ML041170063
| ML041170063 | |
| 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: ML041170063 (52) | |
See also: IR 05000400/2004301
Text
HARRIS EXAM
50-400/2004-301
-
FEBRUARY 23 27,2004
& MARCH 4,2004 (WRITTEN)
Harris
Draft
SRQ
Written
2004
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: I
Given the following conditions:
Whiie operating at 100% power, a drop in PKZ pressure resulted in a Reactor Trip
and Safety Injection.
PRZ level is currently indicating > 100%.
PRZ pressure has stabilized at 1400 psig.
Containment pressure is 3.6 p i g and stable.
RCPs have been stopped.
R V t I S Full Range is indicating 20%.
Core Exit Thermocouples are indicating 745'1:.
RC:S Wide Range Hot Leg Temperatures are indicating 6SO'I:.
Which of the following conditions currently exists'?
a. A PKZ steam space break has occurred and core heat removal is ADEQUAI'E
b. A PRZ steam space break has occurred arid core heat removal is INADEQUAIE
c. An RCS hot ieg break has occurred and core heat removal is ADEQUATE
d. An RCS hot leg break has occ.urred and core heat removal is INADEQL!A?'E
ANSWER:
b. A I'KZ steam space break has occurrcd and core heat removal is INADEQGATE
Post Validation Rwision
Harris NKC Written Examination
Senior Reactor Operator
Llata Sheets
QUESTION NUMBER: 1 TIEWGROUP: 1:1
IOCFR55 CONTENT: 41(b) 43(b) 5
KA: 000008AA2.30
Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident:
Inadequate core cooling
ORJECTIVE: EOP-3.10-4
Given the following EOP steps, notes, aud cautions, describe the associated basis
c. RVLIS level of 39 percent (C. I )
DEVELOPMENT REFERENCES: ECP-FRP-C. 1
C'SFST-Core Cooling
REFERENCES SIJPPII,PEDTO APPLICANT: None
.
OUESTYON SOIJRCE:
LA
NEW flSKGNIFICANT1.Y MODIFIED
LA
n
bl
DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / D I m c r : N~\V
NRC EXAM HISTORY: None
DISTR4CTOR .JUSTYFICACTIOIV(CORRECT ANSWER \I'd):
a. Plausible since the break is located in the PRZ steam space, but heat removal is not adequate.
d b. 'the RCS is superheated and in excess of 700"F, which indicates that inadequate heat rerncwal is
occuiiing. The break is in the PKZ steam space as indicated by the pressurizer being full.
c. Plausible since RCS temperatures are stable, hut the break is in the stearn space and heat removal is
not adequate.
d. Plausihle since RCS heat removal is not adequate, but the break is in the steam space.
DIFFICULTY ANALYSIS:
C0iW"IEPIENSIVE / ANALYSIS KNOWLEDGE /RECALL
DIFTICLJ1,TYRATIXG: 3
EXPLANATION: Must analyze plant conditions to determine location of hreak, determine that
temperature indications support superheated conditions and that heat removal is
inadequate
Post Validation Revision
IIarris NRC Written Examindtion
Senior Reactor Operator
QUESTION: 2
Which of the following describes a condition which would require Emergency Uoration
and the bases for taking this action?
a. e 'l'wenty minutes following a Main Feedwater Pump trip, Control Rods are
determined to be below the rod insertion limit
- Control the reactivity transient associated with a steam line break
h. e Twenty minutes following a Main Feedwater Pump trip, Control Rods are
determined to he helow the rod insertion limit
e Control the reactivity transient associated with an inadvertent dilution
c. * During a reactor startup, the Reactor achieves criticality with Bank C rods at
105 steps
- Control the reactivity transient associated with a stearn line break
d. * During a reactor startup, the Reactor achieves critic.aIitywith Bank C rods at
105 steps
- Control the reactivity transient associated with an inadvertent dilution
AKSWEW:
c. e During a reactor startup. the Reactor achieves criticality with Bank C rods at
IO5 steps
Control the reactivity transient associated with a steam line break
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Opcrator
Data Sheets
QUESTION NUMBER: 2 TIEWGROUP: li2
llOCFR55 CONTENT: 41(b) 43(b) 2
KA: 000024G2.2.25
Knowledge of bases in technical specifications for limiting conditions for operations and safety limits
(Emergency norat ion)
OBJECTIVE: CVCS-3.0-R4
Given a (.VCS coniponentipa~anieter,state whether the componentiparameter is Tech Spec related
DEVELOPMENT REFERENCES: IS Bases 3i4.1.1
.4OP-002 ED
tip-004
REFERENCES SIJPPLIED TO APPLICANT: ?\one
QUE.STIOK SOURCE: NEW SIGNIFICANTLY MODIFIED [3DIRECT
BANK NUMBER FOR SIGNIFICANTLY iV1C)DIFIED / DIRECT: AOP-3.2-Kl 001
NRC EXAM HISTORY: None
DISTRACTOR JIJSTIFICACTION (CORRECT ANSWER dd):
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 Roration would
be required. Additionally, in Modes 1 & 2 , SDM is required to control the reactivity transient
associated with a s t e m line break. However, it is not required during transient conditions, allowing
the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> t o restore rod position.
La. Plausibic 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
he rcyuired. However, it is not required during transient conditions, ailowing 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 c. Emergency boration is required if SDM is not met. Criticality at steady spate conditions is considered
to he a loss of SDM. In Motlcs I & 2, SDM is required to control the reactikity transient associated
with 3 steam line break.
d. Plausihle since Emergency boration is required if SI)M is not met. Criticality at steady state
conditions is considered to he a loss of SDM. However, the concern for an inadvertent dilution is
related to a shutdown condition.
ICKJLTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEIIGE i RECALL
UIFFICULIY RATIXG: 2
EXPL,AN,4IION: Knowledge of the requirements for initiating Emergency Boration and the bases
for these actions.
Post Validation Revisioii
IIarris NRC Written Examination
Senior Reactor Operator
QCESTION: 3
Given the following conditions:
e Ihe plant has been operating at I@@% power for the past three ( 3 ) months.
- CSIP 1.4-SA is operating.
o CSIP 1B-SB has just been restored to a normal alignment following maintenance on
the pump impeller.
o When CSIP 1B-SR is started the operator notes that suction pressure appears nornial,
while discharge pressure, discharge flow, and pump current are oscillating.
Which ofthe following is the most likely cause of these CSIP 1B-SI3 indications?
a. Inadequate venting was performed during clearance restoration
b. The CSIP 1B-SB discharge valve was inadvertently left closed during clearance
restoration
c. A failure of the CSIP 1B-SB miniflow isolation valve has resulted in gas binding
(I. A failure ofthe (XI IR-SB miniflow isolation valve has resulted in all pump
flow being recirculated to the VCT
ANSWER:
a. inadequate venting was perfonned during clearance restoration
Post Validation Revision
Haris NRC Written Examination
Senior Rnctor Operator
Data Sheets
QtXSTION NUMBER: 3 TIEWGROUP: 2: I
IOCFRS C o w r m T : 41(b) 43fb) 5
EL\: 006A2.04
Ability to (a) predict the impacts ofthe following malfunctions or operations on the ICCS; and (b) based
on those predictions, use procedures to correct, control, or mitigate the consequences ofthose
inalfiinctions or operations: Improper discharge pressure
OBJECTIVE: AOP-3.2-4
Given a set of plant conditions and a copy of AOP-002, determine if the possibility of gas hinding the
CSIPs exists and the coirectiue action to be taken
DEVELOPMEST REFERENCES: OP-IO7
SOEK 97-1
REFERENCES SUPPLIED TO APPLICANT:
~ None
QUESTION SOURCE: NEW SIC~MFICANTLYMODIFIED DIRECT
BARK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: Rew
NRC EXAM HISTORY: None
DISTRACTOW SI!STIFPCACTBON (CORRECT ANSWER dd):
d a. Gas binding o f 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 VdhC
would cause discharge pressure to be high.
e. Plausible since gas binding is cause of these indications, but will not occur as a result of pump recirc
valve being open.
d. Plausible since a failed open recirc valve will cause indicated flow to be low since flow i s rneasu~ud
dowstreatn of the recirc valve. hut discharge pressure and current would be at or near normal.
DIFFICULTY ANALYSIS:
COMPREEIENSIVE i ANALYSIS KNOWLEDGE i RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must analyze given pump conditiuns to determitie failure mode and then
determine likely cause of gas binding of the pump
Post Validation Revision
Harris NKC Written Examination
Senior Reactor Operator
QUESTION: 4
Given the following conditions:
e The unit is operating at 100% power, with C.;ontrol Bank D rods at 215 steps.
e ALB 13-7-1, ROD CONIROI, URGENT ALARM, is in AIAKM due to a failure in
Power Cabinet I AC.
o Rod Control is in MAN.
e- A turbine trip occurs, but the Reactor f'ails to trip either automatically or manually.
Which of the following actions should the Reactor Operator be directed to take'?
a. Place the Rod Control BANK SELECTOR in AUTO and allow rods to itisett
b. Maintain the Rod Control K4NK SELECTOR in MAN and manually insert rods
c. Place the Kod Control BANK SELECTOK in RANK U and manually insert rods
d. Maintain rods at 2 15 steps
ANSWER:
d. Maintain rods at 21 5 steps
Post Validation Kevision
IIarris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 4 TIEWGROUP: 22
10CFR55 CONTENT: 4B(h) 43(b) 5
KA: 001G2.4.h
Knowledge of symptom based E01' 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 1.Jser's Guide general information
z. Use of "Bank Select" during an AI'WS
DEVELOPMENT REFERENCES: E( )P-USERS GUIDE
EOP-FRP-S. I
REFERENCES SUPPLIED TO APPLICANT: None
QuESTIcrN SOUIPCE: NEW' SIGNIFICANTLY MODIFIED DIRECT
BANK NUM5ER FOR SIGNIFICANTLY MODIFIED / DIRECT New
NRC EXAM EPIS'IORY: None
DISTRACTOR JLSTIFICACTION (CORRECI' ANSW'ER +d):
a. Plausible since this is an RNO action for a failure of the reactor to trip. but will not be successful due
to the urgent failure in rod control.
b. 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.
c. Plausible since this will allow Bank D rods to tmwe inward, and is the only method of iuserting rods
with the rod coutrol failure, hut should not be used due to the potential to cause unanalyzed flux
shapes.
4 d. Due to the urgent failure, rods will not nmve in AIJTO o r MAN, Although they urill move in BANK
D with this particular failure, niovitig r d s in individual banks may result in unanalyzed flux shapes
which could result in hrl damage.
DIFFICULTY ANALYSIS:
Q~OMPRFXBENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must aualyze the effect of an urgent rod control failure a i d then apply the
failure results to the plant conditions to determine the proper actions
Post Validation Revision
Harris NRC Written Examination
Seniot Reactor Operator
QUESTION: 5
Four Operators worked the following schedule in the Control Room over the past six
days:
I-IOIJRS WORKED (Shift turnover lime not included. Do NOT assume any hours
worked before or after this period.)
OPERATOR DAY B DAY 2 DAY 3 DAY 4 DAY 5 DAY6
1 I0 14 off 12 12 12
2 14 12 14 10 off 11
3 off off off 13 I1 14
4 I1 13 14 off II 12
Which of the operators would be permitted to work a 12-hour shift on Day 7 W'IIHO1iT
requiring permission to exceed nonnal o w t i m e limits?
a. Operator 1
b. Operator 2
c. Operator3
d. Operator 4
ANSWER:
a. Operator 1
Post Validation Revision
Harris NRC Written Examination
Senior Keactor Operator
Data Sheets
QUESTION NCMBER: S TIEWGROIJP: 3
lQCFR55CONTENT: 41@) 43(h) 5
KA: 2.1.2
Knowledge of operator responsibilities during all modes ofplant operation
OBJECTIVE: PP-2.0-SI
$FATE the requirements contained in Administrative Controls Section, including requirenients for
the following:
e Unit staff, including overtime limitations
I)E\ELCPPMENT REFERENCES: AP-012
REFERENCES SUPPLIED TO APPIKANT:
~ None
QUESTION SOIJRCE: NEW SIGNIFICAIVTLY MODIFIED DIRECT
BANK NUMBF:R FOR S1GNIFICANTI.Y RIODIEIED / DIRECT: Robinson NRC 200 I
NRC EXAM IIISTORY: None
DISTRACTOR JI;STIFICACTICPN (CORRECT ANSWER dd):
d a. Working a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 of 18,and 72
hours in I days, both of which are permissible.
b. Plausible since this operator would not e?tc~edthe 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 recent day
off, but 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.
E. Plausible since this operator would not exceed the 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> in 41 day limit and has several recent days
off, but wouid work more than 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in 48 which exceeds limit.
(8. Ilausible since this operator would riot 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 arid has had a recent day
off. but 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.
DIFFECXJLTY AIVALYSBS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIFFICULTY RATING: 3
EXPLANATION: Kequired to compare given data to administrative litnits to dctermine which
operator would remain within acceptable overtime limits
Post Validation Revision
Hairis NRC Written Examination
Senior Reactor Clperator
QBJESTIBN: 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 Oow.
e SG levels are < 25% and the required AFW ilow cannot be established, so the
opcrators enter FOP-ERP-H. 1, Response to Loss of Secondary Heat Sink.
MCS pressure is 175 psig.
Ail S G pressures are between 300 psig and 350 psig.
Which of the following actions is to be taken?
a. Continue in EOP-FRP-H. 1 since FOP-FRP-H. 1 has a higher priority than PATH-I
and attempt to establish AFW or Main Feedwater flow.
b. (ontintie in FOP-FRP-11. I since EOP-FKP-H.1 has a higher priority than PATH-I
and initiate KCS feed and bleed.
c. Keturn to E,OP-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 FOP-PATH- I at Entry Point C since a secondary heat sink is NOT
required following a large break LOCA.
ANSWER:
c. Return to IiOP-PA?II-l at the step that was in elfect since a secondary heat sink is
KOT required following a large break LOCA.
Post Validation Revision
Ifarris NKC Written Examination
Senior Reactor Operator
P d M $lieetS
QIJE.:s'I'ION NUMBER 6 TIEWGROIJP: lil
10CFR55 CO?XENT: 4P(b) 43(b) 5
ai: 00001 1G2.4.6
Knowledge of symptom based EOP mitigation strategies. (Large Break 1,OCA)
OBJECTIVE: EOP-3.11-4
Given the following EOP steps, notes, and cautions, describe the associated basis
e. Requirements fur a heat sink (W. I )
DEVE1,OPMENI' REFERENCES: E0P-FRP-K. 1
REFEKEBCES SUPPI.1F.D TO APPLICANT: None
QrJKSTION SOURCE: NEW SIGNIF'ICANT1,Y MODIFIED DIRECT
BASK NUMBER FOR S K CANTLY MODIFIED / DIRECT: EOP-3. I 1-KI 003
NRC EXAM HISTORY: Sone
1)PSTRACTOH JCJSTIFICACTION (CORRECT ANSWER d'd):
a. Plausible since these are actions that are taken upon entry iuto FRP-H. 1, but a secondary heat sink
would not be required with RC'S pressure <' SG pressure.
b. Plausible since these are actions that might be taken upon entry into FRP-H.I. but a secondary heat
sink would not be required with RCS pressure 'c SG pressure.
I
V c. Since RCS pressure is less than SI a. If a safety injection occurs while implementing FW-S. 1, proper operation of SI equipment is veritkd
while continuing with FRP-S.I.
b. Plausible since PATII-I provides instructions for a response lo safety injection, but FRP-S. I must be
performed until completion.
c. Plausible since PATH-I provides instructions for a response to safety iujection. but FRP-S. I must be
performed until completion.
d. Plausible sirice a safety injection will result in a loss of MFW: hut AFW flow is capable of providing
niininium required flow.
I c u c r Y ANALYSIS:
COMPREHENSIVE i ANALYSIS m O W L E D G E / RECALL
DIFFICULTY RATING: 2
EXPLANATION: Knowledge of procedural requirements in EPP-FRP-S. I
Post Validation Revision
IIairis NRC Written Examination
Senior Reactor Operator
QUESTION: 9
Given the following conditions:
- The plant is in Mode 3 with all Shutdown Rods withdrawn.
e 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 all Shutdown Bank Rods are fully withdrawn using Detnand Position
Indication
b. Determine that all Shutdown Bank Rods are fully withdrawn using the movable
incore detectors
c. Commence a boration ofthe RCS to ensure adequate Shutdown Margin
d. Open the Reactor Trip Breakers
ANSWER:
d. Open the Reactor Trip Breakers
Post Validation Revision
Ilanis NRC: Written Examination
Senior Reactor Operator
Data Sheets
QtJESTION NUMBER: 9 TIEWUGROIJP: 211
POCFR55 COXTENT: 41(b) 43(b) 5
KA: 01442.02
Ability to (a) predict the impacts ofthe following malfunctions or operations on the RF'IS; and (b) based
on those on those predictions, use procedures to correct, control, or mitigate the consequences of those
malfunctions or operations: E.oss of power to the RPIS
OBJECTWE: RODCS-3. I -K4
Given a copy of 'Technical Specifications 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 ACTION statements
DEVELOPMENT REFEKE:NCES: TS 3.1.3.3
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOIJRCE: NEW SIGNPFICANT1,Y MODIFIED DIRECT
BANK NUMREK FOR SIGNIFICANTLY MODIFIED !DIHECT: 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 while operating in
Mode 1 or 2, but u-ith both indications lost in Mode 3 the Reactor Trip Breakers are to be opened.
b. Plausible since this would he 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 Trip Breakers are to be opened.
r . Plausible since loss of indication of L N P I may lead to belief that SDM cannot be verified, which
would require Emergency Boratiou.
d (1. With both IIRPI indications inoperable in Mode 3 , 4 , or 5, TS requires that the Reactor Trip Breakers
be opened imrtiediately.
HCULTY AR'ALYSBS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICTJL'I'Y K4TING: 2
EXPLANATION: Knowledge of Tech Spec immediate action requirements in the event of a loss
of both DRPI indications
Post Validation Revision
IImis KR( Written Examination
Senior Reactor Operator
QUESTION: 10
.4 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. 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> as the Balance of Plant Operator during Mode 5 AKD 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as the
(ontrol 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 hfode 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
ANSWER:
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 Revision
Hatris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUEs'rION NUBIBER: I0 T%ER/GROUP: 3
80CFR55 CONIENI': 41(b) 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 be relieved during the shiti turnover
process
DEVELOPMENT REFERENCES: <)?vfM-OO1
REFERENCES SUPPLIED TO APPLICAWI': Xme
QUESTBOiV SOURCE: NEW SIGN%FICANI%,Y MODIFIE:I)
OR SIGNIFICANTLY 1C1ODIPIE:D / DIRECT:
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFICACI'ION (CORRECT AXSWER .v"d):
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 CO or BOP position. but only those hours
when the plant is above 200°F are acceptable.
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 200°F are acceptable.
c. 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 C:O or DOP position and this has the most
hours in the CO position, but only those hours when the plant is above 200"I" are acceptable.
/' 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 HOP position when the plant is above 2.00"F
DIE'FBCCLTY ANALYSIS:
COMPREHENSIVE / ANALYSIS F KNOW'1,EDGE / RECALI.
DKFFICULTY RATING: 2
EXPLANATION: Must recall requitxment for activating an inactive license from OMM-OOI
Post Validation Revision
IIarris NKC Writtan Examination
Senior Radctor Operatoi
QUESTION: 11
Following a loss of off-site power during recovery from a SGTR, the crew is required to
transition from EPP-019, Post SGTR Cooldown Using Steam Dump, to either:
e EPP-017, Post SGTR Cooldown Using Backfill
e EPP-018, Post SGTR Cooldown CJsing Blowdown
Which ofthe following describe how RCS and SG pressure contrd in EPP-OI 7 compares
to that in EPP-0 18?
a. e EPP-Oi7 maintains RCS pressure below the niptured SG pressure
e EPP-01 8 maintains KCS pressure below the ruptured S G pressure
b. e EPP-017 maintains RCS pressure below the ruptured S G pressure
m EPP-OI 8 maintains RCS pressure the same as the ruptured SG pressure
c. e EPP-017 maintains RCS pressure the same as the ruptured S G pressure
e EPP-018 maintains RCS pressure below the ruptured S G pressure
d. e EPE-017 maintains RCS pressure the sanie as the ruptured SCi pressure
e EPP-018 niaintains RCS pressure the same as the ruptured SG pressure
ANSWER:
b. e EPP-017 maintains RCS pressure below the ruptured SG pressure
e EPP-018 inaintains KCS pressure the same as the ruptured SG pressure
Post Validation Revision
Hams NRC Written Examination
Senior Reactor Operator
Data Sheets
QILJESTHON NUMBER: I 1 TIER/GROUP: i!I
10CFRSS CONTENT: 41(b) 43(b) 5
KA: 0OtJ038EA2.08
Ability to determine or interpret the following as they apply to a S W R : Viable a h n a t i v e s for placing
plant in safe condition when condenser is not available
OBJECTIVE: EOP-1.4- I
Describe the purpose of the following EOPs including the type of event for which they were designed and
the major actions perfornied
- EPP-0 I7
- EPP-0 18
- EPP-0 19
DEVELOPMENT REFERENCES: EPP-0 17
EPP-0 18
REFERENCES SUPPLIED TO APPLICANT: Nonc
QIJESTION SOURCE: SIGNIFICANTLY MODIFIED DIRECT
CAN11,Y MODIFIED DIRECT: ,! -3.4 010
NRC EXAM HISTORY: Harris 2002
DISTRACTOR .JUSTIFIC:ACTION (CORRECT ANSWER dd):
a. Plausible since EPP-017 maintains pressnre below ruptured SG pressure, but EPP-018 maintains
pressure the Same as the ruptured SG pressure.
d b. EPP-017 maintains pressure below S(i pressure to allow backfill from the SG to the RCS, while EPP-
018 maintains pressure the same as SG pressure to niininiize S G leakage.
c. Plausible since either EPP-0 14 or EPP-0 I 8 maiutains pressuix below SG pressure and either EPP-0 I7
or EPP-018 maintains pressure the same as SG pressure, hut this distracter has the correct condition
reveresed.
d. Plausible since EPP-0 I8 maintains pressure the same as the ruptured SG pressure, but P M17
maintains pressure below ruptured SG pressure.
n
DIFFICII1,TY ANALYSIS:
COMPREIIENSPVE / ANALYSIS
DIFFICULTY RATIXTG: 3
KNOW12EDGE IRECALI,
EXPLANATION: Knowledge of differeut mitigation strategies for EPP-017 and EPP-0 I8
Post Validation Revision
IIarris NRC Written Exsmination
Senior Reactor Operator
QI!ESTION: 12
A I.OCA occurred several hours ago. Only one ( i ) Containment Spray Pump is running
due to actions taken in EPP-0 12, Loss of Emergency Coolant Recirculation.
A transition has just been made to FRP-J. 1, Response to High Containment Pressure.
Containment Pressure is 14 psig.
Whish of the following actions should be taken?
a. Start the second Containment Spray Pump if Containment pressure docs NOi
decrease below 10 psig before exiting FRP-.I. 1.
b. Start the second Containment Spray Pump since pressure is ahove 10 psig.
C. Continue operation with one Containment Spray Pump regardless of any increase
in Containment pressure.
d. Continue operation with one Containment Spray Pump unless Containment
pressure begins increasing, then start the second pump.
ANSWER:
c. Continue operation with one Containnlent Spray Pump regardless of any increase
in Containment pressure.
Post Validation Revision
Harris NRC Written Exanunation
Senior Reactor Operator
Data Sheets
QUESTION NUMBER I2 TIEWGRODP: 112
lOCPR55 CONTENT: 41(b) 43(b) 5
KA: WE13E42.2
Ability to determine and interpret the following as they apply to the (High Containment Pressure)
Adherence to appropriate procedures and operation within the limitations i ~the
i facilitys license and
amendments
OBJECTIVE: EOP-3.13-5
Given the following EOP steps, notes, and cautions, describe the assuciated basis: b. C N M I spray
operation (EPP-012 or FRP-J.l)
DEVELOPMENT REFERENCES: EOP-FRP-J. 1
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY B¶ODIFIED/ 1)IRECT: EIOP-3.13-R4 008
NRC EXAM HISTORY: None
DISTRACTOR JUSTPFICACTION (CORRECT ANSWVEK dd):
a. Plausible since this would be a normal action directed by FRP-J.1
&. Plausible since this would be a normal action directed by FRP-J. 1
4 c. EPP-012 directs the operators to run Containment Spray Pumps based upon Containment pressure and
Fan Cooler operation. These actions are taken to minimize RWST depletion. This configuration is to
he maintained even if FRP-J. I is itnplernented.
68. Plausible since woiild better serve the intent of EPP-0 12. but wuuld be contradictory to the inlenr uf
FRP-J. 1 which bas a higher priority concerning the operation ofthe Spray Pumps.
DIFFICULTY ANALYSIS:
COMPREHENSWE / ANALYSIS 0 ELVOWLEDGE / RECALL
DIFFLCULTY RATING: 3
EXPLANATION: Must compare the relative actions in the 2 procedures and make a judgement of
which condition takes precedent
Post Validation Revision
IIarris NRC Written Examination
Senior Reactor Operator
QUESTIQN: 13
During operation at 100%power, an inadvertent SI occurs on 'B' Train ONLY.
Which of the following actions is required?
a. Manually actuate SJ on 'A' Train and continue in PATH-1
b. Continue in PATH-I noting which 'A' Train ESF equipment is NOT running
c. Start ONLY the 'A' Train of E S I equipment for which the redundant 'B' 'Train
cyuipnient failed
d. Transition directly to EI'P-008, SI Termination
ANSWER:
a. Manually actuate SI on 'A' Train and continue in PATH-I
Post Validation Revision
IIarris NRC: Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 13 TIEWGROIJP: 2: I
10CFR55 CONTENT: 41@) 43(b) 5
ICI: 013.42.01
Ability to (a) predict the impacts ofthe following malfunctions or operations on the ESFAS; and (b)
based on those predictions, use procedures to correct. control, or mitigate the consequences of those
malfunctions or operations: LOCA
OBJECTIVE: IE-3. IO-K4
Describe the expected operator actions associated with an imminent RPS or ESFAS actuation
DEVELOPMENT REFERENCES: EOP User's Chide
REFERENCES SUPPLIED TO APPLICANT: None
QtJESTlON SOURCE: NEW SIGNIFKANTLY MODIFBED DIRECT
BANK NIMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: IE-3. IO-R4 001
NRC EXAM IIISTORY: Harris 2 0 0
DISTRACTOR JUSTIPICACTION (CORRECT ANSWER d'd):
4 a. Preferred method of manual actuation although it would be acceptable to start / reposition all
equipment which would be actuated regardless of the perceived need since diagnostics have not yet
been performed.
b. Plausible since only a single train actuation is analyzed, but efforts are to be made to initiate both
trains.
c. Piausible since starting equipment as needed would provide adequate protection, but since diahqIoStiCS
have not yet been completed the equipment required may not yet be known.
d. Plausible since one of the goals following an inadvertent SI is to terminate SI as soon as criteria arc
niet to prevent overfilling / pressurizing the RCS, but procedures are written assuming both trains
started.
ICULTY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE I RECALL
DIPFICI!LTY RATING: 3
EXPLANATION: Required knowledge of procedural requirements for a single train of ESF
actuation
Post Validation Revision
IIarris NKC Written r;xaniinatio,n
Senior Reactor Operator
QUESTION: 14
Given the following conditions:
- 1CS-235, Charging Line Isolation, was closed to establish a clearance boundary for
maintenance on ICs-238.
1CS-235 had to be manually torqued shut.
E 1CS-235 is a Limitorye SMB-OO!SR-OO motor-operated valve.
Prior to declaring lCS-235 operable after the clearance is removed, the valve must be I..
a. wrified to have the torque switch calibrated correctly.
b. stroked with the control switch.
c. monitored for seat leakage.
d. n~anuallystroked h l l open
ANSWER:
b. stroked with the control switch.
Post Validatioii Revisiun
Harris KRC Written Examination
Senior Reactor Operator
Data Sheets
QIJESTIQN NUMBER: 14 TIE:R/GRQUP: 3
IQCFR55CONTENT: 41(b) 43(b) 5
KA: 2.2.19
Knowledge of maintenance work order requirements
QBJECTIVE: PP-2.41
Identify the primary functions and explain the responsibilities of the Work Coordination Centre
DEVELOPMENT REFERENCES: OMM-0 14
REFERENCES SUPPLIED r8 APPLICANT: None
QITESTION SOURCE: NEW SIGNHEICANT1,Y MODIFIED DIRECT
BANK NUMBER FQR SIGNIFICANTLY MODIFIED / DIRECT: E00 028
NRC EXAM HISTORY: Harris 2000
DISTRACTOR JUSTIFICACHQN (CQRRECr ANSWER dd):
a. Plausible since the valve has been manually torqued onto the seat, but the requirement is that the valve
must he stroked electrically from the coutrol switch.
v b. ,411 Iiniitorque SMB-OOISB-00 motor operated valves, if manually operatrd, are required to be stroked
electrically from the control switch to he declared operable.
E. Plausible since over torqueing a valve may result i u seat leakage, hut the requirement is that the valve
must be stroked electricalty from the control switch.
d. Plausible since the valve \vas manually torqued clostU, hut the requirement is that the valve must he
stroked electrically from the control switch.
DIFFICULTY AXALYSIS:
COMPREHENSIVE / ANA1,YSIS KNQWI.EDGE 1 RECALL
DIFFICXJLIY RATING: 3
EXII,AMAIION: Knowledge of administrative post-work practices required
Post Validation Revision
Harris NRC Writtcn Examination
Senior Reactor Operator
QUESTION: 15
Given the following conditions:
0 Following 21 Reactor Trip and Safety Injection, a transition has eventually been made
to EOP-EPP-0 15, l.!ncontrolled Depressurization of All S t e m Generators.
e Both Main and Auxiliary Feed Flow have been isolated to all SGs.
a Directions have just been given to locally isolate steam flows from all SGs.
e SC; A pressure appears to have stabilized at approximately 100 psig, while the other
S G s have completely depressurized.
Which of the following actions should be taken?
a. Transition to FOP-EPP-014, Faulted SG Isolation, since this is indication that
SG A has been isolated.
b. Continue in FOP-EPP-01.5 and re-establish AFW flow to S G A at ininimuni
flow.
c. Transition to EOP-PATH-2 if local radiation surveys indicate primary-to-
sccotidary leakage is occurring.
d. Iransition to FOP-EPP-008, SI Termination, to prevent overpressurizing the
RCS.
ANSWER:
c. Transition to EOP-PAIH-2 if local radiation surveys indicate primary-to-
secondary leakage is occurring.
Post Validation Rcvision
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESllQX NUMBER: 15 'P'IEWGROUR lil
IOCFR55 CONTENT: 41(h) 43(b) 2
Lk 000040G2.1.32
Ability to explain arid appiy all system limits and precautions. (Stearn I i n e Rupture - Excessive Heat
Transfer)
OBJECTIVE: EOP-3.9-7
Given a step, caution. or note from an emergency procedure, state its purpose
DEVELOPMENT REFERENCES: EUP-EI'P-0 15
REFERENCXS SUPPLIED TO APPLICANT: None
QUESTION SOUI1CE: NEW SIGN1FIICANTL.Y MODIFIED DIRECT
HANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NHC EXAM HISTORY: None
DISTRACTOR JUSTPFPCACI'ION (CORRECT ANSWER d'd):
a. Plausible since once a SG is confirmed to be isolated in FFP-OLS, a foldout page item directs a
transition to EPP-014.
h. Plausible since without indications of a SG tuhe leak, actions would be taken to remain in EPP-0 I5
and mainhin feed Row at minimum.
4 e. A S G may be suspected to be ruptured if it fails to d q out following isolation of feed flow. Local
checks for radiation can be used to confinn p r i n ~ a i y - t o - ~ ~ c o i ileakage.
da~-
d. Plausible since a desired goal after isolating a faulted SG is to terminate SI as soon as conditions are
met to prevent overfilling and overpressmizing the RCS.
DIFFICULTY ANALYSIS:
CQMPREIIENSPVE / ANALYSIS KYOWLEDGE i RECALL
DIFFICIjLTY RATING: 3
EXPLANATION: Must analyze the cause of the failure of the SG to depressurize and then
determine thc correct actions based on the analysis.
Post Validation Rwision
Harris NRC Written Exanlinetion
Senior Reactor Operator
QUESTION: 16
The unit has tripped due to a I D C X and ESF equipment has failed to start. As a result,
EOP-FRP-C.2, Response to Degraded Core Conling, has been entered.
A depressurization 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 fi,r Integrity
has occurred.
Which of the following actions should be taken?
a. Immediately transition to EOP-FRP-P. 1, Response to Imminent Pressurized
?herma1 Shock Conditions
b. Stop the YG depressurization and, if the red path does not clear, transition to EOP-
FKP-P. 1 . Response to Imminent Pressurixd Thermal Shock Conditions
c. Complete EOP-FRP-C.2 and then transition to EOP-FW-P. 1, Response to
Imminent Pressurized Thermal Shock Conditions. if the red path still exists
d. Complete the SKf depressurization and then transition to EOP-FRP-P. I, Response
to Imminent Pressurized Thermal Shock Conditions, if the red path still exists
ANSWER
c. Complete EOP-FRP-C.2 and then transition to EOP-FIIP-1. i . Response to
Imminent Pressurized Thennal Shock Conditions, if the red path still exists
Post Validation Revision
Harris NKC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NIJMBER: 16 TIEWGROUP: Ii2
10CPK55 CONTENT: 41(b) 43(b) 2
KA: WE06(i?. I .32
Ahiiity to explain and apply all system limits and precautions. (Degraded Core Cooling)
OBJECTIVE: EOP-3.104
Given the following EOP steps, uotes, and cautions, describe the associated basis
g. Stopping SG depressurization at 80 p i g (C.2)
DEVELOPMENT REFERENCES: EOP-FKP-C.2
REFERENCES SUPPLIED TO APPLICANT: None
Qt!ESTIOX SOURCE: NEW SIGNBFICANFLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICAYlZY MODIFIED / I9IRECT: New
NRC EXAM HISTORY: None
DISTRACTOR .IUSTIFICACTION (COIPRECr ANSWER .Id):
a~ Plausible since the red path for integrity has a higher priority than the orange path that caused entry
into EOI-FRP-C:.2, hut under thsse particular conditions a transition should not occur until completion
of the EOP-FRP-C.2.
h. Plausible since the red path for integrity has a higher priority than the orange path that caused entry
into EOP-FRP-C.2, but under these particular conditions a transitinn should not occur until completion
ofthe EOP-FRP-C.2.
d E. During the depressurization, a red path may occur due to injecting the accumulators. A transition
should not be made until the entire procedure has been completed.
d. Plausible since the red path for inte~grityhas a higher priority than the orange path that caused entry
into EOP-FIW-C.2, hut under these particular conditions a transition should tint occur until completion
of the EOP-FKIC.2.
DIFFICULTY ANALYSIS:
COMPKEIIENSIVE / ANALYSIS KNOWLEDGE / RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must analyze plant conditions to determine that the cause ofthe red path is the
depressurization and that, under these specific conditions, an immediate
transition is not wairanted
Post Validation Kcvision
Harris NKC Written Examination
Senior Reactor Operator
Given the following conditions:
e The unit is in Mode 3.
a Instrument Buses IUP-1B-SI1 and 1DP-IB-SIV are both de-energized.
e Maintenance reports that Instrument Bus IDP-IB-SI1 is ready to be re-energized
In order to prevent an inadvertent Safegaards Actuation, which of the following must be
verified prior to re-energizing the bus and why?
a. Train A Logic Input Error Inhibit must be verified to be in IIWIBIT due to the
proper coincidence for an actuation being available
b. Train A Logic Train Output must be verified to be in TESI to prevent an
inadvertent Safeguard Actuation due to the loss of the SI BLOCK Signals
c. Train B Logic Input Etror Inhibit must be verified to be in INHIBIT due to the
proper coincidence for an actuation being available
d. Train B 1,ogic Train Output must be verified to be in TEST to prevent an
inadvertent Safeguard Actuation due to the loss ofthe SI BLOCK Signals
ANSWER:
d. Train 1%Logic Train Output must be verified to be in T E S I to prevent an
inadvertent Safebmard Actuation due to the loss ofthe SI BLOCK Signals
Post Validation Revision
Harris NRC' Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 14 TIEWGROUP: 21 1
10CFRS5 CONTENT: 41(h) 43(b) 2
KA: 06262.2.22
Ktiowledge of limiting conditions for operations and safety limits. (.4C Electrical Ilistribution)
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 SSFS train.
I)E:VELOPMENT REFERENCES: OP-156.02
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANT1,Y MODIFIED i DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JUSTIFPCACTION (CORRECT ANSWER d'd):
a. Plausible since the loss of both trains of power will provide the proper coincidence, hut power must be
available to the output relays to actuate. Placing the input error inhibit in INHIBIT at this time will
not prevent an actuation since the logic is already made up. Also the incorrect Train.
h. Plausible since the loss of both trains of power causes the SI BIock signals to he lost and when either
of the supplies is restored, power will be available to the output relays to cause an actuation. however
this occurs on Train 'H' for this event.
c. Plausible since the loss of both trains of power will provide the proper coincidence, but power must be
available to the output relays to actuate. Placing the input error inhibit in INHIHI'I' at this time will
not prevent an actuation since the logic is already niade up.
.! d. The loss ofboth trains of power causes the SI Block signals to he lost. When either ofthe supplies is
restored, power will be available to the output relays to cause an actuation.
DIFFICULTY ANALYSIS:
COMPREIIENSIVE / ARALYSIS KNOWLEDGE i RECALL
DIFFICULTY RATING: 3
EXPLAYATION: Must determine train of SSPS affected by the loss of power and then analyze the
effect of partially restoring power
Post Validation Revision
Harris NKC Written Examination
Senior Reactor Operator
QUESTION: 18
The I Jnit-SCO arid Superitltendent-Shift Operations are discussing invoking
I OCFR51).54(x) during the intplernentation of the Emergency Operating Procedures due
to a condition arising which is NOT addressed by the procedures or Technical
Specifications.
Which of the following conditions must be met when invoking 1OCFR50.54(x)?
a. The action must be approved by an additional Iicensed Senior Reactor Operator
when the action is necessary to prevent equipnient damage.
b. The action must he approved by the Superintendent-Shift Operations prior to
taking the action.
c. The KRC 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 Operalions prior to
taking the action.
Post Validation Kevision
Harris NRC Written Examination
Senior Keactor Operator
Data Sheets
QUESTION NUMBER: 18 TIEWGROUP 3
10CFR55 CQNTENT: 41(b) 43(b) 3
KA: 2.2.10
Knowledge of the process for determining if the margin of safety, as defined in the basis of any technical
specification is reduced by a proposed change, test or experiment
OBJECTIVE: P1'-2.0-S2
LTS'I the actions required by the individual who authorizes a deviation from the Technical Specifications
or license conditions
DEVELOPMENT REFERENCES: PRO-KGGC-0200
REFERENCES SUPPLIED TO APPLICANT: None
QtJESTION SOURCE: NEW SIGNHFICtlRTLY MODIFBED DIRECT
BANK NIIMBER FOR SIGXIFICANTLY MODIFIED i DIRECT: INPO 233 I 8
NRC EXAM IIISTORY: None
DISIRACTOR JUSTPFICACTION (CORRECT ANSWER d'd):
a. Plausible since IOCFK5054(x) requires that a licensed SRO approve any actions which deviate from
license conditions prior to performance, but the actions must be to protect the health and safety ofthe
public.
t' b. The minimum level of approvai per PRO-NGGC-0200 is the Superintendent-Shift Operations. but it
can be approved by any personnel holding an SRO license above this position also.
6. Piausible since the NRC must be notified, but the notification requirenients 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 SRO license, but the
actions must he to protect the health and safety ofthe public.
DIFFICULTY ANALYSIS:
COMPREIIENSIVE: / ANALYSIS KNO\I'LF:DGE i RECALL
DIFFICULTY RATING: 2
EXPIANATIOX: Requires knowledge of requirements for process of performing actions nnt
described in any licensing hasis documents.
Post Validation Revision
Harris NRC Written Exanimation
Senior Reactor Operator
QUESTION: 19
Given the following conditions:
- Following a Loss ofAll Power, EDG IA-SA has been restarted and partially loaded.
A transition has been made to EOP-EPP-003, 1x)ss of All AC Power Kecovery with
SI Kequired.
- EDC 1.4-SA is currently loaded to 4.5 MWe and 3.5 MVAR.
K7hich of the following would result in an LJNACCEPTABLE loading condition for EDG
a. Pick up an additional 0.5 h4.1U7e
- Pick up an additional 0.1 MVAR
b. * Pick up an additional I .O MWe
e Pick up an additional 0.5 MVAK
c. * Pick up an additional 1 .S MWe
e Pick up an additional 1.O MVAR
ti. e Pick up an additional 2.0 MWe
- Pick up an additional 1.2 14.IVAR
ANSWER:
c. e Pick up an additional 1.5 MVVe
- Pick up an additional I .O MVAR
Post Validation Kevisioii
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NIMREIP: 19 TIEWGROUP: 1!1
10CFR55 CONTENT: 416b) 43(b) 5
KA: 00005OAA2.14
Ability to detennine and interpret the following as they apply to the Loss of Offsite Power: (lperational
status of EDiGs (A and 13;)
OEJECTIVE: EOP-3.7-6
Given a step, caution, or note from EOP-001, EOP-002, or EOP-003, state its purpose
DEVELOPMENT REFERENCES: OP-l S5, Attachment 9
F:OP-EPP-003
REFERENCES SUPPLIED TO APPLICANT: OP- 155, Attachment 9
QUESTION SOURCK: NEW SIGNIFICANTLY MODIFPED DIRECT
RANK NUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NRC EXAM HISTORY: None
DISTRACTOR JILJSTIFICACT1[ON (C33RRECT ANSWER .Id):
a. Plausible since new loading will be 5.0 MWe and 3.6 MVAK. which is just within acceptable limits.
b. Plausible since new hading will he 5 3 MWe and 4.0 MVAR, which is just within acceptable limits,
d E. New loading will be 6.0 MWe and 4.5 MVAR, which is outside acceptable limits.
at. Plausible since new loading will be 6.5 MWe and 4.7 MVAK. which is just within acceptable limits,
~DIFFICIJLTY ANALYSIS:
~~
COMPREIIENSIVE / ANALYSIS ICVOWLEDGE /RECALL
DIFFICULTY RATING: 3
EXPLANATION: Must analpz EDG operability curve to determine whether additional MWc and
MVAR loading is urithin acceptahle limits
Post Validation Revision
IIarris NRC Written Examination
Senior Reactor Operator
QUESTION: 20
h reactor trip occurred due to a loss of offsite power. The plant is being cooled down on
RIIR per EPP-006. Natural Circulation Cooldown with Steam Void in Vessel with
0 KCS cold leg temperatures are 190°F.
0 Steam generator pressures are 50 psig.
- RVLIS upper range indicates greater than 100%.
- Three CRUX4 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 2M"F prior to cotnplete RCS
depressurization.
c. RCS and SG temperatures are equalized prior to any subsequent RCI' restart
d. RCS temperatures do not increase during the required 29 hour3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> vessel soak period.
AYVSWER:
b. all inactive portions of thc RCS are below 200°F prior to complete RCS
depressurization.
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QIIESTION NI!NBER: 20 TIEWGROUP: 112
lOCFR55 (IONTENT: 41(b) 13(b) 2
MA: WE09(i2.1 3 2
Ability to explain and apply all system limits and precautions. (Natunl Circulation Operations)
OBJECTIVE: EOP-3.8-2
Demonstrate the below-assumed operator knowledge from the SHNPP Step Deviation Document and the
WOG ERGS that support perfonnance of EOP actions: Iieterniining that upper head and SG U-tube
temperatures are below 200 "F
DEVELOPMENT REFERENCES: EOP-EPP-006
REFERENCES SUPPLIED TO APPLICANT: None
QCESTION SOURCE: NEW SIGNHFPCANTLY IIfODIFIEB DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECl: EOP-3.8 006
NRC EXAM HISTORY: None
DHSTRAQ:TORJUSTKFICACTION (CORRECT ANSWER d'd):
a. Plausible since this action would have been performed in this procedure, hut niust be completed prio~
to depressurizing the RCS below 19(m psig:.
./ h. S G pressure above 0 psig indicates that the SGs are above 200°F. Depressurizing the 1 - 3 3 undcr this
condition will result in additional void formation in the SG u-tubes.
e. Plausible since RCP operation throughout NC Cooldown is desirable, but will not be performed at this
point in the procedure.
d. Plausible since a soak period is addressed, but only if continued operation of CKIIM fans had not been
maintained.
I)IFFICUI,TY ANALYSIS:
COMPREHENSIVE / ANALYSIS KNOWLEDGE /HECALL
DIFFICULTY RATISG: 3
EXPIANATION: Must analyze the conditions and detennine that the entire RCS is not below
200°F and the effect of depressurizing under these conditions.
Post Validation Revision
FIarris NRC Wiittcii Examination
Senior Reactor Operator
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 WIPE and the entry does NOT require specific Site Etnergency
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 Site Emergency
Coordinator authorization.
d. 25 Rem E D E and the entry requires specific Site Emergency Coordinator
authorization.
ANSWER:
b. 10 Rem TEDE and the entry requires specific Site Emergency Coordinator
authorization.
Post Validation Revision
Hanis NRC Written Examination
Senior Reactor Operator
Ihta Sheets
QUESTION NUMBER: 2 I TIEMUGROUP: 3
10GPR55 CONTENT: 41@) 43(h) 4
KA: 1.3.7
Knowledge ofthe process for preparing a radiation work pemiit
OWJEClIVE: EP2O-2h
Identify the tyyes of prntcctive actions for HNP personnel (both on and off-site) and who is rcspniisible
for directing them.
m v E L o m E w r REFE.RENCES: PEP-330
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICAN'IT,Y MODIFIED DIRECT
BANK SUMBER FOR SIGNIFICANTLY MODIFIED I DIRECT: New
NHC EXAM HISTORY: None
DISiZkPCTOR SUSTIFICAGTION (CORRECT ANSWER J'd):
a. Plausible since I O rem 'TEDE for protecting valuable company property. hut S - S O approval is
required.
9' h. Exposure is limited to 10 rem TEDE is the limit required for this activity and S - S O approval is
required.
c. Piausihle since 25 rem THIF is the limit required for lifesaving efforts. hut the h i t to protect
equipment atid property is LO rem 'TEDE.
d. Plausible since 25 rem TEDE is the limit required for lifesaving effoits, but the litnit t n protect
equipment and property is 10 rem TEDE.
ICIJLTY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE I RECALL,
DIFFICULTY RATING: 3
EXPLANATION: Requires knowledge of the emergency exposure limits and approval
requirements
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
QIJESTION: 22
Given the following conditions:
e Power is cui-rently at 32% during a plant startup.
Instrument Rut: IDP-IR-SIV deenergized as a result o f a fault in PIC' CAR-4.
0 PIC CAR-4 has been isolated from Instrument Bus SIV and will be deenergized for
approximately eight (8)hours while repairs are being made.
Which of the following actions must be taken?
a. Place ail PIC CAB-4 Reactor Trip instruments in the tripped condition
b. Place all PIC CAB-4 ESI: instrutnents in the tripped condition
c. Place all MFW Regulating Valves in MANUAL
d. Perform a plant shutdown
ANSWER:
d. Perfotm a plant shutdown
Post Validation Revision
Hairis NRC Written Esaniination
Senior Reactor Operator
Data Sheets
QUESTION NL1MBE.R 22 TIERGROUP: Iil
IOCE'R55 CONTEXT: $I(b) 43(h) 2
KA: 000057G2.2.22
Knowledge of limiting conditions for operations and safety limits. (Loss of Vital C \i Instrument I h s )
0WEC:TICT: AOP-3.24-4
Uetemiine the following: a. Consequences of the loss of all power to PIC C A B 4
DEVELOPMENT REFERENCES: AOP-024
TS Table 3.3-3, pg 3-IX and 3-27
TS 3.0.3, pg 0-1
REFERENCES S u P r L I m TO APPLICANT: xone
QUESTION SOURCE: 0 NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED I DLRECT: AOP-3.24-K4 00 1
NRC EXAM HISTORY: None
D1STIIAC:TQ)RJUSTIFPCACTION (CORRECT 4NSWER t'"d):
a. Plausible since instrunlent failures require bistables tripped, but they are deenergized to actuate and
are already tripped since no power is available.
b. Plausible sitice instnnnent faiinrees require bistables tripped, but they are deenergizd to actuate and
are already tripped since no power is available.
c. Plausihle since this is the immediate operator action for a loss of Instrument Bus SIII, not SIV
4 d. Loss of all power to PIC' CAB-4 will result in 3 bistable channels of Steam I i n e Pressure becoming
inoperable. The 'ISaction is io trip the bispables within one hour, but the bistables are energized to
actuare. \Vithout power awilable, this action cannot he perfoimed and TS 3.0.3 becomes applicable.
ICULTY ANALYSIS:
COMPREIIENSIVE I ANALYSIS 0 KNOWLEDGE /RECALL
DIFFICULTY RATING: 4
EXPLANATION: Must recognize that energized io actuate histables cannot be placed in tripped
condition without power, thus an entry into 1's 3.0.3 is required. and must
determine the required TS 3.0.3 actions
Post Validation Revision
Harris NRC: Written Examination
Senior Reactor Operator
QUESTION: 23
During the performance of EOP-PATH-2, the STA reports that the following two (2)
YELLOW path Critical Safety Function Status Trees (CSFST) exist:
- Integrity
e Heat Sink
Which of the following describes how these YELLOW paths are to be addressed and i or
impletnentcd?
a. Both must be addressed and implemented, with Heat Sink having a higher priority
than Integrity, as soon as EOP-PA?-2 actions are completed provided IICIother
higher priority CSFSI conditions exist
b. Both must be addressed, but implemented at the discretion of the Superintendent-
Shift Uperations, prior to exiting from the EOP network
c. Both must be addressed and implanented, with Heat Sink having a higher priority
than Integrity, prior to exiting from the EOP network
tl. Both must be addressed. but implemented at the discretion ofthe Superintendent-
Shift Operations, as soon as FOP-PATII-2 actions are completed provided no
other higher priority (SFST conditions exist
ANSWER:
h. Both must he addressed, but implemented at the discretion of the Superinlendcnt-
Shitt Operations, prior lo exiting from the FOP network
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER 23 'I'IEWGROUP: 3
Iiii IMPORTAXCE: RO sa0 4.0
IOCFR55 CONTENT: 4B(b) 43Bb) S
Kh: 2.4.22
Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations
OBJECTIVE: EOP-3.19-2
Describe Control Room usage o f status trees as it relates to the following
a. Priority of status trees
b. Rules of usage
DEVELOPMENT REFERENCES: EOP User's Guide
REFERENCES SUPPLIED TO AQPLICANT: None
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: New
NRC.' EXAM HISTORY: None
DISTRACTOR JUSTIPICACTION (CORRECT ANSWER d3d):
a. Plausible since they are to be addressed, but only prior to leaving the EOP network and are not
required to he implemented.
4 b. All YELI~.OW-conditionCSFSTs should be addressed prior to exiting the EOP network. However, the
operator is allowcd to decide if and when to implement. and whether to complete any YELLOW-
condition I'KP.
E. Plausible since they are to be addressed, but only prior to leaving the EOP network and are not
required to be implemented.
d. Plausible since they are to be addressed, but only prior to leaving the EOP network and an' not
required to be implemented.
D I B ; w x L r Y ANALYSIS:
COMPREIIENSIVE / ANALYSIS KNOWLEDGE / REXALI,
I)%FFICUI,TYRATING: 2
EXPLANATION: Knowledge o f the iu$ementation criteria for yellow CSFSTs as directed by
plant procedures
Post Validation Revision
Harris NRC Written Examination
Senior Reactor Operator
QUESTION: 24
Following a loss of a11 AC 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 t)C ioad shedding occurs within 30 minutes of the loss of all
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 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 30 minutes of the loss of all
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 ofthe loss of all
A S power
ANSWER:
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 tninutes of the loss of all
AC' power
Post Validation Revision
Harris NKC Written Examination
Senior Reactor Operator
Data Sheets
QUESTION NUMBER: 24 TIEWGROUP: lil
10CF1355 CONTENT: 41(b) 43(b) 2
KA: 00005862.2.25
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 EOP-003, state its purpose
DEVEL.0PMENT REFERENCES: Tech Spec Bases 3.8.2, pg 8-2
EOI'-EPP-00 I
ADEL-1.P-2.6
REFERENCES SIJPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFIED DIRECT
BANK NUMBER FOR SIGNIFICANTLY IWODIFIED / DIRECT: AIlEL2-6-S I 00 I
NRC E:XAM HISTORY: None
DIS'FWAC'L'OR JUSTIFICACTION (CORRECT ANSWER +(I):
a. PLausihlr since this is the time limit which requires actions being taken in accordance with Technical
Specifications, hut the design oftlie hatteries is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
h. Plausihle since this is the time limit u-hich 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 />.
c. Plausible since the design ofthe 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.
v' d. Batteries are designed to can); 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
carry bus or charge hatter)., assuming that required load shedding occurs withiu I hour.
ICUL'FY ANALYSIS:
COMPREHENSIVE I ANALYSIS KNOWLEDGE / RECALL
UIFFICGLTY RATING: 3
EXPLANATION: Knowledge of tech spec basis arid design of safety-related batteries
Post Vdidation Rrvisiou
Harris NKC Written Examination
Senior Reactor Operator
QUESTION: 25
Which of the following actions would be INAPPROPRIATE to pcrforni prior to
direction in an EQP?
a. Isolating AFW flow to a single faulted S G
h. Throttling AFW flow to control a rupkred SG level within the required level band
c. Securing a ('SIP to prevent overfilling the pressurizer following an inadvertant SI
d. Shutting the MSIVs tu 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 Kcvision
IIarris NRC Written Examination
Senior Reactor Operator
Data Sheets
n
QrJESTION NUMBER: 25 TIERKROUP: 5
1OCFR55 CONTENT: 41(h) 43fb) 5
KA: 2.4.14
Knowledge of general guidelines for EOP flowchart use
OBJECTIVE: FOP-LP-3.19-1
Descrihe Control Room usage of the EOP network as it relates to the following: a) Ierforniing steps out
of sequence
DEVELOPMENT REFERENCES: EOP Ksers Chide
REFERENCES SUPPLIED TO APPLICANT: None
QUESTION SOURCE: NEW SIGNIFICANTLY MODIFPED DIRECT
BANK NUMBER FOR SIGNIFICANTLY MODIFIED / DIRECT: EOI-3.19-RI 0 18
NRC EXAM HISTORY: None
DISTRACTOR JUSTIPICACTKON (CORRECT ANSWER dd):
a. Plausible siuce this is a numbered step in PATH-I which are normally required to he performed in
sequence, but the EOP Users Guide addresses this as being acceptable.
b. Plausible since this is a numbered step in PATK 1 which are normally required to he performed in
sequence, but the EOP Users Guide addresses this as being acceptable.
I
V 6. Perfotming steps out of sequence is allowed, but must be done with caution to prevent masking
symptoms or defeating the intent ofthe EOI being used. Although terminating SI early might he
beneficial to prevent filling the pressurizer if the only event is a spurious SI, this may result in further
degradation of the plant if another undiagnosed event is in progress.
d. Plausible since this is a numhered step in PATH-1 which are normally required to be perfomled in
sequence, but the EOP f.Jsers Guide addresses this as being acceptable.
I)IFFICUI,TY ANALYSIS:
COMPREIIENSIVE / ANALYSIS KNOWLEDGE /RECALL
1)IFFICULTY RATING: 3
EXP1,ANATION: Must differentiate between those actions which could potentially result in
degradation ofthe plant iftaken out o f sequence and those actions which would
likely have little impact on the operators abilities to diagnose other events.
Post Validation Revision