ML041170060
| ML041170060 | |
| 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: ML041170060 (203) | |
See also: IR 05000400/2004301
Text
INITIAL L
HARRIS EXAM
50-4OO12OQ4-301
-
FEBRUARY 23 27,2004
& MARCH 4,2004 (WRITTEN)
Harris
Draft
Written
2004
Harris NRC Written Examination
Senior Reactor Operator
QBJESTION: 1
Given the following conditions:
While operating at 10004 power, a drop in 1'KZ pressure resulted in a Reactor Trip
and Safety Injection.
PRZ level is currently indicating ::~IO0%.
PRZ pressure has stabilized at 1400 p i g .
Containnicnt pressure is 3.6 psig and stable.
KCPs have been stopped.
RVLIS Full Range is indicating 209.0.
Core Exit Thermocouples are indicating 745°F.
RCS Wide Kange Hot Leg lemperatures are indicating 680°F.
Which of the following conditions currently exists'?
a. A PRZ steam space break has occurred and core heat removal is ADEQIJA'TE
b. A PRZ steam space break has occurred and core heat removal is INADEQITATE
c. An RCS hot leg break has occurred and core heat removal is ADEQUATE
d. An RCS hot leg break has occurred and core hcat removal is INADEQUATE
ANSWER:
b. A PRZ steam space break has occurred and core heat removal is INADEQUA'W
I'ost Validation Revision
IIams NKC Written Examiliation
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: I 'TIENGROUP: iii
10CFR55 CONTENT: 41(b) 4303) 5
MA: 00000XAA2.30
Ability to determine and interpret the following as they apply to rhe Pressurizer Vapor Space Accident:
Iuadequate core cooling
OBJECTIVE: EOP-3.10-4
Given the following EOP steps, notes, and cautions, describe the associated basis
c. RVLIS level of 39 percent (C.1)
DEVELOPMENT REFERENCES: FOP-FRP-C:.1
CSFS'I-Core Cooling
REFERENCES SIJPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
OR SIGNIFICANTLY MODIFIED / DPWECT: New
NRC EXA,M m s m R Y : NOIE
DISTRACTOR JUSTIFICACTPON (CORRECT ANSWER d'd):
a. Plausible since the break is located in the PRZ steam space, but heat removal is not adequate
.! b. The R ( T is superheated and in excess of 70OoF,which indicates that inadequate t e a t r e m o d is
occurring. The break is in the PRZ stearn space as indicated by the pressurizer being fill.
e. Plausible since RCS temperatures are stable, hut the break is in the steam space and heat removal is
not adequate.
(8. Piausible since RCS heat removal is not adequate, but the break is in the steam space.
DIFI~ICXJLTYANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EX:PI.ANATIBN: Must analyze plant conditions to determine location of break, detemiine that
tetnperature indications support superheated conditions and that heat removal is
inadequate
Post Validation Revision
IIaris NRC Written Examination
Senior Reactor Operator
QUESTION: 2
Which of the following describes a condition which would require Emergency Boration
and the bases for taking this action?
a. e Twenty minutes following a Main Feedwater Pump trip, Control Rods are
determined to be below the rod insertion h i t
e Control the reactivity transient associated with a steam line break
b. a Twenty minutes following a Main Feedwater Pump trip, Control Rods are
determined to be below the rod insertion limit
e Control the reactivity transient associated with an inadvertent dilution
c. e During a reactor startup, the Reactor achieves criticality with Rank C rods at
105 steps
e ('ontrol the reactivity transient associated with a steam line break
d. e During a reactor startup. the Reactor achieves criticality with Rank C rods at
105 steps
0 Control the reactivity transient associated with an inadvertent dilution
ANSWER:
c. e During a reactor startup, the Reactor achievcs critical@ with Bank C rods at
IO5 steps
0 Control the reactivity transient associated with a steam Line break
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION SUMBER: 2 TPEWGROUP: 112
80CFR55 CONTENT: 41(b) 43(b) 2
MA: 00002462.2.25
Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.
(Emergency Boration)
OBJECTIVE: CVCS3.0-R4
Given a CVCS componeiitiparameter, state whether the componeiitiparameter is Tech Spec related
DEVELOPMENT REFERENCES: TS Bases 314. I . I
AOP-002 BD
GP-004
REFERENCES SUPPLIED TO APPLICANT: None
QUESTIQN SOURCE: NEW SIGNIFICANTLY MODIFIED
BANK NUMBER FOR SIGNIFICAXT1,Y MODIFIED i D1RE:C:T: AOP-3.2-RI 00 I
NRC EXAM IIISTORY: Kone
DISTRACTOR JUSIIFICACTION (CORRECT ANSWER Jd):
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 />, instead of 20 minutes, Emergency Boration would
be required. Additionally, in Modes I R: 2, SDM is required to control thc reactivity transient
associated with a steam line break. IIowever, 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. Plansibic 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 />, instcad of 20 minutes, Emergency Boration would
bc 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.
d s. Emergency boration is required if SDM is not met. Criticality at steady state conditions is considercd
to he a loss of SDM. In Modes I k 2 , SDM is required to control the reactivity transient associated
with a steam line break.
d. Plausible since Emergency boration is required if SDM is riot met. Criticality at steady state
conditions is considered to be a loss of SDM However, the concern for an inadvertent dilution is
related to a shutdown condition.
DIFFYCUI,TY ANALYSIS:
COMPREHENSIVE i ANALYSIS KNOWI.EDGE i RECALL
DIFFICULTY RtlTING: 2
EXPLANATION: Knowledge of the requirements for initiating Emergency Boration and the bases
for these actions.
Post Validation Kevision
Harris NRC Written Iixarnitiatiuii
Senior Keactor Operator
QUESTION: 3
Given the following conditions:
- The plant has been operating at 100% power for the past three (3) months.
e CSIP 1A-SA is operating.
e C'SIP IB-SB has just been restored to a normal alignment following maintenance on
the pump impeller.
e When CSIP IB-SB is started the operator notes that sucl~onpressure appears normal,
while discharge pressure, discharge flow. and pump current are oscillating.
Which of the following is the niost likely cause of these CSIF I R-SB indications?
a. Inadequate venting was performed during clearance restoration
b. 'Ihe CSIP 1B-SB discharge valve was inadvertently left closed during clearance
restoration
c, A failure of the CSIP 1R-SU miniflow isolation valve has resulted in gas binding
d. A failure ofthe <'SIP 1R-SR tniniflow isolation valve has resulted in all pump
flow being recirculated to the VCT
ANSWER:
a. Inadequate venting was performtvl during ciearanee restoration
Post Validation Revisiun
Hairis NRC Written Examitlation
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 3 TIEWGROUP: 2i I
lOCFR55 CONTENT: 41(b) 43(b) 5
K4: 006.4204
Ability to (a) predict the impacts of the following malfunctions or operations on the EC'CS; and (b) based
on those predictions, use procedures to correct. control, or mitigate the consequences of those
malfunctions or operations: Improper discharge pressure
OBJECIIVE: AOP-1.2-4
Given a set of plant conditions and a copy of AOP-002, determine ifthe possibility of gas binding the
CSIPs exists and the corrective action to be taken
UEVEL.OPMENT REFERENCES: UP-107
SOEK 97-1
REFERENCES SUPPLIED TO APPLICANT': None
QUEKIION SQCRCE: SIGNIFICANTLY MODIFIED DPRECI
CANTLY MODIFIED / DIRECT:
NRC EXAM WISTBRY: None
DISTRACTOR JUSTIFICACTHON (CORRECT ANSWER +d):
d a. Gas binding of a pump results in lower than expected pressure, flow, and current. Likely cause is
improper venting of pump when restoring from post maintenance activities.
b. Plausible since improper alignment would result in low flow and current, but a closed discharge valve
would cause discharge pressure to be high.
c. Plausible since gas binding is cause ofthese indications, but will not occur as a result ofpump recirc
valve being open.
d. I'kausihle since a Fdiled open recirc valve wiil cause indicated flow to be low since flow is measured
dowstream of the recirc valve, hut discharge pressure and current would be at or nrar normal.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / AXALYSIS MNOWL.EDGEI RE.C:AB,L
DIFFICULTY ItlTINC: 3
EXPLANATION: Must analyze giveti pump conditions to determine hilure mode and then
determine likely cause of gas binding ofthe pump
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 4
Given the following conditions:
e The unit is operating at 100% power, with Control Bank D rods at 2 15 steps.
e ALB 13-4-1, ROD CONTROL IIRGENT AI.ARM, is in ALARM due to a failure in
Power Cabinet 1AC.
e Rod ControI is in MAN
e A turbine trip occurs, but the Reactor fiils to trip either automatically or manually.
Which of the following actions should the Reactor Operator be directed to take?
a. Place the Rod Control K4NK SELECTOR in AUTO and allow rods to insert
b. Maintain the Rod Control BANK SELECTOR in MAN and manually insert rods
c. Place the Rod (hntrol BANK SEI,ECTOII in HANK D and manually insert rods
d. Maintain rods at 215 steps
4WSWER:
d. Maintain rods at 2 I 5 steps
Post Validation Revision
Harris NRC' Written Examination
Senior Reactvr Operator
Data Slicets
QUESTION NUMBER: 4 THEWGROUP: 22
llOCFR55 CONTENT: 41(b) J3@) 5
KA: 001G2.4.h
Knowledge of symptom based EOP mitigation strategies. (Control Rod Drive)
OBJECTIVE: EOP-3.19-4
Given a set of conditions during EOP implementation. determine the correct response or required action
based upon the EOP User's Guide general infonnation
z. IJse of "Bank Select" during an ATWS
DEVELOPMENT REFERENCES: IIOP-USERS GUIDE
EOP-FW'-S. i
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MQDIFIEI) DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
QHS'I'RACTOK .IUSTIFICACTION (CORRECT ANSWER +d):
a. Plausible since this is an RNO action for a failure of the reactor to trip. hut will not be successful due
to the urgent failure in rod control.
19. 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.
c. Plausible since this will allow Bank D rods to move inward, and is the only nlcthod of inserting rods
with the rod control failure, but should not be used due to the potential to cause unanalyzed flux
shapes.
d d. Due to the urgent failure: rods will not move in AI.JT0 or MAE. Altlzough they will move in RANK
1) with this particular failure, moving rods in individual banks may result in nnanalyzed flux shapes
which could result in fuel damage.
DIFFICI!LTY ANALYSIS:
COMPREIIENSIVE I AXALYSIS KNOWLEQGE I RECALL
DIFFICXLTY RtPTHNG: 3
EXPIANATION: Must analyze the effect of an urgent rod control failure and then apply the
failure results to the plant conditions to deteimine the proper actions
Post Validation Revision
Harris NRC Written Examination
Senior Rcactor Operator
QUESTION: 5
Four Operators worked the following schedule in the Control Room over the past six
days:
IIOUKS WORKED (Shift turnover time not included. Do NOT assume any hours
worked before or after this period.)
QPEWATOR DAY1 DAY2 DAY3 DAY4 DAYS DAY6
1 IO I4 off 12 12 12
2 14 12 14 IO off 11
3 off off off 13 11 14
4 11 13 14 0ff I1 12
Which of the operators would be permitted to work a 12-hour shift on Day 7 WKTIPOUT
requiring permission to exceed nonnal overtime limits?
a. Operator I
b. Operator 2
c. Operator 3
d. Operator4
ANSWE R:
a. Operator 1
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
L)ata Sheets
QUESTION NUMBER: 5 TIEWGROUP: 3
10CPR55 CONTENT: 41&) 43(b) 5
KA: 2.1.2
Knowledge of operator responsihililies during all niodes of plant operation
OBJECTIVE: PP-2.0-SI
STATE the requirements contained in Administrative Controls Section, including requirements for
the following:
- Unit staff3including overtime limitations
DEVELOPMENT REFERENCES: AP-0 12
REFERENCES SUPPLIED TO APPLICANT: None
QUESI'ION SOURCE: NEW SHGNIFICANTLU MO1)IFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: Kobinson NRC 2001
NRC: EXAM HISTORY: None
DISTRACTOR JUSTIFICACTBON (CORRECT ANSWER d'd):
\' a. Working a 12 h c w shift on Day 7 would result in this operator working 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> out of48, and 72
liours in 7 days, both of which are permissible.
12. Plausible since this operator would not exceed the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> out of48 limit and has had a recent day
off, but would work 73 liours in 7 days which exceeds limit.
c. Plausible since this operator would not exceed the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in 7 day limit and has several recent days
off, but would work more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 48 which exceeds limit.
d. Plausible since this operator would not exceed the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> out of 48 limit and has had a receut day
off; hut would work 73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> in 7 days which exceeds limit.
D~FFIC~ILTY ANAL.YSHS:
COMPREIHENSIVE / ANALYSIS KNOWLEDGE / RECAI,I,
DIFFICULTY k4TINC;: 3
EXPLANATION: Required to compare glven data to adniinistrarive limits to determine which
operator would remain within acceptable overtime limits
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 6
Given the following conditions:
e A Reactor Trip with SI occurs.
e The operators perform the immediate action steps, verify ECCS flow, and check
AFW flow.
operators enter EOP-FRP-11.1 Response to Loss of Secondary Heat Sink.
e KCS pressure is 175 psig.
e All SG pressures are between 300 psig and 350 psig.
Which of the following actions is to be taken!
a. Continue in EOP-FRF-11. I since EOP-FRP-11. I has a hib :r priority than PATI 1
and attempt to establish .4FW or Main Feedw,ater flow.
b. Continue in EOP-FRP-I1.I since EOP-FRP-H.1 has a higher priority than PATII-1
and initiate RCS feed and bleed.
c. Return to EOP-PATII-I at the step that was in effect since a secondary heat sink is
NOT required following a large break LOCA.
d. Return to EOP-PATH-1 at Entry Point C since a secondary heat sink is NOT
requird following a large break I.OCA.
IPNSWEK:
c. Return to EOP-PAI H-1 at the step that was in effect since a secondary heat sink is
NOT required following a large break LOCA.
Post Validatiori Revision
Harris NRC Written Examination
Senior Reactor (jperator
Data Sheets
QUESTION NUMBER: 6 THERKIPOUP: lil
IOCFRSS CONTENT: 48(b) 43(b) 5
KA: 00001 l(i2.4.6
Knowledge of syniptoni based EOP mitigation strategies. (Large Break LOCX)
OBJE.ClWE: EOP-3.11-4
Given the following EOP steps, notes, and cautions, describe the associated basis
e. Requirements for a heat sink (11.1)
DEVEI,OPMENT REFERENCES: EOP-I'W-1.I
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIPICAXTLY MODIFIED
BANK NUMBER FOR SIGNIFICANTLY MODIFIED i DIRECT:
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
a. Plausible since these are actions that are taken upon entry into I'RP-11.1. but a secondary heat sink
would not he required with RCS pressure <' SG pressure.
h. Plausible since these are actions thar might be taken upon entry into FRP-H.1, but a secondary heat
sink would riot he required with RCS pressure -: SG pressure.
4 c. Since RCS pressure is less than SG pressure, a secondary heat sink is not required since the SG would
act as a heat source rather than a heat sink. Return is to procedure and step in effect.
d. Plausihle since RCS pressure is less than SG pressure and a swondary heat sink is not required
Return is to procediire and step in effect, not Pintry Point C.
PCULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE !RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must interpret first that a secondary lieat sink is not required based on RCS
pressure being greater than S G pressure and then must recognize the entry point
conditions for returning to PATH-I
Post Validation Revision
IIarrts NRC Written Examination
Senior Reactor Operator
QUES'I'IQN: 7
Given the following conditions:
e The Reactor has been taken critical and power is being increased.
e NIS IR channels N3S and N36 are both indicating 5 x IO- amps.
e NIS SK channel N3 1 is indicating 8 x 10" cps.
e Ilue to improper adjustment of the high voltage setting, NIS SR c.hanne1N32 is
indicating 7 x 10' cps.
Power should be stabilized ..
a. at or above amps, and the SR High Flux trip should then he blocked.
b. at the current power level, and the SR High Flux trip should then be blocked.
c. at or above IV amps. but the SR High Flux trip should NQT be blocked.
d. at the ciinrent power level, but the SR IIigh Flux trip should KOT be blocked.
ANSWER:
d. at the current power lever, but the SR High Flux trip should NOT he blocked.
Post Validation Revision
Harris NRC Written Examination
Senior Rcactor Operator
Data Sheets
QtJESTION NUMBER: 7 TIENGROUP: 112
10CFK55 CONTENT: 4I(b) 43(b) 5
KA: 0000324A2.09
Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear
Instrumentation: Effect of improper HV setting
OBJEC.TIVE: GP-3.4-1
Recognize off-normal conditions during a reactor start-up, including
a. Availability of excore nuclear instrumentation channels (SK,IR, I'R)
DEVELOPMENT REFERENCES: GP-004
ALB-0 12-4-5
REFERENCES SUPPLIED TO APP1,ICXNT: Xone
QIJESTION SOUIPCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANILY MODIFIED I DIRECT: New
NRC EXAM HISTORY: None
1)IS'FRACTCPR JUSTIFICIACTION (CORRECT ANSWER v"d):
a. Plausible since power niust he increased above 10 l o amps before blocking trips, but increasing power
to this level will result in SR high flux trip.
b. Plausible since power cannot be increased above amps, but the block ofthe SR high flux trip is
interlocked at this power level.
c. Plausihle since the SI< high flux trip is not pennitted to be blocked without at Icast I decade of overlap
hetween SR and IK, but increming power above I O I" amps will result in a SR high flux trip.
I
3 cl. Ixss than I decade of overlap exists hetureen SR and BR channel before trip would occur. Increasing
power to allow blocking SR would result in trip before reaching power level and attempting to block
at current power level will not be successful.
DIFFICULTY ANALYSIS:
CCPillPREIlENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULI'Y RATING: 3
EXPLANATION: Must determine that increasing power above IO."' amp will result in a rcactcir
trip due to SR high flux, and that attempting to block the SW high flux trip
below 10 I" amps will not be successful. Required to not block SR high flux trip
if < 1 decade of overlap exists.
Post Vaiidation Revision
Harris NRC Written Fxamination
Senior Reactor Operator
QIJESTION: 8
Given the following conditions:
EOP-FRP-S. I , Response to Nuclear Power Generatio~v.AT\h,is bcing
impleniented.
4 An SI actuation has O C C U I T ~ .
8 The Foldout page is applicabie.
Which of the following actions should be taken?
a. Continue with FJOP-FRP-S. I while verifying proper operation of safeguard
equipment
b. Continue with EOP-FRP-S. I until the reactor is tripped or niade subcritical, then
imniediately exit to EOP-PATH-I
c,. Transition to EOP-PATH-1 and verify all autornatic actions required for an SI
have occurred, then return to EOP-FRP-S. 1 only when directed by PATH-1
d. Reset SI and FW isolation as soon as possible to restore feed flow to the steam
generators, then continue with EOP-FRP-S. 1
ANSWER:
a. Continue with EOP-FRP-S. 1 while verifying proper operation of safeguard
equipment
Post Validation Revision
Harris NRC U'ritten Examination
Senior Reactor Operator
Data Sheets
QUESTIOX NUMBER: 8 mcwGRouP: 2!'I
10CFR55 CONTENT: 41(b) 43(b) 5
ILa: 012G2.4.6
Knowledge of symptom based EOP mitigation strategies. (Reactor Protection)
OBJECTIVE: FOP-3. I S
Uescrihe the purpose of the following ROPs including tyFe of event for which they were designed and the
major actions performed
- FEW-S.1
DEVELOPMENT REFERENCXS: EOP-FW-S. 1
EOP User's Guide
REFERENCES SlJPPLIED TO APPLICANT: None
QUESTION SOURCE: 6] NEW SIGNIFICANTLY MODIFIED DIRECT
BASK NUMBER FOR SIGNIFICANTLY MOIXFIED I DIRECT: EOP-l.ls 021
NHC: EXAM HISTORY: Harris NRC 2000
DISTRACTOR .FUSTIFICACTI(4N (CORRECT ANSWER d'd):
-\I a. If a safety injection occurs while implementing FRP-S. 1. proper operation of SI equipment is verified
while continuing with FKP-S. I .
b. Plausible since PA?H-I provides instructions for a response to safety injection, but FKP-S.I niust be
performed until completion.
c. Plausible since PATH-I provides instiuctions for a response to safety injection, but FKP-S. 6 must be
performed until completion.
d. I'lriusible since a safety injection will result in a loss of'MFW, but AFW flow is capable of providing
minimum required flow.
PCULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE IRECALI.
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of procedural requirements in EPP-FRP-S. I
Post Validation Rcvision
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 9
Given the following conditions:
a The plant is in Mode 3 with all Shutdown Rods withdrawn.
a All power is lost to the Digital Rod Position Indication display and CANNOT be
restored.
Which of the following actions is to be taken'?
a. Verify that a11 Shutdown Bank Rods are fully withdrawn using Demand Position
indication
b. Determine that all Shutdown Bank Rods are fully withdrawn using the movable
incore detectors
c. Commence a boration of the RCX to ensure adequate Shutdown Margin
d. Open the Reactor Trip Breakers
ANSWER:
d. Open the Reactor 'Trip Breakers
Post Validation Revision
Harris NRC Written Examination
Seniot Reactor Operator
Data Sheets
QtJESTION NUMBER: 9 TIEWGRQUP 2i I
IOCFR55 CONTENT: 41(b) 43(b) 5
%(A: 014.42.02
Ability to (a) predict the impacts of the following malfunctions or operations on the WIS; and (b) based
on those on thosc predictions, use procedures to correct, control, or mitigate the consequences of those
malfunctions or operations: Loss of power to the IWIS
OBJECTIVE: RODCS-3.1-R4
Given a copy of 'Technical Spccitications and a plant mode, determine if rod position indication
components and actual rod positions meet their Limiting Conditions for Operation; if they do not, then the
applicable .4C'I'ION statements
DEVELOPMENT REFERENCES: 'rs 3.1.3.3
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOCRCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
a. Plausible since this would be required in the event o f a loss o f a single indication whiie operating in
Mode 1 or 2 , but with both indications lost in Mode 3 the Reactor Trip Breakers are to be opened.
b. I'lausible since this wouid be required in the event of a loss of a single indication while operating in
Mode 1 or 2, but with both indications lost in Mode 3 the Reactor 'I'rip Breakers are to be opened.
c. Plausible since Loss of indication of UKI'I may lead to belief that Sr)M cannot be verified; which
would require Emergency Ihration.
d d. With both DRPI indications inoperable in Mode 3,4, or 5, 'TSrequires that the Reactor Trip Breakers
be opened immediately.
DIFFICULTY ANALYSIS:
CQMPREIIENSHVE / ANALYSIS KNO\VLEDGE I RECALL
DIFFICBJLTY RATISG: 2
EXPLANATIQN: Knodedge of Tech Spec immediate action requirements in the event of a Loss
of both DRPI indications
Post Validation Revision
Harris NRC' Written Examination
Senior Reactor Opelator
QUESTION: 10
A licensed Reactor Operator has failed to meet the required number of hours this past
calendar quarter to maintain an active license.
Assuming all other requirements have been met to activate the license. 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 Control Operator during 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 Control
Operator during Mode 4
b. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> as the Balance of Plant Operator during Mode 5 AND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as the
Control Operator during Mode 4
c. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> as the Control Operator during Mode 5
d. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> as the Balance of Plant Operator during Mode 4
AN sw Ew :
d. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> as the Balance of Plant Operator during Mode 4
Post Validation Kevision
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 10 TIEHIGROUP: 3
10CFR55 CONTENT: 41@) 43(b) 5
KA: 2.1.1
Knowledge of conduct of operations requirements
OBJECTIVE: PP-3.1-1
Given a situation, STATE whether or not an off-going operator may he reelieved during the shift turnover
process
DEVE1,OPMENT HEFERENCES: OMM-001
REFERENCES SUPPLIED TO APPLICANT: None
QCKSTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
IMWK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC EXAM IIISTOHY: None
DISTRACTOR JUSTIFBCACTION (COHRECT -4NSWER dd):
a. 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 (0or HOP position, but only those hours
when the plant is above 200°F are acceptahle.
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 C:O or BOP position, but only those hours
when the plant is above 200F arc acceptable.
e. Plausible since this meets the required 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> for the CO or BOP position aud this has the most
hours in the CO position, but only those hours when the plant is above 200°F are acceptable.
t d. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> are required in either the CO or BOP position when the plant is above 200°F.
DIFFICZl1,TY ANALYSIS:
0 COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 2
EXPL.ANATION: Must recall requirement for activating an inactive license from OMhl-00 I
Post Validation Revision
Harris NRC Written Examination
Senior Kcactor Operator
QCESTIQN: 1I
Following a loss of off-site power during recovery from a SGTR, the crew is required to
transition from EPP-019, Post SGIR Cooldown Using Steam Dump, to either:
e EPP-017, Post SGTR Cooldown Using Hac.kfill
EPP-018, Post SGTR Cooldown Using Blowdown
Which of the following describe how RCS and S G pressure control in EPP-017 compares
to that in EPP-018?
a. o EPP-017 maintains KCS pressure below the ruptured S G pressure
e EPP-0 I8 maintains RCS pressure below the nlptured SG pressure
. o EPP-017 maintains KCS pressure below the ruptured S G pressure
8 EPP-OI 8 maintains KCS pressure the same as the niptured SG pressure
c. 8 EPP-017 maintains KCS pressure the same as the ruptured SG pressure
e EPP-OL 8 maintains RCS pressure below the ruptured SG pressure
d. e EPP-017 maintains RCS pressure the same as the ruptured SG pressure
o EPP-018 maintains RCS pressure the same as the ruptured SG pressure
ANSWER:
. o
o
EPP-017 maintains RCS pressure below the ruptured SG pressure
EPP-018 tnaintains RCS pressure the same as the niptured SC; pressure
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Opemtor
Data Slicets
QUESTION NUMBER: 1I TIEWGROUP: 111
IOCFR55 CONTENT: 41(b) 43@) 5
KA: 000038EA2.08
Ability to determine or interpret the following as they apply to a SGTR: Viable alternatives for placing
plant in safe condition when condenser is trot aliailable
OBJECTIVE: EOP-3.4- I
Describe the purpose of the following EOPs including the type of event for which they were desiped atid
the major actions performed
- EPP-0 17
- EPP-0 I 8
- wP-0 I9
DEVELOPMENT REFERENCES: EPP-0 17
EI'P-0 I 8
REFERENCES SUPPLIED TO APP1,ICANT: None
QUESTION SOURCE: NEW SIGNIFKANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 8PP-3.4 0 I0
NRC EXAM HISTORY: Harris 2002
DISTRACX'OR JUSTIFICACTION (CORREC'I' ANSWER d'd):
a. Plausible since EPP-0 I7 maintains pressure below ruptured SG pressure, but EPP-0 IS mainpains
prcssurt: the same as the ruptured S G pressure.
d b. PP-017 maintains pressure helow SG pressure to allow backfill from the SG to the KCS,while EPP-
018 maintains pressure the same as SG pressure to minimize SG leakage.
c. Plausible since either EPP-014 or EPP-018 maintains pressure below SG pressure and either EPP-014
or EPP-018 maintains pressure the same as Sci pressure, but this distracter has the correct conditioti
reveresed.
d. Plausible since EPP-0 18 maintains pressure the same as the ruptured SG pressure, hut EPP-0 17
maintains pressure below ruptured SG pressure.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / REC:AI,I,
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of different mitigation strategies for EPP-0 I7 and EPP-0 I8
Post Validation Kevisioti
Harris NKC Written Examination
Senior Rcactor Operatoi
QUESTION: 12
A LOCA occurred several hours ago. Only one (1) Containment Spray Pump is running
due to actions taken in EPP-012, 1,oss of Emergency Coolant Recirculation.
A transition has just been made to FKP-J. 1 Response to High Containment Pressure.
Containment Pressure is 14 p i g .
Which of the following actions should be taken?
a. Start the second Containment Spray Pump if Containment pressure does NOT
decrease below I O psig before exiting FRP-J. 1.
b. Start the second Containment Spray Pump since pressure is above 10 psig.
C. Continue operation with one Containment Spray Pump regardless of any increase
in Containment pressure.
d. Continue operation with one Contaimnent Spray Pump unless Containment
pressure begins increasing, then start the second pump.
ANSWER
c. Continue operation with one Containment Spray Pump regardless of any increase
in Containment pressure.
Post Validation Revision
Harris NRC Written Exdminatioll
Senior Reactor 0pe:ator
Data Sheets
QUESTION NUILgBEW: 12 TIEWGROCP: 1i2
10CPH55 CONTENT: 41(b) 43(b) 5
KA: WE14EA2.2
Ability to determine and interpret the following as they apply to the (High Containnient Pressure}
.4dherence to appropriate procedures and operation within the limitations in the faeilitys license and
amendments
OBJECTIVE: EOP-3.13-5
Given the following E.OP steps, notes, and cautions, describe the associated basis: h. CNMT spray
operation (EPP-012 or FW-J.1)
DEVELOPMENT REFERENCES: EOP-FW-B. 1
KEFERERCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED f D%RECI: EOP-3.13-R4 OOX
NRC EXAM HISTORY: None
DISTMCITOR SUSTIFICACTION (CIORRECT ANSWER dd):
a. Plausible since this \n.ould be a normal action directed by FIU-J. I
b. Plausible since this would be a nonnal action directed by FRP-J.1
d 6. EIP-012 directs the operators to run Containment Spray Pumps based upon Containment pressure and
Fan Cooler operation. lhese actions are taken to minimize RWST depletion. This configuration is to
be maintained even if FIZP-J. I is implemented.
d. Plausible since would better serve the intent of EPP-012, but would be contradictory to the intent of
FKP-J.i which has a higher priority concerning the operation of the Spray Pumps.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
nmwx;r,n~RATING: 3
EXPLANATIOK: Must compare the relative actions in the 2 procedures and make a judgement of
which condition takes precedent
Post Validation Revision
Elanis NRC Written bxamination
Senior Reactor Operator
QGESTION: I3
During operation at 100% power, an inadvertent SI occurs on 'R' Train ONLY.
Which of the following actions is required?
a. Maiiuaily actuate SI on 'A' Train and continue in PATII-I
b. Continue in PATH-I noting which 'A' Train ESF equipment is NOT running
c. Start ONLY the 'A' Train of ESF equipment for which the redundant 'B' Train
equipment failed
d. Transition directly to EPF-OOXI SI 'l'ermination
ANSWER:
a. ktanually actuate SI on 'A' Train and continue in I'AI'lB- 1
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QIJESTIONNU,MBER: 13 TIEWGROUP: 2i
10CPH55 CONTENT: 48(h) 43(b) 5
KA: 013A2.01
Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b)
based on those predictions, use procedunx to correct, control, or mitigate the conscquences of those
malfunctions or operations: L(3CA
OBJECTIVE: IE-3. IO-K4
Describe the expected operator actions associated with an imminent RPS or ESFAS actuatiun
DEVELOPMENT REFERENCES: EOP Users Guide
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: XEW SIGNIFICANTLY ILlODPFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: IE-3.10-R4 001
NHC EXAM HISTORY: Ilamis 2000
DISTRACTOR JUSTlFICACTION (CORRECT ANSWER dd):
d a. Prefcned method ofniannal actuation although it would be acceptable to start ;reposition all
equipment which would be actuated regardless ofthe perceived need since diagnostics have not yet
been perfornied.
b. Plausible since only a single train actuation is analyzed, but efforts are to be made to initiate both
trains.
E. Ilausible since starting equipment as needed would provide adequate protection, but since diagnostics
have not yet been completed the equipment required map not yet be known.
(8. Plausible since one of the goals following an inadvertent Si is to terminate SI as soon as criteria are
met to prcvent overfilling !pressurizing the KCS, but procedures are written assuming both trains
started.
DIFFHCULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE / RECALL
DIFFICIJI.TY RATING: 3
EXPLANATION: Required knowledge of procedural requirements for a single train of ESF
actuation
Post Validation Revision
Harris NKC' Written Examination
Senior Reactor Opetator
QUESTION: 14
Given the following conditions:
e 1CS-235, Charging Line Isolation, was closed to establish a clearance boundary for
maintenance on ICs-238.
e 1CS-235 had to be inanuailg torqued shut.
e 1CS-235 is a Liinitorque SMR-OO!SR-OO motor-operated valve.
Prior to declaring 1CS-235 operable after the clearance is removed, the valve must be ...
a. verified to have the torque switch c.alibrated correctly
b. stroked with the control switch.
c. monitored for seat leakage.
d. manually stroked full open.
ANSWER:
b. stroked with the control switch.
Post Validation Revision
IIarris NRC Written Exarninatioii
Senior Rcactor Operator
Data Sheets
QUESTION NUMBER. 14 TIEWGROUP: 3
1QCFR55CONTENT: 4I(b) 43Eb) 5
KA: 2.2.19
KnowIedge of maintenance work order requirements
OBJECTIVE: PP-2.4-1
Identify the primary functions and explain the responsibilities of the Work Coordination Centet
DEVELOPMENT REFERENCES: OMh.1-014
REFERENCES SIJPPLIED TO APPLICANT: None
QUMTIOX SOURCE: SIGNIFICA?XI,Y MODIFIED DIRECT
CANTLY MODIFIEI) / DIRECT: 028
NRC EXAM HISTORY: Hareis 2000
DISTKACTOR JIWTFHFICACTION (CORRECT ANSWER 4'61):
a. Plausible since the valve has been nianually torqued onto the seat, hut the requirement is that the valve
must he stroked electrically from the control switch.
d b. All Lirtiitorque SMI3-00!SB-00 motor operated valves, if manually operated, are required to be stroked
electrically from the control switch to he declared operable.
c. %"ible since over torqueing a \tahe may result in seat leakage, but the requirement is that the valve
must he stroked electrically from the control switch.
d. Plausible since the valve was manually torqued closed, hut the requirement is that the valve must be
stroked elecrricaliy from the control switch.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS I(NOWLE1)GE / RECALI,
DIFFICULTY RATING: 3
EXPLAKATION: Kriowlc.dge of administrative post-work practices required
Post Validation Revision
Harris NR( Written Examination
Senior Rcdctor Operator
QUESFHBN: 15
Given the following conditions:
o Following a Reactor Trip and Safety Injection, a transition has eventually been made
to EOP-EPP-015, Uncontrolled Depressurization of All Steam Generators.
o Both Main and Auxiliary Feed Flow h a w been isolated to all SGs.
o Directions have just been given to locally isolate steam flows from all SGs.
o SG A pressure appears to have stabilized at approximately 100 psig, while the other
X i s have completely depressurized.
Which of the following actions should be taken?
a. Transition to EOP-EPP-014. Faulted S G Isolation, since this is indication that
SG A has been isolated.
b. Continue in EOP-EPP-015 and re-establish AFW flow to SG A at minimum
flow.
c. Transition to EOP-IATK-2 if lwa1 radiation surveys indicate priinary-to-
secondary leakage is occurring.
d. Transition to EOP-T;PP-OOR, SI Termination, to prevent overprcssurizing the
RCS.
AWSWEK:
c. Transition to EOP-PATH-2 if local radiation surveys indicate primary-to-
secondary leakage is occurring.
Post Validation Rei2ision
Harris NRC' Written Fxamiiiation
Senior Keactor Operator
Data Sheets
QI!ESTION NUMBER: 15 TIERIGROUE 1/1
80CPR55 COYIENT: 41(b) 43(b) 2
Iw: 000040Ci2.I.32
Ability to explain and apply all system limits and precautions. (Steam 1 . k Rupture - Excessive Heat
Transfer)
OBJECTIVE: EOP-3.9-7
Given a step, caution, or note from an emergency procedure, slate its purpose
DEVELOPMENI' REFERENCES: EC)P-EPP-015
REVERENCES SUPPLIED TO APPI,ICANT: None
BANK NUMBER FOR SIGNIFICANT1,Y MODIFIEI) / DIRECT: New
NRC EXAM IIISTORY: None
DlSTRACTOR JUSTHFYCAC.TION(CORRECT ANSWER d'd):
a. Plausible since once a SG is confirmed to be isolated in EPP-0 15, a foldout page item directs a
transition to EPP-014.
b. Plausible since without indications o f a SC; tube leak, actions would he taken to remain in EPP-015
and maintain feed flow at minimum.
4 c. A SCi may he suspected to he iuptured if it fails to dry out following isolation of feed flow. Local
checks for radiation can be used to confirm ~ r i n i a r ~ ~ t o - s ~ cleakage.
otida~
41. Plausible since a desired goal after isolating a faulted S G is to terminate SI as soon as conditions are
met to prevent overfilling and overpressurizing the IICS.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNQWLEDGE I RECALL
DIFFICULTY KhTlllL'G: 3
EXPLANATION: Must analyze the cause ofthc failure of the SG to depressurize and then
determine the correct actions based on the analysis.
Post Validation Revision
I h - i s NRC Written Exaniination
Senior Keactor Operator
QUESTION: 16
The unit has tripped due to a I.OCA and ESF equipment has failed to start. As a result,
EOP-FRP-C.2, Response to Degraded Core Cooling, has bcen entered.
A depressuridon of the Steam Generators (SGs) to 80 psig is being performed, in
accordance with the procedure, when the STA reports that a Red Path condition for Integrity
has occurred.
Which of the following actions slriould be taken?
a. Inirnediately transition to EOP-FRP-P. 1, Response to imminent Pressurized
Thermal Shock Conditions
b. Stop the S!G depressurization and, if the red path does not clear, transition to EOP-
FRP-P. I , Response to Imminent Pressurized Thermal Shock Conditions
c. Coniplete FOP-FRP-C.2 and then transition to EOP-FRP-P. I , Response to
Imminent Pressurized Thermal Shock Conditions, if the red path still exists
d. Complete the S!G depressurization and then transition to EOP-FRP-P. 1. Response
to Imminent Pressurized Thermal Shock Conditions, if the red path still exists
ANSWER:
c. Complete FOP-FW-C.2 and then transition to FOP-FRY-P. I , Response to
Imminent Pressurized Thermal Shosk Conditions, if the red path still exists
Post Validation Revision
Harris NKC Written Examination
Senioi Reactor Operator
Data Sheets
QUESTION NUMBER: 16 TIEWGROUP: 1!3
1QCFR55CONTENT: 41(b) 43(b) 2
KA: WEObG2.1.32
Ability to explain and apply a11 system limits and precautions. (Degraded Core Cooling)
OBJECTIVE: IiOP-3.10-4
Given the following EOP steps, notes, and cautions, describe the associated basis
g. Stopping SG depressurization at 80 p i g (C.2)
DEVEEOPMENT REFERENCES: EOP-FRP-C.2
REFERENCES SUPPLIED TO APP1,ICANT: Kone
QUESTION SOBJRCE: SIGNIFICANTLY IL1B)DBFIED DIRECT
CAXTLY MOMFIED /DIRECT: New
NRC EXAM HISTORY: None
DETRACTOR JI!STIFHCACTION (CORRECT ANSWER %Id):
a. Plausible since the red path for integrity has a higher priority than the orange path that caused entry
into I<OP-FKI-C.2, but under these particular conditions a transition should not occur until completion
of the EOP-FKP-C.2.
b. Plausible since the red path for integrity has a higher priority than the orange path that caused entry
into EOP-FKP-C.2, but under these particular conditions a transition should not occur until conipletion
of the EOP-FKP-C.2.
d c. During the depressurization, a red path may occur due to injecting the accumulators. A transition
should not be matie 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
into EOP-IXP-C.2. but uiidcr these particular conditions a transition should not occur uritil completion
of the EOP-FKP-C.2.
DIFFICULTY AWALYSHS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPIANATION: Must analyze plant cotiditions to determine that the cause of the red path is the
depressurization and that, under these specific conditions, an immediate
transition is not warranted
Post Validation Kevision
Harris NRC Written Exammation
Senior Reactor Operator
QUESTION: 17
Given the following conditions:
e The unit is in Mode 3.
0 Instrument I3uses IDP-IB-SI1 and IDP-IB-SIV are both de-energized.
0 Maintenance reports that Instrument Bus IDP-IB-SI1 is ready to be re-energized.
In order to prevent an inadvertent Safeguards Actuation, which of the following tnust be
verified prior to re-energizing the bus and why?
a. Train A Logic Input Error inhibit must be verified to he in INHIBIT due to the
proper coincidence for an actuation being available
b. Train A Logic Train Output must be verified to be in T E S l to prevent an
inadvertent Safeguard Actuation due to the loss of the SI BLOCK Signals
c. Train R Logic Input Error Inhibit must be verified to be in INHIBII due to the
proper coincidence for an actuation being available
d. Train B Logic Train Output must be vcrified to be in TESI to prevent an
inadvertent Safepard Actuation due to the loss ofthe SI BLOCK Signals
ANSWER:
d. Train 3 Logic Train Output must be verified to be in TFST to prevent an
inadvertent Safeguard Actuation due to the loss ofthe SI BLOCK Signals
Post Validation Revision
Waris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NWMBER: 17 IIEWGROIJP: 2: I
IOCPRS5 CONTENT: 4I(b) 43b) 2
KA: 062G2.2.22
Knowledge of limiting conditims for operations and safety limits. (.4CElectrical Distribution)
OBJECTIVE: ESFAS-3.0-4
Given applicable logic diagrams and a set of plant conditions. predict how loss of any of the four
instrument buses will affect the ESFAS output functions of each SSPS train.
DEVELOPMENT REFERENCES: OP-156.02
REFERENCES SGTPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY b1ODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIPICANTLY MODIFIED / DIRECT: New
NRC EXAM TORY: None
DPSTRACTOW .JUSTYFICAC:TlQSN(CORRECT ANSWER kd):
a. Plausible since the loss of hoth trains of power will provide the proper coincidence, but power must he
available 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. Also the incorrect Train.
b. Plausihle since the loss of both trains of power causes the SI Block signals to he lost and when either
ofthe supplies is restored, power will he available to the output relays to cause an actuation, however
this occurs on Train B for this event.
c. Plausihie since the loss of both trains of power will provide the proper coincidence, hut power must he
available to the output relays to ac.tuate. Placing the iuput error inhihit in INHIBIT at this time will
riot prewnt an actuation since the logic is already rnade up.
4 d. The loss of both trains of power causes the SI Block siguais to he lost. When either of the supplies is
restored, power will he available to the output relays to cause an actuation.
I)IFFICI!LTY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWIXDGE /RECALL
DIFF%CI!LTYILITING: 3
EXPLANATION: Must deteimine train of SSPS affected by the 10% of power and then analyze the
effect of partially restoring power
Post Validation Revision
f Iarris NKC Written Examination
Senior Reactor Operator
QUESTION: 18
The Unit-SCO and Superintendent-Shift Operations are discussing invoking
1OC'FR50.54(x) during the implementation of the F'mergency Operating Procedures due
to a condition arising which is NVQ'I'addressed by the procedures or Technical
Specifications.
\J%ich of the foilowing conditions must be met when invoking 1OCFR50.54(x)?
a. The action must be approved by an additional licensed Senior Keactor Operator
when the action is necessary to prevent equipment damage.
b. The action must be approved by the Superintendent-Shift Operations prior to
taking the action.
c. The NRC must concur with the action to be taken prior to the action actually being
taken.
d. The action must be approved by the Manager-Operations when the action is
necessary to protect plant personnel.
ANSWER:
b. The action must be approved by the Superintendent-Shift Operations prior to
taking the action.
Post Vaiidation Revision
Harris NRC Written Examination
Senior Reactor Operator
Dzta Sheets
QUESTION NUAMBER: 18 TIEW<;ROUP: 3
10CFR55 CONTENT: 41@) 43(b) 3
KA: 2.2.10
Knowledge of the process for determining if the margin ofsafety, as defined in the basis ofany t e c h n i d
specification is reduced by a proposed change, test or experiment
OBJECTIVE: PP-2.0-S2
LIST the actions required by the individual who authorizes a deviation from the Technical Specifications
or license conditions
DEVELOPMENT REFERENCES: PRO-NGGC~-0200
REFERENCES SUPPLIED TO APPLICANT: XOIK
QUESrXON SOURCE: SIGNIPICANTLY MODIFIED DIRECT
CANTLY MODIFIED / DIRECT: 0 23318
NRC EXAM HISTORY: None
DIS~RAC~TOK JUSTIPIC.4CTION (CORRECT ANSWER >I'd):
a. Plausible since lOCFRS0.54(x) esquires that a licensed SRO approve any actions which deviate from
license conditions prior to perfomlance. but the actions must be to protect the health and safety of the
public.
4 b. The niinimurn level of approval per PRO-NGGCX200 is the Supcrintendent-Shifr Operations, but it
can be approved by any personnel holding an SRO license above this position also.
c. Plausible since the XRC must be notified, but the notification requirements are within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per AP-
617.
d. Plausible since the Manager-Operations can approve a deviation if he holds an SKO license, but the
actions must be to protect the health and safety of the public.
m r L m ANALYSIS:
COMPREHENSIVE./ ANALYSIS KNOWLEDGE /RECAI,I.
DIFFlCULTY RATING: 2
EXPLANATION: Requires knowledge of requirements for process of perfoiming actions not
described in any licensing basis documents.
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
QBJESTYON: I9
Given the following conditions:
e t d partiall: loaded.
Following a INSS of All Power, EDG 1A-SA ha b :n r e s ~ ~ r and
e A transition has been made to FOP-EPP-003. "Loss of All AC Power Recovery with
SI Required."
e EDG IA-SA is currently loaded to 4.5 MWe and 3.5 MVAR.
Which of the following would result in an UNACCEPTABLE loading condition for EDG
a. 0 Pick up an additional 0.5 MWe
e Pick up an additional 0.1 MVAK
b. e Pick up an additional 1 .O MWe
e Pick up an additional 0.5 MVAR
c. 0 Pick up an additional I .5 MWe
0 Pick up an additional 1 .O MVAR
d. e Pick up an additional 2.0 MWe
0 Pick up an additional 1.2 MVAR
ANSWER:
c. e Pick up an additional 1.5 MR7e
0 Pick up an additional 1.O MVAR
Post Validation Revision
IIarris NRC: Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: I O TIEWGROUB: lil
10CFR.55 CONTENT: 41(b) 43(b) 5
KA: 0011056.4A2.14
Ability to deterniine and interpret the following as they apply to the Loss of Offsitc Power: Operational
status of EDKis (A and B)
OBJECI'IVE: EOP-3.7-6
Given a step, caution, or note from K?P-001, EOP-002, or EOP-003, state its purpose
DEVIELOPIIPENT REFERENCES: OP-155, Attachment 9
EOP-!PI'-003
REFERENCES SUPPLIEI) TO APPLICANT: OP-155, Attachment 9
QUESTION SOURCE: XEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR S'IGNIFICANT1,Y ILIODIPIEI) I DIRECT: N~\v
NRC EXAM I m m R Y : NOIK
DISTRACTOR .IUSTIFICAC:TIION (CORRECT ANSWER +d):
a. Plausible since new loading will be 5.0 MWe and 3.6 MVAR, which is just within acceptable limits.
b. Plausible since new loading will he S.5 MWe and 4.0 MVAR, which is just within acceptable limits.
4 c. New loading will he 6.0 MWe and 4.5 MVAR, which is outside acceptable limits
d. Plausible since new loading will be 6.5 MWe and 4.7 MIVAK, which is just within acceptable limits.
DIFFICUL'I'Y ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOW'IXDGPS I RECALL
DIFFICULTY RATING: 3
EXPEANATIOY: Must analyze EDG operability curve tu determine whether additional MWe and
MVAK loading is within acceptable limits
Post Validation Revision
Iarris NKC Written Examination
Senior Reactor Operator
QUESTIOK: 20
h reactor trip occurred due to a loss of offsite power. The plant is being cooled down on
RIIK per EPF-006. Natural Circulation Cooldown with Steam Void in Vessel with
e RSS cold leg temperatures are 190'F.
e Steam generator pressures are 50 ps~g.
e RVLIS upper range indicates greater than 100%.
e Three CRDM fans have been running during the entire cooldown.
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 RCS are below 200°F prior lo complete KCS
depressurization.
c. RCS and SG temperatures are equalized prior to any subsequent RCP restart.
d. RCS ternperatures 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.
AFJSWER:
b. all inactive portions ofthe RCS are below 200'F prior to complete RCS
depressurization.
Post Validation Revision
Harris NRC Written Esainination
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER: 20 TIEWGROUP Ii2
lOCFRS5 CONTEXT: 41(b) 436b) 2
K k . WE09G2.1.32
Ability to explain and apply all system limits and precautions. (Natural Circulation Operations)
OBJECTIVE: EOP-3.8-2
Dcmotistrate the helowassumed operator knowledge from the SHNPP Step Ihkation 1)ocument and the
WOG EI1Gs that support perhniance of EOP actions: Determining that upper head and SG IJ-tube
temperatures are below 200 F
DEVELOPMENT REFERENCES: FOP-EPP-MJB
REFERENCES SUPPLIED TO APPLICANT: None
QIJESTION SOURCE: SIGNIFICANTLY ?&ODIFI[EU [3DIRECT
CANTLY MODIFIED / DIRECT: EOP-3.S 006
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTHON (CORRECT ANSWER ?Id):
a. Plausible siiice this action would kave been performed in this procedure, but must he completed prior
to depressurizing the RCS below 1900 psig.
4 b. SG pressure above (I 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, hut will not be performed at this
point in the procxdure.
d. Plausible since a soak period is addressed, but only if continued operation of CKDM h u s had not been
maintained.
DHFFICWLTY ANALYSIS:
COMPREHENSIVE / ANALYSIS 0 KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPI.ANATIOX: Must analyze the conditions and determine that the entire RCS is not below
200°F and the effect of depressurizing uiider these conditions.
Post Validation Revision
Harris NKC Written Examination
Senior Reactor Opcrator
QUESTION: 21
During an emergency, a worker has been directed to enter a high radiation area and
perform a repair necessary for the protection of valuable property.
In accordance with PEP-330, Radiological Consequences, the workers exposure
should be limited to ., .
a. 10 Rem TEDE and the entry does NOT require specific Site Emergency
Coordinator authorization.
b. 10 Rem TEDE and the entry requires specific Site Emergency Coordinator
authorization.
c. 25 Rem TEDE and the entry does NOT require specific Sitc Emergency
Coordinator authorization.
d. 25 Rem TEDE and the entry rtyuires specific Site Emergency <:oordinator
authorization,
ANSWER:
b. 10 Rem TEDE and the entry requires specific Site Emergency Coordinator
authorization.
Post Validation Revision
Harris KRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 2 I TIEWGROUP: 3
BOCFR55 CONTENT: 41(b) 43(b) 4
%<A: 2.3.7
Knowledge of the process for preparing a radiation xvork permit
OBJECTIVE: EPZO-2A
Identify the types of protective actions for HiVP persotmrl (both on and off-site) and who is responsible
for directing them.
DEVELOPMENT REFERENCES: PEP-330
ROFEMENCES SIJPPLIEIP TO APP1,ICANT:
~~ None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNWIICANTLYMODIFIED I DLRECT: New
NRC EXAM HISTORY: None
DISTRACTOR .IUSTFIF%CACTION (CORRECT .4NSWER +d):
a. Plausible since 10 rem TEDE for protecting valuable company property, hut S-SO approval is
required.
d h. Exposure is limited to 10 rem TEDE is the limit required for this activity and S-SO approval is
required.
c. Ildusibk since 25 rem T E W is the limit required for lifesaving efforts, but the limit to protect
equipment and property is 10 rein IEDE.
d. Plausible since 25 rem TEDE is the limit required for lifesaving efforts, but the Limit to protect
equipment and property is 10 rem TELX.
DIFFICCLTY ANALYSIS:
CY3MPREHENSIVE 1 ANN,YSIS KNOWLEDGE / RECAI.1,
DIFFICULTY RATING: 3
EXPLANATION: Requires knowledge of the emergency exposure limits and approval
requirements
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
Given the folhwing conditions:
m Power i s currently at 32% during a plant startup.
e Instrument Bus IDP- I R-SIV deenergiied as a result of a fault in PIC CAB-4.
- PIC ('AB-4 has been isolated from Instrument Bus SIV atid will be deenergiied fix
approximately eight (8) hours while repairs are being made.
Which ofthe following actions must be taken'?
a. Place all PIC CAB-4 Reactor Trip instruments in the tripped condition
b. Place all PIC C A B 4 ESP instruments in the tripped condition
c. Place all MFW Regulating Valves in MANUAL
d. Perform a plant shutdown
ANSWER:
d. Perform a plant shutdown
Post Validation Revision
Harris N I X Written Examination
Senior Reactor Operator
L)ata Sheets
QUESTION NUMBER: 22 TIEWGROUP: l/l
10CFR55 CONTENT: 41(b) 43(b) 2
K h : 000057Ci2.2.22
Knowledge of limiting conditions for operations and safety limits. (Loss of Vital AC Instrument Bus)
OBJECTIVE: AOP-3.24-4
Dcterniine the following: a. Consequences of the loss of all power to PIC CAR-4
DEVELOPMENT REFERENCES: AOP-024
TS Table 3.3-3, pg 3-18 and 3-27
TS 3.0.3, pg 0-1
REFERENCES SUPP1,IE.DTO APPLICANT: None
QUESTION SBtJRCE: NEW SIGNIFICANTLY MODIFPED DIRECT
RANK IVLMBER FOR SIC; CANTEY MODIFIED / DIRECT: AOP-3.24-R4 00 I
NRC EXAM HISTORY: None
IPISTR4CrOR .JUSTIIFICACTION (CORRECT ANSWER Jd):
a. Piausible since instrunlent failures require bistables tripped, hut they are deenergized to actuate and
are already tripped since no power is available.
b. Plausible since instrument failures require bistables tripped, but they are dcenergized to actuate and
are already tripped since no power is available.
c. Piausible since this is the immediate operator action for a loss of Instrument Bus SUI, not SIV.
\ d. Loss of all power to PIC CAB-4 will result in 3 bistable channels of Steam Line Pressure becoming
inoperable. The TS action is to trip the bistables within one hour, but the bistables are energized to
actuate. Without power available, this xtion cannot be perfomicd and TS 3.0.3 becomes applicable
DIFFICULTY ANALYSIS:
COMPREIIENSIVE / AXALYSIS KNOWLEDGE / RECALL
DIFFICUI.1Y RATING: 4
EXPLANATION: Must recognize that energized to actuate bistables cannot be placed in tripped
condition without power, thus an entry into TS 3.0.3 is required, arid must
determine the required TS 3.0.3 actions
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Opermr
QUESTION: 23
During the performance of FOP-PATH-2, the S'1 A reports that the following two (2)
YEI.I.OW path Critical Safety Function Status Trees (CSFST) exist:
e Integrity
e Ileat Sink
Which ofthe following describes how these YEILOVV paths are to he addressed and !or
implemented?
a. Both must be addressed and implemented, with Heat Sink having a higher priority
than Integrity, as soon as FOP-PATII-2 actions are completed provided no other
higher priority CSFST conditions exist
b. Both must be addressed, hut implemented at the discretion of the Superintendent-
Shift Operations, prior to exiting from the N I P 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, hut implemented at the discretion ofthe Superintendent-
Shift Operations, as soon as EOP-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 Supsrintendent-
Shift Operations. prior to exiting from the FOP network
Post Validation Kevision
Harris NRC Written Examination
Senior Reactor Opcrator
Data Sheets
QUESTION NUMBER: 23 'I'IEWGROUP: 3
10CFR55 CONTENT: 41@) 43(b) 5
Kti: 2.4.22
Knowledge of the bases for prioritizing safety functions during abnornial!eiiiergeney operations
OBJECTIVE: EOP-3.19-2
Describe C:ontrol Room usage of status trees as it relates to the following
a. Priority of status trees
b. Rules of usage
DEVELOPMENT REFERENCES: FOP 1Jser's Guide
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRE.CT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRAC~'ORJUSTIFICACTION (C:ORRECTANSWER +d):
a. Plausible since they are to he addressed, but only prior to leucing the 1;OP network and are not
required to be implernented.
d b. All YELLOW-condition CSFSl's should be addressed prior to exiting the E,OP network. However, the
operator is allowed to decide if and urhen to iniplernent. and whether to complete any YELLOW-
condition FRP.
c. Plausible since they are to be addressed, but only prior to leacing the EOP network and arc not
required to be implemented.
d. Plausible since they are to he addressed, but only prior to leaving the HOP network and are not
required to be implemented.
DWFHCULTY ANALYSIS:
COR'IPREMENSIVE / ANALYSIS KNOWIXDGE /RECALL
DIFPICXJLTY HATING: 2
EXPIANATION: Knowledge of the iniplenientation criteria for yellow CSFSTs as directed by
plant procedures
Post Validation Revision
Harris NRC Written Examination
Senior Reac.tor Operator
QUESTION: 24
Following a loss of all h C power, how long are the safety-related 125 VDC batteries
DESIGNED to allow equipment operation'!
a. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, assuming DC h a d shedding occurs within 30 minutes ofthe loss ofail
AC power
b. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, assuming DC load shedding occurs within 60 minutes of the loss of all
AC power
c. 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />?, assuming DC load shedding occurs within 30 minutes ofthe loss of311
AC power
d. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, assuming DC load shedding occurs within 60 minutes of the loss of all
AC power
ANSWER:
d. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, assuming I)C load shedding occurs within 60 minutes of the ioss of all
AC power
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUEST ION NUMBER: 24 TIER/GROUP: iil
10CPR55 CONTENT: 4101) 43(b) 2
K.4: 000058G2.2.25
Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.
(L.oss of DC Power)
OBJECTIVE: EOP-3.7-6
Given a step, caution, or note from EOP-001, EOP-002. or ECP-003, state its purpose
DEVELOPMENT REFERENCES: Tech Spec Bases 3.8.2, pg 8-2
mP-EPP-00 I
ADEL-LP-2.6
REFERENCES SUPPLIED TO APPLICANT: None
QUESTKON SOIIRCIF,: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: ADI!.L2-6-S I 00 I
NRC EXAM HISTORY: Kone
DISTRACTOR SUS'IIFICACTION (CORRECT ANSWER d'd):
a. Plausible since this is the time limit which requires actions being taken in accordance with Technical
Specifications, but the design of the batteries is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b. PlausihIe since this is the tinie limit which requires actions being taken in accordance with Technical
Specifications: hut the design of the batteries is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
c. Plausible since the design of the hatteries is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, but the design assumes that DC Load shedding
occurs within 60 minutes, not 30.
d d. Batteries are designed to carry required safety related loads for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without AC input to
cany bus or charge battery, assuming that required load shedding occurs within I hour.
DIFFIC:UI.TY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE:/ RECALL
DIFFICULTY RA'TING: 3
EXPLANATION: Knowledge of tech spec basis and design of safety-related batteries
Post Vaiidation Revision
Harris KRC Written Examination
Senior Rcactor Operatot
QUESTION: 25
Which of the following actions would be INAPPROPRIATE to perfomi prior to
direction in an EOP?
a. Isolating AFW' tlow to a single faulted SG
b. Throttling AFW flow to control a ruptured SG level within the required level hand
c. Securing a CSIP to prevent overfilling the pressurizer foliowing an inadvertant SI
d. Shutting the MSIVs to isolate a steamline break which has not resulted in an SI
ANSWER:
c. Securing a CSIP to prevent overfilling the pressurizer following an inadvertant SI
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QWESTION NUMBER: 25 TIEWGROUB: 3
lQCFR55CONTENT: 41(b) 43(b) 5
KA: 2.4.14
Knowledge of general pidclines for EOP flowchart use
OBJECTIVE: EOP-LP-3.19-1
Describe Control Room wage of the EOP network as it relates to the following: a) Peilbrming steps out
of sequence
DEVELOPMENT REFERENCES: KOP Users Guide
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
RANK NUMBER FOR SIGNIFICANTLY hIODIFIE1) / DIRECT: EOP-3.19-RI 0 18
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER dd):
a. Plausible since this is a numbered step in P N H - I which are normally required to he performed in
sequence, but the EOI IJsers Guide addresses this as being acceptable.
b. Plausible since this is a numbered step in PATH-I which are nonnaily required to be perfonned in
sequence, hut the EOP Users Guide addresses this as being acceptable.
v E. Performing steps out of sequence is allowed. but must be done with caution to prevent masking
symptoms or defeating the intent of the JiOP being used. Although terminating SI early might be
heneficial to prevent filling the pressurizer if the only event is a spurious SI, this may result in further
degradation of the piant if another undiagnosed eveut is in progress.
d. Plausible since this is a nnmbered step in PAIH-1 which arc nonnally required to be perfixmed in
sequence, but the EOP Users Guide addresses this as being acceptable.
DIFFICULTY ANALYSIS:
COMPREIIENSLVE / ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICU1,TY RATIXG: 3
EXPLANATION: Must differentiate between those actions which could potentially result in
degradation ofthe plant if taken out of sequence and those actions which would
likely have little impact on the operators abilities to diagnose other events.
Post Validation Revision
. -
HARRIS EXAM
50-40812004-304
-
FEBRUARY 23 27,2004
& MARCH 4,2004 (WRITTEN)
'I
Harris WRC Written Examination
Reactor Operator
QUESTION: 1
Following a Reactor Trip, the RCS temperature is being controlled by the S t e m Dump
Control System at 540°F. FOP-EPP-004, Reactor Trip, directs that the WCS be
maintained at 557°F.
Given the following range of instruments, if the Steam Dump Control System is placed in
the Sleam Pressure mode, what approximate setpoint is required to maintain RCS
temperature at 557F?
Steam header pressure full range: 0-1300 psig
- Steam generator pressure full range: 0-1300 psig
- Turbine main steam pressure full range: 0-1500 psig
a. 16%
b. 24%
C. 73%
d. 84%
ANSWER:
d. 84%)
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Datr Sheets
QUES~IQNNJMRER: I TIEWGROUP: 111
KAIMPORTANCE: RO 3.7 SRQ
10CFR55 CONTER'I': 41@) 4 43w
KA: 000004EA1.10
Ability to operate and monitor the following as they apply to a reactor trip: SIG pressure
OBJECTIVE: SDCS-3.0-4
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
DEVE1,OPMENT REFERENCES: EOP-EPP-004
REFEIUZNCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY ,~.ZODIF'PP:D /DIRECT: S D C S - ~ 4004
NRC EXAM HISTORY: None
I)PSTR4CTOH JUSTIFHCACTION (CORRECT AKSWER d'd):
a. Plausible if the incorrect instrument is used to determine the range of the instrument and the
calculation is performed incorrectly (1500 - 1092 i 1500).
b. Plausible ifthe correct instrument is used to determine the range ofthe instrument, but the calculation
is performed incorrectly (1300 - 1092 I 1300).
c. Plausible if the incorrect instrument is used to determine the rangc of the instrument (1092 I 1500).
d d. The equivalent steam pressure for the required RCS temperature is approximately 1092 p i g . This
calculates to be a setpoint of 84% (1092 / 1300).
DIFFICULTY ANALYSIS:
COMPREEIENSWE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY wrmc;: 3
EXPLAX.4TION: Must detennine required steam pressure for RCS temperature and then calcuiate
setpoint
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 2
With the plant at 100 percent steady-state condition, the following occurs:
m AtB-06-7-3. TOTAL MAKEUP WATER FLOW DEVIATION, alanns.
m A1.B-06-8-4. BORIC ACID FLOW DEVIATION, a i m s .
0 VCT level is at 19.5% and decreasing at the same rate it has beer1 for the last few
days.
Which ofthe following procedures should be addressed?
a. AOP-002, Emergency Roration
b. AOP-003, Malfunction of Keactor Makeup Control
e. AOP-016, Excessive Primary Plant Leakage
d. AOP-017, Loss of Instrument Air
ANSWER:
b. AOP-003, Malfimction of Reactor Makeup Control
Post Validation Revision
H a m s NRC Written Excamination
Reactor Operator
Data Sheets
QCES'rION NUMBER 2 TIEWGHOUP: 1/1
10CFR55 CONTENT: 41(b) 10 43(W
0: 00002262.4.4
Ability to recognize ahnonnal indications for system operating parameters which are entry-level
conditions for emergency and abnormal operating procedures. (Loss of Reactor Coolant Makeup)
OBJECTIVE: AOP-3.341
IDENTIFY symptoms that require entry into AOP-003. Malfunction of Reactor Makeup Control
DEVELOPMENT RJ3FERENCES: AOP-003
REFERENCES SUPPLIED TO APPLICANT: Kone
QUESTION SOURCE: NEW SIGNHFICAXTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: AOP-3.3-KI 002
NRC EXAM HISTORY: Hams NRC 2000
DISTRACTOR JVSTIFPCACTHON (CORRECT ANSWE.R d'd):
a. Plausible since Emergency Roration entry conditions include any condition which is a result of an
unexplained reactivity addition, which candidate may consider this to he.
4 b. 'I'hese are entry conditions for Reactor Makeup Control malfunction
e. Plausible since CVCS leakage, if suspected, would cause entry into AOP-016.
d. Plausible since nontial horation flowpaths are not available during a loss of instrument air event.
I)IFFICUI,TY ANALYSIS:
COMPREIIENSPVE I ANALYSIS KNQWLE.DCEI RECALL
DIFFICULTY RATING: 2
EXPLANATIOX: Knowledge of entry requirements for loss of reador makeup
Post Validation Revision
Harris NKC Written Examination
Reactor Operator
QUESTION: 3
Given the foilowing conditions:
e The plant is operating at 50% power.
e PT-457, Channel I I I Pressurizer Prcssure, has failed and all associated bistables are in
the tripped condition.
e Power is subsequently lost to UPS Bus IDP-1A-S1.
Which of the following describes the effkct of this loss of power on the Phase A
Containment Isolation valves?
a. NO Phase A Containment Isoration valves will close
h. ONLY Train A Phase A Containment Isolation valves will close
c. ONLY lrain B Phase A Containment Isolation valves wit1 close
d. Ail Phase A Containment Isolation valves will close
ANSWER:
c. ONLY Train B Phase A Containment Isolation valves will close
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 3 TIER/CROUP 2i I
10CFR55 CONTENT: 41(b) 5 43(W
Eka: 013K3.03
Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following: Containment
OBJECTIVE: ESFAS- 3.0-4
PREDICT how loss of any ofthe four instrument buses will affect the ESFAS outpiit functions of each
SSPS train
DEVELOPMENT REFERENCES: AOP-024
SD-103
REFERESCES SUPPLIED TO APPLICANT: None
QUESTHON SOURCE: NEW SHGNIFICA"T1,Y MODIFIED DIRECT
BANK NUMBER FOR SHGSIFHCANTIiUMODIFIED i DIFUCCT: ESFAS-3.0-K4 00 1
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFHCXCTIOS (COmECT ANSWER d'd):
a. Plausible since Train SA slave relays will not actuate, hut Train SB relays will still actuate.
b. Plausible since one train of Phase A will not actuate. hut the train that will not actuate is Train SA.
d e. A loss of Bus II)P-lA-SI under these conditions will result in a 2!3 sisnal to both trains of ESFAS,
resulting in an SI and Phase A signal. lrain SA slave relays; however, are powered from IDP-IA-SI
and are energized to actuate, so 'l'rain SA slaves will not perform their function.
d. Plausible since SK and Phase A signals will be generated on both trains ofE.SFAS, but Train SA slave
relays wiil not actuate due to not having power.
DIFFICULTY ANALYSIS:
COMPRRIEIIEXSIVF:1ANALYSIS KNOWLEDGE I RECALL
DIFFICUIJY RATING: 3
EXpI.AXATION: Analyze the effect of a loss ofpower on the actuation signals and determine
which power supplies power which output relays
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 4
Given the following conditions:
e The unit is operating at 30% power.
e A dropped Control Rank 'C' rod has just been re-aligned.
a While attempting to operate the ROD CCPN'I'KOL ALARM RESET, the operator
inadvertently operates the ROD CXNTROL START-BJP RESET.
Which of the following describes the effect of operating the incorrect reset?
a. ,411 Control Bank 'C' rods drop into the core, causing an automatic reactor trip
b. All rods, including Control Hank and Shutdown Bank rods, drop into the core,
causing an automatic reactor trip
c. All rods remain in their current position and there is NO effect on the Rod Control
System circuitry
d. All rods remain in their current position, but the Rod Control System circuitry
senses all rods are lb-ullyinserted
ANSWER:
d. All rods remain in their current position, but the Rod Control System circuitry
senses all rods are fiiily inserted
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QIJESTION NUMBER: 4 TIENGROUP: I12
10CFR55 CONTENT: $I@) 6 W O
KA: 000003AA1.02
Ability to operate and i or monitor the following as they apply to the Dropped Control Kod: Controls and
components necessary to rec~verrod
OBJIETA'E: ROI?CS-3.0-K7
DISCUSS the effects of manipulating each of the following rod control-related switches
0 ROD CONTKOI, START-UP RESET switch
e ROD CONTROH.AI.AKM RIISET switch
DEWXOPMENT REFERENCES: AOP-OO1
ROL)CS-3.0
REFERENCES SUPPLIED TO APPLICANT: None
QtJEST'nON SOURCE: SHGNHFICANTLY MODIFIED c]DIRECT
CANTLY MODIFIED /]DIRECT: RQDCS-3.0-R7 001
KRC EXAM HISTORY: None
DISTRACTOR K'STHFICACTION (CORRECT ANSWER +d):
n. Plausible since improper operation of correct switch could result in rods dropping into core, but
operated switch only resets starting points for rod control circuitry.
b. Plausible since improper operation of correct switch could result in rods dropping into core, hut
operated switch only resets startina points for rod control circuitry.
e. Plausible if misconception that effect is nothing if perfonned at power since switch is normally only
operated prior to withdrawing any rods: but operated switch resets starting points for rod control
circuitry.
d d. Operating switch at power does not affect actual rod position, but resets rod control such th.at circuitry
senses rods are at "full inserted" position.
ICIJLIY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY R4TING: 3
EXPLANATION: Knowledge of the function of rod control system controls
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 5
FRP-J. I, Response to High Containment Pressure, monitors the status of the ESW
Booster Pumps.
Which of the following is the concern if ESW Booster pumps fail to start while high
containment pressure conditions exist?
a. ESW Pump mnout
b. Flooding of safety equipment in containment
c. Loss of containment cooling capability
d. Radioactivity release to the environment
ANSWER:
d. Radioactivity release to the environment
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 5 TIER/GRQPUP: 1I2
10CFR55 CONTENT: 41(10) 8:9 4300
KA: WT14EA1.2
Ability to operate and i or monitor the following as they apply to the (High Containment Pressure)
Operating behavior characteristics of the facility
OBJECTIVE: Bl3-3.13-R3
Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
m ESW booster pump operation
DEVELOPMENT WEFERFNCES: FIW-J. 1
FW-J.1 Step Deviation Basis
REFERENCES SI!PPLIED TO APPLICANT: None
QUEsTION SOURCE: NEW SIGNIFICANTLY IkZODHFIED DHKECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIWECT: EOP-3.13 012
NRC EXAM HISTORY Harris NRC 2002
D r s m . 4 c r o R JUSTIFICACTION (CORRECT AKSWER Jw):
a. Plausible since ESW flow is supplying containment as well as other loads, but actions taken if an
ESW booster pump is not running arc not for ninout concerns?but rather to raise ESW pressure inside
containment.
b. Plausible since an ESW rupture inside containment could result in flooding of containment, but the
booster pump is checked running to ensure ESW pressure is adequate inside containment.
e. Plausible since the ESW booster pump supplies containment cooling units, but the ESW pump is
capable of supplying the loads without the booster pump, hut not at the required pressure for inside
containment.
4 d. ESW Booster Pumps are required to be running to ensnre ESW pressure is > containment pressure
following a LOCA to prevent any leakage in the ESW system to cause the leakage to be to
contninnient and not to the ESW system.
DIFFICXJLTY ANALYSIS:
COMPREHENSIVE / ANALYSIS MNOWI.EBPGE I WECALL
DIFFICULTY RATING: 3
E.XPIANAIION: Knowledgc of EOP procedural bases
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 6
Given the following conditions:
e EOI-FRP-II.I, Response to a Loss of Secondary Heat Sink, is being implemented.
o RCS bleed and feed has been initiated when Auxiliary Feedwater (AFW) capability is
restored.
e All SGs are completely dry and depressurized.
Which of the following describes which SGs are to be fed under these conditions?
a. Feed ONLY one (1) SG to enstire RCS cooldown rates are established within
Technical Specification limits
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 of RCS bleed and feed as .soon as possibk
ANSWER:
b. Feed ONLY one (1) SG to limit the possibility o f a SG tube rupture to a single SG
Post Validation Revision
Harris NRC Written Examination
Kenctor Operator
Data Sheets
QUESTION NCJMBER: 6 TIEWGWOUP: 1/1
KAIMPORTANCE: MO 3.6 SRO
IOCFW55 CONTENT: 41(b) 4/10 43(b)
KA: 000054AK1.02
Knowledge of the operational implications ofthe following concepts as they apply to I m s of Main
Feedwater (MFW): Effects of feedwater introduction on dry S/G
OBJECTIVE: EOP-3.11-4
Given the foilowing EOP steps, notes. and cautions, DESCRIRE the associated basis
e Feed restoration
BPEVELOPhTENT REFERENCES: EOP-FRP-11.1
REFERENCES SUPPLIED TO .UPLLICXNT: None
QLTESTHON SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NHC EXAX IIISTORR None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER .Id):
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 S G .
d b. Flow should only be established to one dry SG so that ifexcess thermal shock causes Pdilure, the
failure is limited to one SG.
e. Plausible since RCS subcooling is a desirable condition to achieve, but only one SG at a time is fed.
d. Plausible since terminating KCS bleed and feed is a desirable condition to acl~ieve,hut only one SG at
a time is fed.
1CLTIW ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICIJLTY RATING: 3
EXPLANATION: Knowledge of the requirements for feeding a dly S G arid the reasons for these
actions
Post Validation Revision
Harris NRC Written Exaniination
Reactor Operator
QUESTION: 7
Given the following conditions:
PZR level is 53% and stable.
a VCT level is 23 and stable.
a 1,etdou.n flow is 45 gpm (FI-150.1).
a RCP seal injection flows are:
RCP A at 8.3 gpm
RCP B at 7.9 gpm
RCP C at 7.8 gpm
a RCP seal return f h v s are:
RCP A at 2.8 gpm
RCP B at 3.1 gpin
RCP C at 2.9 gpm
Which of the following would be the expected flow indication on FI-122A. 1, Charging
Fl(w.*, assuming NQ RCS leakage?
a. 21 g p
b. 30gpm
d. 54gpm
ANSWER:
b. 30gpm
Post Validation Revision
Harris NRC Written Examination
Reactor Operatur
Data Sheets
QUESTION NBTICEWER: 7 TIEHPIGROUP: 2il
10CFR.55 COYIENT: 48(b) 3 43w
Ktk 003A4.01
Ability to manually operate and/or monitor in the control room: Seal injection
OBJECTIVE: CVCS-3.0-Rl
Given appropriate C 3 T S information, PERFORM a CVCS flow balance without reference to procedures
DEVELOPMENT REFERENCES: SD-107
CPL-2165-Sl305
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNHFICANIIJY MODIFIED DIRECT
BANK NEWBEH FOR SIGNIFICANT1,Y MODIFIED / DIRECT CVCS-K 1 00 1
NRC EXAM MIS1ORk: Noire
DISTRACTOR SUSTIFICACTION (CORRECT ANSWER $d):
a. Plausible if misconception is that seal leakoff flow is ignored, but leakoff flow is not.required to be
made up (45 24 = 21). However, seal leakoff flow is required to be included.
~
1 b. Charging flow should equal letdown flow (45 m m ) less seal injection flow (24 gpni) plus seal return
flow (9 a m ) (45 - 24 + 9 = 30).
c. Plausible if misconception that seal injection flow is measured as part of charging flow and seal
leakoff must be subtracted, but seal injection is required to be inciuded (45 - 9 = 36).
d. Plausible if misconception that seal injection flow is measured as paIt of charging flow, but seal
injection is required to be included (45 + 9 = 54).
ICULTY .W&ISPS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Calculation of expected charging flow indication based on given CVCS
parameters
Post Validation Revision
Hams NRC Written Examination
Reactor Operator
QUESTION: 8
EOP-EPP-008, SI Termination, directs resetting SI.
Which of the following describes the effect of operating only ONE (1) of the two (2) SI
RESET switches at this step instead of both?
a. e Bypass - Permissive Light Panel light 4-1, SI ACTUAlE, would blink due to
only one train of SSPS having an SI signal
e Bypass - Permissive Light Panel light 5-1,SI RESET - AUTO SI BLOCKED,
would biink due to only one train of SSPS having SI reset
b. e Bypass - Permissive Light Panel light 4-1, SI ACTUAIX, would extinguish
due to neither train of SSPS having an Si signal
e Bypass -Permissive Light Panel light 5-1, SI RESET - AtrTCP SI BLOCKED,
would light due to both trains of SSPS having Si reset
c. 0 Bypass - Perniissive Eight Panel light 4-1, SI ACTUATE, would blink due to
only one train of SSPS having an SI sibma1
e Bypass - Permissive Light Panel light 5-1, SI RESET - AUTO SI BLOCKED,
would light due to both trains of SSPS having auto SI blocked
d. 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 Panel iight 5-1, SI RESET AUTO SI BLOCKED,
s-
would light due to both trains of SSPS having auto SI blocked
ANSWER:
a. e Bypass -. Permissive Light Panel light 4- I , SI ACTUATE, would blink due to
only one train of SSPS having an SI signal
e Bypass - Pemiissivc Light Panel light 5-1, SI RESET . AUTO
~ SI BLOCKED,
would blink due to only one train of SSPS having SI reset
Post Validation Revision
Harris NRC Written Exsmination
Reactor Operator
Data Sheets
QUESTKON NUMBER: 8 TIEWGROUP: 211
1QCFR55CONTENT: 41(b) 7 Wb)
IM: 006K4.11
Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following: Reset of SIS
OBJECTIVE: SIS-3.0-R4
DETERMINE SIS status from the following
- Eypass-Permissive Light Box
DEVELOPMENT REFERENCES: SD-103
REFERENCES SUPPLIED TO APPIKANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR s I m I F I c m r I , Y MODIFIED DIRECT: INPO 1073
NRC EXAM HISTORY: None
DISTRACTOR JUSTPFICACPHON (CORRECT ANSWER +d):
d 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 SSIS for both the reset and the actuation signals so both lights would blink.
h. 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 be reset so window 4-1 would be responding correctly, but
window 5-1 would also be hlinking due to the disparity between trains.
d. Pkdusibk since the SI Actuation switch only requires a single switch to actuate SI, but the reset
switches are train-related.
DIFFICULTY ANALYSIS:
C:OlllPREHENSNE I ANALYSIS LVOW1,EDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Comprehend the effect of only operating a single train switch on SSPS and how
the indications would be affected
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 9
Given the following conditions:
e Containment Pressure Channel I, PT-950A, is in TEST for surveillance testing
purposes.
e Containment Pressure Channel 111, PT-952A, is failed low.
e A large break &OCAoccurs and actual Containment Pressure reaches 21 psig.
Which of the following describes the response of the Containment Spray system?
a. NEITHER train of Containment Spray will automatically actuate
b. ONLY Train 'A'of Containment Spray will automatically actuate
c. ONLY Train 'B' of Containment Spray will automatically actuate
d. BOTH trains of Containment Spray will automatically actuate
ANSWER
d. BOTH trains of Containment Spray wiil automatically actuate
Post Validation Revision
Harris NKC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 9 TPERIGROGP: 2)1
POCFH55 CONTENT: 41(b) 7i9 43@)
Ktf: 013K6.01
Knowledge of the effect o f a loss or malfunction on the following will have on the ESFAS: Sensors and
detectors
OBJECTWE: CSS-Rl
STATE the conditions that will cause a containment spray actuation signal (CSAS) including coincidence
and setpoints
DEVELOPMENT REFERENCES: SD-103
REFERENCES SUPPLIED TO APPL.ICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIFOZCT
BANK NUI\IBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CSS-RI 003
NRC EXAM HISTORY Harris NRC 2002
DISTRACTOR SIJSTHFICACTION (CORRECT ANSWER dd):
a. Plausible since CSAS is energized to actuate and 2 of the channels are in a deenergized condition, hut
the remaining 2 channe1.s will cause an actuation of both trains of Spray.
b. Plausible since CSAS is energized to actuate and 2 ofthe channels are in a deenergized condition, hut
the chamnek input both trains of SSPS.
c. Plausible since CSAS is energized to actuate and 2 ofthe channels are in a deenergized conditi.cn, but
the channels input both trains of SSPS.
d d. CSAS is energized to actuate and although 2 of the channels are in a deenergized condition, the
remaining 2 channels will cause an actuation of both trains of Spray.
DIFFICULTY ANALYSIS:
COMPREIIEWSIVE I ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Compreherision of the cffcct of failed channels and testing on ESF signals
Post Vslidation Revision
Harrix NKC Written Examination
Reactor Operator
QUESTION: io
Given the following conditions
e The plant is operating at 43% power.
o i20VAC Vital Bus IDF-1B-SI1 deenergizes,
Outward rod motion is inhibited by ...
a. C- 1 ,Intermediate Range rod stop.
b. C-2, Power Range rod stop.
c. C-3, QTAT rod stop.
d. C-4, QPAT rod stop.
ANSWER:
b. C-2, Power Range rod stop.
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 10 TIEWGHOUP: 212
10CFR55 CONTENT: 48(b) 6 43w
KA: OClK4.07
Knowledge of CRDS design feature(s) anct'or interlock(s) which provide for the following: Rod stops
OBJECTIVE: NIS-3.0-9
DISCUSS the operation of the following NI trip-related Cunetions:
b. SR, IR and PR (low) trip blocks
DEVELOPMENT REFERENCES: OF-105
KEFEHENCES SUPPLIED 1'0APPLICANT:
QUESTION SOURCE: NEW a None
SIGNIFICANTLY MODIFIED
BANK NKJMRER FOH SIGNlFICANTLY LMODIFIED1%)ERECT: NHSR6 003
DIIPECT
NHC EXAM HISTORY: Nonc
DISTRACTOR .KJSTIFICAC'HON (CORRECT ANSWER d'd):
a. Plausible since this causes 3 rod stop, and coincidence is 1!2, but 1R rod stop is blocked above P-10 by
manual operator action. Must have 2/4 PK below P-IO io reset.
.\, b. PR rod stop is 1/4 coincidence. With $2-SB deenergized, PR N-42 is tripped.
c. Plausible since causes rod stop. but coincidence is Z 4 instead of 1/4
d. Plausible since causes rod stop, but coincidence is U4 instead of 1!4.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY wrnw: 3
EXp1,ANATION: Analyze effect of loss of power on NIS and rod control and detennine effect of
single channel tripped
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 11
The hasis for the operation of the Electric Hydrogen 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 ofmetals in Containrrient and release of hydrogen from the RCDT.
c. release of hydrogen from the PRT and the radiolytic decomposition ofwater.
d. radiolytic decomposition of water and the corrosion of metals in Containment.
ANSWER:
d. radiolytic decomposition of water and the corrosion of metals in Containment.
Post Validation Revision
Harris NRC X7rittenExamination
Reactor Operator
Data Sheets
QUESTIOK NI:MBEM: i I THIERIGRBtJP: 2!2
BOCFRSS CONTENT: 48(b) 10 43w
IKA: 028G2.2.22
Knowledge of limiting conditions for operations and safety limits. (Hydrogen Recombiner and Purge
Controi)
OBJECTIVE: KR-3.0-1
STATE the purpose and fimction of the Hydrogen Recombiner System, including the following
components:
6 Electric hydrogen recombiner
DEVELOPMENT REFERENCES: 'IS 3.6.4.2 Basis
SD-125
IP-BR-3 .O
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: b]NEW
p.9
u
R
SIGNIFICANTLY MODIFIED DPRFXT
BANK NUMBER FOR SIGNIFICANTLY MODIPIED / DIRECT: IIR 0 1
NMC E m ¶ HIS'I'ORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER +d):
a. Plausible since Electric Hydrogen Recombiners are designed to remove hydrogen in containment
folIowing a 1,OCA due to generation from the zirc-water reaction, but not due to release from the
I'KT.
b. Plausible since Electric Hydrogen Reconibiners are designed to remove hydrogen in containment
following a LOCA due to generation from the corrosion of metals in containnient! but not due to
release from the RCDT.
c. Plausible since Electric Hydrogen Recombiners are designed to remove hydrogen in containment
following a LOCA due to generation from the radiolytic decomposition of water, but not due to
release from the PRT.
4 d. The Hectric Hydrogen Recombiners are designed to F C ~ O Vhydrogen
~ in containment following a
1,OCA due to generation from the zire-water reaction, radiolytic decomposition of water, and
comosion of met& in containment.
DIPPICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE f RECALL
DIFFICULTY Ff.A'rING: 3
EXPLANATION: Knowledge of Tech Spec basis for hydrogen recombiners
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 12
EOP-EPP-00 1, Loss of AC Power to 1.4-SA and IB-SR Buses, is being performed.
Concurrent to the loss of power, a small break LOCA occurred.
The crew has completed the following actions when off-site power is restored to 6.9 KV
Bus 1.4-SA:
e Sequencers have been de-energized
6 Safeguards pumps autostarts have been disabled
6 RCP seals have been isolated
6 MSIVs arid FWIVs have been closed
e Depressurization of SGs to I80 psig has commenced
Which of the following actions is the FIRST to be taken following the restoration of off-
site power?
a. Start an ESWpump
b. Start a CSIP
c. Stabilize SG pressures
ci. Initiate SI
AK§WEW:
c. Stabilize SG pressures
Post Validation Revision
IIarris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 12 TIEWGROUP lil
10CFR55 CONTENT: 41(b) 10 4Xb)
MA: 000055EAi.07
Ability to operate and monitor the following as they apply to a Station Blackout: Restoration of power
from offsite
OBJECTIVE: EO?-3.7-5
Given a title of a continuous action step from a foldout and a list of plant conditions, DETERMINE if
implementation is required
DEVELOPMENT REFERENCES: EOP-EPP-00 1
REFERENCES SUPPLIED TO AIPP1,IC:ANT: None
QUESTION SOURCE: NEW GHGNIFICAWLY MODIFIED
BANK NZMBER FOR S I 6 CANTLY MODIFIED / DIRECT:
NHC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER @d):
a. Plausible since if the power source was an EDG instead of offsite power, it would he important to
provide cooling flow to the EDG.
b. Plausible since a small break LOCA exists and RCS inventory is being lost, but the first action is to
stabilize SG pressure.
d e. Upon restoration of power to at least one bus, the first action taken is to stabilize S G pressures
d. Plausible since a small break ILKX exists and RCS inventory is being iost, but the first action is to
stabilize SG pressure.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS JXNBWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of required actions when power is restored following a loss of all
AC power
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 13
While performing an Operating Procedure, the Reactor Operator comes to a step which
states:
Request Chemistry to sample the RHT for boron concentration.
The Reactor Operator believes the step is NOT essential to achieving the purpose for
which the procedure is being used and that the omission ofthe step does NOT violate the
precautions and limitations of the Operating Procedure.
Which of the following is the MINHMUM requirement(s) that must be met to allow
marking the step NjA?
a. Step must be initialed by the Reactor Operator prior to perfornlance
b. m Step must be initialed by the Reactor Operator prior to performance
e A written explanation ofwhy the step is N!A must be prowded in the
Comments section of the procedure
c. m Step must be initialed by the SCO prior to performance
d. m Step must be initialed by the SCO prior to performance
m A written explanation of why the step is N!A must be provided in the
Comments section of the procedure
ANSWEW:
d. Step must be initialed by the SCO prior to performance
A written explanation ofwhy the step is N/A must be provided in the
Comments section of the procedure
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 13 TIE19/GM OU P 3
10CFR55 CONTENT: 41(b) 10 43W
KA;: 2.1.23
Ability to perfom specific system and integrated plant procedures during all modes of plant operation
BBJEC1'IVE: PP-2.0-2
L)ISCI!SS the requirements in FRO-NGGC-0200 concerning the following:
e Procedure user's responsibilities
DEVELOPMENT REFERENCES: FRO-NCiGC-0200
REFERENCES SUPPI,I(EDTO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM BISTORW None
DHSTRACI'OR JUSTIFICACTION (CORRECT ANSWER d'd):
a. Plausible since the RO discovered the cause for marking the step N!A, but a supervisor must initial the
step prior to performance and a written explanation must be provided in the Comments section.
b. Plausible since a written explanation must be provided in the Comments section, but a supervisor must
initial the step prior to performance.
e. Plausible since a supervisor niust initial the step prior to perforniance, but a written explanation must
be provided in the Comments section.
4 d. The step is initialed by the responsible supervisor prior to performance and a written explanation is
provided in the Comments section.
ICUI,TY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULI'Y RATING: 2
EXPLANATION: Knowledge of use of N.4 during procedure usage
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUES'I'IQN: 14
Given the following conditions:
e Spent resin is being sluiced from the Cation Demineralizer to a Spent Resin Storage
Tank.
e The operator reports that it appears that a pipe in the overhead of's hallway is plugged
with resin.
o HP reports the results of a radiation survey as follows:
e 2500 m r h on contact with pipe
e 1200 m r h @ 18 inches from the pipe
o 5 m r h r at floor level below the pipe
Which one ofthe following describes the required radiological postings?
a. NO pcrstings are required because a ladder is required to access the pipe area
b. Very High Radiation Area with red flashing light
c. High Radiation Area with a red flashing light
d. High Radiation Area, but NO red flashing light required
ANSWER:
c. High Radiation Area with a red flashing light
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 14 TIEWGRBUP: 3
10CFR55 CONTENT: 41(b) 12 43BW
KA: 2.3.2
Knowledge of facility A L A M program
OBJECTIVE: W-35-13
Define the following terms as defined in IOCFRZO:
DEVELOPMENT REFERENCES: AI'-504
REFERENCES SUPP1,IE:D TO APPLICANT: None
QUESTION SOURCE: SIGNHP1CA"FLY MODIFIED DIRECT
CANTLY MODIFIED I IIIKECT: 0 20657
NRC EXAM HISTORY: Harris NRC 2002
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER J'd):
m. Plausible since normal access to the area requires a ladder. hut it is accessible so a red warning light is
required.
b. Plausible since a common error is defining the difference between a very high radiation area and a
high radiation area, but this is a high radiation area requiring a red warning light.
4 c. Accessible areas where rad levels exceed 1000 mWhr are required to be locked. Where it is not
practicai to lock the area, a red warning light shall he in place to warn personnel.
(8. Plausible since this is defined as a high radiation area, but a red warning light is also required.
ICULTY ANALYSIS:
COklPREIIENSWE / ANALYSIS KNOWLEDGE I RIXhLI,
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of radiological posting requirements
Post Validatioii Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 15
The Containment Fan Coil Units (AIi-37 / 38 / 39) provide area cooling to ...
a. the reactor vessel supports and reactor coolant leg nozzles.
b. the clearance between the reactor vessel and primary shield wall.
c. the reactor coolant pumps.
d. contairment for norma1 operation and accident conditions.
ANSWER:
c. the reactor coolant pumps.
Post Validation Revision
Harris hXC Written Examination
Reactor Operator
Data Sheets
QUESTION NLWBER: 15 TIEWGROUIP: 211
10CFR55 CONTENT: .bI(b) 9 43(h)
H(A: 02262. I .28
Knowledge o f t h e purpose and function of major system components and controls. (Containment
Cooling)
OWJECTBVE: CCS-3.0-A1
Sl'ATE the purpose of the following five Containment Cooling Subsystems
e Containment Fan Coil [Jnits
DEVELOPMENT REFERENCES: SD-169
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY M0DIFIE.D DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CCS-AI 010
NRC EXAM HISTORY: None
DIS'TIPAC'FOR JUS'TIFICACTION (CORRECT ANSWER .I'd):
a. Plausible since containment cooling components provide cooling to these components, but it is
provided by the reactor supports cooling system not the containment fan coil units.
h. Plausible since containment cooling components provide cooling to these components, but it is
provided by the primary shield cooling units not the containment fan coil units.
4 c. Air is drawn from containment space, through the cooling coils, to the fan suction. Cooling air from
the fan coil unit is directed to the reactor coolant pump subcompartments.
d. Plausible since containment cooling components provide cooling to these components, but it is
provided by the containment fan coolers not the containment fan coil units.
DIFFICULTY ANALYSIS:
COl\gPREHENSA'E / AKALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of purpose of Containment Fan Coil Units
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 16
Given the folIowing conditions:
e Following a Reactor Trip and Safety Injection due to a leak in the PRZ steam space,
the Critical Safety Function Status Trees (CSFST) are being monitored.
e The CSFST for RCS Inventory first checks PRZ level and then checks the Reactor
Vessel Level Indicating System (RVLIS).
If PW. level is indicating greater than 92%, why is a check of RVI,IS then perform'?
a. Determine if the cause of the high PRZ level is excessive RCS inventory or
voiding in the Reactor Vessel head
b. Determine if SI termination criteria i s met to allow reducing the excessive RCS
inventory
c. Determine if Adverse Containment conditions have caused erroneous indications
ofthe PRZ level instmnients
d. Determine if the cause of the high PRZ level is excessive RCS inventory or
expansion due to an RCS heatup
ANSM'KR:
a. Determine if the cause of the high PRZ level is excessive RCS inventory or
voiding in the Reactor Vessel head
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 16 TIEWGROCP: 1:1
POCFR.55 CONTENT: 41(b) 7 43m
KA: 000008(;2.1.28
Knowledge of the purpose and function of major system components and controls. (Pressurizer Vapor
Space Accident)
QBJECTIVE: ICCM-3.0-1
LIST the two major functions of the Inadequate Core Cooling Monitor (ICXM)
DEVELOPMENT REFERENCES: FOP Rackground for Inventory Status Tree
LPEOP-3.12
REFERENCES SUPPLIED TQ APPLICANT: None
QUESTIQN SOZJRCE: NEW SIGNIFICANTLY MODIFIED
OR SIGNIFICANTLY MODIFIED / DIRECT:
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
d a. Once a determination has been made that P E lwei is full, RVLIS is then used to confirm whether the
cause ofthe full PRZ is excessive inventory or voiding in the head rcgion.
h. Plausible since RVLIS is used throughout the EOP network to determine if SI termination criteria has
been met, but in this instance it is used to determine the cause of the high PRZ level.
c. Plausible since a steam space break in the PRZ will affect the level indications, but RVLIS is used to
determine the cause ofthe PRZ high level condition.
d. Plausible since RVEIS is part of the Inadequate Core Cooling Monitoring System and a heat up of the
RCS will cause expansion ofthe RCS, but but RVIM is used to determine the cause of thc PW. high
level condition.
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
I)IFFHCUI,TY RATING: Knowledge of the purpose of monitoring RVLIS during accident conditions
EXPLANATION:
Post Validation Revision
IrillTis NRC Written Examination
Reactor Operator
QUESTION: 17
Given the following conditions:
e The plant is shutdown for work on Reactor Coolant Pump seals.
e The Reactor Vessel Head is still installed.
e The running Residual Heat Removal (RIIR) pump trips and the crew is unable to start
the standby RHR pump.
e Time to reach core boiling is determined to be 26 minutes.
e Time to reach core boil-off is determined to be 53 minutes.
Of the following two (2) methods of RCS makeup, 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 prefened over starting a CSIP since
starting a CSIP under these conditions would violate Technical Specifications
b. Gravity feed from the RWST to the RCS is preferred over starting a CSIP since
Reactor Makeup to the CSIP may be insufficient to makeup for core boil-off
c. Starting a CSIP is preferred over gravity feed from the RWST since gravity feed
flow may he insufficient to makeup for core boil-off even if the RCS is
depressurized
d. Starting a CSlP is prefemd over gravity feed from the RWST since the RCS may
be pressurized and prohibit gravity flow
ANSWER:
d. Starting a CSIP is preferred over gravity feed fiom the R W S l since the RCS may
be pressurized and prohibit gravity flow
Post Validation Revision
Harris NRC Written Examination
Resctor Operator
Data Sheets
QUESTION NUMBER: 17 TIEWGROUP: l/I
IQCFR55CONTENT: 4l(b) 8/10 43th)
K k 000025AK3.01
Knowledge of the reasons for the following responses as they apply to the Loss ofResidua1 Heat
Removai System: Shift to alternate flowpath
OBJECTIVE: AOP-3.20-3
Given a set of entry conditions and a copy of .4OP-020, IETEKMhE the appropriate response
DEVELOPMENT REFERENCES: AOP-020
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNHFICANI'LY BIODIETED DIRECT
BANK NUh5RE.R FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTR4CTOR .RJSTIFHCACI'ION (CORRECT ANSWER +(I):
a. Piausible since TS requires that a C S P he made inoperabie before these plant conditions are
established, but GP-008 requires that at least one CSHP be lhctional under these conditiuns.
b. Plausible since the CSII' can provide more flow than Reactor Makeup is capable of providing, hut the
suction source for the CSIP would be the RWST.
c. Plausible since starting a C S P is preferred to gravity feed, but only because the KCS may be
pressurized. If the RCS is depressurized, gravity k e d will provide adequate flow.
d d. If the RCS is pressurized, gravity flou. may be insufficient to provide adequate makeup to the RCS.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I IRECALL
mmcucri~RATING: 3
EXPLANATION: Analysis of plant conditions to determine appropriate response and reason for
response
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 18
Given the f~ololowingconditions:
e Containment temperature is 96 "F.
e Containment Fan Coolers (AH-1 2 / 3 / 4) are operating in the Normal Cooling
i\/lQde.
e i\ loss of offsite power occurs and thc plant responds as expected.
'The Containment Fan Coolers should he aligned with one ( i ) fan associated with each
fan cooier operating in ...
a. high speed and discharging to the concrete airshaft
b. high speed itrid discharging to the post-accident discharge duct
c. low speed and discharging to the concrete airshaft
d. low speed and discharging to the post-accident discharge duct
ANSWER:
a. high speed and discharging to the concrete airshaft
Post Validation Revision
Harris NKC Written Emmination
Reactor Operator
Data Sheets
QIJESTIONNUMBER: 18 TIEWGROUP Ill
10CFR55 CONTENT: 41(b) 9 43w
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-3.0-R4
PKEDICT the response(sj of the Containment Cooling Subsystems to the foliowing signals.
- 1,OSP
DEVELOPMENT REFERENCES: SI)-169
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CCS-R4 001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFHCACTION (CORRECT ANSWER dd):
4 a. Orie 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, but the discharge is to the concrete airshaf?
not the post-accident discharge duct.
e. Plausible since this fan response is the response to a LOCA start signal and they do discharge to the
concrete airshaft, but the fans operate in high speed following a loss of offqite power.
d. Plausible since this is the response to a LOCA stari signal, hut the fans operate in high speed and they
discharge to the concrete airshaft following a loss of offsite power.
IClJLTY ANALYSIS:
COMPREHENSnE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge ofthe response ofthe containment fan cooler fans to a loss of
offsite power
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
cpmsrIorv: 19
\;\ihich ofthe following can all be used to confirm that an inoperable / stuck rod is to be
considered misalibaecl?
a. a Delta-H indication
Power range channels
6 Reactor vessel level indication
a Core AT
b. = Delta-I indication
a Power range channels
a QPTR calculation
a Core outlet thermocouples
c. e Core AT
e Power range channels
a QPTR calculation
6 Core outlet thermocouples
d. a Delta-I indication
a Power range channels
a Reactor vessel level indication
6 Core outlet thennocouples
ANSWER:
b. a Delta-I indication
e Power range channels
a QPTR calculation
a Core outlet thermocouples
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION N I M U E R 19 TIEW/GROUP: 112
KAIMPORTANCE: I90 2.5 sa0
10CFR55 CONTENT: 41(b) 6/10 43(b)
KA: OOOOOSAK2.02
Knowledge of the interrelations between the Inoperable i Stuck Control Rod and the following: Breakers,
relays, disconnects, and control room switches
OBJECTIVE: ROP-3.1-3
LIST the indications of a misaligned rod specified in AOP-001, Attachment I, Indications of Miwligned
Rod
DEVEI ,OPMENT REFERENCES: AOP-OO 1
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW S m m c A i v r L Y MODIFIED DIRECT
BAWK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: AOP-3.1-R3 002
AOP-3.1-R3 003
NRC EXAM HISTORY: None
DISTRACTOR JWXIFICACTION (COMECT ANSWER d'd):
a. Plausible since delta-I and QPTR are used to determine a misaligned rod, hut RVkIS and core A'I' are
not used.
d b. Delta-I, QPTR, power range channels, and core outlet thermocouples are all used to determine a
miwdigned rod.
c. Plausible since QPTR, power range channels, and core uutlet thermocouples are all used to determine
a misaligned rod, hut core AT is not used.
d. Plausible since delta-I, power range channels, arid core outlet thermocouples are all used to determine
a misaligned rod, hut KVLIS is not used.
DIFFICULTY ANALYSIS:
DIFFICULTY RATING: 3
EXPIANATION: Knowledge of the indications of a misaligned rod per procedure
Post Velidation Revision
IIarris NRC Written Examination
Keactor Operatcr
QUESTION: 20
Given the following conditions:
- Following an accident, EOP-FRP-C. 1 "Response to Inadequate Core Cooling," is
~
being performed.
e ERFIS is inoperable.
e Plant parameters are as follows:
e ICCM highest TC = 672" F
a RCS WR Itmjxrature (highest) = 688" F
- RCS pressure PT-440 = 1535 psig
e RCS pressure PT-402 = 1635 psig
a CNMT pressure PT-95 1 = 4.5 psig
What value of superheat should be reported?
a. 63'F
b. 71 '1:
c. 49'F
d. 87 'F
ANSWER:
a. 63 OF
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Datu Sheets
QUESTION NWfBER: 20 TIIERIGROUP: i /2
P(DCPR55CONTENT: 41(b) 5 43@1
IBA: 000074EA2.01
Ability to determine or interpret the following as they apply to a Inadequate Core Cooling: Subcooling
margin
OBJECTIVE: EOP-3.19-4
Given a set of conditions during EOP implementation, IiETERiMINE the correct response or required
action based upon the EOP User's Guide general information
e Determining an KCS subcooling value
DEVELOPMENT KEFEKENCES: EOP-IJsers Guide
KEFEFENCES SUPP1,IED TO AFT'LHCANT:
~
Steam Tables
QUESTION SOUKCE: NEW SIGNIFICANTLY MODIFIED DIRECT
RANK KUMIEK FOR SIGNIFICAN'rLY MODIFIED 1DIRECT: EOP-3.19-R4 003
NRC EXAM IIISTORY: None
DISTRACTOR JUSTIFICAQTHON(CORRECT ANSWER \I7@:
4 a. When ERFIS is not available, the highest ICCM temperature should be used. If EWIS is not
available and adverse containment conditions exist, PT-402 should be used for pressure. Saturation
temperature for 1635 psig is 609 OF, so the amount of superheat is 63 "E' (672-609).
b. I'lausible since the superheat determined using the %CCMtemperature and saturation for the lowest
RCS pressure of 1535 psig (not used because of adverse containment conditions) is 71 "F (672-601).
E. Plausible since the superheat determined using the hot leg temperature (not used if ICCM is available)
arid saturation for the PT-402 pressure of I635 psig is 79 "E' (588-609).
d. Plausible since the superheat detemiincd using the hot leg temperature (not used if ICCM is available)
and saturation for the lowest KCS pressure of 1535 psis (not used because of adverse containment
conditions) is 87 "F (688-601).
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANAL.YSIS KNOWLEDGE /RECALL
DIFFICUI,TY RATLVG: 3
EXPLANATION: Knowledge of instruments to use and calculation of subcooling by applying
steam tables
Post Validation Revision
Harris hXC Written Examination
Reactor Operator
QUESTION: 21
A failure of Containment Cooling causes equilibrium Containment temperature to
increase from 105 OF to 130 "F.
Assuming no change in Tave, P M pressure, or Letdown flow rate, how will this effect
ICs-231, FK-122.1 CHARGING FLOW?
a. It will throttle open slightly during the course of the temperature change and then
retuni to its original position
b. It will throttle closed slightly during the course of thc temperature change and then
return to its original position
e. It will throttle open slightly during the course of the temperature change and
remain in that position
d. It will throttle closed slightly during the course ofthe temperature change and
remain in that position
ANSWER:
b. It will throttle closed slightly during the course of the temperature change and then
return to its original position
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 21 TIEWGROUP: 2i 1
lOCFR55 CONTENT: 4l@) 719 43m
KA: 022K3.02
Knowledge of the effect that a loss or tnalfunction of the CCS will have on the following: Containment
instrumentation readings
O B ~ E c r w E : CVCS-3.0-R3
DESCRIBE the controls and interlocks of remotely operated CVCS valves, including the following:
e CVCS controllers, including transfers between automatic and manual control, setpoint determination
and adjustment, and output control
DEVELOPMENT REFERENCES: SD-I 00.3
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: SIGNIFICANTLY MODiFIED
C m m Y MODIFIED DIRECT:
NRC EXAM IIISTORW None
DISTWACIOR .JLrSTIFICACTION (CORRECT ANSWER dd):
10. Plausible since a cotitainnient temperature increase will affect indicated pressurizer level, but
indicated level will increase so charging flow would decrease.
d b. As containment temperature increases, indicated pressurizer level increases due to heating of the
reference leg. This would result in a smaller Awhich would indicate that pressurizer level is high.
Charging flow will decrease to lower actual level and then return to its original value to match
letdown flow.
6. Plausible since a containment temperature increase will affect indicated pressurizer level, but
indicated level will increase so charging flow would decrease.
d. Plausible since a containment temperature increase will affect indicated pressurizer level and charging
Row will decrease, but the flow will return to its original value to match letdown flow.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNO\VEDGE / RECALL
DIFFICUI,TY RATING: 4
EXPLANATION: Analyze the effect ofthe temperature change on pressurizer level and then
determine how this change affects the operation ofFK-122.1
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QIJESTION: 22
Given the following conditions:
e The unit is operating at 12% power.
e The following RCP vibrations are observed:
INDICATION WCP 6A' RCP 'B' KCP 'CY
Frame Vibration 3.6 mil and ? at 2.8 mil and stable 4 mil and ? at
0.3 mil per hr 0.1 mil per hr
Shaft Vibration 12 mil md ? at 7 mils and stable 14 mils and ? at
0.3 mil per hr 0.6 mils per hour
Which of the following describes the actions required for this condition?
a. Stop RCP 'A' and initiate a plant shutdown
b. Trip the reactor, stop RCP 'A', and go to PATH- 1
c. Stop RCP 'C' and initiate a plant shutdown
d. Trip the reactor, stop RCP IC', and g o to PATH-1
ANSWER:
a. Stop RCP 'A' and initiate a plant shutdown
Post Validation Revision
Harris NRC Written Examination
Keactor Operator
Data Sheets
QUF,STBON NUMBER 22 TIEWGROUP: 2: 1
10CFR55 CONTENT: 4P(b) 3/10 43(b)
IGa. 003A1.01
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) asscciated
with operating ?he RCPS controls including: RCP vibration
OBJECTIVE: AOP-3.18-3
Given a set of plant conditions and a copy of AOP-018, DE'FEKMTNE the appropriate response
DEVELOPMENT REFERENCES: AOP-018
REEE,NENCES SUPPLIED TO APPLICANT: AOP-01 8, Attachment 1
QUESTION SOURCE: NEW SIGNIFKCANTEY MODIFIED DIECT
BANK NUMBER FOR SIGNIFICANTLY RIODIFIED /DIRECT: AOP-3.18 01 9
NRC EXAM HISTORY: None
DISTRACTOW JIJSTHFICACTION (COKRECl ANSWER d'd):
4 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 he shutdown.
b. Plausible since these would be the correct actions if the plant was in Mode 1. but the plant is in Mode
c. Plausibie since these are the correct actions, but 'C' RCP has not reached any trip limits while 'A' RCP
has.
d. PlausibIe since these would be the correct actions if the plant was in Mode I, but 'C' RCP has not
reached any trip limits while 'A' RCP has and the plant is in Mode 2.
DIFFICUL'IT ANALYSIS:
COMPWEIIFX'"E /ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis to determine which RCP must be stopped and comparison to power
level to determine proper action
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 23
Given the following conditions:
a The Pressurizer Relief Tank (PRT) is being cooled by recirculation through the
Reactor Coolant Drain Tank Heat Exchanger per OP-100, Reactor Coolant System,
and 01-120.08, Radioactive Equipment Drain System.
e IED-143, RCDT RECIRC ISOIATIOM, loses its air supply.
1ED- 143 will fail open, ...
a. but NOT effect the PRT cooldown because it is already open during this
evolution.
b. slowing down the cooling of the PRT due to starting a recirculation of the RCDT.
c. but NOT affect the PRT cooldown because 1ED- 139 is shut.
d. causing a lowering level in the PRT as coolant is diverted to the RCDT.
ANSWER
d. causing a lowering level in the PRT as coolant is diverted to the RCDT
Post Validation Revision
Harris NRC Written Examination
Keactor Operator
Data Sheets
QUEST1O.Y NUMBEW: 23 TIEWGROUP: 2/1
B O C F RC~o~m m r : 41(b) 3 @@I
KA: 007K4.01
Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following: Quench tank
cooling
OBJECTIVE: PZR-3.0-3
Given il flow diagram of the PRT or associated subsystems and the appropriate procedure, correctly
ALIGN the P W for filling, draining, recirculation, or cooldown
DEVEI,OPMENF REFERENCES: C A R 2165-S-1313
REFERENCES SUPPLIED TO APPIJCANT: CAR 21654-1313
QUEsTHON SOURCE: NEW SIGNIFICAVTLY MODIFIED DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIEI) I DIRECT: PZR-I33 002
NRC EXAM HISTORY: None
DISTRACTOR JUSTHFICACTHON (CORRECT ANSWER +d):
8. Plausible since IEIP-143 does fail open. but it is not open during the cooldown evolution
b. Plausible since 1ED-143 does fail open and will cause the KCDT to recirc, but it is not open during
the cooldown evolution.
C. Plausible since IEIP-143 does fail open, but iED-I39 is not closed during the cooldown evolution.
d d. iED-143 failing open will result in the PRT level decreasing and the KCDT level increasing as water
is transferred from the IRT to the RCUT.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE 1RECALL
DIFFICULTY RATING: 3
EXPLANATION: Anaiysis of the effect of a valve failure on PRT cooldown, having knowledge of
valve alignment required
Post Validation Revision
Harris NIPC Written Examination
Reactor Operator
QUESTION: 24
While operating at 100% power, 125 VDC bus DP-1B-SW is isolated due to a fault.
Which of the following identifies two (2) Technical Specification Action Statements that
must be entered as result of the bus fault?
a. 0 3.4.1.2, AFW Modes 1,2, and 3, due to the TDAFW pump being inoperable
as a result o f a loss ofpower to one (i) of the steam supply valves
D 3.4. I. I, Reactor Coolant Loops and Coolant Circulation, due to the RCPs
being inoperable as a result o f a loss of tripping power to the motor breakers
b. e 3.4.1.1, Reactor Coolant Loops and Coolant Circulation, due to the RCPs
being inoperable as a result of a loss of tripping power to the motor breakers
e 3.6.5, RCS Leak Detection, due to RM-3502A being inoperable as a result of
the sample isolation valves automatically closing
c. e 3.7.1.2, AFW Modes I, 2 , and 3, due to the TDAFW- pump being inoperable
as a result of ti loss o f power to one (1) of the steam supply valves
e 3.8.1 .l, AC Sources Operating, due to the E D 6 being inoperable as a result
of a loss ofpower to the EDG governor control circuit
d. 3.8.1. I, AC Sources Operating, due to the EDG being inoperable as a result
of a loss of power to the EDG governor control circuit
e 3.6.5, RCS Leak Detection, due to KM-3502A being inoperable as a result o f
the sample isolation valves automatically closing
ANSWER
c. e 3.7.1.2, AFW Modes I, 2 , and 3, due to the TDAFW pump being inoperable
as a result o f a loss ofpower to one (1) ofthe steam supply valves
e 3.8.1 .l, AC Sources Operating, due to the EDG being inoperable as a result
of a loss of power to the EDG governor control circuit
Post Validation Revision
Harris NRC: Written Examination
Reactor Operator
nata Sheets
QUESTION NUMBER: 24 TIWGROUP: lil
KAIMPORTANCE: RQ 3.4 SRQ
IOCPR55 CONTENT: 41@) 8 4Xb)
KA: 064K2.03
Knowledge of kDG bus power supplies to the following: Control power
OB3ECTIVE: AOP-3.25-3
Given plant conditions, DISCUSS the following notes, cautions, and procedural steps as they apply
The effects of a loss of a DC bus on equipment operability (Le., DG, sequencer, and TL) AFW)
DEVELOPMENT REFERENCES: AOP-025
REFERENCES
~~~ -~ SUPPLIED TO MPLICANE None
QBJESTION SOURCE: NEW SIGNIFICANTLY MODIFBED
RANK NUMBER FOR SIGKIFICANTLY MODIFIED / DIRECT: AOP-3.25-R3 004
NRC EXAM HISTORY: None
PIISTRACTOR JUSTIFICACTPON (CORRECT ANSWER .Id):
a. Plausible since the TDAFW pump is inoperable and the KCPs use TIC power for the tripping coils, but
tripping the RCPs is not part of the operability requirement.
b. Plausible since the RCPs use DC power for the tripping coils and the RCS leak detection sample
valves will isolate on a loss of power, but tripping the RCPs is not pa^ of the operability requirement
and the sample isolation valves close on a CVIS due to a loss of AC power.
d e. The TDAFW pump is inoperable due to a loss of power to a steam supply valve and the EDG is
inoperable due to a loss of power to the governor, as well as to the generator excitation control circuit
and sequencer.
d. Plausible since the EDG is inoperable and the RCS leak detection sample valves will isolate on a loss
of power, hut the sample isolation vaIves close on a CVIS due to a loss of AC power.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECXLL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of the effect o f a loss ofDC power on equipment operability and
knowledge of the BSLCOs affected
Post Validation Revision
Harris iWC: Written Examination
Reactor Operator
QUESTION: 25
Given the folIowing indications during a plant s t m p :
e Power Kange Channel N-41 26.0%
e Power Range Channel N-42 24.5%
e Power Range Channel N-43 24.5%
e Power Range Channel N-44 25.0%
Loop'A' AT 25.5%
e Loop'B' AT 25.5%
e Loop'C' AT 25.5%
e Turbine Load 24.5%
Which of the following power levels should he reported as being actual reactor power?
a. 24.5%
b. 25.0%
C. 25.5%
d. 26.0%
ANSWER:
c. 2 5 9 0
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 25 TIEWGROUP 212
1QCFR55CONTENT: 4L(h) 3 43(W
MA: 002K5.10
Knowledge o f the operational implications of the Following concepts as they apply to the RCS:
e Relationship 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
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
OR SIGNIFICANTLY MODIFIED / DIRECT: NIS-RIO 004
NRC EXAM HISTORY None
DHSTRACTOR JUSTIFICACTION (CORRECT ANSWER '/'(I):
a. Plausible since this is the lowest given power level and may be considered to be the most
conservative, but GP-005 provides guidelines for which power level should be considered.
b. Plausible since this is the average NIS power level, hut the highest as identified by GP-005
requirements is the average loop AT^
d c. Until a calorimetric is perfomled 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. Plausibie since this is the highest given power level and may be considered to be the most
conservative, but CP-00s provides guidelines for which power level should be Considered.
ICULTY A N a Y S I S :
COMPREHENSIVE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Calculatiun of average power indications and detemlination of most
consewativc value
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 26
AH-82A, NORMAL PURGE SCJPPLY FAN AH-82A, fails to start when the control
switch is placed in START.
Which of the following intsrloclcs would prevent the fan from starting?
a. Normal Purge Inlet and Discharge Valves are open
b. AW-82A fan inlet damper is closed
c. Fan inlet air temperature is low
d. Containment differential pressure is zero
ANSWER
d. Containment differential pressure is zero
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 26 TIEWGROUP: 2i2
10CFR55 CONTENT: 41(b) 9 43m
KA: 029A1.03
Ability to predict and/or monitor changes in parmeters to prevent exceeding design limits) associated
with operating the Containment Purge System controls inchding: Containment pressure, temperature,
and humidity
OBJECTIVE: CVS-3.0-I22
LOCATE the controls and MII.AIN the interlocks associated with the following major components
NCPMU units, including AH-82 fans
DEVELOPMENT REFERENCES: OP-168
REFERENCES SUPPLIED TO MPLBCANT: None
QIJESrION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK IVUMEER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC: EXAM HISTORY None
DISTR4CTOR JUSTIFICACTIBN (COIPRECT ANSWER dd):
a. Plausible since the valves are interrocked to close if fan N M 2 A is stopped, but are manually opened
prior to the start of the fan.
&. Plausible since the inlet damper is interlocked to open when the fan i s started, Rut are closed when the
fail is started.
c. Plausible since a low inlet air temperature will cause an alarm condition: hut will not prevent the fan
from starting.
d d. Fan AH-S2A will only start if containment AP is more negative than -0.400 N O .
DIFFICULTY ANALYSIS:
COMPREIIENSRE /ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of interlocks associate with containment purge fans
Post Validation Revision
Harris hRC Written Examination
Reactor Operator
QUESTION: 27
Given the following conditions:
6 The plant is at the Point of Adding Heat (POAII) when a SG PORV fails open.
6 RCS temperature decreases and stabilizes at 548 "F.
Which of the following predicts the plant response and the operator actions requircd?
a. Reactor power increases; withdraw control rods and dilute, in a controlled
manner, to restore RCS temperature to program within 15 minutes
b. Reactor power increases; trip the reactor if RCS temperature CANNOT be
restored above 551 O 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 increases; trip the reactor if RCS temperature CANNOT be
restored above 551 *F ia a controlled manncr within IS minutes
Post Validation Revision
Hamis NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 27 TIEWGROUP 211
1QCFR55CONTENT: 41(b) 6/10 43(b)
KA: 039.42.05
'4bility 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 mitigate the consequences of those malfunctions or
operations: increasing steam demand, its relationship to increases in reactor power
OBJECTIVE: E-3.10-1
Apply the philosophies of OMM-001 and PLF-629 regarding safe and conservative decisions that must
be made by a control room crew
DEVELOPMENT REFERENCES: OMM-Q01
IE-LP-3.60 (Salem Event, SOER 94-01)
REFERENCES SUPPLIED TO APPLICANT: None
QgJESTION SOURCE: NEW SIGNIF1CA"TLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTR4CTOR ~S'HFICACTPON(CORRECT AKSWER d'd):
a. Plausible since reactor power will increase, but temperature is not to be restored using two different
methods of reactivity control simultaneously and the 15 minute limit is to restore temperature above
55 1 O F , not to program.
4 b. The first operator action should be to attempt to stop the cause (e&, secure the overfeeding) of the
transient. 'i emperature may then be recovered by using control rods in a slow and controlled manner.
Temperature has to be restored to greater than 55 1 "F within 15 minutes due to the requirements of TS
c. Plausible since the 15 minute time limit is associated with restoration, but the reactor does not become
subcritical.
d. Plausible since the reactor is to he tripped if it becomes subcritical due to a malfunction per OMM-
001, but the reactor does not become subcritical.
I)IFFI'ICULTY ANALYSIS:
COMPKEIIENSIVE I ANALYSIS KNOWLEDGE [RECALL
DIFFICULTY HATING: 3
EXP1,ANATION: Analyze the plant response to an increase in steam demand and determine
appropriate actions
Post Validation Revision
Harris N R C : Written Examination
Reactor Operator
QUESTION: 28
The plant is operating at 100% power with the following conditions:
T d Ambient Temp CT Basin 'Temp
I500 35 O F 64 "F
1900 20 "F 60 OF
2300 10 O F 58 "F
Which of the following describes the correct CT Deicing Gate Valve alignment for these
conditions?
1900 2300
a. Full Open Full Open
b. Full Open Half Open
C. Half Open Full Open
d. Half Open Half Open
ANSWER:
b. Full Open Half Open
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 28 TIEWGROUP: 3
BCCFH55 CONTENT: 41(b) 10 43(b)
m: 2.1.25
Ability to obtain and interpret station refexme materials such as graphs, monographs, and tables which
contain perfonnance data
OBJECTIVE: CT-R3
Given OP-141, Attachment 5 , ANALYZE a set ofadverse weather conditions and DESCRIBE the
operation of the Cooling Tower System to prevent ice damage to the fill material
DEVEI,OPMENT REFERENCES: OP-141
REFERENCES SUPPIJED TO APPLICANT: OP-141, Attachment 5
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGXTIFICANT1,YMODIFIED / DIRECT: CT-R3 001
NRC EXAM HISTORY: Harris NRC 2000
DISTRACTOR .JUSTIFICACTION (CORRECT ANSWER .Id):
8. Plausible since valves should he open at 1900, hut are required to be changed to half open at 2300,
d lp. 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 full open to half
open.
d. Plausible since valves should be half open at 2300, hut should be full open at 1900 due to no change
from 1500.
DIFPICULTY ANALYSIS:
COMPKEIHENSIVE I ANALYSIS 0 KNOWLEDGE / RECALL
DIFFICULW RATING: 3
EXPLANATION: Application of given data to curve to determine required operation of deicing
valves
Post Validation Kevision
Harris NRC Written Examination
Rcilctor Operator
QUESTION: 29
Which of the following conditions requires processing a Radioactive Gaseous Release
BATCH permit'?
a. Manual operation of the Containment Vacuum Relief System
b. Resetting and starting the Containment Pre-Entry Purge following an automatic
isolation
C. Startup of the standby Airborne Radioactive Removal fan (S-I) following a trip of
the running fan
d. Swapping the operating Normal Containment Purge Makeup (AH-82) fans frum
Train A to Train B
ANSWER:
b. Resetting and starting the Containment Pre-Entry Purge following an automatic
isolation
Post Validation Revision
Hmis NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 29 TIEWGROUP 3
10CFR55 CONTENT: 41(b) 12 43m
KA: 2.3.11
Ability to control radiation relases
OBJECTIVE: CVS-3.0-R4
EXPLAIN the conditions which require a radioactive release permit prior to operating components
associated with the Containment Ventilation System
DEvELOpMENr REFERENCES: OP-168
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODPFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED 1 DIRECT: CVSR4 00 1
NRC EXAM HISTORY: None
DISTRACTOR JUSTEPICACTION (CORRECT ANSWER +a):
1p. Plausible since the Containment Vacuum Relief System interfaces with Containment atmosphere, but
it supplies air to Containment and does not require a release permit to operate.
s' b. For initial start-up of the Containment Pre-Entry Purge (start of an outage) or if purge was secured for
radiological reasons, a Ratch release pennit is required
c. Plausible since if this was the initial startup ofthe system a batch release wrouId be required, but once
the system is in operation for a period oftime only a continuous release permit is required.
8. Plausible since if this w'as the initial SVdI'tllp of the system a batch release would be required, hut once
the system is in operation for a period of time only a c.ontinuous release permit is required.
DIFFICL%TY ANALYSIS:
GQMPREIPENSNE I ANALYSIS KNOWLEDGE I KECALI,
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the radiological release permits required
Post Validation Revision
Hanis NRC Written Examination
Reactor Operator
QUESTPOW: 30
Which of the following two (2) conditions are both identifial hy EOP-EPP-Oi3, LOCA
Outside Containment, as being used to identify that the LOCA has been isolated?
a. e RCS pressure increasing
E- RAD local room temperalures
b. RAE3 iocal room temperatures
E- RAB radiation levels decreasing
e. e RAR radiation levels decreasing
- Local observation of the isolation
d. e RCS pressure increasing
E- Local observation of the isolation
ANSWER:
d. E- RCS pressure increasing
m 1,ocai observation of the isolation
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NU-MBER: 30 TIERKROUP: 1/1
KAIMPORTANCE: RO 3.5 SRQ
10CFR55 CONTENT: 41(b) 10 43(W
KA: WE04EK1.2
Knowledge ofthe operational implications ofthe following concepts as they apply to the (LOCA Outside
Containment) Normal, abnormal and cmergency operating procedurcs associated with &OCA Outside
Containment)
OBJECTIVE: EOP-2.3-R4
Using appropriate plant procedures and prints, determine the following:
e Transitions to other EOPs
DEVELOPMENT REFERENCES: EOP-EPP-0 13
REFERENCES SUPPIJED TO APPLICANT: None
QUESTIQN SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DPKECT: EOP-3.3 024
NRC EXAM HISTORY: None
DISTRACTOR SIJSTIFICACTION (CORRECT ANS\IER dd):
a. Plausible since RCS pressure increasing is one of the indications used, hut pressurizer level may not
be indicative of actual RCS inventory or the leak being isolated and is not used in IIPP-013.
b. Plausihle since these 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 not
used in EPP-013.
c. Plausible since local observation is one of the indications used, but KAB radiation levels may he
elevated for some time after isolation and is not used in EPP-013.
4 d. EPP-013 determines that the ILEA outside containment is isolated if RCS pressure is increasing and
if local observation confirms the isolation.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge ofthe conditions required by EPF-013 to determine that a LOCA
outside containment is isolated
Post Validation Revision
Harris NRC Written Examination
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 RCP runout in the event of a large break LOCA
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 srnali break in the RCS
Post Validation Revisioti
Harris NKC Written Examination
Reactor Opcrator
Data Sheets
QUESTION NUMBER: 3 1 TIEWGROUP: 111
10CFH55 CONTENT: 4f(b) 3/10 43(b)
KA: 000009EK3.23
Knowledge of the reasons for the following responses as the apply to the small break LOCA: RCP
tiipping requirements
OBJECTIVE: BD-3.1-1
Analyze the Reactor Coolant Pump (RCP) trip criteria. This analysis should include, at the minimum, the
following topics:
The reason for purposely tripping the KCPs under c e r t h accident conditions
DE\'E;LOPMEN'F REFERENCES: Generic Issues of ERG Background - Executive Voiume
LP-BD-3.1
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: BD-3.1 001
NHC EXAM IIISTORY: None
DISTRACTOR JUSTIFICACTIBN (CORRECT ANSWER d'd):
a. Plausible since for most accidents it is desirable to have RCPs available, particularly those cases where
an inadqnate core cooling condition might exist.
b. Plausible since little work is required by the RCPs in the event o f a large break L O W , but this would
result in a lower pump current. not a runout condition.
d C. Tripping the RCPs during the early stages of a small break L.OCA limits the amount of mass lost out
the break, thereby increasing the mass available for heat removal in the event the purrips werc not
tripped but tripped at a later time.
d. Plausible since RCPs are not designed to pump a two-phase mixture and it would be desirable to
protect the pumps from damage.
DIFFICULTY ANALYSIS:
COMMPREIIENSIVE I ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 2
EXPLANATION Knowledge of the reasons for tripping RCPs during a small break LOCA
Post Validation Revision
Harris MRC Written Examination
Reactor Operator
QUESTION: 32
WhiIe operating at 100% power, a failure ofthe Pressurizer Pressure AUTO controller
(PK-444A) occurs and the Reactor Operator takes manual control ofthe controller.
While restoring from the failure, in order to maintain PRZ pressure at 2235 psig, the
Reactor Operator should adjust PK-444A to setpoint of approximately ...
a. 31%.
c. 69%.
d. 89%.
ANSWER:
c. 67%.
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER. 32 TKEKIGROUP: 1/1
1UCFRSS CONTENT: 41(b) 3/7 43(W
Kk 000027AK2.03
Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the
following: Controllers arid positioners
OBJECTIVE: PZRPC-3.0-2
E)ITL.AIN how the pressurizer pressure control setpoint is both determined and adjusted for a desired
KC:S pressure
DEVELOPMENT REFERENCES: SD-100.3
QUESTION SOURCE:
- -
REFERENCES SUPPLIED TO APPLICANT:
u NEW
None
SIGNIFICANTLY MODIFIED u DIRECT
BANK NUMBER FOX SIGNIFICANTLY MODIFIED / DIRECT: PZRPC-R5001
NRC EXAM HISTORY: None
DISTRACTOX JUSTIFICACTlON (CORRECT ANSWER d'd):
a. Plausible ifthe setpoint is calculated by dividing 535 psig (2235-1700) by 1700 p i g (Low end of
span), with a result of 31.4%.
b. Plausible since this is the mid-point of the 0-100% scale
4 E. PK-444A setpoint is detennined by calculating the percent of span of the contmller. Controller span
is 800 psig (1700 - 2500). 2235 psig - 1700 psig := 535 psig. 535 psigl800 psig = 66.9%
d. PIausible if the ratio of 2235 psig to the upper end of the span (2500 psig) is calculated, with a result
of 89.4%.
DIFFICULTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE I REGALL
DLFIWUII.TY RATING: 3
EXPLANATION: Calculation of required setpoint for pressurizer pressure control
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 33
Which one of the following correctly describes how and why the speed of the Condensate
Booster Pumps (CBPs) is varied?
a. Changing the coupling impeller vane pitch to maintain a constant 430 psig feed
pump suction pressure
b. Changing the coupling impeller vane pitch to maintain desired flow from the
CBPs to the feed pumps
c. Varying the miount of oil to the coupling between the pump and motor to
inaintain a constant 430 psig at the fed pump suction
d. Varying the amount of oil to the coupling between the pump and motor to
maintain a desired flow from the CBPs to the feed pumps
ANSWER:
c. Varying the amount of oil to the coupling between the pump and motor to
maintain a constant 430 psig at the feed pump suction
Post Validation Revision
Ifanis NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 33 TIERIGROUB: 211
10CFR55 CONTENT: 41(b) 4 4Xb)
KA: 056G2.1.28
Knowledge ofthe purpose and function of major system components and controls. (Condensate)
OBSECTTVE: CFW-3.0-4
DESCRIBE lhe basic constmetion and operation ofthe following CFW System components /
subsystems
o CRP Variable Speed Fluid Coupling (VSPC)
DEVELOPMENT REFERENCES: SD-134
REFERENCES SUPPLIED TO APPLICANT: None
Qt?ESTIOX SOURCE: NEW SIGNIFIC.4NTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIIPECT: CFW-R.? 001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORXECT ANSWER d'd):
8. Plausible since the variable speed coupling maintains 430 psig at the feed pump suction, but it is
maintained by using oil between the motor and pump coupling.
b. Plausible since this is a means of providing a variable flow rate, but the CBPs used a variable speed oil
4 e. An oil bath between the motor and pump coupling causes the pump tto operate at a variable speed to
rnahtain a constant 430 psig suction 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 430 psig suction at the feed pump rather than a
constant flow rate.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS #IQ\VLEI>GE. /RECALL
DIFFICULTY HATING: 3
EXPLANATION: Knowledge of the operation of the CBPs
Post Validation Revisicn
Harris NRC Written Examination
Reactor Operator
QUESTION: 34
Given the following conditions:
0 The plant is operating at 100% power.
A tube leak has been detected on 'W'SG.
The Condenser Vacuum Pump Rad Monitor, REM- 1TV-3534, and H-X- 15 curves are
being monitored every 15 minutes to estimate the leak rate.
CYPE is operating with NQ motivating air.
Which of the following readings noted on REM-1TV-3534 is the MINIMUM reading
that would require a plant shutdown per Technical Specifications?
a. 5.40 E -7
b. 6.00 E -7
c. 1.08 E -6
d. 1.80 E -6
ANSWER:
c. 1.08 E -6
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 34 TIEWGROUP 1/2
KAIMPORTANCE: KO 3.2 SRO
10CFH55 CONTENT: 41(b) 7;10 43(b)
KA: 000037AA2.10
Ability to determine and interpret the foliowing as they apply to the Steam Generator Tube Leak Tech-
Spec limits for RCS leakage
OBJECTIVE: AOP-3.16
For a primary-to-secondary leak, DESCRIBE when a power reduction or unit shutdown is required.
DEVELOPMENT REFERENCES: AOP-0 16 (unknown)
Curves PI-X-15aihic
REFERENCES SUPBIdED TO APPLICANT: Curves H-X-ISalbic
QUESTION SOURCE: NEW SPGNIFICM'TI,Y MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: Hams NRC 2000-80
NRC EXAM HISTORY: Hams NRC 2000
DHSTRACTOW .PUSTIFICACTPON (COHWECT ANSWER d'd):
1%. Plausible since this exceeds would exceed I'SAL 2 limits if operating on full motivating air (curve H-
X-ISa), but the incorrect curve is used.
b. PIausihle since this exceeds would exceed PSAI, 2 limits if operating on intermediate motivating air
(curve H-X-ISb), but the incorrect curve is used.
4 e. Lowest level that would exceed 75 gpd (PSAL 2) which would require a TS shutdown
d. Plausible since this exceeds the PSAL 3 limit which would require a 'I'S shutdown, but this is not the
lowest level that would require the shutdown.
DIFFICULTY ANALYSIS:
COMPREHENSEVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Interpretation of plant data on KCS leakage curve and comparison to procedural
reyuirenients
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTIQN: 35
FRP-J.2, Response to Containment Flooding, directs that the containment sump be
sampled for activity, and then to notify the operations staff of sump level and the sample
results.
What action will the operations staff be considering based on this information?
a. Isolation of the Cold Leg Accumulators
b. Isolation of the CNMT spray additive tank
e. Shift to Hot Leg Recirculation
d. Transfer of sump water to tanks outside containment
ANSWER
d. Transfer of sump water to tanks outside containment
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESlXOX KkrhXBEW: 35 TIEWGROUP: I 12
KAIMPORTANCE: KO 2.4 SRO
IIICFRR55 CONTENT: 41@) 9/10 43(b)
Kh: WE15EK1.2
Knowledge of the operational implications of the following concepts RS they apply to the (Containment
Flooding) Nonnal, abnormal and emergency operating procedures associated with (Containment
Flooding)
OEECTIVE: EOP-3.13-4
Given the following EOP steps, notes, and cautions, DESCRIRE the associated basis
0 Sampling the CNMT sump for activity (J.2)
DEVELOPMENT REFERENCES: EOP-FW-J.2
LP-IOP-3.13
REFERENCES SUPPLIED TO APPLICANT: None
QUESrION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BAKK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOP-3.13 010
NHC EXAM HISTORY: None
DPSIRACTOR JUSTIFIICACTION(CORRECT ANSWER dd):
a. Plausible since if flooding has occurred it is likely that a large RCS leak has also occurred and the
accumulators have dumped to containment and would no longer be needed, but this sample is to
detennine whether the water can be transferred.
b. Plausible since if flooding has occurred 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. but
this sample is to detennine whether the water can be transferred.
e. Plausible since a shift to hot leg recirc from the sumps will eventually be required in the event of a
large break LOCA, but this sample is to determine whether the water can be transferred.
4 d. The containment sump is sampled to determine if excess water can be transferred to storage tanks
located outside containment.
DIFFICULTY AKALYSIS:
COMPREIIENSIVE / ANALYSIS KNOWLEDGE I RFXALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of purpose for sampling sumps following flooding inside
containment
Post Validation Revision
Hank NRC Written Examination
Reactor Operator
QUESTION: 36
Given the following conditions:
KHR Pump A-SA is tagged out.
Following a large break LOCA, the crew was performing EOP-EPP-010, Transfcr to
Cold Leg Recirculation.
1SI-301, CONTAINMENT SUMP TO RHR PUMP B-SB, failed to open and the
crew transitioned to EOP-EPP-012, Loss of Emergency Coolant Recirculation.
Both Containment Spray Pumps automatically transferred to the Containment Sump.
Two (2) Containment Fan Chlers are operating.
Containment pressure is 12 psig and decreasing slowly.
W i l e performing EPP-012 the Reactor Operator notes that WWST level is 2%)with
both CSIPs, both Containment Spray Pumps, and RHR Pump B-SR operating.
Which ofthe following actions are to he taken?
a. Stop the RHR pump ONLY
b. Stop both CSIPs and the RHR pump ONLY
c. Stop both CSIPs, the RIHR pump, and one Containment Spray pump ONLY
d. Stop both CSIPs, the RKR pump, and both Containment Spray pumps
ANSWER:
b. Stop both CSIPs and the RIlR pump ONLY
Post Validation Revision
Harris NKC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 36 TIER/GROUP 1/1
10CFRS5 CONTENT: 41(b) 7 43(b)
MA WEllEKl.1
Knowledge of the operational implications ofthe following concepts as they apply to the (Loss of
Emergency Coolant k2circulation) Components, capacity, and function of emergency systems
OBJECTIVE: EOP-2.3-S2
Predict how each ofthe following could impact efforts to maintain core cooling during a LOCA
e Failure of valves to realign for cold-Ieg recirculation
DEVELOPMENT REFERENCES: EOP-CPP-012
REFERENCES SBJPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGSIFICANTLY MODIFIED DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOP-3.3-K5 004
NRC EXAM HISTORY: None
DISTRACIQR JUSTIFICACTION (CORRECT ANSWER dd):
a. Plausible since the KHR pump is still aligned to the RWST and must be stopped, but the CSIPs are
also aligned to the RWST and must likewise be stopped.
4 b. The RIIR pump and the CSPs are still aligned to the RWST and must be stopped when the RWST
empty alarm is received at 3% level.
c. Plausible since the RIIR 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 FWR pump and the CSIPs must be stopped, but the spray pumps can continue to
operate since they are no longer aligned to the RWST.
PCULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLE.DGEI RECALL
DIFFICULTY RATING: 3
EXPIANATION: Analyze plant conditions to determine which pumps are taking a suction from
the RWW to determine the pumps which are to be stopped
Post Validation Kevision
Harris NRC Written Examination
Reactor Operator
QUESTION: 37
Given the following plant conditions:
e The plant is operating at 100% power.
e 1CS-7,45 GPM Letdown Orifice A, and lC§-S,hO GPM Letdown Orifice B, are
closed.
e 1CS-9,60 GPM Letdown Orifice C, is open.
e The Reactor Makeup System is setup properly and is in AUTO.
e VGT level transmitter, LT-112, fails high.
Assuming NQ operator action, which of the following describes the piant response?
a. Charging Pump suction is eventually lost as VCT level decreases
h. ICs-120 (LCV-l15A), Letdown VCT/Hold Up Tank, aligns to the VCT and N O
automatic makeup will occur
c. ICs-120 (LCV-1 15A), Letdown VCT/Hold Up Tank, aligns to the HUT and a
CONTINIJOUS makeup to the VCT will occur
d. ICs-120 (LCV-I 15A), L,etdown VCT/Hold Up Tank, aligns to the HUT and
INTERMITTEXT makeups at normal setpoints will occw
ANSWER:
d. IC§-120 (LCV-I 15A), Letdown VCT/Hold IJp Tank, aligns to the HUT and
INTEKMITTENT makeups at normal setpoints will occur
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 37 TIEWGROUP: 211
IOCFWS CONTENT: 41(b) 7 4303)
KA: 00441.06
Ability to predict andfor 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 failures
c. LT-I 12 or LT-115 faiiure (high or low)
DEVELOPMENT REFERENCES: AOP-003
REFERENCES SUPPLIED TO APPLICAWT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Harris NRC 2002-32
NRC EXAIM HISTORY: Harris NRC 2002
DISTRACTOR JLJSTIFHCACTION (CORRECT ANSWER dd):
5. Plausible since this would occur with no operator action if the high failure were 1.T-I 15 instead of LT-
I12
b. Plausible since a low failure of LT-I 15 would result in this response
e. Plausible since this would occur if letdown were in excess or equal to makeup capability. but letdown
is less than makeup capability under the given conditions.
d d. LT-I 12 failing high causes LCV-115A to fully divert to the HUT tank at 60 gptn letdown flow. VCT
level decreases and automatic makeup raises level at 120 gpm, causing niakeup to stop until level
drops again.
DIFFICULTY ANALYSIS:
COMYREIIENSIVE/ ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPI,ANATION Requires analysis of plant response to failures in CVCS given initial plant
conditions
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 38
The plant is operating at 100% power with all e q u i p m t operable and properly aligned.
Which of the following describes changes to the Component Cooling Water System
alignment following a Safety Injection signal?
a. CCW to the Gross Failed Fuel Detector and Primary Sample Panel isolates
b. Both CCW pumps start and the Non-Essential header isolates
c. CCW to and from the RCP Motor Coolers isolates
d. Both CCW pumps start and the Thermal Banier Wx Return isolates
ANSWER:
a. CCW to the Gross Failed Fuel Detector and Primary Sample Panel isolates
Post Validation Revision
Iiarris NRC Written Examination
Reactor Operator
Data Sheets
QUESlION NUMBER: 38 TIEWGROUP: 211
10CFR55 CONTENT: 4I(b) 4 43w
KA: 008A3.08
Ability to monitor automatic operation of the CCWS, including: Automatic actions associated with the
CXWS that occur as a result ofa safety injection signal
OBJECI'ZVE: CCWS-3.0-K2
STATE how the CCWS responds during each ofthe following conditions:
Safety Injection signal
DEVELOPMENT REFERENCES: SD-145
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: SIGNIFICANTLY MODIFIED B DIRECT
cmmy MODIFIED DIRECT: ccws-ru002
NRC EXAM IIISTORY: None
DISTHACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
d a. On an SI signal, both the GPFD and sample panel receive isolation signals.
b. 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 ciose on a Phase B signal, but Phase 5 is not generated by
an SI signal.
d. Plaiisible since the pumps will get a start signal, but the thermal banier heat exchangers are only
isolated on a Phase 5 signal.
ICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPIANATION: Knowledge ofthe response of CCWS to an SI signai
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 39
Given the following conditions:
6 The plant is operating at 23% power.
e Steam pressure channel PT-475 is selected for control of SG A
a Stcam pressure transmitter PT-475 fails high.
Assuming NO operator action, which of the following statements describes the response
ofthe Steam Generator Water Level Control System (SGWLCS)?
a. An increase in steam flow from SG A is sensed and responds by increasing
IFW-140, MN FW A REG BYP FK-449.1, position to increase feed flow to SG
A and level increases
b. An increase in s t e m flow fmm SG A is sensed and responds by increasing
1FW-133, MAIN FW A REGULATOR FK-478, position to increase feed flow to
SG A and level increases
c. A decrease in steam flow from SG A is sensed and responds by decreasing IFW-
140, MN FW A REG BYP FK-479.1. position to decrease feed flow to SG A
and level decreases
d. A decrease in steam flow from SG A is sensed and responds by decreasing IFW-
133, MAIN FW A REGULATOR FK-478, position to decrease feed flow to SG
A and level decreases
ANSWER:
b. An increase in steam flow from SG A is sensed and responds by increasing
1FW-133, MAIN FW A REGULATOR FK-448, position to increase feed flow to
SG A and level increases
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMEiER: 39 TIEWGROUP !
26
1QCFR55CONTENT: 41(b) 4 430~1
KA: 059.44.08
Ability to manually operate and monitor in the control room: Feed regulating valvc controller
OBJECTIVE: SGWLC-3.0-2
Given the status of the various S G W K related control switch positions and controllers, PREDICT holv
a malfunction of the following will effect the SGWIC System:
0 S G pressure channels
DEVE1,OPMENT KEFERENCES: SD-126.02
REPEREWES SUPPLIED TO APPLICANT: None
QLJESTION SOURCE: NEW SIGNIHCAKTLY MODIFIED 0 DIRECT
BANK RUMBER FOR SIGNIFICANTLY MODIFIED / UIRECT: SGWLC-RZ 002
NRC EXAM HISTORE None
DISTRACTOR JUSTlFlCACTION (CORRECTANSWER dd):
a. Plausible since steam pressure failing high causes the steam flow to increase, resulting in SF > FF, but
the feed reg valve is in operation at this power level.
d b. Steam pressure failing high causes the steam flow to increase, resulting in SF > FF. The feed reg
valve. in operation at 15% power, opens to cause FF and level to increase.
E. Plausible since steam pressure failing causes the steam flow to changee,resulting in a SF - FF
mismatch, but the feed reg valw will open to increase FF.
d. Plausible since steam pressure failing causes the steam flow to cliange, resulting in a SF - FF
mismatch, but the feed reg valve will open to increase FF.
ICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXP1,ANATIQN: Analyze the effect of the failure on the control system and recognize which
valve will be controlling at the power level given
Post Validation Revision
Harris hXC Written Examination
Reactor Operator
QUESTION: 40
The plant is operating at 80% power with rod control in automatic and pressurizer
pressure at 2240 p i g .
After a rapid power reduction the plant is stabilized at 40% power, when the Reactor
Operator notes thc followinp conditions:
e Pressurizer pressure is 2275 psig and slowly decreasing.
e Pressurizer level is 45% and slowly decreasing.
a Both pressurizer spray valves indicate mid-position.
e All pressurizer backup heaters are de-energized.
These conditions are indicative of.. .
a. a normal plant response following an outsurge from the pressurizer.
h. 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.
ANSWER:
c. a failure in the Pressurizer Level control circuitry, which failed to energize the
hackup heaters.
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 40 TIEWGROUP: 2!2
10CFR55 CONTENT: 41(b) 3!7 43(b)
KA: 011K6.04
Knowledgc of the effect of a loss or malfunction on the following will have on the PZR I,CS: Operation
of PZR level controllers
OBTECTNE: PZKIC-3.0-5
EXPLAIN how the system controls pressurizer level, including the input parameters and the components
that receive output signals
DEVELOPMENT KEFERENCES: SD-I 00.3
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGKIPICANTLY MODIFIED / DIRECT: PZRLC-R7 00 1
NAC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
8. Plausible sirice the response is correct, with the exception of the pressurizer heaters not being
energized, fix an ontsurge from the pressurizer.
b. Plausible since a downpower shouid result in an insurge which would cause the spray valves to open,
but the heaters should also he energized.
d e. A rapid downpower transient will result in an insurge to the pressurizer. This should result in the
conditions noted, including a high pressurizer level causing the heaters to be energized even during a
high pressure condition causing the spray valves to be open. The heaters not being energirxd with
level more than 5% high is indicative o f a level control system failure.
d. Plausibic since the rcsponse is corrcct, with the exctlption o f :he pressurizer heaters not k i n g
energized, for an insurge to the pressurizer.
ICULTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS KYOWLEDGE /RECALL
DIFFICULTYRATING: 3
E.XPLANATIQN: Analysis of the expected plant response and the actual plant response to an
insurge into the pressurizer
Post Validation Revision
IIanis NRC Written Examination
Reactor Operator
Given the following conditions:
e Feed flow lo the s t e m &eneratorsis being transferred from the Auxiliary Feedwater
(AFW) System to the Main Feedwater (MFW) System in accordance with OP-134-1,
Feedwater System.
e The Motor-Driven AFW Pumps are operating with all Flow Control Valves throttled
in mid-position.
e The Turbine Driven AFW Pump is in standby with all Flow Control Valves full open.
e MFW Pump A is operating with the Feed Reg Bypass Valves throttled slightly
open.
AI1 AFW and MFW Isolation Valves are open.
If a condition occurs which results in a valid AFW Isolation Signal, how will the
following AFW and MFW valves on SG B respond?
T D A W FLOW MFW ISOLATION MFW FEED KEG
CONTROL VALVE VALVE BYPASS VALVE
a. Remain Open Remain Open Rmain Open
b. Remain Open Close Close
C. Close Remain Opcn Remain Open
d. Close Close CIOSC
ANSWER:
d. Close Close Close
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 41 TIEWGROUP: 212
10CFR55 CONTENT: 41(b) 4 43(W
KA: 035KI.01
Knowledge of the physical connections andor cause-effect relationships bdween the S/GS and the
following systems: MFWIAFW systems
OBJECTIVE: AFS-3.0-Ii2
IESCRIBE the controls and interlocks of AFW System valves and controllers
DEVEI,OPMENT REFERENCES: OP-137
SD-134
SI)-103
IPEFEIZENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW fl SIGNIFICANTLY MODIFIED
OR SIGNIRICAFiTLY MODIFIED / DIRECT:
NRC EXAM HISTORY: None
DISTRACTOW JUSTIPICACTION (CORRECT ANSWER +d):
a. Plausible since a AFW isolation signal closes the motor-operated isolation valves so these valves
could remain open, but they also close.
b. Plausible since the F W valves wili go closed, but the TDAFW Pump flow control valves also close.
e. Plausible since the TDAFW Pump flow control valves will go closed, but the FW valves slso close.
4 d. AFW isolation occurs on a Main Steam Isolation Signal, which will also close the F W valves,
concurrent with a high steam line differential pressure signal.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWL.EDGEI RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of the cause of AFW isolation signal, its effect on AFW, as well as
other signals generated which cause FWI
Post Validation Revision
Hanis NRC Written Examination
Reactor Operator
QUESTION: 42
Given the following conditions:
E A reactor trip with SI has occurred.
E The immediate action steps, ECCS flow verifications, and AFW flow verifications
are performed.
E SG levels are 10% and the required AF\V flow CANNOT be estabiished.
E FRP-H. 1, Response to Loss of Secondary Heat Sink, is entered.
m RCS pressure is checked and determined to he less than intact SG pressure.
Which of the folIowtrig describes the pImt conditions?
a. A large break LOCA is in progress AND a secondary heat sink is required
b. A large break LOCA is in progress AND a secondary hcdt sink is NOT required
c. A small break LOCA is in progress AND a secondary heat sink is required
d. A small break LOCA is in progress AND a secondary heat sink is NOT required
ANSWER:
b. A large break LOCA is in progress AND a secondary heat sink is NOT required
Post Validation Revisioii
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESFION NUMBER 42 TIEWGROUP: lil
KAIMPORTANCE: HO 3.9 SRO
10CFR55 CONTENT: 41(b) 7 43m
K k WE05EK2.2
Knowledge of the interrelations between the (Loss of Secondaly Heat Sink) m d the following: Facilitys
heat removal systems, including primary coolant, eniergency coolant, the decay heat renioval systenrs,
and relations between the proper operation of these systems tu the operation of the facility
OBJECTIVE: EOP-3.11
Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
d. Requirements for a heat sink
DEVELOPMENT REFERENCES: FKP-I. 1
LP-EOP-3. I 1
REFERENCES SUPPLIED TQ APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOP-3.11 023
NRC EXAM HISTORY: Harris NRC 2000
DISTRACTOR JUST%F%CACTION (CORRECT ANSWER dd):
a. Plausible since a large break LOCA has occurred, but a secondary heat sink is riot required.
d b. With KCS pressure less than SG pressure a large hreak LOCA has occurred and adequate heat
removal will occur from SVbreak flow.
c. Plausible since a I D C A has occurred, but the LOCA is a large break and a secondary heat sink is not
required.
d. Plausible since a secondary heat sink is not required, but the LOCA is a large break.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Comparison of heat removal systems and plant conditions to determine
requirements
Post Validation Revision
Harris hXC Written Examination
Reactor Operator
QUESTION: 43
Given the following conditions:
e The plant had been operating at 100% for three (3) weeks when a Reactor Trip
occurred.
o Six (8) hours following the trip, a reactor startup is planned.
Which one ofthe following is PROHIBITED at SHNFP as a result of industry wide
premature criticality events?
a. A startup rate in excess o f + 0.3 dpm
b. Delaying the startup until xenon begins to decay
c. Operators performing the EXSPACK estimated critical conditions (ECC)
d. A difference of 400 pcm hetween the POWERTRAX and EXSPACK ECCs
ANSMJEW:
d. A differenec of 400 pcm hetween the POWERTRAX and EXSPACK ECCs
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTIOX NUMBER: 43 TIERKKOUP: 3
10CFK55 CONTENT: 41(b) 10 43W
Kn: 2.2.1
Ability to perform prc-startup procedures for the facility, including operating those controls associated
with plant equipment that could affect reactivity
OBJECTIVE: GP-3.4-6
SWMAKIZE at least three conditions which have contributed to premature criticality events within the
industry; also SUR04ARIZE actions taken at SHNPP to prevent similar occurrences
DEVELOPMENT REFERENCES: GP-004
REFEWNCXS SUPPLIED TO APPLICANT: None
QGESTION SOURCE: SIGNIFICANTLY MODIFIED 0 DIRECT
CANTLY MODIFIED / DIRECT: ( 3 - 3 . 4 01 1
NKC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
8. Plausible since cxcessive startup rates can contribute to lack of reactivity control, but limitations are
placed on startup rate after criticality is achieved.
b. Plausible since xcnon decay will he adding positive reactivity to the core while the startup is being
perfomicd, but is accounted for in the time after trip in the ECC.
c. Plausible since SIMPP required any manual ECC calculations be perfomed by Reactor Engineering,
but EXSPACK is normally perfomed by Operations.
4 d. The threshold for perfomring a reactor startup following a power history of >RO% equilibrium power
is 250 pcm difference between P O W E R T M and EXSPACK and 500 pcm for transient history and
steady state below 80%.
ICIJLTY ANALYSIS:
COMPREHENSIVE / ANALYSL9 KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge of the administrative requirements prior to criticality being
achieved
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 44
While reviewing the MCB annunciators prior to relieving the off-going shift, YOU note
that an annunciator has a RED bar attached to it.
This indicates that the annunciator is in alarm due to ...
a. the aimn being defeated.
b. the associated system being tested.
c. the alarm window itself being inoperable with a Work Request to repair it written.
$. the associated system being under clearance.
ANSWER:
d. the associated system being under clearance.
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 44 TIEWGROUID: 3
KAIMPORTANCE: HO 3.3 sa0
iOCFR55 CONTENT: 41(b) 5 43(b)
KA: 2.4.31
Knowledge of annunciators alarms and indications, and use of the response instructions
OBJECTWE: PP-2.0-R3
DISCUSS the requirements in OMM-001/AP-002/hP-100 concerning the following:
k. MCI3 annunciators
DEVELOPMENT REFERENCES: OMM-00 1
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: 0N E W SIGXIFICANTLY MODIFIED c]DIRECT
BANK NUMBER FOR SIGNIFICANTLY MQDIFIED /DIRECT PP-3.0-R3 003
NRC EXAM HISTORY: None
DISTHACI'OR JUSTIFICACTKON (CORRECT ANSWER d'd):
a. Plausible since a color is used for coding this condition, but the abann being defeated is
indicated by black color coding.
b. Plausible since a color is used for coding this condition, but the associated system is being
tested is indicated by pink color coding.
E. Plausible since a color is used for coding this condition, the alarm window itself being
inoperable with a Work Request to repair it written is indicated by yellow color coding.
4 d. Red color coding indicates that the associated system is under ctearance
DIFFICULTY ANALYSIS:
0 COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFPICUI,TY RATING: 2
EXPLANATION: Knowledge of the color coding for alarm conditions
Post Validation Revision
IIarris NRC Written Examination
Reactor Operator
QUESTIQN: 4.5
Given the following conditions:
e A Reactor Trip occurred fiom 100% power.
e The plant stabilized at 5.57 OF for several minutes.
e Shortly thereafter, a Safety Injection signal actuated.
Which of the following describes the effect of this sequence on the Main Feedwater
System?
a. e A f t a the Reactor Trip occurred, the SGs could be fed using the Feedwater Reg
Bypass Valves
e After the SI occurred, the SGs could be fed using the Feedwater Reg Bypass
Valves
b. e After the Reactor Trip occurred, the SGs could be fed using the Main
Feedwater Reg Valves or the Feedwater Reg Bypass Valves
After the SI occurred, Main Feedwater could NOT be used to feed the SGs
c. e After the Reactor Trip occurred. the SGs could he fed using the Feedwater Reg
Bypass Valves
e After the SI occurred, Main Feedwater could NOT be used to feed the SGs
d. e After the Reactor Trip occurrd, the SGs could he fed using the Main
Feedwater Reg Valves or the Feedwater Reg Bypass Valves
m After the SI occurred, the SGs could he fed using the Feedwater Reg Bypass
Valves
ANSWER:
c. e After the Reactor Trip occurred, the SGs could he fed using the Feedwater Reg
Bypass Valves
e After the SI occurred, Main Feedwater could NOT be used to feed the §Cis
Past Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 45 TIEWGHOUP: 2i I
1QCFR55CONTENT: 41(b) 4 43@)
KA: 059K4.19
Knowledge of MPW design feature($)and/or interlock(s) which provide for the following.:Automatic
ORECCTIVE: AFW-3.0-A6
EXPLAIN the response of major CFW System valves to the following signalskonditions
hPain Peedwater Isolation Sigrxal (MFES)
Reactor trip (P-4) coincident with low Tavg(< 564OF)
DEVELOPMENT REFERENCES: SD-153
REFERENCES SUPPLIED TO APPLICANT: None
QKESTION SOURCE: NEW' c]SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR s m w I c A l r l T r , Y MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER +d):
a. Plausible since on a reactor trip with low Tave (564 "F), the SGs can still be fed with the bypass
valves, but on an SI or high-high SG level MEW can no longer supply the SGs.
b. Plausible since the SGs can no longer be fed using MFW on an SL but on a reactor trip only the
bypass valves can he used to feed the SGs.
4 e. On a reactor trip with low Tave (564 OF), the SGs can still he fed with the bypass valves, but on an SI
or high-high SG level MFW can no longer supply the Scis.
d. 1'PausibIe since on a reactor trip with low 'l'ave (564 "F). the SGs can still be fed with the bypass
valves, hut not the main feed reg valves, and on an SI or high-high SG level MFW can no longer
supply the SGs.
DIPFICLrLI'Y AKALYSIS:
COiMPRHIENSIVE / ANALYSIS c]KNOWLEDGE / RECALL
DIFFICULTY KATING: 3
EXPLANATION: Comprehension that on a reactor trip where the plant stabilizes at no-load
temperature, the P-4 with Low 'rave signal allows feeding with the hypass and
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 46
Which of the following describes the design of Phase A and a Phase B Containment
Isolation signals?
a. a Phase A O m limits radioactive releases following a LOCA
a YRase B O x limits radioactive releases following a LOCA or secondary
system break inside Containment
h. a Phase A limits radioactive releases A X minimizes Containanent
overpressurization following a LOCA
a Phase 3 limits radioactive releases Amminimizes Contitinment
overpressurization following a LOCA or secondary system break inside
Containmcnt
c. a Phase A O m l i m i t s radioactive releases following a LOCA
a Phase B limits radioactive releases following a LOCA A x p r e v e n t s an
excessive RCS cooldown following a secondary system break inside
Containment
d. 8 Phase A limits radioactive releases Amminimizes Containment
overpressurization following a LOCA
a Phase J3 limits radioactive releases following a LOCA Amprevents an
excessive RCS cooidown following a secondary system break inside
Containment
ANSWER:
a. a Phase A ONI,Ylimits radioactive releases following a LOCA
a Phase B O x l i m i t s radioactive releases following a LOCA or secondary
system break inside Containment
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER. 46 TIENGROUP: 1/1
10CFR55 CONTENT: 41(b) 9 43@)
Kk 00001 1EK3.00
Knowledge of the reasons for the following responses as the apply to the Large Break LOCA: Actuation
of Phase A and B during LOCA initiation
OBJECTIVE: CIS-3.0-1
STATE the purpose ofthe Containment Isolation System
DEVELOPMENT REFERENCES: SI)-1 14
REFERENCES SZJPPLIEDTO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: CIS 006
CIS 009
NRC EXAM HISTORY None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER d'd):
4 a. Phase A serves to limit the release of radioactive materials to atmosphere following a LOCA. Phase 13
acts to limit radioactive releases by actuating on a kOCA or a steam or fecdwater line break inside
containment.
b. PlausibIe since both Phase A and Phase B act to limit the release of radioactive 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 by spray actuation, main steam
line isolation, and feed water isolation.
d. Plausible since both Phase A and Phase H act to limit the release of radioactive materials to
atmosphere, but ovepressurimtion and RCS cooldowns are limited by spray actuation, main steam
line isolation, and feed water isolation.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge ofpurpose of Phase A and Phase B signals
Post Validation Revision
Harris NRC Written Examinaticn
Reactor Operator
QUESTION: 47
An entry into FRP-S.1, Response to Nuclear Power GeneratiodATWS, has been made
from PATH- 1. The following conditions currently exist:
e The reactor trip breakers are closed.
e Rods are being inserted manually.
e Control Bank D is at 12 steps.
e Power Range Instruments are all indicating 8%.
e Intermediate Range SUR is NEGATIVE
Which of the following conditions must be met in FRP-S.1 allow a return to PATH-I?
a. One of the reactor trip breakers must be opened
b. Both ofthe reactor trip breakers must be opened
c. Power Range indication must be reduced below 5%
d. Control Bank A must be inserted fully
ANSw ER :
c. Power Range indication must be reduced below 5%
Post Validation Revision
Ilarris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NCMBER: 47 TIEWGROUP 1/1
KAIMPORTANCE: RO 4.4 SRB
IOCFR55 GOhTENT: 4l(h) 7/10 43(b)
KA: 000029EA2.0 1
Ability to determine or interpret the following as they apply to a ATWS: Reactor nuclear instrumentation
OIBJECTWE: EOP-3.1-3
DEMONSTRATE the below-assumed operator knowledge from the SHhTP Step Deviation Documents
and WOG ERGS that support performance of EQP actions:
a. Verification of reactor trip
DEVELOPMENT REFERENCES: EOP-FW-S. 1
REFERENCES SUPPLIED TO APPLICANT: None
QhiE.STION SOURCE: REW SIGNIFICANTLY MODIFiED DIRECT
BANK hWMBER FOR SIGNIFl[CANTI,Y MODIFIED f DIRECT: EOP-3.15-135002
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (COBRECT ANSWER .\l'd):
a. Plausible since this would cause the reactor to be tripped, but it is not required to be done to exit PRP-
s.1.
b. Plausible since this would cause the reactor to be tripped, but it is not required to be done to exit FRP-
s.1.
d c. Exiting FW-S.1 requires that PR N E be less than 5% and IR NIS startup rate be negative. Reactor
trip breaker position is not a condition for exiting the procedure, although actions are taken to open the
breakers.
d. Plausible since this would cause the reactor to be adequately shutdown, but it is not required to be
done to exit FRP-S.I.
D w F I c m x Y ANALYSIS:
0 COMPREHENSIVE / ANALYSIS ICUOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowled&eof the procedural requirements to exit FRP-S. 1
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 48
Given the following conditions:
m A plant cooldown is being performed.
e All Steam Generators (S6s) are currently at approximately 50 psig.
Auxiliary Feed Water (AFW) Pump A-SA is being used to feed the SGs.
e The supply breaker on 120 VAC IDP-IA-SI for 1AF-i9, AUX FW MOTOR PMP
A-SA DISCHARGE \'&V, trips open.
Which of the following describes the effect of this loss of power on the operation of
AFW Pump A-SA?
3. Operates at shutoffhead
b. Operates on minimum recirculation flow
c. Operates on maximum recirculation flow
d. Operates at runout conditions
ANSWER:
d. Operates at runout conditions
Post Validation Revision
Harris NRC Written Examination
Reactor Opentor
Data Sheets
QUESTION NUMBER: 48 TIEWGROUP: 2/ 1
MAIMPORTANCE: RO 2.5 SRQ
10CFR55 CONTENT: 41(b) 4 4309
Kk 061K6.01
Knowledge of the effect of a loss or malfunction of the following will have on the AFW con~ponents:
Controliers and positioners
OBJECTIVE: AFS-3.OR5
DFXKIBE how the AFW system is impacted by a loss of l2Ovac unintermptible power supplies (SI, S I ,
SIII, SIV)
DEVELOPMENT REFERENCES: SD-137
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOUIPCE: NEW SIGNIFICANTLY k1ODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: AFS-A3 001
AFS-A3 007
NRC EXAM HISTORY: None
DISTRACTOR m w I F I c A 6 : T I o N (CORRECT ANSWER +@:
a. Plausible since p o w a is lost to the discharge valve, but the vr~lvcfails open causing flow to increase.
b. Plausible since power is lost to the discharge valve, but the valve fails open causing flow to increase.
e. Plausible since the valve fails open and flow increases, but the pump does not mn on recirculation
flow.
./ (8. The loss of power causes AFW Pump A-SA to reach rnnout conditions due to IAF-19 failing open
and having the SGs at such a low prcssure.
DIFFICULTY ANALYSIS:
COMPREHENSRE / ANALYSIS 0 KNOWLEDGE /RECALL
DIr,-FICUI,TY RATING 3
EXPIANATION: Analysis of the effect o f a failure of the PCV after determining the fail position
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 49
Given the following conditions:
The plant is in Mode 5.
a ALB-008-1-4, RWMU STORAGE TANK MINIMUhUHIGH LEVEL, a l m s .
m WWMU tank level is decreasing with NO VCT makeup in progress.
Which one of the following procedures would be the most appropriate to implement?
a. AOP-003, Maifunction of Reactor Makeup Control
b. AOP-08, Accidental Reiease of Liquid Waste
c. AOP-016, Excessive Primary Plant Leakage
d. AOP-020, Loss of Reactor Coolant Inventory / RHR While Shutdown
ANSWER:
b. AOP-008, Accidental Release of Liquid Waste
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 49 TIEWGROUP: 1/2
1QCFR55CONTENT: 41(b) 10/13 43(b)
KA: 00005962.4.4
Ability to recognize abnormal indications for system operating parameters which are entry-level
conditions for emergency and abnormal operating procedures. (Accidental Liquid Radwaste Release)
OBJECTWE: AOP-3.8
1I)ENTIFY symptoms that require entry into AOP-008, Accidental Release of Liquid Waste
DEVELOPMENT REFERENCES: AOP-008 (unknown)
QUESTION SOURCE:
-
REFERENCES SUPPLIED TO APPLICANT
u NEW u
I--.I
None
SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SPGNIFI'ICANT1,YMODIFIED i DIRECT: AOP-3.8 001
NRC EXAM HISTORY: Hams NKC 2000
DISTRACTOR SUSTIFICACTION (CORRECT ANSWE.R d'd):
8 . Plausible since KMUW tank supplies makeup to VCT, but AOP-003 addresses conditions regarding
valve / transmitter failures, not loss oftank source.
d b. Entry conditions have been met for AOP-008.
e. Plausible since F&lIJW tank supplies makeup to RCS and candidate may imply that loss of supply
results in a loss ofprimary inventow, but conditions are met for entry into AOP-008.
d. Plausible since RMUW tank supplies makeup to RCS and candidate may imply that loss of supply
results in a loss of primary inventory with plant shutdown, but conditions are met for entry into AQP-
008.
DIFFICULTY ANALYSIS:
COMPREHENSIVE i ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of entry requirements ftx accidental liquid release
Post Validation Revision
Hamis NRC Written Examination
Reactor Operator
QUESTION: 50
Which of the following actions would be most effective in responding to a Pressurized
Thermal Shock condition in accordance with EOP-FRP-P. 1, Response to Pressurized
Thennal Shock?
a. Close the block valve for any open PRZ PORV
b. Start a RCP once SI has been terminated
c. Direct an operator to locally isolate any stuck open SG PORV
d. Direct an operator to locally open any failed closed BIT outlet valve
ANSWER:
c. Direct an operator to locally isolate any stuck open SG PQRV
Post Validation Revision
Harris NR(: Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 50 TIEFUGROUP I12
10CFR55 CONTENT: 4l@) 10 43W
- A: WE08G2.1.30
Ability to locate and operate components, including locai controls. (Pressurized Thermal Shock)
OBJECTWE: EOP-3.14-1
DESCRIBE the purpose ofthe following EOPs including the type of event for which they were designed
and the major actions performed
a FRP-P. 1, Response to Imminent Pressurized Thermal Shock
- -
DEVELOPMENT REFERENCES: EOP-FRP-P.I
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED u DJHECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC E?XV HISTORY: None
DISTRACTOR JUSTIFICACTION ( C Q W C T ANSWER d'd):
a. Plausible since closing the block valve for a stuck open PW, PORV is an action taken in FW-S.1,
though it is performed to maintain RCS inventory and will cause pressure to increase which would
cause the severity of a PTS event to worsen.
b. Plausible since an RCP is started in VRP-S.1 to cause mixing of any SI water with the RCS, but only if
SI cannot be terminated.
d c. 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 the BIT out6et valves are supposed to bz opened during an SI condition, but the I3IT is
isolated in FRP-S.1 and local action is taken to close the vaive in the event it cannot be closed
remotely.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of plant conditions during a PTS event to determine the most
appropriate course of action
Post Validation Revision
Hams NFK Written Examination
Reactor Operator
QUESTION: 51
Given the following conditions:
e An operator has been sent to rack out a 480 VAC breaker by the SCO.
e Inadvertently, the incorrect cubicle is opened and the control power fuses are
removed from the wrong breaker.
Which of the following describes how the breaker is affected by the removal of the
control power fuses?
a. All Main Control Board indications will he lost for the breaker and if the breaker
is closed, it will trip and CANNOT be closed until control power is restored
b. All Main Control Board indications will be lost for the breaker and if the breaker
is open>it can only be closed mechanically locally
c. Main Control Board indication will still be available for the breaker, but ifthe
breaker is closed, it will trip and CANNOT be closed until control power is
restored
d. Main Control Board indication will still be available for the breaker, but if the
breaker is open, it can only be closed mechanically locally
ANSWER:
h. All Main Control Board indications will be lost for the breaker ,and ifthe breaker
is open, it can only be closed rncchanically lociilly
Post Validation Revision
Hairis NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NIMBER: 5 1 TIEWGROIOP: 2; I
IOCFR55 CONTENT: 41(b) 7 43(W
KA: 062A4.04
Ability to manually operate and/or monitor in the controi room: 1,ocal operation of breakers
OBJECTIVE: 480V-3.O-RI
State the function of breaker control power and discuss the effects o f a loss of breaker control power
DEVELOPMENT REFERENCES: OP-156.02
REFERENCES SUPP1,IED TO APPLICANT: None
QUESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: 48OV-RI 001
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
a. Plausible since MCB indication will be lost, but the breaker will not trip open on the loss of control
power.
4 h. A loss of control power will cause MCB indication to go out and cause the remote operation ofthe
breaker to be defeated.
e. Plausible since a loss of control power to causes a loss of the ability to operate the breaker, but the
breaker will not trip and MCB indication will he lost.
d. Plausible since the breaker can only be operated locally, but the loss of control power will result in a
loss of MCB indication.
DIFFICWLTY ANALYSIS:
0 COMPREHENSIVE / ANALYSES KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowlcdge ofthe effect of a 108s of control power to a 48OV breaker
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 52
Which ofthe following situations would likely result in an inadvertent dilution event
during Mode I 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. = RCP thermal barrier heat exchanger leak
= AOP-OI 6, Excessive Primary Plant Leakage
b. o The boric acid pump trips during an automatic makeup
AOP-004, Malfunction of Reactor Makeup
e. o A mixed bed demineralizer that was last in service three weeks ago is
mistakenly placed in service at the end-of-cycle
= AOP-033, Cheniistry Out of Tolerance
d. o A tube leak in the Seal Water heat exchanger
m AOP-014, Loss of Component Cooling Water
ANSWER:
d. = A tube leak in the Seal Water heat exchanger
0 AOP-014, Loss of Component Cooling Water
Post Validation Kevision
Harris NRC: Written Examination
Reactor Operator
Data Sheets
QUESTION NGMBER 52 TIEWGROUP 216
10CFR55 CONTENT: 41(b) 6/10 43(b)
KA: 004A2.06
Ability to ( a ) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based
on those predictions, use procedures to correct, control, or mitigate the consequences of those
malfunctions or operations: Inadvertent boratioddilution
OBJECTIVE: E-3.12-3
Identify systems whose operation may alter RCS boron concentration and discuss how operation of these
systems may affect boron concentration
DE\EI;OPMEWT REFERENCES: SOER 94-2
AOP-0 14
REFERENCES SUPPLIED TO APPLPCANT: None
QUESTION SOURCE: SIGNIFICANTLY MODIFIED DIRECT
CANTLY MODIFIED / DIRECT: IE-3.12-R3001
NRC E.XAI\I HISTORY: None
DISTUCTOR JUSTIFICACTION (CORRECT ANSWER $d):
a. Plausible since the thermal barrier interfaces with a non-borated system (CCW), but leahge would be
out of the RCS to CCW and would not affect RCS boron concentration.
b. Plausible since boric acid is required for the proper blended flow, but an automatic makeup would be
terminated automatically in the event of a boric acid pump trip.
c. Piausible since boron concentration will change in CVCS, but this would result in an inadvertent
bordtion rather than a dilution.
d d. A seal water HX leak will result in CVCS being diluted by CCW. This failure is to be addressed by
DIFFICULTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS 0 KNOW1,EDGE /RECALL
DIFFICULTY RATING: 3
EXP1,ANATION: Analyze the effect of each failure on KCS boron concentration and determine
the required procedure to address the failure
Post Validation Revision
IIturis NRC Written Exaniination
Reactor Operator
QUESTION: 53
While establishing a bubble in the PRZ per GP-002, Normal Plant Heatup From Cold
Solid to Hot Subcritical MODE 5 to MODE 3, letdown pressure control valve ICs-38
(PK-145.1), Low Pressure Letdown Pressure Controller, opens.
Which of the following describes why PK-145.1 opens?
a. Thermal expansion of liquid in the pressurizer
b. Change in CCW heat load
c. Spray valves are shut while drawing a bubble
d. Switchover of letdown to orifices from RHR-CVCS cross-connect
ANSWER:
a. Thermal expansion of liquid in the pressurizer
Post Validation Revision
Harris NRC Written Exaniiriation
Reactor Operator
Data Sheets
QUESTION NUMBER: 53 TIEWGROUQ: 2: 1
KAIMPORTANCE: HO 2.9 SRO
10CFR55 CONTENT: 41(b) 7 43W
KA: 010K1.06
Knowiedge of the physical connections andor cause-effect relationships between the PZR PCS and the
following systems: CVCS
OBJECTIVE: CiP-3.2-2
DISCUSS drawing a bubble in the pressurizer, including
b. The parameters used to determine when the bubble has bee11 drawn
DEVELOPMENT REFERENCES: GP-002
1.P-(3-3.2
REFERENCES SUPPLIED TO APPLICANT
~~ None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED mmcr
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EWQ 1081 1
NHC EXAM HISTORY: Hams NWC 2002
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
d a. Thermal expansion of the liquid doe to the heaters being energized results in a pressure increase in the
RCS. PK-145.l opens to maintain letdown pressure, resulting in increased letdown flow.
b. Pballsible since the letdown heat exchanger is cooled by CCW, but temperature has little effect on the
response ofPK-145.1.
e. lkdusible since the spray valves are shut while a bubble is being drawn, but PK-145.1opens to
maintain letdown pressure, not RCS prcssure.
d. Plausible since KHR letdown may be placed in service at low temperature and pressure conditions, but
is not in service while drawing a bubble.
DIFFICCJLTY ANALYSIS:
~~
COI\IPHEIIENSNE / AKALYSIS KNOWLEDGE /RECALL
DIFFICULTY MATING: 3
EXPLANAFION: Comprehension ofthe eEects of drawing a bubble on CVCS components
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 54
125 VDC battery 1A-SA is rated for 1170 amp-hours at a 4-hour discharge rate.
I f DC load shedding is performed such that the loading on the battery is reduced from
292 amps to 146 amps, how long should the battery be available to supply the remaining
loads?
b. More than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, hut 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 />
Post Validation Revision
Hamis NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 54 TIEWGROUP: 2/1
lOCFR§§ CONrENT: 41(b) 8 43w
Kik 063A1.01
Ability to predict and/or monitor changes in parameters associated with operating the 1 X electrical
systeni controls including: Battery capacity as it is affected by discharge rate
OBJECTIVE: DCP-3.0-A3
STATE the fiinction and EXPLAIN the basic operation of the following major components of the DC
Power System:
e Batteries
DEVELOPMENT REFERENCES: EOP-EPP-00 1
ADEL-LP-2.6
DCP-LP-3.0'
MEFERENCES SUPPLIED TO AF'PLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED 0 DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
BHSTHACTOR JGSTIFICACTION (CORRECT ANSWER d'd):
8. Plausible since the battery is rated for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, but at a discharge rate of approximately 293 amps per
hour and decreasing the discharge rate would increase the capacity.
h. Plausihle since the discharge rate has been decreased which would extend the capacity of the battery
for a period of time, but the time would be more than doubled.
C. Plausible since the discharge rate has been halved, so it would appear that the capacity would be
doubled, but it is a n o n - h e x relationship.
d d. Reducing the discharge rate on a battery increases the battery capacity in a non-linear function such
that decreasing the discharge rate by half, increases the capacity by more than double.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KVOWLEDGE /RECALL
m F m x z r y RATING: 4
EXPLANATION: Calculation of the nominal discharge rate of a battery and comprehension of the
effect of reducing discharge rate on battery capacity
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 55
Given the following conditions:
0 The plant has experienced a -rrge Break Loss of Coolant Accident during a reactor
startup.
All equipment functioned as designed and the crew has reached the point in PATH-1
where monitoring Critical Safety Function Status Trees is required.
Which one of the following statements describes the IMMEDIATE result that voiding in
the downcomber region would have on the Source Range instrumentation and procedure
used to mitigate these plant conditions?
a. 0 The displacement of downcomber water would increasc the neutron leakage
and result in a higher source range counl rate.
m The crew should continue in PATH-1 rather than transition to EOP-FRP-S.2,
Resporise to Loss of Core Shutdown.
b. o A decrease in downcomber water density would reduce fission and result in a
lower source range count rate.
0 The crew should transition to EOP-FRP-S.2, Response to Loss of Core
Shutdown, rather than continue in PATH- I .
c. 0 The displacement of boron from the downcomber region would increase
fission and result in a higher source range count rate.
0 The crew should continue in PATI-1-1 rather than transition to EOP-FRP-S.2,
Response to Loss of Core Shutdown.
d. 0 A decrease in downcomber water density would reduce fission and result in a
lower source range count rate.
0 The crew should continue in PATH-1 rather than transition to EOP-FRP-S.2,
Response to Loss of Core Shutdown.
ANSWER:
a. The displacement of water would increase the neutron leakage and result in a
higher sonrce range count rate.
o The crew should continue in PATH-1 rather than transition to EOP-FRP-S.2,
Response to Loss of Core Shutdown.
Post Validation Revision
Hams NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NCJMBER: 55 TIEWGROUP 2l2
ICCFRSS CONTENT: 41(h) 2 43(b)
MA: 015A2.05
Ability to (a) predict the impacts of the following malfunctions or operations on the NE; and (b based on
those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions
or operations: Core void formation
OBJECTIVE: BD-3.10-7
Explain the NIS response to different void fractions in the core and downcomer region
DEVELOPMENT REFERENCES: IIO-BD-3.10
REFERENCES SUPPLIED TO APPLICANR None
QUESTION SOURCE: 17 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGXWICANTLY MODIFIED / DIRECT: HNPO 20608
NRC EXAM HISTORY: None
DISTRACTOH JUSTIFICACTION (CORRECT ANSWER d'd):
4 a. Downcornber voiding results in higher source range indication due to increased leakage. The crew
should continue in PATH-I rather than transfer to FRP-S.2 since entry conditions to F a - S . 2 are a
Yellow path condition.
b. Plausible since a severe decrease in core water density would result in less moderation and a lower
power Icvel, but downcornher density has little effect on core reactivity.
E. 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 severe decrease in core water density would result in less moderation and a lower
power level, but downcomber density has little effect on core reactivity.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE i RECALL
DIFFICULTY MATING: 3
EXPLANATION: Analysis ofthe effects of core voiding on SK indication and knowledge of the
procedure hierarchy during the pcrfonnance of the EOPs
Post Validation Revision
Hamis NRC Written Examination
Reactor Operdtor
QUESTION: 56
Given the following conditions:
e A transition has just been made to FRP-S. 1, Response to Nuclear Power Generation
/ATWS, from PATH-I.
e The Reactor Operator is manually inserting control rods.
e Ail Turbine Throttle Valve (TV) and Turbine Governor Valve (GV) indications show
the RED light OFF and the GREEK light ON, with the exception of TV-3 and (37-2
which have both the RED fight and GREEN light ON.
o Turbine speed is decreasing, and is currently 1680 rpm.
e The Main Steam Isolation Valve (MSIV) Bypass valves are closed.
Which of the following actions should be taken next?
a. Verify all AFW pumps running
b. Manually trip the Turbine from the MCB
c. Pface both Turbine DEH pumps in PULL-TO-LOCK
d. Shut all MSiVs
ANSWER
b. Manually trip the Turbine from the MCR
Post Validation Revision
Hams NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NBIILIBER: 56 TIER/GROt.JP 22
KAIMPORTZTANCE: RO 2.8 SHO
10CFR55 CONTENT: 4l(b) 10 4309
KA: 045A4.06
Ability to manually operate andor monitor in the control room: Turbine stop valves
OBJECTIVE: FOP-3.15-4
Given the following EOP steps, notes, and cautions, DESCRIBE the associated basis
e Order of preference for turbine trip steps from the MCR
DEVELOPMENT REFERENCES: EOP-PKP-S. I
REFERENCES SKJFPLKED TO APPLICANT None
QCESTION SOURCE: NEW SIGNHPICANTLY MODIFIED DIRECT
BANK hWMBER FOR SIGNWICANTLY MODIFIED / DIRECT EOP-3.15-RZ 001
NRC EXAM HISTORY: None
DISTRACTOR SLTSTIFPCACTION (CORRECT ANSWER dd):
a. Plausible since GV-2 and TV-3 are associated with opposite steam chests and it may be assumed that
as long as the GVs are closed for 1 steam chest a i d the TVs are closed for the other steam chest with
turbine speed decreasing, and starting AFW is the next step in the procedure, however the turbine
should not be considered to be tripped.
4 b. Verification of a turhine trip requires either all 4 TVs be closed or all 4 GVs be closed. If one set of
these valves are not all closed, then the RNO directs manually tripping the turbine from the MCB.
c. Plausible since the turhine should not be considered to he tripped based on indications, and this is an
RNO action, bnt 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
RNO action, but should not be perfnnned until a manual trip from the MCB is attempted.
DIFFICULTY ANALYSIS:
0 COMPREHENSIVE / AP4ALYSIS 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
Post Validation Revision
IIarris NTC Written Examination
Reactor Operator
QUESTION: 57
Given the following conditions:
e The Main Control Room has been evacuated and control transferred to the Auxiliary
Control Panel (ACP).
AOP-004, Remote Shutdown, is being perfonned when a loss of offsite power
coincident with a Safety Injection signal occur.
Which of the following describes the response of the plant?
a. The Emergency Diesel Generators automatically start and the sequencers load the
ED@ due to the undervoltage signal
b. The Emergency Diesel Generators automatically start and the sequencers load the
EDGs due to the safety injection signal
e. The Emergency Diesel Generators automatically start, but must he manually
loaded with the required loads
d. The Emergency Diesel Generators must be manually started and manually loaded
with the required loads
ANSWER:
a. The Emergency Diesel Generators automatically start and the sequencers lndd the
EDGs due to the ~ n d e r v o h g esignal
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 57 TIENGROUP 211
10CFR55 CONTENT: 41(b) 8 Wb)
KA: 064A3.07
Ability to monitor automatic operation of the EDiG system, including: Load sequencing
OBIECTTVE: AOP-3.4-R5
DISCUSS how a transfer to the auxiliary control panel would affect the following inputs to the ESF
sequencers
- Safety injection signal
Safety bus undervoltage signal
DEVEIBPMENT REFERENCES: AOP-004
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY I\.ZQDIIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: AOP-3.4-R6 001
NRC EXAM HISTORY: None
DISTEPnC'I'QR JUSTIFICACTION (CORREC'I' Ah'SWER d'd):
4 a. The EDGs should automatically start OR 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 Ut' signal.
c. Plausible since the EDG will automatically start, but loading will he based upon the LW signal.
d. l'lausihle since many automatic functions are defeated when control is transferred to the ACP. but the
EDG will automatically start and loading will be based upon the UV signal.
DIFFICULTY ANALYSIS:
DIFFICULTY RATING: 3
EXPLANATIQN: Analysis ofthe effect o f a transfer to the ACP on the EDG and sequencer
operation
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 58
Given the following conditions:
An I&C technician reports that both of the Control Room Normal Outside Air Intake
Isolation radiation monitors bave failed detectors.
e It will take somewhere between four (4) and eight (8) hours to replace the detectors.
m i c h of the following states the action which must be taken within one (1) hour, in
accordance with Technical Specification 3.3.3. l ?
a. Establish operation of the Control Room Emergency Filtration System in the
Recirculation Mode of Operation
b. Initiate the preplanncd altcmate method of radiation monitoring
C. Return the monitors to service, or be in Hot Standby within the next six ( 6 )hours
d. Pcrforni a surveillance test on the Control Room Emergency Filtration System, or
be in Hot Standby within the next six ( 6 )hours
ANSWER
a. Establish operation of the Control Room Emergency Filtration System in the
KecircuIation Mode of Operation
Post Validation Revision
Harris NRC W-&en Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 58 TIEWGROUP: 3
1 0 ~ ~ COXTENT:
~ 5 5 4i(b) ioiii 43(b)
KA: 2.2.24
Ability to analyze the affect ofmaintenance activities on LCO status
OBJECTIVE: KMS-11
DEMONSTRATE knowledge of the Technical Specifications associated with the Radiation Monitoring
System:
a. RECOGNIZE the LCO limits associated with action statements of one hour or less
DEVELOPMENI REFERENCES: TS Table 3.3-6
REFERENCES SUPPLIED TO APPLICAIVT: None
QUESTION SOURCE: NEW c]SIGNIFICANTIiY MOI9IFIED DIRECT
BANK NUMBER FOH SIGNIFICANTLY MODIFIED / DIRECT: INIO I R442
NRC EXAM HISTORY: Hanis NRC 2002
1)KSTRACTOR JlJSTIFICACTION (CORRECT ANSWER dd):
4 a. With no outside air intakes available, maintain operation of the Control Room Emergency Filtration
System in the RecircuIaticn Mode of Operation.
b. Plausible since this would permit monitoring of the Control Room environment, but is not directed
c. Plausible since this is a typical action requirement in Tech Specs for inoperable equipment, but does
not apply to this condition.
d. Plausible since this is a typical action requirement time limit in Tech Specs for inoperable equipment,
but does not apply to this condition.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 4
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
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 59
Given the following conditions:
8 A LOCA has occurred inside Containment.
8 Containment pressure is 5.5 psig.
8 RCS Wide Range Pressure indications are:
(BLACK BEZELED INSTRUMENTS)
PI440 = 1060 psig
PI-441 = 1040 p ~ i g
(YELLOW BEZELED INSTRUMENTS)
PI-402 = 980 psig
PI-403 = 980 psig
PI-402A 700 psig
RCS pressure should be reported as ..,
a. 700 psig.
b. 980psig.
C. 1040 psig.
d. 10GOpsig.
ANSWER:
b. 980psig.
Post Validation Revision
Ilarris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 59 TIEWGROUP 3
10CFR55 CONTENT: 41(b) 6 43W
KA: 2.4.3
Ability to identi@ post-accident instrumentation
OWECTn7E: EOP-3.19
DESCRIBE Control Room usage of EPPs, foldouts, and FWs as it relates to the following:
g. Use of RCS wide-range pressure indication
DE\EI,BPBIENT REFERENCES: EOP Users Guide
IlEFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICAhTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOP-3.19 034
NRC EXAM HISTORY: Harris NRC 2005
DISTRACTOR JIIJSTIFICACTION(CORRECT ANSWER dd):
a. Plausible since yellow bezeled instniments are qualified for post-accident monitoring. The lowest
qualitid instrunlent following an accident should be used unless the narrow range instrument PI-
402A is on scale with KCS pressure helow 700 psig.
4 h. Yellow bezeled instruments are qualified for post-accident monitoring. The lowest qualified
instrument following an accident should be used unless the narrow range instrument PI-402A is on
scale with RCS pressure below 700 p i g .
c. Plausible since this is the lowest black bezeled instrument, but yellow bezeled instruments should be
used due to post-accident conditions.
d. Plausible since this is the highest black bezeled instrument, but yellow bezeied instruments should be
used due to post-accident conditions.
DIFFICCZTY tAL\SIs:
COMPREHENSIVE /ANALYSIS 0 KNOWLEDGE / RECALL
DIFFICULTY HATING: 3
EXPLANATION: Analysis of plant conditions to determine required actions during adverse
containment
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 60
Which of the following is the MINIMUM required operable components to meet
Technical Specification 3.8.2.1, DC Sources - Operating, Limiting Condition for
Operation?
a. 0 IA-SA 125-V emergency batterybank
0 1B-SB 125-V emcrgency battery bank
b. 0 1A-SA 125-V emergency battery bank
1B-SB 125-V emcrgency batterybank
1A-SA 125-V full capacity charger
c. IA-SA 125-V emergency batterybank
1R-SB 125-V emergency battery bank
e 1A-SB 125-V full capacity charger
e 1B-SA L25-V full capacity charger
d. c IA-SA 125-V emergency battery bank
1B-SB 125-V emergency battery bank
IA-SA 125-V full capacity charger
- 1A-SB 125-V full capacity charger
B IB-SA 125-V full capacity charger
a IB-SB 125-V full capacity charger
ANSWER:
c. e iA-SA 125-V emergency battery bank
e i B-SB 125-V emergency battcry bank
1A-SB 125-V full capacity charger
0 1B-SA 125-V full capacity charger
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 60 TIEWGROUP: 211
IOCFR55 CONTENT: 41(b) 10 43(W
KA: 00005862.1.33
Ability to recognize indications for system operating parameters which are entry-level conditions for
tcciuiical specifications. (Loss of DC Power)
OBJECTIVE: DCP-3.0-RI
Given the name of a component in the DC power system, state whether or not that component is
Technical Specification related
DEVE1,OPMENT REFERENCES: TS 3.8.2.1
REPERENCFS SUPPLIED TO APPLICANT: None
QUESTIQN SOURCE: 0 NE.W SIGNIFICANTLY MODIFIED DIRECT
BANK NLJIRERFOR s ~ ~ x w ~ C m r ~ Y / DIRECT:
MODIFIED DC-KI no1
NRC EXAM HISTORY: None
DISTRAC'FOR JUSTIFICACTION (CORRECT ANSWER d'd):
a. Plausible since both battery banks are listed, but each train also requires a full capacity charger.
b. Plausible since both battery hanks and a single full capacity charger are listed, hut each train also
requires its own full capacity charger.
./ E. Minimum requirements are 125 VDC Emergency Battery Bank IA-SA and either fir11 capacity
charger IA-SA or IH-SA AND 125 VDC Emergency Battery Bank ID-SB and either full capacity
charger IA-SB or 1U-SB.
d. Plausible since this would meet the requirements of the 'IS,but this includes all components and is not
the minimum required.
ICULTY ANALYSIS:
COMPREHESSIVE i mALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 2
EXPL.NATION: Knowledge ofthe TS requirements for Dc power
Past Validation Revision
Harris MKC Written Examination
Reactor Operator
QUESTION: 61
Given the following conditions:
e The plant is operating at 100% power when ALB-OlO-l-lB, RCP A UPPER OIL
RSVR LOW-LEVEL, alarm is received.
e The operator checks the computer points for GD AOP-018 and finds RCP A motor
thrust-bearing temperature at 195F and RCP A tipper radial bearing at 185F with
both slowly increasing.
Which of the following actions are required?
a. Stop RCP Aand initiate a rapid plant shutdown in accordance with AOP-038,
Rapid Downpower
b. Manually trip the reactor and go to PATH-I, stopping RCP Aas time permits
c. Continue monitoring RCP A temperatures, tripping the reactor and entering
PATH-I if RCP A temperatures exceed 300F
d. Stop RCP A, manually trip the reactor and go to PATH-I
ANSWER:
b. Manually trip the reactor and go to PATH-1, stopping RCP Aas time permits
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 61 TIERGROUP: 1/1
KAIMPORTANCE: RO 3.4 SHO
1QCFH55CONTENT: 41(b) 3/10 43(b)
KA: 000015iI7AA2.08
Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions
(Loss of RC Flow): When to secnre RCPs on high bearing temperature
OBJECTIVE: AOP-3.18-3
Given a set of plant conditions and a copy of AOP-018, DETERMINE the appropriate response
DEVELOPMENT REFERENCES: AOP-018
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: MEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: AOP-3.18 019
NRC EXAM HISTORY: None
DISTRACTOR ~ S T I F I C A C T I O N(CORRECT ANSWER +d):
a. Plausible since the RCP is to be stopped, but must be stopped immediately which requires that the
reactor be tripped.
d b. RCP motor temperatures require the punip be stopped. With power above 48%, the reactor must be
tripped prior to tripping the RCP.
c. Plausible since this is a trip setpoint for stator winding temperature:, but the pump must be tripped
immediately based on the given temperatures.
d. Plausible since these are the correct actions, but the reactor should be tripped first and the pump
stopped when time permits.
ICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of RCP motor temperature tripping requirements
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 62
Given the following conditions:
e Path-2 is being performed due to an SGTR.
e The Main Stcam Isolation Valves (MSIVs) on the intact SGs are open.
e The MSIV on the ruptured SG is closed.
a The Condenser is available for Steam Dump operation.
e A cooldown to 485 OF from 557 OF at the maximum rate is required.
Which one of the following describes the method to accomplish this cooldown?
a. Fully open the Stcam Dumps as fast as possible
b. Fully open the Steam Dumps as fast as possible without causing main steam line
isoiation
c. Fully open the intact SG PORVs as fast as possible
d. Fully open the intact SG PORVs as fast as possible without causing a main s t e m
line isolation
ANSWER:
b. Fully open the Steam Dumps as fast as possible without causing rnain steam line
isolation
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 62 TIEWGROUP: 1/1
10CPR55 CONTENT: 41(b) 10 43m
KA: 000038EA1.36
Ability to operate arid monitor the following as they apply io a SGTR Cooldown of RCS to specified
temperature
OBJECTWE: EOP-3.19-R4
Given a set of conditions during EOP implementation, DETERMINE the correct response or required
action based upon the EOP Users Guide general information
e Dumping steam at maximum rate
-u u
DEVELOPMENT REFERENCES: EOP Users Guide
REFERENCES SZJPPLIEDTO MPPI,ICANT: None
n
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT:
NRC EXAM HISTORY None
DIS1RACTOR JUSTIFICACTION (CORRECT ANSWER dd):
a. Plausible since the maximum cooldown rate can be achieved using maximum steam dump flow. but
causing too great a rate ofpressure drop will result in the MSIVs going closed which is undesirable.
4 b. During a SGTR cooldown an attempt should be made to maximize s t e m dump demand without
causing SC; pressure to decrease fast enough to cause a main steam line isolation.
c. Plausible since this action would be taken if the MSIVs on the intact SGBwere already closed, but
with the M S N s open it is desirable to use steam dumps.
d. Plausible since causing the MSIVs to close is not desirable, but steam dumps should be used if
available.
ICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS
DIFFICULTY IWTING: 3
a KNOWLEDGE / RECALL
EXPLANATION: Knowledge of the EOP Users Guide requirement for performing a maximum
rate cooldown
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 63
Given the followiiig conditions:
e While transferring resin, it is noted that RM-1WR-4644A, SPENT RESIN PUMP 1-
4A, radiation monitor is indicating 10 mR&.
e The monitor is physically located 20 feet away from a suspected clog in the pipe
which is the source ofthe monitor indication.
What is the radiation level 5 feet from the pipe? (ASSUME THE CLOG IN THE PIPE IS
A POINT SQIJRCE)
a. 20mRem/hr
b. 4QmRem/hr
c. 80 mRem/hr
d. 160mRem/hr
ANSWER.
d. 160rnRemhr
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 63 TIEWGRO[?E 211
10CFR55 CONTENT: 41(b) 11/12 43(b)
KA: 073K5.02
Knowledge of the operational implications as they apply to concepts as they apply to the P F N system:
Radiation intensity changes with source distance
OBJECTIVE: RP-3.5-21
Calculate dose rates at different distanccs from point sources and line sources
DEVELOPMENT REFERENCES: Rk-LP-3.5
REFERENCES SUI'PIJED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT:
NRC EXAM IIISTORY: None
DIS1'RACTOR JUSTIFICACTION (CORRECT AWSWER d'd):
a. Piausible if the square root of the distances is taken, instead of squared as they should be (IOmR/hr x
2 0 " ~ft = 20 mR//hr x 5'" ft).
b. Plausible if the distances are not squared as they should be (iOmR/hr x 20 fi = 40 mRhr x 5 ft).
e. Plausible if a mathematical error is made (value selected as a distmcter due to the progression of other
numbers in distracters).
4 d. Using the formula 11d1*=12d;, the intensity ofthe source at 5 feet is calculated to be 160 mRem/hr.
ICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Calculation of distance using inverse square for radiation
Post Validation Revision
Marris NRC Written Examination
Reactor Operator
QUESTION: 64
Given the following conditions:
0 The Control Room has been evacuated due to a fire.
0 AOP-004, Remote Shutdown, is being performed.
0 The crew is detcrmining the amount of boric acid required to be added to the RCS.
Which of the following describes the reason for adding boric acid during the performance
of Section 3.1, Remote Shutdown Due to Fire, of AOP-004?
a. Ensure adequate shutdown margin is maintained for the first 12 ~ O U F Sfollowing
the plant trip
b. 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. Ensure adequate shutdown margin is maintained in the event that a cooldown to
Cold Shutdown conditions is required
d. Ensure adequate shutdown margin is maintained in the event that pressurizer is
required to be raised to 90% ievel
ANSWER:
c. Ensure adequate shutdown margin is maintained in the event that a cooldown to
Cold Shutdown conditions is required
Post Validation Rei.,ision
Harris NRC: Written Examination
Reactor Operator
Data Sheets
QUESTION NLMBEIP: 64 TIEWGROUP: 1/2
10CFRSI CONTENT: 41(b) 6 43(W
KA: 000068AK3.13
Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation:
Perfcnning a shutdown margin celculation, inciuding boron needed and boration time
OBJECTIVE:
Given a set of plant conditions and a copy of AOP-004, Remote Shutdown, DETERMINE the
appropriate course of action
DEVELOPMENT REFERENCES: AOP-004-BI)
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW 0 SIGNIFICANTLY MODIFIED DIRECT
BANK IVUI\IBERFOR SIGNIPICAVTLY IMODIFIED/ mmcr: N ~ W
NRC EXAM HISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER .Jd):
a. Plausible since shutdown margin is changing for the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following a plant trip, hut shutdown
margin is increasing.
la. Plausible since shutdown margin will decrease as the core approaches xenon-fke conditions, but the
boration is performed in the event of a cooldown is required.
d c. A boration is only perfonned in the event that a cooldown is required to be performed during the
perforntance of AOP-004.
d. Plausibie since AOP-004 allows increasing PRZ level to 90% level, but this is done to accommodate
performance of a boration.
DIFFICULTY M A L Y S 6 :
COMPREHENSIVE / ANM,YSIS KNOWLEDGE / RECALL
DIFFICIJLTY RAlRvG: 3
EXPLANATION: Knowledge ofthe reason for performing a horation whiIe operating the plant
from the shutdown panel
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 65
Given the following conditions:
e The unit is operating at 50% power.
e LT-460, Channel 111Pressurizer Level, has failed and all associate, istables are in
the tripped condition.
e Power is subsequently lost to UPS Bus IDP-IA-SI.
Which train(§) of Reactor Protection will actuate, if any?
a. Neither train
b. Train SA ONLY
c. Train SB ONLY
d. Both trains
ANSWER:
d. Bothtrains
Post Validation Revisioii
Harris NKC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 65 TIEWGRQUP: 2: I
KAIMPQRTANCE: RQ 3.3 SWQ
1QCFR55CONTENT: 41(b) 7 43w
KA: 012K2.01
Knowledge of bus power supplies to the following: KIPS channels, components, and intercoimections
OBJECTWE: AOP-3.24-2
RECOGNIZE automatic actions that are associated with loss of an instrument bus or loss ofNNS UPS
I)E\ELOPMENT REFERENCES: AOP-024
SD-103
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICNCLY MODIFIED DIRECT
BAKK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: Iarris NRC 2000-29
NRC EXAM HISTORY: Harris NRC 2000
DPSTRAC71OR JUSTIFICACTIQN (CORRECT ANSWER dd):
a. Plausible since some ESF features arc energized to actuate, hut RPS features are all de-energized to
actuate
b. Plausible since an ESF actuation u ~ ~not l doccur on hath trains if required since slave relays require
power to actuate, but KPS is de-energized to actuate.
c. Plausible since an ESF actuation would not occur on both trains if required since slave relays require
power to actuate. hut KPS is de-energized to actuate.
d d. Bus 1A-SI suppries Channel H pressurizer level and a loss of this supply wiIl result in 2 channels being
tripped. Each channel inputs both trains of W S , so both trains of R19S will actuate.
DIFFICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analysis of abnormal conditions regarding Ioss of power to determine response
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 66
Given thc following conditions:
o An earthquake has caused damage to the Main Reservoir dam
Main and Auxiliary Reservoir levels are both currently 240 feet and stable.
o AOP-022, Loss of SenriceWater, is being performed for a IAM of Ultimate Heat
Sink.
e Emergency Sewice Water (ESW) pumps have been aligned to the Main Reservoir.
e One ( I ) Noma1 Service Water (NSW) pump is operating.
Which of the following pumps are required to be operating to provide water to the SSE
Fire Protection Header once the ESW header is aligned to the fire protection header?
a. ONLY an ESW pump
b. An ESW pump AND an ESW Booster pump
c. ONLY a second NSW pump
d. A second NSW pump AND an ESW Booster pump
ANSWER:
b. An ESW pump AND an ESW Booster pump
Post Validation Revision
Iiarris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 66 TIEWGROUP: 21 1
10CFR55 CONTENT: 41(b) 4 43w
KA: 076K1.15
Knowledge of the physical connections and/or cause-effect relationships between the S W S and the
following systems: FPS
OBJECTIVE: FP-3.0-3
STATE the sources of fire water available to the plant including automatic actuation signals
DEVELOPMENT REFERENCES: AOP-022
OH-I39
REFERENCES SUPPLIED TO ABPIXANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED
OW SIGNIFICANTLY MODIFIED I DIRECT:
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
a. Plausible since an ESW pump is started, but an ESW Booster pump is also required.
d b. An ESW pump, aligned to the Main. Reservoir, is started, along with an ESW Booster pump to supply
thc SSE fire protection header.
c. Plausible since the first NSW pump is not required to be tripped provided cooling tower basin level is
adequate and NSW supplies the ESW header (which can supply the fire protection header), but an
ESW pump is required.
d. Plausible since an ESW Rooster pump is required to supply the fire header, hut an ESW pump is
required to supply the booster pump.
DIFFICULTY ANALYSIS:
COlcIPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLATATION: Knowledge of the system alignments avaiiable to supply the fire header
Post Validation Revision
Hmis NRC Written Examination
Reactor Operator
QUESTION: 67
Given the following conditions:
e The plant is being cooled down to 140°F for maintenance which will NOT require the
RCS be opened.
e The crew is in the process of placing the first Residual Heat Removal (RRR) train in
service for RCS cooling.
o Current boron concentrations are as follows:
RHR (train to be placed in service) boron 1021 ppm
Required Shutdown Margin boron 1200 pprn
Cold Shutdown boron 1750 ppm
o Refueling boron 2261 ppm
Before the RHR train can be placed in service for RCS cooling, WHR boron
concentration must be increased by a MINIMUM o f . ..
a. 179ppm.
b. 320ppm.
c. 729ppm.
d. 124Qppm.
ANSWER:
a. 179ppm.
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 67 TIER/GR0UID: 21 I
1OCFR55 CONTENT: 41(b) 8 43W
K& 005K5.09
Knowledge of the operational implications of the following concepts as they apply the RHKS: Dilution
.and boration considerations
OBJECTIVE: RHRS-2.0-12
APPLY precautions and limitations of OP-111,KGNS to Hypothetical System Configurations
DEVE1,OPMENT IWFERENCES: OP-llI
REFERENCES SUPPLIED TO APPLICANT:
~ ~ ~~~ None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DIS'I'RACTOH JUSTIFICACTION (CORRECT ANSWER tl'd):
d a. RHK boron must be greater than or equal to the required SDM or the required reheling concentration.
The boron concentration requirements will be dependent on the intended use of the KHR System.
Using the RHR 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 RHR and RCS boron concentration, but only the
e. Plausible since this is the difference between RHR and Cold Shutdown boron concentration, but only
the required SDM boron is needed.
d. Plausible since this is the difference between RIIR and rcfuelirig boron concentration, and refueling
conditions occur at 140°F.but only the required SDM boron is needed.
DIhTICUI,TY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DnmcucrYw m G : 3
EXPLANATION: Application of actual versus required boron concentration - must determine
minimum limiting requirement
Post Validation Revision
Harris WRC Written Examination
Reactor Operator
QUESTION: 68
Given the foilowing conditions:
A liquid waste discharge from a Treated Laundry and Hot Shower (TL&HS) Tank is
in progress.
REM-1WL-3540, Treated Laundry and Hot Shower Tank Pump Discharge Monitor,
goes into high alarm.
Which of the following terminates the discharge'?
a. The running TL&HS Tank Pump will automaticalEy trip.
b. An operator must take manual action to shut the TL&HS Tank Pump Discharge
Isolation Valve.
c. The running TL&HS Tank Pump Recirc Valve will automatically open.
d. The TL&HS Tank Pump Discharge Isolation Valve will automatically close.
ANSWER:
d. The TI.&WS Tank Pump Discharge Isolation Valve will automatically close.
Post Validation Revision
Hareis NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 68 TIEFUGROUP: 2i2
IOCFR55 CONTENT: 41(b) 7/13 43@)
KA: 068A3.02
Ability to monitor automatic operation ofthe Liquid Radwaste System including: Automatic isolation
OBJECTWE: LWPS-LP-3.0-7
UESCRE3E the automatic protection features associated with discharges to the environment from the
LWPS
DEVELOPMENT REFERENCES: AOP-005
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED D I ~ ~
HANK NUMBER FOR SIGNIFICANTLY MODIFIED /DIRECT: Kh4S-A6 005
NRC EXAM HISTORY: None
DISTRACTOR JUSTIPICACTION (CORRECT ANSWER .I'd):
Plausible since the pump will stop the discharge, but there is no auto trip due to high rad.
h. Plausible since manual isolation will stop the discharge, but an auto isolation will not require operator
action.
c. Plausible since placing the tank in recire will stop discharge, but only because of the isolation valve,
as the recirc valve does not have an auto function.
./ d. On a high rad level as sensed by REM 3540, the discharge isolation valve will automatically close,
terminating any release in progress.
DIFFICULTY ANhLYsrs:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of liquid radwaste design and operation
Post Validation Revision
Harris NKC Written Examination
Reactor Operator
QUESTION: 69
Assuming NO operator actions, which ofthe 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. VCT levcl increases due to minimum charging and the letdown pressure control
valve failing open
d. VC'T level increases due to minimum charging and the letdown orificc isolation
valves failing open
ANSWER:
a. VCT level decreases due to maximum charging and letdown isolation valves
closing
Post Validation Revision
Harris NRC Written Examinatian
Reactor Operator
Data Sheets
QUESTION NUMBEM: 69 TIER/GROUP: 21 1
10CFR55 CONTENT: 41(h) 4 43(W
KA: 048K3.02
Knowledge of the effect that a loss or malfunction of the KAS 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, DETEKMINE the appropriate response.
DEVELOPMENT REFERENCES: AOP-0 17
REFERENCES SUPPLIED TO APPLICANT
QUESTION SOURCE: NEW a None
SIGNTFXCtU'TILY MODIFIED
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: CVCS-R3 008
0 DIRECT
NRC EXAM HISTORY: None
DISTRACTOR JUS'THFICACTION (CORRECT ANSWER d'd):
d a. Charging flow control fails open and letdown isolation valves fiail closed on a loss of air, so VCT level
will decrease.
b. I'lausibie since VCT level will decrease, but it will be due to letdown isolating, not diverting water to
the hold up tank.
C. Plausible since the letdown pressure control vahe faiis open on a loss of air, but the letdown isolation
valves fail closed, isolating ietdown.
d. Plausible since the charging Bow control valve and the letdown orifice valve all fail on a loss of air,
hut fail in the opposite direction as that which would cause this response.
DPFFICUI,TY ANALYSIS:
COMPREIIENSWE / ANALYSIS KNOWLEDGE /RECALL
DIPFICULTY RATING: 3
EXPI.ANATION: Analyze the response of GVCS aAer determining the fail position of various
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 70
Given the following conditions:
o Following a plant trip, EOP-EPP-004, Reactor Trip Response, is being performed.
o The crew is verifying Natural Circulation conditions as a result of a loss of power to
all RCPs.
m Five (5) core exit thermocouples are failed.
How do the failed core exit therniocouples affect indications used to verify Natural
Circulation?
a. * The Core Exit Temperature indications will be HIGHER than actual
o RCS Subcooling will indicate MORE subcooling than actual
b. e The Core Exit Temperature indications will be HIGHER than actual
RCS Subcooling will indicate LESS subcooling than actual
c. Core Exit Temperature indications will indicate LOWER than actual
o RCS Subcooling will indicate MORE subcooling than actual
d. e Core Exit Temperature indications will indicate the SAME as actual
e RCS Subcooling will indicate the SAME subcooling as actual
ANSWER:
d. Core Exit Temperature indications will indicate the SAME as actual
o RCS Subcooling will indicate the SAME subcooling as actual
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QCESTION NUMBER: 70 TIEWGHOUP: 2l2
IOCPR55 CONTENT: 41(b) 5 4300
KA: 017K3.01
Knowledge of the effect that a loss or malfunction of the ITM system will have on the following: Natural
circulation indications
OWECTNE: ICX3Vf-4.0-R6
DESCRIBE how the plant's subcooling monitor infonnation is processed
DEVELOPMENT REFERENCES: SD-106
ICCM-LP-3.0
mFERENCES SUPPLIED TO APPLICANT: None
QUESlION SOURCE: NEW SIGNIFICANTLY RIODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT New
NRC EXAM HISTORY: None
DISTRACTOR SUSTIFICACTION (CORRECT ANSWER d'd):
a. Plausible since the thermocouples are failed, but a failed thermocouple indicates 507: (low) and not
high.
b. Plausible since the thermocouples are failed, but a failed thermocouple indicates 50'F (low) and not
high.
e. Plausible since the failed theirnocouples indicate 50°F (low), but the ICCM indication uses the highest
thennocouples and not the lowest.
d d. The failed thermocouples will not be used to process the indication by the ICCM, so them will be no
change on core exit temperatures and subcooling margin.
DIFFICULTY ANALYSIS:
COMPREIPENSNE I ANALYSIS KNOWLEDGE I RFCAI,L
DIFFICULTY RATING: 3
EXPLANATION: Analyze the effect of failed thermocouples on temperatures and subcooling
margin
Post Validarioti Revision
Hwris NRC Written Examination
Reactor Operator
QUESTION: 71
Which of the following EOP network procedures, containing NO Immediate Operator
Actions, may be directly entered and the conditions under which it may be entered?
a. EOP-FRP-1.1, Response to Pressurizer High Level, when it is desirable to
restore pressurizer level following a malfunction of the Pressurizer Level Control
System and NO accident is in progress
b. EOP-FRP-I. I, Response to Pressurizer High Level, when it is desirable to
restore pressurizer level following an inadvertent Safety Injection actuation with
the plant in Mode 3
C. EPP-005, Natural Circulation Cooldown. when it is desirable to cuoidown with
RCPs unavailable and NO accident is in progess
d. EPP-005, Natural Circulation Cooldown, when it is desirable to cooldown with
RCPs unavailable due to a loss of offsite power with the plant in Mode 3
ANSWER:
c. EPP-005, Natural Circulation Cooldown, when it is desirable to cooldown with
RCPs unavailable and NO accident is in progress
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 41 TIEWGROFTP 3
KAIMPORlANCE: RO 4.3 SHO
10CFH55 CONTENT: 41(h) 10 43m
KA: 2.4.1
Knowledge of EOP entry conditions and immediate action steps
OBJECTIVE: EOF-3.19-1
DESCRIBE Control Room usage of the EOP network as it relates to the following
e Entry into EOP network
DEVFLOPMENT REFERENCES: EOP Users Guide (page 13)
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW , SIGNIFICANTLY MODKFFIED DIRECT
BANK NLJMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: Mew
NRC E.XhM " T O R Y : None
DISTRACTOR JUSTIFKCACTION (CORRECT ANSWER d'd):
8. Plausible since FKP-1.1 is used to lower PFZ leve1 during the perfomlance of the EOP network, but is
entered only by operator judgement when a yellow path condition is encountered.
b. Plausible since FW-1.1 is used to lower PKZ level during the perfomlance of the EOP network, but is
entered only by operdtor judgement when a yellow path condition is encountered.
4 c. EPP-005, "Natural Circulation Cooldown," may be entered whenever a natural circulation cooldown is
required and an accident is not in progress.
d, Plausible since EFP-005 may be entered whenever a natura1 circulation cooldown is required, but no
accident can be in progress.
DIFFICUI.TY ANALYSIS:
0 COMPREFIENSIVE I ANALYSIS KNO\VLEDGE /RECALL
DIFFICUL7L1' RATENG: 2
EXPLAK4'rION: Knowledge of EOPs which can be entered directly
Post Validation Revision
Hmis NRC Written Examination
Reactor 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
b. Containment leakage could bc in excess of design basis lcakage
c. Hydrogen recomhiner efficiency is significantly reduced at the pressure
d. Containment temperature is high enough to prevent adequate core cooling
ANSWER
b. Containment leakage could be in excess of design basis leakage
Post Validation Revision
Harris NIPC Written Examination
Reactor Operator
Data Sheets
QUEsTION NUMBER 92 mwcxtom 2/1
IQCFRSCONTENT: 41(b) 9/10 43(b)
KA: 10362.4.6
Knowledge of symptom based EOP mitigation strategies. (Containment)
OBJECTIVE: EOP-3. 13-4
Given the following EQP steps, notes, and cautions, DESCMBE the associated basis
CSF decision points
DEVELOPMENT REFERENCES: EOP-LP-3.13
REFERENCES SUPPLIED TO APPLICANT:
QUESTION SOURCE: NEW n None
SIGNIFICANTLY MODIFIED
RANK NIJMMBER FOR SIGNIFICANTLY MODIFIED / DIRECT:
DIRECT
EOP-3.13-194 001
NRC EXAM IIISTORY None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
a. Plausible since this is above the postulated pressure following a large break LOCA or steam break, but
containment failure is not expected to occur until several times this value.
d b. 45 psig is above the pressure that design containment leakage rates assumed in off-site radiological
analysis.
e. Plausible since the recombiners are located in containment and are conceivably affected by the high
presmrc, but the 45 psig is selected based an exceeding design leakage rates.
d. Plausible since core cooling in the event of a LOCA is based upon transfemng heat to the injection
water which is then collectcd in containment for recirc, but the 45 psig is selected based on exceeding
design leakage rates.
DIFFICULTY ANALYSIS:
COMPREIIENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Knowledge ofthe basis far CSFST decision points for containment pressure
Post Validation Revision
IIarris MRC Written Examination
Reactor Operator
QUESTION: 73
Which of the following would require that Independent Verifrcation be performed in
accordance with OPS-NGGC- 1303, "Independent Verification'?"
a. During Mode 5, a valve in the Containment Spray system is being repositioned for
testing and the OP lineup will be completed prior to Mode 4 entry
b. During Mode 1, a valve in the Main Steam system is being placed under clearance
and the valve is only accessible with a manlift
C. During Mode 4, a valve in CVCS inside containment is being positioned for
draining and the valve is located in an area where the temperature is I34OF
d. During Mode 3, a valve in CVCS is being placed mder clearance and the valve is
located in a radiation field of 175 mRcmihr with an estimated verification time of
6 minutes
ANSWER:
h. During Mode 1, a valve in the Main Steam system is being placed under clearance
and the valve is only accessible with a manlift
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER 73 TIENGROUP: 3
1OCFR55 CONTENT: 41(b) 10/12 43(b)
KA: 2.2.13
Knowledge of tagging and clearance procedures
OBJECTIVE: PI-3.1 I-H8
DISCUSS the forlowing i t e m concerning independent verification
a. When it is used
e. When it may be waived for A L M A
DEVELOPMENT REFERENCES: OPS-NGGC-I 303
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGN1FICANTL.Y MODIFIED DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: PP-3. i 1-R8 003
NRC EXAM HISTORY: Harris NRC 2002
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER dd):
a. Plausible since the valve is easily accessible, but rV is not required since plant conditions do not
require the system to be operable and spray valve positions are covered by a lineup.
4 b. IV requirements may be waived if radiation exposures : 10 &em would result, if area temperatures
in excess of 1200F exist, or ifplant conditions do not require the equipment to be operable and a varve
lineup controls the position.
c. Plausible since the valve is k i n g repositioned for draining, but area temperature conditions pennit
waiving the Ri requirements.
d. Plausible since the valve is being repositioned for a clearance, but projected radiation exposures
permit waiving the Hv requirements (17.5 mR > 10 a).
DIFFICULTY mmysrs:
COMPREIIENSIVE / ANALYSIS KNOWLEDGE /RECALL
DrFFIcxrcrY RATING: 3
EXPLANATION: Application of plant conditions to implementation of administrative procedural
requirements
Post Validation Revision
IIarris NRC Written Examination
Reactor Operator
QUESTION: 74
Given the following conditions:
Following an accident, EOP-EPP-015, Uncontrolled Depressurization of All Steam
Generators, is being performed.
Due to the excessive cooldown rate, the operators have reduced AFW flow to all
stcam generators (SG) to mininium as they continue attempts to isolate the SGs.
Which of the following describes the expected plant response to the AFW flow reduction
and what actions are to be taken as SG pressures decrease?
a. RCS hol leg temperatures will eventually begin to increase and the crew will then
transition to EOP-EPP-008, Safety Injection Termination
b. RCS hot leg temperatures will eventually begin to increase and the crew will then
be required to increase AFW flow to maintain RCS hol leg temperatures stable or
decreasing
c. The SGs will eventually approach dry conditions and no longer be required and
the crew will then transition to EQP-EPP-008, Safety Irijection Termination
d. Tke SGs will eventually approach dry conditions and the crew will then be
required to isolate AFW flow to all SGs
ANSWER:
b. RCS hot leg temperatures will eventually begin to increase and the crew will then
be required to increax AFW flow to maintain RCS hot leg temperatures stable or
decreasing
Post Validation Revision
Harris NRC Written Examination
Reactor Operator
Data Sheets
QUESTION NUMBER: 74 TIEWGROUP: 1/1
10CFR55 CONTENT: 41(b) 4/10 43@)
I<A: WEl2EK2.1
Knowledge of the interrelations between the (Uncontrolled Depressurization of all Steam Generators) and
the following: Components, and functions of control and safety systems, including instmmentation,
signals, interlocks, failure modes, and automatic and manuai features
OBJECTIVE: EOP-3.9-4
Given actions taken in these emergency procedures, PREDICT the plant response to these actions
DEVELOPMENT REFERENCES: EOP-EPP-0 15
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SlGNIFICAh'TLFLY MODIFIED DIRECT
BANK NLIWBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACTION (CORRECT ANSWER d'd):
a. Plausible since hot leg temperatures will eventuaIly increase, but the comct action is to stabilize
temperature by increasing AE'W flaw and adjusting steaming rate, if possible.
d b. As SCi pressure and steam flow decrease, RCS hot leg temperatures will eventually stabilize and may
increase. Adjusting feed flow and steam dump will control RCS hot leg temperatures.
e. Plausible since the minimum APW flow will result in the SGs nearing dryout conditions, but as hot
leg temperature begins to increase the correct action is to stabilize temperature by increasing AFW
flow and adjusting steaming rate, if possibie.
d. Piausible since the niininium AFW flow will result in the SGs nearing dryout conditions, but as hot
leg tcniperature begins to increase the correct action is to stabilize temperature by increasing AFW
flow and adjusting steaming rate, if possible.
ICULTY ANALYSIS:
COI\IPREII%NSIVE/ ANALYSIS KNOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Analyze SO response to decreasing pressure and reduced AFW flow and
determine correct response
I'oat Validation Revision
Harris NRC Written Examination
Reactor Operator
QUESTION: 75
Which ofthe following conditions would permit securing Containment Spray per EOP-
PATH-1 Guide?
a. a Actuation caused by a LOCA
4 Time since LOCA occurred is 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
Containment pressure is 9 p i g
b. a Actuation caused by il LOCA
o Time since LOCA occurred is 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />
a Containment pressure is 5 psig
c. a Actuation caused by a Steam Line Break
e Time since Steam Line Break occurred is 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />
4 Containment pressure is 5 psig
d. 4 Actuation caused by a Steam Line Break
a Time since Steam Line Break occurred is 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
0 Containment pressure is 9 psig
ANSWER
c. 4 Actuation caused by a Steam Line Brcak
4 Time since Steam Line Break occurred is 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />
- Containment pressure is 5 psig
Post Validation Revision
-.,. ~ ...,.,...,.............
Harris NRC Written Examination
Reactor Operator
Rata Sheets
QUESTION NUMBER 75 'HEWGROUP: 21I
10CFIW CONTENT: 41(b) 7/9 4Yb)
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, contro1,or mitigate the consequences ofthose
malhnctions or operations: Safe securing of containment spray (when it can be done)
OJUECTIVE: EOP-3.1-7
Given the following EOP steps, notes, and cautions, DESCRIBE the associated b a i s
d. Criteria for securing of ChMT spray
DEVELOPMENT REVERENCES: EOP Guide 1
REFERENCES SUPPLIED TO APPLICANT: None
QUES'GI~NSOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / 1)IRECT: EOP-3. I-R5 006
NRC EXAM HISTORY: Hams m C 2002
DISTRACTOR JCJSTIFICACTION (CORRECT ANSWER J'd):
a. Plausible since the minimum time requirement of4 hours of spray has been met, but pressure must be
reduced below 8 psig prior to resetting.
b. Plausible since pressure is below 8 psig, hut must meet the time requirements prior to resetting.
4 c. Containment pressure must be below 8 p i g to ensure it can automatically reactuate if pressure goes
back above 10 p i g , and in the event of a LOCA the system must operate for > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, but in a steam
break early termination is desired.
d. Plausible since the minimum time requirement of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of spray (for 6,MK:A) has been met, but
pressure must be reduced below 8 psig prior to resetting.
DIFFICULTY AllALYSIs:
c]COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Kaowledge of the bases for EQP steps
Post Validation Revision