ML090120832
| ML090120832 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/12/2008 |
| From: | FirstEnergy Nuclear Operating Co |
| To: | David Silk Operations Branch I |
| Hansell S | |
| Shared Package | |
| ML081060562 | List: |
| References | |
| TAC U01628 | |
| Download: ML090120832 (9) | |
Text
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
1 The plant is operating at 100% power with all systems in NSA EXCEPT:
Power Range Channel N44 has been declared inoperable.
Power Range Channel N44 has been removed from service IAW AOP-2.2.1 C, Power Ranqe Channel Malfunction.
Power Range Channel N43 NOW fails HIGH.
All systems function as designed.
No Operator Actions have been taken Which, of the below listed First Out Annunciators (ANN. A5), will alarm in the FIRST 45 seconds AFTER N43 fails High?
(1 ) A5-1 D 2/3 Loops Overtemp AT Reactor Trip (2) A5-2A Reactor Protection System Train A Trouble (3) A5-5G Reactor Trip Due To Turbine Trip (4) A5-6B Turbine Anti-Motoring Turbine Trip (5) A5-6D Turbine Trip Due To Reactor Trip (6) A5-7D Generator Trip Due To Turbine Trip A.
1,3,5&6ONLY B.
2, 4, & 6 ONLY C.
3,5&6ONLY 1,2,3 & 4 ONLY Answer C
ExplanationlJustification:
A.
- 6.
Incorrect. N-44 does NOT input into OTAT trip setpoint calculation, therefore this alarm will NOT be energized.
Incorrect. Candidate may confuse rod control urgent alarm with protection system trouble. Rod control urgent will energize on the trip. Anti-motoring would alarm if the output breakers did not open. However, stem of the question states that all systems functioned as designed.
3 and 5 will both be alarmed.
Correct. IAW 20M-1.4.ABB. 1.4.AA1, 1.4.AAD 26.4.AAF. 26.4.AAI and 35.4.AAF Incorrect N-44 does NOT input into OTAT trip setpoint calculation, therefore this alarm will NOT be energized. Candidate may confuse rod.ontrol urgent alarm with protection system trouble. Rod control urgent will energize on the trip. Anti-motoring would alarm if the output breakers dic not open. However, stem of the question states that all systems functioned as designed. 5 and 6 will both be alarmed.
C.
D.
K/A Sys #
K/A System K/A Category WA Statement 000007 Reactor Tria Knowledqe of the interrelations between a reactor trip and Reactor trip status panel Question Source:
New References provided to Candidate None Objective #:
Task ID#:
the following:
KIA #
EK2.03 KIA Importance 3.5 Exam Level RO Level Of Difficulty: (1-5)
Question Cognitive Level:
Higher Comprehension Technical
References:
10 CFR Part 55 Content:
1.4.AAD. 26.4.AAI and 35.4.AAF (CFR 41.7 145.7)
Page 1 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 2.
The plant is operating at 100% power with all systems in NSA.
e A PRZR vapor space accident occurs.
PRZR pressure drops to 1200 psig.
The Highest Steam Generator pressure is 1000 psig.
HHSl flow is 800 gpm and stable.
All systems functioned as designed.
NO Orange or Red path conditions exist.
The crew is performing the actions of E-I, Loss of Reactor or Secondary Coolant.
At Step 2, Check if RCPs should be stopped, the crew is directed to Stop ALL RCPs.
WHY MUST the RCPs be stopped at this time?
The RCPs are tripped to:
A.
prevent possible pump damage by running the RCPs under highly voided conditions in order to save the pumps for potential future use.
B.
prevent excessive depletion of RCS water inventory which might lead to severe core uncovery if the RCPs were tripped later in the event.
C.
remove their added heat input, thereby ensuring the steam generators will be capable of performing the subsequent RCS cooldown.
D.
ensure the RCP seal package is not damaged by the excessive temperature or steam voiding associated with this event.
Answer B
planation/Justification:
A.
- 6.
C.
Incorrect This is the reason they are stopped in FR-C.2 Correct. IAW with E-1 step 2 basis and RCP trip generic issue Incorrect. Plausible since RCPs are tripped in FR-H.l to remove their heat input, but it is done to extend the effectiveness of the remaining inventory. This is also plausible since tripping the pumps will remove the heat input and removing the heat input will ensure the only heat tha! will be required to be removed is from decay heat alone BUT this is not the reason why they are tripped in E-I.
Incorrect. This is the consequence of losing both seal injection and RCP thermal barrier cooling.
D.
KIA Sys #
KIA System KIA Category KIA Statement 000008 KIA #
AK3.04 KIA Importance 4.2 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
E-I step 2 bases; 20M-53B.5.GI-6 page 6 Objective #:
Task ID#:
10 CFRPart55Content:
(CFR41.5,41.10/45.6/45.13 Pressurizer Vapor Space Accident Knowledge of the reasons for the following responses as they apply to the Pressurizer Vapor Space Accident:
RCP tripping requirements 2"d paragraph Page 2 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 3.
The plant is operating at 100% power with all systems in NSA.
A small break LOCA occurs coincident with a loss of offsite power.
All systems function as designed EXCEPT EDG #2 fails to start and CANNOT be started.
10 minutes after the event began; the crew is performing recovery actions IAW ES-1.2, Po ;t LOCA Cooldow n and Depressurization.
IAW ES-1.2 step 1 Reset SI, the RO depresses the Safety Injection Signal Train A AND TI ain B reset pushbuttons.
AFTER the Safety Injection Signal Train A AND Train B reset pushbuttons have been depressed, 1 i e following plant conditions exist:
PRZR pressure is 1350 psig and slowly rising.
RCS Subcooling is 95°F and stable.
4KV bus 2DF is de-energized.
Annunciator A12-1 C Auto Safety Injection Blocked is flashing (white then dark).
Annunciator A I 2-1 D Safety Injection Signal is flashing (white then dark).
Based on these conditions:
What is the current status of the automatic Safety Injection Actuation system AND what is the significance of annunciators A12-1 C and AI 2-1 D flashing?
A.
ONLY one Train of Safety Injection has reset; the flashing annunciators indicate a status difference between the two trains of automatic Safety Injection actuation.
B.
ONLY one Train of Safety Injection has reset; the flashing annunciators indicate the Purple Train od' electrical power will not respond to an automatic Safety Injection actuation signal.
C.
BOTH Trains of Safety Injection have reset; the flashing annunciators indicate pressurizer pressure is still below the low pressure automatic safety injection setpoint.
D.
BOTH Trains of Safety Injection have reset; the flashing annunciators indicate automatic safety injtxtion actuation will not occur until the reactor trip breakers are re-closed.
Answer A
ExplanationlJustification:
A.
B.
C.
D.
K/A Sys #
WA System K/A Category WA Statement 000009 Small Break LOCA Ability to operate and monitor the following as they apply ESFAS to a small break LOCA:
K/A #
EA1.13 KIA Importance 4.4 Exam Level RO Level Of Difficulty: (1-5)
Question Cognitive Level:
Higher Comprehension Question Source:
New References provided to Candidate None Technical
References:
ES-1.2 step 1 background page 7 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR 41.7 / 45.5 / 45.6)
Correct. IAW ES-1.2 step 1 background document page 7 Incorrect. Right status of SI actuation system; Wrong significance of flashing alarms, inoperable electrical trains are indicated by the BlSl : ystem NOT the flashing of annunciators A12-IC and ID.
Incorrect. Only one train of SI has reset. SI will reset even though an SI signal is still present due to the retentive memory circuit and the P 4 contact development.
Incorrect. Only one train of SI has reset. Closing the reactor trip breakers and re-arming automatic SI is indicted when A12-IC goes DARI.
Page 3 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 4.
The plant is operating at 100% power with all systems in NSA.
Annunciator A2-5C, Reactor Coolant Pump Vibration AledDanger Alarms.
B RCP shaft vibration is 16 mils and stable B RCP frame vibration is I mil and stable The crew enters AOP-2.6.8, Abnormal RCP Operation.
While performing the actions of AOP-2.6.8, Abnormal RCP Operation, the following additional alarrx and indications are received:
A2-5D, Reactor Coolant Pump Seal Vent Pot Level High/Low (RCP 21 B Seal Pot LL]
High, computer address point L0508D)
A2-4D, Reactor Coolant Pump Seal Trouble (RCP 21 B Seal Lk Off CHS-FT155B LON, computer address point F0128D)
RCP 218 Seal Lk Off CHS-FTl55B is.80 gpm and stable Based on these alarms and indications, which B RCP seal has failed?
A.
- I seal B.
- 2seal C.
- 3seal D.
Low pressure seal Answer B
ExplanationlJustification:
A.
Incorrect. If #I seal had failed seal leak-off flow would be high NOT low.
Correct. IAW 20M-7.4.AAH, 6.4.AAE and AOP-2.6.8 Incorrect. If #3 seal had failed the seal vent pot level low would be indicated NOT high.
Incorrect. The low pressure seal is not functional when the motor is coupled to the pump.
- 0.
WA Sys #
WA System KIA Category KIA Statement 000015/17 RCP Malfunctions Knowledge of the interrelations between the Reactor RCP seals WA #
AK2.07 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate None Technical
References:
20M-7.4.AAH, 6.4.AAE Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR 41.7 145.7)
Coolant Pump Malfunctions (Loss of RC Flow) and the following:
Page 4 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 5.
The plant is in Mode 6.
Preparations to flood the refueling cavity are underway.
RCS water level is ONE (1) foot below the top of the reactor vessel flange and stable.
All RCS loop isolation valves are CLOSED.
RCS temperature is 100°F and stable.
RCS is vented to atmosphere.
It has been 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> since the reactor was shutdown.
RHR Pump 2RHS*P21A is operating and RHR Pump 2RHS*P21 B is in Standby.
RHR Pump 2RHS*P21A TRIPS and RHR Pump 2RHS*P21 B WILL NOT start.
The crew enters AOP-2.10.1, Residual Heat Removal System Loss.
At step 11 of AOP-2.10.1, Residual Heat Removal System Loss, the crew is directed to estimate th?
time to RCS saturation.
Using the attached AOP-2. IO. 1 figures and attachments, ESTIMATE the time to RCS saturation.
The estimated time to RCS saturation is A.
4.4 minutes B.
16 minutes C.
25 minutes D.
37.5 minutes nswer C
ExplanationlJustification:
A.
B.
C.
D.
KIA Sys #
KIA System KIA Category KIA Statement 000025 Loss of RHR System NIA Knowledge of abnormal condition proce -lures.
KIA #
2.4.11 WA Importance 4.0 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Application References provided to Candidate Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.10 /43.5/45.13)
Incorrect. This is the heatup RATE, NOT the time to saturation. This rate is necessary to calculate the time to saturation and if a candidate !;tops before completing the calculation, they will choose this answer.
Incorrect. This is the number for 140°F starting temperature.
Correct. IAW figure I C and attachment I Incorrect. If candidate uses figure 1B instead of figure I C they will calculate this value.
AOP-2.10.1 figures IA, IB, IC, 2A, 28, 2C, 3, & att. 1 Technical
References:
AOP-2.10.1 figure I C and Attachment Page 5 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 6.
The plant is operating at 100% power with all systems in NSA.
Primary Component Cooling Water Pump 2CCP*P21 C is on clearance and unavailable.
Primary Component Cooling Water Pump 2CCP*P21A is running.
Primary Component Cooling Water Pump 2CCP*P21 B is in Standby.
2CCPP21A TRIPS and cannot be re-started.
2CCP*P21 B FAILS to automatically start and cannot be manually started.
IAW the guidance provided in AOP-2.15.1, Loss of Primary Component Cooling Water, what sequence of actions is now REQUIRED?
A.
Trip the reactor, Trip the RCPs, complete the immediate operator actions of E-0, THEN isolate letdl iwn B.
Trip the reactor, complete the immediate operator actions of E-0, Trip the RCPs, THEN isolate letdljwn.
C.
Trip the RCPs, isolate letdown, Trip the reactor THEN complete the immediate operator actions of 5 0.
D.
Trip the RCPs, Trip the reactor, isolate letdown, THEN complete the immediate operator actions of E-0.
Answer B
ExplanationIJustification:
A.
E.
C.
- 0.
Incorrect. Plausible since all the actions are correct but the sequence is incorrect.
Correct. IAW AOP-2.15.1 step 2 RNO. AT BVPS the topic of AOP use in conjunction with EOP use has been addressed by providing the xtions to be completed within the AOP. This is done by providing WHEN statements within the AOP. (2.6.8, 2.15.1, 2.6.7)
Incorrect. Plausible since all the actions are correct but the sequence is incorrect.
Incorrect. Plausible since all the actions are correct but the sequence is incorrect.
K
/
i WA Category WA Statement 000026 Loss of Component NIA Cooling Water KIA #
2.4.8 WA Importance 3.8 Exam Level RO Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Knowledge of how abnormal operating procedures are used in conjunction witt; EOPs.
Level Of Difficulty: (1-5)
Question Cognitive Level:
Lower Fundamental Technical
References:
10 CFR Part 55 Content:
AOP-2.15.1 step 2 RNO (CFR: 41. I O / 43.5 I 45.13)
Page 6 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 - 1 7.
The plant is operating at 100% power with all systems in NSA.
A valid reactor trip signal is received HOWEVER the reactor does not Automatically trip and it CANNOT be tripped from the control room.
The crew enters FR-S.1, Response to Nuclear Power Generation-ATWS.
VCT level is 40% and stable While attempting to initiate emergency boration, Emergency Boration lsol Vlv 2CHS*MOV350 CANNOT be opened.
IAW the guidance provided in FR-S.l, what are the MINIMUM control switchhalve positions REQUIRED to ensure boration flow to the RCS.
(Assume the Boric Acid Flow To Blender Flow Totalizer is greater than I000 gallons)
(1) 2CHS*FCV113B Boric Acid Blender Disch To Chg Pumps - OPEN (Red light ONLY)
(2) 2CHS*FCV113B Boric Acid Blender Disch To Chg Pumps - CLOSED (Green light ONLY)
(3) 2CHS*FCV113A Boric Acid To Boric Acid Blender - OPEN (Red light ONLY)
(4) 2CHS*FCV113A Boric Acid To Boric Acid Blender - CLOSED (Green light ONLY)
(5) 2CHS*SOV206 Alt Emergency Boration Vlv - OPEN (Red light ONLY)
(6) Boric Acid Makeup Blender Control - Red light ONLY (7) Boric Acid Makeup Blender Control - Green light ONLY (8) Blender Mode Selector Switch - AUTO (9) Blender Mode Selector Switch - BORATE A.
1, 4, 5, 6 & 9 1, 4, 5, 7 & 8 C.
2, 3, 5, 6 & 9 D.
Answer C
2, 3, 5,7 & 8 ExplanationlJustification:
A.
B.
Incorrect. In order to get flow 113A MUST be opened. Plausible if the candidate has a misconception that SOV206 bypasses 113A instead tjf bypassing 1138.
Incorrect. In order to get flow 113A MUST be opened and a demand signal must be present neither are met in this choice. Plausible if the candidate does not realize that SOV206 requires the blender controls to initiate flow. Since MOV350 does not require the blender, a candid.ite may believe that SOV 206 functions in the same manner.
Correct. IAW FR-S.1 and VONDs showing flowpaths. The alternate boration valve 2CHS*SOV206 will ONLY provide boric acid flow when i.11 of the following are present: valve open, FCVI 13A open, mode selector in borate, flow totalizer signal present..
Incorrect. No demand signal is present. Plausible if the candidate believes a demand signal will be present with the selector switch in auto, however the VCT level must be low to get the demand signal. Also plausible if the candidate does not realize that SOV206 requires the blet der controls to initiate flow. Since MOV350 does not require the blender, a candidate may believe that SOV 206 functions in the same manner.
C.
D.
KIA Sys #
WA System WA Category WA Statement 000029 ATWS WA #
EA2.05 WA Importance 3.4 Exam Level RO Level Of Difficulty: (1-5)
Ability to determine or interpret the following as they apply to a ATWS:
System component valve position indic; tion Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Higher Comprehension Technical
References:
10 CFR Part 55 Content:
FR-S.1 step 3 RNO (CFR 43.5 /45.13)
Page 7 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 8.
The plant is operating at 100% power with all systems in NSA.
0 A Steam Generator Tube Rupture occurs.
The crew enters the EOP network.
The crew is currently implementing E-3, Steam Generator Tube Rupture.
The RCS has been cooled to 500°F in preparation for equalizing RCS pressure with the ruptured SG pressure.
The Unit Supervisor directs you to depressurize the RCS AND while maintaining a minimum of 20°F of Su bcooling.
At the current RCS temperature, what is the lowest RCS pressure can be without violating the 20°F of Subcooling requirement?
A.
-666 psig
- 6.
-695 psig C.
-798 psig D.
-827 psig Answer C
ExplanationIJustification:
A.
B.
C.
D.
Incorrect. Plausible This would be the saturation pressure for 500°F. (680.86 psia - 14.7psi)
Incorrect. Plausible, if candidate attempts to determine pressure for 500°F and mistakenly adds 14.7 psi to 680.86 psia Correct. Saturation pressure for 520°F is 812.53 minus 14.7 psi yields 797.83.
Incorrect. Plausible if candidate mistakenly adds 14.7 psi to the saturation pressure.
,A Sys #
KIA System KIA Category KIA Statement 000038 Steam Gen. Tube Rupture Knowledge of the operational implications of the following concepts as they apply to the SGTR:
Use of steam tables KIA #
EK1.O1 KIA Importance 3.1 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate Steam tables Technical
References:
E-3, Steam tables Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR 41.8 / 41. I O / 45.3)
Page 8 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 9.
The plant is operating at 100% power with all systems in NSA.
A Steamline break outside containment occurs.
The MSlVs fail to close and they cannot be manually closed.
All other systems functioned as designed.
All 3 SGs depressurize to atmospheric pressure.
All 3 SG NR levels are 10% and slowly dropping.
All RCS cold leg temperatures stabilize at 220°F It has been 30 minutes since the steam break occurred.
Indicated PRZR level is 0%.
Indicate AFW flow is 0 gpm and stable.
. RCS Subcooling is 200°F.
The operating crew has entered FR-P.l, Response to Imminent Pressurized Thermal Shock Condilron due to the excessive cooldown rate and all RCS cold leg temperatures being below the Reference Transition Nil Ductility Temperature ( R T N ~ ~ )
of 245°F.
Which One (1) of the below listed actions will limit the overall stress on the Reactor Vessel?
A.
Depressurize the RCS B.
Restore AFW flow to 7340 gpm C.
Maximize safety injection flow D.
Stop all running RCPs Inswer A
cxplanationlJustification:
A.
B.
C.
Correct. IAW FR-P.l bases page 4 one of the major actions to limit the RPV stress is to depressurize the RCS.
Incorrect. AFW Flow shall only be raised to 50 gpm NOT 340 gpm 50 gpm to each SG ensures they remain wetted and limits stress on the i G tubes but NOT stresses on the vessel. ANY increase in AFW flow will increase vessel stresses.
Incorrect. IAW FR-P.l bases page 4 one of the major actions is to terminate SI when the criteria are met. Terminating is done to minimize We cold water effects on the vessel downcomer region. Large break. LOCA strategies include maximizing SI flow. The situation presented by exceecing RTNDTis an exception to Large break LOCA strategies. With PRZR level at 0% a candidate may believe it necessary to maximize SI flow un.'l level is restored.
Incorrect Stopping RCPs would potentially increase overall vessel stress by allowing the cold SI water contact the vessel downcomer regior without any mixing. Therefore, RCPs are left running in FR-P.l until support conditions are no longer available, and then they are secured.
D.
.~
KIA Sys #
KIA System WA Category KIA Statement 000040 Steam Line Rupture Knowledge of the operational implications of the following Nil ductility temperature WA #
AK1.04 WA importance 3.2 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
FR-P.l bases page 4 Objective #:
concepts as they apply to Steam Line Rupture:
Task ID#:
10 CFR Part 55 Content:
(CFR 41.8 / 41. I O / 45.3)
Page 9 of 100
IO.
A.
- 6.
C.
D.
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 The plant is operating at 100% power with all systems in NSA.
An inadvertent Reactor trip occurs WITH a coincidental loss of all 4KV AC power.
All other systems operate as designed.
Twenty minutes after the trip, which ONE (1) of the following sets of parameters indicate that natur: I circulation of the RCS has been established?
SG Pressures Core Exit TCs 590 O F and rising 1060 psig and rising 1060 psig and stable 577 O F and stable 1035 psig and dropping 590 OF and rising 1035 psig and dropping 577 O F and stable Tcold 558 OF and droppin!
558 OF and stable 550 O F and stable 550 OF and droppins..
Answer D
ExplanationlJustification:
A.
B.
C.
Incorrect. CETs rising.
D.
Incorrect. SG Press and CETs are rising and Tcold above Tsat of SG. (Tsat for 1060 psig = 553F)
Incorrect. Tcold above Tsat of SG. (Tsat for 1060 psig = 553F)
Correct. All parameters stable or dropping and Tcold at Tsat of SG.
NOTE:
See SRO question #80 explanation as to why this question has been evaluated to be different enough from SRO question #80 to be used on the same exam.
WA Sys #
WA System K/A Category K/A Statement 10055 Station Blackout Ability to determine or interpret the following as they apply to a Station Blackout:
RCS core cooling through natural circu ltion cooling to SIG cooling WA #
EA2.02 WA Importance 4.4 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate Steam Tables Technical
References:
Steam Tables; EOP Attachment A-I.7 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR 43.5 145.13)
Page 10 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 11.
The plant is operating at 100% power with all systems in NSA.
An inadvertent Reactor trip occurs WITH a coincidental loss of offsite power.
All other systems operate as designed.
Both Emergency 4KV Buses are being powered by their respective diesel generators.
The crew performs the actions of E-0, Reactor Trip or Safety Injection, and transitions into 33-0.2, Natural Circulation Cooldown.
IAW ES-0.2, Natural Circulation Cooldown, what is the MINIMUM required steam generator water Lwei that must be maintained to provide a stable heat sink during the natural circulation cooldown?
A.
WR level of at least 14%
B.
NR level of at least 12%
C.
NR level of least 35%
D.
NR level of at least 50%
Answer C
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. This is the minimum water level for loss of heat sink in FR-H.1.
Incorrect. This is the minimum water level for maintaining the thermal blanket during SGTR recovery.
Correct. IAW step 5 of ES-0.2 and step 5 bases.
Incorrect. This is the Maximum water level for natural circulation cooldown in ES-0.2.
WA Sys #
WA System WA Category WA Statement 000056 Loss of Off-site Power Ability to determine and interpret the following as they apply to the Loss of Offsite Power:
Necessary S/G water level for natural circulation JA #
AA2.88 WA Importance 4.1 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Objective #:
Task ID#:
Technical
References:
10 CFR Part 55 Content:
ES-0.2 step 5 and step 5 bases.
(CFR: 43.5 145.13)
Page 11 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 12.
A.
B.
C.
D.
The plant is operating at 100% power with all systems in NSA.
A loss of Vital Bus 2 has occurred as a result of a failure in the inverter.
The static switch has FAILED to automatically transfer to the backup power supply (MCC2 E06).
The Unit Supervisor has directed you to restore power to Vital Bus 2 using the Manual Bypass Swit :h. In order to accomplish this manual transfer, the Manual Bypass Switch must be placed in the position?
Byp to Alt Line Alternate Source To Load Bypass (Standby)
Bypass (Isolate)
Answer B
Explanation/Justification:
A.
Incorrect. This is a Unit 2 Essential Bus switch NOT a vital bus switch position. Plausible since the switch name does exist at Unit 2 and b; ;ed on the nomenclature of the switch it could be mistaken for the correct switch position. Byp to alt line implies that it will somehow align the alter-iate source.
Correct. IAW AOP-2.38.1B step 6e RNO.
Incorrect. This is the correct manual transfer switch position for Unit 1 NOT Unit 2. Plausible distractor for a Unit 2 exam since auxiliary opr,,ators are trained for both Units, and Unit 2 RO candidates are selected from either Unit.
Incorrect. This is the incorrect manual transfer switch position for Unit 1. Plausible distractor for a Unit 2 exam since auxiliary operators are trained for both Units, and Unit 2 RO candidates are selected from either Unit..
B.
C.
D.
KIA Sys #
KIA System KIA Category KIA Statement 000057 Loss of Vital AC Inst. Bus Ability to operate and I or monitor the following as they Manual inverter swapping apply to the Loss of Vital AC Instrument Bus:
A #
AA1.01 KIA Importance 3.7 Exam Level RO Level Of Difficulty: (1-5) duestion Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Lower Fundamental Technical
References:
10 CFR Part 55 Content:
AOP-2.38.1 B step 6e RNO (CFR 41.7 145.5 I45.6)
Page 12 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 13.
The plant is operating at 5% power with all systems in NSA.
A loss of 125VDC Bus 2-1 has occurred.
Step 2 of AOP-2.39.1A, Loss Of 125VDC Bus 2-1 instructs the operating crew to control RCS temperature and pressure using the Steam Generator Atmospheric Steam Dump Control valves
[2SVS*PCVIOl A(B)(C)] OR the Residual Heat Release Valve [2SVS*HCV104].
Under these conditions, WHY are THESE valves used to control RCS temperature?
Because:
A.
The condenser will NOT be available due to loss of all cooling tower pumps.
- 6. The condenser will NOT be available due to closure of all steam generator MSIVs.
C.
Rod control will NOT be available due to an URGENT failure alarm.
D.
Rod control will NOT be available due to a NON-URGENT failure alarm.
Answer B
ExplanationIJustification:
A.
B.
C.
Incorrect. Loss of DC control power to 4KV breakers will not cause the breaker to trip; it will render the automatic trip circuit inoperable.
Correct. IAW AOP 2.39.1A step 2 and Automatic actions listed on page 1.
Incorrect. Urgent failure alarms will block all automatic and manual rod motion. However, rod control power is not powered by this DC bus.
Plausible since at this low power level, temperature is being controlled by rod motion and NO direct reactor trip will occur as a result of this I C bus failure. However step 1 of the AOP will require the manually tripping of the reactor which makes rods control unavailable for temperaturo control. Therefore the steam dump valves must be used.
Incorrect. Non-Urgent failure alarms will NOT block rod motion. However, Non-urgent alarms are generated from a loss of any 24VDC pows.-r but the DC power is not provided by this DC bus. Plausible since at this low power level, temperature is being controlled by rod motion and NO direct reactor trip will occur as a result of this DC bus failure. However step 1 of the AOP will require the manually tripping of the reactor which m::kes rods control unavailable for temperature control. Therefore the steam dump valves must be used.
D.
KIA Sys #
K/A System K/A Category K/A Statement 000058 Loss of DC Power KIA #
AK3.02 WA Importance 4.0 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Memory References provided to Candidate None Objective #:
Task ID#:
10 CFRPart55Content:
(CFR41.5,41.10/45.6/45.1)
Knowledge of the reasons for the following responses as they apply to the Loss of DC Power:
Actions contained in EOP for loss of D(.
power Technical
References:
AOP 2.39.1A step 2 and Auto actions 0' I page
- 1.
Page 13 of 'I00
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 14.
The plant is operating at 100% power with all systems in NSA.
A Service Water/Normal Intake Structure Loss has occurred.
A Steam Generator Blowdown Test Tank discharge is in progress.
Step 2 of AOP-2.30.1, Service WatedNormal Intake Structure Loss instructs the operating crew to secure any liquid waste discharges IF service water header pressure cannot be restored above 34 )sig.
Under these conditions, WHY are liquid waste discharges secured?
Because:
A.
The required liquid waste discharge dilution water flow cannot be assured.
B.
The liquid waste discharge radiation monitor will be inoperable.
C.
Liquid waste discharge flow control will be unavailable D.
Steam generator cleanup ion exchanger temperature control cannot be assured.
Answer A
ExplanationlJustification:
A.
- 6.
C.
D.
Correct. IAW OM Fig. 31-1 and 25-4 Dilution water for liquid waste discharges is provided by the service water system.
Incorrect. The liquid waste discharge radiation monitor is not cooled by river water and will remain operable during loss of service water.
Incorrect. Air will still be available to the flow control valve since domestic water is manually aligned to cool the air compressors. Therefore t':er will be no loss of air.
Incorrect. Steam Generator Blowdown Test Tank ion exchangers are used to clean-up the water before they are prepared for discharge NU r during the discharge. Additionally, the need to cool evaporator distillate at Unit 2 has been removed since the evaporators have been retirec: in place.
'A Sys #
KIA System WA Category KIA Statement d00062 Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water:
KIA #
AK3.03 KIA Importance 4.0 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
OM Fig. 31-1 and 25-4 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR 41.4, 41.8 / 45.7 )
Loss of Nuclear Service Water Guidance actions contained in EOP for. oss of nuclear service water Page 14 of 100
Beaver Va
- 15.
The plant is in Mode 3 with all s ley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
stems in normal alignment for this mode.
The reactor trip breakers are OPEN.
A Loss of station instrument air occurs.
Station instrument air header pressure is 0 psig.
What impact will this loss of station instrument air have on the following CVCS functions:
Charging will (1)
- letdown will (2)
- RCP seal injection will (3)
- .RCP seal return will (4)
I A.
(1) isolate (2) isolate (3) isolate (4) isolate B.
(1) isolate (2) remain in service (3) remain in service (4) isolate C.
(1) remain in service (2) isolate (3) remain in service (4) remain in service D.
(1) remain in service (2) remain in service (3) isolate (4) remain in service Answer C
ExplanationlJustification:
A.
B.
C.
D.
K/A Sys #
K/A System K/A Category K/A Statement 000065 Loss of Instrument Air Incorrect. Charging, seal injection and seal return remain in service.
Incorrect. Charging and seal return remain in service and letdown will isolate.
Correct. IAW AOP-2.34.1 attachment 2.34.1-1 Ch 7 fail positions.
Incorrect. Letdown will isolate and seal injection will remain in service.
j-Ability to determine and interpret the following as they apply to the Loss of Instrument Air.
Failure modes of air-operated equipme t JA #
AA2.08 K/A Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
AOP-2.34.1 attachment 2.34.1-1 Ch 7 f til Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR:
43.5 I45.13) positions.
Page 15 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 16.
Following a reactor trip and safety injection, the crew is performing actions of E-0, Reactor Trip Or Safety Injection.
The following conditions exist:
All SG pressures are 1000 psig and stable.
All SG NR levels are approximately 35% and stable.
AFW flow is 380 gpm and stable.
RCS pressure is -1 000 psig, lowering slowly.
RCS temperature is 545"F, stable.
Auxiliary Building - 710 Area Radiation Monitor [2RMP-RQ203] is in HIGH alarm.
Auxiliary Building - 735 Area Radiation Monitor [2RMP-RQ204] is in HIGH alarm.
Auxiliary Building - 735 Area Radiation Monitor [2RMP-RQ205] is in HIGH alarm.
Containment pressure is 13.45 psia and stable.
PRT conditions are NORMAL.
CNMT sump level and radiation are NORMAL.
Which ONE (1) of the following procedures MUST be entered to mitigate this event?
A.
- q.
ECA-1.2, LOCA Outside Containment.
C.
E-I, Loss Of Reactor Or Secondary Coolant.
D.
ES-1.2, Post-LOCA Cooldown And Depressurization.
Answer B
Explanation/Justification:
A.
B.
C.
D.
Incorrect. RCS pressure is dropping.
Correct. Per E-0 step 20.
Incorrect. All CNMT parameters are normal.
Incorrect. Entry would be from E-I, which would not be used.
K/A Sys #
KIA System K/A Category K/A Statement WlE04 LOCA Outside Ability to determine and interpret the following as they apply to the (LOCA Outside Containment)
KIA #
A2 1 K/A Importance 3 4 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension Facility conditions and selection of apprc mate procedures during abnormal and emergr icy operations Containment References provided to Candidate None Objective #:
LP 3SQS-53.4 Task Obj. # 3 ID#:
Technical
References:
10 CFR Part 55 Content:
EOP E-0 diagnostic steps (CFR: 43.5 / 45.13)
Page 16 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 17.
The plant is operating at 100% power with all systems in NSA.
A LOCA occurs.
Reactor trip and safety injection actuation occur.
The crew is performing actions of E-I, Loss Of Reactor Or Secondary Coolant.
Cold leg recirculation capability cannot be verified and the crew transitions to ECA-1.1, Lo: s Of Emergency Coolant Recirculation.
At step 18 of ECA-1.I, the crew is instructed to stoplstart charging pumps to establish MINIMUM SI flow to remove decay heat.
Wide range RCS pressure is 1000 psig and stable.
The 5 hottest Core exit thermocouples are 550°F and stable.
What is the reason for establishing MINIMUM SI flow in this procedure step?
A.
Prevent a potential ORANGE path for RCS integrity.
B.
Prevent PRZR overfill and subsequent RCS overpressurization.
C.
Delay SI accumulator injection and subsequent isolation.
D.
Delay Refueling Water Storage Tank (RWST) depletion.
Answer D
ExplanationlJustification:
A.
B.
Incorrect. Potential Orange or Red paths on RCS integrity are prevented by limiting the RCS cooldown to 100 "Flhr in this procedure Incorrect. PRZR overfill and subsequent RCS overpressurization are the reasons for securing SI flow in ES-1.2 which would be appropriati for a SMALL break LOCA but NOT a concern for LARGE break LOCAs. For large break LOCAs the PRZR will not overfill and RVLlS is used fot water inventory indications.
Incorrect. A major objective of ECA-1.I is to C/D and depressurize to get the accumulators to inject their inventory.
Correct. IAW ECA-1.1 step 18 bases.
C.
I.
K/A Sys #
K/A System K/A Category K/A Statement WIEI 1 Loss of Emergency Knowledge of the reasons for the following responses as they apply to the (Loss of Emergency Coolant Recirculation) emergency situations.
Manipulation of controls required to obr:3in desired operating results during abnori ial and Coolant Recirc.
K/A #
EK3.3 K/A Importance 3.8 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
ECA-1.1 step 18 bases Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.5 141.10, 45.6, 45.13)
Page 17 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 18.
The plant is operating at 100% power with all systems in NSA.
A small break LOCA occurs inside containment.
The reactor trips and safety injection actuates.
4KV Emergency Bus 2DF is de-energized.
Quench Spray Pump 2QSS*P2lA TRIPPED cannot be started.
Containment pressure is 20 psig and slowly rising.
AFW flow is 100 gpm to each SG.
SG NR levels are 25%.
A & 6 SG NR levels are slowly dropping.
C SG NR level BEGINS rising in an uncontrolled manner.
The crew is performing E-I, Loss Of Reactor Or Secondary Coolant step 15 Verify Cold le!
Recirculation Capability.
Cold leg recirculation capability CANNOT be verified.
Based on these plant conditions, what procedural transition is Required?
Transition into:
A.
ECA-1.1, Loss Of Emergency Coolant Recirculation B.
FR-H.1, Response To Loss Of Secondary Heat Sink C.
E-3, Steam Generator Tube Rupture D.
FR-Z.l, Response To High Containment Pressure Answer B
cplanationlJustification:
A.
B.
C.
D.
K/A Sys #
K/A System K/A Category K/A Statement WIE05 Incorrect. Although the conditions have been met for ECA-1,I entry, FR-H.1, FR-Z.l and E-3 entry conditions are also present and have a I,igher priority.
Correct. FR-H.4 red path entry conditions are present since AFW flow is only 300 gpm and NR levels in all SGs is less than the required 31 :/o adverse CNMT level.
Incorrect. Although the conditions have been met for E-3 entry based on LHP criteria, FR-H.1, and FR-Z.l entry conditions are also presen and have a higher priority.
Incorrect. Although the conditions have been met for FR-Z.l entry, FR-H.l entry conditions are also present and have a higher priority. -
Loss of Secondary Heat Sink Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink)
K/A #
EA2.1 K/A Importance 3.4 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
F-0.3 status tree and EOP users guide age 9 Objective #:
10 CFR Part 55 Content:
(CFR: 43.5 I45.13)
Facility conditions and selection of appr lpriate procedures during abnormal and emerc mcy operations.
paragraph 5.1 Obj.# 2.c Page 18 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 19.
The Plant is operating at 50% power BOL with all systems in NSA.
Control Bank D is at 175 steps.
Control Bank D Demand step counters are at 175 steps.
Control Rod Group Selector Switch is in the MAN position.
Turbine control is in First Stage Out.
The following VALID control room alarms are received:
A4-9F Rod At Bottom A4-4F NIS Power Range Comparator Deviation A4-4G NIS Power Range Neutron Flux Rate High A4-3C Tavg Deviation from Tref A4-1 D Pressurizer Control Pressure Highllow A4-1 E Pressurizer Control Press Deviation Highllow No operator actions have been taken.
Based on these conditions, what will be the status of the following parameters 5 minutes after the E dent began?
e RCS Tavg will be (1) than 558°F.
PRZR Pressure will be (2) ~-
than 2235 psig.
e PRZR Backup Heaters will be (3)
Reactor power will be (4) 50%.
A.
- 1. lower
- 1. lower C.
- 1. higher D.
- 1. higher
- 2. lower
- 2. lower
- 2. higher
- 2. higher
- 3. energized
- 3. energized
- 3. de-energized
- 3. de-energized
- 4. lower than
- 4. equal to
- 4. lower than
- 4. equal to Answer A
ExplanationlJustification:
A.
Correct. All of the alarms listed will alarm for either high or low conditions EXCEPT A4-9F Rod At Bottom. Since A4-9F has energized and i: valid, the candidate will need to identify there is a dropped rod event in progress. This also will eliminate higher Tavg and PRZR pressure from consideration. With the turbine in first stage out reactor power will be lower since the governor valves will not reposition to adjust for the lo.Jer steam pressures. Reactor power would be the same if the turbine was in First stage In. BOL was selected to provide the most definite chactges to the listed parameters.
Incorrect.. With the turbine in first stage out reactor power will be lower since the governor valves will not reposition to adjust for the lower steam pressures. Reactor power would be the same if the turbine was in First stage In.
Incorrect. Tavg and PRZR pressure will be lower. The PRZR B/U heaters will be energized.
Incorrect. Tavg and PRZR pressure will be lower. The PRZR B/U heaters will be energized. Reactor power will be lower.
B.
C.
D.
WA Sys #
K/A System WA Category WA Statement 000005 Inoperable/Stuck Control N/A Ability to verify that the alarms are const tent WA #
2 4 46 WA Importance 4 2 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis Eod with the plant conditions References provided to Candidate None Objective #:
Task ID#:
Technical
References:
10 CFR Part 55 Content:
AOP-2.1.8 symptoms (ran on Unit 2 sin ulator to confirm all parameters)
(CFR: 41. I O 143.5 I45.3 145.12)
Page 19 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 20.
The plant is in Mode 2 with a reactor startup is in progress. All systems are in normal alignment for this condition.
The reactor trip breakers are closed with the shutdown banks withdrawn.
Control rod withdrawal is in progress.
Two control bank A rods fail to move when required, and become misaligned by 15 steps BOTH Source Range detectors SIMULTANEOUSLY become inoperable.
Reactor power is 1 X I O 4 CPS and stable.
What are ALL of the IMMEDIATELY Required Technical Specification actions?
1. Suspend operations involving positive reactivity additions.
- 2. Open the reactor trip breakers.
- 3. Initiate action to restore one source range neutron flux monitor to OPERABLE status.
A.
1 and4
- 6.
1 and2 C.
2and 3 D.
3and4 Answer B
ExplanationIJustification:
A.
B.
Incorrect. #I is correct. #4 is the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requirement for the misaligned rods; it is NOT an immediate requirement.
Correct. IAW TS 3.3.1 conditions H and I.
Incorrect. #2 is correct. #3 is the immediate requirement for both source ranges inoperable in Mode 6.
Incorrect. #3 is the immediate requirement for both source ranges inoperable in Mode 6 and #4 is the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requirement for the misalignec! rods; it is NOT an immediate requirement.
P K/A Sys #
WA System WA Category WA Statement 000032 Loss of Source Range NI NIA Knowledge of less than or equal to one 'lour Technical Specification action statemer :s for systems.
K/A #
2.2.39 KIA Importance 3.9 Exam Level RO Level Of Difficulty: (1 -5)
Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate None Technical
References:
TS 3.3.1 Table 3.3.1-1 and conditions and I.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 I 41.IO I43.2 145.13)
Page 20 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 21.
The plant is operating at 100% power with all systems in NSA.
A 50 gpm Steam Generator Tube leak develops.
The crew enters AOP 2.6.4, Steam Generator Tube Leakage.
A controlled shutdown to Mode 3 has been completed.
It has been determined that the leaking Steam Generator shall be cooled and depressurize j using the Backfill method.
Which of the below listed attributes are advantages to using the Backfill method over other metho Is?
- 1. Facilitates processing of contaminated primary coolant.
- 2. Minimizes Radiological releases.
- 3. Fastest means to cool the leaking Steam Generator.
- 4. Minimizes boron dilution of the primary coolant.
A.
1 and 2 B.
1 and4 C.
2and 3 D.
3and4 Answer A
ExplanationlJustification:
A.
B.
- d.
KIA Sys #
WA System KIA Category KIA Statement 000037 Steam Generator Tube Correct. IAW E-3 step 43 background (AOP does not have a background document). The AOP used the EOP background to develop attactments to address cooling of the leaking SG).
Incorrect. Boron dilution will NOT be limited by this method.
Incorrect. The fastest means to cooldown the SG is the steam dump method.
Incorrect. These are advantages of the blowdown method.
Knowledge of the reasons for the following responses as they apply to the Steam Generator Tube Leak:
Use of feed and bleed process.
Leak KIA #
AK3.04 WA Importance 2.5 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
E-3 step 43 background Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR 41.5,41.10 / 45.6 / 45.13)
Page 21 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
Which ONE (1) of the following constitutes a loss of an OPERABLE containment?
While in...
MODE 3, it is discovered that Containment Atmosphere Purge Makeup valve will NOT OPEN.
MODE 4, a review of Integrated Leak Rate test results show that leakage is NOT WITHIN LIMITS.
MODE 5, it is discovered that the Phase ' B isolation valve for CCP to the RCPs will NOT CLOSE.
MODE 6, it is discovered that one of the Emergency Airlock (EAL) doors will NOT CLOSE.
Answer B
ExplanationlJustification:
A.
- 6.
Incorrect. Purge Makeup Valve should not normally be open in Mode 3 and would not be a loss of OPERABLE containment if it does not OPEN.
Correct. IAW Technical Specification 3.6.1 and its' bases. (Operable containment equates to CNMT integrity with the new ITS). In order to maintain the containment operable leakage must be maintained less than or equal to 1.O La. The situation posed in the stem of the questior would requires the containment to be restored to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and at BVPS licensed ROs are required to know from memory all ' S LCO corditions that would require less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> actions. (See attached LP objective)
Incorrect. An OPERABLE Containment is not required in Mode 5.
Incorrect. In Mode 6, 1 airlock door may remain open.
C.
D.
K/A Sys #
KIA System KIA Category KIA Statement 000069 Loss of CTMT Integrity Ability to determine and interpret the following as they (W/E 14) apply to the Loss of Containment Integrity:
K/A #
AA2.01 KIA Importance 3.7 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis qeferences provided to Candidate None Technical
References:
TS 3.6.1 and bases.
Loss of containment integrity ijective #:
LPJSQS-CONT ITS Task ID#:
10 CFR Part 55 Content:
(CFR: 43.5 / 45.13)
Obi. # 5 Page 22 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 23.
The plant is operating at 100% power with all systems in NSA.
A small break LOCA occurs inside containment.
The crew performs the actions of E-0, Reactor Trip Or Safety Injection and transitions into GI, Loss Of Reactor Or Secondary Coolant.
Consider the below listed criteria:
What are ALL of the criteria that MUST be met before a transition to ES-1.I, SI Termination can be made?
- 1. BOTH RX trip breakers must be open
- 2. One train of CIA must exist
- 3. Cold leg recirculation capability must exist
- 4. The RCS must be subcooled
- 5. A secondary heat sink must exist
- 6. RCS pressure must be stable or rising
- 7. PRZR level must be indicating on span
- 8. A RCP must be operating A.
B.
C.
Answer D
Explanation/Justification:
1, 3,4, 5, & 8 ONLY 1, 2, 4, 6, & 8 ONLY 2, 3, 5, & 7 ONLY 4, 5, 6, & 7 ONLY A.
- 8.
C.
D.
Incorrect. IAW the background document for ES-1.I the only 4 criteria that must be met are items 4, 5, 6, and 7. The other 4 items are nice 10 have during recovery efforts, but they are not required for SI termination. Additionally, cold leg recirculation capability isnt even checked befwe a transition into ES-1.I is allowed. The 4 correct criteria (RCS Subcooling, heat sink, RCS pressure rising, and indicated PRZR level) combina to indicate that the RCS is in a safe state with adequate core cooling.
Incorrect. IAW the background document for ES-1.I the only 4 criteria that must be met are items 4, 5, 6, and 7. The other 4 items are nice to have during recovery efforts, but they are not required for SI termination. Additionally, cold leg recirculation capability isnt even checked bebre a transition into ES-1.I is allowed. The 4 correct criteria (RCS Subcooling, heat sink, RCS pressure rising, and indicated PRZR level) combinc to indicate that the RCS is in a safe state with adequate core cooling.
Incorrect. IAW the background document for ES-1.I the only 4 criteria that must be met are items 4, 5, 6, and 7. The other 4 items are nice to have during recovery efforts, but they are not required for SI termination. Additionally, cold leg recirculation capability isnt even checked before a transition into ES-1.I is allowed. The 4 correct criteria (RCS Subcooling, heat sink, RCS pressure rising, and indicated PRZR level) combine to indicate that the RCS is in a safe state with adequate core cooling.
Correct. IAW the background document for ES-1.I the only 4 criteria that must be met are items 4, 5, 6, and 7. These 4 criteria (RCS Subcooling.
heat sink, RCS pressure rising, and indicated PRZR level) combine to indicate that the RCS is in a safe state with adequate core cooling.
KIA Sys #
KIA System KIA Category KIA Statement WIE02 SI Termination Knowledge of the interrelations between the (SI Termination) and the following Facilitys heat removal systems, tncludin 4 primary coolant, emergency coolant, the decay heat removal systems, and relatic 1s between the proper operation of these systems to the operation of the facility KIA #
EK2 2 KIA Importance 3 5 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
ES-I 1 background page 1 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR 41 7 145 7)
Page 23 of 1110
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 24.
The plant has been operating at 100% power with all systems in NSA for the past 100 days.
An inadvertent Turbine trip occurs coincident with a loss of offsite power.
IAW the Plant Technical Specifications, which of the following components MUST operate to preve It Steam Generator Overpressure (> 1 10% of design pressure)?
- 1. Atm Stm Dump Control Valves
- 2. Steam Generator Safety Valves 3. Residual Heat Release Valve
- 4. Turb Driven AFW Pump Stm Supply Valves A.
1&2ONLY
- 6.
2ONLY C.
3&4ONLY D.
4ONLY Answer 0
ExplanationlJustification:
A.
Incorrect. Atm Stm Dump Control Valves are -r decay he:
removing there is a
'emovc NOT overpressure protection. They will limit SG pressure by virtue of decay heat, but they are NOT required by UFSAR. Since.,,e Atm Stm Dump Control Valves remove decay heat and thus limit pre sure common misconception that they are required to prevent SG overpressure.
Correct. IAW Tech Spec bases 3.7.1 page 1 1 paragraph and page 2 1" paragraph. At BVPS ROs are expected to know the purpose and 1 ases Tech spec components. (See attached objective)
Incorrect. The residual heat release valve is designed for decay heat only and no credit is taken for limiting SG overpressure. Steam supplic i to the turbine driven AFW pump are required for maintenance of the heat sink NOT overpressure protection.
Incorrect. Steam supplies to the turbine driven AFW pump are required for maintenance of the heat sink NOT overpressure protection.
B.
C.
D.
.,A Sys #
K/A System WA Category WA Statement WlE13 Steam Generator Over-pressure Knowledge of the interrelations between the (Steam Generator Overpressure) and the following:
Components and functions of control an I safety systems, including instrumentatio I, signals, interlocks, failure modes, and automatic and manual features.
KIA ?Y EK2.1 KIA Importance 3.0 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
Tech Spec bases 3.7.1 page 1 1"parag.aph Objective #:
LP 3SQS-PLTSYS Task ID#:
10 CFR Part 55 Content:
(CFR: 42.7 145.7) and page 2 1" paragraph.
ITS Obj. # 2 Page 24 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 25.
The plant is operating at 100% power with all systems in NSA.
A large break LOCA occurs inside containment.
The crew is implementing the actions of E-I, Loss Of Reactor Or Secondary Coolant.
The STA then reports the following CSF status:
YELLOW-Core Cooling-FR-C.3, Response To Saturated Core Cooling (Based on core exit temperatures less than 729°F and RVLIS greater than 40% full range).
ORANGE-Containment-FR-Z.2, Response to Containment Flooding (Based on a containn vent sump level of 189 inches).
YELLOW-Containment-FR-Z.3, Response To High Containment Radiation Level (Based on a containment radiation level of 76 R/hr).
The crew transitions to FR-Z.2, Response to Containment Flooding and completes all of the action: of this procedure. The STA then reports THE SAME CSF status that was reported earlier.
What procedural transition, if any, is now Required?
A.
Return to Step 1 of FR-Z.2, Response to Containment Flooding.
B.
Return to step in effect of E-I, Loss Of Reactor Or Secondary Coolant.
C.
Go to FR-C.3, Response To Saturated Core Cooling.
D.
Go to FR-Z.3, Response To High Containment Radiation Level.
nswer B
ixplanationlJustification:
A.
B.
C.
D.
Incorrect. Plausible since normally the EOP usage rules do not allow a transition out of a red or orange path procedure until the symptoms t,ave been corrected. However, FR-Z.2 is an exception and the crew is directed to return to step and procedure in effect.
Correct. IAW F0.5 bases for step 4 page 7 knowledge paragraph.
Incorrect. Plausible. Core cooling is a higher priority status tree terminus than either containment radiation or flooding. However, the terminL.s is only yellow, and transition to this procedure is optional not required. The question specifically asks for required transition.
Incorrect. Plausible. Containment radiation is a higher priority status tree terminus than returning to E - I. However, the terminus is only yello v, and transition to this procedure is optional not required. The question specifically asks for required transition.
WA Sys #
WA System KIA Category WA Statement WlE15 Containment Flooding Knowledge of the operational implications of the following concepts as they apply to the (Containment Flooding).
Normal, abnormal and emergency oper; sting procedures associated with (Containment Flooding).
K/A #
EK1.2 KIA Importance 2.7 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate None Technical
References:
F0.5 bases for step 4 page 7 knowledge paragraph.
Objective #:
LP-3SQS-53.3 Task ID#
10 CFR Part 55 Content:
(CFR: 41.8 141.10,45.3)
Obj. # 3 Page 25 of 130
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 26.
The plant is operating at 100% power with all systems in NSA.
A small break LOCA occurs inside containment.
All systems function as designed.
The crew is implementing the actions of ES-1.2, Post LOCA Cooldown and Depressurizatir in.
All RCPs have been secured.
Both trains of RVLIS are 00s.
SI, CIA, and CIB have all been reset.
While depressurizing the RCS to minimize subcooling in step 24 of ES-1.2, the following plmt conditions are observed:
PRZR level is 45% and rapidly rising.
RWST level is 395 inches and slowly dropping.
CNMT pressure is 15 psig and slowly dropping What procedural transition, if any, is now Required?
A.
Continue with step 24 of ES-1.2, Post LOCA Cooldown and Depressurization.
B.
Go to ES-0.4, Natural Circulation Cooldown With Steam Void In Vessel (Without RVLIS).
C.
Go to ES-1.3, Transfer To Cold Leg Recirculation.
D.
Go to FR-Z. 1, Response To High Containment Pressure.
Answer C
ExplanationlJustification:
A.
Incorrect. Plausible if the candidate does not recognize the need to transition to ES-1.3 based on RWST level below 400 inches.
Incorrect. Plausible since RCPs are off and a natural circulation cooldown is in progress AND PRZR level rapidly rising is indicative of bubb1.t formation in the upper head region. However, the transition to ES-0.4 can only be made from ES-0.2 where there is no other accident in pro Iress.
Correct. IAW ES-1.2 LHP item 4. ES-1.2 bases page 1 4Ih paragraph.
Incorrect. Plausible since CNMT pressure is above 11 psig. However, this is incorrect because both QS pumps are operating and this is on1 a yellow path procedure.
C.
D.
KIA Sys #
WA System WA Category WA Statement WlE03 LOCA Cooldown -
Ability to determine and interpret the following as they apply to the (LOCA Cooldown and Depressurization)
Facility conditions and selection of apprc piate procedures during abnormal and emergk ncy operations.
Depress.
WA #
EA2.1 WA Importance 3.4 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate None Objective #:
LP-3SQS-53.3 Task ID#:
Obj. # 3 Technical
References:
10 CFR Part 55 Content:
ES-1.2 LHP item 4. ES-1.2 bases page ' dth paragraph.
(CFR: 43.5 145.13)
Page 26 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 27.
The plant has been operating at 100% power with all systems in NSA for the past 100 days.
An inadvertent reactor trip occurs coincident with a loss of offsite power.
All systems function as designed.
The crew is implementing the actions of ES-0.2, Natural Circulation Cooldown.
RCS temperature is 350°F and stable.
RCS Subcooling is 200°F and stable.
RCS Pressure 1200 psig and stable.
RCS cooldown rate is 20"F/hr and stable.
Alarm A I 1 -5G CRDM Shroud Fan Auto-StartlAuto-Stop is received. ALL CRDM shroud fans have tripped and cannot be restarted.
What ramifications will the loss of these CRDM Shroud Fans have on the continued performance of ES-0.2, Natural Circulation Cooldown?
A.
Further RCS cooldown (below 350°F) cannot continue UNTIL a suitable RX vessel head soak has t een performed.
B.
Further RCS depressurization (below 1200 psig) cannot continue UNTIL a suitable RX vessel head soak has been performed.
C.
Immediately INCREASE RCS pressure 100 psig to RAISE RCS subcooling.
D.
Immediately DECREASE RCS pressure 100 psig to LOWER RCS subcooling Answer B
ExplanationlJustification:
Incorrect. The restriction to perform a head soak only applies when the RCS is below 350°F. However, cooldowns below 350°F are still allowed when CRDM fans are unavailable.
Correct. IAW ES-0.2 step 13 and background.
Incorrect. Raising pressure 100 psig is a technique employed by ES-0.4 natural circulation procedure when the head void growth becomes '00 large.
Incorrect. Minimizing Subcooling is a technique employed when RX vessel stresses are the concern but NOT when RX vessel head voids :e the concern. Decreasing pressure may actually cause a void to form.
B.
C.
D.
K/A Sys #
KIA System K/A Category K/A Statement WIE09 Natural Circ.
Knowledge of the operational implications of the following Annunciators and conditions indicating concepts as they apply to the (Natural Circulation Operations) the (Natural Circulation Operations).
signals, and remedial actions associated with KIA #
EK1.3 KIA Importance 3.3 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
ES-0.2 step 13 and background.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.8 / 41.10,45.3)
Page 27 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 28.
The plant is operating at 40% power with all systems in normal alignment for this power level.
B RCP breaker OPENS due to a mechanical failure.
What impact will this OPEN breaker have on the reactor protection system (RPS)?
A reactor trip signal will...
A.
NOT be generated. At this power level it takes 2/3 RCP breakers open to generate a reactor trip sicslal.
B.
NOT be generated. At this power level it takes 213 RCS Loops Low Flow generate a reactor trip sigi ial.
C.
Be generated by the single open RCP breaker.
D.
Be generated by the single RCS loop flow low.
Answer D
ExplanationlJustification:
A.
B.
C.
D.
K/A Sys #
K/A System K/A Category K/A Statement 003 Knowledge of the effect that a loss or malfunction of the RCPS will have on the following:
KIA #
K3.04 K/A Importance 3.9 Exam Level RO Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
UFSAR logic Figure 7.3-10 Incorrecr. It is true that it takes 2/3 breakers open to generate a trip signal at this power level. However, the single loop flow low will general: a trip signal.
Incorrect. It only takes a single loop flow low to generate a trip signal.
Incorrect. It is true that a trip signal will be generated. However, it is not generated from the RCP breaker opening Correct. IAW UFSAR logic Figure 7.3-10
~
Reactor Coolant Pump RPS Level Of Difficulty: (1-5) ijective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 / 45.6)
Page 28 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 29.
In the CVCS
- 1. Which ONE (1) of the below listed components is designed to prevent flashing at the downstream side of the letdown orifices?
- 2. How is this accomplished?
A.
- 1. Letdown Orifice lsol Vlvs [2CHS*AOV200A(B)(C)]
- 2. Close on high temperature downstream of the orifices B.
- 1. Non-Regen HX Disch Press Control Vlv [2CHS*PCV145]
- 2. Maintains pressure downstream of the orifices above saturation pressure C.
- 1. Non-Regen HX Disch Diverting Vlv [2CHS*TCV143]
- 2. Maintains pressure downstream of the orifices above saturation pressure D.
- 1. Non-Regen HX Temp Control Vlv [2CCP*TCV144]
- 2. Maintains letdown temperature downstream of the orifices below saturation temperature.
Answer B
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. These valves do not have a high temperature isolation signal, although this would prevent flashing. They will isolate on CIA.
Correct. IAW 20M-7.1.C page 8 Is' paragraph Incorrect This valve is downstream of the Non-regen HX and diverts water away from the demineralizers to protect them from high tempers ure. It does NOT maintain pressure downstream of the orifices.
incorrect This valve does cool the letdown water, BUT it is too far downstream to prevent flashing at the letdown orifices.
KIA Sys #
WA System KIA Category KIA Statement 004 Chemical and Volume Control Knowledge of CVCS design feature(s) and/or interlock(s) which provide for the following Temperature/pressure control in letdowr line prevent boiling. lifting reliefs, hydraulic s lock, piping damage, and burst KIA #
K4 11 KIA Importance 3 1 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
20M-7 1 C page 8 Is' paragraph Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR 41.7)
Page 29 of 200
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 30.
The plant is in Mode 4 with RCS temperature at 210°F.
RHR Pump 2RHS*P21A is on clearance.
Train "B" of RHS is in service and being used for an RCS cooldown at 25"F/hr.
All Train "B" RHS components are arranged in their normal alignment for plant cooldown.
RHR HX Flow Control Valve [2RHS*HCV758B] is 30% OPEN.
RHR HX Bypass Valve [2RHS*FCV605B] Is 50% OPEN.
As a result of poor Foreign Material Exclusion (FME) practices, a rubber Anti-C boot becomes lodgcd in the tube side of the "B" RHS Heat Exchanger. The boot BLOCKS 25% of the tubes in the heat exchanger.
IF the RCS cooldown is to CONTINUE at 25"F/hr, the reactor operator will be required to A.
Manually CLOSE RHR HX Bypass Valve [2RHS*FCV605B] and allow RHR HX Flow Control Valve
[2RHS*HCV758B] to automatically throttle OPEN to maintain total RHS system flow.
B.
Manually OPEN RHR HX Bypass Valve [2RHS*FCV605B] and allow RHR HX Flow Control Valve
[2RHS*HCV758B] to automatically throttle CLOSED to maintain total RHS system flow.
C.
Manually CLOSE RHR HX Flow Control Valve [2RHS*HCV758B] and allow RHR HX Bypass Valve
[2RHS*FCV605B] to automatically throttle OPEN to maintain total RHS system flow.
D.
Manually OPEN RHR HX Flow Control Valve [2RHS*HCV758B] and allow RHR HX Bypass Valve
[2RHS*FCV605B] to automatically throttle CLOSED to maintain total RHS system flow.
Answer D
.planation/Justification:
A.
- 6.
C.
D.
Incorrect Manually closing 2RHS*FCV605B will force more water through the HX. However, 2RHS*FVC758B does not have automatic flow control and will NOT automatically throttle open to maintain RHS system flow.
Incorrect. Manually opening 2RHS*FCV605B will divert water away from the HX which would cause a heatup. Also, 2RHS*FVC758B does rot have automatic flow control and will NOT automatically throttle open to maintain RHS system flow.
Incorrect. These actions will slow the RCS cooldown.
Correct. IAW OM figure 10-2. 2RHS*FCV605B has the automatic flow control, and 2RHS*FCV758B is a manually adjusted valve to control tiow through the RHS HX.
P KIA Sys #
KIA System KIA Category KIA Statement 005 Knowledge of the effect of a loss or malfunction on the following will have on the RHRS:
KIA #
K6.03 KIA Importance 2.5 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate None Technical
References:
OM figure 10-1 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 145.7)
Residual Heat Removal RHR heat exchanger Page 30 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 31.
The plant is operating at 100% power with all systems in NSA.
Low Head SI Pump 2SIS*P21A becomes inoperable due to a bearing failure on the pump.
In the event of a Large break LOCA, how will this failure impact Train A ECCS performance?
BEFORE transfer to cold leg recirculation there will be AFTER transfer to cold leg recirculation there will be (1 1 (2)
AND (Assume all other systems function as designed during the Large break LOCA)
A.
- 1. NO Low Head SI flow.
- 2. Low Head SI flow available via Recirc spray pump 2RSS*P21 C AND High Head SI flow will be available.
B.
- 1. Low Head SI flow available via Recirc spray pump 2RSS*P21 C.
C.
- 1. NO Low Head SI flow.
D.
- 1. Low Head SI flow available via Recirc spray pump 2RSS*P21 C.
- 2. Low Head SI flow will be available via Recirc spray pump 2RSS*P21 C AND High Head SI flow will be available.
Qnswer A
dxplanationlJustification:
A.
B.
C.
D.
K/A Sys #
K/A System K/A Category 006 K/A #
K6.03 K/A Importance 3.6 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
EOP Attachment A-0.7 and ES-1.3 and JOND Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 / 45.7)
Correct. IAW EOP Attachment A-0.7 and ES-1.3 and VOND 11-1 & 13-1 Incorrect. Before transfer to cold leg Recirc there will be NO LHSl flow; after transfer to cold Recirc there will be flow via 2RSSP21C Incorrect. After transfer to cold leg Recirc there will be LHSl flow via 2RSSP21C and it will provide suction to the HHSl pump.
Incorrect. Before transfer to cold leg Recirc there will be NO LHSl flow.
.~
K/A Statement Safety Injection Pumps Emergency Core Cooling Knowledge of the effect of a loss or malfunction on the following will have on the ECCS:
11-1 & 13-1 Page 31 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 32.
The plant is operating at 100% power with all systems in NSA.
ChargingIHigh Head Safety Inj Pump 2CHS*P21A is running.
Charging/High Head Safety Inj Pump 2CHS*P21 B is in standby.
Charging/High Head Safety Inj Pump 2CHS*P21C is NOT racked onto any bus.
A DBA Large break LOCA occurs inside containment.
All systems function as designed.
RWST level has dropped to below 369 inches.
The SI Recirc Mode signal has been actuated.
All automatic actions associated with the SI Recirc Mode signal have been initiated and corn p leted.
How will Charging/High Head Safety Inj Pump 2CHS*P21A SUCTION and DISCHARGE be impact?d by the receipt of this SI Recirc Mode signal?
Suction will be aligned to the Discharge will be aligned to (1)
(2) -
AND A.
- 1. RWST
- 2. Cold leg injection B.
- 1. 2RSS*P21 C Recirc Spray Pump discharge
- 2. Cold leg injection C.
- 1. RWST
- 2. Hot leg injection D.
- 1. 2RSS*P21 C Recirc Spray Pump discharge
- 2. Hot leg injection Answer B
ExplanationlJustification:
A.
B.
C.
D.
KIA Sys #
KIA System KIA Category KIA Statement 006 KIA #
A4.05 KIA Importance 3.9 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
EOP attachment A-0.7 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 145.5 to 45.8)
Incorrect Suction will aligned to 2RSS*P21 C discharge flowpath.
Correct. IAW EOP attachment A-0.7.
Incorrect Suction will aligned to 2RSS*P21C discharge flowpath AND discharge will remain aligned to the cold legs Incorrect Discharge will remain aligned to the cold legs.
Emergency Core Cooling Ability to manually operate andlor monitor in the control room:
recirculation Transfer of ECCS flowpaths prior to Page 32 of 1CO
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 33.
The plant is operating at 100% power with all systems in NSA.
Over the course of 100 minutes, PRT level indicator 2RCS-LI-470 increases from 73% to 9%
The VCT level change indicates that all of this RCS leakage is into the PRT Based on this PRT level rise, this leakage would be classified as (1 1 and WOL id be
-~
(2) the Technical Specification LCO limit.
A.
(1) Identified (2) Above B.
(1) [Jnidentified (2) Above C.
(1) Identified (2) Within D.
(1) Unidentified (2) Within Answer C
ExplanationlJustification:
A.
Incorrect. Correct leakage classification. Wrong limit. The limit is 3 gpm and the level ri equal s to - 6.7 gpm B.
C.
D.
Incorrect. Wrong leakage classification. Wrong limit. The limit is 10 gpm and the level rise equates to - 6.7 gpm Correct. Correct leakage classification. Correct limit. The limit is 10 gpm and the level rise equates to - 6.7 gpm Incorrect. Wrong leakage classification. Correct limit. The limit is 10 gpm and the level rise equates to - 6.7 gprn KIA Sys #
KIA System KIA Category 007 Pressurizer Relief/Quench NIA Tank KIA #
2.2.42 KIA Importance 3.9 Exam Level 2uestion Source:
New eferences provided to Candidate 2RCS-TK-22 PRT Tank Level curve Objective #:
Task ID#:
KIA Statement Ability to recognize system parameters that are entry-level conditions for Technical Specifications RO Level Of Difficulty: (1-5)
Question Cognitive Level:
Higher Application Technical
References:
2RCS-TK-22 PRT Tank Level curve, 1 S 10 CFR Part 55 Content:
(CFR 41 7 / 41 10 / 43 2 I43 3 / 45 I )
3 4 13 Page 33 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 34.
The plant is operating at 75% power BOL with all systems in normal alignment for this power level.
Rod control is in Automatic All RCS Tavg indications are Matched with Tref VCT level is 45% and stable.
With NO INITIAL change in turbine load, control rods begin to slowly step INWARD.
Which ONE (I) of the below listed failures will cause this inward rod motion?
A.
The Loop A Tc transmitter slowly failing HIGH.
- 6.
A Loop A TH transmitter slowly failing HIGH.
C.
Primary Grade Water To Boric Acid Blender [2CHS*FCV114A] failing OPEN.
D.
Non-Regen HX Disch Temp Control Vlv [2CCP*TCV144] failing OPEN.
Answer D
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. If the Tc transmitter fails high, then Tavg for that loop will rise. However, this will NOT cause rods to move since the Tavg signal tu.l rod control is median selected.
Incorrect. If a Th transmitter fails high, then Tavg for that loop will rise. However, this will NOT cause rods to move since the Tavg signal to od control is median selected.
Incorrect. This would appear to be a potential dilution path. However, with blender setup in NSA, FCVI 13B and FCVl14B are both CLOSE isolating any potential flowpath.
Correct. IAW VOND 15-5 grid F-4 and LP GPF.C4 page 30. Colder water will allow the demineralizers to absorb more boron.
A Sys #
K/A System WA Category KIA Statement 008 Component Cooling Water N/A Ability to diagnose and recognize trend: in an accurate and timely manner utilizing the appropriate control room reference matt rial WA #
2 4 47 WA Importance 4 2 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Higher Comprehension Technical
References:
10 CFR Part 55 Content:
VOND 15-5 grid F-4 and LP GPF.C4 pa le 30.
(CFR: 41.IO 143.5 145.12)
Page 34 of I C 0
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 35.
The plant is in Mode 2 preparing for a turbine startup all systems in normal alignment for this mode Reactor power is 3% and stable.
The PRZR Master Pressure Controller output is at 42% demand signal.
PRZR pressure is 2235 psig and stable.
Both PRZR spray valve controllers are in AUTOMATIC.
PRZR Spray Valve 2RCS*PCV455A is 20% OPEN.
PRZR Spray Valve 2RCS*PCV455B is CLOSED.
The PRZR control heater is in AUTO (Red Target).
The Steam Dump Control Mode Selector switch is in the Stm Press position.
The Main Stm Manifold Press Controller [2MSS-PK464] is in AUTOMATIC with a setpoint )f 1000 psig.
RCS temperature is 547°F and stable.
Main Stm Manifold Stm Press [2MSS-PT464] transmitter THEN fails LOW How will the PRZR Pressure control system INITIALLY respond to this failure?
(Assume NO operator actions)
PRZR Master Pressure Controller output will (1) and cause PRZR Spray Valve 2RCS*PCV455A to (2)
A.
(1 ) RISE above 42%
(2) OPEN more than 20%
- 8.
(1) RISE above 42%
(2) fully CLOSE (1) DROP below 42%
(2) OPEN more than 20%
D.
(1) DROP below 42%
(2) fully CLOSE Answer A
ExplanationlJustification:
A.
Correct IAW 20M-6.4.IFattachment 2 and 20M-21.5.A.12. A low failure of 2MSS-PT464 will cause the steam dumps to close as the controlL:r is trying to maintain 1000 psig and the input is now zero. Closing the steam dumps will cause an RCS heatup which will raise RCS pressure. V. ith rising RCS pressure the PRZR master controller output will rise and spray valve 455A will OPEN farther to drop pressure.
Incorrect. Spray valve 455A will OPEN.
Incorrect. Master controller output will rise.
Incorrect.. Master controller output will rise and Spray valve 455A will OPEN.
B.
C.
D.
K/A Sys #
KIA System K/A Category K/A Statement 010 Pressurizer Pressure Control Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including:
KIA #
AI.06 K/A Importance 3.1 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
20M-6.4.1Fattachment 2 and 20M-21.51.12.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.5 / 45.5)
RCS heatup and cooldown effect on pre:.sure Page 35 of 1011
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 36.
The plant is operating at 100% power with all systems in NSA.
PRZR Channel I Press 2RCS-PT455 has failed HIGH.
The control room crew has tripped all associated bistables IAW 20M-6.4.IF, Instrument Fa lure.
PRZR Control Pressure [2RCS-PT445] THEN fails HIGH.
What will be the INITIAL plant response to this additional failure?
A.
PRZR Spray Valve 2RCS*PCV455A & 2RCS*PCV455B will OPEN.
B.
PRZR PORV 2RCS-PCV455C will OPEN.
C.
PRZR PORVs 2RCS-PCV455D & 2RCS-PCV456 will OPEN.
D.
High PRZR Pressure Reactor Trip will ACTUATE.
Answer C
ExplanationlJustification:
A.
B.
C.
D.
WA Sys #
WA System WA Category WA Statement 010 Pressurizer Pressure Ability to monitor automatic operation of the PZR PCS, PZR pressure WA #
A3.02 WA Importance 3.6 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension rieferences provided to Candidate None Technical
References:
20M-6.4.IF attachment 2.
Incorrect. This would be the INITIAL response if 2RCS-PT444 failed High.
Incorrect. This would be the next response if 2RCS-PT444 failed High.
Correct. IAW 20M-6.4.IF attachment 2.
Incorrec:. Failures are one control channel and one protection channel, therefore NO reactor trip.
Control including:
3jective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 I45.5)
Page 36 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 37.
What are the MINIMUM conditions required to cause the Automatic Reactor Trip Logic for RCS Lciop Low Flow to change from a 2/3 coincidence to a 1/3 coincidence?
Raising power on...
A.
2/4 Power range channels from 8% to 12%.
B.
2/4 Power range channels from 28% to 32%.
C.
314 Power range channels from 8% to 12%.
D.
3/4 Power range channels from 28% to 32%.
Answer B
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. This power change would Arm the trip. Below P-10 the low flow trip is NOT active.
Correct. IAW UFSAR Figs 7.3-9 and 7.3-10.
Incorrect. Plausible since this is the logic to change the logic back to 113 but the setpoint is wrong.
Incorrect. Plausible This will cause the logic to change BUT it s not the minimum.
1 KIA Sys #
WA System KIA Category KIA Statement 012 Reactor Protection KIA #
K4 06 KIA Importance 3 2 Exam Level RO Level Of Difficulty: (1-5)
Knowledge of RPS design featurets) and/or interlock(s) which provide for the following Automatic or manual enable/disable of IPS trips Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Lower Fundamental Technical
References:
UFSAR Figs 7.3-9 and 7.3-10.
10 CFR Part 55 Content:
(CFR: 41.7)
Page 37 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 38.
The plant is operating in Mode 1 with all systems in NSA.
A secondary calorimetric has just been completed. The calorimetric indicates that reactor power is 99.6%. All power range channels are OPERABLE.
Power Range indications are as follows:
N41 -99.8%
N42-99.5%
N43-99.0%
N44 - 100.0%
What Power Range gain adjustments are Required?
A.
Lower ONLY N44 indicated power to 599.6%
B.
Lower N41 AND N44 indicated power to 599.6%
C.
Raise N42 AND N43 indicated power to 299.6%
D.
Raise N41, N42 AND N43 indicated power to 100%
Answer C
ExplanationlJustification:
A.
- 6.
C.
D.
Incorrect. Adjustments not required for Nis with indicated power above actual power.
Incorrect. Adjustments not required for Nis with indicated power above actual power.
Correct. IAW 20M-54.4.Cl page 14 2"d bullet.
Incorrect. Although this would be conservative, it is NOT required to raise N41.
~
K/A Sys #
KIA System KIA Category KIA Statement 01 2 Reactor Protection Ability to predict and/or monitor Changes in parameters Trip setpoint adjustment (to prevent exceeding design limits) associated with operating the RPS controls including:
KIA #
A I.01 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
20M-54.4.Cl page 14 2"d bullet.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.5 I45.5)
Page 38 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 Which ONE (1) of the following is the power supply to the Train B Solid State Protection System (SSPS) slave relays?
Vital Bus 1 Vital Bus 2 125VDC Bus 1 125VDC Bus 2 Answer B
ExplanationlJustification:
A.
B.
C.
D.
WA Sys #
KIA System KIA Category KIA Statement 01 3 Engineered Safety Knowledge of bus power supplies to the following:
ESFASkafeguards equipment control KIA #
K2.01 WA Importance 3.6 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Memory References provided to Candidate None Technical
References:
AOP-2.38.1B page 21 item 7.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7)
Incorrect. This is the power supply to the Train A slave relays.
Correct IAW AOP-2.38.1B page 21 item 7.
Incorrect. Slave relay power is provided by AC Vital bus 1 Incorrect. Slave relay power is provided by AC Vital bus 1 Features Actuation Page 39 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 40.
The plant is operating at 100°/~ power with all systems in NSA.
In ADDITION to bus undervoltage and motor electrical protection, which of the below listed signals will directly TRIP Containment Air Recirculation Fan 2HVR-FN201A?
- 1. 2HVR-FN201A hi-hi vibration signal.
- 2. CIA signal.
- 3. CIB signal.
- 4. Safety Injection signal.
- 5. Containment Sump water level high signal.
A.
1,3, & 4 ONLY B.
1, 4, & 5 ONLY C.
2, &5ONLY, D.
2&3ONLY Answer B
Explanation/Justification:
A.
B.
C.
D.
Incorrect. Will Not trip on CIB and it will trip on sump level high.
Correct. IAW 20M-44C.l.D page 3 2"d paragraph.
Incorrect. Does not trip on CIA.
Incorrect. Does not trip on CIA or CIB.
WA Sys #
WA System KIA Category WA Statement 022 Containment Cooling Ability to manually operate and/or monitor in the control CCS fans WA #
A4.01 WA Importance 3.6 Exam Level RC' Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
20M-44C.1.D page 3 2"d paragraph.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 145.5 to 45.8) room:
Page 40 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 41.
The plant is operating at 100% power with all systems in NSA.
A large break LOCA occurs inside CNMT.
When the main generator tripped, the 2A Normal 4KV bus FAILED to transfer to the off-site power source (SSST).
CNMT pressure reaches 20 psig and is stable.
RWST level reaches 360 inches and is slowly dropping.
All ESF equipment operates as designed.
Based on these conditions, how many HHSllCharging and Recirculation Spray pumps will be DISCHARGING DIRECTLY into the reactor vessel?
HHSIICharqing Recirculation Spray A.
1 0
- 6.
1 1
C.
D.
Answer C
2 2
2 4
ExplanationlJustification:
A.
- 6.
C.
- 9.
Incorrect. EDG functioned therefore both trains of emergency power are available. If candidates believes only one train available and do NUT recognize that RWST level is below the transfer to Recirc setpoint, then they will choose this answer.
Incorrect. EDG functioned therefore both trains of emergency power are available. If a candidate believes only one train available, then the, will choose this answer.
Correct. IAW 20M-13.1.B page 3 3 paragraph.
Incorrect.. HHSI pumps are correct. However, two of the 4 Recirc spray pumps are re-aligned to inject into the core. All 4 pumps will be runr :ng.
but only 2 are injecting into the core. The other 2 continue to inject into the CNMT spray header.
KIA Sys #
K/A System K/A Category K/A Statement 026 Containment Spray Knowledge of the physical connections and/or cause-ECCS effect relationships between the CSS and the following systems:
KIA #
K1.01 K/A Importance 4.2 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Objective #:
Task ID#:
Technical
References:
10 CFR Part 55 Content:
20M-13.1.B page 3 3 paragraph.
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
Page 41 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 42.
The plant is operating at 100% power with all systems in NSA.
Recirculation Spray Cooler 2RSS*E21 A becomes INOPERABLE.
What Technical Specification actions are REQUIRED?
Entry into Technical Specification LCO:
A.
3.6.7 Condition C ONLY
- 6. 3.6.7 Condition C AND 3.6.8 Condition A ONLY C.
3.6.6 Condition A AND 3.6.7 Condition C ONLY D.
3.6.6, 3.6.8 Condition A, AND 3.6.7 Condition C.
Answer A
ExplanationlJustification:
A.
B.
C.
D.
Correct IAW Tech Spec 3.6.7 condition C Incorrect. Candidates who believe chemical addition is through the Recirc spray system will select this answer since the chem. Add subsy:.iem would also be inoperable. However, the chem. Add system injects into the QS system NOT the Recirc spray system.
Incorrect. Candidate may believe that loss of this heat exchanger also impact one train of QS. However, the QS dispersion ring is a separs e header and is NOT impacted by a loss of the RS heat exchanger.
Incorrect. Candidates who believe chemical addition is through the Recirc spray system will select this answer since the chern. Add subsy: :em would also be inoperable. However, the chem. Add system injects into the QS system NOT the Recirc spray system. Candidate may ALSC believe that loss of this heat exchanger will impact one train of QS. However, the QS dispersion ring is a separate header and is NOT imp: :ted by a loss of the RS heat exchanger.
KIA Sys #
K/A System KIA Category KIA Statement 026 Containment Spray NIA Ability to apply Technical Specification: for a system.
JA #
2.2.40 KIA Importance 3.4 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Application References provided to Candidate Tech Spec 3.6.7 condition C.
Objective #:
Tech Spec 3.6.6, 3.6.7, 3.6.8 Task ID#
Technical
References:
10 CFR Part 55 Content:
(CFR: 41. I O I43.2 143.5 / 45.3)
Page 42 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 43.
The plant is operating at 40% power with all systems in normal alignment for this mode.
0 Rod Control is in MANUAL.
A FULL load rejection occurs.
The reactor trip breakers remain CLOSED.
Tavg - Tref deviation indicates 6°F.
All systems function as designed.
Which ONE (1) of the following describes how the Steam Dump system will be operating for these cond i tions?
Steam Dump...
A.
Bank 1 will be PARTIALLY OPEN. All other Steam Dump Banks will be CLOSED.
- 6. Bank 1 will be FULL OPEN. Bank 2 will be PARTIALLY OPEN. Banks 3 & 4 will be CLOSED.
C.
Banks 1 & 2 will be FULL OPEN. Bank 3 will be PARTIALLY OPEN. Bank 4 will be CLOSED.
D.
Banks 1,2, & 3 will be FULL OPEN. Bank 4 will be PARTIALLY OPEN.
Answer A
ExplanationlJustification:
A.
- 6.
C.
D.
Correct. IAW 20M-21.5.A.12 and 13 Incorrect. Banks 3 & 4 response is correct but Bank 1 will only be partially open and bank 2 will be closed. The temperature error is not larg enough to fully open bank 1 and partially open bank 2.
Incorrect. Bank 4 response is correct but Bank 1 will only be partially open and bank 2 will be closed. The temperature error is not large enc..igh to fully open bank 1 and partially open banks 2 and 3.
Incorrect. Bank 1 will only be partially open and banks 2 & 3 will be closed. The temperature error is not large enough to fully open bank 1 a,,id partially open banks 2, 3, & 4.
J A Sys #
KIA System KIA Category KIA Statement 039 KIA #
A4.07 KIA Importance 2.8 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher An a I ys is References provided to Candidate None Technical
References:
20M-21.5.A.12 and 13.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 I45.5 to 45.8)
Main and Reheat Steam Ability to manually operate and/or monitor in the control room:
Steam dump valves Page 43 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 44.
The plant is operating at 100% power with all systems in NSA.
Which ONE (1) of the below listed failures will cause the associated Main Feed Regulating valve t 1 INITIALLY throttle CLOSED.
(Assume NO operator Action)
A.
An associated level transmitter fails HIGH.
B.
The selected steam flow transmitter fails HIGH.
C.
The selected feed flow transmitter fails LOW.
D.
The associated steam pressure transmitter fails LOW.
Answer D
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. Level is median selected. Therefore, a single failure either way will not impact MFRV operation. IF level were NOT median select this failure would cause the MFRV to throttle closed.
Incorrect. This will cause the valve to throttle OPEN.
Incorrec?. This will cause the valve to throttle OPEN.
Correct. IAW 20M-24.4.IF attachment 4 page 31 2"d NOTE. Steam pressure is used to compensate steam flow indication for density, and t ZLS the same effect as steam flow. Therefore, a pressure transmitter failing low will cause the SGWLC system to see low steam flow with respect tc. feed flow. This anticipatory signal will drive the MFRV closed in an attempt to match feed flow to the steam flow. This will be the initial response. Since the SGWLC system is level dominant, when level drops as a result of this initial response, the MFRV will be driven open again in an attemp: to restore level back to program value.
WA Sys #
KIA System WA Category WA Statement 059 Main Feedwater Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following:
WA #
K4 08 KIA Importance 2.5 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Objective #:
Task ID#
10 CFR Part 55 Content:
(CFR: 41.7)
Feedwater regulatory valve operation (c I basis of steam flow, feed flow mismatch Technical
References:
20M-24 4.IF attachment 4 page 31 2"d I OTE Page 44 of 10,)
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 45.
The plant is operating at 100% power with all systems in NSA.
SG Feed Pump 21 B Recirculation Valve 2FWR-FCVI 50B inadvertently fails full OPEN.
(1) With NO operator action, what impact will this failure have on continued power (2) IAW AOP-2.24.1, Loss of Main Feedwater, what actions will be required in response to operations?
this failure?
A.
( I ) The standby condensate pump will automatically START AND the Main feed pumps will NOT trh on (2) Start the SG Startup feed pump [2FWS-P24].
low suction pressure.
- 6.
(1) The standby condensate pump will automatically START AND the Main feed pumps will NOT trh on (2) Place the keylock switch for 2FWR-FCVI 50B to CLOSE.
low suction pressure.
C.
(1) The standby condensate pump will NOT automatically START AND one Main feed pump WILL 'rip (2) Start the SG Startup feed pump [2FWS-P24].
on low suction pressure.
D.
(1) The standby condensate pump will NOT automatically START AND one Main feed pump WILL ;rip (2) Place the keylock switch for 2FWR-FCVI 506 to CLOSE.
on low suction pressure.
Answer B
ExplanationlJustification:
A.
C.
Incorrect. Part 1 is correct. Part 2 is incorrect but plausible since the OLD version of the AOP did address starting this pump. However, for the failure of'2FWR-FCV150B in the OPEN position, starting this pump would NOT alleviate the low pressure condition.
Correct. Part 1 is correct IAW 20M-22A.4.AAD. Part 2 is correct IAW AOP-2.24.1 step 4 RNO.
Incorrect. Normal condensate pump discharge pressure for 100% power operations is just above the auto start setpoint on low pressure. The failure of 2FWR-FCV150B in the OPEN position will cause the standby condensate to reach its' auto start setpoint. Part 2 is incorrect but plausible since the OLD version of the AOP did address starting this pump. However, for the failure of 2FWR-FCV150B in the OPEN position, startins, this pump would NOT alleviate the low pressure condition.
Incorrect. Normal condensate pump discharge pressure for 100% power operations is just above the auto start setpoint on low pressure. Tbe failure of 2FWR-FCV150B in the OPEN position will cause the standby condensate to reach its' auto start setpoint. Part 2 is the correct response.
D. 6 KIA Sys #
KIA System WA Category WA Statement 059 Main Feedwater Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Failure of feedwater control system KIA #
A2.11 K/A Importance 3.0 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Higher Comprehension Technical
References:
10 CFR Part 55 Content:
AOP-2.24.1 step 4 RNO and 20M-22A.4.AAD.
(CFR: 41.5 / 43.5 I45.3 / 45.13)
Page 45 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 46.
The plant is operating at 25% power with all systems in NSA for this power level.
EDG #I is on clearance for a lube oil change-out AND maintenance has just removed 211 lube oil from the crankcase.
An inadvertent reactor trip occurs COINCIDENT with a loss of offsite power.
All SG levels Shrink to 25% NR as a result of the trip.
All systems function as designed.
No Operator actions have occurred.
Based on these conditions:
Which auxiliary feed pumps, if any, will be running A.
NO AFW pumps B.
ALL AFW pumps C.
ONLY the 6 AFW pump D.
BOTH the Steam driven AFW pump AND B AFW pump Answer D
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. Steam driven AFW pump will start on 2/3 RCP bus undervoltage BUT not on lo S/G water level. B AFW pump will start on the trif of all running main feed pump signal BUT not on lo S/G water level.
Incorrect. A AFW pump will NOT have power and therefore will NOT start.
Incorrect. Steam driven AFW pump will start on 2/3 RCP bus undervoltage BUT not on lo S/G water level.
Correct. IAW 20M-24.1.C page 5 1 paragraph and 20M-24.1.D pages 16-18. To get this correct, a student will need to know the Start sigr,als and power supplies to the AFW pumps. They will also need to know that a loss of offsite power at 25% power will cause the last running mi n feed pump to trip and cause 2/3 RCP bus undervoltage P
KIA sys #
KIA System WA Category WA Statement 061 Auxiliary/Emergency Knowledge of bus power supplies to the following:
AFW electric drive pumps WA #I K2.02 WA Importance 3.7 Exam Level RO Level Of Difficulty: (1-5)
Feedwater Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Higher Comprehension Technical
References:
10 CFR Part 55 Content:
20M-24.1.C page 5 1 paragraph. 20h 24.1.D pages 16-18 (CFR: 41.7)
Page 46 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 47.
The plant is operating at 100% power with all systems in NSA.
A Steam Generator Tube Rupture (SGTR) occurs on the B Steam Generator.
All systems function as designed.
The crew has entered E-3, Steam Generator Tube Rupture.
B Steam Generator NR level is 35% and rising.
At step 5 of E-3, the crew is attempting to isolate AFW flow to the B Steam Generator.
21 B SG AFW Throttle Valve 2FWE*HCV1 OOC will NOT close and CANNOT be closed from the control room.
IAW E-3, Steam Generator Tube Rupture step 5 what actions will be REQUIRED to isolate AFW flc JV to the B Steam Generator?
- 1. Reset the SI signal
- 3. Secure 2FWE*P23A motor driven AFW pump
- 4. Secure ZFWEP23B motor driven AFW pump A.
1, 2, and 3 B.
2, 3, and 4 C.
1 and4ONLY D.
1 and 3ONLY Answer A
xplanationlJustification:
A.
B.
C.
D.
KIA Sys #
KIA System K/A Category KIA Statement 061 Auxiliary/Emergency Ability to predict and/or monitor changes in parameters (to SG Level Correct. IAW E-3 step 5 RNO. (predict which controls will need to be operated to prevent overfilling the SG) In order to get this correct, stud ant will need to know the NSA alignment of the turbine driven AFW pump and the fact that 2FWE*HCV100C is a train A valve.
Incorrect. Plausible since this would appear to isolate ALL AFW flow however unless the SI signal is reset, the motor driven AFW pumps wil not STOP.
Incorrect. Plausible since this will isolate the B header, but the A header is still being supplied by the turbine driven pump.
Incorrect. Plausible since this will isolate the A motor driven header, but the A header is still being supplied by the turbine driven pump.
Feedwater prevent exceeding design limits) associated with operating the AFW controls including:
KIA #
Al.01 KIA Importance 3.9 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
E-3 step 5 RNO Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.5 / 45.5)
Page 47 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 48. The plant is operating in Mode 3.
All systems are in normal alignment for this mode EXCEPT Primary Component Cooling Water Pump 2CCP*P21C is racked onto the 2AE bus and is running with its control swi ch in AFTER START (Red Target).
Primary Component Cooling Water Pump 2CCP*P21A is racked onto the 2AE bus with its control switch in AFTER STOP (Green Target).
Primary Component Cooling Water Pump 2CCP*P21 B is racked onto the 2DF bus with its control switch in AFTER STOP (Green Target).
A loss of offsite power occurs and all systems function as designed.
AFTER the EDGs have completed sequentially loading all equipment, WHICH Primary Componen Cooling Water Pump(s) will be running or are REQUIRED to be manually started?
A.
ONLY 2CCP*P21 B.
B.
ONLY 2CCP*P21A AND 2CCP*P21 B.
C.
ONLY 2CCP*P21 C AND 2CCP*P21B.
D.
ALL Primary Component Cooling Water Pumps.
Answer B
ExplanationlJustification:
A.
B.
C.
- 3.
WA Sys #
KIA System WA Category WA Statement 062 Incorrect. 2CCP*21A will also start.
Correct. IAW 20M-15.1.D page 3 1 paragraph and 20M-15.1.D page 6 last paragraph.
Incorrect. 2CCP*21C will NOT start because 2CCP-21A is NOT in disconnect position on the 2AE bus. 2CCP*21A will start.
Incorrect. 2CCP*21C will NOT start because 2CCP-21A is NOT in disconnect position on the 2AE bus.
AC Electrical Distribution Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Aligning standby equipment with correc emergency power source (DIG)
WA #
A2.11 WA Importance 3.7 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension.
References provided to Candidate None Objective #:
Obj. # 22 Technical
References:
10 CFR Part 55 Content:
20M-15.1.D page 6 last paragraph (CFR: 41.5 143.5 145.3 145.13)
Page 48 of 100
- 49.
A.
B.
C.
D.
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 The plant is operating at 100% power with all systems in NSA.
Battery Charger *2-1 FAILS and its associated output breaker OPENS.
All systems function as designed.
Based on these conditions, what will be the status of 125VDC Switchboard 2-I?
ENERGIZED by 120VAC Vital Bus 1 ENERGIZED by station Battery *2-1 DE-ENERGIZED until the spare charger is installed as a replacement.
DE-ENERGIZED until Vital Bus 1 Manual Bypass Switch is placed to Bypass.
Answer B
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. #I inverter automatically receives DC power on loss of input power; it does NOT output power to the DC SWBD.
Correct. IAW 20M-39.1.B page 3 3rd paragraph Incorrect. Installing the spare charger will restore the AC power to the SWBD. However, the battery will provide 125VDC power in the inter n.
Incorrect. The battery will provide 125VDC power. Placing the Vital bus manual bypass switch to Bypass will restore AC power to a Vital bi s that failed to transfer thru its static switch.
NOTE At Unit 2 there is NO position labeled Bypass. This is a Unit 1 term. Bypass position was used in this question to avoid giving any hints to hie candidate on how to answer question #12.
K/A Sys #
K/A System WA Category WA Statement 063 DC Electrical Distribution Knowledge of the physical connections andlor cause-AC electrical system effect relationships between the DC electrical system and the following systems:
K/A #
K1.02 WA Importance 2.7 Exam Level RO Level Of Difficulty: (1-5) uestion Source:
New Question Cognitive Level:
Lower Fundamental Aeferences provided to Candidate None Technical
References:
20M-39.1.E page 3 3d paragraph Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.2 to 41.9 145.7 to 45.8)
Page 49 of 100
- 50.
A.
B.
C.
D.
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 The plant is operating at 100% power with all systems in NSA.
SI and CIA actuate.
An inadvertent reactor trip occurs COINCIDENT with a loss of offsite power.
BOTH EDGs FAIL to start and cannot be started.
The operating crew enters the appropriate Emergency Operating Procedure to address these conditions.
- 30 minutes after the reactor trip, CNMT pressure rises to 5.0 psig.
What will be the status of the following CIA components AFTER this CIA actuation?
- 1. Letdown orifice isolation valves.
- 3. CNMT Instrument Air Compressor suction isolation valves.
- 4. Non-Regen Heat exchanger Letdown inlet valve.
- 1. Closed.
- 2. Closed.
- 3. Closed.
- 4. Closed.
- 1. Closed.
- 2. Open.
- 3. Closed.
- 4. Open.
- 1. Closed.
- 2. Open.
- 3. Open.
- 4. Closed.
- 1. Open.
- 2. Open.
- 3. Open.
- 4. Open.
Answer C.
ExplanationlJustification:
A.
- 6.
C.
D.
KIA Sys #
WA System WA Category WA Statement 064 Emergency Diesel WA #
K3.02 WA Importance 4.2 Exam Level RO Level Of Difficulty: (1-5)
Incorrect. This is the position they should all be in if power is available. However, items 2 and 3 are 480V motor operated valves that w o u l ~
have already been OPENED when power was lost. Without power to close the valves, they would remain open.
Incorrect. Item 2 is a 480V motor operated valve that would have already been OPEN when power was lost. Without power to close the vi1 ve it would remain open. Item 4 is DC powered, and DC power is still available (batteries are designed for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) to close the valve.
Correct. IAW EOP attachment A-0.2 pages 6-9.
Incorrect. Items 1 & 4 are DC powered, and DC power is still available (batteries are designed for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) to close these valves. -
Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following:
ESFAS controlled or actuated system:
Generator Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Fundamental Question Cognitive Level:
Lower Technical
References:
10 CFR Part 55 Content:
EOP attachment A-0.2 pages 6-9.
(CFR: 41.7 / 45.6)
Page 50 cf 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
51 A.
B.
C.
D.
The plant is in Refueling Mode with all systems aligned for core off-load.
While lowering a spent fuel assembly into the Spent fuel pool, the assembly ruptures and releases ALL of the gases from ALL of the rods in that assembly ONLY.
NO other fuel assemblies have been damaged.
Fuel Pit Bridge Radiation Monitor 2RMF-RQ202 goes into HIGH alarm.
Based on these conditions, will Fuel Building Vent Radiation Monitor 2RMF-RQ301AIB ALSO go i-rto a HIGH alarm condition? Why or Why Not?
NO, Fuel Pit Bridge Radiation Monitor 2RMF-RQ202 is designed to detect gamma radiation (GM tune)
AND Fuel Building Vent Radiation Monitor 2RMF-RQ301 A/B is designed to detect beta radiation (scintillation).
NO, The iodine and xenon released from the fuel assembly WILL BE sufficiently scrubbed out by tt e water above the assembly.
YES, Fuel Pit Bridge Radiation Monitor 2RMF-RQ202 is designed to detect beta radiation (scintillat m)
AND Fuel Building Vent Radiation Monitor 2RMF-RQ301NB is designed to detect gamma radiatio I (GM tube).
YES, The iodine and xenon released from the fuel assembly WILL NOT BE sufficiently scrubbed 01 t by the water above the assembly.
Answer D
ExplanationlJustification:
A.
Incorrect.. The type of detectors is correct. However, if the gases released are emitting enough gamma radiation to actuate the high alarm fc ' r 2RMF-RQ-202, then there will be more than enough Xe and iodine to actuate the high alarm on 2RMF-RQ1301, Incorrect. Some iodine will be scrubbed by the 23 feet of water, but enough iodine and other gases will be present to actuate the high alarm on 2RMF-RQ1301.
Incorrect. Yes the alarm will actuate, but not because of detector types which are not correct.
Correct. IAW 20M-43.1.C page 28 AND UFSAR section 15.7.4.3. The analyzed fuel handling accident in the fuel pool will result in an offsitl' dose.
The 2RMF-RQ1301 radiation monitor will detect this release and actuate the alarms. AOP-2.49.1 for the fuel handling accident also lists bo-)
monitors as symptoms of the event.
C.
D.
7 WA Sys #
WA System KIA Category KIA Statement 073 Process Radiation Knowledge of the operational implications as they apply to concepts as they apply to the PRM system:
WA #
K5.01 WA Importance 2.5 Exam Level RO Level Of Difficulty: (1-5)
Radiation theory, including sources, typ is, units, and effects Monitoring Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Lower Fundamental.
Technical
References:
10 CFR Part 55 Content:
20M-43.1.C page 28 AND UFSAR sect )n 15.7.4.3 (CFR: 41.5 145.7)
Page 51 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 52, The plant is operating at 100% power with all systems in NSA.
Service Water Pumps 2SWS*P21A AND B are BOTH in service.
Service Water Pump 2SWS*P21C is on clearance and unavailable.
A and B Primary Plant Component Cooling Water Heat Exchangers are BOTH in set ice.
A and B Secondary Plant Component Cooling Water Heat Exchangers are BOTH in service.
A large Service water leak develops at the inlet to the A Primary Plant Component Cooling Water Heat Exchanger. The leak causes the following Service water header pressure indications:
Service Water Header Press 2SWS-PI113A is 30 psig and stable.
Service Water Header Press 2SWS-PI113B is 40 psig and stable.
(1 ) IF these Service Water Header Pressures are sustained for greater than 1 minute, what will be the impact on Secondary Plant Component Cooling Water Heat Exchanger operations?
(2) IAW AOP-2.30.1, Service Water/Normal Intake Structure Loss, what actions will be REQUIRED IF BOTH Service Water Header Pressures drop below 34 psig and cannot be restored above? 34 psig?
A.
(1) ONLY the A Secondary Plant Component Cooling Water Heat Exchanger will be ISOLATED.
(2) Manually trip the reactor and Go to E-0, Reactor Trip or Safety Injection.
B.
(1) ONLY the A Secondary Plant Component Cooling Water Heat Exchanger will be ISOLATED.
(2) Perform an emergency shutdown IAW AOP-2.51.I, Emergency Shutdown.
C.
(1) NEITHER Secondary Plant Component Cooling Water Heat Exchanger will be ISOLATED.
(2) Manually trip the reactor and Go to E-0, Reactor Trip or Safety Injection.
D.
(1) NEITHER Secondary Plant Component Cooling Water Heat Exchanger will be ISOLATED.
(2) Perform an emergency shutdown IAW AOP-2.51.I, Emergency Shutdown.
Answer C
ExplanationlJustification:
A.
Incorrect. 2SWS*MOV107A will auto close when pressure is less than 34 psig for greater than 45 seconds. However, this only isolates the i header. The 6 header will continue to supply BOTH secondary plant component cooling water heat exchangers. Manually trip the reactor i:-
correct.
Incorrect. 2SWS*MOV107A will auto close when pressure is less than 34 psig for greater than 45 seconds. However, this only isolates the A header. The 6 header will continue to supply BOTH secondary plant component cooling water heat exchangers. Manually trip the reactor i:.
required action, NOT perform an emergency shutdown. Performing an emergency shutdown is appropriate if service water cannot be restore 1 to the secondary side AND it has been restored to the primary side.
Correct. IAW AOP-2.30.1 automatic actions on page 1 & VOND 30-1 grid G-6 and 7; Part 2 IAW AOP-2.30.1 step 2 RNO e.
Incorrect. 2SWS*MOV107A will auto close when pressure is less than 34 psig for greater than 45 seconds. However, this only isolates the 1 header. The B header will continue to supply BOTH secondary plant component cooling water heat exchangers. Manually trip the reactor is required action, NOT perform an emergency shutdown. Performing an emergency shutdown is appropriate if service water cannot be restore I to the secondary side AND it has been restored to the primary side.
B.
C.
D.
9 K/A Sys #
KIA System KIA Category KIA Statement 076 Service Water Ability to (a) predict the impacts of the following Loss of sws malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
KIA #
A2.01 KIA Importance 3.5 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Jjective #:
Obi.# 5 Technical
References:
AOP-2.30.1 automatic actions on page 1 *.
VOND 30-1 grid G-6 and 7; Part 2 IAW Ai )P-2.30.1 step 2 RNO e.
10 CFR Part 55 Content:
(CFR: 41.5 143.5 14513 145113)
Page 52 of 1OCl
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 53.
The plant is operating at 100% power with all systems in NSA.
A Large Break LOCA occurs inside CNMT.
CNMT pressure rises to 35 psig.
All equipment functions as designed.
Which of the below listed components will NOW be cooled by Service Water?
- 1. CNMT Air Recirc Coolers
- 2. Charging pump lube oil coolers
- 3. Primary Plant Component Cooling Water Heat Exchangers
- 4. Recirculation Spray Heat Exchangers
- 5. Secondary Plant Component Cooling Water Heat Exchanger
- 6. Rod Control Area A/C Units A.
3, 5 : and 6 ONLY B.
1, 3, and 5 ONLY C.
I, 2,4, and 6 ONLY D.
2, and 4 ONLY Answer D
ExplanationlJustification:
A.
- 6.
\\.
Incorrect. Primary and Secondary plant component cooling water heat exchangers are isolated. Rod Control area is NOT isolated at the he ider, but the inlet MOVs will only open when temp exceeds 107°F. Even if they did open, the outlet valve is manually closed so there will be no fl )w.
Incorrect. CNMT air Recirc coolers are NOT isolated at the header. However, service water is a backup to normal cooling and MUST be m:.iually aligned. Primary and Secondary plant component cooling water heat exchangers are isolated Incorrect. CNMT air Recirc coolers are NOT isolated at the header. However, service water is a backup to normal cooling and MUST be ni>iually aligned. Rod Control area is NOT isolated at the header, but the inlet MOVs will only open when temp exceeds 107°F. Even if they did ope1 the outlet valve is manually closed so there will be no flow. Charging pump lube oil coolers and Recirc spray heat exchangers are correct.
Correct. IAW VOND 30-1 grid D-6; 30-2 grid D-I; EOP Attachment A-0.5 D.
L
=
=
KIA Sys #
KIA System KIA Category KIA Statement 076 Service Water Ability to manually operate and/or monitor in the control Emergency heat loads KIA #
A4.04 KIA Importance 3.5 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental room:
References provided to Candidate None Objective #:
Task ID#:
Technical
References:
10 CFR Part 55 Content:
VOND 30-1 grid D-6; 30-2 grid D-1 ; EO1 Attachment A-0.5 (CFR: 41.7 145.5 to 45.8)
Page 53 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 54.
The plant is operating at 100% power with all systems in NSA EXCEPT Station Air Compressor [2:;AS-C21 B] is on clearance and unavailable.
A large leak develops in the station service air header.
Station air header pressure begins to drop.
As station air header pressure continues to drop, at what setpoint will each of the below listed autolnatic actions occur:
- 1. Diesel-Driven Air Compressor 21AS-C21 - AUTOMATIC START
- 2. Condensate Polishing Air Compressor 2SAS-C22 - AUTOMATIC START
- 1. 82 psig
- 2. 90 psig
- 3. 86 psig B.
- 1. 86 psig
- 2. 90 psig
- 3. 82 psig C.
- 1. 82 psig
- 2. 86 psig
- 3. 90 psig D.
- 1. 90 psig
- 2. 86 psig
- 3. 82 psig Answer A
ExplanationlJustification:
A.
- 6.
C.
D.
Correct IAW 20M-34.2.B page 2 pressure setpoints. The candidate will need to know the sequence of starting (which one first, 2"d and 3rd, ?ut will NOT need to have these three setpoint memorized. All of these automatic actions are geared towards maintaining Instrument Air available.
Incorrect. Wrong setpoints for 21AS-C21 and 2SAS-AOV105.
Incorrect. Wrong setpoints for 2SAS-C22 and 2SAS-AOVI 05.
Incorrect. All setpoints are wrong.
~P
(
WA Category KIA Statement Air pressure m a Instrument Air WA #
A3.01 WA Importance 3.1 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
20M-34.2.B page 2 pressure setpoints Objective #:
10 CFR Part 55 Content:
(CFR: 41.7 I45.5)
Ability to monitor automatic operation of the IAS, including:
Obi.# 15 Page 54 of 100
- 55.
- 4.
B.
C.
D.
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
Which ONE (1) of the below listed components DIRECTLY receives a CIA signal to CLOSE?
HEPA Filter House No. 1 Outlet Damper 2HVS*MOD211A Pri Comp Clg Wtr Supply Hdr B lsol2CCP*MOV175-1 Control Room ACU Outside Air Intake DMPR 2HVC*MOD201A Regen HX Normal Charging Disch Vlv 2CHS*MOV310 Answer B
ExplanationlJustification:
A.
- 8.
C.
D.
KIA Sys #
K/A System K/A Category K/A Statement 103 Containment Ability to monitor automatic operation of the containment Containment isolation KIA #
A3.01 K/A Importance 3.9 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Memory References provided to Candidate None Technical
References:
EOP Attachment A-0.2 page 7 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 / 45.5)
Incorrect. This damper receives a CIA OPEN signal.
Correct. IAW EOP Attachment A-0.2 page 7 Incorrect. This damper receives a CIB signal.
Incorrect, This valve receives a SI signal.
P system, including:
Page 55 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 56.
The plant is operating at 100% power with all systems in NSA.
All Rods are indicating 228 steps on DRPI.
The following alarms and indications are THEN received in the control room:
Annunciator A4-8B, Rod Control System Non-Urgent Alarm - has ALARMED ALL 48 DRPl General Warning (GW) LED lights - Flashing.
DRPl Rod Deviation 1, R, & 2 LED lights - LIT.
DRPl Data A Failure 1, 2, & 3 LED lights - Flashing.
DRPl Central Control Failure 1, 2, & 3 LED lights - NOT LIT.
DRPl Urgent Failure 1, 2, & 3 LED lights - NOT LIT.
DRPl Data B Failure 1, 2, & 3 LED lights - NOT LIT.
All DRPl Rod Bottom (RB) lights - NOT LIT.
All Rods are STILL indicating 228 steps on DRPI.
Reactor power REMAINS at 100% and stable.
Based on these conditions:
- 1.
Rod positions will be indicated every steps.
- 2. IAW ARP A4-8B, Rod Control System Non-Urgent Alarm, the REQUIRED action is to Place the Accuracy Mode selector switch to the position.
A.
- 1. 6
- 2. A, + B B.
- 1. 6
- 2. B Only 1.12
- 2. A + B D.
- 1. 12
- 2. B Only Answer D
ExplanationlJustification:
A.
Incorrect. Indications given in the stem of the question indicate that Data A has failed. If the GVAC power to the A coils is lost, these indica ons would be present. The presence of the Non-urgent alarm results in half-accuracy mode. This means DRPl will indicate every 12 steps instc ad of every 6. Placing the switch to A+B is incorrect. The NSA position is A+B. Candidate may confuse this A+B switch with the SSPS A+B positic 11 where the NSA position is for 1 train to be in A+B and the other train is in A or B only.
Incorrect. Indications given in the stem of the question indicate that Data A has failed. If the GVAC power to the A coils is lost, these indica? ons would be present. The presence of the Non-urgent alarm results in half-accuracy mode. This means DRPl will indicate every 12 steps instc ad of every 6. Placing the switch to B ONLY is correct. The NSA position is A+B. Candidate may confuse this A+B switch with the SSPS A+B pc,ition where the NSA position is for 1 train to be in A+B and the other train is in A or B only.
Incorrect. Indications given in the stem of the question indicate that Data A has failed. If the GVAC power to the A coils is lost, these indicai-ons would be present. The presence of the Non-urgent alarm results in half-accuracy mode. This means DRPl will indicate every 12 steps insttmd of every 6. Placing the switch to A+B is incorrect. The NSA position is A+B. Candidate may confuse this A+B switch with the SSPS A+B positic 1 where the NSA position is for 1 train to be in A+B and the other train is in A or B only.
Correct. 1,AW ARP A4-8B (20M-1.4.AAK page 3) Ran on simulator to verify all indications for Data A failure.
B.
C.
D.
KIA Sys #
KIA System KIA Category KIA Statement 014 Rod Position Indication Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on those on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Loss of power to the RPIS KIA #
A2.02 KIA Importance 3.1 Exam Level RO Level Of Difficulty: (1-5)
?stion Source:
New
.erences provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Higher Comprehension Technical
References:
ARP A4-8B (20M-1.4.AAK page 3) 10 CFR Part 55 Content:
(CFR: 41.5 / 43.5 / 45.3 I45.13)
Page 56 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 57.
The plant is in Mode 3 with the RX trip breakers CLOSED and the shutdown banks withdrawn. All systems are aligned normally for this plant condition.
BOTH Source range channels are indicating 500 CPS and stable.
2013T-2.3, Nuclear Source Range Channel Test MUST be performed, for N32 ONLY before the Sti irtup can proceed.
During the performance of this surveillance, what control room actions will be REQUIRED to preve It the reactor from tripping on Source Range High Flux?
Place the N32 SR drawer:
A.
High Flux at Shutdown switch to the BLOCK position.
B.
Level Trip switch to the BYPASS position.
C.
HV Manual On/Off switch to the HV ON position.
D.
Operation Selector switch to the I O 4 CPS position.
Answer B
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. This does NOT block the High Flux trip; rather it enables the High Flux at S/D alarm.
Correct. IAW 20ST-2.3 page 37 step 7.
Incorrect. This ensures the HV power to the detector. It does NOT block the trip by holding the power to SSPS relays.
Incorrect. This injects a test signal equal to I O 4 CPS. It does NOT keep the signal from exceeding I O 4 CPS.
KIA Sys #
KIA System KIA Category KIA Statement 01 5 Nuclear Instrumentation Ability to manually operate and/or monitor in the control Trip bypasses WA #
A4.03 WA Importance 3.8 Exam Level RO Level Of Difficulty: (1-5) room:
Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Higher Comprehension Technical
References:
10 CFR Part 55 Content:
20ST-2.3 page 37 step 7.
(CFR: 41.7 / 45.5 to 45.8)
Page 57 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 58.
The plant is operating at 100% power with all systems in NSA.
The Main Turb First Stage Press Sensor Select switch is in the PM 446 position.
IF the Main Turb First Stage Press Sensor Select switch is placed in the PM 447 position, what irr :)act will this have on plant operations?
The Press 2MSS-PT447 transmitter INSTEAD of 1 A First Stage STM Press 2MSS-PT446 transmitter.
will NOW be coming from 1B First Stage STM A.
Tref signal to Steam Dumps B.
Steam Dump Load Rejection Arming signal C.
AMSAC Bypass permissive D.
Tref signal to Rod control.
Answer D
ExplanationlJustification:
A.
B.
C.
D.
K/A Sys #
K/A System K/A Category KIA Statement 016 Non-nuclear Ability to manually operate and/or monitor in the control NNI channel select controls
.A #
A4.01 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)
Jestion Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
20M-24.4.IF attachment 5 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 145.5 to 45.8)
Incorrect. This signal is not selectable.
Incorrect. This signal is always provided by the 447 transmitter.
incorrect. AMSAC bypass requires both 446 and 447 input. Not selectable.
Correct. IAW 20M-24.4.IF attachment 5 Instrumentation room:
Page 58 of 100
- 59.
A.
- 9.
C.
D.
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 Which ONE (1) of the following is NOT a source of hydrogen inside containment following a Design Bases Large break LOCA of an RCS cold leg?
Pressurizer Relief Tank gas space.
Zirc.- water reaction between the fuel clad and the reactor coolant.
Corrosion of aluminum and zinc by the ECCS water.
Radiolysis of water in the core and CNMT sump.
Answer A
ExplanationlJustification:
A.
Correct. PRT gas space gas is nitrogen NOT hydrogen. Also, any gas in the PRT gas space will remain in the PRT during a Large Cold leg LOCA. There is a common misconception that the hydrogen within the RCS will degas as it enters the PRT and therefore become a source :)f hydrogen. However, the choice as written is addressing the nitrogen overgas pressure on the PRT. Candidates may disregard this choice b [sed on the hydrogen degassing of the RCS.
Incorrect. This is a source of hydrogen in containment.
Incorrect. This is a source of hydrogen in containment.
Incorrect. This is a source of hydrogen in containment.
B.
C.
D.
KIA Sys #
K/A System WA Category KIA Statement 028 Hydrogen Recombiner and Furge Control KIA #
K5.03 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New References provided to Candidate None Technical
References:
Lesson Plan 2SQS-46.1 slide 7 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.5 145.7)
Knowledge of the operational implications of the following concepts as they apply to the HRPS:
Sources of hydrogen within containmen, Question Cognitive Level:
Lower Fundamental Page 59 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 A large break LOCA has occurred and the following plant conditions exist:
. RVLIS is available.
All RCPs are STOPPED.
The RCS is 50°F SUPERHEATED.
Which ONE (1) of the following plant conditions will REQUIRE a RED PATH entry into FR-(2.1, Response To Inadequate Core Cooling?
The two hottest core exit TC is 1250"F, ALL the other core exit TCs are 700°F AND RVLIS Full ranqe level is 33%.
The two hottest core exit TC is 1 250"F, ALL the other core exit TCs are 700°F AND RVLIS Dynam c range level is 33%.
The three hottest core exit TCs are 750"F, ALL the other core exit TCs are 700°F AND RVLIS Full range level is 33%.
The three hottest core exit TCs are 750"F, ALL the other core exit TCs are 700°F AND RVLIS Dynamic range level is 33%.
Answer C
Explanation/Justification:
A.
Incorrect. The three MAX TCs are NOT greater than 1200°F NOR are they greater than 729°F. RVLIS level is low enough to require entry 11 TCs are hot enough.
B.
Incorrect. The three MAX TCs are NOT greater than 1200°F NOR are they greater than 729°F. Wrong RVLIS range.
P.
Correct. IAW EOP status tree F-0.2.
Incorrect. The three MAX TCs are NOT greater than 1200°F. Wrong RVLIS range.
KIA Sys #
K/A System K/A Category K/A Statement 017 In-Core Temperature Monitor System (ITM) room:
RCSlRCP operation during inadequate ore Ability to manually operate and/or monitor in the control Temperature values used to determine cooling (Le., if applicable, average of fiv, highest values)
K/A #
A4.02 K/A Importance 3.8 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
EOP status tree F-0.2.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 / 45.5 to 45.8)
Page 60 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 61.
The Unit is in Mode 6. A fuel assembly is being lowered into the core.
IF the fuel assembly BINDS against another fuel assembly, downward motion of the hoist will be automatically stopped to prevent fuel assembly damage.
What manipulator crane interlock provides this protection?
A.
TubeDown B.
Underload C.
Overload D.
Bridge-Trolley-Hoist Answer B
ExplanationlJustification:
A.
B.
C.
D.
KIA Sys #
K/A System K/A Category K/A Statement 034 K/A #
K4.01 K/A Importance 2.6 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
LP 3SQS-6.13 slide 49. (2RP-3.3)
Incorrect. Tube down interlock will stop hoist downward motion when the hoist is all the way down.
Correct. IAW LP 3SQS-6.13 slide 49. (2RP-3.3)
Incorrect. Overload will stop UPWARD motion if an assembly is binding while moving upward.
Incorrect. Bridge-Trolley-Hoist interlock will only allow motionlmovement in one direction at a time.
7 Fuel Handling Equipment Knowledge of design feature(s) andlor interlock(s) which provide for the following:
Fuel protection from binding and droppi rg ijective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7)
Page 61 of 100
- 62.
A.
- 6.
C.
D.
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
The plant is operating at 100% power with all systems in NSA.
An inadvertent turbine trip occurs.
The B reactor trip breaker FAILS to OPEN.
All other systems function as designed.
Without any operator action, where will RCS temperature automatically stabilize?
541 F 547°F 550°F 554°F Answer C
ExplanationlJustification:
A.
- 6.
C.
D.
Incorrecr. This is where RCS would stabilize if it were relying on the steam dump 10-10 Tavg interlock to stop a cooldown.
Incorrec:. This is where RCS would stabilize if it were being controlled by the Rx trip controller. However, with B trip breaker still closed, tP ?
steam d\\JmpS will function on the load rejection controller which has a 3°F deadband before it will open the steam dumps.
Incorreci.. This is where RCS would stabilize if it were relying on the SG safeties to control temperature. This would be necessary if the stea n dumps were NOT armed. However, the A reactor trip breaker opening will arm the dumps.
U P
K/A Sys #
K/A System WA Category KIA Statement 04 1 Steam DumplTurbine Knowledge of the Physical connections andlor cause-RCS E3ypass Control effect relationships between the SDS and the following systems:
YIA #
K1.05 K/A Importance 3.5 Exam Level RCI Level Of Difficulty: (1-5) estion Source:
New Question Cognitive Level:
Higher Comprehension
.,eferences provided to Candidate None Objective #:
Task ID#:
Technical
References:
20M-21.5.A.12.
10 CFR Part 55 Content:
(CFR: 41.2 to 41.9 145.7 to 45.8)
Page 62 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 63.
The plant is operating at 100% power with all systems in NSA.
Condensate Bypass Vlv 2CNM-AOVI 00 inadvertently OPENS.
What effect will this have on plant operations?
A.
Feedwater inlet Temperature to the Steam Generators will DROP.
B.
Condenser hotwell level will RISE.
C.
Main feed Pump Suction pressure will DROP.
D.
Turbine Plant Demineralized Water Storage Tank will RISE.
Answer A
ExplanationlJustification:
A.
B.
C.
D.
K/A Sys #
K/A System K/A Category K/A Statement 056 Condensate Knowledge of the physical connections and/or cause-MFW effect relationships between the Condensate System and the following systems:
Correct. IAW VOND 22A-2 grid B-5 Incorrect. This would be true if the bypass around the normal LCV was failed open. (LCV103).
Incorrect. This would be true if the condensate pump Recirc valve was failed open. (FCVIOI).
Incorrect. This would be true if the bypass around the normal condensate pump reject MOV was failed open. (LCVIOI)
.~
WA #
K1.03 WA Importance 2.6 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
VOND 22A-2 grid B-5 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.2 to 41.9 I45.7 to 45.8)
Page 63 of 100
64 B.
C.
D.
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 Which ONE (1) of the below listed set of conditions are the MINIMUM REQUIRED conditions to ac uate annunciator A I 2-4C, Condenser Unavailable?
2 out of 2 Condenser Pressure transmitters below 19.5 of Hg vacuum 4 out of 4 Circulating Water Pumps NOT running OR 2 out of 2 Condenser Pressure transmitters above 19.5 of Hg vacuum 3 out of 4 Circulating Water Pumps NOT running OR 1 out of 2 Condenser Pressure transmitters below 19.5 of Hg vacuum 4 out of 4 Circulating Water Pumps NOT running OR 1 out of 2 Condenser Pressure transmitters above 19.5 of Hg vacuum 3 out of 4 Circulating Water Pumps NOT running OR Answer C
ExplanationlJustification:
A.
Incorrect. This will actuate the alarm, but it is NOT the MINIMUM required conditions. It only takes %transmitters. Common Validator misconception based on the wording of the alarm window at Unit 2. The window alerts the operators to conditions that make the condense!
UNAVAILBLE, all of the logic diagrams and training are geared towards what it takes to make the condenser AVAILABLE.
Incorrect. It only takes % transmitters, vacuum is below NOT above. Also MUST have 4/4 circ pumps OFF.
Correct 20M-26.4.ABM page 3 setpoints.
Incorrect. Vacuum IS below NOT above, Also MUST have 4/4 circ pumps OFF.
B.
C.
D.
K/A Sys #
K/A System WA Category WA Statement 075 Circulating Water Knowledge of circulating water system design feature(s)
Heat sink K/A #
K4.01 WA Importance 2.5 Exam Level RO Level Of Difficulty: (1-5) and interlock(s) which provide for the following:
Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Lower Memory Technical
References:
10 CFR Part 55 Content:
20M-26.4.ABM page 3 setpoints.
(CFR: 41.7)
Page 64 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 65.
The plant is operating at 100% power with all systems in NSA.
10 Ton C02 Storage Tank 2FPD-TK22 MUST be removed from service for maintenance.
10 Ton C02 Storage Tank 2FPD-TK23 is able to supply C02 to the System 2 Zones.
24 Ton C02 Storage Tank 2FPD-TK24 is able to supply COP to the System 2 Zones.
IAW 20M-33.4.G, C02 Fire Protection System Startup And Storage Tank Fill, how will the C02 sys em be re-aligned to maintain operability of the system?
Align the (1 1 for service, then place the Smoke Detection Panel 2FPS-PNL-XL3 MAIN/RESERVE switch to (2)
, AND isolate (3)
A.
B.
C.
D.
(1 ) 24 Ton C02 Storage Tank 2FPD-TK24 (2) MAIN (3) ONLY 10 Ton C02 Storage Tank 2FPD-TK22 (1 ) 24 Ton C02 Storage Tank 2FPD-TK24 (2) RESERVE (3) BOTH 10 Ton C02 Storage Tanks 2FPD-TK22 & 23 (1 ) 10 Ton C02 Storage Tank 2FPD-TK23 (2) MAIN (3) ONLY 10 Ton C02 Storage Tank 2FPD-TK22 (1 ) 10 Ton C02 Storage Tank 2FPD-TK23 (2) RESERVE (3) BOTH COP Storage Tanks 2FPD-TK22 & 24 nswer B
ExplanationlJustification:
A.
B.
C.
D.
WA Sys #
K/A System WA Category WA Statement 086 Fire Protection Ability to predict and/or monitor changes in parameters (to FPS lineups Incorrect. Switch must be placed to reserve for 24 ton tank master valve to function. All other items are correct.
Correct IAW 20M-33.4.G page 25 step 5. This is NOT minutia, rather it tests the candidates ability to predict what line-up changes are ne1 d to prevent operating the C02 outside of the required alignment.
Incorrect. 24 ton unit must be aligned for service, Smoke Detection Panel 2FPS-PNL-XL3 MAlNlRESERVE switch must be placed to resei Je, and BOTH 10 ton units must be isolated.
Incorrect. 24 ton unit must be aligned for service, items 2 and 3 are correct.
prevent exceeding design limits) associated with Fire Protection System operating the controls including:
WA #
AI.05 WA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
20M-33.4.G page 25 step 5 Rev. 10.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.5 I45.5)
Page 65 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 66.
The plant is in Mode 5 preparing to enter Mode 4.
Valve alignments are being performed on a Safety-Related system.
The REQUIRED NSA position of a manually operated globe valve is 2 Turns OPEN.
The valve must be in this position PRIOR to Mode 4 entry.
The valve has MINIMAL safety significance.
The valve list REQUIRES Concurrent verification for this valve.
The second verifier will receive 5 mR performing the Concurrent verification.
The valve has NO remote valve indication.
The valve CANNOT be verified in the correct position by the performance of a functional test.
e IAW the guidance provided in 1/20M-48.3.D, Administrative Control Of Valves And Equipment, hov, will the Concurrent verification for this valve be addressed?
A.
The Shift Manager shall waive the Concurrent verification for this valve based on MINIMAL safety significance and HIGH radiation exposure to the second verifier.
B.
The First verifier places the valve in the required position; WITHIN 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the second verifier verifit-s the valve in the required position.
C.
The First verifier places the valve in the required position; the second verifier remains OUTSIDE the line of sight of the first verifier THEN verifies the valve in the required position.
D.
The First verifier places the valve in the required position WHILE the second verifier observes the first verifier placing the valve in the required position.
Answer D
ExplanationlJustification:
Incorrect. The shift manager may waive the independent verification of a safety related valve if it has minimal safety significance and will re! ult in 10mR exposure to the second verifier. This valve only has 5 mR exposure. Also since this valve requires a number of turns, the only individ ial that can waive the concurrent verification is the operations manager.
Incorrect. These are the requirements for independent verification of Tech Spec related actions that support current plant conditions. Since his valve is required for Mode 4 entry, it is NOT required for the current plant Mode.
Incorrect. These are the requirements for independent verification NOT concurrent verification. Additionally, this valve must be concurrently verified since independent verification would negate the original condition.
Correct. IAW 1/20M-48.3.D 1II.C and VI.A.9.a. The valve requires concurrent verification and it cannot be waived by the shift manager.
Concurrent verification is specifically defined for valves that require a number of turns. This definition specifically states that the second veri'ier will observe the original manipulation.
- 6.
C.
D.
KIA Sys #
WA System KIA Category KIA Statement NIA Generic Conduct Of Operations Knowledge of how to conduct system lir <?ups, W A #
2.1.29 WA Importance 4.1 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
IAW 1/20M-48.3.D 1II.C and VI.A.9.a.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41. I O I45.1 145.12) such as valves, breakers, switches, etc.
Page 66 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 67.
You are a Licensed Reactor Operator at Beaver Valley.
You have been Licensed for Five and one-half years.
Your License renewal medical examination (NRC Form 396, Certification Of Medical Examination By Facility Licensee) is due to the NRC Regional Administrator in 6 months.
Your License is Active and you are currently assigned as the Unit 2 Reactor Operator.
Your License contains NO medical restrictions.
You have been experiencing some difficulties with your distant vision.
0 I(
On your first relief day, your personal physician (a licensed optometrist) determines that your distant vision has permanently degraded and you will NOW be required to wear corrective lenses at all times.
IAW IOCFR 50.74, Notification of Change In Operator or Senior Operator Status, when are you are REQUIRED to notify the NRC Regional Administrator of this change in your medical status?
A.
Immediately.
B.
Prior to assuming your next shift.
C.
Within 30 days of the diagnosis.
D.
Within 60 days of the diagnosis.
Answer C
ExplanationlJustification:
A.
B.
Incorrect. You should immediately begin wearing the corrective lenses, but not required to report for 30 days.
Incorrect. You should begin wearing the corrective lenses prior to your next shift, but not required to report for 30 days.
Correct. IAW IOCFR 50.74, 55.25, 55.23, NRC form 396. (Beaver Valley specific OE CR 07-2231 1)
Incorrect. Must be within 30 days.
KIA Sys #
K/A System K/A Category K/A Statement NIA Generic Conduct Of Operations Knowledge of individual licensed opera )r responsibilities related to shift staffing, : iJch as medical requirements, no-solo ope.ation, maintenance of active license status, 1 OCFR55, etc.
KIA #
2.1.4 K/A Importance 3.3 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New References provided to Candidate None Technical
References:
IOCFR 50.74, 55.25, 55.23, NRC form 96 Question Cognitive Level:
Lower Fundamental Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41. I O 143.2)
Page 67 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 68.
The plant is operating at 90% power with all systems in NSA.
e Control Bank D is at 229 steps.
Core Burnup is 3000 MWD/MTU.
RCS Boron Concentration is 1250 ppm.
Equilibrium Xenon concentration conditions exist.
Tavg is equal to Tref.
Q How many gallons of dilution water will be needed to raise power to 95% and keep Tavg equal to Tref?
Assume the Boron Correction factor is 1.O and disregard any changes in Xenon concentration.
A.
20 gals.
- 6.
420 gals.
C.
520 gals.
D.
720 gals.
Answer C
ExplanationlJustification:
A.
B.
C.
Incorrect. If the candidate makes a math error and stops after determining the change in boron concentration, they will choose this answer.
Incorrect. If the candidate does all of the calculations correctly and but sloppily uses the correct nomograph for dilution, they will choose thi:
answer. Sloppy use of the nomograph means to inaccurately apply the straight edge to the nomograph.
Correct. CB-28 3000 md/mtu equals boron worth of -6.0 pcm/ppm. CB-21 1250 ppm power defect for 90-95% is 90 pcm. 90pcm/-6pcm/ppn, equals -15 ppm. Must reduce RCS boron 15 ppm to compensate for power defect. Using CB-33 nomograph determine volume of water nee.led
-520 gals.
NOTE: Alternate method of using CB-33 formula 8069FT3 I.02264FT3/lbm/8.33 X In(Ci/Cf)= 516 gals. Candidate will NOT have the t.,ble that specifies the volume of the RCS and will therefore need to use the nomograph to determine the answer.
Incorrect. If the candidate does all of the calculations correctly but DOES NOT realize the correct nomograph for dilution is NOT linear, they will choose this answer.
WA Sys #
KIA System KIA Category KIA Statement NIA Generic Conduct Of Operations Ability to use procedures to determine tl e effects on reactivity of plant changes, si :h as reactor coolant system temperature, secondary plant, fuel depletion, etc.
KIA #
2.1.43 KIA Importance 4.1 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate Objective #:
Task ID#:
I O CFR Part 55 Content:
(CFR:
41.10 / 43.6 / 45.6)
Curve book curves CB-28, 21, &
33 Technical
References:
Curve book curves CB-28, 21, & 33 Page 68 of 100
69 H.
- 9.
C.
D.
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
What is the Technical Specification basis for the Reactor Core Safety Limit?
There must be a least a 95% probability at a 95% confidence level that the :
Hot fuel rod in the core does not experience DNB or centerline fuel melting.
Integrity of the Reactor Coolant System will be protected against overpressurization.
Core will be protected against rapid increases in neutron flux.
Maximum clad oxidation does not exceed 17% of clad thickness.
Answer A
ExplanationlJustification:
A.
B.
C.
D.
KIA Sys #
K/A System K/A Category K/A Statement N/A Generic Equipment Control Knowledge of the bases in Technical Correct. IAW Tech Spec bases 2.1.1 page B 2.1.1-2 Incorrect. This is the bases for the other Tech spec Safety limit. Plausible since limiting heat input would limit the possibility of RCS overpressurization, however this is NOT the bases for the core safety limit it is the bases for the RCS pressure safety limit.
Incorrect. This is the Tech Spec bases for the high positive rate trip. Setpoint.
Incorrect This is an ECCS acceptance criteria NOT the Tech Spec bases for the core safety limit.
~
Specifications for limiting conditions for operations and safety limits.
K/A #
2.2.25 K/A Importance 3.2 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Memory References provided to Candidate None Technical
References:
Tech Spec bases 2.1.1 page 6 2.1.1-2 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.5 / 41.7 / 43.2)
Page 69 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 70.
Refer to the drawing of a typical valve control circuit for a 480 VAC motor-operated valve (see figurv below).
With NO initiating condition present, the valve is currently OPEN. If the S I pushbutton is depressed the valve wi I I and when the S I pushbutton is subsequently released the valve will TYPICAL VALVE CONTROL CIRCUIT A.
remain open; remain open B.
close; remain closed C.
remain open; close D.
close; open Answer B
ExplanationIJustification:
A.
B.
C.
D.
K/A Sys #
KIA System K/A Category K/A Statement NIA Generic Equipment Control Ability to obtain and interpret station elet ':rical and mechanical drawings.
K/A #
2.2.41 K/A Importance 3.5 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
BVPS Bank Question 13933 Question Cognitive Level:
Higher Analysis References provided to Candidate None Technical
References:
Print reading skills Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.10 145.12 145.13)
Incorrect. Wrong initial response; wrong subsequent response.
Correct. Right initial response; right subsequent response.
Incorrect. Wrong initial response; right subsequent response.
Incorrect. Right initial response; wrong subsequent response.
Page 70 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 71.
You have been assigned the task of venting a radioactive system that is located in a Locked High Radiation Area (LHRA).
When you open the vent valve you receive an UNEXPECTED dose rate alarm on your electronic alarming dosimeter (EAD).
IAW NOP-WM-7025, High Radiation Area Program, what are your Required actions for these conditions?
A.
Immediately notify Radiation Protection (RP) and stay in the area to await further instructions.
B.
Close the vent valve and report the alarm to the control room supervisor and Radiation Protection (RP).
C.
Immediately exit the area and perform whole body frisk.
D.
Close the vent valve and immediately exit the area.
Answer D
Explanation/Justification:
A.
- 6.
C.
D.
WA Sys #
WA System K/A Category K/A Statement N/A Generic Radiation Control Knowledge of radiological safety princip 9s Incorrect. These are the correct actions personnel contamination.
Incorrect. These would be appropriate actions for an alarming air monitor.
Incorrect. Frisking is required before exiting the RCA but not necessarily required as part of LHRA exit.
Correct. IAW NOP-WM-7025 step 4.2.12 on page 6 and 7.
pertaining to licensed operator duties, s ch as containment entry requirements, fuel hasidling responsibilities, access to locked high-radiation areas, aligning filters, etc.
KIA #
2.3.12 K/A Importance 3.2 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Lower Fundamental Technical
References:
10 CFR Part 55 Content:
NOP-WM-7025 step 4.2.12 on page 6 a,d 7.
(CFR: 41.12 145.9 / 45.10)
Page 71 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 72.
What type of radiation detector is used in the In-Containment High Range Area Radiation Monitors 2RMR*RQ206(207) AND WHY is this type detector used for this application?
A Proportional; provide early warning of the presence of radiation.
B.
Geiger-Mueller; provide early warning of the presence of radiation.
C.
Ion Chamber; will not saturate in high radiation fields D.
Beta/Gamma Scintillator; will not saturate in high radiation fields Answer C
ExplanationlJwstification:
A.
B.
C.
D.
Incorrect. This is the type of detector used in the source range instrument where it is important to monitor the presence of a small populatioi of neutrons and what the population is doing with respect to time..
Incorrect. This is the type of detector used in most area monitors, however in order to meet the Reg. guide 1.97 criteria for post accident mc nitors, Ion chambers are needed to avoid saturating the detector from the extremely high radiation fields that the monitors are designed to detect.
Correct. IAW 20M-43.1.C page 51. In order to meet the Reg. guide 1.97 criteria for post accident monitors, Ion chambers are needed to av rid saturating the detector from the extremely high radiation fields that the monitors are designed to detect.
Incorrect. These are the type detectors used in the process radiation monitoring system. They are not gas field tubes therefore they will NO saturate in high fields.
KIA Sys #
KIA System KIA Category KIA Statement NIA Generic Radiation Control Knowledge of radiation monitoring systc ns, such as fixed radiation monitors and a12 ms, portable survey instruments, personnel monitoring equipment, etc.
KIA #
5.3.15 K/A Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)
Memory Question Source:
New Question Cognitive Level:
Lower References provided to Candidate None Technical
References:
20M-43.1.C page 51 jective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.12 143.4 145.9)
Page 72 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
- 73.
The plant is in Mode 4 with a plant shutdown in progress.
RHS has just been placed in service.
RHR Pump 2RHS*P21A is in service.
RHR Pump 2RHS*P21 B is out of service.
PRZR level is 15% and stable.
21C RCP is in service.
Annunciator AI-5H Residual Heat Removal System Trouble (RHR TRN A FLW RHS*FT605A LOW computer address point F0600D) - Alarms The following control room indications NOW exist:
RHR Train A Flow [2RHS-F1605A] is oscillating between 0 and 1400 gpm.
21A RHR Pump Amps [2RHS-l12IA] are erratically oscillating.
21A RHR HX Bypass Vlv [2RHS*FCV605A] is erratically oscillating.
21 C RCP Amp [2RCS-I 121 C] indicates 688 amps and stable.
PRZR level remains at 15% and stable.
In order to address these conditions, what procedure are you Required to enter?
A.
AOP-2.6.8, Abnormal RCP Operation C.
AOP-2.10.1, Residual Heat Removal System Loss
- 9.
AOP-2.10.2, Loss of RHS While At Reduced Inventory/Midloop Conditions Answer C
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. This procedure entry would be appropriate if the RHR system was displaying these symptoms due to a loss of inventory. There at ? no indications that a loss of inventory is progress, PRZR level is 15 and stable.
Incorrect. Entry into the procedure would be appropriate if the RCP was displaying the erratic amps and flow. Since the RHP pump is displaiing the erratic amps and flow and RCP amps are stable entry into this procedure is NOT appropriate or required.
Correct. IAW symptoms listed in AOP-2.10.1.
Incorrect. Although all of the symptoms listed in the stem are symptoms in this AOP also, you must also be at reduced inventory or midloop iefore entry is required. With PRZR level stable at 15%, the plant is NOT at reduced inventory OR midloop.
K/A Sys #
K/A System KIA Category K/A Statement NIA Generic Emergency ProcedureslPlan K/A #
2.4.4 K/A Importance 4.5 Exam Level RO Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Ability to recognize abnormal indication: for system operating parameters that are e try-level conditions for emergency and abnl rmal operating procedures.
Level Of Difficulty: (1-5)
Question Cognitive Level:
Higher Comprehension Technical
References:
10 CFR Part 55 Content:
Symptoms listed in AOP-2.10.1, (CFR: 41.10 I43.2 145.6)
Page 73 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 74.
Which of the below listed Abnormal Operating Procedures contain Immediate Operator Actions?
AOP-2.1.3, RCCA Control Bank Inappropriate Continuous Movement AOP-2.1.7, Rod Position Indication Malfunction AOP-2.1.8, Rod Inoperability AOP-2.6.4, Steam Generator Tube Leakage AOP-2.24.1, Loss of Main Feedwater AOP-2.26.1, Turbine and Generator Trip AOP-2.36.1, Loss of All AC Power When Shutdown AOP-2.36.2, Loss of 4KV Emergency Bus A.
1, 2, 6, 7, & 8 ONLY B.
2, 3, 4, 5, & 7 ONLY C.
1, 3,6, & 8 ONLY D.
3, 4, 5, & 7 ONLY Answer C
ExplanationlJustification:
A.
B.
C.
D.
WA Sys #
WA System WA Category WA Statement w IA Generic Emergency Procedures/Plan Knowledge of abnormal condition proce iures.
9#
2.4.1 1 WA Importance 4.0 Exam Level RO Level Of Difficulty: (1-5) duestion Source:
New Question Cognitive Level:
Lower Memory References provided to Candidate None Technical
References:
AOPs 2.1.3, 2.1.8, 2.26.1, & 2.36.2.
Objective #:
10 CFR Part 55 Content:
(CFR: 41. I O I 43.5 145.13)
Incorrect 2 and 7 do NOT have IMAs; 3 does Incorrect 3 is the only one with IMAs.
Correct. IAW AOPs 2.1.3, 2.1.8, 2.26.1, & 2.36.2.
Incorrect 4, 5, & 7 do NOT have IMAs; 1, 6, & 8 have IMAs.
Obj. # I Page 74 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1
- 75.
The plant is operating at 25% power with all systems in normal alignment for this power level.
A Steam Generator Tube Rupture occurs.
RCS pressure slowly drops to the Low PRZR pressure reactor trip setpoint.
Ruptured SG NR level is 20% and slowly rising.
The BOP operator wishes to pre-emptively isolate feed flow to the ruptured SG.
IAW the guidance provided in 1/20M-53B.2, User's Guide, how will this pre-emptive action be accomplished?
The BOP operator is REQUIRED to:
A.
Complete the Immediate actions of E-0, Reactor or Safety Injection, THEN obtain concurrence from the SM/US, THEN isolate feed flow to the ruptured SG.
B.
Complete the Immediate actions of E-0, Reactor or Safety Injection, THEN isolate feed flow to the ruptiired SG, THEN at the first crew brief inform the SM/US of the preemptive actions taken.
C.
Isolate feed flow to the ruptured SG, THEN complete the Immediate actions of E-0, Reactor or Safe y Injection, THEN at the first crew brief inform the SM/US of the preemptive actions taken.
D.
Obtain concurrence from the SM/US, THEN isolate feed flow to the ruptured SG, THEN complete the Immediate actions of E-0, Reactor or Safety Injection.
Answer A
ExplanationlJustification:
A.
Correct. IAW 1/20M-53.B.2 item 10 on page 7. Preemptive actions can only be performed after completing the IMAs and after obtaining Sh /US concurrence.
Incorrect. Must obtain permission first. This is the requirements for any automatic action that failed to occur.
Incorrect. This would be the appropriate response to completing an automatic action that failed to occur EXCEPT the actions were completrtd out of order Incorrect. Completing the lMAs MUST is accomplished first.
C.
D.
K/A Sys #
K/A System KIA Category KIA Statement N/A
(;en e ri c Emergency Procedures/Plan Knowledge of crew roles and responsit dies KIA #
2.4.13 K/A Importance 4.0 Exam Level RO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
1/20M-53.8.2 item 10 on page 7 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41. I O / 45.12) during EOP usage.
Page 75 oi 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 SRO ONLY
- 76.
The plant is operating at 100% power with all systerns in NSA.,D,~>~\\
&.-\\efi+ sop A LOCA occurs coincident with a loss of offsite power..
The crew has entered procedure ECA-1.I, Loss Of Emergency Coolant Recirculation due to the inability to verify cold leg recirculation capability.
At step 13 the crew is attempting to perform an RCS cooldown to Mode 5 at 1 OO"F/hr.
All systems respond as designed EXCEPT the ERFS.G ailed to Automatically start.
IF the RCS cooldown cannot be manually established from the control room:
(1) What local actions will be REQUIRED to perform the cooldown?
(2) What would be the consequences of NOT performing these actions?
AND Direct Local operators to:
A.
(1) Open SG Atm steam Dump Valves [ZSVS*PCVlOlA(B)(C)] IAW EOP Attachment A-1.1 1, Manual (2) RCS depressurization will NOT be permitted and the time to RWST depletion will be shortened.
Handpump Operations Of Hydraulically Actuated Valves.
B.
(1) Open SG Atm steam Dump Valves [ZSVS*PCV'IOlA(B)(C)] IAW EOP Attachment A-1.1 1, Manud (2) RCP seal integrity will be lost and the core will eventually uncover.
Hand pum p Operations Of H yd raul ical I y Actuated Valves.
C.
(I) Perform EOP Attachment A-I.18, ERFS Diesel Generator Startup THEN start the Station and CPJMT (2) RCS depressurization will NOT be permitted and the time to RWST depletion will be shortened.
air compressors.
D.
(1) Perform EOP Attachment A-I.18, ERFS Diesel Generator Startup THEN start the Station and CflMT (2) RCP seal integrity will be lost and the core will eventually uncover.
air compressors.
Answer A
ExplanationIJustification:
A.
B.
C.
D.
Correct. Action directed by ECA-1.1 step 13 and ECA-1.1 bases page 3 item 3, the cooldown is being done to allow RCS depressurization t 1 limit breakflow and prolong the time to RWST depletion.
Incorrect. Correct action but incorrect consequence for not completing the action. These are the consequences for a complete loss of ALL P ;.
However, in this question EDG power is still available.
Incorrect These would be correct actions if the SG Atm steam Dump Valves were air operated valves. These valves are hydraulic valves. C irrect consequence.
Incorrect These would be correct actions if the SG Atm steam Dump Valves were air operated valves These valves are hydraulic valves. Tt ese are the consequences for a complete loss of ALL AC. However, in this question EDG power is still available.
KIA Sys #
WA System KIA Category KIA Statement 00001 1 Large Break LOCA NIA Knowledge of local auxiliary operator ta: ks
=
during an emergency and the resultant operational effects.
KIA #
2.4.35 K/A Importance 4.0 Exam Level SRQ Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate None Technical
References:
ECA-1.I bases, EOP Att. A-1.1 1 Objective #:
Task ID#:
I O CFR Part 55 Content:
(CFR: 41.10 I43.5 / 45.13)
Page 76 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 77.
The plant is operating at 26% power with all systems in normal alignment for this power level Annunciator A2-4D Reactor Coolant Pump Seal Trouble is in alarm (computer address RCP 21 B SEAL LK OFF HIGH).
RCP21 B No. 1 seal leakoff flow indicates >6 gpm (off-scale high).
RCP21B No. 2 seal leakoff flow is less than 0.1 GPM.
VCT pressure is 26 psig.
RCP21 B seal injection flow is 9.5 gpm and stable.
RCP seal return temperature is 156°F and stable IAW AOP 2.6.8, Abnormal RCP Operation, which ONE (1) of the following actions and sequence of actions are you REQUIRED to direct the crew to perform?
A.
Stop RCP 21 B, THEN shut the Seal Water Leakoff Vlv [2CHS*MOV303B] within 3 - 5 minutes of securing the pump. THEN initiate an Emergency Shutdown to Hot Standby in accordance with AOP 2.51.I, Emergency Shutdown.
B.
Trip the reactor and go to E-0, Reactor Trip or Safety Injection. Complete the immediate actions of E.-0 THEN Stop RCP 21 B, THEN shut the Seal Water Leakoff Vlv [2CHS*MOV303B]
within 3 - 5 minutes of securing the pump.
C.
Monitor seal return temperature, THEN maintain seal injection flow to RCP 21 B greater than 9 gpm.
THEN initiate an Emergency Shutdown to Hot Standby in accordance with AOP 2.51.I, Emergency Shutdown.
J.
Trip the reactor. THEN Stop RCP 21 B and go to E-0, Reactor Trip or Safety Injection. Complete the immediate actions of E-0, THEN shut the Seal Water Leakoff Vlv [2CHS*MOV303B]
within 3 - 5 minutes of securing the pump.
Answer B
ExplanationlJustification:
A.
B.
C.
D.
K/A Sys #
KIA System K/A Category K/A Statement 00001 5/17 RCP Malfunctions Incorrect. Wrong Sequence and wrong procedural guidance.
Correct. IAW AOP-2.6.8 step 2.9 RNO Incorrect. Wrong actions and wrong procedural guidance.
Incorrect. Wrong sequence of correct actions.
=
Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):
When to secure RCPs on loss of coolin( or seal injection K/A #
AA2.10 K/A Importance 3.7 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
BVPS Unit 1 Bank (1LOT7 Audit Exam)
Question Cognitive Level:
Higher Comprehension References provided to Candidate None Objective #:
Task ID#:
Technical
References:
10 CFR Part 55 Content:
AOP-2.6.8 step 2.9 RNO, 20M-7.4.AAt (CFR 43.5 / 45.13)
Page 77 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 78.
The plant is operating at 100% power with all systems in NSA EXCEPT:
The control switch for PZR PORV 2RCS*PCV455C is in the CLOSE position and its associaed block valve is closed and de-energized.
While performing 20ST-6.4, Measurement Of Seal Injection Flow, TOTAL RCP seal injection flow i:
discovered to be 33 gpm.
IMMEDIATELY after this discovery:
Pressurizer pressure transmitter 2RCS*PT444 fails HIGH.
NO operator actions have been taken.
PRIOR to any reactor trips occurring, how will this failure impact RCP seal injection flow?
RCP seal injection flow will (1)
IF a small break LOCA were to NOW occur, the amount of ECCS flow that WOULD BE diverted frotn the ECCS injection path will be (2) the range assumed in the safety analysis.
A.
B.
C.
D.
(1 ) increase (2) within (1 ) increase (2) outside of (1) decrease (2) outside of (1 ) decrease (2) within Answer B
ExplanationlJustification:
A.
B.
Incorrect. Correct impact on seal injection flow. Incorrect impact on assumed ECCS flow.
Correct. RCS pressure will drop, which allows seal injection flow to increase. However since the Total RCP seal injection flow is 33 gpm an I above the LCO limit of 28 gpm, ECCS injection flow will be outside the values assumed in the accident analysis. In order to obtain the corrc i:t answer, the student will need to know the TS LCO for seal injection, the bases for the limit, the plant response to PT444 failing high and tht control scheme for PCV455C. This is SRO material since they must recognize that 33 gpm is above the TS allowable LCO value of 28 gpn and they must realize the impact on ECCS flow during an accident..
Incorrect. incorrect impact on seal injection flow. Correct impact on assumed ECCS flow.
Incorrect. Incorrect impact on seal injection flow. Incorrect impact on assumed ECCS flow.
C.
D.
KIA Sys #
WA System KIA Category K/A Statement 000027 Pressurizer Pressure Ability to determine and interpret the following as they RCP injection flow Control System Malfunction apply to the Pressurizer Pressure Control Malfunctions:
Level Of Difficulty: (1-5)
K/A #
AA2.14 KIA Importance 2.9 Exam Level SRO Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate None Technical
References:
Technical Specification 3.5.5 and basc s; 20M-6.4.IF Attachment 2 Objective #:
Task ID#:
30 CFR Part 55 Content:
(CFR: 43.5 / 45.13 I43.b.l)
Page 78 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 79.
The plant is operating at 100% power with all systems in NSA.
A Steam Generator Tube Rupture occurs on the B Steam Generator.
All systems function as designed.
The crew has just entered E-3, Steam Generator Tube Rupture.
A and C Steam Generators are intact.
The following plant conditions exist:
Total AFW flow is 900 gpm and stable.
CNMT pressure is 1 psig and stable.
A and C NR SG level are 0%.
B NR SG level is 5% and rising.
The reactor operator requests permission to perform pre-emptive actions and isolate all AFW flow to the B Steam Generator.
IAW the guidance contained in E-3, Steam Generator Tube Rupture, what direction are you REQUIRED to give the reactor operator AND what is the bases for this direction?
A.
Isolate feed flow to the B Steam Generator, the required heat sink will be maintained by A and C:
B.
Isolate feed flow to the B Steam Generator, B Steam Generator overfill must be avoided to limit \\he radiological consequences.
Continue feeding the B Steam Generator until NR level is >12%, B Steam Generator is required for a heat sink.
D.
Continue feeding the B Steam Generator until NR level is >12%, B Steam Generator tubes muse remain covered to avoid SG depressurization.
Answer D
ExplanationlJustification:
A.
B.
C.
D.
WA Sys #
KIA System K/A Category KIA Statement 000038 Steam Generator Tube NIA Knowledge of the specific bases for E )Ps.
Incorrect. Feed flow is NOT to be isolated UNLESS NR of 212% has been reached (pre-emptive requirement of EOP users guide page 7 iem 10 4Ih bullet) Heat sink requirement is correct.
Incorrect.. Feed Row is NOT to be isolated UNLESS NR of >12% has been reached (pre-emptive requirement of EOP users guide page 7 item 10 4Ih bullet), SG overfill is a concern but not at the expense of allowing a ruptured SG to depressurize.
Incorrect. Correct direction, however B SG will not be needed as a heat sink with A and C intact.
Correct. IAW E-3 step 5 bases page 65 2nd bullet.
Rupture (SGTR)
WA #
2.4.18 K/A Importance 4.0 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
E-3 step 5 bases page 62 2nd bullet.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.10 I43.1 I45.13)
Page 79 of 100
I 80 A.
B.
C.
- 0.
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 SRO ONLY The plant is operating at 100% power with all systerns in NSA.
. A Reactor trip occurs coincident with a loss of offsite power.
The steam driven AFW pump failed to start and cannot be started.
All other plant equipment responded as designed.
. 30 minutes AFTER the reactor trip the following plant conditions exist:
. All SG pressures are 1005 psig and stable.
. RCS Subcooling is 35°F and stable.
. Loop ATs are indicating upscale and stable.
All SG NR levels are 45% and slowly dropping.
Total AFW flow is 100 gpm and stable.
(1) When natural circulation has been established, what will be the status of Tcoldand Thot?
(2) IAW EOP Attachment A-I.7, Natural Circulation Verification, what directions are you REQUIRED to give the crew in order to enhance natural circulation?
(1) Tcoldwill be at 547°F and Thot will be stable or rising (2) Raise SG NR levels by increasing AFW flow.
(1) Tcoldwill be at 512°F and Thot will be stable or dropping.
(2) Raise the rate at which steam is being dumped.
(1) Tcoldwill be at 512°F and Thot will be stable or rising.
(2) Raise SG NR levels by increasing AFW flow.
(1) Tcoldwill be at 547°F and Thot will be stable or dropping.
(2) Raise the rate at which steam is being dumped.
Answer D
ExpIanationlJustification:
A.
B.
C.
D.
Incorrect. Right Tc response; Wrong Th response and wrong enhancement directions. Tc will be at saturation temperature for SG pressure Th will be rising as natural circulation is being developed BUT it will be stable or dropping once it has been developed. If Th is still rising then I atural circulation has not been developed. Raising SG levels may seem plausible however, NR SG levels are within the band of 35-50%. The din ction for enhancing is to raise the steam dump rate.
Incorrect. Wrong Tc response; Right Th response and Right enhancement directions. Tc will be at saturation temperature for SG pressure,vhich is 1005 psig (547°F). Th will be rising as natural circulation is being developed BUT it will be stable or dropping once it has been develope1 is still rising then natural circulation has not been developed. 512°F corresponds to 35°F below Tsat. A non-discriminating candidate may 1 this is where Tc will be for these conditions.
Incorrect. Wrong Tc response; Wrong Th response and wrong enhancement directions. Tc will be at saturation temperature for SG pressu e. Th will be rising as natural circulation is being developed BUT it will be stable or dropping once it has been developed. If Th is still rising then,ratural circulation has not been developed. Raising SG levels may seem plausible however, NR SG levels are within the band of 35-50%. The dir::ction for enhancing is to raise the steam dump rate. 512°F corresponds to 35°F below Tsat. A non-discriminating candidate may think this is wh ':re Tc will be for these conditions.
Correct. Tc will be at saturation temperature for SG pressure which is 1005 psig (547°F). Th will be rising as natural circulation is being de 'eloped BUT it will be stable or dropping once it has been developed, If Th is still rising then natural circulation has not been developed. IAW EOP Attachment A-I.7 Natural Circulation Verification the direction for enhancing natural circulation is to raise the rate of dumping steam.
NOTE:
This question is NOT too similar to RO question # I O. RO question # I O requires candidate to know only trends and the concept C~ Tc being equal to or less than saturation pressure in the SGs. This SRO question requires the SRO candidate to calculate TC and addresses Th which is NOT addressed in the RO question. This question also requires the SRO candidate to determine what act ons will be directed to enhance natural circulation.
P WA Sys #
K/A System K/A Category K/A Statement 000056 Loss of Off-site Power Ability to determine and interpret the following as they apply to the Loss of Offsite Power KIA #
AA2 19 WAlmportance 4 2 ExamLevel SRO Nestion Source:
New References provided to Candidate Steam Tables Technical
References:
EOP Attachment A-I.7 Natural Circul lion Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR 43 5 I 4 5 13)
T-cold and T-hot indicators (wide ranc ?)
Level Of Difficulty: (1-5)
Question Cognitive Level:
Higher Comprehension Verification Page 80 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 81.
The plant is operating at 100% power with all systems in NSA.
A LOCA OUTSIDE containment occurs.
At step 20 of E-0, Reactor Trip Or Safety Injection, the crew enters ECA 1.2, LOCA Outsid(:
Containment.
At the completion of ECA 1.2, the crew has been UNABLE to locate and isolate the break.
The following plant conditions NOW exist:
All SG pressures are 800 psig and stable.
All SG NR levels are 35% and slowly rising.
All Secondary radiation monitors are consistent with pre-event values.
CNMT Pressure is -1.O psig and stable.
CNMT sump level is consistent with pre-event values.
CNMT radiation is consistent with pre-event values.
RCS Subcooling is 40°F and slowly dropping.
AFW flow is 700 gpm and stable.
RCS Pressure is 1125 psig and slowly dropping.
PRZR level is 12% and slowly dropping.
Auxiliary Building Radiation levels are rising.
Auxiliary Building sump levels are rising.
Based on these conditions:
What procedural transition is REQUIRED?
A.
E-0, Reactor Trip Or Safety Injection
- d.
ECA-1. I, Loss Of Emergency Coolant Recirculation.
C.
E-I, Loss Of Reactor Or Secondary Coolant.
D.
ES-I.2, Post-LOCA Cooldown And Depressurization.
Answer B
ExplanationIJustification:
A.
B.
C.
D.
KIA Sys #
WA System WA Category WA Statement WlE04 LOCA Outside Incorrect. Plausible since many of the procedures in the EOP network have the crew returning to procedure and step in effect. There are al.O procedures that have the crew do this even if the procedure was ineffective in correcting the problem.
Correct. IAW ECA-1.2 step 4 RNO. SRO level since this requires a candidate to have a detailed understanding of what the required transit, )n would be when ECA-1.2 is essentially ineffective. ROs would NOT be required to have this detailed knowledge.
Incorrect. Plausible since E-I would be the appropriate entry if RCS pressure were rising. Since RCS pressure is NOT rising, ECA-1.1 is tt ?
appropriate entry procedure to enter.
Incorrect. Plausible since plant conditions support entry into ES-I.2 from E-I but NOT from ECA-1.2.
Ability to determine and interpret the following as they apply to the (LOCA Outside Containment)
WA #
EA2.1 KIA Importance 4.3 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate None Technical
References:
ECA-1.2 step 4 RNO Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 43.5 145.13)
Facility conditions and selection of app,opriate procedures during abnormal and emer,iency operations.
Containment Page 81 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 SRO ONLY
- 82.
The Plant is operating at 100% power with all systems in NSA.
Control Bank D is at 229 steps.
Control Bank D Demand step counters are at 229 steps.
Control Rod Group Selector Switch is in the MAN position.
Plant Parameters are NOW as follows:
Tavg is 575°F and slowly dropping.
RCS Pressure is 2230 psig and slowly dropping.
Reactor power has dropped to 96% and is slowly rising.
PR N-41 Negative Rate Trip bistable is LIT All other PR Negative Rate Trip bistables are NOT LIT Control Bank D Demand step counters remain at 229 steps.
Based on these conditions:
What procedure contains the REQUIRED guidance to address these plant conditions?
A.
E-0, Reactor Trip Or Safety Injection.
B.
AOP 2.2.1 C, Power Range Channel Malfunction.
C.
AOP 2.1.7. Rod Position Indication Malfunction.
D.
AOP 2.1.8, Rod Inoperability.
m w e r D
ExplanationlJustification:
A.
B.
C.
Incorrect. No entry conditions for E-0 have been met. The PR rate coincidence is 2/4 and only one rate bistable has been actuated.
Incorrect. Plausible since the negative rate bistables have actuated. However, with power dropping and temp dropping there MUST be son e negative p being added (dropped rod).
Incorrect. Entry into this procedure is required ONLY if there is no evidence of a plant response to the alarms. In this question, the plant ha.,
responded to a dropped rod with corresponding temp/pressure/power change. This procedure may be entered as part of the initial diagnos,ics, however entry into this procedure is not REQUIRED.
Correct. IAW symptoms listed for AOP 2.1.8, the alarms and plant response are consistent with a dropped rod. AOP 2.1.5 Dropped rod ha. been deleted, and AOP 2.1.8 now addresses a dropped rod in Part A. SRO candidate must evaluate the given conditions and those that are NO..
present to determine that a rod has dropped, and is in fact at zero steps.
NOTE: The stem is worded using the word contains due to the allowance for entry into AOP 2.1.7 which would then diagnose AOP 2.1.8 a,: the correct procedure to address these conditions. If the stem asked what procedure entry is required, then there would be 2 possible answer: Only AOP 2.1.8 Contains the appropriate guidance.
D, P
WA Category KIA Statement Ability to determine and interpret the following as they apply to the Dropped Control Rod:
K/A Sys #
KIA System 000003 Dropped Control Rod KIA #
AA2.01 K/A Importance 3.9 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Analysis References provided to Candidate None Technical
References:
AOPs 2.1.3, 2.1.7, & 2.1.8 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 43.5 / 45.13)
Rod position indication to actual rod pc sition Page 82 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 83.
The Plant is operating at 100% power with all systems in NSA.
RCS activity is high, right at the Technical Specification Limits, due to leaking fuel elements.
At T = 0, RCP 21 C Thermal barrier heat exchanger develops a leak AND 21 C RCP Therma Barrier Outlet Is01 Vlv [2CCP*AOV107C] FAILS to isolate and CANNOT be closed.
Thermal barrier outlet flow is 60 gpm and stable.
At T = 1 minute, The following alarms and indications are received:
Annunciator A4-5C Radiation Monitoring Level High - Alarms Component Cooling Heat Exchanger Radiation Monitor 2SWS-RQI 101 AND Component Cooling Service Water Radiation Monitor 2SWS-RQI102 are BOTH in - HIGH Alarm Radiation Monitor 2SWS-RQI101 is reading 9.0 X IO-* pCi/ml.
Radiation Monitor 2SWS-RQI102 is reading 9.0 X pCi/ml No Reactor Trip or SI signals have been actuated.
No Reactor Trip or SI signals are required.
If all of these conditions continue until T = 20 minutes, What is the highest Emergency Plan Classification REQUIRED, if any, at T = 20 minutes?
(Assume NO Dose projections will be available until T = 50 minutes).
A.
No Emergency Plan Classification is required.
B.
Unusual Event.
Alert.
D.
Site Area Emergency Answer C
ExplanationlJustification:
A.
B.
C.
Incorrect. Candidate could choose this based on RCS identified leakage being less than 25 gpm (Tab 2.6) based on thermal barrier outlet ' ow rising from a nominal 45 gpm to 60 gpm.
Incorrect. Candidate could choose this based on [2SWS-RQI102] being greater than 2 times the ODCM setpoint (Tab 7.2), however this N JST be for a period of greater than 60 minutes to be a UE.
Correct. IAW Tab 7.2 and the bases for Tab 7.2. [2SWS-RQI101] is 200 times the ODCM setpoint and this has been for greater than 15 m lutes.
The candidate must refer to the EAL Tab 7.2 and the corresponding table 7-1 and apply the given data to the EAL matrix. The keys to the question are to recognize that the given radiation monitors are indicators used to determine if an EAL criterion has been exceeded AND to recognize that the numbers in table 7-1 have been exceeded. After analyzing and applying this information, the candidate may still choose UE since tt.is value has been exceeded, but NOT for > 60 minutes. Making the correct EAL determination demonstrates the SRO ability.
NOTE:
At BVPS UNPLANNED releases that are not covered by a Radioactive Waste Discharge Authorization (RWDA) are considered accidental releases.
Incorrect. Candidates could choose this if they incorrectly declare both the Fuel barrier and RCS barrier to be potential losses.
D.
KIA Sys #
KIA System KIA Category KIA Statement 000059 Accidental Liquid N/A Ability to diagnose and recognize tren s in an KIA #
2.4.47 KIA Importance 4.2 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Application References provided to Candidate EALs.
Technical
References:
EAL Tab 7.2 and table 7-1 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.10 / 43.5 I45.12)
RadWaste Rel.
accurate and timely manner utilizing tl e appropriate control room reference m; rerial.
Page 03 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 84.
The Plant is operating at 100% power with all systems in NSA.
A major fire has started in the control room.
The fire is NOW out of control and the fire brigade has not been able to extinguish the fire.
Which ONE (1) of the following methods/locations will be used to bring the unit to cold shutdown?
Direct the crew to conduct a :
A.
Natural circulation cooldown from the Alternate Shutdown Panel (ASP).
B.
Natural circulation cooldown from the Emergency Shutdown Panel (ESP).
C.
Forced circulation cooldown from the Alternate Shutdown Panel (ASP).
D.
Forced circulation cooldown from the Emergency Shutdown Panel (ESP).
Answer A
Explanation/Justification:
A.
- 6.
Correct. IAW 20M-56C.l.B page 2 3" paragraph and 20M-56C.4.A page 2 1" paragraph and 20M-56C.4.B page 3 Znd item.
Incorrect. Implementation of 20M-56C, Alternate Safe Shutdown From Outside The Control Room requires the ASP to be activated. The EF'P is activatec for small fires, toxic fumes, etc. At BVPS natural circulation cooldowns from the ESP would be implemented IAW AOP 2.33.1A, Cf Introl Room Inaccessibility. This would be implemented for small fires NOT Major fires. It is the responsibility of the SRO to determine which procedure to implement, and the determinant is MAJOR vs small. For this question the stem clearly states MAJOR and it is an expectation that all SROs are familiar with this distinction.
Incorrect. RCPs are tripped before leaving the control room.
Incorrect. RCPs are tripped before leaving the control room.
C.
D.
A Sys #
K/A System WA Category WA Statement il00067 Plant Fire On-site NIA Knowledge of "fire in the plant" procedL es KIA #
2.4.27 WA Importance 3.9 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
20M-56C.l.B page 2 3d paragraph ani: 20M-56C.4.A page 2 1" paragraph and 20b 56C.4.B page 3 2"' item.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.10 / 43.5 145.13)
Page 84 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 85.
The Plant is operating at 100% power with all systems in NSA.
The control room crew is performing 20ST-43.6, Containment High Range Area Monitor Channel Test.
During the surveillance, the HIGH alarm setpoint for In-Containment High Range Area moni-or
[2RMR*RQ206] is found to be set at 2.6 X IO4 R/hr.
Background radiation is 100 mr/hr.
Based on these conditions, what is the MINIMUM Technical Specification/LRM action, if any, that is REQUIRED?
A.
No Technical Specification/LRM action is required.
B.
Restore the required alarm channel to OPERABLE status within 30 days.
C.
Adjust the alarm setpoint to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
D.
Declare the radiation monitor alarm inoperable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Answer C
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. The alarm setpoint is out of range, and must be adjusted.
Incorrect. This would be the required action if the MONITOR was inoperable.
Correct. in this case, the LRM provides specific actions for the alarm setpoint being out of range. Therefore, the minimum required action is o adjust the setpoint as specified in LRM 3.3.15 Condition A.1.
Incorrect. This action in an option BUT it must be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> NOT 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
'A Sys #
WA System WA Category WA Statement 0061 ARM System Alarms Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms:
Required actions if alarm channel is oul #of service WA #
AA2.06 WA Importance 4.1 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New References provided to Candidate LRM 3.3.1 5 Condition A.l Objective #:
Question Cognitive Level:
Higher Application Technical
References:
10 CFR Part 55 Content:
20ST-43.6, LRM 3.3.15, TS 3.3.3 Task ID#:
(CFR: 43.5 / 45.13 I43.b.l & 2)
Page 85 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 SRO ONLY
- 86.
A plant heatup/startup is in progress with RCS average temperature at 325°F.
Other plant conditions are as follows:
CharginglHHSl Pump 2CHS*P21 B is INOPERABLE.
Charging/HHSI Pump 2CHS*P21 C is on clearance for maintenance.
A risk assessment for this condition has NOT YET been performed.
Recirculation Pump 2RSS*P21 D becomes INOPERABLE.
Which ONE (1) of the following describes the Technical Specification REQUIRED Actions in order tc continue the plant heatup?
A.
Restore the 2RSS*P21 D recirculation pump and 2CHS*P21 B Charging/HHSI Pump to OPERABLE status BEFORE exceeding 350°F.
B.
Restore ONLY the Charging/HHSI Pump 2CHS*P21 B to OPERABLE status BEFORE exceeding 35 J"F.
C.
Restore the 2RSS*P21 D recirculation pump and 2CHS*P21 B Charging/HHSI Pump to OPERABLE status BEFORE exceeding 375°F.
D.
Restore ONLY the Charging/HHSI Pump 2CHS*P21 B to OPERABLE status BEFORE exceeding 37'5°F Answer A
- planationlJustification:
B.
C.
D.
WA Sys #
WA System WA Category WA Statement 006 Emergency Core Cooling N/A Knowledge of conditions and limitations in the facility license.
WA #
2.2.38 K/A Importance 4.5 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
BVPS Bank 56320 Question Cognitive Level:
Higher Application References provided to Candidate Objective #:
Task ID#:
10CFRPart55Content:
(CFR: 41.7/41.10/43.1 145.13)
Correct. IAW LCO 3.5.2 both pumps must be operable before transitioning above 350°F. At Unit 2 RSS* P21C and D provide the LHSl func' on during recirculation phase.
Incorrect. Both pumps are required before exceeding 350°F.
Incorrect. The 25 degree allowance in Note 2 of TS 3.5.2 is only applicable to the charging pump. The Recirc spray pump is required before exceeding 350°F.
Incorrect. Both pumps are required before exceeding 350°F.
LCO 3.5.2 and LCO 3.5.3 Technical
References:
LCO 3.5.2 and Bases; LCO 3.5.3 and E ses; and LCO 3.0.4 Page 86 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 87.
The Plant is operating at 100% power with all systems in NSA.
The control switches for all PORV Motor Operated Is01 Vlvs [2RCS*MOV535, 536, 5371 arc in AUTO.
An inadvertent reactor trip occurs.
2 Minutes later Pressurizer Spray Valve 2RCS*PCV455A fails OPEN and is stuck OPEN.
The PORV Motor Operated lsol Vlvs [2RCS*MOV535, 536, 5371 will AUTOMATICALLY close when (1)
E-0, Reactor Trip or Safety Injection REQUIRES you to direct the crew to stop Reactor Coolant Pumps (2)
A.
(1) 2/3 PZR Protection channels decrease to less than 2000 psig (2) 21A and 21C B.
(1) 2/3 PZR Protection channels decrease to less than 2185 psig (2) 21A and 21C C.
(1) 2/3 PZR Protection channels decrease to less than 2000 psig (2) 21B and 21C D.
(1 ) 2/3 PZR Protection channels decrease to less than 21 85 psig (2) 21 B and 21 C iswer B
ExplanationlJustification:
A.
B.
C.
D.
WA Sys #
WA System WA Category WA Statement 010 Pressurizer Pressure Ability to (a) predict the impacts of the following Spray valve failures Incorrect. Wrong setpoint for auto closure. 2000 psig is the P-I 1 permissive. At Unit 1 the P-I 1 interlock performs this function. At Unit 2 thl function is performed at 2185 psig.
Correct. IAW 20M-6.4.IF auto close feature is 213 protection channels below 2185 psig. IAW E-0 step 12b RNO secure the 21A and C RCf s.
Incorrect. Wrong setpoint for auto closure. Wrong pumps for the PCV455A failure.
Incorrect. Correct setpoint for auto closure. Wrong pumps for the PCV455A failure.
Control malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
KIA #
A2.02 WA Importance 3.9 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Objective #:
Task ID#:
Technical
References:
10 CFR Part 55 Content:
E-0 step 12b RNO; 20M-6.4.IF (CFR: 41.5 / 43.5 / 45.3 / 45.13)
Page 07 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 88.
The Plant is operating at 100% power with all systems in NSA.
e An Instrument Power fuse for Power Range NIS Channel 2 (N42) Blows.
( 1) What will be the status of the SSPS Rx trip relay for N42 Overpower Trip High Range?
(2) What are ALL of the applicable Reactor Trip System (RTS) Instrumentation Functions that \\!ill REQUIRE Technical Specification action? (Choose from the list below)
- a. Power range neutron flux -High
- b. Power range neutron flux -Low
- c. Power range neutron flux High positive rate
- d. Overtemperature AT
- e. P8 Power range neutron flux interlock
- f.
P9 Power range neutron flux interlock
- g. P I 0 Power range neutron flux interlock A.
(1 ) Tripped (2) a, c, d, e, f, and g B.
(1) lripped (2) a, b, c, d, e, f, and g C.
(1 ) NOT Tripped (2) a, c, d, e, f, and g iswer A
ExplanationIJustification:
A.
Correct. Loss of control power OR instrument power will cause the bistable to trip. It is a common misconception that only a loss of control )ewer will cause the bistable to trip since control power is what powers the drawer. However, the bistable relay driver will input a trip for loss of eiher power supply. The six TS actions are applicable.
Incorrect. Action 2b is not applicable since reactor power is above the P-10 interlock.
Incorreci. Loss of control power OR instrument power will cause the bistable to trip. It is a common misconception that only a loss of contrc power wdI cause the bistable to trip since control power is what powers the drawer. However, the bistable relay driver will input a trip for 10s of either pcwer supply. The six TS actions are applicable.
Incorrect. Loss of control power OR instrument power will cause the bistable to trip. It is a common misconception that only a loss of contrc power will cause the bistable to trip since control power is what powers the drawer. However, the bistable relay driver will input a trip for 10s of either power supply. Action 2b is not applicable since reactor power is above the P-10 interlock.
B.
C.
D.
WA Sys #
WA System WA Category WA Statement 01 2 Reactor Protection System Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures io correct, control, or mitigate the consequences of those malfunctions or operations:
Loss of Instrument Power KIA #
A2.02 WA Importance 3.9 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
BVPS Bank #61405 Modified to make closed book Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
TS Section 3.3.1 and AOP 2.2.1C Syml tom Objective #:
10 CFR Part 55 Content:
(CFR: 41.5 I43.5 / 45.3 / 45.5)
Page 88 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 SRO ONLY
- 89.
The Plant is operating at 100% power with all systems in NSA.
No automatic actions occur.
Battery *2-2 is being charged, per maintenance request.
Emergency Switchgear Exhaust Fan 2HVZ*FN262A is running.
Emergency Switchgear Exhaust Fan 2HVZ*FN262B is in Auto.
The running Battery Room Exhaust Fan 2HVZ*FN216A TRIPS.
Annunciator A I 0-7H Battery Room Exhaust Fan Auto-Start/Auto-Stop is received.
Based on these conditions:
(1) What impact will this have on the battery rooms?
(2) IAW ARP Al0-7H, Battery Room Exhaust Fan Auto-Start/Auto-Stop, what actions are you REQUIRED to direct the crew to perform in order to address this alarm condition?
A.
(1) Oxygen concentrations will buildup.
(2) Start Emergency Switchgear Exhaust Fan 2HVZ*FN262B.
B.
(1) Hydrogen concentrations will buildup.
(2) Start Emergency Switchgear Exhaust Fan 2HVZ*FN262B.
C.
(1) Oxygen concentrations will buildup.
(2) Start Battery Room Exhaust Fan 2HVZ*FN216B.
D.
(1) Hydrogen concentrations will buildup.
(2) Start Battery Room Exhaust Fan 2HVZ*FN216B.
Answer D
~planationlJustification:
A.
Incorrect. Battery charging generates hydrogen gas NOT Oxygen gas. The excess hydrogen gas could buildup to explosive levels if the exbwst system is not functioning. Starting the Emergency Switchgear Exhaust Fan 2HVZ*FN262B may seem like a viable solution since the batter) rooms are located in emergency switchgear. However, the Emergency Switchgear supply and Exhaust Fans provide fresh cool air to the emergency switchgear area and the battery room exhaust fans will pull this air into the battery room and exhaust it to outside. The procedur 11 guidance is to start the redundant battery room exhaust fan NOT start the redundant Emergency Switchgear Exhaust Fan.
Incorrect. Right impact: wrong procedural actions. Starting the Emergency Switchgear Exhaust Fan 2HVTFN262B may seem like a viable solution since the battery rooms are located in emergency switchgear. However, the Emergency Switchgear supply and Exhaust Fans prov.le fresh cool air to the emergency switchgear area and the battery room exhaust fans will pull this air into the battery room and exhaust it to 01 lside.
The procedural guidance is to start the redundant battery room exhaust fan NOT start the redundant Emergency Switchgear Exhaust Fan.
Incorrect. Wrong impact; correct actions.
Correct. Correct impact; correct actions.
- 8.
C.
D.
WA Sys #
KIA System KIA Category KIA Statement 063 L
DC Electrical Distribution Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems:
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Loss of ventilation during battery chargi g KIA #
A2.02 KIA Importance 3.1 Exam Level SRO Level Of Difficulty: (1 -5)
Question Source:
New Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
20M-44F.AAH; 20M-44F.1.B page 3 oi '$
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.5 143.5 145.3 145.13)
Page 89 of 100
Beaver Valley Unit 2 NRC Written Exam (2~016)
SRO ONLY The Plant is operating at 15% power with all systems in normal alignment for this power level.
SGWLC is being maintained Automatically by the SG Feedwater Bypass Control Vlvs
[2FWS*FCV479(489)(499)].
Annunciator A6-3C Station Instrument Air Receiver Tank Trouble is received.
Station Instrument Air Header Pressure is 80 psig and slowly dropping.
A local operator reports that the station instrument air dryers have malfunctioned, and both dryers are venting.
(1) IAW AOP 2.34.1, Loss Of Station Instrument Air, what directions are you REQUIRED to givc the local operator to address the degrading Station Instrument Air Header Pressure?
(2) IF Station Instrument Air Header Pressure continues to drop below 30 psig, how will the SG Feedwater Bypass Control Vlvs FAIL?
(1) Place the Instrument Air Bypass filters in service THEN isolate the Instrument Air dryers.
(2) OPEN.
(1) Place the Instrument Air Bypass filters in service THEN isolate the Instrument Air dryers.
(2) CLOSED.
(1) Supply Station Instrument Air with Containment Instrument Air by OPENING CNMT instrument Air backup supply Valve [ZIAC-MOVI 311 and CNMT instrument Air supply Is01 Valve [ZIAC-MOVI 301.
(2) OPEN.
(1) Supply Station Instrument Air with Containment Instrument Air by OPENING CNMT instrument,iir backup supply Valve [ZIAC-MOVI 311 and CNMT instrument Air supply Is01 Valve [ZIAC-MOVI C'O].
(2) CLOSED.
Answer B
ExplanationlJustification:
A.
B.
C.
D.
Incorrect. Correct actions; wrong failure mode for the SG Feedwater Bypass Control Vlvs Correct. IAW AOP-2.34.1 place bypass filters in service and SG Feedwater Bypass Control Vlvs fail closed.
Incorrect. These are the actions for loss of containment instrument air. Containment instrument air can be supplied by station instrument ai by opening these valves and these are the directions given in AOP 2,34.2; wrong failure mode for the SG Feedwater Bypass Control Vlvs Incorrect. These are the actions for loss of containment instrument air. Containment instrument air can be supplied by station instrument ai by opening these valves and these are the directions given in AOP 2,34.2; correct failure mode for the SG Feedwater Bypass Control Vlvs KIA Sys #
KIA System KIA Category KIA Statement 078 instrument Air Ability to (a) predict the impacts of the following malfunctions or operations on the IAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Air dryer and filter malfunctions KIA #
A2.01 KIA Importance 2.9 Exam Level SKO Level Of Difficulty: (1-5)
Question Source:
New References provided to Candidate None Objective #:
Question Cognitive Level:
Higher Comprehension Technical
References:
10 CFR Part 55 Content:
(CFR: 41.5 I43.5 / 45.3 / 45.13)
AOP 2.34.1 step 3 and NOTE prior to t :ep 7 Task ID#:
Page 90 d 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 91.
The Plant is operating at 100% power with all systems in NSA.
A Reactor Trip coincident with a loss of offsite power occurs.
Both trains of RVLIS are NOT functioning.
All other systems functioned as designed.
RCS Hot leg temperatures are 450°F and stable.
The crew is performing a Natural Circulation Cooldown IAW ES-0.4, Natural Circulation Cooldown With Steam Void in Vessel (Without RVLIS).
Throughout this procedure, the plant is depressurized in several discrete phases. AFTER each depressurization phase there is a check of pressurizer level to ensure it is less than 90%.
During this check of pressurizer level, IF Pressurizer level is greater than 90% :
(1) What directions are you REQUIRED to give the crew to address the pressurizer level situation?
(2) What is the basis for this action?
A.
(1 ) Maximize letdown flow.
(2) Prevent a water solid RCS and the resultant loss of pressure control.
B.
(1) Raise RCS pressure 100 psig using PRZR Heaters.
(2) Prevent a water solid RCS and the resultant loss of pressure control.
C.
(1) Maximize letdown flow.
(2) Partially or wholly collapse the Rx vessel void.
- 7. (1) Raise RCS pressure 100 psig using PRZR Heaters.
(2) Partially or wholly collapse the Rx vessel void.
Answer D
ExplanationlJustification:
A.
Incorrect. This is the required action and basis for high pressurizer level while in ES-0.3 where the technique employed for the RCS cooldc <vn is dramatically different, In ES-0.3 charging and letdown are controlled throughout the cooldown to keep PRZR level below 90%. In ES-0.4 cl arging and letdown are set PRIOR to the cooldown and thereafter NOT adjusted. PRZR level rise is then used to monitor void growth.
Incorrect. Right action; wrong basis.
Incorrect. Wrong action; right basis.
Correct IAW ES-0.4 step 9 and basis.
- 6.
C.
D.
KIA Sys #
KIA System KIA Category K/A Statement 01 1 Pressurizer Level Control N/A Ability to perform specific system and P
integrated plant procedures during all I lodes of plant operation.
KIA #
2.1.23 K/A Importance 4.4 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Higher Analysis Technical
References:
10 CFR Part 55 Content:
ES-0.4 step 9 and basis; (CFR: 41. I O / 43.5 / 45.2 / 45.6)
Page 91 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 92.
The Plant is operating at 100% power with all systems in NSA.
Gaseous Waste Storage Tanks [2GWS-TK25A-G] pressures are 10 psig and stable.
Gaseous Waste Surge Tank [2GWS-TK21] pressure is 62 psig and stable.
The Waste Gas Storage Tanks Radiation Monitor [2GWS-RQ104] is out of service.
Oxygen Analyzer [2GWS-OAI OOA] is out of service.
RCS Coolant activity is 25 pCi/ml.
It is desired to fill the Gaseous Waste Storage Tanks IAW 20M-19.4.G, Filling Unit 2 Gaseous Wast?
Storage Tanks From Unit 2 Surge Tank.
While filling the Gaseous Waste Storage Tanks, under these conditions, what LRM/ODCM compensatory actions are REQUIRED?
At least once per:
A.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; take grab samples and analyze for BOTH Oxygen concentration and radioactive content.
B.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; take grab samples and analyze for Oxygen concentration ONLY.
C.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; take grab samples and analyze for Oxygen concentration and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for radioactive conttmt.
D.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; take grab samples and analyze for Oxygen concentration ONLY.
Answer A
Explanation/Justification:
A.
C.
D.
Correct. IAW LRM 3.3.12 condition B.l and ODCM attachment 0 surveillance 4.1 1.2.5.1.
Incorrect. At Unit 2 Both Oxygen and radioactive content must be sampled and analyzed. If the candidate does NOT correctly apply the OD( M surveillance, then this distractor would appear plausible. Unit 1 does NOT have to perform this surveillance if RCS activity is below 100 pCi/i 11.
Incorrect. This oxygen sample time is the time limit if BOTH oxygen analyzers were 00s. Right actions for radioactive content.
Incorrect. This oxygen sample time is the time limit if BOTH oxygen analyzers were 00s. If the candidate does NOT correctly apply the OD1 :M surveillance, then this distractor would appear plausible. Unit 1 does NOT have to perform this surveillance if RCS activity is below 100 pCih 11. -
WA Sys #
KIA System WA Category WA Statement 07 1 Waste Gas Disposal NIA Ability to interpret and execute proceduri steps.
WA #
2.1.20 WA Importance 4.6 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Application References provided to Candidate Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.10 / 43.5 / 45.12)
'/ ODCM section 3.0.3; LRM 3.3.12 Technical
References:
LRM 3.3.12 condition B.l and '/-ODC-3. 13 attachment 0 surveillance 4.1 1.2.5.1 Page 92 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 93.
The Plant is in Mode 5.
Train A is the declared protected Train.
System Station Transformer 2A is supplying the 2A and 2AE 4KV Buses.
The Deluge valve for System Station Transformer 2A inadvertently actuates and sprays the transformer.
Operators locally isolate the Deluge valve.
System Station Transformer 2A remains in service.
IAW %-ADM-1900, Fire Protection Program, what actions, if any, are REQUIRED for the isolated Deluge valve?
A.
No actions required.
- 6.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, establish an hourly fire watch patrol with backup fire suppression capability and establish controls to prohibit transient combustibles in the affected area.
C.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, establish an hourly fire watch patrol in the affected area with backup fire suppression capability, and to check for proper cooling, no oil leakage, or any abnormal conditions.
D.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, establish an hourly fire watch patrol in the affected area to check for proper cooling, 10 oil leakage, or any abnormal conditions.
Answer D
Explanation/Justification:
A.
- 6.
Incorrect. Candidate may think that since the 2A transform is not safety related that no actions are required.
Incorrect. These are the required actions for safety related equipment that is protected by the C02 system.
Incorrect. This is a combination of C02/water and safetyhon-safety related actions.
Correct. IAW Att. B of %-ADM-1900 item 3b for non-safety related equipment that is required to be operable. Candidate must realize that the 2A transformer is non-safety related and is required to be operable in Mode 5 with Train A protected.
WA Sys #
K/A System WA Category KIA Statement 086 Fire Protection N/A KIA #
2.1.7 WA Importance 4.7 Exam Level SRO Question Source:
New References provided to Candidate
%-ADM-1900 Objective #:
Task ID#:
Ability to evaluate plant performance anc make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Level Of Difficulty: (1-5)
Question Cognitive Level:
Higher Application Technical
References:
10 CFR Part 55 Content:
%-ADM-1900 Attachment B item 3b (CFR: 41.5 / 43.5 I45.12 / 45.13)
Page 93 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 94.
The plant is operating in Mode 6 with all systems in normal alignment for this Mode.
0 0
Core Re-loading activities are in progress.
There are 100 fuel assemblies in the core.
Source Range Channel N32 fails low.
Source Range Channel N31 remains OPERABLE.
Both trains of Gammametrics are INOPERABLE Which ONE (1) of the below listed evolutions can STILL be performed WITHOUT violating the Technical Specification required actions for Source Range Instrumentation?
A.
Removing a SPENT fuel assembly from its fully lowered core position and placing it into the fuel trar sfer cart.
B.
Moving an underwater camera from one core location to another to verify proper seating of fuel ass em b I i es.
C.
Removing a temporary secondary source device that was installed in the center core location to ass st in plotting IIM data.
D.
Moving a fuel assembly from a temporary core location into the final core location that is adjacent to source range channel N31.
Answer B
ExplanationlJustification:
A.
Incorrect. Even though this would lessen the overall reactivity of the core, it would violate the TS action for one inoperable source range chai nel.
This is a Core Alteration since it is fuel movement within the vessel with fuel in the vessel.
Correct. Loss of HV power supply will render N32 inoperable. TS action is to IMMEDIATELY suspend core alterations. Core alterations are defined as movement of fuel, sources, or reactivity control components within the vessel WITH fuel in the vessel. Underwater cameras are n me of these therefore, this evolution would be permitted. In order to obtain the correct answer, a student will need to know from memory the defi iition of core alteration and apply this definition to the situations posed in the question. At BVPS the refueling SROs are responsible for authorizinc, all core alterations and ensuring all TS requirements have been met before the evolution can commence.
Incorrect. Even though this is removing a source device, it would violate the TS action for one inoperable source range channel. This is a Co e Alteration since it is movement as a source within the vessel with fuel in the vessel.
Incorrect. Even though the assembly is already in the core in its temporary location, it would violate the TS action for one inoperable source I inge channel. This is a Core Alteration since it is fuel movement within the vessel with fuel in the vessel.
C.
D.
KIA Sys #
KIA System KIA Category KIA Statement NIA Generic Conduct Of Operations Knowledge of procedures and limitations KIA #
2.1.36 WA Importance 4.1 Exam Level SRO Level Of Difficulty: (1-5) involved in core alterations.
Question Source:
New References provided to Candidate None Objective #:
Task ID#:
Question Cognitive Level:
Higher Comprehension Technical
References:
10 CFR Part 55 Content:
Tech Spec 3.9.2; Tech Spec Definition 01 Core Alteration.
(CFR: 41.10 I43.6 / 45.7)
Page 94 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 SRO ONLY
- 95.
The Plant is operating at 100% steady state power with all systems in NSA. All Primary an(
plant chemistry parameters are within Technical Specification/LRM limits.
Secondary The,shift chemist reports the following STABLE Primary and Secondary plant chemistry conditions, based on the LATEST sample:
Secondary Specific Activity is 0.025 pCi/gm DOSE EQUIVALENT 1-131 RCS Specific Activity is 25.0 pCi/gm DOSE EQUIVALENT 1-131 RCS Dissolved Oxygen is 0.15 ppm RCS Chlorides are 0.10 ppm RCS Fluorides are 0.10 ppm Based on these chemistry conditions, what Technical SpecificatiodLRM actions are REQUIRED at 'his time:'
Within:
A.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> be in Mode 3 and within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> be in Mode 5.
B.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> be in Mode 3 with Tavg less than 500°F.
C.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore DOSE EQUIVALENT 1-1 31 to within its limit.
D.
48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> restore RCS Dissolved Oxygen to within the steady state limit.
Answer B
ExplanationIJustification:
- 6.
C.
D.
KIA Sys #
K/A System K/A Category KIA Statement N/A Generic Conduct Of Operations Knowledge of primary and secondary pla it Chemistry limits.
KIA #
2.1.34 KIA Importance 3.5 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Higher Application References provided to Candidate TS 3.4.16 condition C.l and TS figure 3.4.16-1.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.10 / 43.5 / 45.12)
Incorrect. These are the required actions if Oxygen, chloride, or fluoride, are outside their transient limits. Oxygen is outside the steady state Amit but within the transient limit.
Correct. IAW TS 3.4.16 condition C.l since DOSE EQUIVALENT 1-131 is in the unacceptable region of TS figure 3.4.16-1.
Incorrect. Time limit is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> NOT 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Incorrect. Time limit is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> NOT 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
TS 3.4.16, 3.7.13; LRM 3.4.2 Technical
References:
Page 95 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 SRO ONLY
- 96.
The Plant is operating at 100% power with all systems in NSA.
All PORVs and associated block valves are OPERABLE.
Technical Specification LCO 3.4.1 1 requires each PORV and associated block valve to be OPERABLE.
Every 92 days, Surveillance 20ST-6.6, PORV Isolation Valve Test and Position Check, is performed to meet this requirement.
While performing 20ST-6.6:
2RCS*MOV535 PORV Motor Operated Is01 Vlv CLOSES but WILL NOT OPEN.
Maintenance finds a bad power supply breaker to the MOV, and replaces the entire breakei assembly at the MCC.
ALL of their required work package instructions have been completed.
The tagout has been lifted, 2RCS*MOV535 is ENERGIZED and CLOSED.
2RCS*MOV535 is ready for operations' post-maintenance testing.
For these conditions:
What MINIMUM post-maintenance testing will be REQUIRED to verify compliance with Technical Specification LCO 3.4.1 1 ?
(For each of the below actions, assume all valve stroke times and indications are within accept ab I e I i m its)
A.
Open 2RCS*MOV535, no other actions required.
- 9.
Open 2RCS*MOV535; then Close; then re-open.
J.
Cycle the associated PORV through one complete cycle, then open 2RCS*MOV535 D.
Cycle the associated PORV through one complete cycle, then open 2RCS*MOV535; then Close; then re-open.
Answer B
ExplanationlJustification:
A.
Incorrect. The surveillance requirement is for a complete cycle. Opening the valve would ONLY meet half of a cycle. If the candidate believe that the other half was satisfactorily performed earlier, then the candidate would select this choice. Since maintenance was on the breaker the v i ve must be again cycled through a complete cycle (ope0 and closed)
Correct. The surveillance requirement is for a complete cycle. Since maintenance was on the breaker the valve must be again cycled throug I a complete cycle (open and closed)
Incorrect. Since maintenance was on the breaker the valve must be again cycled through a complete cycle (open and closed). The LCO addresses both the PORV and the block valve, but maintenance was only performed on the block valve. NO requirement to perform any surveillance activities for the PORV.
Incorrect. Right actions for the block valve. Wrong actions for the PORV. The LCO addresses both the PORV and the block valve, but maintenance was only performed on the block valve. NO requirement to perform any surveillance activities for the PORV.
B.
C.
D.
KIA Sys #
WA System KIA Category KIA Statement N/A Ganeric Equipment Control Knowledge of pre-and post-maintenanci operability requirements.
KIA #
2.2.21 K/A Importance 4.1 Exam Level SRO Level Of Difficulty: (1-5)
Analysis Question Source:
New Question Cognitive Level:
Higher References provided to Candidate None Technical
References:
Technical Specification 3.4.1 1 SR 3.4.1' 1; 20ST-6.6 Acceptance Criteria page 5 ar 1 6 Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.10 / 43.2)
Page 96 of 1 [IO
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY The Plant is operating at 100% power with all systems in NSA.
A turbine runback occurs.
All systems respond as designed.
The crew is stabilizing the plant in accordance with the appropriate procedure.
Control Bank "D" Group Counters are at 180 steps.
On DRPI, one Control Bank 'ID" rod indicates 196 steps; all other rods indicate 182 steps.
The affected rod has a blown movable gripper fuse and has been determined to be trippablci.
Power stabilizes at 85%
Which ONE (1) of the following describes the Technical Specification implications of this event?
A.
The rod is INOPERABLE AND NOT within alignment limits; Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure acceptable power distribution limits are maintained.
B.
The rod is INOPERABLE AND NOT within alignment limits; Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure Shutdown Margin is maintained.
C.
The rod is OPERABLE, BUT NOT within alignment limits; Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure acceptable power distribution limits are maintained.
- 0.
The rod is OPERABLE, BUT NOT within alignment limits; Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure Shutdown Margin is maintained.
Answer C
-planation/Justification:
Incorrect. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required by T.S. 3.1.4 Condition A, but rod is not inoperable if it is trippable. If the rod was untrippable, then SDM would b affected. Power distribution limits are the correct reason. Common misconception is that a rod is INOPERABLE if it is misaligned. This misconception stems for the OLD Technical Specifications where misaligned rods WERE INOPERABLE.
Incorrect. Would be true if the rod was untrippable Correct. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required by T.S. 3.1.4 Condition B. Misalignment limits are based on impact on power distribution limits.
Incorrect. Correct call on operability, but the concern for the situation presented is not shutdown margin
- 6.
C.
D.
WA Sys #
WA System WA Category WA Statement NIA Generic Equipment Control Ability to determine operability and/or KIA #
2.2.37 WA Importance 4.6 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
Bank 1LOT7 NRC Exam Question Cognitive Level:
Higher Comprehension References provided to Candidate None Technical
References:
TS 3.1.4, Condition B, and basis Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.7 143.5 145.12) availability of safety related equipment.
Page 97 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY
- 98.
The Plant is operating at 100% power with all systems in NSA.
A 0.02 g earthquake occurs resulting in the following plant conditions:
NO RPS or ESF actuations occur.
A vehicle transporting a High Integrity Container (HIC) of spent resin topples near the In-PI; nt Admin Building (IPAB).
2RMP-RQ210, Chem Sample Panel Area Radiation Monitor (718 PAB) alarms HIGH and is reading 80 mr/hr and stable. (Background level was 2 mr/hr) 2RMS-RQ223 PAF Area Monitor alarms HIGH. (Background level was 0.2 mr/hr) 2RMS-RQ223 is reading 20 mr/hr and stable.
If all of these conditions continue until T = 18 minutes, what is the highest Emergency Plan Classification REQUIRED, if any, at T = 18 minutes?
A.
No E-Plan classification required B.
Unusual Event C.
Alert D.
Site Area Emergency Answer C
ExplanationlJustification:
- 3.
C.
Incorrect. Plausible if the candidate ONLY considers the 2RMP-RQ210 reading which is below the UE threshold 1000 times background Incorrect. Plausible since earthquake would be classified as a UE.
Correct. IAW Tab 7.3 Alert indicator #I. In order to obtain the correct answer the candidate must have knowledge of the fixed radiation monixs that provide indications for emergency action levels in the E-plan. Not all fixed radiation monitors are used for indications of EAL entry. This knowledge and application of the knowledge is an SRO function. SROs are expected to know which radiation monitors are used in the EALs and apply their readings to the appropriate classification. The SRO must recognize that the radiation monitor readings posed in the stem of the question are VALID indicators of required EAL entry.
Incorrect. Plausible since both Tabs 7.3 and 5.1 refers to the fission product matrix, and a student could incorrectly apply the radiation moni.)r readings or earthquake values to obtain a Site area emergency.
D.
WA Sys #
WA System WA Category WA Statement NIA Generic Radiation Control WA#
2.3.15 WA Importance 3.1 Exam Level SRO Question Source:
New References provided to Candidate EALs Objective #:
Task ID#:
Knowledge of radiation monitoring systet IS, such as fixed radiation monitors and alar IS, portable survey instruments, personnel monitoring equipment, etc.
Level Of Difficulty: (1-5)
Question Cognitive Level:
Higher Analysis Technical
References:
10 CFR Part 55 Content:
EAL Tab 7.3 Alert indicator #I (CFR: 41.12 143.4 145.9)
Page 98 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1 SRO ONLY
- 99.
The Unit has sustained a main steam line break affecting all 3 SGs.
The crew is currently performing ECA 2.1, Uncontrolled Depressurization Of All Steam Generators.
The crew has throttled AFW flow to 50 gpm to each SG to minimize the RCS cooldown. Safety lnje 3ion Termination Criteria have NOT been met.
The following conditions exist:
SG A SG B SG C 19% WR slowly decreasing 18% WR slowly decreasing 26% WR slowly increasing Pressure 320 psig decreasing 310 psig decreasing 380 psig increasing Which one of the following describes the required action and the reason for the action?
A.
Transition to E-2, Faulted Steam Generator Isolation because there is an intact SG available.
B.
Transition to FR-H.1, Loss Of Secondary Heat Sink because there is a RED condition on the Heat Sink Status Tree.
C.
Transition to E-3, Steam Generator Tube Rupture because there is an unexplained increase in SG lavel.
- 3.
Continue with ECA 2.1, Uncontrolled Depressurization Of All Steam Generators, because Safety Injection termination is not complete.
Answer A
ExplanationlJustification:
A.
Correct. IAW LHP action of ECA-2.1 requires transition to E-2 when any one SG pressure increases. At BVPS LHP items include unexpectel response conditions such as the one addressed in this question.
B. Incorrect. Plausible, however Operator action reduced feed. Caution prior to Step 3 indicates FR-H.l would not be entered.
C.
Incorrect. Plausible. One SG is higher than the others, but does not constitute uncontrolled or unexplained increase D. Incorrect. SI termination has not been started yet, so transition to E-2 can be made.
WA Sys #
K/A System K/A Category K/A Statement N/A Generic Emergency Procedures/Plan Knowledge of the operational implicatio IS of K/A #
2.4.20 K/A Importance 4.3 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
BVPS Bank 46175 Question Cognitive Level:
Higher Comprehension.
References provided to Candidate None Technical
References:
ECA-2.1 LHP Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41.10 I43.5 / 45.13)
EOP warnings, cautions, and notes.
Page 99 of 100
Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )
SRO ONLY 100.
A.
B.
C.
D.
A large Steam break accident inside containment has occurred.
e e
e e
a Containment pressure peaked at 20 psig.
All Equipment functioned as designed EXCEPT all seal injection flow has been lost.
SI, CIA, and CIB have all been reset.
SWS has been restored to the CCP heat exchangers.
CCP flow has been restored.
While performing EOP Attachment A-I.2, Establishing RCP CCP Cooling and Seal Injectioi, the Reactor Operator is unable to OPEN 21A RCP Thermal Barrier Outlet Is01 Vlv
[2CCP*AOVIO?A], using the benchboard control switch.
In order to OPEN 21A RCP Thermal Barrier Outlet lsol Vlv [2CCP*AOVl07A] it will be necessary 0 defeat the CLOSE signal to 21A RCP Thermal Barrier Outlet lsol Vlv [2CCP*AOV107A].
IAW EOP Attachment A-I.2, Establishing RCP CCP Cooling and Seal Injection:
What directions are you REQUIRED to give the local operator to defeat the CLOSE signal to 21A RCP Thermal Barrier Outlet Is01 Vlv [2CCP*AOV107A]?
Install jumpers across the opening contacts of the valves control circuit.
Remove the valves associated secondary process rack power supply card.
Remove the valves associated control circuit power supply fuse.
Install jumpers across the contacts of the high discharge flow transmitter.
swer B
ExplanationlJustification:
A.
B.
C.
D.
K/A Sys #
WA System KIA Category KIA Statement N/A Generic Emergency ProcedureslPlan Knowledge of local auxiliary operator ta:..ts Incorrect. Although this may open the valve, it is NOT IAW EOP attachment A-1.2.
Correct. IAW EOP attachment A-I.2 step 4.a.3.
Incorrect. This action will fail the valve closed.
Incorrect. This action will only defeat the high flow signal BUT NOT the high pressure and it is NOT IAW EOP attachment A-I.2 P
during an emergency and the resultant operational effects.
K/A #
2.4.35 K/A Importance 4.0 Exam Level SRO Level Of Difficulty: (1-5)
Question Source:
New Question Cognitive Level:
Lower Fundamental References provided to Candidate None Technical
References:
EOP attachment A-I.2 step 4.b.3.
Objective #:
Task ID#:
10 CFR Part 55 Content:
(CFR: 41. I O / 43.5 I45.13)
Page 100 of 100