ML040790069
| ML040790069 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 10/30/2003 |
| From: | Ernstes M Operator Licensing and Human Performance Branch |
| To: | Scarola J Carolina Power & Light Co |
| References | |
| 50-400/04-301 50-400/04-301 | |
| Download: ML040790069 (152) | |
See also: IR 05000400/2004301
Text
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U.S. Nuclear Regulatory Commission
Site-Specific
RO Written Examination
Applicant Information
Instructions
Use the answer sheets provided to document your answers. Staple this cover sheet on top
of the answer sheets. To pass the examination you must achieve a final grade of at least
80.00 percent. Examination papers will be collected six hours after the examination starts.
Applicant Certification
All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature
Results
Examination Value Points
Applicant's Score Point!
Applicant's Grade Percen
Harris NRC Written Exaniination
Reactor Opcrator
VuEsrroN: 1
Following a Reactor Trip, the RCS temperature is being controlled by the Steam Dump
c'ontrol System at 557°F.
Given the following range of instruments. if the linit-SCO dirwts that Steam Dump
Control System be placed in the Steam Pressure mode, what approximate setpoint is
required to maintain tlc RCS temperature at 480'F?
e Sturn header pressure full range: 0-1300 psig
Steam generator pressure full range: 0-1300 psig
Turbine main steam pressure full range: 0-1500 psig
a. 37%
b. 42%
c. 58%
d. 63%
ANSWER
b. 42%
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 1 TIER/GROtJF 111
10CFR55 CONTENT: 41th) 7 43(W
KA: 000007EA1.10
Ability to operate and monitor the following a5 they apply to areactor trip: S/G pressure
OBJECTIVE: SDCS-3.0-4
Explain how the steam dump valves are automatically modulated in the steam pressure control mode,
including control alignment$, setpoint determination and adjustment, and the normal setpoint at power
DEVELOPMEXT REFERENCES: Steam Tables
OP-126 pg. 8
REFERENCES SUPPLIED TO APPLICANT: Steam
QUESTION SOWRCE: 17 NEW SIGNIFIC LY MODIFIED nDIAECT
BANK KUMBER FOR SIGNIFICANTLY IFIED /DIRECT: SDCS-R4 004
NRC EXAM HISTORY. None
DlSTKACTOR JUSTIFICACTION (CORRECT Ah3
a. Plausible if the incorrect instrument is used to dete e range ofthe instrument (551 / 1500)
X b. The equivalent steam pressure for the required RCS temperature is approximately 551 pig. This
calculates to be a setpoint of4256 (551 / 1300).
c. Plausible if the correct instrument is used to determine the range of the instrument, but the calculation
is performed incorrectly (1300 - 551 / 1300).
d. Plausible ifthe incorrect instrument is used to det the range ofthe instrument and the
calculation is performed incorrectly (1500 - 551 i
DKFFICIiLTY ANALYSIS:
fl COMPRJCHENSIVE / ANALYSIS KNOWLED
DIFFICULTY RATING: 3
EXPLANATION: Must determine required stem pressure for RCS temperature and then calculate
setpoint
Harris NRC Written Jixamination
Reactor Operator
QUESTION: 2
Given the following conditions:
e The plant is operating at 95% power during a power ramp.
The Reactor Operator attempts to perform a normal dilution for temperature control
in accordance with OP- 107, Chenrical and Volume Control System.
e ICs-151, RMW TO HOKIC AGID BI,ENUER I:CV-114B, fails to open.
Which of the foilowing actions should he taken?
a. Continue in OP-107, Chemical and Volume Control System. and perform an
Alternate Dilution
b. Increase turbine load per GP-005. Power Operation, to adjust RCS temperature
c. Go to AOP-003, Malhction of Reactor Makeup Control, and perform an
Alternate Dilution
d. Go to AOP-003. Malfunction of Reactor Makeup Control, and perform a locd
htanual Dilution
ANSWER:
d. Go to AOP-003, Malfunction of Reactor Makeup Control, and p r f o m Y local
Manual Dilution
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 2 TIEWGROUP: 1:1
1OCFH55 CONTENT: 41(b) 10 4309
KA: 000022G2.4.4
Ability to recogni7e ahnormal indications for system operating parameters which are entry-level
conditions for emergency and abnornial operating procedures. (Loss of Reactor Coolant Makeup)
OBJECTIVE: A01-3.3-R1
IIIENTIFY symptoms that require entry into AOP-003, Malfunction of Reactor Makeup Control
DEVELOPMENT REFERENCES: AOP-003, pg 12-13,25-26
REFERENCES SUPPLIED TO APPLICANT None
QUESTION SOURCE: [3 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUAMBERFOR SIGNIFICANTLY MODIFIED / DIRECT: AOP-3.3-R1 1
NRC EXAM HISTORY None
DISTRACTOR JZiSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since alternate dilution is a viable method of diluting the RCS, but with ICs-151 failed
closed, alternate dilution will not function &her.
b. Piausihle since adjusting turbine load will result in a change in RCS tempcrature, but temperature is
low requiring dilution, and raising turbine load will further low-er it.
c. Plausible since altcrnate dilution is a viable method of diluting the KCS, but with ICs-151 failed
closed, alternate dilution will not function either.
X d. With 1CS-I 51 closed, the only option available to dilute is to perform a local manual dilution.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICIJLTY RATING: 3
EXPLANATION: Analysis of the effect of a failure on the ability ofthe KMU systeltl
Harris NRC Written Examination
Reactor Operator
QUESTION: 3
Given the following conditions:
- F k plant is operating at 50% power.
- PT-457, Channel 111Pressurizer Pressure, h and all associated bistables are in
the tripped condition.
Power is subsequently lost to IJPS Bus D P -
Which of the following describes the effect of fpower on the Phase A
Containment Isolation valves?
a. NO Phase A Containment Isolation v a l ~
b. ONLY Train A Pliase A Containment I alves will close
c. ONLY Train B Phme A Containment Ks
d. All Phase A Containment Isolation valv
ANSWER:
c. ONLY Train 3 Phase A Containment Is valves will close
H n i s NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 3 TIEWGROUP: 2/ 1
10CFR55 CONTENT: 41(h) 7 43w
KA: 013K3.03
Knowledge ofthe effect that a loss or malfunction of the ESFAS will have on the following: Containment
OBJECTIVE: FSFAS- 3.0-4
PREDICT how loss of any ofthe four instrument buses Will affect the ESFAS output functions of each
SSPS train
DEVELOPMENT REFERENCES: AOP-024 pg 22
SD-103 pg9, 11, 13
REFERICNCES SUPPLIED TO APPLICANT: None
QUESTIOX SOURCE: 0 NEW SIGNIFICAXTLY MODIFIED DIRECT
RANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DlRECT: ESFAS-3 .O-R4 00 1
NRC EXAM HISTORY: None
DISTRACTOR .JUSTIFICACTION(CORRECT ANSWER Xd):
a. Plausible since Train SA slave re.lays will not actuate, but Train SB relays will still actuate..
la. Plausible since one train of Phase A will not actuate, but the hain that will not actuate is Train SA,
X c. A loss of Bus IDP-1,441 under these conditions will result in a 2 3 signal to both trains of ESIAS,
resulting in an SI and Phase A signal. Train SA slave relays, however. me powered from IDP-1 A-SI
and are energized to actuate, so Train SA slaves will not perfom their function.
d. Iiausiblc since SI and Phase A signals will be generated on both trains of ESFAS, but Train SA slave
relays will not actuate due to not having power.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE 1RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze the effect of a loss ofp0we.r on the actuation signals and determine
which power scpplies power which output relays
Harris NRC Written Examination
Reactor Operator
QUESTION: 4
Given the following conditions:
'Ihe unit is operating at 30% power.
0 A droppcd Control Rank 'C' rod has just been re-aligned.
m While attempting to operate the ROD CONrROI, AIAF3vt WSEI', the operator
inadvertently operates the ROD CONTKOI. STAKT-UP RESET.
Which of the following describes the effect of operating the incorrect reset?
a. All Control Hank 'C' rods drop into the core, causing an automatic reactor trip
h. All rods, including Control Bank and Shutdovm Rank rods, drop into the core,
causing an automatic reactor trip
e. All rods remain in their current position and there is NO effect on the Rod Control
System circuitry
d. All rods remain in their current position, but the Rod Control System circuitry
indicates all rods arc fully inserted
ANSWER:
d. AI1 rods remain in their current position, hut the Rod Control System circuitry
indicates all rods are fully inserted
Harris NKC Written b;xamination
Reactor Operator
Data Sheets
QUESTION NIIMHEK: 4 TIEWGROUP: 112
10CFH55 CONTENT: 41(b) 7 43(W
IG1: 000003AA1.02
Ability to operate and / or monitor the following as they apply to the Dropped C.ontrol Rod: Controls and
components necessary to recover rod
OBJECTIVE: KODCS-3.0-R7
IIISCIISS the effects of manipulating each of the following rod control-related switches
0 KO11 CONTKOT, START-UP RESET switch
KO11 CONTKOI, ALARM RESET switc.h
DEVELOPMENT REFERENCES: AOP-001, pg 1 1
KODCS-3.0, pg 69
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DLRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: RODCS-3.0-R7 001
NRC EXAM HISTORY. None
DISTRACTOR JUSTIFICACTION (CORRECTANSWER X'd):
a. PIausihle since improper operation of correct switch could result in rods dropping into core, but
operated switch only resets starting points for rod control circuitry.
I). Plausible since improper operation of correct switch could result in rods dropping into core, but
operated switch only resets starting points for rod control circuitry.
E. Plausible if misconception that effect is nothing if performed at power since switch is normally only
operated prior to withdrawing any rods, but operated switch resets starting points for rod control
circuitry.
X d. Operating switch at power does not affcct actual rod position, but resets rod control such that circuitr).
senses rods are at "full inserted" position.
DIFFICIJLTY ANALYSIS:
COMPREWENSIVI3 / ANALYSIS KNOWLEDGE IRECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the function of rod control system controis
._ _ __
Harris NRC Written Examination
Reactor Operatw
QUESTION: 5
Given the following conditions:
- A large steam hr& has occurred inside Containment.
During the performance of PAIII-1, the crew determined Containment pressure to he
18 psig and they verified proper operation of the Contaiimient Spray Systeni.
A transition hzs just been mads to EPP-014. Faulted Steam Generator Isolation.
Containment pressure is now 22 psig.
Which of the following actions shouid be taken regarding the increase in Containment
pressure?
a. Continue to monitor Containment pressure and transition to F W J . 1 . Response to
High Containment Pressure, if it exceeds 45 psig
b. Continue to monitor Containment pressure and transition to FRP-J. 1, Response to
High Containment Pressurc, if it remains above 10 p i g for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
c. Transition to FW-J.1, Response to IIigh Containment Pressure, to allow
verification of proper operation of the Containment Fan Cooler fans
d. Transition to ERP-J. 1, Response to High Containment Pressure, to allow
verification of proper operation of the Emergency Service Water Booster Pumps
AXSWER:
a. Continue to monitor Containment pressure and transition to FRP-J. 1, Response to
High Containment Pressure, if it exceeds 45 psig
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 5 TIEWGROW: 112
10CFR55 CONTENT: 41(b) 7 ww
KA: WE14EA1.2
Ability to operate a n d / o r monitor the following as they apply to the (High Containment Pressure)
Operating behavior characteristics of the facility
OBJECTIVE: 3.13-4
Given the following EOP steps, notes, and cautions. DESCRIBE the associated basis
8 C S F decision points
DEVELOPNIENT REFERENCES: CSFST-Containment
FRP-J.l
BD-3.13-H0, pg 5-6
REFERENCES SUPPLIED TQ APPLICANT: Xone
QUESTION SOURCE: X NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
UISTRACTOR JUSTIPICACTIQN (CORRECT .ANSWER Xd):
X a. Provided containment pressure is between 10 and 45 psig and at least one spray pump has been
veritied operating and providing flow, a transition is not required to FRP-J.1 per the CSFST as this is
only a yellow path.
b, Plausible sinw two containment spray pumps should reduce containment pressure and the liner acts as
a gas membrane to maintain leakage within limits for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at nearly design pressure, but a
transition would not be required unless containment pressure were. to exceed 45 psig.
e. Plausible since the containment fan coolers assist the containment spray system in reducing
containment pressure, but these conditions result in a yellow path only, allowing the crew to focus on
more time critical tasks, such as isolating a faulted SG.
d. Plausible since the ESW booster pumps are checked in FKP-J.l to ensure radiological releases are
minimized, but these conditions result in a yello\~path only, allowing the crew to foc.us on more time
critical tasks, such as isolating a faulted SG.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE / RKCALL
DIFFICULTY RATING: 3
EXPLANATION: Comprehension of the priority of actions to ke taken regarding contaiument
pressure
Hmis NRC Written Examination
Reactor Operator
QUESTION: 6
Given the following conditions:
- FRP-ELI, Response to a I,oss of Secondary Heat Sink, is being implemented.
e KCS bleed and feed has been initiated when Auxiliary Feedwater (AFW) capability is
restored.
e All SGs are completely dry and depressurized.
Which of the foilowing describes the strategy used to re-establish feed under these
conditions?
a. Feed ONLY one (1) SG to ensure KCS cooldown rates are established within
Technical Specification limits
b. Feed ONLY one (1) SG to limit the possibility o f a SG tube ruptnre to a single SG
c. Feed ALL SGs to establish subcooling conditions in the RCS as soon a s possible
d. Feed ALL SGs to allow termination oERCS bleed and feed as soon as possible
AVSWER:
13. Feed ONLY one (1) SG to limit the possibility of a SC; tube rupture to a single S G
Harris NRC Wrincn Exmiintion
Reactor Operator
Data Shects
QUESTXON NUMBER 6 TIEWGROIX': 1/1
lOCFKS5 CONTENT: 41(b) 8/10 43(b)
KA: 000054AK1.02
Knowledge ofthe operational implications of the folIowing wnwpts as they apply to Loss of Main
Feedwater (MFW): Effects of feedwater introduction on dry S/G
OBJECTIVE: 3.1 1-4
Given the followiiig EOP steps, notes, and cautions, DESCRIBE the associated basis
- Feed restoration
DE.VELOPME.NTKEFEKENCES: FRP-13.1, pg 47
LP-3.11, pg 12
REFERESCES SUPFLIED TO APPLICANT: None
QIJESTION SOXJTKCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORKECT ANSWER X'd):
a. Piausihle since feed is established to only one dry SG, but the reason is to ensure any subsequent
failures due to thermal shock are limited to a single SG.
X b. Flow should only be established to one dry Xi so that if excess thermal shock causes failure, the
failure is limited to one SG.
c. Plausible since RCS subcooling is a desirable condition to achieve, but only one SG at a time is fed,
d. Plausible since terminating RCS bleed and feed i s a desirable condition to achieve. but only one SG at
a time is fed.
UIFFICCLTY ANALYSIS:
0 COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge ofthe requirements for feeding a dry SG and the reasons for the.se
actions
Harris NRC Written Examination
Reactor Operator
QCESTION: 7
Given the following conditions:
0 The unit is at 100% power.
Thc running CSIP trips at 0930.
0 AOP-018, Reactor Coolant Pump Abnormal Conditions, actions have been
completed and the standby CSIP is started at 0933.
Which of the foollouring actions should be taken to establish seal cooling to the RCPs in
accordance with A0IP-O18?
a. Adjust HC-186.1~RCI SEAL WTR INJ FLOW, to estabbiish 8 to 13 gprn seal
injection flow
b. Adjust HC-186.1, RCP SEAL>WTK INJ FI,OW, to establish a 1F per minute
cooldown rate of the seals until 8 to 13 gpm seal injection flow is established
c. Locally adjust ICs-340 / 381 i422, RCP A , R / C SEN, INJ MANUAL ISOI.. to
establish 8 to 13 gprn seal injection flow
d. Locally adjust 1CS-340 / 381 i422. RCP A i B / C SEAL INJ MAKUAI, ISOL, to
establish a 1°F per minute cooldown rate ofthe seals until 8 to 13 gpm seal
injection flow is established
ANSWER
a. Adjust HC-186.1, RCP SEAL WITR INJ FLOW. to establish 8 to 13 gpm seal
injection flow
Harris NRC Written Exmination
Reactor Operator
Data Sheets
QUESTION NUMBER: 7 TIEWGROUP: 2/1
1QCFR55CONTENT: 41(b) 7 43m
KA: 003A4.01
Ability to manually operate andor monitor in the control room: Seal injection
OBJECTIVE: AOP-3.18-3
Gnen a set of plant conditions and a copy of AOP-018, DETliRvqINE the appropriate response
DEVELOPMENT REFERENCES: AOP-Ol8, P 38
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW SIGNlFICANTLY MODIFIED DIRECT
RANK N1JTMBE.R FOR SIGNIFICANTLY MODIFIED I DIRECT: Harris I .OCT BO4 073
NRC EXAM HISTORY: None
DISTRACTOR .IVSTIFICACTION (CORRECT ANSWER Xd):
X a. With seal injection flow lost for less than 5 minutes, seal injection can he established by adjusting HC-
186.1 without concern for damage to the. seals.
b. Plausible since this action would he taken if seal injection flow was lost for more than 5 minutes, but
it is not necessary to consider the cooldown rate if lost for less than 5 minutes.
c. Piausihle since these actions would be taken if the cause of the loss of seal injection flow was other
than a tripped CSIP and the flow was lost for lcss than 5 minutes, but with the loss of the CSIP as the
cause, I<:-186.1 is used.
d. Plausible since these actions would be taken ifthe cause of the loss of seal injection flow was other
than a hipped CSIP and the flow was lost for more than 5 minutes, but with the loss of the CSIP as the
cause, HC-lX6.I is used.
DIPFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPL4NATION Comprehension oftlle effect of a short term loss of seal injection flow to the
fIarri7 XRC U7rinnExmination
Reactor Operator
QUESTION: 8
F,PP-008, SI Termination, directs resetting SI.
Which of the following describes the effect of operating only ONE (1) of the two (2) Si
RESET switches at this step instead of both!
a. Bypass - Permissive Light Panel light 4-1, SI AC1IJAIE. wou1d blink due to
only one train of SSPS having an SI signal
Bypass -Permissive Light Panel light 5-2, SI RESET - AUTO SI BI,OCKEI>,
would blink due to only one train of SSPS having SI reset
b. Bypass - Pe.rmissive Light Panel light 4-1, SI AC?TJAIE, would extinguish
due to neither train of SSPS having an SI signal
0 Bypass Permissive Light Panel light 5-1, SI RESET - ~.4tJTO . SI BLOCKED,
would light due to both trains of SSPS having SI reset
c. e Bypass -Permissive Light Panel light 4-1, SI ACTCATE, would blink due to
only one train of SSPS having an SI signal
0 Bypass -Permissive Light Panel light 5-1, SI RESEI - AUTO SI BLOCKED,
would light due to both trains of SSPS having auto SI blocked
d. Bypass -Permissive Light Panel light 4-1, SI ACTIJATE, would extinguish
due to neither train of SSPS having an SI signal
Bypass - Permissive Light Panel light 5-1, SI RESET - ALTI-0 SI BI.OCKED,
would light due to both trains of SSPS having auto SI blocked
ANSWER:
a. Bypass -Permissive Light Panel light 4-1, SI ACIUATE, would blink due to
only one train of SSPS having an SI signal
e Bypass - Permissive Light Panel light 5-1. SI W,SE?- AITIO SI BLOCKED.
would blink due to only one train of SSPS having SI reset
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 8 TIicwGROuP: 211
10CFR55 CONTENT: 41@) 7 43W
KA: 006K4.11
Kiiowlcdge of ECXS design featuds) and/or interlock(s) which provide for the following: Kcset of SIS
OBJECTIVE:: SIS-3.0-R4
I)E?ERMINF SI§ status front the following
0 Bypass-Permissive L.ight Box
DEVELOPMENT REFERENCES: SD-103pg 13
Fuiictional Diagrams Safeguard
Actuation Signals Sheet 8
EOP17-21 Ilandout
SOER 94-1 Related Industry Events
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTIONSOURCE: II]NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT. INPO IO73
NRC EXAM HISTORY: None
DISTRACTOR JVSTIFICACTION (CORRECT ANSWER Xd):
X a. Operating only one switch only resets SI in a single train of SSPS. This would result in a disparity
between the two trains of SSPS for both the reset and the actuation signals so both lights would blink.
b. Plausible since the SI htuation switch only requires a single switch to actuate SI, but the reset
switchcs are train-rekdted.
c. Plausible since only train of SI wouid be reset so window 4-1 would he responding correctly, but
window 5-1 would also be blinking due to the disparity between trains.
d. Plausible since the SI Actuation switch only requires a single switch to actuate SI, hut the reset
switches are train-related.
DIFFICULTY ANALYSIS:
COMPREHENSI\E I ANALYSIS c] KNOWLEDGE I RECALL
DXFFICIJLTYwrIiw;: 3
EXPLANATION: Comprehend the effect of only operating a single train switch on SSPS and how
the indications would be affected
rams NKC Written Examination
Reactor Operator
QUESTION: 9
Given the following conditions:
m The unit is at 100% power.
Power has been lost to IDP-1A-SIII. Instrument Bus Ill. and actions are being taken
in accordance with AOP-024. 1,oss of Inintemptible Power Supply.
PT-953, Contaiiment Pressure Channel IV, then fails high.
Wliich ofthe following describes the effect on the Safety Injection (SI) and Containment
Spray Actuation Signal (CSAS) systems?
a. Neither an SI nor a CSAS .rvould occur
b. An SI would occur; a CSAS woiild NOT occur
c. An SI would NOT occur; a WAS would occur
d. Both an SI and a CSAS urould occur
ANSWER:
b. An SI would occur; a CSAS would NOT occur
IIwris NRC Written Examination
Reactor Operator
Data Shectv
QUESTION NUMBER: 9 TIEWGROUP: 2! 1
10CFR55 CONTENT: 42(b) 7 43m
KA: 013K6.01
Knowledge of the effect o f a loss or malfunction on the following will have on the ESFAS: Sensors and
dete.ctors
0B.JECTIVE: CSS-RI
Given a set of plant conditions or the status of each bistable light box, DETERMINE which of the
following ESFAS signals are active
Safety injection (SL)
Containnient Spray Actuation
DEVELOPMENT REFERENCES: S1>-103, pg 1 I , 64,68
REPE.RENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: Harris L(3Cx- 139
NRC EXAM HISTORY: None
DISTRACTOR dIJSTIVICACTIQN (CORRECT ANSWER Xd):
a. Plausible since CSAS is energked to actuate and i channel is in a deenergized condition so CSAS will
not occur. but the 2 failed channels will cause an SI actuation.
X b. An SI actnation (deenergized to actuate) will occur, but a CSAS (energized to actuate) will not occur
unless another energized channel senses a high pressure condition.
e. Plausible since one of the two signals is energized to actuate and the other is deenergized to actuate,
but SI i s deenergize to ac.tuate and CSAS is energized to actuate.
d. Ilausible since the 2 failed c.hannelswill cause an SI actuation, hut CSAS is energked to actuate and 1
channel is in a deenergized condition so CSAS will not occur.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOVLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Comprehension of the effect of multiple failed channels on ESFAS signals
Harris NRC Written Examination
Reactor Operator
QUESTION: IO
Given the following conditions:
The plant is operating at 43% power.
0 l2OVAC Vital Bus IDP-113-SI1 deenergizes.
Outward rod motion is inhibited by . . .
a. C-4, OPAY rod stop.
b. C-4, OTAT rod stop
c. C-2, Power Range rod stop.
6. C- 1,Intermediate R a ~ g rod
e stop.
ANSWER:
e. C-2, Power Range rod stop.
Harris NRC Written lixamination
Reactor Operator
Data Sheets
QUESTION NIJMBER: 10 TIERGROUP: 22
10CFR55 CONTENT: 41(b) 7 43w
KA: 001 K4.07
Knowledge of CRIX design feature(s)andor interlock(s) which provide for the following: Rod stops
OBJECTIVE: NIS-3.0-9
DISCUSS the operation of the following NI trip-reiated functions:
b. SR,IR and PR (low) trip blocks
DEVELOPMENT REFERENCES: OP- 105 pg 26
AOP-024 pg 6
REFERENCES SUPPL1E.DTO APPLICANT: None
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NIJMBEN FOR SIGNIFICANTLY MODIFIED /DIRECT. NIS-K6 003
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (COIPKECT ANSWER Xd):
a. Plausible since causes rod stop, but coincidence is 2 4 instead of 1/4.
b. Plausible since causes rod stop, but coincidence is 2!4 instead of 114.
X e. PR rod stop is 14! coincidence. With S2-SB deenergizeed, PR X-42 is tripped.
d. Ptausible since this causes a rod stop, and coincidence is 1/2, but IR rod stop is blocked above P-I 0 by manual
operator action. Must have 2/4 PR beiow P-10 to reset.
DIFFICDLTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS [I]KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze effect of loss of power on XIS and rod control and determine effect of
single channel tripped
Harris NRC Written Examination
Reactor Operator
QUESTION: 11
The basis for the operation of the 1:lectricHydrogen Recombiners is to minimize
hydrogen concentration build up in Containment following a I,OCA due to the ...
a. Arc-water reaction and release of hydrogen from the PKT.
b. corrosion of metals in Containment aid release of hydrogen from the RCDT.
c. release ofhydrogen from the PKT and the radiolytic decomposition of water.
d. radiolytic decomposition of water and the corrosion of metals in Containment.
ANSWER
d. radiolytic decomposition ofwater and the corrosion of metals in Containment.
Harris NKC Written Examination
Reactor Operator
Data Sheets
QliESTION NUMBER 1 1 TIEWGRQW: 212
K h IMPORTANCE: RO 3.4 SRQ
10CPR55 CONTENT: 41(b) None 43(b) 2
Kh: 028G2.2.22
Knowledge of limiting conditions for operations and safety limits. (Hydrogen Recombiner and Purge
Control)
OBJECTIVE: IIK-3.0-1
STATE the purpose and function of the Hydrogen Recombiner System, including the following
components:
Electric hydrogen recombiner
DEVELOPMENT REFERENCES: TS 3.6.4.2 Hasis
SD-125 pg 21
12-IIK-3.0 pg 5
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: h%W 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGXIFICANTLY MODIFIED / DIRECT: HR 01
NRC EXAM HISTORY: None
DISTRACTOR J[JSTIFICACTION (CORRECT ANSWER Xdd):
a. Piausibk since Electric Hydrogen Kccombiners are designed to re.move hydrogen in containment
following a 1,OCA due to generation from the zirc-water reaction, hut not due to release from the
Im.
b. Plausible since Electric Hydrogen Recombiners are designed to remove hydrogen in c.ontainment
following a ILEA due to generation from the corrosion of metals in containment, but not due to
releaye from the KCD?.
E. Plausibie since Electric IIydrogen Recombiners art designed to remove hydrogen in containment
foilowing a LOCA due to generation from the radiolytic decomposition of water, hut not due to
release from the PR?.
X d. The Electric I1ydroge.n Kecombiners are designed to remove hydrogen in containment following a
1,OCA due to generation from the zirc-water reaction, radiolytic decomposition of water, and
c.orrosion of metals in containment.
DLFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of l e c h Spec basis for hydrogen recombiners
Harris NRC Written Examination
Reactor Operator
QUESTION 12
EPP-001, Loss ofAC Power to 1A-SA and ID-SB Buses, is being performed.
Concurrent to thr loss of power, a small break LCXA occurred.
The c.rew has conipleted the following actions when off-site power is restored to 6.9 KV
BUS1A-SA:
e Sequencers h a w been de-energized
e Safeguards pumps autostarts have been disabled
e RCP seals have been isolated
e MSIVs and F W V s have been closed
e Ikpressurization of SGs to 180 psig has commenced
Which of the following actions is the FIRST to be taken following the restoration of off-
site power?
a. Start an ISSW pump
b. StartaCSIP
c. Stabilize S G pressures
d. Initiate SI
ANS\VEK:
c. Stabilize SG pressures
Harris XRC Written Examination
Reactor Operator
Data Sheets
QI:ESTION NUMBER: 12 TIEWGROWP: 11:
10CFR55 CONTENT: 41(h) 7 43m
KA: 000055EAI.07
Ability to operate and monitor the following as they apply to a Station Blackout: Restoration of power
from offsite
OBJECTIVE: 3.7-5
Given a title of a continuous action step from a foldout and a list of plant conditions, DEIERN4lFX if
implerncntation is required
DEVELOPMENT KEFEREYCES: EPP-001 pg 3 5 , 3 S
REFERENCES SUPPLED TO APPLICANT: None
QITESTION SOURCE: X NEW SIGNIFICANTLY MODIFIED DIRECT
RANK NURIBER liOK SIGNIFICANTLY MODIRIED i DIRECT: New
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTKON (CORWLT ANSWE.R Xd):
a. Plausible sinc.e if the power source was an EDG instead of oKqite power, it would be important to
provide cooling flow to the EDG.
b. Plausible since H sniall break LOCA exists and RCS inventory is being lost, but the first action is to
stabilize SG pressure.
X c. Upon restoration of power to at least one bus, the first adion taken is to stabilize S G pressures.
d. Plausible since a srnall break LCXA exists and RCS inventory is being lost, but the first action is to
stabilize SG pressure.
DIFFICIJLTY ANALYSIS:
C:OI\IPREHENSIVE / ANALYSIS KNOWLEDGE i RECALL
DIFIWIJLTY RnTmG: 3
EXPLANATION: Knowledge of required actions when power is restored following a loss of all
AC power
IIarris NRC Written Exnmination
Reactor Operator
QUESTION: I3
mhiie performing an Operating Procedure, the Reactor Operator comes to a step which
states:
Request Chemistry to sample the RHT for boron concentration.
The Reactor Operator believes the step is NOT essential to achieving the purpose for
which the procedure is being used and that the omission of the step does NOT violate the
precautions and limitations of the Operating Procedure.
Which of the following is the MINIMUM requirement(s) that must be met to allow
marking the step /A?
a. 0 Step mwt be initialed by the Reactor Operator prior to perlbrmance
h. 0 Step must be initialed by the Reactor Operator prior to performance
0 A written explanation of why the step is N/A must be provided in the
Comments section of thc procedure
c. 0 Step must be initialed by the SC:O prior to performance
a. .
0
Step must be initiaied by the SCO prior to performance
A written explanation of why the step is N/A must be provided in the
Comments section of the procedure
ANSWER:
d. Step must be initialed by the SCO prior to perfornradlce
0 A written explanation of why the step is N!A must be provided in the
Comments section of the procedure
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 13 TIEWGROUP: 3
10CFR55 CONTENT: 41(h) None 43(b) Xone
KA: 2.1.23
Ability to perform specific system and integrated piant procedures during all modes of piant operation
OBJECTIVE: PP-2.0-2
DISCUSS the requirements in PRO-NGGC-0200 concerning the following:
Procedure user's responsibilities
DEVELOPMENT REFERENCES: PRC)-XCtCiC-O200 pg 11-12
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 X NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NCMBER FOH SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JWTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since the RC) disc,overedthe cause for marking the step N!A, but a supervisor must initial the
step prior to performance and a written explanation must be. provided in the Comments section.
b. Plausible sinc.e a written explanation must he provided in the Comments section, but a supervisor must
initial the step prior to performance.
c. Plausible since a supervisor must initial the step prior to performance, but a written explanation must
he provided in the Comments section.
X d. The step is initialed by the responsible supervisor prior to perfom~anceand a written explanation is
provided in the Comments section.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DKPFICULTY RATING: 2
EXPLANATION: Knowledge of use of W A during procedure wdge
.. ~ __ .I._......
Harris NRC Wrimn Examination
Reactor Operator
QUESTION: 14
A new Progress Energy employee was working at another nuclear utility for the first six
( 6 ) months of this year. His occupational total effective dose equivalent (TEDE) at the
other utility has been documented 8s being 500 mRem for this year.
What is maximum additional T H E that he win receive during the remaining six ( 6 )
months of the year as a Progress Energy employee without exceeding his Annual
Administrative Dose Limit, assuming no extensions are approved'?
a. 1500 mRem
b. 2000 mKem
c. 3500 mRem
d. 4500mKem
ANSWER:
b. 2000mRem
Harris NRC Written Examination
Reactor Operator
Data Sheets
QIJESTION NUMBER: 14 TIEWGROUP: 3
10CFR55 CONTENT: 41(b) 12 43m
KA: 2.3.2
Knowledge of facility ALAKA program
OB.ECTI\E: RP-3.5- 14
State the 10CFK20 and corporate occupational dose limits for individuals
DEVELOPMENT REFE.RENCES: NGGM-PM-002, pg 11
REFERENCES SIJPPIJED TO APPLICANT: None
QUESTIQN SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NIJMBER FOR SIGNIIFICAKTLY MODIFIED / DIRECT: PP-3.7-RI 002
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since. the annual Progress Energy dose limit is 2 Rem and he has already rcc.eived 500
mKem this year, but occupational dcse from another utility is not considered in the 2 Kern lirnitatioii
unless he would exceed 4 Rem combined for the 2 utilities.
X b. Personnel annual Progress Energy TEDE shall not exceed 2 Kern and 4 Rem total dose if non-
Progress Energy cccupdtional dose for the current year is determined.
E. Plausible since he is permitted to receive a total of 4 Rem between the 2 utilities and he already has
500 mRem, but the more limiting is the 2 Rem Progress Energy dose.
d. Plansible since 500 mRem and 4500 mRcm would equal the employees legal limit of 5000 mRem,
but this is greater than the administrative limit of 2000 mKem.
DIFFICU1,TY ANALYSIS:
COMPREHENSIVE / ANALYSIS ICNOWLEDGE /RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of administrative dose limits
IIarris NRC Written Examination
Reactor Operator
QUESTION: 15
Given the following conditions:
e A small break 1,OCA has occurred.
e Containment pressure is 3.8 p i g and increasing.
Containment temperature is 137 Ob' and increasing.
The expected Containment Cooling Fan alignment will be one (1) fan in each
Containment Fan Cooler hit running in .. .
a. high speed with the post-accident dampers slut.
b. high speed with the post-accident dampers open.
c. low speed with the post-accident dampers shut.
d. low speed with the post-accident dampers open.
ANSWER
d. low speed with the post-accident danipers open.
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 15 TIENGROUP: 2'1
1OCFR55 CONTENT: 41(b) 7 43W
KA: 02262.1.2X
Knowledge. ofthe purpose and function of major system components and controls. (Containment
Cooling)
OBJECTIVE: C:CS-3.0-K2
PREIXC'I' the response(s) of the Chtainment Cooling Snhsystems to the following signals
s1
DEWLOPMENT REFERENCES: SD-169, p 14
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOIJRRCE: NEW SIGNIFICANTLY MODIFIED [ziDIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CCS-R4 001
NRC EXAM HISTORY: Xone
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since this alignment is an alignment that would be used following a luss of oEsite power, hut
the Si alignment has the fans in low speed.
b. Plausible since this alignment is an alignment that would be used followringa loss of offsite power
with thc dampers aligned for the SI alignment, but the SI alignment has the fans in low speed.
Plausible since the fans arc aligned per the SI alignment, but the dampers are aligned per the loss of
offsite power alignment.
X d. Following an SI actnation, the containment fan coolers shifl to low speed and the post-accident
dampers open.
DIFFICULTY ANALYSIS:
0 COMPREHENSIVE /ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the response of Containment Cooling tfl an SI signal
IIarris NRC Written Examination
Reactor Operator
QUESTION: 16
Following a Reactor Trip and Safety injection due to a RCS leak, the Critical Safety
Function Status Trees (CSFST) are being monitored.
When monitoring the CSFS? for RCS Inventory, if PKZ level is indicating greater than
92%, why is a check of RVLIS then performed?
a. Determine if the cause of the high I E level is excessive RCS inventory or
voiding in the Reactor Vessel head
b. Determine if SI termination criteria is met to allow reducing the excessive RCS
inventory
c. Determine if Adverse Containment conditions have cilused erroneous indications
of the PRZ level instruments
d. Determine if the cause of the high PRZ level is excessive KCS inventory or
expansion due to an RCS heatup
AILSWEK.
a. Determine if the cause of the high IFU level is excessive RCS inventory or
voiding in the Reactor Vessel head
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NVMHER: 16 TIEWGROUP: 111
1QCFR55CONTENT: 41(b) 7 43m
KA: 000008G1.1.28
Knowledge ofthe purpose and function of major system components and controls. (Pressurizer Vapor
Space Accident)
OBJECTIVE: 1CCM-3.0-1
LIST the two major function5 of the Inadequate Core Cooling Monitor (ICCM)
DEVELOPMENT REFERENCES: EOP Background for Inventory Status Tree, F-0.6, p 8
LP3.12, pg 7
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: 3.12 001
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER Xd):
X a. Once a determination has been made that PRZ level is full RVLIS is then used to confirm whether the
cause ofthe full PRZ is excessive inventory or voiding in the head region.
h. Plausible since RVLIS is used throughout the EOP network to determine if SI termination criteria has
been met, but in this instance it is used to determine the cause of the high PRZ level.
e. Plausible since a steani space hreak in the PRX will affect the level indications, hut RVLIS is used to
deterniine the cause of the PRZ high level condition.
d. Plausible since RVLlS is part of the Inadequate Chre (holing Monitoring System and a heat up of the
RCS will cause expansion ofthe RCS, but but KVLIS is used to determine the cause of the P W high
level condition.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE /RECALL
DIFFICIJLTY R4TING: Knowsledge ofthe purpose of monitoring RVLIS during accident conditions
EXPLANATION:
Harris NKC Written Examinatinn
Keactor Operator
QUESTION: 17
Given the following conditions:
The plant is shutdonn for work on Reactor Coolant Pump seals.
The Reactor Vessel Iread is still installed.
8 The running Residual Heat Kemoval (MI<) pump trips and the crew is unable to start
the standby RHR pump.
Time to reach core boiling is determined to be 26 minutes.
Time to reach core boil-off is determined to be 53 minutes.
Of the following two (2) methods of RCX makeup. in accordance with AOP-020, Loss
of RCS Inventory or Residual Heat Removal While Shutdown, which of the foliowing is
the PREFERRED method of makeup and why is it preferred over the other method?
a. Gravity feed from the RWST to the RCS is preferred over starting a CSIP since
starting a CSIP under these conditions would violate Technical Specifications
b. Gravity feed from the RUST to the RCS is preferred over starting a CSIP since
Reactor Makeup to the CSIP may be insufficient to makeup for core boil-off
C. Starting a CXP is preferred over gravity feed from the RWST since gravity feed
flow may be insufficient to makeup fbr core boil-off even if the RCS is
depressurized
d. Starting a CSIP is prefemd over gravity feed from the KWSI since the KCS nmy
be pressuriLed and prohibit gravity flow
ANSWER:
d. Starting a CSIP is preferred over gravity feed from the RWST since the RCS may
be pressurized and prohibit gravity flow
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NU-IBER: 17 TIERKROUP: 1/1
10CFR55 CONTENR 41(b) 8/10 43(b)
KA: 00002SAK3.01
Knowledge of the reasons for the following responses as they apply to the Loss of Residual Heat
Remcival System: Shift to alternate flowpath
ORJECTI\E: AOP-3.20-3
Given a set of entry conditions and a copy of AOP-020, DETERMINE the appropriate response
DEVELOPMENT REFERENCES: AOP-020, pg 9
A01-020-BD, pg 19
FUWERENCES SIJPPLIED TO APPLICANT: None
QUESTION SOURCE: X NEW SIGNIFICANTLY MODIF1E.U DIRECT
BANK NIJTMBERFOR SIGMFiCANTLY MODIFIED / DIRECT: New-
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFKCACTION (CORKECT ANSWER Xd):
a. Plausible since 1s requires that a CSIP he made inoperable before these plant conditions are
established, but GP-OOX requires that at least one CSIP he functional under these conditions.
h. Piausible sinc.e the CSIP can provide more flow than Reactor Makeup is capable ofproviding, but the
suction source for the CSlP would he the RWST.
6. Plausible since starting a CSIP is preferred to gravity feed. but only hecause the RCS may he
pressurized. If the RCS is depressurized, gravity feed will provide adequate flow.
X d. If the RCS is pressurized, gra1:ity flow may he insufficient to provide adequate makeup to the RCS.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE I RECALL
DIFFICIJLTY RATING: 3
EXPLANATION: Analysis of piant conditions to determine appropriate response and reason for
response
Harris XRC Writeen Examination
Reador Operator
QUESTION: 18
Given the following conditions:
Containment temperature is 96 "F.
- Containment Fan Coolers (AII-1 / 2 / 3 / 4 we operating in the Normal Cooling
Mode.
A loss of offsite power occurs and the piant responds as expected.
The Containment Fan Coolers should be aligned with one (1) fan associated with each
fan cooler operating in ...
a. high speed and discharging to the concrete airshaft
b. high speed and discharging to the post-accident discharge duct
c. low speed and discharging to the concrete airshaft
d. low speed and discharging to the post-accident discharge duct
AXSWER:
a. high speed and discharging to the concrete airshaft
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 18 TIERKROUP: Lil
KAIMFORTANCE: KO 2.7 SRO
10CFR55 CONTENT: 41(b) None 43(b) 5
Kri: 000056AA2.09
Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Operational
status of reactor building cooling unit
OBJECTIVE: C C S - ~ . O - K ~
PREDICT the response(s) of the Containment Cooling Subsystems to the following signals.
0 LOSP
DEVELOPMENT REFERENCES: SI)-169 pg 14
REFERENCES SUPPLIED TO APPLICANT None
QuesTIoN SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK XUM5E.K FOR SIGNIFICANTLY MODIFIED / DIRECT: CCS-R4 001
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICXCTION(CORRECT ANSWER Xd):
x One fan per unit will start on high speed and discharge to the c.oncrete airshaft.
h. Plausible since one fan per unit will start on high speed, hut the discharge is to the concrete airshaft
not the post-accident discharge duct.
c. Plausible since this fan response is the response to a LOCA start signal and they do discharge to the
c.oncrete airshaft, but the fans operate in high speed following a loss of ofkite power.
d. Plausible since this is the response to a I J X A start signal, but the fans operate in high speed and they
discharge to the concrete airshaft following 8 loss of ofisite power.
DIFFICULlY ANAL.YSIS:
[7 COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL.
DIFFICULTY RATING: 3
EXPIANATION: Knowledge of the response ofthe c.ontainnient fan cooler fans to a loss of
offsite power
Harris NRC Written Examination
Reactor Operator
QUESTION: 19
Given the following conditions:
The crew has determined that control rod F- 10 in Control Bank D is misaligned by I 8
steps.
Actions are being performed in accordance with AOP-001, Malfunction of Rod
Control and Indication System.
The crew will attempt to align control rod F- 10 and thc remaining rods in Control Bank D
by placing the Rod Selector Switch to . . .
a. HANK D and opening the lift coil disconnect switches for the remaining rods in
Control Bank D.
b. MANUAL, and opening the lift coil disconnect switches for the remaining rods
Control Rank D.
c. HAKK D and opening the lift coil disconnect switch for control rod F-10.
d. MANIJAI, and opening the lift coil disconnect switch for control rod F-10.
ANSWEN:
a. BANK D and opening the lift coil disconnect switches for the remaining rods in
Control Hank D.
Harris NRC Written Examination
Kcactor Operator
Data Sheets
QUESTION NUMBER 19 TIEWGROUP: 1 i2
10CFR55 CONTENT: 41(b) 7 43@)
KA: 000005AK2.02
Knowledge of the interrelalions between the Jnoperable I Stuck Control Rod and the foliowing: Breakers,
relays, disconnects, and control room switches
OILTECTIVE: AOP-3.1-6
Given a set of plant conditions and a copy of AOP-001.1>ElERMINT, the appropriate response.
DEVELOPMENT REFERENCES: AOP-001 pg 17-1 8
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE:
0X NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
X a. The affected individual bank position should be selected and tbe inoperable rod will he attempted to
be moved by opening the lift coil disconnect switches for the remaining rods in the bank.
b. Ilausible siuce the inoperable rod will be atrempted to he moved by opening the lift coil disconnect
switches for the remaining rods in the bank, but the affected individual bank position should be
selected.
E. Plausible since the affected individual bank position should be selected, but the inoperable rod will be
attempted to he moved by opening the lift coil disconnect switches for the. remaining rods in the bank.
d. Plausible since. the inoperable rod is in Bank 11, but movement should be attempted by using the
individual bank select position.
DIFPICIJLTY ANALYSIS:
COMPREHENSIW /ANALYSIS KNOWLEDGE /RECALL
UIFFICTJLTY RATING: 3
EXPLANATION. Knowledge of the means for a misaligned rod per procsdnre
Harris NRC Written Exainir~atiori
Reactor Operator
QTXSTION: 20
Given the following conditions:
0 I:,RFIS is inoperable.
0 Plnnt parmmeters are as foilows:
0 ICCM highest TC' = 672' F
RCS U'R temperature (highest) = 688" F
RCS pressure PT-440 = 1535 psig
0 RCS pressure PT-402 1635 psig
7
e CNhlI' pressure PT-9.51 = 4.5 psig
What value of superheat should be reported?
a. 63 *F
b. 71 "I'
e. 79 "F
d. 89°F
ANSWER:
a. 63 "F
Harris KKC Written lixamination
Rcactor Operator
Data Sheets
QUESTION NUMBER: 20 TIEWGRBUP: 112
10CFR55CONTENT. 41(b) None 13(b) 5
KA: 000074EA2.01
Ability to determine or interpret the following as they apply to a Inadequate Core Cooling: Suhcooling
margin
OBJECTIVE: 3.19-4
Given a set of conditions during EOP implementation, DETERMINE the correc.t respnse or required
action based upon the EOP User's Guide general information
Deterniining an RCS subcooling value
DEVgLOPMENT REFERENCES: Users Guide, pg 27,34-35
REFERENCES SUPPLIED TO APPLICANT: S t e m Tables
QUESTION SOIJRCE: 0 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOK SIGNIFICANTLY MODIFIED / DIRECT: 3 . 1 9 4 4 003
NHC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (C.!ORRECT ANSWER X'd):
X a. When EKFlS is not available, the highest ICCM temperature should be used. If EWIS is not
available and adverse containment conditions exist, P1'-402 should be used for pressure. Saturation
temperature for 1635 psig is 609 "F, so the amount ofsupcrheat is 63 '1: (642-609).
b. Plausible since the superheat determined using the ICCM temperature and saturation for the lowest
RCS pressure of 1535 p i g (not used because ofadverse wntainment conditions) is 71 "F (672-601)
c. Plausible since the superheat determined using the hot leg temperature (not 1ise.d if ICCM is available)
and saturation for the IT-402 pressure of 1635 psiig is 79 "I: (688-609).
d. Plausible since the superheat determined using the hot leg temperature (not used if ICCM is available)
and saturation for the Inwest RCS pressure of 1535 psig (not used because of adverse containment
conditions) is 87 "F (688-601).
DIFFICIJLTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of instruments to use and calculation of subcooling by applying
steam tables
Harris NKC Written Examination
Reactor Operator
A failure of a Containment Fan Cooler Unit. while the system was aligned to maximum
cooling mode. causes equilibrium Containment temperature to increase from 119 'F to
126 O F .
How does Pressurizer ievel indication change due to this increase in Contdrment
temperature?
a. Level indicates higher than actual due to reference leg density decreasing
b. 1,evel indicates lower than actual due to reference leg density decreasing
e. Level indicates higher than actual due to reference leg density increasing
d. Level indicates lower than actual due to reference leg density increasing
ANSWER:
a. Level indicates higher thnn actual due to reference leg density decreasing
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 21 TIEWGROUP: 2/1
IUCFR55 CONTENT: 41(b) 7 43w
KA: 022K3.02
Knowledge of the effect that a loss or malfunction of the CCS will have on the following: Containment
instrumentation readings
OBJECTIVE: PZKLC-3 .O-4
I)ESC'URI how various errors would affect the pressurizer lcvel indication in the Main Control Room
DEVELOPMENT RKFE.KENCES: 1.P-PZKLC-3.0 pg 10
REFERENCES SUPPLIED TO APPLICANT: None.
QlJEsTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NITMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. Reference leg density decreases as containment temperature increases which causes level lo indicate
higher than actual.
b. Plausible since reference leg density changes as containment temperature inweases which causes level
to indicate different than actual.
c. Plausible since reference leg density changes as containment temperature increases which cduses level
to indicate different than actual.
d. Plausible since reference leg density changes as containment temperature inc.reases which causes level
to indicate different than actual.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFPICIJLTY RATING: 3
EXPLANATION: Analyze the effect of thc temperature change on pressurkzer level
Harris NRC Written Exmiination
Reactor Operator
QUESTION: 22
Given the following conditions:
'The unit is operating at 12Y0power.
The following KCP vibrations are observed:
INDICATION RCP 'A' Iii2.22 l!izx!
Frame Vibration 3.6 mil and ? at 2.8 mil and stable 4 init and 1' at
0.3 mil per hr 0.1 mil per hr
Shaft Vibration 12 mil aid ?' at 7 mils and stable I 4 mils and 1'at
0.3 mil per hr 0.6 mils per hour
Which of the following describes the actions required for this condition?
a. Stop RCP 'A' and initiate a plant shutdown
b. 'Trip the reactor, stop RCP 'A', and go to PAI'II-1
c. Stop RCI' 'C' and initiate a plant shutdown
d. 'Trip the reactor. stop RCP I C ' , and go to PATH-I
ANSWER:
a. Stop RCP 'A' and initiate a plant shutdown
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NIJMBEK: 22 TIER/GROUP: 21 1
KAIMPORTANCE: RO 2.9 SKO
10CFR55 CONTENT: 41(b) 5 43w
KA: 003A1.01
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated
with operating the RCPS controls including: RCP vibration
OBJECTIVE: AOP-3.18-3
Given a set of plant conditions and a copy of AOP-018, DETERMINE the appropriate response
DEVELOPMENT REFERENCES: AOP-018, p 28
REFERENCES SIJPPLIED TO APPLICANT: AOP-018, Attachment 1 (Sheet 2 of 2 ONLY)
QUESTION SOXJTRCE: 0 NE.W SIGNIFICANTLY MODIFIED 0 DIRECT
BANK XUMBEK FOR SIGNIFICANTLY MODIFIED I DIRECT: AOP-3.18 0 17
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. 'A' RCP Vibration has exceeded limits and the pump must he stopped. With the plant in Mode 2, a
reactor trip is not required, hut the plant must he shutdown.
h. Plausible since these would be the correct actions ifthe plant was in Mode 1, but the plant is in Mode
2.
E. Plausible since these are the correct actions, hut 'C' RCP has not reached any trip limits while 'A' KCP
has.
d. Plausible since these would he the correct actions if the plant was in Mode 1, hut 'C' RCP has not
reached any trip limits while 'A' RCP has and the plant is in Mode 2.
DIFIVCULTY ANALYSIS:
COMPRKIIENSIVE I ANALYSIS 0 KNOWLEDGE I RECALL
DIFFICGLTY RATING: 3
EXPLANATION: Analysis to determine whic.h RCP must be stopped and comparison to power
level to determine proper action
IIarris NRC Written Examination
Reactor Operator
QUESTION: 23
ALB-009-8-1~PRESSURIZER RELIEF TANK HICrII-LOW LEVEL PRESS OK TEMP,
alarms due to a high temperature condition.
Which of the following describw how the Pressurizer RelieTTank (PRI) is normally
cooled. in accordance with OP-100, Reactor Coolant System?
a. Recirculate the IRT through the Reactor Coolant Drain Tank heat exchanger,
using Component Cooling Water to cool the heat exchanger
b. Recirculate the IRT through the Reactor Coolant Drain Tank heat exchanger>
using Service Water to ccwl the heat exchanger
c. Drain the IKT to the Reactor Coolant Drain Tank while nxlking up to the PRT
from the Ilemineralized Water Storage Tank
d. Drain the P R T to the Kcactor Coolant Drain Tank while making up to the PRT
from the Reactor Makeup Water Storage rank
ANSWER
a. Recirculate the IKT through the Reactor Coolant Drain Tank heat exchanger,
using Component Cooling Water to cool the heat exchanger
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBEK: 23 TIEWGROUP: 211
lQCFR55CONTENT 41(b) 7 43m
KA: 007K4.01
Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following: Quench tank
cooling
OBJECTIVE: PZK-7 0-3
Given a flow, diagram of the PRT or associated suhsystems and the appropriate procedure, correctly
AI lGN the PRT for filling, draining, recirculation, or cooldown
DEVELOPMENT REFERENCES: APP-ALB-009, pg 29
OP-100, pg 30
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOLICE: X NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. Normal cooling of the PR'I is accomplished by recirculating the PKT water through the RCDT heat
exchanger, which is cooled by CCW.
h. Plausible since normal c.ooling ofthe PKT is accomplished by recirculating the PRT water through the
K C D i heat exchanger, hut it is cooled by CCW, not SW.
c. Plausible since a rapid cooldown of the I'RT would be accomplished by draining to the RCDT and
making up to the PRT, hut the makeup source is RfvlUW, not the DWST.
d. Plausihle since this method would he. used for a rapid cooldown of the PKT, hut is not the normal
cooldown method used.
DIFFICULTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS KNOWLEDGE / RECAIL
DIFFICULTY RATIXG: 2
EXPLANATION: Knowledge of the design method of cooling the PRT
Harris NRC Written Examination
Reactor Operator
QUESTION: 24
Which ofthe foliowing describes the effect ofa loss of 125 VIIC Bus DP-111-SA'?
a. Emergency Diesel Generator A-SA loses excitation power
b. Poww is lost to the Emergency Fscape Air Lock
e. hkster relays in SSPS 'Train A lose power
d. Main Turbine DC Rearing Oil Pump loses power
ANSWER:
a. Emergency Diesel Generator 4-SA loses excitation power
IIarris NRC Written Fkmination
Reactor Operator
Data Sheets
QUESTION NUMBER 24 TIEWGROUP: 1/1
10CFR55 CONTENT: 41(b) 7 'w)
KA: 064K2.03
Knowledge of EDG bus power supplies to the following: Control power
OBJECTIVE: AOP-3.25-3
Given plant conditions, DISCIJSS the following notes, cautions, and procedural steps as they apply
The effects of a loss of a DC bus on equipmcnt operability
DEVELOPMENT REFERENCES: AOP-3.25, p 39
REFERENCES SUPPLIED TO APPLICANT: None
QUESTrON SOURCE: [7X NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUAMBERFOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER X'd):
X a. IW-1 A-SA supplies the EDG governor and generator excitation control circuits.
b. Piansible since the emergency escape air lock is powered *om DC, but not the emergency DC bus.
c. Plausible since SSPS receives iuput from the emergency Dc bus and the master relays operate on DC,
but the emergency bus only supplies the Rx Trip Breaker shunt trip power and the master relays are
powered by 48 vdc which is produced in SSPS via the instrument buses.
d. Plausible since the DC hearing oil pump is powered by DC and is one of the only I)C Loads
speciflcally addressed in the EOPs, hut it is powered by the non-safety related 250 VIIC.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE I RECALL
DIPFICUL'TY RATING: 2
EXPLANATION: Knowledge of the source of control power to the EDGs
Harris NRC Written Examination
Reactor Operator
QUESTION: 25
Given the foilowing indications during a plant startup being performed in accordance
with GP-005, Power Operation:
Power Kange Channel N-41 26.0%
Power Range Channel E-42 24.5%
Power Range Channel E-43 24.5%
Power Range Channel N-44 25.0%
1,00p <Ai\T 25.5%
Loop R AT 255%
Loop c1 1\T 25.556
Turbine L.oad 24.5% (DEII units converted to percent load)
Which of the following power Ievels shodd be reported as being aetLd reactor power?
a. 24.5%
h. 25.056
c. 25.5%
d. 26.0%
ANSWER:
e. 25,S?h
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 25 TIEWGKOUP: 2!2
10CFRS5 CONTENT: 41(b) 5 43Qo
KA: 002K5.10
Knowledge of the operational implications ofthe following concepts as they apply to the RCS:
Relationship between reactor power and RCS differential temperature
OBJECTIVE: XIS-3.0-13
Discuss the cautions associated with monitoring NI power levels during plant start-up and power
operations
DEVELOPMENT REFERENCES: GP-005, pg 12
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: NIS-R 10 003
NRC EXAM HISTORY: None
DISIRACTOR .JUSTIFICAGTION (CORRECT ANSWER Xd):
a. PlausibIe sincc this is the lowest given power ievel and may be considered to be the most
conservative, but (3.005 provides guidelines for which power le.vel should be cunsidered.
b. Plausible since this is the average NIS power level, but the highest as identified by GP-005
requirements is the average loop AT.
X e. CJntil a calorimetric is performed at 30% power, r u e reactor power shall be assumed q u a l to the
highest of the following indicators: average Power Range NI value, average percent Ar, or h i a h
Turbine. load
d. Plausible since this is the highest given power level and may be considered to be the most
conservative, but GP-005 provides guidelines for which power level should be considered
DIFFICULTY ANALYSIS:
COMPREIXEXSIVE / ANALYSIS KNOWLEDGE I RECAI,L
DIFFICIJLTY RATING: 3
EXPLANATION: Calculation of average power indications and determination of most
conservative value
Harris NRC Written Exanination
Reactor Operator
QUESTION: 26
AII-UA, NORMAL PURGE SUPPI,Y FAN AI-I-82A9fails to start when the control
switch is placed in ST,4RT.
Which of the following interlocks would prevent the fan from starting?
a. Normal Purge Inlet and Discharge Valves are open
h. AII-82A fan inlet damper has failed to open
c. Electric heating coil breaker is tripped
d. Containment differential pressure is zero
ANSWER
d. (htainment differential pressure is zero
Harris NRC Written Exmination
Reactor Operator
Dirta Sheets
QUESTION NUMBER: 26 TIEWGROUP: 22
lOCFR55 CONTENT: 41(b) 5 w)
KA: 029A1.03
Ability to predict a n d h monitor changes in parameters to prevent exceeding design limits) assoc.iated
with operating the Containment Purge System controls including: Containment pressure, temperature,
and humidity
OBJECTIVE: CVS-3.0-R2
LOCATE the controls and EXPLAIN the interlocks associated with the following major components
0 NCPMU units, including AH-SZ fans
DEVELOPMENT REFERENCES: 01-168, p 8
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: X NEW SIGNIFICANTLY MODIFIED [7 DIRECT
BANK XUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NKC EXAM HISTORY: %ne.
DISTMCTOR JUSTIFICACTION (COIPRECT ANSWER Xd):
a. Plausible since the valves are interlocked to close iffan AH-824 is stopped, but are manually opened
prior to the start of the fan.
In. Plausible since the inlet damper is interlocked to open when the fan is started, but are closed when the
fan is started.
c. Plausible since the heating coils are interlocked with the fan operation, but the heaters are enabled to
operare when the fan is running and do not prevent the fan from starting.
X d. Fan AH-82A will only start if containment AP is more negativc than 4.400 INWG.
DIFFICULTY ANALYSIS:
u ~~
nKNOWLEDGE RECALL
U
/
DIFFKTJLTY RATING: 3
EXPLANATION: Knowledge of interlocks associate with containment purge fans
Harris NRC Written Examination
Reactor Operator
QUESTION: 27
Given the following conditions:
0 The plant is at the Point of Adding Heat (POAH) when a SG PORV fails open.
0 RCS temperature decreases and stabilizes at 548 F.
Which of the following predicts the plant response and the operator actions required in
accordance with CrP-004, Reactor Startup?
a. Reactor power increases; withdraw control rods and dilute, in a controlled
manner. to restore R(S temperature to program within 15 minutes
b. Reactor power increases; trip the reilctor if RCS temperature CANNOT be
restored above 55 1 I in a controlled manner within 15 minutes
c. The reactor becomes subcritical: trip the reactor if criticality CANNOT be
restored in a controlled manner within 15 minutes
d. The reactor becomes suhcriticai; immediately trip the reactor
ANSWER:
b. Reactor power increases; trip the reactor if RCS temperature CANNOT be
restored above 55 1 F in n controlled manner within 15 minutes
Harris NUC Wrilten Examination
Reactor Operator
Data Sheets
QUESTION NUMBER. 27 TIER/GR<KJR 2/1
IC4 IMPORTANCE: KO 3.3 SRO
10CFR55 CONTENT: 41(b) 5 43@)
KA: 039.42.05
Ability to (a) predict the impacts ofthe followiog malfunctions or operations on the MRSS; and (b) based
on predictions, use procedures to correct, control, or mitigate the consequences of those maltilnctions or
operations: increasing ste.am demand, its relationship to increases in reactor power
OBJECTIVE: IE-3.10-1
Apply the philosophies of OMM-001 and PLP-629 regarding safe and conservative decisions that must
be made by a control room crew
DEVELOPMENT REFERENCES: GP-004 pg 9 P & L # 19
OMM-001 pg 66-67
IE-LP-3.10 (Salem Event, SOEK 93-0 I )
REFERENCES SUPPLIED TO APPLICANT: None
QUE.STION SOIJRCE: NlCW SIGNIFICANTLY MODIFIED [7 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JI:STIFICACTION (CORRECT ANSWER Xd):
a. Ilausiblc since reactor power will increase, but temperature is not to be restored using two different
methods of reactivity control simultaneously and the 15 minute limit is to restorc tempcrature above
551 F, not to program.
X b. l h e first operator action should be to attempt to stop the cause (e.g., secure the overfeeding) ofthe
transient. Temperaturc. may then be recovered by using control rods in a slow and controlled manner.
temperature has to be restored to greater than 551 F within 15 minutes due to the requirements of TS
3.I.I.4.
c. Plausible since the 15 minute time limit is associated with restoration, hut the reactor does not become
subcritical.
d. Plausible since the reactor is to be tripped if it becomes subcritical due to a malfunction pw OMM-
001, but the reactor does not become subcritical.
DIFFICIJLTY ANALYSIS:
fl
u-
C:OMPREIIENSIVE I ANALYSIS n
u
KIVOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze the plant response to an increasc in steam demand and determine
appropriate actions
Harris NRC Written Examination
Reactor Operator
QUESTION: 28
The plant is operating at 100% pourer with the following conditions:
Anihient Temp
. I .
CT Basin I
1500 35 "F 64 "F
1900 20 60 "F
2300 10 O F 5s "F
Which of the following describes the correct CT Deicing Gate Valve aiignrnent for these
conditions?
m 23.m
a. Full Open Full Open
h. Full Open IIalf Open
c. Half Open Full Open
d. Haif Open Half Open
ANSWER:
h. Full Open Half Open
IIarris NRC Written Examination
Re3ctor Operator
Data Sheets
QUESTION NIIMRER 25 TIEWGROIJP: 3
10CTRSS CONTENT: Bl(b) 10 43@)
KA: 2.1.25
Ability to obtain and interpret station reference materials such as graphs, monographs, and tahles which
contain performance data
OBJECTIVE: CTK3
Given OF-141, Attachment S, ANAI,YZE a set of adverse weather conditions and DESCRIBE the
operation of the Cooling Tower System to prevent ice d-amage to the fill material
DEVE.LOPMENTREFERENCES: OP-141, pg SO Attac.hment 5
REFERENCES SIJPPLIED TO APPLICANT: OP-141, Attachment 5
QUESTIQN SQURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGIVIFICANTLY MODIFIED / DIRECT: CT-FG 001
NRC EXAM HISTORY: None
1)ISTRACTOR 3USTIFICACTION (CORRECT ANSWER Xd):
a, Plausible since valves should be open at 1900, hut are required to he changed to half open at 2300.
X b. At 1 500 conditions call for valves to be full open, at I900 conditions call for no change in position,
and at 2300 conditions c.all for change to half open.
c. Plausible since valves should be changed hetween 1900 and 2300, hut should go from full open to half
open.
d. Plausible since valves should be half open at 2300, but should be full open at 1900 due to no change
from 1500.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Application of given data to curve to determine required operation of deicing
valves
Hturis NRC Written Exmiination
Reactor operator
QCESTION: 29
Following a transition to PATH-2 for a SGTR in A SG, which of the following actions
are taken to minimize or prevent radiological ele eases through the SG PORV?
a. Increase A SG IOR\ setpoint on PK 308A1 SA to 90% (1 170 p i g )
b. Increase A SC; PORV setpoint on PK 308A1 SA to 88% (1 145 p i g )
c. Place A SG PORV PK 308A1 SA in MhVUAL with zero output
d. Mvlanually isolate A SG PORV by closing IMS-59
ANSWER:
b. Increase A SG POR\T setpoint on PK 308A1 SA to 88% (1 145 psig)
Harris h7(C Written Examination
Reactor Operator
Uaia Sheets
QUESTION NIJMBER: 29 TIEWGROCR 3
10CFR55 CONTENT: 416b) None 43(b) Sone
KA: 2.3.11
Ability to control radiation releases
DEMONSTRATE the below-assumed operator knolowledge from the HNP Step Deviation Documents
and the WOG ERGS that suppnrt performance of EOP actions
Method of isolating SGTR
DEVELOPMENT REFERENCES: P A m - 2 pg x
REFERENCES SUPPLIED TO APPLICANT: None
QUESIXON SOIJRCE: 0
X NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NKMBER FQR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since PORV setpoint is adjusted, but should bc adjusted to 1145 psig and 1170 psig is the
first safety setpoint.
X b. The SG PORV is to be set at 88% to minimkc the likelihood of a release, but lower than the SG safety
sctpoints.
c. Plausible since this action would be taken if the S G were faulted instead ofruptured.
d. Plausible since this action would be taken if the SG PORV were to fail open, but this would also cause
the safeties to be challenged and should not be performed unless necessary.
DIFFICIJLTY ANALYSIS:
COIVIPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of steps required to isolate a SGTR
Harris XRC' Written kxaminatioii
Reactor Operator
QUESTION: 30
Which of the following two (2) conditions are both identified by EPP-013, "1,OCA
Outside Containnient," as being uycd to identify that the 1,OCA has been isoiatccl'!
a. RCS pressure increasing
e KAB local room temperatures
b. RAW local room temperatures
RAD radiation levels decreasing
c. e R4B radiation levels decreasing
e 1,ocal observation of the isolation
d. RCS pressure increasing
o Local observation of the isolation
ANSWER
d. e RCS pressure increasing
Local observation of the isolation
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NIJMBER: 30 TIERGROUP: lil
K A T ~ O R T A N C E : no 3.5 sno
1QCFR55CONTENT 41(b) 8/10 43(b,3
KA: WE04EKI .2
Knowledge of the operational implications ofthe following concepts ac they apply to the (I.0CA Outside
Containment) Normal, abnormal and emergency operating proccdures associated with (LOCA Outside
Containment)
OBJECTIVE: 2.3-R4
Using appropriate plant procedures and prints, determine the foliowing:
Transitions to other EOPs
DEITLOPMENT REFERENCES: EPP-013 pg 5
REFEREiSC;ES SUPPLIED TO APPLICANT: None
QtJESTION SOURCE: 0 XEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIG CARTLY MODIFIED /DIRECT: 3.3 024
NRC EXAM HISTORY: None
1)ISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since RCS pressure increasing is one of the indications used. but local temperatures are not
used in EPP-0 13.
b. Plausible since these may both be indications that might support that the leak is isolated, but
pressurizer level may not be indicative of actual KCS inventory or the leak being isolated and is not
used in IiPP-013.
e. Plausible since local observation is one of the indications used, but RAB radiation levels may be
elevated for ~ o m time
e after isolation and is not used in RPP-013.
X d. EPP-013 determines that the LOCA outside containment is isolated ifRCS pressure is increasing and
if local ohsenration confirms the isolation.
DIFFICULTY ANALYSIS:
0 COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICIJLTY RATIXG: 3
EXPLANATION: Knowledge of the conditions required by FPP-01.1 to determine that a LOCA
outside containment is isolated
Harris NKC Written Examination
Reactor Operator
Which of tlic forlowing is the reason for purposely tripping the Reactor Coolant Pumps
(KCPs) under accident conditions'?
a, Ensure RCPs are available later in the event if they should be needed in response
to an inadequate core cooling condition
h. Prevent RCP rmout in the event of a large break LOCA
c. Prevent excessive depletion oERCS inventory through a small bresak in the RC'S
d. Prevent damage to RCI's due to pumping a two-phase mixture event
ANSWER:
c. Prevent excessive depletion of RCS inventory though a small break in the RCS
Harris NRC Written Ex.unination
Reactor Operator
Data Sheets
QUESTION NUMBER: 3 1 TIEWGROUP: 1/1
1UCFR55 CONTENT: 41(b) 5/10 43(b)
KA: 000009EK3.23
Knowledge of the reasons for the follow~ingresponses as the apply to the small break 1,OCA: RCI
tripping reqnirements
OBJECTIVE: BD-3.1- 1
Analyie the Reactor Coolant Pump (RCP) trip criteria. This analysis should include, at the minimum, the
foiluwing topics:
The reason for purposely tripping the RCPs uuder certain accident conditions
DEVELOPMENT REFERENCES: Generic Issues of ERG Background - Executive Volume
LP-BD-3.1 pg 8
REFERENCES SUPPLIED TO APPLICANT: Sone
QUESTION SOURCE: NEW SIGNFICANTLY MODIFIED DIRECT
BANK NIJMREK FOR SIGNKFICANTLY MODIFIED 1 DIRECT: ED-3.1 001
NRC EXAM HISTORY: None
DISTR4CTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since for most accidents it is desirable to have RCPs availabk, particnlarly those cases where
an inadequate core cooling condition might exist.
b. Plausible since little work is r e q u i d by the RCPs in the event of a Large break LOCA, hut this would
result in a lower pump current, not a IUnoUt condition.
X c. Tripping the RCPs during the early stages of a small break LOCA limits the amountof mass lost out
the break, thereby increasing the mass available for heat removal in the event the pumps were not
tripped but tripped at a latcr time.
d. Plausible since KCPs are not designed to pump a two-phase mixture and it would be desirable to
protect the pumps from damage.
DIFFICULTY ANALYSIS:
0 COMPREHENSIVE 1 ANALYSIS KYOWLEDGE 1 RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of the reasons for tripping RCPs during G small hre.ak LOCA
Harris NRC Written Examination
Reactor Operator
QUESTION: 32
Given the following conditions:
o The unit is in Mode 3 at normal operating pressure
o Pressurizer Pressure Control is in AUTO.
e Pressurizer Pressure Channel Pr-445 fails high.
e PW. Pressure Channel indications are:
0 PI-444 2050psig
e PI-445 2500psig
0 PI-455 2050psig
PI-456 1950psig
0 PI-457 205cIpsig
Assuming NO operator actions, which of the following describes the expected conditions
of the PRZ Pressure PORVs and Spray Valves?
PRZPORVS IRC-116and 1RC-1180pen
o PRZ Spray Valves PCV-444C and PCV-444D open
0 PRZPORVS 1RC-116 and 1RC-118 closed
o PW. Spray Valves PCV-444C and PCV-444D open
o PRZ PORV IRC-116 and IKC-118 open
e PRZ Spray Valves PCV-444C and FCV-444D closed
d. e PRZ P O W 1RC-114 open
o FW, PORVs 1RC-116 and 1RC-118 closed
o F W Spray Valves PCV-444C and PCV-444D closed
ANSWER
c. o P W PORV IRC-114 closed
0 PRZPORV 1RC-116andlRC-118open
o PRZ Spray Valves PCV-444C and PCV-444D closed
The noun names were pfovided for the following valves:
IRC-I 14, PRZ P O W PCV-444B
IRC-I 18, PFZ PORV PCV-444A
Harris NRC Written Examination
Reactor Operator
Data Sheets
QtJESTION NUMBER: 32 TIEWGROUP: lil
KAI~RTANCE: no 2.6 SRO
10CFR55 CONTENT: 41(b) 4 4309
KA: 000024AK2.03
Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the
following: Controllers and positioners
OBJECTIVE: PXKPC-3.0-3
Given the status of the various pressurizer pressure channels, the position of various presvure eontrol-
related control switch positions and the status of Controllers PK-444A, PK-444C, and PK-444D,
PREDICT the responses of the following functions:
Pressuriter spray valves
Pressurizer Power-Operated Relief Valves (PORVs)
Pressurizer pressure permissive P-l 1
DEVELOPMENT REFERENCES: SI)-100.3, pg 12, 16,38-39
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK mMmn FOR SIGNFKCANTLY MODIFIED / DrnEcr: PZRPC-~3 003
NRC EXAM HISTORY: None
DISTRACTOR .JWTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since PORVs 116 and 118 will open until actual pressure drop5 below 2000 psi& bur the
spray vaives are axitrolled by the other channel and will not open.
b. Plausible since thiq would he the response of the systeni if the failed channel wa$ 444, but with 445
failed, none of these components are affected.
X c. PT-445 controls only POKVs 116 and 118. The PORVs will open and remain open until 2!3 of the
protrction channels 455/456!457 decrease below the P-1 1 setpoint of 2000 psig. Spray valves are
controlled by channel 444.
d. Plausible since the spray valves will remain closed, but 445 controls PORVs 116 and 118, not 114.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KXOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis ofthe failure and current plant conditions to determine expected
response of pressure control
Harris NKC Written Eximiiiation
Reactor Operator
QUESTION: 33
Which one of the following correctly describes how and why the Variable Speed Fluid
Coupling (VSFC) varies the speed of the Condensate Booster Pumps (CBPs)?
a. V S K oil is bypassed around the hydraulic coupling as necessary to maintain a
constant feed pump suction pressure
b. VSFC oil is bypassed around the hydraulic coupling as necessary to maintain the
CBP recirc valves closed
c. VSFC hydraulic coupling is varied as necessary to maintain a constant feed pump
suction pressure
d. VSFC' hydraulic coupling is varied as necessary to maintain the CBP rccirc vaIves
closed
ANSWER:
c. VSFC hydraulic coupling is varied as necessary to maintain a constant feed pump
suction pressure
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 33 TIERKROUP: 2/I
10CFR55 CONTENT: 41(b) 7 4309
KA: 056G2.1.28
Knowledge of the purpose and function of major system components and controls. (Condensate)
OBJECTIVE: CFW-3.0-4
DESCRIBE the basic construction and operation of the following CFW System components /
subsystems
CUP Variable Speed Fluid Coupling (VSFC)
DEVE1,OPMENT REFERENCES: SD-134, p 7,17
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW 0 SIGNIFICANTLY MODIFIED H DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: CFW-R3 001
NRC EXAM HISTORY: None
DISTRACTOR .TIJSTIFIC?ACTION(CORRECT ANSWER Xd):
a. Plausiblc since oil adjusts the hydraulic coupling to maintain a constant suc.tion pressure at the feed
pump. but the oil does not bypass the hydraulic coupling
b. Plausible since oil adjusts the hydraulic coupling, but it does not bypass the hydraulic coupling and
does not maintain the CRP recirc valves closed.
X c. An oil bath between the motor and pump conpling causes the pump to operate at a variable speed to
maintain a constant suction pressure at the feed pump.
d. P h s i h l e since an oil bath between the motor and pump coupling causes the pump to operate at a
variable speed, but it is designed to maintain a constant suction pressure at the feed pump rather than
the CBP recirc valves closed.
DIFFICULTY ANALYSIS:
0 COMPREEIENSI\X / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge ofthe operation ofthe CBPs
Harris NRC Written Examination
Reactor Operator
QUESTION 34
Given the following conditions:
- 'The plant is operating at 100% power.
A tube leak has been detected on 'B' SG.
The Condenser Vacuum Pump Rad Monitor, REM-1TV-3534. and 11-X-15 curves are
being monitored every 15 minutes to estimate the leak rate.
CVPE is operating with NO motivating air.
Which ofthe following readings noted on REM-ITV-3534 is the MINIMUM reading
that would require a plant shutdown per Technical Specifications'?
a. 5.40 E - 7
b. 6.00 E -7
C. 1.08E-6
d. 1.80 E -6
ANSWER:
C. 1.08E-6
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 31 TIEWGROUP: 112
1OCFR55 CONTENT: 41(b) Nom 43(b) 5
Kn: 000037AA2.10
Ability to determine and interpret the following as they apply to the Steam Generator Tuhe Leak: Tech-
Spec limits for RC'S leakage
OBJECTIVE: AOP-3.16
For a primary-to-Fecondary leak, DESCRIBE when a power reduction or unit shutdown is required.
DEVELOPMENT REFERENCES: AOP-016 pg 15
Curves ILX-15dbIc
REFERENCES SWPLIED TO APPLICANT: Curves 11-X- 1Sdb/c
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: Harris NRC 2000-80
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since this exceeds would exceed PSAL 2 limits if operating on full motivating air (curve H-
X-l Sa), but the incorrect curve is used.
b. Plausible since this exceeds would exceed PSAL 2 limits if operating on intermediate motivating air
(curve I-X-15b), but the incorrect curve is used.
X e. Lowest level that would exceed 75 gpd (PSAL, 2 ) which would require a TS shutdown.
d. Plausible siuce this exceeds the PSAL 3 limit which would require a TS shutdown, but this is not the
lowest level that would require the strutdown.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATLYG: 3
EXPIANATION: Interpretation of plant data 011 RCS leakage curve and comparison i o pmcedural
requirements
Harris NRC Written Examination
Reactor Operator
QUESTION: 35
FRP-J.2, Response to Containment Flooding, directs that the containment sump be
sampled for activity. and then to notify the operations staff of sump level and the sample
results.
Receiving this information will allow a decision to be made on which of the following
actions?
a. If the Containment Spray System may be secured
b. If the CNWI spray additive tank should be isolated
c. If Iimergency Service Water to containment should he isolated
d. If sump water may he transferred to tanks outside containment
ANSWER
d. If sump water may k transferred to tanks outside containment
Harris NKC Written Exmination
Reactor Operator
Data Sheet3
QUESTION NCJMBER: 35 TIEWGROUP: 112
10CFR55 CONTENT: 41(b) 8/10 13(b)
KA: WE15EK1.2
Knowledge ofthe operational implications of die following concepts as they apply to the (Containment
Flooding) Normall abnonnal and emergency operating procedures associated with (Containment
Flooding)
OBJECTIVE: 3.13-4
Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
Sampling the CNM? sump for activity (5.2)
DE\ELOPMENT REFERENCES: FKI-J.2, pg 4
LP-3.13, pg 12
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NIJMBER FOR SIGNFICANTLY MODIFIED / DIRECT: 3.13 0 10
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the plant operations staff does makc the determination when Containment Spray is to
be secured, but this sample is to determine whether the water can be transferred.
h. Plausible since ifflooding has occun.ed it is likely that a large KCS leak has also occurred and the
spray chemical addition tank miry have emptied to containment and would no longer be needed, but
this sample is to determine whether the water can be transferred.
c. Plausible since a potential sonrce of flooding is the ESW system to the fan coolers, but this sample is
to determinc whether the water can he transferred and ESW isolation would be determined by the
operating crew based on ESW indications.
X d. The c.ontainment sump is samplcd to determine if excess water can be transfemd to storage tanks
located outside containment.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of purpose for sampling sumps following flooding inside
containment
Harris NRC Written Examination
Renctor Operator
QUESTION: 36
Given the following conditions:
RHR Pump A-SA is tagged out.
Following a large break I,OCA, the crew was performing EPP-010, Transfer to Cold
Leg Recirculation.
1SI-301. CONTtZINMENT SUMP TO RIIR PUMP B-SB, failed to open and the
crew transitioned to EPP-012, 1,oss of Emergency Coolant Recirculation.
Both Containment Spray Pumps automatically transferred to the Containment Sump.
Two (2) Containment Fan Coolers are operating.
Containment pressure is 12 psig and decreasing slowly.
While performing EPP-012 the Reactor Operator notes that RWST level is 2% with
both CSIPs, both Containment Spray Iumps, and RIIR Pump B-SB operating.
Which of the following actions are to be tciken?
a. Stop the RIIR pump ONLY
b. Stop both CSIPs and the RHR pump ONLY
c. Stop both CSIls, the RHK pump. and one Containment Spray pump ONLY
d. Stop both CSIPs, the RHR pump, and both Containment Spray pumps
ANSWER:
b. Stop both CSIPs and the RIIR pump ONLY
Harris NKC Written I3xmiination
&actor Operator
Data Sheets
QUESTION NUMBER: 36 TIEWGKOUP: i/i
10CFR55 CONTENT: 41(b) 8/10 43(b)
K4: WEllF.Kl.1
Knowledge of the. operational implications of the following concepts as they apply to the (Loss of
Emergency Coolant Recirculation) Components, capacity, and function of emergency systems
ORJECTIVE: 2 . 3 - S
Predict how e z h of the fullowing could impact efforts to maintain core cooling during a LOCA
Failure of valves to realign for cold-leg recirculation
UEVEIOPMENT REFERENCES: EPP-012 pg 42
REFERENCES SUPPLIED TO APPLICA&T None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.3-RS 004
NRC EXAM HISTORY None
DISTKACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible sinc.e the KHR pump is still aligned to the RNST and must be stopped, but the CSlPs are
also aligned to the RWSI and must likewrise he stopped.
X b. Ihe RHR pump and the CSIls arc still aligned to the RWS? aud must he stopped when the RWST
empty alarni is received at 3% level.
c. Plausible since the RHK pump and the CXPs must he stopped, hut the spray pumps can continue to
operate since they are no 1onge.ra1igne.d to the RWST.
d. Ilausiblc since the RHR pump and the CSIPs must he stopped, but the spray pumps can continue to
operate. since they arc no longer aligned to the RWSI.
DIFPIC:CI,TY ANALYS.IS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze plant conditions to determine which pumps are taking a suction from
the KWST to determine the pnmps which are to be stopped
Harris NKC Written Examination
Reactor Operator
QUESTION: 37
LT-115, VCT I.eve1, has failed LON7. The Unit-SCO directs the Reactor Operator to
maintain VCT level between 20% arid 70%.
Wiich ofthe following describes how VCT level will be maintained in accordance with
AOP-003, Malfunction of Reactor Makeup Cmtrol?
a. When level lowers to 20%. automatic makeup will begin raising level
Uhen level increases to 70%, ICs-120 (I,CV-l124), Letdomin VCI/Kold Up
Tank, wrill begin diverting letdown to the Hold Up Pa&
b. * When level lowers to 20%, the operator must start a manual makeup to raise
V C I level
When level increases to 70%, ICs-120 (LCV-l12A), Letdown VCT/Hold 1Jp
Tank, will begin diverting letdown to the Hold IJp T d
c. e Wien level lowers to 20%, automatic makeup will begin raising level
e When level increascs to 70%, the operator must align 1CS-120 (I,CV-l12A),
1,etdcwn VCTiHold Up Tank,to the IIold 1Tp Tar&
d. * Uhen level lowers to 2O%oa,the operator must start a manual makeup to raise
VCT level
- When level increases to 40%, the operator must align 1CS-120 (LCV-l12A),
Letdown VCT/Hold IJp Tank. to the Hold Up Tank
ANSWER.
b. e When level lowers to 20%, the operator must start a manual makeup to raise
VCT level
e When level increases to 90%. IC§-120 (LCV-l12A), Letdown VCT/IIoId LJp
Pank, will begin diverting letdown to the Hold Up Tank
EIarris XRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 37 TIEWGROUP: 211
10CFR55 CONTENT 41(h) 5 4300)
KA: 004A1.06
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated
with operating the CVC.S controls including: VCT level
OBJECTIVE: CVCS-RS
PREDICT the response of the CVCS to the following failures
e. LT-112 or LT-115 failure (high or low)
IjEVELOPMENT REFERENCES: AOP-003, pg 5-6, 16
REFERENCES SUPPLIED TO APPLICANT: None
QUESTIQN SOURCE: 0 X NEW SIGNIFICANTLY MODIFIED 0 DIRE.CT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New-
NRC EXAM HISTORY None
DlSTKACTOR JUSTIFICACTION (COKKECT ANSWER Xd):
a. Plausible since LT-112 wiil stili control CS-120 properly, causing a divert to the HIJT, but the
operator must perform a manual blended flow due to the failure of LT-115.
X h. A low failure of LT- I I5 will disable auto makeup capabilities which will required the operator to
perform a manual blended flow and the modulate divert to the H I J l is controlled by LT-I 12.
c. Plausible since operator action is required to perform one of the two evolutions, hut the automatic
makeup, not the divert, must be controlled by the operator.
d. Plausible since a low failure of LT-1 IS will disable auto makeup capabilities which will required the
operator to perform a manual bknded flow, but the modulate divert to the HLJT is controlled by LT-
112.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KXOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of plant response to failures in CVCS to determine the proper operator
response
Harris KRC Written Examination
Reactor Operator
QUESTION: 38
Ihe plant is operating at 100% power with all equipment operable and properly aligned.
W3kh ofthe folrowing describes changes to the Component Cooling Water System
alignment following a Safety Injection signal?
a. CCW to the Gross Failed Fuel Detector and Primary Sample Panel isolates
b. Both CCW pumps start and the Non-Essential header isolates
e. CCW to and from the RCP Motor Coolers isolates
d. Both CCW punips start and the Thermal Barrier Hx Return isolates
AVSWER:
a. CCW to the Gross Failed Fuel Detector and Primary Sample Panel isolates
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NZJMBER: 38 TIEWGROW: 211
10CPRS5 CONTENT: 41(b) 7 43w
KA: 008A3.08
AbiliQ to monitor automatic operation of the CCWS, including: Automatic actions associated with the
CCWS that occur as a result of a safety injection signal
OBJECTWE: C:CWS-3.0-R2
STATE how the CCWS responds during each of the following conditions:
- Safety 1njec.tionsignal
DEVELOPMENT REFERENCES: SD- I45 pg 16-17
REFERENCES SUPPLIED TO APPLICANT: Xone
QIZSTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NLWRER FOR SIGNIFICANTLY MODIFIED /DIRECT CCWS-R2 002
NRC EXAM HISTORY: None
DISTKPlCTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. On an SI signal, both the GFFD and sample panel receive isohtion signals
h. Plausible since the pumps will get a start signal, but only the GFFD and sample panel in thc non-
essential header are isolated.
c. Plausible since the C:CW to RCP isolations close on a Phase. B signal, hut Phase R is not generatcd by
an SI signal.
d. Plausible since the pumps will get a start signal; but the thermal barrier heat exchangers are only
isolated on a Phase H signal.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the response of CCWS to an SI signal
Harris NRC Written Examination
Reactor Operator
QUESTION: 39
Given the following conditions:
- The plant is operating at 13% power.
Steam pressure charnel PI-475 is selected for control of SG A.
- Steani pressure transmitter PT-475 fails high.
Assuming NO operator action, which of the following statements describes the response
of the Skanr Generator Water Level Control System (SGWLCS)?
a. An increase in steam flow from SG A is sensed and responds by increasing
lF\-I40, MN FW A REG BYP FK-479.1, position to increase feed flow to S G
A and level increases
h. An increase in steam flow from SG A is sensed and responds by increasing
1FW-133, MAIN FW A KEGIJLATOR FM-478, position to increase feed flow to
SG A and level increases
c. A decrease in steam flow from SG A is sensed and responds by decreasing 1FW-
140, MN FW A REG BYP FK-479.1, position to decrease feed flow to SG A
and level decreases
d. A decrease in steam flow from SG A is sensed and responds by decreasing IFW-
133. MAIN FW A RI:,GULA?OK FK-478, position to decrease feed flow to SG
A and level decreases
ANSWER:
h. An increase in steam flow from SG A is sensed and responds by increasing
1FW-133. MAIN FW A REGULATOR FK-478, position to increase feed flow to
SG A and level incxases
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUEsTION NUMBER 39 TIEWGROIJP: 2!1
KAlMPORTANCE: RO 3.0 SKO
1OCFR55 CONTENT: 41(h) 7 43m
KA: 059A4.08
Ability to manually operate and monitor in the control room: Feed regulating valve controller
OBJECTIVE: SGW'I,C-3.0-2
Given the stahis of the various SGWLC related control switch positions and controllers, PREDICT how
a malfunction of the fallowing will effect the SQWLC System:
SG pressure channels
DEVELOPMENT REFERENCES: SD-126.02 pg 4 , 8
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
OR SIGNIFICANTLY MODIFIED / DIRECT: SG\VI~,C-S002
NRC EXAM HISTORY: None
DISTRACTOR .KX3TIFIC:ACTIQN (CORRECT ANSWER X'd):
a. Plausible since steam pressure failing high causes the steam flow to increase, resulting in SF >: FF, but
the feed reg valve is in operation at this power level.
X b. Steam pressure failing high causes the steam flow to increase, resulting in SF 2, FF. The fce.d reg
valve, in operation at 15% power, opens to cmse FF and level to increase.
e. Plausible since steam pressure failiug c a m s the steam flow to change, resulting in a SF - FF
mismatc.h. but the feed reg valve will open to increase FF.
d. Plausible since steam pressure failing causes the steam flow to change, resulting in a SF - FF
mismatc.h, but the feed reg valve will open to increase FF.
ICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWL.EDGE/ RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze the effect ofthe failure on the contrnl system and rec.oguize which
valve will be controlling at the power level given
Harris NRC' Written Examination
Reactor Operator
QIJESTION: 40
The plant is operating at 80% poiva with rod control in automatic and pressurizer
pressure at 2240 psig.
After a rapid power reduction the plant is stabilized at 40% power, when the Reactor
Operator notes the following conditions:
e Pressurizer pressure is 2275 psig and slowly decreasing.
Pressurizer levei is 43% and slowly decreasing.
- Both pressurizer spray valves indicate mid-position.
All pressurizer backup heaters are de-energized.
These conditions are indicative o f . . .
a. a normal plant response fbllowing an outsurge from the pressurizer.
17. a failure in the Pressurizer Pressure control circuitry, which opened the spray
valves.
c. a failure in the Pressurizer Level control circuitry, which failed to energize the
backup heaters.
d. a normal pl'ant response following iltl insurge into the pressurizer.
ANSWER:
c. a failure in the Pressurizer Levei control circuitry, which failed to energize the
backup heaters.
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NbUt4BE.R: 40 TIEWGROW: 212
KAIMPORTANCE: KO 3.1 SRO
10CFR55 CONTENT: 41(h) 7 43W
KA: 01IK6.04
Knowledge of the effect o f a loss or malfunction on the following will have on the %K I C s : Operation
of PZR levcl controllers
OBJECTIVE: PZRLC-3 .O-5
EXPLAIN how the system controls pressurizer level, including the input parameters and the components
that receive output signals
DEVELOPMENT REFERENCES: SD-100.3 pg 14-15
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: [7 NEW H SIGNIFICANTLY MODIFIED [7 DIRECT
BANK NJMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: PZRLC-R4 001
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the response is correct, with the exception of the pressurizer heaters not being
energized, for an outsurge from the pressurizer.
h. Plausible since a downpower should result in m insurge which would cause the spray valves to open,
but the heaters should also be energized.
X c. A rapid dowrnpower transient will result in an insurge to the pressurizer. This should result in the.
conditions noted, including a high pressurizer level causing the heaters to be energized even during a
high pressure condition causing the spray valves to be open. The heaters not being energized w3ith
level more than 5% high is indicative of a level control system failure.
d. Ilausible since the response is correct, with the exception of the pressurizer heaters not being
energized, for an insurge to the pressurizer.
DIFFICULTY ANALYSIS:
H COMPREHENSIVE /ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis ofthe expected plant response and the actual plant response to an
insurge into the pressurizer
Harris NRC Written Examination
Reactor Operator
QUESTION: 41
The operators are performing a start up to full power with Main Feedwater Punip B under
clearance.
Which of the following causes an immediate start signal to ONLY the Motor Driven
AFU' Pumps?
a. 0 SG A level is 18%)
- SCi I3 level is 39%
SG C level is 38vo
I,oss of Emergency Bus IA-SA
h. SGAlevelis?4%
- SG B l e d is 33%
SG C level is 22%
o Loss of Emergency Bus 1B-SR
c. o S G A level is 25%
SG B level is 26%
S G C level is 27%
Main Feedwater Pump A trips
d. * S(i A level is 24Yo
- SG 3 level is 23%
SG C level is 28%
o Main Feedwater Pump A trips
AXSWER:
c. S G A level is 25%
- SC; I3 Ievel is 26%
SG C level is 27Y0
Main Feedwater Pump A trips
Harris NKC Written Examination
Reactor Operator
LMa Sheets
QUESTION NUMBER 41 TIEWGROUP: 212
10CFR55 CONTENT: 41(b) 2-9 43@B
KA: 035K1.01
Knowledge of the physical connec.tions and/or cause-effect relationships between the SiGS and llie
following systems: MFW/AFW systems
OBJECTIVE: AFS-3.0-135
State the automatic start signals associated with the:
MDAFW pumps
TDAIW pumps
DE.VEI,OPME.NTREFERENCES: SD-137 pg 12-13
REFERENCES SUPPLIED TO APPLICANT: None
QITESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED 0 DIRECT
RAXK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: Xew,
NRC EXAM HISTORY: Xone
DISTRACTOR SIJSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the SG levels will came a start of only the MDAFW Pumps, hut the loss of the
emergency bus starts the train related MIPAFW Pump and the TDAFW Pump.
b. Plausible since the S G levels will cause a start of o d y the MDAFW Pumps, hut the loss of the
emergency bus starts the train related MDAFW Pump and the TDAIW Pump.
X c. With all 3 SG levels above 2S%, no start signals occur, however the trip of MFW Pump A will cause
both MDAFW Pumps to start since the I3 MFW Pump is already secured.
d. Plausible siuce the trip of MFW Pump A will cause both MDAFW Pumps to start since the B MFW
Pump is already secured, but 2 SG levels below 25% start the TDAFW Pump and the MDAFW
Pumps.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY RATING: 7
EXPLANATION: Analysis of conditions which determine AFW pnmp starts
Harris NRC Written Examination
Reactor Operator
QUESTION: 42
In accordance with PRP-1.1. "Response to Loss of Secondary Heat Sink," why must an
RCS bleed and feed path be immediately established when the conditions for a total loss
of heat sink are diagnosed?
a. The increase in steam production in the core will overpressurize the RCS,
increasing the likelihood ofthe PRZ safety valves opening and an increased loss of
RCS inventory
b. The increase in RCS temperature will increase RCS pressure and decrease SI flow,
increasing the likelihood of core uncovery
c. The loss of natural circulation will result in SI flow being directed to the reactor
vessel without mixing with the RCS, increasing the likelihood oftlrermal shock of
the reactor vessel
a. ?'he increase in IZCS temperature will increase primary-to-secondary AP,
increasing the likelihood of a SGI'R
ANSWER:
b. The increase in RCS temperature will increase RCS pressure and decrease SI flow,
increasing the likelihood of core unwvery
Harris NRC Wrinen Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 42 TIEWGROUP: L!I
KAIMPORTANCE: KO 3.9 SRO
10CFR55 CONTEXT: 4%(h) 7 43m
W WE105EM2.2
Knowledge of the interrelations between the (Loss of Secondary Heat Sink) and the following: Faciiitys
heat removal systems, including primary c.oolant; emergency coolant, the decay heat removal systems.
and relations between the proper operation ofthese systems to the operation of the facility
OBJECTIVE: 3.11-R4
Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
Prompt initiation of Bleed and Feed
DEVELOPMENT REFERENCES: FRP-ELI, pg 19,22
LP-3.11, pg 10-12
RF,FERENCES SUPPL1E.DTO APPLICANT: Xone
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.1 i-K4 015
NRC EXAM HISTORY. None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Piausible since an increase in KCS pressure could result in the safely valves lifting if the POR\s were
to fail, but steam production in the core is not likely to be occurring at the onset of the Loss of heat
sink.
X b. Failure to establish KCS bleed and feed when required will result in an increase in KCS temperature
which will cause an increase in RCS pressure. This will result in decreased SI flow and core
uncovery.
c. Plausible since a heat sink is required for natural circulation and a concern in PKP-P.1 is that cold SI
flow could cause thermal shock of the reactor vessel, but core uncovery doe to a loss of SI flow as
pressure incremes uill also reduce the SI flow that could cause thermal shock.
d. Plausible since an increase in primary-to-secondary AP could result in a SGR, but the concern is that
an increase in temperature and pressure could result in less SI flow and core imcovery.
DIFFICULTY ANALYSIS:
17 COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICLZTY RATING: 3
FXPLANATION: Knowledge of the effect of delaying RCS bleed and feed during a loss of heal
sink
Harris NRC Written Examination
Reactor Operator
QUESTION: 43
Given the following conditions:
The plant had been opemting at 100% for three (3) weeks when a Reactor Trip
occLlned.
Six (6)hours forlowing the trip, a reactor startup is plmned.
Which one of the following is PROHIBITED at SHh'PP as a result of industry wide
premature criticality events?
a. A difference of400 pcm between the POWERTRAX and EXSPACK ECCs
b. Operators performing the EXSPACK estimated critical conditions (ECC)
c. Delaying the startup until xenon begins to decay
d. A startup rate in excess of + 0.3 dpm
ANSVVER
a. h difference of400 pcm between the POWEKI'RAX and EXSPACK ECCs
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 43 TIERKROW: 3
10CFRSS CONTENT: Jl(b) None 43<b) None
KA: 2.2.1
Ability to pcrform pre-startup procedures for the facility, including operating those controls associated
with plant equipment that could affect reactivity
OBJECTIVE: GP-3.4-6
SIJMMAKI7X at least three conditions whic.h have contributed to premature criticality events within the
industry; also SUMMARIZE actions taken at SHNPP to prevent similar occurrences
DEVELOPMENT REFERENCES: GP-004 pg 10
REFERENCES SUFPL.IED TO APPLICANT: Xone
QEESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED 1DIRECT: GP-3.4 01 1
NRC EXAM HISTORY: None
DISTRACTOR HJSTIFICACTION (CORRECT ANSWER X9d):
X a. The threshold for performing a reactor startup following a power history of >80% quilihrium power
is 250 pc.m difference hehveen POWERTRAX and EXSPACK and 500 pcm for transient history and
steady state helow 80%.
b. Plausible since SHNPP required any manual E.CC calculations he performed by Reador Engineering,
hut EXSPACK is normally performed hy Operations.
E. Plausible since xenon decay will he adding positive reactivity to the core while the startup is being
performed, but is accounted for in the time after trip in the ECC.
d. Plausible since excessive startup rates can contribute to lack of reactivity control, hut limitations are
placed on startup rate after criticality is achieved.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KSOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the administrative requirements prior to criticality heing
achieved
IIasris NRC Written Ekaniination
Reactor Operator
QUESTION: 44
Given the following conditions:
0 The plant was operating at 80% power.
Actions of AOP-010. Feedwater Malfunctions, due to a trip of Main Feedwater
Pump A.
m The crew is using transient annunciator response.
Which oftiie following annunciators is the Unit-SCO required to be informed of in
accordance with OMM-00 1 Conduct of Operations?
~
a. A1.B-05-7-4,CCWPUMP AIRIP ORCLOSE CKT TROUB1,E
b. ALB-04-1-1, CHARGING PIJMP DISCHARGE IIEADERIII 1 LO FLOW
c. CThIP-4-2, CLG TWR M-LJ 11IMP 1 TRIP OR START FAII,
d. A1,B-23-2-11, STEAM TUNNEL HIGH TEMP
ANSWER
a. ALB-05-7-4, CCW PUMI A TRIP OR CLOSE CKI TROIJBLE
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NCZVIRER: 44 TIEWGROUP: 3
KAIMPORTANCE: KO 3.3 SRO
lOCFR55 CONTENT: 41(b) 10 43(W
KA: 2.4.31
Knowledge of annunciators alarms and indications, and use of the response instructions
OBJECTIVE: PP-2.O-R3
DISCUSS the requirements in OMhI-001/AP-002AP-IOO concerning the following:
k MCB annunciators
DEVELOPMENT REFERENCES: OMM-001 pg IO
REFERENCES SUFPLIED TO APPLICANT: None
QITESTION SC)URCE: 0X NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUAMBERFOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
X a. Required to he informed of this annunciator due to a required entry into an additional AOP.
b. PLansible since this could indicate a ieak in the RCS,but no AOP entry conditions are met.
c. Plausible since this could indicate a loss of CW cooling flow, but no ACIP entry conditions are met
d. Plausible since this could indicate a steam leak. but no AOP entry conditions are met
DIFFICUI,TY ANALYSIS:
COMPREHENSIVE /ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of relative importance and requirements to prioritize annunciators
Ilanis NRC Written Examination
Reactor Operator
QIJESTION: 45
Given the following c.onditions:
0 A Reactor Trip occurred from 100% power.
0 The plant stabilized at 557 "F for several minutes.
0 Shortly thereafter, a Safety Injection signal actuated.
Which of the lollowing describes the effect of this sequence on the Main Feedwater
System?
a. * A k r the Reactor Trip occurred, the SGs could be fed using the Feedwater Reg
Bypass Valves
0 Alter the SI occurred, the SGs couid be fed using the Feedwater Reg Bypass
Valves
b. After the Reactor Trip occurred, the SGs ccluid be fed using the Main
Feedwater Reg Valves or the Feedwater Reg Bypass Valves
After the SI occurred, Main Feedwater could NOT he used to feed the SGs
c. After the Reactor Trip occurred, the SGs could be fed using the Feedwater Keg
Bypass Valves
0 After the SI occurred, Main Feedwater could NOT bc used to feed the SGs
d. 0 After the Reactor Trip occurred, the SGs could be fed using the Main
Feedwater Reg Valves or the Feedwater Reg Bypass Valves
Alter the SI occurred, the SGs could be fed using the Feedwater Reg Bypass
Valves
ANSWEW:
c. 0 After the Reactor Trip occurred, the SGs could be fed using the Feedwater Reg
Bypass Valves
0 After the SI occurred, Main Feedwater couid NOT be used to feed the SGs
Harris NRC Written Exmiination
Keactor Operator
Lhta Sheets
QUE.STIONNUMBER 45 TIEWGROUP: 2: 1
10CFRSS CONTENT: 41(b) 7 43(W
KA: 059K4.19
Knowledge of MFW design feature(s) and/or interlock(s) which provide for thc following: Automatic.
OBJECTIVE: AFW-3.0-A6
EXPLAIN the response of major CFW System valves to the following signalsiconditioiis
Main Feedwater Isolation Signal (MFIS)
Reactor trip (1'-4) coincident with low Tal,E(e: 564OF)
DEVELOPMENT REFERENCES: SD- 103 pg 26
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTIOS SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NIJMBEH FOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFLCACTION (CORRECT ANSWER X'd):
a. Plausihle since on a reactor trip with low Taw (564 "F), the SGs can still he fed with the bypass
valves. hut on an SI or high-high SG level MFW can no Longer supply the SGs.
b. Plausible since the SGs can no longer be fed using MFW on an SI, but on a reactor trip only the
bypass valves can he used to feed the SGs.
X c. On a reactor trip with low Tave (564 'I;), the SGs can still he fed with the bypass valves, but on an SI
or high-high SG level MFW can no longer supply the SGs.
d. Plausible since on a reactor trip with low Tave (564 'I.'), the SGs can still he fed with the bypass
valves, but not the main feed reg valves. and on an SI or high-high SG level MFW can no longer
supply the SGs.
DIFFICULTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICUL.TYRATING: 3
EXPLANATION: Comprehension that on a reactor trip where the plant stabilizes at no-load
temprratnrc, the P-4 with Low 'I'ave signal allows feeding with the bypass and
Harris NKC Written Examination
Reactor Operator
QUESTION: 46
Which of the following describes the design of Phase A and a Phase B Containment
Isolation signals?
a. Phase A ONLY limits radioactive releases following a I B C A
- Phase B ONLY limits radioactive releases following a LOCA or secondary
system break inside Containment
h. * Phase A limits radioactive releases minimizes Containment
overpressurimtion following a LOCA
Phase B limits radioactive releases kJQ minimizes Containment
overpressuriiation following a LOCA or secondary system b r e k inside
Containment
c. * Phase A m limits radioactive releases following a LOCA
Phase H limits radioactive releases following a LOCA AND prevents an
excessive RCS cooldown following a secondary system brcak inside
Containment
d. a Phase A limits radioactive releases minimizes Containment
overpressurization following a LOCA
a Phase 3 limits radioactive releases following a LOCA kd.D prevents an
excessive RCS cooldown following a secondary system break inside
Containment
ANSWER:
a. Phase. A a limits radioactive releases follomiog a LOCA
a Phase B limits radioactive releases following a LOCA or secondary
system break inside Containment
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 46 TIEWGROUP: 1/1
10CFR55 CONTENT 41(b) 5/10 43@)
KA: 00001IEK3 06
Knowledge of the reasons for the following responses as the apply to the Large Brcak LOCA. Actuation
of Phase A and B during LOCA initiation
OBJECTIVE: CIS-3 .0- 1
STATE the purpose ofthe Containment Isolation System
DEVELOPMENT REFERENCES: SI)-1 14 pg 4-5
REFERENCES SUPPLIED TO APPLICANT: None
QTXSTION SOURCE: [7 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CIS 006
CIS 009
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
X a. Phase A serves to limit the release of radioactive materials to atmosphere following a LQCA. Phase H
acts to limit radioactive releases by actuating on a LOCA or a steam or feedwater line break inside
containment.
h. Plausible since both Phase A and Phase 3 act to limit the relcase of radioactive materials to
atmosphere, hut overpressurization is limited by spray actuation. main steam line isolation, and feed
water isolation.
c. Plausible since both Phase A and Phase R act to limit the release of radioactive materials to
atmosphere, but overpressurization and RCS c.ooldownsare limited by spray actuation, main steam
line isolation, and feed water isolation.
d. Plausible since both Phase A and Phase 5 act to limit the release of radioactive materials to
atmosphere, but overpressurization and RCS cooldowns are limited by spray actuation, main steam
line isolation, and feed water isolation.
DIFFICULTY ANALYSIS:
0 COMPREHENSIVE /ANALYSIS KNOWLEDGE: I RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of purpose of Phase A and Phase B signals
Harris NKC Written Exmiination
Reactor Operator
QUESTION: 47
An entry into FRP-S.1. Response to Nuclear Power Cieneration/AIWS. has been made
from PATH-1. The following conditions currently exist:
0 The reactor trip breakers are closed.
- Rods are being inserted manually.
e Control Bank D is at 12 steps.
Power Range Instruments are all indicating 8%.
Intermediate Kange SLR is NEGATIVE
Which of the following conditions is required by FRP-S.1 to allow a return to PATH-I?
a. One of the reactor trip breakers must be opened
b. Both of the reactor trip breakers must be opened
c. Power Range indication must be reduced below 5%
d. Control Bank A mwt be inserted filly
ANSWER:
c. Power Range indication must be reduced below 5%
Harris NKC Written Examination
Reactor Operator
Ddta ShfXtS
QtJESTION NUMBER 47 TIEWGROUP: lil
1OCFR55 CONTENT: 41(b) Xone 43(b) 5
KA: 000029EA2.01
Ability to determine or interpret the following as they apply to a ATWS: Reactor nudear instrumentation
OBJECTIVE: 3.1-3
DF,MONS'rRATE the below-assumed operator knowledge from the SIINPP Step Deviation Documents
and WOG ERGS that support perf0rmanc.e of EOP actions:
a. Verification of reactor trip
DEVELOPMENT REFERENCES: FKP-S.1, pg 14
KEFERENCES SGPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: 3.15-R5 002
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since this would cause the reactor to be tripped, but it is not required to he done to exit FRP-
s.l.
b. Plausible since this would cause the reactor to be tripped, but it is not required to be done to exit FRP-
S.1.
X c. Exiting FRP-S. 1 requires that PK NIS be less than 5% and IR NIS startup rate be negative. Reactor
trip breaker position is not a condition for exiting the procedure, although actions are taken to open the
breakers.
d. Plausible since this would cause the reactor to be adequately shutdown, but it is not required to be
done to exit FRP-S.1.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the procedural requirements to exit FRP-S.1
Harris NRC Written Examination
Reactor Operator
QUESTION: 48
Criven the following conditions:
A plant cooldown is being performed.
o All Steam Generators (SGs) are currently at approximately 50 psig.
Auxiliary Feed Water (AFW) Pump A-SA is being used to feed the SGs.
o The supply breaker on 120 VAC 1DP-1A-SI for 1AF-19, AUX FW MOTOR PMP
A-SA DISCHARGE VLV, trips open.
Assuming NO operator actions. which of the following describes the effect of this loss of
power on the operation of AFW Pump A-SA?
a. Operates at shutoff head
h. Operates on minimum recirculation flow
c. Operates on maximum recirculation flow
d. Operates at runout conditions
ANSWER:
d. Operates at runout conditions
Harris NRC Written Examination
Reactor Operator
Ddtita Sh&S
QUESTION NIJIMREK: 48 TIEWGROTJP: 2/1
10CFR55 CONTENT: 41(b) 7 4309
KA: 061K6.01
Knowledge of the effect of a Loss or malfunction of llie following will have on the AFW components:
Controllers a d positioners
OBJECTIVE: AES-3.0-R5
DESCRIBE how the AFVJ systcm is impacted hy a loss of 120vac uninterruptibte power supplies (,SI,SII,
SIII, SIV)
DEVELOPMENT REFERENCES: SD-137, PLJ8-9
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: AFS-A3 001
AFS-A3 007
NRC EXAM HISTORE None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since power is lost to the discharge valve, but the valve fails ope11causing flow to increase
b. Plausible since power is iost to the discharge valve, but the valve fails open cawing flow to inwease.
c. Plausible since the valve fails open and flow- increases, but the pump does not run on recirculation
flow.
X d. The. loss of power causes AFW Pump A-SA to reach runout conditions due to 1AF-19 failing open
and having the SGs at such a low pressure.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS K??O\VLEDGE/RECALL
DIFFICIJLTY RATING: 3
EXPLANATION: Analysis of the effect of a failure ofthe PCV after determining the fail position
Hmis NRC Written Examination
Renctor Operator
QUESTION: 49
Given the following conditions:
concentration.
A batch liquid release from the Secondary Waste Sample Tank (SWST) to the
cooling tower discharge is in progress.
Which of the following sets of conditions would require entry into AOP-008, Accidental
Release of Liquid Waste?
a, AI,H-004-?-2, REEUELHNG WATER STORAGE LOW LEVEL, alarms.
e KWST level is at 94% and slowly decreasing.
h. * ALE-019-1-4. KQIWELL HIGII-LOW LEVEL. alarms.
Iiotwcll level is at 14% and slowly decreasing.
c. e An A 0 reports a ieak in the NSW System b i d e the Turbine Building.
FI-9301.1, NSU Discharge Flow, indicates high.
d. ALE-005-6-1, CCW SIJRGE TANK HIGH-LOW LEVEL, alarms.
0 CCW Surge ranklevel is 39% and slowly decreasing.
ANSWER:
a. * ALB-004-2-2. REFUELING WATER STOR4GE LO\a IXVIiI.. alarms.
RWST level is at 94% and sloivly decreasing.
Harris NRC Written Exmiination
Reactor Operator
Data Sheets
QUESTION NUMBEIZ: 49 TIERlGROtJP: 112
1OCFR55 CONTENT: 41(b) 10 43w
KA: 00005YG2.4.4
Ability to recogni7e abnonnal indications for system operating parameters which me entrylevel
conditions for emergency and abnormal operating procedures. (Accidental Liquid Radwaste Release)
OBJECTIVE: AOP-3.8
DENTIFY symptoms that require entry into AOP-008, Accidental Release of Lquid Waste
DEVELOPMENT REFERENCES: AOP-008, p 3
AOP-022, p 3
ALB-005, p 39
REFERENCES SUPPLIED TO APPLICAST: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: .40P-3.8001
NRC EXAM HISTORY: Xone
DISTRACTOR JIJSTIFICACTIOK (CORRECT ANSWER Xd):
X a. Under these conditions no water should be taken out of the KWST, so the decreasing level and alann
will require e n t v into AOP-008.
b. Plausible since water is being released to the Turbine Building, but actions are taken per AOP-010 to
address this.
c. Plausible since water is k i n g released to the Turbine Building, but actions taken in response to a SW
leak are per AOP-022.
d. Plausible since water is being lost from the CCW system, but actions taken in response to a CCW leak
are per AQP-014.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of entry requirements for accidental liquid release
QUESTION: 50
Which of the following actions would be most effective in responding to a Pressurized
Thermal Shock condition in accordance with FRP-P. 1, Response to Pressuriied ?hernial
Shock?
a. From the MCR, close the block valve for any open PRZ PORV
b. From the MCR, isolate any stuck open steam dunip valve
c. Direct an operator to the steam tunnel to locally isolate any stuck open SG PORV
d. Direct an opcratcir to the steam tunnel to locally isolate any stuck open MSIV
ANSWER:
c. Direct an operator to the steam tunnel to locally isolate any stuck open SG POKV
Harris NRC Written Examination
Reactor Operata
Data Sheets
QIJESTION NUMBER: 50 TIEWGROUP: 112
10CFR55 CONTENT: 41(b) 7 43W
KA: WEOIJG2.1.30
Ability to locate and operate components3 including local controls. (Pressurized Thermal Shock)
OBJECTIVE: 3.14-1
DESCRIHF the purposg ofthe following EOPs including the Qpe of event for which they were designed
and the major actions performed
FRP-I>.1 , Response to Imminent Pressurized Thermal Shock
DEVELOPMENT REFERENCES: FRP-P.l, pg 6
KEFE:RE.NC;ES SUPPI,IF,D TO APPLICANT: None
QEESTION SOURCE: 0 X NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOH SIGNIFICANTLY I\lODIFIED / DIRECT: New
NRC EXAM HISTORY None
DISTRACTOR JITSTIFLCACTION (CORRECT ANSWER X'd):
a. Plausible since closing the block valve for a stuck open PRZ PORV is an action taken in FRP-J.l,
though it is performed to maintain RCS inventory and will cause pressure to increase which would
cause the severity of a PIS event to worsen.
b. Plausible since a stuck open steam dump valve would contribute to the cooldown associated with a
PTS event, but individual steam dump valves cannot be operated from the MCB.
X e. A stuck open SG PORV would contrihute to the cooldown associated with a PI'S event. L,ocally
isolating the S G PORV would stop any cooldown caused by the SG PORV.
d. Plausible since locally closing a stnck open MSIV would assist in terminating a cooldown, but the
MSIV is located in the RAW and not the steam tunnel.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of plant conditions during a PTS event to determine the most
appropriate course of action
Harris NRC Written Examination
Reactor Operator
QUESTION: 51
Given the following conditions:
e RIiR Pump 1A-SA is operating chiring a plant heat up
The RHR Pump 1A-SA control power fuses blow.
Which of the following describes how the Main Control Board pump indication and local
breaker control is affectcd by the loss of the control power fuses?
a. Main Control Board red i green running indications will be lost
- The breaker will trip
Local open / closed light indication and local breaker control will be lost until
control power is restored
b. Main Control Board red / green running indications will be lost
The breaker remains closed
Local open / dosed light indication will be lost, hut local breaker control is
possibie without the control power
c. * h.lain Control Board red / green running indications will be available
- The bre&er will trip
e Local open / closed light indication is available, but local breaker control is
possible without the control power
d. e Main Control Board red / green running indications will be available
The breaker remains closed
Local open i closed light indication is available, hut local breaker control *ilI
he lost until control power is restored
ANSWER
b. * Main Control Board red / green running indications will be lost
e l h e breaker remains closed
e Local open I closed light indication will be lost, but local breaker control is
possible without the control power
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 5 1 TIE:R/GROIJP: 2/ 1
1OCRRRsS CONTENT 41(b) 4 43@)
KA: 062A4.04
Ability to manually operate and/or monitor in the control room: Local operation of breakers
OBJECTIVE: 480V-3.0-KI
State the function of breaker control power and discuss the effects of a loss of breaker control power
DEVELOPMENT REFEKENCES: OP-156.02, p 10,61
480V-I,P-3.0, p 11
REFERENCES S1JPPLIE.DTO APPLICANT: Xone
QliESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 480V-Ri 001
NRC EXAM HISTORY: None
DISTRACTOR .JUSTIFICACTION (CORRECT ANSWE.R Xd):
a. Plausible since MC.R and loc.al indication will be lost, but the breaker will not trip open on the loss of
control power and local breaker control is still possible.
X b. A loss of control power will cause MCB and loc.al indication to go out, but the breaker remains closed
and local breaker control is still possible.
c. Ilausible since local breaker operation is still available. but the breaker will not trip and MCB and
local indication will be lost.
d. Ilausihle since the breaker remains closed, but the loss of control power will result in a loss of MCB
and local indication and the breaker can still he locally operated.
DIFFICULTY ANALYSIS:
COMPRFHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowlecige oftlie effect of a loss of control power to a 480V breaker
Harris WRC Written Examination
Reactor Operator
QUESTION: 52
Which of the following situations would result in an inadvertent dilntion of the RCS
during Mode 1 operation and. after the crew has adjusted core reactivity to compensate
for the change in boron concentration. which procedure would be used to address the
cause of the event?
a. 0 RCP thermal barrier heat exchanger leak
0 ,201-016, Excessive Primary Plant Leakage
b. e 4 tube Ieak in the CVCS I.etdovm heat exchanger
AOP-014. Loss of Component Cooling Uater
c. e A mixed bed demineralizcr thdt was last in .senice t h e e weeks ago is
mistakenly placed in service at the end-of-cycle
o 40P-033, Chemistry Out of Tolerance
d. o A tube leak in the Seal Water heat cxchanger
0 AOP-014. Loss of Component Cooling Water
ANSWER
d. e A tube Ieak in the Seal Water heat exchanger
e AOP-014. Loss of Component Cooling Water
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 52 TIEWGROUP: 2!1
10CFR55 CONTENT: 41(b) 5 4303)
KA: 004A2.06
Ability to (a) predict the impacts o f the following malfunctions or operations on the CVCS; and (b) based
on those predictions, use procedures to correct, control, or mitigate the consequences of those
malfunctions or operations: Inadvertent boratioddiliition
OBJECTIVE: IE-3.12-3
Identify systems whose ope.ration may alter RCS boron concentration and discuss how operation of thew
systems may affect boron concentration
DEVELOPMENT REFERENCES: SOER 94-2, p 11-12
AOP-014, p 3,20
AOP-14-BD, p 20
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: IE-3.12-IU001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausihle since the thermal harrier interfaces with a non-borated system (CCW), hut leakage would be
out of the KCS to CCW and would not affect RCS boron concentration.
b. Plausible since CCW cools the heat exchanger and would dilute the RCS if leakage from CCU were
to occur, but letdow-n is at a higher pressure than CCW.
6. Plausible since boron concentration will change in CVCS, but this wrould result in an inadvertent
horation rather than a dilution.
X d. A seal water IIX leak will result in CVCS being diluted by CCW. This failore is to be addressed by
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze the effect of each failure on RCS boron Concentration and determine
the required procedure to address the failure
Harris NRC Written Examination
Reactor Operator
QLJESTION: 53
Given the following conditions:
e The plant is in Mode 4.
e The RCS in a solid plant condition.
lUIR pump 1,4-SA is in service.
In accordance with GP-007, Normal Plant Cooldown, which ofthe following actions
should be taken to raise PRZ pressure to a new steady-state value?
a. Throttle 1CS-28, HC-142.1 RHR LETDOWN, in the shut direction
b. Shut lCS-4,45 GPM LETDOWN ORIFICE A
c. Adjust the setpoint for 1CS-38, PK-145.1 LTDN PRESSIJU, to cause the valve
to go in the shut direction
d. Adjust the setpoint for 1CS-23 I , FK-122.1 CHARGING FLOU, to cause the
valve to go in the open direction
ANSWER
c. Adjust the setpoint for 1CS-38, PK-145.1 LiUN PRESSURE, to cause the valve
to go in the shut direction
Harris NRC Written Examination
Reactor Operatoi
Data Sheets
QIJESTION NIJMREK: 53 TIEWGROUP: 211
10CFR55 CONTENT 4l(b) 2-9 43w
K4: OIOK1.06
Knowledge of the physical connections and/or causeeffect relationships between the PZR PCS and the
following systems: CVCS
OBJECTIVE: GP-3.7-2
With regard to RCS cooldown, DESCRIBE the following per GP-007
The two methods used to c.ontro1 RCS pressure, including the elcments of each
DEVELOPMENT REFERENCES: GP-007, p 41
KEFERENCES STJPPL1E.DTO APPLICANT: Xone
QUESTION SOURCE: 0 NEW SIGKIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Harris LOCT S84
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANS\VE.R Xd):
a. Plausible since this would cause an increase in RCS pressure, hut 1CS-38 will respond to cause
pressure to lower again.
b. Plausible since this would cause an increase in KCS pressure, but ICs-38 will respond to C ~ U S C
pressure to lower again.
X e. Adjusting the setpoint of ICs-38 wsill cause the backpressure on the RHR pump and the. RCS to
increase and is the method of control used.
d. Plausible since this would cause an increase in RCS pressure, hut 1CS-38 will respond to cause
pressure to lower again.
DIFFICuLrY ANALYSIS:
COMPREHEXSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICIJLTY RATING: 3
EXPLANATION: Comprehension of the effects of adjusting CVCS components on PRZ pressure
Harris NRC Written Examination
Keacror Operator
QUESTION: 54
125 VDC battery 1A-SA is currently loaded at 292 amps and is expected to be discharged
in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
II'DC load shedding is perfbrmed such that the loading on the battery is reduced from
292 amps to 146 amps. how long should the battery be available to supply the rcmaining
loads?
a. 4hours
b. More than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, but Iess than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
c. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
d. More than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
ANSWER:
d. More than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
EIarris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION IVIZWRER: 54 TIEWGROUP: 211
10CFR55 CONTENT: 41(b) 5 43W
KA: 063A1.01
Ability to predict and/or monitor changes in parameters associated with operating the DC electrical
system controls including: Battery capacity as it is affected by discharge rate
OBJECTIVE: DCP-3.0-A3
STATE the function and EXPLAIN the basic operation of the following major components of the I X
Power System:
Batteries
DEVELOPMENT REFERENCES: EPP-001, p S5
ADEI,-LP-2.6, p 3
DCP-LP-3.0, p 8
REFERENCES SUPPLIED TO APPLICANT: None
QKESTION SOURCE:
0X NEW SIGNiFICANTLY MODIFIED 0 DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR ;TZTSTIFICACTION(CORRECT ANSWER Xd):
a. Iiausible since the battery is rated for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, hut at a discharge rate of approximately 193 amps per
hour and decreasing the discharge rate would increase the capacity.
b. Plausible since tbe discharge rate has been decreased which would extend the capacity of the battery
for a period of time. but the time would he more than doubled.
c. Plausible since the discharge rate has been halved, so it would appear that the capacity would he
doubled, hut it is a non-linear relationship.
X d. Reducing the discharge rate ou a battery increases the battery capacity in a non-linear function such
that decreasing the discharge rate by half, increases the capacity by more than double.
DIFFICULTY ANALYSIS:
fl COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE I RECALL
DIFFICULTY RATING: 4
EXPLANATION: Calculation of the nominal discharge rate of a battery and comprehension of the
effect of reducing discharge rate on battery capacity
IIarris NRC Written Examination
Reactor Operator
QUESTION: 55
Given the following conditions:
The plant has experienced a Large Brcak Loss of Chlant Accident during a reactor
startup.
o All equipment functioned as designed and the crew has reached the point in PATH-1
where monitoring Critical Safety Function Status Trees is required.
Which one of the following statements describes the IMMEDIATE result that voiding in
the downcomber region would have on the Source Range instrumentation and procedure
used to mitigate these piant conditions?
a. The displacement of douncomkr water would increase the neutron leakage
and result in a higher SOL KC^ range count rate.
o Thc crew should continue in PATEI-1 rather than transition to FRP-S.2,
Response to Loss of Core Shutdown.
b. e A decrease in downcomber water density would reduce fission and result in a
lower source range count rate.
The crew should transition to FW-S.2, Response to Loss of Core Shutdown,
rather than continue in PATII-1.
c. The displaccment of boron from the downcomber region would increase
fission and result in a highcr source range count rate.
The crew should continuc in PATH-1 rather than transition to FIV-S.2,
Response to Loss of Core Shutdown.
d. e A decrease in downcomber water density would reduce fission and result in a
lower source range count rate.
The crew should continue in PATH-1 rather than transition to FW-S.2,
Response to Loss of Core Shutdown.
ANSWEK:
a. The displacement of downcomber water would increase the neutron leakage
and result in a higher source range count rate.
o The crew should continue in PATH- 1 rather tlran transition to FW-S.2,
Response to Loss of Core Shutdown.
Harris NRC Writton Examination
Reactor Operator
Data Sheets
QUESTION NTJMBER: 5 5 TIENGROUP: 2!2
10CFR55 CONTENT: 41(b) 5 43m
KA: 01SA2.0.5
Ability to (a) predict the impacts of the following malfunctions or operations on the N S ; and (b based on
those predictiow, use procedures to correct, control, or mitigate the consequences of those malfunctions
or operations: Core void formation
OBJECTIVE: RD-3.10-7
Explain the NIS response to different void fractions in the core and downcomer region
DEVELOPMENT REFERENCES: HO-BD-3.10 pg 26-24
REFEREXCES SUPPLIED TO APPLICANT: None
QLTSTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIKECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: INPO 20608
NRC EXAM HISTORY: None
DISTRACTOR RJSTIFICACTION (CORRECT ANSWER Xd):
X a. Downcomber voiding results in higher source range indication due to increased leakage. The crewr
should continue in PAIR-1 rather than transfer to FRP-S.2 since entry c.onditions to FRP-S.2 are a
Yellow path condition.
b. Plausible since a severe decrease in core water density would result in less moderation and a lower
power level, but downcomber density has little effect on core reactivity.
c. Plausible sincc displacing core boron would result in a higher power level, but downcomber density
has little effect on core reac.tivity.
d. Plausible sincc a severe decrease in core water density would result in less moderation and a lower
power level, but downcomber density has little effect on core reactivity.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY HATING: 3
EXPLANATION: Analysis of the effects of core voiding on SR indication and knowledge ofthe
procedure hierarchy during the performance of the EOPs
Harris NKC Written lixarnination
Reactor Operator
QIJESTION: 56
Given the following conditions:
a A transition has just been made to FW-S.1, "Response to Nuclear I'ower Generation
'The Reactor Operator is manually inserting control rods.
0 All Turbine Throttle Valve (TV) and Turbine Governor Valve (GV) indications show
the RED light OFF and the GREEN light ON, with the exception of TV-3 and GV-2
which have both the RED light and GWEN light ON.
- Turbine speed is decreasing, and is currently 1680 rpm.
The Main Steam Isolation Valve (MSIV) Bypass valves are closed
Which of the follow4ng actions should he taken next?
a. Verify all AFW pumps running
b. Manually trip the Turbine from the MCB
c. Place both Turbine IlIiH pumps in PULL,-1'O-I,OCK
d. Shut all MSIVs
ANSWER:
b. Manually trip the Turbine from the MCB
Harris NRC Written Exmiination
Reactor Operator
Data Sheets
QUESTION NUMBER: 56 TIEWGROUP: 22
1OCFR55 CONTENT: 41(h) 7 4W)
KA: 045.44.06
Ability to manually operate and/or monitor in the control room: Turbine stop valves
OBJECTIVE: 3.1 5-4
Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
Order of preference for turbine trip steps from the MCB
DEVELOPMENT REFERENCES: FW-S.1 pg 4
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED [7 DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.15-K2 001
NRC EXAM HISTORY: None
DISTRACTOR J[;STIFICACTION (CORRECT ANSWER Xd):
a. Plausible since GV-2 and TV-3 are assoc.iated with opposite steam chests and it may be assumed that
as iong as the GVs are closed for 1 steam chest and the iVs are closed for the other steam che.st with
turbine speed decreasing, and starting AFU is the next step in the proccdure, however the turbine
should not be considered to be tripped.
X b. Verification of a turbine trip requires either all 4 TVs be closed or all 4 GVs be closed. If one set of
these valves are not all closed, then the RNO directs manually tripping the turbine from the MCB.
c. Plausihle since the turbine should not be considered to be tripped based on indications, and this is an
RNO action, but should not be perfonned until amanual trip from the MCB is attenipted.
d. Piausible since the turbine shouid not be considered to be tripped based on indications, and this is an
RNO action, but should not be performed until a manual trip from the MCN is attempted.
DIFFICULTY ANALYSIS:
17 COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the required indications for a turbine trip and the priority for
tripping the turbine if a trip cannot be verified
Harris NIK Written Examination
Reactor Operxtor
QUESTION: 57
Given the foilowing conditions:
e The Main Control Room has been evacuated aid control transferred to the Auxiliary
Control Panel (ACP).
AOP-004, Remote Shutdown, is being performed when a ioss of offsite power
coincident with a Safety Injection signal occur.
Which of the following describes the response of the plant?
a. The Emergency Diesel Generators automatically start and the sequencers load the
EDGs due to the undervoltage signal
b. The Emergency Diesel Generators automatically s t w and the sequencers load the
EDGs due to the safety injection signal
c. The Emergency Diesel Generators automatically start, hut must be manually
loaded with the required loads
d. The Emergency Diescl Generators must he manually started and manually loaded
with the required loads
ANSWER:
a. The Emergency Diesel Generators automaticaliy start and the sequencers ioad the
ED& due to the undervoltage signal
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 57 TIEWGROUP: 211
1UCPR55 CONTENT: 41(b) 7 43ib)
K4: 064A3.07
Ability to monitor automatic operation of the FWG system, including: Load sequencing
OBJECTIVE: AOP-3.4-Ii5
IIISCXJSS how a transfer to the auxiliary control panel would affect the following inputs to the EST;
sequencers
Safety injec.tion signal
- Safety bus undervoltage signal
DEVELOPMENT KF,FERENCES: AOP-004 pg 91
AOP-004-BD pg 26
SD-155.02 pg 6-9
REFERENCES STJPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DILRECT: AOP-3.4-K6 001
NRC EXAM HISTORY: Ncne
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
X a. The EDGs should automatically start on the UV condition and the 1IV signal will still cause the
sequencer to operate. Only the SIAS input to the sequencer is defeated upon transfer to the ACP.
h. Plausible since the EUG will automatically start, but loading will be based upon the UV signal.
c. Ilausible since the EDG will automatically stcart,hut loading will be based upon the UV signal
d. Plausible since many automatic fcmctions are defeated when conteol is transferred to the ACP, but the
EIX3 will automatically start and loading will be based upon the UV signal.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLALVATION: Analysis of the effect of a transfer to the ACP on the EDG and sequencer
operation
Harris NRC Written Examination
Reactor Operator
QUESTION: 58
Given the following conditions:
The unit is operating at 100% power.
e k'ollowing maintenance on 1A-SA Emergency Iliesel Generator (EDG), it is
determined that a common mode failure exists which renders both EDGs innperable.
Which of the fdlowing actions are required to be taken within one (1) hour ofdeciaring
both EDGs inoperable'?
a. Verify and rec.over required functions
b. Restore one (1) ofthe EDGs to operabre status
c. Verify off site power availability
d. Initiate actions to pIace the unit in IIot Standby
ANSWER:
c. Verify off site power availability
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 58 TIERIGROUP: 3
1OCFR55 CONTENT: 41(h) None 43(b) 2
Ea: 2.2.24
Abilit) to analyze the affect of maintenance activities on LCO status
OBJECTIVE: DE3.0-20
Given a plant mode of operation and the applicable LCO-related parameters for an EIX;. IWTtiRMINE
if a Technical Specification one-hour (or less) action statement applies
DEVELOPMENT REFERENCES: TS 3.8.1.1, PB 314 8-3
OST-1023, pg 1-2
REFERE,NCE.SSUPPLIED TO APPLICANT: None
QUESTION SOTJRCE: 17 NEW H SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Harris LOCT 385
h%C EXAM HISTORY: None
DISTRACTOR JUSTIFTCACTION (CORRECT ANSWER Xd):
a. Plausible since it is an action for the EDG operability, however it is not a requirement to verify nor
recover in a 1-hour time frame.
b. Plausihle since restoration of one EDG to operable sLitns is required, but it is required to be performed
within 2 honrs not 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
X c. OST-1023 is required to he performed within one hour to verify off site power capability
d. Plausible since TS 3.0.3 would be required to be entered if an additional loss of off site capability also
existed, but with only the 2 EDGs inoperable this is not required.
DIFFICULTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS H KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> actions required by Technical Specifications
IIarris NRC Written Examination
Reactor Opcrator
QUESTION: 59
Given the fdowing conditions:
The plant has experienced a small break LOCA
The crew has transitioaed to EPP-009, Post LOCA Cooldown and
1)epressurization.
0 The ERFIS computer is failed.
o Containment pressure peaked at 8 psig, but is now 4.5 psig and decreasing dowry
Present pressure indications are:
PI-455.1, PR% IRESSURE CHI = 1800 psig
o 11-456, PRZ PRESSURE CII I1 = 1770 psig
0 PI-457, PK7, PRESSURE CII 111 = 1740 psig
e PI-402.1, IZCS WIDE K4NGE PRE.SSIJRE = 1840 psig
o PI-404, RCS WIDE RANGE IRESSIJRE = Failed High
Which of the following will be used to determine the primary plant pressure?
a. Use PI-457 down to 1700 psig and use PI-402.1 below 1700 psig
b. Use PI-456 down to 1700 p i g and use PI-402.1 below 1900 psig
c. Use PI-455.1 down to 1400 psig and use PI-402.1 below 1700 psig
d. Use PI-402.1 at all pressures
ANSWER
d. lJse PI-402.1 at all pressures
Harris NRC Written Examination
Reactor Operator
Data Sheets
QIJESTION NIJMRER: 59 TIEWGROUP: 3
10CFR55 CONTENT: 41(b) 6 4303)
Kn: 2.4.3
Ability to identi@ post-accident instrumentation
OBJECTWE: 3.19
DESCRIBE Control Room usage of EPPs, foldouts, and FRPs as it relates to the following:
g. Use of RCS wide-range pre\sure indication
DEVELOPMENT REFERENCES: EOP [Jsers Guide pg 27,38
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUAMBERFOR SIGNIFICANTLY MODIFIED /DIRECT: IIarris LOCI' 816
NRC EXAM HISTORY: None
DISTRACTOR .TUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since PI-454 is the lowest reading ofthe pressures and would be the most conservative, but
with adverse containment conditions the post-accident instrument PI-402. I is to be used.
h. Plausible since PI-456 is the highest reading of the pressures and would likely provide the highest indication
until 1700 psig is reached, but with adverse containment conditions the post-accident instrument PI-402.1 is to
be used.
c. Plausible since PI-455 is the median reading of the pressures and would likely provide the average
indication until 1700 psig is reached, but with adverse containment conditions the post-accident
instrument PI-402.1 is to be. used.
X d. Adverse containment conditions still exist so the post-accidcnt instrument, PI-402.1 is to be used at all
pressures.
DIFFICULTY AYALYSIS:
1 COMPREHENSIVE /ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of plant conditions and instrument failures to determine indications to
use during adverse containment
Iinrris NKC Written Examination
Reactor Operator
QlJESTION: 60
Assuming that all other equipment is operable, which of the following would require an
entry into Technical Specification 3.8.2.1, I X
Sources .. Operating (Modes 1-4). action
statements?
a, EMERGENCY BUS A-SA TO AIJX BCS D TIE BREAKER 105 SA trips open
and EDG IA-SA automatically starts and loads
b. 480V EMPX(iEKCY BUS 1.43-SA main feeder breaker trips open
c. BATTERY CHARGER 1A-SA is placed under clearance
d. EMERGENCY BAITERY IA-SA is placcd on a float charge
ANSWER:
b. 480V IiMERGEiYCY BUS lA3-SA main feeder breaker trips open
Harris NRC Written Examination
Reactor Opwatw
Data Sheets
QUESTION NCJTMBEK: 60 TIEWGROUP: 2/ 1
1QCFR55CONTENT 41(b) None 43(b) 2i3
KA: 000058G2.1.33
Ability to recognize indications for system operating parameters which are entry-level c.onditions for
technical specifications. (Loss of DC Power)
OBJECTIVE: DCP-3.0-RI
Given the name o f a component in the DC power system, state whether or not that component is
'l'echnical Specification related
DEVELOPMENT REFERENCES: TS 3.8.2.1, p 3!4 8-12
SD-156, p 24
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOIIJRCE: 0 X NEW 0 SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMl3E.R FOR SIGNIFICANTLY MOD1FIE.D/ DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since the Program A sequencer (LOSP) will strip some MCCs which supply DC battery
chargers, but the A-SA and the B-SA battery chargers will remain capable of maintaining power to the
A-SA battery.
X b. A loss of 480V Emergency AC Bus 1A3-SA will result in a loss of both MCCs IA21-SA and 1.431
S.4, which would cause both A train battery chargers to be inoperable.
e. Plausible since removing a battery charger from service would resuit in a 'TSentry if the other charger
is also out of service, but a single charger will not result in an entry to an action statement.
d. I'lausible since a float charge is a surveillance requirement and most surveillances make the associated
equipment inoperable, but the normal configuration oflhe battery is on a float charge.
DIFFICULTY ANALYSIS:
COMPIIEIIENSIWZ / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EWLANATIOX: Analysis of the eff%t o f a loss of AC power requiring a TS entry for DC power
QUESTION: 61
Given the following conditions:
The plant is operating at 100% power when AI,13-010-1-lB. RCP A IJPPER OIL
KSVR LOW;-LEVEL, alann is received.
The operator chechs the computer points for GI> AOP-018 and find5 RCP A motor
thrust-bearing temperature at 195°F and KCP A upper d i a l bearing at 185F with
both slowly increasing.
Uhich of the following actions are required?
a. Stop KCI Aand initiate a rapid plant shutdown in accordance with AOP-038,
Rapid Downpower
b. Manually trip the reactor and go to PATII-1, stopping RCP Aas time perniits
c. Continue monitoring RCP A temperatures, tripping the reactor and entering
PATH-1 if RCP A temperatures exceed 300*F
d. Stop RCP A, manually trip the reactor and go to PATH-]
ANSWER:
b. Manually trip the reactor and go to PATH-1; stopping RCP Aas time permits
Harris NRC Written Examination
Reactor Operator
Data Sheets
QDESTION NUMBER: 6 1 TIEWGROUP: 1/1
10CFR55 CONTENT: 41(b) None 43(b) 5
KA: 00001 5117AA2.08
Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions
(Loss of RC Flow.): When to secure KCPs on high bearing temperature
OBJECTIVE: AOP-3.18-3
Given a set of plant conditions and a copy of AOP-018, DBI'EKILIINE the appropriate response
DEVELOPMENT REFERENCES: AOP-018 pg 21,27
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOCRCE: 0 NEW 4 SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: AOP-3.18 019
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTIOW (CORRECT ANSWER X'd):
a. Plausible since the KCP is to be stopped, but must be stopped immediately which requires that the
reactor be tripped.
X h. RC'P motor temperatures require the pump be stopped. With power abo\e 48%, the reactor must be
tripped prior to tripping the RCP.
c. Plausible since this is a trip setpoint for stator winding temperature, but the pump must be tripped
immediately based on the given temperatures.
d. Piausible since these are the correct actions. but the reactor should be tripped first and the pump
stopped when time permits.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY HATING: 2
EXPLANATION: Knowledge of RCP motor temperature tripping requirements
IIarris NRC Written Examination
Reactor Operator
QIJESTION: 62
Given the following conditions:
Pdth-2 is being performed due to an SGTR.
The MSIV on the ruptured SG is mechanically stuck open.
The Main Steam Isolation Valves (MSIVs) on the intact SGs are closed.
The Condenser is available for Steam Dump operation.
e A cooldown to 485 "F from 557 "I; at the maximum rate is required.
Which of the following describes the method to accomplish this cooldown in accordance
with PATII-2 and the EOP User's Guide?
a. Fully open the Steam Dumps as fast as possible
b. Fully open the S t e m 1)umps as fast as possible without causing a main steam line
isolation
c. Fdly open the intact SG PORVs as fast as possible
d. Fully open the intact SCi PORVs as fast as possible without causing a main steam
line isolation
ANSWER
c. Fully open the intact SG PORVs as fast as possible
Harris NRC Written Examination
Reactor Opemtor
Data Sheets
QUESTION NUMBER 62 TIEPUGROUP lil
1QCFR55CONTENT: 41(b) 7 43@)
KA: 000038E.Ai 3 6
Ability to operate and monitor the follow-ing as they apply to a SGIK: Cooldown of RCS to specified
teruperature
OBJECTIVE: 3.19-R4
Given a set of conditions during E.OP implementation, DETE.RMINE the correct response or required
action based upon the EOP Users Guide general information
- Dumping steam at maximum rate
DEVELOPMENT REFERENCES: EOP Users Guide, p 38
PATH-2 Guide, p 8, I O
REFERENCES SUPPLIED TO APPLICANT: None
QTJESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIKECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOP-3.10-R4 001
NRC EXAM HISTORY: None
DISTRACTOR JCJSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the maximum cooldown rate can be achieved using maximum steam dump flow, but
causing too great a rate of pressure drop will result in the MSHVs going closed which is undesirable
and it is also undesirable to use steam dumps when the ruptured SC MSIV is open.
h. Plausible sinc.e the maximum cooldowrn rate is desirable using maxinlum steam dump flow without
causing too great a rate of pressure drop will result in the MSIVs going closed, but it is also
undesirable to use steam dumps when the ruptured SG MSIV is open.
X e. During a SGTK cooldown only the intact SGs should be used to cooldown the KCS and sinc.e the
MSIVs on the intact SGs are closed, tire PORVs should be used. The \ d v e s should be opened as fast
as possible since generation of an MSIV signal is not a c.oneern.
d. Plausible since causing the MSIVs to close is not desirable when steam dumps arc being used, but
when already using POKVs to dump steam this is not a eonc.ern.
DIFFICULTY AXALYSIS:
0 COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
1)IFFICULTY RATING: 3
EXPLANATION: Knowledge of the EOP Users Guide requirement for performing a maximum
rate cooldown
Harris NRC Written Examination
Reactor Operator
QCESTION: 63
Given the following conditions:
After transferring resin, it is noted that RM-lWR-36344., SPEKT RESIN PUMP
1-44., radiation monitor is indicating 10 mRern;hr.
e The monitor is physically located 20 feet away from a suspected clog in the pipe
which is the source of'the monitor indication.
An operator must hang a clearance on a valve that is located 5 feet from the suspected
clog in the pipe.
What is the dose rate in the area where the operator wiil be hanging the clearance?
(ASSUME THE CLOG IN THE PIPE IS A POINT SOURCE)
a. 20mRcmh
b. 40mRenvhr
c. 80mRemh
d. 160 mRem/hr
ANSWER
d. 160 mRem,hr
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 63 TIEWGROW: 2; 1
10GFR55 CONTENT: 41(b) 5 43m
KA: 073K5.02
Knowledge of the operational implications as they apply to concepts as they apply to the PKM system:
Radiation intensity changes with source distance
OBJECTIVE: KP-3.5-21
Cdculate dose rates at different distances from point sources and line sources
DEVE.LOPMENT REFERENCES: RP-LP-3.5 pg 22 and
Attachment 1 pg 7
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE:
0X 3TW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLYMODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFKCACTION (CORRECT ANSWER Xd):
a. Plausible if the square root of the distances is taken, instead of squared as they should be (IOmKihr x
201i2fi:::2()mR/hrx51,Q fit),
b. Ilausible if the distances are not squared as they should be (1OmRhr x 20 ti = 40 mWhr x 5 ft).
c. Plausible if a mathematical error is made (vdue selected as a distractcr due to the progression of other
numbers in distracters).
X d. Using the formula Ildi2=I2d;, the intensity ofthe source ;rt 5 feet is calculated to be 160 mRemihr.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICTJLTY RATING: 3
EXPLANATION: Calculation of distance using inverse square for radiation
Harris NRC Writtsn Examination
Reactor Operator
QUESTION: 64
Given the following conditions:
- The Control Room has been evacuated due to a fire.
- AOP-004, Remote Shutdown, is being performcd.
The crew has located the most recent OST-1036, Shutdown Margin Calculation.
and determined that 5.000 gallons of boric acid must be added to the RCS.
a Boric Acid Tank level is 77%.
What lcyel will the Boric Acid T.mk be at when the 5,000 gallons of boric acid are added
to the KCS AND why is there a concern ahout required shutdown margin during the
performance of AOP-004?
a. Final Boric Acid Tank level should be approximately 62% to ensure adequate
shutdown margin is maintained in the event that access to the Control Room is
prevented until the core has reached xenon-free conditions
b. Final Boric Acid Tank level should be approximately 56% to ensure adequate
shutdown margin is maintained in the event that access to the Control Room is
prevented until the core has reached xenon-free conditions
c. Final Boric Acid lank level should be approximately 62% to ensure adequate
shutdown margin is maintained in the event that a cooldown to Cold Shutdown
conditions is required
d. Final Boric Acid Tank level should be approximately 56% to ensure adequate
shutdown margin is maintained in the event that a cooidown to Cold Shutdolm
conditions is required
ANSWER:
c. F i d Boric Acid Tank level should be approximately 62Y0 to ensure adequate
shutdown margin is maintained in the event that a cooldown to Cold Shutdown
conditions is rquirwl
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 64 TIEWGROUP: 112
10CFR55 CONTENT: 41(b) 5/10 43(b)
KA: 000068AK3.13
Knowledge of the reasons for the following responses as they apply to the Control Room Fmcuation:
Performing a shutdown margin calc.ulation, including horon needed a i d horation time
OBJECTIVE:
Given a set of plant conditions and a copy of AOP-004, Remote. Shutdown, DETERMINE. the
appropriate course of action
DEVELOPMENT REFERENCES: AOP-004-BD pg 47
Curve 11-2
REFERENCES SUPPLIED TO APPLICANT: Curve I>-2
QUESTION SOURCE: X NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER xw:
a. PIausible since the RAT level will be at 62% following the 5,000 gallon addition, hut shutdown
margin is a concern in the event of a cooldown.
b. PIausiblc since errors occur when the graph is read, but the B A I level will he at 62% and shutdown
margin is a concern in the event ofa cooldown.
X e. A boration is only performed in the event that a cooldown is required to be performed during the
perfomlance of AOP-004. Using C m - e D-2, 77% level corresponds to 27,000 gallons. Adding 5,000
gallons to the RC.S w-ill leave 22,000 gallons, which corresponds to a BAT level of62%.
a. Plausible since a boration is only performed in the event that a cooldown is required to be performed
during the perfomlance of AOP-004, hut BAT level wlill indicate 62% and not 56%.
DIFFICIJLTY ANALYSLS.:
COMPREHENSIVE I ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the reason for performing a boration while operating the plant
from the shutdown panel and the ability to apply a plant curve
Harris NRC Written Exanuiation
Reaslor Operator
QUESTION: 65
Given the following conditions:
e The reactor is critical at IO-' anips.
?'he Channel I inverter output breaker trips.
b'hich of the following occurs as a result of the breaker tripping?
a. Reactor power remains at 10.' amps and Power Range Channel N-42 deenergizcs
b. Reactor power remains at 10.' amps and Power Range Channel N-41 deenergizes
c. The reactor trips due to Intermediate Range Channel K-36 deenergizing
d. The reactor trips due to Intermediate Range Channel K-35deenergizing
ANSWEK:
d. The reactor trips due to Intermediate Range Channel N-35deenergizing
Elarris NRC Written Examination
Reactor Operator
Data Sheets
QIJESTION NUMBER 65 TIEWGROUP: 211
10CFR55 CONTENT: 41(b) 7 4309
KA: 012K2.01
Knowledge of bus power supplies to the following: RPS channels, components, and interconnections
OBJECTIVF.: AOP-3 24-2
RECOGNIZE automatic actions that are associated with loss of an instrument bus or loss of "NS UPS
DEVELOPMENT REFERENCES: AOP-034, p 33,25,29,34
REFERENCES SIJPPLIED TO APPLICAXT: None
QIJESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODlFIED /DIRECT: Harris LOCT 453
KRC EXAM HISTORY None
DISTRACTOR JUS'ITFICACTIOX (CORRECT ANSWER X'd):
a. Plausible since a loss of power would result in a loss of PR Channel, but the trip occurs due to a loss
of N-3 5 I
b. Plausible since a loss of power would result in a loss of PR Channel, but the trip occurs due to a loss
ofN-35.
c. Plausible since a reactor trip would occur due to N-36 if instriiment bus I1 were lost, but the reactor
trips on a loss of instrument bus I due to a loss of N-35.
X d. A reactor trip would occur doe to N-35 failing if instrument bus I being lost.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RArIN(:: 2
EXPLANATION: Analysis of the efect of a loss of instrument bus power on plant conditions
Harris NKC Written Examination
Reactor Operator
QUESTION: 66
Given the following conditions:
- An earthyu&e has caused damage to the Main Reservoir dam.
Main and Auxiliary Reservoir levek are both currently 240 feet and stable.
- AOP-022, "I oss of Service Water," is being performed for a Loss of Ultimate IIeat
Sink.
a Fmergency Service Water (ESW) punips have been aligned to the Main Reservoir.
a One (1) Normal Service \J'ater (NSW) pump is operating.
Which of the following pumps are required to be operating to provide water to the SSE
Fire Protection IIeadcr once the ESW header is aligned to the fire protection header'?
a. ONLY an ESW pump
b. An ESW pump AND an ESW Booster pump
c. ONLY a second NSW pump
d. A second NSW pump AND an P;SW Booster pump
ANSWER
b. An ESW pump AND an ESW Booster pump
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 66 TIERIGRBUP: 211
10CFR55 CONTENT: 41(b) 2-9 43w
KA: 076KI.15
Knowledge of the physical connections andlor cause-effect relationships between the S W S and the
following systems: FPS
OBJECTIVE: FP-3.0-3
STATE the sources of fire water available to the plant including automatic actuation signals
DEVELOPMENT REFERENCES: AOP-022 pg 30
O1-139 pg 24
REFERENCES SUPPLIED TO APPLICAXT: None
QIJESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: FP 020
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xcl):
a. Plausible since an ESW pump is started, but an ESW Booster pump is also rqnired.
X b. An IlSW pump, aligned to the Main Reservoir, is started, along with an ESW Rooster pump to supply
the SSE fire protection header.
c. Plausible since thc first N S W pump is not required to be tripped provided cooling tower basin level is
adequate and N S W supplies the ESW header (which can supply the fire protection header), but an
ESW pump is required.
d. Plausible since an ESW Booster pump is required to supply the f r e header, but an ESW piitnp is
required to supply the booster pump.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the system alignments available to supply the fire header
Harris NRC Written Exariiinatiori
Reactor Ope1ator
QUESTION: 61
Given the following conditions:
The plait is being cooled down to 140°F for maintenance which will NOT require the
RCS be opened.
The crew is in the process of placing the first Residual k a t Removal (RHR) train in
service for RCS cooling.
a Current boron concentrations are as follows:
KHR (train to be placed in service) boron 1021 pprn
- Kequired Shutdown Margin boron I200 ppm
a RC'S boron I341 ppm
e Cold Shutdown boron 1450 ppm
Refueling boron 2261 ppm
Before the KHR train can he placed in service for RCS cooling, RHR boron
concentmtion must be increased by a MINIMUM o f . . .
a. 179ppm.
h. 320ppm.
c. 729ppm.
d. 1240ppm.
ANSWER
a. 179ppm.
Harris NRC Written Examination
Reactor Operator
Data Sheet.,
QUESITON NUMBER: 67 TIEWGROUP: 2!1
IOCFR55 CONTENT: 41@) 5 43m
KA: 005K5.09
Knowledge of the operational implications ofthe following concepts as they apply the RI-IRS: Dilution
and boration considerations
OBJECTIVE: RI IRS-2.0- I2
APPLY precautions and limitations of OP-11 I, RHRS to Hypothetical System Configurations
DEVELOPMENT REFERENCES: 01- 1 I 1 pg 7
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: S e w
NRC EXAM HISTORY: Xone
DISTRACTOR JKJSTIFICACTIOK (CORRECT ANSWER Xd):
X a. KHR boron must he greater than or equal to the required SDM or the required refueling concentration.
The boron concentration requirements will be dependent on the intended use of the RHR System.
IJsing the RHR system for c.ooldown purposes requires that the boron concentration he greater than or
equal to the required shutdown margin.
b. Plausible since this is the difference between RIIR and RCS boron concentration, but only the
c. Plausible since this is the difference between RHR and Cold Shutdown boron concentration, but only
the required SDM boron is needed.
d. Plairsible since this is the. difference between RHR and refueling boron concentration, and refueling
conditions occur at 140F, hut only the required SDM horou is needed.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Application of actual versus required boron concentration - must determine
minimum limiting requirement
Hmis NRC Written Examination
Reactor Operator
QUESTION: 68
Given the following conditions:
e A iiquid waste discharge from a Tre&d Laundry and Hot Shower (TL&HS) Tank is
in progress.
REM-1WL-3540. Treated Laundry and Hot Shower Tank Pump Discharge Monitor.
goes into high alarm.
Which of the following terminates the discharge?
a. The running TL&HS Tank Punip will automatically trip
b. 31,HS-301. Treated I,&HS Tks Discharge to Cooling Tower Blowdown. will
automatically close
c. 31,HS-293, Flow Control Valve Treated L&HS Ik to Cnviro. will automatically
close
d. 3LHS-396. TL&IIS Pank Pump Discharge Isolation Valve, mill automatically
close
ANSWER
d. 31.HS-336, TL&HS Tank Pump Discharge Isolation Valve, will autoinatically
close
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 68 TIEWGROUF': 2i2
10CFR55 CONTENT 41(b) 7 43w
Kri: 068A3.02
Abiliw to monitor automatic operation of the Liquid Radwaste System including: Automatic isolation
OBJECTWE: LU'PS-LP-3.0-7
DESCRIBE the automatic protection features associated with discharges to the environment from the
LWPS
DEVELOPMENT REFERENCES: AOP-005, p 17-28
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DJXFT1
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: RMS-A6 005
NRC EXAM HISTORY None
DISTRACTOR .ICISTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since the pump will stop the dischmge, but there is no auto trip due to high rad.
b. Plausible since closing this valve will stop the discharge. but this valve does not receive an automatic
closure bignal.
c. Plausible since this valve is in the flow path and will stop the discharge, but this valve does not
re.ceivr an automatic closure signal.
X d. On 3 high rad level as sensed by REh4 3540, the discharge isolation valve will automatically close,
terminating any release in progress.
DIFFICULTY ANALYSIS:
0 COMPREHE.NSIVE/ AXALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of liquid radwaste design and operation
Harris NRC Written Examination
Reactor Operator
QUESTION: 69
Assuming NO operator actions, which ofthe following describes the effect of a Ioss of
instrument air on Volume Control Tnnk (VCT) level?
a. VCT level decreases due to maximum charging and letdown isolation valves
closing
b. VC?' level decreases due to maximum charging and letdown being diverted to the
Hold Up Tank
c. VCT level increases due to minimum charging and the letdowi pressure control
valve failing open
d. VCT level increases due to minimum charging and the letdown orifice isolation
valves failing open
ANSWER
a. VCT level decreases due to maximum charging and letdown isolation valves
closing
Harris KRC Written Examination
Reactur Operator
Data Shsets
QUESTION NU.MBEK: 69 TIE.R/GROUP: 2/1
1QCFR55CONTENT: 41(b) 7 13(b)
KA: 078K3.02
Knowledge ofthe effect that a loss or malfunction of the IAS will have on the following: Systems having
pneumatic valves and controls
OBJECTIVE: AOP-3.17-4
Given a set of entry conditions, and a copy of AOP-017, DETERlMINE the appropriate response.
DEVELOPMENT REFEKENCES: AOP-017 pg 37
REFERENCES SUPPL.IEDTO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: CVCS-R3 008
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. Charging flow control fails open and letdown isolation valves fail closed on a loss of air, so VCT level
will decrease.
b. Plausible since VCT level will decrease, but it wilI be due to letdown isolating, not diverting water to
the hold up tank.
c. Plausible since the Letdown pressure control valve fails open on a loss of air, but the letdown isolation
valves Fail closed, isolating letdown.
d. Plausible since the charging flow c.ontrol valve and the letdown orifice valve all fail on a loss of air,
but fail in the opposite direction as that which would cause this response.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANAL.YSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze the response of CVCS after determining the fail position of various
IIarris NRC Writlen Examination
Reactor Operator
Given the following conditions:
Following a plant trip, EPP-004, Reactor Trip Kcsponsc, is being performed.
- The crew is verifying Natural Circulation conditions as a result of a loss of power to
all KCPs.
- Five (5) core exit thermocouples are failed.
How do the failed core exit thermocouples affect indications used to veri@ Ndtnral
Circulation?
a. The Core Exit Temperature indications will be HIGHER than actual
- RCS Subcooling will indicate MORE subcooling than &ctual
b. The Core Exit Temperature indications will he HIGHER than actual
KCS Subcooling will indicate LESS subcooling than actual
c. Core Exit Tcnipcrature indications will indicate LOWER than actual
RCS Subcooling will indicate MOKE subcooling than actual
d. * Core Exit Temperature indications will indicate the SAME as actual
- RCS Suhcooling will indicate the SAME subcooling as actual
ANSWER:
d. Core Exit Temperature indications will indicate the SAME as actual
RCS Subcooling will indicate the SAME subcooling as actual
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 70 TIE.R/GROUP: 212
10CFR55 CONTENT: 41(b) 7 43&)
KA: 017K3.01
Knowledge of the effect that a loss or malfunction of the ITM system will have on the following: Natural
circulation indications
OBJECTIVE: ICCM-3.0-R6
DESCRIBE how the plant's subcooling monitor information is processed
DEVELOPMENT REFERENCES: SD-106 pg 5,14
ICCM-LP-3.0 pg 1 I , 14-15
REFERENCES SUPPLIED TO APPLICANT: Nom
QUESTION SOURCE: 0 X NEW SIGNIFICANTLY MODIFIED 0 DlRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTKACTOR JUSTIPICACTION (CORRECT ANSWER X'd):
a. Plausible since the thermocouples are failed, but a failed thermocouple indicates 50°F (low) and not
high.
b. Plausible since the thermocouples are failed, but a failed thermocouple indicates 50'F (low) and not
high.
c. Piausible since the failed thermocouples indicate 50°F (low), but the ICCM indication uses the highest
thermocouples and not the lowest.
X d. The failed thermocouples will not be used to process the indication by the ICCM, so there will be no
change. on core exit temperatures and subcooling margin.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS
U U
DIFFICIJLTY RATING: 3
EXPLANATION: Analyze the effect of failed thermocouples on temperatures and subcooling
margin
Harris NRC Written Examination
Reactor Operator
QUESTION: 71
Which of the following EOP network procedures may be directly entered and which
associated action is to be performed without direction from the Unit-SCO?
a. FW-S. I. Response to Nuclear Power Generation 1 A?VIS
0 Initiate emergency boration of the RCS
b. FKP-H.1, Response to Loss of Secondary Heat Sink
0 Attempt to start an AFW Pump
c. EPP-001, Loss of AC Pnwer to 1A-SA and IW-SB Ruses
Manually trip the turbine if still online
d. EPP-005, Natural Circulation Cooldown
Attempt to start an KCI
ANSWER
c. EPP-001. Loss of AC Power to 1A-SA and IN-SI3 Buses
Manually trip the turbine if still online
Harris NRC Written Examination
Reactor Operator
Data Sheets
QEESTION NUMBER 71 TIEWGROUP: 3
KA: 2.4.1
Knowledge of EOP entry conditions and immediate action steps
OBJECTIVE: 3.19-1
DESCXIRE Control Room usage of the EOP network as it relates to the following
Entry into EOP network
DEVELOPMENT REFERENCES: EOP-E.PP-001 pg 3
EOP Users Guide pg 13
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0X NEW 0 SIGNIFICANTLY M0DIFIE.D 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRF.CT: Xew
XRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since FRP-S. 1 contains immediate actions, but is entered only by direction in PATH-1.
h. Plausible since FRP-H.l is a high importance procedure, but is only entered when directed by other
proccdures.
X e. EPP-001 can be entered directly and contains immediate operator actions to manually trip the turbine.
d. Plausible since EPP-005 may be entered whenever a natural circulation cooldown i s required, but it
contains no immediate operator actions.
DIFFICULTY ANALYSIS:
COMPREIIENSI\E / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATIXG: 2
EXPLANATION: Knowledge of EOPs which can be entered directly
Harris NRC Written Examination
Reactor OpbTatoI
QUESTION: 72
Which of the following is a reason that containment pressure greater than 45 psig is
considered an extreme chaiienge to the containment critical safety function'?
a. Containment structural failure is imminent
b. Containment leakage could be in excess of design basis leakage
c. Hydrogen recombiner efficiency is significantly reduced at the pressure
d. Containment tenlperature is high enough to prevent adequate core cooling
ANSWER
b. Containment leakage could be in excess of design basis ieakage
Harris NRC Written fixamination
Reactor Operator
Data Sheets
QUESTION NUMBER: 72 TIEWGROUP: 211
1QCFR55CONTENT: 41(b) 10 43w
K4: 10362.4.6
Knowledge of symptom b a e d EOP mitigation strategies. (Containment)
OBJECTIVE: 3.13-4
Given the following EOP steps, nutcs, and cautions, DESCRIBE the associated basis
CSF decision points
DEVELOPMENT REFERENCES: 1.P-3.13 pg 7
REFERENCES SUPPLIED TO APPLICANE None
QtJESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DJRECT: 3.13-R4 001
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since this is above the postulated pressure following a large break LOCA or steam break, but
containment failure is not expected to occur until several times this value.
X b. 45 psig is above the pressure that design containment leakage rates assumed in off-site radiological
analysis.
e. Plausible since the resombincrs are located in containment and are conceivably affected by the high
pressure, but tlie 45 psig is selected based on exceeding design kakage rates.
d. Plausible since core cooling in the event of a LOCA is based upon transferring heat to the injection
water which is then collected in containment for recirc, but the. 45 psig is selected based on exceeding
design leakage rates.
DIFFICULTY ANALYSIS:
17 COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
E:WL..WATION: Knowledge of the basis for CSFST decision points for containment pressure
IIarris NRC Written Examination
Reactor Operator
VuEsrIm: 73
Assuming the plant is at 100% power steady-state conditions, which ofthe following
would require independent verification instead of concurrent verification'!
a. Kemclval of control power fuses for a clearance on RHR pump 1B-SB
h. Pcrfonnance of PIC portions of OW-Ipp due to the failure of PRZ pressure
transmitter PT-455
c. Installing a jumper in PIC-02 for a surveillance test
d. Lifting leads in Rod Control Power Cabinet 1BIl for troubleshooting
ANSWER:
a. Removal of control power fuses for a clearance on RHR pump IB-SB
Karrrs NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 73 TIEWIGROUP: 3
1QCFR55CONTENT: 41(b) 10 43(W
U: 2.1.13
Knowrledge of tagging and clearance procedures
OBJECTIVE: PP-3.11-7
DF.FINE concurrent verification and independent verification
DEVELOPMENT REFERENCES: OPS-NGGC-1303, pg 12.-13
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: Harris LOCT 635
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
X a. Concurrent verification is not needed on 480V breakers as they would have independent verification
since no adverse action would occur as a result of removing the fuses.
b. Plarrsibie since an OWP directs these actions, but concurrent verification is required since the incorrect
switch operation could result in an RPS or ESF actuation.
c. Plausible since a surveillance test directs these actions, but concurrent verification is required since the
incorrect switch ope.ration could result in an W S or ESF actuation.
d. Plausible since a work order directs these actions, but concurrent verification is required since the
incorrect switch operation could result in an RPS.
DIFFICULTY ANALYSIS:
COMPWX:IIENSIVE/ ANALYSIS KNOWLEDGE I RECALL
DIFFICIJLTY RATING: 3
EXPLANATION: Knowledge ofthe conditions when conc.urrent verification is not permitted
Harris NRC Written Examination
Reactor Operator
QUESTION: 74
Given the following conditions:
Following an accident, IIPP-015, Uncontrolled Depressurization of All Steam
Generators, is being pcrformed.
The operators have reduced AFW flow to all s t e m generators (SGf to minimum as
they continue attempts to isolate the SGs.
Which of the following describes the expected plant response to the AFW flow reduction
and what actions arc to be taken as SG pressures decrease?
a. IZCX hot leg temperatures will eventually begin to increase and the crew will then
transition to IPP-008, Safety Injection Termination
b. RCS hot leg temperatures will evenbilly begin to increase and the crew will then
increase AFW flow while continuing in EPP-015,TJncontrolled Depressurization
of All Steam Geiierators.
c. The SGs will eventually become completely depressurized and the crew will then
transition to EPP-014, Faulted Steam Generator Isolation.
d. The SGs will eventually become completely depressurized and the crew will then
transition to EPP-008, Safety Injection Temiriation.
ANSWER
b. RCS hot leg temperatures will eventually begin to incrcase and the crew will then
increase AFW flow while continuing in EPP-015, 1TncontrolIed Depressurization
of All S t e m Generators.
IIarris NRC Written Examination
Rtxxtor Operator
Data Shczts
QCESTION NUMBER 74 TIEWGROUP: 1/1
KAIMPORTANCE: Rc) 3.4 SRO
10CFR55 CONTENT: 41@) 7 43W
KA: WE12EK2.1
Knowledge of the interrelations between the (Uncontrolled Depressurization of all Steam Generators) and
the following: Components, and functions of control and safety systems, including instrumentation,
signals, interlocks, failure modes, and automatic and manual features
OBJECTIVE: 3.9-4
Given actions teaken in these emergency procedures, PREDICT the plant response to these actions
DEVELOPMEXT REFERENCES: EPP-015,p 8
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 X NE.W SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORIECT ANSWER X'd):
a. Plausible since hot leg temperatures will eventually increase, but the correct action is to stabilize
temperature by increasing AFW flow and adjusting ste'uning rate, if possible.
X h. As SC; pressure and steam flow decrease, RCS hot leg temperatures will eventualIy stabilize and may
increase. .4djusting feed flow and steam dump wiil control RCS hot leg temperatures.
e. Plausible since if no Sci can be isolated the SGs will completely depressurize and there is a foldout
page to transition to EPP-014 if SG pressure increases (will be stable when depressurized), and the
crew will eventually end np in GP-007.
'
d. Plausible since if no SG can be isolated the SGs will completely depressurize and RCS pressure will
increase due to SI flow so the operators would desire to terminate SI, but the crew will eventually end
up in GP-007.
DIFFICULTY ANALYSIS:
COMPREHENSIb'E / ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze SG response to decreasing pressure and redoccd AFW flow and
determine correct response
Harris NKC Written fixamination
Reactor Operttor
QUESTION: 75
The crew is implementing EPP-012, Loss of Emergency Coolant Recirculation. They
are now determining Containment Spray requirements with the following conditions:
Containnient pressure 12 psig
RWST level 3a?
Cuntainnieiit Fan Coolers running 3
Containment Spray Pumps running 2
Which ofthe following actions should bL:taken?
a. Start an additional Containment Fan Cooler
b. Secure both Containment Spray Pumps
c. Secure one Containment Spray Pump
d. Secure one Containment Fan Cooler
ANSWER.
b. Secure both Containnicnt Spray Pumps
IIarris NRC IVriit&nExamination
Reactor Operator
Data Sheets
QIJESTIQN NUMBER: 75 TIEWGROUP: 211
10CFR55 CONTENT: 41(b) 5 43w
KA: 026.42.08
Ability to (a) predict the impacts of the following malfunctions or operations on the C S S ; and (b) based
on those predictions, use procedures to correct, control, or mitigate the Consequences of those
malfunctions or operations: Safe securing of containment spray (when it c.m he done)
OBJECTIVE: 3.3-3
Given the title of an EOP foldout item, state the parameters to monitored for implemeutation.
DEVELOPMENT REFERENCES: EPP-012, p 3, 14
REFERENCES SUPPLIED TO APPLICAXT: None
QUESTIOX SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.3-n5004
NRC EX4M HISTORY: None
DISTRACTOR JLJSTIFICACTION(CORRECT ANSWER Xd):
a. Plausible since thc more Containment Fan Coolers that are running in EPP-Oi2, the fewer spray
pumps are required but no actions direct starting additional fans.
X b. With KWS?level below 3% all pumps taking a suction off the RWST must be secured, including the
Containment Spray Pumps.
c. Plausible since this action would be taken per EPP-012 if the RWST still had sufficient water, but
with the KWST empty all pumps must be stopped.
d. Plausible since ac,tion is taken to stop equipment that is used to remove heat from containment, but the
pumps are stopped, not the fans.
DIFFICULTY ASALYSIS:
COMPREHENSIVE /ANALYSIS Kh;OWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the conditions for securing containment spray