ML041170063

Feb-March 2004 Exam 50-400/2004-301 Draft SRO Written Exam

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.

RVtIS 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.

b.

A PKZ steam space break has occurred and core heat removal is ADEQUAI'E

A PRZ steam space break has occurred arid core heat removal is INADEQUAIE

An RCS hot ieg break has occurred and core heat removal is ADEQUATE

An RCS hot leg break has occ.urred and core heat removal is INADEQL!A?'E

c.

d.

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

KA IMPORTANCE:

RO

SRO

4.1

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,PED TO APPLICANT:

None

OUESTYON SOIJRCE:

NEW fl

SKGNIFICANT1.Y MODIFIED n

DIRECT

LA

L A

bl

.

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.

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.

c.

DIFFICULTY ANALYSIS:

C0iW"IEPIENSIVE / ANALYSIS

DIFTICLJ1,TY RATIXG:

3

EXPLANATION:

KNOWLEDGE /RECALL

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

h.

e

e

c. *

d. *

'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

Twenty minutes following a Main Feedwater Pump trip, Control Rods are

determined to he helow the rod insertion limit

Control the reactivity transient associated with an inadvertent dilution

During a reactor startup, the Reactor achieves criticality with Bank C rods at

Control the reactivity transient associated with a stearn line break

105 steps

During a reactor startup, the Reactor achieves critic.aIity with Bank C rods at

Control the reactivity transient associated with an inadvertent dilution

105 steps

AKSWEW:

c.

e

During a reactor startup. the Reactor achieves criticality with Bank C rods at

Control the reactivity transient associated with a steam line break

IO5 steps

Post Validation Revision

Harris NRC Written Examination

Senior Reactor Opcrator

Data Sheets

QUESTION NUMBER: 2

TIEWGROUP:

li2

KA IMPORTANCE:

RO

SRO

3.7

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 [3 DIRECT

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 stem line break. However, it is not required during transient conditions, allowing

the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore rod position.

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.

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.

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.

d c.

d.

ICKJLTY ANALYSIS:

COMPREHENSIVE / ANALYSIS

UIFFICULIY RATIXG:

2

EXPL,AN,4IION:

KNOWLEIIGE i RECALL

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

• CSIP 1.4-SA is operating.

o

Ihe plant has been operating at I@@% power for the past three (3) months.

CSIP 1B-SB has just been restored to a normal alignment following maintenance on

the pump impeller.

When CSIP 1B-SR is started the operator notes that suction pressure appears nornial,

while discharge pressure, discharge flow, and pump current are oscillating.

o

Which ofthe following is the most likely cause of these CSIP 1 B-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

KA IMPORTANCE:

RO

SRO

3.8

IOCFRS CowrmT:

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

REFERENCES SUPPLIED TO APPLICANT:

None

SOEK 97-1

~

QUESTION SOURCE:

NEW

SIC~MFICANTLY MODIFIED

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 V d h C

would cause discharge pressure to be high.

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 is rneasu~ud

dowstreatn of the recirc valve. hut discharge pressure and current would be at or near normal.

e.

DIFFICULTY ANALYSIS:

COMPREEIENSIVE i ANALYSIS

DIFFICULTY RATING:

3

EXPLANATION:

KNOWLEDGE i RECALL

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

e

The unit is operating at 100% power, with C.;ontrol Bank D rods at 215 steps.

ALB 13-7-1, ROD CONIROI, URGENT ALARM, is in AIAKM due to a failure in

Power Cabinet I AC.

Rod Control is in MAN.

A turbine trip occurs, but the Reactor f'ails to trip either automatically or manually.

o

e-

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:

2 2

KA IMPORTANCE:

RO

SRO

4.0

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.

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 or 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.

c.

DIFFICULTY ANALYSIS:

Q~OMPRFXBENSIVE

/ ANALYSIS

DIFFICULTY RATING:

3

EXPLANATION:

KNOWLEDGE I RECALL

Must aualyze the effect of an urgent rod control failure aid 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-IOI JRS 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

I 0

14

off

12

12

12

2

14

12

14

10

off

11

3

off

off

off

13

I 1

14

4

I 1

13

14

off

I I

12

Which of the operators would be permitted to work a 12-hour shift on Day 7 W'IIHO1iT

requiring permission to exceed nonnal owtime 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

KA IMPORTANCE:

RO

SRO

4.0

lQCFR55 CONTENT:

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~ed 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 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

DIFFICULTY RATING:

3

EXPLANATION:

KNOWLEDGE I RECALL

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

e

A Reactor Trip with SI occurs.

The operators perform the immediate action steps, verify ECCS flow, and check

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 SG pressures are between 300 psig and 350 psig.

AFW Oow.

e

Which of the following actions is to be taken?

a.

b.

c.

d .

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.

(ontintie in FOP-FRP-11. I since EOP-FKP-H.1 has a higher priority than PATH-I

and initiate KCS feed and bleed.

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.

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

EL4 IMPORTANCE:

RO

SRO

4.0

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.

Since RCS pressure is less than S a.

I

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.

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.

c.

I c u c r Y ANALYSIS:

COMPREHENSIVE i ANALYSIS

DIFFICULTY RATING:

2

EXPLANATION:

mOWLEDGE / RECALL

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:

e

The plant is in Mode 3 with all Shutdown Rods withdrawn.

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

KA IMPORTANCE:

RO

SRO

3.6

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 ofa loss ofa 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.

With both IIRPI indications inoperable in Mode 3,4, or 5, TS requires that the Reactor Trip Breakers

be opened imrtiediately.

d (1.

HCULTY AR'ALYSBS:

COMPREHENSIVE / ANALYSIS

DIFFICTJL'I'Y K4TING:

2

EXPLANATION:

KNOWLEDGE / RECALL

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

KA IMPORTANCE:

RQ

SRO

3.8

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-OO 1

REFERENCES SUPPLIED TO APPLICAWI':

X m e

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:

F KNOW'1,EDGE / RECALI.

COMPREHENSIVE / ANALYSIS

DKFFICULTY RATING:

2

EXPLANATION:

Must recall requitxment for activating an inactive license from OMM-OO I

Post Validation Revision

IIarris NKC Writtan Examination

Senior Radctor Operatoi

QUESTION:

1 1

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

e

Which ofthe following describe how RCS and SG pressure contrd in EPP-OI 7 compares

to that in EPP-0 18?

EPP-017, Post SGTR Cooldown Using Backfill

EPP-018, Post SGTR Cooldown CJsing Blowdown

a.

e

e

EPP-Oi7 maintains RCS pressure below the niptured SG pressure

EPP-01 8 maintains KCS pressure below the ruptured S G pressure

b.

e

m

EPP-017 maintains RCS pressure below the ruptured S G pressure

EPP-OI 8 maintains RCS pressure the same as the ruptured SG pressure

c.

e

e

EPP-017 maintains RCS pressure the same as the ruptured S G pressure

EPP-018 maintains RCS pressure below the ruptured S G pressure

d.

e

e

EPE-017 maintains RCS pressure the sanie as the ruptured SCi pressure

EPP-018 niaintains RCS pressure the same as the ruptured SG pressure

ANSWER:

b.

e

e

EPP-017 maintains RCS pressure below the ruptured SG pressure

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

KA IMPORTANCE:

RO

SRO

4.4

10CFRSS CONTENT:

41(b)

43(b)

5

KA:

0OtJ038EA2.08

Ability to determine or interpret the following as they apply to a SWR: Viable ahnatives 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.

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 SG 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 M 17

maintains pressure below ruptured SG pressure.

d b.

DIFFICII1,TY ANALYSIS:

n

COMPREIIENSPVE / ANALYSIS

KNOW12EDGE IRECALI,

DIFFICULTY RATIXTG:

3

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.

b.

C.

d.

Start the second Containment Spray Pump if Containment pressure docs NOi

decrease below 10 psig before exiting FRP-.I. 1.

Start the second Containment Spray Pump since pressure is ahove 10 psig.

Continue operation with one Containment Spray Pump regardless of any increase

in Containment pressure.

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

MA IMPORTANCE:

RO

SRO

3.8

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 ~ i

the facilitys license and

amendments

OBJECTIVE:

EOP-3.13-5

Given the following EOP steps, notes, and cautions, describe the assuciated basis: b. CNMI 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

DIFFLCULTY RATING:

3

EXPLANATION:

0 ELVOWLEDGE / RECALL

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 ESI 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

KA IMPORTANCE:

RO

SRO

4.6

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.

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.

c.

ICULTY ANALYSIS:

COMPREHENSIVE / ANALYSIS

DIPFICI!LTY RATING:

3

EXPLANATION:

KNOWLEDGE I RECALL

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.

1 CS-235 is a Limitorye SMB-OO!SR-OO motor-operated valve.

E

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~anually stroked hll 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

KA IMPORTANCE:

RQ

SRQ

3.1

IQCFR55 CONTENT:

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 iu 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.

Both Main and Auxiliary Feed Flow have been isolated to all SGs.

Directions have just been given to locally isolate steam flows from all SGs.

SC; A pressure appears to have stabilized at approximately 100 psig, while the other

SGs have completely depressurized.

e

a

e

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

K4 IMPORTANCE:

RO

SRO

3.8

IOCFR55 CONTENT:

41(h)

43(b)

2

Lk 000040G2.1.32

Ability to explain arid appiy all system limits and precautions. (Stearn Iine 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 SG 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 prin~aiy-to-~~coiida~-

leakage.

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 IDCX 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:

I i2

KA IMPORTANCE:

RO

SRO

3.8

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 pig (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.

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~grity has 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.

d E.

DIFFICULTY ANALYSIS:

COMPKEIIENSIVE / ANALYSIS

DIFFICULTY RATING:

3

EXPLANATION:

KNOWLEDGE / RECALL

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

a

e

The unit is in Mode 3.

Instrument Buses IUP-1B-SI1 and 1DP-IB-SIV are both de-energized.

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 TESI 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

KA IMPORTANCE:

RO

SRO

3.4

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

DIFFICULTY RATING:

3

EXPLAYATION:

KNOWLEDGE i RECALL

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 1 OCFR50.54(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.

The KRC must concur with the action to be taken prior to the action actually being

taken.

c.

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

MA IMPORTANCE:

RO

SRO

3.3

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 I8

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

DIFFICULTY RATING:

2

KNO\\I'LF:DGE i RECALL

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

1 A-SA?

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

Kri IMPORTANCE:

RO

SRO

4.6

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:

QUESTION SOURCK:

NEW

SIGNIFICANTLY MODIFPED

DIRECT

OP- 155, Attachment 9

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

RVLIS.

0

0

KCS cold leg temperatures are 190°F.

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

ICI IMPORTANCE:

RO

SRO

3.8

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:TOR JUSTKFICACTION (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

DIFFICULTY RATISG:

3

EXPIANATION:

KNOWLEDGE /HECALL

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 EDE 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

K.4 IlbIPORTANCE:

RO

SRO

3.3

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 IO rem 'TEDE for protecting valuable company property. hut S- SO approval is

required.

9' h. Exposure is limited to 10 rem TEDE is the limit required for this activity and S- SO approval is

required.

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 tn protect

equipment and property is 10 rem TEDE.

c.

ICIJLTY ANALYSIS:

COMPREHENSIVE I ANALYSIS

DIFFICULTY RATING:

3

EXPLANATION:

KNOWLEDGE I RECALL,

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

0

Power is cui-rently at 32% during a plant startup.

Instrument Rut: IDP-IR-SIV deenergized as a result ofa fault in PIC' CAR-4.

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

K A IMPORTANCE:

RO

SRO

4.1

IOCE'R55 CONTEXT:

$I(b)

43(h)

2

KA:

000057G2.2.22

Knowledge of limiting conditions for operations and safety limits. (Loss of Vital i\\C

Instrument Ihs)

0WEC:TICT:

AOP-3.24-4

Uetemiine the following: a. Consequences of the loss of all power to PIC CAB4

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)R JUSTIFPCACTION (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 Iine Pressure becoming

inoperable. The 'IS action 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

DIFFICULTY RATING:

4

EXPLANATION:

0 KNOWLEDGE /RECALL

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 IICI other

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

Both must be addressed and implanented, with Heat Sink having a higher priority

than Integrity, prior to exiting from the EOP network

c.

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 of 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-condition CSFSTs 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.

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.

E.

D I B ; w x L r Y ANALYSIS:

COMPREIIENSIVE / ANALYSIS

I)%FFICUI,TY RATING:

2

EXPLANATION:

KNOWLEDGE / REXALI,

Knowledge of 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

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

b.

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

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

d.

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

KA IMPORTANCE:

RO

SRO

3.7

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):

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.

a.

ICUL'FY ANALYSIS:

COMPREHENSIVE I ANALYSIS

UIFFICGLTY RATING:

3

EXPLANATION:

KNOWLEDGE / RECALL

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

KA IMFORlANCE:

RO

SRO

3 3

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.

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

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