ML15007A171
| ML15007A171 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 12/12/2014 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | |
| References | |
| Download: ML15007A171 (147) | |
Text
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 1 QUESTION 1 Rev 1 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295001 AK2.02 IR 3.2 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following: Nuclear boiler instrumenta tion. Proposed Question:
The plant was operating at 100% power when a transient occurs resulting in the following conditions:
RPV level 35 inches and stable Reactor Power is 73% and stable Total Core Flow is 51.5 x 10 6 lbm/hr and stable The cause of this transient was the receipt of a signal from the ___.
A. ATWS/ARI logic.
B. EOC-RPT logic.
C. Recirc Flow Control Valve Runback logic D. Recirc Pump Cavitation Interlock Circuitry Proposed Answer:
C Explanation A. The ATWS logic uses high RPV pressure and/or low RPV water level to trip the Recirc pumps to OFF
. Core flow would indicate the Recirc pumps are still running.
B. This logic uses Turbine stop valve/first stage pressure to trip the Recirc pumps to SLOW, however, Slow speed Recirc Pumps would have altered Reactor power and core flow to less than given in stem
. C. Correct - The Reactor power and the Total Core Flow are consistent with a FCV Runback
. D. The cavitation interlocks will trip the Recirc pumps to slow speed; Reactor power and core flow are inconsistent with Recirc pumps in slow speed
. Technical Reference(s):
R-STM-005 3, Rev 13 , p. 40 of 7 6 AOP-0024, Thermal Hydraulic Stability, Rev 25 Attachment 1
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-005 3, Obj 2 Question Source:
Bank # October 2000 NRC exam #26 Question History:
Last NRC Exam October 2000 Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 2 QUESTION 2 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295004 AK1.04 IR 2.8 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF DC POWER : Effect of battery discharge rate on capacity
. Proposed Question:
A 125 VDC bus has experienced a trip of its battery charger supply breaker. Efforts to restore the battery charger to service have been unsuccessful. Assuming battery loading remains un
-changed; predict the change in bus voltage under these conditions.
Voltage will _____.
A. decrease sharply and then level off.
B. decrease in a linear fashion.
C. slowly lower and then drop sharply.
D. stair step down as individual cells are deplet ed. Proposed Answer:
C Explanation A. See "C" B. See "C" C. Correct - Battery voltage will decrease slowly then drop sharply as battery voltage reversal occurs as seen in SER 3
-99. D. See "C" Technical Reference(s):
RPPT-STM-0305-ILO Rev 2, Slide 95; SER 3-99 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0305, Obj 8 Question Source:
Bank # March 2014 NRC exam
- 48 Question History:
Last NRC Exam March 2014 Exam #48 Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.8 Comments: Appeared on one of last 2 NRC exams (1 of 2
)
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 3 QUESTION 3 Rev 1 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295005 AK2.07 IR 3.6 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: Reactor pressure control Proposed Question:
The plant is performing a Startup, operating at 30% power when a break occurs in th e EHC Hydraulic Supply line to the #2 Main Stop Valve Actuator. The Standby EHC pump starts on low pressure and the main turbine trip s. In this transient, with no operator action, the Turbine Bypass Valves will _(1)_ and the Reactor will scram due to _
(2)_. A. (1) open ;
(2) high pressure
. B. (1) fail open ;
(2) MSIV closure after a low steam line pressure.
C. (1) fail closed ;
(2) high pressure
. D. (1) open ;
(2) turbine stop valve fast closure
. Proposed Answer:
A Explanation A. Correct - BPVs remain available for automatic pressure control due to an independent hydraulic system; due to the turbine trip, pressure will rise and the BPVs will try to control pressure by opening
. If power were above 30.92%, then the Control Valve Fast closure scram would be enabled to send a trip signal to RPS, but this feature is not enabled so pressure will rise until the high pressure scram setpoint
. B. BPVs remain available for automatic pressure control due to an independent hydraulic system
. C. BPVs remain available for automatic pressure control due to an independent hydraulic system
. D. See A. Technical Reference(s):
R-STM-0509, pp. 15, 54 of 81 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0509, Obj 10d Question Source:
Bank # RBS-OPS-3 4 78 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.7 Comments: Replaced question to better match the KA
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 4 QUESTION 4 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295006 AK3.03 IR 3.8 Knowledge of the operational implications of the following concepts as they apply to SCRAM: Reactor pressure response Proposed Question:
A reactor plant has been operating at 100% power for an extended period when a scram occurs. Operators are executing AOP
-0001, Reactor Scram. For a period of several minutes after the scram, there will be a _(1)_ due to _(2)_. A. (1) drop in RPV level; (2) Recirc pumps shifting to slow speed B. (1) rise in RPV level; (2) rise in void content C. (1) rise in reactor pressure; (2) decay heat D. (1) drop in reactor pressure; (2) automatic operation of SRV's Proposed Answer:
C Explanation A. Part 1 is correct, but it is due to an automatic feature of the feed water level control system; not because of Recirc pumps shifting.
B. Part 1 is seen in plant operation due to the setpoint set down feature of the Feedwater level control system; it is not because of void content.
C. Correct - reactor pressure does go up and must be controlled with one of the following: the turbine, turbine bypass valves, or safety relief valves (IAW AOP
-1); the reason pressure rises for several minutes after a scram is because of decay heat: 7% for first 8
-10 seconds, then drops to about 1% after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
D. Both parts are incorrect: for part 1 see "
C" Technical Reference(s):
GFES - Reactor Operational Physics, pp. 63
-64 of 83; AOP
-0001 Proposed references to be provided to applicants during examination:
None Learning Objective:
GFES Reactor Operational Physics, Obj 31 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.1 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 5 QUESTION 5 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295016 A A1.06 IR 4.0 Knowledge of the operational implications of the following concepts as they apply to CONTROL ROOM ABANDONMENT: Reactor water level Proposed Question:
The Main Control Room must be abandoned due to Halon actuation; there is not a fire in the MCR. In accordance with AOP
-0031, Shutdown from Outside the Main Control Room, what systems will be initiated for RPV level control?
A. LPCS, HPCS, and RCIC B. HPCS, and RCIC only C. RCIC only D. All Division 1 ECCS systems Proposed Answer:
A Explanation A. Correct
- Subsequent actions for abandonment of the control room without a fire direct initiation of HPCS, LPCS, and RCIC.
B. See "A" this answer does not include LPCS C. See "A" this answer does not include LPCS or RCIC D. See "A" this answer does not include HPCS, and should not include RHR
-A Technical Reference(s):
AOP-0031, Rev 322 p.10 of 122 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-OPS-AOP-0031, Obj 4 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 6 QUESTION 6 Rev 1 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295018 A A2.04 IR 2.9 Ability to determine and/or interpret system flow as i t applies to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER
. Proposed Question:
With the plant operating at 100%, the following annunciators are received:
TURB CMPNT CLG WATER SYSTEM LOW PRESSURE TURB CMPNT CLG WTR SYS SURGE TK LOW LEVEL On H13-P870, the unit operator observes:
All 3 CCS pumps are running with elevated amps MWS-AOV132 TPCCW SURGE TK MAKE
-UP VALVE is OPEN Which of the following is the cause of the above?
A. CCS piping failure has caus ed a loss of inventory
. B. GSN-PCV1C , MANIFOLD REGULATOR (Nitrogen Pressure Control Valve to TPCCW Surge Tank) has failed closed. C. CCS-LT113, SURGE TANK LEVEL TRANSMITTER has failed low.
D. CCS-PV111, MINIMUM FLOW AND PRESSURE CONTROL VALVE, has failed OPEN. Proposed Answer:
A Explanation A. Correct
- A large leak would cause all the indications, low pressure
- starts stby pump, max amps as the pumps operate in runout condition, and loss of inventory causes the surge tank low level condition and opening of the makeup valve
. B. The TPCCW surge tank is blanketed with 18 psig nitrogen for NPSH and for corrosion control; if this valve fails closed it would lower system pressure, but not surge tank level
. C. The surge tank level transmitter failing low would cause the surge tank low level alarm and the opening of the surge tank makeup valve, but would not cause all pumps to run with maximum amps, nor would it cause a system low pressure alarm.
D. This would lower system pressure, but would not account for a low Surge Tank level.
Technical Reference(s):
ARP-P870-5 5-B01, C02, PID 09
-07A Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-116 Obj 2, 5 Question Source:
Bank # RBS Audit Exam, March 2010 #6 Question History:
Last NRC Exam NA Cognitive Level:
Memory o r Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.4 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 7 QUESTION 7 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295019 G2.4.47 IR 4.2 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. Proposed Question:
The plant is operating at 100% power when the following alarms annunciate:
P870-51A-A01, INSTRUMENT AIR COMPR ESSOR TROUBLE P870-51A-B01, SAS COMPRESSOR ALIGNED TO IA SYSTEM The Unit Operator notices the following:
Instrument Air Header Pressure on IAS
-PI105 is trending downward at a rate of 1 psig every 10 seconds.
A and B IAS compressors amber lights lit; C IAS compressor red light is lit Assuming the situation continues to degrade, which of the following represents the correct sequence of events?
A. Feedwater Reg Valves Lock
-up; then SAS-AOV133, Service Air Header Block Valve, closes; then MSIV's fail shut B. Instrument Air to the Lower Fuel Pools Gate Seals is replaced by the passive nitrogen back
-up system; then Feedwater Reg Valves Lock
-up; then Condensate and Heater Drain Pump Recirc valves fail open C. P870-51A-B02, "Instrument Air Header Pressure Low" Annunciator Alarms; then SAS-AOV133, Service Air Header Block Valve, closes; then Feedwater Reg Valves Lock
-up D. Feedwater Reg Valves Lock
-up; then P870-51A-B02, "Instrument Air Header Pressure Low" Annunciator Alarms; then MSIV's fail shut Proposed Answer:
C Explanation A. (also B. and D.) See "C" C. Correct - the cross tie valve opens at 113 psig; the low air header pressure alarm and the service air block valve closing both occur at 110 psig ; FRVs lock
-up at 85 psig; MSIV's go closed at 65 psig; Fuel Pool Gates switch to nitrogen at 32 psig.
Technical Reference(s):
AOP-0008, Loss of Instrument Air, Rev 37, p. 4 of 21 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-0527, Obj 4 Question Source:
Modified Bank
- RBS December 2008 NRC Exam #8 Question History:
Last NRC Exam December 2008 NRC Exam #8 Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 8 QUESTION 8 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295021 AK1.03 IR 3.9 Knowledge of the operational implications of the following concepts as they apply to LOSS OF SHUTDOWN COOLING : Adequate core cooling
Proposed Question: The plant has been in MODE 4 for 2 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, cooling down for a refueling outage with the following conditions
- RHR-B is in Shutdown Cooling mode of operation Recirc Pump A is tagged out Recirc Pump B is running in slow speed Reactor Coolant Temperatur RPV level +38" and steady RHR Pump B trips on motor overload and RHR Pump A will not start.
Which of the following operator actions would assure adequate core cooling?
A. Align RWCU for Alternate Shutdown Cooling
. B. Align for Main Steam Line Flooding C. Align SPC/ADHR in Configuration 1 D. Raise RPV level to greater than 75 inches Proposed Answer:
C Explanation A. B. C. Correct - (Configuration 1 takes a suction from SDC and discharges through LPCI
-C injection line)
D. Raising RPV level to +75 ensures natural circulation to prevent stratification and inadvertent pressurization, but does not assure adequate core cooling Technical Reference(s):
OSP-0041, Alternate Decay Heat Removal, Rev 306 pp 8, 18, 40 AOP-0051, Loss of Decay Heat Removal, Rev 312 pp 5&6 of 30 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-543, Obj 5 Question Source:
Bank 2010 Grand Gulf NRC exam Q#40 Question History: Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CF R Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 9 QUESTION 9 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295023 AK2.05 IR 3.5 Knowledge of the interrelations between REFUELING ACCIDENTS and the following:
Secondary containment ventilation Proposed Question:
Refueling operations are in progress when an irradiated fuel bundle is dropped in the spent fuel pool. The refueling team on the lower bridge reports bubbles rising from the dropped fuel bundle and the following annunciators are received in the MCR:
H13-P863-75A-H01, DIV I Fuel Bldg Exh PAM Gaseous Radn Alarm RMS-DSPL230-1GE005, Fuel Buil d Stack/Vent Exhaust A
- High RMS-DSPL230-2GE005, Fuel Build Stack/Vent Exhaust A
- High Which of the following describes the Fuel Building ventilation lineup after the conditions given above?
A. Fuel Building Ventilation is completely isolated B. Supply air is via normal supply fans and exhaust is through the Div 1 charcoal filter trains C. Supply air is via Fuel Receiving Area and exhaust is through both Div 1 and Div 2 charcoal filter trains D. Supply air is via Fuel Receiving Area and exhaust is through the Div 1 charcoal filter trains Proposed Answer:
C Explanation A. A high radiation condition in the fuel building does not isolate the ventilation system; it isolates the normal supply and exhaust fans, and starts the charcoal filtration system.
B. A high radiation condition in the fuel building isolates the normal supply and exhaust fans, and starts the charcoal filtration system.
C. Correct - RMS-RE5A is the instrument that drives all three annunciators; this instrument will start the Div 1 filter train only, however a low flow signal will also start the Div 2 train
. D. This instrument will send a start signal to the Div 1 filter train only, but both will start (see C)
Technical Reference(s):
R-STM-0406, pp 19 & 41 of 50, ARP
-H13-P863-75A-H01 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0406, Obj 11 Question Sourc e: New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 10 QUESTION 10 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295024 EK3.0 6 IR 4.0 Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE
- Reactor Scram Proposed Question:
What is the reason for the reactor scram that occurs due to a High Drywell Pressure? A. To minimize the possibility of fuel damage due to a reactor coolant pressure boundary leak by reducing the amount of energy being added to the coolant.
B. To ensure the Pressure Suppression function of the containment is maintained in
the event Emergency Depressurization is required.
C. To ensure that offsite dose limits are not exceeded during a reactor coolant pressure boundary leak.
D. To avoid clearing of the suppression pool vents due to high drywell pressure.
Proposed Answer: A Explanation A high drywell pressure condition results due to a leak of the primary system. Due to the loss of coolant, an inability to cool the fuel may result. A reactor scram occurs to minimize the energy being produced.
The pressure suppression function of the containment is based on containment pressure not drywell pressure. Offsite dose limits are prevented from being exceeded by the high drywell pressure containment isolation, not the high drywell pressure reactor scram. Although the scram signal will reduce the energy being leaked into the drywell, and may avoid clearing of the suppression pool vents, this is not the reason for the scram.
Technical Reference(s):
R-STM-0 508, RPS, Rev 6 p. 46 of 59 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0508 Obj 2 Question Source:
Bank # 2008 NRC Exam Q#11 Question History:
Last NRC Exam RBS 2008 Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content: 55.41.b.6 Comments: Rejected KA:
EK3.04 ; Randomly selected new KA statement EK3.06 (RBS has no guidance to Emergency Depressurize for a High DW D/P
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 11 QUESTION 11 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295025 EA1.05 IR 3.7 Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: RCIC: Plant
-Specific Proposed Question:
The plant is operating at 93% power.
A RCIC lube oil change has just been completed and RCIC has been started with the "slow-roll startup" section of SOP
-0035 in the RPV Pressure Control lineup (CST to CST). All MSIV's go closed and the reactor scrams.
What is the status of RCIC after the scram?
A. RCIC will isolate on steam supply line high flow.
B. RCIC will re
-align to inject to the RPV.
C. RCIC will isolate on high steam line pressure.
D. RCIC will remain in a CST to CST lineup.
Proposed Answer:
D Explanation A. High steam line flow can be a cause for RCIC isolation, but MSIV isolation does not cause this.
B. RCIC initiation occurs at RPV Level 2 (-43") C. Low Steam Line Pressure, not high, can cause a RCIC isolation D. Correct - There is no logic setpoints reached to cause a realignment of RCIC Technical Reference(s):
R-STM-0209, Rev 10, pp. 26,27,43 of 52
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0209, Obj 7, 12 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 12 QUESTION 12 Rev 1 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295026 EA2.03 IR 3.9 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor pressure Proposed Question:
Following a loss of offsite power, the plant is experiencing an ATWS. Suppression Pool level is currently 19 feet, 11 inches
. In accordance with the Pressure leg of EOP
-1A, RPV Control, ATWS, which of the following pressure band/suppression pool temperature combinations would result in the need for emergency depressurization?
RPV Pressure Supp Pool Temperature A. 800 - 1090 psig B. 800 - 1090 psig C. 500 - 700 psig D. 500 - 700 psig Proposed Answer:
B Explanation B. Correct - The pressure leg of EOP
-1A has an override step in RPA
-4 that directs maintaining RPV pressure below the HCTL: Suppression pool level, stated as 19'11" in the stem, requires the examinee to use the 19'6" line on the HCTL curve (not allowed to interpolate). The pressure bands are common given bands for the given plant condition and the top of the band must be used to determine if the HCTL will be reached. The lowest s.p. temperature that would require ED for 700 psig Technical Reference(s):
EOPs, Heat Capacity Temperature Limit Curve Proposed references to be provided to applicants during exam:
Heat Capacity Temperature Limit Curve
Learning Objective:
RLP-HLO-517, Obj 2 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 13 QUESTION 13 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295027 G2.4.20 IR 3.8 Knowledge of the operational implications of EOP warnings, cautions, and notes Proposed Question:
A. direct pressurization of containment from RCIC turbine exhaust during operation.
B. damage to the RCIC turbine during operation due to reduced lube oil cooling.
C. inability to monitor suppression pool temperature due to instrument being off scale high.
D. inability to monitor suppression pool level due to temperature being greater than instrumentation environmental ratings.
Proposed Answer:
B Explanation A. Direct pressurization is a concern with a high suppression pool level, not temperature (EOP Caution 4)
B. Correct- the lube oil and control oil for RCIC is cooled by process flow; the max allowable cooling water temperature for RCIC lube oil is 140 C. Monitoring
-555-4203) D. See explanation in C; EOP Caution 7 warns of the inability to trust suppression pool temperatures with suppression pool water level too low.
Technical Reference(s):
EOP-1 Caution 3 , Bases p. B 10 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-OPS-HLO-511, Obj 6
Question Source: New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 4 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.9 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 14 QUESTION 14 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295028 EK1.01 IR 3.5 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Reactor water level measurement
Proposed Question:
EOP-1 Caution 1, part 2 identifies RPV levels above which RPV level instruments may be used when the containment or drywell temperature near the reference legs is at the specified limits. At these elevated RUN TEMPERATURES, the instrument would ___
. A. continue to indicate level on-scale when actual RPV level went off
-scale high (above the indicating range).
B. fail off-scale low.
C. continue to indicate level on
-scale when actual RPV level went below the variable leg tap.
D. provide erratic level indication when actual RPV level went off-scale low due to loss of the variable leg.
Proposed Answer:
C Explanation A. Caution 1 deals with low RPV level, not high B. see C. C. Correct- indicated level would be on
-scale while actual level would be below the variable leg tap.
D. Erratic indication would be a sign of boiling in the reference leg.
Technical Reference(s):
EOP Caution 1, Bases p. B 2 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-OPS-HLO-511, Obj 6 Question Source:
Bank # RBS-NRC-706 Question History:
Last NRC Exam Audit Exam March 2008 Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.2 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 15 QUESTION 15 Rev 0 Examination Outline Cross-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295030 EK2.08 IR 3.5 Knowledge of the interrelations between LOW SUPPRESSION POOL WATER LEVEL and the following: SRV discharge submergence
Proposed Question:
For normal SRV operations , which of the following suppression pool level s is based on direct pressurization of containment air space?
A. 13 feet B. 15 feet, 5 inches C. 21 feet, 3 inches D. 21 feet, 6 inches Proposed Answer:
A Explanation A. Correct - This is the elevation of the top of the SRV discharge device below which opening of an SRV may cause pressurization of the containment air space B. 15'5" is 2 feet above the horizontal vents; this level is associated with a leak from the DW passing through the horizontal vents
. C. 21'3" is the SRV tail pipe level limit; operation of an SRV above this limit may cause damage to the SRV discharge lines.
D. 21'6" is based on the Pressure Suppression Pressure (PSP) Curve; the PSP pressure is a function of suppression pool level. Technical Reference(s):
EOP Bases p. B-6-56 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-514, Obj 5 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowle dge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.9 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 16 QUESTION 16 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295031 EK2.03 IR 4.2 Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following: Low pressure core spray Proposed Question:
An STP is being performed that directs the operator to close E21
-MOV F011, LPCS Min Flow Valve; all other valves are in their normal, standby positions. A valid RPV low level, Low Pressure Core Spray (LPCS) initiation signal is received.
Assuming no operator action, how will the LPCS system respond?
A. The LPCS Min Flow Valve will remain closed until the initiation signal is reset.
B. The LPCS Min Flow Valve will open until system flow rises above 875 gpm.
C. The LPCS Pump will not receive a start signal because of the pump protection logic permissive not being met.
D. The LPCS Pump will start and will eventually overheat
. Proposed Answer:
B Explanation A. The reset pushbutton is in the LPCS initiation sequence logic; the min flow valve is not part of this logic.
B. Correct - the design logic for the min flow valve is to OPEN when the pump breaker is closed and the sensed flow is less than 875 gpm.
C. The pump start logic does not include a permissive from the min flow valve D. The pump would overheat without a flow path providing at least 500 gpm; the min flow valve will open with the given conditions.
Technical Reference(s):
R-STM-0205 p p.13-14,16 of 33 ; ST P-205-4 201, Rev 301 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0205, Obj 4,9
Question Source:
Modified Bank #
RBS-OPS-5554 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 17 QUESTION 17 Rev 0 Examination Outline Cross
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Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295037 EA1.06 IR 4.1 Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Neutron monitoring system Proposed Question:
The plant was operating at 80% power when a transient in the Fancy Point Switchyard resulted in a Main Turbine trip. All control rods did not fully insert.
Current parameters:
Reactor Power 22.5% Reactor Pressure 900 psig RPV Level -50 inches (Wide Range
) Feed Flow 2.79 Mlbm/hr IRM and SRM detectors have been inserted Main Steam Bypass Valves are both full open Two SRVs are open To avoid exceeding the Heat Capacity Temperature Limit Curve, the CRS has directed pressure lowered to 700 psig using SRVs.
Immediately following the opening of SRVs, indicated reactor power will ___.
A. rise due to the lowering of the reactor coolant temperature adding positive reactivity.
B. rise due to the water level inside the core rising, causing more neutron moderation.
C. lower due to voiding as the water in the core flashes to steam.
D. lower due to the moderator temperature rising with the low flow in the core.
Proposed Answer:
C Explanation A. reactor power would lower; later as pressure stabilizes, and the water that flashed to steam is replaced with Feedwater, coolant temperature will add positive reactivity, but not immediately.
B. reactor power would lower; voiding causes less moderation C. Correct - lowering pressure will flash the water in the core and therefore causing void volume to rise. Neutron moderation would decrease, thereby lowering powe
- r. D. reactor power will lower, but it is due to voiding, not a temperature change Technical Reference(s):
GFES Components BC070 Proposed references to be provided to applicants during examination:
None Learning Objective:
BC07022 Question Source:
Bank # RBS-NRC-569 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.1 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 18 QUESTION 18 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295038 EA2.03 IR 3.5 Ability to determine and/or interpret the following as they apply to HIGH OFF
-SITE RELEASE RATE: Radiation levels Proposed Question:
The radioactivity release control leg of EOP
-0003, Secondary Containment and Radioactive Release Control is entered when radiation monitors reach the rate levels corresponding to the ___ action level defined in the Site Emergency Plan
. A. NOUE B. ALERT C. SITE AREA EMERGENCY D. GENERAL EMERGENCY Proposed Answer:
B Explanation A. NOUE is below the entry level for EOP
-0003 B. Correct- Entry into EOP
-0003 radioactive release leg corresponds to the ALERT action level; radiation monitors' setpoints are set accordingly.
C. See B D. See B Technical Reference(s):
EOP-00 03 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-515, Obj 2 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 19 QUESTION 19 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 600000 G2.4.11 IR 4.0 Knowledge of abnormal condition procedures Proposed Question: The plant is operating at 100% power. Standby Gas Treatment Filter Train "A" is being operated for surveillance testing.
The Auxiliary Building Operator reports smoke coming from the "A" Standby Gas Treatment Filter Train and the filter train case is glowing red.
Which one of the following describes the method to combat a fire in the Standby Gas Treatment Filter Train?
A. The Fire Protection System will initiate the automatic deluge system and fill the filter train with water.
B. The Fire Protection System will automatically open a deluge isolation valve , however, valves must be manually opened to admit water to the filter train.
C. The Fire Protection system Deluge Valve will have to be manually initiated via the pull station to admit water to the filter train.
D. The Fire Protection System at the filter train must be manually valved in to admit water to the filter train.
Proposed Answer:
D Explanation A. There is no automatic fire suppression for the GTS train B. The isolation valves must be manually opened, however there is no auto fire suppression for GTS C. Some deluge systems on site have this capability but not for the GTS train D. The isolation valves must be manually opened, no deluge valve actuation is needed.
Technical Reference(s):
Pre-Fire Strategy AB
-141-531 SGTS Filter A Room Fire Area AB
-14 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0257, Obj 4 Question Source:
Bank # RBS
-NRC-654 Question History:
Last NRC Exam NA Cognitive Level: Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 20 QUESTION 20 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 700000 AK1.03 IR 3.3 Knowledge of the operational implications of the following concepts as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Under-excitation
Proposed Question:
Under-excitation of the Main Generator results in ___
. A. Turbine rotor overheating.
B. Generator field overheating.
C. the Voltage Regulator shifting to Manual.
D. Generator armature overheating.
Proposed Answer:
D Explanation A. A loss of generator field causes "turbine torque oscillation" which leads to the Turbine rotor overheating. B. Over-excitation causes the Generator field to overheat.
C. An exciter field overcurrent causes the Voltage Regulator to shift to Manual.
D. Correct - Under-excitation causes overheating of the armature.
Technical Reference(s):
R-STM-0310, Rev 8
- p. 29 of 76 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0310, Obj 3,10 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.41b.5 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 21 QUESTION 21 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 295007 AK2.01 IR 3.5 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: Reactor/turbine pressure regulating system Proposed Question:
The plant is starting up and is currently operating at 60% power when the A Recirc. Flow Control valve ramps open
. How will the EHC pressure control system respond to this condition?
The EHC pressure control system will cause generator load to
_(1)_ and reactor pressure will
_(2)_ A. (1) rise (2) lower B. (1) rise (2) remain steady C. (1) lower (2) lower D. (1) lower (2) remain steady Proposed Answer:
B Exp lanation As the FCV opens at this power level reactor power will rise and reactor pressure will follow. The pressure regulating system will open the turbine control valves causing generator load to rise. Reactor pressure will remain the same as the turbine control valves open to control reactor pressure. B is the correct answer.
Technical Reference(s):
R-STM-0509, Rev 14 p.51 of 81 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0509, Obj 10 Question Source: New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.5 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 22 QUESTION 22 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 295008 AA2.05 IR 2.9 Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL: Swell Proposed Question:
River Bend was operating at full power with RCIC tagged out. A loss of offsite power occurred and HPCS recovered RPV level. Operators have closed the HPCS injection valve with an RPV water level of +20 inches. Current conditions are as follows:
MSIV's are closed Reactor pressure is 700 psig and rising at 10psig per minute Time after scram is 5 minutes Drywell pressure is 0.3 psid
What is the expected RPV water level response over the next 10 minutes and why?
A. RPV level will rise due to swell from decay heat B. RPV level will rise due to feed regulating valve leak by C. RPV level will lower due to cool down D. RPV level will lower due to injection being secured Proposed Answer:
A Explanation A. Correct
- Level rises due to expansion cause by the heat up from decay heat B. Feedwater is not available due to the loss of offsite power C. RPV water level will rise not lower D. RPV water level will rise not lower
Technical Reference(s):
RLP-HLO-16 8, Rev 06, Thermodynamics: Steam, pp. 35
-37 of 73 Proposed references to be provided to applicants during examination:
None Learning Objective: HLO-167, Obj 2 Question Source:
Modified Bank #
RBS-NRC-01108 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.14 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 23 QUESTION 23 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 295011 G2.4.04 IR 4.5 Ability to recognize abnormal indications for system operating parameters that are entry
-level conditions for emergency and abnormal operating procedures: High Containment Temperature Proposed Question:
Based on the following conditions, which one of the following describes the Emergency Operating Procedure(s) that should be entered?
Reactor Power 0% (all rods in)
RPV Pressure 970 psig RPV Level 20 inches Containment Temperature 92 Containment Pressure 0.25 psig Drywell Temperature Drywell Pressure 0.75 psid Annulus Differ. Pressure
-1.3 in WC A. EOP-1, RPV Control ONLY B. EOP-1, RPV Control AND EOP
-2, Primary Containment Control C. EOP-2, Primary Containment Control ONLY D. EOP-2, Primary Containment Control AND EOP
-3, Secondary Containment and Radioactivity Release Proposed Answer:
C Explanation A. No EOP-1 Entry Condition exists
. B. No EOP-1 Entry Condition exists C. Correct
- EOP-0002 Entry Condition is Containment Temperature above 90 D. No EOP-3 Entry Condition exists
. Technical Reference(s):
EOP-1, -2, and -3 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-514, Obj 3 Question Source:
Modified Bank # RBS-NRC-444 Question History: Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments: KA rejected because there is no Alarm or ARP for Cont. High Temperature.
Redrew from the 50 possible choices in the G2.4 category to get 2.4.4
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 24 QUESTION 24 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 295012 AK1.02 IR 3.1 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE : Reactor power level control Proposed Question:
Prior to Emergency Depressurization on High Drywell Temperature
, EOP-1 is directed to be entered to assure, if possible, ___. A. that RPV pressure is lowered as low as possible by steam loads before RPV depressurization is initiated.
B. that the reactor is scrammed and shutdown by control rod insertion before RPV depressurization is initiated C. that RPV water level is raised as high as possible by Feed water before RPV depressurization is initiated D. that the suppression pool is cooled and lowered by RHR before RPV depressurization is initiated Proposed Answer:
B Explanation A. While a lower pressure is desirable it is not specifically directed by the EOP prior to ED B. Correct- Per EOP Bases the reactor is scrammed and shutdown if possible to minimize the amount of energy sent to containment C. While a higher level is desirable to minimize inventory loss it is not specifically directed by the EOP prior to ED D. While a cooler suppression pool is desired it is not directed by EOP
-0001 Technical Reference(s):
EOP BASES p. B-8-6 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-514, Obj 5 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.9 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 25 QUESTION 25 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Ti er # 1 Group # 2 K/A # 295013 AK2.01 IR 3.6 Knowledge of the interrelations between HIGH SUPPRESSION POOL TEMPERATURE and the following: Suppression pool cooling
Proposed Question:
A safety relief valve has opened and AOP
-0035, SAFETY RELIEF VALVE STUCK OPEN has been entered. Following reduction in reactor power to 89%, the CRS has directed you to place RHR A in suppression pool cooling mode.
Why is suppression pool cooling initiated?
A. to obtain localized suppression pool temperature B. to minimize containment atmospheric activity C. to reject suppression pool level D. to establish bulk mixing of the suppression pool Proposed Answer:
D Explanation A. T.S. and TRM actions are based on average temperature not localized temperature B. SP cooling does not significantly affect containment activity levels and the reason for establishing SP cooling is to establish bulk mixing and to cool the pool C. While reject may become required, the reason for establishing SP cooling is to establish bulk mixing and to cool the pool D. Correct
- RHR A is placed in suppression pool cooling mode to establish bulk mixing of the suppression pool such that an average suppression pool temperature may be obtained
Technical Reference(s):
AOP-0035, Safety Relief Valve Stuck Open, Rev 19 p. 5 of 10 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-OPS-AOP035, Obj 4
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 26 QUESTION 26 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 295029 EK3.03 IR 3.04 Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL WATER LEVEL : Reactor SCRAM Proposed Question:
EOP-2, Suppression Pool Level Control, require s entry into EOP
-1, RPV Control if suppression pool level cannot be maintained below 21 feet. Entering EOP
-1 ___. A. will allow action s to take place that support terminating injection of water from sources external to containment.
B. will force a scram and shutdown of the reactor; this will make the requirement of maintaining Suppression Pool Level below 21 feet no longer applicable.
C. will allow for a reactor shutdown prior to flooding the drywell.
D. is required so the required emergency depressurization can be accomplished.
Proposed Answer:
A Explanation A. Correct
- EOP bases states that prior entry to EOP
-1 requires control of RPV level and reactor power and therefore, facilitates making the determination if a system taking suction from a source external to the containment is needed for RPV injection or to shutdown the reactor B. EOP-2 is still applicable when the reactor is shutdown C. EOP-1 entry on SP level is based on SRV component loading and injection source termination.
D. EOP-1 entry on SP level is based on SRV component loading and injection source termination Technical Reference(s):
EOP Bases p. B-8-2 1 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-514, Obj 5 Question Source:
Bank # RBS-NRC-20 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 27 QUESTION 27 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 295032 EA1.01 IR 3.6 Ability to operate and/or monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE : Area temperature monitoring system Proposed Question:
The plant is at 100% power.
Alarm P601/21A/H02 , AIR TEMP MON R608 RCIC RM TEMP HIGH is received. The CRS has directed you to monitor and trend the value for the high area temperature.
What temperature monitoring equipment will be used to obtain this information?
A. ERIS computer data via the Plant Data Server System B. H13-P632 Vent Diff Temperature recorder C. H13-P632 Area Temperature recorder D. Local temperature monitoring by a building operator Proposed Answer:
C Explanation A. The ERIS computer does not monitor this area of secondary containment B. This recorder only supplies unit cooler differential temperature indication C. The alarm response procedure directs the use of H13
-P632 Area Temperature recorder D. While this is not disallowed the alarm setpoint for this alarm is 144ºF and is a high dose area. This would be a safety and rad concern to enter the area to obtain temperatures. Technical Reference(s):
P601-21A-H02 PDS data point list Proposed references to be provided to applicants during examination:
None Learning Objective:
None Identified Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 28 QUESTION 28 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 203000 K4.0 3 IR 3.2 Knowledge of RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following: Pump minimum flow protection Proposed Question:
The reactor has been shut down due to a transient.
Current plant conditions are: Reactor pressure is 528 psig and slowly decreasing Reactor water level is
-120 inches and slowly decreasing Drywell pressure is 1.
78 psid and slowly increasing All low pressure ECCS systems are running as designed What is the status of the RHR C system? A. Injection valve (F042C) is open and pump minimum flow valve F064C is closed.
B. Injection valve (F042C) is open and pump minimum flow valve F064C is open. C. Injection valve (F042C) is closed and pump minimum flow valve F064C is closed
. D. Injection valve (F042C) is closed and pump minimum flow valve F064C is open
. Proposed Answer:
D Explanation A. and B are not correct, the injection valve, F064C will not open until reactor pressure is below 487 psig.
C. With the injection valve closed, system flow will be 0 gpm, the min flow valve is a normally open valve and is interlocked to open when flow is <1100 gpm.
D. Correct, see C Technical Reference(s):
R-STM-0 204 , RHR , Rev 11 pp. 12,14,15 of 63 Proposed references to be provided to applicants during examination: None Learning Objective:
RLP-STM-0204 , Obj 6 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.7 Comments: Rejected KA K4.09
- for inability to write psychometrically sound question; Randomly selected K4.03 Wording of KA suggests How do design features/interlocks provide for STPs
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 29 QUESTION 29 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 205000 K5.02 IR 2.8 Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) : Valve operation Proposed Question:
The plant is shut down. RHR A is operating in Shutdown Cooling mode (SDC) with an RPV water level of 75 inches.
Operation from the main control room of which of the following valves would lead to an OPDRV (Operations with a Potential to Drain the Reactor Vessel)?
A. E12-F004A, RHR PUMP A SUP PL SUCTION VALVE B. E12-F006A, RHR PUMP A SDC SUCTION VALVE C. E12-F024A, RHR PUMP A TEST RTN TO SUP PL D. E12-F042A, RHR PUMP A LPCI INJECTION ISOL VALVE Proposed Answer:
C Explanation A. Interlocks prevent opening of E12
-F004A with E12
-F006A already open B. closure of E12
-F006A would cause to the RHR pump to trip but would not cause an OPDRV C. Correct
- There are no interlocks preventing operation and opening of E12
-F024 would direct RPV water to the suppression pool D. Open of E12
-F042A would redirect water through the LPCI injection line but would not cause an OPDRV Technical Reference(s):
R-STM-0204, RHR, Rev 11 pp. 16,21 of 63 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0204, Obj 6 Question Source:
Bank # RBS-NRC-182 Question History:
Last NRC Exam NRC Exam 7/1997 Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 30 QUESTION 30 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 209001 K6.03 IR 3.3 Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM : Torus/suppression pool water level Proposed Question:
The Low Pressure Core Spray (LPCS) pump should not be run with suction from the suppression pool if level is less than 13 ft 3in, except in accordance with the EOPs because of the need to ____.
A. install an EOP
-0005 enclosure to allow operation
. B. assure adequate net positive suction head
. C. provide cooling to the pump bearings
. D. maximize pump injection time following a LOCA
. Proposed Answer:
B Explanation A. There are NO interlocks between suppression pool level and LPCS operation B. Correct- SOP-0032 Precaution and Limitation 2.10 Lists 13'3" as the necessary level to assure NPSH C. The pump bearings are NOT cooled by system flow D. While a high suppression pool level will allow longer pump run time prior to loss of NPSH it is not t he reason behind the limitation of 13'3" Technical Reference(s):
SOP-0032, LPCS, Rev 23 P&L 2.10 (p. 3 of 30)
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0205, Obj 8
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 31 QUESTION 31 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 209001 A3.03 IR 3.5 Ability to monitor automatic operations of the LOW PRESSURE CORE SPRAY SYSTEM including: System pressure Proposed Question:
The Low Pressure Core Spray (LPCS) system is running in the test return to the suppression pool mode at 5050 gpm.
A leak has caused drywell pressure to increase to 1.95 psid.
Reactor water level is
-62 inches Reactor pressure is 450 psig Which of the following identifies the expected AUTOMATIC response?
A. The Test Return Valve to the Suppression Pool (E21
-F012) closes, Injection Isolation Valve (E21
-F005) opens and discharge pressure rises
. B. The Test Return Valve to the Suppression Pool (E21
-F012) closes
, Injection Isolation Valve (E21
-F005) opens and discharge pressure lowers
. C. The Test Return Valve to the Suppression Pool (E21
-F012) remains open , Injection Isolation Valve (E21
-F005) remains closed and discharge pressure rises. D. The Test Return Valve to the Suppression Pool (E21
-F012) remains open, the Injection Isolation Valve (E21-F005) remains closed and discharge pressure lowers. Proposed Answer:
A Explanation A. Upon a LOCA signal of 1.68 drywell diff pressure the LPCS system automatically aligns to inject into the RPV, the test return valve closes, if open, the injection valve opens, if RPV pressure is below 487 psig and the min flow valve opens if flow drops below 875 gpm. Since the stem RPV pressure is below the injection valve interlock pressure but above the pump shut off head (282 psig) the injection valve will open but the pump will not flow into the RPV. The min flow valve will open, flow will drop from the 5050 gpm and discharge pressure will rise. For these reasons all other choices are incorrect.
B, C, D. See A Technical Reference(s):
R-STM-0205, LPCS, Rev 5 pp.
17-18 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0205, Obj 9 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension o r Analysis 4 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 32 QUESTION 32 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 209002 A1.08 IR 3.1 Ability to predict and/or monitor changes in parameters associated with operating the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) controls including: System lineup: BWR
-5,6 Proposed Question:
Following a loss of normal Feedwater, HPCS initiated and is restoring level. At 0 inches in the RPV
, the operator close s the HPCS injection valve. How will the HPCS Discharge Pressure and HPCS Flow Rate parameters change?
Discharge Pressure Flow Rate A. Rise Rise B. Rise Lower C. Lower Rise D. Lower Lower Proposed Answer:
B Explanation A. See B B. Correct
- as the pumps discharge is constrained to the min flow capacity of 500 gpm, discharge pressure will rise and flow rate will lower C. See B D. See B Technical Reference(s):
R-STM-0203, HPCS, Rev 8 pp.15,16,20 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0203, Obj 9 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.8 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 33 QUESTION 33 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 211000 K5.06 IR 3.0 Knowledge of the operational implications of the following concepts as they apply to STANDBY LIQUID CONTROL SYSTEM: Tank level measurement Proposed Question:
An ATWS has occurred and the CRS has directed SLC injection. The hard card directs you to monitor SLC tank level and to report when Hot Shutdown Boron Weight is achieved.
If the initial SLC tank level is 2666 gallons, which of the following is the highest tank level at which Hot Shutdown Boron Weight should be reported as having been injected?
A. 1986 gallons B. 2036 gallons C. 1866 gallons D. 1966 gallons Proposed Answer:
D Ex planation A. See D B. See D C. See D D. Correct
- Note above tank levels in OSP
-0053 states that "When tank level falls between values, then the smaller value should be used", with an initial value of 26 66 gallons the level must be the highest level that is below 1974 gallons which is 1966 gallons
Technical Reference(s):
OSP-0053, Emergency & Transient Response, Attachment 13, Rev 22 Proposed references to be provided to applicants during examination:
OSP-53, Attach. 13 p. 2 of 2 only
Learning Objectiv e: RLP-STM-0201, Standby Liquid Control, Obj 1 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 34 QUESTION 34 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 211000 A2.03 IR 3.2 Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A.C. power failures Proposed Question:
Following a LOSS of RSS #1 Plant conditions are as follows:
All control rods did not insert Reactor power is 38% Division 1 DG started but tripped Division 2 DG is in standby Division 3 DG is running supplying its bus Which of the following is a correct statement with regard to SLC operation?
A. SLC injection is required using SLC Pump A B. SLC injection is required using SLC Pump B C. EOP-0005 Enclosure 15 for alternate SLC is required D. SLC injection is not required Proposed Answer:
B Explanation A. No power is available to SLC pump A B. Correct - 38% power exceeds the drain and bypass valve capacity so SLC injection is required; power is only available to Div 2 (SLC pump B)
C. Power is available to SLC pump B, Alternate SLC injection is not warranted.
D. See B Technical Reference(s):
R-STM-0201, SLC, Rev 7 pp. 22,32 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-513, Obj 3 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 35 QUESTION 35 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 212000 A3.06 IR 4.2 Ability to monitor automatic operations of the REACTOR PROTECTION SYSTEM including
- Main turbine trip: Plant
-Specific Proposed Question:
Following a main turbine trip from 100% power, the reactor will scram first on which of the following signals?
A. Low water level due to shrink B. High pressure due to control valve closure C. High neutron flux due to pressure rise D. Turbine stop valve position Proposed Answer:
D Explanation A. See D B. See D C. See D D. Correct
- This scram anticipates the rise in reactor pressure, neutron flux and heat flux resulting from the loss of heat sink on a turbine trip
Technical Reference(s):
R-STM-0508, RPS, Rev 6 p. 47 of 59
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0508, Obj 2
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.6 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 36 QUESTION 36 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 215003 A4.0 5 IR 3.4 Ability to manually operate and/or monitor in the control room:
Trip bypasses Proposed Question:
A plant startup is in progress. Reactor power is on range 5 of the IRMs.
IRM-F amplifier card fails, causing an INOP condition on IRM
-F and a 1/2 Scram.
The CRS directs you to bypass IRM
-F and reset the 1/2 Scram.
To bypass the IRM you should:
A. Rotate the IRM draw mode switch to the STANDBY position B. Fully withdraw IRM
-F and down range to range 1 C. Rotate the IRM draw mode switch to the TRIP/TEST position D. Place the IRM bypass Select Switch to the IRM
-F position Proposed Answer:
D Explanation A. Rotating this switch to standby generates an INOP condition for the IRM and does not change the status B. This will only bypass the IRM if the reactor mode switch is in RUN C. See A D. Correct
- This is the only method for the control room operator to bypass IRM
-F in these conditions Technical Reference(s):
SOP-0074, Neutron Monitoring, Rev 306 R- STM-0503 , Neutron Monitoring, Rev 8 p. 42 of 112
Proposed references to be provided to applicants during examination:
None Learning Objective:
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.2 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 37 QUESTION 37 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 215004 K2.01 IR 2.6 Knowledge of electrical power supplies to the following:
SRM channels/detectors Proposed Question:
The power supply to SRM '
C' detector is ___. A. RPS A B. RPS B C. VBS-PNL01 A1 D. VBS-PNL01B1 Proposed Answer:
C Explanation A. Correct - RPS A is the power supply to SRM A & C B. RPS B is the power supply to SRM B & D C. VBS-PNL01A1 is the power supply to neutron monitoring recorders D. VBS-PNL01B1 is the power supply to neutron monitoring recorders Technical Reference(s):
R-STM-0503, Neutron Monitoring, Rev 8 p.84 of 112 Proposed references to be provided to applicants during examination
- None Learning Objective:
R-STM-0503, Obj 7 Question Source:
Modified Bank # Nov 2010 Audit Q#36 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.2 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 38 QUESTION 38 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 215004 G2.4.4 IR 4.5 Knowledge of annunciator alarms, indications, or response procedures. SRMs Proposed Question:
During a reactor startup, the ATC has been withdrawing SRMs in accordance with GOP-0001, Plant Startup. All SRMs except C indicate full out; SRM C is reading offscale high and indicates "driving out." T he following alarm is in:
SRM UPSCALE OR INOPERATIVE Which of the following additional conditions will initiate a Control Rod Withdraw Block?
A. Mode switch in RUN and all IRMs on Range 8 B. Mode switch in STARTUP/HOT STBY and all IRMs on Range 7 C. Mode switch in STARTUP/HOT STBY and SRM C Bypassed D. Mode switch in RUN and SRM C Selector switch in Standby Proposed Answer:
B Explanation A. and D. The rod withdrawal block is bypassed when the mode switch in run B. Correct
- If IRMs are not on Range 8 or above in Mode 2, an SRM UPSC or INOP initiates a rod withdrawal block .
C. The rod withdrawal block is bypassed when the affected SRM is bypassed Technical Reference(s):
ARP-P680-05-C05 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0503, Obj 5,7 Question Source: Modified Bank # RBS
-NRC-53 Question History:
Last NRC Exam NA (original question was on the 1997 NRC exam)
Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.6 Comments: Rejected KA: There is no AOP or EOP entry conditions having to do with SRMs.
The only G2.4 category applicable to SRMs is G2.4.31
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 39 QUESTION 39 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 215005 K1.09 IR 3.6 Knowledge of the physical connections and/or cause
-effect relationships between AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and the following:
Reactor recirculation system: BWR
-5,6 Proposed Question:
Core flow rate is provided to the APRMs for use by the flow control trip reference cards to establish control rod withdrawal blocks and scram trip setpoints, this flow rate is obtained from which of the following sources?
A. the calibrated jet pumps B. the above and below core plate differential pressure C. the elbow taps on both recirculation loops D. the steam flow feed flow differential summer Proposed Answer:
C Explanation A. See C B. See C C. Correct
- A d/p signal is provided to the APRMs from elbow taps on both recirculation loops D. See C Technical Reference(s):
R-STM-0503, Neutron Monitoring, Rev 7 p. 60 of 112
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0503, Obj 27 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.2 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 40 QUESTION 40 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 217000 K2.01 IR 2.8 Knowledge of electrical power supplies to the following:
Motor operated valves Proposed Question:
The normal power supply to E51
-F063, RCIC STEAM SUPPLY INBD ISOL VALVE is
____. A. EHS-MCC2L B. EHS-MCC2D C. ENB-MCC1 D. BYS-SWG01B Proposed Answer:
B Explanation A. EHS-MCC2L is an alternate source of power to E51
-F063 utilized for AOP
-0031 B. Correct
- EHS-MCC2D is the normal power source to E51
-F063 C. DC supply ENB
-MCC1 is the source of power to many E51 MOVs but NOT to E51
-F063 D. DC supply BYS
-SWG01B is the source of power to the RCIC gland seal compressor NOT to E51
-F063 Technical Reference(s):
R-STM-0209, RCIC, Rev 10 p. 42 of 52 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0209, Obj 13 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.8 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 41 QUESTION 41 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 218000 K3.02 IR 4.5 Knowledge of the effect that a loss or malfunction of the AUTOMATIC DEPRESSURIZATION SYSTEM will have on following:
Ability to rapidly depressurize the reactor Proposed Question:
The plant is operating in the Emergency Operating Procedures following a significant transient. During the transient, a short resulted in the loss of ENB
-PNL02A. Plant conditions require Emergency Depressurization per the Emergency Operating Procedures.
Which of the following represents the method that should be used to accomplish Emergency Depressurization?
A. At H13-P601, open 7 ADS/SRVs.
B. At H13-P631, open 7 ADS/SRVs.
C. Arm and depress the Division 1 ADS Manual Initiate pushbuttons. D. Alternate depressurization methods listed in the EOPs should be utilized due to SRV failure.
Proposed Answer:
B. Explanation: A. Loss of power to ENB
-PNL02A prevents use of Div 1 SRV solenoids; the switches on P601 are Div 1.
B. Correct
- Div 2 solenoids are still available to open the SRVs.
C. Div 1 solenoids are de
-energized.
D. This would only be required if SRVs could not be opened. Div 2 solenoids are still available.
Technical Reference(s):
R-STM-0202, ADS, Rev 2 pp. 20,21 of 36 R-STM-0109, Main Steam, Rev 13 p. 12 of 95 Proposed references to be provided to applicants during examination:
None Learning Objective:
R-STM-0202 Obj. 6,12 Question Source:
N ew Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledg e Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41b.3 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 42 QUESTION 42 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 218000 A4.02 IR 4.2 Ability to manually operate and/or monitor in the control room:
ADS logic initiation Proposed Question:
The plant has scrammed due to a loss of offsite power.
Neither HPCS nor RCIC will start.
Approximately 5 minutes later, RPV water level decreases below
-143 inches, the "DIV 2 ADS LOGIC TIMER INITIATED" annunciator illuminates.
The Unit Operator is directed to "INHIBIT ADS" per EOP
-0001. Later the Operator ARMS and Depresses the ADS B MANUAL INITIATION pushbuttons.
What is the response of the ADS System in this situation?
ADS will initiate: A. immediately, if any Div 2 low pressure ECCS subsystem pressure permissive is satisfied.
B. in 105 seconds, if any Div 2 low pressure subsystem pressure permissive is satisfied.
C. immediately, regardless of low pressure ECCS subsystem status.
D. In 105 seconds, regardless of low pressure ECCS subsystem status.
Proposed Answer:
A Explanation A. Correct
- ADS Logic only requires subsystem pressure permissive prior to manual initiation B. See A C. See A D. See A Technical Reference(s):
R-STM-0202, ADS, Rev 2 p.38 of 41 (Figure 2)
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0202, Obj 7 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.8 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 43 QUESTION 43 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 223002 K4.01 IR 3.0 Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT
-OFF design feature(s) and/or interlocks which provide for the following:
Redundancy Proposed Question:
Why do systems penetrating containment which are part of the reactor coolant boundary have redundant isolation valves? A. To allow one valve to be removed from service for maintenance activities.
B. To ensure that multiple failure will not prevent a t least a single valve isolation of the process line C. To allow for manual isolation from the control room in case the automatic isolation does not work D. To ensure that a single failure will not prevent at least a single valve isolation of the process line.
Proposed Answer:
D Explanation A. See D B. Multiple failures could still prevent penetration isolation C. Both valves must be automatic isolations per design criteria D. Correct
- 10CFR50, General Design Criteria 55 Technical Reference(s):
R-STM-0058, CRVICS, Rev 9 p. 5 of 63
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0058, Obj 11 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.3 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 44 QUESTION 44 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 239002 K5.04 IR 3.3 Knowledge of the operational implications of the following concepts as they apply to RELIEF/SAFETY VALVES :
Tail pipe temperature monitoring Proposed Question: The plant had been operating at 100% power when a severe over
-pressure transient resulted in ALL Safety Relief Valves (SRV) opening in the "relief" mode.
RPV pressure peaked at 1200 psig Current reactor pressure is 500 psig and lowering One SRV remains stuck open Which of the following describes the resulting tailpipe temperature trend as the plant cools down and depressurizes through the stuck open SRV? (Assume containment pressure is 0 psig and remains constant.)
SRV tailpipe temperature will
___ and will then slowly fall, following reactor pressure during the depressurization below 500 psig.
A. Start at 285°F, currently 280°F B. Start at 285°F, currently 330°F C. Start at 330°F, currently 280°F D. Start at 330°F, currently 320°F Proposed Answer:
B Explanation A, Correct starting point, however tail pipe temperature will rise to 330ºF as pressure falls.
B. Correct
- At onset of this event the tail pipe temperature would be 285ºF ( the isentropic value for 1200 psig and ~25 psia)(14.7 plus 10 feet of water over the tail pipe quencher). Tail pipe temperature will rise to 330ºF as pressure fa lls to 500 psig.
C and D cannot be true due to temperature dropping Technical Reference(s):
HLO-168 Steam Tables Proposed references to be provided to applicants during examination:
Mollier Diagram (large)
Learning Objective:
Question Source:
Modified Bank #
RBS-NRC-459 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content
- 55.41.b.14 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 45 QUESTION 45 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 259002 K6.03 IR 3.1 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM : Main steam flow input Proposed Question:
While operating at 100% power, the 'C' Main Steam Line flow transmitter fails low.
What is the expected response of the Feedwater level control system with no operator actions taken?
A. RPV level will lower until it reaches Level 3 and the reactor scrams.
B. RPV level will lower and stabilize at a new lower level above Level 3.
C. RPV level will rise and stabilize at a new higher level below Level 8.
D. RPV level will rise until it reaches Level 8 and the reactor scrams.
Proposed Answer:
B Explanation A. Only first part is correct; see B B. Correct - A loss of 1 out of 4 steam flow transmitters results in a sensed reduction in steam flow of 25%; this sensed reduction will cause the FWLC system to close the Feed Reg Valves and consequently RPV level will lower. Level will stabilize below its normal level (but above the Level 3 scram setpoint) because the FWLC system is level dominant.
C. This is the expected response for a single steam line transmitter failing high D. This is the expected response for a single feed flow transmitter failing low Technical Reference(s):
R-STM-0107, Feedwater, Rev 27 p. 71 of 105
Proposed references to be provided to applicants during examination:
None Learning Objective: RLP-STM-0107, Obj B14 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.4 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 46 QUESTION 46 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 261000 A1.01 IR 2.9 Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including:
System flow Proposed Question:
A leak has occurred in the Drywell and Drywell pressure is 2.14 psid and stable. Both trains of Standby Gas Treatment (GTS) started and consequently OSP-0053 Hard Card Attachment 21, Operating Auxiliary Building Ventilation was used to reduce to one train of SBGT running; GTS-A is running. Subsequently, a loss RSS#1 occurred and the Div 1 Diesel Generator failed to start.
What is the status of GTS
-B following the loss of RSS#1?
A. GTS-B may only be started in High Volume Purge mode due to power failure.
B. GTS-B must be manually initiated due to manually securing.
C. GTS-B will restart automatically due to low flow in GTS
-A. D. GTS-B will restart automatically due to undervoltage trip of GTS
-A. Proposed Answer:
C Explanation A. Power is only lost to Div 1; Div 2 is still available B. Securing per the hard card places GTS
-B in standby; C. Correct - DW 1.68 is still locked in, so when low flow occurs in Div 1 due to the loss of power, then the Div 2 will automatically restart.
D. The interlock that will start GTS
-B is low flow in the running GTS; not undervoltage.
Technical Reference(s):
R-STM-0257, Standby Gas Treatment, Rev 5 p. 15 of 28 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0257, Obj 5,12
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 47 QUESTION 47 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 262001 A2.11 IR 3.2 Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Degraded system voltages
Proposed Question:
The plant is operating at 100% power when alarm P808
-86A, Grid Trouble is received.
SPI-REC102 on P808 indicates 230 KV system voltage at Fancy Point to be 223.3 KV.
The SOC contacts the control room and indicates that a transient has occurred and Fancy Point stability cannot be determined at this time.
The synchroscope for Division I D/G is indicating erratically.
(1) What is the impact to plant equipment?
(2) What actions should be taken to mitigate this event?
A. (1) Amps will increase on the running equipment.
(2) Perform a normal start of the D/G, parallel it to offsite power and disconnect the bus from the grid
. B. (1) Amps will increase on the running equipment.
(2) Perform an emergency start of the D/G and open the safety related bus supply breaker.
C. (1) Amps will decrease on the running equipment.
(2) Perform an emergency start of the D/G and open the safety related bus supply breaker. D. (1) Amps will decrease on the running equipment.
(2) Perform a normal start of the D/G, parallel it to offsite power and disconnect the bus from the grid
. Proposed Answer:
B Explanation Plant indications show that Fancy Point is NOT stable(second note on page 64 of AOP
-64) A. Would be correct if Fancy Point was stable B. Correct per steps:5.5.3.1 and 5.5.3.4 C. and D. Running equipment amps will rise as voltage lowers. Normal voltage is 230KV the stem gives 223.3KV Technical Reference(s):
AOP-006 4 , Degraded Grid , Rev 6 p 8 of 11 Proposed references to be provided to applicants during examination:
AOP-0064 Degraded Grid
, Rev 6 Learning Objective:
_________________________ (As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content:
55.41.b.5 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 48 QUESTION 48 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 262002 A3.01 IR 2.8 Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including: Transfer from preferred to alternate source Proposed Question:
BYS-INV01A displays the following information:
Rectifier Output 0 amps Battery Output Normal at 135 VDC Inverter Output Normal at 122 VAC What is the status of the loads normally supplied by this inverter?
A. Loads are currently de
-energized B. Loads are currently being supplied by the battery via the inverter C. Loads are currently being supplied by the normal AC source via the inverter D. Loads have automatically swapped to the alternate AC source Explanation A. The inverter output has power, so the loads have power B. Correct - The normal AC source is unavailable based on the rectifier output; output of the battery and inverter reveal that the battery is supplying power to the inverter and the inverter is powering the loads C. Rectifier output reveals that the normal AC power source is de
-energized D. Alternate AC power source is provided through a Manual Bypass Switch (which has not been turned)
Proposed Answer:
B Technical Reference(s):
R-STM-0300, AC Electrical Distr., Rev 26 pp. 28
-29 of 105 Proposed references to be provided to applicants during examination:
None Learning Objective: RLP-STM-0300, Obj 12
Question Source:
Bank # RBS Nov 2008 Audit Q#49
Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 49 QUESTION 49 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 263000 A4.01 IR 3.3 Ability to manually operate and/or monitor in the control room:
Major breakers and control power fuses: Plant
- Specific Proposed Question:
A breaker test is being perform ed on ENS-ACB03, E12
-C002A RHR A PUMP breaker. The breaker will be in the TEST position with control power fuses INSTALLED.
T he breaker will then be CLOSED.
Which of the following represents the expect ed control room light indications for the RHR Pump A breaker when the test conditions mentioned above are established?
A. Red light OFF, Green light OFF, White light OFF B. Red light OFF, Green light OFF, White light ON C. Red light ON, Green light OFF, White light OFF D. Red light ON, Green light OFF, White light ON Proposed Answer:
C Explanation A. B. and D. See C C. While in the test position the control room breaker indication for the red and green lights will be the same as the local indication however the white power available light will not be lite even with the control power fuses installed unless the breaker is in the 'connect' position. All other combination of light indication s are incorrect
Technical Reference(s):
OSP-0052, Breaker Racking, Rev 18 p Proposed references to be provided to applicants during examination:
None Learning Objective:
Not available Question Source:
Bank # RBS-OPS-07756 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 50 QUESTION 50 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 264000 K1.05 IR 3.2 Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY GENERATORS (DIESEL/JET) and the following:
Emergency generator fuel oil supply system Proposed Question:
The correct fuel oil supply flowpath for operation of the Div 1 Standby Emergency Diesel Generator is from its associated storage tank thr ough the _____, to the injectors.
A. fuel oil transfer pump, strainer, duplex filter, fuel oil day tank B. strainer, fuel oil transfer pump, fuel oil day tank, duplex filters C. strainer, fuel oil transfer pump, fuel oil day tank, booster pump D. fuel oil transfer pump, strainer, fuel oil day tank, booster pump Proposed Answer:
D Explanation A. the filter is located after the day tank B. the strainer is after the transfer pump C. the strainer is after the transfer pump D. Correct - The full flowpath of fuel oil is storage tank, transfer pump, day tank, duplex strainer, booster pump, duplex filter, injector pump, injectors
Technical Reference(s):
R-STM-0309S, Rev 14, pp. 17 & 20 of 117
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0309S, Obj 2
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.8 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 51 QUESTION 51 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 264000 G2.1.31 IR 4.6 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.
Emergency Diesel Generators Proposed Question:
The Div 2 Emergency Diesel Generator is operating and tied to ENS
-SWG 1B. The unit operator rotates the DG Voltage Regulator Control Switch to raise. Which indicati on shows the largest change (1) if offsite is paralleled to the bus, and (2) if offsite is NOT paralleled to the bus.
(1) (2) A. Reactive Load (KVAR)
Frequency (Hz)
B. Real Load (KW)
Voltage (VAC)
C. Reactive Load (KVAR)
Voltage (VAC)
D. Real Load (KW)
Frequency (Hz)
Proposed Answer:
C Explanation A. (1) is correct however, Frequency is controlled by the Vovernor control switch when not in parallel B. Generator load (KW) is adjusted with the Governor Control switch when the EDG is in parallel C. Correct - The voltage regulator control switch varies excitation current and will adjust generator reactive load when in parallel and adjusts voltage when not in parallel.
D. Generator load (KW) is adjusted with the Governor Control switch when the EDG is in parallel Technical Reference(s):
R-STM-0309S, Rev 13 pp. 43
-44 of 117 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0309S, Obj 8 Question Source:
Modified Bank #
RBS-OPS-3255 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 52 QUESTION 52 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Gro up # 1 K/A # 300000 K2.0 1 IR 2.8 Knowledge of electrical power supplies to the following:
Instrument air compressor Proposed Question:
The Instrument air compressors A, B, and C are powered from ___, ___, and ___
respectively
. A. NJS-SWG 1G, 1H, and 1F B. NJS-SWG 1E, 1J, and 1K C. NHS-MCC 1L, 1M, and 1M D. NHS-MCC 1L, 1L, and 1M Proposed Answer:
A Explanation A. Correct
- these are the power supplies to the IAS compressors B. these are the power supplies to the Service Air compressors C. these are the power supplies to the IAS trim coolers D. these are the power supplies to the SAS trim coolers Technical Reference(s):
R-STM-0121, Rev 16 p. 9 of 69
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0121, Obj 10 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.4 Comments: rejected KA K2.02; re
-selected K2.01
- RBS has diesel operated emergency air compressor
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 53 QUESTION 53 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 400000 K3.01 IR 2.9 Knowledge of the effect that a loss or malfunction of the CCWS will have on the following: Loads cooled by CCWS Proposed Question:
The Division 2 isolation valves in the Reactor Plant Component Cooling Water System (CCP) have closed on a 56 psig isolation signal.
What equipment is still receiving cooling water from the CCP system?
A. Recirc pumps, RWCU NRHXs, RWCU pumps B. RWCU pumps, RHR Pump A Seal Coolers, RWCU NRHXs C. RWCU pumps, Reactor Plant Sampling panel, Fuel Pool Cooler 1A D. Recirc pumps, RHR Pump A Seal Coolers, Reactor Plant Sampling panel Proposed Answer:
A Explanation A. Correct
- The Safety loop isolates which secures cooling to the RHR Pump Seal Coolers, Fuel pool coolers, and CRD pumps. Cooling is not isolated to the three loads listed.
B. RHR Pump Seal Coolers (both divisions) are isolated as part of the safety loop C. Fuel Pool Coolers (both divisions) are isolated as part of the safety loop D. RHR Pump Seal Coolers (both divisions) are isolated as part of the safety loop Technical Reference(s):
R-STM-0115, Rev 6 pp. 12
-14 of 35 and Figure 1 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0115, Obj 4,8,11 Question Source:
Bank # RBS-OPS-1789 Question Histor y: Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 54 QUESTION 54 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 201003 A4.02 IR 3.5 Ability to manually operate and/or monitor in the control room: CRD mechanism position: Plant
-Specific Proposed Question:
The ATC operator just selected a control rod to move in accordance with an approved reactivity maneuver plan when the Control Rod Withdrawal Block Alarm annunciates.
The RC & IS Operator Control Module (OCM) also displays the following: DATA FAULT pushbutton is backlit CH DATA 1 / CH DATA 2 are consecutively lighting and extinguishing CHANNEL DISAGREE is backlit The rod position displayed on the full core display for the selected rod is flashing alternately between 08 and
- - .
What operator action would allow this rod to be moved?
A. Isolating the rod electrically within the Rod Gang Drive Cabinet B. Depressing the DATA MODE pushbutton on the OCM C. Depressing the DATA SOURCE pushbutton on the OCM D. Depressing the ENTER SUBST pushbutton on the OCM Proposed Answer:
D Explanation A. Electrically isolating the rod in RGDC will prevent rod movement; bypassing the rod in RACS (Rod Action Control System would allow rod movement.
B. This pushbutton works in conjunction with the Data Source pushbutton to allow you to choose either single channel or alternating channel(s) to display on the full core display; this would make the alternating rod position indication to stop, but not clear the rod block.
C. This pushbutton allows you to choose which of the the 2 channels to display on the full core display; pushing this button would show first one channel then pushing it again would show the other.
D. Correct - This pushbutton forces the single channel of bad data to be replaced with the other channels good data; thereby allowing rod movement.
Technical Reference(s):
R-STM-0500, Rev 3 pp. 19
-25 of 46; SOP-0071, Rev 29, Section 5.12 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0500, Obj 7,9 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.6 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 55 QUESTION 55 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 201005 A1.01 IR 3.2 Ability to predict and/or monitor changes in parameters associated with operating the ROD CONTROL AND INFORMATION SYSTEM (RCIS) controls including:
First stage shell pressure/turbine load: BWR
-6 Proposed Question:
Given the following plant conditions:
45% 480 MWe Power ascension is in progress. The next step of the Reactivity Maneuvering Plan is to select and continuously withdraw control rod 28
-49 from position 12 to position 24.
Just prior to withdrawing the rod, the Main Turbine First Stage Shell Pressure transmitter output signal fails upscale.
When the withdraw button is pushed, control rod 28
-49 will ___.
A. remain at position 12 B. withdraw to position 16 and settle C. withdraw to position 20 and settle D. withdraw to position 24 and settle Proposed Answer:
B Explanation A. If turbine first stage shell pressure was indicating failed low, the Rod Pattern Controller logic would be enabled and could cause a rod withdraw blow (depending on the current rod pattern)
B. Correct- the rod withdrawal limitations are dependent on reactor power as sensed by First Stage Shell Pressure. An upscale failure would indicate reactor power above the high power setpoint to RC&IS. The Rod Withdrawal Limiter will then limit rod withdrawals to 2 notches (12-16). C. See B D. See B Technical Reference(s):
R-STM-0500, Rev 3 p.16 of 46 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0500, Obj 22 Question Source:
Bank # RBS-NRC-665 Question History:
Last NRC Exam Dec 2008 Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.6 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 56 QUESTION 56 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Gro up # 2 K/A # 202001 K4.02 IR 3.1 Knowledge of RECIRCULATION System design feature(s) and/or interlocks which provide for the following:
Adequate recirculation pump NPSH Proposed Question:
When total Feedwater flow drops below the Reactor Recirc system interlock level, the Reactor Recirc Pumps downshift from fast speed to slow speed. This interlock ___.
A. prevents flow velocity effects on the wide range level indication.
B. prevents thermal stress on the Recirc pump C. prevents cavitation in the Recirc pumps D. adds negative reactivity in anticipation of an imminent reactor scram Proposed Answer:
C Explanation A. the interlock associated with wide range flow velocity effects is the low reactor level interlock B. the interlock associated with thermal stress is the Steam Dome to Vessel Bottom Head delta T C. Correct - Cavitation is prevented by inhibiting high speed operation of pumps with feed flow < 19.9%
D. the EOC-RPT interlock anticipates a reactor scram based on a turbine trip causing a pressure transient, which will rapidly add positive reactivity.
Technical Reference(s):
R-STM-0053, Rev 13 p. 27 of 76
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0053, Obj 17
Question Source:
Modified Bank
- RBS-NRC-168 Question History:
Last NRC Exam 1995 Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.2 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 57 QUESTION 57 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 202002 K1.09 IR 3.1 Knowledge of the physical connections and/or cause
-effect relationships between RECIRCULATION FLOW CONTROL SYSTEM and the following: Reactor water level Proposed Question:
The plant is operating at 100% power. The "A" narrow range level channel is selected as the input to the Feedwater Level Control System. A leak has developed in the reference leg of the "A" narrow range level transmitter.
The ATC operator promptly placed the Master Feedwater Level Controller in MANUAL. As a result of this condition, both Recirc pumps will ___.
A. remain at their present speed, and the Recirc Flow Control Valves will runback to 60% drive flow position.
B. transfer to SLOW speed operation, and the Recirc Flow Control Valves will remain at their present position.
C. transfer to SLOW speed operation, and the Recirc Flow Control Valves will runback to 60% drive flow position.
D. remain at their present speed, and the Recirc Flow Control Valves will remain at their present position.
Proposed Answer:
D Explanation A. first part is correct, but the reference leg leak will cause a false high level indication; FCV's will not runback due to a high level B. and C. The reference leg leak will cause a false high level indication, Recirc pumps will not downshift due to a high RPV level.
D. Correct - Recirc Pumps and FCV's logic receive level signal from the narrow range selected (A); the reference leg leak will cause transmitter differential pressure to rise, therefore the Recirc system will not receive a low RPV level (Recirc pumps downshift at level 3 and FCVs runback at level 4)
Technical Reference(s):
R-STM-0107B, Rev 27 pp. 58
-59 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0107B, Obj 10 Question Source:
Bank # Dec 2008 NRC Exam Q#57 Question History:
Last NRC Exam Dec 2008 Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.2 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 58 QUESTION 58 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 216000 K6.01 IR 3.1 Knowledge of the effect that a loss or malfunction of the following will have on the NUCLEAR BOILER INSTRUMENTATION: A.C. electrical distribution Proposed Question:
A loss of RPS Bus A causes which RPV level indication on the 680 panel to fail downscale?
A. B21-R604, Wide Range meter B. C33-R606A, Narrow Range Channel A meter C. C33-R608B, Narrow Range recorder D. C33-R608R, Upset Range recorder Proposed Answer:
A Explanation A. Correct
- power supply to the wide range meter comes from RPS
-A and fails downscale B. power supplied by inverter 1VBN-PNL01B1 C. power supplied by inverter 1VBN-PNL01B1 D. The narrow range/upset range recorder has a single power supply (1VBN
-PNL01B1) Technical Reference(s):
R-STM-0051, Rev 5 p. 23 of 47 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0051, Obj 13; RLP
-STM-0508, Obj 7 Question Source:
Bank # RBS-NRC-808 Question History:
Last NRC Exam Feb 2003 Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 59 QUESTION 59 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 219000 K2.02 IR 3.1 Knowledge of electrical power supplies to the following:
Pumps (RHR-Supp Pool Cooling)
Proposed Question:
The electrical power supplies to (1) RHR Pump C, and (2) its' associated line fill pump is ___ and ___.
A. (1) ENS-SWG1A ; (2) ENS
-SWG1A B. (1) ENS-SWG1B ; (2) EJS
-SWG1A C. (1) ENS-SWG1B ; (2) EJS
-SWG1B D. (1) ENS-SWG1A ; (2) ENS
-SWG1B Proposed Answer:
C Explanation A. Both part 1 and 2 are incorrect; See C B. First part is correct; See C for part 2 C. Correct - RHR pumps B & C power supply is ENS
-SWG1B; the line fill pump for Div 2 is EJS
-SWG1B D. Part 1 is incorrect; See C Technical Reference(s):
R-STM-0204, p. 25 of 63 ; EE
-001AC Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0204, Obj 11 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 1 0 CFR Part 55 Content:
55.41.b.8 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 60 QUESTION 60 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 223001 G2.1.28 IR 4.1 Knowledge of the purpose and function of major system components and controls. Primary Containment Systems Proposed Question:
The function of the ___ System is to control hydrogen concentration that would be generated during a Design Basis Accident. If hydrogen is detected in the drywell or containment, proper operation of this system alone is sufficient to consume the hydrogen being generated.
A. Containment Hydrogen Purge B. Hydrogen Igniter C. Hydrogen Mixing D. Hydrogen Recombiner Proposed Answer:
B Explanation A. This system designed as a backup to the H 2 Recombiners; there is no EOP/SAP guidance for its use B. Correct - The H 2 igniters are designed to handle 75% of the Metal
-Water Reaction rate; they are designed to mitigate the consequences of a generation event more severe than a design basis LOCA. The Igniter system bounds the DBA scenario used in sizing the H 2 Mixing and H 2 Recombiner Systems.
C. Primary function is to mix containment and drywell atmospheres thereby temporarily diluting the H 2 D. Recombiners are started 14 days after a DBA Technical Reference(s):
EOP Bases, Rev 16, p. B 30, 31 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0057, Obj 20
Question Source:
Modified Bank #
RBS-OPS-06288 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.9 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 61 QUESTION 61 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 239001 K3.15 IR 3.5 Knowledge of the effect that a loss or malfunction of the MAIN AND REHEAT STEAM SYSTEM will have on following:
Reactor water level control Proposed Question:
The plant is operating at 70% reactor power when one inboard MSIV fails closed.
RPV level will __
_. A. (1) increase and stabilize at a higher level.
B. (1) decrease and stabilize at a lower level.
C. (1) decrease and then return to normal level.
D. (1) increase and then return to normal level.
Proposed Answer:
C Explanation A. RPV level indication measures water level in the downcomer region and due to steam flow across the dryers, level is approximately 7 inches higher in the downcomer than the core. When an MSIV closes, pressure will go up momentarily and concurrently collapse some of the steam void in the core. Water levels equalize and then FWLC will compensate and get level back to the the setpoint on the tape set B. First part is correct, but RPV level is being controlled by FWLC system and will return to whatever the tape set is dialed to (setpoint setdown is not activated)
C. Correct - RPV level initially drops due to a pressure transient causing voids to collapse; water from downcomer flows into the core and then FWLC compensates and level returns to normal.
D. See A Technical Reference(s):
RPPT-H LO-0316, Slide 46 of 72
Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0109, Obj 2, 19; RLP-HLO-0316, Obj 1, 2 Question Source:
Bank # RBS-NRC-308 Question History:
Last NRC Exam June 1995 Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.5 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 62 QUESTION 62 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 241000 A2.01 IR 3.5 Ability to (a) predict the impacts of the following on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of turbine inlet pressure signal
Proposed Question:
The plant is starting up and reactor power is currently 8%. EHC Pressure Regulator Channel A is selected for pressure control. Pressure Regulator Channel B is in TEST.
If the Averaging Manifold Pressure transmitter for Pressure Regulator Channel A fails to 0 psig, the Turbine Control Valves will __(1)__ and the steam Bypass Valves __
(2)__. A. (1) fully o pen; (2) remain closed B. (1) fully close; (2) fully open C. (1) fully close; (2) remain closed D. (1) fully open; (2) fully open Proposed Answer:
C Explanation A. The pressure regulator sensing a low pressure, will close (not open) the TCVs trying to raise pressure.
B. TCVs will close causing pressure to rise, but the BPVs use the same pressure transmitter and therefore would remain closed (not open)
. C. Correct
- Turbine Control Valves will close and the BPVs use the same pressure transmitter and therefore would also remain closed D. The pressure regulator sensing a low pressure, will close (not open) the TCVs trying to raise pressure
Technical Reference(s):
R-STM-0509, Rev 14 p.51 of 81 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0509, Obj 10 Question Source:
Modified Ban k # RBS-LOR-1252 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.5 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 63 QUESTION 63 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 272000 K4.02 IR 3.7 Knowledge of RADIATION MONITORING System design feature(s) and/or interlocks which provide for the following:
Automatic actions to contain the radioactive release in the event that the predetermined release rates are exceeded Proposed Question:
The plant is operating at 100% power. Annulus Pressure Control system is in operation with HVR-FN16A, Annulus Pressure Control (APC) Fan A running.
RMS-RE11A, Div 1 Annulus Exhaust Radiation Monitor goes into High Alarm (reading greater than 3.89E
-5 µci/cc). Which of the following describes the ventilation lineup after this event?
A. HVR-FN16A trips; Both Standby Gas Treatment Trains are running B. HVR-FN16A stays running; Both Standby Gas Treatment Trains are running C. HVR-FN16A stays running; ONLY Standby Gas Treatment Train A is running D. HVR-FN16A trips; ONLY Standby Gas Treatment Train A is running Proposed Answer:
A Explanation A. Correct
- RMS-RE11A sends a signal to trip FN16A and start GT S-FNA; GTS-FNB will start due to a low APC System flow B. RMS-RE11A sends a signal to trip FN16A C. RMS-RE11A sends a signal to trip FN16A D. Part 1 is correct; Part 2 is wrong because GTS
-FNB will start due to a low APC System flow Technical Reference(s)
- ARP-DSPL230-1GP011, Rev 9 R-STM-0511, Radiation Monitoring, Rev 15 p. 46 of 48 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0511, Obj 4,6
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.11 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 64 QUESTION 64 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 288000 A3.01 IR 3.8 Ability to monitor automatic operations of the PLANT VENTILATION SYSTEMS including:
Isolation/initiation signals Proposed Question:
Containment Unit Coolers 1A and 1B are running.
Which of the following signals will cause (1) chilled water to the Containment Unit Coolers to isolate, and (2) Service Water to the Containment Unit Coolers to align?
A. (1) High Negative D/P between Containment and Annulus (greater than
-12") ; (2) RPV level 2 (-43") B. (1) High Drywell D/P (1.68) ; (2) RPV level 1 (-143") after 60 sec time delay C. (1) RPV level 1 (-143") ;
(2) RPV level 1 (-143") after 60 sec time delay D. (1) RPV level 2 (-43") ;
(2) RPV level 2 (-43") Proposed Answer:
B Explanation A. Part 1 is correct ; SW aligns on RPV level 1 after 60 time delay B. Correct - chill water isolates on RPV level2, DW D/P, and
-12"; Service Water does not align until level 1 is reached and then after a time delay of 60 seconds to allow separation between CW and SW.
C. Part 1 is incorrect - Level 2 signal isolates CW; D. Part 1 is correct
R-STM-0403, Reactor Building HVAC, Rev 8 pp.
7-11 of 50 Proposed references to be provided to applicants during examination: None Learning Objective:
RLP-STM-0403, Obj 6 Question Source:
Modified Bank # RBS-OPS-2307 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.7 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 65 QUESTION 65 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 290003 K5.01 IR 3.2 Knowledge of the operational implications of the following concepts as they apply to CONTROL ROOM HVAC
- Airborne contamination (e.g., radiological, toxic gas, smoke) control Proposed Question:
A high-high radiation condition has been detected by RMS
-RE13A and 13B, Main Control Room Local Intake A and B.
Which of the following represents the effect on the Control Building HVAC system?
HVC-MOV1A/B HVC-AOD19C/D/E/F CR AHU Outside Supply HVC Local Air Intake A. OPEN OPEN B. OPEN CLOSE C. CLOSE OPEN D. CLOSE CLOSE Proposed Answer:
C Explanation A high-high rad detected by RMS
-RE13 will cause the following: HVC
-MOV1 to close, HVC
-AOD19s to open, and HVC
-FN1 to start, Making C correct.
Technical Reference(s):
ARP-P863-74A-H03 and -H08, Rev 24 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0402, Obj 7 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.11 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 66 QUESTION 66 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.1.7 IR 4.4 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation Proposed Question: OSP-0053, Emergency and Transient Response Support Procedure, contains Post
-Scram Level Control Strategies.
In non-ATWS conditions, when MSIVs are closed, the prescribed level band is _(1)_. In an ATWS condition, with Reactor power at 30% and the Turbine on-line, the prescribed level band is _(2)_. A. (1) 10 to 51 inches; (2) -60 to -140 inches B. (1) -20 to 51 inches; (2) -100 to -140 inches C. (1) 10 to 51 inches; (2) -100 to -140 inches D. (1) -20 to 51 inches; (2) -60 to -140 inche s Proposed Answer:
D Explanation A. Part one would only be correct if MSIVs were open B. Part 1 is correct, part two would be correct if SRVs were required for pressure control C. see A D. Correct - The prescribed level band for a non
-ATWS condition is normally 10 to 51", except when certain plant conditions described in the EOP bases are present. One such condition is closure of MSIVs where the expanded band is prescribed. In an ATWS when conditions are not present that would challenge containment (SRVs are not required for pressure control), then the level control band should remain
-60 to -100 inches.
Technical Reference(s):
OSP-0053, Rev 22 pp. 16-19 of 74 EOP Bases pp. B-6-12 & -13; B-7-21 & -22 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-OPS-0512, Obj 5 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 67 QUESTION 67 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.1.31 IR 4.6 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup Proposed Question:
The main turbine has just tripped while operating at 40% power.
The operator will see ___. A. all 8 Scram Pilot Solenoid Valve indicating lights out on H13-P691. B. Turbine Bypass Valves indicating open on H13
-P680. C. ENS-SWG1A aligned to NNS
-SWG1B on H13-P808. D. the main generator output breakers tripped and the exciter field breaker closed on H13-P680. Proposed Answer:
B. Explanation: A. The 8 scram pilot lights are on H13
-P680. P691 contains the RPS trip unit that input to the trip logic.
B. Correct
-Bypass valve indications are on H13
-P680 and they would be open as a result of the turbine trip.
C. ENS-SWG1A is normally aligned to NNS
-SWG1A. This transient would not change that lineup.
D. Exciter field breaker would also be tripped for this transient.
Technical Reference(s):
AOP-0001, Reactor Scram, Rev 28 p. 4 of 10; AOP-0002, Turbine Trip, Rev 26 p. 6 of 10 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0310 Obj. 8
Question Source:
Bank # RBS-NRC-01183 Question History:
Last NRC Exam RBS Dec. 2010 Q# 67
Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41b.4 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 68 QUESTION 68 Re v 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.1.40 IR 2.8 Knowledge of refueling administrative requirements Proposed Question:
During refueling activities, Main Control Room personnel are required to maintain constant communication with the
___. A. Fuel Movement Supervisor B. Bridge Driver/Fuel Handler C. IFTS Operator D. Refuel Senior Reactor Operator Proposed Answer:
D Explanation: Required by FHP
-0003, Roles and Responsibilities of Refuel SRO Step 2.1 Technical Reference(s):
TR 3.9.11, Communications during core alts EN-FAP-OU-108, Fuel Handling Process, Rev 5 pp. 3,4 of 39 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-OPS-0410, Obj 1
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 4 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 69 QUESTION 69 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.2.6 IR 3.0 Knowledge of the process for making changes to procedures Proposed Question:
In accordance with RBNP
-001, DEVELOPMENT AND CONTROL OF RBS PROCEDURES, a Comment PAR, may be used for which of the following?
A. To make minor changes that result in a change of intent.
B. To correct typographical errors to a procedure that is being implemented.
C. To make suggestions for future procedure improvements.
D. To change acceptance criteria.
Proposed Answer:
C Explanation A. A procedure "Revision" is required for changes that result in a change of intent.
B. This requires an Editorial change.
C. Correct
- This choice describes a "Comment".
D. A change in acceptance criteria also results in a change of intent, so would require a "Revision" Technical Reference(s):
RBNP-001, Development and Control of Procedures, Rev 35 p. 19 of 43 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-0202, Obj 1 Question Source:
Bank # RBS-OPS-1 667 Question History:
Last NRC Exam RBS April 2010 Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 70 QUESTION 70 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.2.14 IR 3.9 Knowledge of the process for controlling equipment configuration or status Proposed Question:
A decon technician has called the RO requesting permission to open an MWS supply valve to obtain water for deconning purposes. What action is required?
A. The RO can authorize the manipulation and logging is not required as long as the tech calls back before the end of shift stating that the valve is closed B. The OSM/CRS must authorize the manipulation and it must be logged in the control room log or manipulated device log book.
C. The RO can authorize the manipulation and it must be logged in the control room log or manipulated device log book D. The OSM/CRS must authorize the manipulation and logging is not required as long as the tech calls back before the end of shift stating that the valve is closed.
Proposed Answer:
B Explanation A. See B B. Correct OSP
-0014 requires OSM/CRS approval and logging for configuration control purposes when manipulating plant components outside of procedures C. See B D. See B Technical Reference(s):
OSP-0014, Administrative Control of Equipment, Rev 304 p. 6 of 16
Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 71 QUESTION 71 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.2.35 IR 3.6 Ability to determine Technical Specification Mode of Operation Proposed Question:
The following plant conditions exist:
The mode switch position is Shutdown Refueling operations have been completed and the last head bolt has been tensioned What mode of operation is the plant in?
A. Mode 2 Startup B. Mode 3 Hot Shutdown C. Mode 4 Cold Shutdown D. Mode 5 Refueling Proposed Answer:
C Explanation A. See C B. See C C. Correct
- Per Tech Spec Definitions Table 1.1
-1 D. See C Technical Reference(s):
Tech Spec Definitions Table 1.1
-1 Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 72 QUESTION 72 Rev 0 Examination Outline Cross
-
Reference:
Leve l RO SRO Tier # 3 K/A # G2.3.11 IR 3.8 Ability to control radiation releases Proposed Question:
Due to a steam leak, the Main Steam Tunnel area temperatures have caused automatic isolations to occur as designed. Because of the location of the leak, no LOCA signals have been generated by RPV level or DW pressure. An ALERT has been declared due to offsite release rate.
Which one of the following will reduce the UNMONITORED release rate?
A. Shutdown the Turbine Building Ventilation System, if operating. B. Shutdown the Fuel Building Ventilation System, if operating
. C. Start the Turbine Building Ventilation System, if not operating.
D. Start the Fuel Building Charcoal Ventilation System, if not operating.
Proposed Answer:
C Explanation A. This action would raise the unmonitored release rate B. The fuel building normal and charcoal filtration systems are both monitored; this would not affect unmonitored release C. Correct - IAW EOP-0003 D. See B Technical Reference(s):
EOP-0003 p. B-10-3 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-515, Obj 6 Question Source:
Bank # RBS-N RC-7 97 Question History:
Last NRC Exam RBS 2003 NRC Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 73 QUESTION 73 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.3.13 IR 3.4 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc Proposed Question:
For which of the following evolutions is the licensed operator in the control room procedurally required to notify Radiation Protection prior to performance?
A. Suppression Pool reject to Radwaste with Residual Heat Removal (RHR)
B. Placing Heater Drain pumps in the PUMP FORWARD mode C. Startup of Circulating Water Blowdown D. Reactor Core Isolation Cooling (RCIC) slow roll startup Proposed Answer:
D Explanation: A. See D. (SOP
-0031) B. See D. (SOP
-0010) C. See D. (SOP
-0006) D. Although notifying RP prior to any of the 4 choices demonstrate good teamwork, only RCIC slow roll startup specifically requires notification per SOP
-0035. Technical Reference(s):
SOP-0035, RCIC, Rev 47 p. 16 of 76 Proposed references to be provided to applicants during examination:
Non e Learning Objective:
RLP-STM-0209 Obj 10c Question Source:
Bank # RBS Oct 2012 Audit Q#72 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 4 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 74 QUESTION 74 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.4.12 IR 4.0 Knowledge of general operating crew responsibilities during emergency operations Proposed Question:
To which of the following Emergency Response Organization facilities is the Fire Brigade assigned to when not performing fire fighting functions?
A. Operations Support Center (OSC)
B. Technical Support Center (TSC)
C. Emergency Operating Facility (EOF)
D. Joint Information Center (JIC) Proposed Answer:
A Explanation A. Correct
- per EIP-2-016. B. C. and D. See A Technical Reference(s):
EIP-2-016, Operations Support Center, Rev 29 p. 5 of 31 Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
Bank # RBS Oct 2008 Audit Q#75 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments:
December 2014 River Bend Station NRC Initial License Examination Reactor Operator 75 QUESTION 75 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.4.26 IR 3.1 Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage Proposed Question:
According to EN
-OP-115, Conduct of Operations, a Fire Brigade of at least _(1)_ members shall be maintained on site at all times; the fire brigade members shall NOT include the
_(2)_. A. (1) Four ;
(2) Shift Technical Advisor (STA)
B. (1) Five ;
(2) Duty Manager C. (1) Five ;
(2) Shift Technical Advisor (STA)
D. (1) Four ;
(2) Duty Manager Proposed Answer:
C Explanation A. EN-OP-115 requires a minimum of 5 Fire Brigade Members B. Part 1 is correct; The Duty Manager is not designated as being precluded by procedure C. Correct - The fire brigade shall not include the OSM, CRS, STA, ATC, and 1 NEO D. EN-OP-115 requires a minimum of 5 Fire Brigade Members Technical Reference(s):
EN-OP-115, Conduct of Operations, Rev 15 p. 78 of 89 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-0206, Obj 7 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.41.b.10 Comments: Similar to Bank Question #RBS
-OPS-3382 December 2014 River Bend Station NRC Examination Senior Reactor Operator 76 QUESTION 76 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295004 G2.2.44 IR 4.4 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions (Partial or Complete Loss of D C) Proposed Question:
During normal plant operation, several alarms are received on H13
-P601 Insert 16, including:
DIV III 125VDC SYSTEM TROUBLE DIV III BATTERY CHARGER TROUBLE In addition to these alarms, the (2) 125VDC white power available lights on H13
-P601 High Pressure Core Spray area are extinguished.
Which of the following procedures should the CRS enter to mitigate this condition?
A. SOP-0030, HIGH PRESSURE CORE SPRAY SYSTEM B. SOP-0049, 125 VDC SYSTEM C. AOP-0014, LOSS OF 125 VDC D. AOP-0042, LOSS OF INSTRUMENT BUS Proposed Answer:
C. Explanation:
A. This procedure is for normal operation of the High Pressure Core Spray System. The stem indicates abnormal conditions. B. This procedure is for normal operation of the 125 VDC System. The stem indicates abnormal conditions.
C. Correct
- The conditions in the stem indicate that 125 VDC to the HPCS has been lost. The correct procedure for this condition is AOP
-0014. D. This procedure deals with the loss of various uninterruptible power supplies. These UPS systems are associated with Div 1, Div 2 and non
-safety related components.
Technical Reference(s):
AOP-0014 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-OPS-0532 Obj 2 Question Source:
Bank# April 2010 NRC Exam Question History:
Last NRC Exam April 2010 Q#77 Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.43.b.5 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 77 QUESTION 77 Rev 1 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295006 G2.4.
34 IR 4.2 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (SCRAM) Proposed Question:
The plant has experienced an ATWS. The following conditions exist at the 680 panel
- The Mode Switch is in Shutdown ARI has been initiated Approximately 50% of the withdrawn rods fully inserted All eight white scram solenoid lights are extinguished The SDV Vent and Drain Valve position green lights are on and red lights are off Annunciator 680
-05A-C08, SCRAM Pilot Valve Air Header Low Pressure NOT lit In accordance with EOP Enclosure 26, Control Rod Insertion Method Determination, w hich procedure should be directed to allow insertion of control rods
? A. EOP Enclosure 10, De
-Energizing SCRAM Solenoids B. EOP Enclosure 11, Venting SCRAM Air Header C. EOP Enclosure 13, Opening Individual SCRAM Test Switc hes D. EOP Enclosure 17, Venting CRD Overpiston Volumes Proposed Answer:
B A. The stem conditions are indicative the failure of the scram air header to vent; scram solenoids are already d e-energiz e d. B. Correct - this enclosure will vent the scram air header and allow the rods to be inserted.
C. and D. These two enclosures are used after the scram air header is depressurized.
Technical Reference(s):
- ATWS; EOP-000 5 Enclosure 26
, Control Rod Insertion Method Determination Proposed references to be provided to applicants during examination:
None. Learning Objective:
RLP-HLO-0513, Obj 5 Question Source:
Modified Bank#
RBS-OPS-06265 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.43.b.5 Comments: Original KA was G2.4.2, re-rolled (drew G 2.4.34) ; The original KA was not conducive to writing an SRO level question.
December 2014 River Bend Station NRC Examination Senior Reactor Operator 78 QUESTION 78 Rev 1 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295021 AA2.04 IR 3.6 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING : Reactor water temperature Proposed Question:
The plant is Shutdown in Mode 4 following 400 days of continuous operation.
A spurious Group 3 isolation occurs and the logic cannot be reset
. Reactor water temperature rises to 20 5 What is the time limit for notification to the NRC
? A. Immediate notification (within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
B. 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification C. 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification D. 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification Proposed Answer:
A Explanation A Correct - the conditions are indicative of a NOUE; a group 3 isolation would cause a loss of shutdown cooling; temperature rose enough for a mode change; notification is required immediately B 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is a valid notification time but not for this condition
. C. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> would be for an valid ESF actuation/isolation D. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a valid notification time but not for this condition
. Technical Reference(s):
EIP-2-001 , Classification of Emergencies, Rev 24, Attach 10 EIP-2-002 , Classification Actions, Rev 31, p. 7 of 20 Proposed references to be provided to applicants during examination: EIP-2-001, Attach 10 only Learning Objective:
(As available)
Question Source:
Bank# GGNS 2013 AUDIT Q# 79 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content:
55.43.b.5 Comments: Replaced question to make this SRO level
December 2014 River Bend Station NRC Examination Senior Reactor Operator 79 QUESTION 79 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295028 G2.4.21 IR 4.6 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (High Drywell Temperature)
Proposed Question:
A transient occurred that caused the Control Room Team to enter the Alternate Level Control Leg of EOP
-1. While executing the ALC steps, a large break LOCA, ALL RPV level instruments indicated off-scale low.
Five seconds later, the Fuel Zone Level instruments returned on scale. The following conditions now exist:
Containment temperature 91°F (at EL 119 ft)
Drywell temperature 285°F (at EL 145 ft)
RPV Pressure 10 psig Fuel Zone Level indication
-290 inches
, rising and falling rapidly With the se conditions, Fuel Zone Level indication
____ be used to determine RPV level ; the CRS should direct ____.
A. CANNOT ; Emergency Depressurizing per EOP
-1 B. CANNOT ; exiting EOP
-1 and entering EOP
-4 C. CAN ; restoring and maintaining RPV level above
-162 inches per EOP
-1 D. CAN ; exiting the ALC leg and entering the Steam cooling leg of EOP
-1 Proposed Answer:
B Explanation A. Incorrect per override RC
-2 of EOP-1 B Correct per override RC
-2 EOP-4 must be entered C, D - Incorrect, FZ level indication is unreliable due to boiling in the instrument line run Technical Reference(s):
, Rev 26 Proposed references to be provided to applicants during examination:
Caution 1 of E OP-1 Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CF R Part 55 Content:
55.43.b._ Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 80 QUESTION 80 Rev 1 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295031 EA2.04 IR 4.8 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling Proposed Question:
The following conditions exist:
- Reactor Pressure is 943 psig
- RHR pumps A and C are lined up for injection
- HPCS injection has commenced
- Reactor water level is
-190" and slowly lowering Which of the following is correct concerning cooling of the core, and Which procedural action should the SRO direct next?
A. Adequate core cooling exists and Steam Cooling is required
. B. Adequate core cooling exists and Spray Cooling is required
. C. Adequate core cooling does not exist and Emergency Depressurization is required. D. Adequate core cooling does not exist and Containment Flooding is required
. Proposed Answer: C Explanation A. -200 is the water level for steam cooling without injection B. -211 is the water level for spray cooling, at 943 psig HPCS will not develop 5000 gpm C. Correct
- adequate core cooling does not exist, ED is necessary to allow low pressure ECCS to inject D. Containment flooding would be required if adequate core cooling could not be restored and maintained
. Technical Reference(s):
EOP Bases Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History: Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.43.b.5 Comments: Revised question to include procedural choice to make this SRO level
December 2014 River Bend Station NRC Examination Senior Reactor Operator 81 QUESTION 81 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 600000 AA2.16 IR 3.5 Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: Vital equipment and control systems to be maintained and operated during a fire Proposed Question:
The control room has received a report of a fire on the east side crescent area of the auxiliary building. The reactor has been shut down and the current conditions are:
Reactor Recirc Pumps both in slow speed Reactor pressure 125 psig and lowering RPV level 36" and steady What direction should the CRS give concerning Shutdown Cooling (SDC) operation?
A. Place RHR-B in normal SDC per SOP
-0031, RHR System B. Place RHR-A in normal SDC per SOP
-0031, RHR System C. Place RHR-C in alternate SDC per AOP
-0020, Alternate Decay Heat Removal D. Place RHR-B in alternate SDC per AOP
-0020, Alternate Decay Heat Removal Proposed Answer:
B Explanation A. C. and D. Per AOP-0052, Attachment 1 Div 2 equipment is not available B. Correct- Per AOP-0052, Attachment 1 this equipment is available
Technical Reference(s):
AOP-0052, Fire Outside the MCR, Rev 25 Attachment 1 Proposed references to be provided to applicants during examination:
Attachment 1 of AOP
-52 Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.43.b.5 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 82 QUESTION 82 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 700000 G2.2.37 IR 4.6 Ability to determine operability and/or availability of safety related equipment (Generator Voltage and Electric Grid Disturbances)
Proposed Question:
The plant is operating at rated power.
A grid disturbance occurs that causes: Alarm P808
-XSR1C VOLTAGE LOW ENS-ACB06 ENS-SWG1A PFD Supply BKR trips on under voltage Div I DG starts and ties to the bus An attempt to re
-close ENS-ACB06 was unsuccessful A CR has been initiated concerning the status of the ENS
-ACB06 ENS-SWG1A PFD Supply BKR.
What Operability Code should be assigned to the initial screening of this CR in PCRS?
AND Which T.S., if any, should be entered?
A. INOPERABLE and T.S. 3.8.1. condition A B. OPERABLE and no T.S. entry C. INOPERABLE
-OP EVAL and T.S. 3.8.1. condition A D. OPERABLE-COMP MEAS and no T.S. entry Proposed Answer:
A Technical Reference(s):
AOP-0064, Degraded Grid, Rev 6 ENS-DC-199, Offsite Power Supply Design Requirements, Attach 9.3
Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.2 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 83 QUESTION 83 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 295015 AA2.02 IR 4.2 Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM : Control rod position Proposed Question:
Following a high drywell differential pressure scram signal, the following conditions exist:
1 42 control rods indicate Full
-In 3 control rods are not full in and are at various positions Reactor power indicates 0%
SRM s are inserted and count rates are lowering (1) What is the status of the reactor and (2) what procedure does the CRS direct
? A. The reactor will remain shutdown without boron
- Enter EOP-1, RPV Control B. It can NOT be determined that the reactor will remain shutdown under all conditions without boron
C. It can NOT be determined that the reactor will remain shutdown under all conditions without boron
-1A is NOT applicable with reactor power less than 5%.
D. The reactor will remain shutdown without boron; Enter EOP
-1, RPV Control and transition to EOP
-1A, RPV Control, ATWS Proposed Answer:
B Explanation A. and D. With more than one rod out, it can not be determined that the reactor will remain shutdown. IAW EOP Bases, "For the current fuel design and core load, the MSBWP is all control rods fully inserted
." B. Correct - With more than one rod out, it can not be determined that the reactor will remain shutdown under all conditions without boron; A high drywell differential pressure scram signal is an EOP
-0001 entry condition, transition to EOP
-1A is required due to override step RC
-2 in EOP-1. C. First part correct, however, EOP
-1 override RC
-2 directs exiting EOP
-1 and entering EOP-1A Technical Reference(s):
EOP Base s p. B-6-5 Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
Modified Bank #
2008 Audit Q#25 Question History: Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.43.b.5 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 84 QUESTION 84 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 295029 G2.2.25 IR 4.2 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits (High Suppression Pool Wtr Lvl)
Proposed Question:
Technical Specifications (TS) require suppression pool water level to be
< 20 ft 0 inches. What is the TS basis for this upper limit?
A. Above 20 ft 0 inches
, excessive pressure could build up in the drywell before the drywell horizontal vents are cleared during a DBA LOCA.
B. Water levels above 20 ft 0 inches would cause low pressure ECCS systems to operate outside of their design NPSH values.
C. Water levels above 20 ft 0 inches would overflow the drywell weir wall and cause inaccurate drywell leakage indications due to excessive pedestal sump discharge flow. D. Above 20 ft 0 inches, SRV tail pipe pressure could build up to o high during SRV operation and damage the tail pipes.
Proposed Answer: D Explanation A. A higher DW pressure buildup would occur as suppression pool level rises but this is within design ana l ysis and "excessive" pressure buildup would not occur. This condition is not mentioned in the base
- s. B. Levels above 20 ft would raise NPSH of the low pressure ECCS pumps but would not cause them to operate outside of design parameters.
C. The weir wall would overflow at 21 ft 3 inches and cause the pedestal sump to cycle but this is not mentioned as a basis for the upper limit D. Correct - TS 3.6.2.2 Bases Statement; If the suppression pool water level is too high, it could result in excessive clearing loads from S/RV discharges".
Technical Reference(s):
Tech Spec 3.6.2.2 and Bases Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.2 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 85 QUESTION 85 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 500000 EA2.03 IR 3.8 Ability to determine and / or interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: Combustible limits for drywell Proposed Question:
The plant has experienced a LOCA. The following conditions exist:
RPV pressure is 250 psig RPV level is +10 inches and slowly rising Containment pressure is 1.5 psig Drywell temperature is 200°F Drywell hydrogen concentration is 2 percent Containment hydrogen concentration is 2 percent Based on this, the SRO should ________.
A. Enter EOP-2 at Hydrogen Control and Verify start of Div. I and Div. II H 2 monitoring systems B. Continue EOP
-1 at Pressure Control and initiate shutdown cooling C. Enter EOP-2 at Hydrogen Control and operate all hydrogen igniters D. Exit all EOPs and enter the SAPs Proposed Answer:
C Explanation With hydrogen concentration above 1.5%, all igniters should be operated A - Hydrogen mixing not authorized in EOP
-2 B - Shutdown cooling interlock not clear D - Drywell / Containment H 2 concentration is not high enough Technical Reference(s):
EOP Bases p. B-8-26 & 27 Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.2 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 86 QUESTION 86 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 205000 A2.03 IR 3.2 Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C. failure Proposed Question:
The plant is shutdown, in Mode 4 with RHR A in SDC mode and no Recirc Pumps running. A leak in the drywell causes RPV Level to lower. An operator begins to line up RHR A for Injection mode.
When E12-F006A, RHR Pump A SDC Suction Valve begins to close, electrical power is lost to the valve. The operator closes E12
-F008 and -F009, RHR Shutdown Cooling Outboard and Inboard Isolation valves.
The CRS will direct ___ ; and RHR A Shutdown Cooling is ___.
A. Manually closing E12
-F006 A ; Operable B. Aligning RHR Injection through E12
-F0053A, RHR A SDC Injection Valve
- Operable C. Manually closing E12
-F006A ; Inoperable D. Aligning RHR Injection through E12
-F0053A, RHR A SDC Injection Valve
- Inoperable Proposed Answer
A Explanation A. Correct
- The RHR Suppression Pool Injection Suction Valve (F004) cannot be opened with the F006 valve not full closed, but the F006 valve can be manually closed thereby allowing the F004 valve to be opened; Without opening closing the F006 and opening the F004, there is no suction path available to allow for injecting in any path.
RHR is operable because the LCO statement reads: "An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, two heat exchangers in series, and the associated piping and valves. Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat."
Technical Reference(s):
AOP-0051, Loss of Decay Heat Removal, Rev 312 Tech Spec Bases 3.4.10 p. 3.4
-48 and 49 Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.43.b.2 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 87 QUESTION 87 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 212000 G2.2.12 IR 4.1 Knowledge of surveillance procedures (RPS)
Proposed Question:
The plant is operating at 90% power. ST P-508-0201, Manual Scram Channel Functional Test and LSFT is being performed.
The last division is being restored.
When the operator attempted to reset the half
-scram, status light H13
-P680-11E1, RPS DIV 4 SCRAM SOV VALVES OPEN
- CR 4B white light is NOT on. All other SOV status lights are lit.
A loss of RPS Bus B occur s; what is the expected plant response and what AOP(s) should be entered?
A. All of the control rods will scram ;
AOP-0001, Reactor Scram only B. Approximately 1/4 of the control rods will scram ;
AOP-0001, Reactor Scram only C. Approximately 1/4 of the control rods will scram ;
AOP-0001, Reactor Scram, and AOP
-0010, Loss of RPS Bus D. None of the control rods will scram ;
AOP-0010, Loss of RPS Bus only Proposed Answer:
D Explanation A. See D B. See D C. See D D. Correct - The loss of RPS B will only cause a half
-scram signal; The Div 4 SOV is powered by RPS
-B, so a loss of RPS B would have de
-energized the CR4B which is already de
-energized; IF the transient had been a loss of RPS A, it would have caused approximately 1/4 of the rods to scram due to de-energizing the second SOV in Div 4 and actions would need to be taken to insert a manual scram and enter AOP
-1. AOP-0010 must be entered in order to restore power to the bus and reset isolation signals.
Technical Reference(s):
R-STM-508, RPS, Rev 5 p. 23 of 59; STP
-508-0201, Rev 14 pp 19
-20 of 26 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0508, Ob j 8 Question Source:
Modified Bank #
RBS-NRC-937 Question History:
Last NRC Exam Sept 2004 NRC Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.43.b.5 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 88 QUESTION 88 Rev 1 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 215005 A2.02 IR 3.7 Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Upscale or downscale trips
Proposed Question:
GOP-0001 Plant Startup requires verification that all APRMs are >5% prior to taking the mode switch to RUN.
This step was signed and then a shift turnover occurred.
The mode switch is taken to run
. APRM s B & C are both reading <5%.
All other APRMs are operable and are reading >5%.
The result of this condition is a ___; the CRS will direct ___.
A. reactor scram
. B. Control Rod Withdrawal Block
- bypassing APRMs B & C to clear the rod block per EN-OP-115, Conduct of Operations C. Control Rod Withdrawal Block
- be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in accordance with Tech Specs
. D. half scram
- transfer the Mode Switch back to START & HOT STBY per GOP-0002, Power Decrease/Plant Shutdown
. Proposed Answer:
B Explanation A. There will be a rod withdraw block not a scram.
B. Correct - With mode switch in run and APRMs <5%, a rod withdraw block will occur; not a scram. With all other APRMs operable, B & C can be bypassed to clear the rod block. Tech Specs allow bypassing one channel from each trip system. Rods can be withdrawn to bring B & C above 5%.
C. First Part is correct, but this Tech Specs required action is for when one or more required channels are inoperable; Tech Specs only requires 3 channels per trip system (which in the given condition, is met)
D. A half scram is not present for two APRMs < 5%.
Technical Reference(s):
SOP-0074, Neutron Monitoring, Rev 306 p.20 of 74 R-STM-0503, RPS, Rev 7, pp. 63,68,106 of 112 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0503, Obj 26,28 Question Source:
Bank # 2008 RBS AUDIT Q#89 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.43.b.5 Comments: Question revised to force procedure selection to make it SRO level December 2014 River Bend Station NRC Examination Senior Reactor Operator 89 QUESTION 89 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 239002 A2.06 IR 4.3 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Reactor high pressure Proposed Question:
While operating at 100% power
, a loss of condenser vacuum occurs..
After a reactor scram, the following plant conditions exist: Reactor pressure spiked to 1140 psig Main Turbine has tripped Condenser vacuum 0" (1) What is the current pressure control band for SRV 51D?
(2) What direction does the CRS give to control Reactor pressure?
A. (1) 1063 psig to 956 psig (2) Stabilize pressure with SRVs, then establish a band of 500 to 1090 pounds in accordance with OSP
-0053, Emergency and Transient Response.
B. (1) 1063 psig to 956 psig (2) Install Enclosure 9, Defeating MSIV Isolation Interlocks, and open MSIV s C. (1) 1 103 psig to 9 66 psig (2) Stabilize pressure with SRVs, then establish a band of 500 to 1090 pounds in accordance with OSP
-0053, Emergency and Transient Response.
D. (1) 1103 psig to 966 psig (2) Install Enclosure 9, Defeating MSIV Isolation Interlocks, and open MSIVs Proposed Answer:
A Explanation A. Correct
- set points for 51D ; SRVs are the only available system due to low vacuum B. First part is correct; Encl 9 is only allowed for alternate pressure control in this condition C.. and D. First part is incorrect
- these are the LLS set points for SRV 51C Technical Reference(s):
R-STM-0109, Main Steam, Rev 13 pp. 36,37,60 of 95 EOP-1, RPV Control Proposed references to be provided to applicants during examination:
None Le arning Objective:
RL P-STM-0109, Obj 22 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 3 10 CFR Part 55 Content:
55.43.b.5 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 90 QUESTION 90 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 262001 G2.2.25 IR 4.2 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits (AC Electrical Distribution)
Proposed Question:
Due to a loss of EJS*LDC2A, you have entered TS 3.8.9 condition A (One or more Division I or II AC electrical power distribution subsystems inoperable
). The required action is to Restore Division I and II AC electrical power distribution subsystems to OPERABLE status.
The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> completion time for this LCO action is based on:
A. the unit is more vulnerable to a complete loss of AC power B. the plant is significantly more vulnerable to a complete loss of all uninterruptible power. C. the plant is significantly more vulnerable to a complete loss of DC power.
D. the plant is significantly more vulnerable to a complete loss of D/G power. Proposed Answer:
A Explanation A. Correct
- copied from the bases for condition A B. This applies to condition B C. This applies to condition C D. This applies to TS 3.8.1 condition A Technical Reference(s):
Tech Spec Bases 3.8.9 p. B 3.8
-81 Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 4 Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.2 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 91 QUESTION 91 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 201005 G2.1.23 IR 4.4 Ability to perform specific system and integrated plant procedures during all modes of plant operation (RCIS)
Proposed Question:
GOP-001, Plant Startup is in progress currently in Section C (Approach to Critical: Instructions). Pulling control rods began two hours ago. The dedicated reactor operator has just informed you that a single control rod was pulled out of sequence (4 notches). What will you direct the reactor operator to do?
A. Manually scram the reactor and enter AOP
-0001, Reactor Scram B. Perform a MON calculation to determine margin to Preconditioning Envelope per GOP-0001, Plant Startup
. C. Select and continuously insert the control rod to position 00 per AOP-0061 Mispositioned Control Rod
. D. Suspend all rod motion and enter AOP
-0061, Mispositioned Control Rod.
Proposed Answer:
D Explanation A. This action is directed in AOP
-61 when more than one rod has scrammed or been fully inserted.
B. This would be a required action if a control rod is found in an unintended position or a single rod scram were to have occurred.
C. This is the required action for a drifting rod.
D. Correct - GOP-1 Caution for step 11 directs IF an out of sequence rod should be withdrawn, THEN suspend rod motion, notify the OSM, and reference AOP
-0061. AOP-61, Step 5.2 directs returning the rod to its required position and notify Reactor engineering.
Technical Reference(s):
GOP-0001, Plant Startup, Rev 82 p. 34 of 103 AOP-0061, Mispositioned Control Rod, Rev 7 p. 5 of 8 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-HLO-0711, Obj 4 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.43.b.5 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 92 QUESTION 92 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 204000 A2.
14 IR 3.2 Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System High Temperature
Proposed Question:
A partial loss of Reactor Plant Component Cooling Water (CCP) has resulted in reduced cooling to the Reactor Water Cleanup System (RWCU).
How will the RWCU system respond, and what actions are required to mitigate that response? A. G33-F001, RWCU Pump s Inbd Suction Valve will close when NRHX outlet temperature 1 40°F ; Transfer resin to the holding tank until CCP restored per SOP
-0090. B. G33-F001, RWCU Pump s Inbd Suction Valve will close when NRHX outlet temperature 140°F ; Secure RWCU per SOP-0090. C. G33-F004, RWCU Pumps Outbd Suction Valve will close when NRHX outlet temperature 1 4 0°F ; Transfer resin to the holding tank until CCP restored per SOP
-0090. D. G33-F004, RWCU Pumps Outbd Suction Valve will close when NRHX outlet temperature 140°F ; Secure RWCU per SOP
-0090. Proposed Answer:
D Explanation A. See D B. See D C. Part 1 is correct; See D for Part 2 D. Correct - G33-F001 and F004 are interlocked to close on SLC injection, but only F004 is interlocked to close on Filter/Demin inlet temperature; The ARP requires either reducing flow through demins (Alarm at 130°F) or securing the system Technical Reference(s):
ARP-680-01-B01 & -C01; R-STM-0601, RWCU, Rev 8 pp. 15
-16, 25-26 of 51 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0601, Obj 7 Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.43.b.5 Comments: Original KA was A2.02, re
-rolled (drew 2.14). RBS does not have Low Press RWCU
.
December 2014 River Bend Station NRC Examination Senior Reactor Operator 93 QUESTION 93 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 2 K/A # 290001 A2.04 IR 3.7 Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
High airborne radiation
Proposed Question:
A RWCU leak has occurred in the RWCU pump room and control room attempts t o isolate the leak have failed. Work management is establishing a team to enter the Aux Building to attempt to drive the isolation valves closed from the motor control centers.
Area temperature monitors show the following readings:
RWCU PUMP ROOM The Digital Radiation Monitoring System (DRMS) shows the following indications:
RMS-RE110, AUX BLDG VENT
- Alarm (RED)
RMS-RE213, RHR EQUIPMENT ROOM A
- Normal (Green) RMS-RE214, RHR EQUIPMENT ROOM B
- Normal (Green)
RMS-RE215, RHR EQUIPMENT ROOM C
- ALERT (Yellow)
RMS-RE217, HPCS PENETRATION AREA
- Normal (Green)
RMS-RE218, LPCS PENETRATION AREA
- Normal (Green)
RMS-RE219 RCIC EQUIPMENT ROOM
- ALERT (Yello w) What impact could these radiation levels have and what procedural actions should be directed A. High Radiation levels will have no impact they only provide indication that water from a primary system may be discharging into secondary containment
- the RPV should be emergency depressurized
. B. High radiation levels may preclude personnel access and lead to equipment failure; the Aux building should be isolated per OSP
-0053 Attachment 21 hard card
. C. High Radiation levels may preclude personnel access and lead to equipment failure; the RPV should be emergency depressurized
. D. High Radiation levels will have no impact they only provide indication that water from a primary system may be discharging into secondary containment; the Aux building should be isolated per OSP-0053 Attachment 21 hard card
. Proposed Answer:
B December 2014 River Bend Station NRC Examination Senior Reactor Operator 94 Explanation A. While high radiation levels are indication that there may be a primary system discharging into secondary containment there is still the potential for radiation levels to preclude access and cause equipment failure. Criteria for ED have not been met.
B. Correct
- High radiation levels have the potential to preclude access and cause equipment failure, with RMS-RE110 in alarm the auxiliary building should be isolated and exhaust processed through the standby gas system to minimize release to the public.
C. Criteria for ED have not been met.
D. While high radiation levels are indication that there may be a primary system discharging into secondary containment there is still the potential for radiation levels to preclude access and cause equipment failure
Technical Reference(s):
EOP BASES Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History: Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 4 10 CFR Part 55 Content:
55.43.b.5 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 95 QUESTION 94 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.1.20 IR 4.6 Ability to interpret and execute procedure steps Proposed Question:
Refueling is in progress. Reactor Recirculation and Reactor Water Cleanup System are secured for system maintenance.
How is TR 3.4.13, Chemistry satisfied for reactor coolant conductivity during this portion of a refueling outage?
A. The conductivity of the reactor coolant is recorded continuously.
B. An in-line conductivity measurement is obtained every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
C. An in-line conductivity measurement is obtained every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D. Not required to be met in this mode.
Proposed Answer:
C. Explanation: A. With both Reactor Recirculation and RWCU secured, continuous monitoring is not available.
B. This would be required in Modes 1, 2, & 3.
C. Correct
- Continuous monitoring is not available so an inline conductivity measurement must be obtained every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. D. Required at all times.
Technical Reference(s):
TR 3.4.13 Proposed references to be provided to applicants during examination:
TR 3.4.13 Learning Objectiv e: RLP-STM-0601 Obj. 9 Question Source:
Bank # Dec 2010 NRC Q#95 Question History:
Last NRC Exam Dec 2010 NRC Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis 2 10 CFR Part 55 Content:
55.4 3.b.2 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 96 QUESTION 95 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.1.36 IR 4.1 Knowledge of procedures and limitations involved in core alterations Proposed Question:
During a refuel outage, the Outage Control Center notifies the main control room that core alterations are scheduled to commence next shift.
Select the procedure below which the CRS utilizes to ensure that all requirements for core alterations have been met.
A. STP-000-0005, DAILY REFUELING LOGS B. GMP-0102, REACTOR VESSEL DISASSEMBLY C. FHP-0001, CONTROL OF FUEL HANDLING AND REFUELING OPERATIONS D. FHP-0003, REFUEL PLATFORM OPERATION Proposed Answer:
C. Explanation: A. See C. B. See C. C. Correct
- FHP-0001 Attachment 2 contains a list of Applicalbe Mode 5 Tech Specs which includes (section 2) a list of requirements to commence core alterations.
D. See C.
Technical Reference(s):
FHP-0001, Control of Fuel Handling & Refueling Ops, Rev 35 pp. 4, 35 of 40 Proposed references to be provided to applicants during examination:
None Learning Objective:
RLP-STM-0055 Obj 6 & 9 Question Source:
Bank # Nov 2012 NRC Q#95 Question History:
Last NRC Exam Nov 2012 NRC
Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.7 Comments: Appeared on one of last 2 NRC exams (2 of 2):
December 2014 River Bend Station NRC Examination Senior Reactor Operator 97 QUESTION 96 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.2.11 IR 3.3 Knowledge of the process for controlling temporary design changes Proposed Question:
An event has occurred which presents an imminent threat to the safety of the plant.
In order to mitigate the event a temporary modification must be installed as directed by the Shift Manager. Whose concurrence is required to implement the emergency temporary modification?
A. Another SRO B. The RBS Vice President (VP)
C. The General Manager Plant Operations (GMPO)
D. The Engineering Director or his designee Proposed Answer:
D Explanation A. See D B. See D C. See D D. Correct - Per EN-DC-136, the concurrence of the Engineering Director, or designee is required Technical Reference(s):
EN-DC-136, Temporary Modifications, Rev 10 Section 5.3 Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.3 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 98 QUESTION 97 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.2.38 IR 4.5 Knowledge of conditions and limitations in the facility license Proposed Question:
Which of the Limiting Conditions for Operation (LCOs) listed below provide guidance concerning when changes in mode or other specified conditions are allowed when an LCO is not met?
A. LCO 3.0.2 B. LCO 3.0.3 C. LCO 3.0.4 D. LCO 3.0.6 Proposed Answer:
C. Explanation:
A. 3.0.2 establishes that upon discovery to me et an LCO, the associated ACTIONS must be met.
B. 3.0.3. establishes the actions that must be met when an LCO is not met and the actions and completions times are not met and no other actions apply.
C. Correct D. 3.0.6 establishes exception to 3.0.2 for support systems that have an LCO in the Tech Specs.
Technical Reference(s):
LCO 3.0.4 Bases Proposed references to be provided to applicants during examination:
N one Learning Objective:
RLP-HLO-416 Obj 15.
Question Source:
Bank # Nov 2010 Audit Q#96 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 4 Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.1 Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 99 QUESTION 98 Rev 1 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.3.14 IR 3.8 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities Proposed Question:
Which of the following statements explains the significance of the Maximum Safe Area Radiation Level being reached in the areas identified in EOP
-0003, Secondary Containment and Radioactive Release Control?
A. This is the value that requires entry into EOP
-0003. B. This is the value that requires isolating all systems discharging into the area except systems (1) required for damage control, and (2) required to be operated by EOPs. C. This value in any one area requires the plant to be shutdown per GOP
-0003. D. This value along with a primary system discharging into secondary containment requires entry into EOP
-0003. Proposed Answer:
D Explanation A. The maximum normal operating value is the entry condition for EOP
-0003. B. The max normal operating value is used as the stop sign before this action in EOP-3. C. EOP-0003 requires two areas to be at the Maximum Safe value prior to shutting down per GOP
-0002 Plant Shutdown D. Correct -
Technical Reference(s):
EOP Bases Proposed references to be provided to applicants during examination:
None Learning Objective:
(As available)
Question Source:
New Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.5 Comments: Distractor B replaced due to plausibility December 2014 River Bend Station NRC Examination Senior Reactor Operator 100 QUESTION 99 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.4.16 IR 4.4 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines Proposed Question:
The plant is at 100% power when a manual reactor scram is inserted due to a steam leak on Moisture Separator Reheater B.
The Control Room Supervisor should: A. Enter AOP-0001, REACTOR SCRAM, AOP
-0002, TURBINE TRIP and EOP-0001, RPV CONTROL concurrently. EOP
-0001 may be exited when it is determined that an emergency no longer exists.
B. Enter AOP-0001, REACTOR SCRAM and AOP
-0002, TURBINE TRIP. After AOP immediate actions and required subsequent actions are complete, enter EOP-0001, RPV CONTROL. EOP
-0001 may be exited when it is determined that an emergency no longer exists.
C. Enter EOP-0001, RPV CONTROL and exit when it is determined that an emergency no longer exists. Then enter AOP
-0001, REACTOR SCRAM and AOP-0002, TURBINE GENERATOR TRIP.
D. Enter AOP-0001, REACTOR SCRAM, AOP
-0002 TURBINE GENERATOR TRIP, and EOP
-0001, RPV CONTROL concurrently. EOP
-0001 may only be exited when directed by exit steps in the procedure. Proposed Answer:
A. Explanation:
A. Correct
- The AOPs and EOP are entered concurrently. The EOP is exited when the emergency no longer exits.
B. The listed procedures are executed concurrently.
C. The listed procedures are executed concurrently
. D. First part is correct, but the EOP may be exited when the emergency no longer exists. Exit steps normally only are provided for escalating conditions.
Technical Reference(s):
OSP-0009, Section 10.2 Proposed references to be provided to applicants during examination:
None Learning Objective:
R-LPOPS-HLO218 Obj 4; R
-LP-OPS-HLO520 Obj 2; R
-LPOPS-HLO512 Obj 4 Question Source:
Bank # Oct 2012 Audit #99 Question History:
Last NRC Exam NA Cognitive Level:
Memory or Fundamental Knowledge 2 Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.5 Comments: Comments:
December 2014 River Bend Station NRC Examination Senior Reactor Operator 101 QUESTION 100 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 3 K/A # G2.4.44 IR 4.4 Knowledge of emergency plan protective action recommendations Proposed Question:
Following the declaration of a General Emergency, Protective Action Recommendations (PARs) are issued by ______________.
A. Parish Officials.
B. State Officials.
C. Federal Emergency Management Agency.
D. Emergency Director
. Proposed Answer:
D. Explanation: A. See D. B. See D. C. See D. D. The Emergency Director issues PARs. Offsite agencies determine whether or not to act upon the recommendation.
Technical Reference(s):
EIP-2-007, Protective Action Recommendations, Rev 25 p. 3 of 9 Proposed references to be provided to applicants during examination:
None Learning Objective:
None identified.
Question Source:
Bank # RBS-Audit 2010 #99
Question History:
Last NRC Exam RBS NRC 2007 Cognitive Level:
Memory or Fundamental Knowledge 3 Comprehension or Analysis 10 CFR Part 55 Content:
55.43.b.5 Comments:
RBS December 2014 NRC Initial Written Exam Modified Question
- Parent Questions Q# Parent Question 7 RBS December 2008 NRC Exam #8 16 RBS-OPS-5554 22 RBS-NRC-01108 23 RBS-NRC-444 37 Nov 2010 Audit Q#36 38 RBS-NRC-53 44 RBS-NRC-459 51 RBS-OPS-3255 56 RBS-NRC-168 60 RBS-OPS-06288 62 RBS-LOR-1252 64 RBS-OPS-2307 77 RBS-OPS-06265 83 2008 Audit Q#25 87 RBS-NRC-937 Exam Question # 7 Modified from Parent Q# RBS December 2008 NRC Exam #8 QUESTION 8 Rev 0 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 1 K/A # 295019 AA1.03 Importance Rating
3.0 Proposed
Question:
Due to severe weather in the area around River Bend Station, a loss of offsite power has occurred. All three diesel generators have started and are supplying safety related loads.
What is the status of the Plant Air Systems?
- a. The Service Air compressors will be de
-energized. Instrument Air compressors will be supplied by a safety related power source.
- b. The Instrument Air compressors will be de
-energized. SAS
-AOV134 cross
-connect valve will open to maintain IAS header pressure with the SAS compressors.
- c. Instrument Air and Service Air compressors will both be de
-energized. The diesel air compressor will automatically start to supply plant air needs.
- d. Instrument Air and Service Air compressors will both be de
-energized. The diesel air compressor must be manually started to supply plant air needs.
Proposed Answer:
D.
Exam Question # 16 Modified from Parent Q# RBS-OPS-5554 As part of a routine surveillance, valve timing is being performed on the Low Pressure Core Spray (LPCS) system valves. Currently all valves are in their normal standby position except for E21
-MOVF001, Suppression Pool Suction Valve, which is closed.
A valid high drywell pressure, Low Pressure Core Spray (LPCS) initiation signal is received. Assuming no operator action, which of the following statements identifies the expected response of LPCS?
A. LPCS Pump will start and run with pump suction valve E21-MOVF001 closed.
B. LPCS Pump will start and then immediately trip because it has no suction path.
C. LPCS Pump will not receive a start signal because the suction path permissive is not met. D. E21-MOVF001 receives an automatic open signal and the LPCS Pump starts when the valve is full open.
Answer: A Exam Question # 22 Modified from Parent Q# RBS-NRC-01108 River Bend was operating at full power with RCIC tagged out. A trip of all Reactor Feedwater Pumps caused RPV level to drop and a Reactor scram to occur on low RPV water level. HPCS recovered RPV level and E22
-F004, HPCS INJECT ISOL VALVE was closed when RPV water level reached +35 inches. Conditions are currently:
Vessel is isolated from the condenser Reactor pressure 700 psig (rising at 10 psig per minute)
Time after scram 5 minutes CRD pumps both tripped Drywell pressure 0.3 psid If no additional operator actions are taken other than the ones listed, what is the expected RPV water level response over the next 10 minutes and why
? A. rise due to swell from decay heat.
B. rise due to Feed Reg. Valve leakage exceeding heat requiements.
C. lower due to cooldown.
D. lower due to steam loads reducing RPV water inventory.
Answer: A Exam Question # 23 Modified from Parent Q# RBS-NRC-444 Given the following plant conditions:
- Reactor Power 0% (all rods in)
- Reactor Level +33 inches
- Reactor Pressure 890 psig
- Drywell Pressure 1.8 psid
- Drywell Temperature 138°F
- Containment Temperature 88°F
- Containment Pressure 0.35 psig - Annulus Differential Pressure
-4.5 in.WC Based on the above conditions, which one of the following describes the Emergency Operating Procedures that should be entered?
A. EOP-1 ONLY B. EOP-1 and 2 C. EOP-2 ONLY D. EOP-2 and 3 Answer: B Exam Question # 37 Modified from Parent Q# Nov 2010 Audit Q# 36 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 2 Group # 1 K/A # 215004 IR 2.6 Knowledge of the electrical power supplies to SRM channels/detectors.
Proposed Question: The power supply to SRM 'A' detector is ______________.
A. VBS-PNL01A B. VBS-PNL01B C. RPS A D. RPS B Proposed Answer:
C.
Exam Question # 38 Modified from Parent Q# RBS
-NRC-53 During a reactor startup the ATC has been withdrawing SRM detectors per GOP
-0001. All SRMs except A indicate full out. SRM A has an upscale high and an upscale High
-high trip indicated and is reading off
-scale high. The P680 indications show the detector is "driving out." The ATC should . . .
A. Immediately insert control rods to return SRM A readings on
-scale. B. Insert a Div I half scram and continue with the plant startup.
C. Check the SRM A drive power fuses, if the problem is not corrected, obtain reactor engineering assistance..
D. Since the other drive OUT lights are on, SRM A drive has power therefore contact I & C for assistance.
Answer: C Exam Question # 44 Modified from Parent Q# RBS
-NRC-459 Given the following conditions:
- The plant had been operating at 100% power
- A severe over
-pressure transient resulted in ALL Safety Relief Valves (SRV) opening in the "relief" mode and then lifting in their "safety" mode
- RPV pressure peaked at 1200 psig
- All valves have closed (reseated) with the exception of one SRV that remains open in its "safety" mode
- The required actions of AOP
-0035, "Safety Relief Valve Stuck Open", have been taken including scramming the reactor but the SRV has not closed.
Which of the following describes the resulting tailpipe temperature trend as the plant cools down and depressurizes through the stuck open SRV? (Assume containment pressure is 0 psig and remains constant.)
SRV tailpipe temperature will:
A. start at 260 °F, rise to approximately 290 °F and then will slowly fall following reactor pressure during the depressurization below 500 psig.
B. start at 525 °F and will slowly fall following reactor pressure during the depressurization.
C. start at 285 °F, rise to approximately 325 °F and then will slowly fall following reactor pressure during the depressurization below 500 psig. D. start at 305 °F and will slowly fall following reactor pressure during the depressurization.
Answer: C Exam Question # 51 Modified from Parent Q# RBS
-OPS-3255 The "A" Standby Diesel Generator is operating in parallel with the offsite power supply. If the operator takes the DG Voltage Regulator Control Switch to raise, which of the following indications should be showing the LARGEST change?
A. Frequency (HZ)
B. Real Load (KW)
C. Engine Speed (RPM)
D. Reactive Load (KVAR)
Answer: D Exam Question # 56 Modified from Parent Q# RBS
-NRC-168 If total feedwater flow drops below the reactor recirculation system interlock level, the Reactor Recirculation pumps will downshift to slow speed.
What is the PRIMARY reason for this interloc k? A. pump cavitation.
B. flow control valve cavitation.
C. excessive axial thrust on the pump.
D. inaccurate wide range level indication.
Answer: B Exam Question # 60 Modified from Parent Q# RBS
-OPS-06288 The primary means of hydrogen control following a LOCA utilizes:
A. Hydrogen Igniters B. Hydrogen Recombiners C. Containment Purge subsystem D. Drywell Purge subsystem Answer: A Exam Question # 62 Modified from Parent Q# RBS
-LOR-1252 The plant is operating at 80% power with EHC Pressure Regulator Channel A selected for pressure control. Pressure Regulator Channel B is in TEST.
Which one of the following describes the plant response if the Pressure Regulator Channel A Averaging Manifold Pressure transmitter fails to 0 psig? A. Turbine control valves fully open, steam bypass valves fully open, reactor pressure lowers until MSIVs close.
B. Turbine control valves fully close, steam bypass valves fully open, reactor pressure rises until a reactor scram occurs.
C. Turbine control valves fully open, steam bypass valves remain closed, reactor pressure lowers until MSIVs close.
D. Turbine control valves fully close, steam bypass valves remain closed, reactor pressure rises until a reactor scram occurs.
Answer: D Exam Question # 64 Modified from Parent Q# RBS
-OPS-2307 The HVN Chilled Water system Supply and Return valves to Containment (HVN
-MOV127, 128, 129, 130, 102) will AUTO
-CLOSE upon __________.
A. RPV Level 1 (-143") B. RPV Level 2 (-43") C. Drywell to Annulus d/p greater than 12 inches D. Opening of Service Water supply and Return valves to Containment Answer: B Exam Question # 77 Modified from Parent Q# RBS
-OPS-06265 The plant has experienced an ATWS. The following conditions exist at P680:
All eight white scram solenoid lights are extinguished Annunciator P680 C08 SCRAM PILOT VLV AIR HEADER LOW PRESSURE is alarming SDV Vent and Drain valve position lights indicate all four valves are closed Approximately 20% of the withdrawn control rods fully inserted CRD cooling water differential pressure has been maximized ARI has been inititated Which of the following methods for alternate control rod insertion should be attempted next? A. venting the scram air header.
B. resetting and reinitiating ARI.
C. removing the scram solenoid power fuses.
D. resetting the scram and intiating a manual scram.
Answer: D Exam Question # 83 Modified from Parent Q# 2008 Audit Q# 25 Examination Outline Cross
-
Reference:
Level RO SRO Tier # 1 Group # 2 K/A # 295015 AA2.02 Importance Rating
4.1 Proposed
Question:
Following a Level 3 scram signal, the following conditions exist:
143 control rods indicate full
-in 2 control rods are not full
-in Reactor power indicates 0%
Which of the following concerning reactor status and required procedure implementation is correct? a. It can NOT be determined that the reactor will remain shutdown under all conditions without boron. Transition to EOP
-0001A is required.
-0001A is NOT applicable.
- c. It can NOT be determined that the reactor will remain shutdown under all conditions without boron, but with reactor power less than 5%, EOP
-0001A is NOT applicable. d. It can NOT be determined that the reactor will remain shutdown under all conditions without boron, but no EOP entry condition exists.
Proposed Answer:
A.
Exam Question # 87 Modified from Parent Q# RBS
-NRC-937 The plant is operating at 85% power. The Manual Scram Pushbutton Surveillance is being performed. After arming and depressing the last Manual Scram Pushbutton for DIV 4, the half scram is reset. However, a failure of the K14D relay contacts to reclose when re-energized results in the "B" RPS DIV 4 SCRAM SOV VALVES OPEN light above the DIV 4 Manual Scram Pushbutton on P680 remaining out. All other RPS scram solenoid valve white lights on P680 are lit.
If a loss of RPS Bus A occurs at this time, which one of the following is the expected response and appropriate action to be taken?
A. None of the control rods will scram. Transfer RPS Bus A to Alternate power per AOP
-0010, Loss of RPS Bus.
B. Approximately 1/4 of the control rods will scram. Manually scram the reactor and enter AOP-0001 and AOP
-0010. C. Approximately 1/4 of the control rods will scram. Transfer RPS Bus A to Alternate power per AOP
-0010, Loss of RPS Bus.
D. All of the control rods will scram. Manually scram the reactor and enter AOP
-0001 and AOP-0010. Answer: B Proposed handouts for the Draft Written Exam submittal follow this coversheet.
CONTINUOUS USE ATTACHMENT 13PAGE 2 OF 2INITIATING STANDBY LIQUID CONTROLOSP-0053 REV - 022 PAGE 47 OF 74STANDBY LIQUID CONTROL INJECTION REQUIREMENTSTANK LEVEL PRIOR TANK LEVEL AFTER TANK LEVEL AFTER TO INJECTION INJECTION OF 69 lb Boron INJECTION OF 166 lb Boron (approximately 16 min inj time) (approximately 38 min inj time)
GAL GAL GAL NOTE WHEN tank level falls between values, THEN the smaller value should be used.1531 905 01550 924 19 1600 974 69 1700 1074 169 1800 1174 269 1900 1274 369 2000 1374 469 2100 1474 569 2200 1574 669 2300 1674 769 2400 1774 869 2500 1874 969 2600 1974 1069 2700 2074 1169 2800 2174 1269 2900 2274 1369 3000 2374 1469 3100 2474 1569 3200 2574 1669 3300 2674 1769 3400 2774 1869 3500 2874 1969 3600 2974 2069 3700 3074 2169 3800 3174 2269 3900 3274 2369 4000 3374 2469 4100 3474 2569 4200 3574 2669
HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETYOPERATING MODES:GENERAL EMERGENCYSITE AREA EMERGENCYALERTNOUESYSTEM MALFUNCTIONOPERATING MODES:GENERAL EMERGENCYSITE AREA EMERGENCYALERTNOUE L o s s o f A C P o wSG1Prolonged loss of all offsite and all onsite ACpower to emergency bussesEmergency Action Level(s)
- 1. a. Loss of all offsite and all onsite AC power toDiv I, II and III ENS busses ANDb. Either of the following:Restoration of at least one emergency busin < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely ORSS1Loss of all offsite and all onsite AC power toemergency busses for >15 minutesEmergency Action Level(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition has exceeded,or will likely exceed, the applicable time.
1.LossofalloffsitepowerandallonsiteACSA1AC power capability to emergency bussesreduced to a single power source for >
15minutes such that any additional singlefailure would result in station blackoutEmergency Action Level(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition has exceeded,orwilllikelyexceed,theapplicabletime.SU1Loss of all offsite AC power to emergencybusses for >15 minutesEmergency Action Level(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition has exceeded, orwill likely exceed, the applicable time.
1.LossofalloffsiteACpowertoDivIandII12345D12345DPlant Modes (white boxes indicate applicable modes)1Power Operations2Startup3Hot Shutdown4Cold Shutdown5RefuelDDefueledEMERGENCY CLASSIFICATION USER AID 1 MODES 1 -3ABNORMAL RADIATION LEVELS / RADIOLOGICAL EFFLUENTOPERATING MODES:GENERAL EMERGENCYSITE AREA EMERGENCYALERTNOUE R a d iAG1Offsite dose resulting from an actual or IMMINENTrelease of gaseous radioactivity > 1000 mR TEDE or5000 mR thyroid CDE for the actual or projectedduration of the release using actual meteorologyEmergency Action Level:(1 or 2 or 3)NOTE:The Emergency Director should not wait until theapplicable time has elapsed, but should declare the eventas soon as it is determined that the condition will likelyexceed the applicable time. If dose assessment results areavailable, the classification should be based on EAL #2instead of EAL #1. Do not delay declaration awaitingAS1Offsite dose resulting from an actual or IMMINENTrelease of gaseous radioactivity > 100 mR TEDE or 500mR thyroid CDE for the actual or projected duration ofthe releaseEmergency Action Level(s)(1 or 2 or 3)NOTE:The Emergency Director should not wait until theapplicable time has elapsed, but should declare the eventas soon as it is determined that the condition will likelyexceed the applicable time. If dose assessment results areavailable, the classification should be based on EAL #2instead of EAL #1. Do not delay declaration awaiting AA1Any release of gaseous or liquid radioactivity to theenvironment > 200 times the ODCM limit for >
15minutesEmergency Action Level(s): (1 or 2 or 3
)NOTE:The Emergency Director should not wait until theapplicable time has elapsed, but should declare theevent as soon as it is determined that the releaseduration has exceeded, or will likely exceed theapplicable time. In the absence of data to thecontrary, assume that the release duration hasexceeded the applicable time if an ongoing release is AU1Any release of gaseous or liquid radioactivity to theenvironment > 2 times the ODCM limit for >60 minutesEmergency Action Level(s):(1 or 2 or 3)NOTE:The Emergency Director should not wait until theapplicable time has elapsed, but should declare the eventas soon as it is determined that the release duration hasexceeded, or will likely exceed the applicable time. In theabsence of data to the contrary, assume that the releaseduration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.12345D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 DEIP-2-001 CLASSIFICATION USER AIDATTACHMENT 10PAGE 1 of 2 1 2 3 4 5 D S e c u r i t y HG1HOSTILE ACTION resulting in loss ofphysical control of the facilityEmergency Action Level(s): (1 or 2)1.A HOSTILE ACTION has occurred such thatplant personnel are unable to operateequipment required to maintain safetyfunctions OR2.A HOSTILE ACTION has caused failure ofSpe n t F ue lCoo lin gSyste m s a n dHS1HOSTILE ACTION within the PROTECTEDAREAEmergency Action Level(s)
- 1.A HOSTILE ACTION is occurring or hasoccurred within the PROTECTED AREA asreported by the RBS security shiftsupervisionHA1HOSTILE ACTION within the OWNERCONTROLLED AREA or airborne attackthreatEmergency Action Level(s): (1 or 2)1.A HOSTILE ACTION is occurring or hasoccurred within the OWNER CONTROLLEDAREA as reported by the RBS security shiftsupervision OR 2.A va lidated n otificat i on fr o m N R C o f a nHU1Confirmed SECURITY CONDITION orthreat which indicates a potential degradationin the level of safety of the plantEmergency Action Level(s):(1 or 2 or 3)1.A SECURITY CONDITION that does NOTinvolve a HOSTILE ACTION as reported bythe RBS security shift supervision OR2.A credible site specific security threatnotification 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D w e r ORRPV level cannot be maintained> -162 inches.
1.Loss ofalloffsitepowerandallonsite ACpower to Div I, II and III ENS busses for >
15minutes orwilllikelyexceed,theapplicabletime.1. a. AC power capability to Div I and Div IIENS busses reduced to a single powersource for >15 minutes ANDb. Any additional single failure will result in astation blackout 1.Loss ofalloffsite ACpower toDiv Iand IIENS busses for >15 minutes F a i l u r e o f R e a c t o r P r o t e c t iSG3Automatic scram and all manual actions failto shutdown the reactor and indication of anextreme challenge to the ability to cool thecore existsEmergency Action Level(s)
- 1.a.An automatic scram failed to shutdown thereactor ANDb. All manual actions do not shutdown thereactor as indicated by reactor power >
5%ANDc.Either of the following exist or have occurreddue to continued power generation:SS3Automatic scram fails to shutdown the reactorand the manual actions taken from the reactorcontrol console are not successful in shuttingdown the reactorEmergency Action Level(s)
- 1.a. An automatic scram failed to shutdown thereactor ANDb. Manual actions taken at the reactor controlconsole do not shutdown the reactor asindicated by reactor power >
5%SA3Automatic scram fails to shutdown thereactor and the manual actions taken fromthe reactor control console are successful inshutting down the reactorEmergency Action Level(s)
- 1. a. An automatic scram failed to shutdown thereactor ANDb. Manual actions taken at the reactor controlconsole successfully shutdown the reactor asindicated by reactor power < 5%
i o l o g i c a l E f f l u e n t sdose assessment results.1.VALID reading on any of the radiation monitors in TableR1 > the GENERAL EMERGENCY reading for >
15minutes OR2.Dose assessment using actual meteorology indicatesdoses > 1000 mR TEDE or 5000 mR thyroid CDE at orbeyond the SITE BOUNDARY OR3.Field survey results indicate closed window dose rates >1000 mR/hr expected to continue for >60 minutes; oranalyses of field survey samples indicate thyroid CDE >5000 mR for one hour of inhalation, at or beyond theSITE BOUNDARYdose assessment results.1.VALID reading on any of the radiation monitors in TableR1 > the SITE AREA EMERGENCY reading for >
15minutes OR2.Dose assessment using actual meteorology indicates doses> 100 mR TEDE or 500 mR thyroid CDE at or beyond theSITE BOUNDARY OR3.Field survey results indicate closed window dose rates >100 mR/hr expected to continue for >60 minutes; oranalyses of field survey samples indicate thyroid CDE >500 mR for one hour of inhalation, at or beyond the SITEBOUNDARYdetected and the release start time is unknown.1.VALID reading on any of the radiation monitors inTable R1 > the Alert reading for >15 minutes OR2.For RMS-RE107 effluent monitor:EITHERVALID reading > 200 times the alarm setpointestablished by a current radioactivity discharge permitfor >15 minutes ORVALID reading > 1.27E-01 µCi/ml for >15 minutes OR3.Confirmed sample analysis for gaseous or liquidreleases indicate concentrations or release rates > 200times the ODCM limit for >15 minutes1.VALID reading on any of the radiation monitors in TableR1 > the NOUE reading for >60 minutes OR2.VALID reading on RMS-RE107 effluent monitor > 2times the alarm setpoint established by a currentradioactivity discharge permit for >60 minutes OR3.Confirmed sample analyses for gaseous or liquid releasesindicate concentrations or release rates > 2 times theODCM limit for >60 minutes AA2Damage to irradiated fuel or loss of water level that hasresulted or will result in the uncovering of irradiatedfuel outside the reactor vesselEmergency Action Level(s): (1 or 2)1.A water level drop in the reactor refueling cavity, spentfuel pool or fuel transfer canal that will result iniditdflbid AU2UNPLANNED rise in plant radiation levelsEmergency Action Level(s): (1 or 2)1. a. UNPLANNED water level drop in a reactor refuelingpathway as indicated by any of the following:Water level drop in the reactor refueling cavity, spentfuel pool, or fuel transfer canal indication on ControlRPl 870 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 DSpentFuelCoolingSystemsandIMMINENT fuel damage is likely for afreshly off-loaded reactor core in pool 2.AvalidatednotificationfromNRC of anairliner attack threat within 30 minutes of the site OR3.A validated notification from NRC providinginformation of an aircraft threat D i s c r e t i o n a r y HG2Other conditions exist which in the judgmentof the Emergency Director warrantdeclaration of a General EmergencyEmergency Action Level(s):1.Other conditions exist which in the judgmentof the Emergency Director indicate thatevents are in progress or have occurred whichinvolve actual or IMMINENT substantialcore degradation or melting with potential forloss of containment integrity or HOSTILEACTION that results in an actual loss ofphysical control of the facility. Releases canbe reasonably expected to exceed EPAProtective Action Guideline exposure levelsoffsite for more than the immediate site areaHS2Other conditions exist which in the judgmentof the Emergency Director warrantdeclaration of a SITE AREA EMERGENCYEmergency Action Level(s)
- 1.Other conditions exist which in the judgmentof the Emergency Director indicate thatevents are in progress or have occurred whichinvolve actual or likely major failures ofplant functions needed for protection of thepublic or HOSTILE ACTION that results inintentional damage or malicious acts: (1)toward site personnel or equipment that couldlead to the likely failure of or: (2) thatprevent effective access to equipment neededfor the protection of the public. Any releasesare not expected to result in exposure levelswhich exceed EPA Protective ActionGuideline exposure levels beyond the SITEBOUNDARYHA2Other conditions exist which in the judgmentof the Emergency Director warrantdeclaration of an ALERTEmergency Action Level(s)
- 1.Other conditions exist which in the judgmentof the Emergency Director indicate that eventsare in progress or have occurred which involveactual or likely potential substantialdegradation of the level of safety of the plantor a security event that involves probable lifethreatening risk to site personnel or damage tosite equipment because of HOSTILE ACTION.Any releases are expected to be limited tosmall fractions of the EPA Protective ActionGuideline exposure levelsHU2Other conditions exist which in the judgmentof the Emergency Director warrantdeclaration of aNOUEEmergency Action Level(s)
- 1.Other conditions exist which in the judgmentof the Emergency Director indicate thatevents are in progress or have occurred whichindicate a potential degradation of the levelof safety of the plant or indicate a securitythreat to facility protection has been initiated.No releases of radioactive material requiringoffsite response or monitoring are expectedunless further degradation of safety systemsoccurs 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D i o n S y s t e m pgCore cooling is extremely challenged asindicated by RPV level cannot be restoredand maintained > -186 inches ORHeat removal is extremely challenged asindicated by RPV pressure and SuppressionPool temperature cannot be maintained inthe EOP Heat Capacity Temperature Limit(HCTL) Safe Zone L o s s o f D C P o w e rSS4Loss of all vital DC power for >15 minutesEmergency Action Level(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition has exceeded,or will likely exceed, the applicable time.1.< 105 VDC on all vital DC busses >
15minutesSS6SA6 SU6 A b n o r m a l R a d i a t i o n L e v e l s i rra di a t e d f ue l becom ing uncovere d OR2. A VALID reading on any of the following radiationmonitors due to damage to irradiated fuel or loss ofwater level:RMS-RE140 2000 mR/hrRMS-RE141 2000 mR/hrRMS-RE192 2000 mR/hrRMS-RE193 2000 mR/hrRMS-RE5A 1.64E+03Ci/secRMS-RE5B (GE) 5.29E-04Ci/ml R oom Pane l 870Personnel observation by visual or remote means.
ANDb. UNPLANNED VALID area radiation monitor alarm onany of the following:RMS-RE140RMS-RE141RMS-RE192RMS-RE193 OR2.UNPLANNED VALID area radiation monitor readingsor survey results indicate a rise by a factor of 1000 overnormal* levelsNOTE:For area radiation monitors with ranges incapable ofmeasuring 1000 times normal* levels, classification shallbe based on VALID full scale indications unless surveysconfirm that area radiation levels are below 1000 timesnormal* within 15 minutes of the area radiation monitorindications going full scale.* Normal can be considered the highest reading in the past24 hours excluding the current peak value.
AA3Rise in radiation levels within the facility that impedes C o n t r o l R o o m E v a c u a t i o nHS3Control room evacuation has been initiatedand plant control cannot be establishedEmergency Action Level(s)
- 1. a. Control room evacuation has been initiated ANDb. Control of the plant cannot be established inaccordance with AOP-0031, Shutdown fromOutside the Main Control Room, within 15minutesHA3Control room evacuation has been initiatedEmergency Action Level(s)
- 1.AOP-0031, Shutdown from Outside the MainControl Room requires Control Roomevacuation F i r eHA4FIRE or EXPLOSION affecting theoperability of plant safety systems required toestablish or maintain safe shutdownEmergency Action Level(s)
- 1.FIRE or EXPLOSION resulting in VISIBLEDAMAGE to any of the structures or areas inTable H2 containing safety systems orcomponents or Control Room indication ofdegraded performance of those safety systemsHU4FIRE within the PROTECTED AREA notextinguished within 15 minutes of detection orEXPLOSION within the PROTECTEDAREAEmergency Action Level(s)
- (1 or 2)NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the duration has exceeded, orwill likely exceed, the applicable time.
1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 DBOUNDARYTable R1 EAL THRESHOLDMethodGENERALDRMS ThresholdSITE AREADRMS ThresholdALERTDRMS ThresholdNOUEDRMS ThresholdMain PlantPrimarySecondary4GE125 4.70E+08Ci/secN/A4GE125 4.70E+07Ci/secN/A4GE125 3.06E+07Ci/sec1GE126 2.82E-01Ci/ml4GE125 3.06E+05Ci/sec1GE126 5.26E-03Ci/mlFB VentPrimarySecondary4GE005 6.70E+07Ci/secN/A4GE005 6.70E+06Ci/secN/A4GE005 2.19E+06Ci/sec5GE005 2.82E-01Ci/ml4GE005 2.19E+04Ci/sec5GE005 4.65E-03Ci/mlRW Vent Pi4GE006258E+06Ci/sec4GE006258E+04Ci/sec L o s s o f A n n u n c i a t i o n/I n d i c a t i o nInability to monitor a SIGNIFICANTTRANSIENT in progressEmergency Action Level(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition has exceeded,or will likely exceed, the applicable time.1. a. UNPLANNED loss of > approximately 75%of the following for >15 minutes:Control Room safety system annunciation ORControl Room safety system indication ANDb. A SIGNIFICANT TRANSIENT is inprogress ANDc. Compensatory indications are unavailableUNPLANNED loss of safety systemannunciation or indication in the controlroom with either (1) a SIGNIFICANTTRANSIENT in progress, or (2)compensatory non-alarming indicators arenot availableEmergency Action Level(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition has exceeded,or will likely exceed, the applicable time.1. a. UNPLANNED loss of > approximately 75%of the following for >15 minutes:Control Room safety system annunciation ORControl Room safety system indication ANDb. Either of the following:A SIGNIFICANT TRANSIENT is inprogress ORCompensatoryindicationsunavailableUNPLANNED loss of safety systemannunciation or indication in the ControlRoom for >15 minutesEmergency Action Level(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition has exceeded, orwill likely exceed, the applicable time.1.UNPLANNED loss of > approximately 75%of the following for >15 minutes:a. Control Room safety system annunciation ORb. Control Room safety system indicationoperation of systems required to maintain plant safetyfunctionsEmergency Action Level: (s)1.Dose rate > 15 mR/hr in any of the following areasrequiring continuous occupancy to maintain plantsafety functions:Main Control RoomCAS1.FIRE not extinguished within 15 minutes ofControl Room notification or verification of aControl Room FIRE alarm in any Table H2structure or area OR2. EXPLOSION within the PROTECTEDAREA T o x i c o r F l a m m a b l e G a s sHA5Access to a VITAL AREA is prohibited due totoxic, corrosive, asphyxiant or flammablegases which jeopardize operation of operableequipment required to maintain safeoperations or safely shutdown the reactorEmergency Action Level(s)
- NOTE:If the equipment in the stated area wasalready inoperable, or out of service, beforethe event occurred, then this EAL should notbe declared as it will have no adverse impacton the ability of the plant to safely operate orsafely shutdown beyond that already allowedby Technical Specifications at the time of theevent.HU5Release of toxic, corrosive, asphyxiant orflammable gases deemed detrimental toNORMAL PLANT OPERATIONSEmergency Action Level(s)(1 or 2)1.Toxic, corrosive, asphyxiant or flammablegases in amounts that have or could adverselyaffect NORMAL PLANT OPERATIONS OR2.Report by West Feliciana Parish forevacuation or sheltering of site personnelbased on an offsite event 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D P r imarySecondaryN/AN/A4GE006 2.58E+06Ci/sec5GE006 6.84E-02Ci/ml4GE006 2.58E+04Ci/sec5GE006 6.84E-04Ci/mlFISSION PRODUCT BARRIEROPERATING MODES
- GENERAL EMERGENCYSITE AREA EMERGENCYALERTNOUE F P B L o s s/P o t e n t i a l L o s sFG1Loss of ANY two barriers AND loss or potential loss of thirdbarrierEmergency Action Level(s)
- 1. Loss of any two barriers ANDLoss or potential loss of the third barrierFS1Loss or potential loss of ANY two barriersEmergency Action Level(s):1.Loss or potential loss of any two barriersFA1ANY loss or ANY potential loss of EITHER fuel clad or RCSEmergency Action Level(s):1.Any loss or any potential loss of fuel clad ORAny loss or any potential loss of RCSFU1ANY loss or ANY potential loss of containmentEmergency Action Level(s):1. Any loss or any potential loss of containmentFUEL CLAD (FC) BarrierREACTOR COOLANT SYSTEM (RC) BarrierPRIMARY CONTAINMENT (PC) BarrierParameterLossPotential LossParameterLossPotential LossParameterLossPotential LossFC1PrimarycoolantactivitylevelCoolant activity> 300 µCi/gm dose equivalent I131None RC1DrywellpressureDrywell pressure > 1.68 psidwith indications of reactorcoolant leak in drywell.None PC1Primary containmentconditions1. Rapid unexplained loss of PCpressure following initialpressure rise OR1. PC pressure > 15 psig andrising OR2aPChydrogenintheCompensatoryindicationsunavailable R C S L e a k a g eSU7RCS leakageEmergency Action Level(s): (1 or 2)NOTE:A relief valve that operates and fails toclose per design should be consideredapplicable if the relief valve cannot be isolated.1.Unidentified or pressure boundary leakage > 10 gpm OR2.Identified leakage > 35 gpm L o s s o f CSU8Loss of all onsite or offsite communicationcapabilitiesEmergency Action Level(s): (1 or 2)1.Loss of all of the following onsitecommunications methods affecting the ability toperform routine operations:Plant radio system li12345D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D s e s1.Access to Main Control Room, AuxiliaryBuilding, or 95' Control Building isprohibited due to toxic, corrosive, asphyxiantor flammable gases which jeopardizeoperation of systems required to maintain safeoperations or safely shutdown the reactorHA6Natural or destructive phenomena affectingVITAL AREASEmergencyActionLevel(s):(1or2or3or4or5or6)1.a. Seismic event > Operating Basis Earthquake(OBE) as indicated by:Annunciator "Seismic Tape RecordingSystem Start" (P680-02A-D06)
ANDEventIndicatoronERS-NBI-102iswhite ANDReceiptofEITHER 1 OR 2:1.Annunciator"SeismicEventHigh"(P 680-02 A-C 06)HU6Natural or destructive phenomena affectingthe PROTECTED AREAEmergency Action Level(s):(1 or 2 or 3 or 4 or 5)1.Seismic event identified by any 2 of thefollowing:Seismic event confirmed by activatedseismic switch as indicated by receipt ofEITHER a OR b:a.Annunciator "Seismic TapeRecording SYS Start" (P680-02A-D06)b.Event Indicator on ERS-NBI-102 iswhiteEarthquakefeltinplant 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 Dactivitylevel OR2. PC pressure response notconsistent with LOCAconditions 2.a.PChydrogen intheunsafe zone of HDOLcurve ORb. DW hydrogenconcentration > 9%
OR3. RPV pressure andsuppression pool temperaturecannot be maintained belowthe HCTL FC2Reactor vesselwater levelRPV water level cannot berestored and maintained above-186 inchesRPV water level cannot berestored and maintainedabove -162 inches or cannotbe determined RC2Reactor vesselwater levelRPV water level cannot berestored and maintainedabove -162 inches or cannotbe determinedNone PC2Reactor vesselwater levelNoneEntry into PC floodingprocedures SAP-1 and SAP-2FC3PrimarycontainmentradiationmonitorsContainment radiation monitorRMS-RE16 reading > 3,000 R/hrNone RC3RCS Leak Rate1.UNISOLABLE mainsteam line break asindicated by the failure ofboth MSIVs in any one lineto close ANDHigh MSL flowannunciator (P601-19A-A2)OR1. RCS leakage > 50 gpminside the drywell OR2. UNISOLABLE RCSleakage outside PC asindicated by exceedingeither of the following:a.Max Normal OperatingTemperature (Table F2)PC3Primarycontainmentisolation failure orbypass1. a. Failure of both valves inany one line to close ANDb. Direct downstreampathway to theenvironment exists afterPC isolation OR2. Intentional PC venting perEOPs or SAPsNone o m m u n i c a t i o n P lant pag ing systemSound powered phonesIn-plant telephones OR2.Loss of all of the following offsitecommunications methods affecting the ability toperform offsite notifications:All telephonesNRC phonesState of Louisiana RadioOffsite notifications system and hotline C l a d d i n g D e g r a d a t i o nSU9Fuel clad degradationEmergency Action Level(s): (1 or 2)1.Offgas pre-treatment radiation monitor reading> the Table S1 Dose Rate Limit for the actualindicated offgas flow indicating fuel claddegradation > T.S. allowable limits OR2.Reactor coolant sample activity value indicatingfuel clad degradation > T.S. allowable limits> 4.0 µCi/gm dose equivalent I-131 OR> 0.2µCi/gm dose e quivalent I-131 for > 48 N a t u r a l o r D e s t r u c t i v e P h e n o m e n a (P 680-02 A-C 06)2. Annunciator "Seismic Event High-High" (P680-02A-B06)
ANDamberlight(s) on panel NBI-101 ANDb. Earthquake confirmed by any of thefollowing:Earthquake is felt in plantNational Earthquake CenterControl Room indication of degradedperformance of systems required forsafe shutdown of the plant OR2.Tornado striking resulting in VISIBLEDAMAGE to any of the Table H2 structuresor areas containing safety systems orcomponents or Control Room indication ofdegraded performance of those safetysystems OR3.Internal flooding in Auxiliary Building 70 ftelevation resulting in an electrical shockhazard that precludes access to operate ormonitor safety equipment or Control Roomindication of degraded performance of thosesafety systems O REarthquakefelt inplantNational Earthquake Center OR2.Tornado striking within the PROTECTEDAREA boundary OR3. Internal flooding that has the potential toaffect safety related equipment required byTechnical Specifications for the currentoperating mode in any Table H1 area OR4. Turbine failure resulting in casing penetrationor damage to turbine or generator seals OR5. Severe weather or hurricane conditions withindication of SUSTAINED high winds >
74mph within the PROTECTED AREAboundaryTable H2Structures Containing Functionsor Systems Required for SafeShutdownReactorBuildingStandby CoolingTowerAuxiliaryBuildingDiesel GeneratorBuildingControlBuildingTunnels (B, D, E, F, G)Table H1Uncontrolled Flooding ThresholdArea Water LevelAffectedLocation /ParameterMax Safe OperatingValue / IndicatorAuxBldg6inchesabovefloor 1 2 3 4 5 DTable S1Flow(cfm)Dose Rate(mR/hr)<159000>15-178000>17-207000>20-255000>25-304000>30-602000 OR2. Indication of anUNISOLABLE HPCS,feedwater, RWCU or RCICbreak OR3. Emergency RPVdepressurization is required ORb.Max Normal AreaRadiation (Table F2)
OR3. UNISOLABLE RCSleakage outside PC asindicated by exceedingeither of the following:a. Max Safe OperatingTemperature (Table F1)
ORb. Max Safe Area Radiation(Table F1)
RC4DrywellradiationDrywell radiation monitorRMS-RE20 reading > 100R/hr due to reactor coolantleakageNone PC4Primary containmentradiation monitorsNoneContainment radiation monitorRMS-RE16 reading> 10,000 R/hrFC4EmergencyDirectorjudgmentAny condition in the opinion ofthe Emergency Director thatindicates loss of the Fuel CladbarrierAny condition in the opinionof the Emergency Directorthat indicates potential lossof the Fuel Clad barrier RC5EmergencyDirectorjudgmentAny condition in the opinionof the Emergency Directorthat indicates loss of the RCSbarrierAny condition in the opinionof the Emergency Directorthat indicates potential loss ofthe RCS barrierPC5Emergency DirectorjudgmentAny condition in the opinion ofthe Emergency Director thatindicates loss of the PrimaryContainment barrierAny condition in the opinion ofthe Emergency Director thatindicates potential loss of thePrimary Containment barrierTABLE F1TABLE F2µg qhours I n a d v e r t e n t C r i t i c a l i t ySU10Inadvertent criticalityEmergency Action Level(s)
- 1.UNPLANNED sustained positive periodobserved on nuclear instrumentation T E C H S P E C T i m e L i m i t E xSU11Inability to reach required operating modewithin Technical Specification limitsEmergency Action Level(s)
- 1.Plant is not brought to required operating modewithin Technical Specifications LCO ActionStatement timeEVENTS RELATED TO ISFSIOPERATING MODES:4.Turbine failure-generated PROJECTILESresulting in VISIBLE DAMAGE to orpenetration of any of the Table H2 structuresor areas containing safety systems orcomponents or Control Room indication ofdegraded performance of those safetysystems OR5.Vehicle crash resulting in VISIBLEDAMAGE to any of the Table H2 structuresor areas containing safety systems orcomponents or Control Room indication ofdegraded performance of those safetysystems OR6.Hurricane or high SUSTAINED windconditions >74 mph within thePROTECTED AREA boundary and resultingin VISIBLE DAMAGE to any of the TableH2 structures or areas containing safetysystems or components or Control Roomindication of degraded performance of thosesafety systems 1 2 3 4 5 D 1 2 3 4 5 D12345DBuilding G)Fuel BuildingAuxBldgCrescent Area70' EL 6inchesabovefloor(must be verified locally)HPCS Room70'EL4 inches above floor(P870-51A-G4)RHR A Room70'EL4 inches above floor(P870-51A-G4)RHR B Room70'EL4 inches above floor(P870-51A-G4)RHR C Room70'EL4 inches above floor(P870-51A-G4)LPCS Room70'EL4 inches above floor(P870-51A-G4)RCIC Room70'EL4 inches above floor(P870-51A-G4)>60-1401000>140-200700PC 3 Loss of Primary ContainmentParameterArea TemperatureMax Safe Operating ValueArea Radiation LevelDRMS Grid 2Max Safe OperatingValueRHR A equip area200º F1213 9.5E+03 mR/hrRHR B equip area200º F1214 9.5E+03 mR/hrRHR C equip areaN/A1215 9.5E+03 mR/hrRCIC room200º F1219 9.5E+03 mR/hrMSL Tunnel200º FN/ARWCU pump room 1 (A) / 2 (B)200º FN/ARC 3 Potential Loss of RCSParameterArea Temperature(isolation temperature alarm)Area Radiation LevelDRMS Grid 2Max NormalOperatingValueRHR A equip area117º F (P601-20A-B4)1213 8.2E+01 mR/hrRHR B equip area117º F (P601-20A-B4)1214 8.2E+01 mR/hrRHR C equip areaN/A1215 8.2E+01 mR/hrRCIC room182º F (P601-21A-B6)1219 1.20E+02 mR/hrMSL Tunnel173º F (P601-19A-A1/A3/B1/B3)N/ARWCU pump room 1 (A) / 2 (B)165º F (P680-1A-A2/B2)N/A x c e e d e dGENERAL EMERGENCYSITE AREA EMERGENCYALERTNOUE C a s k D a m a g eE-HU1Damage to a loaded cask CONFINEMENTBOUNDARYEmergency Action Level(s)
- 1.Damage to a loaded cask CONFINEMENTBOUNDARY 1 2 3 4 5 DEIP-2-001 ATT 10 USER AID 1 Revision DEIP-2-001 REV -24 PAGE 157 of 158 EMERGENCY CLASSIFICATION USER AID 2 MODES 4, 5, and DefueledCOLD SHUTDOWN / REFUELINGOPERATING MODES:GENERAL EMERGENCYSITE AREA EMERGENCYALERTNOUECG1Loss of RCS/RPV inventory affecting fuelclad integrity with containmentchallengedEmergencyActionLevel(s):(1or2)CS1Loss of RCS/RPV inventory affecting coredecay heat removal capabilityEmergency Action Level(s): (1 or 2 or 3)
CA1Loss of RCS inventoryEmergency Action Level(s): (1 or 2)NOTE:TheEmergencyDirectorshouldnotwait CU1RCS leakageEmergency Action Level(s)
- NOTE:TheEmergencyDirectorshouldnotwait 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 DATTACHMENT 10PAGE 2 of 2EIP-2-001 CLASSIFICATION USER AID12345DPlant Modes (white boxes indicate applicable modes)1Power Operations2Startup3Hot Shutdown4Cold Shutdown5RefuelDDefueledABNORMAL RADIATION LEVELS / RADIOLOGICAL EFFLUENTOPERATING MODES:GENERAL EMERGENCYSITE AREA EMERGENCYALERTNOUEAG1Offsite dose resulting from an actual or IMMINENTrelease of gaseous radioactivity > 1000 mR TEDE or5000 mR thyroid CDE for the actual or projectedduration of the release using actual meteorologyAS1Offsite dose resulting from an actual or IMMINENTrelease of gaseous radioactivity > 100 mR TEDE or 500mR thyroid CDE for the actual or projected duration ofthe release AA1Any release of gaseous or liquid radioactivity to theenvironment > 200 times the ODCM limit for >
15minutesEmergency Action Level(s): (1 or 2 or 3
)AU1Any release of gaseous or liquid radioactivity to theenvironment > 2 times the ODCM limit for >60 minutesEmergency Action Level(s):(1 or 2 or 3)NOTE:The Emergency Director should not wait until the12345D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 DHAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETYOPERATING MODES:GENERAL EMERGENCYSITE AREA EMERGENCYALERTNOUE12345D HG1HOSTILE ACTION resulting in loss ofphysical control of the facilityEmergencyActionLevel(s):(1or2)HS1HOSTILE ACTION within the PROTECTEDAREAEmergencyActionLevel(s)
- HA1HOSTILE ACTION within the OWNERCONTROLLED AREA or airborne attackthreatHU1Confirmed SECURITY CONDITION orthreat which indicates a potential degradationin the level of safety of the plant 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D L o s s o f R C S/R P V I n v e n t o r yEmergencyActionLevel(s): (1 or 2)1. a. RPV level < -162 inches (TAF) for >
30minutes ANDb. Any containment challenge indication inTable C1 OR2. a. RCS level cannot be monitored with coreuncovery indicated by any of thefollowing for >30 minutes:RMS-RE16 reading > 100 R/hrErratic source range monitorindicationUnexplained rise in floor orequipment sump level, SuppressionPool level, vessel make-up rate orobservation of leakage or inventoryloss ANDb. Any containment challenge indication inTable C1NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition will likelyexceed the applicable time.1.With CONTAINMENT CLOSURE notestablished, UNPLANNED RPV level < -49inches OR2.With CONTAINMENT CLOSUREestablished, RPV level < -162 inches (TAF)
OR3. RCS level cannot be monitored for >30 minutes with a loss of RCSinventory as indicated by any of thefollowing:RMS-RE16 reading > 100 R/hrErratic Source Range MonitorindicationUnexplained rise in floor orequipment sump level, SuppressionPool level, vessel make-up rate orobservation of leakage or inventory lossNOTE:TheEmergencyDirectorshould notwaituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition will likelyexceed the applicable time.1.UNPLANNED loss of RCS inventory asindicated by RPV level < -43 inches (Level 2)OR2.RCS level cannot be monitored for >
15minutes with a loss of RCS inventory asindicated by anunexplained rise in flooror equipment sump level, SuppressionPool level, vessel make-up rate orobservation of leakage or inventory lossNOTE:TheEmergencyDirectorshould notwaituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition will likelyexceed the applicable time.1.RCS leakage results in the inability tomaintain or restore RPV level > +9.7 inches(Level 3) for >15 minutesTable C1Containment ChallengeIndications R a d i o l o g i c a l E f f l u e n t sEmergency Action Level:(1 or 2 or 3)NOTE:The Emergency Director should not wait until theapplicable time has elapsed, but should declare the eventas soon as it is determined that the condition will likelyexceed the applicable time. If dose assessment results areavailable, the classification should be based on EAL #2instead of EAL #1. Do not delay declaration awaitingdose assessment results.1.VALID reading on any of the radiation monitors in TableR1 > the GENERAL EMERGENCY reading for >
15minutes OR2.Dose assessment using actual meteorology indicatesdoses > 1000 mR TEDE or 5000 mR thyroid CDE at orbeyond the SITE BOUNDARY OR3.Field survey results indicate closed window dose rates >1000 mR/hr expected to continue for >60 minutes; oranalyses of field survey samples indicate thyroid CDE >5000 mR for one hour of inhalation, at or beyond theSITE BOUNDARYEmergency Action Level(s)(1 or 2 or 3)NOTE:The Emergency Director should not wait until theapplicable time has elapsed, but should declare the eventas soon as it is determined that the condition will likelyexceed the applicable time. If dose assessment results areavailable, the classification should be based on EAL #2instead of EAL #1. Do not delay declaration awaitingdose assessment results.1.VALID reading on any of the radiation monitors in TableR1 > the SITE AREA EMERGENCY reading for >
15minutes OR2.Dose assessment using actual meteorology indicates doses> 100 mR TEDE or 500 mR thyroid CDE at or beyond theSITE BOUNDARY OR3.Field survey results indicate closed window dose rates >100 mR/hr expected to continue for >60 minutes; oranalyses of field survey samples indicate thyroid CDE >500 mR for one hour of inhalation, at or beyond the SITEBOUNDARYNOTE:The Emergency Director should not wait until theapplicable time has elapsed, but should declare theevent as soon as it is determined that the releaseduration has exceeded, or will likely exceed theapplicable time. In the absence of data to thecontrary, assume that the release duration hasexceeded the applicable time if an ongoing release isdetected and the release start time is unknown.1.VALID reading on any of the radiation monitors inTable R1 > the Alert reading for >15 minutes OR2.For RMS-RE107 effluent monitor:EITHERVALID reading > 200 times the alarm setpointestablished by a current radioactivity discharge permitfor >15 minutes ORVALID reading > 1.27E-01 µCi/ml for >15 minutes OR3.Confirmed sample analysis for gaseous or liquidreleases indicate concentrations or release rates > 200times the ODCM limit for >15 minutesapplicable time has elapsed, but should declare the eventas soon as it is determined that the release duration hasexceeded, or will likely exceed the applicable time. In theabsence of data to the contrary, assume that the releaseduration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.1.VALID reading on any of the radiation monitors in TableR1 > the NOUE reading for >60 minutes OR2.VALID reading on RMS-RE107 effluent monitor > 2times the alarm setpoint established by a currentradioactivity discharge permit for >60 minutes OR3.Confirmed sample analyses for gaseous or liquid releasesindicate concentrations or release rates > 2 times theODCM limit for >60 minutes S e c u r i t yEmergencyActionLevel(s): (1 or 2)1.A HOSTILE ACTION has occurred such thatplant personnel are unable to operateequipment required to maintain safetyfunctions OR2.A HOSTILE ACTION has caused failure ofSpent Fuel Cooling Systems andIMMINENT fuel damage is likely for afreshly off-loaded reactor core in poolEmergencyActionLevel(s):1.A HOSTILE ACTION is occurring or hasoccurred within the PROTECTED AREA asreported by the RBS security shiftsupervisionEmergency Action Level(s): (1 or 2)1.A HOSTILE ACTION is occurring or hasoccurred within the OWNER CONTROLLEDAREA as reported by the RBS security shiftsupervision OR2.A validated notification from NRC of anairliner attack threat within 30 minutes of the siteEmergency Action Level(s):(1 or 2 or 3)1.A SECURITY CONDITION that does NOTinvolve a HOSTILE ACTION as reported bythe RBS security shift supervision OR2.A credible site specific security threatnotification OR3.A validated notification from NRC providinginformation of an aircraft threat D i s c r e t i o n a r y HG2Other conditions exist which in the judgmentof the Emergency Director warrantdeclaration of a General EmergencyEmergency Action Level(s):1.Other conditions exist which in the judgmentof the Emergency Director indicate thatevents are in progress or have occurred whichinvolve actual or IMMINENT substantialcore degradation or melting with potential forloss of containment integrity or HOSTILEACTIONthatresultsinanactuallossofHS2Other conditions exist which in the judgmentof the Emergency Director warrantdeclaration of a SITE AREA EMERGENCYEmergency Action Level(s)
- 1.Other conditions exist which in the judgmentof the Emergency Director indicate thatevents are in progress or have occurred whichinvolve actual or likely major failures ofplant functions needed for protection of thepublic or HOSTILE ACTION that results inintentionaldamageormaliciousacts:(1)HA2Other conditions exist which in the judgmentof the Emergency Director warrantdeclaration of an ALERTEmergency Action Level(s)
- 1.Other conditions exist which in the judgmentof the Emergency Director indicate that eventsare in progress or have occurred which involveactual or likely potential substantialdegradation of the level of safety of the plantor a security event that involves probable lifethreatening risk to site personnel or damage toHU2Other conditions exist which in the judgmentof the Emergency Director warrantdeclaration of aNOUEEmergency Action Level(s)
- 1.Other conditions exist which in the judgmentof the Emergency Director indicate thatevents are in progress or have occurred whichindicate a potential degradation of the levelof safety of the plant or indicate a securitythreat to facility protection has been initiated.No releases of radioactive material requiring 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D L o s s o f R C S/R P V I n v e n t o r y CU2UNPLANNED loss of RCS/RPV inventoryEmergency Action Level(s): (1 or 2)NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition will likelyexceed the applicable time.1. UNPLANNED RCS level drop as indicatedby either of the following:a. RCS water level drop below the RPV flangefor >15minutes when the RCS level bandis established above the RPV flange ORb. RCS water level drop below the RPV levelband for >15minutes when the RCS levelbandisestablishedbelowtheRPVflangeCONTAINMENT CLOSURE notestablishedExplosive mixture inside containmentUNPLANNED rise in containmentpressureSecondary containment area radiationmonitor above EOP Max SafeOperating Value below:AreaDRMSGrid 2Max SafeOperatingValueRHREquipRm A12139.5E+03 mR/hrRHREquipRm B12149.5E+03 mR/hrRHREquipRm C12159.5E+03 mR/hr 1 2 3 4 5 D A b n o r m a l R a d i a t i o n L e v e l s AA2Damage to irradiated fuel or loss of water level that hasresulted or will result in the uncovering of irradiatedfuel outside the reactor vesselEmergency Action Level(s): (1 or 2)1.A water level drop in the reactor refueling cavity, spentfuel pool or fuel transfer canal that will result inirradiated fuel becoming uncovered OR2. A VALID reading on any of the following radiationmonitors due to damage to irradiated fuel or loss ofwater level:RMS-RE140 2000 mR/hrRMS-RE141 2000 mR/hrRMS-RE192 2000 mR/hrRMS-RE193 2000 mR/hrRMS-RE5A 1.64E+03Ci/secRMS-RE5B (GE) 5.29E-04Ci/ml AU2UNPLANNED rise in plant radiation levelsEmergency Action Level(s): (1 or 2)1. a. UNPLANNED water level drop in a reactor refuelingpathway as indicated by any of the following:Water level drop in the reactor refueling cavity, spentfuel pool, or fuel transfer canal indication on ControlRoom Panel 870Personnel observation by visual or remote means.
ANDb. UNPLANNED VALID area radiation monitor alarm onany of the following:RMS-RE140RMS-RE141RMS-RE192RMS-RE193 OR2.UNPLANNED VALID area radiation monitor readingsor survey results indicate a rise by a factor of 1000 overnormal* levelsNOTE:For area radiation monitors with ranges incapable ofmeasuring 1000 times normal* levels, classificationshall be based on VALID full scale indications unlesssurveysconfirmthatarearadiationlevelsarebelow 1 2 3 4 5 D 1 2 3 4 5 DACTIONthatresults in anactual loss ofphysical control of the facility. Releases canbe reasonably expected to exceed EPAProtective Action Guideline exposure levelsoffsite for more than the immediate site areaintentionaldamage ormaliciousacts:(1)toward site personnel or equipment that couldlead to the likely failure of or: (2) thatprevent effective access to equipment neededfor the protection of the public. Any releasesare not expected to result in exposure levelswhich exceed EPA Protective ActionGuideline exposure levels beyond the SITEBOUNDARYsite equipment because of HOSTILE ACTION.Any releases are expected to be limited tosmall fractions of the EPA Protective ActionGuideline exposure levelsoffsite response or monitoring are expectedunless further degradation of safety systemsoccurs C o n t r o l R o o m E v a c u a t i o nHS3Control room evacuation has been initiatedand plant control cannot be establishedEmergency Action Level(s)
- 1. a. Control room evacuation has been initiated ANDb. Control of the plant cannot be established inaccordance with AOP-0031, Shutdown fromOutside the Main Control Room, within 15minutesHA3Control room evacuation has been initiatedEmergency Action Level(s)
- 1.AOP-0031, Shutdown from Outside the MainControl Room requires Control RoomevacuationHA4FIRE or EXPLOSION affecting theoperability of plant safety systems required toestablish or maintain safe shutdownHU4FIRE within the PROTECTED AREA notextinguished within 15 minutes of detection orEXPLOSION within the PROTECTEDAREA 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 Dband isestablishedbelowtheRPVflange OR2. RCS level cannot be monitored with a lossof RCS inventory as indicated by anunexplained rise in floor or equipmentsump level, Suppression Pool level,vessel make-up rate or observation ofleakage or inventory loss L o s s o f D e c a y H e a t R CA3Inability to maintain plant in cold shutdownEmergency Action Level(s):(1or 2)1.An UNPLANNED event results in RCStemperature >200 °F > the specifiedduration in Table C2 OR2.An UNPLANNED event results in RCSpressure rise > 10 psig due to a loss of RCScooling CU3UNPLANNED loss of decay heat removalcapability with irradiated fuel in the RPVEmergency Action Level(s): (1 or 2)NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition will likelyexceed the applicable time.1.An UNPLANNED event results in RCStemperature exceeding 200°F OR2.Loss of all RCS temperature and RCS/RPVlevel indication for >15 minutes 1 2 3 4 5 D 1 2 3 4 5 DTable C2RCS Reheat Duration ThresholdsRCSContainmentClosureDurationTable R1 EAL THRESHOLDsurveysconfirmthatarearadiationlevels arebelow1000 times normal* within 15 minutes of the arearadiation monitor indications going full scale.* Normal can be considered the highest reading in the past24 hours excluding the current peak value.
AA3Rise in radiation levels within the facility that impedesoperation of systems required to maintain plant safetyfunctionsEmergency Action Level: (s)1.Dose rate > 15 mR/hr in any of the following areasrequiring continuous occupancy to maintain plantsafety functions:Main Control RoomCAS 1 2 3 4 5 D F i r eEmergency Action Level(s)
- 1.FIRE or EXPLOSION resulting in VISIBLEDAMAGE to any of the structures or areas inTable H2 containing safety systems orcomponents or Control Room indication ofdegraded performance of those safety systemsAREAEmergency Action Level(s)
- (1 or 2)NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the duration has exceeded, orwill likely exceed, the applicable time.1.FIRE not extinguished within 15 minutes ofControl Room notification or verification of aControl Room FIRE alarm in any Table H2structure or area OR2. EXPLOSION within the PROTECTEDAREA T oHA5Access to a VITAL AREA is prohibited due totoxic, corrosive, asphyxiant or flammablegases which jeopardize operation of operableequipment required to maintain safeoperations or safely shutdown the reactorEmer genc y Action Level (s):HU5Release of toxic, corrosive, asphyxiant orflammable gases deemed detrimental toNORMAL PLANT OPERATIONSEmergency Action Level(s)(1 or 2)1.Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversel y 1 2 3 4 5 D 1 2 3 4 5 DEVENTS RELATED TO ISFSIOPERATING MODES:
R e m o v a l L o s s o f A C P o w CA5Loss of all offsite and all onsite AC power toemergency busses for >15 minutesEmergencyActionLevel(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition will likelyexceed the applicable time.1.Loss of all offsite and all onsite AC power to CU5AC power capability to emergency bussesreduced to a single power source for >
15minutes such that any additional singlefailure would result in station blackoutEmergency Action Level(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition will likelyexceed the applicable time.
1 2 3 4 5 D 1 2 3 4 5 D12345DIntactN/A60 minutes*NotintactEstablished20 minutes*NotEstablished0 minutes*If and RCS heat removal system is inoperation within this time frame and RCStemperature is being reduced, the EAL is notapplicable.MethodGENERALDRMS ThresholdSITE AREADRMS ThresholdALERTDRMS ThresholdNOUEDRMS ThresholdMain PlantPrimarySecondary4GE125 4.70E+08Ci/secN/A4GE125 4.70E+07Ci/secN/A4GE125 3.06E+07Ci/sec1GE126 2.82E-01Ci/ml4GE125 3.06E+05Ci/sec1GE126 5.26E-03Ci/mlFB VentPrimarySecondary4GE005 6.70E+07Ci/secN/A4GE005 6.70E+06Ci/secN/A4GE005 2.19E+06Ci/sec5GE005 2.82E-01Ci/ml4GE005 2.19E+04Ci/sec5GE005 4.65E-03Ci/mlRW VentPrimarySecondaryN/AN/A4GE006 2.58E+06Ci/sec5GE006 6.84E-02Ci/ml4GE006 2.58E+04Ci/sec5GE006 6.84E-04Ci/ml o x i c o r F l a m m a b l e G a s s e sgy()NOTE:If the equipment in the stated area wasalready inoperable, or out of service, beforethe event occurred, then this EAL should notbe declared as it will have no adverse impacton the ability of the plant to safely operate orsafely shutdown beyond that already allowedby Technical Specifications at the time of theevent.1.Access to Main Control Room, AuxiliaryBuilding, or 95' Control Building isprohibited due to toxic, corrosive, asphyxiantor flammable gases which jeopardizeoperation of systems required to maintain safeoperations or safely shutdown the reactor gyaffect NORMAL PLANT OPERATIONS OR2.Report by West Feliciana Parish forevacuation or sheltering of site personnelbased on an offsite eventHA6Natural or destructive phenomena affectingVITAL AREASEmergencyActionLevel(s):(1or2or3or4or5or6)1.a. Seismic event > Operating Basis Earthquake(OBE) as indicated by:HU6Natural or destructive phenomena affecting thePROTECTED AREAEmergency Action Level(s):(1 or 2 or 3 or 4 or 5)1.Seismic event identified by any 2 of thefollowing:Seismic event confirmed by activated 1 2 3 4 5 D 1 2 3 4 5 DGENERAL EMERGENCYSITE AREA EMERGENCYALERTNOUE C o n f i n e m e n t B o u n d a r y d a m a g e/l o s sE-HU1Damage to a loaded cask CONFINEMENT BOUNDARYEmergency Action Level(s)
- 1.Damage to a loaded cask CONFINEMENT BOUNDARY e r pDiv I and Div II ENS busses for >
15minutes1. a. AC power capability to Div I and Div IIENS busses reduced to a single powersource for >15 minutes ANDb. Any additional single power source failurewill result in station blackout L o s s o f D C P o w e r CU6Loss of required DC power for >
15minutesEmergency Action Level(s)
- NOTE:The Emergency Director should not waituntil the applicable time has elapsed, butshould declare the event as soon as it isdetermined that the condition will likelyexceed the applicable time.1.< 105 VDCon required Vital DC bussesfor >15 minutes I n a d v e r t e CU7Inadvertent criticalityEmergencyActionLevel(s)
- 1 2 3 4 5 D 1 2 3 4 5 D 1 2 3 4 5 D N a t u r a l o r D e s t r u c t i vAnnunciator "Seismic Tape RecordingSystem Start" (P680-02A-D06)
ANDEventIndicatoronERS-NBI-102iswhite ANDReceiptofEITHER 1 OR 2:1.Annunciator"SeismicEventHigh"(P680-02A-C06)2. Annunciator "Seismic Event High-High" (P680-02A-B06)
ANDamberlight(s) on panel NBI-101 ANDb. Earthquake confirmed by any of thefollowing:Earthquake is felt in plantNational Earthquake CenterControl Room indication of degradedperformance of systems required forsafe shutdown of the plant OR2.Tornado striking resulting in VISIBLEDAMAGE to any of the Table H2 structuresor areas containing safety systems orcomponents or Control Room indication ofdegraded performance of those safetysystemsseismic switch as indicated by receipt ofEITHER a OR b:a.Annunciator "Seismic Tape RecordingSYS Start" (P680-02A-D06)b.Event Indicator on ERS-NBI-102 iswhiteEarthquake felt in plantNational Earthquake Center OR2.Tornado striking within the PROTECTEDAREA boundary OR3. Internal flooding that has the potential to affectsafety related equipment required by TechnicalSpecifications for the current operating modein any Table H1 area OR4. Turbine failure resulting in casing penetrationor damage to turbine or generator seals OR5. Severe weather or hurricane conditions withindication of SUSTAINED high winds >
74mph within the PROTECTED AREAboundaryTable H2Structures Containing Functionsor Systems Required for SafeShutdownReactorStandbyCoolingTableH1 e n t C r i t i c a l i t yEmergencyActionLevel(s):1.UNPLANNED sustained positive periodobserved on nuclear instrumentation L o s s o f C o m m u n i c a t i o n s CU8Loss of all onsite or offsite communicationscapabilitiesEmergency Action Level(s): (1 or 2)1.Loss of all of the following onsitecommunication methods affecting theability to perform routine operations:Plant radio systemPlant paging systemSound powered phonesIn-plant telephones OR2.Loss of all of the following offsitecommunication methods affecting theability to perform offsite notifications:All telephonesNRC phonesState of Louisiana RadioOffsite notification system andhotline 1 2 3 4 5 D v e P h e n o m e n a O R3.Internal flooding in Auxiliary Building 70 ftelevation resulting in an electrical shockhazard that precludes access to operate ormonitor safety equipment or Control Roomindication of degraded performance of thosesafety systems OR4.Turbine failure-generated PROJECTILESresulting in VISIBLE DAMAGE to orpenetration of any of the Table H2 structuresor areas containing safety systems orcomponents or Control Room indication ofdegraded performance of those safetysystems OR5.Vehicle crash resulting in VISIBLEDAMAGE to any of the Table H2 structuresor areas containing safety systems orcomponents or Control Room indication ofdegraded performance of those safetysystems OR6.Hurricane or high SUSTAINED windconditions >74 mph within thePROTECTED AREA boundary and resultinginVISIBLEDAMAGEtoanyoftheTableReactorBuildingStandbyCoolingTowerAuxiliaryBuildingDiesel GeneratorBuildingControlBuildingTunnels (B, D, E, F, G)Fuel BuildingTable H1Uncontrolled Flooding ThresholdArea Water LevelAffectedLocation /ParameterMax Safe OperatingValue / IndicatorAux BldgCrescent Area70' EL6 inches above floor(must be verified locally)HPCS Room70'EL4 inches above floor(P870-51A-G4)RHR A Room70'EL4 inches above floor(P870-51A-G4)RHR B Room70'EL4 inches above floor(P870-51A-G4)RHR C Room70'EL4 inches above floor(P870-51A-G4)LPCS Room 4 inches above floorhotlineEIP-2-001 ATT 10 USER AID 2 Revision DEIP-2-001 REV -24 PAGE 158 of 158 inVISIBLEDAMAGE toany oftheTableH2 structures or areas containing safetysystems or components or Control Roomindication of degraded performance of thosesafety systems70'EL(P870-51A-G4)RCIC Room70'EL4 inches above floor(P870-51A-G4)
CONTINUOUS USE ATTACHMENT 1PAGE 1 OF 8AVAILABLE SAFE SHUTDOWN SYSTEMSAOP-0052 REV - 025 PAGE 9 OF 62Fire Area:AB-1 AB-2 AB-3 AB-4 AB-5AB-6 AB-7 AB-10 AB-13 West Side HPCS & RHR C, RCIC LPCS & MS Tunnel Standby Gas Crescent HPCS HatchRHR B & RWCU RHR A LPCS Hatch D-Tunnel South End Treatment BElectrical PowerDiv. I X X X X X XDiv. II X X X X X X XDiv. III X X X X X XOffsiteRPV Level ControlHPCS (Div. III) X X X X XRCIC (Div. I) X X XLPCS (Div. I) X X X X X XLPCI-RHR C (Div. II) X X X X X XRPV Pressure ControlDiv. I SRVs X X X X X XDiv. II SRVs X X X X X X XDecay Heat RemovalRHR A Suppression Pool Cooling X X X X XRHR A Alternate Shutdown Cooling X X X X XRHR A Normal Shutdown CoolingRHR B Suppression Pool Cooling X X X X X XRHR B Alternate Shutdown Cooling X X X X X XRHR B Normal Shutdown CoolingMech./Environ. SupportDiv. I Standby Service Water X X X X X XDiv. I HVAC Systems X X X X X XDiv. I Containment/RPV Monitoring X X X X X XDiv. II Standby Service Water X X X X X X XDiv. II HVAC Systems X X X X X X XDiv. II Containment/RPV Monitoring X X X X X X XNormal Service WaterEquipment availability dependsupon fire location in Fire AreaLocal Manual Action Required?N Y Y N N N Y N N CONTINUOUS USE ATTACHMENT 1PAGE 2 OF 8AVAILABLE SAFE SHUTDOWN SYSTEMSAOP-0052 REV - 025 PAGE 10 OF 62Fire Area:AB-14 AB-15 AB-17 AB-18 C-1 C-2 C-3 C-4 Standby GasEast Side Cont. Vent D-Tunnel Cable ChaseCable ChaseRm. North ofACU Rooms Treatment A Crescent Flt. Train Rm.Cable ChaseI (NE) II (SE) ACU RoomsEast WestElectrical PowerDiv. I X X X X X X XDiv. II X X partial X XDiv. III X X X X X XOffsiteRPV Level ControlHPCS (Div. III) X X X X XRCIC (Div. I) X X X XLPCS (Div. I) X X X X X X XLPCI-RHR C (Div. II) X X X XRPV Pressure ControlDiv. I SRVs X X X X X X XDiv. II SRVs X X X XDecay Heat RemovalRHR A Suppression Pool Cooling X X X X X X XRHR A Alternate Shutdown Cooling X X X X X X XRHR A Normal Shutdown Cooling XRHR B Suppression Pool Cooling X X X XRHR B Alternate Shutdown Cooling X X X XRHR B Normal Shutdown CoolingMech./Environ. SupportDiv. I Standby Service Water X X X 1 pump 1 pump X XDiv. I HVAC Systems X X X X X X XDiv. I Containment/RPV Monitoring X X X X X X XDiv. II Standby Service Water X X X XDiv. II HVAC Systems X X X XDiv. II Containment/RPV Monitoring X X X XNormal Service WaterEquipment availability dependsupon fire location in Fire Area XLocal Manual Action Required?N Y N Y Y Y N N N CONTINUOUS USE ATTACHMENT 1PAGE 3 OF 8AVAILABLE SAFE SHUTDOWN SYSTEMSAOP-0052 REV - 025 PAGE 11 OF 62Fire Area:C-5 C-6 C-7 C-9 C-10 C-11 C-13E C-13W C-14 Cable Area Remainder Post-Accid. Cable ChaseCable ChaseVestibule HVK Chiller HVK Chiller Div. II Stby. S. of ACU of El. 70' Rad. Monit. III (NW) IV (SW) (NW Corner)East Side West Side Swgr. Rm.Electrical PowerDiv. I X X X XDiv. II X X X X XDiv. III X X XOffsiteRPV Level ControlHPCS (Div. III) X X XRCIC (Div. I) X X XLPCS (Div. I) X X X XLPCI-RHR C (Div. II) X X X X XRPV Pressure ControlDiv. I SRVs X X X XDiv. II SRVs X X X X XDecay Heat RemovalRHR A Suppression Pool Cooling X X X XRHR A Alternate Shutdown Cooling X X X XRHR A Normal Shutdown CoolingRHR B Suppression Pool Cooling X X X X XRHR B Alternate Shutdown Cooling X X X X XRHR B Normal Shutdown CoolingMech./Environ. SupportDiv. I Standby Service Water 1 pump 1 pump X 1 pumpDiv. I HVAC Systems X X X XDiv. I Containment/RPV Monitoring X X X XDiv. II Standby Service Water X X X X XDiv. II HVAC Systems X X X X XDiv. II Containment/RPV Monitoring X X X X XNormal Service WaterEquipment availability dependsupon fire location in Fire AreaLocal Manual Action Required?Y Y N N N N N N Y CONTINUOUS USE ATTACHMENT 1PAGE 4 OF 8AVAILABLE SAFE SHUTDOWN SYSTEMSAOP-0052 REV - 025 PAGE 12 OF 62Fire Area:C-15 C-16 C-17 C-18 C-19C-21 C-22 C-24 C-27 Div. I Stby. Remote Control Bldg.ENB A Batt. ENB B Batt. HPCS Batt. HPCS Remainder East Pipe Swgr. Rm. S/D Room Vent. Room & Inv/Chgr. & Inv/Chgr. & Chgr. Rms.Swgr. Rm. of El. 116' ChaseElectrical PowerDiv. I X X X XDiv. II X X X X X XDiv. III X X X XOffsiteRPV Level ControlHPCS (Div. III) X X XRCIC (Div. I) X X X XLPCS (Div. I) X X XLPCI-RHR C (Div. II) X X X X X XRPV Pressure ControlDiv. I SRVs X X X XDiv. II SRVs X X X X X XDecay Heat RemovalRHR A Suppression Pool Cooling X X X XRHR A Alternate Shutdown Cooling X X XRHR A Normal Shutdown Cooling XRHR B Suppression Pool Cooling X X X X X XRHR B Alternate Shutdown Cooling X X X X X XRHR B Normal Shutdown CoolingMech./Environ. SupportDiv. I Standby Service Water X X 1 pump XDiv. I HVAC Systems X X X XDiv. I Containment/RPV Monitoring X X X XDiv. II Standby Service Water X X X X X XDiv. II HVAC Systems X X X X X XDiv. II Containment/RPV Monitoring X X X X X XNormal Service WaterEquipment availability dependsupon fire location in Fire AreaLocal Manual Action Required?N N Y N N Y N Y N CONTINUOUS USE ATTACHMENT 1PAGE 5 OF 8AVAILABLE SAFE SHUTDOWN SYSTEMSAOP-0052 REV - 025 PAGE 13 OF 62Fire Area:C-29 C-30 DG-1 DG-2 DG-3DG-4 DG-5 DG-6 DG-7 NW Stairwell Div. II Fuel Div. III Fuel Div. I Fuel Div. II DieselDiv. III DieselDiv. I Diesel Div.II Diesel Vestibule No. 1 Storage Tk. Storage Tk. Storage Tk. Gen. Room Gen. Room Gen. Room Elec. TunnelElectrical PowerDiv. I X X X X X XDiv. II X X X X X XDiv. III X X X X X X XOffsiteRPV Level ControlHPCS (Div. III) X X X X X X XRCIC (Div. I) X X X X X XLPCS (Div. I) X X X X X XLPCI-RHR C (Div. II) X X X X X XRPV Pressure ControlDiv. I SRVs X X X X X XDiv. II SRVs X X X X X XDecay Heat RemovalRHR A Suppression Pool Cooling X X X X X XRHR A Alternate Shutdown Cooling X X X X X XRHR A Normal Shutdown CoolingRHR B Suppression Pool Cooling X X X X X XRHR B Alternate Shutdown Cooling X X X X X XRHR B Normal Shutdown CoolingMech./Environ. SupportDiv. I Standby Service Water X X X 1 pump X XDiv. I HVAC Systems X X X X X XDiv. I Containment/RPV Monitoring X X X X X XDiv. II Standby Service Water X X X X X XDiv. II HVAC Systems X X X X X XDiv. II Containment/RPV Monitoring X X X X X XNormal Service WaterEquipment availability dependsupon fire location in Fire AreaLocal Manual Action Required?N N N N N N N N N CONTINUOUS USE ATTACHMENT 1PAGE 6 OF 8AVAILABLE SAFE SHUTDOWN SYSTEMSAOP-0052 REV - 025 PAGE 14 OF 62Fire Area:ET-1 ET-2 ET-3 ET-4 ET-5 ET-6 FB-1 MG-1 B-Tunnel B-Tunnel T-Tunnel T-Tunnel B-Tunnel C-Tunnel Fuel LFMG East (DG) West (FB) West El.67'6"West El.95' South Building BuildingElectrical PowerDiv. I X X X X X X XDiv. II X X X X X X XDiv. III X X X X X X XOffsiteRPV Level ControlHPCS (Div. III) X X X X X XRCIC (Div. I) X X X X X XLPCS (Div. I) X X X X X X XLPCI-RHR C (Div. II) X X X X X X XRPV Pressure ControlDiv. I SRVs X X X X X X XDiv. II SRVs X X X X X X XDecay Heat RemovalRHR A Suppression Pool Cooling X X X X X X XRHR A Alternate Shutdown Cooling X X X X X X XRHR A Normal Shutdown CoolingRHR B Suppression Pool Cooling X X X X X X XRHR B Alternate Shutdown Cooling X X X X X X XRHR B Normal Shutdown CoolingMech./Environ. SupportDiv. I Standby Service Water X X X X X X XDiv. I HVAC Systems X X X X X X XDiv. I Containment/RPV Monitoring X X X X X X XDiv. II Standby Service Water X X X X X X XDiv. II HVAC Systems X X X X X X XDiv. II Containment/RPV Monitoring X X X X X X XNormal Service WaterEquipment availability dependsupon fire location in Fire AreaLocal Manual Action Required?N Y N N N N Y N CONTINUOUS USE ATTACHMENT 1PAGE 7 OF 8AVAILABLE SAFE SHUTDOWN SYSTEMSAOP-0052 REV - 025 PAGE 15 OF 62Fire Area:PH-1 PH-2 PH-3 PH-4 PH-5 PT-1 PT-2 Stby. CoolingStby. CoolingRemote Air Div. II SCT Div. I SCT E, F & E-Tunnel Twr. (Div. I) Twr. (Div. II) Intake RoomFans Fans G-Tunnels (SWP/CCP)Electrical PowerDiv. I X X X X X XDiv. II X X X X XDiv. III X X X X X X XOffsite XRPV Level ControlHPCS (Div. III) X X X X X X XRCIC (Div. I) X X X X X XLPCS (Div. I) X X X X X XLPCI-RHR C (Div. II) X X X X XRPV Pressure ControlDiv. I SRVs X X X X X XDiv. II SRVs X X X X XDecay Heat RemovalRHR A Suppression Pool Cooling X X X X X XRHR A Alternate Shutdown Cooling X X X X X XRHR A Normal Shutdown CoolingRHR B Suppression Pool Cooling X X X X XRHR B Alternate Shutdown Cooling X X X X XRHR B Normal Shutdown CoolingMech./Environ. SupportDiv. I Standby Service Water X X X XDiv. I HVAC Systems X X X X X XDiv. I Containment/RPV Monitoring X X X X X XDiv. II Standby Service Water X X X XDiv. II HVAC Systems X X X X XDiv. II Containment/RPV Monitoring X X X X XNormal Service Water XEquipment availability dependsupon fire location in Fire AreaLocal Manual Action Required?N N N N N N N CONTINUOUS USE ATTACHMENT 1PAGE 8 OF 8AVAILABLE SAFE SHUTDOWN SYSTEMSAOP-0052 REV - 025 PAGE 16 OF 62Fire Area:RC-2 RC-3 RC-4 RC-6 RDW-1 MS Tunnel, NReactor BldgReactor BldgAnnulus Area Drywell of Imp. Wall East All ElevWest All ElevEast West East WestElectrical PowerDiv. I X X X XDiv. II X X X XDiv. III X X X X XOffsiteRPV Level ControlHPCS (Div. III) X X X X X XRCIC (Div. I)LPCS (Div. I) X X X X X XLPCI-RHR C (Div. II) X X X XRPV Pressure ControlDiv. I SRVs X X X XDiv. II SRVs X X X XDecay Heat RemovalRHR A Suppression Pool Cooling X X X XRHR A Alternate Shutdown Cooling X X X XRHR A Normal Shutdown CoolingRHR B Suppression Pool Cooling X X X XRHR B Alternate Shutdown Cooling X X X XRHR B Normal Shutdown CoolingMech./Environ. SupportDiv. I Standby Service Water X X X XDiv. I HVAC Systems X X X XDiv. I Containment/RPV Monitoring X X X XDiv. II Standby Service Water X X X XDiv. II HVAC Systems X X X XDiv. II Containment/RPV Monitoring X X X XNormal Service WaterEquipment availability dependsupon fire location in Fire AreaLocal Manual Action Required?N N N N N N N Chemistry TR 3.4.13RIVER BEND TR 3.4-13 Revision 5 (33i)TR 3.4.13 CHEMISTRYTLCO 3.4.13 The chemistry of the reactor coolant system shall be maintained within the limits specified in Table 3.4.13-1.APPLICABILITY: At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIMEA. In MODE 1, with the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.4.13-1.A.1 Restore to within limits.72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sB. Required Action A.1 and associated Completion Time not met. OR Conductivity or chloride concentration exceeds the limit specified in Table 3.4.13-1 while in MODE 1 for > 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> in any 365 day period.B.1 Be in Mode 2.
6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sC. In MODE 2 and 3 with the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.4.13-1.C.1 Restore to within limits.48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (continued)
Chemistry TR 3.4.13RIVER BEND TR 3.4-14 Revision 5 (33ii)ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIMED. Required Action C.1 and associated Completion Time not met.
OR The identification while in MODE 1, that conductivity exceeds 10
µmho/cm at 25°C or chloride concentration exceeds 0.5 ppmD.1 Be in Mode 3.
ANDD.2 Be in Mode 4.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hoursE. At all times other than MODE 1, 2 or 3, with the conductivity or pH exceeding the limit specified in Table 3.4.13-1.E.1 Restore the conductivity and pH to within the limit.
72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sF. At all times other than MODE 1, 2 or 3, with the chloride concentration exceeding the limit specified in Table 3.4.13-1.F.1 Restore chloride concentration to within limit.
24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sG. Required Action F.1 and associated Completion Time not met.G.1 Perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system.
Prior to exceeding 200 F RCS temperature.
Chemistry TR 3.4.13RIVER BEND TR 3.4-15 Revision 39 (33iii)SURVEILLANCE REQUIREMENTS
NOTE------------------------------------
The reactor coolant shall be determined to be within the specified chemistry limit by performance of the following:
SURVEILLANCE FREQUENCYTSR 3.4.13.1 Determine reactor coolant to be within the specified chemistry limit by analyzing a sample of the reactor coolant for chlorides.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND--------NOTE------
When conductivity is greater than the limit in Table 3.4.13-1.
8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sTSR 3.4.13.2 Determine reactor coolant to be within the specified chemistry limit by analyzing a sample of the reactor coolant for conductivity.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.TSR 3.4.13.3 ----------------NOTE--------------------- Not required to be met when conductivity is 1.0 mhos/cm @25°C. ----------------------------------------- Determine reactor coolant to be within the specified chemistry limit by analyzing a sample of the reactor coolant for pH.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND--------NOTE------
When conductivity is greater than the limit in Table 3.4.13-1.
8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sTSR 3.4.13.4 -----------------NOTE-------------------- Not required to be met when obtaining in-line conductivity measurements per TSR 3.4.13.5 ----------------------------------------- Record the conductivity of the reactor coolant.Continuously Continued Chemistry TR 3.4.13RIVER BEND TR 3.4-16 Revision 5 (33iv)SURVEILLANCE FREQUENCYTSR 3.4.13.5 ------------------NOTE------------------- Not required to be met when the continuous recording conductivity monitor is operable. ----------------------------------------- Obtain an in-line conductivity measurement.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in MODE 1, 2 or 3 AND 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sTSR 3.4.13.6 --------------------NOTE----------------- Not required to be met when obtaining in-line conductivity measurements per TSR 3.4.13.5. ----------------------------------------- Perform a CHANNEL CHECK of the continuous conductivity monitor with an in-line flow cell.7 days AND------NOTE-------
When conductivity is greater than the limit in Table 3.4.13-1.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> TABLE 3.4.13-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS CONDUCTIVITY MODE CHLORIDES (mhos/cm @25°C) pH 1 0.2 ppm 1.0 5.6 pH 8.6 2 and 3 0.1 ppm 2.0 5.6 pH 8.6 At all other times 0.5 ppm 10.0 5.3 pH 8.6