Tier # 2 2 Question # 48 Group i# 1 1 K/A # 062.K2.0 1 Importance Rating 3.3 3.4 Proposed Question:
Initial Conditions:
0 The plant is at 100% power.
0 34C and D are being powered from the RSSTs due to CONVEX testing.
The following sequence of events occurs:
0 The plant trips due to a loss of 4 160V bus 34B.
0 All other busses remain energized.
Which of the following lists contains energized loads?
A. Load Center 32M, "B" Control Building Chiller, and "B" Circulating Water pump.
B. "B" Screenwash pump, "D" Circulating Water pump, and "B" CDS Chiller.
C. " B Reactor Coolant Pump, "B" Condensate pump, and Motor Driven Main Feed Pump.
. D. Load Center 32B, "B" TPCCW pump, and "B" MSR Drain Pump.
Proposed Answer: B Explanation (Optional): "B' is correct since loads off 34B include "B" Screen Wash pump, "B" CDS chiller, "B" Heater Drain pump, "B", "D" & "F" Circ Water Pumps, and Load Center's 32H, J,K, L, M, N,P &Q. "A" is wrong since the control bldg chiller is powered from 34D. "C" is wrong, since these are 6.9KV loads. " D is wrong, since Load Center 32B is powered fiom 34A.
Technical Reference@): Electrical Drawing EE-1H (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-05024 Given the following failure of the 4 kV Distribution system or a portion (As available)
Objective: of the system, determine the effects on the system and on interrelated systems: 4 kV bus deenergization on safety-related and non-safety loads Question Source: Bank #68083 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.4 1.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-401-5
.\-,
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 49 Group # 1 1 K/A # 063.A2.0 1 Importance Rating 2.5 3.2 Proposed Question:
With the plant at 100% power, the following sequence of events occurs:
- 1. A battery trouble annunciator is received on MB8.
- 2. The PEO sent out to investigate reports a DC Ground on Battery 301B-2 with a 1 12 Volt differential.
- 4. The STA calls up a Sequence of Events Report to assist in diagnosing the ground location.
- 5. The extra senior licensed operator starts investigating for recently started equipment to assist in diagnosing the ground location.
In accordance with AOP 3551,what strategy will the crew use to isolate the ground?
A. The crew will check the in-service battery charger first, followed by the inverter, and if at any time the STA or extra SRO suspects a source for the ground, the crew will complete the step in progress and then try to isolate the suspected equipment by proceeding to the applicable step in AOP 3551.
B. The crew will check the inverter frst, followed by the in-service battery charger, and if at any time the STA or extra SRO suspects a source for the ground, the crew will complete the step in progress and then try to isolate the suspected equipment by proceeding to the applicable step in AOP 3551.
C. The crew will check the in-service battery charger frst, followed by the inverter, and if at any time the STA or extra SRO suspects a source for the ground, the crew will continue in AOP 355 1 and wait for the appropriate step to check the suspected equipment.
D. The crew will check the inverter first, followed by the in-service battery charger, and if at any time the STA or extra SRO suspects a source for the ground, the crew will continue in AOP 355 1 and wait for the appropriate step to check the suspected equipment.
Proposed Answer: A Explanation (Optional): The AOP checks the battery charger first. An alternate charger is available to place in service if the charger is grounded (Byand Dwrong). Per the note prior to step 1 the crew will complete the step in progress and then go directly to the equipment isolation step if a ground source is suspected (A correct, C wrong).
Technical Reference(s): AOP 355 1, Note prior to step 1, and steps 1 - 4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MC-05887 Describe the major action categories within AOP 355 1 (As available)
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.5 Comments:
29 of 34 NUREG-I021, Revision 9
~~
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.e Tier # 2 2 Question # 50 Group # 1 1 WA # 064.K3.01 Importance Rating 3.8 4.1 Proposed Question:
With the plant at 100% power with the A Charging Pump running, the following sequence of events occurs:
- 1. A fire breaks out in the A Sequencer.
- 2. The power supply to the sequencer trips, and the sequencer de-energizes.
- 3. The fire extinguishes when the sequencer deenergizes.
- 4. The fire brigade is responding.
- 5. The crew has entered EOP 3509 Fire Emergency.
- 6. A loss of off-site power occurs.
Before operator action is taken, how will the plant respond to the loss of offsite power?
A. The A EDG will not start. All required B train loads will sequence on, except no charging pump will be running.
B. Both diesels wiIl start, but the A EDG output breaker will not close. All required Bytrain loads will sequence on, except no charging pump will be running.
C. Both dieseIs will start and their output breakers will close, but A train loads will not energize since they will not automatically sequence on. All required Btrain loads will sequence on, including the B Charging pump.
Both diesels will start, and all required equipment will start; however, the A Train equipment will load immediately, since the loads were not stripped.
Proposed Answer: B Explanation (Optional): When a LOP occurs, all loads are stripped except Load Centers and charging pumps, if their breaker is closed prior to the LOP. Since the B Charging Pump was not running, it will not start (Cwrong). The A Train sequencer will not start the EDG, strip loads, close the EDG output breaker (Dwrong), or sequence loads on; but the LOP signal will start the EDG directly fiom the switchgear (Bcorrect, A wrong).
Technical Reference(s): P&ID 104A (Attach if not previously provided)
ESK 5CS, 5CT, 5CU, 5CV Proposed references to be provided to applicants during examination: None Learning MC-044 I6 Describe the operation of the emergency diesel load sequencers under the (As available)
Objective: following normal, abnormal, and emergency conditions: Automatic operation; LOP only, SIS only.. .
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.7 Comments:
J 29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO
.L - Tier # 2 2 Question # 5 1 Group # 1 1 IUA # 073.A1.01 Importance Rating 3.2 3.5 Proposed Question:
With the plant initially at 100% power, the following sequence of events occurs:
- 1. The reactor trips due to a failed reactor trip breaker.
- 2. The crew enters ES-0.1 Reactor Trip Response.
- 3. A tube leak occurs in the CSG.
- 4. Radiation monitor ARC21-1 goes into ALERT.
- 5. The RO reports all primary plant parameters appear to be stable.
- 6. The crew enters AOP 3573 Radiation Monitor Alarm Response.
- 7. The RO is directed to call up a SGTR Trend at the RMS console to look for indication of a primary to secondary leak.
Which type of trend should the RO call up on the RMS console, and how will the crew confirm the leak?
A. 1-minute trend, with the leak indicated on 3MSS-RE80C C SG N-16 Monitor.
B. 1 -minute trend, with the leak indicated on 3SSR-REO8 Blowdown Radiation Monitor.
C. 10-minute trend, with the leak indicated on 3MSS-RE77 C Main Steamline Radiation Monitor.
./ D. 10-minute trend, with the leak indicated by SG activity sample.
Proposed Answer: D Expianation (Optional): D is correct and A and B are wrong, since a 10 minute trend is required post trip, as N-16 gammas decay away within one minute after the trip. A and Care wrong, since the leak occurred post-trip, so N-16 detectors will not show the leak. Dis correct, since AOP 3573 will send the crew to AOP 3576 Steam Generator Tube Leak, and this procedure will direct chemistry to obtain a grab sample from the SGs.
Technical Reference(s): AOP 3573, Att. A, page 12 (Attach if not previously provided)
AOP 3576, step 3 Proposed references to be provided to applicants during examination: None Learning Objective: MC-04919 Describe the major parameter changes associated with SGTRs. (As available)
Question Source: Modified bank #75660 Parent question attached Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 I .5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Original Bank Question #75660 L
The plant had been operating normally at 100% power when the reactor has tripped and the crew is progressing through ES-0.1 when a tube leak occurs in the C Steam Generator.
Which Radiation monitor would provide the earliest indicationthat a leak exists?
A. 3ARC-RE2 1 Condenser Air Ejector Radiation Monitor B. 3MSS-RE77 C Main Steamline Radiation Monitor C. 3MSS-RE79 Terry Turbine Exhaust Radiation Monitor D. 3SSR-RE08 Steam Generator Blowdown Radiation Monitor Answer: A c
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.i /r-Tier # 2 2 Question # 52 Group # 1 1 WA # 076.KI . I 7 Importance Rating 3.6 2.7 Proposed Question:
From which Heat Exchangers would the Process Radiation Monitoring System be able to detect activity entering the Service Water System?
A. RPCCW Heat Exchangers
- 3. TPCCW Heat Exchangers C. CCE Heat Exchangers D. RSS Heat Exchangers Proposed Answer: D Explanation (Optional): D is correct, since Radiation Monitors 3SWP*RE60A and B sample the Service Water System downstream of the RSS Heat Exchangers. A, B , and C are wrong since these Heat Exchangers do not have Rh4S monitors associated with them, but are plausible since they are heat exchangers cooled by the Service Water System.
~. Technical Reference(s): P&ID 133A, B, C, and D (Attach if not previously provided)
-\_-
Proposed references to be provided to applicants during examination: None Learning MC-00165 Describe the function and location of the following Radiation (As available)
Objective: Monitors.. . SWP*RE-60A/B Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.2 to 41.9 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Ou KW+ @J Level RO SRO of Tier# 2 2 Question # 53 Group # 1 1
/OS/ 6.1 WA # 078.K3.02 p 4 L d , 7 Importance Rating 3.4 3.6 Proposed Questi /eceJ v cLvo The reactor is manually tripped due to a complete loss of instrument air.
How will the loss of instrument air affect the control of Tave after the trip?
A. The Steam Generator Safety valves and Atmospheric Dump Bypass Valves will be available.
B. The Steam Generator Safety valves and Condenser Steam Dumps will be avaiIable.
C. The Steam Generator Safety valves and Atmospheric Steam Dumps will be available.
D. Only the Steam Generator Safety Valves will be available.
Proposed Answer: A Explanation (Optionai): A is correct, and B and C are wrong since the condenser steam dump valves and the atmospheric dump valves are pneumatic, failing closed on a loss of air. D is wrong, since the Atmospheric Dump Bypass Valves are Motor Operated Valves.
- 1 Technical Reference(s): AOP 3562, page 3. (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-05324 Given a failure of the partial or complete, of the plant air systems, Objective: determine the effects on the systems and interrelated systems. (As available)
Question Source: Bank #60670 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.4 I .7 Comments:
29 of 34 NUREG-1021, Revision 9
I 1 LOSS OF INSTRUMENT AIR AOP 3562 Page 3 Rev. 4 Svstern ResDonse CNS - Normal makeup fails open DGS - Containment header isolates FWA - Terry turbine starts FWS - Feedwater control and bypass valves fail closed
- Feedwater isolation trip valves can not be opened due to failure of air -driven hydraulic pump GWS - PRT vent fails closed
- Containment isolation valves close
- Process vent fan dampers close MSS - Atmospheric steam dumps fail closed
- Condenser steam dumps fail closed PRT - Tank outlet valve closes RHR - Temperature control is lost
- WCCW to RHR heat exchangers fails as is due to SOV lockup valves.
- RHR flow control valve fails open
- RHR bypass flow control valve fails open SFC - Purification flow lost
- Makeup valve to spent fuel pool fails closed SWP - Dilution water supply isolation valves for the hypochlorite pumps fail closed WTS - Acid/causticmix tanks drain to TIC- 1-1 Ventilation - Most dampers fail open
- System remains operating
- Outside air mixing is terminated
- Recirculation dampers open
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO L l Tier # 2 2 Question ## 54 Group # 1 1 WA # 103.K4.06 Importance Rating 3.1 3.7 Proposed Question:
With the plant at IOOYo power, an inadvertent Containment Isolation Phase A (CIA) occurs. The reactor has NOT tripped and Safety Injection has NOT actuated.
How has the CL4 signal affected the Chemical and Volume Control System (CHS) Containment paths?
A. None of these CHS System valves have closed, since the SIS signal is what isolates the CHS CTMT paths.
B. The Letdown Isolation Valves (3CHS*CVS152 and 8 160) have closed only.
C. The Letdown Isolation Valves (3CHS*CV8152 and 8160) and the RCP Seal Return Isolation Valves (3CHS*MV8100 and 81 12) have closed only.
D. The Letdown Isolation Valves (3CHS*CV8152 and 8160), RCP Seal Return Isolation Valves (3CHS*MV8100 and 8 1 12), and Charging Flow Control Isolation Valves (3CHS*MV8105 and 8 106) have closed.
Proposed Answer: C Explanation (Optional): The CIA signal closes the Letdown Isolation Valves and the RCP Seal Return Isolation Valves v-
~
(C correct, A and B wrong). The Charging Flow Control Isolation Valves are closed by the SIS signal (,,D wrong, A, B, and D plausible).
Technical Reference(s): P&ID 104A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-04199 Describe operation of the Controls and Interlocks associated with the (As available)
Objective: following Chemical and Volume Control Systems components.. . Letdown Line Containment Isolation Valves.. . Charging Header Isolation Valves.. . Reactor Coolant Pump Seal Water Return Containment Isolation Valves.. .
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
~~
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 013.Pl.07 Examination Out Level RO SRO fVd,%+ /fiQ
- i-b LA,,
Question # 55 caen3&
45s4dcg fq ut%
-hop EKAS Tier #
Group #
KIA #
2 1
013.A1.09 2
1 Importance Rating 3.4 3.7 Proposed Questic dvdr Initial Conditions:
0 The plant is at 100% power, beginning of life.
Rod Control is in MANUAL.
0 Tave is on program.
RE has requested the crew to slowly withdraw control bank D rods to full out after MTC testing.
The crew is to allow MTC to control reactor power, without borating during the rod withdrawal.
The RO slowly withdraws control bank D rods, resulting in the following:
0 RCS Narrow Range That increases by 4F.
PZR pressure control system maintains RCS pressure stable.
0 AFD remains in the program band.
How do the OTAT and OPAT trip setpoints respond?
A. OTAT setpoint DECREASES.
~- OPAT setpoint DECREASES.
B. OTAT setpoint DECREASES.
OPAT setpoint DOES NOT CHANGE.
C. OTAT setpoint DOES NOT CHANGE.
OPAT setpoint DECREASES.
D. OTAT setpoint DOES NOT CHANGE.
OPAT setpoint DOES NOT CHANGE.
Proposed Answer: A Explanation (Optional): OTAT setpoint is penalized by Tave increasing (A correct), pressure decreasing, or AFD increasing. Tave penalizes OPAT above 587°F (A correct) or Tave increasing. Distractors are plausible since a change in pressure or AFD affects only one of the two signals, pressure is stable, and a change in PRNI doesnt change either setpoint.
Technical Reference(s): Tech Spec Table 2.2-1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MC-05493 Describe the operation of the following RPS controls and (As available) interlocks.. . OTAT.. . OPAT.. .
Question Source: Bank #64304 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 L, Comments:
29 of 34 NUREG-I 021,Revision 9
i JTABHdE 2.2-1(Continuedl TABLE NOTATIONS z
i?
r NOTE 1:OVERTEMPERA" AT m
4 2
m i
2
?where: AT is measured Reactor Coolant SystemAT, OF; AT0 is loop specific indicated AT at RATED THERMAL POWER,OF; w
(1 +zp)
(1+~ ~ is8the)function generated by the lead-lag compensator on measured A T 21 and 22 are the time constants utilized in the lead-lag compensator for AT,21 2 [*] sec, T~ 5 [*) sec; : I K, r: [*I I Kz 2 [*]/OF; 1 (1 + y f is the function generated by the lead-lag compensator for TaVg; (1 +y)
~4 and 5 are the time constants utilized in the lead-lag compensator for Tavg,24 2 [*I see, I [*] sec; T is measuredReactor Coolant System average temperature, OF; T' is loop specific indicatedTavgat RATED TNERMAT, POWER,S [*]OF; K3 2 [*]/psi P is measured pressurizer pressure, psia; P' is nominal pressurizerpressure, 2 [*] psia; s is &e Laplace transform operator, sec-*;
(The values denoted with [*] are specified in the COLR.)
f LE 2.21 (Continued)
(Continued) and fi (AI) is a fimction of the indicated differencebetween top and bottom dekctors.ofthe power range neutron ion chambers;with nominal gains to be selected based on measured instrumentresponse during plant startup tests calibrations such that:
(1) FOX qt - qb between [*I% and [*I%, fi, (AI) 1[*I, where Q and ql, percent RATED THERMAL POWER h I the upper and lower halves of the core, respectively, and qt -+ e,is the total T H E W P O m R in percent RATEDTHERMALPO~
For each percent that the magnitude of Q - qb exceeds I*]%, the AT Trip Setpoint shallbe automatically (2) reduced by 2 [*I% of its value at RATED T H E W POWER.
For each percent that the magnitude of q qb exceeds [*I%, the AT Trip Setpoint shall be automatically I
(3) reduced by 2 [*I% of its value at RATED THERMAL POWER. I NOTE 2: The maximum channel as left trip setpoint shall not exceed its computed trip setpoint by more than the following:
(1) 0.4% AT span for the AT channel (2) 0.4% AT span for the T ,, channel (3) 0.4% AT span for the pressurizer pressure channel (4) 0.8% AT span for the f(AI) channel Z
P (The values denoted with [*] are specified in the COLR.) I 8
E P
1 s
F W L E NOTATION8 2
0 NOTE3: OVERPOWER AT z
m i
3 u
. Where: AT is measured Reactor Coolant System AT, OF; AT0 is loop specific indicated AT at MTED THEXMAL POWER, OF; (1 +TIS)
( 1+ 2
- s. is the function generated by the lead-lag compensatoron measured AT T~ and 22 are the time constants utilized in the lead-lag compensatorfor AT, TI1 [*] sec, q I [*I sec; I
& 2 [*I; I ICs >[*]/OF for increasing Tavg and K5 S [*] for decreasing Tavg; I (275) is the function generated by the rate-lag compensatorfor Tavg; (1 + 77s) 2, 52 [*] sec; is the time constant utilized in the rate-lag compensator for Tavg, I T is measuredaverage Reactor Coolant System temperature, OF; T"is loop specific indicated Tavg at RATED THERMAL POWER,S [*]OF; I
& 1 [*]/OF when T > T" and S [*]/OF when T S TI'; I s is the Laplace transform operator, sec-*; 5 w
.$ (The values denoted with r*l are mecified in the COLR .)
UI l! LE NOTATIONS (Continued)
NOTE 4 The maximum channel as left trip setpoint shallnot exceed its computed trip setpoint by more than 0.4% AT span for 9
W the AT channel and 0.4%AT span for the Tavgchannel.
w NOTE 5: Setpoint is for increasing power.
NOTE 6: Setpoint is for decreasing power.
Z
?
"8 L
LJl W
~- ~~
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
,=
Tier # 2 2 Question # 56 Group # 2 2 KIA # 014.K4.03 Importance Rating 3.2 3.4 Proposed Question:
With the plant at 80% power, the crew is preparing to recover a dropped Control Bank A rod.
At what point will the rod bottom (RB)LED and the ONE ROD BOTTOM annunciator on MB4 clear for the applicable rod?
A. Both the RE3 LED and the ONE ROD BOTTOM annunciator will clear when GRPI indicates 4 steps.
B. The RE3 LED will clear when GRPI indicates 10 steps, and the ONE ROD BOTTOM annunciator will clear when GRPI reaches 12 steps.
C. The RE3 LED will clear when GRPI indicates 4 steps, and the ONE ROD BOTTOM annunciator will clear when GRPI reaches 10 steps.
D. Both the RJ3 LED and the ONE ROD BOTTOM annunciator will clear when GRPI indicates 12 steps.
Proposed Answer: A Explanation (Optional): A is correct, since with Data A and Data B operating, accuracy of the system is f 4 steps.
-. . When coil B-1 is completely penetrated at 3 steps (4 steps due to DRPI accuracy), the 6 step LED will light (Byand D wrong). The ONE ROD BOTTOM annunciator will clear with Bank demand > 12 steps (it is at approximately 220 steps) and no rods on the bottom, which occurs at 4 steps (A correct, and C wrong). The 10 step setpoint is plausible since this is related to the DRPI accuracy with the loss of Data A or 3. The 12 step setpoint is plausible since this is the setpoint for Bank Demand position that enables the ONE ROD BOTTOM annunciator. This also is the Bank A rod height where the ROD BOTTOM annunciators normally clear on a reactor startup (all bank A rods off the bottom and bank demand >I2 steps).
TechnicaI Reference(s): AOP 3552, Att. C , NOTE prior to step 3 (Attach if not previously provided)
ARP 3353.MB4C, 5-10 LSK 25-5B Proposed references to be provided to applicants during examination: None Learning MC-05482 Describe the operation of the following Rod Position Indication System (As available)
Objective: Controls and Interlocks.. . Rod position indicators.. .
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 c-Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 57 Group # 2 2 K/A # 0 16.A2.04 Importance Rating 2.5 2.6 Proposed Question:
With the plant at 100% power, with turbine load control on Load Set, when the following sequence of events occurs:
- 1. C-16 comes in, and the EHC HOLD light illuminates.
- 2. I&C investigates, and the cause is found to be improper instrument voltage.
- 3. The voltage problem is repaired, and the crew is preparing to restore from HOLD operation.
What voltage problem could have caused C-16 to actuate, and how will the crew restore from HOLD operation?
A. Voltage faiIed low to a Narrow Range Tho,instrument. The crew will lower the Load Limit pot until the HOLD light clears.
B. Voltage failed high to a Narrow Range Tho,instrument. The crew will depress any Load Decrease rate pushbutton.
C. Voltage failed low to the selected Turbine Impulse Pressure instrument. The crew will Iower the Load Limit pot until the HOLD light clears.
D. Voltage failed high to the selected Turbine Impulse Pressure instrument. The crew wilI depress any Load Decrease rate pushbutton.
Proposed Answer: A Explanation (0ptional):C-I6 comes in if Auctioneered Low Tave is 20°F lower than Tref. This can be caused by Tave failing low (A correct, B7wrong), or Tref failing high (Cwrong). The C-16 HOLD condition is cleared by lowering the Load Limit pot to take control away fi-om Load Set (A correct, D wrong). B and D are plausible since the HOLD pushbutton is included in the row of Load Decrease Rate pushbuttons on the EHC insert.
Technical Reference(s): 3323A, section 4.1 I (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-07084 Describe the operation of the controls and interlocks associated with the (As available)
Objective: following EHC controls system load control unit circuits ... load set circuit...
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 and41.10 Comments:
29 of 34 NUREG-I 021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
\/ Tier # 2 2 Question # 58 Group # 2 2 KIA # 0 17.A2.02 Importance Rating 3.6 4.1 Proposed Question:
A small break LOCA is in progress, and plant radiation levels indicate significant fuel damage has occurred.
Current ICCRVLMS status is as follows:
0 Several CETCs are flashing and read > 5000°F.
0 The US tells the crew that 5000°F indicates an open condition, likely due to damage from the event in progress.
0 The remaining CETCs indicate about 1400°F.
0 Subcooling margin indicates 1,999"F.
0 RCS pressure is 1200 psia.
What is the status of the ICCRVLMS?
A. The CETC data channels reading >5000"F have been automatically removed from scan. The CETCs reading 1400" are in an out-of-range condition. Actual core exit temperature is higher than 1400°F.
B. The CETC data channels reading >5000"F have been automatically removed from scan. The CETCs reading 1400" are providing accurate data.
C. The CETC data channels reading >5000°F are providing faulty data to the subcooling indication and should be L,'
removed from scan. The CETCs reading 1400" are in an out-of-range condition. Actual core exit temperature is higher than 1400°F.
D. The CETC data channels reading >5000"F are providing faulty data to the subcooling indication and should be removed from scan. The CETCs reading 1400" are providing accurate data.
Proposed Answer: D Explanation (Optional): "A" and "B" are wrong, since the CETCs must be manually removed from scan. "D' is correct, and "C" is wrong, since CETCs are able to read well above 1400'F. "C" is plausible, since the subcooled margin display indicates a value of 1999°F when an out-of-range condition exists, and this would indicate very high CETC temperatures if it were accurate.
Technical Reference(s): OP 3301K, Cautions prior to step 4.1.1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: Steam Tables Learning MC-04832 Describe the operation of the ICC system during the following.. . Loss of (As available)
Objective: core cooling accident.. .
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.4 1.5 55.43.5 Comments:
29 of 34 NUREG-1021,Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO L
Tier # 2 2 Question # 59 Group # 2 2 WA # 027.A4.01 Importance Rating 3.3 3.3 Proposed Question:
The crew is responding to a small break LOCA, and the following conditions exist:
CDA actuated 14 minutes ago.
CTMT pressure is 16.5 psia and decreasing.
Alarm lights for both CTMT high range rad monitors (3RMS*RE04A and REOSA) are LIT.
The crew is at E-1 step 7 Check if CTMT Spray Should Be Stopped.
The ADTS has determined that the crew will operate only the A QSS pump to reduce CTMT radiation levels.
RWST level is 800,000 gallons.
Based on current plant conditions, what are the minimum actions physically required by the RO in order to align the QSS pumps so that only the A QSS pump wilI be spraying containment?
A. Reset CDA, take the A QSS pump to START and back to AUTO, and open the A QSS pump discharge spray valve.
B. Reset CDA, take the B QSS pump to STOP and back to AUTO, and close the B QSS pump discharge spray valve.
- 4 C. Reset SIS and CDA, take the A QSS pump to START and back to AUTO, and open the A QSS pump discharge spray valve.
D. Reset SIS and CDA, take the B QSS pump to STOP and back to AUTO, and close the B QSS pump discharge spray valve.
Proposed Answer: B Explanation (Optional): The QSS pumps are already running fkom the CDA signal with adequate RWST level (Aand C wrong). Only the CDA signal needs to be reset in order to allow stopping of the pumps (B correct, D wrong).
Technical Reference@): E-1, steps 7 and 22. (Attach if not previously provided)
LSK 24-9.4F and 27-12F Proposed referehces to be provided to applicants during examination: None Learning MC-05 171 Describe operation of the following containment depressurizationsystem (As available)
Objective: components controls and interlocks: Quench Spray System (QSS)...
Question Source: Bank # 76237 Question History: Millstone 3 2002 LOUT NRC Exam Questiob Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO L. Tier # 2 2 Question # 60 Group # 2 2 WA # 034.K4.02 Importance Rating 2.5 3.3 Proposed Question:
Current Conditions:
0 Fuel movement in the spent fuel pool is about to begin.
The spent fuel bridge operator is directed to move the spent fuel bridgehrolley and pick up the spent fuel-handling tool.
The tool is stored in its normal storage location 0 Currently the spent fuel bridge hoist is up.
Which of the following interlocks will require action by the bridge operator?
A. The geared limit switch hoist f i l l up interlock will need to be bypassed using the Bypass Enable key switch.
B. The low speed interlock must be bypassed to travel at any other speed to the north wall.
C. The Traverse Travel Limits interlock will need to be averted using the avert pushbutton.
D. The Slack Cable interlock will need to be averted using the avert pushbutton.
L-Proposed Answer: D Explanation (Optional): UPPERLOWER GEARED LIMIT switch stops the hoist at ends of travel. This will not need to be enabIed (A wrong). B is wrong, since low speed switches have been removed, speed is programmed into the bridge. The TRAVERSE TRAVEL LIMIT INTERLOCK stops bridge/trolley motion when the hoist approaches restricted areas adjacent to the pool walls. This interlock illuminates the traverse travel lamp, which may be AVERTED. C is wrong since this is not in effect, as the tool is stored on the north wall. The SLACK CABLE INTERLOCK (200 pounds) stops the hoists downward motion and illuminates the slack cable lamp. It may be AVERTED, and in this case needs to be averted since there is no load on the hoist (Dcorrect).
Technical Reference(s): OP3303A, Attachment 1. (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None
~~~b~ MC-0454 1 Describe the operation of the following Fuel Handling System (As available)
Objective: equipment, controls, and interlocks... Spent fuel handling tool, Spent fuel bridge crane and hoist.. .
Question Source: Bank #69795 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-?021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO h.-
Tier # 2 2 Question # 61 Group # 2 2 WA # 035.K6.02 Importance Rating 3.1 3.5 Proposed Question:
With the plant initially stable at 50% power, a transient occurs, the BOP reports the following parameters for the A SG, with similar trends on the other SGs:
0 SG Pressure: 1000 psia and decreasing 0 SGLevel: 52% and increasing 0 SG Steam Flow: 2.0 mpph and increasing Based on plant conditions, what event is in progress?
A. A dilution event is in progress.
B. A turbine runback is in progress.
C. A Steam Generator tube has ruptured.
D. A SG Atmospheric dump valve has failed open.
Proposed Answer: D
\-
Explanation (Optional): D is correct, since a failed open atmospheric dump valve will increase steam flow for all SGs via the common main steam header. Increased steam flow will cause SG pressures to drop and swell to raise SG Narrow Range Ievels. A is wrong, since SG pressures are decreasing, but plausible since a dilution would cause increased steam flow. B is wrong, since this would decrease steam flow, but plausibIe, since this would affect all of the parameters listed in the stem. C is wrong, since a tube rupture will not cause SG pressure to decrease, but is plausible since it would cause SG level to increase.
Technical Reference(s): Westinghouse Transient Analysis Text, page 8.33 (Attach if not previously provided)
Proposed references to be provided to appIicants during examination: None Learning MC-0488 1 Describe the major parameter changes associated with increased heat (As available)
Objective: removal by the secondary system.
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO L
Tier # 2 2 Question ## 62 Group # 2 2 WA # 045.K3.01 Importance Rating 2.9 3.2 Proposed Question:
With the plant at 100% power, the main turbine quill shaft starts to fail, and Main Shaft Oil Pump (MSOP) discharge pressure starts to decrease.
Assuming the quill shaft fails completely and MSOP pressure drops to zero, how will the turbine lube oil system maintain lubrication to the Main Turbine, and what will be the effect on the rest of the plant?
A. The Turning Gear Oil Pump will automatically start, maintaining lubrication to the main turbine, preventing a reactor trip.
B. The Turning Gear Oil Pump will automatically start, maintaining lubrication to the main turbine, but the reactor will trip.
C. The Auxiliary Oil Pump will automatically start, maintaining lubrication to the main turbine, preventing a reactor trip.
D. The Auxiliary Oil Pump will automatically start, maintaining lubrication to the main turbine, but the reactor will trip.
Proposed Answer: B Explanation (Optional): The AOP and TGOP receive AUTO-START signals when MSOP discharge pressure drops to 190 psig, with the TGOP supplying the Main Turbine and the AOP supplying the Main Feed Pumps (Cand D wrong, but plausible). The reactor will trip, since low MSOP discharge pressure automatically trips the turbine at I00 psig with turbine speed >75% (B correct, A7wrong). Ais plausible since the TGOP start will maintain bearing oil pressure >24.5 psig, which is the setpoint for the Bearing Oil Pressure Low turbine trip.
Technical Reference(s): P&ID 141A and C (Attach if not previously provided)
LSK 16-2A and B OP 3353.MB7B, 2-2 and 3-1 Proposed references to be provided to applicants during examination: None Learning MC-05770 Describe the operation or the following Main Turbine Controls and (As available)
Objective: Interlocks... Shaft Driven Lube Oil Pump Pressure.. .
Question Source: Bank #69539 Question Cognitive Level: Memory or Fundamental Knowledge I O CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5
\- Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Question # 63 Group # 2 2 K/A # 075.K1 .OS Importance Rating 3.2 3.2 Proposed Question:
While preparing to perform a thermal backwash of the B Circulating Water Bay, the crew starts the C Service Water Pump and stops the A Service Water Pump.
Why is the A Service Water Pump removed from service?
A. This ensures lube water will not be lost to the A Train Service Water Pump.
B. This ensures adequate NPSH is available for the A Train Service Water Pump.
C. This prevents excessive DP across the A Traveling Water Screen.
D. This minimizes the increase in A Train Service Water temperature.
Proposed Answer: D Explanation (Optional): D is correct, since during a thermal backwash of the B Circ Water Bay, hot returning circulating water is discharged in the vicinity of the LAService Water Pump suction. RPCCW high temperature alarms can result. A is wrong, but plausible, since lube water is supplied by the service water system for the Circ Pumps, not I vice-versa. B is wrong, but plausible, since Circ Pumps are automatically tripped off in bays with Service Water pumps on high screen DP (not backwash) in order to preserve NPSH. C is wrong since excessive DP has not been a problem during backwash operations, but the Circ pump is supplying its condenser bay, as well as backwashing the adjacent bay.
Technical Reference(s): OP 3325A, precautions 3.18-3.22 (Attach if not previously provided)
OP 3325A, section 4.4.
Proposed references to be provided to applicants during examination: None Learning MC-04282 Describe the major administrativeor procedural precautions and limitations (As Objective: placed on the operation of the main circulating water.. . systems, and the basis for each. available)
Question Source: Bank #665 13 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.4 and41.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination
~~ ~-~~-
Question Worksheet Form ES-401-5
~
RO SRO 2 2 2 2 0- ze\sor~ KIA # 001.K6.02 Importance Rating 2.8 3.3 I& C-KV3 Proposed Quesi Current Plant Conditions:
rn A plant startup is in progress in accordance with OP 3203 Pianl Startup.
Reactor power is 25%.
0 All control systems are in AUTOMATIC.
Control Bank D rods are at 150 steps.
The lower detector for power range nuclear instrument channel N4 I fails HIGH, causing channel total power indication to increase to 70%.
Assuming no operator action is taken, what will be the response of the rod control system?
A. Rods will insert, and then withdraw.
B. Rods will drive in some distance then stop.
C. Rods will drive all the way in.
D. Rods will not move.
Proposed Answer: A Explanation (Optional): A is correct, since the power mismatch circuit initially senses nuclear power increasing relative to turbine power, resulting in rods driving in. This causes Tave to decrease, creating a temperature mismatch.
The power mismatch signal will decay away, and the temperature error will cause rods to withdraw to restore Tave to program. B is plausible since this would be the correct answer ifthe failure happened at 100% power, since C-2 overpower rod stop (I of 4 channels) would prevent rod withdrawal on the NI failure. Cis plausible, since this would be the correct answer on a temperature instrument failing low, since it is not rate dependent. D is plausible, since power is above 20% C-1 IR rod block setpoint, but this is procedurally blocked when power is raised above P-10.
Technical Reference(s): Functional Sheet #9 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MC-05229 Given the folIowing conditions, determine the effects on the (As available)
NIS system, and on interrelated systems.. . Power Range Instrument failure in MODE 1 above P IO...
Question Source: Bank #75447 Question Cognitive Level: Comprehension or Analysis I O CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
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ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
-.I Tier # 2 2 Question ## 65 Group # 2 2 K/A # 086.A4.05 Importance Rating 3 .O 3.5 Proposed Question:
With the plant operating at 100% power, the following sequence of events occurs:
- 1. The SIMPLEX Color Graphics Unit is alarming.
What action can the operators perform at the Color Graphics Unit during this event?
A. Check for the RSST A Water Flow alarm, confirming that the A RSST Deluge Valve has automatically opened.
B. Check for the RSST A Water Flow alarm, confirming that one or more of the closed-head sprinkler heads has melted.
C. Check the status of the Deluge Valve timer, verifying that the Deluge Valve automatically closes at the required time.
D. Command the ARSST Deluge Valve to open, commencing deluge spray to the A RSST.
Proposed Answer: A
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-I Explanation (Optional): A is correct, since the indications available to the control room are the fire alarms from the heat detectors, cutout valve status, and water flow. B is wrong, since the RSST is protected by an open head deluge sprinkler system, but plausible, since the closed head, wet pipe sprinkler is the most widely used fire water system at MP3, and water flow is monitored for the RSST (ZP9B, Zone 9). C is wrong, since the RSST deluge valve does not automatically close, but plausible since the filter bank deluge valves do have a timed deluge. D is wrong, since deluge valve operation capability has been disabled due to an inadvertent initiation. D is plausible, since deluge valve operation capability previously existed.
Technical Reference(s): OP 3341A, Section 1.2 and step 4.17 (Attach if not previously provided)
OP 3341D, Tables 12, 13, and 17 sheet 17 OP 3341D. stem 4.3.4 and 4.4.1 Proposed references to be provided to applicants during examination: None Learning MC-04597 Describe the individual operations which can be performed at the Main (As available)
Objective: Fire Protection ConsoIe.
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.-..--, Tier # 3 3 Question # 66 Group ## 1 I K/A # GEN.2. I .23 Importance Rating 3.9 4.0 Proposed Question:
Which action is specifically required of the RO during a reactor startup?
A. Independently review the ECP for accuracy.
B. Be aware of the effect of transient Samarium on predicted rod height.
C . Use diverse indications to monitor the core's response.
D. Remain cognizant of 1/M plots.
Proposed Answer: B Explanation (Optional): "B" is correct, and "A", "C", and "D"are wrong since "B" is the only answer not specified in DNAP 1410, section 3.8.9 as an RO responsibility during an approach to criticality.
TechnicaI Reference(s): DNAP 1410, section 3.8.9 (Attach if not previously provided)
-- Proposed references to be provided to applicants during examination: None Learning MC-06343 Understand reactivity management principles as outlined in DNAP 1410, Objective: Reactivity Management. (As available)
Question Source: Bank #77880 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 5 5.4 1.10 Comments:
29 of 34 NUREG-I021, Revision 9
ES-401 Written Examination Question Worksheet FOITII ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier ## 3 3 Question # 67 Group # 1 1 KIA # GEN.2.1.24 Importance Rating 2.8 3.1 Proposed Question:
With the plant shutdown, the following sequence of events occurs:
The BOP operator attempts to close a 4KV feeder breaker at MB8.
The breaker does not close.
0 The green light remains lit.
A PEO goes to the breaker and reports that the local green light is lit.
The PEO also reports that the white auxiliary circuit light is NOT lit.
Using ESK SA, attached to the back of this exam, what caused these indications?
A. A UL fuse is blown.
B. A UT fuse is blown.
C. A UC fuse is blown.
D. There is an open in the trip coil.
-L-' Proposed Answer: C Explanation (Optional): The UC fuses supply the closing circuit and the white auxiliary circuit light, but not the position indicating lights ( V correct). The UL fuses power the elevator motor, but the breaker is already racked up, as evidenced by the indicating lights ("A" wrong). The UT fuses feed the UC fuses, but they also feed the indicating lights, which still indicate ("B"wrong). An open trip coil would not prevent the green light from indicating, but it also would not prevent the breaker from closing ("D"wrong).
Technical Reference(s): ESK 5A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: ESK 5A Learning MC-04325 identify the following information on electrical control schematics (As available)
Objective: (ESKs). .. Circuit Flowpaths...
Question Source: Bank#076131 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.45.12 & 13 Comments:
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ES-401 Written Examination Ouestion Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO I
Tier # 3 3 Question # 68 Group # 1 1 WA # GEN.2.1.25 Importance Rating 2.8 3.1 Proposed Question:
A plant is being cooled down to 400°F in accordance with OP 3208 Plant Cooldown.
Current conditions are as follows:
Core Burnup is 6,000 MWD/MTU.
0 RCS Temperature is 450°F.
The core is Xenon Free.
Using the SHUTDOWN MARGIN curves attached to this exam, what is the minimum amount of boron required to ensure adequate SHUTDOWN MARGIN in accordance with OP 3209B Shutdown Margin?
A. 128Oppm.
B. 1450 ppm.
C. 1610ppm.
D. 2060ppm.
-v.
Proposed Answer: C Explanation (Optional): C is correct since curve RE-B-02 Shutdown Margin - MODE 3 shows 1610 ppm boron required at 350°F at 6,000 MWDMTU burnup. A is plausible, since 1280 ppm is obtained if the 557°F curve is used. B is plausible since 1450 ppm is obtained if 450°F is interpolated between the curves on the graph, but this is not allowed per OP 3209B. The lowest temperature curve must be used. D is plausible since this value is obtained if the MODE 4 curve is used.
Technical Reference(s): Shutdown Margin Curves RE-B-02 and 03 (Attach if not previously provided)
OP 3209B, page 3, third paragraph OP 3209B, step 4.2.3.c Proposed references to be provided to applicants during examination: Curves RE-B-02 and 03 Learning Objective: MC-03475 Perform a S D Margin Calculation in MODES 3,4, or 5. (As available)
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFRPart 55 Content: 55.41.10 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
I I
2521 2-ER-04-0020, Rev. 0 RE-B-02 Page C( of Z/
MP3-10-00 Shutdown Margin MODE 3 Loops Filled 0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 Core Burnup (MWDIMTU)
Preparer/Dat Reviewer/Dat Page 1 of 2 Approver/Dat
RE-B-03 Rev. 0 25212-ER-04-0020, MP3-I 0-00 Page // of ZZ 2500 n
E P
8 2000 1500 1000 500 0
0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 Core Burnup (MWD/MTU)
Page 1 of 2
This procedure is performed for MODEs 3,4 or 5 by determining current or anticipated plant conditions and verifying, for the condition analyzed, that adequate RCS boron concentration is, or will be, available. If adequate RCS boron concentration is not available for current conditions, the Operator is directed to either perform a xenon correction if in MODE 3 or immediate borate. If adequate RCS boron concentration will be available for anticipated conditions, the Operator is directed to perform normal boration or perform a xenon correction if anticipating MODE 3. Credit for samarium concentration is nQt taken when .I performing a shutdown margin or reactivity determination for MODEs 3, 4 and 5.
For convenience, the MODE 3 shutdown banks out curve is inchded on the MODE 3 Shutdown Margingraph and directions for the cuves use is included in this procedure. The w e ensures the MODE 3 Technical Specification for less than 0.99 with the shutdown banks out is met.
When referred to this procedure by another procedure to perform a SHUTDOWN MARGIN determination and the shutdown banks are withdrawn, this curve should be used and treated the same as a Shutdown Margin m e .
When determining the value of a parameters reactivity or worth, visual interpolation between curves on a graph may be employed. If interpolation is not used, ensure the most conservative curve is utilized for the application. Do interpolate between curves on separate pages, instead use the most conservative value for the given condition. Do not interpolate between required boron concentration curves on the Shutdown Margin graphs.
1.3 Applicability This procedure may be performed in any MODE to determine SHUTDOWN MARGIN requirements met.
1.4 Frequency This procedure is performed at the normal surveillance frequency or as directed by other procedures which may require determination of adequate SHUTDOWN MARGIN in response to degraded equipment or plant conditions.
OP 3209B Rev. 010-02 Continuous 3 of 20
4.2 Shutdown Margin Determination for MODES 3,4, and 5 4.2.1 VERIFY General Prerequisites completed.
4.2.2 Refer To OP 3209B-002 and RECORD time and date.
NOTE With DRPI deenergized, verification of rod position is based upon the rod position prior to deenergizing the Digital Rod Position Indication System.
4.2.3 E determining SHUTDOWN MARGIN for current plant conditions, Refer To OP 3209B-002 and PERFORM the following in the Current Conditions section:
- a. CIRCLE the current MODE, and if applicable, the RCS loops condition (Refer To Definition 2.3.4).
- b. Enter the present core burnup value obtained from the plant process computer.
- c. PERFORM the following to determine temperature condition:
- 1) IF in MODE 3, AND maintaining no load Tavg, CIRCLE 557°F.
- 2) E in MODE 3, AND not maintaining no load Tavg, CIRCLE 350E
- 3) IF in MODE 4 or 5, CIRCLE the applicable MODE.
- d. IF performing SHUTDOWN MARGIN determination at increased surveillance frequency due to an inoperable control rod or an inoperable SMM, PERFORM the following:
- 1) REQUEST Chemistry sample the RCS for boron concentration and every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.
- 2) ENTER the most current 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RCS boron concentration sample result.
- 3) Go To step 4.2.3.f.
OP 3209B Rev. 010-02 Continuous 10 of 20
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Y
Tier # 3 3 Question # 69 Group # 2 2 WA # GEN.2.2.11 Importance Rating 2.5 3.4 Proposed Question:
Which of the following temporary installations in the plant would require a temporary modification tag to be installed per WCIO Temporary Modtjkations?
A. A temporary heater being plugged into a 120V outlet by the EHC skid to maintain EHC fluid temperature.
B. A hose being connected from a Service Air connection to a temporary sump pump in the east condenser pit.
C. Plastic sheeting being instaIled over a Domestic Water System MOV to protect it from rainwater.
D. A leaking relief valve being gagged shut on the A Auxiliary Boiler steam line.
Proposed Answer: D Explanation (Optional): D is correct, since gagging relief valves requires temporary modification controls. A and B are wrong since connections to service connections are not considered Temporary Modifications. C is wrong, since tarpaulins or plastic sheeting does not require Temporary Modification controls. A, B7,and C are plausible since each of these are instailations that are temporary in nature.
- - Technical Reference(s): WC- 10, Attachment 1 (Attach if not previously provided)
WCI 0,Section I . I .4 Proposed references to be provided to applicants during examination: None Learning Objective: MC-05 104 Outline the process for Temporary Modification installation. (As available)
Question Source: Bank #74354 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 55.43.3 and 43.5 Comments:
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ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.*,- Tier # 3 3 Question # 70 Group # 2 2 KIA # GEN.2.2.13 Importance Rating 3.6 3.8 Proposed Question:
Near the end of an outage, a PEO is sent out to clear a large tagout in Containment, and the following sequence of events occurs:
- 1. The PEO brings a copy of the tagout, rather than the original, in order to prevent the tag sheet from becoming contaminated.
- 2. While clearing tags, it becomes apparent that configuration control has been lost on the system.
- 4. While exiting Containment,the PEO discards the contaminated tag in the rad waste barrel and marks on the tagout order that the tag was destroyed.
- 5. The PEO transfers his signatures to the original tagout document.
- 6. The PEO reports to the tagging authority that one tag was contaminated, and that a loss of configuration control on the system was evident.
Were the PEO's actions proper while clearing the tags?
A. No. The PEO was required to use the original tagout sheet to document the clearing of tags as they were being removed.
B. No. The PEO was required to stop clearing tags and contact the Tagging Authority as soon as he discovered the
- loss of configuration controI.
C. No. The PEO was required to stop clearing tags and contact the Tagging Authority as soon as he discovered one of the tags was contaminated.
D. Yes. The PEO dealt with the tagout copy, the configuration control issue, and the contaminated tag properly.
Proposed Answer: B Explanation (Optional): "Bl' is correct and "D" wrong since the PEO is required to STOP and CONTACT the tagging authority if loss of configuration control is evident. "A" is wrong since tagout copies are allowed, and the PEO properly transferred his signatures to the original document. "C" is wrong since the PEO properly disposed of the contaminated tag. The distractors are plausible since they involve unusual circumstances, and the PEO takes corrective actions to all events.
Technical Reference(s): WC2, section 1.20 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-06332 Identify the standard clearance practices for each of the following (As available)
Objective: categories: General Practices.. .
Question Source: Bank #67360 Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41. I O Comments:
29 of 34 NUREG-1021, Revision 9
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ES-40 1 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Question # 7 1 Group # 2 2 IUA # GEN.2.2.33 Proposed Question: Importance Rating 2.5 2.9 Initial Conditions:
The plant is at 80% power.
0 Bank D ControI Rods are at 186 steps.
Preparations are being made for a load increase.
The following sequence of events occurs:
- 1. The US directs the RO to place rods in MANUAL for the up power.
- 2. The RO inadvertently places the Bank Selector Switch in the CBD (Control Bank D) position.
- 3. The reactor power is then taken to loo%, all rods out (Bank D at 228 steps) condition.
- 4. Rod control is restored to AUTO, and the ROs error was not detected.
If NO hrther action is taken with respect to rod control, what will be the effect when the plant is subsequently shutdown?
A. Control Bank C rods will begin to insert early, with Bank D rods well above their proper setpoint.
B. Control Bank C rods will begin to insert late, kith Bank D rods well below their proper setpoint.
-.- C. The calculated RIL will cause the ROD CONTROL LIMIT LO LO annunciator to come in with Bank D rods well above the proper setpoint.
D. The calculated RIL will cause the ROD CONTROL LIMIT LO LO annunciator to come in with Bank D rods well below the proper setpoint.
Proposed Answer: A Explanation (Optional): The Bank Overlap Circuit determines when the individual banks start to insert in MANUAL or AUTO, and the P/A converter feeds the RIL computer and Bank D Full Withdrawal Limit C- 1 I. In Bank Select, the Bank Overlap Unit is fiozen, but the P/A Converter still receives the outward demand signal. Also, the RIL setpoint is based on AT, which is also working correctly (Cyand D wrong). With the Bank Overlap Unit frozen during the rod withdrawal, it did not detect Bank D rod height increasing from 186 steps, so when rods are inserted, it will improperly sense Bank D at 1 15 steps and start driving in Bank C rods early (A correct and Bywrong).
Technical Reference(s): Functional Drawing #9 (Attach if not previously provided)
OP 3302A, Precaution 3.6 Proposed references to be provided to applicants during examination: None Learning MC-05477 Describe the operation of the following Rod Control System controls and (As available)
Objective: interlocks... Bank Selector Switch...
Question Source: MPO Exam Bank Question History: 2000 Kewaunee 1 NRC Exam Question Cognitive Level: comprehension or Analysis 10 CFR Part 55 Content: 55.41.6 55.43.6 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
-_I Tier # 3 3 Question # 72 Group # 3 3 WA # GEN.2.3.4 Importance Rating 2.5 3.1 Proposed Question:
Current Conditions:
0 A Site Area Emergency has been declared due to a LOCA outside CTMT.
The LOCA is into the ESF building and a pathway to the environment exists.
0 Limited makeup to the RWST is available.
0 An operator is sent in to the ESF building to locally isolate the leak.
0 This action has all of the required approvals.
This action will result in a significant reduction in offsite dose, protecting a large population.
The operators current TEDE dose is as follows:
Exposure for current year: 200 mrem.
0 Current total lifetime exposure: 1200 mrem.
What is the maximum emergency exposure this operator may receive while performing this action?
A. 4800 rnrem TEDE B. 10000 mrem TEDE C. 23800 mrem TEDE D. 25000 mrem TEDE Proposed Answer: D Explanation (Optional): Emergency exposure limits for lifesaving or protection of large populations is 25 rem (D correct). A is plausible, since this dose would bring the workers annual dose to 5 rem, which is the maximum TEDE dose allowed per year. B is plausible, since this dose is the emergency dose authorized for the protection of valuable property. C is plausible, since this dose would bring the workers dose to 25 rem for the year, which is the emergency limit, but this limit is independent of previous dose.
Technical Reference(s): MP-26-EPI-FAP09 Attachment 3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning EP-00688 State the radiation exposure guidelines which have been established for (As available)
Objective: emergencies and the considerations for applying those guidelines.
Question Source: Bank #I74358 Question History: Millstone 3 2000 NRC exam.
Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.4 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO L
Tier # 3 3 Question # 73 Group # 3 3 WA # GEN.2.3. I 1 Importance Rating 2.7 3.2 Proposed Question:
With the pIant at 100% power and a SLCRS Fan running, the SLCRS Filter Fan Discharge to Millstone Stack radiation monitor 3HVR*RE19B goes into alarm.
What two subsequent actions are required to be performed to check for potential sources of the radioactive release?
A. Check the Condenser Air Ejector Discharge Radiation Monitor 3ARC-RE21 trend, and check the CTMT Purge Exhaust Fans running.
B. Check the RPCCW Radiation Monitor 3CCP-RE31 trend, and check the CTMT Purge Exhaust Fans running.
C. Check the Condenser Air Ejector Discharge Radiation Monitor 3ARC-RE21 trend, and check the CTMT Vacuum Pumps running.
D. Check the RPCCW Radiation Monitor 3CCP-RE3 1 trend, and check the CTMT Vacuum Pumps running.
Proposed Answer: C Explanation (Optional): C is correct since both the Condenser Air Ejectors and the CTMT Vacuum Pumps discharge to Gaseous Waste, which goes to the Millstone Stack. A and B are wrong, since the Containment Purge System exhausts to the Turbine Bldg stack, but plausible since it draws on CTMT, as does the CTMT vacuum pumps. B and D are wrong, since RPCCW is monitored by CCP-RE31, which is not alarming. B and D are plausible, since RPCCW overflows to the Aux Bldg, which is drawn on by SLCRS.
Technical Reference(s): AOP 3573, Attachment A, Page 10 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-04763 Describe the operation of the HVC/HVK systems under the (As available)
Objective: following...High radiation detected by HVC*RE16A or B.. .
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.8,41.10,41.11, and41.12 55.43.4 Comments:
29 of 34 NUREG-1021. Revision 9
RADIATION MONITOR AOP 3573 Page 10 of 12 ALARM RESPONSE Rev. 013-01 Attachment A Process Radiation Monitors Monitor Sample Stread Automatic/Subsequent Possible Cause Actions HVRI 8- 1,2 Waste Disposal Building exhaust to 1 Using OP 3314B, "Waste suction of Waste Disposal Building Dis osal and Fuel Building exhaust fans (FN8A and FW8B) dAC," Change Waste Disposal BuiIdmg ventilation from unfiltered to filtered mode.
2 Check trend history and alarm status of monitors (FloorplanrWSTEO4and WSTE24) RMS17, RMS18, RMS19, RMS20, and RMS25.
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HVRISA-1-5" SLCFG filter fan discharge to 1 Check trend history and Millstone stack. alarm status of monitors TRM 3.3.3.6.1 ARC21 and GWS48 as source of radioactivity.
HVRI 9B-1*
2 Check Ctmt Vacuum pumps REMODCM V.C.2. running as source of radioactivity.
3 Monitor HVR*19Aand take appro riate adion using MP-%-EPI-FApo6-003, Unit 3 Emergency Action Levels (Offsite Releases).
4 E 3HYR19A is inoperable AND 3HVR19B is offsde high, THEN Request HP obtain a closed-window ion chamber reading of the plume immediately downwind of the release point.
An asterisk (*) indicates the monitor is safety related or Class 1E.
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 c- Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Question # 74 Group ## 4 4 KIA # GEN.2.4. I 1 Importance Rating 3.4 3.6 Proposed Question:
The crew is responding to a loss of instrument air per AOP 3562 Loss of Instrument Air when the reactor trips.
In accordance with OP 3272 EOP Users Guide, which of the following actions is allowed in this situation?
A. Prior to entering E-0 Reactor Trip or Safety Injection, the US directs the RO by reading out-loud the step of AOP 3562 that operates Pzr heaters, assisting with RCS pressure control. Later, the remaining actions of AOP 3562 are performed in parallel with ES-0.1 Reactor Trip Response.
B. The US hands AOP 3562 to the extra senior licensed operator, who directs the BOP by reading out loud the steps of AOP 3562 while the US directs the RO to perform the immediate actions of E-0 Reactor Trip or Safety Injection.
C. After exiting E-0 Reactor Trip or Safety Injection, the US directs the RO and the BOP by reading out-loud all of AOP 3562. The crew then performs ES-0.1 Reactor Trip Response.
D. After entering ES-0.1 Reactor Trip Response, the US directs the RO and the BOP by reading out-loud steps from ES-0.1 and AOP 3562 in parallel. Only the steps of AOP 3562 that are necessary to ensure success of ES-0.1 are performed, without completing a11 of AOP 3562.
Proposed Answer: D Explanation (Optional): It is acceptable to perform the actions of an AOP in parallel with an ERG derived EOP provided the actions of the ERG-derived procedure receives priority (C wrong) and all the actions of the AOP are not initiated before completing all immediate actions of the ERG derived procedure (A and B wrong). It is not necessary to perform all steps in the parallel procedure. Only those steps necessary to ensure success of the ERG derived procedure need to be performed (D correct).
Technical Reference(s): OP 3272, Section 1.7. (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-04837 The crew operates the plant in compliance with all applicable plant (As available)
Objective: procedures.. .
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.10 Comments:
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ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO L-Tier # 3 3 Question # 75 Group # 4 4 WA # GEN.2.4.45 Proposed Question: Importance Rating 3.3 3.6 With the plant operating at 100% power, an earthquake occurs, and the following annunciators are received.
0 EARTHQUAKE (MBI) 0 RCP HI RANGE LKG FLOW HI (MB3)
TDFW PP A SUCTION PRESSURE LO (MB5)
GEN SEAL OIL PUMP DIS PRES LO (MB7)
GEN SEAL OIL TO H2 DP LO (MB7) 0 CRDM CLG FAN A AUTO TRIP (VPl A)
The board operators make the following reports:
0 #1 seal leakoff for the A RCP is 6.5 gpm and increasing slowly, and #I seal inlet temperature is stable.
0 Feed Pump suction pressure is 260 psig and stable.
0 Generator hydrogen pressure is 28 pounds and decreasing slowly.
The A CRDM Cooling Fan amber light is lit.
0 CRDM Shroud Inlet Temperature is 97F and stable.
Which of the below listed annunciators is the highest priority, and why?
A. RCP HI RANGE LKG FLOW HI, since a downpower is required.
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B. TDFW PP A SUCTION PRESSURE LO, since a downpower is required.
C. GEN SEAL OIL TO H2 DP LO, since a reactor trip is required.
D. CRDM CLG FAN A AUTO TRIP, since a reactor trip is required.
Proposed Answer: C Explanation (Optional): An annunciator requiring a reactor trip is a higher priority than one requiring a downpower (A and B wrong). A and B are plausible since these annunciators require prompt action. C is correct, and D is wrong, since one tripped CRDM Fan does not require a reactor trip, while Seal oil low DP with hydrogen pressure <30 psig does require a trip. D is plausible, since in the past, loss of CRDM cooling required a trip.
Technical Reference(s): OP 3353.MB3B, 2-10 (Attach if not previously provided)
OP 3353.MB5A73-6 OP 3353.MB7A, 1-5 OP 3353.VP1A, 4-7 Proposed references to be provided to applicants during examination: None Learning MC-04703 Given a plant condition or equipment malfunction relating to the GMO (As available)
Objective: system, determine when the turbine is required to be tripped, or when the generator must be shutdown.
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 55.43.5 c Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO c--
Tier # 1 Question # 76 Group # 1 WA # EPE.038.EK1.02 Importance Rating 3.5 Proposed Question:
Current Plant Conditions:
A SGTR has occurred on the A SG.
0 The crew has entered E-3 Steam generator Tube Rupture.
After cooldown and depressurization, ECCS flow was terminated.
0 Normal charging and letdown have been established.
0 Pressurizer level is 55% and slowly increasing.
0 Ruptured SG narrow range level is 65% and slowly decreasing.
RCS pressure is 1100 psia and slowly decreasing.
Using E-3, step 27 Control RCS Pressure And Charging Flow To Minimize RCS-to-Secondary Leakage attached to the back of this exam, what are all of the required actions to be taken?
A. Turn on the pressurizer heaters.
B. Increase charging flow.
C. Depressurize the RCS AND decrease charging flow.
=__--
D. Maintain RCS and ruptured SG pressures equal.
Proposed Answer: A Explanation (Optional): It is desired to use charging flow to compensate for letdown and coolant shrinkage so that RCS and SG inventories remain constant. SG level provides the most direct indication of primary to secondary leakage, and since SG level is decreasing, backflow is occurring, so RCS pressure will be raised. The break size for this event was such that adequate PZR inventory currently exists, so heaters will be used rather than increased charging to raise RCS pressure (A correct). B, C, and D are plausible since they a11 are actions that may be taken at step 27, depending on conditions.
Technical Reference(s): E-3 Step 27 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: E-3, step 27 Learning Objective: MC-04371 Describe the major action categories within EOP 35 E-3. (As available)
Question Source: Modified Bank #606 17 Parent question attached.
Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.8 and 41.10 Comments:
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ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Original Question #606 I7
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Current Plant Conditions:
A SGTR has occurred After cooldown and depressurization, ECCS flow was terminated Normal charging and letdown have been established Pressurizer level is 61% and slowly increasing Ruptured SG level is 65% and increasing Using E-3, step 27 ControI RCS Pressure And Charging Flow To Minimize RCS-to-Secondary Leakage attached to the back of this exam, what are all of the required actions to be taken?
A. Increase charging flow AND maintain RCS and ruptured SG pressures equal.
B. Turn on the pressurizer heaters.
C . Maintain RCS and ruptured SG pressures equal.
D. Depressurize the RCS AND decrease charging flow.
Answer: D
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29 of 34 NUREG-?021,Revision 9
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ES-401 Written Examination Question Worksheet Examination Outline Cross-reference: Level RO SRQ
-.-: Tier # 1 Question # 77 Group # 1 WA # APE.025.AKl .O1 Importance Rating 4.3 Proposed Question:
INITIAL PLANT CONDITIONS:
Plant is in MODE 5 on the A Train of RHR 0 B Electrical Distribution Train is deenergized for maintenance, and cannot be immediately restored All RCS loops are full 0 RCS temperature is 180°F 0 RCS pressure is 200 psia 0 All Steam Generators are in Wet-Layup No RCPs are running The A RHR Pump fails due to a motor bearing failure.
Which of the following actions are required?
A. Open both PORVs, fill the RCS using one Charging Pump from the RWST.
B. Open both PORVs, fill the RCS using one SI Pump from the RWST.
C. Establish conditions for natural circulation and open the steam generator atmospheric dump valves.
L-D. Start one Reactor Coolant Pump, and open the steam generator atmospheric dump valves.
Proposed Answer: C Explanation (Optional): C is correct, since the RCS is already full and steam generators are available. The procedure has conditions established for natural circulation, RCS pressure is increased to ensure subcooled natural circulation cooling, and the steam generators are used to dump steam. A and B are wrong, since bleed and feed is only used if natural circuIation cooling is unsuccesshl. The charging pump is the preferred feed source, and the SI Pump is the backup source of feed. Dis wrong, since forced cooling is only used if a RCP is already running.
Technical Reference(s): EOP 3505 Attachment B, Steps 9-14 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-04351 Describe major action categories within EOP 3505, Loss of Shutdown (As available)
Objective: Cooling and/or RCS Inventory Question Source: Bank #6429 1 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.8 and 10 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 I Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.L.'
Tier # 1 Question # 78 Group # 1 WA # WE12.GEN.2.1.7 Importance Rating 4.4 Proposed Question:
PLANT CONDITIONS The plant tripped 25 minutes ago due to an inadvertent MSI.
The BOP operator has adjusted AFW flow to a total of 600 gpm to all SGs.
One Safety on each Steam Generator has stuck open.
Narrow Range level is offscaIe low in all Steam Generators All equipment is operating as designed Wide range cold leg temperature is 440°F.
Hot and Cold leg temperatures are decreasing slowly.
The crew has transitioned to the appropriate contingency action procedure.
Based on current plant conditions, what action will the US direct with AFW flow?
A. Do not adjust AFW flow.
B. Establish 100 gpm AFW flow to each Steam Generator.
C. Control flow to each Steam Generator to stabilize RCS temperature.
-w D. Stop AFW flow to all Steam Generators.
Proposed Answer: B Explanation (Optional): The RCS has cooled down by about 110°F in 25 minutes. This requires AFW to be throttled to 100 gpm per SG ("B" correct). "A" is wrong, but plausible, since this is desired if cooIdown rate is 4 0 ° F . "C" is wrong, but plausible, since this is required if hot leg temperatures are increasing. "D" is wrong, since this is not allowed per the CAUTION prior to step 2. " D is plausible, since a cooldown is in progress.
Technical Reference(s): ECA-2.1 step 3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MC-0388 1 Describe the major action categories within EOP 35 ECA-2.1. (As available)
Question Source: Bank #6499 1 Question Cognitive Level: Comprehension or Analysis I O CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-I021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.. Tier # 1 Question # 79 Group # 1 KIA # APE.056.GEN.2.4.7 Importance Rating 3.8 Proposed Question:
Current conditions:
A loss of offsite power event is in progress.
The crew is preparing to perform a cooldown using ES-0.2 Natural Circulation Cooldown.
Only one CRDM fan is available.
How does having only 1 CRDM fan impact the natural circulation cooldown?
A. The crew will go to ES-0.3 Natural Circulation Cooldown With Steam Voidin Vessel (with RVLMS).
B. The crew will maintain the natural circulation cooldown rate LESS THAN 5O0F/hr, and subcooling will be maintained at 52°F.
C. The crew will maintain the natural circulation cooldown rate LESS THAN 80" F h , and subcooling will be maintained greater than 132°F.
D. The crew will align a reactor vessel head vent letdown path and then conduct the cooldown in the same manner as if two CRDM fans are running.
-- Proposed Answer: D Explanation (Optional):
D is correct based on ES-0.2 step 4 and 1 1.a. RNO. The added head-cooling path prevents drawing a void in the head.
A is wrong since ES-0.3 is used when an increased cooldown rate is required (Note prior to ES-0.2, step 1 I).
B is wrong since ES-0.2 requires 82°F subcooling. " B is plausible since 50°F cooldown rate is correct, and the 52" subcooling number is the requirement in ES-0.3.
C is wrong since a 50°F/hr cooldown rate requirement exists whether or not 2 CRDM fans are available, and 132°F subcooling is only a requirement if the head vent path can not be aligned (ES-0.2. step 9 and 11).
Technical Reference(s): ES-0.2 Natural Circ Cooldown steps 4,9, and 1 I (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-05944 Discuss the basis of major procedure steps and/or sequence of steps in (As available)
Objective: EOP 35 ES-0.2, Natural Circulation Cooldown.
Question Source: Bank # 73834 Question Cognitive Level: Comprehension or Analysis I O CFR Part 55 Content: 55.41.10 55.43.5 Comments:
29 of 34 NUREG-I 021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # .1 Question # 80 Group # 1 WA # WE1 I .EA2.2 Importance Rating 4.2 Proposed Question:
With the plant at 100% power, the following sequence of events occurs:
1 . A LOCA outside Containment occurs, resulting in a reactor trip and safety injection.
- 2. Over the next 10 minutes, RCS pressure increases to and cycles at 2350 psia, with PZR PORVs cycling.
- 3. The crew is responding using ECA-1.2 LOCA Outside Containment.
- 4. RWST level is 900,000 gallons and slowly decreasing.
- 5. Pressurizer level is 65% and increasing.
- 6. While attempting to isolate the break, the final valve the crew closes in attempt to isolate the break is RHR pump "A" cold leg injection valve (3SIL*MV8809A).
Which procedure will the crew transition to from ECA-I .2?
A. E-1 Loss ofReactor or Secondaly Coolant.
-.- C . ES-1.3 Transfer to Cold Leg Recirculation.
D. ECA-I. 1 Loss of Emergency Coolant Recirculation.
Proposed Answer: D Explanation (Optional): "D" is correct and "A" wrong, since it can be determined that the break is not isolated with no change in the PORV cycling rate. "A"is plausible since step 5 directs transition to E-1 if pressure is increasing, and pressure can't increase due to the PORVs cycling. The NOTE prior to step 1 must be applied while pressure is cycling on the PORVs. "B" is plausible, since with the break isolated, the crew would transition to E-l first, and then to ES-1 . l .
'IC" is wrong because during a LOCA outside CTMT, a loss of recirculation capability exists.
Technical Reference(s): ECA-1.2, Note prior to step 1, and steps 4 and 5 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-03878 Discuss conditions which require transition to other procedures from (As available)
Objective: EOP 35 ECA-1.2.
Question Source: Modified Bank #75669 Parent attached.
Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Original 75669 With the plant at 100% power, the following sequence of events occurs:
- 1. The reactor trips and safety injection actuates.
- 2. Over the next 10 minutes, RCS pressure decreases to and stabiIizes at 1600 psia.
- 3. The crew is responding using ECA-1.2, "LOCA Outside Containment."
- 4. While attempting to isolate the break, the final valve the crew closes in attempt to isolate the leak is RHR pump "A" cold leg injection valve (3SIL*MV8809A).
- 5. After 3SIL*MV8809A closes, RCS pressure holds steady at 1600 psia.
Which procedure will the crew transition to fi-om ECA-I .2?
A. E-1 "Loss of Reactor or Secondary Coolant" B. ES-I .I "SI Termination" C . ES- 1.2 "Post LOCA Cooldown and Depressurization".
D. ECA-I. 1 "Loss of Emergency Coolant Recirculation".
Answer: D
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29 of 34 NUREG-?021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
- Tier # 1 Question ## 8 1 Group # 1 KIA # APE.027.AA2.04 Importance Rating 4.3 Proposed Question:
With the plant at 100% power, a malfimction in the Pressurizer Master Pressure Controller results in RCS pressure slowly decreasing. Current conditions are as follows:
0 The RO has taken manual control of PZR pressure.
Reactor power: 100%
0 RCS pressure: 2200 psia and stable.
0 Pressurizer level: 64% and stable.
0 RCS temperature: 585°F and stable.
What ACTION, if any, is required?
A. Technical Specifications require no hrther ACTION. All LCOs are satisfied.
B. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> restore applicable primary plant parameters to within limits, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
C. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> restore applicable primary plant parameters to within limits, or reduce THERMAL POWER to less than 5% within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
L-D. Be in at Ieast HOT STANDBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Proposed Answer: C Explanation (Optional): A is wrong, and c is correct, since Tech Spec 3.2.5 DNB Parameters applies. Ais plausible since the RCS pressure DNB setpoint has been recently moved out of Tech Specs, into the COLR. B is wrong, but plausible, since this is the action for Tech Spec 3.4.3.1 Pressurizer level, which is abnormally high, but not out of spec. The limit is 6% of program (which is 61.5%). D is plausible, since this is the action for an inoperable pressurizer.
Technical Reference(s): Tech Spec 3.2.5,3.4.3.1 (Attach if not previously provided)
OP3204 (DNB limits)
Proposed references to be provided to applicants during examination: Tech Spec sections 3/4 (not including section 3/4.3)
Learning MC-03403 given a plant condition requiring the use of the OP-3204 procedure, (As available)
Objective: identify applicable technical specification ACTION requirements.
Question Source: Bank #70267 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.2 and 43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
-1 Tier # 1 Question # 82 Group # 1 WA # APE.062.GEN.2.2.22 Importance Rating 4.1 Proposed Question:
Initial Conditions:
The plant is in MODE 3.
0 Preparations are being made to conduct a reactor startup.
The following sequence of events occurs:
- 1. The SERVICE WTR PUMP DIS PRES LO Annunciator comes in on MBl .
- 2. The Outside Rounds PEO reports a lot of water spraying in the A Service Water cubicle.
- 3. The RO starts 3SWP*PlD.
- 4. The RO stops 3SWP*PlA and 3SWP*PlC and places them in PULL-TO-LOCK.
- 5. The US enters AOP 3560, Loss of Service Water.
What ACTION is required by Technical Specifications?
A. Verify offsite sources are OPERABLE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and remain in MODE 3 indefinitely until the A train of Service Water is returned to service.
B. Verify offsite sources are OPERABLE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, maintain the plant in MODE 3 and restore the A train of Service Water in the next 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
C . Continue with the plant startup and restore the A train of Service Water in the next 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
D. Cool down the plant to COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Proposed Answer: B Explanation (Optional): The crew should enter LCO 3.7.4 due to loss of A Train SWP, and 3.8.1.1 due to loss of cooling to the A EDG. With one train of Service Water INOPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (The plant is already in MODE 3) and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. With one EDG INOPERABLE, verify offsite sources within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. B is correct, and A is wrong, since Offsite sources must be checked per 3.8.1 . I , and the plant can only remain in HOT STANDBY for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with only one train of Service Water. C is wrong, since entry into an operational MODE is not alIowed when an ACTION requires a shutdown. D is wrong, since the plant is allowed to remain in its current MODE for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while initiating repair efforts.
Technical Reference(s): Tech Specs 4.0.4,3.7.4 and 3.8.1.1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: Tech Spec sections 314 (not including section 314.3)
Learning MC-03930 Given a plant condition requiring the use of AOP-3560, identifl the (As available)
Objective: applicable Technical Specificationsand action requirements.
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.2 Comments:
29 of 34 NUREG-I021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
Tier ## 1 Question # 83 Group # 2 WA # APE.OO1.GEN.2.4.30 Importance Rating 3.6 Proposed Question:
With the plant at 98% power, the following sequence of events occurs:
I. Control bank D rods start to rapidly withdraw with no demand signal present.
- 2. The RO fails to rapidly place rods in MANUAL.
- 3. The C-2 rod block comes in, stopping rod motion.
- 4. The STA reports instantaneous calorimetric power indicates 3500 MWth.
- 5. The STA also reports that the 4-minute average calorimetric power indicates 3450 MWth.
6 . The US directs a manual reactor trip.
- 7. The RO trips the reactor, and reports that all rods are on the bottom.
What non-emergency event reportability requirement exists to the NRC?
A. The failure of the rod control system.
B. The instantaneous power level exceeding 102%.
C. The 4-minute average power level exceeding 102%.
-.-- D. The manual reactor trip.
Proposed Answer: D Explanation (Optional): D is correct since a manual reactor trip is reportable if it is not part of a pre-planned sequence (even though it was initiated prior to reaching an automatic reactor trip setpoint). None of the other events are reportable. A is plausible, since the rod control system affects reactivity, and the C-2 control interlock failed. B is plausible, since instantaneous power did exceed 102%. C is plausible, since the 4-minute average power exceeding 102% would be reportable per OP 3204, section 4.3.1 .a.5), but the limit is 3479 MWth.
Technical Reference(s): NUREG 1022, page 47 and 48. (Attach if not previously provided)
RAC 14, page 53 and 55 OP 3204, section 4.3.1 Proposed references to be provided to applicants during examination: None Learning MC-00016 (SRO) Given a plant condition or equipment malfunction, use provided (As Objective: reference material to determine.. . required federal andor state reporting requirements.. . available)
Question Source: New Question Cognitive Level: Memory or Fundamental KnowIedge 10 CFR Part 55 Content: 55.43.5 Comments:
i 29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO i..-
Tier # 1 Question # 84 Group # 2 K/A # APE.003.GEN.2.4. I 1 Importance Rating 3.6 Proposed Question:
With the plant steady at 60% power, BOL, and Control Bank Drods in AUTO at 160 steps, the following sequence of events occurs:
0 A single rod in ControI Bank A drops into the core.
0 The RO promptly places rods in MANUAL.
0 The IT REF/AUCT TAVE DEVIATION annunciator comes in on MB4.
The US enters AOP 3552 Maljirnction of the Rod Drive System.
What action will the US direct the crew to take to minimize the initial Tave/Tref deviation?
A. Initiate dilution.
B. Initiate boration.
C. Manually withdraw rods.
D. Decrease turbine load.
Proposed Answer: D Explanation (Optional): D is correct, since Tave will decrease with the dropped rod. Tave will be less than Tref. C is wrong, since the AOP does not allow rod motion to restore Tave to Tref. C is plausible, since Tave will be less than Tref after the rod drops. After Tav/Tref deviation has been minimized via decreasing turbine load, the crew will borate or dilute as necessary to maintain Tav/Tref (A and B wrong, but plausible).
Technical Reference(s): AOP 3552, step l a through If. (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MC-03901 Describe the major action categories contained within AOP 3552. (As available)
Question Source: Bank #65044 Question Cognitive Level: Comprehension or Analysis IO CFR Part 55 Content: 55.41. I O 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5
-\.: Examination Outline Cross-reference: Level RO SRO Tier # 1 Question # 85 Group # 2 KIA # APE.059.AA2.05 Importance Rating 3.9 Proposed Question:
After commencing a discharge of the A Waste Test Tank (WTT), the Radioactive Liquid Waste EMuent Radiation Monitor 3LWS-RE70 goes into ALARM due to high radiation.
What automatic actions, if any, have occurred, and what actions will the US direct?
A. Liquid Waste to Discharge Tunnel Stop Valve 3LWS-HV77 has automaticallyclosed. The US will direct the Rad Waste PEO to close 3LWS-V54 (3LWS-HV77 inlet isolation valve); and the chemistry department to sample the A WTT and issue a new discharge permit if sample results show activity greater than the original permit. Issue a CR.
B. Liquid Waste to Discharge Tunnel Stop Valve 3LWS-HV77 has automatically closed. The US will direct the Rad Waste PEO to close 3LWS-V54 (3LWS-HV77 inlet isolation valve); and purge 3LWS-RE70 for 5 minutes. If the LWS-RE70 alarm clears during the purge, direct the PEO to open 3LWS-V54 and continue the discharge.
C. No automatic actions have occurred. The US will direct the Rad Waste PEO to locally trip 3LWS-HV77 (Liquid Waste to Discharge Tunnel Stop Valve); and the chemistry department to sample the AWTT and issue a new discharge permit if sample results show activity greater than the original permit. Issue a CR.
.-_ - D. No automatic actions have occurred. The US will direct the Rad Waste PEO to locally trip 3LWS-HV77 (Liquid Waste to Discharge Tunnel Stop Valve); and purge 3LWS-RE70 for 5 minutes. If the LWS-RE70 alarm clears during the purge, direct the PEO to open 3LWS-V54 and continue the discharge.
Proposed Answer: A Explanation (Optional): 3LWS-HV77 tripping closed will automatically terminate the discharge, and the crew will close manual valve 3LWS-V54 (C and D wrong). If sample results show activity similar to the original sample, a 5-minute flush is performed, and the discharge is recommenced (Band D wrong, but plausible). If sample results show higher activity than the original permit, a CR is issued and a new discharge permit is issued (A correct, C plausible).
Technical Reference@): OP 3335D, Attachment 1 (Attach if not previously provided)
AOP 3573 Actions for LWS70-1 Proposed references to be provided to applicants during examination: None Learning MC-04867 Describe the major administrative or procedural precautions an (As available)
Objective: limitations placed on the operation of the LWS system, and the basis for each.
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-I 021, Revision 9
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ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
-u Tier # 1 Question # 86 Group # 2 KIA # WE14.EA2.1 Importance Rating 3.8 Proposed Question:
The following sequence of event occurs:
- 1. A LOCA has occurred.
- 2. The crew is performing actions in E- 1 Loss Of Reactor Or Secondary Coolant.
- 4. The STA reports the following parameters:
Containment pressure is 24 psia and slowly increasing.
Containmentradiation is 100 R/hr and slowly increasing.
Containment sump level is 15 feet and slowly increasing.
What action is the crew required to take?
A. Continue in E-l Loss Of Reactor Or Secondav Coolant.
B. Transition to FR-Z. 1 Response to High Containment Pressure.
C. Transition to FR-2.2 Response to Containment.Flooding.
--- D. Transition to FR-Z.3 Response to High Containment Radiation Level.
Proposed Answer: B Explanation (Optional): CTMT orange paths are from CTMT pressure of 23 psia (Bcorrect, A wrong), or CTMT high sump level 15.75 feet (C wrong). CTMT radiation is above the setpoint of 1OR/hr, but is only a yellow path (D wrong).
Technical Reference(s): CTMT CSF Status Tree (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MC-04666 Identify plant conditions which require entry into EOP35 FR-Z. 1. (As available)
Question Source: Bank #76222 Question History: Millstone 3 2002 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
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ES-40 I Written Examination Question Worksheet Fortn ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
.-.-- Tier # 1 Question # 87 Group # 2 WA # W/E03.A2.02 Importance Rating 4.1 Proposed Question:
A LOCA is in progress, and the crew has entered E-1 Loss OfReactor Or Secondary Coolant. A11 equipment operated as expected. Current conditions are as follows:
The crew has just stopped quench spray pumps per E-1 step 7 Check If Ctmt Spray Should Be Stopped.
RCS pressure is 260 psia and stable.
Containment temperature is 145F RCS subcooling is 150F.
The pressurizer is empty.
RHR pump flow is 200 gpm.
RWST level is 800,000 gallons.
In which procedure will the crew be directed to stop the first SIH pump?
A. E-1 Loss ofprimary or Secondary Coolant.
C. ES-1.2 Post LOCA Cooldown and Depressurization.
-d D. ES-I .4 Transfer To Hot Leg Recirculation.
Proposed Answer: C Explanation (Optional): C is correct, since the crew will transition to ES-1.2 at E-1 step 12 RNO. A is wrong, since E-1 does not direct the stopping of SIH pumps. A is plausible, since it is the procedure the crew is currently performing. B is wrong, since the pressurizer is empty. B is plausible, since the other 3 SI Termination criteria are met. D is wrodg, since the RWST is at 800,000 gallons, and is only being lowered by about 2,000 gpm. It will take about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to reach the RWST switchover setpoint. Dis plausible, since RCS pressure is <300 psia, RHR flow exists, and SIH pumps are stopped in ES-1.4 during switchover to hot leg recirculation.
Technical Reference(s): E-I, steps 6,9, 12, and 13. (Attach if not previously provided)
ES-1.2, step 14.
Proposed references to be provided to appIicants during examination: None Learning Objective: MC-05528 Identify plant conditions that require entry into EOP 35 ES-1.2. (As available)
Question Source: Modified Bank #63999 Parent question attached Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: . 55.43.5 Comments:
29 of 34 NUREG-1021. Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Original question #63999
'-I Due to a Small Break LOCA a plant trip and Safety Injection has occurred.
The following conditions exist:
- All automatic equipment responds as expected
- Containment temperature is 185°F
- Containment High Range Monitors REO4A and O5A read SR/hr
- CTMT Histogram indicates several monitors in ALERT.
- RCS pressure is 1800 psia and stable
- Core exit thermocouple temperature is 540°F
- Pix level is 24% and slowly increasing Assuming conditions do not significantly change, in which of the following procedures would the crew be directed to stop one charging pump?
A. In E-0, Reactor Trip or Safety Injection.
B. In E- 1, Loss of Primary or Secondary Coolant.
C. In ES-I .l, SI Termination.
D. In ES-1.2, Post LOCA Cooldown and Depressurization.
Answer: D 29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO c--- Tier # 2 Question # 88 Group # 1 KIA ## 003.GEN.2.1.23 Importance Rating 4.0 Proposed Question:
CURRENT PLANT CONDITIONS:
The plant had been cooled down to 120°F.
A plant heatup is in progress per OP 320 1 Plant Heatup.
RCS temperature is currently 170°F.
0 All RHR./RCS loop isolation valves are open.
COPPS is ARMED.
Train A of RHR is in service.
The pressurizer is solid.
In accordance with OP 3201 Plant Heatup, how many RCPs are in operation?
A. None B. One C. Two v
D. Four Proposed Answer: B Explanation (Optional): Per OP 3201, steps 4.2.5 and 4.2.6, the first RCP is started when temperature is <160°F (A is wrong but plausible, and B is correct), and per steps 4.3.3 and 4.3.4, the remaining RCPs are started as needed to increase the heatup rate when COPPS has been blocked above 230°F (C and D wrong, but plausible).
Technical Reference(s): OP 3201, steps 4.2.5,4.2.6,4.3.3, and 4.3.4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MC-05430 Describe the operation of the RCPs under ... 4 Loop Operation... (As available)
Question Source: Bank #65950 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Ll Tier # 2 Question # 89 Group # 1 WA # 026.A2.03 Proposed Question: Importance Rating 4.4 The following sequence of events occurs:
T=O: The reactor trips and the crew enters E-0 Reactor Trip or Safely Injection.
T+7 minutes: During the brief at E-0 step 16, the Containment Status Tree turns ORANGE based on CTMT pressure.
T+17 minutes: The crew transitions from E-0 to FR-Z. 1 Response To High CTMT Pressure.
T+17 minutes: The STA reports CTMT pressure is at 37 psia, CDA did NOT actuate, and both Quench Spray pumps are NOT running.
Ti-18 minutes: The RO attempts to manually actuate CDA. CDA will NOT actuate.
T+19 minutes: The RO attempts to start both Quench Spray pumps.
T+I 9 minutes: The RO reports that the B Quench Spray pump did NOT start.
How will current conditions affect the implementation of FR-Z.1 ?
A. The crew will manually start all RSS pumps when the CTMT Recirc Pump AUTO Start Signal annunciator ihminates on Main Board 2.
- 3. RSS Pumps A and B will be placed in PULL TO LOCK, and both sequencers will be placed in TEST 2 with RSS Pumps C and D placed in INHIBIT:
C. A PEO will be dispatched to the ESF Building to align one RSS pump to spray Containment from the RWST.
D. The RSS pumps will be manually started when CTMT wide range sump level is greater than I .5 feet.
Proposed Answer: D Explanation (Optional): D is correct, since with limited QSS flow, an adequate suction source for the RSS pumps must be manually verified prior to allowing the pumps to start (FR-Z. 1, step 7.e. RNO and step 8.a). RSS Pumps normally utilize an 1I-minute timer to prevent starting without adequate sump level, but in this case RSS pumps will not automatically start since CDA could not be actuated (Awrong, but plausible). B is wrong, but plausible; since these actions are taken in ES-1.3, step 3.d. RNO if the A and B RSS Pumps are not available. C is wrong, but plausible, since this action would be taken if no Quench Spray pumps were running (step 7.e. RNO).
Technical Reference(s): FR-Z. 1, Step 7 and 8 (Attach if not previousIy provided)
ES-1.3, step 3.d. RNO Proposed references to be provided to applicants during examination: None Learning Objective: MC-04667 Describe the major action categories with EOP 35 FR-Z.l. (As available)
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 55.43.5 Comments:
29 of 34 NUREG-I 021,Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Question I# 90 Group I# 1 K/A # 039.GEN.2.4.41 Proposed Question: Importance Rating 3.6 Initial Conditions:
0 The plant is being cooled down due to a tube leak in the D SG exceeding technical specification limits.
RCS cold leg temperatures are 400°F.
The following sequence of events occurs:
1 . A large steamline break occurs in the D SG steamline in the Main Steam Valve Building upstream of the MSIVs.
- 2. An RCS Integrity Red Path is received.
Using the EAL tables, what is the classification for this event?
A. Unusual Event - D-1 .
B. Alert - C-I.
C. Site Area Emergency - C-2.
D. General Emergency - A.
Proposed Answer: C Explanation (Optional): C is correct, and A, B , and Dwrong, since the RCS barrier has been potentially lost due the RCS integrity red path (RCBl), and the CTMT barrier has been lost with primary to secondary leakage > Tech Specs and an unisolable secondary release to the environment in progress (CNB4). A is plausible, since 3RMS*RE04A/B have increased unexpectedly (RUI). B is plausible, since a steam break outside CTMT is normally classified as Alert - C-1 (BA2). D is plausible, since 3RMS*RE04A/OSA have increased, indicating a problem with the fie1 clad (FCB3).
Technical Reference(s): MP-26-EPI-FAP06-003 (EAL Tables) (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EAL Tables Learning MC-06064 Determine reportability and classification requirements for a given plant (As available)
Objective: condition or event.
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO 4-Tier # 2 Question ## 91 Group # 1 WA # 064.GEN.2.1.12 Importance Rating 4.0 Proposed Question:
With the Plant at 100% power the following conditions exist:
0 The Emergency Diesel Generator (EDG) Monthly Surveillancewas successfully performed 30 days ago for the A EDG 0 Preparation is being made to perform the Monthly Surveillance for the A EDG Due to a fuel oil leak, the B EDG becomes inoperable. Engineering determines that the potential exists for this to be a common mode failure with the A EDG. Engineering will be contacting the vendor, and estimates the common cause failure possibility will be resolved within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
What action, if any, is required with regard to the A Emergency Diesel Generator?
A. Do not start the AEDG until the B EDG is OPERABLE.
B. Start the A EDG as required by the LCO ACTION requirements. The A EDG does not need to be loaded onto the Emergency Bus.
C. Start the A EDG to satisfy LCO ACTION requirements. Load the A EDG to 4800 to 5000 kW for at least 60 minutes.
\v D. Restore the AEDG to OPERABLE within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Proposed Answer: B Explanation (Optional): LCO ACTION 3.8.1.1 .b applies. This requires Surveillance4.8.1.1.2.a.5 to be performed, which is to start the A EDG (B correct, A wrong). A is plausible, since the A EDG run would not be required if the common cause failure concern could be resolved within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The surveillance requirement is only that the A EDG is started; it does not have to be loaded (Cwrong). C is plausible since loading the EDG is part of surveillance 4.8.1.1.2.a.6. D is wrong, but plausible, since this action is required for two inoperable EDGs, and engineering has reported that the potential exists for a common mode failure.
Technical Reference(s): TS 3.8. I . I .b, Action b (Attach if not previously provided)
TS 4.8.1.1.2.a.5 Proposed references to be provided to applicants during examination: Tech Spec sections 314 (not including section 3/43)
Learning MC-04405 Given a plant condition or equipment malfunction, use provided reference (As Objective: material to.. . Evaluate Technical Specification applicability and determine required actions available)
Question Source: Bank #69329 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55-43.2 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO
..-/ Tier # 2 Question # 92 Group # 2 WA # 002.GEN.2.4.47 Proposed Question: Importance Rating 3.7 The Plant is in Mode 3 with a cooldown in progress in accordance with OP 3208 "Plant CooIdown". Due to a Tech Spec ACTION requirement, the crew is attempting to cooldown at the maximum allowed administrative cooldown rate limit. The following data has been logged over the last hour:
TIME RCS TEMP RCS PRESS I500 550.O"F 2075 psia 1515 53 1.5"F 1700 psia 1530 513.O"F I500 psia 1545 49321°F 1350 psia I600 474.0'F 1125 psia Which of the folIowing actions will be taken at Time 1600?
A. Maintain current cooldown rate, since it is at the administrative limit.
-.._, B. Decrease the cooldown rate, since it exceeds the administrative limit, but not the Tech Spec limit.
C . Stop the cooldown, since it exceeds both the administrative limit and the Tech Spec limit.
D. Increase the cooldown rate, since it is below the administrative limit.
Proposed Answer: B Explanation (Optional):
The cooldown rate is checked both over the last hour and over the last 15 minutes. The administrative limit is 75°F in any one-hour period, or 1.25"F per minute ( I 8.75"F in 15 minutes). The Tech Spec limit is 80"F/hr, or 1.33"F per minute (20°F in 15 minutes). Cooldown over the last hour was 76"F, and AT for current I5 minutes is 1 9S°F in 15 minutes, or 1.30 "F/min. Based on plant conditions the cooldown rate should be decreased to approx. 1.25"F/min.
Technical Reference(s): SP 3601G, step 4.2.3.d & e. (Attach if not previously provided)
Form 3601G.2-1, pages 3 & 4 OP 3208, step 4.2.9.b Proposed references to be provided to applicants during examination: Tech Spec sections 3f4 Learning MC-04837 The crew operates the plant in compliance with all applicable (As available)
Objective: plant procedures and technical specifications Question Source: Modified Bank # 69844 Parent question attached.
Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-401-5 Original Question 6944 I?.* The Plant is in Mode 3 with a cooldown in progress in accordance with OP 3208 Plant CooIdown. Due to a Tech Spec ACTION requirement, the crew is attempting to cooldown at the maximum allowed administrative cooldown rate limit. The following data has been logged over the last hour:
TIME RCS TEMP RCS PRESS 1500 549°F 2075 psia 1515 534°F 1700 psia 1530 517°F 1500 psia 1545 502°F 1375 psia 1600 489°F 1250 psia Which of the following actions should be taken at Time 1600?
A. Maintain current cooldown rate, since.it is at the administrative limit.
B. Decrease the cooldown rate, since it exceeds the administrative limit, but not the Tech Spec limit.
C . Stop the cooldown, since it exceeds both the administrative limit and the Tech Spec limit.
D. Increase the cooldown rate, since it is below the administrative limit.
L, Answer: D 29 of 34 NUREG-I021, Revision 9
1 4.2 Log of Reactor Coolant System Cooldown Rate
/
NOTE Temperature changes during cooldown should be controlled to spread the temperature change in one hour as evenly as possible over the time period.
4.2.1 VERIEY General Prerequisites complete and INITIAL on SP 36016.2-001.
NOTE
- 1. RCS Tc may be monitored using computer points (RCS-T413B7 RCS-T423B, RCS-T433B, and RCS-T443B) or the corresponding MB indications.
- 2. RCS pressure may be monitored using computer points (RCS-P403 or RCS-P405 and when on scale RCS-P455A*, RcS-P456, RCS-P457, or RcS-P458) or the corresponding MB indications.
- 3. A computer printout may be used as a backup record of RCS cooldown, but SP 3601G.2-001 must be utillzed during the cooldown and serve as the primary cooldown record.
4.2.2 At the commencement of cooldown, Refer To SP 3601G.2-001, and PERFORM the following:
- a. RECORD the following:
Date Time Lowest RCS WR cold leg temperature and source (computer ID point or MB channel No.)
RCS pressure and source (computer ID point or MB channel No.)
SP 3601G.2 Rev. 008-03 8 of 17
- b. To determine the maximum allowed pressure, PERFORM the following:
L/
- 1) PERFORM one of the following:
1 Refer To Technical Specification Figure 3.4-3 and DETERMINE the maximum allowable pressure for the recorded RCS temperature SELECT computer point CVMAX, MAX ALLOWABLE PRESSURE, and OBTAIN pressure I 2) RECORD the maximum allowed pressure.
- c. IF RCS pressure is greater than allowed, Refer To T/S 3.4.9.1 and DETERMINE ACTION required.
4.2.3 Refer To SP 36016.2-001, and PERFORM the following every 15 minutes during cooldown:
- a. RECORD the following:
Time c
Lowest RCS WR cold leg temperature and source (computer ID point or MB channel No.)
1 RCS pressure and source (computer ID point or MB channel No.)
- b. To determine the maximum allowed pressure, PERFORM the following:
- 1) PERFORM one of the following:
Refer To Technical Specification Figure 3.4-3 and DETERMINE the maximum allowable pressure for the recorded RCS temperature SELECT computer point CVMAX, MAX ALLOWABLE PRESSURE, and OBTAIN pressure
- 2) RECORD the maximum allowed pressure
- c. IF RCS pressure is greater than allowed, Refer To T/S 3.4.9.1 and DETERMINE ACTION required.
7 1
Level of Use SP 36016.2 Rev. 008-03 9 of 17 I
d, PERFORM the following to determine the current cooldown rate:
i CALCULATE the temperature difference ( O F ) between' current RCS temperature and previous RCS temperature.
CALCULATE the time elapsed (min) between current and previous reading.
C A L C m T E the current cooldown rate ("F/min) by dividing the temperature difference by the time elapsed and RECORD SP 36016.2-001.
-IF existing cooldown rate ("F/min) exceeds limits required for soak time, SUSPEND cooldown for the required soak time.
- e. WHEN four logging intervals (1hour) are completed, PERFORM the following, and every 15 minutes thereafter:
- 1) CALCULATE the temperature difference between the current RCS temperature (Tc) and the previous hour's RCS temperature (T,) and RECORD the current hour's cooldown ( O F ) .
- 2) Refer To SP 36016.2-001, "Acceptance Criteria," and DETERMINE the RCS cooldown limit ( O F ) for the current plant conditions and RECORD on the form.
- 3) IF the hourly cooldown is greater than the cooldown limit, Refer To T/S 3.4.9.1 and DETERMINE ACTION required.
4.2.4 IF a hold point is established during cooldown the target temperature is attained, PERFORM the following:
- a. WHEN two additional logging intervals indicate RCS temperature has stabilized within the desired band, SUSPEND cooldown rate logging.
- b. IF continued cooldown desired, Go To step 4.2.2.
4.2.5 WHEN cooldown completed, SEND SP 3601G.2-001 to the SM for review.
-End of Section 4.2 -
SP 3601G.2 Rev. 008-03 General 10 of 17
RCS Heatup and Cooldown Rate Heatup Soak Time Determination Condition I
RATE
("F/min)
ACTION I < 0.66"F/min No soak required.
Any T, 5 160°F Stop headp.
2 0.66 "F/min Determine soak time.*
When soak complete, restart heatup.
< 1.33"F/min No soak required.
~~ ~~
All T, > 160°F Stop heatup.
2 1.33"F/min Determine soak time.*
I When soak complete, restart heatup.
Cooldown Soak Time Determination Condition RATE ACTION
("F/min)
I .
< 1.33"F/min I No soak required.
Stop heatup.
2 1.33"F/min Determine soak time.*
When soak complete, restart cooldown.
No soak required.
Any T, < 180°F Stop heatup.
2 0.33"F/min Determine soak time.*
When soak complete, restart cooldown
- Soak Time is measured from the time heatup or cooldown is stopped.
Soak Time = Number of "F/min heatup or cooldown rate was exceeded multiplied by (minutes) the time elapsed (minutes) between current and previous reading SP 36016.2-001 Rev. 005 Page 3 of 4
RCS Heatup and Cooldown Rate ACCEPTANCE CRITERIA PARAMETER CONDITION T/S ACCEPTANCE CRITERIA Normal Heatup Within heatup limit line shown RCS Temperature and on T/S Figure 3.4-2 Pressure Within cooldown limit line showi Normal Cooldown on T/S Figure 3.4-3 Any Tc I 160°F S 40°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period RCS Heatup Limit All Tc > 160°F 5 80°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period - e All Tc 2 160°F 5 80°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period ,
Any Tc< 160°Fand RCS pressure between high & low 5 20°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period limit lines on T/S Figure 3.4-3 RCS Cooldown Limit Any Tc < 1 6 0 " F d RCS pressure maintained less than I40°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period the lower limit line on T/S Figure 3.4-3 Performed by: - Date:
Performed by: Date:
Performed by: Date:
Performed by: Date:
Performed by: Date:
Performed by: Date:
Performed by: Date:
Performed by: Date:
SP 36016.2-001 Rev. 005 Page 4 of 4
v CAUTION v
- 1. Exercise extreme care when drawing steam to avoid a rate compensated low steam line pressure SI. Steam line pressure will change more rapidly if fewer than four steam generators are supplying the main steam header.
- 2. RCS temperature must not be reduced below 520°F until main steam line low pressure SI is blocked (P -11).
- 3. If performing cooldown with seals isolated on one or more RCPs due to a loss of all seal cooling, the cooldown rate is limited to 60°F in any one hour period. I 4.2.9 INITIATE RCS cooldown using atmospheric relief valves or relief bypass valves as follows:
L a. E the atmospheric relief valves are available, PERFORM the following:
- 1) PLACE each SG ATMOSPHERIC RELIEF W controller in h4ANUAE:
3MSS-PIC20A1, ATMOSPHERIC RELEF W 3MSS-PIC20B1, ATMOSPHERICRELIEF W 3MSS-PIC2OC1, ATMOSPHERIC RELIEF W 3MSS-PIC20D1, ATMOSPHERIC RELIEF W
- 2) Slowly INCREASE output on each SG ATMOSPHERIC RELIEF W controller, one controller at a time, to establish a uniform cooldown rate not to exceed the administrative limit of 75°F in any one hour period (or 60°F in any one hour period, if applicable).
. Rev.020-10
- b. IF using the atmospheric relief bypass valves is desired, I@
under %ATMOSPHERIC RELIEF BYPASS, slowly throttle OPEN each atmospheric relief bypass valve, one valve at a time, to establish a uniform cooldown rate not to exceed the administrative limit of 75°F in any one hour period (or 60°F in any one hour period, if applicable):
3MSS*MOV74A, SG 1 3MSS*MOV74B, SG 2 3MSS*MOV74CYSG 3 3MSS*MOV74D, SG 4
- c. MAINTAIN all four steam generator pressures approximately equal to ensure uniform RCS cooldown.
- d. E desired, SET a low limit value for computer point alarm CVRH, Hourly RCS HUR/CDR as desired to warn of an approach to the cooldown rate limit (set as a negative value).
I 4.2.10 PERFORM the following activities as necessary during cooldown:
I 4 a.
b.
MAINTAIN steam generator narrow range levels between 45% and 55%.
Refer To OP 3319B, Condensate Makeup and Drawoff System, and PERFORM filling the Condensate Surge Tank.
- c. Refer To OP 3319B, Condensate Makeup and Drawoff System, and PERFORM filling the Condensate Storage Tank
- d. Refer To OP 3322, 4Auxilia~ Feedwater System, and PERFORM one of the following:
Filling the DWST from water treatment vendor Filling the DWST from the Condensate System Filling the DWST from the Domestic Water System Rev. 020- 10
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO i
Tier # 2 Question # 93 Group # 2 KIA # 01 1.GEN.2.1.7 Proposed Question: Importance Rating 4.4 Initial Conditions:
0 A plant cooldown is in progress in accordance with OP 3208 Plant Cooldown.
Both trains of RHR are in service in the cooldown mode.
Pressurizer level is stable at 55%, being maintained by 3RCS-LK459 in AUTO.
RCS cold leg temperatures are 250°F and decreasing.
RCS pressure is 350 psia and stable.
PZR temperature and surge line temperature are both stable at 430°F.
AI1 Pressurizer heaters are energized.
The RO reports that pressurizer surge line temperature has started decreasing, indicating 420°F.
What adverse plant condition exists, and what action will the US direct?
A. Spray flow has initiated with excess AT across the Pressurizer spray nozzle. The US will direct the extra senior licensed operator to notify Engineering and initiate a CR.
B. Spray flow has initiated with excess AT across the Pressurizer spray nozzle. The US will direct the RO to deenergize pressurizer heaters to restore AT to within limits within 30 minutes.
u-C. A pressurizer insurge is in progress. The US will direct the RO to adjust 3RCS-LK459 PZR LVL to decrease charging flow.
D. A pressurizer insurge is in progress. The US will direct the RO to adjust 3RCS-PK131 L/D PRES CNTL to decrease letdown flow.
Proposed Answer: C Explanation (Optional): In this situation, the pressurizer level control system is being used to maintain PZR level constant with spray flow adding water to the PZR at a rate greater than the net charging rate to the RCS as the RCS contracts during the cooldown. This establishes a continuous PZR outsurge, preventing a PZR insurge and the associated thermal transient. If net charging flow increases above the 35 gpm spray flow, an insurge occurs, as evidenced by the surge line temperature drop. The US must either increase letdown flow (D wrong, but plausible) or decrease charging flow (C correct). There is a 182°F temperature difference between the RCS and the PZR, which is within the 200F spray nozzle administrative limit. A lists actions required if the 200°F limit is exceeded, and B lists the actions related to the TRh4 320°F limit. (A and B wrong, but plausible).
Technical Reference(s): OP 3208, steps 4.3.28 and 4.3.29 (Attach if not previously provided)
OP 3208 basis document, steps 4.3.28 and 4.3.29 TRM 3.4.9.2.C Proposed references to be provided to applicants during examination: None Learning MC-03444 Describe the major action categories contained within the OP 3208 (As available)
Objective: procedure.
Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
~ ~
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
'v' Tier # 3 Question # 94 Group # 1 K/A # GEN.2. I .7 Importance Rating 4.4 Proposed Question:
During a storm with rod control in manual, screen DP starts increasing on the "A" Circulating Water Pump. The following sequence of events occurs:
- 1. The "A"Circulating Water Pump trips.
- 3. Rods are manually inserted below RIL to maintain Tave within 5'F of Tref.
- 5. When power reaches 25%, the STA notices that Tave has lowered below the minimum temperature for criticality.
- 6. The RO pulls rods out continuously to raise temperature to within the program band, and above RIL.
- 7. The "B" Circulating Water Pump trips, and condenser backpressure in the "A" condenser bay increases to 6 inches Hg absolute.
- 8. With power at 28%, the US directs the RO to trip the reactor, and enters E-0 Reactor Trip or Safety Injection.
What improper action did the crew take during this event?
A. The US improperly failed to enter AOP 3566 Immediate Boration when rods inserted below RIL.
B. The RO improperIy withdrew control rods continuously to restore temperature and rod height.
C. The crew was required to trip the turbine and enter AOP 3550 Turbine Trip rather than trip the reactor.
D. The crew was NOT required to trip the turbineheactor until condenser backpressure reached 7.5 inches Hg.
Proposed Answer: B Explanation (Optional): This event is based on the Salem marsh grass event in SOER 94-1 Non-Conservative Decisions.
Operators inappropriately pulled rods continuously with an unstable secondary plant, resuIting in a SIS. "A" is wrong, since AOP 3575 provides adequate guidance for immediate boration with rods below RIL. "B"is correct, since operators are required to operate control rods only in a deliberate, carefully controlled manner while closely monitoring the reactor's response (DNAP 1410, section 3.4.4.e). Also, unexpected reactivity changes shall be thoroughly investigated and resolved prior to increasing reactor power (DNAP 1410, section 3.4.6). In this case, the RO is withdrawing rods quite a distance to raise temperature fiom below minimum temperature for criticality up to program band at 30% power. "C" is wrong, since C-9 has been lost, and a reactor trip is required. "D" is wrong, since when power drops below 30%, the turbine trip setpoint drops fiom 7.5" to 5" Hg abs.
Technical Reference(s): DNAP 1410, section 3.4.4.e and 3.4.6. (Attach if not previously provided)
AOP 3559 foldout page.
AOP 3575 Note prior to step 1.
Proposed references to be provided to applicants during examination: None Learning Objective: MC-01925 Demonstrate the ability to make conservative decisions. (As available)
Question Source: Bank #77879 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
-. Tier # 3 Question # 95 Group # 1 WA # GEN.2. I . 12 Importance Rating 4.0 Proposed Question:
Current Plant Conditions:
The plant is critical at 1 O-*amps.
0 The Work Control SRO has just discovere that a weekly surveillance on a piece of equipment coverec Technical Specificationswas last completed 9 days ago.
The equipment has an ACTION STATEMENT allowing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore the equipment to OPERABLE or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
0 The crew will be able to complete the surveillancewithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
How will Technical Specificationsbe applied to the overdue surveillance?
A. The crew needs to conduct the surveillance, but the startup may continue since the overdue surveillance is within its maximum allowable surveillance interval per Surveillance Requirement 4.0.2.
B. The crew will log into the applicable ACTION STATEMENT, but since the surveillance can be completed within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ACTION STATEMENT, the startup can continue into MODE 1.
C. The crew will delay logging into the ACTION STATEMENT for up to twenty-four hours to allow time to perform the surveillance.
-L-D. The crew will continue the startup if the SURVEILLANCE REQUIREMENT portion of the applicable TECHNICAL SPECIFICATION states, The provisions of Specification 4.0.2 are not applicable.
Proposed Answer: C Explanation (Optional): A is wrong, since the maximum allowable surveilIance interval per Requirement 4.0.2 is 25%
o f the surveillance interval (7 days x 1.25 = 8.75 days). B is wrong, since LCO 3.0.4 prohibits entry into an OPERATIONAL MODE when the conditions for the LCO are not met and the associated ACTION requires a shutdown if they are not met. C is correct per SurveillanceRequirement 4.0.3, which allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time of discovery to complete the surveillance. D is wrong, since The provisions of Specification 4.0.2 are not applicable means that the 25% allowable extension is not allowed for this equipment.
Technical Reference(s): Tech Spec 3.0.4 (Attach if not previously provided)
Tech Spec 4.0.2 and 4.0.3 Proposed references to be provided to applicants during examination: Tech Spec sections 3/4 (not including section 314.3)
Learning MC-04788 Given a plant condition or equipment malfunction, use provided (As available)
Objective: reference material to.. . Evaluate Technical Specification applicability.. .
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.2 Comments:
29 of 34 NUREG-1021, Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO
-- Tier # 3 Question # 96 Group # 1 WA # GEN.2.1. I4 Importance Rating 3.3 Proposed Question:
With the plant at 100% power, which of the following changes in system status would require a notification of plant personnel via the plant paging system prior to starting the equipment?
A. The planned starting of a Turbine Building Exhaust Fan.
B. The planned starting of the C Main Condensate Pump.
C . Starting the BScreen Wash Pump due to a clogged strainer on the A pump.
D. Starting the Motor Driven Main Feed Pump due to a trip of the A TDMFP.
Proposed Answer: B Explanation (Optional): E is correct since plant paging is required for the starting of major pumps or fans. This is defined as 6.9KV (condensate pumps), 4.16 KV, or large 480V loads that are controlled fiom the control room. A and C are wrong, since these are specifically identified as pumpslfms not requiring notification of plant personnel.
D is wrong, since a planned start of the MDMFP requires a plant page, but this is an unplanned start. A, C,and D are plausible, since these are large loads.
i--
Technical Reference(s): MP- 14-OPS-GDL200,section 3.16 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MC-05573 Outline the duties and responsibilities of the Unit Supervisor. (As available)
Question Source: Modified Bank #75636 Parent attached.
Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.5 Comments:
Original Bank #75636 Which of the following planned changes in plant status would NOT require a notification of plant personnel via the plant paging system?
A. Starting the B Screen Wash Pump due to a dogged strainer on the A pump.
B. Starting the Motor Driven Main Feed Pump in preparation for raising power above 50%.
C. While pulling rods to criticality, a constant positive startup rate is obtained with no rod motion.
D. Starting the C CAR fan in CTMT while refueIing is in progress.
Answer: A 29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO I
i Tier # 3 Question # 97 Group ## 2 WA # GEN.2.2. I4 Importance Rating 3 .O Proposed Question:
Current pIant conditions:
The plant is at 100% power.
Charging line flow control valve 3CHS*FCV121 has failed open.
Operations department is preparing to operate the plant in an alternate configuration.
0 The lineup will throttle closed a manual valve in line with 3CHS*FCV121, and align 3CHS*HCV190A to provide some flow, allowing main board control of charging demand.
This configuration is not controlled by an approved procedure.
Is a 10CFR50.59 safety evaluation required for this change? Why or why not?
A. Yes, since the system is performing differently from its FSAR description.
B. Yes, since otherwise, the Charging System will have to be decIared INOPERABLE.
C. No, since all components are operating in their design configuration.
D. No, since a CR is all that is required to document a configuration adverse to plant reliability.
??I-.
Proposed Answer: A Explanation (Optional): A is correct, and C is wrong, since this alignment is not a normal mode of operation, and alternate configurations requires RE to perform a 50.59 safety screen. B is wrong, since the Charging System is still capable of performing its safety function. D is wrong, since the CR is required, but would be what drives the 50.59 screening.
Technical Reference(s): MP-ICOPS-REF01 SO2 Section 2.8 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning (As available)
Objective: MC-06335 Identify when a 10CFR50.59 safety evaluation screening is required.
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.3 Comments:
29 of 34 NUREG-1021, Revision 9
ES-401 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 3 Question # 98 Group # 2 KIA # GEN.2.2.26 Importance Rating 3.7 Proposed Question:
Current Conditions:
The plant is in MODE 6.
Fuel movement was suspended for repairs to the Spent Fuel Bridge Crane.
Repairs to the Spent Fuel Bridge Crane are complete.
Westinghouse Source Range Channel N3 1 is out of service for surveillance testing.
Both Gammametrics Channels are OPERABLE.
Both fuel building exhaust filter trains are OPERABLE.
The A Train Fuel Bldg Filter is running, and the B Train is not running.
The he1 building roll-up door is being intermittently opened under administrative control.
The refieling team has established communications with the control room, and has requested permission to move the next fuei bundle from the fuel building to the core.
Are administrative conditions met to recommence fuel movement?
A. No, Source Range Channel N3 1 must be restored to OPERABLE.
B. No, both trains of Fuel Building Filters must be running.
-- C. Yes, but only if the roll-up door is closed during actual fuel movement.
D. Yes, all requirements are met; fuel movement can recommence.
Proposed Answer: D Explanation (Optional): A portion of this question tests new license amendment 219. The new fuel handling accident analysis does not credit fuel-buildingfilters, or fuel building integrity. Band C are wrong, since Iicense amendment 219 deleted Tech Spec 3.9.12, which allowed intermittent fuel building boundary breaches in MODE 6 under administrative control, and required only one train of Fuel Building Filters to be running. A is wrong, since Gammametrics detectors can suffice for Source Range Counts. Since requirements are met, D is correct. A, B, and C are plausible, since each of these conditions are less than optimal.
Technical Reference(s): SP 3672.1, page I O of 13 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: Tech Spec sections 314 (not including section 3/4.3)
Learning MC-06495 Describe the stop work requirements with regards to fuel (As available)
Objective: movement...
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge I O CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021,Revision 9
ES-40 1 Written Examination Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO u Tier # 3 Question # 99 Group # 4 WA# GEN.2.4.36 Importance Rating 2.8 Proposed Question:
A reactor trip has occurred involving a radiation release, an ALERT, C-1 has been declared, and SERO has been activated.
What action is the responsibility of the chemistry technician once he arrives in the control room?
A. Perform the initial dose assessment using IDA.
B. Conduct in plant surveys and sample analysis.
C. Recommend which repair teams will require HP accompaniment.
D. Access OFIS to obtain and provide data to the TIC on the status of the offsite dose release.
Proposed Answer: A Explanation (Optional):
A is correct since this is the responsibility of the Chemistry Technician. B is wrong, since this is the responsibility of RMT1 (HP Tech). C is wrong since this is an OSC responsibiIity. D is wrong, since the CRDC will be accessing OFIS and establishing communication with the TIC.
v Technical Reference(s): MP-26-EPI-FAPOI, section 1.4.4 and Att. 2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning MC-02534, The Shift Manager and Unit Supervisor will perform a11 administrative (As available)
Objective: actions necessary to protect the public in accordance with emergency pian procedures.
Question Source: Bank #76283 Question History: 2002 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-I021, Revision 9
ES-40 I Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO
--- - Tier # 3 Question # 100 Group # 4 WA # GEN.2.4.40 Importance Rating 4.0 Proposed Question:
An ALERT has been declared, and the Shift Manager has assumed the role of the Control Room - Director of Station Emergency Organization (CR DSEO).
Which of the following tasks CAN the CR-DSEO delegate to the Station Duty Officer?
A. Notify the NRC of a 50.54(x) invocation.
B. Authorize off-site notifications.
C. Approve the station evacuation.
D. Issue KI tablets to the control room staff.
Proposed Answer: A Explanation (Optional): A is correct, since the CRDSEO can delegate this responsibility. B , C, and Dare wrong since these tasks cannot be delegated. All distractors are plausible since these are all responsibilities of the CR DSEO.
L Technical Reference(s): MP-26-EPI-FAPO1, Attachment 2, sheet 1 (Attach if not previously provided)
MF-26-EPI-FAPO 1-001, Section F, step 9 Proposed references to be provided to applicants during examination: None Learning EP-00208 List the responsibiIities of the shift manager while serving as CR Director (As available)
Objective: of Station Emergency Operations (CR-DSEO).
Question Source: New Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.43.5 Comments:
29 of 34 NUREG-1021, Revision 9
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Attachment 2 Responsibilities (Sheet 1 of2)
- 1. Control Room Director of Station Emergency Operations (CR-DSEO)
The CR-DSEO is responsible for the following activities, which cannot be delegated, until relieved by the EOF DSEO:
e Assuming command and control of station emergency response e Classifying events e Authorizing off-site notifications e Initiating station emergency response e Authorizing mitigation and repair activities e Approving evacuations e Authorizing emergency exposures e Approving off-site Protective Action Recommendations L
e Issuing KJ I 2. Manager of Control Room Operations (MCRO)
The MCRO is responsible for the following activities:
e Recommending corrective actions to the ADTS e Providing current plant status to the ADTS e Recommending event classifi&tion changes to the ADTS e Coordinating actions to mitigate degradation of plant systems with the ADTS e Coordinating Control Room actions and equipment operability and repair team activities with the MOSC 2/
I MP-26-EPI-FAFO1 Rev. 001-03 12 of 13
Section F: Routine and Follow-up Activities 0 9. E suspension of safeguards and $50.54(x) action is invoked, ensure that the NRC is notified of the departure as soon as possible (but within one hour) using the ENS.
0 10. Direct the RMT #1 to perform control room and plant habitability surveys and sampling.
NOTE The State of CTAocal agencies are responsibIe for issuing KI to offsite @
responders assigned to the station.
0 11. E necessary, issue KI tablets to control room staff in accordance with EPI-FAPO9-003, KITablet Issue Authorization and Tracking Sheet, and log time of issue on SERO 0 Log Sheet (non-delegable).
0 12. Conduct periodic briefings with the control room staff.
0 13. events have been controlled to the point where termination of the emergency can be considered, Refer To EPI-FAP06, Classification and PARS, for guidance.
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MI-26-EPI-FAPO 1-001
- Rev. 001-06 Page 14 of 15