ML23160A077
| ML23160A077 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 05/18/2023 |
| From: | NRC/RGN-III |
| To: | Constellation Energy Generation |
| Gregory Roach | |
| Shared Package | |
| ML22007A049 | List: |
| References | |
| 50-010/23-301, 50-237/23-301, 50/249/23-301 50-010/OL-23, 50-237/OL-23, 50-249/OL-23 | |
| Download: ML23160A077 (1) | |
Text
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 1 of 150 08 March 2023 1
ID: 14655 Points: 1.00 Unit 2 was operating at near rated power when BOTH Recirc Pumps tripped.
Prior to the transient, ACTUAL RPV water level was ___(1)___ than INDICATED Fuel Zone RPV water level.
After the transient, the difference between ACTUAL RPV water level and INDICATED Fuel Zone RPV water level will get ___(2)___.
A.
(1) lower (2) smaller B.
(1) lower (2) larger C.
(1) higher (2) smaller D.
(1) higher (2) larger Answer:
A Answer Explanation Fuel Zone indications are affected by anything that causes flow through the monitored jet pump.
Therefore, Recirc pump flow causes the Fuel Zone instruments to be inaccurate in the non-conservative direction (read higher than actual).
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 2 of 150 08 March 2023 Question 1 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
14655 User-Defined ID:
14655 Cross Reference Number:
Topic:
01 - 295001.A1.07 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE216LN001.06
Reference:
OP-DR-103-102-1002, Operator Aid #250, TSG K/A:
295001.A1.07 3.4 K/A:
Ability to operate and/or monitor the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation:
Nuclear boiler instrumentation system CFR: 41.7 / 45.6 Level: High Safety Function: 1 & 4 Pedigree: Bank History: None Explanation:
A. Correct - Fuel Zone indications are affected by anything that causes flow through the monitored jet pump. Therefore, Recirc pump flow causes the Fuel Zone instruments to be inaccurate in the non-conservative direction (read higher than actual).
B. Incorrect - First part is correct. The difference will get smaller following the transient. Plausible because there are indicators in the plant that the difference between indicated and actual would get larger.
C. Incorrect - Indicated level on the Fuel Zone instruments is higher than actual level due to the forced flow through the monitored jet pump. Plausible because the candidate must determine how the indicators work and interpolate what the indications would be.
D. Incorrect - Indicated level on the Fuel Zone instruments is higher than actual level due to the forced flow through the monitored jet pump. The difference will get smaller following the transient.
Plausible because the candidate must determine how the indicators work and interpolate what the indications would be and there are indicators in the plant that the difference between indicated and actual would get larger.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 3 of 150 08 March 2023 2
ID: 23477 Points: 1.00 DSSP 0100-CR, HOT SHUTDOWN PROCEDURE - CONTROL ROOM EVACUATION, requires verifying MO 2-1301-1, U2 ISOLATION CONDENSER RX OUTLET ISOL VLV, open.
This is done by placing the ISOL COND RX INLET VLV 2-1301-1 selector switch in VLV1 and taking the ISOL COND RX INLET VLVS 2-1301 local control switch to OPEN, inside local panel 2202-76, in the 2/3 D/G room.
Which one of the following describes the reason for placing the ISOL COND RX INLET VLV 2-1301-1 selector switch in VLV1?
A.
To disconnect local control circuits from the valve.
B.
To disconnect Control Room control circuits from the valve.
C.
To isolate wire runs to meet divisional physical separation criteria.
D.
To bypass the Group V Isolations to ensure the valve can be operated.
Answer:
B Answer Explanation Placing the 2-1301-1 valve in the VLV1 position, removes the control circuit from the Main Control Room, in case of fire/evacuation.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 4 of 150 08 March 2023 Question 2 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
23477 User-Defined ID:
23477 Cross Reference Number:
Topic:
02 - 295016.K2.08 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29501LP083
Reference:
DSSP 0100-CR, 12E-2507B K/A:
295016.K2.08 3.9 K/A:
Knowledge of the relationship between Control Room Abandonment and the following systems or components:
Isolation condensers CFR: 41.7 / 45.8 Level: Memory Safety Function: 7 Pedigree: Bank History: None Explanation:
A. Incorrect - With the selector switch in the VLV1 position, this removes the Control Room circuits from the valve. Plausible because if the switch was in the NORMAL position it would remove the local control circuits from the valve.
B. Correct - Placing the 2-1301-1 valve in the VLV1 position, removes the control circuit from the Main Control Room, in case of fire/evacuation.
C. Incorrect - Operating the switch does not change the physical routing or location of equipment. Plausible because division separation is important to the safety of the plant.
D. Incorrect - Plausible because Group V Isolations can be bypassed when directed by DEOPs by utilizing DEOP 500-2 if the Iso Cond would be needed.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 5 of 150 08 March 2023 3
ID: 27483 Points: 1.00 Unit 2 is operating at near rated power.
2B Instrument Air Compressor is OOS.
3C Instrument Air Compressor is lined up to Unit 3.
Annunciator 923-1 F-4, U2 INST AIR PRESS LO, alarmed.
Subsequently, IA Pressure has returned to normal.
What in-plant action must be directed to return Unit Air Systems to normal alignment?
A.
Close the 2-4701-501A, U2 SERV AIR TO INST AIR X-TIE MANUAL ISOL VLV, North of 2A IA dryer.
B.
Verify 2-4608-500, U2 TO U1 SERV AIR XTIE BYP VLV, closed across from AEER Halon cylinders.
C.
Close AO 2-4701-500, U2 SERV AIR TO INST AIR X-TIE VLV, by depressing RESET on control box West of U2 Main IA Receiver.
D.
Close AO 2-4701-500, U2 SERV AIR TO INST AIR X-TIE VLV, by depressing RESET on control box North of 2B Instrument Air Compressor.
Answer:
C Answer Explanation With the U2 Instrument Air Pressure Lo alarm in the U2 service air to instrument air auto x-tie opens. The reset button at the Main Instrument Air receiver must be depressed to return system lineup to normal
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 6 of 150 08 March 2023 Question 3 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
27483 User-Defined ID:
27483 Cross Reference Number:
Topic:
03 - 295019.A1.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE278LN001.08
Reference:
DOA 4700-01; DAN 923-1 F-4 K/A:
295019.A1.02 3.2 K/A:
Ability to operate and/or monitor the following as they apply to Partial or Complete Loss of Instrument Air: System valves CFR: 41.7 / 45.6 Safety Function: 8 Level: High Pedigree: Bank History: 18-1 NRC Explanation A. Incorrect - The U2 Service air to instrument air manual isolation vlv is a normally open valve that would not be repositioned. This is plausible because this would be correct if pressure did not return to normal.
B. Incorrect - Plausible because per DAN 923-1 F-4 U2 Instrument air pressure LO, if U2 SA is crosstied to U1 the the U2 to U1 Service Air Xtie Byp Vlv must be verified closed. But, this is not part of a normal lineup.
C. Correct - With the U2 Instrument Air Pressure Lo alarm in the U2 service air to instrument air auto x-tie opens. The reset button at the Main Instrument Air receiver must be depressed to return system lineup to normal D. Incorrect - the action is correct. The location is incorrect but plausible because the U2 Service air to instrument air xtie vlv is located near the 2A IAC not the 2B IAC.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 7 of 150 08 March 2023 4
ID: 28075 Points: 1.00 Unit 2 is operating at 50% power.
Which of the following changes would result in Control Rod Worth becoming LESS NEGATIVE (absorb less neutrons)?
A.
Lowering void fraction B.
Lowering fuel temperature C.
Rising Xenon concentration D.
Rising moderator temperature Answer:
C Answer Explanation A rise in Xenon concentration results in a smaller number of thermal neutrons available to be absorbed by the control rods. Therefore, Control Rod Worth will become less negative.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 8 of 150 08 March 2023 Question 4 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
2.00 System ID:
28075 User-Defined ID:
28075 Cross Reference Number:
Topic:
04 - 292005.K1.09 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: BR0SIr4_Control_Rods Obj 9
Reference:
Generic Fundamentals BR0SIr4_Control_Rods K/A:
292005.K1.09 2.6 K/A:
Control Rods - Explain direction of change in the magnitude of CRW for a change in moderator temperature, void fraction, control rod density, and xenon CFR: 41.1 PRA: No Level: Memory Safety Function: N/A Pedigree: Bank History: None Explanation:
A. Incorrect-Plausible because changing void fraction will have an effect on CRW, but lowering the void fraction causes CRW to become more negative.
B. Incorrect - Plausible because a change in fuel temperature does have a small effect on CRW, but a drop in temperature will cause CRW to become more negative.
C. Correct - A rise in Xenon concentration results in a smaller number of thermal neutrons available to be absorbed by the control rods.
Therefore, Control Rod Worth will become less negative.
D. Incorrect - Plausible because changing moderator temperature will effect CRW, but rising temperature will make CRW more negative.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 9 of 150 08 March 2023 5
ID: 24256 Points: 1.00 Unit 3 is in Mode 5 with fuel moves in progress. SRM counts begin to steadily go up and continue to go up over a 5 minute period in the quadrant containing the fuel moves.
How are fuel moves affected?
A.
Fuel moves may continue. SRM response is normal for these conditions.
B.
Fuel moves may continue. The grapple may be raised, but NOT lowered.
C.
Stop ALL fuel moves. The grapple may be lowered, but NOT raised.
D.
Stop ALL fuel moves. Do NOT attempt to raise or lower the grapple.
Answer:
D Answer Explanation Per DOA 0800-03, True criticality is indicated by a sustained increase in count rate, over 15 to 20 seconds, of the SRM closest to the Fuel Assembly/Bundle OR Control Rod being moved. The other SRMs may also begin to increase as neutron population increases throughout the core. Immediate actions of DOA 0800-03 require operators to suspend fuel moves and NEITHER raise nor lower the grapple.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 10 of 150 08 March 2023 Question 5 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
24256 User-Defined ID:
24256 Cross Reference Number:
Topic:
05 - 295023.K1.04 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE272LN002.06
Reference:
DOA 0800-03, DFP 0800-01 K/A:
295023K1.04 3.4 K/A:
Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Refueling Accidents: Fuel positioning CFR: 41.8 to 41.10 PRA: Yes Level: High Safety Function: 8 Pedigree: Bank History: 14-1 NRC Explanation:
A. Incorrect - Per DOA 0800-03, inadvertent criticality has occurred and fuel moves must be suspended. Plausible because during fuel moves SRM counts will change.
B. Incorrect - Per DOA 0800-03, inadvertent criticality has occurred and fuel moves must be suspended. plausible because it must be determined that the indications are a True criticality and not caused by instrument noise.
C. Incorrect - Per DOA 0800-03, inadvertent criticality has occurred and fuel moves must be suspended. Plausible because part 1 is correct and part 2 would put the fuel back in the core if it was allowed by procedure.
D. Correct - Per DOA 0800-03, True criticality is indicated by a sustained increase in count rate, over 15 to 20 seconds, of the SRM closest to the Fuel Assembly/Bundle OR Control Rod being moved.
The other SRMs may also begin to increase as neutron population increases throughout the core. Immediate actions of DOA 0800-03 require operators to suspend fuel moves and NEITHER raise nor lower the grapple.
Justification for HIGH order: The candidate must determine inadvertent criticality has occurred and then determine the correct operational implications.
REQUIRED
REFERENCE:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 11 of 150 08 March 2023 6
ID: 28103 Points: 1.00 Unit 2 is operating at near rated power, when Bus 23-1 trips due to an overcurrent condition.
What INITIAL Containment impacts are there with this loss and why?
A.
The Drywell to Torus D/P rises due to Drywell temperature rising.
B.
The Drywell to Torus D/P lowers due to Drywell temperature rising.
C.
The Drywell to Torus D/P rises due to Torus temperature rising.
D.
The Drywell to Torus D/P lowers due to Torus temperature rising.
Answer:
A Answer Explanation The overcurrent on Bus 23-1 causes it to fully de-energize (EDG cannot close onto it). With Bus 23-1 de-energized, Bus 28 becomes de-energized. With Bus 28 de-energized, four of the Drywell Coolers (A, B, F, G) lose power, causing temperature to rise in the Drywell. As temperature rises, a corresponding rise in Drywell pressure will occur and therefore Drywell to Torus D/P will rise.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 12 of 150 08 March 2023 Question 6 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28103 User-Defined ID:
22589 Cross Reference Number:
Topic:
06 - 295012.K1.01 Comments:
Objective: 262LN001.12
Reference:
DANs 902-5 G-5, 902-4 G-17 K/A:
295012.K1.01 4.0 K/A:
Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to High Drywell Temperature: Drywell pressure CFR: 41.8 to 41.10 PRA: No Level: High Safety Function: 5 Pedigree: New History: N/A Explanation:
A. Correct - The overcurrent on Bus 23-1 causes it to fully de-energize (EDG cannot close onto it). With Bus 23-1 de-energized, Bus 28 becomes de-energized. With Bus 28 de-energized, four of the Drywell Coolers (A, B, F, G) lose power, causing temperature to rise in the Drywell. As temperature rises, a corresponding rise in Drywell pressure will occur and therefore Drywell to Torus D/P will rise.
B. Incorrect - Drywell temperature would rise causing Drywell pressure to rise and then Drywell to Torus D/P would RISE. The second part is correct. Plausible because the candidate will have to analyze how the D/P detectors operate and determine how pressure will be affected.
C. Incorrect - The first part is correct. Torus temperature and pressure will not be IMMEDIATELY affected therefore would not affect Drywell to Torus D/P. Plausible because if Torus temperature were to rise then that would affect the Drywell to Torus D/P and the candidate would have to analyze how the D/P detectors operate and determine how pressure will be affected.
D. Incorrect - If Torus temperature were to rise, which would cause Torus pressure to rise that would cause the Drywell to Torus D/P to lower but Torus temperature and pressure will not be IMMEDIATELY affected therefore would not affect Drywell to Torus D/P. Plausible because if Torus temperature were to rise then that would affect the Drywell to Torus D/P and the candidate would have to analyze how the D/P detectors operate and determine how pressure will be affected.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 13 of 150 08 March 2023 7
ID: 28019 Points: 1.00 Unit 3 is operating at 80% power. DOS 2300-10, HIGH PRESSURE COOLANT INJECTION SYSTEM IST COMPREHENSIVE/PRESERVICE PUMP TEST, is in progress.
SBGT is in operation Torus Cooling is running IAW DOP 1500-02, TORUS WATER COOLING MODE OF LOW PRESSURE COOLANT INJECTION SYSTEM, with 3A & 3C LPCI pumps and 3A & 3D CCSW pump operating After approximately 10 minutes of running Torus Bulk Temperature is 93°F and rising at 1°F per minute.
In 13 minutes what action is required?
A.
Secure HPCI B.
Enter DEOP 200-1 and continue the surveillance C.
Increase Torus Cooling flow and continue the surveillance D.
Slow the HPCI turbine to minimum speed until Torus Temperature is <95°F Answer:
A Answer Explanation IAW TS 3.6.2.1 when Suppression pool average temperature >105°F AND thermal power is >1% RTP AND performing testing that adds heat to the suppression pool then the required action is to suspend all testing that adds heat to the suppression pool IMMEDIATELY, therefore securing HPCI is the proper action.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 14 of 150 08 March 2023 Question 7 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28019 User-Defined ID:
28019 Cross Reference Number:
Topic:
07 - 295013.A1.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 223LN001.8
Reference:
DOS 2300-10, DEOP 200-1, TS 3.6.2.1 K/A:
295013.A1.02 4.1 K/A:
Ability to operate and/or monitor the following as they apply to High Suppression Pool Water Temperature: Systems that add heat to the suppression pool CFR: 41.7 / 45.6 PRA: Yes Level: High Safety Function: 5 Pedigree: Bank History: 20-2 NRC Explanation:
A. Correct - IAW TS 3.6.2.1 when Suppression pool average temperature >105°F AND thermal power is >1% RTP AND performing testing that adds heat to the suppression pool then the required action is to suspend all testing that adds heat to the suppression pool IMMEDIATELY, therefore securing HPCI is the proper action.
B. Incorrect - DEOP 200-1 will be entered when Torus Bulk Temperature >95°F but the surveillance will be stopped. Plausible because there have been multiple times when actually performing the surveillance and having to enter DEOP 200-1 while continuing with the surveillance.
C. Incorrect - Max Torus Cooling will be placed in service when Torus Bulk Temperature >95°F as directed by DEOP 200-1 and DOP 1500-02 but the surveillance will be stopped. Plausible because if the math is done incorrectly and it is determined to be <105°F then Max Torus Cooling would be established and the surveillance would continue.
D. Incorrect - HPCI is required to be secured if Torus Bulk Temperature >105°F. Plausible because there are multiple HPCI surveillances that require HPCI to ran at minimum speed for extended periods of time.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 15 of 150 08 March 2023 8
ID: 27433 Points: 1.00 Unit 2 was operating at 100% power.
Annunciator 902-8 D-6, 480V BUS 20, 25 THRU 29 DC POWER FAILURE, is in alarm.
ONLY the control power feed breaker from the Normal 125 VDC bus to Bus 29 has tripped.
Bus 29 and 125VDC Bus 2B-1 remain energized.
Then, a Loss of Coolant Accident occurs.
HPCI has automatically started and is injecting at rated flow Reactor pressure is at 200 psig and lowering Reactor water level is -160 inches and lowering Which LPCI pumps, if any, are injecting into the reactor?
A.
NO LPCI pumps are injecting B.
ONLY the 2A and 2B LPCI pumps are injecting C.
ONLY the 2C and 2D LPCI pumps are injecting D.
ALL LPCI pumps are injecting Answer:
D Answer Explanation 125 VDC control power to Bus 29 is used for all breakers on Bus 29, including the breaker going to MCC 28-7/29-7. Without control power, all of the breakers on those buses will remain in the state they lost power in. When the LOCA occurs, the Unit will scram and the Unit Aux Transformer (UAW) will de-energize. All electrical loads will automatically fast transfer to the Reserve Aux Transformer (RAT),
without any loss of power. The LPCI Injection valves are powered from swing MCC 28-7/29-7, which is normally aligned to Bus 29. When Bus 29 loses control power, the MCC will no longer have control power to open the breaker from Bus 29. Because all LPCI pump buses are energized from the RAT and MCC 28-7/29-7's source bus has power, all LPCI pumps and LPCI injection valves operate correctly. With Reactor pressure less than 325 psig, the injection valves will open and all LPCI pumps will inject.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 16 of 150 08 March 2023 Question 8 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
27433 User-Defined ID:
27433 Cross Reference Number:
Topic:
08 - 203000.K2.03 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE203LN001.08
Reference:
DOP 1500-03 K/A:
203000.K2.03 3.7 K/A:
RHR/LPCI: Injection Mode: Knowledge of electrical power supplies to the following:
Initiation logic.
CFR: 41.7 PRA: Yes Safety Function: 2 & 4 Level: High Pedigree: Bank History: 16-1 NRC, 18-1 NRC Explanation:
A. Incorrect - Plausible if the loss of control power to Bus 29 will cause the breakers on Bus 29 to trip. A failure of Bus 29 to supply MCC 28-7/29-7 would cause the LPCI injection valves to remain shut.
B. Incorrect - Plausible if assumed that 125 VDC power to the LPCI B loop initiation logic was lost. If LPCI B loop logic power were lost, only the A and B LPCI pumps would start. 125 VDC Bus 2B-1 provides power to LPCI B initiation logic. 125 VDC Bus 2B-1 provides control power to Bus 29.
C. Incorrect - Plausible if 125 VDC power to the LPCI A loop initiation logic was lost. If LPCI A loop logic power were lost, only the C and D LPCI pumps would start. 125 VDC Bus 2A-1 provides power to LPCI A initiation logic.
D. Correct - 125 VDC control power to Bus 29 is used for all breakers on Bus 29, including the breaker going to MCC 28-7/29-7. Without control power, all of the breakers on those buses will remain in the state they lost power in. When the LOCA occurs, the Unit will scram and the Unit Aux Transformer (UAT) will de-energize. All electrical loads will automatically fast transfer to the Reserve Aux Transformer (RAT), without any loss of power. The LPCI Injection valves are powered from swing MCC 28-7/29-7, which is normally aligned to Bus 29. When Bus 29 loses control power, the MCC will no longer have control power to open the breaker from Bus 29. Because all LPCI pump buses are energized from the RAT and MCC 28-7/29-7's source bus has power, all LPCI pumps and LPCI injection valves operate correctly. With Reactor pressure less than 325 psig, the injection valves will open and all LPCI pumps will inject.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 17 of 150 08 March 2023 9
ID: 28104 Points: 1.00 Unit 2 is in Mode 4.
RPV level is at +30 inches.
Recirc loop temperatures 200°F and steady.
2A and 2B RBCCW pumps are operating on Unit 2.
MO 2-3704, RBCCW OUTLET VLV, was timed opened for 16 seconds.
2C SDC Heat Exchanger is out of service.
What would be the effect if 2A and 2B Shutdown Cooling (SDC) Pumps are started in the COOLING mode with their discharge valves 60% open?
A.
RPV water level will rise.
B.
Recirculation loop temperature will rise.
C.
SDC will be at the maximum cooling limit for this condition.
D.
BOTH RBCCW pumps will trip on low discharge pressure.
Answer:
C Answer Explanation Per the Limitations and Actions of the DOP 1000-03, to achieve MAXIMUM cooling for this condition, the 2-3704 is timed open for 16 seconds and per the CAUTION prior to step G.1 the MAXIMUM valve position for the SDC pump discharge valves for Unit 2 is 60%.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 18 of 150 08 March 2023 Question 9 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
4 Difficulty:
3.00 System ID:
28104 User-Defined ID:
13663 Cross Reference Number:
Topic:
09 - 205000.K5.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE205001.08
Reference:
DOP 1000-03 K/A:
205000.K5.02 3.5 K/A:
Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Shutdown Cooling System (RHR Shutdown Cooling Mode):
Valve operation CFR: 41.5 / 45.3 Level: Memory Safety Function: 4 Pedigree: Bank History: None Explanation:
A. Incorrect - Plausible because if the 2-3704 was open enough then RPV temp would go up and subsequently RPV level would go up.
B. Incorrect - Plausible because Recirc loop temp would increase if the valve was NOT opened for 16 seconds - indicating less cooling and subsequent increase in RPV temperature.
C. Correct - Per the Limitations and Actions of the DOP 1000-03, to achieve MAXIMUM cooling for this condition, the 2-3704 is timed open for 16 seconds and per the CAUTION prior to step G.1 the MAXIMUM valve position for the SDC pump discharge valves for Unit 2 is 60%.
D. Incorrect - Plausible because there are systems that if the discharge valve is open too much then the pump(s) will runout and potentially trip.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 19 of 150 08 March 2023 10 ID: 13791 Points: 1.00 Unit 2 was operating at near rated power when the following sequence of events occurred:
Time = 0 seconds:
A spurious Group 1 signal occurs Time = 5 seconds:
RPV pressure peaks at 1070 psig Time = 12 seconds:
RPV pressure drops to 1025 psig The Reactor scrammed on ___(1)___ and the Isolation Condenser ___(2)___ initiated to control RPV pressure.
A.
(1) MSIV closure (2) automatically B.
(1) MSIV closure (2) will be manually C.
(1) High RPV pressure (2) automatically D.
(1) High RPV pressure (2) will be manually Answer:
B Answer Explanation Scram caused by MSIV closure BEFORE the high RPV pressure signal was received as a result of the Group 1 isolation. The IC does not initiate until RPV pressure is sustained above setpoint (1047 to 1063) for nominal time of 15 seconds. Pressure reaches 1070 within 5 seconds, then goes below 1047 within the next 7 seconds - therefore no automatic initiation - standby lineup.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 20 of 150 08 March 2023 Question 10 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
13791 User-Defined ID:
13791 Cross Reference Number:
Topic:
10 - 207000.A4.07 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE207LN001.06
Reference:
Electrical Print 12E-2502A, 12E-2506, 12E-2507, 12E-2512, DOP 1300-02, DAN 902-5 D-4, DAN 902-4 A-15 K/A:
207000.A4.07 4.2 K/A:
Isolation (Emergency) Condenser - Ability to manually operate and/or monitor in the control room: System initiation CFR: 41.7 / 45.5 to 45.8 PRA: Yes Level: High Safety Function: 4 Pedigree: Bank History: 06-1 NRC Explanation:
A. Incorrect - First part is correct. Second part is plausible because the Isolation Condenser does have an automatic initiation setpoint.
B. Correct - Scram caused by MSIV closure BEFORE the high RPV pressure signal was received as a result of the Group 1 isolation.
The IC does not initiate until RPV pressure is sustained above setpoint (1047 to 1063) for nominal time of 15 seconds. Pressure reaches 1070 within 5 seconds, then goes below 1047 within the next 7 seconds - therefore no automatic initiation - standby lineup.
C. Incorrect - First part is plausible because High RPV pressure is a scram signal. Second part is plausible because the Isolation Condenser does have an automatic initiation setpoint.
D. Incorrect - First part is plausible because High RPV pressure is a scram signal. Second part is correct.
REQUIRED
REFERENCE:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 21 of 150 08 March 2023 11 ID: 24024 Points: 1.00 Unit 2 was operating at near rated power.
Bus 29 tripped on overcurrent Unit 2 Reactor was scrammed on low RPV water level.
The SRO directs you to inject SBLC as an Alternate Injection System per DOP 1100-02, INJECTION OF SBLC Hard Card.
What would be the expected system response after completion of the hard card actions?
A.
NO injection flow.
B.
~40 gpm flow from the 2A pump ONLY.
C.
~40 gpm flow from the 2B pump ONLY.
D.
~80 gpm flow from BOTH pumps.
Answer:
B Answer Explanation The 2A pump is the only pump with an electrical power supply, and it delivers ~40 gpm.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 22 of 150 08 March 2023 Question 11 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
4 Difficulty:
2.00 System ID:
24024 User-Defined ID:
24024 Cross Reference Number:
Topic:
11 - 211000.K2.01 Num Field 1:
0.00 Num Field 2:
0.00 Text Field:
Comments:
Objective: DRE211LN001.02
Reference:
211000.K2.01 3.6 K/A:
Standby Liquid Control System - Knowledge of electrical power supplies to the following: SLCS pumps.
CFR: 41.7 Level: High PRA: No Safety Function: 1 Pedigree: Bank History: None Explanation:
A. Incorrect - Only the 2B pump would without power and therefore 2A pump is still available to inject. Plausible if the candidate did not understand the power supplies B. Correct - The 2A pump is the only pump with an electrical power supply, and it delivers ~40 gpm.
C. Incorrect - With a loss of Bus 29, MCC 29-1 would be lost. MCC 29-1 is the power supply to both the 2B pump and the 2B squib valve, which would not be able to deliver any flow. Plausible if the candidate did not understand the power supplies D. Incorrect - With a loss of Bus 29, MCC 29-1 would be lost. MCC 29-1 is the power supply to both the 2B pump and the 2B squib valve, which would not be able to deliver any flow. Plausible if the candidate did not understand the power supplies REQUIRED
REFERENCES:
None Question 11 Table-Item Links
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 23 of 150 08 March 2023 12 ID: 14030 Points: 1.00 If the reactor mode switch is in RUN, which one of the following conditions will cause a DIRECT trip of one RPS trip system (i.e., will cause a half scram)?
A.
Reactor power at 10% with MSIVs 1C & 2D less than 90% open.
B.
Reactor power at 10% with Turbine Stop Valves 3 & 4 less than 90% open.
C.
Reactor power at 45% with MSIVs 1A & 1D less than 90% open.
D.
Reactor power at 45% with Turbine Stop Valves 2 & 3 less than 90% open.
Answer:
A Answer Explanation MSIVs C and D do not meet the 5 alive concept, so at this power a 1/2 scram would result.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 24 of 150 08 March 2023 Question 12 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
14030 User-Defined ID:
14030 Cross Reference Number:
Topic:
12 - 212000.K5.02 Num Field 1:
0.00 Num Field 2:
0.00 Text Field:
Comments:
Objective: 212LN001.06
Reference:
12E-2464, 2465, 2466; DAN 902(3)-5 D-14 K/A:
212000.K5.02 4.1 K/A:
Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Reactor Protection System: Logic channel arrangements CFR: 41.5 /45.3 Safety Function: 7 PRA: No Level: High Pedigree: Bank History: 19-1 NRC Explanations:
A. Correct - MSIVs C and D do not meet the 5 alive concept, so at this power a 1/2 scram would result.
B. Incorrect - TSV 3 and 4 do not add up to 5, but at 10% power, Turbine Stop Valves would not cause a 1/2 scram (38.5% bypass).
Plausible because TSV do not add up to 5 (5 alive)
C. Incorrect - MSIVs A and D meet the "5" alive requirement (no half scram). Plausible because candidate must determine for this power level which combinations are OK.
D. Incorrect - TSV 2 and 3 meet the "5" alive requirement (no half scram). Plausible because candidate must determine for this power level which combinations are OK.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 25 of 150 08 March 2023 13 ID: 14674 Points: 1.00 The SRM DRIVE IN push button must be ___(1)___ to drive the SRM detectors into the core, and the SRM DRIVE OUT push button must be ___(2)___ to drive the SRM detectors out of the core.
A.
(1) continually held (2) continually held B.
(1) continually held (2) momentarily depressed C.
(1) momentarily depressed (2) continually held D.
(1) momentarily depressed (2) momentarily depressed Answer:
C Answer Explanation The DRIVE IN switch is a seal in circuit and need NOT be held in. The DRIVE OUT switch is NOT a seal in circuit
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 26 of 150 08 March 2023 Question 13 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
2.00 System ID:
14674 User-Defined ID:
14674 Cross Reference Number:
Topic:
13 - 215004.A4.04 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE215LN004.11
Reference:
DOP 0700-01 K/A:
215004.A4.04 3.2 / 3.2 K/A:
Source Range Monitor System - Ability to manually operate and/or monitor in the control room: SRMS drive control switches CFR: 41.7 Safety Function: 7 PRA: No Level: Memory Pedigree: Bank History: 05-1 NRC, 18-1 NRC Explanation:
A. Incorrect - Since the drive in button has a design contact that locks in, the button does not need to be held in. Plausible because part 2 is correct and part 1 would be correct for drive out.
B. Incorrect - Since the drive in button has a design contact that locks in, the button does not need to be held in. Plausible because part 1 would be correct for drive out and part 2 would be correct for drive out.
C. Correct - The DRIVE IN switch is a seal in circuit and need NOT be held in. The DRIVE OUT switch is NOT a seal in circuit D. Incorrect - The drive out push button does not have a maintain contact feature, so it needs to be held continuously. Plausible because part 1 is correct and part 2 would be correct for drive in.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 27 of 150 08 March 2023 14 ID: 28105 Points: 1.00 Both Units were at rated power when the Feed Breaker to Bus 2A-1 tripped.
What is the status of the ADS system?
A.
DIV 1 ADS Logic is available ADS valve circuitry is powered from the alternate power source B.
DIV 1 ADS Logic is available ADS valve circuitry is powered from the normal power source C.
DIV 1 ADS Logic is unavailable ADS valve circuitry is powered from the alternate power source D.
DIV 1 ADS Logic is unavailable ADS valve circuitry is powered from the normal power source Answer:
C Answer Explanation 2A-1 is the only power source to DIV 1 Logic and the normal source to the ADS valve circuitry.
Therefore, DIV 1 logic is unavailable and the ADS valve circuitry will swap over to 2B-1.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 28 of 150 08 March 2023 Question 14 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28105 User-Defined ID:
28020 Cross Reference Number:
Topic:
14 - 218000.K6.06 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 218LN001.12
Reference:
DOA 6900-T1, DAN 902-3 C-15, 12E-2461 & 12E-2462 K/A:
218000.K6.06 4.0 K/A:
Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Automatic Depressurization System: DC power CFR: 41.7 / 45.7 PRA: No Level: High Safety Function: 3 Pedigree: New History: N/A Explanation:
A. Incorrect - 2A-1 is the only power source to DIV 1 Logic, therefore is unavailable. Plausible because parts 2 is correct and part 1 must identify a single source failure.
B. Incorrect - 2A-1 is the only power source to DIV 1 Logic, therefore is unavailable. ADS valve circuitry is normally powered from 2A-1.
Plausible because must identify normal and alternate power supplies and transfer logic.
C. Correct - 2A-1 is the only power source to DIV 1 Logic and the normal source to the ADS valve circuitry. Therefore, DIV 1 logic is unavailable and the ADS valve circuitry will swap over to 2B-1.
D. Incorrect - 2A-1 is the only power source to DIV 1 Logic, therefore is unavailable. ADS valve circuitry is normally powered from 2A-1 and will automatically switch to its alternate power supply. Plausible because part 1 is correct. Part 2 must recognize the normal and alternate power supply and transfer logic.
REQUIRED
REFERENCE:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 29 of 150 08 March 2023 15 ID: 22386 Points: 1.00 Unit 3 is operating at near rated power, when the following occurs:
Bus 34-1 experiences an overcurrent condition.
A fire in 250 VDC Turbine Building MCC 3 causes the MCC to become de-energized.
What effect does this have on the ESS Bus?
A.
The ESS ABT will transfer power to the ESS Bus, from the Inverter to MCC 38-2, via a Transformer.
B.
The ESS ABT will transfer power to the ESS Bus, from the Static Switch to MCC 38-2, via a Transformer.
C.
The ESS Static Switch will transfer power to the ESS Bus, from the Inverter to Bus 35, via a Voltage Regulator.
D.
The ESS Static Switch will transfer power to the ESS Bus, from the Inverter to Bus 36, via a Voltage Regulator.
Answer:
D Answer Explanation Upon a loss of Bus 34-1 (overcurrent), Bus 39 becomes de-energized. Subsequently with a loss of the TB 250 VDC MCC 3, the Inverter loses power. With no power into the Inverter, the Static Switch will transfer power to the ESS Bus, from the Inverter to Bus 36, via a Voltage Regulator.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 30 of 150 08 March 2023 Question 15 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
2.00 System ID:
22386 User-Defined ID:
22386 Cross Reference Number:
Topic:
15 - 262002.K2.01 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE262LN001.06
Reference:
DAN 902-8 E-8, DOP 6800-01 K/A:
262002.K2.01 3.3 K/A:
Uninterruptable Power Supply (AC/DC) - Knowledge of electrical power supplies to the following: Static switch/inverter CFR: 41.7 PRA: No Level: High Safety Function: 6 Pedigree: Bank History: None Explanation:
A. Incorrect - The ESS ABT will transfer power to the ESS Bus from Bus 38-2 if there is no other power available first. The ESS ABT is downstream of the Static Switch not the Inverter. Plausible because this is a power supply to the ESS Bus.
B. Incorrect - The ESS ABT will transfer power to the ESS Bus from Bus 38-2 if there is no other power available first. Plausible because this is a power supply to the ESS Bus.
C. Incorrect - The ESS Static Switch will transfer power from the Inverter to Bus 36 not Bus 35. Plausible because Bus 25 is a Unit 2 power supply to its ESS Bus.
D. Correct - Upon a loss of Bus 34-1 (overcurrent), Bus 39 becomes de-energized. Subsequently with a loss of the TB 250 VDC MCC 3, the Inverter loses power. With no power into the Inverter, the Static Switch will transfer power to the ESS Bus, from the Inverter to Bus 36, via a Voltage Regulator.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 31 of 150 08 March 2023 16 ID: 24191 Points: 1.00 A locked throttle valve in a safety related system is to be positioned three turns CLOSED FROM FULL OPEN.
Per OP-AA-108-101-1001, COMPONENT POSITION DETERMINATION, which of the following verification techniques is required for the valve's position?
A.
Peer Check B.
Alternate Verification C.
Concurrent Verification D.
Independent Verification Answer:
C Answer Explanation Per OP-AA-108-101-1001 for component position determination, concurrent verification is required for throttle valve position.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 32 of 150 08 March 2023 Question 16 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
2.00 System ID:
24191 User-Defined ID:
24191 Cross Reference Number:
Topic:
16 - Generic 2.1.29 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29900LK051
References:
OP-AA-108-101-1001, HU-AA-101 K/A:
Generic 2.1.29 4.1 / 4.0 K/A:
Knowledge of how to conduct system lineups, such as valves, breakers, or switches CFR: 41.10 / 45.1 / 45.12 PRA: No Level: Memory Pedigree: Bank History: 14-1 NRC Explanation:
A. Incorrect - Peer Check is the act of checking the correct component identification and discussing subsequent component manipulation prior to action being taken and for a locked throttle valve this would not be appropriate. Plausible because peer checking in an approved method of verification.
B. Incorrect - Alternate Verification would be utilizing other indications (ie flow as the valve is operated, visual inspection of valve, etc.) and for a locked throttle valve this would not be appropriate. Plausible because the proper position of some pieces of equipment could be verified using this method.
C. Correct - Per OP-AA-108-101-1001 for component position determination, concurrent verification is required for a locked throttle valve position.
D. Incorrect - Independent Verification would not be used for the throttle valve; once the valve is in the position it is required to be in then it should not be moved again. In order to complete an Independent Verification, the second checker would have to move the valve.
Plausible because IV is an approved method of other valve positions, just not throttle valves.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 33 of 150 08 March 2023 17 ID: 28072 Points: 1.00 Unit 2 is operating at rated power.
High rainfall causes grass and debris to accumulate on the 2/3 Cribhouse bar racks.
Condenser vacuum is degrading.
NPSH to Condensate pumps is lowering.
Which of the following would indicate that cavitation was occurring?
- 1. Fluctuating pump discharge pressure
- 2. Fluctuating pump flow rate
- 3. Steadily rising motor currents
- 4. Excessive pump noise A.
1, 2, and 3 B.
1, 2, and 4 C.
1, 3, and 4 D.
2, 3, and 4 Answer:
B Answer Explanation Degrading condenser vacuum will affect the amount of condensate depression achieved by the condenser. as conditions worsen the eye of the impeller of the condensate pumps will reach the point where cavitation will occur. Indications of cavitation are:
- a. Fluctuating pump discharge pressure
- b. Fluctuating pump flow rate
- c. Fluctuating pump motor current
- d. Excessive pump noise (pump sounds like it is pumping rocks).
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 34 of 150 08 March 2023 Question 17 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28072 User-Defined ID:
28072 Cross Reference Number:
Topic:
17 - 256000.291004.K1.01 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: BC02Sr4 pumps-03
References:
BC02Sr4 Pumps K/A:
256000.291004.K1.01 3.2 K/A:
Condensate - Identification, symptoms, and consequences of cavitation CFR: 41.3 Safety Function: 2 Level: Memory Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because parts 1 and 2 are correct. Motor amps will fluctuate versus steadily increasing.
B. Correct - Degrading condenser vacuum will affect the amount of condensate depression achieved by the condenser. as conditions worsen the eye of the impeller of the condensate pumps will reach the point where cavitation will occur. Indications of cavitation are:
- a. Fluctuating pump discharge pressure
- b. Fluctuating pump flow rate
- c. Fluctuating pump motor current
- d. Excessive pump noise (pump sounds like it is pumping rocks)
C. Incorrect - Plausible because motor amps will rise but also lower (fluctuate). Parts 1 and 3 are correct.
D. Incorrect - Plausible because parts 1 and 3 are correct. Part 2 motor amps will rise but also lower (fluctuate).
Required
Reference:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 35 of 150 08 March 2023 18 ID: 13321 Points: 1.00 A licensed NSO is being administered a JPM at a Unit 2 CRD accumulator as part of their Annual Requal Exam.
A continuous 2 minute siren sounds followed by an announcement directing all personnel NOT having emergency assignments, to report to the CLOSEST assembly area.
To what area must the NSO report?
A.
Main Control Room B.
Operation Support Center (OSC)
C.
Unit 2 Turbine Building Trackway D.
Administration Building Lunchroom/Foyer Area Answer:
C Answer Explanation Per EP-AA-1004, upon hearing a 2 minute continuous siren (EP assembly siren) all personnel not having emergency assignments have been instructed to assemble in pre-designated assembly areas. Refer to figure 4-2. Per figure 4-2, the closest area from the Unit 2 accumulator banks is the Unit 2 turbine building main corridor.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 36 of 150 08 March 2023 Question 18 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
2.00 System ID:
13321 User-Defined ID:
13321 Cross Reference Number:
Topic:
18 - Generic 2.4.39 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29501LP083
Reference:
EP-AA-1004 K/A:
Generic 2.4.39 3.9 K/A:
Knowledge of RO responsibilities in emergency plan implementing procedures CFR: 41.10 Safety Function: N/A Level: Memory Pedigree: Bank History: 05-1 NRC, 18-1 NRC Explanation:
A. Incorrect - Plausible because this would be the correct answer for a NSO if they were on-shift, but not while performing training activities.
B. Incorrect - Plausible because this is where the shift operators go that are not in tech spec required positions C. Correct - Per EP-AA-1004, upon hearing a 2 minute continuous siren (EP assembly siren) all personnel not having emergency assignments have been instructed to assemble in pre-designated assembly areas. Refer to figure 4-2. Per figure 4-2, the closest area from the Unit 2 accumulator banks is the Unit 2 turbine building main corridor.
D. Incorrect - Plausible because this would be correct if the NSO was outside the plant.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 37 of 150 08 March 2023 19 ID: 13812 Points: 1.00 Unit 2 was operating at near rated conditions, when the following occurred:
Unit 2 experienced a Loss of Off-Site Power (LOOP)
Unit 2 experiences a Reactor Scram Annunciator 902-8 E-4 2/3 DG OVERLOAD alarmed Annunciator 902-8 D-4 2/3 DG GROUND FAULT alarmed Annunciator 902-8 E-5 4KV BUS 24-1 OVERCURRENT alarmed 2/3 DIESEL GENERATOR KILOWATT meter reads 2850 Kilowatts What INITIAL action is the NSO required to take?
A.
Dispatch an EO to open 2/3 D/G to Bus 23-1 ACB B.
Trip ALL loads connected to 2/3 EDG. If the fault clears, then close breakers one at a time to locate ground fault to prevent damage to the 2/3 EDG C.
Trip ALL loads connected to 2/3 EDG. If the fault clears, then close breakers one at a time to locate ground fault to prevent damage to the load AFTER Off-Site power restored D.
Trip all UNNECESSARY loads connected to 2/3 DG. If the fault clears, then close breakers one at a time to locate ground fault to prevent damage to the load AFTER Off-Site power restored Answer:
D Answer Explanation With the 2/3 EDG running via an AUTO start signal (LOOP), the actions required are to trip ALL unnecessary loads connected, then close breakers one at a time to locate ground fault to prevent damage to the load when Off-Site power restored.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 38 of 150 08 March 2023 Question 19 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
13812 User-Defined ID:
13812 Cross Reference Number:
Topic:
19 - 295003.K2.03 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE264LN004.10
Reference:
DAN 902-8 D-4, DAN 902-8 E-4 K/A:
295003.K2.03 3.9 K/A:
Knowledge of the relationship between Partial or Complete Loss of AC Power and the following systems or components:
AC electrical distribution system CFR: 41.7/45.8 PRA: Yes Safety Function: 6 Level: High Pedigree: Bank History: None Explanation:
A. Incorrect - No procedural guidance exists to open this breaker locally. Additionally, opening this breaker will result in a loss of ALL AC power. Plausible because is this was a manual start of the EDG then the breaker should have auto tripped.
B. Incorrect - Tripping of all loads connected to 2/3 EDG is incorrect.
Damage to 2/3 EDG is not of concern. Plausible because all UNNECESSARY loads are tripped and then closed in one at a time.
C. Incorrect - Tripping of all loads connected to the 2/3 EDG is incorrect. Damage to equipment upon restoration of off-site power is of concern. Plausible because all UNNECESSARY loads are tripped and then closed in one at a time.
D. Correct - With the 2/3 EDG running via an AUTO start signal (LOOP), the actions required are to trip ALL unnecessary loads connected, then close breakers one at a time to locate ground fault to prevent damage to the load when Off-Site power restored.
Justification for HIGH order: The candidate is forced to evaluate the effects of EDG automatic in addition to actions taken.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 39 of 150 08 March 2023 20 ID: 28123 Points: 1.00 Unit 2 was operating at near rated power when the following ATS power supplies were lost:
MCC 28-1 125 VDC Bus 2A-1 What is the expected plant response, if any?
A.
NO SCRAM signal will be received due to the automatic swap of ATS AC and DC power supplies to the alternate source.
B.
ONLY a HALF SCRAM will occur due to loss of ATS Division 1 power supplies (A1 and A2 RPS Logic)
C.
ONLY a HALF SCRAM will occur due to loss of ATS Division 2 power supplies (B1 and B2 RPS Logic)
D.
A FULL SCRAM will occur due to the loss of ATS Division 1 power supplies (A1 and B1 RPS Logic).
Answer:
D Answer Explanation The loss of 125 VDC 2A-1 and MCC 28-1 will make up the RPS logic for a full scram due to a loss of both DIV 1 and DIV 2 ATS power for one channel of Reactor Water Level Low.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 40 of 150 08 March 2023 Question 20 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
0 Difficulty:
0.00 System ID:
28123 User-Defined ID:
28123 Cross Reference Number:
Topic:
20 - 295004.A1.04 (1)
Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE263LN002.12
Reference:
DAN 902-4 H-20, DAN 902-4 G-20, DOA 6900 -02, DOP 6800-05 K/A:
295004.A1.04 3.7 K/A:
Ability to operate and/or monitor the following as they apply to Partial or Complete Loss of DC Power: DC electrical loads CFR: 41.7/45.6 PRA: Yes Safety Function: 6 Level: High Pedigree: New History: None Explanation:
A. Incorrect - The loss of 125 VDC 2A-1 and MCC 28-1 will make up the RPS logic for a full scram. Plausible because there are many systems that have reserve power that will auto swap to protect from power loss or trips (example, ADS valve power). Additionally, because ATS has both AC and DC power to both divisions, has multiple instruments that are not RPS related, and because the candidate must know what the loss of each power supply would cause.
B. Incorrect - The loss of 125 VDC 2A-1 and MCC 28-1 will make up the RPS logic for a full scram. Plausible because there are systems where the power supply appears swapped (example, RPS Bus vs RPS MG Set, and Medium Range 'A"), and because the candidate must know what the loss of each power supply would cause.
C. Incorrect - The loss of 125 VDC 2A-1 and MCC 28-1 will make up the RPS logic for a full scram. Plausible because there are systems where the power supply appears swapped (example, RPS Bus vs RPS MG Set, and Medium Range 'A"), and because the candidate must know what the loss of each power supply would cause.
D. Correct - The loss of 125 VDC 2A-1 and MCC 28-1 will make up the RPS logic for a full scram due to a loss of both DIV 1 and DIV 2 ATS power for one channel of Reactor Water Level Low.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 41 of 150 08 March 2023 21 ID: 28023 Points: 1.00 A hydraulic ATWS has occurred on Unit 2.
The NSO will ___(1)___.
The MINIMUM electrical safety precautions required to perform this task are ___(2)___.
A.
(1) de-energize scram solenoids (2) safety glasses, long sleeve electrical safety coat, and rubber gloves B.
(1) de-energize scram solenoids (2) all metal removed, safety glasses, long sleeve FR shirt, and rubber gloves C.
(1) install jumpers, perform repeated scrams/resets (2) safety glasses, long sleeve electrical safety coat, and rubber gloves D.
(1) install jumpers, perform repeated scrams/resets (2) all metal removed, safety glasses, long sleeve FR shirt, and rubber gloves Answer:
D Answer Explanation Per DEOP 400 For a hydraulic scram, jumpers need to installed prior to repeated scrams for CRD insertion. Per SA-AA-129 - the voltage in the area being worked in for DEOP 500-5 the minimum PPE is all metal removed, long sleeve arc-rated FR shirt and FR pants or FR coverall, safety glasses and rubber gloves.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 42 of 150 08 March 2023 Question 21 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
2.00 System ID:
28023 User-Defined ID:
28023 Cross Reference Number:
Topic:
21 - 295006 G.2.1.2 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 295L105
Reference:
SA-AA-129, DEOP 400-5, DEOP 500-05 K/A:
295006 G.2.1.2 4.1 / 4.4 K/A:
Knowledge of operator responsibilities during any mode of plant operation: SCRAM CFR: 41.10 / 43.1 / 45.13 PRA: No Level: High Safety Function: N/A Pedigree: New History: N/A Explanation:
A. Incorrect - Need to remove all metal. Plausible because pulling scram solenoid fuses is the correct for an electrical ATWS.
B. Incorrect - Jumpers need to installed. Plausible because pulling scram solenoid fuses is the correct for an electrical ATWS and part 2 is correct.
C. Incorrect - Need to remove all metal. Plausible because part 1 is correct. The PPE is not a correct list.
D. Correct - Per DEOP 400 For a hydraulic scram, jumpers need to installed prior to repeated scrams for CRD insertion. Per SA-AA-129 - the voltage in the area being worked in for DEOP 500-5 the minimum PPE is all metal removed, long sleeve arc-rated FR shirt and FR pants or FR coverall, safety glasses and rubber gloves.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 43 of 150 08 March 2023 22 ID: 28106 Points: 1.00 U3 was operating at 98% rated power when an ERV opened fully and is now stuck full open.
What is a negative effect of this condition ONE MINUTE LATER?
A.
NPSH for the ECCS pumps is lowering B.
ECCS pump motor currents would rise C.
Reactor power is exceeding 100%
D.
Drywell to Torus DP is rising Answer:
A Answer Explanation The open ERV is causing Torus temperature to rise. Rising torus temperature lowers the available NPSH; causing the pumps to come closer to cavitation.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 44 of 150 08 March 2023 Question 22 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
0 Difficulty:
0.00 System ID:
28106 User-Defined ID:
28106 Cross Reference Number:
Topic:
22 - 295026.A2.01-1 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 223LN001.3h
Reference:
295026.A2.01 4.1/4.0 K/A:
Ability to determine and/or interpret the following as they apply to Suppression Pool High Water Temperature: Suppression pool water temperature CFR: 41.10/43.5/45.13 Safety Function: 5 Level: Memory Pedigree: Bank History: None Explanation: Rising temperature lowers the available NPSH - closer to cavitation. Higher temp for torus water would tend to raise pressure in the torus, thus lowering drywell to torus dp, with little or no impact on drywell pressure. As torus temp goes up, water density goes down, therefore, motor amps go down (less work).
A. Correct - The open ERV is causing Torus temperature to rise. Rising torus temperature lowers the available NPSH; causing the pumps to come closer to cavitation.
B. Incorrect - As torus temp goes up, water density goes down, therefore, motor amps go down (less work). Plausible because this would be correct if torus temperature was going down, and because the student has to understand the pump laws, and the impact of rising water temperature on pump amps.
C. Incorrect - One minute later reactor power and the total steam flow (hotel loads, turbine and ERV) on the plant would stabilize at approximately the original value. Plausible because an ERV capacity is 140 MWT, which is approximately 4.7 % reactor power, and because on a pressurized water reactor, an open ERV would cause power to rise.
D. Incorrect - Higher temp for torus water would tend to raise pressure in the torus, thus lowering drywell to torus dP with little or no impact on drywell pressure. Plausible because drywell to torus dP would change, just in the opposite direction.
Required
References:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 45 of 150 08 March 2023 23 ID: 28024 Points: 1.00 An electrical fire has been reported in the 2/3 Emergency Diesel Generator (EDG) room. The automatic fire suppression system actuated as designed. Fire response personnel are planning to enter the room to confirm that the fire is extinguished.
This fire would be classified as Class ___(1)___.
A significant hazard associated with entering the room is the potential for suffocation due to ___(2)___.
A.
(1) A (2) CO2 discharge B.
(1) A (2) Halon C.
(1) C (2) CO2 Discharge D.
(1) C (2) Halon Answer:
C Answer Explanation The fire is electrical in nature, so it is Class C. The EDG room is protected by automatic CO2 suppression, so suffocation is a major hazard to consider prior to entry.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 46 of 150 08 March 2023 Question 23 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28024 User-Defined ID:
28024 Cross Reference Number:
Topic:
23 - 600000.K1.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29501LK080
Reference:
SA-AA-0301 K/A:
600000.K1.02 3.4 K/A:
Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Plant Fire on Site: Firefighting methods for each type of fire CFR: 41.8/41.10/45.3 PRA: No Safety Function: 8 Pedigree: Bank History: None Level: Memory Explanation:
A. Incorrect - Class A pertains to paper/wood fires and this fire was electrical in nature. The second part of the answer is correct.
Plausible because (1) there are four classes of fire (A, B, C and D).
The student must be able to distinguish between the classes. (2) the second part of the answer is correct.
B. Incorrect - Class A pertains to paper/wood fires and this fire was electrical in nature. In addition, the room is NOT protected by halon.
Plausible because (1) there are four classes of fire (A, B, C and D).
The student must be able to distinguish between the classes. (2) the AEER is protected by CO2 and Halon.
C. Correct - The fire is electrical in nature, so it is Class C. The EDG room is protected by automatic CO2 suppression, so suffocation is a major hazard to consider prior to entry.
D. Incorrect - Class C is correct; however, the room is not protected by Halon. Plausible because (1) the first part of the answer is correct.
(2) the AEER is protected by CO2 and Halon.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 47 of 150 08 March 2023 24 ID: 28025 Points: 1.00 Unit 2 is operating at full power when the following occurs:
Steam line rupture has occurred in the HPCI room and a Group IV Isolation does not initiate.
HPCI pump room temperature is 155°F.
HPCI pump room radiation level is 100 mr/hr.
Per OP-DR-103-102-1002, STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION, which of the following room coolers must be started in addition to U2 HPCI room cooler?
- 1. Unit 2 LPCI/CS room cooler
- 2. Unit 3 LPCI/CS room cooler
- 3. Unit 3 HPCI room cooler A.
3 ONLY B.
1 and 2 ONLY C.
1 and 3 ONLY D.
1, 2, and 3 Answer:
D Answer Explanation With the HPCI room temperature exceeding Max normal of 150 degrees, DEOP 300-1, SECONDARY CONTAINMENT CONTROL, entry is required. Room coolers are to be started. Per OP-DR-103-102-1002, STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION, when operating room coolers, start all available LPCI and HPCI room coolers on both units.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 48 of 150 08 March 2023 Question 24 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28025 User-Defined ID:
28025 Cross Reference Number:
Topic:
24 - 295032.K2.01 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29502LK052
References:
DEOP 300-1 K/A:
295032.K3.03 3.5 K/A:
Knowledge of the relationship between High Secondary Containment Area Temperature and the following systems or components: Area/room coolers CFR: 41.7/45.8 PRA: No Safety Function: 9 Level: Memory Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because the U3 HPCI should be started, but LPCI room coolers must be started as well.
B. Incorrect - Plausible because both units LPCI room coolers should be started but U3 HPCI room cooler must be started as well even though it has an auto start on high room temperature.
C. Incorrect - Plausible because both U2 LPCI and U3 HPCI room coolers must be started, but U3 LPCI room cooler must be started as well.
D. Correct - With the HPCI room temperature exceeding Max normal of 150°F, DEOP 300-1, SECONDARY CONTAINMENT CONTROL, entry is required. Room coolers are to be started. Per OP-DR-103-102-1002, STRATEGIES FOR SUCCESSFUL TRANSIENT MITIGATION, when operating room coolers, start all available LPCI and HPCI room coolers on both units.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 49 of 150 08 March 2023 25 ID: 14645 Points: 1.00 The following plant conditions exist after a transient with both units at power:
There is 1 inch of water on the Unit 2 HPCI Room floor - Water level is steady.
The Unit 2 West Corner room floor is covered in water (< 1 inch and level is steady).
Why are the sump pumps operated under these conditions per DEOP 300-1, SECONDARY CONTAINMENT CONTROL?
A.
To maintain equipment operability.
B.
To maintain site release rates below 10 CFR 100 limits.
C.
To quantify the leakage rate to determine Tech Spec required actions.
D.
To ensure environmental conditions are maintained for EQ Instrumentation.
Answer:
A Answer Explanation The basis for pumping the sumps on high water level in Secondary Containment is to maintain operability of equipment in the area, and to maintain the areas in a condition permitting safe entry by personnel. The normal water level in the corner rooms is 'none', i.e., dry floors. The Max Safe level is 8 inches. Water level above Max Safe will jeopardize equipment and prevent personnel entry as electrical conduit and junction boxes will be submerged.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 50 of 150 08 March 2023 Question 25 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
14645 User-Defined ID:
14645 Cross Reference Number:
Topic:
25 - 295036 G.2.4.22 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29502LK052
Reference:
EOP-DEOP-SAMG TB, DAN 902-4 C-19, DOA 0040-02, DEOP 300-1 K/A:
295036 G.2.4.22 4.0/4.7 K/A:
Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations: Secondary Containment High Sump/Area Water Level CFR: 41.7/41.10/43.5/45.12 PRA: No Safety Function: 5 Level: Memory Pedigree: Bank History: None Explanation:
A. Correct - The basis for pumping the sumps on high water level in Secondary Containment is to maintain operability of equipment in the area, and to maintain the areas in a condition permitting safe entry by personnel. The normal water level in the corner rooms is
'none', i.e., dry floors. The Max Safe level is 8 inches. Water level above Max Safe will jeopardize equipment and prevent personnel entry as electrical conduit and junction boxes will be submerged.
B. Incorrect - Plausible because pumping the sumps will control spread of contamination, it is not the DEOP 300-1 basis.
C. Plausible because the sumps are being pumped and DW sumps are able to quantify volume. The leakage cannot be quantified via the sumps, D. Plausible because of the EQ instruments located in the corner rooms, but water level does not put the EQ instruments at risk (those that need to operate in high humidity are leak tight).
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 51 of 150 08 March 2023 26 ID: 28122 Points: 1.00 Unit 2 is operating at approximately 86% power:
Total core flow is 51 Mlbm/hr Total Recirc flow is 47% of rated drive flow Individual APRM readings are as follows:
APRM 1: 85%
APRM 2: 84%
APRM 3: 85%
APRM 4: 85%
APRM 5: 89%
APRM 6: 86%
What is the response, if any, to the above conditions?
(Reference provided)
A.
NO trip signals present B.
Rod Block signal ONLY C.
Rod Block and HALF Scram signals present D.
Rod Block and FULL Scram signals present Answer:
B Answer Explanation For the given conditions, the Flow Biased setpoints are lower than the fixed rod block and SCRAM setpoints. They are listed below:
Flow Biased Rod Block Setpoints:
TS: 56W+55.4 Actual: 56W+54 Flow Biased SCRAM Setpoints:
TS: 56W+67.4 Actual: 56W+66 Per the given conditions, "W" = 47; thus a rod block is present for APRM 4 (.56(47)+54 = 80.3%).
Conditions are below the scram setpoint (.56(47)+66% = 92.3%). "W" is defined in the COLR, Section 6 as the "% of drive flow required to produce a rated core flow of 98 Mlb/hr. The highest power reading on RPA channel A is 85%. This is greater than the Rod Block setpoint of 80.3%, but less than the SCRAM setpoint of 92.3%.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 52 of 150 08 March 2023 Question 26 Info Question Type:
Multiple Choice Status:
Active Time to Complete: 3 Difficulty:
3.00 System ID:
28122 User-Defined ID:
28035 Topic:
26 - 215005.A3.03 Comments:
Objective: 215LN005.06
Reference:
Tech Spec 3.3.1.1, TRM 3.3.a, DAN 902(3)-5 A-7 K/A:
215005.A3.03 3.6 K/A:
Ability to monitor automatic operation of the Average Power Range Monitor/Local Power Range Monitor System, including: Meters and recorders.
Safety Function: 7 CFR: 41.7 Level: High Pedigree: Bank History: None Explanations:
A. Incorrect - A rod block would occur based on the conditions listed in the stem. Plausibility:
Plausible because the answer would be correct if APRMs were indicating slightly lower (below 80.3%), and because the student may miscalculate the trips values, or use the fixed trip setpoints (which are much higher).
B. Correct - For the given conditions, the Flow Biased setpoints are lower than the fixed rod block and SCRAM setpoints. They are listed below:
Flow Biased Rod Block Setpoints:
TS: 56W+55.4 Actual: 56W+54 Flow Biased SCRAM Setpoints:
TS: 56W+67.4 Actual: 56W+66 Per the given conditions, "W" = 47; thus a rod block is present for APRM 4 (.56(47)+54 =
80.3%). Conditions are below the scram setpoint (.56(47)+66% = 92.3%). "W" is defined in the COLR, Section 6 as the "% of drive flow required to produce a rated core flow of 98 Mlb/hr. The highest power reading on RPA channel A is 85%. This is greater than the Rod Block setpoint of 80.3%, but less than the SCRAM setpoint of 92.3%.
C. Incorrect - A half scram would not occur because none of the APRM 4 - 6 readings are above the SCRAM setpoint. Plausibility: This is plausible because the B RPS channel APRMs are higher than the A RPS channel, and because the values are actually above the SCRAM setpoint for single loop flow biased trips. (0.56(47) + 61 = 87.3%).
D. Incorrect - A full SCRAM would not occur, since ALL APRMs are below the flow biased setpoints. Plausibility: Plausible a Rod Block will occur, and a SCRAM would occur if power were slightly higher. Additionally, if the student does not properly calculate the setpoints, then both a rod block and SCRAM would be expected to occur.
REQUIRED
REFERENCES:
Tech Spec 3.3.1.1, TRM 3.3.a with 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less blanked out None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 53 of 150 08 March 2023 27 ID: 28036 Points: 1.00 Both units are at rated power.
Both Reactor Building Ventilation Systems are isolated for isolation damper repairs B SBGT train is in START A SBGT train is in STANDBY Unit 3 experiences a sustained total loss of instrument air All equipment operated as designed What is the preferred action regarding the SBGT system?
A.
Start 2/3A Train of SBGT AND secure 2/3B Train B.
Take manual control of B SBGT flow control damper C.
Verify the backup air supply to the SBGT system opens D.
Verify automatic start of A SBGT train AND trip of B SBGT train Answer:
A Answer Explanation SBGT is required to be running due to Reactor Building Ventilation being off. DOA 4700-01 states that on a loss of air, it is preferred to operate the unaffected SBGT train since the flow control damper will fail open on the affected SBGT train which may cause it to exceed its design flow.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 54 of 150 08 March 2023 Question 27 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28036 User-Defined ID:
28036 Cross Reference Number:
Topic:
27 - 261000 K1.10 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 261LN001-8
References:
DOA 4700-01 K/A:
261000 K1.10 3.0 K/A:
Knowledge of the physical connections and/or cause and effect relationships between the Standby Gas Treatment System and the following systems: Instrument air system.
CFR: 41.4 to 41.9/45.7to 45.8 Safety Function: 9 Level: High Pedigree: New History: N/A Explanation:
A. Correct - SBGT is required to be running due to Reactor Building Ventilation being off. DOA 4700-01 states that on a loss of air, it is preferred to operate the unaffected SBGT train since the flow control damper will fail open on the affected SBGT train which may cause it to exceed its design flow.
B. Incorrect - The SBGT flow control damper does NOT have manual throttle capability. Plausible because some dampers can be controlled manually.
C. Incorrect - The SBGT system does NOT have a backup air supply.
Plausible because some valves and dampers in the plant have backup air supply or accumulators to operate equipment on loss of air.
D. Incorrect - A SBGT train will NOT auto start since B SBGT will NOT have low flow or lose its heater. Plausible because if other dampers were affected, it would cause a loss of system flow which would provide an auto start signal to 2/3A.
Required
Reference:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 55 of 150 08 March 2023 28 ID: 28026 Points: 1.00 Operations is performing a LIVE bus transfer between Bus 24-1 and 34-1 and is being conducted in accordance with DOP 6500-08, BUS 24-1 TO BUS 34-1 TIE BREAKER OPERATION, with the following conditions:
The first cross tie breaker has been closed The synch selector switch has been placed to ON for the second cross tie breaker The voltage difference between the busses is 25 Volts The Synch Meter is within 3 degrees of vertical (357 degrees)
Under these conditions, should the second crosstie breaker be closed, and why?
A.
No, because the voltage difference is too high for cross tie operation.
B.
No, because the two buses are too far out of phase for cross tie operation.
C.
Yes, because the voltage and phase limits are met for cross tie operation.
D.
Yes, because the phase limit is met and the voltage limit is NOT applicable when closing the second cross-tie breaker.
Answer:
C Answer Explanation Per the limitations and actions of the DOP, the limits for cross tie operation is within 5 degrees of vertical on the Synch meter and the voltage difference is <50 volts.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 56 of 150 08 March 2023 Question 28 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28026 User-Defined ID:
28026 Cross Reference Number:
Topic:
28 - 262001.A4.06 Num Field 1:
0.00 Num Field 2:
0.00 Text Field:
Comments:
Objective: 26204LK009
Reference:
DOP 6500-08 K/A:
262001.A4.06 3.2 K/A:
Ability to manually operate and/or monitor in the control room:
Instrument switches.
Safety Function: 6 CFR: 41.7/45.5 to 45.8 Level: Memory Pedigree: Bank History: None Explanation:
A. Incorrect - Plausible because The voltage difference between the emergency buses that are crosstied must be < 50 volts (preferably closer to zero volts). At 25 volts the limit is met but not close to zero.
B. Incorrect - Plausible because zero degrees out of phase is preferred, 3 degrees is within the tolerance of 5 degrees of the vertical position.
C. Correct - Per the limitations and actions of the DOP, the limits for cross tie operation is within 5 degrees of vertical on the Synch meter and the voltage difference is <50 volts.
D. Incorrect - Plausible because the phase limit is met, the voltage limit is applicable even though it is in spec for this evolution.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 57 of 150 08 March 2023 29 ID: 28107 Points: 1.00 Unit 2 was operating at near rated power when a complete loss of U2 125 VDC Division 2 occurred.
The Control Room control switch indication lights will be lost for.....
A.
2A LPCI pump B.
2C LPCI pump C.
2A SDC pump D.
2C SDC pump Answer:
B Answer Explanation With a loss of 125 Vdc Division 2, pumps powered from Bus 24-1 will lose remote indication and protection ability. 2C LPCI is the only pump powered from Bus 24-1.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 58 of 150 08 March 2023 Question 29 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
3.00 System ID:
28107 User-Defined ID:
28107 Cross Reference Number:
Topic:
29 - 263000.A1.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE263LN002.12
Reference:
DOA 6900-02, DOA 6900-T1 K/A:
263000.A1.02 3.3 K/A:
Ability to predict and/or monitor changes in parameters associated with operation of the DC Electrical Distribution, including: Lights and alarms CFR: CFR: 41.5 / 45.5 Safety Function: 6 Level: Memory Pedigree: New History: N/A Explanation:
A. Incorrect - 2A LPCI pump is powered from Bus 23-1 and its control power is from Div 1 125 VDC. Plausible because the reserve control power is supplied from U2 125VDC Div 1.
B. Correct - With a loss of 125 Vdc Division 2, pumps powered from Bus 24-1 will lose remote indication and protection ability. 2C LPCI is the only pump powered from Bus 24-1.
C. Incorrect - 2A SDC pump is powered from Bus 23-1 and its control power is from Div 1 125 VDC. Plausible because the reserve control power is supplied from U2 125VDC Div 1.
D. Incorrect - 2C SDC pump is powered from Bus 23-1 and its control power is from Div 1 125 VDC. Plausible because the reserve control power is supplied from U2 125VDC Div 1 and this is also a difference between units. 3C SDC is powered from Bus 34-1 (Div 2) and would be affected if this situation was on Unit 3.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 59 of 150 08 March 2023 30 ID: 28028 Points: 1.00 Unit 2 is operating at 100% power.
The Instrument Air system is lost.
What is the system impact for the feedwater heating system?
A.
Extraction Steam Bypass AOs fail open B.
Extraction Steam Bypass AOs fail as is C.
Extraction Steam Bypass AOs fail closed D.
Heater Emergency Drain AOs fails closed Answer:
A Answer Explanation Per DOA 4700-01, INSTRUMENT AIR SYSTEM FAILURE, on a loss of instrument air or instrument bus, extraction steam non-return valves would close AND the extraction bypass valves would open.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 60 of 150 08 March 2023 Question 30 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
2.00 System ID:
28028 User-Defined ID:
28028 Cross Reference Number:
Topic:
30 - 300000 K3.11 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE260LN001.06
Reference:
DOP 6800-02, DOA 3500-02, M-14, DOA 4700-01 K/A:
239001.K3.11 3.0 K/A:
Knowledge of the effect that a loss or malfunction of the Instrument Air System will have on the following systems or system parameters: Extraction steam system.
CFR: 41.7 Safety Function: 3 Level: Memory Pedigree: Bank History: 18-1 NRC Explanation:
A. Correct - Per DOA 4700-01, INSTRUMENT AIR SYSTEM FAILURE, on a loss of instrument air or instrument bus, extraction steam non-return valves would close AND the extraction bypass valves would open.
B. Incorrect - When power is lost to the solenoids the extraction bypass valves will fail to the open position to prevent a turbine trip due to high feedwater heater level.
Plausibility: Failed "as is" is plausible because Dresden has a number of AOV's that fail as is on a loss of Instrument Air or Instrument bus power to the solenoids. An example is the Off gas chimney isolation valves as well as the SJAE suction vlvs.
C. Incorrect - The extraction bypass AOVs fail open.
Plausibility: This is plausible because there are a number of air operated valves that fail closed when air is taken away when power is lost to the solenoids.
D. Incorrect - Opens on loss of power to the solenoids supplied by instrument bus.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 61 of 150 08 March 2023 31 ID: 28029 Points: 1.00 Both Units were operating at 100% power.
Fire header pressure has reached 91 psig and is dropping at a rate of 3 psig/min.
If the trend remains constant and no operator action is taken, the EARLIEST time at which the U2/3 DFP will have auto started is....
A.
4 minutes later B.
5 minutes later C.
6 minutes later D.
7 minutes later Answer:
A Answer Explanation At the 4 minute point, fire main pressure would be 79 psig, The U2/3 DFP will start at 80 psig.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 62 of 150 08 March 2023 Question 31 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
2.00 System ID:
28029 User-Defined ID:
28029 Cross Reference Number:
Topic:
31 - 510000.K3.09 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE286LN001.11
Reference:
DOA 3900-01, DAN XL3 82-30, DAN 901-2 H-8, DAN 923-1 G-4 K/A:
510000 K3.09 2.7 K/A:
Knowledge of the effect that a loss or malfunction of the Service Water System will have on the following systems or system parameters: Fire protection system.
CFR: 41.7/45.4 Safety Function: 4 PRA: No Pedigree: New Level: Memory History: N/A Explanations:
A. Correct - At the 4 minute point, fire main pressure would be 79 psig, The U2/3 DFP will start at 80 psig.
B. Incorrect - At the 5 minute point, fire main pressure would be 76 psig.
Plausible because the U2/3 DFP will already be started the U1 DFP starts at 75 psig.
C. Incorrect - The Unit 1 DFP will auto start at 75 psig. This has been met. Plausible because this is the EARLIEST time for U1 DFP to start.
D. Incorrect - At the 7 minute point, fire main pressure would be 70 psig. The U2/3 and U1 DFP's will have started earlier. Plausible because at this time the U1 screen wash pumps will also have started at 70 psig.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 63 of 150 08 March 2023 32 ID: 28030 Points: 1.00 Unit 3 was operating at near rated power, when the following occurred:
An automatic scram occurred, with an ATWS resulting Reactor Power is currently 42%
Drywell pressure is 1.30 psig and rising RPV water level was terminated and prevented to -55 inches per DEOP 400-5 Maximum Torus Cooling has been established Then Drywell pressure exceeds 2.0 psig.
(1) What effect, if any, does this have on Torus Cooling?
(2) What action, if any, must be done to re-establish Max Torus Cooling?
A.
(1) None, because the HX BYPASS VLVs will be interlocked closed (2) No manipulations are required B.
(1) None, until RPV pressure drops below 350 psig, at which time the HX BYPASS VLVs will open (2) The HX BYPASS VLVs are required to be re-closed after they have opened C.
(1) Cooling will be reduced, because the HX BYPASS VLVs will open and be interlocked open for 30 seconds (2) The HX BYPASS VLVs are required to be re-closed after interlock has timed out D.
(1) Cooling will be reduced, because the HX BYPASS VLVs will open and be interlocked open until RPV pressure drops below 350 psig (2) The HX BYPASS VLVs are required to be re-closed after they have opened Answer:
C Answer Explanation The HX BYPASS VLVs were closed to establish Max Torus Cooling - there was no initiation signal present. When DW pressure exceeds the ECCS initiation setpoint, the valves receive an open signal, and are interlocked open for 30 seconds. After interlock times out, the valves are re-closed via the control switch to re-establish max torus cooling.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 64 of 150 08 March 2023 Question 32 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28030 User-Defined ID:
28030 Cross Reference Number:
Topic:
32 - 219000.A3.01 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE203LN001.03
Reference:
DOP 1500-02 attach C, DOP 1500-03 K/A:
219000.A3.01 3.9 K/A:
Ability to monitor automatic operation of the RHR/LPCI:
Torus/Suppression pool Cooling Mode, including: Valve operation.
CFR: 41.7/45.7 Safety Function: 5 Level : High Pedigree: Bank History: 2007 NRC Explanation:
A. Incorrect - The HX BYPASS valves were manually closed to establish max torus cooling. Plausible because without a valid initiation signal the valves would remain closed.
B. Incorrect - The actions will take place at +2 psig drywell not 350 psig RPV pressure. Drywell pressure initiation logic does not use RPV pressure (used by low-low level initiation logic). Plausible because the actions are correct for a 2# signal.
C. Correct - The HX BYPASS VLVs were closed to establish Max Torus Cooling - there was no initiation signal present. When DW pressure exceeds the ECCS initiation setpoint, the valves receive an open signal, and are interlocked open for 30 seconds. After interlock times out, the valves are re-closed via the control switch to re-establish max torus cooling.
D. Incorrect - Plausible the HX BYPASS valves will open and be interlocked closed for 30 seconds. The HX BYPASS vlvs will be reopened manually.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 65 of 150 08 March 2023 33 ID: 28031 Points: 1.00 The following conditions exist on Unit 2:
2A Off Gas process rad monitor is reading above the Hi-Hi setpoint 2B Off Gas process rad monitor has failed down-scale How will the Off Gas system respond?
A.
Off Gas system isolates immediately B.
SJAE Suction Valves close after 15 minutes C.
Chimney Isolation Valve closes after 15 minutes D.
Chimney Isolation Valve goes closed immediately followed by the SJAE suction valves 15 minutes later Answer:
C Answer Explanation Per DAN 902-54 B-8, One radiation monitor is downscale and the other radiation monitor is upscale OR both radiation monitors are upscale, the Off-gas hold-up line isolation logic timer has started AND 15 minutes later the following valves will close:
AO 2-5406, OFFGAS CHIMNEY ISOL. VLV.
AO 2-5423-500, HOLDUP LINE DRAIN VLV.
SO 2-5437, PRESSURIZED DRAIN TANK OUTLET VALVE, isolating the pressurized drain tank.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 66 of 150 08 March 2023 Question 33 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28031 User-Defined ID:
28031 Cross Reference Number:
Topic:
33 - 271000.K4.08 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE271LN001.06
Reference:
DAN 902-54 B-8, DAN 902(3)-3 C-2 K/A:
271000.K4.08 3.7 K/A:
Knowledge of Offgas System design features and/or interlocks that provide for the following: automatic system isolation.
CFR: 41.7 Safety Function: 9 Level: Memory Pedigree: Bank History: None Explanation:
A. Incorrect - Plausible because the Off Gas system will operate normally for 15 minutes then isolate.
B. Incorrect - The SJAE suction valves are NOT interlocked with the Off Gas PRMs. Plausible because of the number of valves that do isolate.
C. Correct - Per DAN 902-54 B-8, One radiation monitor is downscale and the other radiation monitor is upscale OR both radiation monitors are upscale, the Off-gas hold-up line isolation logic timer has started AND 15 minutes later the following valves will close:
AO 2-5406, OFFGAS CHIMNEY ISOL. VLV.
AO 2-5423-500, HOLDUP LINE DRAIN VLV.
SO 2-5437, PRESSURIZED DRAIN TANK OUTLET VALVE, isolating the pressurized drain tank.
D. Incorrect - Plausible because the Chimney Isolation Valve will close just not immediately, there is a 15 minute timer. The SJAE suction valves are not interlocked with the Off Gas PRMs. Plausible because of the number of valves that do isolate.
Required
References:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 67 of 150 08 March 2023 34 ID: 28110 Points: 1.00 From the following, select which statement describes "Operable-Operability", in accordance with Dresden Technical Specifications.
The condition of a system, subsystem, division, component, or device...
A.
capable of performing its specified safety function(s) independent of its support systems.
B.
that will allow testing, calibration or inspection to assure operation is within Safety Limits and LCOs.
C.
necessary to protect the integrity of certain physical barriers to guard against the uncontrolled release of radioactivity.
D.
capable of performing its specified safety function(s) with its support systems capable of performing their required support function(s).
Answer:
D Answer Explanation Per TS 1.1 the definition of OPERABLE-OPERABILITY is: a system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 68 of 150 08 March 2023 Question 34 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
2.00 System ID:
28110 User-Defined ID:
28110 Cross Reference Number:
Topic:
34 - Generic 2.2.38 (1)
Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE299LN001.1
Reference:
TS 1.1 K/A:
Generic 2.2.38 3.6 / 4.5 K/A:
Knowledge of conditions and limitations in the facility license CFR: 41.7 / 41.10 / 43.1 / 45.13 PRA: No Level: Memory Safety Function: N/A Pedigree: Bank History: Quad Cities 2012 ILT NRC Exam Explanation:
A. Incorrect - It IS required to have all its support systems capable of performing their required support functions. Plausible because there are definitions that require independence.
B. Incorrect - This is the definition of a Surveillance. Plausible because it is a definition listed in TS 1.1.
C. Incorrect - This is the definition of a Safety Limit. Plausible because it is a definition listed in TS 1.1.
D. Correct - Per TS 1.1 the definition of OPERABLE-OPERABILITY is:
a system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
Required
Reference:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 69 of 150 08 March 2023 35 ID: 28074 Points: 1.00 A Unit 3 startup is in progress with the Reactor critical below the point of adding heat (POAH).
Which of the following describes the effect of the fuel temperature (Doppler) coefficient of reactivity?
If fuel temperature RISES, A.
positive reactivity will be added due to a change in leakage from the core.
B.
positive reactivity will be added due to a change in resonance absorption in U-238.
C.
negative reactivity will be added due to a change in leakage from the core.
D.
negative reactivity will be added due to a change in resonance absorption in U-238.
Answer:
D Answer Explanation The fuel temperature coefficient of reactivity is negative, therefore a rise in fuel temperature will result in addition of negative reactivity. This is due to more resonant absorption of neutrons in U-238 in the fuel, which results in fewer neutrons reaching thermal energies and fewer thermal fissions in the Reactor.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 70 of 150 08 March 2023 Question 35 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
2.00 System ID:
28074 User-Defined ID:
28074 Cross Reference Number:
Topic:
35 - 292004.K1.05 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: BC02Ir4_Coefficients Obj 5
Reference:
Generic Fundamentals BC02Ir4_Coefficients K/A:
292004.K1.05 2.9 K/A:
Reactivity Coefficients - Define the fuel temperature (Doppler) coefficient of reactivity CFR: 41.1 PRA: No Level: Memory Pedigree: Bank History: None Explanation:
A. Incorrect - This is due to more resonant absorption of neutrons in U-238 in the fuel. Plausible because leakage from the core is relevant with the moderator temperature coefficient of reactivity.
B. Incorrect - The fuel temperature coefficient of reactivity is negative, therefore a rise in fuel temperature will result in addition of negative reactivity. Plausible because reactivity coefficients can be positive or negative.
C. Incorrect - The fuel temperature coefficient of reactivity is negative, therefore a rise in fuel temperature will result in addition of negative reactivity. Plausible because reactivity coefficients can be positive or negative. This is due to more resonant absorption of neutrons in U-238 in the fuel. Plausible because leakage from the core is relevant with the moderator temperature coefficient of reactivity.
D. Correct - The fuel temperature coefficient of reactivity is negative, therefore a rise in fuel temperature will result in addition of negative reactivity. This is due to more resonant absorption of neutrons in U-238 in the fuel, which results in fewer neutrons reaching thermal energies and fewer thermal fissions in the Reactor.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 71 of 150 08 March 2023 36 ID: 28046 Points: 1.00 An event has occurred at the station resulting in the following trend on Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPING as indicated on PPDS. (READINGS ARE COMBINED SUM)
(HH:MM) 01:00 2.0 E+05 µCi/sec 01:30 7.0 E+05 µCi/sec 02:00 2.0 E+06 µCi/sec 02:30 7.0 E+06 µCi/sec 03:00 2.0 E+07 µCi/sec When is DEOP 0300-02, Radioactivity Release Control, FIRST required if the trend continues unchanged?
(Reference provided)
A.
02:30 B.
03:30 C.
04:30 D.
05:30 Answer:
B Answer Explanation This is the first time off-site release rates are in excess of the ALERT level (2.05E+07µCi/sec) for EALs.
This is the entry condition for DEOP 0300-02.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 72 of 150 08 March 2023 Question 36 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
28046 User-Defined ID:
28046 Cross Reference Number:
Topic:
36 - 295038.G.2.4.6 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29502LK056
Reference:
EP-AA-1004 Addendum 3, DEOP 0300-02 K/A:
295038.G.2.4.6 3.7 / 4.7 K/A:
High Off-site Release Rate: Knowledge of emergency and abnormal operation procedures major action categories.
CFR: 41.10/ 43.5/45.13 Safety Function: 9 Pedigree: Bank Level: High History: None Explanation:
A. Incorrect - Plausible because this would coincide with an Unusual Event.
B. Correct - This is the first time off-site release rates are in excess of the ALERT level (2.05E+07µCi/sec) for EALs. This is the entry condition for DEOP 0300-02.
C. Incorrect - Plausible because this time coincides with exceeding Site Area Emergency limits.
D. Incorrect - Plausible because this time coincides with exceeding General Emergency limits.
Required
References:
EP-AA-1004 Addendum 3 EAL Charts containing RU1-RG1 only None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 73 of 150 08 March 2023 37 ID: 7648 Points: 1.00 The Rod Worth Minimizer and its loaded sequence prevent.....
A.
mispositioning of a control rod, whether in or out of sequence.
B.
a MCPR value of greater than 1.00 if the control rod pattern is out of sequence and a rod drop accident occurs.
C.
peak fuel enthalpy from exceeding 170 cal/gm if the highest allowable worth control rod is involved in a single rod scram.
D.
peak fuel enthalpy from exceeding 280 cal/gm if the highest allowable worth control rod is involved in a rod drop accident.
Answer:
D Answer Explanation The rod pattern that inputted into the RWM are to limit control rod worth and the reactivity addition rate resulting from a control rod drop and thus assure that peak fuel enthalpy would be less than 280 cal/gm.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 74 of 150 08 March 2023 Question 37 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
3.00 System ID:
7648 User-Defined ID:
7648 Cross Reference Number:
LI Topic:
37 - 201006.K5.08 Num Field 1:
0.00 Num Field 2:
0.00 Text Field:
Comments:
Objective: DRE201LN006.1
Reference:
UFSAR 7.7.2, TS bases 2.1.1, NF-DR-721 K/A:
201006.K5.08 3.3 K/A:
Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Rod Worth Minimizer System: Rod pattern limits CFR: 41.5 / 45.3 PRA: No Level: Memory Safety Function: 7 Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because the RWM does not PREVENT a mispositioned control rod, it alarms and issues rod blocks if a control rod is mispositioned.
B. Incorrect - Plausible because the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00).
C. Incorrect - Plausible because the IRMs provide protection against local control rod withdrawal errors and results in peak fuel enthalpy below the 170 cal/gm fuel failure threshold criterion.
D. Correct - The rod pattern that inputted into the RWM are to limit control rod worth and the reactivity addition rate resulting from a control rod drop and thus assure that peak fuel enthalpy would be less than 280 cal/gm.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 75 of 150 08 March 2023 38 ID: 28048 Points: 1.00 Unit 3 is at rated conditions when the following occurs:
Annunciators 903-4 G-3, 3A RECIRC PP SEAL CLG WTR FLOW LO and 903-4 G-7, 3B RECIRC PP SEAL CLG WTR FLOW LO, alarm simultaneously.
MO 3-3702, U3 DW SUPPLY VLV, has gone closed and will NOT re-open.
Annunciators 903-4 E-5, 3A RECIRC PP TEMP HI and F-9, 3B RECIRC PP TEMP HI, then alarm within a few seconds of each other.
(1) What is the FIRST action required?
(2) What is the reason for that action?
A.
(1) Scram and trip both Recirc pumps (2) To prevent damage to Recirc pump seals and bearings B.
(1) Scram and trip both Recirc pumps (2) To prevent damage to RBCCW piping due to thermal expansion C.
(1) Reduce both Recirc pump speeds (2) To prevent damage to Recirc pump seals and bearings D.
(1) Reduces both Recirc pump speeds (2) To prevent damage to RBCCW piping due to thermal expansion Answer:
A Answer Explanation With MO 3-3702 closed, no RBCCW flow is getting to the DW to cool the Recirc Pumps. DOA 3700-01 says IF RBCCW flow is lost and CANNOT be restored within one minute, THEN perform the following: IF the Mode Switch is in RUN, THEN manually scram the reactor AND Enter DGP 2-3, Reactor Scram and perform concurrently with this procedure. Trip the Recirculation Pumps AND enter DOA 0202-01, Recirculation (Recirc) Pump Trip - One or Both Pumps
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 76 of 150 08 March 2023 Question 38 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28048 User-Defined ID:
28048 Cross Reference Number:
Topic:
38 - 295018.A2.01 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 20800LK010
Reference:
DOA 3700-01, DAN 902-4 G-3, G-7, E-5, F-9 K/A:
295018.A2.01 3.7 K/A:
Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of Component Cooling Water:
Component temperatures.
CFR: 41.10/43.5/45.13 Safety Function: 8 Level: High Pedigree: Bank History: None Explanation:
A. Correct - With MO 3-3702 closed, no RBCCW flow is getting to the DW to cool the Recirc Pumps. DOA 3700-01 says IF RBCCW flow is lost and CANNOT be restored within one minute, THEN perform the following: IF the Mode Switch is in RUN, THEN manually scram the reactor AND Enter DGP 2-3, Reactor Scram and perform concurrently with this procedure. Trip the Recirculation Pumps AND enter DOA 0202-01, Recirculation (Recirc) Pump Trip - One or Both Pumps and perform concurrently with this procedure.
B. Incorrect - Plausible because Part 1 is correct, part 2 flow is lost to RBCCW piping in the DW, but it will not heat up that fast as to cause thermal expansion due to the increase temperature due to discharge valves being open and there are relief valves installed in the line to protect them.
C. Incorrect - Plausible because if only the 3A and B RECIRC PP TEMP HI annunciators were alarming, it may have been possible to lower pump speed to clear the alarms. Part 2 is correct D. Incorrect - Plausible because if only the 3A and B RECIRC PP TEMP HI annunciators were alarming, it may have been possible to lower pump speed to clear the alarms. Part 2 flow is lost to RBCCW piping in the DW, but it will not heat up that fast as to cause thermal expansion due to the increase temperature due to discharge valves being open and there are relief valves installed in the line to protect them.
Required
Reference:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 77 of 150 08 March 2023 39 ID: 28050 Points: 1.00 Unit 2 has scrammed from full power.
Drywell Temperature (points 9 and 10 of TR 2-1340-1) is 185°F and steady Reactor Building temperature is 185°F and steady RPV pressure is 850 psig and steady Both Recirc Pumps are tripped Narrow Range level is -43 inches and lowering slowly Fuel Zone level is -84 inches and lowering slowly WR level is -56 inches and lowering slowly What level instruments CAN BE used to accurately evaluate the core submergence?
(Reference provided)
A.
ONLY Fuel Zone Instruments B.
ONLY Wide Range Instruments C.
Fuel Zone OR Wide Range Instruments D.
Narrow OR Medium OR Wide Range OR Fuel Zone Instruments Answer:
A Answer Explanation For the given conditions only the Fuel Zone instruments are considered accurate. Narrow/Medium range is below minimum usable of -39 inches. Wide Range is below minimum useable of -51 inches.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 78 of 150 08 March 2023 Question 39 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
28050 User-Defined ID:
28050 Cross Reference Number:
Topic:
39 - 295009.A2.01 Num Field 1:
0.00 Num Field 2:
0.00 Text Field:
Comments:
Objective: 29501LP001
Reference:
DEOP 0100 K/A:
295009.A2.01 4.1 K/A:
Ability to determine and/or interpret the following as they apply to Low Reactor Water Level: Reactor water level.
CFR: 41.10/43.5/45.13 PRA: No Safety Function: 7 Level: High Pedigree: Bank History: None Explanation:
A. Correct - For the given conditions only the Fuel Zone instruments are considered accurate. Narrow/Medium range is below minimum usable of -39 inches. Wide Range is below minimum useable of -51 inches.
B. Incorrect - Plausible because Wide Range would be correct with Drywell temp between 32 to 100°F.
C. Incorrect - Plausible because Fuel Zone is correct and Wide Range would be correct with Drywell temp between 32 to 100°F.
D. Incorrect - Plausible because the table list values that would be useable for all four level instruments and must determine which are applicable at the current pressure and temperature.
REQUIRED
REFERENCES:
DEOP 0100 with entry conditions blanked out None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 79 of 150 08 March 2023 40 ID: 28117 Points: 1.00 Unit 2 was operating at 100% power when a transient occurred and a manual SCRAM was entered.
ONE HOUR LATER:
ALL rods are fully inserted.
ALL MSIV's are closed.
The ONLY available injection sources are HPCI, 2A CRD, 2B CRD, and LPCI cross-tie.
RPV pressure is 500 psig and steady.
The 2A and 2B CRD Pumps are injecting.
RPV level is 20 inches and slowly rising.
Which of the following actions are appropriate under the given conditions?
A.
Fully depressurize the RPV using the Isolation Condenser ONLY B.
Fully depressurize the RPV using HPCI in pressure control mode ONLY C.
Depressurize to the SDC High Temperature interlock, using the Isolation Condenser D.
Maintain RPV pressure between 150 and 800 psig using HPCI in pressure control mode Answer:
A Answer Explanation With the given conditions, the reactor is shutdown with no Table L3 injection subsystems available.
Therefore, the crew would be in DEOP 0400-06. Per EOP-DEOP-SAMG TB Vol 1, Diamonds symbolize sequential decision points leading to alternative branch paths. Questions inside the diamonds identify the decisions that must be made. With RPV level being maintained using only CRD pumps, HPCI is not needed for RPV injection, therefore the correct path to take at the P-1 diamond is to enter the 31 leg and fully depressurize using the Isolation Condenser, since this does not cause the loss of any RPV water inventory.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 80 of 150 08 March 2023 Question 40 Info Topic:
40 - 206000 G.2.4.19 Comments:
Objective: 29503LK007
References:
DEOP 0010-00, DEOP 0100, DEOP 400-06, EOP-DEOP-SAMG-TB-Vol 1 and 2 K/A:
206000 G.2.4.19 3.4 / 4.1 K/A:
Knowledge of emergency and abnormal operation procedures layout, symbols, and icons: High-Pressure Coolant Injection.
CFR: 41.10/45.13 PRA: No Safety Function: 3 Level: High Pedigree: New History: N/A Explanation:
A. Correct - With the given conditions, the reactor is shutdown with no Table L3 injection subsystems available. Therefore, the crew would be in DEOP 0400-06. Per EOP-DEOP-SAMG TB Vol 1, Diamonds symbolize sequential decision points leading to alternative branch paths. Questions inside the diamonds identify the decisions that must be made.
With RPV level being maintained using only CRD pumps, HPCI is not needed for RPV injection, therefore the correct path to take at the P-1 "diamond" is to enter the "31" leg and fully depressurize using the Isolation Condenser, since this does not cause the loss of any RPV water inventory.
B. Incorrect - The reactor should be fully depressurized, per DEOP 0400-06, but using the IC only. HPCI in pressure control mode would cause a loss of inventory, and although it has been greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after shutdown, and detail P4 conditions are HPCI are met, the "33" leg would not be entered since IC is available. Also, HPCI would not be used in pressure control mode ONLY, since this would prevent injection using HPCI. Plausible because HPCI is available, and because the Table P4 Depressurization Criteria for HPCI are met.
C. Incorrect - The RPV would be fully depressurized using the isolation condenser. Plausible because LPCI cross-tie is available, and if the candidate does not recognize that it does not qualify as a detail L3 injection subsystem, then then cooling down to the SDC interlock is an appropriate action per DEOP 0100.
D. Incorrect - With HPCI available, but not needed for injection, the decision diamond at block P-1 would require that leg "31" be entered. This requires the RPV to be fully depressurized using the IC. Plausible because HPCI is available per the stem, and if the candidate assumes HPCI is needed for injection, the actions in block P-9 require controlling pressure between 150 psig and 800 psig, using HPCI.
RO Justification: Although HPCI is available, it is not needed per the stem of the question.
Therefore, the appropriate leg of DEOP 0400-06 would be leg "31" ("Without HPCI")
Objective 29503LK007 - While executing the pressure control leg without HPCI:
- a. Describe why the isolation condenser is used to depressurize the RPV.
- c. Describe the criteria that must be met to depressurize the RPV.
K/A Justification: Must know what the flowchart symbols mean to answer this question. For example, a diamond is a DECISION POINT and a decision must be made and move on from there. The rounded rectangle in blue is an OVERRIDE and must know that will take precedence when the conditions are met.
REQUIRED
REFERENCES:
DEOP 0400-06, DEOP 0100
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 81 of 150 08 March 2023 41 ID: 28116 Points: 1.00 Unit 2 was at 100% power.
The 'A' SBGT train in PRI and the 'B' SBGT train STBY.
The Unit 2 Reactor Building Floor is being re-sealed with an epoxy which contains Volatile Organic Compounds (VOCs).
THEN the following sequence of events occurred:
A small steam leak developed in the HPCI room.
Annunciator 902-3 A-3, RX BLDG VENT CH B RAD HI HI alarmed.
2 minutes later Bus 29 trips on overcurrent.
Five minutes later the ___(1)___ SBGT would be running.
The VOCs from the newly applied epoxy paint may cause a higher radioactive ___(2)___ release from the plant due to their impact on the running SBGT.
A.
(1) 'A' (2) iodine B.
(1) 'B' (2) iodine C.
(1) 'A' (2) particulate D.
(1) 'B' (2) particulate Answer:
B Answer Explanation The Rx Bldg vent high radiation condition causes the SBGT train in PRI (A) to auto start. When Bus 29 trips, the 2/3 'A' SBGT will trip on loss of heater, and low flow, and 2/3 'B' SBGT will auto start.
The charcoal adsorber bed is designed to remove radioactive and non-radioactive forms of Iodine. Oil based paint/epoxy paint and propane powered equipment adversely affect the charcoal adsorber iodine removal efficiency, so the discharge of the running SBGT would have a higher radioactive iodine concentration. The HEPA filters are designed to remove particulate, and are not adversely affected by the VOC's.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 82 of 150 08 March 2023 Question 41 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
0 Difficulty:
0.00 System ID:
28116 User-Defined ID:
28116 Topic:
41 - 261000.K4.04 Comments:
Objective: DRE261LN001.02
Reference:
DANs 902-3 A-3, 923-5 A-6, DOP 7500-01 K/A:
261000.K4.04 3.4 K/A:
Knowledge of Standby Gas Treatment System design features and/or interlocks that provide for the following: Radioactive particulate filtration CFR: 41.7 PRA: No Level: High Pedigree: New History: N/A K/A Justification: With the Rx Bldg vent high radiation condition this causes SBGT to autostart to filter the radioactive particulate out that is causing the high radiation condition.
Explanation:
A. Incorrect - (1) The 2/3B SBGT would be running (2) the second part of the answer is correct. Plausible because if the candidate does not recognize that Bus 29 supplies power to the 2/3A SBGT fan and heater (via MCC 29-9), then this will appear to be correct and (2) the second part of the answer is correct.
B. Correct - (1) The Rx Bldg vent high radiation condition causes the SBGT train in PRI (A) to auto start. When Bus 29 trips, the 2/3 'A' SBGT will trip on loss of heater, and low flow, and 2/3 'B' SBGT will auto start. (2) The charcoal adsorber bed is designed to remove radioactive and non-radioactive forms of Iodine. Per DOP 7500-01 Limitations and Actions, Oil based paint/epoxy paint and propane powered equipment adversely affect the charcoal adsorber iodine removal efficiency, so the discharge of the running SBGT would have a higher radioactive iodine concentration. The HEPA filters are designed to remove particulate, and are not adversely affected by the VOCs.
C. Incorrect - (1) The 2/3B SBGT would be running. (2) the correct answer is iodine. Plausible because (1) if the candidate does not recognize that Bus 29 supplies power to the 2/3A SBGT fan and heater (via MCC 29-9), then this will appear to be correct and (2) the candidate must understand the purpose of the SBGT system, including the what the HEPA filter and charcoal adsorber are for, and the impact of VOCs on these components based on the limitations and actions in DOP 7500-01.
D. Incorrect - (1) the first part of the answer is correct. (2) the correct answer is iodine. Plausible because (1) the first part of the answer is correct (2) the candidate must understand the purpose of the SBGT system, including the what the HEPA filter and charcoal adsorber are for, and the impact of VOCs on these components based on the limitations and actions in DOP 7500-01.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 83 of 150 08 March 2023 42 ID: 28111 Points: 1.00 Unit 2 was producing 985 MWe when the following alarms come in:
902-3 D-13, 2C ELECTROMATIC RELIEF VLV OPEN 902 4 H 17, VLV LEAK DET SYS TEMP HI 902 4 H 19, ACOUSTIC MONITOR ACTUATED Generator output has dropped to 940 MWe as a result of the transient.
DOA 0250-01, RELIEF VALVE FAILURE, Immediate Operator actions have been completed. The following indications are observed:
Generator output is reading 960 MWe Torus bulk temperature is reading 87.4°F and rising at a trend of 1°F every 5 minutes No lights for the 2-203-3C valve are illuminated on the 902-3 panel 902-21 Acoustic Monitor Channel 3C red and amber lights are illuminated (1) What is the current status of the 2C ERV?
(2) What is the NEXT action to be taken?
A.
(1) Leaking (2) Cycle control switch from OFF to MANUAL and back to OFF B.
(1) Leaking (2) Immediately scram the reactor C.
(1) Stuck open (2) Cycle control switch from OFF to MANUAL and back to OFF D.
(1) Stuck open (2) Immediately scram the reactor Answer:
C Answer Explanation The indications are that the 2C ERV is still open following the Immediate Operator actions, therefore it is stuck. Cycling the control switch is the next action to take to attempt to close the ERV.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 84 of 150 08 March 2023 Question 42 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
0 Difficulty:
0.00 System ID:
28111 User-Defined ID:
28111 Cross Reference Number:
Topic:
42 - 239002.A2.03 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE239LN001.8
Reference:
DOA 0250-01 K/A:
239002.A2.03 4.6 / 4.4 K/A:
Ability to (a) predict the impacts of the following on the Safety Relief Valves and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Stuck-open SRV CFR: 41.5 / 43.5 / 45.6 PRA: Yes Level: High Safety Function: 3 Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because per DOA 0250-01 discussion section the more probable failure mode is a leaking relief valve. The second part is correct B. Incorrect - Plausible because per DOA 0250-01 discussion section the more probable failure mode is a leaking relief valve. Plausible because scramming the reactor is an action of DOA 0250-01, but not until Torus temperature is approaching 110°F.
C. Correct - The indications are that the 2C ERV is still open following the Immediate Operator actions, therefore it is stuck. Cycling the control switch is the next action to take to attempt to close the ERV.
D. Incorrect - The first part is correct. Plausible because scramming the reactor is an action of DOA 0250-01, but not until Torus temperature is approaching 110°F.
Required references: None None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 85 of 150 08 March 2023 43 ID: 28054 Points: 1.00 Unit 2 is operating at rated power when a loss of Instrument Bus to the Bailey system occurs.
What is the impact to the FWLC system?
A.
FWLC is now powered from ESS B.
FWLC is now powered from 24 VDC C.
Loss of FWLC from the control room D.
FWLC swaps from 3 element to single element Answer:
A Answer Explanation Bailey is powered by 4 power supplies, two fed from the Instrument Bus (IB) and two fed from the Essential Service System (ESS). On a loss of a single power source the system automatically transfers to the power supplies fed by the other source and an annunciator annunciates.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 86 of 150 08 March 2023 Question 43 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28054 User-Defined ID:
28054 Cross Reference Number:
Topic:
43 - 259002.K6.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE259LN002-12
References:
DOA 6800-T1 K/A:
259002K6.02 3.3 K/A:
Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Reactor Water Level Control system: AC power Safety Function: 2 CFR: 41.7/45.7 Level: Memory Pedigree: New History: N/A Explanation:
A. Correct - Bailey is powered by 4 power supplies, two fed from the Instrument Bus (IB) and two fed from the Essential Service System (ESS). On a loss of a single power source the system automatically transfers to the power supplies fed by the other source and an annunciator annunciates.
B. Incorrect - Plausible because the FWLC system is powered from 24 VDC power supplies that are auctioneered between Essential Service and Instrument Bus.
C. Incorrect - Plausible because if ESS was lost as well this would cause a loss of Feedwater level control from the control room, other means to control water level would be required.
D. Incorrect - Plausible because Bad Quality Failures causes the Individual and Total Steam (Feed) Flow Signals to indicate incorrectly. If in Three Element Control the FWLC automatically transfers to Single Element.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 87 of 150 08 March 2023 44 ID: 28118 Points: 1.00 Unit 2 was operating at 100% power when a SCRAM occurred, and DEOP 100, RPV CONTROL, was entered.
The RPV was depressurized using the main turbine bypass valves.
RPV water level is steady at 30 inches.
RPV pressure is 100 psig and slowly lowering.
The crew is preparing to unisolate the SDC system and start the 2A and 2B SDC pumps.
When performing DOP 1000-01, FILL AND VENT OF SHUTDOWN COOLING SYSTEM, the SDC system was filled but NOT properly vented.
(1) When starting up the SDC system, the potential exists for RPV water level to rapidly drop up to 5 inches ________.
(2) The RPV Cooldown rate will INITIALLY be controlled by throttling ________ per DOP 1000-03, SHUTDOWN COOLING MODE OF OPERATION.
A.
(1) ONLY (2) the 2A and 2B SDC pump discharge valves with MOV 2 3704, RBCCW OUTLET VLV, fully open B.
(1) ONLY (2) MOV 2-3704, RBCCW OUTLET VLV, with the 2A and 2B SDC pump discharge valves at their maximum open positions C.
(1) AND for a significant water hammer event to occur (2) the 2A and 2B SDC pump discharge valves with MOV 2 3704, RBCCW OUTLET VLV, fully open D.
(1) AND for a significant water hammer event to occur (2) MOV 2-3704, RBCCW OUTLET VLV, with the 2A and 2B SDC pump discharge valves at their maximum open positions Answer:
C Answer Explanation (1) The consequence of not properly venting the SDC system would be the potential for a drop in RPV level when the system is unisolated. DOP 1000-03 also states that "RPV water level may drop up to 5 inches when SDC is un isolated". Additionally, per the precautions of DOP 1000-01, Improper filling and venting of the SDC system can cause water hammer/vibrations in the SDC and RBCCW system piping.
(2) With RPV pressure given as 100 psig, RPV water temperature will be approximately 338°F. Per DOP 1000-03, with RPV water temperature above 212°F, the cooldown rate should be controlled by throttling the SDC Pump Discharge valves, while maintaining MO 2(3) 3704, RBCCW OUTLET VLV, fully open.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 88 of 150 08 March 2023 Question 44 Info Question Type:
Multiple Choice Status:
Active Time to Complete: 2 Difficulty:
1.00 System ID:
28118 User-Defined ID:
28118 Topic:
44 - 205000.K5.04 Comments:
Objective: 205LN001.14
Reference:
205000.K5.04 2.9 K/A:
Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the Shutdown Cooling System (RHR Shutdown Cooling Mode): System venting CFR: 41.5 / 45.3 PRA: No Level: High Safety Function: 4 Pedigree: New History: N/A Explanation:
A. Incorrect - (1) The potential for a significant water hammer event also exists. (2) the second part of the answer is correct. Plausible because (1) the candidate must understand the basis behind why venting is performed, and that failing to properly vent a water system can result in significant water hammer when the associate pumps are started and (2) the second part of the answer is correct.
B. Incorrect - (1) The potential for a significant water hammer event also exists. (2) with the given conditions, the RBCCW outlet valve would be maintained fully open, and the SDC pumps discharge valves would be throttled. Plausible because (1) the candidate must understand the basis behind why venting is performed, and that failing to properly vent a water system can result in significant water hammer when the associate pumps are started and (2) if RPV water temperature was less than 212°F, this would be correct, and because.the candidate must first determine that, with the given conditions, RPV temperature is above 212°F, and then apply the appropriate valve positions for the RBCCW outlet valve and the SDC discharge pumps.
C. Correct - (1) The consequence of not properly venting the SDC system would be the potential for a drop in RPV level when the system is unisolated. DOP 1000-03 also states that "RPV water level may drop up to 5 inches when SDC is un isolated". Additionally, per the precautions of DOP 1000-01, Improper filling and venting of the SDC system can cause water hammer/vibrations in the SDC and RBCCW system piping. (2) With RPV pressure given as 100 psig, RPV water temperature will be approximately 338°F. Per DOP 1000-03, with RPV water temperature above 212°F, the cooldown rate should be controlled by throttling the SDC Pump Discharge valves, while maintaining MO 2(3) 3704, RBCCW OUTLET VLV, fully open.
D. Incorrect - (1) The first part of the answer is correct (2 with the given conditions, the RBCCW outlet valve would be maintained fully open, and the SDC pumps discharge valves would be throttled. Plausible because (1) the first part of the answer is correct and (2) if RPV water temperature was less than 212°F, this would be correct, and because
.the candidate must first determine that, with the given conditions, RPV temperature is above 212°F, and then apply the appropriate valve positions for the RBCCW outlet valve and the SDC discharge pumps.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 89 of 150 08 March 2023 45 ID: 28125 Points: 1.00 Unit 2 is operating at full power.
Suddenly, all the FWRVs locked up. RWL is stable at +30 inches.
Five (5) minutes later:
Drywell pressure begins to slowly rise RWL begins lowering at a rate of 1 inch / 25 minutes (1) IAW DOA 0040-01, SLOW LEAK, what would be the approximate change in Drywell pressure?
(2) Is an LCO entry required?
A.
(1) 0.16 psig (2) Yes B.
(1) 0.16 psig (2) No C.
(1) 0.24 psig (2) Yes D.
(1) 0.24 psig (2) No Answer:
C Answer Explanation The thumb rule for computing drywell leakage is Drywell Pressure will rise approximately 0.03 psig for a 1 gpm leak. There are 200 gallons/inch in the RPV and by dividing that by 25 minutes the leak rate is 8 gpm. Therefore, the corresponding change in Drywell Pressure would be 0.24 psig. With >5 gpm of Unidentified Leakage from the RCS LCO 3.4.4 would be entered.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 90 of 150 08 March 2023 Question 45 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28125 User-Defined ID:
28125 Cross Reference Number:
Topic:
45 - 295010.K3.04 (2)
Comments:
Objective: 29501LK055
Reference:
DOA 0040-01 K/A:
295010.K3.04 3.6 K/A:
Knowledge of the reasons for the following responses or actions as they apply to High Drywell Pressure: Leak investigation CFR: 41.5 / 45.6 PRA: Yes Level: High Safety Function: 5 Pedigree: New History: N/A Explanation:
There are a number of thumb rules that apply to Drywell leakage. The key to this question is the application of the correct thumb rule. The thumb rule for computing drywell leakage is Drywell Pressure will rise approximately 0.03 psig for a 1 gpm leak. The leak size would be 8 gpm based on 200 gallons per inch in the RPV.
A. Incorrect - The thumb rule for Drywell Temperature is that temperature will raise 1°F for a two (2) gpm leak. Plausible if the candidate determines that the leakage is 4 gpm by dividing the 8 gpm by 2 that comes from the Drywell Temperature thumb rule and then multiplying that by the 0.03 psig for the Drywell Pressure thumb rule. Second part is correct.
B. Incorrect - The thumb rule for Drywell Temperature is that temperature will raise 1°F for a two (2) gpm leak. Plausible if the candidate determines that the leakage is 4 gpm by dividing the 8 gpm by 2 that comes from the Drywell Temperature thumb rule and then multiplying that by the 0.03 psig for the Drywell Pressure thumb rule. With >5 gpm of Unidentified Leakage from the RCS LCO 3.4.4 would be entered.
C. Correct - The thumb rule for computing drywell leakage is Drywell Pressure will rise approximately 0.03 psig for a 1 gpm leak.
Therefore, the corresponding change in Drywell Pressure would be 0.24 psig. With >5 gpm of Unidentified Leakage from the RCS LCO 3.4.4 would be entered.
D. Incorrect - First part is correct. With >5 gpm of Unidentified Leakage from the RCS LCO 3.4.4 would be entered.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 91 of 150 08 March 2023 46 ID: 28119 Points: 1.00 Units 2 and 3 were operating at 100% power when a LOOP occurred on BOTH units, and BOTH reactors SCRAMMED.
The 2/3 EDG started and automatically connected to Bus 23-1.
THEN Drywell pressure rose to 3.0 psig on UNIT 2.
ONE (1) MINUTE LATER Drywell pressure rose to 2.5 psig on UNIT 3.
(1) What is the current lineup of the U2/3 EDG?
(2) What action, if any, would be required to energize Bus 33-1 from the U2/3 EDG?
A.
(1) The 2/3 EDG is running idle (2) The 2/3 EDG can NOT be closed on Bus 33-1 due to the ECCS signal on Unit 2.
B.
(1) The 2/3 EDG is running idle (2) FORCE the 2/3 EDG to close onto Bus 33-1 per DGA-12, LOSS OF OFFSITE POWER, Attachment C.
C.
(1) The 2/3 EDG is carrying Bus 23-1 (2) The 2/3 EDG can NOT be closed on Bus 33-1 due to the ECCS signal on Unit 2.
D.
(1) The 2/3 EDG is carrying Bus 23-1 (2) FORCE the 2/3 EDG to close onto Bus 33-1 per DGA-12, LOSS OF OFFSITE POWER, Attachment C Answer:
B Answer Explanation (1) With the given conditions, the 2/3 EDG automatically started and connected to Bus 23-1. It then would remain on Bus 23 when drywell pressure on Unit 2 rose above 2.0 psig. When drywell pressure rose above 2.0 psig on Unit 3, one minute later, the Unit 2 EDG Output Breaker would OPEN, but would NOT close on Bus 33-1, thus causing the 2/3 EDG to run idle. (2) In order to energize Bus 33-1, the 2/3 EDG must be forced to close on Bus 33-1 per DGA-12, LOSS OF OFFSITE POWER, Attachment C.
a) There are two 2/3 EDG output breakers, one each for Bus 23-1 and Bus 33-1. Each of the 2/3 EDG output breakers has is interlocked to prevent it from being closed if there is an ECCS signal on the opposite unit. This is to ensure the 2/3 EDG remains available for the unit experiencing the LOCA should a dual unit loss of offsite power occur.
b) Each breaker has a two position (NORMAL-BYPASS) keylock switch that allows bypassing the ECCS interlock described above. There are TWO of these switches, one on the 902-8 panel and one on the 903-8 panel, which bypasses the interlock to their respective 2/3 EDG output breakers ONLY.
c) This interlock will also cause the breaker to trip open if it is closed. A key lock for each breaker is located on the 902(3) 8 panel and will permit bypassing the interlock and closing the associated breaker.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 92 of 150 08 March 2023 Question 46 Info Topic:
46 - 264000.A2.07 Comments: Objective: 264LN001-05
Reference:
DGA - 12 Attachment C K/A:
264000.A2.07 4.7 / 4.6 K/A:
Ability to (a) predict the impacts of the following on the Emergency Generators and (b) base on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Loss of offsite power.
Safety Function: 6 CFR: 41.5/43.5/45.6 Level: High Pedigree: New History: N/A Explanation:
A. Incorrect - (1) the first part of the answer is correct (2) although the ECCS signal on Unit 2, due to High Drywell Pressure, causes the 2/3 EDG output breaker to Unit 3 to trip, this can be overridden using DGA-12, Attachment C. Plausible because (1) the first part of the answer is correct and (2) the candidate must understand the 2/3 EDG output breaker interlocks, and what actions can be taken to override them. This EDG output breaker is unique, since the U2 and U3 EDG output breakers do not have similar keylock override switches.
B. Correct - (1) With the given conditions, the 2/3 EDG automatically started and connected to Bus 23-1. It then would remain on Bus 23 when drywell pressure on Unit 2 rose above 2.0 psig. When drywell pressure rose above 2.0 psig on Unit 3, one minute later, the Unit 2 EDG Output Breaker would OPEN, but would NOT close on Bus 33-1, thus causing the 2/3 EDG to run idle. (2) In order to energize Bus 33-1, the 2/3 EDG must be forced to close on Bus 33-1 per DGA-12, LOSS OF OFFSITE POWER, Attachment C.
a) There are two 2/3 EDG output breakers, one each for Bus 23-1 and Bus 33-1. Each of the 2/3 EDG output breakers has is interlocked to prevent it from being closed if there is an ECCS signal on the opposite unit. This is to ensure the 2/3 EDG remains available for the unit experiencing the LOCA should a dual unit loss of offsite power occur.
b) Each breaker has a two position (NORMAL-BYPASS) keylock switch that allows bypassing the ECCS interlock described above. There are TWO of these switches, one on the 902-8 panel and one on the 903-8 panel, which bypasses the interlock to their respective 2/3 EDG output breakers ONLY.
c) This interlock will also cause the breaker to trip open if it is closed. A key lock for each breaker is located on the 902(3) 8 panel and will permit bypassing the interlock and closing the associated breaker.
C. Incorrect - (1) the 2/3 EDG will be running idle, since the EDG output breaker to Bus 23-1 would open due to the ECCS signal on Unit 3, and the EDG output breaker to Bus 33-1 would not be allowed to automatically close due to the ECCS signal on Unit 2 (2) although the ECCS signal on Unit 2, due to High Drywell Pressure, causes the 2/3 EDG output breaker to Unit 3 to trip, this can be overridden using DGA-12, Attachment C. Plausible because (1) the 2/3 EDG initially closed on Bus 23-1 and would have remained on Bus 23-1 when Unit 2 Drywell Pressure rose above 2.0 psig. If it had been on Bus 23-1 due to power loss only, with no ECCS signal on either unit, and power were subsequently lost to Bus 33-1, it would remain on Bus 23-1 (2) the candidate must understand the 2/3 EDG output breaker interlocks, and what actions can be taken to override them. This EDG output breaker is unique, since the U2 and U3 EDG output breakers do not have similar keylock override switches.
D. Incorrect - (1) the 2/3 EDG will be running idle, since the EDG output breaker to Bus 23-1 would open due to the ECCS signal on Unit 3, and the EDG output breaker to Bus 33-1 would not be allowed to automatically close due to the ECCS signal on Unit 2 (2) the second part of the answer is correct.
Plausible because (1) the 2/3 EDG initially closed on Bus 23-1 and would have remained on Bus 23-1 when Unit 2 Drywell Pressure rose above 2.0 psig. If it had been on Bus 23-1 due to power loss only, with no ECCS signal on either unit, and power were subsequently lost to Bus 33-1, it would remain on Bus 23-1 (2) the second part of the answer is correct.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 93 of 150 08 March 2023 47 ID: 28059 Points: 1.00 Unit 2 is operating at full power. Annunciator 902-3 G-1, LIQUID PROCESS RAD MONITOR HI, alarms.
(1) Where would the rising radiation levels be sensed at?
(2) Where in the Control Room would the radiation levels be indicated on?
A.
(1) Discharge of the RBCCW pumps (2) 902-2 panel B.
(1) Discharge of the RBCCW pumps (2) 923-1 panel C.
(1) Combined RBCCW heat exchanger header (2) 902-2 panel D.
(1) Combined RBCCW heat exchanger header (2) 923-1 panel Answer:
C Answer Explanation Radiation levels in RBCCW are sensed on the Combined heat exchanger header. The radiation levels are indicated on the 902-2 panel.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 94 of 150 08 March 2023 Question 47 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
1.00 System ID:
28059 User-Defined ID:
28059 Cross Reference Number:
Topic:
47 - 400000.K1.03 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE272LN002.3d
Reference:
DAN 902-3 G-1, M-20 K/A:
400000.K1.03 3.1 K/A:
Knowledge of the physical connections and/or cause and effect relationships between the Component Cooling Water System and the following systems: Radiation monitoring systems CFR: 41.4 to 41.5 / 41.7 to 41.9 PRA: No Level: Memory Safety Function: 8 Pedigree: New History: N/A Explanation:
A. Incorrect - Radiation levels in RBCCW are sensed on the Combined heat exchanger header. Plausible because the RBCCW Pressure Lo is sensed at the discharge of the pumps. The second part is correct B. Incorrect - Radiation levels in RBCCW are sensed on the Combined heat exchanger header. Plausible because the RBCCW Pressure Lo is sensed at the discharge of the pumps. The radiation levels are indicated on the 902-2 panel. Plausible because there are many RBCCW parameters that are on the 923-1 panel.
C. Correct - Radiation levels in RBCCW are sensed on the Combined heat exchanger header. The radiation levels are indicated on the 902-2 panel.
D. Incorrect - The first part is correct. The radiation levels are indicated on the 902-2 panel. Plausible because there are many RBCCW parameters that are on the 923-1 panel.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 95 of 150 08 March 2023 48 ID: 28128 Points: 1.00 A transient occurred on Unit 2 The U2 NSO has depressed both scram buttons and placed the mode switch in Shut Down.
All control rods do NOT go in.
APRMs indicate Reactor power is at 8%.
Channel A and B Scram Solenoid Group Lights are out.
The Following Annunciators are in on the 902-5 panel:
A-1 Scram Valve Air Supply Press Lo A-10 Channel A Manual Trip A-15 Channel B Manual Trip C-1 Scram Inst Vol Hi Lvl Rod Block D-1 West Scram Inst Vol Not Drained E-1 East Scram Inst Vol Not Drained Per DGP 02-03, Reactor Scram, the NSO's next action is to ___(1)___ and based on the above information, this action ___(2)___ be successful in reducing Reactor power.
A.
(1) initiate ARI (2) will B.
(1) initiate ARI (2) will NOT C.
(1) remove RPS fuses (2) will D.
(1) remove RPS fuses (2) will NOT Answer:
B Answer Explanation DGP 2-3 directs the operator to initiate ARI if all rods do not go in. ARI will not be successful since the scram air header is already depressurized and the rods have not gone in. The scram valves got the signal since the scram pilot solenoids are de-energized. The scram inst volume not drained alarm indicates a hydraulic lock on rods.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 96 of 150 08 March 2023 Question 48 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28128 User-Defined ID:
28128 Cross Reference Number:
Topic:
48 - 295037.K2.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE212L002
Reference:
DGP 2-3 K/A:
295037.K2.02 4.0 K/A:
Knowledge of the relationship between SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown and the following systems or components:
Redundant reactivity control system CFR: 41.7 41.8 / 45.8 PRA: No Level: High Safety Function: 1 Pedigree: Bank History: None Explanation:
A. Incorrect - First part is correct. ARI will not be successful since the scram air header is already depressurized and the rods have not gone in. Plausible because ARI is one of the backup systems to scram the control rods.
B. Correct - DGP 2-3 directs the operator to initiate ARI if all rods do not go in. ARI will not be successful since the scram air header is already depressurized and the rods have not gone in. The scram valves got the signal since the scram pilot solenoids are de-energized. The scram inst volume not drained alarm indicates a hydraulic lock on rods.
C. Incorrect - Plausible because removing the RPS fuses would be appropriate if this was an electrical ATWS and if it was only an electrical ATWS the rods would go in being successful in reducing Reactor power.
D. Incorrect - Plausible because removing the RPS fuses would be appropriate if this was an electrical ATWS. Second part is correct.
REQUIRED
REFERENCE:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 97 of 150 08 March 2023 49 ID: 28061 Points: 1.00 Unit 2 is operating at rated power when a transient occurred.
Drywell pressure is 6 psig All MSIVs are closed All rods are in RPV pressure is 600 psig and steady RPV level is lowering The EARLIEST an Emergency Depressurization can be performed is when RPV level reaches ___(1)___
inches.
This is done in order to maintain Peak Cladding Temperature less than ___(2)___ °F.
A.
(1) -143 (2) 1500 B.
(1) -170 (2) 1500 C.
(1) -143 (2) 1800 D.
(1) -170 (2) 1800 Answer:
B Answer Explanation With RPV pressure > 500 psig, TAF is -170". Per EPGs, an ED may be performed when RPV level reaches TAF and MUST be performed prior to reaching MSCRWL (-186" with RPV pressure >500 psig).
The PCT limit that applies is 1500F.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 98 of 150 08 March 2023 Question 49 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice? No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
28061 User-Defined ID:
28061 Topic:
49 - 295031.K3.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29501LK028
Reference:
EOP-DEOP-SAMG TB K/A:
295031.K3.02 4.7 K/A:
Knowledge of the reasons for the following responses as they apply to Reactor Low Water Level: Core submergence.
CFR: 41.5/45.6 Safety Function: 2 PRA: No Level: High Pedigree: Bank History: 14-1 NRC Explanation:
A. Incorrect - With RPV pressure > 500 psig, TAF is -170". ED due to RPV level prior to reaching TAF is not permitted by procedure Plausibility: Candidate must recognize that a corrected value of -170 inches is used for TAF when RPV Pressure is > 500 Psig. The PCT limit of 1500°F listed is correct.
B. Correct - With RPV pressure > 500 psig, TAF is -170". Per EPGs, an ED may be performed when RPV level reaches TAF and MUST be performed prior to reaching MSCRWL (-186" with RPV pressure >500 psig). The PCT limit that applies is 1500°F.
C. Incorrect - With injection sources available, a blowdown would occur at TAF, therefore PCT limit of 1800F is not correct. Also, with RPV pressure > 500 psig, TAF is -170".
Plausibility: Candidate must recognize that a corrected value of -170 inches is used for TAF when RPV Pressure is > 500 Psig. The candidate may not recognize that a PCT limit of 1500°F is correct for a blowdown at TAF, versus 1800°F for steam cooling.
D. Incorrect - With injection sources available, a blowdown would occur at TAF, therefore PCT limit of 1800°F is not correct. The level of -170" is correct for the current pressure.
Plausibility: The candidate may not recognize that a PCT limit of 1500°F is correct for a blowdown at TAF, versus 1800°F for steam cooling.
Justification of HIGH order: The candidate must determine whether to use pressure corrected TAF or not, and identify the correct PCT limit.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 99 of 150 08 March 2023 50 ID: 28115 Points: 1.00 Unit 2 was operating at near rated power, when a LOCA occurred.
Reactor pressure is 100 psig and LOWERING slowly.
Which of the following sets of parameters would result in a USABLE indicated Wide Range RPV Water Level?
Drywell Temperature of ___(1)___ and an indicated Wide Range RPV Water Level of ___(2)___.
(Reference provided)
A.
(1) 250°F (2) -13 inches B.
(1) 275°F (2) -32 inches C.
(1) 320°F (2) 0 inches D.
(1) 370°F (2) +17 inches Answer:
A Answer Explanation Wide Range RPV Level instrument range is +330 inches to -70 inches however it is only usable depending on the Drywell Temperature. Also, if above saturation temperature in the RPV, water level instruments may be unreliable. At 100 psig reactor pressure, saturation temperature is approximately 340°F. At 250°F and -13 inches, the Wide Range RPV Level instrument would still be usable.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 100 of 150 08 March 2023 Question 50 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28115 User-Defined ID:
28115 Cross Reference Number:
Topic:
50 - 295028.A2.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29501LK001
Reference:
DEOP 100 K/A:
295028.A2.02 3.6 / 4.0 K/A:
Ability to determine and/or interpret the following as they apply to High Drywell Temperature: Reactor pressure CFR: 41.10 / 43.5 / 45.13 PRA: No Level: High Safety Function: 5 Pedigree: Bank History: None Explanation:
A. Correct - Wide Range RPV Level instrument range is +330 inches to
-70 inches however it is only usable depending on the Drywell Temperature. Also, if above saturation temperature in the RPV, water level instruments may be unreliable. At 100 psig reactor pressure, saturation temperature is approximately 340°F. At 250°F and -13 inches, the Wide Range RPV Level instrument would still be usable.
B. Incorrect - Drywell temperature between 201°F to 300°F Wide Range RPV Level is usable above -20 inches.
C. Incorrect - Drywell temperature between 301°F and 400°F Wide Range RPV Level is usable above 19 inches.
D. Incorrect - Drywell temperature between 301°F and 400°F Wide Range RPV Level is usable above 19 inches.
All distractors are plausible based on interpreting both the graph of RPV pressure vs Drywell temperature and the chart that determines the criteria for usability.
Required
Reference:
DEOP 100 with entry conditions blanked out None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 101 of 150 08 March 2023 51 ID: 28063 Points: 1.00 A Unit 2 transient has resulted in the following conditions:
Reactor Scram with all rods in RPV completely depressurized Torus bottom pressure is 15 psig Torus water level is 9 feet 3 inches Torus bulk water temperature is 150°F RPV water level is being maintained at -160 inches with the 2B Core Spray pump, operating at rated flow No additional ECCS pumps are available The 2B Core Spray pump may experience pump damage due to violating its ___(1)___ AND ___(2)___.
(Reference provided)
A.
(1) Vortex limits ONLY (2) securing the 'B' Core Spray pump and flooding the containment is required B.
(1) Vortex limits ONLY (2) continuing 'B' Core Spray pump operation regardless of potential pump damage is permitted C.
(1) Vortex AND NPSH limits (2) securing the 'B' Core Spray pump and flooding the containment is required D.
(1) Vortex AND NPSH limits (2) continuing 'B' Core Spray pump operation regardless of potential pump damage is permitted Answer:
B Answer Explanation Vortex limits are being violated but Core Spray is still needed for RPV level. NPSH limits are not being exceeded.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 102 of 150 08 March 2023 Question 51 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
28063 User-Defined ID:
28063 Cross Reference Number:
Topic:
51 - 295030.K3.07 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29502LP005
Reference:
DEOP 100, table V and table W.
K/A:
295030.K3.07 3.8 K/A:
Knowledge of the reasons for the following responses or actions as they apply to Low Suppression Pool Water Level:
NPSH/vortex limits.
CFR: 41.5/45.6 Safety Function: 2 & 4 PRA: Yes Level: High Pedigree: Bank History: 18-1 NRC Explanation: With the Core Spray pump operating at rated flow (5000 gpm) the pump is only violating its vortex limit, NOT the NPSH.
A. Incorrect - Plausible because Only Vortex limits are being violated. If the pump were secured flooding of containment would be required.
The pump is allowed to be operated regardless of NPSH and Vortex limits. DEOP cautions exists to warn of possible equipment damage, but do not direct securing the pump.
B. Correct - Vortex limits are being violated but Core Spray is still needed for RPV level. NPSH limits are not being exceeded.
C. Incorrect - Plausible because Only Vortex limits are being violated.
Continued operation of B Core spray is permitted. If the pump were secured flooding of containment would be required. The pump is allowed to be operated regardless of NPSH and Vortex limits.
DEOP cautions exists to warn of possible equipment damage, but do not direct securing the pump.
D. Incorrect - Only Vortex limits are being violated. Plausible because the second part is correct and first part must be determined from the graphs.
REQUIRED
REFERENCE:
DEOP 100 with entry conditions redacted.
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 103 of 150 08 March 2023 52 ID: 28057 Points: 1.00 A LOOP and subsequent transient occurred on Unit 2 requiring RPV injection with low pressure ECCS systems.
The Unit Supervisor has directed you to maximize injection with Core Spray.
LPCI is unavailable NO high pressure injection sources are available RPV pressure is 90 psig What is the expected TOTAL flowrate of injection into the RPV?
A.
4,500 gpm B.
9,000 gpm C.
18,000 gpm D.
27,000 gpm Answer:
B Answer Explanation Each CS pump is required to generate 4500 gpm flowrate against an RPV pressure of 90 psig.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 104 of 150 08 March 2023 Question 52 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28057 User-Defined ID:
28057 Cross Reference Number:
Topic:
52 - 209001.A1.04 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 209LN001.03
Reference:
TS 3.5.1 SR 3.5.1.5 K/A:
209001.A1.04 4.1 K/A:
Ability to predict and/or monitor changes in parameters associated with operation of the Low-Pressure Core Spray System, including: Reactor pressure CFR: 41.5 Safety Function: 2 PRA: Yes Pedigree: Bank History: 16-1 NRC Level: Memory Explanation:
A. Incorrect - This would be correct for 1 CS pump.
B. Correct - Each CS pump is required to generate 4500 gpm flowrate against an RPV pressure of 90 psig.
C. Incorrect - This would be correct for 4 LP ECCS pumps (i.e. LPCI)
D. Incorrect - This would be correct for all LP ECCS pumps.
REQUIRED
REFERENCE:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 105 of 150 08 March 2023 53 ID: 28114 Points: 1.00 Unit 2 is at 30% Reactor power:
All IRMs are currently on range 2 IRM 13 is bypassed IRM 15 fails upscale to 125 What is the expected plant response, if any?
A.
Rod Block and NO RPS actuation B.
Rod Block and RPS Channel B half-scram C.
NO Rod Block and NO RPS actuation D.
NO Rod Block and RPS Channel B half-scram Answer:
C Answer Explanation With the Mode switch in RUN IRM Hi Rod Blocks and IRM Hi Hi Scram are bypassed.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 106 of 150 08 March 2023 Question 53 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
0 Difficulty:
0.00 System ID:
28114 User-Defined ID:
28114 Cross Reference Number:
Topic:
53 - 215003.K4.10 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 215LN003.02
Reference:
DRE215LN003 K/A:
215003 K4.10 3.6 K/A:
Knowledge of Intermediate Range Monitor System design features and/or interlocks that provide for the following:
Automatically bypassing IRM rod block signals CFR: 41.7 Level: Memory PRA: No Safety Function: 7 Pedigree: Bank History: None Explanation:
A. Incorrect - With the Mode switch in RUN IRM Hi Rod Blocks are bypassed. Plausible because if the Mode switch was not in RUN then a Rod Block would occur.
B. Incorrect - With the Mode switch in RUN IRM Hi Rod Blocks and IRM Hi Hi Scram are bypassed. Plausible because if the Mode switch was not in RUN then a Rod Block and a B Half Scram would occur.
C. Correct - With the Mode switch in RUN IRM Hi Rod Blocks and IRM Hi Hi Scram are bypassed.
D. Incorrect - With the Mode switch in RUN IRM Hi Hi Scram is bypassed. Plausible because if the Mode switch was not in RUN then a B Half Scram would occur Required
Reference:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 107 of 150 08 March 2023 54 ID: 28066 Points: 1.00 Unit 2 is at rated power.
The U2 NSO took the SBLC Control Switch to 1&2 position.
Which of the following valves would receive an isolation signal?
- 1) 2-1201-1 RX OUTLET ISOL
- 2) 2-1201-1A RX OUTLET BYP
- 3) 2-1201-2 INLET ISOL
- 4) 2-1201-3 AUX PP SUCT
- 5) 2-1201-7 RX RETURN A.
1 and 2 ONLY B.
3 and 4 ONLY C.
2, 3, 4 and 5 ONLY D.
1, 2, 3, 4, and 5 Answer:
D Answer Explanation These valves all will be closed following a SBLC initiation.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 108 of 150 08 March 2023 Question 54 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
1.00 System ID:
28066 User-Defined ID:
28066 Cross Reference Number:
Topic:
54 - 211000.A3.06 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE211LN001.6
Reference:
DOP 1100-02 K/A:
211000.A3.06 4.1 K/A:
Ability to monitor automatic operation of the Standby Liquid Control System, including: RWCU system isolation CFR: 41.7 / 45.7 PRA: No Level: Memory Safety Function: 1 Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because these valves would all indicate closed, but not all of them.
B. Incorrect - Plausible because these valves would all indicate closed, but not all of them.
C. Incorrect - Plausible because these valves would all indicate closed, but not all of them.
D. Correct - These valves all will be closed following a SBLC initiation.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 109 of 150 08 March 2023 55 ID: 28067 Points: 1.00 Unit 2 was operating at rated power when 2-203-1D, U2 1D INBD MSIV, failed shut.
RPV level is being controlled at -40" in automatic RPV pressure is 1040 psig and being controlled with the IC All attempts to open the 2-4399-74, CLEAN DEMIN VLV, have failed Per DOP 1300-03, MANUAL OPERATION OF THE ISOLATION CONDENSER, the preferred RPV PRESSURE CONTROL method is:
A.
ADS valve operation B.
HPCI operation in the pressure control mode C.
IC operation with contaminated demin make up D.
IC operation with fire suppression system make up Answer:
B Answer Explanation With the conditions stated in the stem, a PCIS GRP I has occurred due to high steam flow. IC shell side makeup is required due to inventory loss. A reactor scram followed with a failure of all control rods to fully insert. With the failure of MO 2-4399-74 normal shell side makeup using IC makeup pumps is unavailable. Per DOP 1300-03, clean demin is the preferred source with IC makeup pumps unavailable; however, the flowpath for clean demin also requires the use of MO 2-4399-74 thereby making IC with clean demin unavailable. Since RPV level is being maintained by FWLC in auto, HPCI is NOT needed for RPV level control and therefore available in the pressure control mode. IC operation with fire suppression makeup and IC operation with contaminated makeup are listed after HPCI in the order of preference in DOP 1300-03, MANUAL OPERATION OF ISOLATION CONDENSER.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 110 of 150 08 March 2023 Question 55 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
4.00 System ID:
28067 User-Defined ID:
28067 Topic:
55 - 207000.K3.01 Comments:
Objective: 207LN001.08
Reference:
DOP 1300-03 K/A:
207000.K3.01 3.8 K/A:
Knowledge of the effect that a loss or malfunction of the Isolation (Emergency) Condenser will have on the following systems or system parameters: Reactor pressure.
CFR: 41.7/45.4 Safety Function: 4 PRA: Yes Level: High Pedigree: Bank History: 14-1 NRC Explanation:
A. Incorrect - ADS valves are available however they are not the preferred RPV pressure control method at this time.
B. Correct - With the conditions stated in the stem, a PCIS GRP I has occurred due to high steam flow. IC shell side makeup is required due to inventory loss. A reactor scram followed with a failure of all control rods to fully insert. With the failure of MO 2-4399-74 normal shell side makeup using IC makeup pumps is unavailable. Per DOP 1300-03, clean demin is the preferred source with IC makeup pumps unavailable; however, the flowpath for clean demin also requires the use of MO 2-4399-74 thereby making IC with clean demin unavailable. Since RPV level is being maintained by FWLC in auto, HPCI is NOT needed for RPV level control and therefore available in the pressure control mode.
IC operation with fire suppression makeup and IC operation with contaminated makeup are listed after HPCI in the order of preference in DOP 1300-03, MANUAL OPERATION OF ISOLATION CONDENSER.
C. Incorrect - This is the LEAST preferred option for makeup water to the IC.
HPCI would be employed for pressure control prior to the use of the IC with contaminated demin water as shell side makeup.
D. Incorrect - Fire water is the NEXT preferred source of makeup water to the IC shell side, however RPV pressure control should be transitioned to HPCI operation in the pressure control mode before this is employed.
Justification of HIGH order: The candidate must analyze the plant conditions and system status and apply knowledge of procedure precautions and limitations.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 111 of 150 08 March 2023 56 ID: 28064 Points: 1.00 Unit 2 is operating at rated power.
Grid Disturbances cause the following alarms to actuate.
DAN 902(3)-8 C-2, RES AUX TR 22 TROUBLE DAN 902-8 H-10, 4 KV BUS 24-1 VOLTAGE DEGRADED DOA 6100-03, AUX POWER TRANSFORMER TROUBLE, directs reducing power to 65 to 70%
Per DGP 03-01, POWER CHANGES, the PREFERRED method is...
A.
insert CRAM rods ONLY.
B.
immediately reduce core flow ONLY.
C.
insert rods in reverse sequence THEN reduce core flow.
D.
insert CRAM rods and reduce core flow simultaneously.
Answer:
C Answer Explanation Per the DOA for Aux Power Transformer Trouble, actions are to Decrease Unit load, while increasing VARS. Per DGP 03-01 Power Changes Emergency Load Decrease Guidance:
- a. IF FCL is >93%, THEN reduce power by inserting control rods in reverse sequence (preferred) or CRAM rod insertion 90 MWe of generator power OR 9% of APRM power.
THEN
- b. Reduce Reactor power by decreasing core flow to 58 Mlbm/hour (58 to 62 Mlbm/hour)
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 112 of 150 08 March 2023 Question 56 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28064 User-Defined ID:
28064 Cross Reference Number:
Topic:
56 - 700000.A1.04 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE103LN002.3
References:
DGP 03-01, DAN 902-8 H-10, DOA 6100-03 K/A:
700000.A1.04 3.6 K/A:
Ability to operate and/or monitor the following as the apply to Generator Voltage and Electric Grid Disturbances: Reactor Controls CFR: 41.5/41.10/45.5/45.7/45.8 Safety Function: 6 Level: High Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because CRAMs can be used but are not preferred. Quick action required due to 5 minute timer B. Incorrect - Plausible because this would be correct if the Unit was not at full power with FCL greater than 93%
C. Correct - Per the DOA for Aux Power Transformer Trouble, actions are to Decrease Unit load, while increasing VARS. Per DGP 03-01 Power Changes Emergency Load Decrease Guidance:
- a. IF FCL is >93%, THEN reduce power by inserting control rods in reverse sequence (preferred) or CRAM rod insertion 90 MWe of generator power OR 9% of APRM power.
THEN
- b. Reduce Reactor power by decreasing core flow to 58 Mlbm/hour (58 to 62 Mlbm/hr)
D. Incorrect - Plausible because CRAMs could be used but not preferred, core flow would be lowered but not until after FCL less than 93%. Quick action require due to 5 minute timer.
Required
References:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 113 of 150 08 March 2023 57 ID: 28069 Points: 1.00 Which of the following describes the sequencing of the directional control valves during a CRD single notch insert?
During a CRD single notch insert, the drive insert and exhaust valves (121 & 123) open, then.....
A.
settle valve (120) opens, drive insert and exhaust valves (121 & 123) close, drive settles and settle valve (120) closes.
B.
settle valve (120) opens, drive insert valve (121) closes, drive settles, and then exhaust valve (123) closes.
C.
drive insert and exhaust valves (121 & 123) close, drive settles through the cooling water header.
D.
drive insert and exhaust valves (121 & 123) close, drive settles by the stab valves closing.
Answer:
A Answer Explanation The 121 & 123 directional control valves open (for 3 seconds). Two seconds into the sequence, the 120 directional control valve, the "settle" valve, opens for five seconds. Conclusion is that the 121 & 123 valves close and the 120 valve closes 4 seconds later.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 114 of 150 08 March 2023 Question 57 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
3.00 System ID:
28069 User-Defined ID:
28069 Cross Reference Number:
Topic:
57 - 201001.K1.08 Num Field 1:
0.00 Num Field 2:
0.00 Text Field:
Comments:
Objective: 201LN001.06
References:
M-34, DOP 0400-01, 12E-2414 K/A:
201001.K1.08 4.2 K/A:
Knowledge of the physical connections and/or cause and effect relationships between the Control Rod Drive Hydraulic System and the following systems: Reactor manual control system.
CFR: 41.1-3 to 41.5-8/45.1-6/45.8 PRA: No Safety Function: 1 Level: Memory Pedigree: Bank History: None Explanation:
A. Correct - The 121 & 123 directional control valves open (for 3 seconds). Two seconds into the sequence, the 120 directional control valve, the "settle" valve, opens for five seconds. Conclusion is that the 121 & 123 valves close and the 120 valve closes 4 seconds later.
B. Incorrect - Plausible because the 121 and 123 valves do close, just not the sequence described.
C. Incorrect - Plausible because the 121 and 123 valves are correct, drive settles via the drive header.
D. Incorrect - Plausible because the 121 and 123 valves are correct, stab valves are not part of the settle process.
Required
Reference:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 115 of 150 08 March 2023 58 ID: 28070 Points: 1.00 Unit 3 was operating at full power with the U3 EDG OOS.
Bus 34 trips on overcurrent No actions are taken to cross-tie busses Which of the following can be operated?
A.
3A FPC pump B.
3B SBLC pump C.
3C LPCI pump D.
3D Condensate/Condensate Booster pump Answer:
A Answer Explanation 3A FPC pump is powered from Bus 38, therefore has power and can be operated
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 116 of 150 08 March 2023 Question 58 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
2.00 System ID:
28070 User-Defined ID:
28070 Cross Reference Number:
Topic:
58 - 233000.K2.01 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE233LN001.2
Reference:
233000.K2.01 3.1 K/A:
Knowledge of electrical power supplies to the following: Fuel pool cooling pumps CFR: 41.7 PRA: No Level: Memory Safety Function: 9 Pedigree: New History: N/A Explanation:
A. Correct - 3A FPC pump is powered from Bus 38, therefore has power and can be operated B. Incorrect - 3B SBLC pump is powered from MCC 39-1. With U3 EDG OOS and Bus 34 tripped on overcurrent then Bus 34-1 and Bus 39 have no power either.
C. Incorrect - 3C LPCI pump is powered from Bus 34-1. With U3 EDG OOS and Bus 34 tripped on overcurrent then Bus 34-1 has no power.
D. Incorrect - 3D Condensate/Condensate Booster pump is powered from Bus 34.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 117 of 150 08 March 2023 59 ID: 28071 Points: 1.00 Unit 2 is operating at 40% power.
DOS 0250-02, FULL CLOSURE TIMING AND EXERCISING OF MAIN STEAM ISOLATION VALVES, is being performed.
The MSIVs are being timed using a stopwatch and read to the nearest one tenth of a second.
Which of the following would meet the Acceptance Criteria?
A.
4.0 seconds "Switch-to-Light" AND 2.8 seconds "Light-to Light" B.
4.2 seconds "Switch-to-Light" AND 3.8 seconds "Light-to-Light" C.
4.6 seconds "Switch-to-Light" AND 2.5 seconds "Light-to-Light" D.
5.1 seconds "Switch-to-Light" AND 3.3 seconds "Light-to-Light" Answer:
B Answer Explanation Per DOS 0250-02 with the Unit in a HOT condition the limit is < 4.5 seconds "Switch-to Light" AND > 3.0 seconds "Light-to Light"
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 118 of 150 08 March 2023 Question 59 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28071 User-Defined ID:
28071 Cross Reference Number:
Topic:
59 - 223002.291001.K1.09 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE239LN001-08
Reference:
DOS 0250-02 K/A:
223002.291001.K1.09 2.7 K/A:
Primary Containment Isolation/Nuclear Steam Supply Shutoff -
The stroke test for a valve, including the use of a stopwatch CFR: 41.3 Safety Function: 3 Level: Memory Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because part 1 is correct. Part 2 must realize that 3 seconds is a minimum time not a maximum time.
B. Correct - Per DOS 0250-02 with the Unit in a HOT condition the limit is < 4.5 seconds "Switch-to Light" AND > 3.0 seconds "Light-to Light" C. Incorrect - Plausible because part 1 must determine that rounding is only required if timing to 100ths of seconds. Part 2 must realize that 3 seconds is a minimum time not a maximum time.
D. Incorrect - Plausible because this would be a correct answer if the reactor was in COLD conditions.
REQUIRED
REFERENCE:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 119 of 150 08 March 2023 60 ID: 22619 Points: 1.00 Per the UFSAR, why is it NOT permissible to run the Mechanical Vacuum Pump when the Reactor Mode Switch is in the RUN position?
A.
This would bypass the Low Condenser Vacuum scram with the Reactor Mode Switch in RUN.
B.
This would provide an untreated release pathway for non-condensibles to the Main Chimney.
C.
Because of the potential of Hydrogen fires and/or explosions due to the gases being admitted to the main condenser.
D.
Because the common suction line can NOT accommodate the required flow to both the Mechanical Vacuum Pump and the SJAE's.
Answer:
B Answer Explanation Not permissible in RUN due to bypassing the Off Gas System and discharging directly to the 310' chimney (which results in an untreated release pathway for non-condensibles).
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 120 of 150 08 March 2023 Question 60 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
22619 User-Defined ID:
22619 Cross Reference Number:
Topic:
60 - Generic 2.3.11 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE275LN001.03
Reference:
UFSAR 11.3.2.3, DAN 902-7 H-3 K/A:
Generic 2.3.11 3.8 / 4.3 K/A:
Ability to control radiation releases CFR: 41.11/43.4/45.10 PRA: No Level: Memory Pedigree: Bank History: 08-1 NRC, 16-1 NRC Explanation:
A. Incorrect - Plausible because the low condenser vacuum bypass is jumpered out when running the Mechanical Vacuum Pump.
B. Correct - Not permissible in RUN due to bypassing the Off Gas System and discharging directly to the 310' chimney (which results in an untreated release pathway for non-condensibles).
C. Incorrect - Plausible because this is true for the discharge of the pump, not the suction.
D. Incorrect - The mechanical vacuum pump discharges via the gland seal exhaust piping, not the SJAE piping. Plausible because both lines go to the Main Chimney.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 121 of 150 08 March 2023 61 ID: 28043 Points: 1.00 A scram has occurred on U2 due to rising Drywell Pressure.
Drywell Pressure continues to rise even with Drywell Sprays initiated.
If pressure exceeds the Pressure Suppression Pressure limit a blowdown is required.
What is the reason for performing an Emergency Depressurization? To ensure the.....
A.
pressure capability of the primary containment is not exceeded B.
maximum primary containment pressure at which ADSVs can be opened and will remain open is not exceeded C.
RPV is depressurized while pressure suppression capability is still available and to limit further release of energy into the primary containment D.
maximum primary containment pressure at which vent valves sized to reject all decay heat from the containment can be opened and closed is not exceeded Answer:
C Answer Explanation Per DEOP/SAMG Technical Basis If containment sprays cannot be initiated or are ineffective in controlling primary containment pressure below the Pressure Suppression Pressure, a blowdown is required. The blowdown is performed to limit further release of energy into the primary containment and to ensure that the RPV is depressurized while pressure suppression capability is still available.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 122 of 150 08 March 2023 Question 61 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
4 Difficulty:
4.00 System ID:
28043 User-Defined ID:
28043 Cross Reference Number:
Topic:
61 - 295024.K3.04 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29502LK008
References:
295024.K3.05 4.2 K/A:
Knowledge of the reasons for the following responses or actions as they apply to High Drywell Pressure: Emergency depressurization.
CFR: 41.5/45.6 Safety Function: 5 Level: Memory Pedigree: New History: N/A Explanations:
A. Incorrect - Plausible because this is a reason for containment venting to be evaluated due to Primary Containment Pressure Limit.
B. Incorrect - Plausible because this is a reason for containment venting to be evaluated due to Primary Containment Pressure Limit.
C. Correct - Per DEOP/SAMG Technical Basis If containment sprays cannot be initiated or are ineffective in controlling primary containment pressure below the Pressure Suppression Pressure, a blowdown is required. The blowdown is performed to limit further release of energy into the primary containment and to ensure that the RPV is depressurized while pressure suppression capability is still available.
D. Incorrect - Plausible because this is a reason for containment venting to be evaluated due to Primary Containment Pressure Limit.
Required
References:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 123 of 150 08 March 2023 62 ID: 28032 Points: 1.00 Given the following Unit 2 conditions:
Mode 5 with the Reactor Cavity and Dryer Separator flooded to 4 inches below cavity ventilation ducts Main Steam Line (MSL) Plugs are installed in all four MSLs Reactor cavity water temperature is 93°F Unit 2 Fuel Pool water level is 37'9" at 93°F Operating team is maintaining Reactor cavity and Fuel Pool temp between 90°F to 100°F Reactor core alterations are in progress 2A Shutdown Cooling (SDC) Pump is OOS 2B SDC Pump is aligned to the Reactor Vessel with MO 2-3704 positioned at max setting 2C SDC Pump is aligned to the Unit 2 Fuel Pool Cooling System IF 2B SDC Pump tripped on overcurrent with these conditions, which decay heat removal alternatives listed below is a viable option to control reactor water temperatures?
A.
Initiate 2A Core Spray pump B.
Raise RBCCW flow to shell side of SDC heat exchangers C.
Initiate feed and bleed with CRD and main steam line drains D.
Utilize natural circulation from Fuel Pool to Reactor Cavity with Fuel Pool Gates removed Answer:
D Answer Explanation Since 2C SDC loop is aligned to the Fuel Pool Utilizing a natural circulation flow path from Fuel Pool to Reactor Cavity with Fuel Pool Gates removed would provide decay heat removal path immediately on the 2B SDC pump trip.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 124 of 150 08 March 2023 Question 62 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28032 User-Defined ID:
28018 Cross Reference Number:
Topic:
62 - 295021.K1.04 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29501LP036
Reference:
DOA 1000-01 K/A:
295021.K1.04 3.9 K/A:
Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Loss of Shutdown Cooling: Natural circulation CFR: 41.8 to 41.10 PRA: No Level: High Safety Function: 4 Pedigree: Bank History: None Explanation:
A. Incorrect - Initiating a CS pump would flood the cavity and fuel pool into RB vent ducts with no drain path available. Plausible because a CS pump would be able to inject into the RPV with the current conditions.
B. Incorrect - Raise RBCCW flow to shell side of SDC heat exchangers is NOT option to 2C SDC HX since MO 2-3704 is at max setting for two exchanger operation. Plausible because increasing the RBCCW flow is a way to increase cooling.
C. Incorrect - MSL drains are isolated due to MSL plugs being installed.
Plausible because that is a viable path per DOA 1000-01 if the MSL drains were not isolated.
D. Correct - Since 2C SDC loop is aligned to the Fuel Pool Utilizing a natural circulation flow path from Fuel Pool to Reactor Cavity with Fuel Pool Gates removed would provide decay heat removal path immediately on the 2B SDC pump trip.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 125 of 150 08 March 2023 63 ID: 28076 Points: 1.00 Which one of the following identifies (1) a burnable poison that is loaded into the Reactor and (2) describes the purpose of this poison?
A.
(1) Gadolinium (2) Allow more fuel to be loaded into the core B.
(1) Gadolinium (2) Compensate for buildup of fission product poison C.
(1) Samarium (2) Allow more fuel to be loaded into the core D.
(1) Samarium (2) Compensate for buildup of fission product poison Answer:
A Answer Explanation Gadolinium is a burnable poison loaded into the Reactor. Its purpose is to allow more fuel to be loaded into the Reactor.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 126 of 150 08 March 2023 Question 63 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
2.00 System ID:
28076 User-Defined ID:
28076 Cross Reference Number:
Topic:
63 - 292007.K1.01 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: BR07Ir4_Fuel_Depletion Obj 1
Reference:
Generic Fundamentals BR07Ir4_Fuel_Depletion K/A:
292007.K1.01 3.1 K/A:
Fuel Depletion and Burnable Poisons - Define burnable poison and state its use in the reactor CFR: 41.1 PRA: No Level: Memory Pedigree: Bank History: None Explanation:
A. Correct - Gadolinium is a burnable poison loaded into the Reactor.
Its purpose is to allow more fuel to be loaded into the Reactor.
B. Incorrect - The purpose of burnable poisons is not to compensate for building of fission product poisons. Plausible because Reactor operation does simultaneously burnout burnable poisons and create fission product poisons.
C. Incorrect - Samarium is not a burnable poison loaded into the Reactor. Plausible because it is a fission product poison that builds up during Reactor operation.
D. Incorrect - Samarium is not a burnable poison loaded into the Reactor. Plausible because it is a fission product poison that builds up during Reactor operation. The purpose of burnable poisons is not to compensate for building of fission product poisons. Plausible because Reactor operation does simultaneously burnout burnable poisons and create fission product poisons.
REQUIRED
REFERENCE:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 127 of 150 08 March 2023 64 ID: 28079 Points: 1.00 Both units are operating at full power.
Control Room personnel have noticed a smell of smoke in the Control Room It has been determined that the smoke is coming from outside of the Control Room In accordance with DOA 5750-04, SMOKE, NOXIOUS FUMES OR AIRBORNE CONTAMINANTS IN THE CONTROL ROOM, the Control Room will have to be ___(1)___ and must be verified at the
___(2)___ panel.
A.
(1) purged (2) 923-5 B.
(1) purged (2) 2/3-9400-105 C.
(1) isolated and pressurized (2) 923-5 D.
(1) isolated and pressurized (2) 2/3-9400-105 Answer:
D Answer Explanation With smoke coming from outside of the Control Room the proper action is the place CREVs in Isolate and Pressurize. The isolations will be verified at Panel 2/3-9400-105.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 128 of 150 08 March 2023 Question 64 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28079 User-Defined ID:
28079 Cross Reference Number:
Topic:
64 - 290003.A2.06 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE288LN003.8
Reference:
DOA 5750-04 K/A:
290003.A2.06 3.2 / 3.7 K/A:
Ability to (a) predict the impacts of the following on the Control Room Ventilation and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Breaches of control room envelope CFR: 41.5 PRA: No Level: High Safety Function: 9 Pedigree: New History: N/A Explanation:
A. Incorrect - The first part is plausible because purging the Control Room would be correct if the smoke was coming from inside the Control Room. Second part is plausible because Control Room Ventilation is controlled at the 923-5 panel.
B. Incorrect - The first part is plausible because purging the Control Room would be correct if the smoke was coming from inside the Control Room. Second part is correct.
C. Incorrect - First part is correct. Second part is plausible because Control Room Ventilation is controlled at the 923-5 panel.
D. Correct - With smoke coming from outside of the Control Room the proper action is the place CREVs in Isolate and Pressurize. The isolations will be verified at Panel 2/3-9400-105.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 129 of 150 08 March 2023 65 ID: 28081 Points: 1.00 Unit 2 was operating at 30% power, with load ascension in progress, when the following timeline of events occurred:
03:05:00 - Stator Cooling INLET water flow to the Main Generator reached 450 gpm 03:05:03 - DAN 902-7 C-3, TURB STATOR COOLANT RUNBACK, illuminated 03:05:45 - The Unit 2 NSO began reducing reactor power with core flow 03:08:25 - Generator stator amps are observed as 9300 and steady Which of the following describes the additional actions, if any, that would be expected to automatically occur by 03:08:35?
A.
No additional automatic actions occur B.
The standby Stator Coolant pump starts C.
The Main Turbine/Generator trips ONLY D.
The Main Turbine/Generator trips AND the Reactor scrams Answer:
C Answer Explanation When Stator Cooling water inlet flow is <500 gpm, a Stator Runback is received. If Stator amps are NOT
<9121 stator amps within 3.5 minutes, a Turbine trip is initiated.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 130 of 150 08 March 2023 Question 65 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28081 User-Defined ID:
28081 Cross Reference Number:
Topic:
65 - 245000.K3.08 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE253LN001.06
Reference:
DOA 7400-01, DANs 902-7 B-10 & C-3 K/A:
245000.K3.08 3.7 K/A:
Knowledge of the effect that a loss or malfunction of the Main Turbine Generator and Auxiliary Systems will have on the following systems or system parameters: Reactor/turbine pressure regulating system CFR: 41.7 / 45.4 PRA: No Level: High Safety Function: 4 Pedigree: New History: N/A Explanation:
A. Incorrect - The turbine will trip if stator amps are not <9121 amps within 3.5 minutes. Plausible because if amps were <9121 then no additional automatic actions would occur.
B. Incorrect - The standby pump will only AUTO start if the running pump trips or its discharge pressure is <65 psig. Plausible because this is an automatic action that occurs within the Stator Water Cooling system.
C. Correct - When Stator Cooling water inlet flow is <500 gpm, a Stator Runback is received. If Stator amps are NOT <9121 stator amps within 3.5 minutes, a Turbine trip is initiated.
D. Incorrect - The reactor does NOT scram since rated core thermal power was < 38.5%. Plausible because if power were >38.5 then the Bypass Valves would not be able to handle to demand and cause a Reactor scram.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 131 of 150 08 March 2023 66 ID: 28047 Points: 1.00 Unit 2 is operating at rated power when a Scram occurs due to spurious Group I Isolation.
Reactor Power is 5%
Reactor Pressure is 1060 psig and rising slowly ERVs are cycling Which of the following is performed to minimize power transients through changes in the core void fraction during execution of DEOP 400-05, FAILURE TO SCRAM?
A.
Inhibit ADS and initiate IC B.
Verify FWLCS in automatic C.
Initiate IC and open ADSVs to lower RPV pressure to 945 psig D.
Terminate and prevent all RPV injection except boron and CRD to -35 inches Answer:
C Answer Explanation Allowing pressure oscillations can cause significant power transients through changes in the core void fraction. This is controlled via the Iso Condenser and ADSVs, thus not allowing the ADSVs to cycle.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 132 of 150 08 March 2023 Question 66 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28047 User-Defined ID:
28047 Cross Reference Number:
Topic:
66 - 295025.K1.07 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29501LK025
Reference:
DEOP 100, DEOP 400-5, DEOP TB Vol 2 K/A:
295025.K1.07 4.2 K/A:
Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to High Reactor Pressure: Pressure control strategies.
CFR: 41.8 to 41.10 Safety Function: 3 Level: High Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because ADS is inhibited and Iso Condenser is initiated, ADS in inhibit does not lower pressure.
B. Incorrect - Plausible because if all rods were in and FWLC in auto, setpoint setdown would lower level to minimize the impact of shrink and swell.
C. Correct - Allowing pressure oscillations can cause significant power transients through changes in the core void fraction. This is controlled via the Iso Condenser and ADSVs, thus not allowing the ADSVs to cycle.
D. Incorrect - Plausible because if power was greater than 6% then Terminate and Prevent to less than -35 inches would be required to uncover feedwater spargers.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 133 of 150 08 March 2023 67 ID: 7684 Points: 1.00 Unit 2 is at rated power Annunciator 902-5 G-3, RPIS SYS INOP, is alarming Based on these conditions, control rods can.....
A.
be moved IF the RWM is bypassed.
B.
NOT be moved due to a Select Block.
C.
NOT be moved due to a timer malfunction.
D.
NOT be moved due to a rod withdrawal block.
Answer:
B Answer Explanation This alarm is indicative of a loss of RPIS 24 VDC power supply and as a result of that a Select Block is inserted.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 134 of 150 08 March 2023 Question 67 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
2.00 System ID:
7684 User-Defined ID:
7684 Cross Reference Number:
LI Topic:
67 - 201002.K6.04 Num Field 1:
0.00 Num Field 2:
0.00 Text Field:
Comments:
Objective: 201LN002-12
Reference:
DAN 902(3)-5 G-3 K/A:
201002.K6.04 3.5 K/A:
Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Reactor Manual Control System: RPIS CFR: 41.7 / 45.7 PRA: No Level: Memory Safety Function: 1 Pedigree: Bank History: None Explanation:
A. Incorrect - Plausible because if a RWM Block were to be in then this would be correct B. Correct - This alarm is indicative of a loss of RPIS 24 VDC power supply and as a result of that a Select Block is inserted.
C. Incorrect - Plausible because a timer malfunction is another cause of a Select Block.
D. Incorrect - Plausible because a withdrawal block would cause a rod to be able to not be moved out, but loss of power to RPIS does not cause a withdrawal block.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 135 of 150 08 March 2023 68 ID: 28088 Points: 1.00 Unit 2 is at 80% power.
Rod J-6 is being moved from position 36 to 48 Annunciator 902-5 F-3, ROD DRIVE TEMP HI, alarms Temperature on TR-2-340-16, CONTROL ROD DRIVE TEMP, indicates 360°F for J-6 and steady This could cause ___(1)___.
The required action is ___(2)___.
A.
- 1) slow scram time
- 2) drive the rod to position 00 and take OOS B.
- 1) slow scram time
C.
- 1) fast scram time
- 2) drive the rod to position 00 and take OOS D.
- 1) fast scram time
Answer:
B Answer Explanation Per DAN 902(3)-5 F-3 Rod Drive Temp Hi, Rod movement may cause a high temperature alarm if a CRD was already operating near (within 50°F) of its alarm setpoint. Setpoint is 250°F. If temperature is between 350 and 400°F the proper action is to make preparations to scram test the CRD as soon as possible. WHEN rod temperatures are > 400°F, THEN application of scram time penalties will always result in Tech Spec Slow rods per DOS 0300-06. Elevated temperatures result in slower scram times.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 136 of 150 08 March 2023 Question 68 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
4 Difficulty:
4.00 System ID:
28088 User-Defined ID:
28088 Cross Reference Number:
Topic:
68 - 201003.A1.04 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE201LN001.10
References:
DAN 902(3)-5 F-3, DOS 0300-06 K/A:
201003.A1.04 3.0 K/A:
Ability to predict and/or monitor changes in parameters associated with operation of Control Rod and Drive Mechanism including: CRD mechanism temperature.
CFR: 41.1-6/45.1-6 Safety Function: 1 Level: High Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because part 1 is correct. Part 2 would be correct if the rod was declared inoperable due to scram time not meeting tech spec requirement.
B. Correct - Per DAN 902(3)-5 F-3 Rod Drive Temp Hi, Rod movement may cause a high temperature alarm if a CRD was already operating near (within 50 F) of its alarm setpoint. Setpoint is 250°F.
If temperature is between 350 and 400°F the proper action is to make preparations to scram test the CRD as soon as possible.
WHEN rod temperatures are > 400°F, THEN application of scram time penalties will always result in Tech Spec Slow rods per DOS 0300 06. Elevated temperatures result in slower scram times.
C. Incorrect - Plausible because must determine if higher temperature would cause faster or slower rod movement. Part 2 would be correct if the rod was declared inoperable due to scram time not meeting tech spec requirement.
D. Incorrect - Plausible because must determine if higher temperature would cause faster or slower rod movement. Part 2 is correct.
Required reference: None None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 137 of 150 08 March 2023 69 ID: 28089 Points: 1.00 (1) What is a possible consequence of operating a LPCI pump at shutoff head for a prolonged period of time?
(2) This consequence can be avoided by ensuring proper operation of....
A.
- 1) overheating of pump
- 2) minimum flow valve B.
- 1) overheating of pump
- 2) keep-fill system C.
- 1) excessive motor current
- 2) minimum flow valve D.
- 1) excessive motor current
- 2) keep-fill system Answer:
A Answer Explanation Operating a centrifugal pump, such as a LPCI pump, at shutoff head can cause elevated pump temperatures due to inadequate flow. Ensuring proper operation of the minimum flow valve helps to avoid this overheating.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 138 of 150 08 March 2023 Question 69 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
1.00 System ID:
28089 User-Defined ID:
28089 Cross Reference Number:
Topic:
69 - 293006.K1.18 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: BT06Ir4_Statics_Dynamics Obj 34
Reference:
Generic Fundamentals BT06Ir4_Statics_Dynamics K/A:
293006.K1.18 2.7 K/A:
Fluid Statics and Dynamics - Explain how operating a centrifugal pump at shutoff head may cause overheating and describe the methods used to avoid overheating.
CFR:
41.14 PRA: No Level: Memory Pedigree: Bank History: None Explanation:
A. Correct - Operating a centrifugal pump, such as a LPCI pump, at shutoff head can cause elevated pump temperatures due to inadequate flow. Ensuring proper operation of the minimum flow valve helps to avoid this overheating.
B. Incorrect - Ensuring proper operation of the minimum flow valve helps to avoid this overheating. Plausible because the keep-full system must be operated properly to avoid other unwanted conditions with LPCI, such as water hammer and degraded flow rates.
C. Incorrect - Motor current at shutoff head conditions is low, not high.
Plausible because this is correct for runout conditions.
D. Incorrect - Motor current at shutoff head conditions is low, not high.
Plausible because this is correct for runout conditions. Ensuring proper operation of the minimum flow valve helps to avoid this overheating. Plausible because the keep-full system must be operated properly to avoid other unwanted conditions with LPCI, such as water hammer and degraded flow rates.
Required
References:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 139 of 150 08 March 2023 70 ID: 28090 Points: 1.00 The nil-ductility transition temperature is the temperature.....
A.
in which the distortion of the lattice usually occurs.
B.
below which the probability of brittle fracture significantly increases.
C.
when metal splits along certain crystal planes and can rapidly occur.
D.
when mechanisms operating at high temperatures that are cooled by cold fluid break.
Answer:
B
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 140 of 150 08 March 2023 Answer Explanation Nil-ductility transition temperature is the temperature below which metal fails by brittle fracture....
Question 70 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
1.00 System ID:
28090 User-Defined ID:
28090 Cross Reference Number:
Topic:
70 - 293010.K1.02 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: BT10Ir4_Thermal_Limits Obj 8.a
Reference:
Generic Fundamentals BT10Ir4_Thermal_Limits K/A:
293010.K1.02 2.7 K/A:
Brittle Fracture and Vessel Thermal Stress - State the definition of Nil-Ductility Transition Temperature CFR: 41.14 PRA: No Level: Memory Safety Function: N/A Pedigree: Bank History: None Explanation:
A. Incorrect - The distortion of the lattice occurs at high temperatures.
B. Correct - Nil-ductility transition temperature is the temperature below which metal fails by brittle fracture.
C. Incorrect - This is the definition Cleavage Fracture.
D. Incorrect - This is the definition of Thermal Stress.
Required
References:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 141 of 150 08 March 2023 71 ID: 28092 Points: 1.00 Unit 2 was operating at near rated power when an instrument failure requires an in-plant manipulation which will affect reactivity.
Which of the following describes the MINIMUM requirement to perform this evolution?
Communication between the Control Room and ______, in the plant.
A.
any SRO ONLY B.
any SRO with a qualified QNE C.
any ACTIVE LICENSED Operator ONLY D.
an ACTIVE LICENSED Operator with no restriction which would prohibit solo operations Answer:
D Answer Explanation Local Operation of equipment that affects reactivity requires a Active License (RO or SRO), but with the additional caveat of no restrictions that would prohibit solo operations (Per HR-AA-07-101, No Solo Operation: License restriction that prohibits solo operation in the Main Control Room or other specified controlled areas).
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 142 of 150 08 March 2023 Question 71 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28092 User-Defined ID:
28092 Topic:
71 - Generic 2.1.4 Comments:
Objective: 20200LN001.08
Reference:
DOP 0202-16, HR-AA-07-101 K/A:
Generic 2.1.4 3.3 K/A:
Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10 CFR Part 55.
Safety Function: N/A CFR: 41.10/43.2 Level: Memory Pedigree: Bank History: 2009 NRC Explanations:
A. Incorrect - Local Operation of equipment that affects reactivity requires an Active License (RO or SRO) with no restrictions that would prohibit solo operations (Per HR-AA-07-101, No Solo Operation: License restriction that prohibits solo operation in the Main Control Room or other specified controlled areas). Plausible because SRO can direct operations in the field that affect reactivity but must have an active license and no restrictions that would prohibit solo operations.
B. Incorrect - Local Operation of equipment that affects reactivity requires an Active License, but with the additional caveat of no restrictions that would prohibit solo operations (Per HR-AA-07-101, No Solo Operation: License restriction that prohibits solo operation in the Main Control Room or other specified controlled areas). QNE = Qualified Nuclear Engineer. Plausible because the answer is partially correct. Missing the Active License and no solo restrictions.
C. Incorrect - Local Operation of equipment that affects reactivity requires an Active License (RO or SRO), but with the additional caveat of no restrictions that would prohibit solo operations (Per HR-AA-07-101, No Solo Operation:
License restriction that prohibits solo operation in the Main Control Room or other specified controlled areas). Plausible because the answer is partially correct. Missing the no solo restrictions.
D. Correct - Local Operation of equipment that affects reactivity requires an Active License (RO or SRO), but with the additional caveat of no restrictions that would prohibit solo operations (Per HR-AA-07-101, No Solo Operation:
License restriction that prohibits solo operation in the Main Control Room or other specified controlled areas).
Required
References:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 143 of 150 08 March 2023 72 ID: 28127 Points: 1.00 A surveillance is scheduled to be performed this shift that will require authorization of a fire protection impairment permit. The Fire Marshall is not onsite.
(1) Whose authorization is REQUIRED to perform the surveillance?
(2) Whose authorization is REQUIRED for the fire protection impairment permit?
A.
(1) Unit Supervisor (2) Unit Supervisor B.
(1) Unit Supervisor (2) Work Execution Center Supervisor C.
(1) Work Execution Center Supervisor (2) Unit Supervisor D.
(1) Work Execution Center Supervisor (2) Work Execution Center Supervisor Answer:
B Answer Explanation The Unit Supervisor will AUTHORIZE testing, surveillances, outages, and maintenance on all equipment and systems affecting plant safety or place the plant in a degraded mode. The WEC Supervisor will authorize fire protection impairment permits in the absence of the Fire Marshall.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 144 of 150 08 March 2023 Question 72 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
1.00 System ID:
28127 User-Defined ID:
28127 Cross Reference Number:
Topic:
72 - Generic.2.2.12 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 29501LK063
Reference:
OP-AA-101-111 K/A:
Generic.2.2.12 3.7 / 4.1 K/A:
Knowledge of surveillance procedures CFR: 41.10 / 43.2 / 45.13 PRA: No Level: Memory Safety Function: N/A Pedigree: New History: N/A Explanation:
A. Incorrect - First part is correct. The WEC Supervisor will authorize fire protection impairment permits in the absence of the Fire Marshall. Plausible because the Unit Supervisor is required to authorize work.
B. Correct -The Unit Supervisor will AUTHORIZE testing, surveillances, outages, and maintenance on all equipment and systems affecting plant safety or place the plant in a degraded mode. The WEC Supervisor will authorize fire protection impairment permits in the absence of the Fire Marshall.
C. Incorrect - The Unit Supervisor is required to authorize the performance of surveillances. Plausible because the WEC Supervisor implements the daily station work schedule and shall direct and assign those operators assigned to them. The WEC Supervisor will authorize fire protection impairment permits in the absence of the Fire Marshall. Plausible because the Unit Supervisor is required to authorize work.
D. Incorrect - The Unit Supervisor is required to authorize the performance of surveillances. Plausible because the WEC Supervisor implements the daily station work schedule and shall direct and assign those operators assigned to them. Second part is correct.
REQUIRED
REFERENCES:
None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 145 of 150 08 March 2023 73 ID: 28096 Points: 1.00 Unit 2 is operating at 100% power.
A pressure transient occurs causing DAN 902-5 H-5 RPV PRESS HI, to alarm.
The increase in pressure would result in a(an) ___(1)___ in Critical Power.
This would mean ___(2)___ bundle power is necessary to cause transition boiling to occur.
A.
(1) increase (2) less B.
(1) increase (2) more C.
(1) decrease (2) less D.
(1) decrease (2) more Answer:
C Answer Explanation An increase in pressure would result in a decrease in critical power. As pressure increases, the latent heat of vaporization (hfg) decreases, requiring less bundle power necessary to cause transition boiling to occur.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 146 of 150 08 March 2023 Question 73 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
2 Difficulty:
1.00 System ID:
28096 User-Defined ID:
28096 Cross Reference Number:
Topic:
73 - 293009.K1.42 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: BT09Ir4_Thermal_Limits Obj 35.c
Reference:
Generic Fundamentals BT09Ir4_Thermal_Limits K/A:
293009.K1.42 3.3 K/A:
Core Thermal Limits - For the following plant operating or accident conditions, identify which of the three core thermal limits are most limiting: Increase in reactor pressure.
CFR: 41.14 PRA: No Level: High Pedigree: New History: N/A Explanation:
A. Incorrect - This is the basis behind MFLCPR. Plausible because this is another thermal limit ratio given on the OD-20 edit.
B. Incorrect - This is the basis behind MFLCPR. Plausible because this is another thermal limit ratio given on the OD-20 edit.
C. Correct - An increase in pressure would result in a decrease in critical power. As pressure increases, the latent heat of vaporization (hfg) decreases, requiring less bundle power necessary to cause transition boiling to occur.
D. Incorrect - This is the basis behind MFLCPR. Plausible because this is another thermal limit ratio given on the OD-20 edit.
REQUIRED
REFERENCES:
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 147 of 150 08 March 2023 74 ID: 28100 Points: 1.00 A transient has occurred on Unit 2.
Reactor Pressure is 600 psig and lowering at 20 psig/min Drywell Pressure is 12 psig and rising 0.5 psig/min Reactor Water Level -150 inches and lowering 10 inches/min In order to start Torus and Drywell sprays per the Hard Card for LPCI/CCSW OPERATION DURING TRANSIENT SITUATIONS, the 316 and ___(1)___ keylock switches must be placed in MANUAL /
MANUAL OVERRD.
6 minutes later, ___(2)___ are running.
A.
(1) 317 (2) No Sprays B.
(1) 317 (2) Torus and Drywell Sprays C.
(1) 318 (2) No Sprays D.
(1) 318 (2) Torus and Drywell Sprays Answer:
C Answer Explanation With the conditions listed LPCI would have started. Using the Hard Card the first step is to place the 316 and 318 keylock switches to MANUAL/MANUAL OVERRD. Do to the trends, 6 minutes later reactor pressure would be 480 psig so level correction would not be needed. Reactor water level would be -210 inches (less than 2/3 core height). Torus and Drywell sprays will isolate unless the 317 switches are taken to MANUAL OVERRD with US permission.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 148 of 150 08 March 2023 Question 74 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
4 Difficulty:
3.00 System ID:
28100 User-Defined ID:
28100 Cross Reference Number:
Topic:
74 - 230000.K6.08 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: DRE203LN001-11
References:
UFSAR 7.4.1, UFSAR 6.2, DOP 1500-02 K/A:
230000.K6.08 3.3 K/A:
Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the RHR-LPCI: Torus/Suppression Pool Spray Mode: Nuclear boiler instrumentation.
CFR: 41.7/45.7 PRA: No Safety Function: 5 Level: High Pedigree: New History: N/A Explanation:
A. Incorrect - Plausible because the 317 interlock switches could be used with US permission but not needed at current level and pressure. Part 2 is correct.
B. Incorrect - Plausible because the 317 interlock switches could be used with US permission but not needed at current level and pressure. Part 2 must determine the level and pressure based on current trends and understand 2/3 core height logic.
C. Correct - With the conditions listed LPCI would have started. Using the Hard Card the first step is to place the 316 and 318 keylock switches to MANUAL/MANUAL OVERRD. Do to the trends, 6 minutes later reactor pressure would be 480 psig so level correction would not be needed. Reactor water level would be -210 inches (less than 2/3 core height). Torus and Drywell sprays will isolate unless the 317 switches are taken to MANUAL OVERRD with US permission.
D. Incorrect - Plausible because Part 1 is correct. Part 2 must determine the level and pressure based on current trends and understand 2/3 core height logic.
Required references: None None
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 149 of 150 08 March 2023 75 ID: 28102 Points: 1.00 Unit 2 was operating at near rated power when the Turbine tripped.
The Generator has NOT tripped The Generator Field Breaker is Open The team is required to open Generator GCBs ___(1)___ from the ___(2)___ panel.
A.
(1) After 90 seconds (2) 902-8 B.
(1) IMMEDIATELY (2) 923-2 C.
(1) IMMEDIATELY (2) 902-8 D.
(1) After 90 seconds (2) 923-2 Answer:
B Answer Explanation With the generator still on line and the Field Breaker open then the GCBs should be IMMEDIATELY opened from the 923-2 panel per DOA 5600-01 immediate operator action 3.
EXAMINATION ANSWER KEY 22-1 (2023-301) NRC Exam - RO ILT EXAM Page: 150 of 150 08 March 2023 Question 75 Info Question Type:
Multiple Choice Status:
Active Always select on test?
No Authorized for practice?
No Points:
1.00 Time to Complete:
3 Difficulty:
3.00 System ID:
28102 User-Defined ID:
28102 Cross Reference Number:
Topic:
75 - 295005.G.2.4.12 Num Field 1:
Num Field 2:
Text Field:
Comments:
Objective: 24501LK017
Reference:
DOA 5600-01, DOP 6400-13, DGP 02-01 K/A:
295005.G.2.4.12 4.0 / 4.3 K/A:
Main Turbine Generator Trip - Knowledge of operating crew responsibilities during emergency and abnormal operations Safety Function: 3 CFR: 41.10 / 45.12 PRA: No Level: Memory Pedigree: Bank History: 19-1 NRC Explanation:
A. Incorrect - (1) With the generator still on line the GCBs must be opened on the 923-2 panel. (2) This action should be taken immediately. Plausible because (1) GCBs 1-2 and 1-7 are operated from the 902-8 panel under normal circumstances, and (2) the operate is directed to wait 90 seconds if the field breaker remains closed.
B. Correct - With the generator still on line and the Field Breaker open then the GCBs should be IMMEDIATELY opened from the 923-2 panel per DOA 5600-01 immediate operator action 3.
C. Incorrect - IF Generator fails to trip AND Field Breaker opens, THEN IMMEDIATELY open the Generator GCBs from Panel 923-2.
Plausible (1) because part 1 is correct and (2) GCBs 1-2 and 1-7 are operated from the 902-8 panel under normal circumstances.
D. Incorrect - IF Generator fails to trip AND Field Breaker opens, THEN IMMEDIATELY open the Generator GCBs from Panel 923-2.
Plausible (1) because if the Field Breaker remains closed part 1 would be correct. (2) Part 2 is correct.
REQUIRED
REFERENCES:
TRM Control Rod Block Instrumentation 3.3.a Dresden 2 and 3 3.3.a-1 Revision 2 3.3 INSTRUMENTATION 3.3.a Control Rod Block Instrumentation TLCO 3.3.a The control rod block instrumentation for each Function in Table T3.3.a-1 shall be OPERABLE.
APPLICABILITY:
According to Table T3.3.a-1.
ACTIONS
NOTE-------------------------------------------
Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A. For Functions 1.a, 1.b, 1.c, 1.d, 2.a, 2.b, 2.c, 3.a, 3.b, 3.c, and 3.d, one required channel inoperable.
A.1 Restore inoperable channel to OPERABLE status.
7 days B. For Functions 4.a and 4.b one or more required channels inoperable.
B.1 Place inoperable channel(s) in trip.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (continued)
TRM Control Rod Block Instrumentation 3.3.a Dresden 2 and 3 3.3.a-2 Revision 0 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. For Functions 1.a, 1.b, 1.c, 1.d, 2.a, 2.b, 2.c, 3.a, 3.b, 3.c, and 3.d, two or more required channels inoperable.
OR Required Action and associated Completion Time of Condition A not met.
C.1
TRM Control Rod Block Instrumentation 3.3.a Dresden 2 and 3 3.3.a-3 Revision 95 SURVEILLANCE REQUIREMENTS
NOTES---------------------------------------
- 1.
Refer to Table T3.3.a-1 to determine which TSRs apply to each Control Rod Block Function.
- 2.
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains Control Rod Block capability.
SURVEILLANCE FREQUENCY TSR 3.3.a.1
NOTE--------------------
Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.
Perform CHANNEL FUNCTIONAL TEST.
7 days TSR 3.3.a.2 Perform CHANNEL FUNCTIONAL TEST.
NOTE--------------------
Function 4.a Only Perform CHANNEL FUNCTIONAL TEST.
92 days 12 months TSR 3.3.a.3 Calibrate the trip units.
NOTE--------------------
Function 4.a Only Calibrate the trip units.
92 days 12 months (continued)
TRM Control Rod Block Instrumentation 3.3.a Dresden 2 and 3 3.3.a-4 Revision 80 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.3.a.4
NOTES-------------------------
- 1. For Function 1.d, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.
- 2. Neutron detectors are excluded.
Perform CHANNEL CALIBRATION.
184 days TSR 3.3.a.5 Perform CHANNEL FUNCTIONAL TEST.
24 months TSR 3.3.a.6
NOTE--------------------------
- 1. Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.
- 2. Neutron detectors are excluded.
Perform CHANNEL CALIBRATION.
24 months TSR 3.3.a.7
NOTE--------------------
Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.
Perform CHANNEL FUNCTIONAL TEST.
31 days
TRM Control Rod Block Instrumentation 3.3.a Dresden 2 and 3 3.3.a-5 Revision 80 Table T3.3.a-1 (page 1 of 2)
Control Rod Block Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- 1. Average Power Range Monitors
- a. Flow Biased Neutron Flux High 1
4 TSR 3.3.a.2 TSR 3.3.a.4
< 0.56W +
55.4%(a) RTP and < 109.9%
- b. Inoperative 1, 2, 5(b) 4 TSR 3.3.a.2 N.A.
- c. Downscale 1
4 TSR 3.3.a.2 TSR 3.3.a.4
> 3.4 % of RTP
- d. Startup Neutron Flux High 2,5(b) 4 TSR 3.3.a.2 TSR 3.3.a.4
< 14.1 % of RTP
- 2. Source Range Monitors
- a. Detector not full in 2(c)(e) 3 TSR 3.3.a.7 TSR 3.3.a.6 N.A.
5(e) 2 TSR 3.3.a.7 N.A.
- b. Upscale 2(d) 3 TSR 3.3.a.7 TSR 3.3.a.6
< 3.0 x 105 cps 5
2 TSR 3.3.a.7 TSR 3.3.a.6
< 3.0 x 105 cps
- c. Inoperative 2(d) 3 TSR 3.3.a.7 N.A.
5 2
TSR 3.3.a.7 N.A.
(continued)
(a) Allowable Value is < 0.56W + 51.2% RTP and < 109.9% RTP when reset for single loop operation per Technical Specification LCO 3.4.1, "Recirculation Loops Operating."
(b) Required to be OPERABLE only during SHUTDOWN MARGIN demonstrations performed per Technical Specification LCO 3.10.7.
(c) With the Intermediate Range Monitor (IRM) channels are on range 2 or below.
(d) With IRM channels on range 7 or below.
(e) With detector count rate < 100 cps.
TRM Control Rod Block Instrumentation 3.3.a Dresden 2 and 3 3.3.a-6 Revision 80 Table T3.3.a-1 (page 2 of 2)
Control Rod Block Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- a. Detector not full in 2, 5(h) 6 TSR 3.3.a.1 TSR 3.3.a.6 N.A.
- b. Upscale 2, 5(h) 6 TSR 3.3.a.1 TSR 3.3.a.6
< 112/125 of full scale
- c. Inoperative 2, 5(h) 6 TSR 3.3.a.1 N.A.
- d. Downscale 2
(f), 5(h) 6 TSR 3.3.a.1 TSR 3.3.a.6
> 5/125 of full scale
- a. Water Level High (Unit 2) 1, 2, 5 (g) 1 per bank TSR 3.3.a.2 TSR 3.3.a.6
< 28.1 gal Water Level High (Unit 3) 1, 2, 5 (g) 1 per bank TSR 3.3.a.2 TSR 3.3.a.3 TSR 3.3.a.6
< 23.4 gal
- b. Scram Discharge Volume Switch in Bypass 5
(g) 1 TSR 3.3.a.5 N.A.
(f) With IRM channels on range 2 or higher.
(g) With two or more control rods withdrawn. Not applicable to control rods removed per Technical Specification LCO 3.10.4, "Single Control Rod Drive Removed Refueling," or LCO 3.10.5, "Multiple Control Rod Withdrawal Refueling."
(h) With any control rod withdrawn from a core cell containing one or more fuel assemblies.
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-1 Amendment No. 260/253 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.
APPLICABILITY:
According to Table 3.3.1.1-1.
ACTIONS
NOTES -----------------------------------
- 1.
Separate Condition entry is allowed for each channel.
- 2.
When Functions 2.b and 2.c channels are inoperable due to the calculated power exceeding the APRM output by more than 2% RTP while operating at 25% RTP, entry into associated Conditions and Required Actions may be delayed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required channels inoperable.
A.1 Place channel in trip.
OR A.2 Place associated trip system in trip.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 12 hours B. One or more Functions with one or more required channels inoperable in both trip systems.
B.1 Place channel in one trip system in trip.
OR B.2 Place one trip system in trip.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 6 hours (continued)
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-2 Amendment No. 239/232 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions with RPS trip capability not maintained.
C.1 D. Required Action and associated Completion Time of Condition A, B, or C not met.
D.1 E. As required by Required Action D.1 and referenced in Table 3.3.1.1-1.
E.1 Reduce THERMAL POWER to < 38.5% RTP 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> F. As required by Required Action D.1 and referenced in Table 3.3.1.1-1.
F.1 Be in MODE 2.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (continued)
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-3 Amendment No. 185/180 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. As required by Required Action D.1 and referenced in Table 3.3.1.1-1.
G.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> H. As required by Required Action D.1 and referenced in Table 3.3.1.1-1.
H.1
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-4 Amendment No. 260/253 SURVEILLANCE REQUIREMENTS
NOTES -----------------------------------
- 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
- 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains RPS trip capability.
SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.2
NOTE-------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER 25% RTP.
Verify the calculated power does not exceed the average power range monitor (APRM) channels by greater than 2% RTP while operating at 25% RTP.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.3 Adjust the channel to conform to a calibrated flow signal.
In accordance with the Surveillance Frequency Control Program (continued)
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-5 Amendment No. 237/230 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.4
NOTE-------------------
Not required to be performed when entering MODE 2 from MODE 1 until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering MODE 2.
Perform CHANNEL FUNCTIONAL TEST.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.5 Perform a functional test of each RPS automatic scram contactor.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.6 Verify the source range monitor (SRM) and intermediate range monitor (IRM) channels overlap.
Prior to fully withdrawing SRMs SR 3.3.1.1.7
NOTE-------------------
Only required to be met during entry into MODE 2 from MODE 1.
Verify the IRM and APRM channels overlap.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST.
In accordance with the Surveillance Frequency Control Program (continued)
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-6 Amendment No. 237/230 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.9 Calibrate the local power range monitors.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.10 Deleted.
SR 3.3.1.1.11 Perform CHANNEL FUNCTIONAL TEST.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.12 Calibrate the trip units.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.13 Perform CHANNEL CALIBRATION.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.14 Verify Turbine Stop ValveClosure and Turbine Control Valve Fast Closure, Trip Oil PressureLow Functions are not bypassed when THERMAL POWER is 38.5% RTP.
In accordance with the Surveillance Frequency Control Program (continued)
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-7 Amendment No. 237/230 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.15
NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering MODE 2.
- 3. For Function 2.b, not required for the flow portion of the channels.
Perform CHANNEL CALIBRATION.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.16 Perform CHANNEL FUNCTIONAL TEST.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.17
NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For Function 1.a, not required to be performed when entering MODE 2 from MODE 1 until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering MODE 2.
Perform CHANNEL CALIBRATION.
In accordance with the Surveillance Frequency Control Program (continued)
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-8 Amendment No. 237/230 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.18 Perform LOGIC SYSTEM FUNCTIONAL TEST.
In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.19
NOTES------------------
Neutron detectors are excluded.
Verify the RPS RESPONSE TIME is within limits.
In accordance with the Surveillance Frequency Control Program
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-9 Amendment No. 237/230 Table 3.3.1.1-1 (page 1 of 3)
Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- a. Neutron FluxHigh 2
3 G
SR 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.17 SR 3.3.1.1.18 121/125 divisions of full scale 5(a) 3 H
SR 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.17 SR 3.3.1.1.18 121/125 divisions of full scale
- b. Inop 2
5(a) 3 3
G H
SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.18 SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.18 NA NA
- 2. Average Power Range Monitors
- a. Neutron FluxHigh, Setdown 2
2 G
SR 3.3.1.1.1 SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.7 SR 3.3.1.1.9 SR 3.3.1.1.15 SR 3.3.1.1.18 17.1% RTP
- b. Flow Biased Neutron FluxHigh 1
2 F
SR 3.3.1.1.1 SR 3.3.1.1.2 SR 3.3.1.1.3 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19 0.56 W 67.4% RTP and 122% RTP(b)
(continued)
(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.
(b) 0.56 W + 63.2% and 118.5% RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating."
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-10 Amendment No. 239/232 Table 3.3.1.1-1 (page 2 of 3)
Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- 2. Average Power Range Monitors (continued)
- c. Fixed Neutron Flux-High 1
2 F
SR 3.3.1.1.1 SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.18 SR 3.3.1.1.19 122% RTP
- d. Inop 1,2 2
G SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.18 NA
- 3. Reactor Vessel Steam Dome PressureHigh 1,2 2
G SR 3.3.1.1.1 SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19 1045 psig
- 4. Reactor Vessel Water LevelLow 1,2 2
G SR 3.3.1.1.1 SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19 2.65 inches
- 5. Main Steam Isolation ValveClosure 1
8 F
SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19 9.5% closed
- 6. Drywell PressureHigh 1,2 2
G SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18 SR 3.3.1.1.19 1.94 psig (continued)
RPS Instrumentation 3.3.1.1 Dresden 2 and 3 3.3.1.1-11 Amendment No. 239/232 Table 3.3.1.1-1 (page 3 of 3)
Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- 7. Scram Discharge Volume Water LevelHigh
- a. Thermal Switch (Unit 2)
Level Indicating Switch (Unit 3) 1,2 2
G SR 3.3.1.1.1(c)
SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12(c)
SR 3.3.1.1.17 SR 3.3.1.1.18 37.9 gallons (Unit 2) 38.7 gallons (Unit 3) 5(a) 2 H
SR 3.3.1.1.1(c)
SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12(c)
SR 3.3.1.1.17 SR 3.3.1.1.18 37.9 gallons (Unit 2) 38.7 gallons (Unit 3)
- b. Differential Pressure Switch 1,2 2
G SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 37.9 gallons (Unit 2) 38.7 gallons (Unit 3) 5(a) 2 H
SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 37.9 gallons (Unit 2) 38.7 gallons (Unit 3)
- 8. Turbine Stop Valve-Closure 38.5% RTP 4
E SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.14 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19 9.5% closed
- 9. Turbine Control Valve Fast Closure, Trip Oil PressureLow 38.5% RTP 2
E SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.14 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19 466 psig
- 10. Turbine Condenser VacuumLow 1
2 F
SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18 SR 3.3.1.1.19 20.5 inches Hg vacuum
- 11. Reactor Mode Switch Shutdown Position 1,2 1
G SR 3.3.1.1.16 SR 3.3.1.1.18 NA 5(a) 1 H
SR 3.3.1.1.16 SR 3.3.1.1.18 NA
- 12.
Manual Scram 1,2 1
G SR 3.3.1.1.8 SR 3.3.1.1.18 NA 5(a) 1 H
SR 3.3.1.1.8 SR 3.3.1.1.18 NA (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.
(c) Specified SR performance only required for Unit 3.
Dresden Annex Exelon Nuclear December 2021 DR 2-1 EP-AA-1004 Addendum 3 (Revision 13)
HOT MATRIX HOT MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Levels / Radiological Effluents Radiological Effluents RG1 Release of gaseous radioactivity 1 2 3 4 5 D resulting in offsite dose greater than 1,000 mRem TEDE or 5,000 mRem thyroid CDE.
Emergency Action Level (EAL):
Notes:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
- 1.
The sum of readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 2.05 E+09 uCi/sec for > 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 2.
Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
- a.
> 1000 mRem TEDE.
- b.
> 5000 mRem CDE Thyroid.
- 3. Field survey results at or beyond the site boundary indicate EITHER:
- a. Gamma (closed window) dose rates >1000 mR/hr are expected to continue for > 60 minutes.
- b. Analyses of field survey samples indicate
> 5000 mRem CDE Thyroid for 60 minutes of inhalation.
RS1 Release of gaseous radioactivity 1 2 3 4 5 D resulting in offsite dose greater than 100 mRem TEDE or 500 mRem thyroid CDE.
Emergency Action Level (EAL):
Notes:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
- 1. The sum of readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 2.05 E+08 uCi/sec for > 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 2.
Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
- a.
> 100 mRem TEDE.
- b.
> 500 mRem CDE Thyroid.
- 3.
Field survey results at or beyond the site boundary indicate EITHER:
- a.
Gamma (closed window) dose rates >100 mR/hr are expected to continue for > 60 minutes.
- b.
Analyses of field survey samples indicate
> 500 mRem CDE Thyroid for 60 minutes of inhalation.
RA1 Release of gaseous or liquid 1 2 3 4 5 D radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.
Emergency Action Level (EAL):
Notes:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
The pre-calculated effluent monitor values presented in EAL
- 1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
- 1. The sum of readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 2.05 E+07 uCi/sec for > 15 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 2. Dose assessment using actual meteorology indicates doses at or beyond the site boundary of EITHER:
- a. > 10 mRem TEDE.
- b. > 50 mRem CDE Thyroid.
- 3. Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than EITHER of the following at or beyond the site boundary
- a. 10 mRem TEDE for 60 minutes of exposure.
- b. 50 mRem CDE Thyroid for 60 minutes of exposure.
- 4. Field survey results at or beyond the site boundary indicate EITHER:
- a. Gamma (closed window) dose rates > 10 mR/hr are expected to continue for > 60 minutes.
- b. Analyses of field survey samples indic ate
> 50 mRem CDE Thyroid for 60 minutes of inhalation.
RU1 Any release of gaseous or liquid 1 2 3 4 5 D radioactivity to the environment greater than 2 times the ODCM for 60 minutes or longer.
Emergency Action Level (EAL):
Notes:
The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
- 1. Reading on ANY of the following effluent monitors
> 2 times alarm setpoint established by a current radioactive release discharge permit for 60 minutes.
Radwaste Effluent Monitor 2/3-2001-948 OR Discharge Permit specified monitor OR
- 2. The sum of readings on the Unit 2/3 Rx Bldg and Unit 2/3 Chimney SPINGs > 2.34 E+05 uCi/sec for > 60 minutes (as determined by DOP 1700-10 or PPDS - Total Noble Gas Release Rate).
- 3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 2 times ODCM Limit with a release duration of > 60 minutes.
Modes:
1 - Power Operation 2 - Startup 3 - Hot Shutdown 4 - Cold Shutdown 5 - Refueling D - Defueled HOT MATRIX HOT MATRIX
ENTRY CONDITIONS (MODES 1, 2, 3)
RPV Water Level Instruments A
RPV water level instruments may be unreliable due to boiling in the instrument runs when drywell or reactor building temperatures near the instrument runs are above Fig B, RPV Saturation Temperature.
An RPV water level instrument may not be used if the instrument reads at or below its Minimum Usable Indicating Level (Detail C).
Fuel Zone
(+60" to -340")
Medium/Narrow Range All Rx Bldg Temps
(+60" to -60")
Medium/Narrow Range Rx Bldg Temps 171°F or less
(+60" to -60")
Wide Range
(+330" to -70")
-294
-295
-295
-296
-296
-297
-37
-38
-39
-40
-42
-42
-60
-60
-60
-60
-60
-60
-68
-50
-20 19 68 106 32 101 201 301 401 501 to to to to to to 100 200 300 400 500 559 Drywell Temperature (°F)
TR 2(3)-1340-1, Point 9 or 10 C
Minimum Usable Indicating Levels B
RPV Saturation Temperature 0
200 RPV Pressure (psig)
Drywell Temperature (°F)
TR 2(3)-1340-1, Point 9 or 10 OR Reactor Building Temperature (°F) 100 200 300 400 500 600 700 800 900 1000 1100 300 400 500 600 0
Total ECCS Flow (gpm) 9.0 Wide Range Torus Water Level (ft)
ECCS Vortex Limit V
5,000 10,000 15,000 20,000 25,000 30,000 35,000 9.5 10.0 10.5 11.0 0
Primary Containment Water Level (ft) 0 Torus Bottom Pressure (psig)
Primary Containment Pressure Limit D
10 20 30 40 50 60 70 80 90 100 10 20 30 40 50 60 70 93 Alternate Injection Systems (DEOP 500-03)
L2 Standby Coolant Supply 15000 gpm NA CRD crosstie 110 gpm 1655 psig Reactor head cooling 110 gpm 1655 psig SBLC boron tank 40 gpm 1517 psig SBLC test tank 40 gpm 1517 psig Core Spray with CST suction 4650 gpm 360 psig LPCI with CST suction 5000 gpm 290 psig LPCI crosstie 500 gpm 300 psig Condensate Transfer 500 gpm 65 psig Service unit back flush 500 gpm 65 psig Fire System 300 gpm 180 psig ECCS Keep Fill 50 gpm 38 psig HPCI from outside control room (DOA 2300-03) 5600 gpm 1335 psig Portable pumps (FSGs, TSG-3):
- FLEX 1000 gpm 110 psig
- Godwin HS150HH 850 gpm 113 psig System Max Press.
Capacity No / Unknown Yes 2
All rods in to at least 04
?
No / Unknown Yes 3
Will the reactor stay shutdown under all conditions without boron (TSG-3.2.1)
?
FAILURE TO SCRAM:
Exit this procedure Enter DEOP 400-05
=
4 1
Scram While in this procedure:
IF THEN 7
No Injection Subsystem available to inject at design capacity (Detail L3)
Exit this procedure
=
Enter DEOP 400-06 USE ALTERNATE LEVEL/PRESSURE CONTROL:
RPV water level unknown Exit this procedure
=
Enter DEOP 400-01 at h FLOOD THE RPV:
Core damage is occurring (TSG-3.2.2)
Exit all DEOP flowcharts
=
Enter SAMG-1 and SAMG-2 ENTER SAMGs:
Verify needed auto actions while continuing here:
- Isolations
- ECCS start
- Diesel generator start
+
6 LEVEL PRESSURE IF THEN P-1 Drywell pressure above 2.0 psig Before RPV pressure drops to 350 psig, prevent Core Spray and LPCI injection not needed for core cooling.
Anticipate RPV Blowdown requirement Depressurize RPV rapidly using IC and main turbine bypass valves.
p OK to exceed 100°F/hr cooldown.
Cannot hold torus bulk temperature below Fig M, Heat Capacity Limit in DEOP 200-01 Lower RPV pressure to stay below Fig M, Heat Capacity Limit.
p OK to exceed 100°F/hr.
RPV Blowdown Required by any DEOP Go to 4 P-3 Depressurize the RPV using main turbine BPVs.
p Hold cooldown rate below 100°F/hr.
p Use other RPV Pressure Control Systems (Detail P1) if needed.
p If Drywell Pneumatics lost, minimize use of the Target-Rock ADSV (203-3A).
Cool down to cold shutdown using Shutdown Cooling.
p Hold cooldown rate below 100°F/hr.
P-5 IF THEN Shutdown Cooling does not work Hold recirc suction temperature below 350°F using RPV Pressure Control Systems (Detail P1).
P-4 WAIT until:
Shutdown Cooling high temperature interlock clears (recirculation suction temperature below 350°F)
AND further cooldown is required 4
IF THEN L-1 High level is needed for use of Alternate Injection Systems (Detail L2) or Shutdown Cooling Restore and hold level between -143 in. (TAF) and 117 in. (main steam lines).
p Before raising level above 48 in., isolate HPCI steam lines.
p OK to bypass high RPV water level trips (DEOP 500-02).
Cannot stay below Fig D, Primary Containment Pressure Limit Stop injection from outside the primary containment not needed for core cooling.
Restore and hold RPV water level between 8 in. and 48 in. using any of the Preferred Injection Systems (Detail L1).
p Use Alternate Injection Systems if needed (Detail L2).
p Check Detail A to find out which level instruments to use.
L-2 IF THEN Cannot restore and hold level between 8 in. and 48 in.
Go to 2 L-4 Initiate IC.
Restore and hold RPV water level above -143 in. (TAF) using any of the Preferred Injection Systems (Detail L1).
p Use Alternate Injection Systems if needed (Detail L2).
L-5 IF THEN Less than 2 Injection Subsystems (Detail L3) available Line up as many Alternate Injection Systems (Detail L2) as possible while continuing here.
+
Cannot restore and hold level above
-143 in. (TAF)
Go to 3 Cannot hold level above -59 in.
Inhibit ADS.
IF THEN L-3 You can restore and hold RPV water level above 8 in.
Return to 1 2
Transfer to 4 in Pressure while continuing here.
RPV BLOWDOWN REQUIRED:
+
L-7 Restore and hold RPV water level above -143 in. (TAF) using any of the Preferred Injection Systems (Detail L1).
p Use Alternate Injection Systems if needed (Detail L2).
L-8 IF THEN Cannot restore and hold level above -163 in.
AND Neither Core Spray loop flow is at or above 4750 gpm
- 1. Maximize injection using Preferred and Alternate Injection Systems (Details L1, L2).
- 2. IF.. you still cannot restore and hold level above -163 in.,
THEN...ENTER SAMGs:
Exit all DEOP Enter SAMG-1
=
Flowcharts and SAMG-2 Cannot restore and hold level at or above -191 in.
(Blow down)
RPV water level drops to
-163 in.
BEFORE L-6 3
P-2 Stabilize RPV pressure below 1060 psig using main turbine BPVs (DGP 02-03).
p Use other RPV Pressure Control Systems (Detail P1) if needed.
p If Drywell Pneumatics lost, place the Target-Rock switch (203-3A) in AUTO.
RPV Blowdown To 4 Issue Procedure To Be Entered Entered (Y/N)
Enter DGP 2-3 while continuing here.
ENTER SCRAM PROCEDURE:
+
5 DEOP 100 RPV CONTROL Reactor Pressure
< 500 psig
> 500 psig
-143 in.
-163 in.
-191 in.
-215 in.
-170 in.
-186 in.
-209 in.
-229 in.
Top of Active Fuel Minimum Steam Cooling RPV Water Level 2/3 Core Height Minimum Zero-Injection RPV Water Level Action Level Fuel Zone Level Correction (TSG Attachment L)
Preferred Injection Systems L1 Condensate 5700 gpm 375 psig p OK to bypass high RPV water level trip (DEOP 500-02).
Feedwater 8700 gpm 1590 psig CRD 110 gpm 1655 psig HPCI 5600 gpm 1335 psig Exceeding 165°F suction temperature may cause system damage.
p Use CST suction if you can.
- OK to bypass high torus level transfer (DEOP 500-02).
p OK to bypass (DEOP 500-02):
- High area temperature isolation.
- High drywell pressure test return isolation.
Core Spray 4650 gpm 345 psig LPCI 5000 gpm 270 psig p Use HXs as soon as you can.
Exceeding NPSH/Vortex Limits (Figs N1, N2, N3, V) may cause system damage.
System / Details Max Press.
Capacity At or below 6 ft Above 6 ft P-14 Torus water level
?
P-15 Open all 5 ADSVs.
p OK to exceed 100°F/hr cooldown.
P-12 Prevent Core Spray and LPCI injection not needed for core cooling.
P-13 Initiate IC to maximum flow.
Below 2.0 psig At or above 2.0 psig P-11 Drywell pressure
?
1 12 ft 13 ft 14 ft 15 ft 16 ft 0
RPV Pressure (psig) 130 Torus Bulk Temperature (°F)
Heat Capacity Limit M
100 200 300 400 500 600 700 800 900 1000 1100 140 150 160 170 180 190 200 210 220 230 240 250 111 psig Use only when torus level is between 12 ft and 16 ft.
See TSGs for torus levels below 12 ft or above 16 ft.
Torus Level L3 Injection Subsystems Note Condensate 5700 gpm 375 psig Core Spray Pump A 4650 gpm 345 psig Core Spray Pump B 4650 gpm 345 psig LPCI 5000 gpm 270 psig System Capacity Max Press.
P1 RPV Pressure Control Systems Note ADSVsonly if torus water level is above 6 ft.
p Use preferred sequence if you can: A, C, E, D, B 140 MWt 0.54 Mlbm/hr IC (DOP 1300-03) 74 MWt 0.29 Mlbm/hr HPCI (DOA 2300-02)
Exceeding NPSH/Vortex Limits (Figs N3, V) or 165°F suction temperature may cause system damage.
p Use CST suction if you can.
- OK to bypass high torus level transfer (DEOP 500-02).
p OK to bypass (DEOP 500-02):
- High area temperature isolation.
- High drywell pressure test return isolation.
26-37 MWt 0.10-0.14 Mlbm/hr Main steam line drains (DGP 02-03) 14 MWt 0.05 Mlbm/hr RWCU, recirculation mode (DOP 1200-03) p Bypass demins (DOP 1200-03).
p OK to bypass isolations (DEOP 500-02).
10 MWt 0.04 Mlbm/hr RWCU, blowdown mode (DOP 1200-02) 5.4 MWt 0.021 Mlbm/hr Gland Seal Steam (DOP 5600-02) 3.9 MWt 0.015 Mlbm/hr IC vent valves (DOP 1300-03) 1.9 MWt 0.007 Mlbm/hr IC from outside control room (TSG-3) 74 MWt 0.29 Mlbm/hr ADSVs from outside control room (TSG-3)...only if torus water level is above 6 ft.
140 MWt 0.54 Mlbm/hr Main turbine BPVs 112 MWt 0.44 Mlbm/hr System/Details Capacity (MWt)
Steam Flow (Mlbm/hr)
P2 RPV Blowdown Systems (DEOP 500-07)
Note ADSVsonly if torus water level is above 6 ft.
140 MWt 0.54 Mlbm/hr IC 74 MWt 0.29 Mlbm/hr HPCI (DOA 2300-02)
Exceeding NPSH/Vortex Limits (Figs N3, V) or 165°F suction temperature may cause system damage.
p Use CST suction if you can.
- OK to bypass high torus level transfer (DEOP 500-02).
p OK to bypass (DEOP 500-02):
- High area temperature isolation.
- High drywell pressure test return isolation.
26-37 MWt 0.10-0.14 Mlbm/hr Main steam line drains 14 MWt 0.05 Mlbm/hr RWCU, recirculation mode p Bypass demins.
p OK to bypass all isolations (DEOP 500-02).
10 MWt 0.04 Mlbm/hr RWCU, blowdown mode 5.4 MWt 0.021 Mlbm/hr Gland Seal Steam 3.9 MWt 0.015 Mlbm/hr IC vent valves 1.9 MWt 0.007 Mlbm/hr IC from outside control room (TSG-3) 74 MWt 0.29 Mlbm/hr ADSVs from outside control room (TSG-3)...only if torus water level is above 6 ft.
140 MWt 0.54 Mlbm/hr Main turbine BPVs 112 MWt 0.44 Mlbm/hr Shutdown Cooling (DOP 1000-03) 7.9 MWt 0.031 Mlbm/hr Head vent 2.5 MWt 0.008 Mlbm/hr HPCI steam line 1.8 MWt 0.006 Mlbm/hr System/Details Capacity (MWt)
Steam Flow (Mlbm/hr) 0 LPCI / Core Spray Pump Flow (gpm) 50 Torus Bulk Temperature (°F)
LPCI / Core Spray NPSH Limit ECCS Flow Up to 10,750 gpm N1 1000 2000 3000 4000 5000 6000 100 150 200 250
> 15 psig 10 psig 5 psig 3.5 psig 0
LPCI / Core Spray Pump Flow (gpm) 50 Torus Bulk Temperature (°F)
LPCI / Core Spray NPSH Limit ECCS Flow Up to 25,550 gpm N2 1000 2000 3000 4000 5000 6000 100 150 200 250
> 15 psig 10 psig 5 psig 3.5 psig Torus Bottom Pressure (psig) 0 HPCI Flow (gpm) 50 Torus Bulk Temperature (°F)
> 15 psig 10 psig 5 psig 3.5 psig Torus Bottom Pressure (psig) 1000 2000 3000 4000 5000 6000 7000 100 150 200 250 DEOP 100 14 RPV CONTROL Dresden Nuclear Power Station Units 2(3)
EMERGENCY OPERATING PROCEDURE Rev Number Title CATEGORY 1 Applicability: Modes 1,2,3 P-17 WAIT until:
Shutdown Cooling high temperature interlock clears (recirculation suction temperature below 350°F)
AND further cooldown is required Leave ADSV switches in MAN (if opened). Allow RPV to depressurize.
p OK to exceed 100°F/hr cooldown.
p Reducing RPV pressure affects margin to RPV Saturation Temperature.
Check Detail A.
P-16 IF THEN Less than 5 ADSVs can be opened AND Difference between RPV pressure and drywell pressure is greater than or equal to 51 psig Hold RPV pressure less than 51 psig above drywell pressure using other RPV Blowdown Systems (Detail P2).
p OK to exceed 100°F/hr cooldown.
p OK to exceed release rate limits.
p OK to bypass interlocks (DEOP 500-02).
While blowing down:
IF THEN P-10 No Injection Subsystem available to inject at design capacity (Detail L3)
Exit this procedure
=
Enter DEOP 400-06
- 1. Terminate the depressurization.
- 2. USE ALTERNATE LEVEL/PRESSURE CONTROL:
Cool down to cold shutdown using Shutdown Cooling.
P-18 IF THEN Shutdown Cooling does not work Cool down to cold shutdown using other RPV Blowdown Systems (Detail P2).
p OK to bypass interlocks (DEOP 500-02).
Preferred Injection Systems L1 Condensate 5700 gpm 375 psig p OK to bypass high RPV water level trip (DEOP 500-02).
Feedwater 8700 gpm 1590 psig CRD 110 gpm 1655 psig HPCI (TSG-5.2.3, FSG-02) 5600 gpm 1335 psig
! Exceeding 165°F suction temperature may cause system damage.
p Use CST suction if you can.
- OK to bypass high torus level transfer (DEOP 500-02).
p OK to bypass (DEOP 500-02):
- High area temperature isolation.
- High drywell pressure test return isolation.
Core Spray 4650 gpm 345 psig LPCI 5000 gpm 270 psig p Use HXs as soon as you can.
! Exceeding NPSH/Vortex Limits (Figs N1, N2, N3, V) may cause system damage.
System / Details Max Press.
Capacity While in this procedure:
IF THEN 1
Core damage is occurring (TSG-3.2.2)
Exit all DEOPs
=
Enter SAMG-1 and SAMG-2 ENTER SAMGs:
Any Injection Subsystem able to inject at design capacity (Detail L3)
Exit this procedure
=
Enter DEOP 100 RETURN TO RPV CONTROL:
RPV water level unknown Exit this procedure
=
Enter DEOP 400-01 at h FLOOD THE RPV:
Cannot stay below Fig D, Primary Containment Pressure Limit Stop injection from outside the primary containment not needed for core cooling.
START LEVEL L-1 Inhibit ADS.
L-2 Initiate IC.
PRESSURE No Yes P-1 HPCI available and needed for RPV injection
?
IF THEN P-8 HPCI unavailable OR HPCI injection no longer needed Go to z Cannot restore and hold RPV water level above -143 in. (TAF)
AND Difference between RPV pressure and drywell pressure is greater than or equal to 51 psig STEAM COOL:
Go to b RPV depressurization will result in loss of HPCI injection needed for core cooling
- 1. Terminate RPV depressurization.
- 2. Control RPV pressure as low as possible while maintaining adequate core cooling.
p HPCI isolates at 106 psig.
IF THEN P-2 HPCI available AND needed for RPV injection Go to x Cannot restore and hold RPV water level above -143 in. (TAF)
AND Difference between RPV pressure and drywell pressure is greater than or equal to 51 psig STEAM COOL:
Go to b 30 P-9 Control RPV pressure between 150 psig and 800 psig using HPCI (TSG-5.2.3, FSG-02).
! Exceeding NPSH/Vortex Limits (Figs N3, V) or 165°F suction temperature may cause system damage.
p Use CST suction if you can. OK to bypass high torus level transfer (DEOP 500-02).
p Use other RPV Blowdown Systems (Detail P2) if needed.
p OK to bypass interlocks (DEOP 500-02).
p OK to exceed release rate limits.
p OK to exceed 100°F/hr cooldown.
p If Drywell Pneumatics lost, place the Target-Rock switch (203-3A) in AUTO.
Restore and hold RPV water level as high as possible between -11 in. (Feedwater/HPCI sparger) and 117 in. (main steam line).
p If HPCI is operating, control level below 42 in. (HPCI trip).
p Before raising level above 48 in., isolate HPCI steam lines.
p OK to use any available injection systems (Details L1, L2).
p Check Detail A to find out which level instruments to use.
p OK to bypass high RPV water level trips (DEOP 500-02)
L-3 IF THEN Cannot restore and hold level above -143 in. (TAF)
Go to b in Pressure while continuing here.
+
STEAM COOL:
AND Difference between RPV pressure and drywell pressure is greater than or equal to 51 psig Cannot restore and hold level above -163 in.
Difference between RPV pressure and drywell pressure is less than 51 psig
- 1. Maximize injection using Preferred and Alternate Injection Systems (Details L1, L2).
- 2. IF......... you still cannot restore and hold level above -163 in.,
THEN...ENTER SAMGs:
Exit all DEOP Enter SAMG-1
=
Flowcharts and SAMG-2 0
Total ECCS Flow (gpm) 9.0 Wide Range Torus Water Level (ft)
ECCS Vortex Limit V
5,000 10,000 15,000 20,000 25,000 30,000 35,000 9.5 10.0 10.5 11.0 0
Primary Containment Water Level (ft) 0 Torus Bottom Pressure (psig)
Primary Containment Pressure Limit D
10 20 30 40 50 60 70 80 90 100 10 20 30 40 50 60 70 93 0
LPCI / Core Spray Pump Flow (gpm) 50 Torus Bulk Temperature (°F)
LPCI / Core Spray NPSH Limit ECCS Flow Up to 10,750 gpm N1 1000 2000 3000 4000 5000 6000 100 150 200 250
> 15 psig 10 psig 5 psig 3.5 psig 0
LPCI / Core Spray Pump Flow (gpm) 50 Torus Bulk Temperature (°F)
LPCI / Core Spray NPSH Limit ECCS Flow Up to 25,550 gpm N2 1000 2000 3000 4000 5000 6000 100 150 200 250
> 15 psig 10 psig 5 psig 3.5 psig Torus Bottom Pressure (psig) 0 HPCI Flow (gpm) 50 Torus Bulk Temperature (°F)
> 15 psig 10 psig 5 psig 3.5 psig Torus Bottom Pressure (psig) 1000 2000 3000 4000 5000 6000 7000 100 150 200 250 DEOP 400-06 ALTERNATE RPV LEVEL/PRESSURE CONTROL 31 32 To x With HPCI Without HPCI Initiate IC to maximum flow. Allow RPV to depressurize.
p OK to exceed 100°F/hr cooldown rate.
P-3 IF THEN IC unavailable Go to c P-6 Depressurize the RPV using RPV Blowdown Systems (Detail P2).
p Use IC if possible.
p OK to exceed 100°F/hr cooldown rate.
p OK to bypass interlocks (DEOP 500-02).
p OK to exceed release rate limits.
p If Drywell Pneumatics lost, minimize use of the Target-Rock ADSV (203-3A).
P-7 Hold RPV pressure less than 51 psig above drywell pressure using RPV Blowdown Systems (Detail P2).
p OK to bypass interlocks (DEOP 500-02).
p OK to exceed release rate limits.
33 P-5 WAIT until:
Transfer to low pressure motor-driven injection systems is required AND Depressurization Criteria are satisfied (Detail P4, TSG-3.2.4)
Issue Procedure To Be Entered Entered (Y/N)
RPV Water Level Instruments A
! RPV water level instruments may be unreliable due to boiling in the instrument runs when drywell or reactor building temperatures near the instrument runs are above Fig B, RPV Saturation Temperature.
An RPV water level instrument may not be used if the instrument reads at or below its Minimum Usable Indicating Level (Detail C).
Fuel Zone
(+60" to -340")
Medium/Narrow Range All Rx Bldg Temps
(+60" to -60")
Medium/Narrow Range Rx Bldg Temps 171°F or less
(+60" to -60")
Wide Range
(+330" to -70")
-294
-295
-295
-296
-296
-297
-37
-38
-39
-40
-42
-42
-60
-60
-60
-60
-60
-60
-68
-50
-20 19 68 106 32 101 201 301 401 501 to to to to to to 100 200 300 400 500 559 Drywell Temperature (°F)
TR 2(3)-1340-1, Point 9 or 10 C
Minimum Usable Indicating Levels B
RPV Saturation Temperature 0
200 RPV Pressure (psig)
Drywell Temperature (°F)
TR 2(3)-1340-1, Point 9 or 10 OR Reactor Building Temperature (°F) 100 200 300 400 500 600 700 800 900 1000 1100 300 400 500 600 P-4 Control RPV pressure below 1060 psig using RPV Blowdown Systems (Detail P2).
p OK to exceed 100°F/hr cooldown rate.
p OK to bypass interlocks (DEOP 500-02).
p OK to exceed release rate limits.
p If Drywell Pneumatics lost, place the Target-Rock switch (203-3A) in AUTO.
100 RPV Pressure (psig)
-140 RPV Water Level (in.)
Minimum Pre-Depressurization RPV Water Level H
150 200 250 300 350 400
-120
-100
-80
-60
-40
-20 0
20 40 5 ADSVs 4 ADSVs 3 ADSVs Time after shutdown: 1 hr Steam Cooling 35 IF THEN P-10 RPV pressure drops below the maximum injection pressure of any available injection source (Details L1, L2)
AND You cannot restore and hold RPV water level above -163 in.
BLOW DOWN:
Go to n RPV water level rising EXIT STEAM COOLING:
Go to ; (top of Pressure)
P-11 Initiate IC.
P-12 Stabilize RPV pressure using RPV Blowdown Systems (Detail P2).
p OK to exceed release rate limits.
p OK to bypass interlocks (DEOP 500-02).
P-12 WAIT until RPV water level drops to -215 in.
36 At or below 6 ft Above 6 ft P-13 Torus water level
?
P-14 Open all 5 ADSVs.
p OK to exceed 100°F/hr cooldown.
Blowdown Leave ADSV switches in MAN (if opened). Allow RPV to depressurize.
p OK to exceed 100°F/hr cooldown.
p Reducing RPV pressure affects margin to RPV Saturation Temperature.
Check Detail A.
P-15 IF THEN Less than 5 ADSVs can be opened AND Difference between RPV pressure and drywell pressure is greater than or equal to 51 psig Hold RPV pressure less than 51 psig above drywell pressure using other RPV Blowdown Systems (Detail P2).
p OK to exceed 100°F/hr cooldown.
p OK to exceed release rate limits.
p OK to bypass interlocks (DEOP 500-02).
To z To b DEOP 400-06 0
ALTERNATE RPV LEVEL/PRESSURE CONTROL Dresden Nuclear Power Station Units 2(3)
EMERGENCY OPERATING PROCEDURE Rev Number Title CATEGORY 1 Applicability: Modes 1,2,3 P4 Depressurization Criteria (TSG-3.2.4)
Note Depressurization Method Depressurization Criteria HPCI
- More than 1 hr after shutdown, AND
- Motor-driven injection capacity greater than 270 gpm, AND
- More than 1 hr after shutdown, AND
- Motor-driven injection capacity greater than 270 gpm, AND
- At least 3 ADSVs available, AND
< 500 psig
> 500 psig
-143 in.
-163 in.
-191 in.
-215 in.
-170 in.
-186 in.
-209 in.
-229 in.
Top of Active Fuel Minimum Steam Cooling RPV Water Level 2/3 Core Height Minimum Zero-Injection RPV Water Level Action Level Fuel Zone Level Correction (TSG Attachment L)
Alternate Injection Systems (DEOP 500-03)
L2 Standby Coolant Supply 15000 gpm NA CRD crosstie 110 gpm 1655 psig Reactor head cooling 110 gpm 1655 psig SBLC boron tank 40 gpm 1517 psig SBLC test tank 40 gpm 1517 psig Core Spray with CST suction 4650 gpm 360 psig LPCI with CST suction 5000 gpm 290 psig LPCI crosstie 500 gpm 300 psig Condensate Transfer 500 gpm 65 psig Service unit back flush 500 gpm 65 psig Fire System 300 gpm 180 psig ECCS Keep Fill 50 gpm 38 psig HPCI from outside control room (DOA 2300-03) 5600 gpm 1335 psig Portable pumps (FSGs, TSG-3):
- FLEX 1000 gpm 110 psig
- Godwin HS150HH 850 gpm 113 psig System Max Press.
Capacity L3 Injection Subsystems Note Condensate 5700 gpm 375 psig Core Spray Pump A 4650 gpm 345 psig Core Spray Pump B 4650 gpm 345 psig LPCI 5000 gpm 270 psig System Capacity Max Press.
P2 RPV Blowdown Systems (DEOP 500-07)
Note ADSVsonly if torus water level is above 6 ft.
140 MWt 0.54 Mlbm/hr IC 74 MWt 0.29 Mlbm/hr HPCI (DOA 2300-02)
! Exceeding NPSH/Vortex Limits (Figs N3, V) or 165°F suction temperature may cause system damage.
p Use CST suction if you can.
- OK to bypass high torus level transfer (DEOP 500-02).
p OK to bypass (DEOP 500-02):
- High area temperature isolation.
- High drywell pressure test return isolation.
26-37 MWt 0.10-0.14 Mlbm/hr Main steam line drains 14 MWt 0.05 Mlbm/hr RWCU, recirculation mode p Bypass demins.
p OK to bypass all isolations (DEOP 500-02).
10 MWt 0.04 Mlbm/hr RWCU, blowdown mode 5.4 MWt 0.021 Mlbm/hr Gland Seal Steam 3.9 MWt 0.015 Mlbm/hr IC vent valves 1.9 MWt 0.007 Mlbm/hr IC from outside control room (TSG-3) 74 MWt 0.29 Mlbm/hr ADSVs from outside control room (TSG-3)...only if torus water level is above 6 ft.
140 MWt 0.54 Mlbm/hr Main turbine BPVs 112 MWt 0.44 Mlbm/hr Shutdown Cooling (DOP 1000-03) 7.9 MWt 0.031 Mlbm/hr Head vent 2.5 MWt 0.008 Mlbm/hr HPCI steam line 1.8 MWt 0.006 Mlbm/hr System/Details Capacity (MWt)
Steam Flow (Mlbm/hr)