ML24011A187
| ML24011A187 | |
| Person / Time | |
|---|---|
| Site: | Limerick  (NPF-039, NPF-085) |
| Issue date: | 12/14/2023 |
| From: | Constellation Energy Generation |
| To: | Thomas Setzer NRC/RGN-I/DORS/OB |
| References | |
| EPID L-2024-OLL-0000 | |
| Download: ML24011A187 (1) | |
Text
1st 10 Written Exam Questions Submittal Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 1 of 50 05 September 2023 1
ID: 2503945 Points: 1.00 Question #10 Tier 3 Rad Unit 1 is operating at 100% power when the REACTOR ENCL AREA HI RADIATION annunciator, 109-B4, alarms.
Which one of the following describes the possible cause for this alarm?
A.
Reactor fuel pin leak B.
1A Recirculation Pump Seal Failure C.
Steam Leak at the Main Turbine Stop Valves D.
RWCU leak in the 1A RWCU pump room Answer:
D
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 2 of 50 05 September 2023 Answer Explanation ANSWER RWCU leak in the 1A RWCU pump room is correct: A primary coolant leak is always a source of potential high radiation. A leak from the RWCU system into the RWCU pump room will allow radioactive steam loose in the secondary containment and the RWCU pump room in particular. This room is monitored for hi radiation by the Area Rad Monitoring System, Channel 24.
DISTRACTOR Reactor fuel pin leak is wrong: Plausible source of radiation as a fuel pin leak will add a lot of radiation to the main steam which could plausibly be seen by the Rx Encl. Rad Monitors. Actually, assuming there are no steam leaks in the Reactor Enclosure, radioactive steam should be transported directly to the turbine and main condenser and the first indication of a problem would be air ejector rad monitors which are in the Turbine building and have a separate alarming system.
DISTRACTOR 1A Recirculation Pump Seal Failure is wrong: Plausible source of radiation because a leaking Recirc Seal would put radioactive steam into the Drywell which a candidate could assume brings in a Reactor Enclosure Hi Radiaition alarm. This does not. The drywell steam leak would cause a rise on the Drywell Rad Monitor which samples the DW airspace only.
DISTRACTOR Steam Leak at the Main Turbine Stop Valves is wrong: A Steam Leak at the Main Turbine Stop Valves is plausible as steam leaking out would cause local hi radiation signals which would be reported by the ARM system and the candidate could assume they ring in the same alarm 109-B4. However, these ARM channels associated with the Turbine bring in alarm 109-B5, "Turbine Enclosure Area Hi Radiation"
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 3 of 50 05 September 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:
0.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2503945 Version ID:
3179702 User-Defined ID:
Q #10 NEW ALTERNATE Cross Reference Number:
CLOSED Topic:
Secondary Containment Rad Control Num Field 1:
LM Num Field 2:
RO Text Field:
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 4 of 50 05 September 2023 Comments:
References Provided None K/A Justification SRO-Only Justification N/A Additional Information N/A General Data Level RO Tier 3
Group N/A KA # and Rating G 2.3.5 2.9 KA Statement RADIATION CONTROL Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms or personnel monitoring equipment Cognitive level High Safety Function N/A 10 CFR 55 41.11 Technical Reference with Revision No:
ARC-MCR-107-I1 Rev002 Question History: (i.e. LGS NRC-05)
New Question Type: (New, Bank, Modified)
New Revision History:
Training Objective LGSOPS0026A.10
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 5 of 50 05 September 2023 2
ID: 2495854 Points: 1.00 Question #20, Line Item 61 Unit 1 is operating at 100% power when the following alarms are received:
118-H5, "REAC ENCL COOLING WATER HEAD TANK HI/LO LEVEL" 109-B1, "1 REAC ENCL COOLING WATER HI RADIATION" EO reports that the level in the RECW Head Tank is 71 inches up slow RECW Rad monitor indicates 850 cpm up slow SW Rad monitor indicates 63 cpm steady WHICH ONE of the following identifies the location of leakage into or out of the RECW system?
A.
RECW Heat Exchanger B.
Reactor Recirc Pump Shaft Seal Cooler C.
RWCU Regenerative Heat Exchanger D.
Fuel Pool Cooling and Cleanup Heat Exchanger Answer:
B
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 6 of 50 05 September 2023 Answer Explanation
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 7 of 50 05 September 2023 From the stem, the candidate determines that the RECW Head Tank level issue is a HI level issue due to recalling that the level provided by the EO (71") is higher than normal and above the alarm setpoint (70.5").
In conjunction with the RECW High Radiation Alarm, the correct conclusion is that water is leaking into RECW from a higher pressure system with contaminated water.
ANSWER (B)
Reactor Recirc Pump Shaft Seal Cooler is Correct: as described above. With RECW supplied to the RR pump shaft seal coolers, the higher pressure of the Rx water will force it into the RECW in the event of a boundary leak. Reactor coolant can be pressurized up to 1000 psig+ whereas RECW is generally operating around 150 psig DISTRACTOR (A)
RECW Heat Exchanger is wrong: Plausible answer given the potential that the RECW heat exchanger interfaces with Service water as a cooling medium. A transfer from SW to RECW would make the head tank level rise if the candidate thought that SW pressure was higher the RECW pressure. However, The candidate should know that 72 inches in the head tank is above normal. SW
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 8 of 50 05 September 2023 operating pressure is about 125 psig maximum and RECW would leak into the SW system.
DISTRACTOR (C)
RWCU Regenerative Heat Exchanger is wrong: Plausible answer due to one of RECWs cooling loads is the Non Regenerative Heat Exchanger in RWCU (NRHX). This answer is plausible to the candidate who confuses the two heat exchangers and assumes RECW cools the RGHX which is does not. The RGHX is self cooled.
DISTRACTOR (D)
Fuel Pool Cooling and Cleanup Heat Exchanger is wrong: This is a plausible answer as FPC and Cleanup Heat Exchanger can be one of RECWs cooling loads. This is done when the reactor is shutdown however as all other cooling loads are lost during this evolution. This scenario is at power with all RECW normal loads in service. This method of cooling also requires a spool piece installed to function and this is abnormal.
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 9 of 50 05 September 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:
0.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2495854 Version ID:
3176248 User-Defined ID:
Q #20 NEW Cross Reference Number:
CLOSED Topic:
RECW Head Tank Hi Level Num Field 1:
LM Num Field 2:
RO-HIGH Text Field:
LO-I
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 10 of 50 05 September 2023 Comments:
References Provided None K/A Justification None SRO-Only Justification N/A Additional Information None General Data Level RO Tier 2
Group 1
KA # and Rating 400000 K5.02 3.1 KA Statement Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF8 CCS)
COMPONENT COOLING WATER SYSTEM :
Determine source(s) of RCS leakage into CCW Cognitive level High Safety Function 8
10 CFR 55 41.5 Technical Reference with Revision No:
S13.7.A Rev.011 S13.6.D Rev.015 S13.0.B Rev.018 Question History: (i.e. LGS NRC-05)
New Question Type: (New, Bank, Modified)
New Revision History:
Training Objective LGSOPS0013.02
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 11 of 50 05 September 2023 3
ID: 2496056 Points: 1.00 Question #35, Tier 4 Thermo Limerick Unit 2 is operating at full power, 100 full power hours following startup from a refueling.
The power range nuclear instruments have been adjusted to 100% based on a calculated heat balance.
Which one of the following will result in indicated reactor power being lower than actual reactor power?
A.
The feedwater temperature used in the heat balance calculation was 20°F higher than actual feedwater temperature.
B.
The reactor recirculation pump heat input term used in the heat balance was 10% lower than actual.
C.
The feedwater flow rate used in the heat balance calculation was 10% higher than actual flow rate.
D.
The operator miscalculated the enthalpy of the steam exiting the reactor vessel to be 10 Btu/lbm higher than actual.
Answer:
A
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 12 of 50 05 September 2023 Answer Explanation ANSWER The feedwater temperature used in the heat balance calculation was 20°F higher than actual feedwater temperature is correct:
-Using a higher feedwater temperature would result in using a correspondingly high specific enthalpy value. Resulting in a higher feedwater heat transfer value.
Because the feedwater heat transfer rate value is removed from the indicated power value, an increase in the feedwater heat transfer rate results in an indicated power lower than actual power DISTRACTOR The reactor recirculation pump heat input term used in the heat balance was 10% lower than actual is wrong: Plausible because it is part of the Heat Balance equation. If the value for the reactor recirculation pump heat input term was 10% lower than actual, the indicated power would be greater than the actual power as RRP adds energy to the core.
DISTRACTOR The feedwater flow rate used in the heat balance calculation was 10% higher than actual flow rate is wrong: Plausible because it is part of the Heat Balance equation. First by using a higher feedwater flow rate the feedwater heat transfer rate value will increase.
Second, feedwater flow rate also effects the steam heat transfer rate.
Using a higher feedwater flow rate, the steam heat transfer rate value will increase DISTRACTOR The operator miscalculated the enthalpy of the steam exiting the reactor vessel to be 10 Btu/lbm higher than actual is wrong:
Plausible because it is part of the Heat Balance equation. Using a higher specific enthalpy value results in a higher steam heat transfer.
Because the steam heat transfer rate value is added to the indicated power value, an increase in the steam heat transfer rate results in an indicated power higher than actual power
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 13 of 50 05 September 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:
2 Difficulty:
0.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2496056 Version ID:
3170497 User-Defined ID:
Q #35 BANK Cross Reference Number:
CLOSED Topic:
GF Heat Balance Num Field 1:
0.00 Num Field 2:
RO-High Text Field:
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 14 of 50 05 September 2023 Comments:
References Provided None K/A Justification None SRO-Only Justification N/A Additional Information None General Data Level RO Tier 4
Group N/A KA # and Rating 293007 K1.13 2.9 KA Statement HEAT TRANSFER (CORE THERMAL POWER)
Calculate core thermal power using a simplified heat balance Cognitive level High Safety Function N/A 10 CFR 55 41.14 Technical Reference with Revision No:
NRC GF Question History: (i.e. LGS NRC-05)
Bank 2284 Question Type: (New, Bank, Modified)
Bank Revision History:
Training Objective N/A
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 15 of 50 05 September 2023 4
ID: 2498567 Points: 1.00 Question #63, Item #48 Unit 2 RCIC has been manually initiated following a MSIV closure Reactor pressure is being maintained 900 to 1000 psig with HPCI Which of the following indications show RCIC injecting properly into the vessel?
A.
Discharge Pressure = 625 psig RCIC Speed = 3650 RPM RCIC Flow = 605 GPM in Auto B.
Discharge Pressure = 400 psig RCIC Speed = 650 RPM RCIC Flow = 699 GPM in Auto C.
Discharge Pressure = 1070 psig RCIC Speed = 4100 RPM RCIC Flow = 595 GPM in Auto D.
Discharge Pressure = 1210 psig RCIC Speed = 4560 RPM RCIC Flow = 250 GPM in Auto Answer:
C
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 16 of 50 05 September 2023 Answer Explanation ANSWER (C)
Discharge Pressure = 1070 psig RCIC Speed = 4100 RPM RCIC Flow = 595 GPM in Auto is correct: Upon initiatation, the RCIC turbine will speed up to develop sufficient discharge pressure to overcome reactor pressure. As flow increases, the flow controller will adjust speed to maintain the discharge pressure required to keep the flow in the control band which is aroung the 600 gpm setpoint. With Reactor pressure of 900-1000 psig, it will take about 100 psig above reactor pressure to reach rated flow.
DISTRACTOR (A)
Discharge Pressure = 625 psig RCIC Speed = 3650 RPM RCIC Flow = 605 GPM in Auto is wrong: Plausible response to this question as RCIC flow controller setpoint is 600 GPM. However, with a discharge pressure well below Reactor pressure, this much flow is not being injected to the reactor. A injection line break would present similar indications.
DISTRACTOR (B)
Discharge Pressure = 400 psig RCIC Speed = 650 RPM RCIC Flow = 699 GPM in Auto is wrong: These values are wrong as the speed and pressure are very low. It is plausible to the candidate who focuses on the flow value but does not evaluate the other parameters. This is not the answer to the question however which asks for indications of injection.
DISTRACTOR (D)
RCIC Discharge Pressure = 1210 psig RCIC Speed = 4560 RPM RCIC Flow = 250 GPM in Auto is wrong: Plausible answer as the speed and discharge pressure are in the normal range if somewhat on the high side. But the lack of sufficient flow indicates that the system cannot develop the required flow and has raised speed to the maximun but setpoint flow cannot be achieved.
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 17 of 50 05 September 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:
3 Difficulty:
0.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2498567 Version ID:
3170628 User-Defined ID:
Q #63 NEW Cross Reference Number:
CLOSED Topic:
RCIC Turbine Startup Num Field 1:
LM Num Field 2:
RO Low Text Field:
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 18 of 50 05 September 2023 Comments:
References Provided None K/A Justification SRO-Only Justification N/A Additional Information N/A General Data Level RO Tier 2
Group 1
KA # and Rating 217000 A3.02 4.0 KA Statement Ability to monitor automatic operation of the (SF2, SF4 RCIC) REACTOR CORE ISOLATION COOLING SYSTEM including: Turbine startup Cognitive level Low Safety Function 2
10 CFR 55 41.7 Technical Reference with Revision No:
S49.1.D Rev.045 S49.1.C Rev.017 Question History: (i.e. LGS NRC-05)
New Question Type: (New, Bank, Modified)
New Revision History:
Training Objective LGSOPS0049.06
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 19 of 50 05 September 2023 5
ID: 2495873 Points: 1.00 Question #69, Item #62 LGS Units 1 and 2 have experienced a Loss of Offsite Power and the both units scram.
Unit 1 emergency diesel generators have started and are carrying their vital buses with the exception of D12 All Unit 2 emergency diesel generators are carrying their buses While the crew is stabilizing the plant, the 0D ESW pump tripped on overcurrent Consider:
E-10/20, "Loss of Offsite Power""Attachment 3,'Alternate Power Supply for Any 4KV Safeguard Bus Using Any Diesel Generator" S11.6.A, "Transfer C & D ESW Pumps to Alternate Supply What is the status of the D14 EDG 10 minutes later with no operator intervention and what action should the crew take?
D14 Action A.
Tripped on high temperature Perform E10/20 Att. 3 and start the 0B ESW pump B.
Tripped on high temperature Move the 0D ESW breaker to the D14 cubicle per S11.6.A C.
Continues to run Perform E10/20 Att. 3 and start the 0B ESW pump D.
Continues to run Move the 0D ESW breaker to the D14 cubicle per S11.6.A Answer:
A
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 20 of 50 05 September 2023 Answer Explanation ANSWER (A)
Tripped on high temperature Perform E10/20 Att. 3 and start the 0B ESW pump is correct: Following a loss of offsite power (LOOP),
all EDGs should start and load the 4kv vital buses. ESW pumps A and B are powered from D11 and D12. C and D are from D23 and D24.
Any single ESW pump in a loop (A or B) is sufficient to cool all loads. In this scenario, D12 EDG does not start and along with it, 0B ESW.
When the 0D ESW pump trips, no flow is available in the B ESW loop and all running EDGs will lose cooling. The loaded Diesels will rapdily heat up and trip on high temperature without cooling. To correct this situation, power must be restored to the 0B ESW pump and the way to do this is by cross tying D12 to an energized bus and then restoring ESW pump operation.
DISTRACTOR (B)
Tripped on high temperature, Move the 0D ESW breaker to the D14 cubicle per S11.6.A is wrong: Plausible answer to swap cubicles as the procedure allows it as an option in SE-10/20.step 2.6.7.
A candidate could consider that perhaps a different source would eliminate the cause of the ESW trip. This is incorrect however, as a tripped ESW pump either has mechanical or electrical issues causing the overcurrent or a breaker problem which is swapped anyway to the new cubicle. Therefore, the problem is not fixed, only moved.
DISTRACTOR (C)
Continues to run, Perform E10/20 Att. 3 and start the 0B ESW pump is wrong: Plausible answer for the candidate who recalls that the EDG will bypass high temperature trips on a LOCA start and assumes that the same operation occurs on a bus undervoltage start.
With high temp trips bypassed, the EDG would continue to run until it broke. In fact, high temperature trips are not bypassed on a low voltage start and the EDG would trip in approximately 5 to 7 minutes depending on the bus loading.
DISTRACTOR (D)
Continues to run, Move the 0D ESW breaker to the D14 cubicle per S11.6.A is wrong: Plausible answer for the candidate who recalls that the EDG will bypass high temperature trips on a LOCA start and assumes that the same operation occurs on a bus undervoltage start.
With high temp trips bypassed, the EDG would continue to run until it broke. In fact, high temperature trips are not bypassed on a low voltage start and the EDG would trip in approximately 5 to 7 minutes depending on the bus loading. As noted above, moving the breaker is plausible but will not fix the problem.
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 21 of 50 05 September 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:
0.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2495873 Version ID:
3179656 User-Defined ID:
Q #69 NEW Cross Reference Number:
CLOSED Topic:
LOOP ESW pump trip actions Num Field 1:
LM Num Field 2:
RO-High Text Field:
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 22 of 50 05 September 2023 Comments:
References Provided None K/A Justification This question requires the candidate to predict the impact of an ESW pump trip and determine the correct procedure to perform to mitigate the fault. ESW is a service water system which is specific to the Emergency Diesel Generators and when running will supply several additional important plant safety loads which may lose their normal Service Water Supply when offsite power is lost.
SRO-Only Justification N/A Additional Information N/A General Data Level RO Tier 2
Group 1
KA # and Rating 510000 A2.01 3.6 KA Statement Ability to (a) predict the impacts of the following on the (SF4 SWS*) SERVICE WATER SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:
Pump/motor failure Cognitive level High Safety Function 4
10 CFR 55 41.5 Technical Reference with Revision No:
E-10/20 Rev.060 S11.6.A Rev.009 Question History: (i.e. LGS NRC-05)
New Question Type: (New, Bank, Modified)
New Revision History:
Training Objective LGSOPS0011.03
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 23 of 50 05 September 2023 6
ID: 2502984 Points: 1.00 Question #82, Item #81
- SRO ONLY**********
Unit was scrammed from 75% power due to drywell leak Drywell pressure is 7.6 psig Suppression pool spray was directed by the CRS When the HS-051-1F024, "1B RHR Pp Full Flow Test Return Vlv" was taken to open, D124 Load Center breaker tripped on overcurrent Drywell Temperature is 185 °F SP pressure is 2.4 psig up slow Containment Leak Detector Rad Monitor reads 1.50E+3 cpm and steady Consider the following:
T-102, "Primary Containment Control" T-225, "Startup and Shutdown of Suppression Pool and Drywell Spray" GP-8.5, "Isolation Bypass of Crucial Systems" OT-101, "High Drywell Pressure" Which of the following describes the action the CRS should direct, per T-102, that will mitigate the containment pressure rise?
A.
Vent the Drywell per OT-101 Attachment 3 B.
Bypass and restore Drywell Cooling per GP-8.5 C.
Spray the Drywell with 1A RHR per T-225 section 4.7 D.
Spray the Suppression Pool with 1B RHRSW per T-225 section 4.3 Answer:
B
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 24 of 50 05 September 2023 Answer Explanation Given that pool pressure is > drywell the leak must be in the suppression pool air space. There are 2 ways to reach drywell spray in T-102, from containment pressure leg and from drywell temperature leg. Drywell temperature leg DWT-4 stop sign when DW temp cannot be maintained <145 degrees continue. From the containment pressure leg drywell spray is not authorized until pool pressure is > 7.5 psig ANSWER (B)
Bypass and restore Drywell Cooling per GP-8.5 is correct: Given the conditions of SP pressure at 2.4 psig, T-102 would direct us to Spray the SP per PC/P-5. When the D124 LC is lost, 1B RHR cannot be aligned to spray the SP (or anything else). Therefore, the SRO must determine a different mitigative action. The only available action given is to bypass and restore DW cooling per GP-8.5. This action is called out of step DW/T-5.
DISTRACTOR (A)
Vent the Drywell per OT-101 Attachment 3 is wrong. Venting the drywell is a plausible action as it would have the intended result of lowering DW pressure and with the DW isolation signal, the rad reading would be steady, ie. not changing. This action is not directed at this time and venting the drywell is not normally directed unless DW integrity is threatened.(PC/P-10)
DISTRACTOR (C)
Spray the Drywell with 1A RHR per T-225 section 4.7 is wrong:
Plausible answer as this drywell pressure would quickly respond to DW spray and pressure would rapidly lower. The drywell pressure of 7.6 psig is above the value of 7.5 psig in step PC/P-6 to continue to the spray step. Step PC/P-6 refers to suppression pool press, however. Table PC/P-1 makes clear that SP pressure is the preferred pressure and should be used for decision making. T-102 step PC/P-6 is a stop sign which prevents passing until SP pressure is greater than 7.5 psig. Therefore, DW spray is not authorized at this time.
DISTRACTOR (D)
Spray the Suppression Pool with RHRSW per T-225 section 4.3 is wrong: Plausible option as section 4.3 of T-225 is provided to spray the SP. RHRSW is only available to Unit 1 through the B RHR system. therefore, the same loss of power to the B RHR valves would prevent this use.
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 25 of 50 05 September 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:
4.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2502984 Version ID:
3178665 User-Defined ID:
Q #82 NEW Cross Reference Number:
CLOSED Topic:
SRO Only SRV Tailpipe Leak Power loss Num Field 1:
LM Num Field 2:
RO-HIGH Text Field:
LO-I
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 26 of 50 05 September 2023 Comments:
References Provided None K/A Justification None SRO-Only Justification N/A Additional Information None General Data Level SRO Tier 2
Group 2
KA # and Rating 230000A2.05 / 3.8 KA Statement (230000A2.05) Ability to (a) predict the impacts of the following on the (SF5 RHR SPS) RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 /
45.6) AC electrical failures Cognitive level High Safety Function 5
10 CFR 55 CFR: 41.5 / 45.6 Technical Reference with Revision No:
T-102 Rev028 T-225 U/1, Rev. 26 Question History: (i.e. LGS NRC-05)
New Question Type: (New, Bank, Modified)
New Revision History:
New Training Objective LGSOPS1103.3 LGSOPS0093.5
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 27 of 50 05 September 2023 7
ID: 2504203 Points: 1.00 Question #86, Tier 3 Rad
- SRO ONLY **********
Given the following:
A loss of coolant accident has occurred T-101, RPV Control, and T-102, Primary Containment Control, have been entered SE-10, LOCA, has been entered 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> have elapsed since the LOCA signal Plant operators have arrived at the following SE-10 step:
4.25 WHEN greater than three hours have elapsed following the LOCA signal, THEN INJECT SLC per S48.1.B, Standby Liquid Control System Manual Initiation. (CM-4)
WHICH ONE of the following describes the basis for performance of this step?
A.
Provides an additional source of injection to the reactor to help recover RPV water level B.
Ensures sufficient negative reactivity is present to ensure reactor remains shutdown due to changes in core geometry C.
Ensures Hot Shutdown Boron Weight is injected before suppression pool temperature exceeds the Heat Capacity Temperature Limit D.
Minimizes radioactive release by adding sodium pentaborate to the suppression pool to satisfy the methodology for Alternate Source Term Answer:
D
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 28 of 50 05 September 2023 Answer Explanation
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 29 of 50 05 September 2023 Comments: This question is SRO only as it requires knowledge of administrative procedures that specify implementation and coordination of plant emergency procedures regarding radiation hazards that may arise during abnormal plant conditions and knowledge of conditions and limitations in the facility license.
ANSWER (D)
Minimizes radioactive release by adding sodium pentaborate to the suppression pool to satisfy the methodology for Alternate Source Term is correct: Design basis analyses credit SLC injection for limiting the radiological dose following loss of coolant accidents involving core damage. Radiation induced reactions are predicted to convert large fractions of dissolved ionic iodine into elemental iodine and organic iodides which can escape into the containment atmosphere. The rate of these reactions is strongly dependent on suppression pool pH. If the bulk Suppression Pool pH is maintained greater than 7, very little of the dissolved iodine will be converted to volatile forms and most of the iodine fission products will be retained in the suppression pool, thereby preventing iodine re-evolution. Over time, the pH in the Suppression Pool will tend to lower due to the addition of acidic chemicals. The sodium pentaborate solution used in the SLC system is derived from a strong base and therefore raises suppression pool pH DISTRACTOR (A)
Provides an additional source of injection to the reactor to help recover RPV water level is wrong: This is a plausible distracter as injecting SLC is an established step in T-101 for assisting in level control. Since SLC is identified as an Alternate Injection System it would likely be started to augment RPV injection in an earlier step of the Level branch, before RPV water level reaches the top of the active fuel. Also plausible for level control as injection requirements 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after shutdown are much lower and closer to the capacity of the SLC pumps. In this scenario however, SLC is not being used for level control and there is no indication that SLC injection was necessary to cover the core.
DISTRACTOR (B)
Ensures sufficient negative reactivity is present to ensure reactor remains shutdown due to changes in core geometry is wrong:
Pausible: Boration of the reactor coolant is performed to reduce power levels in the core by neutron moderation. This is plausible distracter for those candidates who recall that SLC lowers power and boration of the coolant in the vessel would by extension, aid in maintaining geometry. As noted above, this is not the intent of this step in SE-10.
DISTRACTOR (C)
Ensures Hot Shutdown Boron Weight is injected before suppression pool temperature exceeds the Heat Capacity Temperature Limit is wrong: Plausible to the examinee who mistakes the basis for this injection with the basis for injecting SLC
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 30 of 50 05 September 2023 prior to Suppression Pool Temperature exceeded 110 degrees F from T-101 Step RC/Q-16. This is not the basis for the SE-10 step.
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 31 of 50 05 September 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:
2 Difficulty:
3.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2504203 Version ID:
3178948 User-Defined ID:
Q #86 BANK Cross Reference Number:
CLOSED Topic:
(SRO Only) Pool PH minimze rad Num Field 1:
LM Num Field 2:
SRO-LOW Text Field:
LO-I
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 32 of 50 05 September 2023 Comments:
References Provided None K/A Justification This question requires the candidate to recognize the basis for injecting sodium pentaborate into the reactor well beyond when the reactor is confirmed shutdown in order to limit radioactivity SRO-Only Justification This question requires the SRO candidate to analyze the basis in tech spec 3.1.5. for Alternative Source Term and ph buffering post accident. This is one of the requirements of Nureg-1021 for TS bases knowledge required to analyze TS required actions Additional Information N/A General Data Level SRO Tier 3
Group N/A KA # and Rating (G2.3.14) / 3.8 KA Statement (G2.3.14) RADIATION CONTROL Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities, such as analysis and interpretation of radiation and activity readings as they pertain to administrative, normal, abnormal, and emergency procedures, or analysis and interpretation of coolant activity, including comparison to emergency plan or regulatory limits (SRO Only) (CFR: 43.4 /
45.10)
Cognitive level Low Safety Function 9
10 CFR 55 CFR: 43.4 / 45.10 Technical Reference with Revision No:
U1 TSB 3.1.5 ECR 09-00406 SE-10, Rev. 66
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 33 of 50 05 September 2023 UFSAR Section 9.3.5, Rev.
21 S48.1.B U/1, Rev. 000 Question History: (i.e. LGS NRC-05)
New Question Type: (New, Bank, Modified)
New Revision History:
New Training Objective LGSOPS0048.10
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 34 of 50 05 September 2023 8
ID: 2504026 Points: 1.00 Question #89, Item #23
- SRO ONLY**********
Unit 2 experienced a LOCA.
Drywell pressure peaked at 21 psig Drywell spray has been intitiated.
Reactor Level is being maintained with HPCI Current Drywell Pressure is 0.5 psig down fast.
Current Reactor pressure is 140 psig down slow Suppression Pool level is 26.5 ft up slow Suppression Pool Temperature is 111 °F What action should be directed and what is the basis for that action?
A.
Swap Drywell Spray to an internal suction source B.
Initiate S/D cooling using ONLY those RHR pumps not required to maintain RPV level above +12.5" C.
Secure HPCI to limit heat input into the Suppression Pool D.
Secure Drywell Spray to maintain containment pressure positive Answer:
D
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 35 of 50 05 September 2023 Answer Explanation
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 36 of 50 05 September 2023 The caution listed above from PCC-2 warns against lowering primary containment pressure excessively when using pumps with a SP suction such as HPCI. This would be another reason to limit de-pressurization.
ANSWER (D)
Secure Drywell Spray to maintain containment pressure positive is correct. T-102 step PC/P-4 states: "IF Pri Cont press drops below 0 psig, THEN control Pri Cont press above -5 psig exceeding offsite release rate limits if necessary." And step PCC-8 states "BEFORE DW press drops below 0 psig, THEN terminate DW spray."
These two notes make the point that it is undesireable to let pressure go negative and imperative to maintain it above -5 psig. Anything less than that risks exceeding design limits and breaking the containment.
Also, the caution above on NPSH would argue for limiting depressurization when running RHR and HPCI DISTRACTOR (A)
Swap Drywell Spray to an internal suction source is wrong. T-102 requires internal spray to be selected per step PCC-9 only when the "SAFE side of PCPL curve CANNOT be restored AND maintained."
The containment parameters given are easily recognized within the normal DBA LOCA response, which will not exceed PCPL. From UFSAR 15.6.5.4, Barrier Performance: "The primary containment is designed to maintain pressure integrity in the event of an instantaneous rupture of the largest single primary system piping within the structure while also accommodating the dynamic effects of the pipe break. Therefore, any postulated LOCA would not exceed the containment design limit." This distractor is plausible because Suppression Pool Level is above normal limits of both Tech Specs and T-102, which would require mitigative actions (but would not require cessation of external injection).
DISTRACTOR (C)
Initiate S/D cooling using ONLY those RHR pumps not required to maintain RPV level above +12.5" is wrong. This is wrong because T-101 steps RC/P-9 and -10 state: "WHEN RPV press less than 75
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 37 of 50 05 September 2023 psig AND further cooldown required, THEN Initiate S/D cooling using ONLY those RHR pumps not required to maintain RPV level above
+12.5"." RPV pressure is 140 psig in the stem, and shutdown cooling operation is not required or available. This is plausible because RPV pressure in the of 140 psig is a valid pressure for SDC if the plant is in T-111, "Level Restoration, Steam Cooling" and further cooldown is required due to the LOCA condition. But in T-102, pressure must be below 75 psig.
DISTRACTOR (C)
Secure HPCI to limit heat input into the Suppression Pool is wrong. A Plausible answer for the follwoing reasons:
- 1) T-102 step SP/T-5 states: "BEFORE Supp Pool temp reaches 110°F, enter T-101 AND execute concurrently" This indicates that 110°F is a limiting temperature for the suppression pool.
- 2) TS 3.6.2.1.a.2.b) lists 110°F as a limit for power operation. And
- 3) HPCI operational testing requires HPCI to be shutdown when suppression pool temperature exceeds 105°F.
In this scenario, if HPCI is required to maintain level, it would not be secured and the 110°F limit is an OPCON 1 limit.
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 38 of 50 05 September 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:
0.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2504026 Version ID:
3177933 User-Defined ID:
Q #89 NEW Cross Reference Number:
CLOSED Topic:
(SRO Only) Hi DW Press EPE, notes and cautions Num Field 1:
LM Num Field 2:
SRO-HIGH Text Field:
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 39 of 50 05 September 2023 Comments:
References Provided None K/A Justification This question requires the candidate to recall the information contained in a caution and notes in T-102, Primary Containment EOP for containment spray, and apply it to a set of plant conditions.
SRO-Only Justification This question is at the SRO level because it requires the candidate to assess plant conditions (emergency) and then prescribe a section of a procedure to mitigate or recover or with which to proceed. This also requires detailed procedure knowledge beyond the general progression of an event.
Additional Information N/A General Data Level SRO Tier 1
Group 1
KA # and Rating (295024) (G2.4.20) / 3.8 KA Statement (295024) (EPE 1) HIGH DRYWELL PRESSURE EMERGENCY PROCEDURES/PLAN (G2.4.20) Knowledge of the operational implications of emergency and abnormal operating procedures warnings, cautions, and notes (CFR: 41.10 / 43.5 / 45.13)
Cognitive level High Safety Function 2
10 CFR 55 CFR: 41.10 / 43.5 / 45.13 Technical Reference with Revision No:
T-225 U/2 Rev 26 T-101 Rev 28 T-102 Rev 29 Question History: (i.e. LGS NRC-05)
New Question Type: (New, Bank, Modified)
New Revision History:
New Training Objective LGSOPS1560.6
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 40 of 50 05 September 2023 9
ID: 2504068 Points: 1.00 Question #91, Item #24
- SRO ONLY*********
Both Units are at 100% power when the following occurs:
A Loss of All Offsite Power occurs 1 minute later this is plant status:
All Unit 1 Diesel Generators have failed to start Unit 1 Reactor level is -40" and rising Unit 1 HPCI and RCIC are running and injecting All MSIVs have closed Reactor power is 5% and steady SRVs are auto cycling open and closed WHICH ONE of the following describes 1) the instrument that can be used to determine Unit 1 Reactor pressure per E-1, Station Blackout, and 2) what action should be directed?
Instrument Used Action Directed A.
RCIC Steam Pressure, PI-49-1R602 Manually lower pressure to 990 psig B.
RCIC Steam Pressure, PI-49-1R602 Manually initiate Alternate Rod Insertion C.
"A" PAMS, XR-42-1R623A Manually lower pressure to 990 psig D.
"A" PAMS, XR-42-1R623A Manually initiate Alternate Rod Insertion Answer:
A
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 41 of 50 05 September 2023 Answer Explanation
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 42 of 50 05 September 2023 The Station Blackout is described in section 15.12 of the LGS USFAR Chapter 15 - Accident Analysis.
Station blackout is addressed by Limerick procedure E-1, Loss of all AC Power (Station Blackout)
During a station blackout very few instruments remain available for monitoring the reactor during the accident. They are identified by E-1.
From E-1 BASES The following RPV pressure instruments are available during a station blackout:
PI-42-*R605 WR PI-55-*R602 HPCI (available)
PI-49 *R602 RCIC (available)
From T-117 ANSWER RCIC Steam Pressure, PI-49-1R602, Manually lower pressure to 990 psig is correct: As noted above in the table, RCIC steam pressure on PI-49-1R602 is a valid reading for Rx pressure when AC power is unavailable. "A" PAMS indicator is powered from a safeguard AC bus but without a running EDG, it has no power on a complete LOOP. Because the reactor did not shutdown as evidenced by power at 5%, entry into T-117 would be required. T-117 step LQ/P-2 directs manually opening SRVs to lower pressure to 990 psig to minimize the possibility of the valves sticking open.
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 43 of 50 05 September 2023 DISTRACTOR RCIC Steam Pressure, PI-49-1R602, Manually initiate Alternate Rod Insertion is wrong: Alternate Rod Insertion is plausible as an action to use during an ATWS as it activates the backup scram valves to assist in Rod insertion. However, because the Rx power has lowered to 5% in the stem, RPS and or ARI has already initiated on low level (-38"). Therefore, this would be an ineffective action to take.
DISTRACTOR "A" PAMS, XR-42-1R623A, Manually lower pressure to 990 psig is rong: Plausible to the candidate who remembers that "A" PAMS is a EQ Post Accident indicator but forgets that it loses power on a Station Blackout. Until the D11 bus is restored, it will be de-energized.
DISTRACTOR "A" PAMS, XR-42-1R623A, Manually initiate Alternate Rod Insertion is wrong: Plausible to the candidate who remembers that "A" PAMS is a EQ Post Accident indicator but forgets that it loses power on a Station Blackout.Also as noted above, Alternate Rod Insertion is plausible as an action to use during an ATWS as it activates the backup scram valves to assist in Rod insertion. However, because the Rx power has lowered to 5% in the stem, RPS and or ARI has already initiated on low level (-38").
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 44 of 50 05 September 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:
3 Difficulty:
0.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2504068 Version ID:
3178787 User-Defined ID:
Q #91 NEW Cross Reference Number:
CLOSED Topic:
(SRO Only) Determine RPV pressure instr. available during a Station Blackout and Required Actions Num Field 1:
LM Num Field 2:
SRO-LOW Text Field:
LO-I
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 45 of 50 05 September 2023 Comments:
References Provided None K/A Justification This question requires the candidate to determine that in the stated stem conditions, Reactor Pressure will be rising and then interpret which action must be directed to mitigate the pressure rise SRO-Only Justification This question tests at the SRO level because it requires assessment of plant conditions (normal, abnormal, or emergency) and then prescribing a procedure or section of a procedure to mitigate or recover or with which to proceed Additional Information EOPs control High Reactor Pressure in T-101, unless one of several contingency procedures are in effect. In this question, ATWS RPV Control contingency procedure T-117 is in control of RPV Pressure in lieu of T-101, and all guidance for lowering reactor pressure is contained therein.
General Data Level SRO Tier 1
Group 1
KA # and Rating 295025EA2.01 / 4.7 KA Statement (295025EA2.01) Ability to determine or interpret the following as they apply to (EPE 2) HIGH REACTOR PRESSURE: (CFR: 41.10 /
43.5 / 45.13) Reactor pressure Cognitive level High Safety Function 3
10 CFR 55 CFR: 41.10 / 43.5 / 45.13 Technical Reference with Revision No:
T-117 Rev 23 Question History: (i.e. LGS NRC-05)
New Question Type: (New, Bank, Modified)
New Revision History:
New Training Objective LGSOPS1560.6
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 46 of 50 05 September 2023 10 ID: 2494888 Points: 1.00 Question #99, Item #66
- SRO ONLY**********
A Reactor Startup is in progress on Unit 1.
The RO has completed ST-6-107-884-1, NEUTRON MONITORING SYSTEM OVERLAP VERIFICATION ON STARTUP SRMs are being withdrawn to maintain count rate between the high and low trip setpoints per GP-2 Appendix 1, REACTOR START-UP AND HEAT-UP Plant status is as follows:
IRM Range Reading IRM Range Reading 1A 4
16 1B 3
48 1C 3
42 1D 2
75 1E 3
56 1F 3
7 1G 3
52 1H 3
39 SRM Reading SRM Reading 1A 4.8 E+04 cps 1B 1.1 E+02 cps 1C 1.8 E+05 cps 1D 8.2 E+03 cps Which ARC directs the action required to continue with Control Rod withdrawal?
A.
ARC-MCR-107-F3, IRM UPSCALE B.
ARC-MCR-107-G3, IRM DOWNSCALE C.
ARC-MCR-107-H4, SRM UPSCALE D.
ARC-MCR-107-I4, SRM RETRACTED WHEN NOT PERMITTED Answer:
C
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 47 of 50 05 September 2023 Answer Explanation ANSWER (C)
ARC-MCR-107-F3, SRM UPSCALE is correct. This alarm annunciates > 1.0 E+05 cps on any SRM with the Reactor Mode Switch not in Run. Because SRM 1C is above the setpoint, a rod withdrawal block is generated. This ARC directs action to withdraw the SRMs to lower the countrate below the setpoint and clear the block as long as the IRMs are on range 3 or above.
DISTRACTOR (A)
ARC-MCR-107-F3, IRM UPSCALE is wrong: Plausible answer given that 1D IRM is reading 75 and presumably rising. 75 is quite high and generally, the operators would range up prior to reaching this level.
The actual setpoint is 85/125 however.
DISTRACTOR (B)
ARC-MCR-107-G3, IRM DOWNSCALE is wrong. This alarm annunciates when One or more IRM channel(s) are downscale (above range 1). This is plausible because IRM 1F is very close to the setpoint (5), and could be ranged down.
DISTRACTOR (D)
ARC-MCR-107-I4, SRM RETRACTED WHEN NOT PERMITTED is wrong. This alarm is plausible as 1D SRM is 110 cps and 1D IRM is on range 2. A candidate could misread 1.1 E+02 cps as 11 cps rather than 110 and determine that it was below the setpoint. ARC 107-I4 annunciates when an attempt is made to retract any SRM while less than 100 CPS AND associated IRMs (on the same RPS side) less than range 3. However, the with all SRMs > than 100, this is not an option
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 48 of 50 05 September 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:
0.00 Allow multiple selections?
No Randomize choice order?
No System ID:
2494888 Version ID:
3179629 User-Defined ID:
Q #99 NEW Cross Reference Number:
CLOSED Topic:
(SRO Only) SRM trips, procedures to correct Num Field 1:
LM Num Field 2:
SRO-HIGH Text Field:
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 49 of 50 05 September 2023 Comments:
References Provided None K/A Justification This question requires assessing plant conditions against known setpoints to predict the plant response, and then to determine which procedure must be used to correct and control control rod withdrawals SRO-Only Justification This question is at the SRO level because it requires assessment of plant conditions (normal, abnormal, or emergency) and then prescribing a procedure or section of a procedure to mitigate or recover or with which to proceed Additional Information N/A General Data Level SRO Tier 2
Group 1
KA # and Rating 215004A2.04 / 3.7 KA Statement (215004A2.04) Ability to (a) predict the impacts of the following on the (SF7 SRMS) SOURCE RANGE MONITOR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 /
45.6) Upscale and downscale trips Cognitive level High Safety Function 7
10 CFR 55 CFR: 41.5 / 45.6 Technical Reference with Revision No:
ARC-MCR-108-F3, Rev. 2 ARC-MCR-107 H2, Rev. 2 Question History: (i.e. LGS NRC-05)
New Question Type: (New, Bank, Modified)
New Revision History:
New Training Objective LGSOPS0074.6
Test Answer Key Early NRC Review Questions Test ID: 373313 LM-OPS-EXAM-ILT-2023 Page: 50 of 50 05 September 2023