ML040790077
| ML040790077 | |
| 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: ML040790077 (202) | |
See also: IR 05000400/2004301
Text
HARRIS EXAM
50-400/2004-301
-
FEBRUARY 23 27,2004
& MARCH 4,2004 (WRITTEN)
WRITT
U.S.NucEear Regulatory Commission
Site-Specific
SRQ Written Examination
Applicant information
instructions
Use the answer sheets provided to document your answers. Staple this cover sheet OR top
of the answer sheets. To pass the examination you must achieve a final grade of at least
80.00 percent overall, with a 70.00 percent or better cn the SRO-only items if given in
conjunction with the RO exam; SRO-only exams given alone require an 80.00 percent to
pass. You have eight hours to complete the ccrnbined examination, and three hours if you
are only taking the SRO portion.
Applicant Certification
All work done on this examination is my own. I have neither given nor received aid
Applicant's Signature
Results
80 / SRO-Only I Total Examination Values Points
Applicant's Scores -.-.--l-...-l- Points
Applicant's Grade -1-1- Percent
QUESTION: 1
ature is being controlled by the Steam Dump
Given the follciwi it-SCO directs that Steam Dump
mode, what approximate setpoint is
Steani header press
E Turbine main s t a n pres
a. 37%
b. 42%
c. 58%
d. 63%
ANSWER:
b. 42%
Harris h T C Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 1 TIEWGROUP: 1/1
KAIMPORTANCE: HO 3.7 SRO
lOCFH55 CONTENT: 41(b) 7 4W)
KA: 000007EA1.10
Ability to operate and monitor the following as they apply to a reactor trip: S/G pressure
OR.IECTIVE: SIICS-3 .O-4
Explain how the steam dump valves are automatically modulated in the steam pressure control mode,
including control alignments, setpoint determination and adjustment, and the normal setpoint at power
DEVELOPMENT REFEREKCES: Steam Tables
OP-126 pg. 8
REFERENCES SUPPLIED TO APPLICANT: Steam o tables
QUESTION SOIJRCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: SDCS-R.1004
NRC EXAM IIISTOHY: None
DISTRACTOR .nJSTIFICACTION (CORRECT ANSWER X'd):
a. PLansible ifthe incorrect instrument is used to determine the range ofthe instrument ( 5 5 I / 1500).
X b. The equivalent steam pressure for the required RCS temperature is approximntely 551 psig. This
calculates to be a setpoint of 42% (551 / 1300).
c. Plausible if the correct instrument is used to determine the range ofthe instrument, but the calculation
is perfornied incorrectly (1300 - 551 / 1300).
d. Plausible if the incorrect instniment is used to determine the range of the instrument and the
calculation is performed incorrectly (1500 - 551 / 1500).
DIFFICULTY ANALYSIS:
COWPREHENSIVE / ANALYSIS KNOW1,EDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Miist determine required stearn pressure for RC'S temperature and then calculate
setpoint
Hank NKC Written Examination
Senior Reactor Operator
QIJESTION: 2
Given the following conditions:
e The plant is operating at 95% power during a power ramp.
e The Reactor Operator attempts to perform a normai dilution for temperature Control
in accordance with OP-107. Theinied and Volume Control System.
e lC;S-151, RnlW TO BORIC ACID BLENDER FCV-I 14H, fids to open.
Which ofthe following actions should be taken?
a. Continue in OP-107, Chemical and Volume Control System, and perform m
Alternate Dilution
b. Increase turbine load per GP-005, Power Operation, to adjust RCS temperature
c. Go to AOP-003, Maifhnction of Reactor Makeup Control, and perform an
Alternate Dilution
d. Go to AOP-003, Malfunction of Reactor Makeup Control, and perform a local
Manual Dilution
ANSWER
d. Go to AOP-003. Malfunction of Reactor Makeup Control, and perform il local
Manual Dilution
Hamis NRC WriUm Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 2 TIEWGROIJP: lii
10CFR55 CONTENT: 41(b) 10 43w
KA: 000022G2.4.4
Ability to rec.ognize abnormal indications for system operating parameters which are entry- level
conditions for emergency and abnormal operating procedures. (Loss of Reactor Coolant Makeup)
OBJECTIVE: AOP-3.3-R1
IDENTIFY symptoms that require entry into AOP-003, MaIhnction of Reactor Makeup Control
DEVELOPMENT REFERENCES: AOP-003, pg 12-13,25-26
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: AOP-3.3-RI 1
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible sinc.e alternate dilution is a viable method of diluting the RCS, but with ICs-151 failed
closed, alternate dilution will not function either.
h. Placisible since adjusting turbine load will result in a change in RCS temperature, but temperature is
low requiring dilution, and raising turbine load will further lower it.
c. Plausible since alternate dilution is a viable method of diiuting the RCS, but with ICs-IS 1 failed
c.losed, alternate dilution will not function either.
X d. With ICs-151 closed, the only option available to dilute is to perform a local manual dilution.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of the effect of a failure on the ability of the RMU system
Iimis NRC Wriritteti Examination
Senior Reactor Operator
QCESTION: 3
Given the foliowing conditions:
- The plant is operating at 5o?h power.
PT-457, Channel 111 Pressurizer Pressure, has ailed and ail associated bistabies are in
the tripped condition.
- Power is subsequently lost to I P S Bus IDP-1A-SI.
Which of the foIloi$ing describes the effect ofthis loss of power on the Phase A
Containment Isolation valves?
a. NO Phase A Containment Isolation valves Will close
b. OKLY Train A Phase A Containment Isolation valves wdi close
c. ONI,Y Train B Phase -4Containment Isolation valves will close
d. All Phase A Containment Isolation valves will close
ANSWER:
c. ONLY Train B Phase A Containment Isolation valves will close
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 3 TIERGROUP: 211
10CFR55 CONTENT: 41(b) 7 43w
KA: 013K3.03
Knowledge ofthe effect that a loss or maifiinction ofthe ESFAS will have on the following: Containment
OBJECTIVE:: ESFAS- 3.0-4
PREDICT how loss of any of the four instrument buses will affect the tSI.AS output functions of each
SSPS train
DEVELOPMENT REFERENCES: AOP-024 pg 22
SD-103 pg9, 11, I 3
REFERESCES SUPPLIED TO APPLICANT: Xone
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NCJMRER FOR SIGNIFICANTLY MODIFIED I DIRECT: ESFAS-3.044 00 I
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since Train SA stave relays will not actuate, but Train SR relays will still actuate,
b. Plausible since one train of Phase A wtilI not actuate, but the train that wilI not actuate: is Train SA.
X c. A loss of Bus IDP-1A-SI under these c.onditions will result in a 2!3 signal to hoth trains of ESFAS,
resulting in an SI and Phase A signal. Train SA slave relays. however, are powered from IDP-1A-SI
and are energized to actuate, so Train SA slaves will not perform their fitnction.
d. Plausible sinc.e 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 ASALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY KATING: 3
EXPLANATION: Analyze the effect of a loss of power on the actuation signals and determine
which power supplies power which output relays
Harris NKC Written Examination
Senior Reactor Operator
Given the following conditions:
The unit is operating at 30% power.
A dropped Control Dank 'C' rod has just been re-digned.
While attempting to operate the ROD CONTROL ALARM RESET, the operator
inadvertently operates the ROD CONTROL START-UP RESFl'.
Which of the following describes the effect of operating the incorrect reset?
a. .4ll Control Hank 'C' rods drop into the core, causing an automatic reactor trip
b. All rods, including Control Bank and Shutdown Bank rods, drop into the core,
causing an automatic reactor trip
c. All rods remain in their current position md 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 are fully inserted
AXSWER:
d. All rods remain in their current position, but the Rod Control System circuitry
indicates a11 rods are fully inserted
Harris NRC Written Ihmination
Senior Reactor Operator
Data Sheets
QXJESTION NUMBER: 4 TIEIUGROUP: 1 '2
10CFR55 CONTENT: 4l(b) 7 43W
KA: 000003AA1.02
Ability to operate and / or monitor the following as they apply to the Dropped Control Rod: Controls and
components necessary to recover rod
OBJECTIVE: RODCS-3 0-R7
DISCKSS the effects of manipulating each of the following rod control-related switches
e ROD CONTROL START-UP RESET switch
ROD CONIROL ALARM KESET switch
DEVELOPMENT REFERJZNCES: AOP-001, pg 1 1
KODCS-3.0, pg 69
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: RODCS-3.0-R400 1
NRC E.XAM HISTORY: None
DISTIt4CTOR .KSTJFICACTION (CORRECT ANSWER X'd):
a. Plausible since inrproper operation of correct switch could result in rods dropping into core, but
operated switdi only resets starting points for rod control circuitry.
b. Plausible since impr0pe.r operation of correct switch could result in rods dropping into core, but
operated switch only resets starting points for rod control circuitry.
c. Plausible if misconception that effect is nothing if performed at powe.r since switch is normally only
operated prior to withdrawing any rods, hut operated switch resets starting points for rod control
c.ircuitry.
X d. Operating swirch at power does not affect actual rod position, but resets rod control mch that circuitry
senses rods are at "full inserted" position.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS H KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge ofthe function of rod control system controls
Harris NRC Written Cwamination
Senior Reactor Operator
QUESTION: 5
Given the following conditions:
A large steam break has occurred inside Containment.
During the performance of PATH-I, the crew determined Containment pressure to be
18 psig and they verified proper operation ofthe Containment Spray System.
h transition has just been made to EPP-014, Faulted Steani Generator Isolation.
Containment pressure is now 22 psig.
Which of the following actions should he taken regarding the increase in Containment
pressure?
a. Continue to monitor Contiailment pressure and transition to FW-J. 1, Response to
High Containment Pressure, if it exceeds 45 psig
h. Continue to monitor Containment pressure and transition to FRP-J.1. Response to
High Containment Ircssure, if it remains above 10 psig 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 FRP-J.1. Response to High Containment Pressure, to allow
verification of proper operation ofthe Containment Fan Cooler fans
d. Transition to FRP-J.1, Response to High Containment Pressure, to ailow
verification of proper operation of the Emergency Service Water Booster Pumps
AXSWEK:
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
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER: 5 TIEWGROIJP: 1i 2
10CFR55 CORTENT: 41(b) 7 4309
KA: WE14EAl.2
Ability to operate and i or monitor the following as they apply to the (Iligh containment Pressure)
Operating hehavior characteristics ofthe facility
OBJECTIVE: 3.13-4
Given the following iOP steps, notes, and cautions, IXSCRIHII the associated basis
e CSF decision points
DEVELOPMENT REFERENCES: CSFS'I-Containment
FKP-J.l
HD-3.13-HO, pg 5-6
REFERENCES SUPPLIED TO APPLICANT. None
QUESTION SOIJRCE: 0X NEW SIGNIFICAXTLY MODIFIED 0 DIRECT
BANK NUAMBERFOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTKACTOR JUSTIFICACTIOS (CORRECT ANSWER X'd):
X a. Provided containment prcssure is between 10 and 45 psig and at least one spray pump has been
verified operating and providing flow, a transition is not required to FKP-J. I per the C'SFST as this is
only a ye.lIow path.
b. Plausible since two containment spray pumps should reduce containment presswe and the 1ine.r 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 pressurc, but a
transition would not he required unless containment pressure were to exceed 45 psig.
c. Plausible since the containment fan coolers assist the containment spray system in reducing
containment pressurc, bur 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 F,SW booster pumps are checked in FW-J.1 to ensure radiological releases are
minimized, 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.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Comprehension of the priority of actions to be taken regarding containment
pressure
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 6
Given the following conditions:
FRP-H.1, Response to B I.oss o Secon ry Heat Sink. is being implementel
KCS bleed and feed has been initiated when Auxiliary Feedwater (AFW) q d b i l i t y is
restored.
- All SGs are completely dry and depressurized.
Which of the following describes the strategy used to re-establish feed under these
conditions?
a. Feed ONLY one (1) SG to ensure RCS cooldown rates are established within
Technic.al Specification h i t s
b. Feed ONLY one (1) SG to limit the possibility of a SG tube rupture to a single SG
c. Feed ALL SGs to establish subcooling conditions in the RCS as soon as possible
d. Feed ALL SGs to allow termination ofRC.S bleed and feed as soon as possible
ANSWER:
b. Feed ONLY one (1) SC; to limit the possibility of a SG tube rupture to a single SC;
IIarris NRC Written Ekaniination
Senior Reactor Operator
Date Sheets
QUESTKIN NUMBER: 6 TIEWGROUP: 1/1
lOCFR55 CONTENT: 41(b) 8/10 43(h)
KA: 000054AK1.02
Knowledge of the operational implications ofthe following concepts as they apply to Loss of Main
Feedwater (MFU): Effect5 of feedwater introduction on dry S/Ci
OBJECTIVE: 3.11-4
Given the following FOF steps, notes, and cautions, DESCRIRF the associated basis
e Fced restoration
DEVELOPMENT REFERENCES: FKP-11.1, pg 47
LP-3.11, pg 12
REFERENCES SUPPLIED TO APPLIC.4NT: None
QUESTION SOURCE: 0X NEW 0 SIGNIFICANTLY MODIF1E.D 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Xew
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible 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 he established to one dry SG so that if excess thermaI shock causes failure, the
failure is limited to one SG.
e. Plausible since RCS subcooling is a desirable condition to achieve, hut only one S G at a time is fed.
d. Plausible since terminating RCS bleed and feed is a desirable condition to achieve, but only one SG at
a time is fed.
DIPPICULTY ANALYSIS:
COMPKEWENSIVE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Know,ledge of the requirements for feeding a dry SG and the reasons for these
actions
Harris NKC Written Examination
Senior Keactor Operator
QUESTION: 7
Given the followhg conditions:
The unit is at 100% power.
The running CSIP trips at 0930.
AOP-018, Reactor Coolant Pump Abnormal Conditions, actions have been
completed and the standby CSEP is started at 0933.
Which of the following actions should be taken to establish seal cooling to the RCPs in
accordance with AOP-018?
a. Adjust HC-180.1, KCP SEAL WTR INJ FLOW, to establish 8 to 13 gpm seal
injection flow
b. Adjust HC-186.1. RCP SEAL WTRINJ PLOW. to establish a 1°F per minute
cooldown rate of the seals until 8 to 13 gpm seal injection llow is established
C. J.,ocally adjust 1CS-340 / 381 / 422, KCP A i 3 C SEAL INJ MAN1JAL ISOI., to
establish 8 to 13 gpm seal injection flow
a. Locally adjust 1CS-340 / 381 i 422, RCP A / B / C SEAL INJ MANUAL ISOI,, to
establish a 1F per minute cooldown rate ofthe seals until 8 to 13 gpm seal
injection flow is established
ANSWER:
a. Adjust IIC-I 86.1, RCP SEAL WTR TNJ FLOW, to establish 8 to 13 gpm seal
injection flow
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER 7 TIEWGROUP: 2i 1
1OCFR.55 CONTE.NT: 41(h) 7 43(W
KA: 003A4.01
Ability to manually operate andior monitor in the control room: Seal injection
OBJECTIVE: AOP-3.18-3
Given a set of plant conditions and a copy of 4OP-018, DE'I'I<RM"E the appropriate response
DEVELOPMENT REFERENCES: AOP-OIS, P 38
REFERENCES SIJPPLIED TO APPLICANT: None
QUESTION SOURCE: XEW SIGNIFICANTLY MODIFIED DIKECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Ikrris 1,CPCT BO4 073
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. With seal injection flow lost for less than 5 minutes, seal injection can be established by adjusting IIC-
186.1 without concern for damage to the seals.
h. Plausible since this action would be taken ifseal 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. Plausible since these actions would be taken ifthe cause ofthe loss of seal injection flow was other
than a tripped CSIP and the flow was lost for less than 5 minutes, but with the loss ofthe CSIP as the
cause, IIC-186.1is used.
d. Plausible since these actions would be taken if the eause of the loss of seal injection flow was other
than a tripped CSIP and the flow<was lost for more than 5 minutes, but with the loss of the CSIP as the
cause, NC- 186.1 is used.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFPICC1,TY RATING: 3
EXPLANATION: Comprehension of the effect of a short term loss of seal injec?ion flow to the
RCPS
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 8
EPP-008, "SI Termination," directs resetting SI.
Which ofthe following describes thc effect ofoperating only ONE (1) ofthe two (2) SI
RESET switches at this step instead of both?
a. e Bypass - Permissive 1,ight Panel light 4-1, SI ACTUATE, would blink due to
only one train of SSPS having an SI signal
e Bypass -Permissive Light Panel light 5-1, SI RESET - AIJTO SI BI,OCKED,
would blink due to only one train of SSPS having SI reset
b. e Bypass - Permissive Light Panel light 4-1, SI ACTUATE, would extinguish
due to neither train of SSPS having an SI signal
e Bypass - Permissive Light PmeI light 5-1, SI RESE'I- AUTO SI BLOCKED,
would light due to both trains of SSPS having SI reset
c. * Bypass - Permisskc K,ight Panel light 4-1, SI ACITJATE. would blink due to
only one train of SSPS having an SI signal
Bypass -Permissive Light Panel light 5-1, SI RESET - AIJTO SI BLOCKED,
would Iight due to both trains of SSPS having auto SI blocked
d. Bypass - Permissive Light Panel light 4-1. SI ACTUATE, would extinguish
due to neither train of SSPS having an SI signal
Bypass - Pemiissive 1,ight Panel light 3-1, SI RESET -AUTO SI BLOCKEII.
would light due to both trains ofSSPS having auto SI blocked
ANSWER
a. * Bypass - Permissive Light Panel light 4-1, SI ACTUAIE, would blink due to
only one train of SSPS having an SI signal
Bypass - Permissive Light Panel light 5-1, SI RESET - AUTO SI BIXXKED,
woukl blink due to only one train of SSPS having SI reset
Harris NKC Written Examinaiion
Senior Reactor Operator
Data Sheets
QIJESTION NUMBER: 8 TIEWGROUP: 2/1
K4 IMPORTANCE: KO 3.9 SRO
10CFR55 CONTENT: 41(b) 7 43(W
KA: 006K4.11
Knowledge of EC'CS design feature($) and/or interlock(s) which provide for the following: Kcset of SIS
OBJJWTIVE: SIS-3.0-R4
DETERMINE SIS status from the following
Bypass-Permissive Light Box
DEVELOPMENT REFERENCES: SD-103 pg 13
Functional Diagrams Safeguard
Actuation Signals Sheet 8
EOI'I 4-2 1 H ~ ~ d o u t
SOEK 94-1 Related Industry Events
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIREST
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: INPO 1073
NRC EXAM HISTORY: None
DISTR4CTOR JUSTIFICACTION (CORRECT ANSWER X'd):
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 Actuation switch only requires a single switch to actuate SI, but the reset
switches are train-related.
c. Plausible since only train of SI would he reset so window 4-1 would he responding correc.tiy, but
window 5- 1 would also bc blinking due to the disparity hetween trains.
d. PLausible since the SI Ac.tuation switch only requires a single switch to actuate SI, hut the reset
switches arc train-related.
DIFFICULTY ANALYSIS:
COMPKEIIENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICIJLTY RATING: 3
EXPLAXATION: Comprehend the effect of only operating a single train w i t c h on SSPS and how
the indications would be affected
IIarris NRC Written Exuniination
Senior Reactor Operator
QUESTION: 9
Given the following conditions:
a lhe unit is at 100% power.
-
Power has been lost to IDP-1A-SIII, instrument Bus 111, and actions are being taken
in accordance with AOP-024, 1,oss of Unintemptible Power Supply.
PT-953, Containment Pressure Channel 11. then fails high.
Which ofthe following describes the effect on the Safety Injection (SI) and Containment
Spray Actuation Signal (CSAS) systems?
a. Neither an SI nor a CSAS would occur
b. An SI would occur: a CSAS would NOT occur
c. An SI would NOT occur; a CXAS would occur
d. Both an SI and a CSAS would occur
ANSWER:
b. An SI would occur; il CSAS would NOT occur
Harris NRC Written Examination
Senior Reactor Operator
I h t a Sheets
(ICESTION NIJMBER: 9 TIERKROIJP: 2; 1
lOCFR55 CONTE.NT: 41(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
detectors
OBXECTIVE: CSS-Rl
Given a set of plant c.onditions or the status of each bistable light box, DETERMINE which of tire
following E.SFAS signals are active
Safety Injection (SI)
0 Containment Spray Actuation
DEVELOPMENT REFERENCES: SD-103. pp 11,64,68
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGMFICANTLY MODIFIED /DIRECT: IIarris IAXT 139
NRC E.XAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible sinc.e CSAS is energized to actuate and 1 channel is in a deenergized condition so CSAS will
not oc.cur, but the 2 failed channels will cause an SI actuation.
X b. An SI actuation (detnergized 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 ofthe two signals is energized to actuate and the other is deenergized to ac.tuate,
but SI is deenergize to actuate and CSAS is energized to actuate.
d. Plausible since the 2 failed channels will c.ause an SI actuation, but CSAS is energized to xtuate and 1
channel is in a deenergized c.ondition so CSAS will not oc.cur.
DIFFICULTY ANALYSIS:
COMPRE.HENSIVEI ANALYSIS KNOWLEDGE I RECALL
IIIFFICIJLTY RATING: 3
EXPLANATION: Comprehension of the effect of multiple failed channels on ESFAS signals
Harris NRC Written Examination
Scnior Reactor Operator
QIJESTION: 10
Given the following conditions:
The plant is operating at 43% power.
- 120VAC Vital nus IDP-1B-SI1 denergizes.
Outward rod motion is inhibited by ...
a. C-4, OPA'T rod stop.
h. C-3>OTAT rod stop.
c. C.-2, Power Range rod stop.
d. C-1,Intermediate Range rod stop.
ANSWER
c. C-2, Power Range rod stop.
IIarris NKC Written Examination
Senior Reactor Operator
Data Shtxts
QUESTION NUMBER: 10 TIEWGROUP: 2!2
iOCFR55 CONTENT: 41(b) 7 43(W
KA: 00LK4.07
Knowledge of CRDS design fcature(s) and/or interloc.k(s) which provide for the foliowing: Rod stops
OBJECTIVE: XIS-3.0-9
LXSCXJSS the operation of the following NI trip-related func.tions:
b. SR, IK and PK (low) trip blocks
DEVELOPMENT REFERENCES: OI-lO5 pg 26
AOP-024 pg 6
REFERENCES SUPPLIED TO APPLICANT: None
QrJESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: NIS-R6 003
NHC E.XAMHISTORY None
DISTRACTOR .KSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since causes rod stop, but coincidence is Y4 instead of 1/4
b. Plausibie since causes rod stop, hut coincidence is 214 instead of 1/4
X c. PR rod stop is 1/4 coincidence. With S2-SB deenergized, ER N-42 is tripped.
d. Plausible since this causes a rod stop, and coincidence is 1 2 , but IR rod stop is blocked above P-10 by inanual
operator ac.tion. Must have 214 IR be.low, P-10 to reset.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS 17 KNQWL.EDGEI RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze effect of loss of power on NIS and rod control and determine effect of
single channel tripped
Harris NRC Written fixamination
Senior Reactor Opentor
QUE.STION: 11
?he basis for the operation ofthe Electric IIydrogrn Recombiners is to minimize
hydrogen concentration build up in Containment following a LOCA due to the , ..
a. zirc-water reaction and release of hydrogen from the PRT.
b. corrosion of metals in Containment and release of hydrogen from the RCDT.
c. release of hydrogen from the PKI and the riidiclytic decomposition of water.
d. radiolytic deconiposition of watcr and the corrosion of metals in Containment.
?INSWER:
d. radiolytic decomposition of water and the corrosion of metals in Containment.
Harris NRC Written Examination
Senior Reactor Operator
nata Sheets
QUESTION NUMBER. 11 TIEWGROW: 212
10CFRSS CONTENT: 41(b) None 43(b) 2
KA: 02862.2.22
Knowledge of limiting conditions for operations and safety limits. (IIydrogen Recombiner and Purge
Control)
OBJECTIVE: HK-3 .O- 1
STATE the purpose and function of the Hydrogen Rec.ornbiner System, inc.luding the following
components:
Electric hydrogen recomhiner
DEVELOPMENT REFERENCES: 3.6.4.2 Basis
SD-12s pg 21
LP-IIR-3.0 pg S
REFERENCES SUPPLI
QUE.STI0N SOURCIE: SIGNIFICANTLY MODIFIED H DIRECT
BANK NTJMB DIFIED / DIRECT: HR 01
NRC EX4M HISTORY Xone
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X7d):
a. Plausible since E.lectric Hydrogen Recombiners are designed to remove hydrogen in containment
following a I,OCA due to generation from the zirc-water reaction, hut not due to release from the
PRI'.
b. Plausible since Electric Hydrogen Recombiners are designed to remove hydrogen in containment
following a LOCA due to generation from the corrosion of metals in containment, but not due to
release from the RCDI.
E. Plausible since Electric Hydrogen Recombiners are designed to remove hydrogen in containment
following a 1,OC.Adue to generation fiom the radiolylic decomposition of Water, hut not due to
release from the I'KT.
X d. The Electric IIydrogen Recnmhiners are designed to rcmove hydrogen in containment following a
LOCA due to generation from the zirc-water reaction. radiolytic decomposition of water, and
corrosion of metals in containment.
ICtJLTU ANALYSIS:
COMPRFXENSIVE I ANALYSIS H KNOWI,EDGE /RECALL
DIFFICIJLTY RATING: 3
EXPLANATION: Knowledge of Tech Spec basis for hydrogen rcc.otnbiners
Harris NRC Written Earnination
Senior Reactor operator
QUESTION: 12
EPP-001. 1,oss of AC Power to 1A-SA md IB-SB Buses, is being performed.
Concurrent to the loss of power. a small break LOG\ occurred.
The crew has completed the following actions when off-site power is restored to 6.9 KV
Bus 14-SA:
0 Sequencers haw been de-energized
0 Safeguards pumps autostarts have been disabled
0 KCP seals have been isolated
0 Depressurization 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 ESW pump
h. Start a CSIP
c. Stabilize SG pressures
d. Initiate SI
ANSWER
c. Stabiliz SC;pressures
Harris NKC Written Examination
Senior Ksactor Operator
Data Sheets
QUESTION NUMBER: 12 TIERKROUP: 1!1
10CFR55 CONTENT: 4I(h) 7 43(W
KA: 000055EA1 .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, DETBKMINE if
implementation is reqiiired
UEVE1,OPNIENT REFERENCES: EPP-001 pg 3 5 , 3 S
REFERENCES SUPPLIED TO APPLICkYT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR .TUSTIIIICACTION(CORRECT ANSWER Xd):
a. Plansible since if the power source was an F,DG instead of offsite power, it would be important to
provide cooling flow to the E.DG.
b. Plausible since a small break I,OC4 exists and RCS inventory is being lost, but the first action is to
stabilize SG pressure.
X E. Upon restoration of power to at least one bus, the first action taken is to stabilize SG pressures
d. Plausible since a small break LOCA exists and RCS inventory is being lost, but the first action is to
stabilize SCi pressure.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / AKALYSIS KN0M7LEDGE/ RECALL
DIFFIC3JLTY RATING: 3
EXPLANATION Knowkdpe of required actions when power is re.stored following a loss of all
AC power
Harris NRC Written Examination
Senior Reactor Operator
QIJESTION: I3
While performing an Operating Procedure, the Reactor Operator conies to a step which
states:
Requcst Cheinistry to sample the RI1? 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 o f the Operating Procedure.
Which ofthe following is the MINIMUM requiremeirt(s) that must be met to allow
marking the step NlK?
a. e Step must be initialed by the Reactor Operator prior to performance
b. * Step must be initiaied by the Reactor Operator prior to performance
A written explanation of why the step is N/A must he provided in the
Coniinents section of the procedure
e. * Step must bc initialed by the S C O prior to performance
d. Step must be initialed by the SCO prior to performance
- A written explanation of why the step is N/A must be provided in the
Comments section ofthe procedure
ANSWER:
d. 0 Step must be initialed by the SCO prior to perforinanee
e A written explanation of why the step is N/A must be provided in the
Comments section of the procedure
IIarris NRC Written f:xamination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 13 TIEWGROUP: 3
IOCFR55 CONTENT: 41(h) None 43(h) Xone
KA: 2.1.23
Ability to perform specific system and integrated plant procedures during all mode.s of plant operation
OB.JECTIVE: PP-2.0-2
IZISCXJSS the requirements in PRO-NGGC-0200 concerning the following:
Procedure user's responsibilities
DEVE.LOPMENTREFERENCES: PRO-h'GGC-0200 pg 1 1- 12
REFERESCES SUPPLIED TO APPLICANT: Sone
QUESTION SOURCE: X NEW SIGNIFICANTLY MODIFIED DIRECT
RANK NUhXl3E.RFOR SIGNIFICANTLY MODIFIED / I9IKECT: New
NRC EXAM HISTORE Xonc
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER X'd):
L Plausible since the KO discovered the cause for marking tbe 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 since a written expianation must be provided in the Comments section, but a supenrisor must
initial the step prior to performance.
e. Plausible sinc.e a supervisor must initial the step prior to performance, but a written explanation must
be provided in the Comments section.
X d. The. step is initialed by the responsible supervisor prior to performance and a written expianation is
provided i n the Comments section.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANAI.YSIS KNOWLEDGE / RECALL
DIFFICCLTY RATING: 2
EXPLANATION: Knowledge of use of Ki'A during procedure usage
Harris NRC Written Examination
Senior Reactor Operator
QIJESTION: 14
A new Progress Energy employee was working at another nuclear utility for the first six
(6) months ofthis year. H i s occupational total effective dose equivalent (TEDE) at the
other utility has been documented as being SO0 mRem for this year.
\%at is maximum addirional 'I'EDE that he can 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
h. 2000mRem
c. 3500mRem
ANSWER:
h. 2000mKem
Harris NRC Written Examination
Senior Reactor Operator
Llata Shcets
QUESTION NUMBER: 14 TZFWGROIJP: 3
10CFR55 CONTENT: 41(b) 12 4300
KA: 2.3.2
Knowledge of facility A L A M program
OBJECHVE: RP-3.5- 14
State the 1OCFK20 and corporate occupational dose limits for individuals
DEVELOPMENT REFERENCES: XM-IM-002, pg 11
REFERENCES SUPPLIED TO APPLICANT: Sone
QUESTION SOCRCX: NEW SIGNIFICANTLY MODIFIED 17 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: lP-3.4-Ri 002
NRC EXAM HISTORY: None
DISTRACTOR .IUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the annunl Progress Energy dose limit is 2 Rem and he has already received 500
mRem this year, hut occupational dose from another utility is not considered in the 2 Rem limitation
unless he would exceed 4 Rem cornhitied for the 2 utilities.
X b. Personnel annual Progress Energy T H E shall not exceed 2 Rem arid 4 Rem total dose if non-
Progress E.nergy occupational dose for the current year is determined.
c. Plausible since he is permitted to receive a total of 4 Rem hehveen the 2 utilities and he already has
500 mRem, hut the more limiting is the 2 Rem Progress Energy dose.
d. Ilausihle since 500 mRem and 4500 mRem would equal the employees legal limit of 5000 rnRem,
hut this is greater than the administrative limit of2000 mRem.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY R4TING: 2
EXPLANATION: Knowledge ofadministratiue dose limits
Harris KRC Written Iixamination
Senior Reactor Operator
QUESTION: 15
Given the following conditions:
A small hreak I,OCA has occurred.
Containment pressure is 3.8 psig and increasing.
a Containment temperature is 137 OF and increasing.
The expected Containnient Cooling Fan alignment will he one (1) f3n in each
Containment Fan Cooler L J r ~ running
t in ...
a. high speed with the post-iiccident dampers shut.
h. 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 dampers open.
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QIJESTION NUMBER: 15 TIEWGROUP: 2/1
KAIMPORTANCE: HO 3.2 SRO
iOCFR55 CONTENT: 41(b) 7 43m
KA: 022G2.1.28
Knowledge of the pnrpose and function of major system components and controls. (Containment
Cooling)
OBJECTIVE: CCS-3.0-R2
I'RE.DIC'1 the response(s) of the Containment Cooling Subsystems to the following signals.
8 SI
DEVE1,OPMENT REFERENCES: SD-169, p 14
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW' SIGNIFICANTLY MODIFIED niREcr
BANK NUMBER FOH SIGNIFICANTLY MODIFIED /DIRECT: CCS-R4 001
NRC EX4B.I HISTORY: None
I)ISTR4CTOH JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since this alignment is an alignment that would be used following a loss of offsite power, but
the SI alignment has the fans in low speed.
h. Plausible since this alignment is an alignment that would be used following a loss of offsite pow'er
with the dampers aligned for the SI alignment, but the SI alignment has the fans in low speed.
c. Plausible since the fans are aligned per the SI alignment, but the dampers are aligned per the loss of
offsitc power alignment.
X d. Following an SI actuation, rhe containment fan coolers shift to low speed and the post-accident
dampers open.
DIFFICULTY ANALYSIS:
0 COMPREIIENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the response of Containment Cooling to an SI signal
Harris NRC Written Examination
Senior Reactor Operator
QIJESTION: I6
lollowing a Reactor Trip and Safety Injection due to a RCS leak. the Critical Safety
Function Status Trees (CSFST) are being monitored.
When monitoring the CSFST for KCS Inventory, ifPK% level is indicating greater than
929A0,why is a check of R V I X then perfomied?
a. Determine if the cause ofthe high PRZ 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 caused erroneous indications
of the 1RZ ievel instruments
d. Determine ifthe cause of the high PRZ level is excessive RCS inventory or
expansion due to an RCS heatup
ANSWER
a. Determine if the cause of the high E level is excessive KCS inventory or
voiding in the Reactor VesseI head
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 16 TIEWGROUP: 1!1
10CFR55 CONTENT: 41@) 7 Wb)
KA: 00000SG2.1.28
Knowledge of the purpose and function of major system components and controls. (Pressurizer Vapor
Space Accident)
OBJECTIVE: IC.:CM-3.0-1
1,ISI the two major functions 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 AlTLICANT: None
QIJESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED i D m c r : 3.12 001
NHC EXAM HISTORY: None
DISTRACTOR JUSTIPICACTIION (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 of the full PKZ is excessive inventory or voiding in the head region.
b. Plausihle since RVLIS is used throughout the EOF network to determine if SI termination criteria has
been met, hiit in this insrance it is used to determine the cause ofthe high PRZ le.vel.
e. Plausible since a steam space break in the PKZ will affect the level indications, hut RVLIS is used to
determine the cause of the PRZ high level condition.
d. Plausible since RVLIS is part of the Inadequate Core Cooling Monitoring System and a heat up of the
RCS will cause expansion ofthe KCS, hut hut RVLIS is used to determine the cause of the PRZ high
level condition.
DIFFICULTY ANALYSIS:
[7 COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICITLTY RATING: Knowledge of the purpose of monitoring RVI.IS during accident conditions
EXPLANATION:
Ilarris NKC Written Examination
Senior Reactor Operator
QUESTION: 17
Given the following conditions:
E The plant is shutdown for work on Reactor Coolant Pump seals
E 1he Reactor Vessel IIead is still inslalled.
E Iherunning Residual IIeat Removal (RIIR) pump trips and the crew is unable to start
the standby KHR pump.
Time to reach core boiling is determined to be 26 minutes.
Time to reach core boil-off is determined to be 53 minutes.
Ofthe following two (2) methods of RCS makeup, in accordance with AOP-020, L.oss
of RCS Inventory or Residual Heat Removal While Shutdown, which of the following 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 wouid violate Technical Specifications
b. Gravity feed from the KWST to Ihe 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 CSIP is preferred over gravity feed from the RWST since gravity feed
flow may be insufficient to makeup for core boil-off even if the RCS is
depressurized
d. Starting a CSIF is preferred over gravity feed from the RWST since the RCS may
be pressurir;ed and prohibit gravity flow
ANSWER
d. Starting a CSIP is preferred over gravity feed from the RWST since the KCS nmy
be pressurized and prohibit gravity flow
Harris NKC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 17 TIEWGROUP: lil
10CFR55 CONTENT: 41(b) 8/10 43(b)
KA: 000025AK3.01
Knowledge ofthe reasons for the following responses as they apply to the Loss of Residual Heat
Kemoval System: Shift to alternate flowpath
OBJECTIVE: AOP-3.20-1
Given B set of entry conditions and a copy of AOP-020, DETERMINE the appropriate response
DEVELOPiVENT REFERENCES: AOP-020, pg 9
AOP-020-BD, pg 19
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
1)ISTIIACTOK nJSTIFICXCTION (CORRECT ANSWER X'd):
a. Plausible since TS requires that a CSIP be made inoperahle before these plant conditions are
established, hut ( 2 - 0 0 8 requires that at least one CSIP be functional under these conditions.
b. Plausible since the CSIP can provide more flow than Reactor Makeup is capable of providing, hut thr
suction source. for the CSIP would be the RWST.
c. Plausible since starting a CSW is preferred to gravity feed, but only because the KCS may be
pressurized, I f the RCS is depressurized, gravity feed will provide adequate flow.
X d. If the RCS is pressurized, gravity flow may bc insufficient to provide adequate makeup to the RCS.
DIFFICULTY ASALYSIS:
COMPKEHENSIVE /ANALYSIS KITOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of plant conditions to determine appropriate response and reason for
response
Harris NRC Written Esnniination
Senior Reactor Operator
QUESTION: 18
Given the following conditions:
e Containment temperature is 96 "F.
C'cntainment Fan Coolers (AH- 1 / 2 3 / 4) arc operating in the Normal Cooling
Mode.
A loss of offsite power occurs and the plant 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 spced and discharging to the post-accident discharge duct
c. iow speed and discharging to the concrete ail-shaft
d. low speed and discharging to the post-accident discharge duct
ANSWER:
a. high speed and discharging to the concrete airshaft
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: I8 TIEWGROUP: 1!1
10CFR55 CONTENT: 41(b) None 43(b) 5
KA: 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: CCS-3 .O-R4
PKEDICI' the response(s) of the Containment Cooling Subsystems to the following signals.
L.0SP
DEVELOPMENT REFERENCES: SD- 169 pg 14
REFERENCES SUPPLIED TO APPL,ICANT: None
QUESTION SOGRCF.: 0 NEW SIGNIFICANTLY n m m m DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: CC.SR4 00 1
NRC EXAM HISTORY: None
DISIRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
x a. One fan per unit will start on high speed and discharge to the concrete airshaft.
b. 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.
E. Plausible sinc.e this fan response is the. response to a LOCA start signal and they do discharge to the
concrete airshaft, hut the fans operate in high speed following a loss of offsite power.
d. Plausible since this is the responsc to a LOCA start signal, but the fans operate in high speed and they
discharge to the concrete airshaft following a loss of offsite power.
DIFFICTJI,TYANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the response of the c.ontainment fan cooler fans to a loss of
offsite power
Harris NRC Written Examination
Senior Keactor Operator
QUESIION: 19
Given the following conditions:
The crew has determined that control rod F-10 in Control Bark D is misaligned by 18
steps.
Actions are being performed in accurdancc with AOP-001, Malfunction of Rod
Control and Indication System.
?he crew will attempt to align control rod F-10 and the remaining rods in Chntroi Rank U
by placing the Rod Selector Switch to . . .
a. BANK D and opening the lift coil disconnect switches for the remaining rods in
Controi Bank U.
b. MANUAL and upening the lift coil disconnect switches for the remaining rods in
Control Hank D.
c. BANK D and opening the lift coil disconnect switch for control rod F-10
d. MANUAL and opening the lift coil disconnect switch for control rod E-IO.
ANSWER:
a. BANK 11and opening the lift coil disconnect switches fur the remaining rods in
Control Bank D.
Hamis NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 19 TIEWGROUP: 1I2
10CFR55 CONTENT: 41(b) 7 43@B
Icn: 000005AK2.02
Knowledge of the interrelations between the Inoperable Stuck Control Rod and the following: Breakers,
relays, disconnects, and control room switches
OBJECTIVE: AOP-3.1-6
Given a set of plant conditions and a copy of AOP-001, DETERMINE the appropriate response
DEVE1,OPMENT KEFERFNCES: AOP-001 pg 17-1 8
KEFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED
OR SIGNIFICANTLY MODIFIED / DIRECT:
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION(CORRECT ANSWER Xd):
X a. The affected individual bank position should be selected and the inoperable rod will be attempted to
be moved by opening the liA coil disconne.ct switches for the remaining rods in the bank.
b. Plausible since the inoperahlr rod will be attempted to be moved by opening the lift coil disconnect
switches for the remaining rods in the bank, but the affeded individual bank position should he
selected.
c. Plausible since the affected individual bank position should be selected, but the inoperable rod will he
attempted tO bc moved by opening the lift coil disconnect switches for the remaining rods in the bank.
d. Plausible since the inoperable rod is in Bank D, but movement should be attempted by using the
individual bank select position.
ICULTY ANALYSIS:
COMPREHENSIVE / .ANALYSIS KNOWI.EDGE / HECAI,I,
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the means for a misaligned rod per procedure
IImis NRC Written Examination
Senior Reactor Operator
QUESTION: 20
Given the following conditions:
EKFIS is inoperable.
- Plait parameters are as follows:
- ICCM highest TC = 672' F
- IZCS N7Rtemperature (highest) = 688" F
- RCS pressure PT-440 - 1535 psig
- RCS pressure 1"'-402 = 1635 psig
- CNMT pressure Pl-95 1 = 4.5 psig
What value of superheat should be reported?
a. 63 O F
b. 71 "F
c. 79°F
d. 87OF
ANSWER
a. 63'F
Hmis NRC Written Exmiinatinn
Senior Reactor operator
Data Sheets
QUESTION NKJMBER: 20 TIEWGROUP: 1/2
KAIMPQRTANCE: RQ 4.6 SRQ
10CFR55 CONTENT: 41(b) Noue 43(h) 5
K4: 000074EA2.01
Ability to determine or interpret the following as they apply to a Inadequate Core Cooling: Subcooling
margin
OBJECTIVE: 3.19-4
Given a set of conditions during EOP implementation, DEI'EKMINE the correct response or required
action based upon the EOP User's Guide general information
Determining an KCS subcooling value
DE:VE.LOPMEKT REFERENCES: Users Guide, pg 27,34-35
REFEREXCES SIJPPLIED TO APPLICANT: Steam Tables
QUESTION SOURCE: [7 NEW SIGNIFICANTLY MODIFIED [7 DIRECT
BANK NVMBER FOR SIGNIFICANTLY MODIF1E.D/ DIRECT: 3.19-R3 003
NRC EXAM HISTORR Kone
DISTRACTOR JUSTIF'ICACTION (CORRECT ANSWER X'd):
X a. When ERI'IS is not available, the highest ICCM temperature should be used. If ERI'IS is not
available and adverse containment conditions exist, PT-402 should be used for pressure. Saturation
temperature for 1635 psig is 609 "F, so the amount of superheat is 63 "F (672-609).
b. Plausible since tire superheat determined using the ICCM temperature and saturation for the lowest
KCS pressure of IS35 psig (not used because of adverse containment conditions) is 71 "F (672-601).
c. Plausible since the superheat determined using the hot leg temperature (not used if ICChl is available)
and saturation for the P1'-401pressure of I635 psig i s 79 "F (688-609).
d. Plausible since the superheat determined using the hot leg temperature (not used if1CC:M is available)
and saturation for the Lowest RCS pressure of 1535 psig (not used because of adverse containment
conditions) is 87 OF (688-601).
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / WXALL
DIFFICULTY RATING: 3
EXPLASATION: Knowledge of instruments to use and calc,ulation of subcooling by applying
steam tahles
Harris NKC Written Examination
Senior Reactor Operator
QCESTION: 21
A failure o f a Containment Fan Cooler Unit, while the system was aligned to maximum
cooling mode. causes equilibrium Containment ternperiiture to increase froin 119 "I; to
126 "F.
IIow does Pressurizer level indication change due to this increase in Containment
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
c. 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. Ixvel indicates higher than actual due to reference leg density decreasing
Harris NRC Written Examination
Senior Reactor Operator
DaPu Sheets
QUESTION NUMBER: 2 1 TIER/GROW: 2: 1
10CFR55 CONTENT: 41(b) 7 43W
KA: 022K33.02
Knowledge ofthe effect that a loss or malfunction ofthe CCS will have on the following: Containment
instrumentation readings
OBJECTIVE: PZRLC-3 .O-4
IIESCRIBE how various errors would affect the pressurizer level indication in the Main Control Room
DE.VELOPMENTREFERENCES: LP-P%RI,C-3.0 pg 10
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECX
BARK NUMBEK FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM IIISTORY Kone
DISTRACTOH JUSTIFICACTION (CORRECT ANSWER Xd):
X a. Reference leg density decreases as containment temperature increases which causes level to indicate
higher than ac.tua1.
b. Plausible since reference leg density changes as containment temperature increases which causes level
to indicate different than actual.
c. Plausible since reference 1e.g density changes as containment temperature increases which causes level
to indic.ate different than actual.
d. Plausible since reference leg density changes as containment temperature increases which causes level
to indicate. different than actual.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze the effect of the temperature change on pressurizer level
Harris NKC Written Examination
Senior Reactor Openiwr
QUESTION: 22
Given the following conditions:
The unit is operating at 12Yo power.
?'he following RCP vibrations are observed:
IlVDICATION RCP 'A' RCP 'R' RCP 'C'
Frame Vibration 3.6 mil and 1' at 2.8 mil and stable 4 mil and 1'at
0.3 mil per hr 0.1 mil per hr
Shaft Vibration 12 mil and 1'at 3 mils and s h h k 14 mils and 1'at
0.3 mil per hr 0.6 mils per hour
Which of the Miowing describes the actions required for this condition'?
a. Stop RCP 'A' and initiate a plant shutdown
h. Trip the reactor, stop KCP 'A', and go to PATH-1
c. Stop KCP 'C' and initiate a piant shutdown
d. Trip the reactor, stop RCP IC', and go to PATH-1
ANSWER:
a. Stop RCP ' A and initiate a plant shutdown
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMREK: 22 TIENGROUP: 2il
10CPR55 CONTENT: 41(b) 5 43m
KA: 003.41.01
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated
with operating the KCPS controls including: RCP vibration
OBJECTIVE: AOP-3.18-3
Given a qet of plant conditions and a copy of AOP-018, DETERMINE the appropriate responhe
DEVELOPMENT REFERENCES: AOP-018, p 28
REFERENCES SUPPLIED TO APPLICART: AOP-018, Attachment 1 (Sheet 2 of 2 ONLY)
QIJESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: AOP-3.18 0 I7
NHC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION ( C O W C T ANSWER X'd):
X a. 'A' RCP vibration has exceeded limits and the pump must be stopped. With the plant in Mode 2, a
reactor trip is not required, but the plant must be shutdown.
b. Plausible since these \~.ouldbe the correct actions ifthe plant was in Mode 1, hut the plant is in Mode
L
e. Plausible since these are the correct actions, but 'C' RCP has not reached any trip limits while 'A' RCP
has.
d. Plausible since these would he the correct actions ifthe plant was in Mode 1, but 'C' RCI' has not
reached any trip limits while 'A' RCP has and the plant is in Mode 1.
DIPFICIJLTY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KSOV17LEDGE/RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis to determine which RCP must be stopped and comparison to power
level to determine proper action
Harris NKC Written Exmiination
Senior Reactor Operator
QUESTION: 23
ALB-OO~-R-L PRESSI JRIZEK RELIEF TANK HIGII-LOIX~IXVEL PRESS OK I I:MP,
alarms due to a high temperature condition.
Which of the following describes how the Pressurizer Relief Tank (PRT) is normally
cooled, in accordance with OP-100, Reactor Coolant System?
a. Recirculate the PRY through the Keactor Coolant Drain Tank heat cxchanser.
using Component Cooling Water to mol the heat exchanger
h. Recirculate the PKT through the Reactor Coolant Drain Tank heat exchanger.
using Service Water to cool the heat exckanger
c. Drain the PRI to the Keactor CooIant Drain Tank while making up to the P R l
from the Demineralized Water Storage Tank
d. Ilrain thc IRT tri the Reactor Coolant Drain Tank while making up to the PRT
from the Keactor Makeup Water Storage Tank
ANSWER:
a. Recirculate the PKT through the Keactor Coolant Drain I d heat exchanger,
using Component Cooling Water to cool the heat exchanger
HaiTis NRC Written Examination
Senior Reactor Operator
lhta Sheets
QUESTION NUMBER: 23 TIEWGROIJP: 211
10CFR55 CONTENT: 41(b) 7 43w
K4: 007K4.01
Knowledge of PRTS design feature(sj and/or interlock(s) which provide for the following: Quench tank
cooling
OB.IF,CTIVE: PZK-3.o-3
Given a flow diagram of the P R I or associated subsystems and the appropriate procedure, correctly
A1,IC;N the PRT for filling, draining, recirculation, or cooldown
DEVELOPMENT REFERENCE.!% APP-ALB-009, pg 29
c)P-loO1pg 30
REFERENCES SKJPPLIED TO APPLICANT: None
QUESTION SQIIRCE:
0 X NEW SIGNIFICANTLY MODIFIED
BANK NTiMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
DIKECT
NRC EXAM HISTORY: Noue
DISTRACTOR JIJSTIPICACTION (CORRECT ANSWER Xd):
X a Normal cooling of the PR? i s accomplished by recirculating the PKI water through the RCDT heat
exchanger, which is cooled by CCW.
h. Plausible since noma1 cooling of the PRr is accomplished by recirculating the PRT water through the
RCDT heat exchanger, but it is cooled by CCW, not SW.
e. Plausible since a rapid cooldown ofthe PRT would be acwmplished by draining to the RCDT and
making up to the PRT, but the makeup source is RMUW, not the D W X .
d. Plausible since this method would be used for a rapid cooldown of the PRT, but is not the nonnal
cooldown method used.
DIFFICULTY ANALYSIS:
CX)MPRE.IENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 2
EXPIASATION Knowrledge of the design method of cooling the PR?
Harris NRC: Written 1:xamination
Senior Reactor Operaor
QUESTION: 24
Which of the following describes the effect of a loss of 125 VDC Bus DP-1 A-SA?
a. Eniergency Diesel Generator A-SA loses excitation power
b. Power is lost to the Emergency Escape Air Imck
c. Master relays in SSPS Train A lose power
d. Main Turbine DC Hearing Oil Pump loses power
ANST5'E.R:
a. Emergency Diesel Generator A-SA loses excitation power
IIarris NRC Written Examination
Senior Reactor Operator
Data Sheets
QI!ESTION NUMBER: 24 TIERKROCP: 111
10CFR55 CONTENT: 41(b) 7 43@)
K h : 064K2.03
Koowledge of EDG bus power supplies to the following: Control power
OBJECTIVE: AOP-3 25-3
Given plant conditions?DISCUSS the following notesI cautions, and procedural steps as they apply
The effects of a loss o f a DC bus on equipment operability
DEVELOPMENT REFERENCES: AOP-3.25, p 39
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
X a. DP-IA-SA supplies the ED(? governor and generator excitation control circuits.
h. Plausible since the emergency escape air lock is powered from Dc.,but not the emergency DC bus.
c. Plausible since SSPS receives iuput from the emergency I X bus and the master relays operate on DC,
brit the emergency bus only supplies the Kx Trip Breaker shunt trip power and the master relays arc
powered by 4X vdc which is produced in SSPS via the instrument buses.
d. Plausible sinc.e the TIC baring oil pump is powered by DC and is one of the only 1)C loads
specifically addressed in the I!.OPs, but it is powered by the non-safety related 250 VDC.
DIPFICtJLTY ANALYSIS:
COMPREIIENSIVE / AXALYSIS KYOWLEDGE / RECALL
DIFFICULTY HATING: 2
EXPLANATION: Knowledge of the source of control power to the E.DGs
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 25
Given the following indications during a plant startup being performed in accordance
with GP-005, Power Operation:
- Power Range Channel N-41 26.0%
- Power Range Channel N-42 24.5%
- Power Range Channel N-43 24.5%
Power Range Chnnnel N-44 25.0%
Loop A AT 25.5%
Loop AT 25.5%
0 1,oop C AT 25.5%
- Turbine Load 24.5% (DEI1 units converted to percent load)
Whish of the following power levels should be reported as being actual reactor power?
a. 24.5%
b. 25.0%
c. 25.5%
d. 26.0%
ANSWER.
c. 25.5%
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUhXHER: 25 TIEWGROUP: 2/2
10CFR5S CONTENT: 41(b) 5 43w
KA: 002K.5.10
Knowledge of the operational iniplications ofthe following concepts as they apply to the KCS:
Kelationship between reactor power and RCS differential temperature
OBJECTIVE: NIS-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 SIJPPLIED TO APPLICANT: None
QUESTION SOURCE: II]NEW SIGNIFICANTLY MODIFIED I7 DIRECT
HANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT NIS-Kl0 003
NRC EXAM HISTORY: Kone
D I s r m c r o R JUSTIFICACTION ( c o r n uANSWER XV):
ar. Plausible since this is the lowest given power level and may be considered to he the most
conservative, but GI)-005 provides guidelines for which power level should be considered.
h. Plausible since this is the average NIS power level; but the highest as identified by GP-005
requirements is the average loop AT.
X c. Until a calorimetric is performed at 30?? power, true reactor power shall be assumed equal to the
highest of the following indicators: average Power Range NI value, average percent AT>or Main
Turbine load
d. Ilausible since this is the highest given power level and may be considered to he the most
conservative, hut GP-005 provides guidelines for which power level should be eonside.red.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Calcuhtion of a\wrage power indications and determination of most
conservative value
EIarris NKC Written Examination
Senior Reactor Operator
QUESTION: 26
AII-82.4, NORMAL PLXGE SUPPI,Y FAN AH-82A. Pails to start when the contrd
switch is placed in STAKT.
Mihich of the following interlocks would prevent the fan from starting?
a. Normal Purge Inlet and Discharge Valves are open
b. AII-82A fan inlet damper has failed to open
c. Elec.tric heating coil breaker is tripped
d. Containment differential pressure is zero
ANSWER:
d. Containment differential pressure is zero
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER 26 TIEWGROUP: 212
1OCFR55 CONTEhT: 41(b) 5 43(W
KA: 029r21.07
Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated
with operating the Containment Purge System controls including: Containment pressure, temperature,
and humidity
OI~('TIVE: cvs-3. o m
LOCATE the ccmtrols and EXPLAIN the interlocks associated with the following major components
- NCPMIJ units. including AII-82 fans
DEVELOPMENT REFERENCES: OP-168, p 8
REVERENCES SUPPLIED TO APPLICANT: None
QUESTION SOERCE: 0X NEW 0 SIGNIFICANTLY MODIFIED 0 DIRECT
IUNK NL1MBE.R FOR SIGNIFICANTLY MODIFIED /DIRECT: Xew
NRC EXAM HISTORY: None
1)ISTMACTOK .nTSTIFICAC'rION (CORRECT ANSWER X'd):
s interlocked to close if fan AH-824 is stopped, but are manually ope.ncd
a. I'lausihle since the v a l ~ e are
prior to the start of the fan.
b. Plaosibie s h e the inlet damper is interlocked to open when the fan is started, hut are closed when the
fan is started.
c. Plausible since the heating coils arc interlocked with the fan operation, hut the heaters are enahled to
operate when the fan is running and do not prevent the fan from starting.
X d. Fan AII-82A will only start if containment <APis more negative than -0.400 INWG
DIFFICITLTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICIJ1,TY RATING: 3
EWLANATIOS: Knowledge of interlocks associate with containment purge fans
Harris NRC Written Exmination
Senior Reactor Operator
QUESTION: 27
Given the following conditions:
- RCS temperature decreases and stabilixs at 548 "F.
Which of the following predicts the plant response and the operator actions required in
accordance with GP-004, "Reactor St'artup"?
a. Reactor power increases; withdraw control rods and dilute, in a controiled
manner, to restore RCS temperature to program within 15 minutes
b. Redctor power increases; trip the reactor if RCS temperature CANNOT be
restored above 5 5 1 "F 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 subcritical; immediately trip the reactor
ANSWER:
b. Reactor power incrcases; trip the reactor if RC'S temperature CANNOT be
restored abovc 55 1 "F in a controlled manner within 15 minutes
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QGESTION NUMBER: 27 TIER/GROUP: 211
10CYR55 CONTENT: 41(b) 5 43@)
KA: 039A22.05
Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (h) based
on predictions, use procedures to correct, control, or niitigate the consequences of those malfunctions or
operations: increasing steam demand, its relationship to inc,reases in reactor power
OBJECTIVE: IE-3. IO- 1
Apply the philosophies of OMM-001 arid PLP-629 regarding safe and conservative decisions that must
be made by a control room crew
DEVELOPMENT REFERENCES: GI>-004pg 9 P & L # 19
OMM-001 pg 66-69
IE-LP-3.10 (Salem Event, SOER 94-01)
REFERENCES SEPPLIED TO APPLICAVT: None
QIJESTION SOURCE: 0X NEW SIGNIFICANTLY MODIFIED 0 DIRECT
HANK NIJMBER FOR SIGSIFICANTLY MODIFIED / DIRECT: New-
NRC EXAM HISTORY: None
DISTRACTOR .WSTIE'ICACTION (CORRECT ANSWER X'd):
a. Plausible sinc.e reactor power will incrcase, but temperature is not to he restored using two different
methods of rextivity control sirnuitaneonsly and the 15 minute limit is to restore temperature above
55 1 O F , not to program.
X b. 'The first operator action should be to attempt to stop the cause. (e.&, secure the overfeeding) of the
transient. Temperature may then be recovered by using control rods in a slow and controlled tr1anne.r.
'Temperature has to be restored to greater than 551 "1' within 15 minutes due to the requirements of TS
3.1' I .4.
s. Plausible since the 15 minute time limit is assoc.iated with restoration, but the reactor does not become
subcritical.
d. Plausible since the reactor is to bc tripped if it becomes subcritical due to a malfunction per O m -
001, but the reactor does not become subc.ritical.
DIFFICU1,TY ANALYSIS:
COMPREHENSIVE / ANALYSIS c[ KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze the plant response to an increase in steam demand and determine
appropriate actions
IIarris NRC Written Examination
Senior Reactor Operator
QUESTION: 28
The plant is operating at 100% power with the following conditions:
w Ambient Temp CT Hasin TemQ
1500 35 "F 64 "F
1900 20 "F 60 OF
2300 10 "F 58 OF
Which of the following describes the correct CT Deicing Gate Valve alignment for these
conditions?
1900 2Q.Q
a. Full Open Full Open
b. Full Open IIaIf open
c. Half Open Full Open
d. Half Open Half Open
AXSWER:
b. Full Open Half Open
Harris NRC Written Exaniination
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER: 18 TIENGROUP: 3
10CFH55 CONTENT: 41(b) 10 43w
KA: 2.1.25
Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which
contain performance data
OBJECTIVE: C T - R ~
Given c1P-141, Attachment 5, ANALYZE a set of advcrse weather conditions and DESCRIBE the
operation ofthe Cooling Tower System to prevent ice damage to the fill material
DEVELOPMENT REFERENCES: OP-141, pg 50 Attachment 5
REFEKENCES SUPPLIED TO APPLICANT: OP-141, Attachment 5
QUESTION SOIJRCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NIJMBER FOR SIGMFICANTLY MODIFIED /DIRECT: CT-R3 001
NRC E.XAM HISTORY: None
DISTRACTOR .Jl:STIFIC:ACTION (CORRECT ANSWER Xd):
a. Plausible since valves should be open at 1900, hut are required to he changed to half open at 1-300.
X b. At 1500 conditions call for valves to be full open, at 1900 conditions call for no change in position,
and at 2300 conditions call for change to half open.
c. Plausible since valves should he changed between 1900 and 2300, but should go from fW open to half
open.
d. Plausible since valves should be half open at 2300, hut should he full open at 1900 due to no change
from 1500.
D1FFICI:LTY ANA1,YSIS:
COMPRFBEXSIVE I ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Application of given data to cnwe to detennine required operation of deicing
valves
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 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 releases through the SG PORV?
a. Increase A SC;IORV setpoint on PK 308AI SA to 90% (1 170 p i g )
b. Increase A SG PORV setpoint on PK 308A1 SA to 88% (1 145 psig)
e. Place A SG PORV PK 308A1 SA in MANUAL with zero output
d. Manually isolate A SG PORV by closing 1MS-59
ANSWER:
b. Increase A SG PORV sctpoint on PK 308.41 SA to 88% (I 145 psig)
IIarrin NKC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER: 29 TIEWGROUP: 3
10CFRS5 CONTENT: 41(b) None 43(h) None
KA: 2.3.11
Ability to control radiation releases
OBJECTIVE: EOP-3.2-2
I>Bh.IONSIMTE the below-assumed operator knowledge from the FlNP Step Deviation Documents
and the WOG ERGS that support performance of E.OP actions
e Method of isolating SGTK
DEVELOPMENT REFERENCES: PA?"-? pg 8
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE:
0
X NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Sew
XRC EXAM HISTORY: None
DISTRACTOR JUSTIFICLKTION (CORRECT ANSWER X'd):
a. Plausible since PORV setpoint is adjusted, but should be adjustcd to 1145 psis and 1170 psig is the
first safety setpoint.
X b. The SG PORV is to be set at 8S% to minimize the likelihood o f a release, but lower than the SG safety
setpoints.
e. Plansible since this action w3ouId be taken if the SG were faulted instead of ruptured
d. Plausible since this action would be taken if the SG POKV were to fail open, hut this would also cause
the safeties to be challenged and should not be performed unless necessary.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY HATING: 3
EWLASATION: Knowledge of steps required to isolate a SGTR
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 30
Which of the following two (2) conditions are both identified by IPP-013, LOCA
Outside Containment, as being used to identify that the LOCA has been isolated?
a. e RCX pressure increasing
0 RAB local room temperatures
b. e RAB local room temperatures
0 RAB radiation levels decreasing
c. 0 RAB radiation levels decreasing
I m a l observation of the isolation
d. e R(:S pressure increasing
e Local observation of the isolation
ANSWER:
d. RCS pressure increasing
1,ocal observation of the isolation
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 30 TIEWGROUP: 111
10CFH55 CONTENT: 41(h) S/lO 43(h)
KA: WE04F:K1.2
Knowledge of the operational implications of the following concepts as they apply to the &OCA Outside
Containment) Normal, ahnormal and emergency operating procedures associated with (LOCA Cltitside
Containment)
OBJECTIVE: 1.3-K4
I h n g appropriate plant procedures and prints, determine the following:
0 Transitions to other EOPs
DEVELOPMEXI REFERENCES: EPP-013 pg 5
REFEHENCES SUPPLIED TO APPLICANT Xoae
QEESTION SOURCE: 0 NEW 0 SIGNIFICANTLY MODIFIED DIRECT
RAAK NITMBER FOR SIGNIFICANTLY I\1ODIFIED / DIRJZCT: 3.3 024
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xcl):
a. Plausible since RCS pressure increasing is one of the indications used, hut local temperatures are not
used in EPP-013.
h. Plausible since those may both be indications that might support that the leak is isolated, but
pressurizer Level may not be indicative of actual RCS inventory or the leak being isolated and is nut
used in EPP-0 13.
E. Plausible since local observation is one ofthe indications used, but RAH radiation levels may be
elevated for some time after isolation and is not used in EPP-013.
X d. EII-013 determines that the L.OCA outside containment is isolated if I K S pressure is increasing and
if iocal observation confirms the isolation.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the conditions required by E.PP-013 to determine that a LOCA
outside containment is isolated
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 31
Which of the following is the reason for purposely tripping the Reactor Coolant Pumps
(RCPs) 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
b. Prevent KCP ninout in the event of a large break I B C A
c. Prevent excessive depletion of RCS inventory through a small break in the RCS
d. Prevent damage to RCPs due to pumping a two-phase mixture event
ANSWER
c. Prevent excessive depletion of RCS inventory through a small break in the RCS
Harris NRC Written Exaniination
Senior Reactor Operator
Data Sheet?
QUESTION NTJMBER: 3 1 TIER/GHOUP: I/I
10CFR55 CONTENT: 41(b) 5/10 43(b)
KA: ooono9~~3.23
Knowledge of the reasons for the following responses as the apply to the small break EOCA: RCP
tripping requirenrents
OBJECTIVE: BD-3.1-1
Analyze the Reactor C:oolant Pump (RCP) trip criteria. lliis analysis should include. at the minimum, the
following topics:
0 The reason for purposely tripping the RCPs under certain accident conditions
DEVELOPMENT REFERENCES: Generic Issues of ERG Background - Executive Volume
LP-13D-3.1 pg 8
REFERENCES SUPPLIED TO APPLICANT: None
QtIESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: BD-3.1001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since for most accidents it is desirable to have RCPs available, particularly those cases where
an inadequate core cooling condition might exist.
b. Plausible since little work is required by the RCPs in the event of a large break LOCA, but this would
result in a lower pump current, nut a runout c.ondition.
X c. Tripping the RCPs during the early stages of a small break 1,OCA limits the amount of mass lost out
the break, thereby increasing the inass availabie for heat removal in the event the pumps were not
tripped but tripped at a later time.
d. Plausible sinc.c RC:Is are not designed to pump a two-phase mixture and it would be desirahle to
protect the pumps from damage.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 7
EXPLANATION: Knowledge of the reasons for tripping RCPs during a small break LOCA
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 32
Given the following conditions:
0 The unit is in Mode 3 at normal operating pressure
0 Pressurizer Pressure Control is in AUTO.
0 Pressurizer Pressure Channel PT-445 fails high.
0 PW, Pressure Channel indications are:
0 PI-444 2050psig
e PI-445 2500psig
0 PI455 2050psig
0 PI-456 1950psig
PI-457 2050psig
Assuming NQ operator actions, which of the following describes the expected conditions
of the P U Pressure POKVs and Spray Valves?
a. e PRZ PORV IRC-I 14 closed
PIG? PORVS 1RC-I 16 a d IRC-118 open
0 PRZ Spray Valves PCV-444C and PCV-444D open
b. 0 PRZ PORV IRC-I 14 open
e PRZ PORVs 1RC-116 and IRC-118 closed
e PRZ Spray Valves PCV-444C and PCV-444D open
0 PRZPORV lRC-116and 1RC-118 open
0 PRZ Spray Valves PCV-444C and PCV-444D closed
d. 0 PRZ PORV IRC-I 14 open
e PRZ PORVS1RC-116 and IRC-118 closed
0 PRZ Spray Valves PCV-444C and PCV-444D closed
ANSWER:
c. * PKZ PORV IRC-114 closed
0 PRZPORV lRC-116and IRC-118open
0 PR7, Spray Valves PCV-444C and PCV-444D closed
The noun names were provided for the following valves:
Harris NRC Written Exaniination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER 32 TIEFUGROUP: 1!1
IOCFW55 CONTENT: 41(b) 7 43w
KA: 000027AK2.03
Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the
following: Controllers and positioners
OBJECTIVE: PZRPC-3 .O-3
Given the status of the various pressurizer pressure channels, the position of various pressure control-
related control switch positions and the status of Controllers PK-444A, PK-4442, and PK-444D,
PREDICT the responses of the following functions:
Pressurizer spray valves
- Pressurizer Power-Operated Relief Valves (PORVs)
Pressurizer pressure permissive P-11
DEVELOPMENT REFERENCES: SD-100.3, pg 12, 16,38-39
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOCJRCE: NEW 1 SIGNlFICANTLY MODIFIED DlRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: P%RPC-R3003
NRC EXAM HISTORY None
DISTRACTOR JZJSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since POKVs I I 6 and 118 will open until actual pressure drops be low^ 2000 psig, but the
spray valves are controlled by the. other channel and will not open.
h. Plausible since this would be the response of the system if the failed channel was 444, but with 445
failed, none of these components are affected.
X c. PT-445 controls only PORVs 116 and I18. Tlie PORVs will open and remain open until 2/3 of the
protection channels 455/456/457 decrease below the P-l 1 setpoint of 2000 psi& Spray valves are
controlled by chnnnel444.
(1. Plausible since the spray valves will remain closed. but 445 controls PORVs 116 and 118, not 114.
DIFBICICULTY ANALYSIS:
COMPREHENSRE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EWLANATIOS: Analysis of the failure and current plant conditions to determine expected
response of pressure control
Harris NRC Written Examination
Senior Keactur 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 (CWPs)?
a. VSFC oil is bypassed around the hydraulic coupling as necessary to maintain a
constant feed pump suction pressure
h. 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 CRP recirc valves
closed
ANSWER:
c. VSFC hydraulic coupling is varied as necessary to maintain a constant feed pump
suction pressure
IIarris WRC' Written Examination
Senior Keactor Operator
Data Sheets
QUESTION NUMBER: 33 TIEWGROUY: 211
IOCFR55 CONTENT: 41(b) 7 '4309
KA: 056'32.1.28
Knowledge of the purpose and function of major system components and controls. (Condensate)
OI?.KCTTVE: CFW-3.0-4
DESCRIBE the basic. construction and operation of the following CFW System components 1
subsystems
CBP Variable Speed Firrid Chpling (VSFC)
DEVELOPMENT REFERENCES: SD-134. p 7 , 1 7
KEFERENCES SUPPLIED TO A P P I K A N T None
QIJESTION SOURCE: NEW SIGNIFTCANTLY MODIFIED DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CFW-R3 001
NRC E X A M HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since oil adjusts the hydraulic coupling to maintain a constant suction pressure at the feed
pump, but the oil does not bypass the hydraulic coupling
h. Plausible since oil adjusts the hydraulic coupling, but it does nut bypass the hydraulic coupling and
does not maintain the CDP recirc valves closed.
X c. An oil bath between the motor and pump coupling causes the pump to operate at a variable speed to
maintain a c.onstant suction pressure at the feed pump.
d. Plausible 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 c1ose.d.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KYOWLEDGE / RECALL
DIPFICUtTY RATING: 3
EXPLANATION: Knowledge of the operation ofthe CBPs
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 34
Given the following conditions:
The plant is operating at 100% power.
- A tube leak has been detected on 'ESG.
?'he Condenser Vacuum Pump Rad Monitor, KEM-ITV-3534, and H-X-15 curves are
being monitored every 15 minutes to estimate the leak rate.
CVPE is operating with NO motivating air.
Which oftlie following readings noted on REM-lTV-3534 is the MINIMIJMreading
that would require a plant shutdown per Technical Specifications'?
a. 5.40E-7
b. 6.00E -7
c. 1.08 E -4
d. 1.80 E -6
ANSWER:
C. 1.08 E - 6
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER: 34 TIEWGROUP: 1!
2
10CFRS5 COSTENT 41(b) None 43(b) 5
KA: 000037AA2.10
Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: Tech-
Spec limit? for RCS Leakage
OBJECTIVE: AOP-3.16
For a primarv-to-secondary leak. DESCRIBE when a power reduction or unit shutdown is required.
DEVELOPMENT RF3EREXCE:S: AOP-0 16 pg 15
Curves H-X-ISdhic
REFERENCES SUPPLIED TO APPLICANT: Curves 11-X-I Sdbic
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Harris NRC 2000-80
NRC EXAM HISTORY: None
DISTRACTOR NTTSTIFICACTION (CORRECT ANSWER X'd):
a. PLausible since this exceeds would excecd PSAL 2 limits if operating on full motivating air (curve H-
X-l5a), but the incorrect curve is used.
h. Plausible since this exceeds \vould exceed PSAL 2 limits if operating on intermediate motivating air
(c.urve H-X-l5b), but the incorrect curve is used.
X c. Lowest level that would exceed 7.5 gpd (PSAL 2) which would require a TS shutdowrn.
d. Plausible since this exceeds the PSAL 3 limit which would require a TS shutdown, hut this is not the
lowwt lc.vel that would require the shutdown.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Interpretation of plant data on RCS leakage curve and comparison to procedural
requirements
Harris NRC Written Examination
Senior Reactor Opetator
QUESTION: 35
FRP-J.?, I<esponsc to Containment Flooding, directs that the containment sump be
sampled for activity, a i d then to notil+ 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 CNMT spray additive tank should be isolated
c. If Emergency Service Water to containment should be isolated
d. If sump water may be transferred to tanks outside containment
ANSWER:
d. If sump water may be transferred to tanks outside containment
Harris NRC Nkitten Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 35 TIEWGROUP: 1/2
10CFRSS CONTENT: 41(b) 8/10 43(b)
KA: WElSEK1.2
Knowledge ofthe operational implications of the following concepts as they apply to the (Containment
Flooding) Normal, abnormal and emergency operating procedures associated with (Containment
Plooding)
OBJECTIVE: 3.13-4
Given the following EOP steps, notes, and cautions, 1)ESCKIRE the associated basis
Sampling the CNMT sump for activity (J.2)
DEVELOPMENT REFERENCES: FRP-J.2, pg 4
I.P-3.131pg 12
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIPICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: 3. i 3 o i n
NNC EXAM HISTORY None
DISTRACTOW JTJSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since the plant operations staff does make the determination when Containment Spray is to
be secured, but this sample is to determine whether the water can be transferred.
b. I'lausible since if flooding has occ.urred it is likely that a large RCS leak has also occurred and the
spray chemical addition tank may have emptied to containment and would no longer be needed, hut
this saniple is to determine whether the water can be transferred.
c. Plausible since a potential source of flooding is the ESW system to the fan coolers, but this sample is
to determine whether the water can be transferred and ESW isolation would be determined by the
operating crew based on ESW indications.
X d. The containment sump is sampled to determine if excess water can be transferred to storage tanks
located outside containment.
DIFFICULTY -4NALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 2
EXP1,ANATION: Knowledge of pnrpose for sampling sumps foliowing flooding inside
containment
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 36
Given the following conditions:
RHR Pump A-SA is tagged out.
Following a large break LOCA, the crew was performing EPP-010, Transfer to Cold
Leg Recirculation.
1SI-301, CONTAINMENI SIJMP TO KHR PUMP D-SB, failed to open and the
crew transitioned to lTP-012, Loss of Emergency Coolant Rccirc.ulation.
Both Containment Spray Ipumpsautomatically transferred to the Containment Sump.
Two (2) Containment Fan Coolers are operating.
Containment pressure is I 2 psig and decreasing slowly.
While performing E.PP-012 the Reactor Operator notes that RWST 1eve.lis 2% with
both CSIPs, both Containnient Spray Pumps, and RIIR Punip R-SB operating.
Which of the following actions are to be taken?
a. Stop the RIIR pump ONLY
b. Stop both CSIPs and the RKR pump ONLY
c. Stop both CSIPs, the RIIR pump, and one Containment Spray pump ONLY
d. Stop both CSIPs, the RIIR pump, and both Containment Spray pumps
ANSWER
b. Stop both CSIPs and the RIIR pump ONLY
Harris NKC Written Fixamination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 36 TIER/GROXP: Ill
10CFR55 CONTENT: 41(b) 8/10 43(b)
KA: WEllEK1.1
Knowledge ofthe operational implications ofthe following concepts as they apply to the (Loss of
Emergency Coolant Kecirculation) Components, capacity, and function of emergency systems
OBJECTIVE: 2 . 3 4 2
Predict how each of the followring could impact efforts to maintain core cooling during a ILEA
- Failure of valves to realign for cold-leg recirculation
DEVELOPMENT REFERENCES: EPP-012 pg 42
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.3-RS OW
NRC EXAM HISTORY None
DISTRACTOH JWTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the RHK pump is still aligned to the RWST and must be. stopped, but the CSIPs are
also aligned to the RWST and must likewise be st0ppe.d.
X b. The RHK pump and the CSIPs are still aligned to the KWST and must be stopped when the RIVSI
empty alarm is received at 3% level.
c. Plausible since the RfIR pump and the CSIPs must be stopped, but the spray pumps can continue to
operate since they are no longer aligned to the RWST.
d. Plausible since the RNK pump and the CSTPs must he stopped, but the spray pumps can continue to
operate since they are no longer aligned to the RUST.
DIPFICTJ1,TY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWL.EDGE/ RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze. plant conditions to determine which pumps arc taking a suction from
the KWST to determine the pimps which are to be stopped
Harris NKC Written Examination
Senior Keactor Operator
QUESTION: 37
L?-115, VCT L,evel. has failed LOW. The I:nit-SCO directs the Reactor Operator to
maintain VCT level between 20% and 40%.
Which ofthe following describes how VCT level will be maintained in accordance with
AOP-003. Malfmction of Reactor Makeup Coiitrol~
a. 0 When level lowers to 20%, automatic makeup will begin raising level
0 When Ievel increases to 70%, ICs-120 (LCV-112A). Letdown VCTMold LJp
Tank, will begin diverting letdown to the Hold Up Tank
b. e When Ievel lowers to 20%. the operator must start a manual makeup to raise
VCT level
When level increases to 70%, 1CS-120 (I.CV-l12A), Letdown VCT4Iold Up
Tank, will begin diverting letdown to the Hold Up Tank
e. 0 When l e d lowers to 20%, automatic makeup will begin raising level
0 When level incremes to 70%, the operator must align 1CS-120 (LCV-I l2A),
Letdown VC?/Hold Lip Tank, to the Hold Up Tank
a. 0 When level lowers to 20%. the operator must start a manual makeup to raisc
VCT levei
0 When level increases to 70%, the operator must dign ICs-120 (LCV-l12A),
Letdown VCT/Hold Up Tank, to the Hold C p Tank
ANSWER:
b. * When level lowers to 20%, the operator must start a manual makeup to raise
vc?level
o When level increases to 70%, 1CS-120 (LCV-112A). Letdown VCT/Hdd IJp
Tank, will begin diverting letdown to the Hold Up Tank
Harris NKC Written Examination
Senior Reactor Operiltor
Data Sheets
QVESTION NUMBER: 37 TIEWGROUP: 2/1
lOCFRS5 CONTENT: 41(b) S 43(b)
KA: 004A1.06
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated
with operating the CVCS controls including: VCT level
OBJECTIVE: CVCS-RS
PREDICT the response of the CVCS to the following failure\
c. 1.T-I12 or 1,T-llSfailure (high or low)
DEVELOPMENT REFERENCES: AOP-003, pg 5-6, 16
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW 17 SIGNIFICANTLY MODIFIED 0 DIRECT
OR SIGNIFICANTLY MODIFIED / DBRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUS'HFICACTION (CORRECT ANSWER X'd):
a. Plausible since 1,T-I 12 will still coutrol CS-120 properly, causing a divert to the HUI, hut the
operator must perform a manual blended flow due to the failure of LT-115.
X b. A low failure of LT-I 15 will disable auto makeup capabilities which will required the operatur to
perform a manual blended flow and the modulate divert to the HUT is controlled by LT-112.
c. Plausible since operator action is required to perform one of the two evolutions, but the automatic
makeup. not the divert, must he controlled by the operator.
d. Plausible since a low failure of LT-115 will disable auto makeup capabilities which will required the
operator to perform a manual hlended flow, hut the modulate divert to the H I T is controlled by LT-
112.
1)IFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE I RECALL
DIFFICULTY R4TING: 3
EXPLANATION: Analysis of plant response to failures in CVCS to determine the proper operntor
response
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 38
The plant is operating at 100% power with all equipment operabie and properly aligned.
Which of the following describes changes to the Component Cooling Water System
alignment following a Safety Injection signal?
a. CC'W to the Gross Failed Fuel Detector and Primary Sample Panel isolates
b. Both CCW pumps start and the NonBssentiaI header isolates
c, CCW to and from the RGP Motor Coolers isolates
d. Both CCW punips start and the Thermal Harrier Hx Keturn isolates
ANSWER:
a. CCW to the Gross Failed Fuel Detector and Primary Saniple I'anel isolates
Harris NRC Written Exmiination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 38 TIEWGROITP: 2! 1
10CFR55 CONTENT: 41(b) 7 43(b)
K4: 008A3.08
Ability to monitor automatic operation of the CCWS, including: Automatic actions associated with the
CCWS that occur as a result of a safety injection signal
OBJECTIVE: CCWS-3.0-R2
STATE how the CCWS responds during each ofthe following conditions:
Safety Injection signal
DEVELOPMENT REFERENCES: SD-145pg 16-17
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CCWS-K2 002
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER X'd):
X a. On an SI signal, both the GFFI) and sample panel receive isolation signals.
h. Plausible since the pumps will get a start signal, but only the GFFD and sample panel in the non-
essential header are isolated.
c. Plausible since the CCW to RCP isolations close on a Phase R signal, hut Phase B is not generated hy
an SI signal.
d. Plausible since the pumps will get a start signal, but the thermal harrier heat exchangers are only
isolated on a Phase B signal.
DIFFICULTY ANALYSIS:
COI\fPREHENSIVE / ANALYSIS KNOWLEDG&/RECAL.I,
DIFFICULTY RATING: 3
EXPLANATION: Knowledge ofthe response of CCWS to an SI signal
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 39
Given the following conditions:
The plant is operating at 23% power.
Steani pressure channel PT-475 is selected for control of SG A.
Steam pressure transmitter PT-475 fails high.
Assuming NO opcrator action, which of the following statements describes the response
of the Steam Generator Water Ixvel Control System (SGWLCS)?
a. An increase in steam flow from SG A is sensed and responds by increasing
1FA-140, MiX FW A KEG BYP FK-479.1, position to increase feed flow to SG
A and level increases
b. An increase in steam flow from SG A is sensed and responds by increasing
1FW-133, MAIN FW A KEGULATOR FK-478, position to increase feed flow to
SG A and level increases
c.. A decrease in steam flow fiom SG A is sensed and responds by decreasing IFW-
140, MN FW A REG BYP FK-449.1, position to decrease feed flow to SCJ A
and level decreases
d. A decrease in steam flow from SG A is sensed and responds by decreasing IFW-
133, ILIAlN FW A REGULATOR FK-478, position to demease fcccl flow to SG
A and IeveI decrcilses
ANSWER:
b. An increase in steam flow from SG A is sensed and responds by increasing
1FW-133, MAIN I W A REGULATOR FK-478, psition to increase feed flow to
SG A and level increases
Harris NRC Written Examination
Senior Reactor Operator
DaFa Sheets
QUESTION NUMBER: 39 TIEWGHOUP: 2!1
10CFH55 CONTENT: 41(b) 7 4 W
K4: 059A4.08
Ability to manually operate and monitor in the control room: Feed regulating valve controller
OBJECTIVE: SGWIX-3.0-2
Given the status of the various SGWLC related control switch positions and controllers, PREDICT how
a malfunction of the following will effect the SGWLC System:
S G pressure channels
DEVELOPMENT REFERENCES: SD-126.02 pg 4 , s
REFERENCES SLTPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOK SIGNIFICANTLY MODIFIED /DIRECT: SGWL.C-R2 002
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since steam pressure failing high causes the stei~nflow to increase, resulting in SF > FF, hut
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 > FF. The feed reg
valve, in operation at 1% power, opens to muse FF and level to increase.
c. Plausible since steam pressure failing causes the steam flow to change, resulting in a SF - FF
mismatch, but the feed reg valve will open to inc.rease FF.
d. Plausible since steam pressure failing causes the steam flow to c.hange, resulting in a SF - FF
mismatch, but the feed reg valve will open to increase PF.
DIFFICULrY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY HATING: 3
EXPLANATION: Analyze the effect ofthe failure on the control system and recognke which
valve will be controlling at the power level given
Harris KRC Written Examination
Senior Reactor Operator
QUESTION: 40
The plant is operating at 80% power 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:
Pressurizer pressure is 2275 psig and slowly decreasing.
Pressurizer level is 45Oh and slowly decreasing.
Both pressurizer spray viilves indicate mid-position.
- All pressurizer backup heaters are de-encrgiyed.
These conditions are indicative o f . .
a. a normal plant response following an outsnrge from the pressurizer.
b. 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 plant response following an insurge into the pressurizer.
AILSWER:
c. a failure in the Pressurizer Level control circuitry, which failed to energize the
backup heaters.
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER 40 TIEWGROUP: -/ -
? i?
10CFR55 CONTENT: 41(b) 7 43m
K4: OllK6.04
Knowledge of the effect of a loss or malfunction on the following will have on thc PZK I,CS: Operation
of PZR Level controllers
OBJECTIVE: PZKI,C-3.0-5
EXPLAIN how the system controls pressurizer level, including the input parameters and the components
that rcceive output signals
DEVELOPMENT REFERENCES: SD- 100.3 pg 14-15
REFERE.NCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BAhW NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: PZ.RIGK7 001
NRC EXAM HISTORY: None
DISTRACTOR JITSTIIIICACTION(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.
b. Plausible since a downpower should result in an insurge which would cause the spray valves to open,
hut the heaters should also be energized.
X c. .4rtlpid downpower transient will result in an insurge to the prcssurizer. This should result in the
conditions noted, including a high pressurim 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 with
level more than 5% high is indicative of a level control system failure.
d. Plausible since the respnnse is correct, with the exception of the pressurizer heaters not heing
energized, for an insurge to the pressurizer.
DIFFICIiL.TY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICTJLTY RATING: 3
EXPLANATION: Analysis of the expected plant response and the actual plant response to an
insurge into the pressurizer
Harris KRC' Written Examination
Senior Reactor Operator
QIJESTION: 41
The operators are perforniing a start up to full power with Main Feedwater Pump 3 under
clearance.
Wlich of'the following causes an immediate start signal to ONLY the hlotor Driven
AFW Pumps?
a. SG A level is 18%
0 SG B level is 39%
SC; (' level is 38%
Loss ofEmergency Bus 1.4-SA
b. S G A level is 34%
SG K Ievel is 33%
SG C level is 22%
- Loss of lmergency Bus 1B-SB
C. SG .4 level is 25%
SG B level is 26%
SG C level is 27%
Main Feedwater Pump A trips
d. SG A level is 24%
SG B level is 23941
- S G C level is 28%
Main Feedwater Pump 4 trips
ANSWER:
c. SG 4 level is 25%
- S G B level is 26%
SCi C level is 27%
Main Feedwater Pump A trips
Hmis NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER. 41 TIEWGROW: 2/2
10CFR55 CONTENT: 41(b) 2-9 43(W
KA: 03SK1.01
Know>ledgeof the physical connections and/or cause-effect relationships betwen the S!GS and the
following systems: MFW!AFW systems
OBJECTIVE: AFS-3.0-R5
State the automatic start signals associated with the:
MDAFW pumps
0 IUAFW pumps
DEVELOPMENT REFERENCES: SI)-117 pg 12- 13
KEFEKENCES SUPPLIED TO APPIdCAKT: None
QUESTION SOIJKCE:
0 X NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGhTFICAKTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR .TUSTIIACTIQN (CORRECT ANSWER Xd):
a. Plausible since the SG levels will cause a start ofonly the MDAIW Pumps, but the loss of the
emergency bus starts the train related MDAFW Pump and the 1DAFW Pump.
b. Plausible since the SCi levels will cause a start of only the MDAFW Pumps. but the loss of the
emergency bus starts the train related MDAFW Pump and the TDAFW Pump.
X c. With all 3 S G levels above 25%, no start signals occur, hi3weve.r the trip of MFW Pump A w,ili cause
both MIIAFW Pumps to start since the B MFU Pump is already secured.
d. Plausible sinc.e the trip of MFW Pump A will cause both MDAFW Pumps to start since the E MFW
Pump is already secured, but 2 SG levels below 25% start the TDAFW Pump and the MDAIW
Pumps.
DIFFICULTY ANALYSIS:
COMPREHENSIVF. /ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of conditions which determine AFW pump starts
IIarris NRC Written Examination
Senior Reactor Operator
QUESTION: 42
In accordance with FW-H. I, Response to I.oss of Secondary IEeat 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 of the PFU safety valves opening and an increased loss of
RCS inventory
b. The increase in KCS temperature mill 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 iikelihood of thermal shock of
the reactor vessel
d. ?he increase in RCS temperature mi11 increase primary-to-secondary AP,
increasing the likelihood of a SGTR
ANSWER:
b. The increase in RCS temperature will increase KClS pressure and decrease SI flow,
increasing the likelihood of core uncovery
IIarris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 42 TIEWGROUP: 1!1
10CFR55 CONTENT: 41(b) 7 43ib)
KA: WE05EK2.2
Knowledge ofthe interrelations hetween the (Loss of Secondary Heat Sink) and the following: Facilitys
heat renioval syste.ms, including primary coolant, emergency coolant, the decay heat removal systems,
and relations between the proper operation of these systems to the operation of the facility
OBJECTIVE: 3.1 1-R4
Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
e Prompt initiation of Bleed and Feed
DEVELOPMENT REFERENCES: FRP-H.l, pg 19,22
EP-3.11, pg 10-12
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTION SOIJRCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: 3.1 1-R4 015
NKC EXAM HISTORY: None
DISTRACTOK JUSTIFICACTION (COKRECT ANSWER X'd):
a. Plausible since an increase in RCS pressure could result in the safety valves lifting if the PORVs were
to fail, hut steam production in the core is not likely to be occurring at the onset of the loss of heat
sink.
X b. Failure to estahlish KC:S bleed and feed when required will result in an increase in RCS temperature
which will cause an increase in RCS pressure. This will result in decreased SI flow and core
uncovcry.
c. Plausible since a heat sink is required for natural circulation and a concern in FRP-P.1 is that coid SI
flow could cause thermal shock of the reactor vessel, but core uneovery due to a loss of SI flow as
pressure increases will also reduce the SI flow that could c.ause thhemml shock.
d. Plausible since an increase in primary-to-secondary AP could result in a SGTR, but the concern is that
an increase in temperature and pressure could result in less SI flow and core uncovery.
DIFFICULTY ANALYSIS:
CI COMPREHENSIVE I ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the effect of delaying RCS bleed and feed during a loss of heat
sink
Harris NRC Written Iixamination
Senior Reactor Operator
QUESTION: 43
Given the following conditions:
The plant had been operating at 100% for three (3) weeks when a Reactor Trip
occurred.
- Six ( 6 ) hours following the trip, a reactor startup is planned.
Which one ofthe following is IKONIBITED at SIINPP as a result of industry wide
premature criticality events?
a. A difference of400 pcm between the IOUERTRAX and EXSPACK ECCs
b. Operators performing the EXSPACK estimated critical conditions (ECC)
c. Ilelnying the startup until xenon begins to decay
d. A stcutuprate in excess of - 1 0.3 dpm
ANSWER:
a. h difference of 400 pcm between the IOWERIXAX and EXSPACK ECCs
Harris NKC Written Examination
Senior Reactor Operator
Data Sheets
QCESTION NUMBER 43 TIEWGROUP: 3
10CFR55 CONTENT: 41(h) None 43(b) None
KA: 2.2.1
Ability to perform pre-startup procedures for the facility, including operating those controls associated
with plant equipment that could affect reactivity
OBJECTIVE: GP-3.4-6
SIMMARIZE at least three conditions which have contributed to prematnre criticality events within the
industry; also SI!M?r4ARI%Iiactions taken at SHNPF to prevent similar occurrences
DEVELOPMENT REFERENCES: GP-004 pg 10
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: GP-3.4 01 1
NRC EXAM HISTORY: None
DISTKACTOR .TUSTIFICACTION (CORRECT ANSWER Xd):
X a. The threshold for performing a reactor startup following a power history of 3 0 % equilibrium power
is 250 pcm difference between POWEKTRAX and EXSPACK and 500 pcrn for lrmsient history and
steady state below 80%.
h. Plausible since SI1NPP required any manual ECC calculations he performed by Reactor Engineering,
but IZSPACK is normally performed by Operations.
c. Plausible since xenon decay will be adding positive reactivity to the core white the startup is being
performed, hut is accounted for in the time after trip in the ECC.
d. Plausible since excessive startup rates can contribute to lack of reactivity control, but limitations are
placed on startup rate after criticality is achieved.
DIFFICU1,TY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Know-ledge of the administrative requircrnents prior to criticality be.ing
achieved
Harris NRC Written Examination
Senior Reactor Operator
QIJESTION: 44
Given the following conditions:
The plant was operating at 80% power.
Actions of AOP-010. Feedwater Malfunctions. due to a trip of Main Feedwater
Pump A.
- The crew is using transient annunciator response.
Which of the following annunciators is the lJnit-SCO required to be informed of in
accordance with OMM-001, Conduct of Operations?
21. ALB-05-7-4, CCW PLJMP A TRIP OR CIXHE CKT TROIJBLE
b. ALB-06-1-1, CIIlZKGlNG PUMP DISCHARGE HEADER 11/ I,O FLOW
c. CTM1-4-2, C I X i TWR M-l! PUMP 1 TRIP OR START FAIL
d. AIB-23-2-11. STEAM TUNNIL IIIGK TEMP
ANSWER
a. ALB-05-7-4, CCW PUMP A TRIP OK CLOSE CKT TROIJBLE
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 44 TIENGROUP: 3
KAIMPORTANCE: KO 3.3 SRO
10CFR55 CONTENT: 41(b) 10 Wb)
KA: 2.4.11
Knowledge of annunciators alarms and indications, arid use of the response instructions
OBJECTIVE: PP-2.0-R3
DISCUSS the requirements in OMM-001!AP-002/AP-l00 concerning the following:
h. MCR annunciators
DEVELOPMEXI REFERENCES: OMM-001 pg 10
REFERENCES SUPPHED TO APPLICANT: Xone
QIJESTION SOURCE: X NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
X a. Required to be informed of this annunciator due to a required entry into an additional AOP.
b. Plausible since this could indicate a Ieak in the KCS,but no AOI entry conditions are met.
E. Plausible since this could indicate a loss of CW cooling flow, but no AOP entry conditions are met
d. Plausible since this could indicate a steam leak, but no AOP entry conditions are met.
DIFFICULTY ANALYSIS:
-
COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY HATING: 3
EXPLANATION: Analysis of relative importance and requirements to prioritize annunciators
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 45
Given the following conditions:
A Reactor Trip occurred from 100% power.
a The plant stabilized at 557 "F for several minutes.
o Shortly thereafter, a Safety Injection signal actuated.
Which of the follouring describes the effect of this sequence on the Main Feedwater
System?
a. After the Reactor Trip occurred, the SGs could be fed using the Feedwater Reg
Uppass Valves
After the SI occurred, the SGs could be fed using the Fccdw&x Reg Bypass
Valves
b. a After the Reactor Trip occurred, the SGs could he fed using the Main
Feedwater Reg Valves or the Feedwater Reg Bypass Valves
o After the SI occurred. Main Feedwater could NOT be used to feed the SGs
c. After the Reactor Trip occurred, the SGs could be fed using the Feedwater Reg
Bypass Valves
After the SI occurred, Main Feedwater could N o r be used to leed the SGs
d. m After the Reactor Trip occurrd, the SGs could be fed using the Main
Feedwater Reg Valves or the Feedwater Reg Bypass Valves
After the SI occurred, the SGs could be fed using t l ~ eFeedwater Reg Bypass
Valves
ANSWER:
c. After the Reactor Trip occurred, the SGs could be fed using the Feedwater Reg
Bypass Valves
IIarris NRC' Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER: 45 TIEWGROUP: 2i 1
10CFR55 CONTENT: 41(b) 7 43@)
KA: 059K4.19
Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: Automatic
OB.W,CTIVE: AI:W-3.0-A6
EXPI.AIN the response of major CFW System valves to the following signalsiconditions
Main Feedwater Isolation Signal (MFIS)
Reactor trip (P-4)coincident with low T , , (< 564°F)
DE\'ELOPMENT REFI5RENCE.S: SD-103pg 26
REFERENCES SUPPLIED TO APPLICANT: None
QKJESTION SOURCE: NE\V SIGNIFICANTLY RIOUIFIED DIRECT
BANK NUAMBERFOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER X'd):
a. PIansible since on a reactor trip with low 'lave (564 OF), the SCis can still be fed with the bypass
valves, but on an SI or high-high SG level MFW can no longer supply the SGs.
b. Plausible since the SGs can n o longer be fed using MFW on an SI, but on a reactor trip only the
bypass valves can be used to feed the SGs.
X c. On a reactor trip with low Tave (564 OF), the SGs can stili be 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 "F), the SGs can still be fed with the bypass
valves, bid not the main feed reg valves, and on an SI or high-high SG level MFW c.an no longer
supply the SGs.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLE1)GE I RECAL.L
DIFFICIJLTY RATING: 3
EXPLANATION: Comprehension that on a reactor trip where the plant stabilizes at no-load
temperature, the P-4 with Low 'l'irve signal allows feeding with the bypass and
SI isolates all MFU'
Harris NRC Written I:sdinination
Senior Redctor Operator
QUESTION: 46
Which ofthe following describes the design of Phase A and a Phase B Containment
Isolation signals?
a. Phase A limits radioactive releases following a LOCA
a Phase H limits radioactive releases following a LOCA or secondary
system break inside Containment
b. Phase A limits radioactive releases AND minimizes Containment
overpressurization following a LOCA
e Phase H limits radioactive releases minimizes Containment
overpressurization following a LOCA or secondary system break inside
Containment
c. * Phase A limits radioactive releases following a LOCA
e Phase B limits radioactive releases following a LOCA AN2 prevents an
excessive KCS cooldown following a secondary system break inside
Containment
d. a Phase A limits radioactive releases minimizes Containment
overpressuriiation following a LOCA
Phase I3 limits radioactive releases following a LOCA prevents an
excessive RCS cooldown foliowing a secondary system break inside
Containment
ANSWER
a. Phase A limits radioactive releases following a LOCX
e Phase B limits radioactive releases following a LOCA or secondary
system break inside Containment
Harris NRC Written Examination
Senior Reactor Operator
Daa Sheets
QUESTION NUMBER 46 TIEWGROUP: 11
10CFRS5 CONTENT 41(b) 5/10 43(h)
KA: 00001 1EK3.06
Knowledge of the reasons for the following responses as the apply to the Large Break LOCA: Actuation
of Phase A and B during I,OCA initiation
OBJECTIVE: CIS-3.0-1
STATE the purpose of the Containment Isolation System
DEVE1,OPMENT REFERENCES: SD-114pg 4-5
REFERENCES SIJPPLIED TO APPLICANT: None
QUESTION SOURCE: SEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CIS 006:
CIS 009
NRC EXAM HISTORY: None
DISTR4CTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. Phase A serves to limit the release of radioactive materials to atmosphere following a L.OCA. Phase B
ac.ts to limit radioac.tive releases by actuating on a I,OCA or a steam or feedwater line break inside
containment.
b. Plausible since both Phase A and Phase B ac.t to limit the release ofradioactive materials to
atmosphere, but overpressurization is limited by spray actuation, main steam line isolation. and feed
water isolation.
c. Plausible since both Phase A and Phase B act to Limit the release of radioactive materials to
atmosphere, but overpressurization and RCS cooldowns are Limited hy spray actuation, main steam
line isolation, and feed water isolation.
d. Plausihle sinc.e.both Phase A and Phase 3 act to litnit the release of radioactive materials to
atmosphere, but overpressurization and RCS cooldowns are limited by spray actuation, main stcam
line isolation, and feed water isolation.
DIPFICIJLTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of purpose of Phase A and Phase B signals
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 47
An entry into FRP-S.1, Response to NucIear Power GeneratiodATWS, has been made
from PAIH-1. The following conditions currently exist:
- The reactor trip breakers are closed.
Rods are being inserted manually.
Control Bank D is at 12 steps.
- Power Knnge Instruments are all indicating 8%
- Intermediate Range SIJR is NEGATIVE
Which of the following conditions is required by FRP-S.1 to allow a return to PATH-17
a. One of the. reactor trip breakers must be opened
b. Both ofthe reactor trip breakers must be opened
c. Power Rangc indication must be reduced below 5%
d. Control Bank A must be inserted fully
ANSWER:
c. Power Range indication must be reduced below 5%
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NI:MBEW: 47 TIEWGROUP: lil
ICAIMPORFANCE: RO 4.4 SKO
10CFR55 CONTENT: 41(b) Xone 43(b) 5
KA: 000029CA2.01
Ability to determine OK interpret the following as they apply to a A'I'WS: Reactor nuclear instrumentation
OBJECTIVE: 3.1-3
DEMCINSTRAI'Ethe below-assumed operator knowledge from the SIINPP Step Deviation Documents
and WOG ERGS that support performance of EOP ac.tions:
a. Ve,rification of reactor trip
DEVELQPME.NTREFEMENCES: FRP-S.1, pg 14
REFERENCES SIJPPLIED TO APPLICANT: None
QITESTION SOCKCE: NEW SIGNIFICANTLY MODIFIED 0 DTRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3,15-R5 002
NRC EXAM HISTORY: None
DISTRACTOR JUS'IIFICACTION (CORRECT ANSWER X'd):
a. Plausible since this would cause the reactor to be hipped, hut it is not required to he done to exit FKP-
s.1.
h. Plausible since this would cause the reactor to be tripped, hut it i s not required to he done to exit FRP-
S.1.
X c. Exiting FRP-S. 1 requires that PR NIS be less than 5% and IR NIS startup rate he 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 he adequately shutdown, hut it is not required to he
done to exit FRP-S.1.
DIFFICULTY ANALYSIS:
CXHWPKEWENSIVE / ANALYSIS KNOWLEDGE I RECAI,L
DIFFICULTY RATING: 3
KXF'LANATION: Knowledge of the procedural requirements to exit FRP-S. 1
Harris NRC Written Examination
Senior Reactor Operator
QUESTIOK: 48
Given the following conditions:
A plant cooldown is being performed.
a All Steam Generators (SGs) are currently at approximately 50 psig.
Auxiliary Feed Water (AFW) Pump A-SA is being used to feed the SGs.
a The supply breaker on 120 VAC IDP- I A-S 1 [or 1AF-19. A1JX FW MOTOR P34P
A-SA UISCIIARW VLV, trips open.
Assuming NO operator actions, which of the following describes the effect of this loss of
power on the operation of AFW Punip A-SA?
a. Operates at shutoff head
b. 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 Ewarnin8tion
Senior Keactor Operator
Data Sheets
QUESTION NUMBER: 48 TIEWGROUP: ?/I
10CFR55 CONTENT: 41(b) 7 43W
K4: 061K6.01
Knowledge of the effcct of a loss or malfunction of the following will have on the AFW wmponents:
Controllers and positioners
OBJECTIVE: AFS-3.0-R5
1XSC:KIBE how the AFW- system is impacted by a loss of 120vac uninterruptible power supplies (SI, SIL,
SIII, SIV)
DEVELOPMENT RE.FERENCES: 33-137, pg 8-9
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTION SOIJRCE: NEW SIGXIFICANTLY MODIFIED DIRECT
BANK NUMRKR FOR SIGNIFICANTLY MODIFIED I DIRECT: AFS-A3 001
AI:S-A3 007
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since paver is lost to the discharge valve, but the valve fails open causing flow to increase.
b. Plausible since power is lost to the discharge valve, but the valve fails open causing flow to increase.
c. Plausible since the valve fails open and flow increases, hut the pump docs not run on recirculation
flow.
X d. The loss of power causes AFW Pump A-SA to reach runoot conditions due to 1AF-19 failing open
and having the SGs at such a low pressure.
DIFHCIJLTY ANALYSIS:
COMPREHE:NSIVE I ANALYSIS KNOWLEDGE I RECALL
DIPFICULTY RATISG: 3
F:XPLANATION Analysis oftha effect of a failure ofthe PCV 3fter determining the fail position
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 49
Given the following conditions:
The plant is in Mode 3 during a dilution ofthe KCS to the required boron
concentration.
A batch liquid release from the Secondary Waste Sample Tank (SWMT) 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,B-004-2-2, REFUELING WATER STORAGE LOW LEVEL, a l a ~ ~ ~ ~ s .
- KU'ST level is at 94% and slowly decreasing.
b. e ALB-019-1-4, IIOTWE1,I. HIGII-LOW I,VEL, alarms.
- Hotwell level is at 14% and slowly decreasing.
c. o An A 0 reports a leak in the NSW System inside the Turbine Building.
FI-9301.1, NSW Discharge Flow, indicates high.
d. ALB-005-6-1, CCU' SURGE TAXK HIGH-LOW LEVEL., alarms.
- CCU' Surge Tank level is 39% and slowly decreasing.
ANSWER:
a. e AI,I3-004-2-2, KTFIJELING WATER S'WKAGE LOW I.EVEL, alarms.
- KWST level is at 94% and slowly decreasing.
Harris NRC Written Examination
Senior Reactor Operator
I h t a Sheets
QhIESTION NUMBER 49 TIEWGROW: i !2
10CFR55 CONTENT: 41(b) 10 43(b)
KA: 000059(.;?.4.4
Ability to recognize abnormal indications for system operating parameters which am entry-levcl
conditions foe emergency and abnormal operating procedures. (hccidental Liquid Radwaste Release)
0BJE.CTIVE: ,4C)P-3.8
lDENTIFY symptoms that require entry into AOP-008, Accidental Release ofI,iquid Waste
DEVELOPMENT REFERENCES: AOP-008, p 3
AOP-022, p 3
ALB-005, p 39
REFERENCES SIJPPIJED TO APPLICANT: None
QUESTION SOURCE: 0 NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: AOP-3.8 001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
X a. Under these conditions no water should be taken out ofthe RWSF, so the decreasing level and alarm
will require entry into AOP-008.
b. Plausible since water i7 being released to the Tmbine Buildins. but actions are taken per AOP-010 to
address this
c. Plausible since water is being 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. hut actions taken in response tu a CCW leak
are per AOP-014.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIPFICC1,TY RATING: 2
EXFLAXATION: Knowledge of entry requirements for accidental liquid reiease
Harris NRC Written Examination
Senior Reactor Operator
QKESTION: SO
Which of the following actions would he most effective in responding to a Pressurized
Thermal Shack condition in accordance with FRP-1. 1, Response to Pressurized Thermal
Shack?
a. From the MCR, close the block valve for any open PRZ FORV
b. From the MCB, isolate any stuck open steam dunp valve
c. Direct an operator to the steam tunnel to locally isolate any stuck open Sci PORV
d. Direct an operator to the steam tunnel to locally isohte any stuck open MSIV
ANSWER:
c. Direct an operator to the steam tunnel to locally isolate any stuck open SG PORV
Harris NKC Written Examinatim
Senior Reactor Operatm
Data Sheets
QUESTION NUMBER: 50 TIEWGROUP: 112
10CFR55 CONTENT: 41(b) 7 43@)
KA: WEOSG2.1.30
Ability to locate and operate components, including local controls. (Pressuri7rd Thermal Shock)
OBJECTIVE: 3.14-1
DESCRIBE the purpose of the following EOPs including the type of event for which they were designed
and thc major actions performed
FW-P. 1, Response to Imminent Pressurized Thermal Shock
DEVELOPMENT REFERENCES: FRP-P. 1, pg 6
REFERENCES SUPPLIED TO APPLICANT: Nons
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC E.XAM HISTORY None
DISTRACTOR JZISTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since closing the block valve for a stuck open IRZ PORV is an action taken in FRP-3.1,
though it is pefformed to maintain RCS inventory and will cause pressure to increase which would
cause the severity of a PTS event to worsen.
b. Plausible since a stuck o p m stcam dump valve would contribute to the cooldown associated with a
PIS event, but individual steam dump valves cannot be operated from the MCD.
X E. A stuck open SG PORV would contribute to the cooldown associated with a PTS event. Locally
isolating the SG PORV would stop any cooldown caused by the SG PORV.
d. Plausible since localiy closing a stuck open MSIV would assist in terminating a cooldown, hut the
MSIV is located in the RAD and not the steam tunnel.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICGLTY RATING: 3
EWLAVATION: Analysis of plant conditions during a PTS event to determine the most
appropriate course of action
Harris YRC Written Esmnination
Senior Reactor Operator
QUESTION: 51
Given the following conditions:
e RIIR Pump 1A-SA is operating during a plant heat up
e 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 affected hp the loss of the control power fuses?
a. e Main Control Board red / green running indications will be lost
e The hre,aker will trip
e Lmal open / closed light indication and local breaker control will be lost until
control power is restored
h. e Main Control Hoard red / green running indications will be lost
e The breaker remains closed
e 1,ocal open / closed light indication will be lost, hut local breaker control is
possible without the control power
c. e Main Control Hoard red / green running indications will be available
- The breaker will trip
e Local open /closed light indication is available, but local breaker control is
possible without the control power
d. e hfaia Control Board red / green running indications will be available
e The breaker remains closed
- 1,ocal open / closed light indication is available, but local breaker control will
he lost until control power is restored
ANSWER:
b. Main Control Board red / grmn mming indications will be lost
- The breaker remains closed
e Local open 1 closed light indication will be lost. but local breaker control is
possible without the control power
flarris NRC Written Examination
Senior Reactor Operator
DaCd Sheets
QUESTION NUMBER: 5 1 TIERIGROW': 2/1
10CFH55 CONTENT: 41(b) 7 43(b)
KA: 062A4.04
Ability to manually operate and/or monitor in the control room: Local operation of breakers
OBJECTIVE: 48OV-3.0-Kl
State the function of breaker control pow'er and discuss the effects of a loss of breaker control power
DEVELOPMENT REFERENCES: OP-156.02, p 10: 61
4.8OV-LP-3.0.p I 1
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTION SOURCE: 1 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NU,MBER FOR SIGNIFICANTLY MODIFIED /DIRECT: 480V-Rl OOl
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since MCB and local indication will be lost, hut the breaker will not trip open on the loss of
control powe.r and local breaker control is still possible.
X b. A loss of control power will cause MCB and local indication to gn out. but the breaker remains closed
and local breaker control is srill possible.
c. Plausible since local breaker operation is still available, but the breaker will not trip and MCB and
local indication will be lost.
d. Plausible since the breaker remains c.losed, but the loss of control power will result in a loss of MCD
and local indication and the breaker can still be locally operated.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE 1RECALL
DIFFICULTY RATING: 3
EXPLANATION: hiowledge ofthe effect o f a loss of control power to a 480V breaker
ihmis NRC Written Examination
Senior Reactor Operator
QUESTION: 52
Which of the following situations would result in an inadvertent dilution 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. o RCP thermal barrier heat exchanger leak
AOP-016, Excessive Primary Plant Leakage
b. A tube leak in the CVCS Letdown heat exchanger
0 AOP-014. Loss of Component Cooling Water
c. A mixed bed demineralizer that was last in service three weeks ago is
mistakenly placed in service at the end-of-cycle
e ,401433. Chemistry Out of Tolerance
d. 0 A tube ieak in the Seal Water heat exchanger
0 AOP-014, 1,oss of Ccmponent Cooling Water
ANSWER.
d. o A tube leak in the Seal Water heat exchanger
AOP-014, Loss of Component Cooling Water
Harris NRC Written Excamination
Senior Reactor Operator
Ihta Sheets
QIJESTION NUMBER: 52 TIE:R/GROUP: 2/ 1
lOCFR55 CONTENT: 41(b) 5 43@)
KA: 004A2.06
Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) bmed
on those pre.dictions, use procedures to correct, control, or mitigate the consequences of those
malfunctions or operations: Inadvertent boration!dilution
OBJECTIVE: IE-3.32-3
Identify sysrems whose operation may alter RCS boron concentration and discuss how operation of these
systems may affect bnron concentration
DEVELOPMENT REFERENCES: SOER 94-2,p i 1-12
AOP-014, p 3,20
AOP- 14-BD, p 20
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOCRCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFlED / DIRECT: IE-3.12-R3001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Ilausible since the thermal barrier interfaces with a non-borated system (CCW), but leakage would be
out oftlie RCS to CCW and would not affect RCS boron concentration.
b. Ilausible since CCW cools the heat exchanger and would dilute the RCS if leakage from CCW were
to occiir, but letdown is at a higher pressure than CCW.
c. Piausible since boron concentration will change in CVCS, but this would result in an inadvertent
boration rather than a dilution.
X d. A seal water IIX leak wrill re.sult in CVCS being diluted by CCW. This failure is to be addressed by
DIFFICULTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS 0 KNOWL.EDGE/ RECALL
DIFFICULTY RATING: 3
EXPLANATION: .4nalyze the effect of each failure on RCS boron concentration and determine
the required procedure to address the failure
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 53
Given the following conditions:
The plant is in Mode 4.
Thc KCS in a solid plant condition.
IPHR pump 1A-SA is in service.
In accordance with G1-007, Normal Plant Cooldown, which of the following actions
should be taken to raise PKZ pressure to a new steady-state value?
a. Throttle 1CS-28, IiC-142.1 KHR LETDOWV. in the shut direction
b. Shut ICs-4, 45 GPM 1,ETDOWN ORIFICE A
c. Adjust the setpoint for 1CS-38. PI(-145.1 LTDN PRESSIJRk, to cause the valve
to bo in the shut direction
d. Adjust the setpoint for 1CS-231, FK-122.1 CHARGING FLOW, to cause the
valve to go in the open direction
ANSWER:
c. Adjust the setpoint for 1CS-38, PK-145.1 I.TDN PRESSURE, to cause the valve
to go in the shut direction
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QIJESTION NUMBER: 53 TIBWGROIJP: 2ii
IC1 IMPORTANCE: KO 2.3 SRO
10CFR55 CONTENT: 41@) 2-9 43w
KA: 010Ki.06
Knowledge of the physical connections andior c.ause-effect relationships between the PZR PCS and the
following systems: C.VCS
OBJECTIVE: GP-3.7-2
With regard to KCS cooldown, DESCRIBE the following per GP-007
e Thc two methods used to control RCS pressure, including the elements of each
DE.VELOPMENTREFERENCES: GP-007, p 41
REFERENCES SUPPLIED TO APPLICANT:
QUESTION SOCKCE: NEW a None
SIGNIFICANTLY MODIFIED
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Harris IOCT 5S7
DIHECT
NRC EXAM HISTORY: None
DISTRACTOK JIISTEFICACTION (CORRECT ANSWER Xdf:
a. Plausible since this would cause an increme in RCS pressure, hut ICs-38 will respond to cause
pressure to lower again.
b. Plausible since this would cause an increase in RCS pressure, but 1CS-38 will respond to cause
pressure to lower again.
X c. Adjusting the setpoint of ICs-38 will 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 ICs-38 will respond to cause
pressure to lower again.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLE.DGE/RECALL
DIFFICULTY RATING: 3
EXPLANATION: Comprehension of the effects of adjusting CVCS components on PIC?. pressure
Harris NRC Written Examination
Senior Reactor 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 />.
If DC load shedding is performed such that the loading on the battery is reduced from
292 amps to 146 amps, how long should the batter). be available to supply the remaining
loads?
a. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
b. More than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, but less 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 />
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QTJESTION NUMBER: 54 TIEREROD: 21I
1OCFR55 CONTENT: 41(b) 5 43@)
KA: 063A1.01
Ability to predict anhor 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
STA 1 I-, the function and EXPLAIN the bask operalion of the following major components of the DC
Power System:
Batteries
DEVELOPMENT REFERENCES: EFF-001. p 55
ADEL-LP-2.6, p 3
DCF-LP-3.0, p 8
REFERENCES SWPLLED TO APPLICANT: None
QUESTIOX SOURCE: X NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT New
NHC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Ilau~ihlesince 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 293 amps per
hour and decreasing the discharge rate would increase the capacity.
b. Ilausiblc since the discharge rate has been decreased which would extend the capacity of the batteq
foi a period of time, but the time would be more than doubled.
E. Plausible since the dischargc rate has been halved, so it would appear that the capacity would he
douhled, hut it is a non-linear relationship.
X d. Reducing the discharge rate on a hattery increases the battery capacity in a non-linear function such
that decreasing the discharge rate by half, increases the capacity by more than douhle.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS 0 KNOWLEDGE / RECALL
DIFFICULTY RATING: 4
EXPLANATION: Calculation of the nominal disch ge rate of a hatter). and comprehension ofthe
effect of reducing discharge rate on battery capacity
IIarris NRC Written Examination
Senior Reactor Operator
QUESTION: 55
Given the following conditions:
o The plant has experienced ii Large Break Loss of Coolant Accident during a rcactor
startup.
- All equipment functioned as designed and the crew has reached the point in PA?-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 haw on the Source Range instrumentation and procedure
used to mitigate these plant conditions?
a. m The displacement of downcomber water would increase the neutron leakage
and result in a higher source range count rate.
The crew should continue in PATH-1 rather thm transition to FRP-S.2,
Response to I.oss of Core Shutdown.
b. A decrease in downcomber water density would reduce fission and result in a
lower source range count rate.
The crew should transition to FRP-S.2, Response to Loss of Core Shutdown,
rather than continue in PAIE-1.
c. The displacement of boron from the downcomher region would increase
fission and result in a higher source range count rate.
The crew should continue in PAIH-1 rather than transition to FRP-S.2,
Response to Loss of Core Shutdown.
d. 4 decrease in downcomber water density would reduce fission and result in a
lower source range count ratc.
The crew should continue in PATII-1 rather than trinsition to FRP-S.2,
Response to Loss of Core Shutdown.
ANSWER.
a. o The displacement of downcomber water would increase the neutron leakage
and result in a higher source range count rate.
The crew should continue in PATH-I rather than transition to FW-S.2,
Response to I m s of Core Shutdown.
Harris NKC Written Examination
Senior Reactor Operator
Data Sheets
Q7JESTION NUMBER: 55 TIERKROIE 212
1QCFR55COhTENT: 41(b) 5 $30~)
MA: 015A2.05
Ability to (a) predict the impacts of the following malfunctions or operations on the XIS; and (b based on
those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions
or operations: Core void foimation
OBJECTIVE: 1311-3.10-7
Explain the NIS response to different void fractions in the core and downcomer region
DEVELOPMENT REFERENCES: 110-31)-3.10 pg 26-27
REFERENCES SUPPL1E.DTO APPLICANT: None
QEESTION SOURCE: 0NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: INPO 20608
NRC EXAM IIISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. Downcomber voiding results in higher source range indication due to increased leakage. 'The crew
should continue in PATH-1 rather than transfer to FRP-S.2 since entry conditions to FRP-S.2 are a
Yellow path c.ondition.
b. Plausible sinc.e a severe decrease in c.orc water density would result in Less moderation and a lower
power level, but downcomber density has little effect on core reactivitp.
c. Plausible since displacing core boron would result in a higher power level, but downcomber density
has little effect on core reactivity.
d. Plausible since a scvere decrease in core w-ater density would result in Less moderation and a lower
power level hut domicomber density has little eff'ect on core reactivity.
DIFFICUL.TYANALYSIS:
COMPREHENSIVE / ANALYSIS
u ~~
U
DIFFICULTY RATING: 3
EXPLANATION: Analysis of the effects of core voiding on SR indication and knowledge of the
procedure hierarchy during the performance of the EOl's
Hams NRC Written Emminution
Senior Reactor Operator
QUESTION: 56
Given the following conditions:
A transition has just been made to FRP-S. 1, Response to Nuclear Power Generation
ATWS, from PATII-1.
- The Reactor Operator is manually inserting control rods.
- All Turbine Throttle Valve (TV) and Turbine Governor Valve (GV) indications show
the REI1 light OFF and the GREEN light ON, with the exception of TV-3 and GV-2
which have both the RED light and GREEN light ON.
Turbine speed is decreasing. and is currently 1680 rpm.
- The Main Steam Isolation Valve (MSIV) Bypass valves are closed.
Which ofthe following actions should be taken next?
a. Verify all AFW pumps running
b. Manually trip the Turbine from the MCR
c. Place both Turbine DEII pumps in PIJLL-?O-I,OCK
d. Shut all MSIVs
ANSWER:
b. Manually trip the Turbine from the MCB
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 56 TIEWGROLT: 2/2
10CFR55 CONTENT: 41(b) 7 43w
KA: 04SA4.06
AhiIity to manually operate andior monitor in the control room: Turbine stop valves
OBJECTIVE: 3.15-4
Given the following E,OP steps, notes, and cautiom, DESCRIBE the associated basis
Order of preference for turbine trip steps from the MCB
DEVELOPMENT REFERENCES: FRP-S.1 pg 4
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.15-K2001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Ptausihlc since GV-2 and TV-3 are associated with opposite steam chests and it may he awinied that
as long as the GVs are closed for 1 steam chest and the TVs arc closed for the other steam chest with
turbine speed decrea\ing. and starting AFW is the next step in the procedure, however the turbine
should not be considered to he tripped.
X b. Verification of a turhine trip requires either all 4 TVs he c l o d or all 4 GVs he closed. If one set of
these valves are not all closed, then the RNO directs manually tripping the turbine from the MCB.
c. Plausible since the turbine should not be considered to he tripped based on indications, and this is an
RNO action, hur should not be performed until a manual trip from the MCB is attempted.
d. Plausible since the turbine should not be considered to he tripped based on indications, and this is an
RKO action, hut should not be performed until a manual trip from the MCB is attempted.
DIFFICULTY ANALYSIS:
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
IIarris NRC Written Examination
Senior Reactor Operator
QUESTION: 57
Given the folIowing conditions:
E The Main Control Room has been evacuated and control traisferrcd to the Auxiliary
Control Panel (ACP).
- AOP-004, Remote Shutdown, is being performed when a loss of offsite power
coincident with a Safety Injection signal occur.
Shiclr of the following describes the response of the piant?
a. The Emergency Diesel Generators automatically start and the sequencers load the
EDGs due to the undervoltage signal
b. The Emergency Diesel Generators automatieally start and the sequencers load the
EDGs due to the safety injection signal
c. The Iimergency Diesel Generators automatically start. but must he nianually
loaded with the required loads
d. The Emergency D i e d Generators must be manually started and manually loaded
with the required Ioads
ANSWER
a. The Emergency Diesel Generators automatically start and the sequencers Ioad the
EDGs due lo the undervoltage signal
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 54 TIEWGROUP: 2/1
KAIMPORTANCE: KO 3.6 SRO
IOCFR55 CONTENT: 41(b) 7 43(M
KA: 064.43.07
Ability to monitor automatic operation of the liI)/G system, including: Load sequencing
OBJECTIVE: AOP-3.4-R5
DISCITSS how a transfer to the auxiliary control panel would affect the following inputs to the FSF
sequencers
- Safety injection signal
- Safety bus undervoltagc signal
DEVELOPMENT REFERENCES: AOF-004 pg 91
AOP-004-BD pg 26
SIP-155.02 pg 6-9
REFERENCES SUPPLIED TO APPLICANT: None
0 0
~~
Q7JESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: AOP-3.4-R6 001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. The EDGs should automatically start on the UV condition and the UV signal will still cause the
sequencer to operate. Only the SIAS input to the sequencer is defeated upon transfer to the ACP
b. Plausible since the EDG will automatically start, hut loading will be based upon the UV signal.
c. Plausible since the EDC; will automatically start, but loading will be based upon the IJV signal.
d. Plausible since many automaric functions are defeated when control is transferred to the ACP, but the
EDG will automatically start and loading will he based upon the UV signal.
DIFFICULTY ANALYSIS:
COMF'REHENSIVE / ANALYSIS 0 KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of the effect of B transfer to the ACF on the EIPG and sequencer
operation
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 58
Given the following conditions:
- The unit is operating at 100% power.
- Following maintenance on IA-SA Emergency Diesel Generator (EDG), it is
determined that a cominnn mode failure exists which renders both RDGs inoperable.
Which ofthe following actions are required to be taken within one (1) hour of declaring
both EDGs inoperable?
a. Verify and recover required functions
b. Restore one (1) of the ELlGs to operable status
c. Verif>roff site power availability
d. Initiate actions to place the unit in Not Standby
ANSWER:
c. Verify off site power availability
Harris NRC Written Examination
Senior Reactor Operator
Sheets
QUESTION NUMBER: 58 TIFWGROUP: 3
10CFR55 CONTENT: $I(b) None 43(b) 2
- A: 2.2.24
Ability to analyze the affect of maintenance activities on LCO status
OBJECTIVE: DE-3.0-20
Given a plant mode of operation and the applicable 1,CO-related parameters for an EDG, DETERMINE
if n Technical Specification one-hour (or less) action statement applies
DEVELOPMENT REFERENCES: TS 3.8.1. l , pg 3 4 8-3
OST-1023, pg 1-2
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIPIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIKF.CT: Harris LOCT 385
NRC EXAM HISTORY: None
DISTRACTOR JTJSTIFICACTION(CORRECT ANSWER X'd):
a. Plausible since it is an ac.tion for the EDG operability, however it is not a requirement to verify nor
recove.r i n a I-hour time frame.
b. Plausible since restoration of one EDG to operable status is required, but it is required to be performed
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 be 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 E.DGs inoperable this is not required.
DIFFICULTY ANALYSIS:
0 COMPREHENSIVE / ANALYSIS ICVO\VI,EDGE / 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
Senior knctor Operator
QUESTION: 59
Given the following conditions:
The plant has experienced a small break I,OCA
Ihecrew has transitioned to EPP-009, Post LOCA Cooldown and
Depressurization.
a The ERFIS computer is failed.
a Containment pressure peaked at 8 psig, but is now 4.5 psig and decreasing slowly.
Present pressure indications are:
a PI-455.1, PRZ PKESSURE CH I = 800 psig
PI-456, PRZ PRESSIJRE CH I1 = 770 psig
a PI-457: PW, PWSSIJIW CII I11 = 740 psig
- PI-402.1, RCS WIDE RANGE PRESSURE = 840 psig
o PI-403; RCS WIDE RANGE PRESSURE = Failed High
Which of the following will be used to determine the primary plant pressure!
a. IJse PI-457 down to 1700 psig and use PI-402.1 below 1700 psig
1). Use 11-456 down to 1700 psig and use PI-402.1 below 1700 p i g
c. Use PI-455.1 down to 1700 psig and use PI-402.1 below 1700 psig
d. Use PI-402.1 at all pressures
ANSWER
d. Use PI-402.1 at all pressures
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER 59 TIEWGROUP: 3
KAIMPORTANCE: RO 3.5 SHO
10CFR55 CONTENT: 41(b) 6 43w
KA: 2.4.3
Ability to identify post-accident instrumentation
OBJECTIVE: 3. I9
DESCRIBE Control Room usage of EPPs, foldouts, and FRPs as it relates to the following:
g. Use of KCS wide-range pressure indication
DEVE1,OPMENT REFERENCES: EOP IJsers Guide pg 27,38
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOIJRCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK MITMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: Harris LOCI 846
NRC EXAM HISTORY None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since Pi-457 is the lowest reading ofthe pressures and would he the most conservative, hut
with adverse containment conditions the post-acc.ident instrument PI-402.1 is to he used.
b. Plausible since Pi-456 is the highest reading of the pressures and would likely provide the highest indiclition
until 1700 psig is reached, but with adverse containment conditions the post-accident instrument PI-402.1 is to
he used.
c. Plausible since PI-455 is the median reading ofthe 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 he used.
X d. Adverse containment conditions still exist so the post-accident instrument. PI-402.1 is to hc used at all
pressures.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of plant conditions and instrument failures to determine indications to
use during adverse containment
Harris NRC Written Kxamination
Senior Reactor Operator
QUESTION: 60
Assuming that all other equipment is operable, which of the following would require an
entry into Technical Specification 3.8.2.1, I>C Sources - Operating (Modes 1-4), action
statements?
a, EMERGEIWY BUS A-SA TO AUX BUS I) TIE BREAKER 105 SA trips open
and WIG 1A-SA automatically starts and Ioads
b. 480V EMERGENCY BUS lA3-SA main feeder breaker trips open
e. BATTERY CHARGER 1A-SA is placed under clearance
d. EMERGENCY BATTERY 1A-SA is placed on a float charge
ANSWER:
b. 4801 EMERGENCY BUS lA3-SA main feeder breaker trips open
Harris NKC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER: 60 TIEWGROUP: 21I
10CFR55 CONTEXT: 41(h) None 43(h) 213
KA: 00005SG2.1.33
Ability to recognize indications for system operating parameters which are entry-level conditions for
tcchnical specifications. (Loss of DC Power)
OBJECTIVE: IlCP-3.0-Rl
Given the name of a component in the DC power system, state whether or not that component is
'I'echnical Specification related
DEVELOPMENT REFERENCES: TS 3.8.2.1, p 314 8-12
SD-156,p 24
REFERENCES SWPLIED TO APPLICANT: None
QUESTION SOURCE: 0N E W 0
X SIGNIFICANTLY MODIFIED 0 DIRECT
BAXK XUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY None
I)ISTR4CTOR JIISTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since. the Program A sequencer (L,OSP) will strip some MCCs which suppIy 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 h. A loss of480V Emergency AC Bus 1.43-SA will result in a loss of both MCCs lA21-SA and 1A31-
SA, which would cnuse both A train battery chargers to be inoperable.
c. Plausible since removing a battery charger from service would result in a TS entry iftlie other charger
is also out of service, but a single charger will not result in an entry to an action statement.
d. Plausible since a float charge is a surveillance requirement and most surveillances m.&e the associated
equipment inoperable, but the normal configuration of the battery is on a float chargee.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIPFICUI,TY RATING: 3
EXPLANATION: Analysis of the effect of a loss of AC power requiring a 'IS entry for DC power
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 61
Given the following conditions:
The plant is operating at 100% power when ALB-010-1-1B. RCP A UPPER 011,
RSVK LOW-LEVEL, alann is received.
The operator checks the computer points for GI) AOP-O 18 and finds RCP A motor
thrust-bearing temperature at 195°F and RCP A upper radial bearing at 185F with
both slowly increasing.
Which of the following actions are required?
a. Stop RCP A and initiate a rapid plant shutdown in accordance with AOF-038,
Rapid Downpower
b. Manually trip the reactor and go to PATH-1, stopping RCP A as time permits
c. Continue monitoring RCP A temperatures, tripping the reactor and entering
PATII-1 if RCP A temperatures exceed 300°F
d. Stop RCP A, manually trip the reactor and go to PATII-1
AKSWER:
b. Manually trip the reactor and go to PATH-1, stopping RCP A as time permits
Harris NRC Written Examination
Senior Rcactor Operator
Data Sheets
QUESTION NCJMBER: 61 TIEWGROUP: lil
1QCFR55CONTENT: I l ( b ) None O(b) 5
KA: 000015/17AA2.08
Ability to determine and interpret the following as they apply to the Renctor Coolant Pump Malfunctions
(Loss of RC Flow): When to secure RCPs on high bearing temperature
OBJECTIVE: AOP-3.18-3
Given a set of plant conditions and a copy of AOP-018, DKIERMINE the appropriate respcnse
DEVELOPMENT REFERENCES: AOP-018 pg 21,27
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: AOP-3.18 01 9
NRC EXAM HISTORY None
DIS'I'RACTOR JCISTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since the RCP is to be stopped, but must be stopped immediately wshich requires that the
reactor be tripped.
X b. RCP motor temperatures require the pump be stopped. With power above 4S%, the reactor must be
tripped prior to tripping the KCP.
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. Plausible 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
DIFFICIJLTY RATING: 2
EXPLANATION: Knowledge of RCP motor temperature tripping requirements
Harris NRC Written ExatiLifiation
Senior Reactor Operator
QUESTION: 62
Given the following conditions:
Path-2 is being performed due to an SGTR.
The MSIV on the ruptured SG is mech'micaliy stuck open.
o The Main Steam Isolation Valves (MSIVs) on the intact SGs arc closed.
o The Condenser is available for Steam Dump operation.
A cooldown to 485 "I:from 554 @Fat the maxiinum rate is required.
Which of the following describes the method to accomplish this cooldown in accordance
with PATH-2 and the EOP 1Jser's Guide?
a. Fully open the Steam Dunips as fwt as possible
b. Fully open the Steam Dumps as fast as possible without causing a main steam line
isolation
c. Fully open the intact SCr PORVs as fast as possible
d. Fully open the intact SG 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
Senior Reactor Operator
Data Sheets
QUESTION NUXBER: 62 TIEWGROUP: lil
10CFR55 CONTENT: 4l(b) 7 43@)
KA: 000038EA1.36
Ability to operate and monitor the following as they apply to a SGTR Cooldown of RCS to specified
temperature
OBJECTIVE: 3.I9-K4
Given a set of conditions during EOP implementation, IXTERMINE the correct response or required
action based upon thc EC)P Ilsers Guide general information
Dumping steam at maximum rate
DEVELOPMENT REFERENCES: EOP Users Guide, p 38
PATH-2 Guide, p 8, 10
KEFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOIJRCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: EOP-3.19-K4 001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the maximum cooldown rate can be achieved using mwimum steam dump flow, but
causing too great a rate of pressure drop will result in the MSIVs going closed which is undesirable
and it is also undesirahle to use steam dumps when the ruptured SG MSIV is open.
b. Plausible since the mdximum cooldown rate is desirable using maximum 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 c. During a SGTR cooldow-n only the intad SGs should be used to cooldown the KCS and since tlie
MSIVs oil the intact SCis are closed, the POIiVs should he used. The valves should he opened as fast
as possible since generation of an MSIV signal is not a c.oncem.
d. Plausible since causing the MSIVs to close is not desirable when steam dumps are being used, but
when already using PORVs to dump steam this is not a concern.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KYOWLEDGE / RECALJ,
DIFFICULTY RATING: 3
EXPLANATION: Knowledge ofthe E01 L k r s Guide requirement for performing a maximum
rate cooldown
Harris NKC Written Examination
Senior Reactor Operator
QUESTION: 63
Given the following conditions:
Aficr transferring resin, it is noted that RM-lWR-3644A. SPENT RESIN PLJMP
1-4A, radiation monitor is indicating 10 mRem!hr.
The monitor is physically located 20 feet away from a suspected clog in tlie pipe
which is the source ofthe 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 will be hanging the clearance?
(ASSUME THE CLOG IN 'THE PIPE IS A POINT SOURCE)
a. 20mRemhr
b. 4 0 m R c m h
d. 160inRendhr
ANSWER:
d. 160 rnRemihr
Iarris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 63 TIENGROUP: 2! 1
10CFR55 CONTENT: 4l(b) 5 43(b)
IC4: 073K5.02
Knowledge of the operational implicstioiis they apply to c.oncepts as they apply to the PRM system:
Radiation intensity changes with source distance
OBJECTIVE: RP-3.5-2 I
Calculate dose rates at different distances from point sources and line. sources
DEVELOPMENT REFERENCES: W-LP-3.5 pg 22 and
Attachment i pg 7
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: H XEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible if the square. root of the distances is taken, instead of squared as they stouid be (IOmRihr x
2OIiLft = 20 mR/hr x 52 fi)~
b. Ilausible if the distances are not squared as they should be (lOmWhr x 20 ft :--40 - mRhr x 5 f t)
c. Plausible if a mathematical error is made (value selected as a distracter due to the progression of other
numbers in distracters).
X d. IJsing the formula Ild12==12d~, the intensity ofthe source at 5 fe.et is caiculated to be 160 mRem;hr.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE: / RECALL
DIFFICULTY RATING: 3
EWLANATION: Calculation of distance wing inverse square for radiation
I i m i s NRC Written Examination
Senior Reactor Operator
QUESTION: 64
Given the following conditions:
E The Control Room has been evacuated due to a fire.
0 AOP-004. Remote Shutdown. is being performed.
0 The crew has located the most recent OST-1036, Shutdown Margin Calculation,
and determined that 5,000 galions of boric acid must be added to the RCS.
- Boric Acid Tank level is 77%J.
What level will the Boric Acid Tank be at when the 5,000 gallons of boric acid are added
to the KCS AND why is there a concern about required shutdown margin dnring 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. Find 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 Tank 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 1 ank level should be approximately 56% to ensure adequate
shutdown margin is maintained in the event thdt a cooldown to Cold Shutdown
conditions is required
ANSWER
c. Final Boric Acid Tank level should be approximately 62% to enrue adequate
shutdown margin is maintained in the event that a cooldown to Cold Shutdown
conditions is required
Harris NKC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER. 64 TIERGROUP: 1i2
IOCFRSS CONTENT: 41(b) 5iI0 43(b)
KA: 000068AK3.13
Knowledge ofthe reasons for the following responses a\ they apply to the Control Room Evacuation.
Performing a shutdown margin calculation, including boron needed and boration time
OBJECTIVE:
< h e n a set of plant conditions and a copy of AOP-004. Remote Shutdown, DETERMINE the
appropriate course of action
DEVELOPMENT REFEJENCES: AOP-004-RD pg 47
Curve D-2
REFERENCE.$ SUPPLIED TO APPLICANT: Curve D-2
QKJESTION SOURCE: X NEW SIGNIFICANTLY MODIFIED DIKECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (COKRECT ANSWER Xd):
a. Plausible since the BAT level will be at 62% following the 5,000 galion addition, but shutdowsn
margin is a concern in the event of a cooldown.
b. Plausible since errors occur when the graph is read, but the BAT level will be at 62% and shutdown
margin is a concern in the event of a cooldown.
X c.. A boration is only performed in the went that a cooldown is required to bt: performed during the.
performance of AOP-004. [Jsing Curve 11-2,77% lcvcl corresponds to 27,000 gallons. Adding 5,000
gallons to the RCS will leave 22,000 gallons, which c.orrespondsto a BAT level of 62%.
d. Plausible since a boration is only performed in the event that a cooldown is required to be performed
during the performanc.e of AOP-004, but RAT level will indicate 62% and not 56%.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXI1,ANATION: Knowledge of thc reason for performing a boration while operating the plant
from the shutdown panel and the ability to apply a plant curve
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 65
Given the following conditions:
The reactor is critical at amps.
0 The Channel I inverter output breaker trips.
Which of the following occurs as a result of the breaker tripping?
a. Reactor power remains at 1W8amps and Power Range Channel N-42 deenergizes
b. Reactor power remain? at 10.' amps and Power Range Channel N-41 deenergizes
c. The reactor trips due to Intermediate Range Channel N-36 deencrgizing
d. The reactor trips due to Intermediate Range Channel N-35 deencrgizing
ANSWER
d. Thc reactor trips due to Intermediate Range Channel N-35 deeiicrgizing
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 65 TIE.R/GROUP: 211
10CFR55 CONTENT: 41(b) 7 1w
K4: 012K2.01
Knowledge of bus power supplies to the following: RPS channels, components, atid interconnections
OBJECTIVE: AOP-3 24-2
REC:OGNIZE automatic actions that are associated with loss of an instrument bus or loss of NNS UPS
DEVELOPMENT REFERENCES: AOP-024, p 23,25,29,34
REFEHENCES SUPPLIED TO APPLICANT: None
QGESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIKECT: Harris LOCI 457
NRC EXAM HISTORY None
DISTRACTOH JCISTIFPCACTION (CORRECT ANSWER X'd):
a. Plausible smce a loss ofpower would result in a loss ofPR Channel, but the trip occurs due to a loss
ofN-35
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 instrument bus I1 were lost, but the reactor
trips on a loss of instrument bus I due to il loss ofN-35.
X d. A reactor trip would occur duc to N-35 failing if instrument bus 1 being lost.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS 0 KNOWLEDGE I RECALL
DIFFICULTY RATING: 2
EXPLANATION: Analysis of the effect o f a loss of instrument bus power on plant conditions
Harris NRC Written Examination
Senior Keactor Operator
QUESTION: 66
Given the following conditions:
An earthquake has caused damage to the Main Reservoir dam.
- Main and Auxiliary Reservoir Ievels are both currentIy 240 feet and stable.
- AOP-022, Loss of Service Kater. is being performed for a I.oss of [Jkimate Went
Sink.
- Emergency Semice Water (ESW) pumps have been aligned to the Main Reservoir.
e One (1) Normal Service Water (NSW) pump is operating.
Which of the following pumps are required to be operating to provide water to the SSE
Iire Protecticm Header once the ESW header is aligned to the fire protection header?
a. ONLY an ESW pump
h. An ESW pump AND an FSW Booster pump
c. ONLY a second NS W pump
d. A second NSW pump AND an ESW Booster pump
ANSWER:
b. An ESW pump AND m ESW Booster pump
Harris NKC Written Examination
Senior Keactor Operator
Data Sheets
QUESTION NUMBER: 66 TIEIUGROUP: 2/1
10CFR55 CONTENT: 41(b) 2-9 43(W
KA: 076K1.15
Knowledge of the physical connections and/or cause-effect relationships between the SWS 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
OP-139 pg 27
W,FEKENCES SUPPLIED 'roAPPLICANT: NOIK
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: FP 020
NRC EXAM HISTORY Xone
I)ISTR4CTOR JLTSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since an ESU' piimp is started, but an FSW Booster pump is also required.
X b. An ESW pump, aligned to the Main Reservoir, is started, along with an ESW Roostcr pump to supply
the SSE fire protection header.
c. Plausible since the first NSW pump is not required to be tripped provided coding tower basin level is
adequatc and NSW 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 firs heade.r, but an ESW pump is
required to supply the booster pump.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS ICVO\VI,EI)GE /RECALI,
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the system alignments available to suppiy the fire header
Fiarris NRC Written Examination
Senior Reactor Operator
QUESTION: 67
Given the following conditions:
0 The plant is being cooled down to 140°F for maintenmce which will NOT require the
RCS be opened.
0 The crew is in the process of placing the first Residual Heat Removal (RIIR) train in
service for RCS cooling.
o Current boron concentrations are as follows:
0 KIIR (train to he placed in service) boron 1021 pprn
Required Shutdown Margin boron 1200 ppm
0 Cold Shutdown boron 1750 ppm
Refueling boron 2261 ppm
Bcfore the RIIR train can he placed in service for RCS cooling, RHK boron
concentration must be increased by a MIINIMCM o f . . .
a. 179 ppni.
b. 320ppm.
c. 729ppm.
d. 1240ppm.
ANSWER:
a. 179ppm.
Harris NKC Writteu Examination
Senior Reactor Operator
Data Sheets
QIJESTION NUMBER: 64 TIEWGROUP: 2/1
10CFR55 CONTENT: 41(b) 5 43w
KA: 005K5.09
Knowledge of the operational implications of the following concepts as they apply the RIIRS: Dilution
and boration considerations
OBJECTIVE: KHRS-2.0-12
APP1,Y precautions and limitations of OP-11 I, RHRS to IIypothetic.al System Configurations
DEVE.I,OPMENTREFERENCES: OP-I 11 pg 7
REFERENCES SUPPL.IED TO APPLICANT: Xone
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED nmEcT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. RHK boron must be greater than or equal to the required SIIM or the required refueling concentration.
The boron concentration requirements will he dependent on the inteuded use ofthe RHR System.
Using the R I R system for cooldown purposes requires that the boron concentration be greater than or
equal to the required shutdown margin.
b. Plausible since this is the difference between KHK and RCS boron concentration, but only the
c. Plausible since this is the difference between KHR and Cold Shutdown boron concentration, but only
the required SDM boron is needed.
d. Plausible since this is the difference between RHK and refueling boron concentration, and refueling
conditions occur at 140"F, but only the required SDh4 boron is needed.
DIFFICULTY ANALYSIS:
COMPREIIENSWE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATISG: 3
EXPLANATION: Application of actual versus required boron concentration - must determine
minimum limiting requirement
IIarris NRC Written Examination
Sciiior Reactor Operator
QUESTION: 68
Given the following conditions:
a A liquid waste discharge from a Treated 1,aundry and Hot Shower (TI.&HS) Tank is
in progress.
- REM-1 WL-3530, Treated Iaundr). and IIot Shower Tank Pump Discharge Monitor,
goes into high alarm.
Which of the following terminates the discharge?
a. The running TL&IIS Tank Pump \vi11 automatically trip
b. 3I.HS-301, Treated L&HS 1 b Discharge to Cooling Towcr Hlowrdoxm.will
automatically close
c. 3I.HS-293, Flow Control Valve Treated L&HS Tk to Enviro, wil1 automatically
close
d. 3LHS-396, TL&IIS Tank Pump Discharge Isolation Valve, will autoniaticaIly
close
ANSWER:
d. 3LhIS-396, IL&HS Tank Pump Discharge Isolation Valve, will automatically
close
Harris NKC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 68 TIERKROUP: 212
10CFRSS CONTENT: Il(b) 7 43m
KA: 06843.02
Abiliw to monitor automatic operation ofthe Liquid Radwraste System including: Automatic isolation
OBJECTIVE: I.WPS-LP-3.0-7
DESCRIBE the automatic protection features associated with discharges to the environment from the
LWPS
DEVELOPMENT REFERENCES: AOP-005, p 27-28
REFERENCES SUPPLIED TO APPLIC.ANT: None
QGESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: KMS-A6 005
NRC EXAM HISTORY: None
DISTRACTOR JZISTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the pump will stop the discharge, 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 signal.
E. Plausihlr since this valve is in the flow*path and will stop the discharge, but this valve does not
receive an automatic. closure signal.
X d. On a high rad level as sensed by REM 3540, the discharge isolation valve will automatically close,
terminating any release in progrcss.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of liquid radw-aste design and operation
Harris NKC Written Examination
Senior Reactor Operator
QITESTION: 69
Assuniing NO operator actions, which of the following describes the effect of a loss of
instrument air on Volume Control Tank (VCT) level'?
a. VCT level decreases due to maximum charging and letdown isolation valves
closing
b. VCT level decreases due to maximum charging and letdown being diverted to the
Hold Up Tank
c. VC'T level increases due to minimum charging and the letdown pressure control
valve failing open
d. VCT level increases due to minimum charging and the letdown orifice isolation
~ A v e failing
s open
ANSWER
a. VCT level decreases due to maximuni charging and letdown isolation valves
closing
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 69 TIEWGROUP: 211
10CFR55 CONTENT: 41(b) 7 43m
KA: 078K3.02
Knowledge of the 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: DEEFWIINE the appropriate response.
DEVELOPMENT REFERENCES: AOP-017 pg 37
REFERENCES SUPPLIED TO APPLICANT: Xone
QIJESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUAMBERFOR SIGNIFICANTLY MODIFIED /DIRECT: CVCS-R3 008
NRC EXAM HISTORY: None
DISTRACTOK JUSTIPICACTKON (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 dec.rease, but it will 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 control 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 / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze the response of CVCS aRer determining the fail position of various
CVC'S valves on a loss of IA
IImis NRC Written Examination
Senior Reador Opemtor
QUESTION: 70
Given the following conditions:
e Following a plant trip, EPP-004. Reactor Trip Response, is being performed.
e The crew is verifying Natural Circulation conditions as a result of a loss of power to
all RCPs.
e Five ( 5 ) core exit thennocouples are failed.
How do the failed core exit thermocouples affect indications used to verify Natural
Circulation?
a. e The Core Exit Temperature indications will be HIGHER than actual
KCS Subcooling will indicate MORE subcooling tl~anactual
b. The Core Exit Iemperature indications will be HIGHER than actual
KCS Subcooling wTill indicate LESS subcooling than actual
c. Core Exit Temnperatnre indications will indicate LOWER than actual
e RCS Subcooling will indicate MORE subcooling than actual
d. Core Exit Temperature indications urill indicate the SAME as actual
RCS Subcooling will indicate the SAME subcooring as actual
ANSWER:
d. e Core Exit Temperature indications will indicate the SAME as actual
e RCS Subcooling will indicate the SAME subcooling as actual
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION XUMBER: 70 TIEWGROW: 2/2
KAKWORTANCE: KO 3.5 sno
10CFR55 CONTENT: 41(b) 7 43(W
KA: 017K3.01
Knowledge of the effect that a loss or malfunction of the I l l 4 system will have on the following: Natural
circulation indications
OBJECTIVE: ICCM-3 .0-K6
DESCRIBE) how the plant's subcooling monitor information is processed
DEVELOPMENT REFERENCES: SI)-106 pg 5, 14
ICCM-LP-3 .O pg 1 1, 14- 15
REFERENCES SUPPLIED TO APPLICAVT: None
QUESTION SOURCE: 0X NEW SIGNIFICANTLY MODIFIED 0 DPRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / 1)IRFCT: Sew
NRC EXAM HISTORY ?;one
DISTR4CTOR JIJSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since the thermocouples are failed, hut a failed thernrocouple indicates 5O"F (low) and not
high.
b. Plausible since the thermocouples are failed, hut a failed thermocouple indimes 50°F (low) and not
high.
c. Plausible 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 ASALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze the effect of failed thermocouples on temperatures and subcooling
margin
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 71
Which ofthe following EOP network procedures may be directly entered and which
associated action is to be performed without direction from the Unit-SCO?
a. 0 FW-S. 1, .Response to Nuclear Power Generation / ATUS
0 Initiate emergency boration ofthe RCS
b. 0 FRP-H. 1. Response to Loss of Secondary Heat Sink
- Attempt to start an AFW Pump
c. e EPP-001; 1.0s~ of AC Power to IA-SA and 1B-SB Ruses
e Man~iallytrip the tusbine if still online
d. EPP-005. Natural Circulation Cooldown
- Attempt to start an KCP
ANSWER
c. * EPP-001, Loss ofAC Power to 1A-SA and IB-SB Ruses
Manually trip the turbinc if still online
Harris NRC Written Examination
Senioi Reactor Operator
Data Sheets
QIJEsTION NUAMBER: 7 1 TIEWGROCJP: 3
10CFR55 CONTENT: 41(b) 10 43@)
KA: 2.4.1
Knowledge of EOP entry conditions and immediate action steps
OBJECTIVE: 3.19-1
DESCRIBE Control Room usage of the EOP network as it relates to the following
Entry into EOP network
DEVELOPMENT KEFENNCES: Ii0P-EPP-OOI pg 3
EOP Users Guide pg 13
REFERENCES SIJPPLIED TO APPLICANT: Kone
QUESTION SOIJRCE: 0X NEW SIGNIFICANTLY MODIFIED 0 DlRECT
BANK NUMBER FOK SIGNIFICANTLY MODIFIED I DIRECT: New
NRC EXAM HISTOKY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Ilausible since FRP-S.l contains immediate actions, but is entered only by direction in PAIH-I
b. Plausible since FRP-1.1 is a high importance procedure: but is only entered when directed by other
produres.
X e. EPP-001 can be entered directly and contains immediate operator actions to manually trip the turbine.
8. Plausible since EPP-005 may be entered whenever a natoral circulation cooldown is required, hut it
contains no immediate operator actions.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWL.EDGEI RECALL
DIFFICULTY RATING: 2
EXPLAXATION: Knowledge of EOPs which can he entered directly
Harris NRC Written Examination
Senior Reacfor Operator
QUESTION: 72
Which of the following is a reason that containment pressure greater than 45 psig is
considered an extreme challenge to the containment critical safety function?
a. Containment structural failure is imminent
h. Containment leakage could be in excess of design basis leakage
c. Hydrogen recombiner efficiency is significantly reduced at the pressure
d. Containment temperature is high enough to prevent adequate core cooling
ANSWER:
b. Cmntainment leakage could be in excess of design basis leakage
Harris NRC Written fixamination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER 42 TIERKROUP: 2! 1
KAMPORTANCE: HO 3.1 SRO
10CFR55 CONTENT: 4l(b) IO 43@)
KA: 103Ci1.4.6
Knowledge of symptom based EOP mitigation strategies. (Containment)
o B m C m m 3.13-4
Given the fdlowing EOP steps, notes, and cautions, DESCRIBE the associated basis
CSF decision points
DEVELOPMENT REFERENCES: LP-3.13 pg 7
REFERENCE3 SUPPLIED TO APPIXANT: None
QCESTION SOURCE: 0 NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.13-R4 001
NRC EXAM HISTORY: None
DISTRACTOR JITSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since this is above the postulated pressure following a large break LOCA or steani break, but
containment failure is not expected to 0cc.ur 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. PIausihle since the rccombiners arc Located in containment and are conceivably affected by the high
pressure, bur the 45 psig is selected based on exceeding design leakage 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 psiig is selected based on exceeding
design leakage rates.
DIFFICULTY ANALYSIS:
m
Y
.
nKNOWLEDGE / RECALL
U
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the basis for CSFST decision points for containment pressorc
Harris NRC Written Examinatioii
Senior Reactor Operator
QUESTION: 73
Assunling the plant is at 100% power steady-state conditions, which ofthe following
would require independent verification insteild o f concurrent verification?
a. Removal of control power fuses for a clearance on KHK pump 1I3-SL3
b. Ierformanmce of PIC portions of OWP-RP due to the failure of PRZ pressure
transmitter PI-455
c. Installing a juniper in PIC-02 for a surveillance test
d. Lifting leads in Rod Control Power Cabinet lR1) for troubleshooting
ANSWER
a. Removal of control power fiises for a dearance on RHR punip 1B-SB
Harris NRC Written txamination
Senior Reactor Operator
Data Sheets
QEESTION NUMBER: 73 TIEWGROUP: 3
10CFR55 CONTENT: 41(b) 10 43m
K4: 2.2.13
Knowledge oftagging and clearance procedures
OBJFXTIVE: PP-3.11-7
L)E.FINEconcurrent verification and independent verification
DEVELOPMENT REFERENCES: OPS-NGGC- 1303, pg 12- L 3
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0 NE\% 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NlJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: Harris L O U 635
NRC EXAM HISTORY: None
DISTRACTOR JCISTIFICACTION (CORRECT ANSWER Xd):
X a. Concurrent verification is not needed on 4SOV breakers as they would have independent verification
since no adverse action would occur as a result of removing the fuses.
h. Plausible since an OWP directs these actions, but concurrent verification is required since the incorrect
switch operation could result in an RPS or LSF actuation.
c. IIausible since a surveillance test directs these actions, hut concurrent verification is required since the
incorrect switch operation could result in an KIS or ESF actuation.
d. Plausible since a work order directs these actions, hut concurrent verification is required since the
incorrect switch operation could rcsult in an RPS.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICI!I,TY RATING: 3
EXPLANATION: Knowledge of the conditions when concurrent verification is not permitted
IIwis NRC Written Exmiination
Senior Reactor Operator
QUESTION: 74
Given the following conditions:
- Following an accident, EPP-015, Uncontrolled Depressurization of d l Steam
Generators. is being perfornicd.
The operators have reduced AFW flow to all steam generators (SG) to minimum as
they continue attempts to isolate the SGs.
Uhich of the following describes the expected plant response to the AFW flow reduction
and what actions are to be taken as SC; pressures decrease?
a. RCS hot leg temperatures will eventually begin to inercase and the crew will then
transition to EPP-008, Safety injection Iemination
b. RCS hot leg telnpcratures will eventually begin to increase and the crew will then
increase AFW flow while continuing in EPP-015, TJnccntrolled Depressurization
of All Steam Generators.
c. The SGs will eventually become completely depressurized and the crew wiIl then
transition to EPP-014, Faulted Steam Generator Isolation.
d. Ihe SGs will eventually become completely depressurized and the crew will then
transition to EPP-008, SafetyInjection Termination.
ANSWER:
b. RCS hot leg temperatures will eventually begin to increase and the crew will then
increase A4FWflow while continuing in EPP-015, Vncontrolled Depressurization
of All Steani Generators.
ParisNRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 74 TIERGROUP: i!l
10CFR55 CONTENT: 41(b) 7 43@)
EX: WE12EK2.1
Knowledge of the interrelations between the (Uncontrolled Ikpressurization 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 taken in these emergency procedures, IREDICT the plant response to these actions
DEVELOPMENT REFERENCES: KIT-01 5, p 8
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECTANSWER Xd):
a. Plausible since hot leg tempeeatures will eventually increase, hut the correct ac.tion is to stabilize
temperature by increasing AFW flow and adjusting steaming rate. if possible.
X b. As SG pressure and steam flow- decrease, RCS hot leg temperatures will eventually stabili7.e and may
increase. Adjusting feed flow and steam dump will control KCS hot leg temperatures.
c. Ilausible since if no SG can be isolated the SGs will completely depressurize and there is a foldout
page to transition to EPP-014 if SC; pressure increases (will be stable when depressurized), and the
crew will eventually end up 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:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze S G response to decreasing pressure and reduced AFW flow and
determine correct response
Harris NRC Writtm Examination
Senior Reactor Operator
QUESTION: 75
The crew is implementing EPP-012. 1,oss of Emergency Coolant Recirculation. They
are now determining Containment Spray requirements with the following conditions:
- Containment pressure I 2 psig
- KWST level 3 ?dn
- Containment Fan Coolers running 3
- Containment Spray Punips running 2
Which of the follouring actions should be taken?
Start an additional Containment Fan Choler
b. Secure both Contaimnent Spray Pumps
,-. Secure one Containment Spray Pump
d. Secure one Containment Fan Cooler
ANSWER
b. Secure both Containment Spray Pumps
Harris NRC Written Examination
Senior Keactor Operator
Data Sheets
QUESTION NUMBER. 75 TIEWGROUP: 2: I
10CPR55 CONTENT: 41(b) 5 43w
KA: 026A2.08
Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; 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 can he done)
OBJECTIVE: 3.3-3
Given the title of an FiOP foldout item, state the parameters to monitored for implementation.
DEVELOPMENT REFERENCES: E1'1~-012~ p 3, 14
1cEFERENCE.SSUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW [7 SIGNIFICXXIIY MODIFIED DIRECT
BANK NIJI\.IBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.3-K5004
NRC EXAM HISTORY: None
DISTRACTOR JCTSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since the more Containment Fan Coolers that are running in EPP-012, the fewer spray
pumps are required, but no actions direct starting additional fans.
X b. With RWST level helow 3% all pumps taking a suction off the RWST must be secured, including the
Containment Spray Pumps.
e. Plausible since this action would be taken per EPP-012 ifthe RR7SI' still had sufficient water, hut
with the KWST empty all pumps must he stopped.
d. Plausible since action is taken to stop equipment that is used to remove heat from c.ontainment, hut the
pumps are stopped, not the fans.
DIPFICULTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS KNOWLEDGE /RECALL.
DIFFICULTY RATING: 3
EXPLANATION: Knowledge ofthc conditions for securing containment spray
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 76
Given the following conditions:
- While operating at 100% power, a drop in PRZ pressure resulted in a Reactor Trip
and Safety Injection.
Containment pressure is 3.6 psig and stable.
RCPs have been stopped.
R V I X Full Range is indicating 20%.
Core Exit Thermocouples are indicating 745°F.
P M level is currcntly indicating >100%.
PRZ pressure has stabilized at 1400 psig.
RCS Wide Range IIot Leg Temperatures are indicating 680°F.
Which of the following conditions currently exists)
a. A PRZ steam space break has occurred and a transition to FRP-C. 1 Response to
~
Inadequate Core Cooling, is required
b. A PRZ steam space break has occurred and a transition to FRP-C.2, Response to
Degraded Core Cooling, is required
c. An RCS hot leg break has occurred and a trmsition to FRP-C.1, Response to
Inadequate Core Cooling, is required
d. An RCS hot leg break has occurred and a transition to FW-C.2, Response to
Degraded Core Cooling, is required
ANSWER
a. A PR2, steatn space break has occurred and a transition to FW-C. 1, Response to
Inadequate Core Cooling, is required
Harris NRC Witten Examination
Scnior Reador Operatoi
Data Sheets
QUESTION NUMBER 46 TIERIGROUP: lil
10CFR55 CONTENT: 41(b) 43(b) 5
K4: 000008AA2.30
Ability to determine and interpret the following as they apply to the PressuriLer Vapor Space Accident:
Inadequate core cooling
OBJECTIVE:
DEVE.LOPMENTWXERENCES: FRP-C. 1
CSFST-Core Cooling
REFERESCES ShJPPL1E.DTO APPLICANT: None
QUESTION SOIJRCE: X NEW 0 SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JXISTIFICACTION (CORRECT ANSWER X9d):
X a. The RCS is superheated and in excess of 70OoF,which indicates that entry conditions to FRP-C:. 1 have
been met. The break is in the PRZ steam space as indicated by the pressurizer being full.
b. Plausible since the break is located in the PRZ steam space, but a transition to FRP-C.1 is required.
c. Plausible since RCS temperatures are stable, but the break is in the steam space and a transition to
FRP-C:. I is required.
d. Plausible since RCS heat removal is not adequate, but the break is in the steam space and a transition
to FRP-C.1 is required..
DIFFICULTY AXALYSIS:
COMPREHENSIVE / ANALYSIS 0 KYOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must analyze plant conditions to determine location of break, derermine that
temperature indications support superheated conditions and that entry
requirenients for FRP-C.1 have been met
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 77
Which of the folhwing describes a condition in Technical Specifications and its bases
which would require Emergency Boration in accordance with AOP-002, Emergency
Boration?
a. During the recovery from a Main Feedwater Pump trip, Control Rods are
determined to be below the rod insertion limit
Control the reactivity transient associated with a steam line break
b. * During the recovery from a Main Feedwater Pump trip. Control Rods are
determined to be below the rod insertion limit
Control the reactivity transient associated with an inadvertent dilution
6. During a reactor startup. the Reactor achieves criticality with Bank c rods at
105 steps
0 Control the reactivity transient associated with a steam line break
d. o During a reactor startup, the Reactor achieves criticality with Bank C rods at
105 steps
Control the reactivity transient associated with an inadvertent dilution
ANSWER
c. * During a reactor startup, the Reactor achieves criticality with Bank C rods at
105 steps
- Control the reactivity transient associated with a steam line break
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QI~ESTIONNIJTMRER: 77 TIEWGROUP: 112
lOCFR55 CONTENT: 41(b) 43(b) 2
KA: 000024<;2.2.25
Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.
(Emergency Boration)
OBJECTIVE:
DEVELOPMENT REFERENCES: 7's Bases 314.1.1
AOP-002 BD
GP-004
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIF1K.D I DIRECT: AOP-3.2-R1 001
NRC EXAM HISTORY None
DISTRACTOR JIJISTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since if this condition existed for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, Emergency Boration would be required.
Additionally, in Modes 1 22 2 , SDM k required to control the reactivity transient associated with a
steam line break. However, it is not required during transient conditions, allowing the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to
restore rod position.
b. Plausihlc since ifthis condition existed for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, Emergency Boration would be required.
However, it is not required during transient conditions, allowing the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore rod position.
X E. Emergency boration is required if SDM is not niet Criticality at steady state conditions is considered
to be a loss of SDM. In Modes I & 2, SDM is requiTed to control the reactivity transient associated
with a steam Line break.
d. Plausible since Eniergency boration is required if SDM is not met. Criticality at stwdy state
conditions is considered to be a loss of SDM. However, the concern for an inadvertent dilution is
related to a shutdown condition.
DIFFICULTY ANALYSIS:
COMPREHEXSIVE I ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of the requirements fix initiating Emergency Boration and the bases
for these actions.
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 78 TIEIUGROUP: 21i
lOCFR55 CONTENT: 4I(h) 43(b) 5
KA: 006.42.04
Ability to (a) predict the impac,ts ofthe following malfunctions or operations on the ECCS; and (b) based
on those predictions, use procedures to correct, control, or mitigate the c.onsequences of those
malfunctions or operations: Improper discharge pressure
OBJECTIVE:
DEVELOPMENT HEWERENCES: EPP-OOX
User's Guide
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOIJKCE: 0 NEW SIGNIFICANTLY MODIFIED IMRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: 3,19-R4 35
NRC EXAM HISTORY: None
DISTRACTOR .KISTIFICACI'ION (CORRECT ANSWER X'd):
a. PLausible since the charging flow isolation valves being open can result in pump runout, but even after
the valve is closed a transition is required to PATH-1.
X b. The charging flow isolation valves being open at the same time as the BIT valves can result in pump
runout, but even after the valve is closed a transition is required to PATH- 1.
e. Plausible since the alternate miniflow isohtion valves being open will result in increased flow, but the
valves me dosed at this point and even if open after the valve is closed a tranqition is required to
PATII-I.
d. Plausible since the alternate miniflow isolation valves being open will result in increased flow and 3
transition is required to PATH-I, but the valves are closed at this point.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must analyze given conditions to determine that runout is occurring as a result
offailing to close the charging valves first and also determining that a transition
is still required to PATH-I
H a n k NRC Written Examination
Senior Reactor Operator
QUESTION: 78
Given the following conditions:
e A Reactor Trip and Safety Injection have occurred due to a smdl break LOCA.
e The crew has isolated the break and has just established normal charging in
accordance urith EPP-008. Safety Injection Termination.
e They then check Safety Injection Reinitiation criteria and determine that Safety
Injection flow is required.
e When they open the BIT Outlet valves, 1SI-3 and 1SI-4. they note that CSIP
discharge pressure is low and oscillating.
Which ofthe following actions is to be taken?
a. Isolate the charging line due to the CSIP operating at runout conditions and
continue in EPP-008, Safety Injection Termination.
b. Isolate the charging line due to the CSIF operating at runout conditions and
transition to PATH-1.
C. Isolate the arternate miniflow line due to the CSIP cavitating and continue in EPP-
Safety Injection lcrmination.
d. Isolate the alternate miniflow line due to the CSIP cavitating and transition to
PATII-1.
ANS\\ER:
b. Isolate the charging line due to the CSIP operating at runout conditions and
transition to PATH-1.
Harris KRC Written Examination
Senior Reactor Operator
QUESTION: 79
Given the folhwing conditions:
'1he unit is operating at 100% power, with Control Bank D rods at 21 5 steps.
AI,B 13-7- 1, KO11 CONTROL IJRGEYi' ALAKhI, is in ALARM due to a failure in
Power Cabinet IAC.
Rod Control is in MAN.
A turbine trip occurs, but the Reactor fails to trip either automatically or manually.
Which of the foliowing actions should the Reactor Operator be directed to take in
accordance with FRP-S. 1. "Response to Nuclear Power Generation 1 ATWS"?
a. Place the Rod Control RANK SELEC'I'OK in AITTO and allow rods to insert
b. Maintain the Kod Control BANK SELECTOR in MAN and manually insert rods
c. Place the Rod Control BANK SELECTOK in BANK D and mmnally insert rods
d. Maintain rods at 215 steps
ANSWER:
d. Maintain rods at 21 5 steps
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER: 79 TIEWGROUP: 2!2
10CFR55 CONTEXT: 41(b) 43(b) 5
U: 001G2.4.6
Knowledge of symptom based EOP mitigation strategies. (Control Rod Drive)
OBJECTIVE:
DEVELOPMENT REFERFJVCES: USERS GIJIDE
FW-S.1
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: SEW 0 SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JCJSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since this is an RNO action for a failure of the reactor to trip, but will not be successful due
to the urgent failure in rod control.
b. Piausihle since this is an RNO ac.tion for a failure of the reactor to trip, but will not be successful due
to the urgent failure i n rod control.
c. Plausible since this will allow Bank D rods to move inward, and is the only method of inserting rods
with the rod control failure, but should not be used due to the potential to cause unanalyzed flux
shapes.
X d. Due to the urgent fiilure, rods will not move in AUTO or MAN. Although they will move in BANK
I) with this particular failure, moving rods in individual banks may result in unanalyzed flus shapes
which could result in fuel damage.
DIFFICLTLTY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must analyze the effect of an urgent rod control failure and then apply the
failure results to the plant conditions to determine the proper actions
Harris NRC Written 1:xamination
Senior Reactor Operator
QUESTION: 80
Given the follouring conditions:
8 The unit was operating at 98% power with Heater Drain Pump A under
clearance.
ALB 20-2-2, TIJRBINE KINBACK OPERKIIVR, is in alarm.
ALB 11-4-3, KIACTOR TRIP OVERTEMP AT, is in alarm.
A turbine runback occurs.
Reactor power is currently at 93% and lowering as the turbine runback continues.
Both Main Feedwater Pumps are operating.
IIeater Drain Pump H is operating.
All loop AYs indicate less than the OTKI and OPAT setpoints.
All OTAT and OPAT Trip Status Light Boxes are dark.
Which of the following actions should be taken?
a. Select IURBINE MANUAL due to a runback circuitry failure and go to AOP-
015, Secondary Load Rejection
b. Verify the runback stops at 90% load and go to AOP-015, Secondary Load
Rejection
c. Verify the runback stops when the OTAI condition clears and go to AOP-015.
Secondary Load Rejection
d. Trip the Reactor and go to PATH-1
ANSWER
d. Trip the Reactor and go to PATII-1
I I x ~ i NRC
s Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 80 TIEWGROUP: 3
10CPR55 CONTENT: 41(h) K?(b) S
u: 2.1.1
Ability to evaluate plant performance and make operational judgments based on operating characteristics,
reactor behavior, and instrument interpretation.
OBJECTIVE :
DEVELOPMENT HEFERENCE.S: OMM-00 1
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTKACTOR JUSTIPICACTION (CORRECT ANSWER Xd):
a. Plausihle since there is no apparent cause of the turbine runback, but with the plant in a transient atid a
Reactor Trip First Out annunciator lit, the reactor should be tripped.
R. Plausible since a trip of both IlDPs would result in a runhack to 90%, but with the plant in a transient
and a Reactor Trip First Out annunciator lit, the reactor shouid be tripped.
C. Plausible since the OTAT condition would result in a riinback until the condition clears, hut with the
plant in a transient and a Reactor Trip First Out annunciator lit, the reactor should be tripped.
X d. With the plant in a transient and a Reactor Trip First Out annunciator lit, the reactor should be tripped.
DIFFICCJ,TY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / KECALL
DIFFICULTY HATING: 3
EXPLANATION: Analysis of plant conditions to determine the required actions
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 81
Given the foliowing conditions:
A Reactor Trip with SI occurs.
The operators perform the immediate action steps, verify ECCS flow. and check
AFW flow in accordance with PATH-I.
- Entry requirements for FRP-11.1. Response to I.oss of Secondary IIcat Sink, have
been met.
RCS pressure is 175 psig.
0 All SG pressures are between 300 psig and 350 psig.
Which ofthe following actions is to be taken?
a. Transition to FRP-H.1. Response to I,oss of Secondary Heat Sink, and attempt
to establish AFW or Main Feedwater flow
b. Iransition to FKP-II.1, Response to Loss of Secondary IIeat Sink, and iniriate
RCS feed and bleed.
c. Transition to FRP-1-1.1, Response to I m s of Secondary Heat Sink, and then
return to FATII-1 since a secondary heat sink is NOT required
d. Remain in PATH-1 since a secondary heat sink is NOT required
ANSWER:
6. Transition to FRP-H.l, Response to Loss of Secondary IIeat Sink. and then
return to IATII-1 since a secondary heat sink is NOT required
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER 8 I TiEWGROUP : i!l
10CFR55 CONTENT: 4l(h) 43(b) ti
KA: 00001 i(32.4.6
Knowledge of symptom based EOP mitigation strategies. (Large Break 1,OCA)
OBW,CTI\7E:
DEVELOPMENT REFERENCES: FRP-1.1
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.1 l-RiOO3
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since these are actions that are taken upon entry into FW-11.1, but a secondary heat sink
would not be required with RCS pressure <: SG pressure.
b. Plausible since these are actions that might be taken upon entry into FKP-H.1. but a se.condary heat
sink would not be required with RCS pressure iSG pressure.
X e. Since RCS pressure is less than SG pressure, a secondary heat sink is not required since the SC; would
act as a heat source rather than a heat sink. Return is to procedure and step in effect.
d. Plausible since RCS pressure is less than S(3 pressure and B secondary heat sink is not required
Return is to procedure and step in effect, not Entry Point C .
DIFFICCLTY AN.4LYSIS:
COMPREIIENSIVE / ANAL.YSIS KNOWLEDGE /RECALL
DIFFICIXTY RATING: 3
EXPLANATION: Must interpret first that a sec.ondary heat sink is not required based on RCS
pressure being greater than SG pressure and then must recognize the entry point
conditions for returning to PATH-1
Harris NKC Written Examination
Senior Keactor Operator
QUESTION: 82
Given the following conditions:
The Keactor has been taken critical and power is being increased.
NIS IR channels N15 and N36 are both indicating 5 x lo- imps.
NIS SR channel K31 is indicating 8 x lo3 cps.
a XIS SR channel N7.12 is indicating 7 x lo4 cps.
Power should he stabilized . . .
a. at or above IO- amps, and the SR High Flux trip should then be blocked.
b. at the current power level, and the SR High Flux trip should then be blocked.
c. at or above lo- amps, hut the SIP High Flux trip should NOT he blocked.
d. at the current power level, but the SR High Flux trip should NOT be blocked.
ANSWEK:
d. at the current power level, hut the SR IIigh Flux trip should NOT be blocked.
Ifmi5 NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 82 TIEKUGROUP: 1i 2
10CPR55 CONTENT: 41(b) 43(b) 5
KA: 000032AA2.09
Ahility to determine and interpret the following as they apply to the Loss of Source Range Nuc.lear
Instrumentation: F.ffect of improper HV scttiug
OBJECTIVE:
DEVELOPMENT REFERENCES: GP-004
ALB-0 12-4-5
REFEKE:NCE.SSUPPI.,IEI) TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY RIODIFIEU DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR .IWSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since power must he increased above amps before blocking trips, hut increasing power
to this level will result in SR high flux trip.
b. I'lausihle since power cannot he increased above IO-'" amps, hut the block of the SR high flux trip is
interlocked at this power level.
c. Plausible since the SR high flux trip is not permitted to be blocked without at least 1 decade of overlap
between SR and IR, hut increasing power ahove 10. amps will result in a SK high flux trip.
X d. I,ess than 1 decade of overlap exists hetwcen SR and IR channel before trip would occur. Increasing
power to allow blocking SR would re.sult in trip before reaching power level and attempting to block
at current power level will not he successful.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATISG: 3
EXP1,ANATION: Must determine that increasing power above 10-'"amp will result in a reactor
trip due to SR high flux, and that attempting to block the SR high flux trip
below 10-loampswill not he successful. Required to not block SR high flux trip
if :
. 1 decade of o\,erlap exists.
IIarrk NRC, Written Exmination
Senior Reactor Operator
QUESTION: 83
Given the following conditions:
- The reactor fails to trip when required.
- The Operators take actions per the appropriate procedurc(s) and obtain the required
plantisystcm!'componeiit responses, except that the reactor is still NOT tripped.
Emergency boration CANNOT be initiated because of bloc.kage in the boration flow
paths.
0 All Power Range channels indicate 3%
Startup rate is zero on both Intermediate Range channels.
Which of the following describes the correct operator ac,tions under these conditions
ANI) the primary reason for taking these actions?
a. 0 Return to the procedure and step in effect.
a Power is less than 5% and the IR startup rate is zero.
b. Remain in FRP-S.1, "Response to Nuclear Power Generation / KIWS," and
allow the RC.S to heat up while continuing efforts to establish emergency
boration.
The heatup will insert negative reactivity.
c,. Go to FW-S.2, "Response to I m s of Core Shutdolw~"
This is required by the Subcriticality CSFST based on the current reactor
conditions.
d. Remain in FRP-S. 1, "Response to Nuclear Power Generation / ATWS," and
maintain RCS temperature stable while continuing efforts to establish
emergency boratioii.
- Stable temperatures preclude positive reactivity insertion by cooldown.
ANSWER
b. a Remain in FKP-S.1, "Response to Nuclear Power Generation / A'IWS." and
allow the RCS to heat up while continuing eflorts to establish emergency
boration.
- The heatup will insert negative reactivity.
Harris NKC Written kxamination
Senior Reactor Operator
Data Sheets
QIJESTION NUMBER: 83 'HEWGROUP: 2il
10CFR55 CONTENT: 41(b) 43(b) 5
U: 012G2.4.6
Knowledge of symptom based EQP mitigation strategies. (Reactor Protection)
OBJECTIVE:
DEVELOPMENT REFERENCES: FRP-S. 1
REFERENCES SIJPPLIED TO APPLICANT: None
QUESTION SOIJRCE: [7 SEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 3.15-R.5 11
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since the exit conditions for FRP-S.1 include Power Range channels < 5%, but Intermediate
Kange channels must indicate a negative startup rate to consider the reactor shutdown.
X h. Intermediate range startup rate being zero indicates that the reactor is not shutdown, and since the
reactor is not tripped and emergency boration cannot be established, the RCS is allowed to heat up to
add negative reactivity.
c. Plausible since the entry conditions for FRP-S.2 include Power Range channels < 5% with the
Intermediate Kange channels indicating a zero or positive startup rate, but a transition would not be
made because FRP-S.1 is a higher priority procedure.
d. Ptansible since maintaining temperature stable would maintain p o ~ below a 5%? bun without
emergency boration the only means of adding negative reactivity to the core, as required, is by
allowing the RCS to heat up.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS [7 KNO\VLEDGE / RECALL
DIFFICULTY RATISG: 3
EXPLANATION: .4nalysis of multiple failure conditions to determine proper course of action and
basis
Harris NKC Written Examination
Senior Reactor Operator
QIJESTION: 84
Given the following conditions:
- The plant is in Mode 3 with all Shutdown Rods withdrawn.
The Digital Kod Position Indication emergency power supply is under clearance.
The normal Digital Rod Position Indication power supply has just tripped.
Which of the following actions is to he taken?
a. Due to the loss of all Digital Rod Position Indication. open the Reactor Trip
Breakers in accordance with Technical Specification 3.1.3.3, Position Indication
System -- Shutdown
h. Due to the loss of all Digital Rod Position Indication and Demand Position
Indication. verify that all Shutdown Rank Rods are fully withdrawn using the
movable incore detectors in accordance with AOP-001. Malfunction of Rod
Control
C. Due to the loss of all Digital Rod Position Indication and Demand Position
Indication, commence a horation of the RCS to ensure adequate Shutdown Margin
in accordance with AOP-002, Emergency Roration
d. Due to the loss of all Digital Rod Position Indication, verify that d l Shutdown
Bank Rods iire fully withdrawn using Demand Position Indication in accordance
with AOP-001, Malfunction of Rod Control
ANSWER:
a. Due to the loss of all Digital Kod Position Indication, open the Reactor Trip
Breakers in accordance with Technical Specification 3.1.3.3, Iosition Indication
System - Shutdown
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QIJESTION NUMBER: 84 TIE.WGROIJP: 2
!1
10CFRS5 CONTENT: 41(b) 43(b) 2
KA: 014A2.02
Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based
011 those on those predictions, use procedures to correct, control, or mitigate the consequences of those
malfunctions or operations: Loss of power to the RPIS
OBJECTIVE:
DEVELOPMENT REFERENCES: TS 3.1.3.3
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOIJRCE:
0 NEW
X SIGNIFICANTLY MODIFIED 0 DIRECT
RANK NUMBER FOR SIGNIFICANTLY M0DIFIE.D /DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JCJSTIFICACTION (CORRECT ANSWER Xd):
X a. With both DRPI indications inoperable in Mode 3 , 4 , or 5 . IS requires that the Reactor Trip Breakers
be opened immediately.
b. Ilausible since this would be required in the event of a loss o f a single indic.ation while operating in
Mode 1 or 2, but with both indications lost in Mode 3 the Reactor Trip Breakers are to be opened per
E3.1.3.3.
c. Plausible since loss of indication of DRPI may lead to belief that SDM cannot be verified, which
would require. Emergency Boration, but the Reactor TripBreakers are required to be opened per IS
3.1 3.3.
d. Plausible since this would be required in the event of a loss of a single indication while operating in
Mode. I or 2 , but with both itidications lost in Mode 3 the Reactor Trip Breakers are to be opened per
is3.1.3.3.
DIFFICU1,TY ANALYSIS:
0 COMPREIIENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of Tech Spec immediate action requirements in the eveut of a loss
of both DKPI indications
Harris NRC Written Exmiin&on
Senior Reactor Operator
QUESTION: 85
A Senior Reactor Operator has failed to meet the required number of hours this past
calendar quarter to maintain an active license.
Assuming all other requirements h v e been met to activate the license in the SRCl
position, in accordance with OMM-001, Conduct of Operations, which of the following
watches completed under instruction would satisfy the requirement to allow activation of
the license?
a. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as the Unit-SCO in Mode 4 AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as the S-SO in Mode 5
b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as the Unit-SCO in Mode 5 AND 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> as the S-SO in Mode 4
c. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> as the Unit-SCO in Mode 5 ANI) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as the S-SO in Mode 4
d. 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> as the Unit-SCO in Mode 5
ANSWER
a. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as the Unit-SCO in Mode 4 AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as the S-SO in Mode 5
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 85 TIEWGROL!Pc!p: 3
1QCFRSJCONTENT: 41(b) 10 $309
KA: 2.1.1
Knowledge of conduct of operations requirements
OBJECTIVE: PP-2.0/3.0/5 .O-S1
IDENTIFY the responsibilities of the Unit Senior Control Operator as listed in OMM-001 (SRO
Ohjective)
DEVE.LOPMENTREFERENCES: OMM-00 1
RF,FERENCXS SUPPL.IEDTO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIItEECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIEI) / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> arc required in either the S-SO position in any Mode and in the IJnit-SCQ position when the
plant is above 200°F.
b. Plausible since this exceeds the required 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> for the SRO position, but only those hours when the
plant is above 200°F are acceptable for the IJnit-SCO position and all hours are acceptable for the S-
SO position.
e. Plausible since this exceeds the required 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> for the. SRO position, but only those hours when the
plant is above 200°F are acceptahle for the Unit-SCO position and all hours are acceptable for the S-
SO position.
d. Piausihle since. this exceeds the required 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> for the SKO position, but only those hours whe.n the
plant is above 200'F are acceptable for the IJnit-SC,O position and all hours are acceptable for the S-
SO position.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KYOWLEDGE / RECALL
DIFFLCIJLTY RATING: 2
EXPLANATION: Must recall requirement for activating an inactive license from OMM-001
Harris NRC Written Exmiination
Senior Reactor Operator
QUESTION: 86
Given the following conditions:
Following a steam generator tube rupture (SGTR), the crew has completed PATTI-2
and a hansition has just been made to the SGTR recovery procedure as directed by
PATII-2.
Condenser Available light (C-9) is NOT lit.
Which ofthe following actions should be taken to cool down the RCS?
a. Dump s t m i using the intact SG PORVs and transition to EPP-018, Post-SGTR
Cooldown Using Blowdown
b. Dump steam using the intact SG PORVs and remain in EPP-017, Post-SGTR
Cooldown LJsing Backfill
C. CooIdown using the TDAFW Pump and transition to FPP-017, Post-SGIR
Cooldown Using Backfill
d. Cooldown using the TDAFW Pump and remain in EPP-018, Post-SGTR Cooldown
Using Blowdomn
ANSWER
b. Dump steam using the intact SG PORVs ,and remain in EPP-017, Post-SGIR
Cooldown Using Backfill
IIarris NRC Written Exminiltion
Senior Keactcr Operator
Data Shcets
QUESTION NUMBER 86 TIEFUGROUP: 1/1
10CFR55 CONTENT: 41(h) 43(h) 5
KA: 000038EA2.08
Ability to determine or interpret the foliowing as they apply to a SGTR: Viable alternatives for placing
plant in safe condition when condenser is not availahle
OBJECTIVE:
DEVELOPMENT REFERENCES: PATH-2
EPP-0 17
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0X NEW 0 SIGNIFICANTLY MODIFIED 0 DIRECT
RANK SUMBER FOR SIGNIFICANTLY MODIFiED /DIRECT: Sew
NRC EXAM HISTORY Sone
DISTRACTOR JUUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since steam must he dumped using the SG P(PR\'s or the TDAFW Punq& hut the crew will
remain in the preferred procedure which is EPP-017.
X b. Steani must he dumped using the SG PORVs or the TDAFW Pump and the crew will remain in the
preferred procedure w-hich is EPP-017.
c. Plausible since steam mnst he dumped using the SG PORVs OK the TDAFW Pump, hut the crew will
will not transition to E.PP-017 since they are already in this proc.edure as it is the preferred resovery
procedure.
d. Plausible since steam must he dumped using the SG PORVs or the 'I'DAFW Pump, hut the crew will
remain in the preferred procedure which is EPP-017, not EPP-018.
DIFFICULTY ANALYSIS:
0 COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of mitigation strategies when steam dumps are not availahle and
knowledge of preferred recovery method from SGTR
IIarris NRC Written Examination
Senior Reactor Operator
QUESTION: 87
A LOCA occurrtd several hours ago. Only one (1) Containment Spray Pump is running
clue to actions taken in FPP-012, Loss of Emergency Coolant Recirculation.
A transition has just been made to FW-J. 1, Response to High Containment Pressure.
Containment Pressure is 14 psig.
Which of the following actions should be taken?
a. Start the second Containment Spray Pump if Containment pressure does NOT
decrease below 10 psig bcfore exiting FW-J. 1.
b. Start the second Containment Spray Pump per FW-J.lsince pressure is above 10
psig.
c. Continue operation with one Containment Spray Pump per EPP-012 unless
Containment pressure exceeds design, then start the second pump.
d. Continue operation with one Containment Spray Pump per EPP-012 unless
Containment pressure begins increasing, then start the second pump.
ANSWER
c. Continue operation with one Containment Spray PUIII~per EPP-012 unless
Containment pressure exceeds design, then start the second pump.
1Iarriu NRC Written Exmiination
Senior Reactor Operator
Data Sheets
QIIESTION NUMBER: 87 TIEWGROUP: 112
10CPR55 CONTENT: 41(b) 43(b) 5
KA: HE14EA2.2
Ability to determine and interpret the following as they apply to the (High Containment Pressure)
Adherence to appropriate procedures and operation within the limitations in the facilitys license and
amendments
OBJECTIVE:
DEVELOPMENT REFERENCES: FRP-J. 1
REFEKENCES SUPPLIED TO APPLICANT: None
QUESTION SOIJRCE: 0 NEW SIGNIFICANTLY MODIFIED
BANK NIJMBEX FOR SIGNIFICANTLY MODIFIED /DIRECT: 3.13-R4 008
NRC EXAM HISTOKY: None
DISTRACTOK .IX!STIFICACTION (CORRECT ANSWER Xd):
a. Plausible since this would be a normal action directed by FRP-J. 1.
h. Plausible since this would he a normal action directed by l K P J . 1
X c. EPP-012 directs the operators to run Containment Spray Pumps based upon Containment pressure and
Fan Cooler operation. These actions are taken to minimize RWST depletion. This configuration is to
be maintained even if FRP-J.1 is implemented.
d. Plausible since would better serve the intent of EPP-012, but would be contradictory to the intent of
FRP-J.1 which has a higher priority concerning the operation of the Spray Pumps.
DIFFICIXIY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICGLTY KATING: 3
EXPLANATION: Must compare the relative actions in the 2 procedures and make a jiidgement of
which condition tnkes precedent
IIarris NRC Written Examination
Senior Reactor Operator
QTJESTION: 88
Given the following conditions:
A large break LOCA has occurred.
- The crew hasjust transitioned to EPP-010, lransfeer to Cold Leg Recirculation.
The Reactor Operator reports that RHR Pump 1A-SA trippedjust before the
transition was made to EPP-010.
Which of the following actions is to be taken?
a. Continue with the transfer to Cold Leg Rccirculation and remain in EPP-010.
lransfer to Cold Leg Recirculation. but only align one (1) CSIP for recirculation
b. Continue with the transfer to Cold Leg Recirculation and remain in EPP-010,
Transfer to Cold Leg Recirculation, aligning both CSIPs for recirculation
c. Transition to EN-012, Loss of Emergency Cookant Recirculation, and reduce
injection flow to a single RIiR Pump running and a singlc CSIP running
d. Transition to P,PP-OlS, Loss of Emergency Coolant Recirculation. and secure
both RIIR Pumps while maintaining both CSIPs running
ANSWER:
b. Continue with the transfer to Cold Leg Recirculation and remain in EPP-010,
Transfer to Cold Leg Recirculation. aligning both CSIPs for recirculation
IIarTis NRC Written fixamination
Senior Reaetor Operator
Data Sheets
QUESTION NUMBER 88 TIEWGROUP 2;1
10CFR55 CONTENT: 41(b) 43(b) 5
KA: 013A2.01
Ability to (a) predict the impacts of the following malfunctions or operations on the ESE'AS; and (b)
based on those predictions, use procedures to correct, control, or mitigate the consequences of those
malfunctions or operations: LOCX
OBJECTIVE:
DEVELOPMENT REFERENCES: E w - n i o
REFERENCES SUPPLIED TO APPLICANT: None
QTJESTION SOURCE: 0 X NEW' 0 SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC EXAM HISTORY Kone
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since the normal alignment would have both RHK pumps supplying both CSIPs and
reducing CSIP flow would limit the pumping requirement of a single RHR pump, hut both CSIPs will
still be aligned for recirc.
X h. Thc steps tn align for cold leg recirculation assume only a single train of ESF equipment i s available
and aligns all operable equipment, including both CSIPs.
c. Piausible since EPP-012 would be entered ifthe erewwas unable to establish cold leg recirculation,
but a single RIIR pump will still permit alignment.
d. Piausible since EPP-012 would be entered if the crew was unable to establish cold leg recirculation,
hit a single KHR pump will still permit alignment.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS 0 KNOWLEDGE I RECALL
DIFFICULTY RATING: 4
EXPLANATION: Analysis of the effect of a failure of a single RHR pump on the ability to align
for cold leg recirc and the procedure whic.h should be used to recover
IIarris NRC W-rittcn Examination
Senior Reactor Operator
QUESTION: 89
Which of the fbllowing situations occurring on July 1, 2003. would require that an
I;,yuipment Inoperability Record be filled out in accordance with OMM-014. Operation
of the Work Control Center?
a. The ERFIS computer faiied and was removed from service per the OWP
b. The SWST Discharge Radiation Monitor was removed from service per the OWP
c. Scheduled MST-10052, Pressurizer Level Loop (L-0459). was performed
d. During an OST for the A EDG, the E-86 A-SA, EDG Room Exhaust I:an faiied
ANSWER
d. During an OST for the A EDG. the E-86.4-SA, EDG Room Exhaust Fan Failed
Harris NKC Written Exanination
Senior KeaCtQrOperator
Data Sheets
QtJESTION NUMBER: 89 TIEWGROUP: 3
10CFRSS CONTENT: 41(b) 43(b) 5
KA: 2.2.19
Knowledge of maintenance work order requirements
ORJECTIVE:
DEVELOPMENT REFERENCES: OMM-014
OWP-HVAC
REFERENCES SUPPLIED TO APPLICAXT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NIJMRER FOR SIGNIFICANTLY MODIFIED /DIRECT: PP-2.4-S6 2
YRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since EFWIS affects Tech Spec operability of equipment, but an exception is provided for
EKFIS per OMM-014 if it is addressed by OWP-ERFIS.
b. Plausible since the radiation monitor is addressed in PLP-I 14 for operability ofequipment, but NI
exception is provided for a radiation monitor per OMM-014 if it is addressed by OWP-RM.
c. Plansible since an EIR would be required if this was corrective, bnt scheduled maintenance does not
require it.
X d. The EDG fan affects the operability of the EDG which requires that an EIR be completed.
DIFFICULTY ANALYSIS:
0 COMPREHENSIVE / ANALYSIS KKOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of requirements for gcnernting an EIK
Harris NRC Writtsn Examination
Senior Reactor Operator
QUESTION: 90
Given the following conditions:
A steam bre& has occurred on SG A.
Following a Reactor Trip and Safety Injection, a transition has been macle to EPP-
015. Uncontrolled Depressurization of All Stem1 Generators.
Why should the 1init-SCO direct the operators to attempt to close the Main Steam
Isolation Valves before taking any other actions to isolate the SGs?
a. To minimize the positive reactivity effects associated with the cooldoun of the
RCS due to more than one (1) SG blowing down AND to minimize the pressure
rise inside Containment in the event the steam break is inside Containment
b. To minimize the positive reactivity effects associated with the cooldown of the
RCS due to more than one (1) SG blowing down AND to cnsme an adequate
supply of AFW in the Condensate Storage Tank in the event of a subsequent loss
of all AC power
c. To ensure an adequate supply of AFW in the Condensate Storage Tank in the
event of a subsequent loss of all AC power AND to limit the likelihood o f a
radiological release to the environment in the event of a subsequent SGTR
d. To minimize the pressure rise inside Contaimcnt in the event the steam break is
inside Containment AND to limit the likelihood o f a radiological release to the
environment in the event of B subsequent SGTR
ANSWEK:
a. To minimize the positive reactivity effects associated Kith the cooldown of the
KCS due to more than one (1) SCi blowing down AND to minimize the pressure
rise inside Containment in the event the steam break is inside Containment
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QIXSTION NUMBER: 90 TIEWGROUP: 111
10CFR55 CONTENT: 41(b) 43(h) 2
KA: 000040G2.1.32
Ability to explain and apply all system limits and precautions. (Steam Line Rupture - Excessive Heat
Transfer)
OBJECTIVE:
DEVELOPMENT REFERENCES: EPP-015
TS Bases 314.7.1.5
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC KXAM HISTORY None
D1STRAC:TOR JUSTIFICACTION (CORRECT ANSWER X'd):
X a. The bases for the MSIVs is to prevent an excessive positive reactivity addition due to more than one
SG blowing down through the break and to limit the pressure rise inside containment in the event the
break is located in containment.
h. Plausible since a goal of shutting the MSIV is to prevent an excessive positive reactivity addition due
to more than one SG blowing down through the break, but isolating AFW flow also minimizes the
cooldowri and is not n concern about maintaining CST inventory.
c. Plausible since isolating AFW flow will limit the cooldown of the KCS and the faulted S G is the most
likely SG to rupture due to the differential pressure, but isolating the MSIV is concerned with
excessive positive. reactivity addition and the pressure rise inside containment.
d. Plausible since the MSIV is isolated to minimize the pressure rise inside containment and the faulted
SG is the most likely SG to rupture due to the differential pressure, but isolating the MSIV is
concerned with excessive positive reactivity addition and the pressure rise inside containment.
DIFFICIJLTY ANAL.YSIS:
C:OMPKEIENSIVE/ ANALYSIS KIYOWLEDGE / RECALL
DIFFICIJLTY RATING: 3
EXPLANATION Knowledge ofthe TS bases for the MSIVs
Harris NRC Written Examination
Senior Reactor Operator
QIJESTION: 31
?he unit has tripped due tu a LOCA and ESF equipment has failed to start. As a result,
I RP-C.2, Response to Degraded Core Cooling. has been entered.
A depressurization of the Steam Generators (SGs) to 80 psig is being perfixmed, in
accordance with FRP-C.2, when the STA reports that a Red Path condition for Integrity has
occurred.
WXch ofthe following actions should be taken?
a. Immediately transition to FW-P. 1, Response to Imminent Pressurized Thermal
Shock Conditions
b. Stop the S G depressiuization and, if the red path does not clear, transition to FRP-
1. 1. Response to Imminent Pressurized lhemal Shock Conditions
C. Ckxnplete W - C . 2 and then transition to FRP-P. 1, Response to Imminent
Pressurized Thermal Shock Conditions, if the red path still exists
d. Complete the S/G depressurization and then trnnsition to FRP-P. 1, Response to
Imminent Pressurized Thermal Shock Conditions, if the red path still exists
ANSWER:
c. Complete YRP-(2.2 and then transition to FRP-P. 1, Response to Imminent
Pressurized ?hernial Shock Conditions, if the red path still exists
Ifarris NRC Written Examination
Senior Reactor Operator
Ddtta Sheets
QUESTION NUMBER 91 TIEFUGROUP: 1 !2
10CFR55 CONTENT: 41(b) 43(h) 2
JSA: WEOGG2.1.32
Ability to explain and apply all system limits and precautions. (Degraded Core Cooling)
OBJECTIVE:
DEVELOPMENT REFERENCES: FRP-C.2
REFERENCES SUPPLIED TO APPLICANT: None
QtJESTIONSOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the red path for integrity has a higher priority than the orange path that caused entry
into FRP-C.2, hut under these particular conditions a transition should not occur until completion of
the FKP-C.2.
h. Plausible since the red path for integrity has a higher priority than the orange path that caused entry
into FRP-C.2, hut under these particular conditions a transition should not occur until conipletion of
the FRP-(2.2.
X c. During the depressurization, a red path niay occur due to injecting the ac.cumulators. A transition
should not be made until the entire procedure has been completed.
d. Plausible since the red path for integrity has a higher priority than the orange path that caused entry
inro FRP-C.2, but under these particular conditions a transition should not occur until completion of
the FRP-C.2.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS 0 KNOWLEDGE / RECALL
DIFFICIJLTY RATING: 3
EXPLANATION: Must analyze plant conditions to determine that the cause of the red path is the
depressurization and that, under these specific conditions, an immediate
transition is not warranted
Mmis NRC Written Examination
Senior Reactor Operator
QUESTION: 32
Given the following conditions:
The unit is in Mode 3.
- Instrument Ruses IDP-1B-SI1 and 1DP-113-SIV are both de-energized.
Maintenance reports that Instrument Bus IDP-1R-SI1 is ready to be reenergized.
In order to prevent an inadvertent Safeguards Actctuiition, which of the following must be
verilied prior to re-energizing the bus?
a. Train A Logic Input Error Inhibit must be verified to be in INIIIBIT
b. Train A Logic Train Output must be verified to be in TEST
c. Train R 1,ogic Input Error Inhibit must be verifitd to be in INHIBIT
d. Train 8Logic Train Output must be verified to be in TEST
ANSWER:
d. Train B Logic Train Output must be verified to be in TEST
Harris NRC Written fixaminatinn
Senior Reactor Operator
Data Sheets
QUESTION NUMBER 92 TIEWGROUP 211
1OCFRS5 COWTENT: 41(b) 43(b) 2
KA: 06262.2.22
Knowledge of lirriiting conditions for operations and safety limits. (AC Electrical Distribution)
OBJECTIVE:
DEVELOPMENT REFERENCES: OP- 156.02
REFERENCES SUPPLIED TO APPLICANT: S o n e
QIJESTIOX SOURCE: 0X XEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED [DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the loss of both trains of power will provide the proper coincidence, hut power must be
available to the output relays to actuate. Placing the input error inhibit in INHIRI? at this time will
not prevent an actuation since the logic is already made up. Also the incorrect Train.
h. Plausible since the loss of both trains of power causes the SI Block signals to be lost a d when either
of the supplies is restored, power will be available to the output relays to cause an actuation, howevet
this OCCUKS on Train R for this event.
c. Plausible since the loss of both trains of power will provide the proper coincidence, hut power must be
availahle to the output relays to actuate. Placing the input error inhibit in INHIBIT at this time will
not prevent an actuation since the logic is already made up.
X d. The loss of both trains of power causes the SI Block signals to he lost. When either of the supp1ie.s is
restored, power will he available to the output relays to cause an actuation.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must determine train of SSIS affected by the loss of power and then analyze the
effect ofpartially restoring power
Harris NRC Written Examitlation
Senior Reactor Operator
QUESTION: 93
With the plant in Mode 3. a new system engineer has requested that C X P lD-SB be
started with the discharge valve throttled to 75% open to determine starting current and
flow rate under these conditions.
This evolution is NOT described in current procedures, nor the Updated Final Safety
Analysis Report. The system engineer has developed a temporary change to the
Engineering Surveillance Test for perfnrniing the evolution.
The Superintendent - Shift Operations may . . ,
a. approve the evolution when a written safety evaluation has been performed and
approved.
b. approve the evolution if the Manager-HESS concurs
c. approve the evolution if another SRO coilcurs
d. approve the evolution when Mode 5 conditions are met.
ANSWEH:
a. approve the evolution when a written safety evaluation has been performed and
approved.
Elanis NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 93 TIER/GROEIP: 3
1QCFRS5CONTENT: 41(b) 43(b) 3
KA: 2.2.10
Knowledge of the process for determining ifthe margin of sdfsty, as defined in the basis of any technical
specification is reduced by a proposed change, test or experinlent
OBJECTIVE:
DEVELOPMENT REFERENCES: REG-NGGC-0010
REFERENCES SUPPLIED TO APPLICANT: ?;one
QUESTIQN SOURCE: X'EW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: INFO 19840
NRC EXAM HISTORY: None
DISTRACTOR JESTIFICACTION (CORRECT ANSWER X'd):
X a. Tests or experinie.nts that put the facility in a situation that has not previously been evaluated or that
could affect the capability of SSCs to perforni their intcndcd dcsign functions must he evaluated under
10 CFR 50.59 before they arc conducted.
b. Plausible since a temporary procedure has been written to address the test, but the procedure must
have undergone a safety review prior to implementation even if the Manager-HESS concurs.
c. Plausihle since this is a requirement for implementing 10CI:KS0.54(x) for deviating from plant
procedures to protect the health and safety of the public, but this does not qnalify as a condition under
which this would he implemented.
d. Plausihle since the CSIPs are only required to be operable in Modes 1-4. but since the test falls outside
the licensing basis documents a safety evaluation is required.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Requires knowledge of requirements for process of performing a test not
described in any licensing basis documents.
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 94
Given the foilowing conditions:
e ED(; 1B-SH is under clearancc when a Ioss of offsite power occurs
e EDG 1.4-SA starts and is loaded by the Sequencer.
e The BOP infbrms you that ISW Pump 1A-SA has tripped.
Wliich of the following actions are to he taken?
a. Locally monitor EDG 1A-SA operating parameters and continue in IPAI-I
b. Open E I X IA-SA Output Breaker 106 and transition to EPP-001, Loss of AC
Power to 1A-SA and 1H-SR Buses
c. Emergency stop ED<; IA-SA and transition to EPP-001. Ims of AC Power to
IA-SA and 1B-SB Huses
d. Emergency stop EIPG 1A-SA and continue in PATII-I
ANSWER:
c. Emergency stop EDG 1A-SA and transition to EPP-001, 1,oss of.4C Powcr to
Harris NRC Written Exmiiniition
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 94 TIEWGROUP: 1/1
10CFR55 CONTENT: 41(b) 43(b) 5
KA: 000056AA2.14
Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Operational
status of EIYCrs (A and 13)
OBJECTIVE:
DEVELOPMENT REFERENCES: PATH- 1
Users Guide
REFERENCES SUPPLIED TO APPLICANT: None
QLTSTION SOURCE: [7
X NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since the IJsers tiuide allows operating the EDG unloaded withuut ESW flow provided it is
locally monitored, but this is permissible only after the transition to EPP-001 has be.en made.
b. Plausible since a transition is required to E.PP-001 and this would unload the EDC;. but the EDG is tu
be emergency stopped when E S W is not available.
X e. Without cooling water flow, the E.DG is only permitted to operate for 1 rninute while loaded, but this
decision will be made by SRO to stop EDG since it is the only available power source lo the plant.
d. Plausible since the E.DG must he stopped, but without a source of power to both 1.4-SA and IB-SD
buses entry must be made to EPP-001.
DIFFICULTY ANALYSIS:
COMPREHENSIVE /ANALYSIS 0 KNOWLEDGE / KE:CALL
DIFFICULTY RATING: 3
EXPLANATION: Must malyze E I X conditions to make determination to secure o d y source of
power to plant
Harris NRC Written Examination
Senior Keactor Operator
QUESTION: 95
A reactor trip occurred due to a loss of offsite power. 'The plant is being cooled down on
RIIR per WP-006, Natural Circulation Clooldown with Steam Void in Vessel with
RCS cold leg temperatures are 190'F.
0 Steam generator pressures are 50 psig.
0 RVI,IS upper range indicates greater than 100%.
0 T h e e CRDM fans have been running during the entire c.ooidown.
Steam should be dumped from all SGs to ensure ..,
a. boron concentration is equalized throughout the RCS prior to taking a sample to
verify cold shutdown boron conditions.
b. all inactive portions of the llCS are below 200°F prior to complete KCS
depressurization.
c. RCS and SG temperatures are equalized prior to any subsequent RCP restart
d. RCS temperatures do not increase during the required 29 hour3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> vessel soak period.
ANSWER
b. all inactive portions of the RCS are below ZOOOF prior to conipletc liCS
depressurization.
llarris NRC Written Examination
Senior Reactor Operator
Data Sheets
QIJESTION NUMBER: 95 TIEWGROUP: I12
10CFR55 CONTENT: 41(b) 13(b) 2
KA: WEOOG2.1.32
Ability to explain and apply all system limits and precautions. (Natural Circulation Operations)
OBJECTIVE:
DEVELOPMENT REFERENCES: EPP-006
KEFERE.NCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW' SIGNIFICANTLY MODIFIED DIRECT
BANK NCME3E.RFOR SIGNIFICANTLY MODIFIED /DIRECT: 3.8 006
NRC EXAM HISTORY: None
DISTRACTOR JUICTSTIFICACTION (CORRECT ANSWER X'd):
a. Plausible since this action would have been perfonned in this procedure, but must be completed prior
to depressurizing the RCS below 1900 p i g .
X b. SG pressure above 0 psig indicates that the SGs are above 200°F. Depressurizing the RCS under this
condition will result in additional void formation in the SG u-tubes.
c. Plausible since KCP operation throughout NC Cooldown is desirable, but will not be performed at this
point in the procedure.
d. Plausible since a soak period is addressed, but only if continued operation of CKDM fans had not been
maintained.
DIFFICULTY ANALYSIS:
COMPREIIE.NSIVE/ AXALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must analyze the conditions and determine that the entire RCS is not below
200'1: and the effect of depressurizing under these conditions.
Harris NRC Written Examination
Senior Keactor Operator
QUESTION: 96
Given the following conditions:
Refueling in in progress.
A spent fuel assembly is being moved in the Fuel Handling Building (FHB) when its
damaged by contacting a wall of the pool.
- Spent Fuel Pool area radiation monitor RM-lFK-3566A-SA is in HIGH alarm.
- Spent Fuel Pool area radiation monitor RM-lFR-3567B-SB is in ALERT.
Which of the following actions must the Refueling SRO direct?
a. Immediately evacuate the FHB in accordance with AOP-013. Fuel Handling
Accident
b. Immediately evacuate the FHN in accordance with AOP-005, Radiation
Monitoring System
c,. Place the assembly in a safe location and then evasuate the FIIR in accordance
with AOP-013, Fuel Handling Accident
d. Place the assembiy in a safe location and then evacuate the FIIB in accordance
with AOP-005, Radiation Monitoring System
ANSWER:
a. Immediately evacuate the FKB in accordance with AOP-013, Fuel Handling
Accident
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 96 TIENGROUP: 3
10CFRSS CONTENT: 41(b) 43(b) 4
KA: 2.3.10
Ability to perform procedures to reduce excessive levels of radiation and guard against personnel
exposure.
OBJE:CTIVE:
DEVELOPMENT REFERENCES: AOP-013
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: IIarris LOCT 53 I
XRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER Xd):
X a. Any FIIB monitor in High Alarm or Alert requires immediate evacuation in accordance with AOP-
013.
b. Plausible since any FHR monitor in High Alarm or Alert requires immediate evacuation, but these
actions are performed in accordance with AOP-013, not ACT-005.
c. Plausible since it is desirable to place the fuel assembly in a safe location prior to evacuating. but any
FHR monitor in High AIarm or Alert requires immediate evacuation in acc.ordauce with AOP-013.
d. Plausible since it is desirahle to place the fuel assembly in a safe location prior to evacuating. but any
FHR monitor i n High Alarm or Alert requires immediate evacuation in accordance with AOP-013.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the threshold requirenientr for determining whether the fuel
assembly should be placed in a safe location prior to evacuating
IarrisNRC Written Examindim
Senior Reactor Operator
QUESTION: 97
Given the following conditions:
Power is currently at 32% during a plant startup.
o lnstnment Bus IDP-1H-SIV deenergized as a result of a fault in PIC CAB-4.
o PIC CAB-4 has been isolated from Instrument Bus SIV and will be deenergized for
approximately eight (8) hours while repairs are being made.
In ordcr to put the plant in a safe condition . .
a. place all PIC C A B 4 Reactor Xrip instruments in the tripped condition in
accordance with OWP-RP, Reactor Protection.
b. place all PIC CAB-4 ESF instruments in the tripped condition in accordance with
OIW-ESF, Engineered Sdety Feature Actuation.
c. place all MFW Regulating Valves in MA4N1J.4Lin accordance with AOP-024,
Loss of Uninterruptihle Power Supply.
d. perform B plant shutdown in accordance with GIJ-006, T\;ornial Plant Shutdown
From Power Operation to Hot Standby.
ANSWEk
d. perform a plant shutdown in accordance with GP-006, Normal Plant Shutdown
From Power Operation to IIot Standby.
Harris NRC Written Examination
Senior Reactor Operator
ked Sheets
QIJESTION NUMBER: 97 TIEWGROUP: 111
10CFR55 CWNTENT: 41(b) 43(b) 2
KA: 000057G2.2.22
Knowledge of limiting conditions for operations and safety h i t s . (I.oss of Vital A( Instrument Rus)
OBJECTIVE:
DEVELOPMENT REFERENCES: AOP-024
TS Table 3.3-3, pg 3-1 8 and 3-27
IS 3.0.3, pg 0-1
REFERENCES SWPLIED TO APP1,ICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BASK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: rZOP-3.24-R4 001
NKC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Piausihle since instrument failures require bistahles tripped, but they are deenergized tu actuate and
are already tripped sincx no power is available.
b. Plausible since instrument failures require bistables tripped, hut they are deenergized to actuate and
are already tripped since no power is available.
E. Plausible since this is the immediate operator action for a loss of Instrument Bus SUI, not SIV,
X d. Loss of all power to PIC CAB-4 will result in 3 bistable channels of Steam Line Pressiire becoming
inoperable. The IS action is to trip the histables w-ithin six hours, but the histables are energized to
actuate. Without power availahle, this action cannot be performed and IS 3.0.3 beconies applicable
DIFPICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS 0 KNOWLEDGE I RECALL
DIFFICU1,TY RATING: 4
EXPLANATION: Must recognire that energized to actuate bistable5 catmot be placed in tripped
condition without power, thus an entry into TS 3.0.3 is required, and must
determine the required 1s 3.0.3 actions
Harris NRC Written Examination
Senior Reactor Operatotor
QUESTIOK: 98
During the performance ofIATI1-2, the STA reports that the following two (2)
YELLOW path Critical Safety Function Status Trees (C:SI:sI) exist:
e Integity
e Heat Sink
Which of the following describes how these YELLOW paths are to be addressed and / or
implemented:,
a. Both must be addressed and implemented, with IIeat Sink having a higher priority
than Integrity, as soon as PATH-? actions are completed provided no other higher
priority CSFS? conditions exist
b. Both must be addressed. but implemented at the discretion of the Superintendent-
Shift Operations, prior to exiting from the EOP network
c. Both must be addressed and implemented. with Heat Sink having a higher priority
than Integrity, prior to exiting from the EOP network
d. Both must be addressed, but implemented at the discretion of the Superintendent-
Shift Operatiom as soon as PATH-2 actions are completed provided no other
higher priority CSFST conditions exist
ANSWER:
b. Both must be addressed, but implemented at the discretion of the Superintendent-
Shift Operations. prior to exiting from the EOP network
Harris NRC Written Exmination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER 98 TIEWGROUP: 3
10CFR55 CONTENT: 41(b) 43(b) 5
KA: 2.3.22
Knowledge of the bases for prioritiiing safety functions during abnormaI/emcr&encyoperations
OBJECTIVE:
DEVELOPMENT REFERENCES: EOP Users Guide
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK NCJMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since they are to be addressed, but only prior to leaving the EOP network and are not
required to be implemented.
X b. All YELLOW-condition CSFSls should be addressed prior to exiting the EOP network. However, the
operator is allowed to decide if and when to implement. and whether to complete any YEIJLW-
condition FKP.
E. Plausible since they are to be addressed, but only prior to leaving the EOP nehwrk and are not
required to be. implemented.
d. Plausible since they are to be addressed, but only prior to leaving the EOP network and are not
required to be implemented.
DIFPICIJLTY ANALYSIS:
COMPREHENSIVE /ANALYSIS KNOWLEDGE / RECALL
DIFPICULTY RATING: 2
EXPLANATION: Knowledge of the implementation criteria for yellow CSFSTs as directcd by
plant procedures
Harris NRC Writtcn Examination
Senior Reactor Operator
QUESTION: 99
Given the following conditions:
'Ihe plant is operating at 100% power.
Emergency DC Bus UP-1R-SB deenergizes.
Which of the following describes the effect ofthis loss of power on Technical
Specification (TS) required equipment'?
a. 1CS-11, CVCS NORMAL LTDN ISOL. fails shut, meeting the Containment
isolation 'IS requirements
13. Instrument Buses SI11 and SIV maintain power from their inverters, meeting the
onsite pourer distribution TS requirements
c. 'IDAFW Pump loses one stcam supply availability, requiring AFW TS
implementation
d. RCPs lose tripping capability, requiring RCS TS implementation
ANSWER:
c. 'IIIAFW Pump loses one steam supply availability. requiring AFW TS
implementation
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 99 TIER/<:ROUP: 1!1
10CFRS5 CONTENT: 41(b) 43(b) 2/3
KA: 000058G2.1.33
Ability to rec.ogniz.eindications for system operating parameters which are entry-level conditious for
technical specifications. (Loss of DC Power)
OBJECTIVE:
DEVELOPMENT REFERENCES: AOP-025
.rs 3.7.1.2
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTIOX SOURCE:
0 X NEW SIGNIFICAIVTLY MODIFIEI)
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: New
DIRFXT
NKC EXAM HISTORY: None
DISTKACTOR JIJSTIFICACTION (CORRECT ANSWER Xd):
a. Plausible since ICs-I 1 is a containrnent isolation valve and is required to fail shut, but since it is
inoperable TS 3.6.3 must he entered.
1). PlansibIe since instrument h u e s SI11 and SIV will still have power availahle from the inverter, hut IS
3.S.3.1must he entered since it requires the inverter he supplied by the DC bus.
X c. The TDAFW Pump is inoperable due to a loss of power to steam supply valve MS-72 and ?S 3.7.1.2
actions must he applied.
d. Plausible since Train 13 KCP tripping power is lost, but Train A is available and RSS TSs do not
address RCP tripping capability.
DIFFICULTY ANALYSIS:
0 COMPREIIENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATWG: 3
EXPLANATION: Knowledge of the effect o f a loss of power on the operability of the AFW
system
Harris NRC Written Examination
Senior Reactor Operabr
QUESTION: 100
Which of the following actions, in accordance with the EOP-User's Guide, c,m the SKO
direct prior to reaching the step in the LOP?
a. Isolating AFW flow to a single faulted SG prior to determining if an intact SG
exists
b. Securing AFW flow to a known ruptured SG with lewl below the narrow range
c. Securing a CSIP to prevent overfilling the pressurizcr following an inadvertant SI
d. Forlowing AFU' actuation from SC; shrink, throttling AFW flow control valves to
maintain SCf level in the required band
ANSWER:
d. Eollowing AFW actuation from SG shrink, throttling AFU' flow control ~ a l ~toe s
maintain Sct level in the required band
Harris NRC Written Examination
Senior Reactor Operator
Data Shects
QVESTION NUMBER: 100 TIEWGROUP: 3
10CFK55 CONTENT: 41(b) 10 43@)
KA: 2.4.14
Knowledge of general guidelines for EOP flowchart use
OBJECTIVE: 344-040-H6-02
Evaluate the adequacy of AOPs / EOPs for mitigation capabilities during off-nornial conditions
DEVELOPMENT REFERENCES: EOP Users Guide
REFERENCES SUPPLIED TO APPLICANT: Sone
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NIJMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: 3.1Y-KlO18
?fRC EXAM HISTORY: Sone
DISTRACTOR JUSTIFICACTIOS (CORRECT ANSWER Xd):
a. Plausible since the EOP Users Guide addresses isolating a faulted SG as being acceytabie, but only
after verifying that an intact SG is available.
b. Plausible since the FOP Users Guide addresses isolating a ruptured SG as being acceptable, but only
after verifying ihat level in the SC; is above the tubes.
e. Plausible since terminating SI early might he beneficial to prevent filling the pressuriim if the only
event is a spurious SI, hut this may result in further degradation of the plant if another undiagnosed
event is in progress.
X d. Control of SC; level in the nornial band is an acceptable step to be performed out of sequence per tbe
E.OP IJsers Guide.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY IL4TING: 3
EXPLANATION: Must differentiate hetween those actions which could potentially result in
degradation of the plant if taken out of sequence and those actions which would
likely have little impac.t on the operators abilities to diagnose other events.