ML20237B967

Forwards Listed Matl,For Review,Comment & Approval,Iaw Instructions in NUREG-1021, Operator Licensing Exam Stds for Power Reactors, (Interim Rev 8)

ML20237B967
Person / Time
Site:	Beaver Valley
Issue date:	04/28/1998
From:	Brooks R DUQUESNE LIGHT CO.
To:	Kenny T NRC
Shared Package
ML20237B958	List: ML20237B961 ML20237B967 ML20237B972
References
RTR-NUREG-1021 NUDOCS 9808200128
Download: ML20237B967 (110)

Text

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To: Tom Kenny USNRC From: Rich Brooks Duquesne Light Co. ph Date: April 28,1998

Subject:

Unit i Initial Written Exam Materials in accordance with the instructions in NUREG-1021, " Operator Licensing Examination Standards for Power Reactors,"(Interim Rev. 8) the following materials are being submitted to you for review, comment, and approval:

1. PWR SRO Examination Outline (Form ES-401-3)

2. Generic Knowledge and Abilities Outline (Form ES-401-5)

3. Written Examination Quality Assurance Checklist (Form ES-401-6) 4.100 Written Exam Questions, each with:

A. Cognitive level.

B. K/A cross reference.

C. Learning Objective.

D. Reference.

E. Source and History.

F. Material required for exam (if necessary).

5 Reference material needed for the exam.

These materials are being submitted in support of the NRC initial operator licensing Written exam, at Beaver Valley Power Station Unit 1, scheduled for May 20,1998.

The 100 questions were all validated by an operations crew and

! the exam team.

, l' l We request these materials be withheld from public disclosure until after the completion of the exam.

If you have any questions or require further information please contact me at 412-393-5755 l

yD i

( 9908200128 980817 h j (DR ADOCK 05000334 PDR I

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Facility Beaver Valley 1 Date of Exam: 04/20/98 Exam Level: SRO Tier Group K/A Category Points Point Total K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G 4 4 4 ""' E '

1. 1 5 4 4 3 24 Emergency 2 3 3 '" 3 4 P -

1 2 16

& Abnormal 3 1 2 '

3 Plant ~'

Tier -

Evolutions

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Totals 8 5 7 8 10 5 43 1 3 2 1 1 1 1 2 2 2 2 2 19

2. Plant 2 3 1 5 1 1 2 2 1 1 17 Systems 3 1 1 1 1 4 Tier Totals 6 2 3 7 2 3 4 4 4 3 2 40

3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 6 4 2 5 17 Note: -

Attempt to distribute topics among all K/A Categories: select at least one topic from every K/A category within each tier.

Actual point totals must match those specified in the table.

Select topics from many systems: avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.

Systems / evolutions within each group are identified on the associated outline.

- The shaded areas are not applicable to the category / tier.

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~

~

l FrCility: Beave5aiiey3 ~ ~ ~ Date of ExamI/20/98 Exam Level: SRO

_._. Category KA # Topic  ! Imp. Points Conduct of Operations '2.1.10 Knowledge of conditions and limitations in the facility license. 3.9 1

2.1.12 Ability to apply technical specifications for a system. 4.0 1 l

.2.1.13 Knowledge of facility requirements for controlling vital / controlled 2.9 1 l

. access.

72.1.2 Knowledge of operator responsibilities during all modes of plant 4.0 1 l

[ operation. '

2.1.25 Ability to obtain and interpret station reference materials such as 3.1 1 graphs, monographs, and tables which contain performance data.

5 29 Knowledge of how to conduct and verify valve lineups. 3.3 1 l

Total 6 Equipment Control 2.2.12 Knowledge of surveillance procedures. 3.4 1 2.2.13 Knowledge of tagging and clearance procedures. 3.8 1

.2.2.31 Knowledge of SRO fuel handling responsibilities. 3.8 1

~ ' ~ ~ ~

2.2.6 ~Kn5wledge'of'the process for making changesin procedures as 3.3 1 described in the safety analysis report.

Total 4 R:diatioriCo5rol '2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control 3.0 1

_ requirements.

,_2.3.6 Knowledge of the requirements for reviewing and approving release 3.1 1 l permits.

l Total 2

~

Emerg'eley'Pr'ocedures[ 2.4.14 Knowledge of general guidelines for EOP flowchart use. 3.9 1 Plin 2.4.16 Knowledge of EOP implementation hierarchy and coordination with 4.0 1

, _ . . . _ _ _ _other support procedures.

2.4.16  : Knowledge of EOP implementation hierarchy and coordination with 4.0 1

-other support procedures. ]

2.4.27 Knowledge of fire in the plant procedure. 3.5 1 2.4.9 Kndwledge of low pov.ar I shutdown implications in accident (e g. LOCA 3.9 1 or loss of RHR mitigativn strategies).

Total 5 Ti r 1 Target Point Total (RO/SRO) 17 r

l Tussday, April 21,1998 2:52:26 PM

Question Topic: l Temperature tiending during cooldown A cooldown is in progress. The milestones listed on Figure 1 of 10M-51.4C,(see attached) were reached at 4

the following times:

- i (1) 0800

. (2) 0833

• (3) 0857

• . (4) 0917 What action, if any, is required to be taken to comply with Technical Specifications?

a. RCS cooldown is acceptable to this point. RCS cooldown rate will rot be exceeded if Figure I time limits 'are complied with from this point on.

b. RCS cooldown is acceptable to this point. RCS cooldown rate may be exceeded even if Figure 1 times are complied with from this point on.

c. RCS cooldown exceeded Technical Specifications. RCS temperature must remain constant until 0927. ,

d. RCS cooldown exceeded. Technical Specifications. Cooldown rate must be restored to within .

Technical Specification limits by 0947.

A s: Ia l Exam Level: lS l Cognitive Level: l Application ~ l Explanatio a cf Answer KA: l2.1.2 l RO Value: l3.0 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

KA Conduct of Operations Statement:

Knowledge of operator responsibilities during all modes of plant operation.

R;ference Reference Number Reference Section Page Number (s) Revision Learn.

Obj j l Station Shutdown - lOM 51.4.C IV.A.13.b C 10-11 Iss 4 Rev  !

Cooldown From MODE 3 to 12 ,

MODE 4 l

Beaver Valley - Unit 1 3.4.9.2 3/4 4-22,4-27 Amend Technical Specifications No.179 OM 6,7 & 10 Operational LP-SQS-RX IV.D.4 20 6 Lecture l

Q:estion Source l New l Question Modification Method l Question Source Comments: l M:terial Required for Figure 1 of OM-51.4.C - Blowup curve to max 81/2 x 11 Ex:mination:

[

_ d)p h a. E P (85 e(~ d to

/ ,;

[J'b, PageI

Questin Tcpic: l core Safety Limit curve eval At 20% power, the maximum allowable T.,, is limited by the Reactor Core Safety Limit. The basis for limiting T,,, under these conditions ensures that:

a. DNBR remains greater than or equal to the safety analysis DNBR limit and the average enthalpy at the vessel exit will not exceed saturation .

b. ~ DNBR remains greater than or equal to the safety analysis DNBR limit and the highest enthalpy  ;

anywhere in the core will not equal saturation. I

c. DNBR remains less than the safety analysis DNBR iimit and the average enthalpy at the vessel exit will not exceed saturation.

d. DNBR remains less than the safety analysis DNBR limit and the highest enthalpy anywhere in the core will not exceed saturation..

Ans: la l Exam Level: lS l Cognitive Level: 1 Memory l l Explanatio 2ef Answer KA: l 2.1.10 l RO Value: l2.7 l SRO Value: l3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

l KA Conduct of Operations i St:tement:

Knowledge of conditions and limitations in the facility license.

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Ob)

Reactor Protection System LP-SQS- 1.1 II.C.3 7 6 4.c

, Qxestion Source l New l Question Modification Method l l

Question Source Comments: l l l Material Required for TS Figure 2.1.1 Ex:mination:

l l

Page 2 I

l- - .- - - - - - - - - - - - - - - - -

vnniaJ u opB l Ires.w.s RJ During power operation the Diesel Generator #1 ic declared inoperable. Subsequently the 1B Quench Spray l pump is determined to be inoperable. I i

Assuming all required surveillance are completed satisfactorily, what is the required Technical Specification

. action?.

a. Restore both the IB Quench Spray and Diesel Generator #1 operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in Hot Standby within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. Restore either the 1B Quench Spray pump or Diesel Generator #1 to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in Hot Standby within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. Restore the 1B Quench Spray pump to operable status within one hour or be in Hot Standby within

. the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. .

d. Restore the IB Quench Spray pump .or Diesel Generator # 1 to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or be in Hot Standby within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Ans: ld l Exam Level: lS 1 Cognitive Level: -l Application- l-Explanatio a ef Answer KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l4.0 l Section: } PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

KA Conduct of Operations Setement:

I Ability to apply technical specifications for a system.

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj j Technical Specifications TS 3.0.5,3.6.2.1, 3.8.1.1 i Containment LP-SQS-13.01 5 12 l Depressurization Systems j Q:estion Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l  ;

Material Required for Technical Specifications )

Ex:mination:

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Qzestion Tcpic: l FFD requirements What are the fitness-for-duty requirements, with respect to alcohol, for an unscheduled RO who has been called out?

a. The RO may report to work if he/she has consumed alcohol within the past FIVE hours, but will be required.to pass a breath analysis test.

b. The RO may report to work if he/she has consumed alcohol within the past FIVE hours, but will be rubject to a breath analysis test only if deemed necessary by the NSS.

c. The RO must report to work even if he/she has consumed alcohol within the past FIVE hours, but will be required to pass a breath analysis test.

d. The RO shall not report to work if he/she has consumed alcohol within the past FIVE hours.

A s: ia l Exam level: l S- l Cognitive level: l Application l Explanatio ct af Answe'r KA: l 2.1.13 .l RO Value: l2.0 l SRO Value: l2.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1 SystemfEvc!ution

Title:

KA Conduct of Operations -

St:tement:

Knowledge of facility requirements for controlling vital / controlled access.

R:ference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Fitness-For-Duty Program 1/2 NPDAP 714 IV.2 & 3 2 0 l For Duquesne Light l Employees Conduct Of Operations 1/2LP-SQS-48.1 Vill. 18 10 3,39 l Question Source l New l Question Modification Method l Q:estion Source Comments: l Miterial Required for 1/2 NPDAP 2.14 Enmination:

i l

l Page 4 L___________________________

Question Tople: l TS SDM & Emergency Boration Given the following conditions:

RCS T., - 355 F.

l

• RCS pressure -400 psig

! -* RCS boron concentration.- 1000 ppm -

• Shutdown margin is below Technical Specifications allowable value

• Emergency Boration is initiated at 30 gpm boric acid

• A 70 ppm RCS boron concentration change is required to restore the required SDM Of times listed below, which is the MINIMUM emergency boration time that will ensure the required boric l acid has been added?

l

a. 15 ' minutes

b. 17 minutes -

c. 21 minutes

d. 24 minutes Ars: lc l Exam Level: lS l Cognitive Level: l Application l Explanatio A 73 ppm change at Normal Operating Conditions would require 500 gallons boric acid. The correction factor of c of Answer 1.18 multiplied by 500 would result in 590 gallons of boric acid. 590/30gpm = 19 minutes 40 seconds.

KA: l 2.1.25  ! RO Value: l2.8 l SRO Value: l 3.1 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

KA Conduct of Operations Statement:

Ability to obtain and interpret station reference materials such as graphs. monographs, and tables which contain performance data.

R:ference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Emergency Boration . lOM-7.4.S IV.A S2 1ss 4 Rev 1

Beaver Valley Unit 1 - '

3.1.1.1 3/4 1-1 Amend Technical Specifications No. 91 CVCS LP-SQS-7.1 IV.E 28 12 QIestion Source l New l Question Modification Method l QIestion Source Comments: l M;terial Required for - lOM-7.5 Figures 7-7,7 8 & Table 71.

Exa wation:

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i Page 5

Question Topic: l Permission for d .vi: tion from NSA. l

, In addition to normal requirements for manipulating components, which of the following describes who is l required to approve placing component in other than its Normal System Alignment (NSA)?

i

! a. Two SROs are required to approve the manipulation.

! b. Specific permission is required from the NSS. 4 cc. - W ^ W e' I m'",u l c. Either the NSS or ANSS has to approve the manipulation.

d. The General Manager, Nuclear Operations.

t Ans: Ie l Exam Level: }S l Cognitive Level: l Memory l Explanatio

! a cf Answer l KA: l 2.1.29 l RO Value: l3.4 l SRO Value: l3.3 l Section: l PWG l RO Group: l 1 l SRG Group: l1 l

System / Evolution j l

Title:

)

l l

KA' Conduct of Operations Statement:

Knowledge of how to conduct and verify. valve lineups.

Reference Reference Number Reference Section . Page Number (s) Revision Learn.  !

l Obj l 1 i

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! I i

Question Source l Facility Exam Bank l Question Modification Method l l

[ Question Source Comments: l j l Material Required for l l Ex mination: l l l

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1 Page 6

Q:estion Topic: l Procedure change rules for type of procedure i While et 100% power, an OMCN is to be written to change IOM-7.4.L " Blender Boration Operation." This '

change adds a step that directs placing ONE bank of Pressurizer heaters in MANUAL prior to initiating a boration. An Operations Unit Non-Intent Reviewer has determined that this does NOT change the intent of the procedure.

The on the spot change:

l

a. can be approved by TWO members of management, ONE holding a valid SRO license on Unit 1.

b. becomes effective 14 calendar days following review by the OSC and approval of the GMNO.

c. cannot be made because use of the procedure is not expected in the next 30 days.

d. cannot be made because this is a safety related procedure.

Ais: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio a cf Answer KA: l'2.2.6 l RO Value: l2.3 1 SRO Value: l3.3 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution I

Title:

KA Equipment Control Statement:

Knowledge of the process for making changes in procedures as described in the safety analysis report.

Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Control Of Operating 1/20M-48.2.B C. I .a B 10 Iss 4 Rev Procedures 13 Conduct Of Operations I/2LP-SQS-48.1 1.H.2 4 10 8, 9 Q:estion Source lNew l Question Modification Method l QIestion Source Comments: l Material Required for j Ermination: l l

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l Page 7

Q esti:n Tcple: l Omissions in OSTs A partit! OST is to be performed. Which of the following is an acceptable method of blocking the portions of the OST that are NOT applicable?

a. The ANSS blocks the 1.on-applicable portions.

b. The STA blocks the non-applicable portions and the RO v'erifies they are correct.

c. The system engineer blocks the non-applicable ponions and the ANSS verifies they are correct.

d. The PO blocks the non-applicable portions and the RO verifies they are correct.

Ans: lc l Exam Leveli l S l Cognitive Level: l Memory l Explinatio n cf Answer KA: l 2.2.12 j RO Value: l3.0 l SRO Value: l3.4 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

KA Equipment Control Statement:

Knowledge of surveillance procedures..

Reference Reference Number Reference Section Page Number (s) Revision . Learn.

Obj Adherence and 1/20M-48.2.C VI.B.17 10 iss 3 Rev Familiarization to Operating 18 Procedures Conduct Of Operations 1/2LP-SQS-48.1 10 10 Qrestion Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination:

Il l

l l

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Page 8 L _ _ ___-____- -_ -_ ____ _

Qrestion Tepic: l Caution Trgs Use of a Caution Tag is PROHIBITED for which of the following conditions?

I

a. Special additional manual actions are required to operate the tagged component. I

b. Operation of the tagged component will be affected because a portion of the system is not in NSA. l

c. As a temporary replacement for a component label that has fallen off.

d. As a warning that operation of the component will cause erratic indication.  ;

1 Ars: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio n cf Answer KA: l 2.2.13 l RO Value: l3.6 l SRO Value: l3.8 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

KA Equipment Control Statement:

Knowledge of tagging and clearance procedures.

Reference Reference Number Reference Section Page Numair(s) Revision Learn.

Obj Use of Caution Tags 1/2OM-48.3.L IV.A 1-2,3 Iss 4 Rev 4

Conduct Of Operations 1/2LP-SQF,48.1 V.P 7 10 15 Qrestion Source l Facility Exam Bank l Question Modification Method l Qrestion Source Comments: l M:terial Required for Ermination:

Page 9

Question T pic: l SRO control Which of the following d: scribes c responsibility of the Refueling SRu b.. g iael movement?

The Refueling SRO will:  !

I

a. initial the Fuel Assembly Handling Deviation Report with NSS concurrence.

b. be located on the manipulator crane structure during most fuel handling activities.

c. maintain the DLC Master Copy of the Fuel Handling data Sheets. l i

d. continuously monitor source range count level, tovid bt a jNW - l A s: lb l Eram Level: lS l Cognitive Level: l Memory l Explanatio O cf Answer KA: l2.2.31 l RO Value: l1.6 { SRO Value: l3.8 l C 'ction: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

KA Equipment Control Statement:

Knowledge of SRO fuel handling responsibilities.

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj R' fueling Administrative Book 1 -IRP 12R-1.1 II.D.4.b.15) 10 iss 0 Rev l Section 0

)

Fuel Handling Operations LP-SQS-6.13 III.B 5 5 2.b l

1 Question Source l New l Question Modification Method l Question Source Comments: l l M:terial Required for Ex mination:

I t

Page 10 l

L____

l Q estiomTopic: l High Radiation Definition l Technical Specifications requires radiation areas to be isol:ted by locked doors if the radiation levels are ,

l greater than:

t

a. 100 mrem /hr

b. 500 mrem /hr l c. 1000 mrem /hr I- d. 5000 mrem /hr & c<rdd b a Wew/ crat<4 ausuw.

Ais: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio l e tf Answer l

KA: l2.3.1 l RO Value: l2.6 l SRO Value: l3.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

KA Radiation Control l

Statement:

Knowledge of 10 CFR: 20 and related facility radiation control requirements.

Refirence Reference Number Reference Section Page Number (s) Revision Learn.

, Obj Technical Specifications 6.12 6-23 188

( QIestion Source l New l Question Modification Method l Question Source Comments: l l

Material Required for Verify Section 6 of the Technical Specification is not included in materials Ex:mination:

l l

l Page11

l Question Topic: l SRO action for gas release Given the following conditions:

1

• Reactor power - 100%-

Discharge of Waste Gas Decay Tank [lGW-TK-1 A] is planned for 1000 on 4/22/98 e The RWDA-G had been approved on 1.500 on 4/20/98 e

l The meteorologicalinformation indicates Stability Class A for atmospheric conditions

[ The status of the Gaseous Effluent Monitors is as follows:

l -

Gaseous Waste / Process vent [RM-GW-108A] noble gas channel inoperable Gaseous Waste / Process vent [RM-GW-108B] noble gas channel inoperable Preparation for the release was then delayed until 2300 on 4/23/98 i

l Which of the following describes the status at the new planned time for release (2300 on 4/23/98), assuming i equipment status and other conditions do NOT change?

a. The release can be initiated without restriction.

b. The release can be initiated only if sampling of the release stream is analyzed at least.one per every FOUR hours.

l

c. The release cannot be made because the 72-hour effective time limit for the RWDA-G has elapsed.

d. The release cannot be made because the Stability Class for release is unacceptable. t l Ans: lc l Exam Level: lS l Cognitive Level: l Comprehension l

[ Explanatio c e f Answer <

KA: l2.3.6 l RO Value: l2.1 l SRO Value: l3.1 l Section: l PWG l RO Group: l 1 l SRO Group: l1 l

System / Evolution l l

Title:

KA Radiation Control l Statement:

l Kno vledge of the requirements for reviewing and approving release peimits.

Reference Reference Number Reference Section Page Number (s) Revision Learn.

l' j Obj l

Decay Tank Discharge IOM-19.4.E step 7 NOTE E3 iss 3 Rev 2

Gaseous Waste Disposal LbSQS-19.1 11.G, ODCM 3.3.3.10 17-18 5 9.e System j; Question Source l New l Question Modification Method j Question Source Comments: l Material Required for IOM-19.4.E Examination:

(

Page 12

7 l

Question Topic: - l Time to Core Boiling l

Givzn the following conditions:

j

• The reactor has been shutdown for 2 days.

'

• RCS temperature is 150 'F.

• RCS pressure is atmospheric. l

• . PZR is a normal level for shutdown cooling.

Assume RHR is lost. Which of the following describes the time available until core boiling occurs?  ;

( Using the attached references, AOP 1.10.1 attachments 1,2,3, & 4)

a. Less than 10 minutes.

b. I1 to 20 minutes.

I

c. 21 to 30 minutes.

d. 31 to 40 minutes.

l l l Ans: l ~d l Exam Level: lS l Cognitive Level: l Application l l Explanatio l scf Answer l KA: l2.4.9 l RO Value: l3.3 l SRO Value: l 3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1 Syst:m/ Evolution l

Title:

KA- Emergency Procedures / Plan l Statement:

Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR mitigation strategies).

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj ResidualHeat Removal AOP 1.10.1 11, Attachment 1 iss 3A Syst m Loss Rev5

( Residual Heat Removal LP-SQS-10.1 8 9,10 System l

Q:estion Source l Previous 2 NRC Exams l Question Modification Method l Q estion Source Comments: l  ;

Material Required for AOP 1.10.1 Attachments I,2 3 & 4.  !

Ermination:  ;

i l

Page 13 E----_--_-__ - _ - . - _ _ --.- - - - - - - - - - - _ _ - - - _--- - -- _ _ - - - - _ - _ - - - - - - -

Question Topic: l Impl' mentation of Orange Path Given the f:llowing conditions:

• An unisolable steam line break has occurred on SG "B"

• SG "A" and "C" levels were overfed.

• A reactor trip and SI occur.

• Pressurizer pressure is 1180 psig

• Pressurizer levelis 12%

• T,,,is 400 'F and slowly dropping

• E-0 " Reactor Trip Or Safety Injection", step 9 is being performed.

- The STA informs' crew that B loop Toa is 283 'F and slowly dropping.

What is the EOP flowpath that will be followed given the above conditions?

a. Immediately transition to FR-P.1 " Response To Imminent Pressurized Thermal Shock Condition"

b. Perform actions ofE-0 through diagnosis of steamline break, then transition to E-2 " Faulted Steam Generator Isolation"a idt e-- 6='. .

c. Perform actions of E-0 through diagnosis of steamline break, then transition to FR-P.1 " Response to

' Imminent Pressurized Thermal Shock Condition"

d. Perform actions of E-0 through diagnosis of steamline break, then transition to FR-P.2 " Response to Anticipated Pressurized Thermal Shock Condition" Ans: Ie l Exam Level: lS l Cornitive Level: l Application l Explanation of Answer KA: l 2.4.14 l RO Value: l3.0 l SRO Value: l3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution j

Title:

l KA Emergency Procedures / Plan St'tement: i Knowledge of general guidelines for EOP flowchart use. I Reference Reference Number Reference Section Page Number (s) Revision Learn.

_ Ob]

Subcriticality - Status Tree F-0.4 ORANGE PATH issIB Rev1 Reactor Trip Or Safety IOM-53B.4.E-0 I. Ist paragraph 1 IssIB Injection Background Rev 5 EOP Introduction LP SQS-53.1 B.I 2 1 Question Source l New l Question Modification Method l Question Source Comments: l Material Required for F-0.4 and Att 5-D Examination:

f Page 14 i

L_______.____ _ _ . _ . .

Questiot Topic: l EOP Usage During Critical Safety Function Status Tree monitoring it was determined that TWO functions had Orange P:ths. One of the Orange paths is FR-H.1, Response to Loss of Secondary Heat Sink.

Wliich Critical Safety Function, also Orange, would take precedence over FR-H.17

a. FR-C.1, Response to Inadequate Core Cooling

b. FR-Z.1, Response to High Containment Pressure

c. FR-P.1, Response to Imminent Pressurized Thermal Shock Condition l l

d. FR-I.1, Response to High Pressurizer Level I Avs: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio i c cf Answer KA: l 2.4.16 l RO Value: l3.0 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

KA Emergency Procedures / Plan Stat: ment:

Knowledge of EOP implementation hierarchy and coordination with other suppon procedures.

Ref.rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj EOP Executive Volume - 1/20M-53B.2 til.B 9 iss1B User's Guide Rev 3 EOP Introduction LP SQS-53.1 2 Q:estion Source l Facility Exam Bank l Question Modification Method l Qyestion Source Comments: l Miteriki Required for Enmination:

l-Page 15 w____--____-____________- _ _ _ _ - _ _ _ _

Q::estio2 Tc pic: l Functional Recovery Procedure usage During a loss of all Emergency 4KV AC Power, When are Functional Restoration Procedures implemented?

a. Immediately upon electrical power restoration to I AE or IDF.

b. Immediately upon exiting ECA-0.0 " Loss of all 4KV AC Emergency Power "

c. When directed by ECA-0.1 " Loss of all Power Recovery Without SI Required" or ECA-0.2. " Loss l of all AC Power Recovery With SI Required"

d. When ECA-0.1 " Loss of all Power Recovery Without SI Required" or ECA-0.2. " Loss of all AC Power Recovery With SI Required"is completed.

Ans: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio e cf Answer KA: l 2.4.16 l RO Value: l3.0 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1 l

Syst:m/ Evolution

Title:

KA Emergency Procedures / Plan Stat: ment:

Knowledge of EOP implementation hierarchy and coordination with other support procedures.

Reference Reference Number Reference Section Page Number (s) Revision Learn. l Obj EOP Executive Volume - 1/20M-53B.2 VI.D 15 Iss1B Users Guide Rev 3 i l

'EOP Introduction LP-SQS-53.1 IV.C.4 20 1 1 Question Source l Facility Exam Bank l Question Modification Method l Q:estion Source Comments: l l l

M:terial Required for l Ex:mination:

Page 16 l

L --

Question Topic: l Fire Brigade Responsibility:s During a plant fire, who is responsible for coordinating fire-fighting activiti;s with the offsite fire departm:nt chiefs?

a. The ANSS when acting as the Fire Brigade Chief.

b. The ANSS when acting as the Fire Brigade Captain.

A

c. The affected Unit's NSS. g ).'b W

d. The Nuclear Operator when he/she is acting as the Fire Brigade Captain. ( k^ Q A _

A s: ls l Exam Level: lS l Cognitive Level: l Memory l Explanatio 2 cf Answer i

~

KA: l 2.4.27 l RO Value: l3.0 j SRO Value: l3.5 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution

Title:

KA Emergency Procedures / Plan Statement:

Knowledge of fire in the plant procedure.

t R i rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Fire Protection NPDAP 3.5 lil.N 3 6 I

Conduct of Operations 1/2LP-SQS-48.1 10 1 Q :estion Source l New l Question Modification Method l Qrestion Source Comments: l M:terial Required for Examination:

l I

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_ _ - - - _ - _ - _ _ _ _ - - - _ - - - - - - _ _ - _ _ = _ _ _ _ _

Question Topic: l Rod motion control-If c power mismitch signal is generated by the Rod Control System, which of the following paramet:rs determines the magnitude of the gain imposed by the variable gain unit?

a. Median Tave b.' Median delta T

c. N44 Power

d. Turbine Impulse pressure Avs: ld l Exam Level: lS l Cognitive Level: J Memory l Explanatio c ef Answer KA: l 001 A1.02 l RO Value: l3.1 l SRO Value: l 3.4 l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Control Rod Drive System

Title:

KA Ability to predict and/or monitor changes in parameters associated with operating the Control Rod Drive System Statement: controls including:

T-ref Ref;rence Reference Number Reference Sectica Page Number (s) Revision Learn.

Obj Reactor Control and lOM 1.5.A.51 1 Iss 4 Rev Protection 0 Reactor Control and 10M-1.1.D 13 iss 4 Rev 13 Protection i Full Length Rod Control LP-SQS-1.3 7 Q restion Source l NRC Exam Bank l Question Modification Method l QIestion Source Comments: l Miterial Required for Ex:mination:

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yuestion Uop6c2 l 1Mmpenma gn:nin gmmm gaan Given the following conditions:

• Reactor Power-72%

• Control Rods are at step 210 on Control Bank D  ;

• AOP 1.1.1, Failure of RCCA Control Bank to Move, is implemented due to rod control problems

• The RO incorrectly places the Control Rod Bank Sel Sw in CONTROL BANK D ,

t instead of MANUAL

• Rods are withdrawn 5 steps before this is discovered l

If the Control Rod Bank Sel Sw is placed in Manual at this point, which of the following will occur?  !

a. Upon shutdown, all Control Bank D rods will remain 5 steps withdrawn from the core.

b. Upon shutdown, the ROD BOTTOM / ROD DROP alarm will actuate 5 steps sooner than expected.

t

c. While operating, the Rod Insertion Limit alarms (A4-116 and A4-134) for Control Bank D would actuate 5 steps lower than the actual alarm setpoint positions.

d. While operating, the Bank Demand Position Indication will read 5 steps lower than the Analog Rod i Position Indication.

Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio a c.f Answer KA: l 001 K4.02 l RO Value: l3.8 l SRO Value: l3.8 l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Control Rod Drive System

Title:

KA Knowledge of Control Rod Drive System design feature (s) and or interlock (s) which provide for the following:

Statement:

Control rod mode select control (movement control)

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj reactor Control & Protection IOM-1.1.D Bank Overlap 15 16 iss 4 Rev

-Instrumentation and 1 Controls Full Length Rod Control LP-SQS-1.3 Ill.F.1 13 4 6.a Q esilon Source l New ,

l Question Modification Method l Qrestion Source Comments: l Material Required for Ex mination:

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wnumawpm i hmitmaag3httegLro During a natural circulation cooldown the required number of CRDM fans cannot be started.

During the cooldown, upper head voiding is prevented by:

a. venting the head via reactor vessel head vents.

b. verifying incore thermocouple temperatures are within an allowable range ofloop temperatures.

i

! c. increasing the minimum subcooling margin during portions of the cooldown.

l l

d. periodically injecting cold Safety Injection water into the Hot legs.

Ans: Ic l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio a cf Answer i

KA: l 002 K5.15 l RO Value: l4.2 l SRO Value: l4.6 l Section: lSYS l RO Group: l 2 l SRO Group: l2 l System / Evolution Reactor Coolant System Titit: i KA Knowledge of the operational implications of the following concepts as they apply to the Reactor Coolant System: l Statement:

Reasons for maintaining subcooling margin during natural circulation

! R:f;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj EOP Generic issues LP-SQS-53.2 1 13 Natural Circulation IOM-53 B.4.ES-0.2 1 2-3 iss1B Cooldown Background Rev 4 i

Q:estion Source l Facility Exam Bank l Question Modification Method l Q estion Source Comments: l M:terial Required for Examination:

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Page 20 l

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Question Tcpic: l SG temperature efTect upon start cf RCP Given the following conditions:

• Plant heatup in progress

• RCS temperature - 175 *F

• RCS pressure - 325 psig

• Pressurizerlevel- 28%

• Preparations are underway for the start of the first RCP, RCP 1 A The requirement of having less than 25 F difference between SG temperature and the primary system  ;

temperatures:

a. is not applicable since this is the first RCP to be started,

b. prevents an RCS overpressure event.

c. prevents exceeding RCS heatup rates.

d. prevents exceeding RCS cooldown rates.

Ars: lb l Esam Level: lS l Cognitive Level: l Memory l Explanatio j a c.f Answer KA: l 003 Kl.10 l RO Value: l3.0 l SRO Value: l3.2 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution Reactor Coolant Pump System

Title:

i KA Knowledge of the physical connections and/or cause-effect relationships between Reactor Coolant Pump System Statement: and the following:

RCS Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj i Re:ctor Coolant Pump IOM-6.4.A li.V 3 Iss 4 Rev Startup 7 i RCS - Reactor Coolant LP-SQS-6.3 Ill.A 24 4 12.A Pumps

(' Question Source lNew l Question Modification Method l Question Source Comments: l Material Required for Examination:

1 Page 21

Question Topic: l RCP power supplies The reactor is ct 35% with the electrical busses in NSA. Unit St: tion Service Transformer 1D develops a f; ult opening [4KV ACB 241D] USST ID Supply to IC 4KV Bus and (4KV ACB 341D) USST ID Supply to ID 4KV Bus. The auto bus transfer fails to operate on C & D Bus, Which of the following lists all running RCPs?

a. RCP1A

b. RCP 1 A and IB l, c. ~ RCP IB and IC

d. RCP IC Ans: lb l Exam Level: (S l Cognitive Level: l Memory l Explanatio a cf Answer KA: l 003 K2.01 l RO Value: l3.1 l SRO Value: l3.1 l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Reactor Coolant Pump System

Title:

KA Knowledge of electrical power supplies to the folicwing:

Statement:

RCPS Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj 4KV Distribution System LP SQS-36.1 Ill.B.2 3 I Reactor Coolant System - LP-SQS-6.3 I.C.1 11 4 1 R: actor Coolant Pumps Q/esteon Source l New l Question Modification Method l Q estion Source Comments: l Mat: rial Required for Enmination:

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Question Topic: l TS cval for charging pump Given the following conditions:

• Plant heatup in progress

• RCS temperature - 175 F e RCS pressure - 325 psig

• Charging pump [1CH-P-1B] is in service.

• Charging pump [1CH-P-1 A] is inoperable.

Which of the following describes limitations, if any, if[1CH-P-1C] were to be placed in service on AE Bus, and [lCH-P-1B] were to be removed from service?

a. [1CH P-1B] must be stopped and placed in PULL-TO-LOCK prior to taking [1CH-P-1C) out of PULL-TO-LOCK.

b. [lCH-P-1B] must be stopped and placed in AUTO prior to taking [lCH-P-lC] out of PULL-TO-LOCK.

c. [1CH-P-1B and IC] may be run simultaneously for up to 15 minutes, after which [1CH-P-1B] must be stopped and placed in PULL-TO-LOCK.

d. Both Charging Pumps may be tun without restriction until [lCH-P-1B] is removed from service.

ATs: lc- l Exam Level: lS l Cognitive Level: I Comprehension l Explanatio s cf Answer KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l4.0 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution Chemical and Volume Control System

Title:

KA Conduct Of Operations Statement: j Ability to apply technical specifications for a system. ]

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Beaver Valley - Unit 1 3.4.9.3 3/4 4-27a Amend Technical Specifications No.193 Placing the Spare Charging lOM-7.4 W IV.C W 9-13 Iss 4 Rev 12 Pump into Operation 10 CVCS LP SQS-7.1 IV.A B 28 12 Question Source lNew l Question Modification Method l Question Source Comments: [

Material Required for Technical Specifications Ermination:

Page 23

Question Tzpic: l Eval ofleak in Regen Hx Given the following conditions:

• Reactor power- 90%

• Pressurizer level- 51% stable  ;

e VCTlevel- 30% rising i

=

Letdown flow on [FI-CH-150] - 60 gpm

• Charging flow on [FI-CH-122] - 45 gpm

• Seal Injection flows - 8 gpm (A); 10 gpm (B); 7 gpm (C)

)

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• RCP #1 sealleakoff flows - 4 gpm (A); 4 gpm (B); 2 gpm (C) l Which of the following would result in the conditions above?

a. A leak exists in the Seal Water Heat Exchanger.

b. RCP #1 Seal Bypass Valve [MOV-CH-307] was inadvertently opened.

c. Letdown Pressure Control valve [PCV-CH-145] has failed open.

d. A leak exists in the CVCS Non-Regenerative Heat Exchanger.

Ans: la l Exam Lesel: {S l Cognitive Level: l Comprehension l Explanatio ,

acf Answer KA: l 004 K6.07 l RO Value: l2.7 l SRO Value: l2.8 l Section: lSYS l RO Group: l 1 l SRO Group: l1 j System / Evolution Chemical and Volume ControlSystem l

Title:

l i

KA Knowledge of the of the effect of a loss or malfunction on the following will have on the Chemical and Volume '

Statement: Control System:

Heat exchangers and condensers Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj CVCS LP-SQS-7.1 II.S 13 6 2,9 Question Source lNew l Question Modification Method l i' Question Source C6r,ments: l Material Required for Examination:

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Page 24 L______________...._

Question Topic: l RHR/RCS pressure response Given the f;11owing conditions:

• Plant cooldown is in progress at 20 F/hr

• RCS temperature - 155 F Pressurizer level (LI-1RC-462] Cold Calib - 100%

RHR Pump 1 A is running with flow of 4000 gpm set on [MOV-RH-605] RHR Flowin AUTO

[MOV-RH-758] Residual Heat Removal Hx FCV demand is set at 40%

• [MOV-CH-142] RH LTDN to Non Regen Hx Inle Flow Control Viv demand is set to 75%

Controller for [PCV-CH-145] LP LTDN Back Press Reg Viv is set in MANUAL at the position that is maintaining 50 psig with charging flow balanced If [ HIC-RH-758] controller causes [MOV-RH-758] to close with NO operator action, which of the following are the results for the first 10 minutes?

a. RHR flow will decrease and RCS pressure will decrease,

b. RHR flow will increase and RCS pressure will increase.

c. RHR flow will remain the same and RCS pressure will decrease.

d. RHR flow will remain the same and RCS pressure will increase.

Ans: ld l Eram Level: lS l Cognitive Level: l Comprehension I Explanatio o cf Answer KA: l 005 K3.01 l RO Value: l3.9 l SRO Value: l4.0 l Section: l SYS l RO Group: l 3 l SRO Group: l3 System / Evolution Residual Heat Removal System

Title:

KA Knowledge of the effect that a loss or malfunction of the Residual Heat Removal System will have on the Statement: following:

RCS Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Residual Heat Removal IOM 10.4.A E, F A 8-9 Iss 4 Rev System Startup (Plant 9 cooldown) And Operation RHRS LP-SQS-10.1 D.2.e, f 7-8 8 5.a, b, f; 10 Question Source lNew l Question Modification Method l Question Source Comments: l Material Required for

[

Examination:

Page 25

Question Topic: l Loss cf ONE St Accum Given the following conditions:

l

• Reactor power is 55%

L Accumulator [ISI-TK-1 A] level is 85%

Accumulator [ISI-TK-1 A] pressure is 657 psig SI Accumulator Isolation Valve [MOV-ISI-865A] is closed

! = The lockoutjack is removed a

Reactor shutdown was initiated due to the accumulator conditions Which of the following states the response of the SI Accumulators if a Design Basis LOCA occurs on the Loop B Cold Leg?

a. THREE Accumulators will fully inject into the core.

b. THREE Accumulators will fully inject into the core, provided the operator manually opens [MOV-ISI-865A).

c. ' TWO Accumulators, IB and 1C, will fully inject to the core.

d. ONE Accumulator, IC, will fully inject to the core.

Avs: ld l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio The IB Accumulator will discharge through the break a cf Answer KA: l 006 K6.02 l RO Value: l3.4 l SRO Value: l3.9 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Emergency Core Cooling System

Title:

KA Knowledge of the of the effect of a loss or malfunction on the following will have on the Emergency Core Cooling Statement: System:

Core flood tanks (accumulators)

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj SIS LP-SQS-11.1 Vill.D.7, XI.C.2 18,23 4 7.d,12.a 1 Qnestion Source lNew l Question Modification Method l )

Q:estion Source Comments: l Material Required for Ex mination:

Page 26

Question Topic: l Source ef PRT conditions Reactor is a 100% with di systems in NSA. The operator observes that PRT level has increased.

, Which of the following can caase the level increase?

a. A relief valve on the CCR system inside containment has lifted.

l b. RCP #2 Seal Leak off flow has increased.

t i

c. A PORV is leaking.

f d. RCP #1 Seal Leak off flow has increased.

1 Ans: lc l Exam lxvel: lS l Cognitive Level: l Memory l Explanatio n of answer KA: l 007 A3.01 l RO Value: l 2.7' l SRO Value: l 2.9 l Section: lSYS l RO Group: l 3 l SRO Group: l3 I

l System / Evolution Pressuriur Relief Tank / Quench Tank System

Title:

l KA Ability to monitor automatic operations of the Pressuriur Relief Tank / Quench lank System including: l Statement:

Components which discharge to the PRT I

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Alarm - Pressuriur Relief IO.M-6.4.AAF PC No. 2 AAF 2-3 lss 4 Rev Tank Level High-Low 3 Pressuriar and Pressure LP-SQS-6.4 1.B.2.c 4-5 4 7 ReliefSystems Reactor Coolant System. LP-SQS-6.3 R actorCoolant Pumps ,

i Q"estion Source l New l Question Modification Method l Question Source Comments: l l l

Material Required for Ex:mination: i Page 27

Question Tcpic: l PORV operation

[MOV-RC-535] Pressurizer Power ReliefIsolation Valve is closed due to [PCV-RC-455C] PORV leaking.

[PT-RC-445] Pressurize Pressure has failed downscale.

Select the available automatic overpressure protection, if any.

a. No PORVs will protect against overpressure.

b. Only PCV-RC-455D will protect against overpressure.

I

c. Only PCV-RC-456 will protect against overpressure.

d. Both PCV-RC-456 and 455D will protect against overpressure.

Ais: la l Exam Level: lS l Cognitive Level: l Application l Explanatio i o ef Answer l KA: l 010 K4.03 l RO Value: l3.8 l SRO Value: l4.1 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Pressurizer Pressure Control System

Title:

KA Knowledge of Pressurizer Pressure Control System design feature (s) and or interlock (s) which provide for the Statement: following: 1 Over pressure control Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Instrument Failure Procedure IOM-6.4-IF Figure 22 iss 4 Rev 6

Pressurizer & Pressure Relief LP-SQS-6.4 4 11 System i

Qrestion Source lNew l Question Modification Method l Q7estion Source Comments: l M:terial Required for I OM-6.4-l F Ex"mination:

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Page 28 t

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Questio] Topic: l Pressurizer Level Rx trip Pressurizer Level Control Channel Selector is selected to LT 459 & 460. All plant conditions are stable.

Which of the following will result in a reactor trip due to high pressurizer level?

a. 'At 5% power LT-RC-461 fails low.

b. At 5% power LT-RC-459 fails high.

c. At 25% power LT-RC-460 fails low,

d. At 25% power LT-RC-461 fails low.

ATs: lc l Exam Level: lS l Cognitive Level: l Comprehension l Expt;natio ca ef Answer KA: l 011 Kl.u4 l RO Value: l3.8 l SRO Value: l3.9 l Section: lSYS l RO Group: l 2 l SRO Group: l2 System / Evolution Pressurizer Level Control System

Title:

KA Knowledge of the physical connections and/or cause-effect relationships between Pressurizer Level Control Statement: System and the following:

RPS R;ference Reference Number Reference Section Page Number (s) Revision Learn.

Obj RCS -Instrument failure IOM-6.4.lF ll.a,ll.C.I.a IF 8-9 iss 4 Rev 6 i Pressurizer and Pressure LP-SQs-6.4 1.D. I .f 9-10 4 12 i ReliefSystem I Qrestion Source l New l Question Modification Method l )

Question Source Comments: l M:terial Required for Ex mination:

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Questioz Tcpic: l Eval OTDT & OPDT setpoints on input failure l

During operation at 97% power one Tw instrument is reading 4 degrees higher than other Tw instruments.

All T, temperatures are equal.

Which of the following describes the effect on OPdeltaT and OTdeltaT for the loop with the highest T 3,,?

Loop deltaT will be 1

a. closer to both OPdeltaT and OTdeltaT trip setpoints.

b. closer to its OPdeltaT trip setpoint, but will be farther from its OTdeltaT trip setpoint.

c. ' farther from its OPdeltaT trip setpoint, but will be closer to its OTdeltaT trip setpoint.

1

' d. farther from both OPdeltaT and OTdeltaT trip setpoints.

A s: la l Exam Level: lS l Cognitive Level: l Comprehension l Esplanatio  !

aif Answer KA: l 012 A2.05 l RO Value: l 3.l* l SRO Value: l 3.2* l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Reactor Protection System

Title:

KA Ability to (a) predict the impacts of the following on the Reactor Protection System and (b) based on those St:t: ment: predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Faulty or erratic operation of detectors and function generators Reference Reference Number Reference Section Page Number (s) Revision Learn. i Obj j RCS-Instrument Failure IOM-6.4.lF ll.B,111. IF 32 33,35-36 iss 4 Rev l 6

Reactor Protection Systi m LP-SQS-1.1 V.C.16 25-26 6 8 Reactor Coolant System LP-SQS-6.5 IV.A 17-20 5.a, b Q:estion Source l Facilit/ Exam Bank l Question Modification Method l Q:estion Source Comments: l l M ,terial Required for Examination:

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Q'estiom Tcpic: l Operation of BOTH Bypass Trip Breakers RPS testing is in progress for RPS train B and the status of the breakers are as follows:

l

• Reactor trip beakers (RTA and RTB) closed

• Reactor bypass breaker B (BYB) closed l

s Bypassing both RPS trains simultaneously is prevented by:

f a. tripping only BYA ifit is racked in and its CLOSE pushbutton'is depressed.

b. tripping only BYB if BYA is fully racked in.

c. preventing closure of BYA'ifit is racked in.

d. tripping all reactor trip and bypass breakers if BYA is racked in and its. CLOSE pushbutton is depressed.

Ans: Id l Eram Level: lS I Cognitive Level: l Memory l Explanatio o cf Answer -

KA: .l 012 A3.07 l RO Value: l4.0 l SRO Value: l4.0 .l Section: l.SYS l RO Group:-l 2 j SRO Group: l2 System / Evolution Reactor Protection System

Title:

KA Ability to monitor automatic operations of the Reactor Protection System including:

Statement:

Trip breakers Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj R; actor Control and lOM-1.1.B RP,2nd paragraph 2 iss 4, Prot:ction - Summary Rev.0 Description Reactor Protection System LP-SQS-1.2 11.1 7 6 8, 9 Hardware Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l Material Required for ,

Ex:mination: )

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Qrestion Tcpie: l Containment Pressure logics Containment pressure instrument PT-LM-100C has failed downsca'e. All appropriate actions of lOM-1.4.IF, Instrument Failure Procedure, have been completed.

Subsequently PT-LM-100D fails upscale.

Which of the followirg lists all expected actions?

a. CIA and SI

b. CIA, SI and MSLI 4

. c. CIB and MSLI

d. CIA, CIB, SI and MSLI A's: lb l Exara Level: lS l Cognitive Level: l Comprehension l Explanatio '

e cf Answer KA: l 013 A2.06 j RO Value: l 3.7' l SRO Value: l 4.0 l Section: lSYS [ RO Group: l 1 l SRO Group: l1 Syrtem/ Evolution Engineered Safety Features Actuation System

Title:

KA Ability to (a) predict the impacts of the following on the Engineered Safety Features Actuation System and (b)

Statement: based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Inadvenent ESFAS actuation Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Instrument Failure Procedure 10M 1.4.lf II.C 4 1ss 4 Rev 1  ;

Reactor Protection Trip LP-SQ-1.1 - 6 9 )

, Logics Q:estion Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l M terialRequired for Examination:

Page 32

I Q:estici Tcpic: l Operation following Si signal A steam break has occurred causing an Si on high containment pressure. Reactor Trip Breaker BYA will l NOT open. The crew has transitioned to ES-1.1, Si Termination. If containment pressure remains above the  !

Si setpoint, which of the following will occur if both SI Reset Pushbuttons are depressed?

l l

> , a. . Neither train of SI will reset.. .

b. Only one train of SI. will reset. .

c. Both trains of S1 will reset but one train will immediately reinitiate.

d. Only one train of S1 will reset. The reset train will immediately reinitiate. -

l A* s: lc l Exam Level: lS l Cognitive Level: l Application l ,

Tiplanatio ' ' '

6 af Answer j, 1A: l'Ol3 A3.02 l RO Value: l 4_.1 l SRO Value: l4.2 l Section: l SYS l RO Group: l 1 l SRO Group: .l 1 Sytum/Evolutiora Engineered Safety Features Actuation System

Title:

KA Ability to monitor automatic operations of the Engineered Safety Features Actuation System including:

Stat ement:

Operation of actuated equipment Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj FSAR Logic Diagrams Figure 7.21 Sheet 8 Reactor Protection System LP SQS-1.1 VI.E.1.f 34-351 6 9 Q estion Sc urce l Faci.lity Exam Bank l Question Modification Method l i

Question St urce Comments: l M terial Re tuired for Figure 7.2-1 Sheet 8 Ex:mination.

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Questio'n Tepic: l ROD BOTTOM alarm j During a reactor startup, when does the ROD BOTTOM / ROD DROP alarm (A4-126) become active for each control bank?

The alarm will actuate for a dropped rod for:

a. any Control Bank whenever Control Bank A RPI output is above 20 steps.

b. each Con' trol Bank whenever that Control Bank demand position is above 35 steps.

c. Control Banks A, B and C whenever their Control Bank demand position is above 35 steps, and for Control Bank D whenever Control B' ank D demand position is above 20 steps.

'd. ' Control bank A whenever Con' trol Bank A RPI output is above 20 steps, and for Control Banks B, C and D whenever their Control Bank RPI output is above 35 steps.

Avs: Id l Exam Level: lS l Cognitive Level: l Memory l l

Esplanatio oefAnswer KA: l2.4.31 l RO Value: l3.3 l SRO Value: l3.4 l Section: lSYS l RO Group: l 2 l SRO Group: l1 1 l

System / Evolution Rod Position Indication System - l Title .

1 KA Emergency Procederes / Plan Stat; ment: 1 Knowledge of annunciators alarms and indications, and use of the response instructions.

Refirence Reference Number Reference Section Page Number (s) Revision learn.

Obj Reactor Control & Protection IOM-1.1.B RP1,1st & 2nd 16 iss 4 Rev

. - Summary Description paragraphs 1 RPI and Insertion Limits LP-SQS 1.4 VI.B. C 5-6 5 2.b, c Reactor Control and lOM-1.2.B 1 Pr:tection Setpoints j Q=stion Source l Previous 2 NRC Exams l Question Modification Method l Q estion Source Comments: l ,

l Material Required for i Examination:

f Page 34

i Q:estim Tepic: l determination of NIS counts by IR/SR status l

Given the following conditions: l l-

• Reactor tripped from 100% power l Following transition to ES-0.1 " Reactor Trip Response", Intermediate

{

. Range.NIS is reading 1E-7 amps 4 I

• Five minutes later Intermediate range NIS is reading 2.2E-9 amps How soon following the last reading will Source Range NIS provide correct readings?

a. 4 minutes.

b. 8 minutes.

c. 10 minutes. , , ,

d. 13 minutes.

Ars: la l Exam Level: lS l Cognitive Level: I Comprehension l Explanatio P=Pi 10((T)(SUR)} Determine SUR form IRNIS readings over 5 minutes which gives SUR ='-l/1 dpm (constant

o ef Answer rate). This SUR is used with IR activation setpoint ~ 1E-10 gives time of 4.02 minutes..

KA: l 015 K5.06 l RO Value: l3.4 l SRO Value: l3.7 l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution NuclearInstrumentation System ,

Title:

( KA Knowledge of the operational implications of the following concepts as they apply to the Nuclear instrumen'_ation i Statement: System:

Subcritical multiplications and NIS indications Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Reactor Excore last. System I OM-2.1.C IR 2nd paragraph 9 iss 4 Rev

- Major Components 1 Excore Instrumentation LP-SQS-2.1 IV.C.8 10 5 5, 8 .

System Question Source l New l Question Modification Method l Qrestion Source Comments: l M;terial Required for Examination:

Page 35 u___________________

Qrestion Tepic: l Leak in RVLIS A leak has occurred at the inlet to a RVLIS differential pressure transmitter.

Which of the following describes RVLIS system indication and how the leak will be isolated?

a. RVLIS hydraulic. isolator po.sition will indicate a leak has occurred. Tlle leak will automatically.

isolate.

b. RVLIS hydraulic isolator position will indicate a leak has occurred. The leak can only be isolated by closing a manual isolation valve.

c. RVLIS high volume sensor position will indicate a leak has occurred. The leak will automatically isolate,

d. RVLIS high volume sensor position will indicate a leak has occurred. The leak can only by isolated

~

' by closing a manual isolation valve.

A s: la l Exam Level: lSe l Cognitive i.evel: -l Comprehension l <

Explanatio -

c cf Answer .

KA: l 016 K3.01 . l RO Value: l3.4* l SRO Value: l3.6* l Section: lSYS l RO Group: l 2 l SRO Group: l2 System / Evolution Non-Nuclear instrumentation System

Title:

KA Knowledge of the effect that a loss or malfunction of the Non-Nuclear Instrumentation System will have on the Setement: following:

RCS Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj RVLIS Hydraulic isolator IOM-6.4.AG IV.A.7, 8 AG2 iss 4 Rev Mtifunction 0 RVLSI & Core Cooling LP-SQS-6.7 II.B.e, f; ll.G.c; ll.H 4-5,16-17.,22- 1 6 M nitor 23 Question Source lNew l Question Modification Method l QIestion Source Comments: l M;terial Required for Ex mination:

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, Q:estio2 Tepic: l Eval of Natural Circulation for conditions l

! Given the following conditions:

• A loss of offsite power occurred

!

• A natural circulation cooldown.was initiated A The five hottest.T/Cs aver. age ternperature - 555 F , .

• RCS wide range pressure -1275 psig l

• All RCS Loop Tw - 552 F

• All RCS Loop Tm - 544 F

• All SG pressures - 940 psig Adequate natural circulation flow: (Refer to Att. 6A & 2G) l, a. exists and the RCS is subcooled.

b. does not exist and the RCS is subcooled.

c. exists and the RCS is at saturation.

d. does not exist and tlie RCS is at saturation.

Ans: lb l Exam level: 'l S l Cognitive Level: l Application l Explanatio c ef Answer KA: l 017 A3.01 l RO Value: l 3.6* l SRO Value: l 3.8' l Section: lSYS l RO Group: l 1 l SRO Group: l Syst;m/ Evolution in-Core Temperature Monitor System

Title:

KA ' Ability to monitor automatic operations of the In-Core Temperature Monitor System including:

Stat: ment:

Indications of normal, natural, and interrupted circulation of RCS R;firence Reference Number Reference Section Page Number (s) Revision Learn.

Obj 0 F Plus Subcooling Based 10M-53 A.I .6-A 1 iss 1B on Core Exit TCs Rev 2 N turalCirculation EOP Attachment 2-0 1 2 issIB l Verification Rev 2 EOP Generic issues LP-SQS-53.2 Vlli.C 19 12 Question Source l NRC Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Steam tables, EOP att. 2-G and 6A Examination:

Page 37

Questio) Tepic: j Powir supply following CU j The Containment Air Recirculation fans are in NSA prior to a transient which causes CIB.

i After CIB occurs, what will be the status of the Containment Air Recirculation fans?

1

. .a ' Running in fast speed 1

b. Running in slow speed

c. Tripped but the power supply is energized

d. Tripped with the power supply deenergized Ais: ld l Esam Level: lS l Cognitive Level: l Comprehension l Explanatio c of Anrwer + -

KA: l 022 K2.01 l RO Value: { 3.0' l SRO Value: l3.1 l Section: l SYS l RO Group: l 1 l SRO Group: l1 l

Syst;m/ Evolution Containment Cooling System i

Title:

I l

KA- Knowledge of electrical power supplies to the following:  !

Statiment:

Containment cooling fans  ;

Ref;rence Reference Number Reference Section Page Number (s) Revision- Learn. <

Obj l CNMT Vent - Summary lOM-44C. I .13 CNMT Air 1 Iss 4 Rev i Description Recirculation 0 Cont inment Ventilation LP-SQS-44C.1 II.A.! l-2 4 $,7 Systems QIestion Source l New l Question Modification Method l Qrestion Source Comments: l M;terial Required for Examination:

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Page 38 I

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Question Tcpic: l Quench Spray response to RWST level Oiven the following conditions:

• Reactor trip, SI and CIB occurred from 100% power due to a LOCA

• RWST level has decreased to 3 feet 9 inches l

..t CIB has not been reset..

q l

What would be the status of the Quench Spray (QS) system? i (Assume no operator action has been performed in the Quench Spray system.)

L . '

L a. BOTH QS pumps are running with [MOV lQS-103 A,103B] QSPP Cavitating Venturi Nozzle Bypass Isol,Vivs closed, and TWO QS Chemical Injectio' n' pumps are running.

b. BOTH QS pumps are running.with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle '

l Bypass Isol Vivs closed, and FOUR QS Chemical Injection pumps are running.

c. BOTH QS pum' ps are running with [MOV-lQS-Id3A,103B] QSPP Cavitating Venturi Nozzle Bypass Isol Vivs open, and TWO QS Chemical Injection pumps are running.

. d. BOTH QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle

. Bypass Isol Vivs open, and FOUR QS Chemical Injection pumps are running.

Ass: la i Exam Level: lS l Cognitive Level: l Comprehension l l Explanatio a cf Answer KA: l 026 Kl.01 l RO Value: l4.2 l SRO Value: l4.2 l Section: lSYS l RO Group: l 2 l SRO Group: l1 System / Evolution Containment Spray System

Title:

KA Knowledge of the physical connections and/or cause-effect relationships between Containment Spray System and St:tement: the following:

I- ECCS Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Loss of reactor Or Secondary E-1 step 30 22 1ssIB Coolant Rev 4 i

Transfer to Cold Leg ES-1.3 step 6 6 issIB Recirculation Rev 4 CNMT Depressurization LP-SQS-13.1 V.D. I 17-18 5.b System Question Source l New l Question Modification Method l Q estion Source Comments: l Material Required for Ex:mination: )

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I Page 39 L_____. - _ _ . _

Question Topic: l Recombiner Ops Given the following conditions:

• A LOCA has occurred 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ago

• ONE Hydrogen recombiner is placed in service when hydrogen concentration reaches 0.5%

l .

With a recombiner in operation, containment pressure:

a. should be maintained at approximately 8.9 PSIA, to prevent excessive recombiner flow.

b. will be adequate for recomb.iner operation ifit is maintained between 8.9 PSIA and -3 PSIG

c. should be maintained slightly above atmospheric, to ensure sufficient recombiner flow.

d. should be maintained at approximately -2PSIG, to ensure suflicient re. combiner flow. '

s s.

i Ams: Ic I Esam Icel: lS I Cognitive Level: l Application I i

Explanation i of Answer KA: l 028 Al.01 ~ l RO Value: l3.4 l SRO Value: I3.8 i Section: lSYS l RO Group: l 3 l SRO Group: l2 System / Evolution Hydrogen Recombiner and Purge Control System

Title:

KA Ability to predict anWor monitor changes in parameters associated with operating the Hydrogen Recombine' and Statement: Purge Control System controls including: 1 Hydrogen concentration Ref rence Reference Number Reference Section Page Number (s) Revision Learn.  ;

Oh]

Post DBA Hydrogen Control 10M-46.1.B 4th paragraph 1 Iss 44:

System - Summary Rev.0 l Description  !

Post DBA H2 Control LP.SQS-46.1 11.C.2.d 7 3 8,9 System System } ,

I

. Question Source l New l Question Modification Method l Question Source Comments: l Miterial Required for OM 46.4.A Ex-.mination:

i Page 40 L._ _ _ .__ . . . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ ___________________ _ _

Question Tcpic: l Evaluition of a leak Given the following conditions:

Reactor power is 85%

• Spent Fuel Pool is aligned for cooling

, *. A. leak has. occurred in th.e suction of[FC-P-1 A] Fuel Pool Cooling Pump If the leak remains unisolated, Spent Fuel Pool level should stabilize at:

a. ~25 feet above the top of the fuel.

' b. ~23' feet above the top of the fuel,

c. ~10 feet above the top of the fuel.

'd. the top of the fuel l ~

Aat: l c' l Exam level: 1S - l Cognitive level: l Memory l ,

Explanatio e of Answer KA: l 033 A2.03  : l RO Value: l3.1 l SRO.Value: l3.5 l Section: l SYS .l RO Group: l 2 l SRO Group: l2 System / Evolution Spent Fuel Pool Cooling System -

Title:

KA Ability to (a) predict the impacts of the following on the Spent Fuel Pool Cooling System and (b) based on those Statement: predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Abnormal spent fuel pool water level or loss of water level Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Fuel Pool Cooling and ICM-20.1.B 3 iss 4 Rev Purification 3 Fuel Pool Cooling and LP-SQS-20.1 9 6,9b Purification Question hource l New { Question Modification Method j l Question Source Comments: l M:terial Required for Examination:

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Page 41 L _ _ _ . - - ____ o

Q Y estio3 Tepic: l Transfer Can Operation Which of following d: scribes the interlock between the conveyor car drive and the upenders when i transferring the conveyor car from the transfer canal to the refueling cavity?

I

(

a. Both upenders must be in the down position before the conveyor car can be moved.

~

l li. ' Only tlie upender in the reftidling cafity triust be in the down position before thd conveyor cai can'be moved.

c. Only the up:nder in the transfer canal must be in the down position before the conveyor car can be moved.

d, if upender in the refueling cavity is not in the down position, movement'of the conveyor car can be initiated, however the conveyor car will stop before reaching the upender.

.Aas: la l Exam Level: lS - l Cognitive Level: j Memory l Explanatio '

o cf Answer KA: l 034 K4.02 l RO Value: l2.5 l SRO Value: l3.3 l,Section: lSYS l RO Group: l 3 l SRO Group: l2 System / Evolution Fuel Handling Equipment System

Title:

KA Knowledge of Fuel Handling Equipment System design feature (s) and or interlock (s) which provide for the Statement: following:

Fuel movement Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Fuel Handling Operations LP-SQS-6.13 XI.H.9.e 32 4 8.a IRP-12 R-3.2 II.6.b 2 iss 0 Rev 0

1 QIestion Source l NRC Exam Bank j Question Modification Method l QIestion Source Comments: l M;terial Required for Ex:mination:

l t

Page 42

QTestion Tc pic: l SG leval program Reactor power is 25% and all plant systems are in NSA.

Which failure would decrease feedwater flow to all SGs? 1 l

. a. ONE condenser-steam dump fails open.

b. Heater Drain receiver Level Control Valve [LCV-1SD-106B] fails open.

c.' Turbine First Stage Pressure channel [PT-lMS-446] fails low,

d. Combined Feedwater Header Pressure chann.el [PS-1FW-151] fails high. .

Ars: ic l Exam Level: lS 'l Cognitive Level: l Comprehension l Explanatio a of Answer -

KA: } 035 Kl.01 l RO Value: l4.2 l SRO Value: l4.5 l Section: lSYS l RO Group: l 2 l SRO Group: l '2 System / Evolution Steam Generator System

Title:

KA- Knowledge of the physical connections and/or cause-effect relationships between Steam Generator System and the Statement: following: ,

MFW/AFW systems l

Reference -

Reference Number Reference Section Page Number (s) Revision - Learn. -j Obj j SG Feedwater System - lOM-24.lD SGWLC 78 iss 4 Rev l Instrumentation and Controls 2  !

SG Feedwater System - I OM-24.4.lF Attachment 5, ll.A.2 IF 38 iss 4 Rev instrument Failure 2 ,

l Feedwater System LP-SQS-24.1 lil.E.10.d 14 1.A Question Source l New l Question Modification Method l l Q:estion Source Comments: l M:terial Required for Examination:

l l Page 43 l

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Qzestloa Tepic: l Effect of MS- PT-464 failing high Given the following conditions:

!

• The unit is in MODE 3 preparing for normal plant cooldown Condenser Steam Dump System is automatically controlling T,,, at 547 F in Steam Pressure Mode

• [PT-;MS-464) Mai.n Steam Header Pressure fails high .

Which one of the following describes the effect this will have on the Condenser Steam Dump system?

a. Two banks of steam dumps will open and remain open until manually closed.

b. Two banks of steam dumps will open but shou'Id reclose with no operator action.

( c. All banks of steam dumps will open and remain open until manually closed.

d. All banks of ste'am dumps will open but shotil'd reclose.with'no operator action.

Ass: lb l Eram Level: lS l Cognitive Level: l Comprehension - l Explanatio a of Answer

! KA: l 041 K6.03 l RO Value: l 2.7. j SRO Value: l 2.9 l Section: lSYS l RO Group: j 3 l SRO Group: j3 Systtm/ Evolution Steam Dump System and Turbine Bypass Control

Title:

KA Knowledge of the of the effect of a loss or malfunction on the following will have on the Steam Dump System and l Statiment: Turbine Bypass Control:

Controller and positioners, including ICS, S/G, CIWS Ref;rence Reference Number Reference Section Page Number (s) Revision learn.

Obj l M;in Steam System lOM-21.5.A.24 1 Iss 4 Rev i 0  !

Main Steam System LP-SQS-21.1 4 3 QIestion Source l New l Question Modification Method l Question Source Comments: l Material Reqaired for

Ex

mination:

l I

l Page 44

Qrestio1 Trpic: l NPSH for FW Given the following conditions:

l

• Reactor power- 100% ,

A load rejection occurs and the plant stabilizes at 45% power I

• Load rejection bistables " LOAD REJ 15-50%".and " LOAD REJ GREATER THAN 50%"

hrelit i I

How are the Steam Generator Fe'ed Pumps (IFW-P-1 A, IB) protected from a loss of suction pressure during the load rejection?

a. The Feedwater Heater Bypass Valve [TV-lCN-100] opened and closed FOUR minutes later. j

b. The Heater Drain Receiver Level Control Valve [LCV-ISD-106B] was maintained fully open until LO.W-LOW level was sensed in the Hehter Drain Receiver.

c. The Heater Bypass to Heater Drain Pump Suction Valve (TV-CN-125] opened and closed four.

minutes later.

d. The Co6densate Pumps Recirculation Valve [FCV-1CN-101] closed on the 15-50% load rejection and reopened FIVE minutes later.

A s: la l Exam Level: lS. j Cognitive Level: l Comprehension l Explanatio 2 cf Answer KA: l 056 A1.08 l RO Value: l2.3 l SRO Value: l 2.6' l Section: l SYS l RO Group: l 1 l SRO Group: l1 Syst;m/ Evolution Condensate System

Title:

KA Ability to predict and/or monitor changes in parameters associated with operating the Condensate System controls St:t: ment: including: j MFW pump suction pressure Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Load Rejection AOP 1.35.2 step Il 7 Iss 3A Rev 6 Figure 22 Step Load lOM-22.5.A.6 I iss 4 Rev Rejection Ckt 0 Extraction Steam and Heater LP-SQS-23 Ill.C.7 8-9 12.E Drains Qnstion Source l Other Facility l Question Modification Method l Q estion Source Comments: l M:terial Required for Ex:mination:

i Page 45 L____________________-______

Qrestici Tzpic: l Restoration of FW capability An inadvertent SI signal occurred at 100% power. The condition causing the SI signal is no longer present.

All systems function as designed and RCS conditions stabilize as expected following the inadvertent SI.

Which of the following states the condition (s) that would have to be met to feed via [FCV-lFW- ,

479(489)(499)], SG FW Bypass FCVs? . . l

a. Only the FWI FW BYPASS VALVE RESET pushbuttons would have to be depressed.

b. P-4 would have to be cleared and the FWI FW BYPASS VALVE RESET pushbuttons would have  ;

to be depressed. '

c. SI would have to be reset and the FWI FW BYPASS VALVE RESET pushbuttons would have to .be depressed.

~

d. SI wouId'have 't o be' reset, P-4 woulii have to be cleared and the FWI FW' BYPASS VALVE RESET-

.pu,shbuttons. would have to be depressed. ,

Ars: la l Enam Level: lS l Cognitive Level: l Application l Explanatio - .

o ef Answer KA: l 059 A4.ll l RO Value: l3.1 l SRO Value: l3.3 l Section: l SYS l RO Group: l 1 l SRO Group: l1 Syst:m/ Evolution Main Feedwater System .

Title:

KA Ability to manually operate and/or monitor in the control room:  !

Statiment:

Recovery from automatic feedwater isolation Refirence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Feedwater System LP SQS-24.1 lil.E. I1. 15-16 7 5.j, 7.A.(12)

Re"ctor Protection Systems LP-SQS-1.1 VI.E.5 38-39 6 9 Updated FSAR Figure 7.2-1 sheet 1 & 13 Q:estion Source l Facility Exam Bank l Question Modification Method l Q1estion Source Comments: l Material Required for Figure 7.2-1 sheet 1 & 13 Ex mination:

i Page 46 L-- - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - -- - _ _ _ _ _ __ _J

Qrestion Tepic: l SGWLC inputs Given the following conditions:

• Reactor power is 20%

• Feedwater has been transferred to the Main Feed Regulating Valves

• All. systems areNSA. -

• Narrow Range SG IC levelis 44%

4 [FCV-1FW-499] IC SG FW Bypass Viv is manually opened 15% .

After plant conditions stabilize, which parameter (s) will be different from those prior to [FCV-lFW-499]

opening?

a. Only [FCV-lFW-498] IC Main FW Reg Viv position l ' b. [FCV ~1FW-498] 1C Main FW Reg Viv position and Narrow Range SG IC Level c.. . Only Narrow Range SG IC Level

d. Narrcw Range SG 1C level and Stm Gen 1C Feed Flow indication A s: la l Exam Level: }S l Cognitive Level: l Comprehension l Explanatio nef Answer KA: l 059 Kl.04 l RO Value: l3.4 l SRO Value: l?.4 l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Main Feedwater System

Title:

KA Knowledge of the physical connections and/or cause-effect relationships between Main Feedwater System and the

j. St:tement: following:

S/GS water level control system Refirence Reference Number Reference Section Page Number (s) Revision Learn.

Obj SG Feedwater System - lOM-24,1.D SGWLC 7-8 Iss 4 Rev Instrumentation and Controls 2 Feedwater System LP-SQS-24.1 lll.E.10. 14 15 7 1.A QJestion Source l New l Question Modification Method l Question Source Comments: l M;terial Required for Errmination:

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Page 47

Question Tiple: l Relationship of AFW steam supply & feed supplies to SG Given the following conditions:

• Reactor power- 100%

• A loss of all AC power occurs

• Auxiliary Feed Pump IFW:P-2 startsand runs .

• The steam supply line from SG B to IFW-P-2 ruptures at the connection

- to the main steam line. ,

j

• The steam break prevents access to the Main Steam Valve Room Which of the following describes how the Auxiliary Feed System is affected by the above conditions?

a. All SGs will blowdown through the rupture, and NO auxiliary feed will be available. I

• / b. SG A and SG B will blowdown through the rupture, but NO auxiliary feed will be available.

1 c.' , SG A and SG B will blowdown through the rupture, but auxiliary feed can be established by , l opening the manual steam supply isolation valve from SG C.

d. Only SG B will blowdown through the rupture,'and auxiliary feed can be established from SG A.

Ans: ld l Exam Level: lS l Cognitive Level: l Memory l Explanatio ,

o ef Answer j KA: l 061 K3.02 l RO Value: l4.2 l SRO Value: l4.4 l Section: lSYS l RO Group: l 1 l SRO Group: l1 Syst;m/r. volution Auxiliary / Emergency Feedwater System

Title:

KA Knowledge of the effect that a loss or malfunction of the Auxiliary / Emergency Feedwater System will have on l Stattment: the following:

S/G  :

R:firence Reference Number Reference Section Page Number (s) Revision Learn.

Obj SG Feedwater System IOM 24.1.C Auxiliary Feed Pumps 2-3 iss 4; Rev 2 i Feedwater System LP-SQS-21.1 lil.J.9 20 7 1.B SG Feedwater System LP-SQS 21.1 lil.L.3.a .22 Question Source l New l Question Modification Method l Question Source Comments: l Mat: rial Required for  !

Ex mination:

Page 48 L_-_-______-__

Questio2 Tcpic: l Ovtreurrent effect on breaker cperation The Unit is et 85%. Which of the following conditions will result in bus l AE being maintained deenergized.

! a. [ACB-1 A10] 1 AE Emergency Bus feeder breaker trips on overcurrent.

, b. .l AE Emergency Bus reverse phase PT blows.a fuse. .. ,

l- c. [ACB-41C) 1 A Normal 4KV Bus Feeder Breaker trips on overcurrent.

d. ~ [ACC 41C] 1 A Normal 4KV Bus Feeder Breaker trips on Unit Station Service Tranformer IC Differential Trip.

A5s: Ia l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio acf Answer KA: l 062 K4.01 l RO Value:- l 2.6 l SRO Value: l3.2 l Section: {SYS l RO Gooup: l 2 l SRO Group: l2 System / Evolution A.C. Electrical Distribution

Title:

KA .. Knowledge of A.C. Electrical Distribution design feature (s) and or interlock (s) which provide for the following:

Statement:

Bus lockouts Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj 4160V Emergency Bus I AE lOM-36.4.ACZ lss 3 Rev ACB-1 A10 Auto Trip i Di;sel Generators LP SQS-36.2 - 8 6

~

Q:estion Source l Previous 2 NRC Exams l Question Modification Method l Question Source Comments: j Material Required for Examination:

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Page 49

Questi:n Tipic: l Bre:.kerintirlock(s)

Reactor power is 25% during a startup. Electrical loads have been transferred to the Unit Station Service Transformer (USST).

In order for Bus I A to be setup for Auto Bus Transfer to the System Station Service Transformer, which of the following lists the required position of the Live. Bus Transfer switch and the control switch for. ACB. .

41A7

a. Live Bus Transfer Switch - OFF i ACB 41 A Control Switch - After Close I 1

l

b. Live Bus Transfer Switch - OFF  !

l ACB 41 A Control Switch - After Trip I f I

c. Live Bus Transfer Switch- ON '

ACB 41 A Control Switch - After Close l d. Live Bus Transfer Switch - ON' ACB 41 A Control Switch - After Trip L Avs: la l Exam Level: lS l Cognitive Level: l Memory l Explanatio ccf Answer

[ KA: l 062 K4.01 l RO Value: l2.6 l SRO Value: l3.2 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution A.C. Electrical Distribution

Title:

KA Knowledge of A.C. Electrical Distribution design feature (s) and or interlock (s) which provide for the following:

Stat; ment:

t Bus lockouts Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj 4KV Station Service System I OM-36.1.E 20-21 iss 4 Rev

- Specific Instrumentation 1 and Controls 4KV Distribution LP-SQS-36.1 45 7 3.

Q estion Source l New l Question Modification Method l Question Source Comments: l Mit: rial Required for Ex mination: i l

l I

I-Page 50

Question Tepic: l Response to ground indication DC Bus 1-2 ground voltmeter went from 0 volts to -105 volts. The DC Bus is in NSA for 100% power operations.

1 Which of the following describes the effect the ground will have on DC bus operations?

a. The ground has caused actual voltage to the DC loads to decrease to 105 Volts.

b. The affected battery will discharge significantly faster than designed.

c. The bus will operate as required but the bus reliability has decreased.

d. Another ground on the same polarity of the bus will cause a short circuit.

ass: jc .l Exam Level: lS l Cognitive Level: l Memory l Explanatio a cf Answer l

KA: l 063 A2.01 l RO Value: l2.5 l SRO Value: l 3.2* l Section: ]SYS l RO Group: l 2 l SROGroup: l1 System / Evolution D.C. Electrical Distribution

Title:

KA- Ability to (a) predict the impacts of the following on the D.C. Electrical Distribution and (b) based on those i Statement: predictions, use procedures to correct. control or mitigate the consequences of those abnormal operation:

Grounds -

R:ference Reference Number Reference Section Page Number (s) Revision Learn.

Obj 125 V DC Control System- IOM 39.2 A.16 2 iss 3 Rev Precautions & Setpoints 0 125 V DC Control System IOM 39.1 3 iss 4 Rev 0

125 VDC LP-SQS-39.1 1 Question Source l New l Question Modification Method l ,

Qrestion Source Comments: {

M;terial Required for Ex mination:

i l

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Question Tepic: l Reverse powzr trip of DG Di:sel Generator No.1 is paralleled to 4160V Bus I AE for testing. The operator is in the process of adjusting load and voltage when the Governor Control switch sticks in the LOWER position.

If NO operator action is taken, what will be the Diesel Generator response to this condition?

DG frequency will:

a. decrease and the diesel will trip on reverse power.

b. decrease and the diesel will trip on overcurrent.

c. remain constant but the diesel will ip on reverse power.

d. remain constant but the diesel will trip on overcurrent. , ,

Avs: lc l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio ccf Answer KA: l 064 A1.08 l RO Value: l3.1 l SRO Value: l 3.4 l Section: l SYS JRO droup: l 2 l SRO Group: l2 System / Evolution Emergency DieselGenerators ,

Title:

KA Ability to predict and/or monitor changes in parameters associated with operating the Emergency Diesel .

Statement: Generators controls including:

Maintaining minimum load on ED/G (to prevent reverse power)

Refirence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Transferring Emergency IOM-36.4.Q IV.A.9 & CAUTION Q2 iss 4 Rev Feed 3 Transferring Emergency Busses I AE And 1DF From Em:rgency Feed To Normal Feed Alarm DIESEL IOM-34.ADU A8 127 ADU1 1ss 3 Rev GENERATOR NO.1 I REVERSE POWER Diesel Generators LP-SQS-36.2 VI.B 29 6 Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Ermination:

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^

Questio] Tz pie: l Diesel Gen rator Trips l

A loss of off-site power occurred and the diesel generators are supplying the emergency buses.

j .Which of the following will trip a diesel generator?

.n. The governor control switch in the control room is held in the RAISE position.

b. A governor failure causes engine speed to increase to 1050 RPM.

c. The jacket cooling water pump trips,

d. The' coupling fails on the lube oil pump.

i A:s: lb l Exam Level: lS l Cognitive Level: l Memory l l' "

Explanatio

!' c of Answer

, KA: l 064 K4.02 l RO Value: l3.9 l SRO Value: l4.2 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Emergency Diesel Generators Thie:

KA Knowledge of Emergency Diesel Generators design feature (s) and or interlock (s) which provide for the following:

-Statement:

Trips for ED/G while operating (normal or emergency)

R'.ference Reference Number Reference Section Page Number (s) Revision - Learn. '

Obj Locci- Overspeed Trip IOM 36.4.AFN 1 iss 3 Rev 1

Diesel Generators LP-SQS-36.2 8 6 Technical Specifications 4.8.1.1.2.b.4 3/48.4s Question Source l Facility Exam 13ank l Question Modification Method l Q:estion Source Comments: l M:t: rial Required for Ex:mination. .

)

Page 53

Question Topic: l Drain Tank isolation

( .Given the following conditions:

l l

• Low Level Waste Drain Tank level is 110 inches l = The discharge permit has been approved at discharge rate of 15 gpm

.* The discharge is in progress a.t 15 spm . ,

l What condition will automatically stop the release?

a. Both [TV-LW-105] Liquid Waste Effluent Trip valve and [FCV-LW-104-2] High Range Liquid i Waste Effluent Flow Control Valve closing on high-high radiation signal from [RM-LW-104].

l b. [FCV-LW-104-2] High Range Liquid Waste Effluent Flow Control Valve closing on low flow rate.

l l c. [FCV-LW-104-1] Low Range. Liquid Waste Effluent Elow Control Valve closing on low Waste Drain Tank level.

L L

d. The Lo'w Level Waste Drain pump tripping on low flow rate.

Ans: la l Exam Level: lS l Cognitive Level: l Memory l Explanatio e of Answer KA: l 068 A4.04 l RO Value: }3.8 l SRO Value: l3.7 l Section: l SYS l RO Group: l 1 l SRO Group: l1 l System / Evolution Liquid Radwaste System

Title:

i KA Ability to manually operate and/or monitor in the control room:

Stat; ment:

, Automatic isolation f

Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.  ;

Obj j Liquid Waste Disposal LP-SQS-17.1 II.C.7,8 & 10 11-13 3 2.b ,

Systim l Question Source l New l Question Modification Method l l Question Source Comments: l M terial Required for Ex mination:

l Page 54

Qrestica Tepic: l Annunci: tor Operation Due to a Steam Generator Tube Leak a Condenser Air Ejector Vent Monitor [RM-ISV-100] High alarm occurs causing Annuciator " Radiation Monitoring High"(A4-71) alarm to be received. Annuciator (A4-71) is acknowledged. Which of the following will cause Annuciator " Radiation Monitoring High"(A4-71) to reflash?

a. Condenser Air Ejector Vent Monitor [RM-ISV-100] rising to the High-High alarm setpoint.

b. Steam Generator Blowdown Sample Monitor [RM-ISS-100] rising to the High alarm Setpoint.

c. Steam Generator N-16 Monitor [RM-lMS-102] rising to the High alarm Setpoint.

d. High Capacity Steam Generator Blowdown Monitor [RM-1BD-101] rising to the High alarm Setpoint.

Ars: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio uf Answer KA: l 073 A4.02 l RO Value: l3.7 l SRO Value: l3.7 l Section: lSYS l RO Group: l 2 l SRO Group: l2 I System / Evolution Process Radiation Monitoring System j

Title:

1 KA Ability to manually operate and/or monitor in the control room: j Statement:  !

Radiation monitoring system control panel Ref rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Rad Monitoring System - I OM-43.1.D 10 iss 4 Rev Instrumentation and Cbntrols 3 Radiation Monitoring System LP-SQS-43.1 1 Question Source lNew l Question Modification Method l Q:estion Source Comments: l M:terial Required for Examination:

I l

i l

Page 55

Q estio]T:ple: l Ev~luatirn of av:ilable gir sources l A leak has occurred in the Station Air System in the Fuel Building. [PI-ISA-101] Station Air Main Header and [PI-IIA-106] Station Instrument Air Header pressure indicatien are both lowering.

Wh n Station Air pressure decreases to a specific setpoint, (TV-ISA-105] Station Air Header Trip Valve will:

a. open to supply instrument air loads.

b. open to supply containment air loads.

c. close to ensure all station air will be supplied to the instrument air loads.

d. close to maintain air to all station loads.

Ars: lc l Exam Level: lS l Cognitive Level: I Memory l E5planatio i cef Answer KA: l 078 K4.02 l RO Value: l3.2 l SRO Value: l3.5 l Section: l SYS l RO Group: l 3 l SRO Group: l3 Syst:m/ Evolution Instrument Air System

Title:

KA Knowledge ofInstrument Air System design feature (s) and or interlock (s) which provide for the following:

Stat ment:

Cross-ever to other air systems Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Compressed Air Systems - lO M 34.1.D Station Air Header Trip 5 iss 4 Rev instrumentation and Controls D Valve 0 VOND 34-1 Compressed Air LP-SQS-34.1 IV.A & D 15 5 Qrestion Source l New l Question Modification Method l Q:estion Source Comments: l M;terial Required for Er mination:

Page 56

Questici Trpic: l Containment Building P netrations during refueling Which cf the following is N'OT part of the Technical Specification definition of CONTAINMENT l INTEGRITY  !

l 1

a. The containment leakage monitoring system is OPERABLE. i b; All equipment ha'tches are closed and ~ sealed.

c. The sealing mechanism associated with each penetration is OPERABLE. l

d. The containment leakage rates are within their LCO limits.

Ars: .la l Exam Level: lS l Cognitive Level: l Comprehension l j Explanatio . ,

s cf Answer )

KA: l 103 Kl.02- l RO Value: l3.9 l SRO Value: l 4.l* l Section: l SYS l RO Group: l 3 l SRO Group: l2 Syst m/ Evolution Containment System

Title:

l KA- Knov ledge of the physical connections and/or cause-etTect relationships between Containment System and the Stat; ment: following:

Containment isolation / containment integrity '-

Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Technical Specification 3/4.9.4 3/4 9-4 Containment System LP-SQS-47.1 VI.B 20 4 8.h Question Source lNew l Question Modification Method l QIestion Source Comments: l M:terial Required for Enmination:

L Page 57 i

Q:estion Topic: l Determination of power increase Given the following conditions:

• EOL i Reactor power is 80% steady state

• RCS T,,,is on program

• Control Rod position - 160 steps on Control Bank D  !

• Control Rods begin to withdraw

• When Control Bank D is at 170 steps the Control Rod Bank Sel Sw is placed in MANUAL stopping rod motion If N0 further operator action is taken, what would be the affect on actual power level and RCS T ,, after conditions stabilize?

a. Reactor power and RCS T ,, would both rise equally by an amount equivalent to the reactivity addition.

b. Reactor power would rise by an amount equivalent to the reactivity addition and RCS T ,, would remain approximately 571 F.

c. Reactor power would remain approximately 80% and RCS T,,, would rise by an amount equivalent to the reactivity addition.

d. Neither reactor power nor RCS T,,, would be significantly affected.

i Ans: lc l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio Reactivity addition by rod movement would add power to RCS. Since turbine load controls power level on NIS, a cf Answer RCS would heat up. By using Power defect curves could determine the equivalent power level the reactivity would allow and the associated Tavg at that power will approximate the temperature of the RCS (Use of Power Defect Curves provides an approximation because it includes Fuel temp / Doppler coefficient, but impact is relatively small compared to moderator temp coefficient over area of concern)

KA: l 001 AKl.03 l RO Value: l3.9 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l1 Syst;m/ Evolution Continuous Rod Withdrawal

Title:

KA- Knowledge of the operational implications of the following concepts as they apply to Continuous Rod -

Statement: Withdrawal:

Relationship of reactivity and reactor power to rod movement Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Full Length Rod Control LP-SQS-1.3 4 16 Question Source l New l Question Modification Method l Question Source Comments: l

, Material Required for Ex:mination:

! Page 58

Question Tople: l Operation cf Disconnect Switch l

L Oiven the following canditions:

i

• Reactor power - 5%

-* Control rod F-6 in Control Bank D has fully dropped.

• Recovery of the dropped rod is in progress per AOP 1.1.5 " Dropped RCCA"

• All Disconnect Switches in Control Bank D are in DISCONNECT except for F-6 Which of the following describes alarms that will be received and their effect on recovering the dropped

[ control rod?

a.. An urgent failure will be received, however rod recovery can proceed with the Control Rod Bank Sel Sw in Control Bank D.

b. An urgent failure will be received, however rod recovery can proceed with the Control Rod Bank l Sel Sw in Manual.

L c. A non-urgent failure will be received which will not affect control rod movement.

d. An urgent failure will be received, however rod recovery can proceed after depressing the Rod Control Alarm Reset pushbutton.

A,rs: la l Exam Level: l S' l Cognitive Level: l Comprehension l I l

l Explanatio -

)

I at af Answer KA: l 003 AK2.05 - l RO Value: l2.5 l SRO Value: l 2.8 - l Section: l EPE l RO Group: l 2 l SRO Group: l1 i System / Evolution - Dropped Control Rod -

Title:

KA- Knowledge of the interrelations between Dropped Control Rod and the following:

Statement:

Control rod drive power supplies and logic circuits Reference Reference Number Reference Section Page Number (s) Rcvision Learn.

Obj

. Dropped RCCA AOP 1.1.5 11 5 iss 3A Rev 7 -

Alarm - ROD CONTROL IOM-1.4.AAR A4-105 Corrective AARI- Iss3 Rev SYSTEM URGENT Action NOTE 2

[

FAILURE

! Full Length Rod Control LP-SQS-1.3 II.G.3 & IV.A.3 14 & 16 10;16 i ,

t Question Source - l New- l Question Modification Method - l Question Source Comments: [

M:terial Required for -

Examiastion:

Page 59 i

j

Question Topic: l Operation:1 limits & basis with given stuck rod Given the following conditions:

• Reactor power - 85%

• Load increase is in progress

• Control Bank D is 2 steps above the RIL

= Control rod K-6 indicates 15 steps below the remaining rods in Control Bank D

= Control rod trippability is confirmed

. Shutdown Margin is verified to be satisfied if the NSS decides to continue power operation with the control rod misaligned, which of the following describes required power reduction and the associated reason?

Reactor power must be reduced to at least:

a. 75% power within ONE hour to remain in compliance with Rod Insertion Limit restrictions.

b. 75% power within ONE hour to provide assurance of fuel rod integrity during continued operations.

c. 50% power within FOUR hours to remain in compliance with Rod Insertion Limit restrictions.

d. 50% power within FOUR hours to provide assurance of fuel rod integrity during continued operations.

Ans: Ib l Exam Level: lS l Cognitive Level: l Application l Expt;natio a of Answer KA: l 005 AKl.06 l RO Value: l2.9 l SRO Value: l3.8 l Section: l EPE l RO Group: ll l SRO Group: l1 System / Evolution. Inoperable / Stuck Control Rod

Title:

KA Knowledge of the operational implications of the following concepts as they apply to inoperable / Stuck Control St:tement: Rod:

Bases for power limit, for rod misalignment R:ftrence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Beaver Valley - Unit 1 3. l.3.1 (ACTION C.3) 3/4 1-18-19 . Amend Technical Specifications No.154 Beaver Valley - Unit 1 Bases 3/4.1.3 B 3/4 l-4 Amend Technical Specifications No.141 Full Length Rod Control LP-SQS-1.3 111.1.1 15 15 Q:estion Source l NRC Exam Bank j Question Modification Method l Question Source Comments: l Miterial Required for Technical Specifications Ernmination:

Page 60

QuesttbFoph: l Rteamliismp 0 Meets Given the following conditions:

• , Reactor tripped from 100% power

• Reactor trip breaker (RTB), which provides P-4 input to Reactor Trip .

.Corf aller, CANNOT be opened after the trip

= Reactor trip breaker (RTA) opened Which of the following identifies where the RCS temperature should stabilize prior to placing the Steam Pressure 1.fode Selector Switch in Steam Pressure Mode?

a. 543 F.

. 'b. ' 547 F. .

c. 549 F.

. d. 554 F.

Ans: lc l Exam Level: lS l Cognitive Level:- l Comprehension l Explanatio a of Answer KA: l 007 EA2,03 l RO Value: l4.2 l SRO Value: l4.4 l Section: j EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Reactor Trip

Title:

KA Ability to determine and interpret the following as they apply to Reactor Trip:

Statement:

kcactor trip breaker position Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Main Steam Systems IOM-21.5.A.24 1 iss 4 Rev 1 0 Main Steam System - lOM-21.1.D various 3-6 iss 4 Rev Instrumentation and Controls 1 Main Steam Supply / Steam LP-SQS-21.1 til.D, Ill.E, V.C.5, 12 14,27-28, i .e, 3.a Dump System V.E.1 30-31 l Question Sourec l New l Question Modification Method l l- Q:estion Source Comments: ]

i M:terial Required for Examination: )

l Page 61 1

1 L _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _

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EcumD wptn I upents sImmmm1aasamasa /Artfg A manual reactor trip was initiated at 100%, however the reactor will not trip. Step 1 of FR-S.1 is being performed. Control rods are in AUTOMATIC.

With the turbine tripped, which of the following describes required action concerning control rod insertion?

L. . ,. .

Control rods should be inserted in:

' ^

a. MANUAL even if they are inserting in AUTOMATIC.

1

b. _ AUTOMATIC provided rods are inserting in AUTOMATIC.

c. AUTOMATIC until reactor power is.less than 15% where the rods will'stop, requiring MENUAL insertion.

. d.' ~ AUTOMATIC tintil the Rod Insertion Limit is reached'where the rods will stop, rsquiring' MANUAL insertion.

Ars: lb l Exam Level: lS l Cognitive Lesel: l Comprehension l Explanatio o cf Answer KA: l 007 EK3.01 j RO Value: l4.0 l SRO Value: l4.6 l Section: lEPE l RO Group: l 2 l SRO Group: l2 l System / Evolution Reactor Trip

Title:

l KA Knowledge of the reasons for 11. . following responses as they app!y to Reactor Trip:

Statement:

Actions contained in EOP for reactor trip Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Response To Nuclear Power FR-S.1 step 1, RNO 2 iss1B G:neration- ATWS Rev 4 Rrsponse To Nuclear Power IOM-53.4.F R-S. ] 111.1 Knowledge 57 iss 1B Generation- ATWS Rev 4

Background

EOPs LP-SQS-53.3 1, 3 Q:estion Source l New l Question Modification Method l QIestion Source Comments: l M:terial Required for Examination:

I l

l l Page 62 o_______________

Qxestion Te pici l Evil of vapor space le:A -Tech Spec limit Given the following conditions:

• The reactor is operating at 100% power

• A 1.2 gpm valve packing leak has occurred on [PCV-RC-455B] PRZR Spray Viv e .The Primary Drains Transfer Tank level is increasing ,

Which of the following describes what type ofleakage this is and based on the leak size what action is required per Technical Specifications?

This leak is considered:

a. Primary boundary LEAKAGE that requires Technical Specification entry.

b. Identified LEAKAGE that does not require Technical Specification entry.

.c. Unider.tified LEAKAGE that requires Technical Specification ent.ry,

d. Unidentified LEAKAGE that does not require Technical Specification entry.

Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio c ef Answer j KA: l 2.2.22 l RO Value: l3.4 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Pressurizer Vapor Space Accident ,

Title:

i KA Equipment Control ,

St:tement: l Knowledge of limiting conditions for operations and safety limits.

R:ference Reference Number Reference Section Page Number (s) Revision Learn. l Obj Beaver Valley -Unit i 1.14,3.4,6.2 1-3,3/4 4-13 Technical Specifications RCS LP-SQS-6.5 Vll.A 24 4 8.g QY.estion Source lNew l Question Modification Method l Qxestion Source Comments: l Miterial Required for Ex mination:

Page 63

Question Topie: l Basis for use of ADVERSE Cnmt values Given the following conditions:

• A LOCA has occurred

• Containment pressure increased to 6.0 psig

• Containment radiation has increased to 1.5E+5 R/hr. .,.

Ninety minutes later containment pressure decreases to 3.0 psig and containment radiation has decreased to 4E+4 R/hr. Integrated CNMT radiation dose is 2.3E+5 Rads.

Which of the.following describes whether the use of adverse containment parameters can be discontinued?

a. Use of adverse containment parameters can be discontinued.

b. Continued use of adverse containment parameters is requin:d only due to the containment radiation readings.

c. . Continued use of adverse containment parameters is required only due to the containment pressure conditions.

d. Continued use of adverse containment parameters is required due to both the containment pressure.

and radiation conditions.

~

Ans: la l Exam Level: lS l Cognitive Level: l Application l-Explanatio n of Answer KA: l 009 EK3.16 l RO Value: l3.8 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Small Break LOCA

Title:

KA Knowledge of the reasons for the following responses as they apply to Small Break LOCA:

Statement:

Containment temperature, pressure, humidity and level limits Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Generic Instrumentation IOM-53B.5.Gi-2 II.D 12-13 issIB Rev 2 EOP Generic issues LP-SQS-53.2 X.B.6, 8 22-23 1 15 Qutstion Source lNew l Question Modification Method l Question Source Comments: l Msterial Required for Subcooling Attachment 6-A Examination:

Page 64

vrimucpa i tm<manmannmummerengutcve Given the following conditions:

a A LOCA has occurred e Containment pressure is 9.2 psig and lowering a RCS pressure has stabilized at 325 psig .

• Steam generator pressures are 800 psig and lowering a All ECCS equipment has responded as required Which of the following describes when the RCPs should be tripped?

a. Immediately

b. When the highest steam generator pressure reaches 700 psig.

c. When the highest steam' generator pressure reaches 525 psig.

' d. ~ When the lowest steam generator pressure reaches 700 psig.

A*s: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio c cf Answer KA: l 011 EA1.03 l RO Value: l4.0 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l1 Syst;m/ Evolution Large Break LOCA

Title:

KA Ability to operate and / or monitor the following as they apply to Large Break LOCA:

Stat: ment:

Securing of RCPs Reference Reference Number Reference Section Page Number (s) Revision Learn.

^

Obj Recctor Trip Or Si lOM-53.A.E-0 Foldout IssIB Rev 5 EOP Generic issues LP-SQS-53.2 Terminal Obj.

Question Source l New ] Question Modification Method l Question Source Comments: l M terial Required for Ex:mination:

1 Page 65 l

u._____._

Question Topic: l Determination of RCP/ reactor trip Reactor power is 35%. Which of the following combinations ofloop flow conditions indicates that a reactor trip should have occurred?

a. [F1-1RC-414] RCL 1 A Flow indicates 80%.

[F1-1RC-424] RCL IB Flow indicates.80%. ,

b. [FI-lRC-414] RCL 1 A Flow indicates 80%.

[FI-lRC-415] RCL 1 A Flow indicates 80%.

c. [F1-1RC 414] RCL 1 A Flow indicates downscale.

'[FI-IRC-435] RCL 1C Flow indicates 80%.

d. [FI-l RC-414] RCL 1 A Flow indicates upscale.

[FI-lRC-415] RCL 1 A Flow indicates 80%.

Ans: lb l Exam Level: lS l Cognitivo Level: l Memory l Explanatio n of Answer KA: l 015 AAl.03 l RO Value: l3.7' l SRO Value: l3.8 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Reactor Coolant Pump Malfunctions

Title:

KA Ability to operate and / or monitor the following as they apply to Reactor Coolant Pump Malfunctions:

Statement:

Reactor trip alarms, switches, and indicators Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Reactor Coolant System - 10M-6.4-IF lil.A 27 iss 4 Rev instrument Failure Procedure 6 Reactor Coolant System LP-SOS-6.5 4 5,6 Question Source lNew l Question Modification Method l Question Source Comments: l Material Required for Examination:

Page 66

~

QD:1LTDVC0tn I ffai[ure ofhndietr@ j 1

Given the following conditions: l

- VOLUME CONTROL TANK LEVEL HIGH-LOW (A3-53) has alarmed

- [LI-lCH415] Volume Control Tank Level (VB-A) failed offscale high Actual VCT level will:

a.- remain constant.

b. decrease until automatic makeup initiates.

c. decrease unt.il the ' charging pump suction transfers to the IkWST.

d. decrease until the VCT is empty.

Ass: Id l Exam Level: 1S l Cognitive Level: l Application l Explanatio o ef Answer +

KA: l 022 AA1.08 l RO Value: l3.4 l SRO Value: l3.3 l Section: l EPE - l RO Group: l 2 l SRO Group:] 2 System / Evolution Loss of Reactor Coolant Makeup

Title:

KA Ability to operate and / or monitor the following as they apply to Loss of Reactor Coolant Makeup:

Statement:

VCT level R:ference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Alarm - A3-53 VCT Level l OM-7.4.AAX PC 4,5 A 2-3 iss 4 Rev High Low 0 CVCS - Instrumentation and lOM 7.1.D Auto M/U, LCVs 1-2, 8-9 iss 4 Rev Controls 2 CVCS LP-SQS-7.1 Ill.D.2.b 21 2.g. 6.a Qrestion Source lNew ] Question Modification Method l Q=stion Source Comments: l Material Required for OM Figure 7 39 {

Ex mination:

I

!- Page 67 N _ __--_ ___ -___ _-_

vrumnreps: 1 L:sansalsgLohmIranspem Given the following conditions:

i

• [lCH-P-2A] Boric Acid Transfer Pump is out of service

• . RCS Temperature is 420 F

• SDM is .l.67 delta K/K

• S/D Banks are fully withdrawn if[ICH-P-2B] Boric Acid Transfer Pump trips, HOW will required Technical Specification Shutdown Margin be restored?

a. BORATE, by gravity feed'ing the in-service Boric Acid tank to the blender.

.b. Emergency borate through the Emergency Boration valve [MOV-CH-350].

c. Align the suction of the charging pum'p to the RWST.

' ~

d. ' Open the reactor trip breakers.

A s: lc- l Exam level: l B- l Cognitive Level: l Application . l

~

Explanatio ncf Answer KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l4.0 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Emergency Boration

Title:

KA Conduct Of Operations Statement:

Ability to apply technical specifications for a system.

R'.f;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Technical Specifications 3.1.1.1,3.1.2.2, and 3.1.2.6 CVCS LP-SQS-7.1 6 11 Question Source l New l Question Modification Method l Q estion Source Comments: l Material Required for Technical Specifications Ex;mination:

l

{

Page 68

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QUnIEDUcpm l aiMrgaisylannsegasmiGuiGim Fcilowing a turbine load rejection, control rods are automatically inserted causing ROD CONTROL B ANK l

D LOW-LOW elarm (A4-124) to be received.

Which of the following is the required action by procedure?

a. ' Place the rods in manual and withdraw them until the alarm clears. i l

b. Place the rods in manual and allow temperature to stabilize.

c. Emergency borate.

'd. Borate via the normal flow path until th'e CONTROL B ANK D LOW-LOW alarm clears.

Ars: _jc l Exam Level: IS l Cognitive Level: l Memory l Explanatio asf Answer KA: l2.4.31 l RO Value: l3.3 l SRO Value: l3.4 l Section: lEPE l RO Group: l 1' l SRO Grour: l l_

System / Evolution Emergency Boration

Title:

KA' Emergency Procedures / Plan Statement:

Knowledge of annunciators alarms and indications, and use of the response instructions.

Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Emergency Boration 10M 7.4.S 1 1 iss 4 Rev 1

Rod Control Bank D Low 10M-l.4.ABF 1 iss 3 Rev Low 1 I CVCS LP-SQS-7.1 10.p  ;

QJestion Source l Facility Exam Bank l Question Modification Method l i Q7estion Source Comments: l )

Miterial Required for Ex:mination:

Page 69 I

l _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ ._. .__ _ _ _ _ -

wrm2D UEp23: l LROJMRamWQUanaimmnis  ;

While operating tt 175 *F and the RCS depressurized, the running RHR pump trips. The other RHR pump is available to be immediately started.

s Which of the following describes when the other RHR pump should be started and the basis for this decision? .

l I The second RHR pump should be started:

a. immediately, to avoid any heatup of the RCS. ]

o after investigating ~the cause of the running pump trip, to avoid losing the second pump.

b. ' nly l

l

c. only after observing an R'CS heatup, to avoid unnecessary starts of the RHR pump.

~

d.' sith'ni five minutes, whicliis'tlie most limiting time until boiling will occur.

l A .5: lb l Exam Level:- l S l Cognitive Level: i Memory l l Explanatio e cf Answer )

KA: J 025 AKl.01 l RO Value: . l 3.9 l SRO Value: l4.3 l Section: l EPE l RO Group: l 2 l SRO Group: l2, l System / Evolution Loss of Residual Heat Removal System  !

Title:

i KA Knowledge of the operationalimplications of the following concepts as they apply to Loss of Residual Heat 1 Statement: Removal System:

Loss of RHRS during all modes of operation j R:ference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Risidual Heat Removal AOP 1.10.1 Caution 2 iss 3 A System Loss Rev5 I

OM 53C- AOPs LP-SQS-53.C 5 4 QIestion Source l New l Question Modification Method l Question Source Comments: l Material Required for Ex:mination:

Page 70 L________________________..__

Ques 3MDFoph: l floss c6@CU furing a Coss of power l 1B and IC Component Cooling Water Pumps [lCC-P-1B & IC] are BOTH racked to the Connect position on the DF bus.

Which of the following control switch positions describes when BOTH [lCC-P-lC) and [lCC-P-1B] will l

, fail to restart on.a D/G load sequence signal, following a DF bus undervoltage con.dition?

a. [lCC-P-1B]- After START, [1CC-P-1C]- After START l

b. [lCC-P-1B]- PULL-TO-LOCK, [lCC-P-lC]- After Start

c. [lCC-P-1B]- After STOP, [1CC-P-1C]- PULL-TO-LOCK

d. '[lCC-P-1B)- After STOP, [1CC-P-lC) 'After STOP Anst i b. l Exam Level: JS l Cogtnitive Level: l Comprehension. l Explanatio o of Answer KA: l 026 AA2.02 l RO Value: l2.9 l SRO Value: l3.6 l Section: lEPE l RO Group: { l l SRO Group: l1 System / Evolution Loss of Component Cooling Water

Title:

KA Ability to deterTnine and interpret the following as they apply to Loss of Component Cooling Water:

Statement:

The cause of possible CCW loss R;f;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Reactor Plant Component IOM-15.1.D 3 !ssue 4 and Neutron Tank Cooling Rev1 W:ter (CCRS)

Reactor Plant Component LP SQS-15.1 4 5 and Neutron Tank Cooiing Water (CCRS)

Q: estion Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination:

I

! Page 71 i

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- ~ ~ ~ - - - - - - - - ~

ymm Ucpng l Laing1sIcGImising twem Given the following conditions:

• Reactor power- 100%

• A leak develops on the reference leg for the controlling Pressurizer level sensor How will charging flow respond over next five minutes?

Charging flow will:

a. decrease to the minimum value.

.b. decrease and then return to the initial value,

c. increase to makeup for the loss through the leak. ,

-d.

increase to the maximum flow value.

Avs: l'a l Exam Level: lS l Cognitive Level: l Comprehension l l Explanatio n cf Answer KA: l 028 AKl.01 l RO Value: l 2.8* l SRO Value: l 3.1 l Section: l EPE l RO Group: l 3 l SRO Group: l3 System / Evolution . Pressurizer Level Control Malfunction

Title:

KA Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Statement: Malfunction:

PZR reference leak abnormalities Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Pressurizer and Pressure LP-SQS-6.4 1.D. I .f 9-10 4 14 RelicfSystem Reactor Coolant System - IOM-6.4.lf 12 4 6 Instnament Failure Procedure Question Source l New l Question Modification Method l Question Source Comments: l l M;terial Required for Examination:

)

i

! Page 72 i

l t

QuestE3Wp2 l /AT/ /MEms@G@ /A%yXC Given the following conditions:

l

• Reactor power - 100%

• Both feedwater pumps trip

. . .The reactor fails to trip.

i Which of the following describes when AMSAC should trip the turbine? -

a. Immediately after the feedwater pumps trip.

1 b.- Immediately after feedwater flow decreases below 2S% flow.  ;

1 l c. 150 seconds after the feedwater pumps trip.

, d.- 25 seconds after feedwater flow decreases below 25% flow.'

A ss: ld >l Exam Level: lS- -l Cognitive Level: -l Memory l Explanatio d<f Answe~r j KA: l 029 AA2.09 l RO Value: j4.4 l SRO Value: l4.5 l Section: l EPE l RO Group: l 2 l SRO Group: l1 System / Evolution Anticipated Trantient Without Scram

Title:

KA Ability to determine and interpret the following as they apply to Anticipated Transient Without Scram:

Statement:

l Occurrence of a main turbine / reactor trip Reference Reference Number Reference Section Page Number (s) Revision Learn.

!' Obj ATWS Mitigation System 10M-458.1.B 1,2 iss 4 Rev Actuation Circuitry 0 j AMSAC LP-SOS-45.2 II.D.2.e 4 1 3 l i

Q1estion Source l New l Question Modification Method l Question Source Comments

l Material Required for Ermination:

r Page 73

Question Tc pic: l Evaluition of SR NIS voltage failure What would be the plant response to the following condit;ons?

c The plant is operating at 100% power call systems are NSA cThe "A". train-Source range RESET / BLOCK switch is inadvertently turned to the BLOCK position. ,

a. The reactor would trip, and N31 SR would energize

b. The reactor would not trip, and N31 SR would not energize.

c. ' The reactor would trip, and N31 SR would not energize.

d. The reactor would not trip, and N31 SR would energize Anst lb l Emant Level: -l S l Cognitive level: l Application- l Explanatio e cf Ans ver KA: l 032 AKl.01 l RO Value:-l 2.5 l SRO Value: l 3.l_ l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss of Source Range Nuclear instrumentation

Title:

KA Knowledge of the operational implications of the following concepts as they apply to Loss of Source Range St:tement: Nuclear Instrumentation:.

Effects of voltage changes on performance Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj UFSAR fig. 7.2 sheet 3 4 Rtactor Excore Instrument 10M-2.1.c Ill.E.3 7 iss 4 Rev System 1 Question Source l New l Question Modification Method l Question Source Comments: l M:terial Required for UFSAR fig. 7.2 sheet 3 Examination:

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Page 74

_ - - _ - . - - - _ - _ .- .-. -------------------A

1

- i Question Topic: l Eval of failed IR channel on SU Given the following conditions:

C Plant startup is in progress.  ;

• All power range channels indicate 6% reactor power.

]

' * ' Intermediate channel N-36 fails HIGH. 1

• Reactor power remains at 6%.

l 1

Which of the following describes required operator actions? '

. a. Initiate a reactor trip, en.ter E-0, and FR-S.I. l

b. Immediately commence a controlled reactor shutdown.

c. Raise power.to greater than P10 and block both' intermediate fanges.

d. Continue power operations. ,

Ans: Ib l Exam Level: IS l Cognitive Level: 1 Memory l l l

Explanation of Answer KA: l2.1.1 l RO Value: l3.7 i SRO Value: l 3.8 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss ofIntermediate Range Nuclear lustrumentation ,

Title:

KA Conduct Of Operations St-tement:

Knowledge of conduct of operations requiren nts. ,

Reference Reference Number Reference Section Page Number (s) Resision Learn.

Obi Excore Instrumentation LP-SQS-2.1 V.C.3.c & c 16-17 5 5,8,12 System Cenduct of Operations 1/20M-48.1.B VI.H.5 9 Iss 3 Rev 17 Conduct of Operations 1/2LP-SQS-48. I 6 Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination:

Page 75

QEGILTD VCpB l 1 Fell LKtndlntg 09eMGM MAR 3He OGqcuim A fuel assembly was ruptured during movement in the fuel building.

Which of the following describes how the fuel building evacuation alarm is actuated?

a. The alarm must be manually initiated from the control room.

b. [RM-1RM-206] and (RM-1RM-207] Fuel Pool Bridge Area Monitors will sound the evacuat'.on alarm.

c. [RM.-IVS-103A, B] Fuel Building Ventilation Exhaust monitors will sound the evacuation alarm.

d. The alarm must be manually initiated from either the fuel building or the control room.

Ans: Ic l Exam Level: lS l Cognitive Level: l Memory l Explanatio a cf Answer KA: l 036 AA2.02 l RO Value: l3.4 l SRO Value: l4.1 l Section: l EPE l RO Group: -l 3 l SRO Group: l3 System / Evolution fuel Handlingincidents ,

Title:

KA- Ability to determine and interpret the following as they apply to Fuel Handling incidents:

St tement:

Occurrence of a fuel handling incident R:ference Reference Numher Reference Section Page Number (s) Revision Learn.

Obj trradiated Fuel Damage AOP 1.49.1 C.2 2 Iss 3A Rev 3 OM 53C AOPs LP-SQS-53C.1 5 6 Question Source l Facility Exam Bank l Question Modification Method l QIestion Source Comments: l Material Required for l Examination:

Page 76 t

Questici Topic: l R':sponse of SG leak detrction monitors At what power level will the steam generator leakage N-16 Radiation Monitors [RM-MS-102A,B, & C]

BEGIN to provide valid leak rates, in GPD7

a. 5%

b. '20%

c. 30%

d. 50 %

A:s: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio acf Answer KA: l 037 AA1.06 l RO Value: 13.8' l SRO Value: l 3.9 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution,. Steam Generator Tube Leak l

Title:

KA Ability to operate and / .or monitor the following as they apply to Steam Generator Tube Leak:

Statement:

Main steam line rad monitor meters RIference Reference Number Reference Section Page Number (s) Revision Learn.

Obj 4 R:diation Monitoring 10M-43.1.C 8 iss 4 Rev Systems - Major components 2 OM 53C - AOPs LP SQS-$3C.1 6 Q:estion Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l M terial Required for Examination:

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1 Page 77 w_-_____--___

yuestmo vepB: l GeunumMisimMaceErg;rumGacuse@

i Given the following conditions:

l

• A Steam Generator Tube Rupture has occurred

• 53, Steam Generator Tube Rupture,is being performed

. * . The RCS has been cooled down to.the target temperature.

In order to maintain RCS subcooling, intact steam generator pressure must be maintained:

a. greater than the ruptured generator. l 1

l

b. equal to the ruptured generator.

i

c. greater than the saturation pressure of the RCS.

' d. less th'ah tiie ruptured generator.

lJ 1 Aus: ld } Eram Level: lS l Cugnitive Level: j Application l i l

Explanatio OcfAnswer '

KA: l 038 EA1.36 l RO Value: l4.3 l SRO Value: l 4.5 j Section: . ] EPE l RO Group: l 2 l SRO Group: l2 l

System / Evolution Steam Generator Tube Rupture

Title:

f KA Ability to operate and / or monitor the following as they apply to Steam Generator Tube Rupture:

Stat

ment:

Cooldown of RCS to specified temperature l Refirence Reference Number Reference Section Page Number (s) Revision Learn.

!< Obj l

Steam Generator Tube 82 IssIB 1 Rupture Background Rev 5 EOPs LP-SQS-53.3 3 Qzestion Source l Facility Exam Bank l Question Modification Method l Q: estion Source Comments: l 1 Mat: rial Required for i Ex:mination l

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Page 78 l

l L__________. _ _ _ _ - _ _ ._ _ _ _ _ _ __ __ _ _ . . _ _ _ _ . . . _ _ _ _ _ _ _ _ _ - _ _ _ _ - - - _ _ _ - - O

Question Topic: l Evaluati n cf FW condition , _ ,

~

Given the following conditions:

• A steam break has occurred on SG "A" l'

• A reactor trip was manually initiated

• A Si has NOT been initiated .

• No operator actions have been performed on the feedwater system.

. Only SG "A" narrow range level has decreased below 12%.

RCS T,,, are (A) 542 F, (B) $50 *F, (C) 550 F Which of the fcilowing is the expected status of feedwater?

a. The feedwater regulating valves will be shut. The Turbine Driven AFW pump will be running.

l

b. The feedwater regulating valves will be shut. All AFW pumps will be running.

c. A complete FWI isolation will be initiated. All AFW. pumps will be running.. I

d. The feedwater system will be in the same lineup as },.ior to the reactor trip, except the FRVs will be throttled closed; A s: la l Exam Level: lS l Cognitive Level: l Application l Explanatio  !

etf Answer j KA: l 040 AAl.02 l RO Value: l4.5 l SRO Value: l4.5 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Steam Line Rupture l

Title:

I KA Ability to operate and / or monitor the following as they apply to Steam Line Rupture:

Statement:

Feedwater isolation R:ference Reference Number Reference Section Page Number (s) Revision Learn.

Obj SG Feedwater System - 10M-24.lD Feedwater isolation 2,6 iss 4, instrumentation and Controls Rev.2 Reactor Protection System LP-SQS-1.1 V.E.5 38 6 9 Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination:

Page 79

Qrestici Tepic: l Effect & mitigation techniques Civen the following conditions:

• An uncontrolled depressurization of all steam generators has occurred

• Current RCS cooldown rate is 125 F/hr Which of the following describes how drying out, of the steam generators, is avoided while trying to. limit cooldown rate?

a. A minimum AFW flow to all steam generators is maintained.

b. SGs are intermittently fed to assure that a wide range levels remain above 10%'.

c. Only reducing AFW flow as necessary to reduce the cooldown rate to less than 100 F.

d. APM feed rate is limited 'to m'aintain constant level, provided the level is above f0% wide range.

Ass: l'a l Exam level: lS l Cognitive Level: l Comprehension l Explanatio c cf Answer KA: l 040 AKl.07 l RO Value: l3.4 1 SRO Value: l4.2 l Section: -l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Steam Line Rupture

Title:

KA. Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture:

Statement:

Effects of feedwater introduction on dry S/G Ref;rence Reference Number Reference Section Page Number (s) Revision L carn.

Obj Uncontrolled ECA-2.1 STEP 6 5 iss1B, Depressurization of all SGs rev.4 Uncontrolled 10M-53B.4.ECA 2.1 IV.6 25 iss IB; Depressurization of all SGs Rev.4 j Background l EOPs LP-SQS-53.3 3 )

Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: j Page 80

QIestion Tcpic: l Block of steIm dumps on turbine trip A loss of condenser vacuum has occurred due a leak in the condenser. Main Condenser Steam Dumps are cpen following a turbine trip.

l As vacuum decreases, at what condenser vacuum will Main Condenser Steam Dumps close?

i

a. 25" Hg Vacuum

b. 20" Hg Vacuum

c. 10" Hg Vacuum

d. 5" Hg Vacuum l

ATs: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio a ef Answer KA: l 051 AK3.01 l RO Value: l 2.8' l SRO Value: l 3.l* l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Loss of Condenser Vacuum

Title:

KA Knowledge of the reasons for the following responses as they apply to Loss of Condenser Vacuum:

Statement:

. Loss of steam dump capability upon loss of condenser vacuum R;f;rence . Reference Number Reference Section Page Number (s) Revision Learn.

Obj M in Steam Supply / Steam LP-SQS-21.1 25 4 3 Dump System 10M-26.2.B 10 4 7 Qrestion Source l New l Question Modification M4 thod l Q estion Source Comments: l ML:.terial Required for Exr.mination:

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Page 81 u_________ _ _ _ - _ _ _ - -

Q:estion T pic: l Det:rmin:ti:n of Feedline br:ak j A break has occurred on the feedwater line to SG "A" downstream of[MOV-FW-156A]. Main Feed Line Containment Isolation valve. Containment pressure increases to the Si setpoint.

Following the reactor trip and SI, which of the following SG pressure indications would exist?

s

a. Only SG "A" pressure would be decreasing from the break.

b. All SG pressures would be decreasing from the break via the main steam lines.

c. All SG pressures would be decreasing from the break via the main feedwater lines,

d. All SG pressures would be decreasing from the break via the auxil'ary i feedwater lines.

Avs: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio  ;

e ef Answer KA: l 054 AKl.01 l RO Value: l4.1 l SRO Value: l .43 l Section: l EPE l RO Group: l 2 l SRO Group: l2  !

System / Evolution Loss of Main Feedwater 4

Title:

I KA Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater:

Statement:'

MFW line break depressurizes the S/G (similar to a steam line break)

R:ference Reference Number Reference Section Page Number (s) Revision Learn.

Obj j Main Steam Supply / Steam LP-SQS-21.1 4 1,4g l Dump System  !

Main Steam System IOM-21.1.C 5 1ss 4 Rev 1

VOND 24-1 l

Question Source l New l Question Modification Method l Question Source Comments: l l Material Required for Examination:

Page 82

j i '

A loss of all 4KV busses has occurred. ECA-0.0 has been implemented to the point of placing deenergized

l. equipm:nt in PULL TO LOCK. The 1DF emergency bus has been selected to cross tie to Unit 2.

l  ;

l Which of the following l AE Emergency Bus loads shall remain in the AUTO position and the basis for l l le*.ving that pump in AUTO? -

l l

a. Reactor River Water Pump to assure that the diesel has cooling upon 'startup. ,

, b. Charging Pump to restore seal flow.

u

c. Charging Pump to restore Pressurizer level. -

d. Component Cooling Water Pump to restore cooling to the thermal barrier.

l \

Ans: - l a l Exam Level: lS -l Cognitive level: l Memory l l Explanatio c cf' Answer' l i KA: l 2.4.20 l RO Value: l3.3 l SRO Value: l4.0 l Section: l EPE l RO Group: l 1 l SRO Group: l1

~

System / Evolution Station Blackout

Title:

KA Emergency Procedures / Plan Statement:

Knowledge of operational implications of EOP warnings, cautions, and notes.

Refirence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Loss of All Emergency 4KV 10M-53A.I.ECA-0.0 Caution Step 14 10 1ss1B AC Power Rev 4 Emergency Operating LP-SQS-53.3 1 3 Proc dures Question Source l Facility Exam Bank l Question Modification Method l Qrestion Source Comments: l M:t: rial Required for Ex:mination:

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Page 83 L

Question Tepic: l Purpose of S1 Reset if an Si actuation signal is received when performing ECA-0.0, " Loss of All Emergency 4KV Power", the Si l signal should be: l l

a. reset to prevent lockout of the stub busses.

b. reset to permit manuel loading o'f equipment of an 8mergency bus. l 1

c. allowed to remain active to ensure rapid injection of core cooling water when power is restored. l

d. allowed to remain active to ensure the load sequencer re-initiates when the DG starts.

A s: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio )

acfAnswer KA: l 055 EK3.02 l RO Value: l4.3 j SRO Value:-l 4.6 l Section: lEPE l RO Group: l 1 l SRO Group: l1 System / Evolution Station Blackout

Title:

KA . Knowledge of the reasons for the following responses.as they apply to Station Blackout:

Statement:

Actions contained in EOP for loss of offsite and onsite power Refirence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Loss of All Emergency 4KV ECA-0.0 steps 31 & 37 22 & 25 iss 1B; l AC Power Rev 4 l Less of All Emergency 4KV 10M-53B.4.ECA-0.0 Step 31, Basis 127 iss 1B-AC Power Background Rev 4 l EOPs LP-SQS-53.3 3 Qrestion Source l NRC Exam Bank l Question Modification Method l Q1estion Source Comments: l M;terial Required for Ermination:

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Page 84

Questlom Tepie: l RCS temperatures What is the expected response of RCS Hot and Cold leg temperatures during the first few minutes following ,

a re ctor trip froml00% power COINCIDENT with a loss of offsite power?

l J

a. Hot leg temperatures will rise, and Cold leg temperatures will remain relatively constant, until natural circulation flow is established.

)

b. Hot leg temperatures and Cold leg temperatures will both rise, until natural circulation flow is established.

c. Hot leg temperatures will remain relatively constant and Cold leg temperatures will drop, until natural circulation flow is established.

d. Hot leg temperatures will rise r.d Cc!d leg temperature, will drop, until natural circulation flow is established.

ATs: la l Exam Level: lS l Cognitive L,evel: l Memory l Explanatio c ef Answer KA: l 056 AA2.18 l RO Value: l3.8 l SRO Value: l4.0 l Section: lEPE l RO Group: l 3 l SRO Group: l3 1

l System / Evolution Loss of Off-Site Power

Title:

KA Ability to determine and interpret the following as they apply to Loss of oft-Site Power:

Statiment:

Reactor coolant temperature, pressure, and PZR level recorders RIference Reference Number Reference Section Pagt Number (s) Revision Learn.

Obj Reactor Trip Response ES-0.1 Note before step 3 3-4 IssIB Rev 4 EOPs LP-SQS-53.3 1 6 Question Source l Facility Exam Bank l Question Modification Method }

QI stion Source Comments: l M terial Required for Examination:

l Page 85

I I

Questio2 Tepic: l Effect of a loss of Vital AC on Feedwater Given the following conditions:

• Reactor power is 74%

• Feedwater controlis in automatic

• Loss of a single 120 VAC Vital bus has occurred .

Which of the following descriks the expected response of Main Feedwater Regulating Valves which do l NOT remain in AUTO?

a. The FRVs will immediately fail open.

b. The FRVs will immediately fail closed.

l l c. The FRVs will drift shut.. .

l

d. The FRVs will transfer to either MANUAL or AUTO HOLD.

Ans: ld l Exam Level: lS l Cognitive Level: l Memory l Explanatio l D ef Answer j KA: l 057 AA2.19 l RO Value: l4.0 l SRO Value: l4.3 l Section: l EPE l RO Group: l 1 l SRO Group: l1 I System / Evolution Loss of Vital AC Instrument Bus

Title:

KA Ability to determine and interpret the following as they apply to Loss of Vital AC Instrument Bus:

Statement:

The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus j' Reference Reference Number Reference Section Page Number (s) Revision Learn.

l Obj Alarm Vital Bus I,11,111,IV 10M-38.4.AAA, AAC, 2 Trouble AAE, AAG 120V AC Distribution LP-SQS-38.1 32 6 6 System

~

Question Source l Facility Exam Bank l Question Modification Method l Qeestion Source Comments: l l M:terial Required for Examination:

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Page 86 l

1 1 i

ym1JEpB2 l L%f201012mafLEGBQLCJB Which of the following is the effect that losing 125 VDC Bus I will have on the Reactor Coolant Pumps?

a. One or two RCPs will trip on undervoltage.

b. One or two RCP breakers will ONLY be able to be tripped using the mechanical trip at breaker.

c. Component cooling water will be lost to all RCPs.

d. Seal water flow to the RCPs will be isolated.

Ars: lc l Exam Level: lS l Cognitive Level: j Application l Explanatio a cf Answer KA: 'l 058 AA2.03 l RO value: l3.5 l SMO Value: l3.9 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss of DC Power

Title:

KA Ability to determine and interpret the following as they apply to 1,oss of DC Power:

i Statement:

DC loads lost; impact on to operate and monitor plant systems Reference Reference Number Reference Section Page Number (s) Revision Learn. I Obj OM 39 1.0M-39. 5.B.6 Table 39-6 all Iss 4 Rev 5

Q:estion Source l New l Question Modi 0 cation Method l Q:estion Source Comments: l Miterial Required for IOM-39.5.B.6(28 pages)

Examination:

Page 87

Qrestiol Topic: l Ev:11 of Tcch Spec Given the following conditions:

• Unit 1 is in MODE 6

• Unit 2 is in MODE 1

e. Movement ofirradiated fuel is ongoing in the Unit 1 Containment only

• Monitor RM-IRM-218A Control Room Area - Unit I has failed low What action is required for th : above conditions?

a. No action is required because the monitor is not required to be operable.

b. Within ONE hour the respective Unit 2 control room monitor train shall be verified operable.

c. Within CNE hour, verify that Control Room Area - Unit 1 monitor [RM-1RM-21SB] is operable.

d. Wiiin UNE hour, suspend all operations involving movement ofirradiated fuel.

r s: lb l Exam Level: lS l Cognitive Level: l Applicati>n l Explanatio ocf Answer KA: l 061 AA2.06 l RO Value: l3.2 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l Sp() Group: l2 System / Evolution Area Radiation Monitoring System

Title:

KA Ability to determine and interpr;t the following as they apply to Area Radiation Monitoring System:

Statement:

Required actions if alarm channel is out of service Itference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Beaver Valley - Unit 1 3.3.3.1 Table 3.3-6,1.c, 3/4 3-33-3-35 Amend Technical Specifications Action 41 119 Radiatim Monitoring System LP-SQS-43.1 VI.A 31 4 7.a Q:estion Source l New l Question Modification Method l Question Source Comments: l kieriad Required for Tech Specs Examination:

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Page 88

Quest:0 3 Topic: l Effect cf restoring gir using IIA-90.

During a loss of containment air, which of the following is the possible effect of opening [llA-90]

Instrument Air to Containment Air Isol Valve too quickly?

a. Station Air compressor trips

b. CVCS letdown isolation

c. SG Main FW Feed Reg Vivs failing open I

d. Main Steam Line Trip Valve closure l

A"s: ld l Exam Level: }S l Cognitive level: } Memory l Explanatio a cf Answer KA: l 065 AK3.08 l RO Value: l3.7 l SRO Value: l3.9' l Section: l EPE l RO Group: l 3 l SRO Group: l2 System / Evolution . Loss ofinstrument Air

Title:

KA Knowledge of the reasons for the following responses as they apply to Loss ofInstrument Air:

Statement:

Actions contained in EOP for loss of instrument air l

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj ,

Loss of Containment AOP 1.34.2 Caution before step 4 3 lss 3A l Instrument Air Rev 3 OM 53C - AOPs LP-SQS-53C.1 5 4 Question Source l New l Question Modification Method j Q:estion Source Comments: l M:terial Required for Ex:mination: _

Page 89 l

L __ _ _ ___ __

Q:estion Topic: l Type ef detection / extinguishing eqpt for use Which of the follewing describes the fire protection c.fford:d for the primary process rack area?

a. Carbon Dioxide is released to the area by manual actuation only.

b. Carbon Dioxide i,s released to the area by automatic actuation of smoke detection or by, manual l actuation.

l

c. Halon is released to the area by manual actuation only.

d. Halon is released to the area by automatic actuation of smoke detection or by manual actuation.

Airs: Id l Exam Lesel: lS l Cognitive Level: l Memory l Explanatio n cf Answer KA: l 067 AA1.08 l RO Value: l3.4 l SRO Value: l3.7 l Section: l EPE l RO Group. l 1 l SRO Group: l1 l

System / Evolution Plant Fire on Site

Title:

KA Ability to operate and / or monitor the following as they apply to Plant Fire on Site:.

Statement:

Fire fighting equipment used on each class of fire Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Fire Protection System - IOM-33.1.B Halon paragraphs 1 & 4 5 1ss 4; Summary Description Rev.3 Fire Protection System LP-SQS-33.1 E.l b.4 29 5 3.e Question Source l New l Question Modification Method l QIestion Source Comments: l M;terial Required for Examination:

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Page 90

Question Tc pic: l Pressurizer level control l A fire in the control room has resulted in control room evacuation. Plant control has been transferred to local control panels as required by 10M-56C.1, Alternate Safe Shutdown from Outside the Control Room.

Until a cooldown is initiated from the BIP, pressurizer level is maintained by charging via: I l

a. [MOWRC-556A, B, C] Reactor Coolant Loop Fill Valves to the RCS loops.

l

b. the normal charging connection. l l

c. the RCP seals. )

d. the BIT.

Avs: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio a cf Answer i i

KA: l 068 AAI.30 l RO Value: l3.4 l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1 j System / Evolution Control Room Evacuation l I

Title:

KA Ability to operate and / or monitor the following as they apply to Control Room Evecuation:

Statement:

Operation of the letdown system Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj ,

Alternate Safe Shutdown LP-STA-56C.1 VI.A.3 12 2 6.a I from Outside the Control Room I

Q:estion Source lNew l Question Modification Method ]

Qrestion Source Comments: l M;terial Required for Examination:

l l

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Page 9I l

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Questio': Tople: l Contr:Ilerloc tion j Which of the following identifies the components used by the operator stationed at the BIP (Backup Indicating Panel) to lower pressurizer level?

a. {SOV-lRC-102B] RCVS Reactor Vessel Vent Viv

[SO.V-1RC-103B] RCVS Pressurizer Vent Viv ,

[SOV-IRC-105] RCVS Vent to Containment Isolation Viv j

b. [LCV-lCH-460A and B] Ltdn to Regen Hx Isol

[TV-CH-200B] 60 GPM Ltdn Orifice Cnmt Isol Viv Letdown will flow to the degasifier via [LCV-115A], which has failed to the degasilier position.

c. [MOV-CH-201] Excess Ltdn HX Inlet Isolation Viv l'

[MOV-lCH-137] Excess Ltdn HX Flow Control Viv

d. [PCV-lRC-455D] PZR PORV Relief Viv

[PCV-lRC-456] PZR PORV Relief Viv 1

A:s: la l Exam Level: lS l Cognitive Level: l Memory l Explanatio a tI Answer KA: l 068 AK2.01 l RO Value: l3.9 l SRO Value: l4.0 l Section: lEPE l RO Group: l 1 l SRO Group: l1 System / Evolution Control Room Evacuation

Title:

KA Knowledge of the interrelations between Control Room Evacuation and the following:

St:tement:  !

Auxiliary shutdown panel layout Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Misc. Safety-Related 10M-45.1.B (BIP) Indications 7 lss 4 Rev l Systems - Summary 1 Description  !

l Alternate Safe Shutdown LP-STA-56C.I 12 2 6 j l Outside the Control Room Reactor Coolant System - IC Mc6.1.D 5-6 Instrumentation and Controls Question Source l New l Question Modification Method l Question Source Comments: l M:terial Regnired for l Examination:

l Page 92 I

L_ - -- _ _ _ _ _ _ _

I Question Topic: l Basis for starting an RCP l An RCP is started in FR-C.1, " Response to Inadequate Core Cooling", in order to:

l l

a. allow using RVLIS Dynamic Range indication to determine core void content.

l .b. temporarily improve core cooling until some form of makeup flow to the RCS can be established.

c. enhance the cooling caused by rapid depressurization of the stear. :nerators.

d. establish pressurizer spray flow to reduce RCS pressure to cause Ic ." pressure systems to inject.

A s: Ib i Exam Levti: IS l Cognitive level: l Comprehension l Explanation of A?swer KA: l 074 EK2.01 l RO Value: l3.6 l SRO Value: l3.8 l Section: l EPE l RO Group: l 1 l SRO Group: l1 Syst;m/ Evolution inadequate Core Cooling

Title:

1 l

KA Knowledge of the interrelations between inadequate Core Cooling and the following:

Statement:

RCP l'

Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obi F.esponse to inadequate Core 10M 53D.4.FR-C.1 1 Iss1B Cooling Background Rev 4 Emergency Operating LP-SQS-53.3 1 3 Procedures Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Examination:

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Page 93 L__--__ _ _ _ _ _ _

i Qxestion Tepic: l Acti:ns to I:wer R/A levels Given the following conditions:

• . Reactor power has just been raised from 20% to 100% .

. - Dose Equivalent Iodine has just been reported as 5.0 pei/ gram.

Which of the following explains why operation can continue with Dose Equivalent Iodine above the Technical Specification LCO limit?

a. To allow for CVCS removal of the crud released by the power change,

b. The Technical Specification LCO limit is conservative enough, to allow extended periods (> 7 days) of exceeding the limit.

c. To accommodate the iodine that was released during the power change,

d. The probability of a Large break LOCA occurring during the time period Iodine is above the limit, presents an acceptable risk.

A's: -l c l Exam Level: lS l Cognitive Level: l Memory -l Explanatio ,

o rf Answer KA: l 076 AK3.05 l RO Value: l2.9 l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution - High Reactor Coolant Activity

Title:

KA Knowledge of the reasons for the following responses as they apply to High Reactor Coolant Activity:

Statement:

Corrective actions as a result of high fission-product radioactivity level in the RCS Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Technical Specifications LP-SQS-TS 0 4 Beaver Valley - Unit 1 Bases 3/4 4-4 B 3/4 4-4 Amend No 102 Question Source l NRC Exam Bank l Question Modification Method l Q estion Source Comments: l Material Required for ,

Examination: l l

i l

l I

i Page 94

Question Topic: l Securing Si flow Which cf the following describes the required subcooling requirements before terminating SI in ES-1.1, S1 Termination?

The required subcooling:

a. is based on saturation conditions plus instrument enors.

b. is based on the expected pressure after SI is terminated.

c. is based on the expected temperatures after SI is terminated.

d. provides for a 50 F margin to saturation to avoid reinitiation.

A:s: la l Exam Level: lS l Cognitive Level: l Memory l Explanatio cefAnswer KA: l E02 EK3.2 l RO Value: l3.3 l SRO Value: l3.8 l Section: lEPE l RO Group: l 1 l SRO Group: l1 System / Evolution S1 Termination

Title:

KA Knowledge of the reasons for the following responses as they apply to S1 Termination:

Stat: ment:

Normal, abnormal and emergency operating procedures associated with (Si Termination).

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj S1 Termination /Reinitiation lOM053B.5.GI l1 II.A.1 3 issIB Rev1 EOP Generic issues LP-SQS-53.2 1 LO3 Qrestion Source l New l Question Modification Method l Question Source Comments: l M;.t: rial Required for Examination:

4

! Page 95 l

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I Question Topic: l Basis for required Pressurizer Lev;l A reactor trip and SI have occurred, and the control room operators are responding to a small-break LOCA.

All RCPs are tripped. The operators have proceeded to the recovery stage in ES-1.2, " Post-LOCA Cooldown and Depressurization". A PZR PORV is used to depressurize the RCS until PZR level is greater than 18% [50% ADVERSE CONTAINMENT].

In addition to ensuring that RCS conditions are under adequate operator control, the basis for this pressurizer level ensures:

a. that a reduction in subcooling does not occur when SI flow is reduced.

b. sufficient inventcry such that PZR level does not drop low when an RCP is started.

c. pressurizer level indication is not due to a void in the vessel head. j 1

d. adequate PZR steam space to absorb pressure fluctuations during RCP start.  !

l A:s: lb l Exam Level: lS l Cognitive level: l Comprehension l l Esplanatio o rf Answer KA: l E03 EK2.2 l RO Value: . l 3.7 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l2 )

Syst:m/ Evolution LOCA Cooldown and Depressurization

Title:

l l KA Knowledge of the interrelations between LOCA Cooldown and Depressurization and the following:

Statement:

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.

Obj Post LOCA Cooldown and ES-1.2 step 15 10 IssIB Depressurization Rev 5 Post LOCA Cooldown and IOM 53B.4.ES-1.2 25 issIB l Depressurization Rev 5 )

i EOP Generic issues 1 LP-SQS-53.2 II.B.l. lil.A 5.10 3. 4 l Question Source l NRC Exam Bank l Question Modification Method l l

(, Question Source Comments: l Material Required for Ex:mination:

l l

l Page 96

l Qrest6o2 Tcpic: l Purpose of ECA-1.2 Given the following conditions:

o A small break LOCA has occurred due to - N;.k at some unknown location outside containment.

o Performance of ECA - 1.2 "LOCA Outs' .s Containment" did not isolate the break.

l o At the completion ECA - 1.2 "LOCA Outside Containment", RCS pressure is still dropping l At the conclusion of ECA - 1.2 "LOCA Outside Containment" the operating crew should transition to

a. E-0 "Rx Trip or SI" in order to reverify that all automatic actions have been completed, l

b. E-3 "SGTR", since there are adequate steps within this procedure to deal with these conditions.

c. ES-0.0 "Rediagnosis" in an attempt to diagnosis the break location.

d. ECA-1.1 " Loss of Emergency Coolant Recirculation", in order to deal with the loss of available l inventory for core cooling.

Ars: ld 'l Exam Level: lS l Cognitive Level: l Comprehension l .

Explanatio aaf Answer KA: l E04 EK2.2 l RO Value: l3.8 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l1 System / Evolution LOCA Outside Containment

Title:

KA Knowledge of the interrelations between LOCA Outside Containment and the following:

St tement:

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

R:Terence Reference Number Reference Section Page Number (s) Revision Learn.

Obj LOCA Outside Containment IOM-538.ECA-1.2 1 issIB Bickground Rev 3 Emergency Operating LP-SQS-53.3 1 1 l Procedures ,

l QIestion Source l New l Question Modification Method l Question Source Comments: l Miterial Required for Examination:

Page 97

i Questio] Topic: l Apply procedural direction f r cooldown

• During a natural circulati:n cooldown with RVLIS unavrilible, it is likely that voids will form in the upper ,

head region. ES-0.4 " Natural Circulation Cooldown With Steam Void in the Vessel (Without RVLIS)", l limits the size of these voids in the RCS head region by : l a.' Requiring all CRDM fans to be runnung.

b. Limiting the allowable increase in pressurizer level.

c. Limiting the maximum temperature on Core Exit Thermocouple.

d. Requiring a minimum of 200F subcooling.

A's: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio e of Answer KA: l E10 EA2.2 l RO Value: l3.4 l SRO Value: l3.9 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Natural Circulation with Steam Void in Vessel with/without RVLIS

Title:

KA Ability to determine and interpret the following as they apply to Natural Circulation with Steam Void in Vessel Statement: with/without RVLIS:

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj NaturalCirculation ES-0.4 step 9 8 Iss1B Cooldown With Steam Void Rev 4 in Vessel (Without RVLIS) i EOPs LP-SQS-53.3 3 Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l M:terial Required for l Ex:mination:

l Page 98 l

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Question Topic: l Condition resulting in loss cf recirc Given the following conditions:

• A LGCA has occurred a Due to low RWST level a transfer to Cold Leg Recirculation has occurred.

• All automatic actions for the transfer to Cold Leg Recirculation are complete.

• [ISI-P-1B] LHS1 pump is not available

• Containment pressure - 12.4 psig Which of the following would result in a loss ofinjection flow? l

a. RCS pressure - 450 psig  ;

[MOV-ISI-862A] 1 A LHSI Pump RWST Suct Viv fails open j

b. RCS pressure -250 psig

[MOV-ISI-863A) 1 A LHSI to Chg Pumps Sup Viv fails closed

c. RCS pressure - 380 psig

[CH-P-1 A] 1 A Charging /HHSI Pump trips i

[MOV-ISI-863B] IB LHSI To Chg Pumps Sup Valve fails closed.

I

d. RCS pressure - 180 psig i

[MOV-lSI-885A] 1 A LHSI PP Mini Flow Isol Valve fails open

~Ars: lb l Exam level: IS l Cognitive Level: l Comprehension l Explanatio  ;

Mf Answer I l

KA: l El1 EA2.1 l RO Value: l3.4 l SRO Value: l4.2 l Section: lEPE l RO Group: l 2 l SRO Group: l2 '

System / Evolution Loss of Emergency Coolant Recirculation

Title:

KA Ability to determine and interpret the following as they apply to Loss of Emergency Coolant Recirculation:

St:tement:

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Reference Reference Number Reference Section Page Number (s) Revision Learn.

Obj Transfer To Cold Leg ES-1.3 step 4 3 issIB Recirculation Rev 4 EOP Attachment 1-G IOM 53A.I.1-G step 2 2 issIB Rev 2 EOPs LP-SQS-53.3 6 Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination:

Page 99

Question Topic: l CIB setpoints How long after a CIB sign-1 is received will the qutnch spray and containment spray pumps start?

a. [QS-P-1 A,B] Quench Spray pumps - 5 seconds 1

[lRS-P-2A, B] Outside Recirc Spray Pumps = 120 seconds

{

[lRS-P-1 A, B] Inside Recirc Spray Pumps = 225 seconds

b. [QS-P-1 A,B] Quench Spray pumps - 60 seconds j

[lRS-P-1 A] Inside Recire Spray Pump, [lRS-P.2B] Outside Recirc Spray Pump = 120 seconds j

[lRS-P-1B] Inside Recirc Spray Pump, [lRS-P-2A] Outside Recirc Spray Pump = 210 seconds I

c. [QS-P-1 A,B] Quench Spray pumps - 60 seconds

[lRS-P-1 A, B] Inside Recirc Spray Pumps = 210 seconds

[lRS-P-2A, B] Outside Recirc Spray Pumps = 225 seconds

d. [QS-P-1 A,B] Quench Spray pumps - 5 seconds

[lRS-P-1 A] Inside Recirc Spray Pump, [lRS-P-2B] Outside Recirc Spray Pump = 210 seconds

[lRS-P-1B] Inside Recire Spray Pump, [lRS-P-2A] Outside Recirc Spray Pump = 225 seconds Ans: Id l Exam level: IS I Cognitive level: l Memorv I Explanation of Answer KA: l E14 EKl.3 I RO Value: l3.3 l SRO Value: l34 l Section: l EPE l RO Group: l 1 I SRO Group: l1 System / Evolution High Containment Pressure

Title:

KA Knowledge of the operational implications of the following concepts as they apply to High Containment Pressure:

Statement:

Annunciators and conditions indicating signals, and remedial actions associated with the (High Containment Pressure).

Ref:rence Reference Number Reference Section Page Number (s) Revision Learn.

Ob]

Containment IOM-13.2.B 2 1ss 4 Rev Depressurization System 3 ,

Containment LP-SQS-13.01 27 5 5 Depressurization System l Question Source l New l Question Modification Method l I Question Source Comments: l M:terial Required for Examination:

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