ML20217E738

Forwards Operator Licensing Exam Administered on 970317-20 W/Outline & Initial Exam Submittal Designated for Distribution Under Rids Code A070

ML20217E738
Person / Time
Site:	Indian Point
Issue date:	10/02/1997
From:	Curley V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9710070162
Download: ML20217E738 (201)

Text

. . . _- - - - . ..

October 2, 1997 '

NOTE T0: NRC Document Control Desk Mail Stop 0 5 0 24 FROM: ho r'l bon len . Lic sing Assistant

• Opera';ing Licensirp" Branch, R

SUBJECT:

OPERATOR LICENSING EXAMINATION ADMINISTERED ON March 17-20, 1997 . AT Indian Point 2 '

DOCKET #50- 247 (Written Retake Exam)

On March 17-20. 1997 Operator Licensing Examinations were administered at the referenced facility. Attached, you will find the following informtion for processing through NUDOCS and distribution to the NRC staff, including the NRC POR:

Item #1 - 'a) Facility submitted outline and initial exam submittal, designated for distribution under RIOS Code A070.

b) As given operating examination, designated for distritation under RIOS Code A070, Item #2 - Examination Report with the as given written examination attached, designated for distribution under RIOS Code IE42.

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. GeneralPhysics Corporation 6700 Alexander Bell Drive Columbia, Md. 21046 (410) 290 2300

-_ Knowledge and Abilities Record Form Plant Systems

. indian Point Unit 2- Reactor Operator Written Exam Group i Plan; Systems (23 Points) 001 K4.07 (1) Reactivity Control Control Rod Stops 3.7 001 K5.06 (1) Reactivity Control Effects cf Control Rod Motion on Axial Offset 3.8 004 A2.32 (1) Reactivity Control Expected reactivity changes after valving in a 3.4 new mixed bed dcmineralizer that has not L.bcen preborated .

. . . .003 K3.04 (4)Rx,,H, cat Removal RCP failure effect on RPS 3.9 059 A4.12 (4;Rx Ilcal Removal Initiation of automatic feedwater isolation (Iow 3.4

[ __ . _ __

Tave) 061. j K4.04 _ (4)Rx Heat Removal _ Prevention of AFW pump runout by limi ing 3.1

.L AFW flow l 015 y K.4 05 ,J. Reactor Trip from NIS , 4,3 i 015 "4.07 ', .I.1711nstrumen!alion_

(7)lnstrumentation Permissives 3.3

'"(i;T "t"d4 01 (7)instrumentatio'n inputs to subcooling monitors 3.4 072 A1.01 ] (7)lnstrumentation Changes in ARM readings due to fuel failure 34 068- A4.03 (9) Radioactive Release Stoppage if release limits exceeded 3.9 071 A4.26 q (9 Radioactive Release Authorized waste gas release conducted in 3.1 compliance with radioactive gas discharge

[. +. permit _ ._ _

l004 7 K6.13 ...l (1) Reactivity Control Purpose / Function of boration/ dilution controls 3.1 A l .01 - i (2)lnventon Control _R CS Pressure and Ter. Mature g 4.0

{013. .

p056 [ A2._0_4_ ! (4)Rx llcal Removal Loss of condensate, pumps _2_.6

022. A2.05 i (5) Containment integ. Major leak of CCS (service water) potential 3.1 I

[ impact on containment integrity l015 K3.02, ,3,17]Instru.me-tation , Loss of NIS impact on CRDS , 3.9 1

071 A3.02 - (9) Radioactive Release Radiation monitoring system alarm and 3.6 l actuating signals 001 K4.01 (1) Reactivity Control Rod Position Indication (P/A converter 3.5 operation) 003 K5.01 (4)Rx Heat Removal Relationship between RCS flow rate and the 3.3 nuclear reactor core operating parameters (DNBR) 059- , K 1.04 (4)Rx Heat Removal SG water level control system 3.4 y

001- K6.11 (1) Reactivity Control Location and operation of CRDS fault 2.9 detection (urgent failure)

[ 013 A202  ! (2 Nnventory. Control , Excess Steam Demand . _4.3 j

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I GeneralPhysics Corporation 6700 Alexander Bell Drive Columbia, Md. 21046 (410) 290-2300 Knowledge and Abilities Record Form Plant Systems Indian Point Unit 2 Reactor Operator Written Exam Group 11 Plant Systems (20 Points)

K5,15 002 (2)RCS Inventory Reasons for maintaining subcooling margin 4.2

+ .

dug.ing. natural circulation 002. . . . . _ K4.05 _}.......___(2)RCS Inventory J Detection of RCS Leakage

3. 8,

, 011 A 1.01 } (2)RCS Inventory  ; PZR level and pressure ,,_,[3.5 010 g A3.02 (1)RCS Pressure Cont. J _PZR Pressure (OPS sy stem) 3.6 902, ,,.._, . . ,K 1 :1.1.,.. (4 )Ry. pcat.Rc!ngp,,,.

s.035. . K6 01 _

(41Rs Heat Remoial

,,SG.s/Maip Failure of MSIVs Feedw alets.RCS Integyclationship.. 3.2

,,,}. 4

!006 A1,18 (3)Rx Pressure Cont. Per level and pressure response (S! reduction) 4.0 012. K4.06 (7)lnstrumentation Auto / Manual enable /disabic RPS trips 3.2

.. (permissives) __ _ ,,4

,j}i2. . . . . . A3.06 .J7)lnstrumentation Trip _Logicjbistable status) ,.,j 3.7

, _A2 01 014 i i1) Reactivity Control Loss of offsite power and Loss of Power to 2.8 A202 l RPIS with Reactor Trip (verification) 3.1 A2 05 i 3.9 5.026 A4.05 _ jj$) Containment I,nteg [ Containment Spnq reset switches .,,3)

, K 1.03

(!29., [(8) Plant Sc. nice SystenQ Engineered Safeguards

_ . _ 3.6_

075 A2.03 (8) Plant Service System Safety features and relationship between 2.5

, condenser vacuum, turbine trip, and steam dump (CWP interlock uit HP stm dump)

[ OS6 K4 07 (8,) Plant Service System M T/G at.d T/G Trip 2.5 086 A2.02 (8) Plant Sc.Tice System Low FPS header pressure (sequence of pump 3.0 starts)

039 K4.08 (4)Rx Heat Removal I nterlocks i on MSIV and bypass valves (MSIV " 3.3 i 86 devices) 062 i K101 l (6) Electrical EDds 4.1

, 062 I A3.01 1 (6) Electrical Vital bus amperage (6.9 KV bus amps vs 480 3.0

' i L....... _ _ ...VAC bus amps.)..._.. ,..

[ 063 IK302 j J6) Elect,ncal . j Components using,DC control power !3.5 y ,

064 A2.16 l (6)Electncal Loss of Offsite Power during full load testing f3.3

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General Physles Corporation 6700 Alexander Bell Drive Columbia, Md. -21046 (410) 290 2300 Knowledge and Abilities Record Form Plant Systems l

, Indian Point Unit 2 Reactor Operator Written Exam Group 111 Plant Sys1rms (8 Points) j 005 K2 01 (4)Rx lleat removal Bus power supplies RilR pumps 3.0

[ 041 A4.08 - J4)Rx Heat Removal Steam Dump Valves (IIP) 3.0 045 A I.05 (4)Rx Heat Removal Expected response of primary plant parameters

' 3.8 following a TG trip

$08 h4.01 (8) Plant Scnice System Automatic start of standby pump 3.1 076 K4.06 (4)Rx lical Removal Automatic stan features of SW pumps 2.9 p (blackouO

!103 K4.06 _ jj5) Containment irtec Containment isolation S3 stem 3.1

[103 K302 [(5) Containment Integ. ' Loss of Containment integrity during normal 3.8

[ ,_,  ! operation 005 Al.02 { (4)Rx licat Removal RiiR Dow rate (parameters /lincups which are 3.3 used_to determine min / max Dow) a 4

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1 GeneralPhysics Corporation  !

6700 Alexander Bell Drive Columbia, Md. 21046 (410)290 2300 Knowledge and Abilities Record Form Plant Systems

. Indian Point Unit 2 Reactor Operator Written Exam Group i Emercency and Abnormal Plant Evolutions (16 Points) tKHK)$5 EK 1.02 Loss of all AC _ Natural Circulation Cooling 4.1 3XM174 - EK3.04 Inadequate Core Cool Tripping of RCPs 3.9 p.gKNM)5 AK3.04

_,lnoperable Rodis)

TeclySpec linyits for inoperable rods 3.4 0(MKil5 A A 1.22 RCP Malfunctions RCP seal failurchnalfunction 4.0

~UK124

(

. AK3.01 "Eniergen_edioration i When emergency 1, rapid) boration is reauired 4.1 (KMK)26 AK3.03 Loss of CCW Guidance contained in EOPs for loss of 4.0

_ _ . __ CCW/ nuclear senice water (MMK127 _.AA2 04 Pzt Press Malf. Tech Spec limits RCS pressure (DNB related 3.7

_ _ _ _ _ pa_rameters applicability)

I tKMH140 AA2.01 l Sim Line Rupture Occurrence and lo:ation of a steam line 4.2 I . rupture from pressure and flow indications (MNK151 AA2.02 i Lossof V cuum l Conditions requiring reactor and'or turbine 3.9 i trip

" OtHMI57 " AA2.19 Loss of inst. Bus l Plant automatic actions that will occur4.0 on the

. ........ . _ ...l. loss of ny,i.tal. AC, e,lecir]c,al lostrui,nent Bus

[ OtNd)67 i AK1.01 . Fire on site. . . _ . . i Fire classifications bLtype . , , _2.9 (kK1068  ! AK3.12 CCR Evacuation Required sequence of actions for emergency 4.1 tKHK)69

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evacuation of control room Isolation valves, dampers, and electro- 3.5 g, . _

f _AA1.01 l Loss of Cont.'Integ.

i

.. pneumatic devices (two is true) l 0(M1076 AK3.05 } liigh RCS Activity Corrective actions as a result of hign fission 2.9

[ i product activity levels in RCS (MKKK)5 AK3.01 Inop1 Stuck Rod Boration/cmcrgency boration requirements in 4.0 the event of a stuck rod during a trip (KKK)l5 AK3.07 RCP Failure Ensuring that SG levels are controlled 4.1

, ,propedy for natural circulation cooldou n .

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GeneralPhysics Corporation 6700 Alexander Bell Drive Columbia, Md. 21046 (410) 290 2300 Knowledge and Abilities Record Form Plant Systems indian Point Unit 2 Reactor Operator Written Exam a

Group 11 Emergtney and Abnormal Plant Fvolutions (17 Points)

(KKXX)7 EK3.01 Reactor Trip Actions contained in EOP for cactor trip 4.0 m_ _ -._____

a................~4............. . _ _ _

...@O fpg..steg .lL_

(KKKK19 l EK3.04 SBLOCA Staning additional charging pumps (bases for 4I establishing max charging flow before SI

......................... ..!cd351 ion) , , , , , , , , _ .

,(xkk)II.. . . . . . . . .i .EA2.10 ... LBLOCA Verification of adequate core cooling 4.5 (HMK129 j EK3.06 f ATWS _  ! Reason for tripping turbine 4.2 g

(KKK138 E A 1.36 iSGTR ' Reason for cooldown of RCS to specified i4.3 i

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AK1.17

. _ . . . . . , . . . . . . _ . _ . _ . . _ . . ...... , .l.C".!PCEuurc Cont. Rod Withdrawal

_.{

Difference in response EOL s s BOL (h1TC) 3.4

' (MHMK13 AA202 ' Dropped Rod Signal inputs to sad control sy stem (alarms on 2.7 l , _ _ . . . rod withdra val)

!(KWKK)S AA2.03 l Vapor Space LOCA RCS Pressure / temperature indicators / alarms j 3.9

.-...................;......._......,......................... ..,...[PORV tailppe.te!npl,.

,,4 _ _ _

(MK1022 AK3.02 j Loss of RCS hiakeup 35 Actions in AOl for loss of chargi,ng/ makeup

'j (KXH)25 j A A l .02 Qoss of RilR RCS Im entorv (mid-loop) 3.8 i (HKK)32 AK301 l Loss of SR NIS Stanup termination on fource range loss 3.2 i

. (minimum SR required for S/U) g(KKX)33 AK3.02 ; Loss ofIR NIS i Guidance in EOPs on loss ofIR (manual re. 3.6

...;. encryize. of SR inst) . . . _ . . . _ , _ ...,

000037 . . . . . . . . .AK  ! SG 1.01Tnbc Lenk- - . . . . . . . .' .Leak . . . - .rate

. . . .vs.. . .Pressure drop (full pouct vs. No t 3.5 3._......... .................:........ . . . . . . . . . . . . _ . . . . ..

..lpadL _ _ _ _ _ _ _

,,,._,..4.,

l (MHK)54 AK l.02 i Loss of h1FW EITects of feeduater introduction on dry SG 3.6 l.

I

!h9Ldn;SG) 4

! (HXXI38 . . EA2.03 SGTR ..,,, ... ,Which__SG is ruptured [EOP definition)_ 4.4 (XXiOO9 i EK3.28 SBLOCA h1anual ESF actuation (Sl reinitiation) 4.5 (KKK)I1 ,j EK3 ii [Uii'UU " i'55lh injection /recirculatioA -~ ~"' ~'38 1

4 GeneralPhysics Corporation 6700 Alexander Bell Drive Columbia, Md. 21046 (410) 290-2300 Knowledge and Abilities Record Form Plant Systems indian Point Unit 2 Reactor Operator Written Exam Group 111 Emergency and Abnormal Plant Evolutions (3 Points)

AK3.03 MM PZR Level Malf. False indication of PZR level when PORV or l000028 3.5

.l

} .. .. . .. .. - - - - - - - - - . - - . . . . . _ . . . . . . _ . . . . . . . . . . . . .sgaryal e, l opened and RCS saturated ,,._,._. ...

(MKK136 . ,AK2jl,_I __ _ l Fuct (landl,ing Acc;. _, l'ucl handling cquipment. ..,,, . 2.9 (KHK)56 AK3.02 l Loss of Offsite Power Actions contained in EOP (verification of 4.4 natunal. gireulation flow)

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J GeneralPhysics Corporation 6700 Alexander 13 ell Drive Columbia, Md. 21046 (410) 290-2300 Knowledge and Abilities Record Form Plant Systems Indian Point Unit 2 Reactor Operator Written Exam Plant Wide Generic Responsibilities (13 Points) 294001 Kl.01 Valve Lineups Knowledge of how to conduct and verify valve 3.6 lineups (second verification) 294001 Kl.02 Stop Tags Knowledge of tagging and clearance 3.7 procedures (what to do if tag is damaged or

.. lost) 294001 l Kl.14 Confined Spaces Knowledge of safety procedures related to 31 3 j_ confmed spaces 294001 K1.16 Fire Equipment Knowledge of facility protection requirements 3.5 including fire brigade and portable fire

,.-- ..,- fi .I.!ti,n.g S cquip. ment _ _ _ _ __ _ _ _ __

i 294001 K1.04 ALARA Knowledge of facility ALAR requirements 3.3 . _ __l

,., (RWPs)

GEN __ _

K2.1.3 Shift,, Turnover ___, K,nowpdge_of shift turnover requirements 10,,

4 294001 A l.01 Procedures Abihty to ob in and verify control procedure f3.3

,. COPLQP, Cst, ,,, ,,,, ,,,, , ,J .

294001 A1.04 _ _ _[ . ... ., Communications Ability to use plwt phone, paging system, 3.0

-[ . .

._ d ._ . . _ . . . _ r

.,_ adios (RECS. ENS etc.)

294001 A 1.16 i E Plan Ability to take actions called for in the facility 3.1

Emergency Plan (Emergency Class i

..............J _

Definitions) _

. . . . ,294001 K 1.09 l Safety

, Knowledge of safety procedures related to high 3.4 4 1

, pressure (confined spaces) 1 294001 K l.15 Safety Knowledge of safesy procedures related to 3.4 1

' [ ~

hydrogen (flammability limits) 294001 K l.05 Security Knowledge of facility requirements for 3.1 controlling access to vital / control areas.

(definitions for vital / protected / owner

, ,. J. , controlled) 294001 Al.06 ' Conduct of OPS Ability to maintain accurate, clear and concise 3.4 logs. records, status boards, and reports (log keeping requirements - OOS etc.)

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Reactor Operator Exam Answer Key - 2/24/97

1. C 26. D 51. A 76. A

2. B 27. B 52. B 77. D

3. C 28. C 53. B 78. D 4 C 29. A 54. D 79. A

5. B 30. C 55. R 80. C

6. A 31. B 56. C 81. B

7. D 32. D $7. A 82. C

8. C 33. B 58. D 83. A

9. B 34. C 59. B 84 D j 10. D 35. A 60. C 85. B l  ! 1. A 36. D 61. D 86. A

12. D 37. B 62. A 87. C

13. C 38. B 63. C 88. D

14. B 39. D 64. D 89. B L5. B 40. C 65. B 90. C

16. C 41. A 66. C 91. D

17. D 42, B 67. B I 92. A

18. B 43. D 68. C 93. B

19. A 44. D 69. D 94. D

20. C 45. A 70. A 95. A

21. D 46. D 71. C 96. C

22. B 47. D 72. B 97. B

23. D 48. C 73. D 98. C

- 24. B 49. B 74. B 99, D

25. A 50. D 75. B 100. D j

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indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION i

Consider the following initial conditions when selecting your answer:

• Reactor Power 100 %

• RCS T.,, 559 F e Control Rods Automatic / Bank D @ 215 Steps

.

• RCS Boron 1070 PPM e Chemistry has advised the control room that #21 CVCS Mixed Bed Demineralizer resin is exhausted. The Reactor Operator is directed to coordinate with the Nuclear NPO to place #22 CVCS Mixed Bed Demineralizer in service.

\~ *

1. 22 CVCS Mixed Bed Demineralizer resin was replaced last week. The demineralizer has remained isolated since resin replacement. No operations have been performed on 1

this demineralizer since that time.

Which of the following statements describes the effect, if any, that placing #22 CVCS

Mixed Bed Demineralizer in senice without saturating at existing RCS boron concentration will have on the following RCS parametere

1

• RCS T.,,

• Control Rod Position j -* RCS Boron Concentration

A. RCS T.,, DECREASE Control Rod Position INCREASE RCS Boron Concentration INCREASE B, RCS T.s., - NO CHANGE Centrol Rod Position NO CHANGE RCS Boron Concentration NO CHANGE C. RCS T.,, INCREASE Control Rod Position DECREASE RCS Boron Concentration DECREASE D. RCS T.,, DECREASE Control Rod Position DECREASE RCS Boron Concentration INCREASE-F

Page 1 of 100

Indisn Point Unit 2 l

, Consolidated Edison Ccmpany ofNY l i

1

, REACTOR OPERATOR EXAMINATION Which of the following statements correctly describes the status of the Main and Low

, Flow Feedwater Regulating Valves after a reactor trip from 100% power and subsequent

, five (5) minute cooldown to 530'F due to a stuck open atmospheric steam dump valve?

4 I

A The Main Feed Regulating Valves are CLOSED, and the Low Flow Feed Regulating Valves are 75% OPEN (100% power position).

B. The Main Feed Regulating Valves AND the Low Flow Feed Regulating Valves 4

se CLOSED.

2-C. The Main Feed Regulating Valves are full OPEN due to large level error signal caused by " shrink". The Low Flow Feed Regulating Valves are 75% OPEN

_ (100% power position)

D. The Main Feed Regulating Valves are full OPEN due to large level error signal

caused by " shrink". The Low Flow Feed Regulating Valves are CLOSED.

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3 Page 2 of 100

' indian Point Unit 3 Consolidated Edison Company ofNY REACTOR OPERATOR EXAh11 NATION Consider the following initial conditions when selecting your answer:

• Reactor Power 100 %

• . RCS T.,, 559 F

• Control Rods - Automatic / Bank D @ 215 Steps The narrow range hot leg RTD for Loop 23 fails instantaneously high due to an open circuit. Which of the following statements correctly describes the effect of this failure on the Rod Control System?

l A. Rods automatically insert to restore indicated T.v. to Tur. Rod withdrawal is blocked in AUTOMATIC and MANUAL..

B. No rod motion occurs. Rod insertion and withdrawal are blocked in AUTOMATIC i only. MANUAL rod insertion and withdrawal are available.

C. No rod motion occurs. AUTOMATIC rod insertion and withdrawal are blocked.

M ANUAL rod insertion is available. MANUAL rod withdrawal is blocked.

D, Rods automatically insert to restore T.,, to Tor. AUTOMATIC and MANUAL rod insertion and withdrawal are blocked.

i Page 3 of 100 l

1

indian Point Unit 8 Consolidated Edison Company ofNl' REACTOR OPERATOR EXAMINATION Consider the following event when selecting your answer:

• A small break LOCA has occurred

• The reactor was tripped from 100% power

• A Safety injection was initiated due to low Pressurizer level Emargency Operating Procedure ES-1.2, Post LOCA Cooldown and Depressurization is being used to reduce safety injection flow.

Which of the following statements correctly describes the anticipated respcnse of Pressurizer level and pressure immediately after the fi;st Safety injection Pump is stopped?

- A. Pressurizer level AND pressure will INCREASE due to voiding in the reactor head.

B. Pressurizer level will DECREASE, Pressurizer pressure will INCREASE as water in the Pressurizer flashes to steam.

C Pressurizer level AND pressure will DECREASE due to reduced injection flow.

D. Pressurizer pressure will DECREASE and Pressurizer level will INCREASE due to voiding in the reactor head.

Page 4 of 100

d Indian Point Unit 2 Consolid:ted Edison Company ofNY REACTOR OPERATOR EXAhllNATION Which of the following statements describes the purpose and operation of the hiain Steam isolation Valve (MSlV) 86 Relays?

t

?

A. The MSIV 86 Relay actuates only when the CCR control switch is placed in the .

L CLOSE position to ensure the MSIV stays CLOSED.

i' B. The MSIV 86 Relay actuates when the MSiv does not indicate full OPEN and initiates a Turbine Trip to prevent a Safety Injection signal from being generated, i

C. The MSIV 86 Relay actuates only when a Main Steam Isolation signal occurs to ensure the MSIVs stay CLOSED.

D. The MSIV 86 Relay actuates when the MSIV indicates OPEN to seal in the solenoids to maintain the MSIV in the OPEN position.

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Page 5 of 100

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- Indian Point Unit 3 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider the following event when selecting your answer:

k

- - With the plant at 100% Reactor Power, control rod H 8 (center of core), drops

!' approximately 100 steps (62 inches) from the bank position of 215 steps. Which of the following statements correctly describes the effect,if any, that this event'will have on axial i core power distribution (41)?

i Assume that a turbine runback DOES NOT occur.

. - A. A I will become more negative due to reduced power generation in the top of the i core and increased power generation in the bottom of the core.

1-I B. A I will become less negative due to increased power generation at the top of the j and reduced power generation at the bottom.of the ore.

! C. A I will not change since rod H 8 is at the center of the core and will affect all quadrants equally.

l D. A 1 will initially become more negative then retum to its original value when

! positive reactivity from the power coefficient retums reactor power to its original value. ,

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( - - . . , . . _ . . - . . . ___,_ _ . . . . - .

Indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION During operation at 100% reactor power,6.9 KV Bus I normal supply breaker (UT-1) trips open due to a relay failure and a reactor trip occurs.- From the choices below, select the protection signal which initiated the reactor trip?

A. Two loop loss of flow B. 6.9 KV bus undervoltage C. Reactor Coolant Pump under frequency.

D. Single loop loss of flow Page 7 of 100

~ _ _ _ - _ _ _ . _ _ . _ . _ _ . _ _ . . _ _ ____.m. - __ . - . . .. _ . _ _ . . . - _

indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider the following event when selecting ycur answer:

Following a reactor trip from 100% power due to a loss of both Main Boiler Feed Pumps, the Motor-Driven Auxiliary Feedwater Pumps are being used to control Steam Generator (SG) Levels. Shonly after the AFW system is placed in service the piping downstream of the Auxiliary Feed Regulating valve to #24 SG mptures. Which of the following statements correctly describes the automatic response of the AFW system to this failure?

A. The #24 SG auxiliary feed regulating valve will automatically close. Feed flow to

1. 21, #22 and #23 SG will not be affected.

B. The #22 SG and #24 SG auxiliary feed regulating valves will automatically modulate to prevent pump runout and mai .tain sufficient AFW discharge pressure to maintain auxiliary feed How to #22 SG Feed flow to #21 and #23 SG will not be affected.

C. The #23 SG and #24 SG auxi tary feed regult. ting valves will automatically modulate to prevent pump runout and m,.intain sufficient AFW dischage pressure to maintain auxiliary feed Dow to #23 SG, Feed flow to #21 and #22 SG will not be affected.

D. Since #23 and #24 SGs are supplicd by the same AFW pump, feed flow to both SGs will be completely lost. Feed Dow to #21 and #22 SG will not be affected.

Page 8 of 100

_. _ _ ~ _

Indian Point Unit 2 Consolidated Eduson Company ofNY REACTOR OPERATOR EXAMINATION Due to failure of a pre amplifier la the circuit for NIS Power Range Channel N42, the lower detector output ha failed to zero. Total indicated power from NIS Power Range Channel N42 indicates 51% with the Reactor at 100% power. In order to repair the pre-amplifier the channel must be removed from service by removing the instrument power fuses.-

- Which of the following statements correctly states the effect that this or ution will have

on the Power Range Nuclear Instrumentation 35 stem and Reactor Prott in System?

l j

A. The high flux bistables for N 42 will be inhibited from tripping. A minimum of two of

the remaining three power range channels must sense a high flux condition to trin the reactor.

B. The high flux bistables associated with N42 will trip. A minimum of one of the three remaining power range channels must sense a high flux condition to trip the reactor.

C. The high flux bistables associated with N42 will trip. A minimum of two of the three remaining power range channels must sense a high flux condition to trip the reactor.

D._ The high flux bistables for N42 will be inhibited from tripping. A minimum of one of the three remaining power range channels must sense a high flux condition to trip the reactor.

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Page 9 of 100

indisn Point Unit 2 Consolidated Edison Company ofNY j- REACTOR OPERATOR EXAMINATION Which of the following correctly identifies the instruments that provide the pressure signal for the determination of the subcooling value displayed on Flight Panel (FD)?

A. Pressurizer Pressare (Channels I and 11)

B. RCS OPS System Pressure Instruments (PT-413/433/443)'

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C Pressurizer Pressure (Channels 111 and IV) i-l '- D. RCS Wide Range Pressure Instruments (PT-402/403) i.

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Page 10 of 100

Indian Point Unit 2 Corisolldated Edison Company ofNl' REACTOR OPERATOR EXAMINATION Consider the following indications when selecting your answer:

• Reactor Power 100 %

e T4ve 559'F

• Control Bank D 215 steps / automatic Containment Area Radiation Monitor (R2) Alarming

• Containment Atraosphere Radiation Monitors (R41/42) Normal

• - Containment Area Radiation Monitor (R7) Alarming

• Charging Pump Cell Area Radiation Monitor (R4) Alarming e Sample Cell Area Radiation Monitor (R6) . Alarming

• Normal Plant Vent Radiation Monitors (R43/R44) i From the list below select the event which would explain the indications described above?

A. Reactor core fuel cicmcnt failure B RCS leak in containment C. RCS/CVCS leak in charging cell

, D. RCS/CVCS leak in sample cell Page 11 of 100

Indian Point Unit 2

Consolid:ted Edison Company ofNY REACTOR OPERATOR EXAMINATION While discharging #13 Waste Distillate Storage Tank (WDST) to the river using #14 r- Waste Distillate Transfer Pump (WDTP), an alarm occurs on Panel SAF 1,"R 54 LIQUID WASTE DISTILLATE HI RAD / TROUBLE" due to a high radiation signali i -- - = Which of the followin; correctly identifies the anticipated response of the following Liquid Radwaste System components:

• #13 and 14 WDTPs l e #13 and 14 WDTP Discharge Valves (SOV CT-965-MCV and SOV CT-982 MCV)

Common WDTP Discharge Valve (SOY CT-971-FCV) 4 A #13 WDTP RUNNING l #14 WDTP STOPPED -

l #13 WDTP Discharge Valve (SOV CT 965 MCV) CLOSED

1. 14_WDTP Discharge Valve (SOV CT-932 MCV) CLOSED

, Common WDTP Discharge Valve (SOV CT-971-FCV) OPEN i~

j - B, #13 WDTP STOPPED

1. 14 WDTP STOPPED

1. 13 WDTP Discharge Valve (SOV CT 965 MCV) CLOSED

1. 14 WDTP Discharge Valve (SOV CT-982-MCV) CLOSED j Common WDTP Discharge Valve (SOV CT-971-FCV) OPEN 4

C. #13 WDTP RUNNING

! #14 WDTP STOPPED

1. 13 WDTP Discharge Valve (SOV CT-965-MCV) OPEN

1. 14 WDTP Discharge Valve (SOV CT-982-MCV) CLOSED i Common WDTP Discharge Valve (SOV CT-971-FCV) CLOSED 6

f - D. #13 WDTP STOPPED l #14 WDTP _

STOPPED

, #13 WDTP Discharge Valve (SOV CT-965-MCV) CLOSED E #14 WDTP Discharge Valve (SOV CT-982-MCV) CLOSED

"~

Common WDTP Discharge Valve (SOV CT-971-FCV) CLOSED I

Page 12 of 100

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Indian Point Unit 3 \

Consolidated Edison Company ofNY.

REACTOR OPERATOR EXAMINATION The Senior Reactor Operator has directed you to initiate a Containment Building Pressure Relief. Approximately 15 minutes after the release has been initiated the S-133B, i

Meteorological Data Display on the Accident Assessment Panel stops functioning. Which

of the following actions should you take to compensate for this failure?

! A. Immediately stop the release and have no further releases until the display is repaired.

! B. Record the Plant Vent Radiation Monitor (R44) reading every hour until the release is j terminated,

C. Verify that meteorological data is available and record meteorological data every hour j throughout the remainder of the release.

D. Stop the release and prepare a new release permit using the most adverse meteorological conditions.

4 4

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i Page 13 of 100 5

Indian Point Unit 2

Consolidated Edison Company ofNY v

n REACTOR OPERATOR EXAMINATION I

During power operation the CVCS Automatic VCT Makeup System initiates blended makeup due to the VCT level reaching the low level setpoint. An air leak on the supply i

line to Boric Acid Flow Control Valve, FCV-Il0A, causes air pressarc to the valve I diaphragm to decrease to 0 PSIG and prevents the valve from responding to the Hight panel controller (FIC-110A). The air leak is small enough that it does not have any significant effect on ir.strument air header pressure, t

i Which of the following statements correctly describes the effect, if any, that this failure will have on the boric acid concentration of the Reactor Coolant System?

i 4

- A. RCS Boric Acid Concentration will DECREASE because FCV Il0A fails CLOSED on a loss of air pressure, cc.using blended makeup to have a lower than desired boric acid concentration.

I B. RCS Boric Acid Concentration will INCREASE because FCV-110A fails OPEN on a loss of air pressure, causing blended makeup to have a higher . nan desired boric acid

.I concentration, f

4 C. RCS Boric Acid Concentration will NOT CliANGE because FCV-110A is only used j when the Makeup Selector Switch is in the BORA'l E position.

D. RCS Boric Acid Concentration will NOT CHANGE because FCV-110A is only used

when the Makeup Selector Switch is in the M ANUAL position.

4 i

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t Page 14 of 100 u

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Indian IcInt Unit 2 Cons: lid:ted Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider the following conditions when selecting your answer:

A small break LOCA has occurred e RCS pressure is at 1900 PSIG e

The Safety injection System has automatically actuated due to High Containment Pressure

• ALL ESF equipment has responded as designed e RCS break flow is approximately 800 GPM e No Charging Pumps are operating Select the response which correctly describes the status of the following ESF equipment, and RCS parameters which will exist when stable condilipns are reachei 5 the Reactor Coolant System? Assume that a cooldown HAS NOT been initiated.

• St Pump Status e RHR Pump Status

• RCS Pressure l

e RCS Break Flow A. S1 Pump Status Running - Recire to RWST RHR Pump Status Running - Flow to RCS-RCS Pressure LESS THAN 1500 PSIG and GREATER THAN 150 PSIG RCS Break Flow LESS THAN 800 GPM B. S1 Pump Status Running Flow to RCS RHR Pump Status Running - On Recirculation RCS Pressure LESS THAN 1500 PSIG and GREATER THAN 150 PSIG RCS Break Flow LESS THAN 800 GPM C. 31 Pump Status Running - Recirc to RWST RHR Pump Status Running - Flow to RCS RCS Pressure LESS THAN 150 PSIG RCS Break Flow GREATER THAN 800 GPM D. SI Pump Status Running - Flow to RCS RHR Pump Status Running - Flow to RCS RCS Pressure LESS THAN 150 PSIG RCS Break Flow LESS THAN 800 GPM Page 15 of 100

Indien Point Unit 2 -

Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider the following initial conditions wh n selecting your answer:

e Reactor Power - 1 X 10'* IR amps

• T.,, 547'F

= Plant stanup in progress During a plant startup, after the Main Steam Isolation Valves have been opened and Main Condenser Vacuum has been established, a fault is detected on the Station Auxiliaiy Transformer causing a loss of 138 KV electrical power to the station. Assuming no operator action, which of the following describes the immediate impact that this fiilure will have on the operation of the following:

• Condensate Pumps

• Condenser vacuum

. Reactor i

r A. Condensate pumps Stopped Condenser vacuum Stable Reactor Cntical B. Condensate pumps Running Condenser vacuum Decreasing Reactor Tripped C, Condensate pumps Stopped Condenser vacuum Decreasing Reactor Tripped D. Condensate pumps Running _

Condenser vacuum Stable Reactor Critical Page 16 of 100

--~

i indian Point Unit 2 '

Consolidated Edison Company ofNY

'f REACTOR OPERATOR EXAMINATION During power operatio 1 a' catastrophic failure of the service water supply line to the #23 Containment Fan Cooler Unit iesults in flooding of the containment building including the

- filling of the reactor cavity sump to the 46 ft, elevation of the containment building. How and why could this failure impact the ability of the Emergency Safeguards System (ESP) to perform its design function if a desigr, bases large break LOCA were to occur?

Assume all other ESF equipment operates as designed when the accident occurs.

A Containment pressure may exceed design limits due to loss of cooling water to the Containment Fan Cooler Units which are required to meet the minimum safeguards equipment requirements for containment pressure control.

B, Accumulator isolation valves will be submerged and may fail to open when signaled resulting in failure of the accumulators to inject into the RCS.

C. Cold leg recirculation will be impossible since the recirculation pumps will be submerged preventing the establishment of any recirculation path.

D. Containment pressure and water level may exceed design limits due to reduction in containment free volume before the accident occurred.

Page 17 of 100

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Indian Point Unit 3 Consolidated Edison Cornpany ofNY i

REACTOR OPERATOR EXAMINATION Consider the following initial conditions when selecting your answer:

1 e: Reactor Power 100 % -

-l i e Tm 559'F j e Control rods Bank D @ 215 Steps / Automatic i

During power operation with the above conditions Power Range Channel N41 fails to j- maximum detecto output (120%). Which of the following statements correctly der.cribes j the effect that this failure will have on the Rod Control System?

i A. Control rods will automatically insen to reduce indicated reactor power to equal turbine power.

~

B, Control rods will automatically insert until the rate of change ofindicated reactor power versus turbine power decays to zero.

C. Control rod motion will be inhibited due to the overpower rod stop, D. Control rods will automatically insert due to a turbine runback initiated by the chang:

in indicated reactor power exceeding 5% in 5 seconds.

Page 18 of 100

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>J t\ / Indian Point Unit 2 W ;e Q "' _ Consolid:ted Edison Company ofNY

F ,. ) "

4 REACTOR OPERATOR EXAMINATION While releasing #24 Large cas Decay Tank (LGDT) the Nuclear NPO inadvertently opens RCV 014 (LGDT Release liand Control Valve) to the full open position causing a the R-

- 43/R-44 PLANT VENT HI RAD / TROUBLE alarm to actuate on panel SAF-1 in the -

control room, due to R-44, Plant Vent Gas Monitor, exceeding the High Radiation Gas alarm setpoint.

Which of the following statements describes the impact, if any, that this event will have on the following equipment:

• RCV-014

• - PAB Supply Fan a e PAB Exhaust Fan A. RCV 014 Closes PAB Supply Fan No change PAB Exhaust Fan No Change 4

B. RCV-014 Closes

- PAB Supply Fan Trips PAB Exhaust Fan No Change C. RCV-014 No change i PAB Supply Fan Trips PAB Exhaust Fan Trips D. RCV-014 Closes PAB Supply Fan No Change PAB Exhaust Fan Trips i

i l

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it-i 4

Page 19 of 100

indian l'olzt Unit 2 Consolid:ted lidison Cornpany ofNY REACTOR Ol":ltATOR EXAMINATION During power operation the APPROAClllNO ROD INSERTION LIMIT 12.5" and the ROD INSERTION 1.lMIT 0" alanns are ac:uated on Panel S AF in the control room.

investigation reveals that all control and shutdown rods are fully withdrawn and that the rod insenion limi is t not being violated, indicating that the alanns have been actuated due to a inalfunction.

Which of the following components, if malfunctioning, could cause the alanns to actuate?

A. Individual Rod Position Indicator voltage decreases from 3.45 VDC to 0 VDC.

II. Rod Ilottsn kod Sinp flistable C. Pulse to Analog Converter (P/A Convener)

D. Rod llottom Rad Stop Bypass llistable Page 20 of 1(X)

lxdian P: int Unit 2 Censolidated Edison Company ofNY \

i.

REACTOR OPERATOR EXAMINATION 4

Consider the following initial conditions when selecting your answer:

• Reactor Power 60 %

J

• Tm 554*F j

• Control rods Bank D @ 200 Steps / Automatic 1

During surveillance testing it is discovered that the #22 Reactor Coolant Pump is not performing at design conditions, The flow rate in loop #22 is 7% less than design due to l

degradation of the diffuser ring in the pump casing. Flow in the remaining loops are at or ,

j' above design and total core flow is reduced by 5%. Which of the following statements ,

describes how this condition will affect the following:  ;

• Core A T

• Departure From Nucleate Boiling Ratio (DNBR) i j- A. Core 4 T Lower than design

DNBR Lower than design

B. Core A T Lower than design

,DNBR Higher than design 4

l C. Core A T Higher than design DNBR Higher than design l

t

[ D Core A T Higher than design DNBR Lower than design g,.n' ,,

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i Page 21 of 100

imdlan l'olnt Unit 2 Consolidated Edison Cempany ofNY AEACTOR OPERATOR EXAhllNATION Which of the following correctly identifies the signals that are used to pmvide input to the htBFP Speed Control System for detennining the following parameters:

• Actual Feed Regulating Valve Differential Pressure (AP) e Power e  % Startup Signal A.AP Main Feed lleader Pressure - Main Steam licader Pressure Power Turbine First Stage Pressure (IT-412A)

% Startup Signal MBFP Governor Control Switch B. AP Main Feed lleader Pressure - hiain Steam licader Pressure Power Total Stm Flow (sum of all controlling steam flow channels)

% Startup Signal MBFP Governor Control Switch C.AP Main Feed lleader Pressure - Main Steam lleader Pressure Power Total Stm Flow (sum of all controlling steam flow channels)

% Startup Signal MBFP Foxboro Speed Controller Manual Setting D.AP Main Feed 11eader Pressure - Steam Generator Pressure Power Total Stm Flow (sum of all controlling steam flow channels)

% Startup Signal MBFP Governor Control Switch Page 22 of 100

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Indian P; Int Unit 2 Censolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION While retrieving a dropped control in Control Bank D Group 1, with the unit at 80%

power, a ROD CONTROL URGENT FAILURE alarm is received on panel SDF 1 in the control room. Which of the following statements correctly describes the cause of this alann?

A. Remaining rods in Control Bank D Group 1 are not moving when demanded.

B._ Alarm is caused by the affected rod position deviating by more than 12 steps from bank demand position.

C. Alarm is caused by control bank rods moving with no bank overlap.

D. All rods in Control Bank D Group 2 are not moving when demanded 4

Je=

Page 23 of 100

1:di:n P:lzt Unit 2 Cons:lidated Edison Ccmpany ofNl' REACTOR OPERATOR EXAMINATION Consider the following event when selecting your answer:

e Reactor Power 0%

• T.,, 530*F 6

e Total Steam Flow 0.75 X 10 1bm/hr

• Pressurizer Level 32'7c e Pressurizer Pressure 2235 PSIO The plant is at hot shutdown due to a tube leak in #21 Steam Utnerator(SO). The Main Steam Isolation Valve for #21 SO is closed and the Reactor Operator is performing a cooldown to Cold Shutdown using the condenser steam dumps.

While positioning the condenser steam dump valves, a controller failure causes ALL of the steam dumps to OPEN fully. A Safety injection Signalis received. Directly afterwards the Reactor Operator notes the following parameiers:

• Reactor Power 0%

• T.,, 523*F e Total Steam Flow 0lbm/hr e Pressurizer Level 25 %

• Pressurizer Pressure 2110 PSIO Which of the following ESF actuation signals initiated the Safety injection signal?

A. High Containment Pressure B. liigh Steam Flow coincident with Low T.,,

C. Low Pressurizer Pressure D. Main Steam Line AP (21 SG OREATER TilAN 22,23,24 SO)

Page 24 of 100

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k Indian l'olnt Unit 2 Consolidated lidison Cornpany ofN)*

REACTOR OPERATOR EXAMINATION

~

Select the answer which conectly identifies the automatic reactor trip signals that are blocked when the POWER llELOW P 7 permissive is enabled?

A. Pressurizer Low Pressure Reactor Trip P:essurizer liigh Level Reactor Trip Two Loop Loss of Flow 6.9 KV 13us Undervoltage B. Pressurizer Low Pressure Reactor Trip Pressuriter liigh Level Reactor Trip

Steam Generator Lo Lo Level Trip 6.9 KV Hus Undervoltage C. Pressurizer Low Pressure Reactor Trip Pressuriier Low Level Reactor Trip Two Loop Loss of Flow 6.9 KV Bus Undervoltage D. Pressuriter liigh Pressure Reactor Trip Pressurizer liigh Level Reactoc Trip Steam Generator Lo Lo Level Trip 6.9 KV Bus Undervoltage Page 25 of 100

i i lxdian Point Unit 2

.' Consolidatedlidison Ccmpany ofNY r REACTOR OPERATOR EXAMINATION l

! While performing a natural circulation cooldown using ES 0.2, Natural Circulation i Cooldown, you are directed to depressurite the Reactor Coolant System to 1890 PSIO i after verifying that the RCS Ilot Leg temperatures are less than 550'F. Which of the J following statements correctly describes the reason for the maximum limit on hot leg i temperature before depressurization can commence?

l p A. Ensure that the AT limit between auxiliary spray fluid temperature and the RCS is

, not violated.

l i j B. Ensure that wide range hot leg temperatures are approximately saturation temperature  :

2 for 50 pressure.

t i >

C. Ensure that RCS subcooling is above the RCP Termination Criteria for the E 0 series  ;

j of procedures.  ;

i i

j D. Ensure that adequate subcooling exists to prevent void formation in the rc~ actor head i j when pressure is reduced to 1890 PSIO.

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Page 26 of 100 9 - , , - ,-r,.w--.---..-,,, -,,w.%,cwyy,- ,,-,,,s-,.m,y. -- n,,... ,,_,,m,,,, -

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indien Point Unit 2 Cors:lidated Edis:n Ccmpany ofNl' REACTOR OPERATOR EXAMINATION During power operation an increase in RCS leakage is noted during a routine RCS leak rate surveillance test. The subsequent RCS leakage safety evaluation determines that Reactor Coolant Drain Tank in leakage has increased by the same amount that RCS leakage has increased. Which of the following leakage sources could be ...;Lo.,J to both the increase in RCS leakage and the increase in RCDT in leakage? 1'r (*-ec f

• L. . t at fLc (1/~. ' ,

A. CVCS Letdown Line Relief valve leakage

,>4 t. c fle s> m c B. Reactor Vessel Flange O. Ring ,( ri,c /r t.3c' C. Pressurizer Power Operated Relief Valve (PORV) Leakage D. Reactor Coolant Putnp #3 Seal Leakage Page 27 of 100 l

ledian P: Int Unit 2 Consolidated Edinn Ccmpany ofNY REACTOR OPERATOR EXAMINATION During power operation a Pressurizer insurge results from a step decrease in power from 100% to 90% Which one of the following automatic control actions is designed to prevent a step decrease in RCS pressure when T.,, is restored to program and a resultant outsurge occurs?

A. Charging pump speed increases to maintain Pressurizer level stable.

B. Pressurizer backup heaters automatically DE ENEROlZE when Pressurizer level exceeds program by SE C. Pressurizer backup heaters automatically ENERGlZE when Pressurizer level exceeds program by SE f

D. Letdown line isolation prevents rapid outsurge. , i'],f' f 4' 0* , ye

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indian Polzt Uzit 2 C::solidated Edis:n Ccmpany ofNl' REACTOR OPERATOR EXAMINATION Which one of the following signals are used to autornatically AEM the RCS Overpressurization Systern (OPS) during a plant cooldown to Cold Shutdown?

A. Two out of three (2/3) RCS Wide Range Cold Leg Temperature RTDs LESS TilAN setpoint i

13. Two out of three (2/3) RCS Wide Range llot Leg Ternperature RTDs LESS TilAN setpoint C. Two out of three (2/3) RCS Wide Cold Leg Pressures GREATER TliAN setpoint D. Two out of three (2/3) RCS Wide Cold Leg Pressures LESS TilAN setpoint l

l Page 29 of 100 l

indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION While performing Ernergency Operating Procedure FR 11 1, Loss of Secondary lleat Sink, following a Loss of Coolant Accident (LOCA), you are directed to discontinue use of the procedure if RCS Pressure is LESS Til AN non faulted Steam Generator (SG) pressure (s).

Which of the following stateinents correctly describes why it is not necessary to continue with heat sink restoration if RCS Pressure is LESS TilAN non faulted SG pressure (s)?

A FR 11 1, will initiate RCS 131eed and Feed, A bleed path is not necessary if RCS pressure is low.

B. Since RCS pressure is less than non faulted SG pressure, RilR flow inay be established to remove core decay heat, C, Secondary heat sink is not required in this condition since core decay heat is removed by break flow.

D. Since RCS pressure is less than non faulted SG pressure, an inadequate Core Cooling condition probably exists, which is a higher priority procedure.

Page 30 of 100

\ indian P Int Unit 2 Cc::solidated Edis:n Company ofNl' HEACTOR OPERATOR EXAMINATION i

j During power operation a Main Steam Line Break occurs inside containment on the steam line from #24 Steam Generator (SO). A Safety injection signal is received from the Main Steam Line AP circuit. During recovery operations the control room operators are unable i I to close the Main Steam Line Isolation Valve for #24 SG.

4 1 Which of the following statements is correct regarding the ability of the operators to i

control Reactor Coolant System (RCS) temperature during this event?

I  :

) A. RCS temperature will continue to decrease until ALL sos have dried out. Subsequent  !

l temperature control will be perfonned by limiting Auxiliary Feedwater flow.

B. RCS temperature will continue to decrease until 24 So has dried out. Subsequent

temperature control will be perfonned using the remaining Sos, and auxiliary feed ,

j_ watcr flow.

1

[ C. RCS temperature will continue to decrease until 24 SO has dried out. Subsequent temperature control will be performed using the Atmospheric Steam Dump Valve (s) ,

on the intact sos since an automatic Main Steam Line Isolation was actuated by the Main Steam Line AP Safety injection s!gnal.

J D. RCS temperature will continue to decrease until ALL sos have dried out. Subsequent

temperature control will be perfonned by using RCS Bleed and Feed.

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j Page 31 of 100 f

indian P; Int Unit 2 Cons:lidated Edisin Ccmpany ofNl' REACTOR OPERATOR EXAMINATION Which of the following statements correctly describes the purpose of the POWER ADOVE P 10 permissive?

A. Automatically blocks the Source Range liigh Fin and Intermediate Range liigh Flux reactor trips. Prevents re instatement of the Source Range instruments.

B. Allow operator to manually block the Power Range liigh Flux Low Setpoint and Source Range liigh Flux reactor trips. Prevents re instatement of the Source Range l instruments.

C. Automatically blocks the Power Range liigh Flux Low Setpoint and Intermediate Range liigh Flux reactor trips. Prevents re instatement of the Source Range instmments.

D. Allow operator to manually block the Power Range liigh Flux Low Setpoint and Intermediate Range liigh Flux reactor trips. Prevents re instatement of the Source Range instruments.

Page 32 of 100 l

frdi:n Point Unit 2 Cc::s:lidated Edison Ccmpany ofNY REACTOR OPERATOR EXAMINATION During power operation Channel 1 Pressurizer Picssure Instrument falls high to maximum output (100%) While performing the subsequent actions of the Abnormal Operating Instruction (AOI) the Reactor Operator is directed to trip the following bistables:

P e Pressurizer liigh Pressure Reactor Trip e Pressurizer Low Pressure Reactor Trip e Pressurizer Low Pressure Safety injection e Pressurizer Low Pressure SI Unblock Which of the following correctly states the expected status (illuminated / extinguished) of the associated bistable proving lamp as each bistable trip switch is placed in the trip position?

A Pressurizer !!igh Pressure Reactor Trip Extinguished Pressurizer Low Pressure Reactor Trip Extinguished Pressurizer Low Pressure Safety injection Extinguished Pressurizer 1,ow Pressure SI Unblock lliuminated B. Pressurizer liigh Pressure Reactor Trip Extinguished Pressurizer Low Pressure Reactor Trip Illuminated Pressurizer Low Pressure Safety injection Illuminated Pressurizer Low Pressure SI Unblock Extinguished C. Pressurizer liigh Pressure Reactor Trip tiluminated Pressurizer Low Pressure Reactor Trip Extinguished Pressurizer Low Pressure Safety injection Extinguished Pressurizer Low Pressure SI Unblock illuminated D. Pressurizer liigh Pressure Reactor Trip Extinguished Pressurizer Low Pressure Reactor Trip 111uminated Pressurizer Low Pressure Safety injection Extinguished Pressurizer Low Pressure SI Unblock Extinguished Page 33 of 100

l 1.idlan Pol:*t Unit 2 i

Consolidated Edis:n Ccmpany ofNY REACTOR OPERATOR EXAMINATION i Consider the following event when selecting your answer:

F While the plant was at 100% Reactor Power a loss of the 138 KV offsite power source i

occurred, followed by a Turbine Trip on low vacuum and subsequent reactor trip. The Reactor Operator is performing Step 1 of E 0, Reactor Trip or Safety injection.

f Which of the following statements correctly lists the indications that the Reactor Operator will use to verify that the Reactor trip has occurred?

l a Rod Bottom Lights Reactor Trip Breaker Position '

d I B. Neutron Flux "

Individual Rod Position Indication

C. Reactor Trip Breaker Position Neutron Flux

! - D. Reactor Trip Breaker Position Bank Step Counters i

l l

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1 Page 34 of 100

indian itint Unit 2 C:nsolidated Edis:n Company ofNY REACTOR OPERATOR EXAMINATION While perfonning actions directed by the Emergency Operating Procedures, the Reactor Operator resets the Containment Spray signal. After depressing the reset push buttons, the Reactor notes that the white indicating lights above the buttons illuminate and remain illuminated after the buttons are released.

Which of the following statements correctly describes the reason the lights illuminated when the reset push buttons were depressed and remained illuminated when the buttons were released?

A. The lights illuminated when the buttons were depressed to indicate that the spray signal was reset. The lights remained illuminated indicating that an automatic Containment Spray actuation signal was present.

B. The lights illuminated when the buttons were depressed to indicate that the spray signal was reset. The lights remained illuminated indicating that both containment spray pumps were running.

C. The lights illuminated when the buttons were depressed to indicate that the spray signal could NOT be reset. The lights remained illuminated indicating that an automatic Containment Spray actuation signal was present.

D. The lights illuminated when the buttons were depressed to indicate that the spray signal was reset. The lights remained illuminated indicating that both containment spray pumps were NOT running.

Page 35 of 100

I indian P;lzt Unit 2 Consolidated Edison Ccmpany cfNY

. REACTOR OPERATOR EXAMINATION l

Which of the following correctly ide:itifies the Process Radiation Monitors which are capable of AUTOMATICALLY terminating a Containment Purge if they sense a high radiation condition?

A. R 44, Plant Vent Oas Monitor i

R-43, Plant Vent Particulate Monitor R 42, Containment Oas Monitor l B. R-43 Plant Vent Particulate Monitor R-41,' Containment Particulate Monitor R 42, containment Oas Monitor t

C. R-43. Plant Vent Particulate Monitor R 44, Plant Vent Oas Monitor R-41, Containmerit Particulate Monitor D. R 44, Plant Vent Oas Monitor R 41, Containment Paniculate Monitor  !

R 42, Containment Oas Monitor l

y Page 36 of 100

indlan l'olzt Unit 2 Cons:lidated Edis:n Company ofNl' REACTOR OPERATOR EXAMINATION Ehlle performing a plant cooldown using the Condenser Steam Dumps, a 6.9 KV MOTOR TRIP (COMMON) alarm is actuated. The Reactor Operator notes a step decrease in steam flow at the same time the alarm occurs. No other annunciators are actuated.

Which one of the following 6.9 KV motors is the most likely cause of the alarm?

l A. Reactor Coolant Pump l

B. Circulating Water Pump C.11 cater Drain Pump D. Condensate Pump Page 37 of 100

l 1;;dian Peint *

• sit 2

Consolidated Edis:n Lampany ofNY REACTOR OPERATOR EXAMINATION i

Which of the following lists correctly identifies the plant equipment protected by 4 concentrated foam fire suppression systems?

i i

A. Ilyd 0 Fen Seal Oil Unit i Main and Auxiliary Transformer Oils Systems l l Clean and Dirty Oil Storage Tanks

! Turbine Oil Reservoir i.

l B.11ydrogen Seal Oil Unit Boller Feed Pump Console '

'! Clean and Dirty Oil Storage Tanks Turbine Oil Reservoir

! C. liydrogen Seal Oil Unit Boller Feed Putnp Console i Support Facility Ignition Oil Tanks

- Turbine Oil Resenair i

! D. ' Main Genceator Boiler Feco Pump Consolc I

Clean and Diny oil Storage Tanks

Turbine Oil Reservoir i

a

+

u i

s 1

1 4

Page 38 of 100

~ . - . -

1:dian Point Unit 2 i Censolidated Edis:n Company ofNl'  ?

REACTOR OPERATOR EXAMINATION bue a fire in the transformer yard, fire main pr:ssure is decreasing because of heavy demand Which of the following selections correctly identifies the sequence in which the standby fire panps will automatically stan to maintain fire header pressure?  !

A. Fire Main Booster Pumps A ',$ . "', (' ' ' I ITs DicM P. imp ' ,, f. J 6 gt$$ S:audby hmtge Maintenance Pump t' ,4 g ydA, j t . l B. Fire Diesel Pump A s Standby Pressure Maintenance Pump Fire Main Booster Pumps C. Fire Main Booster Pumps i Standby Pressure Maintenance Pump i Fire Diesel Pump i j D. Standby Pressure Maintenance Pump j Fire Main Booster Pump l !- _ Fire Diesel Pump l i i l i i 1 i 1 }

- Page 39 of 100

indian Polzt Ucit 2 Censolidated Edison Ccmpany ofNY REACTOR OPERATOR EXAMINATION Following a unit trip from 507c power, a station blackout occurs due to a fault on the 138 KV offsite power supply. Which one of the following selections correctly identifies the 480 VAC equipment that will automatically stan WilEN the 480 VAC ss.fcguards busses breakers are re energized? NOTE:NO SI S10NAL(s) EXIST A. Two Non Essential Service Water Pumps Three Essential Service Water Pumps 21 and 23 Motor Driven Auxiliary Feed Water Pumps B. Two Essential Service Water Pumps 21 AND 23 Auxiliary Feed Water Pumps 21,22. AND 23 Component Cooling Water Pumps C. Three Essential Service Water Pumps 21 AND 23 Auxiliary Feed Water Pumps 21,22, AND 23 Component Cooling Water Pumps D. Three Essential Service Water Pumps 21 AND 23 Auxiliary Feed Water Pumps 21 AND 23 Component Cooling Water Pumps Page 40 of 100 indian Poh.t Uzit 2 Cuxnlidated Edison Ccmpany ofNY REACTOR OPERATOR EXAMINATION When loading electrical equipment on the 480 VAC busses while recovering from a unit trip and station blackout you are directed to limit the load on Transformers 5,2,3 and 6, to less than 200 Amps. Which of the following selections conectly identifies the indication (parameter) that will be used to verify compliance with this direction? A. Station Service Transformer liigh Side (6.9 KV) Ammeter (s) B. 480 VAC Bus Ammeters C. 6.9 KV Station Auxiliary Ammeters D. Sum ofindividual equipment ammeters  ! l i i 4 r i Page 41 of 100 ,. -, . . . . . . ... -. ...- - -.. ..-. .---... .- --.....- -..-._ _ - - .- _ -.a - .-.-. indian Point Unh 2 l C:ns:lidated Edis:n Ccmpany ofNY REACTOR OPERATOR EXAMINATION l 1 Following a Safety injection you receive indication that 125 VDC Control Power has been l lost to all equipment powered from 125 VDC bus #21. Which of the following selections describe llOW and Wily this failure WILL or WILL NOT impact your ability to satisfy the following potential EOP requirervents:

• RCP Trip Criteria e Si Reduction

• Close Accumulator isolation Valves A. RCP Trip Criteria No impact, RCPs can always be tripped from CCR Si Reduction - Si pumps must be tripped locally at 480 VAC Switchgear Accumulator Isolation Valves - No impact, control power supplied from individual breaker AC feed B. RCP Trip Criteria - RCPs that have lost control power must be tripped locally Si Reduction - No impact, control power automatically transferred to 23 DC Bus Accumulator Isolation Valves No impact, control power supplied from individual breaker AC feed C. RCP Trip Criteria - RCPs that have lost control power must be tripped locally S1 Reduction - SI pumps must be tripped locally at 480 VAC Switchycar Accumulator Isolation Valves - No impact, control ower supplied from individual breuker AC feed D. RCP Trip Criteria - RCPs that have lost control power must be tripped locally SI Reduction - No impact, control power automatically transferred to 23 DC Bus Accumulator Isolation Valves - Valves that receive control power from 21 DC bus cannot be closed Page 42 of 100

indian P; Int Unit 2 Cens:lidatedlidis:n Company ofNl' REACTC '* ')PERATOR EXAMINATION With the unit at llot Shutdown (liSD) a full load to: is being perfonned on #21 Emergency Diesel Generator (EDO) #22 and #23 EDOs are operable and in AUTOMATIC, After the EDO is fully loaded, a fire in the Station Auxiliary Transfonner causes a loss of 6.9 KV power. Which of the following statements describe the expected response of the #21 EDO and 480 VAC bus SA7 A. #21 EDO trips and restarts when returned to AUTO Bus SA Normal Feeder Breaker opens l Allloads strip l Blackout loads sequence stan B, #21 EDO trips and restarts when returned to AUTO Bus SA Nonnal Feeder Breaker opens All loads except running blackout loads strip Non running blackout loads sequence stan when bus is re-energized C. #21 EDO continues to run - Bus SA Nonnal AND Emergency Feeder Breakers open Bus SA Emergency Feed Breaker closes Blackout loads sequence stan D. #21 EDO continues to nm Bus SA Nonnal Feeder Breaker opens All loads except running blackout loads strip Non running blackout loads sequence start Page 43 of 100 indian P: Int Unit 2 Cc solidated Edis:n Ccmpany ofNl' REACTOR OPERATOR EXAMINATION Consider the following initial conditions when selecting your answer: , 1

• Reactor Power 100 %

e #21 RilR Pump OOS for last six hours (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO action tirne) While operating with the above conditions the control room receives a report that a lubricating oil leak on the oil cooler for #21 EDO has been discovered making #21 EDG INOPERABLE (7 day LCO action time), it is estimated that it will take 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to repair the leak. Which of the following statements correctly describes the actions that must be taken to ensure compliance with technical specifications? A. Complete repairs on both #21 EDG and #21 R11R within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> or place the plant in llot Shutdown B. Complete repairs on #21 EDO within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> or place the plant in 110t Shutdown C. Verify operability of remaining safeguards equipment and continue operation observing Technical Specification limits and action times for equipment out of service. D. Complete repairs on #21 R11R Pump or be in llot Shutdown within seven hours. Page 44 of 100 1:dian iti t Unit 2 l Ccts:lidated Edis:3 C:mpany ofNY REACTOR OPERA (OR EXAMINATION  ! hou have been directed to conduct a plant cooldown from llot Shutdown (llSD) to l 350*F at 50*F/hr using the condenser steam dumps. Approximately one hour after i positioning the steam dump valves in MANUAL > PRESSURE MODE to establish the i desired cooldown rate you notice that the cooldown rate has decreased froin 50'F/hr to { 20*F/hr. Steam dump valve position has not been changed. Which of the following statements conectly describes the reason that the cooldown rate has decreased and the actions necessary to maintain a constant cooldown rate? A. Steam flow has decreased due to reduced steam pressure. Valves must be gradually opened as the cooldown progresses. B. Steam dump pressure controller will not allow Main Steam Pressure to decrease below the dial setting in MANUAL or AUTO. Setpoint must be gradually reduced as the cooldown progresses. C. AT between the steam temperature and the condenser cooling water (circulating water) has decreased. Cooldown rate cannot be increased unless circulating water flow is increased. D. AT between RCS temperature and feedwater temperature has decreased. Feed , flow must be increased to increase heat removal. *- o 1 M - hh.. /(/ o Page 45 of 100 indian IUlzt U lt 2 Cc:s:lidated Edis:n Ccmpa::y cfNl' REACTOR OPERATOR EXAMINATION Consider the following initial conditiota when selecting your answer:

• Reactor Power 10 %

• RCS T.., $48'F e Control Rods Bank D 125 steps / MANUAL e Main Turbine Startup - approaching synchrenous speed e llP Steam Dumps AUTOMATIC / Steam Pressure Mode While performing a plant startup, the turbine trips due to an overspeed condition. Which of the following selections is correct regarding the effect that this malfunction will have on the following RCS parameters:

• RCS T,,,

l l

• Average Loop AT e Pressurizer Pressure

• Pressurizer Level A. RCS T. . INCREASE r^js<[

Average Loop AT INCREASE RCS Pressure INCREASE .[ [ Pressurizer level INCREASE B. RCS T.,, OdCREASE Average Loop AT INCREASE RCS Pressure DECREASE Pressurizer level DECREASE C. RCS T.,, INCREASE Average Loop AT INCREASE RCS Pressure DECREASE Pressurizer Level DECREASE D. RCS T.,, INCREASE Average Loop AT DECREASE RCS Pressure INCREASE Pressurizer Level INCREASE Page 46 of 100 Indian P:l:t U lt 2 Cc:solidated Edison Company ofNl' REACTOR OPERATOR EXAMINATION in Accordance with System Operating Procedures, W11EN RCS temperature is GREATER TilAN or EQUAL TO 350'F, the CCW Pump Auto Start key switch must be in the NORMAL position, and WilEN RCS tempera'ure is LESS T!!AN 350 F, the CCW Pump Auto Start key switch is placed in the BYPASS Position. Which ONE of the following statements correctly describes the reason for placing the CCW Pump Auto Start key switch in the BYPASS position WilEN RCS temperature is LESS Til AN 350'F? A. Block the CCW standby pump auto stan feature to prevent water hammer when the RilR system is in service. B. Allow operation of the CCW system with three pumps running to meet heavy demand imposed by RilR heat load. C Technical Specifications allow defeating the Auto start feature below 350 Fif three CCW pumps are OPERABLE. D. Permit operation of the CCW system with less than three pumps mnning when CCW is flowing through R11R lleat Exchangers. Page 47 of 100 l Indian Point Unit 2 Consolidated Edison Company ofNl' REACTOR OPERATOR EXAMINATION l 1 i During power operation, the control room coordinates with the NPOs to shift the  ! 2 ' ' Essential Service Water Header from the 1-2-3 Header to the 4 5 6 Header. When the necessary valving is completed the SWP Mode Control Switch on the SBF-1 panel is i inadvertently left in the 1 2-3 position. The service water system is operating in the Three  ; i Header Configuration. Which ONE of the following components will be supplied with service water if a Safety injectio:, signalis initiated? A. Instmment Air Compressor Heat Exchangers i B. Emergency Diesel Generators C. CCW Heat Exchangers ] D. Containment Fan Cooler Units 4 i 1 Page 48 of 100 indian Point Unit 2 . Consolidated Edison Company ofNl' REACTOR OPERATOR EXAMINATION Which ONE of the following selections identines ALL of the conditions which will initiate a Containment Ventilation isolation signal? A. High Radiation (containment, R-41, R-42) Containment Isolation Phase B Containment Phase A Isolation Signal Containment Hi Hi Pressure Signal Manual Containment Spray Signal B. liigh Radiation (containment, R-41, R-42) High radiation (plant vent, R-44) Containment Phase A lsolation Signal Containment Hi-Hi Prusure Signal Manual Containment Spray Signal C. High Radiation (containment, R-41, R-42) High radiation (plant vent, R-44) ' Containment Phase A Isolation Signal Containment Hi-Hi Pressure Signal Station Blackout D. High Radiation (containment, R-41, R-42) High radiation (plant vent, R-44) Containment Phase A Isolation Signal High Radiation (containment R-2, R-7) Manuel Containment Spray Signal Page 49 of 100 indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION Which ONE of the following conditions would be considered a loss of containmens integrity during normal operation? A. An automatic containment isolation valve is found to be inoperable in the CLOSED position. B. Weld channel zone pressure indicates 50 psig. C. Personnel hatch inner door indicates OPEN, outer door indicates CLOSED. D. Weld channel seal to the equipment hatch is lost. Page 50 of 100 . . _ _ . _ ._ __ _ -- _ _ __ _ _ . _ _ , _ ...-.___m_ _ _ _ _ ____-_-__m. 1 Indian Point Unit 2 - Consolidated Edison Company ofNY . REACTOR OPERATOR EXAMINATION in accordance with procedure Residual 11 eat Removal (RHR) System flow rate should be

AT LEAST 2500 GPM per RHR Pump /RHR Heat Exchanger.

Which ONE of the following statements is correct regarding the reason for this limitation?

A. Minimize turbulence downstream of the associated RHR Heat Exchanger and

HCV-638 and liCV 640.

B. Prevent the RHR pumps from reaching mnout conditions. C. Prevent vortexing at the RCS Hot Leg loop connection. D. Ensure RHR flow is evenly distributed via the SI manifold to all RCS cold legs. 4 3 4 f i i a N Page 51 of 100 T ' indian Point Unit 2. Consolidated Edison Cempany ofNY REACTOR OPERATOR EXAMINATION

- When using Emergency Operating Procedure (EOP) E-1, Loss of Reactor or Secondary Coolant, you are directed to operate Reactor Coolant Pumps (RCP) in accordance with

_ EOP FR C.2, Response to Degraded Core Cooling, rather than tripping the RCPs as . directed by EOP E-1, if the RVLIS Dynamic Range indication indicates less than the value

obtained from the following table

4 44 % - 4 RCPs I/S 30% - 3 RCPs I/S 20 % - 2 RCPs US i 13 % - 1 RCP I/S ! Which of the following statements is correct regarding the reason for the guidance - l' described above? i , A. RCPs operating under these conditions may seize when tripped preventing their restart p in future recovery actions. B. Tripping RCPs under these conditions could lead to core uncovering and an Inadequate Core Cooling Condition, i-j - C. Tripping RCPs under these conditions would result in an increase of mass flow from the break due to phase separation of the fluid.

D. Tripping RCPs under these conditions will result in a loss of RCS pressure control due

{ to the loss of Pressurizer spray capability, 4 3-1 1 f i- !- Page 53 of 100 Indian Point Unit 2 i- Consolidsled Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider the following indications when selecting your answer:

• Reactor Power 50%  ;

e RCS Tm 553*F i e #23 RCP #2 Seal Standpipe Low Level Alarm actuated ' e #23 RCP Seal injection flow 8.0 GPM e #23 RCP #1 Seal Leakoff flow 3.0 GPM While operating with the above conditions and indications the Reactor Operator reports that the RCS leakage calculation is normal however there has been an INCREASE in total leakage into containment as evidenced by an increase in the pumping frequency of the containment sump. Which ONE of the following failures or malfunctions would support ALL of the above indications and conditions? A. Failure of the #23 RCP #2 Seal ,/' , p B. Failure of the #23 RCP #3 Seal . 4@/' ,F ^ C, Failure of the #23 RCP #1 S:al 8[ ' D. Failure of the #23 RCP Seal Package (#1,#2, and #3 Seal) p/ g, ' + pt . p# [.^[ it , 0 ;/4. )'p0 pi s g s Page 55 of 100 Indian Point Unit 3 Consolidated Edison Company ofNl' REACTOR OPERATOR EXAMINATION While the reactor is in the Refueling Condition the Reactor Operator identifies an unexplained increase in Source Range count rate and a steady positive 0.15 DPM Source Range Startup Rate indication. The Senior Reactor Operator directs the Reactor Operator to initiate boration of the RCS per A3.4, Un.ontrolled Reactivity Addition. Which one of the following boration flow paths and methods is the preferred method for completing this task in accordance with A3.4, Uncontrolled Reactivity Addition? A. RWST via LCV ll2B, Emergency RWST Makeup Stop B. MOV-333, Emergency Boration Stop to charging pump suction C. Normal boration flowpath at maximum rate to charging pump suction D. Normal boration flowpath at maximum rate to Volume Control Tank i Page 56 of 100 Indian Point Unit 2 Consolid:ted Edison Company ofNY REACTOR OPERATOR EXAMINATION Which ONE of the following will NOT result in the automatic start of a Component Cooling Water Pump?

A. A loss of offsite electrical power is followed by Unit Trip AND Safety injection.

B. CCW Header Pressure decteases to 60 psig with two CCW pumps nmning. C. An inadvertent Safety Injection Signal is actuated due to an instrument failure. The 480 VAC busses are ALL energized from offsite electrical power. D. A Station Blackout Signal is actuated due to a loss of offsite power following a Unit Trip. No Safety Injection signal is present. 4 Page 57 of 100 indian Point Unit 2 Cons: lid:ted Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider the following event when selecting your answer:

• The controlling Pressurizer pressure channel (CliANNEL_l) has failed high with the unit at 100% power

• RCS pressure is stabilized at 2115 psig by manually closing the Pressurizer spray valves After the plant is stabilized the Senior Reactor Operator reviews the following Technical Specification Limit:

Reactor Coolant System Pressure. Temperature and Flow Rate The following DNB related parameters pertain to four loop steady state operation at power levels greater than 98% of full rated power: e Reactor Coolant System T n s 587.2'F e Pressurizer Pressure 2 2190 psia e Reactor Coolant System Total Flow Rate 2 331,840 GPM ltem (b), pressurizer pressure,is not applicable during either a thermal power change in excess of 5% of rated thermal power per minute, or a thermal power step change of 10% of rated thermal power. Under the applicable operating conditions, should reactor coolant Ty, or pressurizer pressure exceed the values given in items (a) and (b) the parameter shall be restored to its applicable range within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Which of the following statements correctly describes the appropriate response,if any, the Senior Reactor Operator should take to ensure compliance with this technical specification? A. Specification for Pressurizer Pressure is not applicable because the transient was induced by an instrument failure. B. Specification for Pressurizer Pressure is not applicable because there was no change in reactor power. C. Specification is not applicable because it only applies when power is ;t 98% of ' RATED FULL POWER which is equivalent to 98% of the maximum attainable power 4 level of 108% (High Flux Trip Setpoint) D. Restore RCS pressure to GREATER THAN 2190 psia within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce power to s 98%. Page 58 of 100 Indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION A main steam line rupture has occurred with the plant at hot shutdown resulting in a Main Steam Line AP Safety Injecti_on Signal. After analyzing the following indications determine which ONE of the following selections correctly identifies the portion of the Main Steam System that has ruptured?

• 21 SG Pressure 780 psig (decreasing slowly) e 22 SG Pressure 782 psig (decreasing slowly)

• 23 SG Pressure 340 psig (decreasing rapidly) e 24 SG pressure 775 psig (decreasing slowly)

ALL Main Steam Line Flow Indications 0 lbm/hr . ALL MSIVs indicate OPEN A. 23 SG Main Steam Line Upstream of MSIVs outside containment B. 23 SG Main Steam Line between SG and Flow Element C. 23 SG Main Steam Line Downstream of MSIVs D. 23 SG Main Steam Line between Flow Element and Cor tainment Penetration Page 59 of 100 . . . - - . - - - . - - .- -- . ~ . - . - - . . - . - -, . . - . - . l Indian Point Unit 2 #^$

Consolidsted Edissn Company ofhT

('# T'& REACTOR OPERATOR EXAMINATION

Consider the following event when selecting your answer

• - Reactor Power 15 %

,

• T.,, 547'F 4

• Turbine Startup in progress Turbine at synchronous speed

• HP Steam Dumps l Pressure Mode - AUTOMATIC During plant startup with the above conditions, a failure of the exhaust boot (expansion i joint) between #21 Low Pressure Turbine and the condenser results in a rapid loss of condenser vacuum to atmospheric pressure over a period of 5 minutes.

j Assuming that all plant protection AND control systems function as designed, which ONE of the following statements correctly describes the response (INCREASE / DECREASE /NO CHANGE) of the following parameters as a result of this event.

• Reactor Power

• Turbine RPM e Steam Flow i

• T.,.

A. Reactor Power DECREASE Turbine RPM DECREASE .! Steam Flow NO CHANGE T.v. NO CHANGE 3 4 B. Reactor Power DECREASE Turbine RPM DECREASE

Steam Flow INCREASE T.,. INCREASE i C. Reactor Power DECREASE Turbine RPM DECREASE Steam Flow DECREASE T v. INCREASE D. Reactor Power NO CHANGE i

Turbine RPM DECREASE Steam Flow NO CHANGE i T.v. NO CHANGE Page 60 of 100 4 . . .. - _ . - - -_ -..- ~ .. . . -. .- .- Indien Point Unit 3 Consolidated Edison Company ofNl' REACTOR OPERATOR EXAMINATION During power operation at 100% reactor power a loss of electrical power to #21 Instrument Bus occurs causing the loss of PT-412A, Which ONE cf the following statements describes the response of the liigh Pressure Steam Dump System? Assume that no equipment or instrumentation was out of service before the failure. A. Loss ofload interlock will trip arming the steam dumps. B. Tur will fail to 547'F causing the High Pressure Steam Dumps to OPEN. C. Steam dump actuation will be inhibited due to loss of power to loss of load interlock. D. Tyrwill fail to 547'F. High Pressure Steam Dumps will OPEN IE loss of load interlock trips, 1 4 i Page 61 of 100 indian Point Unit 3 Censolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION The Conventional NPO has reported a small electrical fire on the 5' elevation of the Turbine Building. The following fire fighting equipment is available to the Fire Brigade. Select the equipment which is most suitable for extinguishing a fire of this type? A. Ponable CO 2fire extinguisher B. High pressure water hose C. Water stream portable fire extinguisher D. Dry chemical fire extinguisher Page 62 of 100 Indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION A fire in the control building requires tiiat the Central Control Room be evacuated. You have been designated as the First RO, and the SRO has directed you to trip the Reactor locally. Which of the following selections list in ORDER the locations and equipment from which you would accomplish this task, assuming that you are unsuccessful after each attempt. A. Cable Spreading Room Reactor Trip Breakers Cable Spreading Room - Rod Drive MG Set Breakers 480V Switchgear Room - Rod Drive MG Set Breakers 480V Switchgear Room - Bus 3A and 5A Supply Breakers 6.9 KV Switchgear - Station Service Transformer 3 and 5 Supply Breakers B. 480V Switchgear Room -_ Rod Drive MG Set Breakers 480V Switchgear Room - Bus 2A and 6A Supply Breakers ~ Cable Spreading Room - Reactor Trip Breakers Cable Spreading Room - Rod Drive MG Set Breakers 6.9 KV Switchgear - Station Service Transformer 2 and 6 Supply Breakers C. Cable Spreading Room - Reactor Trip Breakers Cable Spreading Room - Rod Drive MG Set Breakers 480V Switch; ear Room - Rod Drive MG Set Breakers 480V Switchgear Room - Bus 2A and 6A Supply Breakers - 6.9 KV Switchgear - Station Service Transformer 2 and 6 Supply Breakers D. Cable Spreading Room - Reactor Trip Breakers Cable Spreading Room - Rod Control System Power Cabinets 480V Switchgear Room - Rcd Drive MG Set Breakers 480V Switchgear Room - Bus 2A and 6A Supply Breakers 6.9 KV Switchgear - Station Service Transformer 2 and 6 Supply Breakers . Page 63 of 100 Indian Point Unit 2 Consolidated Edison Company ofNY , REACTOR OPERATOR EXAMINATION . While verifying Containment isolation valves are in the correct position following a Safety injection actuation due '.o Hi Hi Containment Pressure, you notice the following "Two is True" indication for MOV-222, RCP Seal Leakoff Containment Isolation Valve:

• Left side oflight Illuminated - AMBER light e Right side oflight Extinguished Which ONE of the following selections is correct regarding the expected position of MOV-222, AND the indicated position of MOV-222 with respect to the "Two is True" indicating lights?

A. Expected Position OPEN Indicated Position OPEN B. Expected Position CLOSED Indicated Position CLOSED C. Expected Position OPEN Indicated Position CLOSED D. Expected Position CLOSED Indicated Position OPEN Page 64 of 100 indian Peint Unit 3 , Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION Which of the following statements identifies the conditions where Technical Specifications requires that RCS activity (for nuclides other than tritium with halflives of more than 30 minutes) be LESS THAN 60/E-bar pei/cc7 A. When there is fuel in the Reactor Vessel or RCS temperature is GREATER THAN 200*F. B, When reactor is critical or RCS temperature is GREATER THAN 500'F. C. When the reactor is critical or RCS temperature is GREATER THAN 350 F. , ,J' D. When the reactor is in " Power Operation Condition" or RCS temperature is GREATER THAN 350 F, o rh j, v ' s 4* Page 65 of 100 indian Point Unit 2 Consolidated Edison Company ofNl' REACTOR OPERATOR EXAMINATION Following a reactor trip you are directed by ES-0.1, Reactor Trip Response, to verify that all control rods are fully inserted. Which ONE of the following rod position indications would meet the criteria for a control rod b'QT being fully inserted? A. Individual IRP! reading 5.5 inches ,,r,

1. Y gd.

,f; * # B. Prueus Computer rod position 7 steps ] - p/; I"f

p e;'

C. Proteus Computer rod position 13 steps D. 0.05 Volts on Digital Volt Meter (DVM) Page 66 of 100 Indi::n P: Int Unit 2 Cons:lidated Edison Company ofNY REACTOR OPERATOR EXAMINATION While conducting a natural circulation cooldown using ES-0.2, Natural Circulation CoolJown, you are directed to control SG levels. Which ONE of the following actions if performed c7'ild temporarily impede or reduce natural circulation flow? A. Steam generator #21 level is allowed to slowly increase to 55% as seen on the narrow range level indicator. B. Auxiliary feed flow to #21 SG is rapidly increased from 50 GPM to 200 GPM C. Steam ger.erator #21 level is allowed to slowly decrease to 35% as seen on the narrow range levelindicator D. Auxiliary feed flow to #21 SG is rapidly decreased from 200 GPM to 50 GPM a Page 67 of 100 Indian P: int Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION While performing the immediate actions for a Reactor Trip using EOP E-0, Reactor Trip or Safety injection, you are directed to de-energize 480VAC busses 2A and 6A to trip the reactor since a reactor trip cannot be verified by available indication. After 480 VAC bus 2A and 6A are re-energized you are directed to depress the Blackout Relay Reset 480V push button on the SC panelIf the Main Generator Output Breakers are CLOSED. Which ONE of the following statements is correct regarding the reason that the Blackout Relay must be reset at this time? A. Resetting the B!ackout Relay at this time removes the 480 VAC Bus 2A Undervoltage Signal from the blackout logic and E the Main Generator Output Breakers are CLOSED (86P and 86BU relays reses) a station blackout signal will be avoided. B. Resetting the Blackout Relay at this time removes the 480 VAC Bus 2A AND 480 VAC Bus 6A Undervoltage Signals from the blackout logic and E the Main Generator Output Breakers are CLOSED (86P and 86BU relays reset) a station blackout signal will be avoided. C. Resetting the Blackout Relay at this time removes the 480 VAC Bus 6A Undervoltage Signal from the blackout logic and E the Main Generator Output Breakers are CLOSED (86P and 86BU relays reset) a station blackout signal will be avoided. D. Main Generator Output Breakers should already be OPEN. Depressing the Blackout Relay Reset 480V push button will trip the 86P and 86BU relays causing the Main Generator Output Breakers to OPEN. Page 68 of 100 Indl3n Point Unit 2 Consolidated Edison Company ofNY l i REACTGR OPERATOR EXAMINATION i Consider the following event when selecting your answer: A small break LOCA has occurred and a Safety injection (SI) Signal has been 4 actuated. 2

• All Si equipment has operated as designed.

EOP ES 1.2, Post LOCA Cooldown and Depressurization has been implementea. While performing the actions required by EOP ES-1.2, you are directed to establish ). maximum charging flow to the RCS. Which ONE of the following statements is correct regarding the reason for this action when performing a Post LOCA Cooldown and

Depressurization?

1

A. Maximum charging flow is established in order to provide maximum auxiliary spray flow capability in the event that Reactor Coolant Pumps are not running and normal

spray is unavailable, ,

B. Maximum charging flow is established to ensure that maximum boration capability { exists. C. Max. mum charging flow is established in an attempt to achieve S1 Terminatien Criteria, thus avoiding the tedious task of SI Reduction. D. Maximum charging flow is established in order to provide sufficient makeup so that SI

pumps can be more readily reduced during the SI Reduction sequence.

s t 1-Page 69 of 100 indian Point Unit 2 Consolidsted Edison Company ofNl' REACTOR OPERATOR EXAMINATION Consider the following conditions when selecting your answer:

• Large Break LOCA has occurred

• Core Exit Thermocouple Temperatures 1210'F

• RVLIS Full Range Indication 30% -

• RCS Subcooling 0F The Watch Engineer repons that the above conditions require entry into the Functional

? Restoration Procedures. Which ONE of the following Functional Restoration Procedures must be implemented? l A. FR-C.1, Response Inadequate Core Cooling ,,J ' i t , p . : .e .a-o, B. FR-C.2, Response to Degraded Core Cooling s , ,s i ' ' s.,e e~^4 , p et s, C. FR-H.1, Response to Loss of Secondary Heat Sink I'. pA' n; ('- (V6 .,o D. FR I.3, Response to Voids in the Reactor Vessel j an~ , y A a',

• 4 4

1 t Page 70 of 100 ' Indian Point Unit 2. Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION - Consider the following events when selecting your answer:

• A tube rupture has occurred in #22 Steam Generator (SG)

. A Safety Injection (SI) signal has been actuated EOP E-3, Steam Generator Tube Rupture has been implemented e #22 SG has been isolated e The Si signal has been reset The SRO has directed you to dump steam from the intact SGs at the maximum rate to establish a Core Exit T emperature of 488*F AND then stop the cooldown. Which ONE of the following statements correctly describes the reason for reducing RCS temperature to this value? A. Reduce RCS pressure by causing an outsurge from the Pressurizer to minimize leakage into the #22 SG. B. Establish sufficient subcocling in the RCS so that the RCS will remain subcooled after pressure is decreased to #22 SG pressure. C. Establish sufficient subcooling in the RCS so that the Reactor Coolant Pumps will not have to be tripped when the RCS pressure is decreased to #22 SG pressure. D. Reduce temperature of RCS fluid leaking into #22SG to reduce #22SG pressure to minimize potential of radioactive release through the atmospheric steam dump valve. Page 72 of 100 indian Point Unit 3 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider the following event when selecting your answer: f -

• Reactor Power 80 %  !

!- e- RCS T.,, 557 F

• Control Rods Control Bank D - 195 steps / Manual

, e RCS Boron Concentration 980 ppm Beginning Of Life (BOL) . After withdrawing control rods to adjust T.,, you note that when you release the In-Hold-Out switch Control Bank D rods continue to withdraw for an additional 20 steps. As a result T.,, increases to 5*F above program. Which ONE of the following statements is TRUE regarding this event? , A. IF the same event occurred at EOL the INCREA.S]in RCS T.,, would have been GREATER. B. IF the same event occurred at EOL the INCREASE in RCS T.,, would have been the SAME. C. IF the same event occurred at EOL tuere would have been NO INCREASE in RCS T., . . D. IF the same event occurred at EOL the INCREASE in RCS T.,, would have been LOWER. Page 73 of 100 Indian P: Int Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider .he following event when selecting your answer: !

• A l'ressurizer Power Operated Relief Valve (PORV) has OPENED and failed to close.

• A Safety injection Signal has been actuated.

.

• RCS Pressure stable 1480 psig

• Pressurizer Relief Tank pressure 30 psig

Pressurizer relief tank temperature

• 140 F

• Pressurizer Level 90 %

• Pressurizer Relief Tank Level 82 %

, Using the indications and conditions provided determine which ONE of the following

temperatures would be indicated on the PORV Downstream Temperature Indicator?

A. 593 F B. 275 F C. 250 F D.140 F t Page 75 of 100 . . ._ . . - _ _ _ . _ . _ -- ___ ____._. __ __ ._ _ . _ . - . _ _ _ - - . - - - . _ - _ _ - ~ . Indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION - Abnormal Operating Instruction (AOI) 3.1, Chemical Volume Control System (CVCS) Malfunctions, states that when preparing to start a Charging Pump, the Charging Pump controller must be placed in manual AND set for approximately 20% before the pump is started? - Which ONE of statements is correct regarding the reason for setting the controller to 20% before starting the Charging Pump? . A. Ensure that a low bearing oil pressure trip does not occur during the charging pump start. B. Minimize starting current on the charging pump motor. C. Balance MANUAL signal with AUTO signal before the pump is started. D. Provided a minimum of 8 GPM seal injection flow to each RCP as soon as the pump is started. i a 1 Page 76 of 100 Indisn Point Unit 2  : Consolidsled Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider the following indications when selecting your answer: 4

• Reactor Startup in progress

• Source Range N31 Count Rate 2X10' cps 5

. -Source Range N32 Count Rate 4X10 cps

• Intermediate Range N35 Current <lX 10'" amps

• Intermediate Range N36 Current <1X10'" amps

• Source Range N31 Startup Rate 0.5 dpm

!

• Source Range N32 Startup Rate 0.1 dpm

. Control Rods Control Bank D - 100 steps / MANUAL i Using ONLY the information provided determine which ONE of the following actions and associated reason is appropriate regarding the continuation of the teactor startup? A. Source Range N32 is reading high due to a failure in the Pulse Height Discrimination ! circuitry and should be considered inoperable. The startup may continue without further action. A B. Source Range N31 is reading low due to a failure in the Pulse Height Discrimination f circuitry and should be considered inoperable. The startup may continue without further action, 4 C. Intermediate Range N35 AND N36 are not responding. Startup may_ continue as long as neutron flux remains in the source range. D. Nuclear instrumentation is NOT indicating as anticipated. The approach to criticality SHALL be stop;,ed AND no actions SHALL be taken which could add positive reactivity until the discrepancy is resolved. Page 78 of 100 Indian Point Unit 2 Consolidcted Edison Company ofNY REACTOR OPERATOR EXAMINATION' A reactor trip has occurred. Approximately 30 minutes after the reactor has tripped the Reactor Operator is performing actions directed by ES-0.I, Reactor Trip Response, when he notes the following indications:

• Intermediate Range N35 IX10* amps (stable)

• Intermediate Range N36 1X10'" amps (stable)

• Source Range N31 0 cps (stable) e Source Range N32 0 cps (stable)

SOURCE RANGE LOSS OF DETECTOR VOLTAGE annunciator illuminated Which ONE of the following statements correctly describes the response,if any, that the Reactor Operator should take regarding these indications? A. Manually re-energize the Source Range NIS by depressing the Train A and Train B Intermediate Range Permissive Override push buttons. B. Manually re-energize the Source Range NIS by depressing the Train A and Train B Power Range Permissive Override push buttons. C. Initiate rapid boric acid injection in accordance with, A 3.4, Uncontrolled Reactivity Addition. D. Monitor Intermediate Range N35, and verify reinstatement of the Source Range NIS when both Intermediate Range instruments are less than IX10 amps. Page 79 of 100 Indian Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION Consider the following indications when selecting your answer: e Reactor Power 100 %

• Tm 559'F e Total RCS Leakage 0.8 gpm (includes SG tube leakage)

• SG Tube Leakage 0.25 gpm The Senior Watch Supervisor has directed the control room operators to perform a reactor shutdown due to increasing secondary side activity caused by SG tube leakage.

Which ONE of the following statements is correct regarding the anticipated change, if any, in total RCS leakage as a result of the plant shutdown? A. Total RCS leakage will increase, due to the increase in SG tube leakage, B. Total RCS leakage will r main the same, increase in SG leakage will be offset by decrease in other RCS leakage. C. Total RCS leakage will decrease, due to the decrease in SG tube leakage. D. Total RCS leakage will remain the same, decrease in SG leakage will be offset by increase in other RCS leakage. Page 80 of 100 l indian P: Int Unit 2

Consolidated Edis

a Company ofNl' REACTOR OPERATOR EXAMINATION While recovering from a Reactor Trip due to a loss of both main feedwater pumps. IIOP FR II.5, Response to Steam Generator Low Level is entered due to indication ' hat water 4 level in #23 SO has decreased 10 0% WIDE RANGE level. In accordance with FR il.5, j the SRO directs you to feed #23 50 at LESS TilAN 100 OPM UNTIL water level is

GREATER Til AN 10% as indicated on #23 SG WIDE RANGE level indication.

i Which ONE of the following statements is TRUE regarding the reason for limiting feed i water flow to #23 SO to LESS TilAN 100 GPM until level is OREATER TilAN 10%7 A. Feed now is limited to prevent a rapid RCS cooldown which could result in challenge I to the INTEORITY critical safety function. i 1 .! B. Feed flow is limited to prevent unnecessary thermal shock to a "liot Dry SO" which ! could result in SO tube failure. C. Feed flow is limited to prevent flashing in the SO which could result in lifting the SO Safety Valves, 1 i D. Feed flow is limited to prevent runout conditions on the #23 Auxiliary Feed Water Pump. 1 I l 4 h l 5 Page 81 of 100 indian IUl::t Unit 2 Ccasolidated Edison Compazy ofNl' REACTOR OPERATOR EXAhllNATION Following a manual Reactor Trip from 10091 power you are perfonning Step 3 of E-0, Reactor Trip and Safety injection," Check If Si Actuated", and note the following indications:

• SI Annunciator Not Lit

• StPumps None Running i

e Pressurizer Pressure 2090 psig (stable)

• Stearnline AP All < 50 psid (stable) e Steam Line flow All < 100,000 lbm/hr (stable)

• Containment Pressure 0.75 psig (stable)

• P:essuriier level 5'/c (stable) e RCS Subcooling hlargin 87'F (stable)

Which ONE of the following actions should you take in response to these indicaCons, and Wily is the action requhed? A. hianually initiate Safety injection due to Pressurizer Low Level. B. hianually initiate Safety injection due to failure of the liigh Steam Flow Si to actuate, C. Check RCS subcooling table, if LESS TilAN required, manually initiate Safety injection due to Low Subcooling. D. Transition to ES.O.1, Reactor Trip Response, Safety injection is NOT required. s 4 Page 83 of 100 indian Point Unit 2 Consolidated Edison Company ofNl' REACTOR OPERATOR EXAMINATION A Large Break Loss of Coolant Accident has occurred. Approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after aligning for Cold Leg Recirculation, the Senior Reactor Operator implements EOP ES-1.4. Transfer to llot Leg Recirculation. Which ONE of the following statements is correct regarding the reason for placing Ilot Leg Recirculation in service at this time? A. Ilot leg recirculation is implemented to sweep non-condensable gasses from the reactor head region. B. Ilot leg recirculation is implemented to cool the reactor head to enable RCS depressurization without additional void forma: ion. C. liot leg recirculation is implemented to refill the reactor vessel and preclude fuel rod damage at the top of the core. D. Ilot leg recirculation is implemented to prevent boron precipitation in the core. Page 84 of 100 i indi:n Point Unit 2 Consolidated Edison Company ofNY REACTOR OPERATOR EXAMINATION Following a Steam Generator Tube Rupture. EOP E 3, Steam Generator Tube Rupture, has been implemented.

The Reactor Operator is controlling RCS temperature using the Condenser Steam Dumps. The Senior Reactor

! Operator directs the Reactor Operator to depressurize the RCS to LESS TilAN RUPTURED SG pressure using the Pressurizer PORVs since the RCPs have been tripped. The Reactor Operator CLQSES the steam dump valves and prepares for depressurizain of the RCS. Before ,I commencing depressuritation the Reactor Operator notes the following indications: I] . e RCS Wide Range Cold Leg Temperatures 480T(increasing slowly) e Core Exit Temperature 488T(increasing slowly) 4

• Pressurlier Level (hot calibrated) 0%

• RCS Prenure 1370 psig (increasing) j e Ruptured SG Lesel 889 (Narrow Range increasing) e Ruptured 50 Pressure 1015 psig (increasing slowly) l i Just prior to closing the Pressurizer PORVs the Reactor Operator notes the following indications:

j e RCS Wide Range Cold Leg Temperatures 50MF(increasing slowly)

e Core Exit Temperature $10T(stable) j' e Pressurlier Level (hot calibrated) 80% (inercasing rapidly) e RCS Pressure 760 psig (increasing)

e Ruptured 50 Level 859 (Narrow Range . decreasing slowly)

! e Ruptured SG Pressure 1005 psig (decreasing slowly) Which ONE of the following statements could explain ALL of the changes in the above indications that have occurred since the Reactor Operator commenced depressurization? A. RCS cold leg temperature has INCREASED due to Dashing of RCS Duld in the hot leg.

PZR level has INCREASED due to increased makeup Dow AND voiding in the Reactor llead.

i Ruptured SG levelis DECREASING due to SG Duid backfilling the RCS. B. RCS cold leg temperature has INCREASED due to REDUCED natural circulation now rate, Pressurizer level has INCREASED due to increased makeup flow AND voiding in the Reactor 11ead, i Ruptured SG level is DECREASING due to SG Duid backfilling the RCS. C, RCS cold leg temperature has INCREASED due to REDUCED natural circulation flow rate. , ! Indicated Pressurizer level has INCREASED to due Dashing of the Guid in the levelinstrument reference leg. l Ruptured SG level is DECREASING due to SG Duld backfilling the RCS, . D. RCS cold leg temperature has INCREASED due to STOPPAGE of natural circulation Dow. 4 Pressurizer level has INCREASED due to increased makeup Dow AND voiding in the Reactor llead. Ruptured SG level is DECREASING due to steaming through the Atmospheric Steam Dump valve, 4 Page 85 of 100 indian P Int Unit 2 i Coxs:lidated Edison Company ofNY RFACTOR OPERATOR EXAMINATION l A refueling operator hasjust placed an irradiated fuel assembly in the containment side j upender when the Refueling SRO notes that refueling cavity water level is decreasing j rapidly. Which ONE of the following actions should be directed by the Refueling SRO? l ! A. Disengage the manipulator from the fuel assembly and lower the upender to the fully lowered position, TiiEN evacuate containment. 1 i B. Withdraw the fuel assembly from the upender and move it to the reactor, TiiEN

evacuate containment.  ;

l , C. Withdraw the fuel assembly from the upender and store it in the manipulator mast, , TilEN evacuate containment. ' D. Disengage the manipulator from the fuel assembly and leave upright, TiiEN evacuate certainment. i l ? 1 4 1 4 1 Page 86 of 100 . 4 , ...__,_f._.r ,_.. ,_y..,--,--,,.,_.p,, g _y,,7.-._, y. ...,_..,,.p, ,7 -y. , ,._ w.p, ... ,__ ,. _ _ , _ , ,__ y_.,- _ , ,,my._,____m_., . .mw._._ , .rm* indian l'olnt Unit 2 Consslidated Edisom Company of NY REACTOR OPERATOR EXAMINATION You have been directed to coordinate the completion of a System Check Off List (COL) for the Safety injection System. You note that the COL requires Independent Verification. Which ONE of the following staternents is TRUE regarding acceptable practice when conducting INDEPENDENT VERIFICATION ofitems contained in a COL that requires independent verification? A. Operator Perfonning the INDEPENDENT VERIFICATION (Second Checker) may perform the verification at the same time as the First Checker and should record the AS FOUND position of each component. B. Operator Perfonning the INDEPENDENT VERIFICATION (Second Checker) must perfonn the verification independently of the First Checker, and should reposition the component IF it is not in the required position. C.. Operator Performing the INDEPENDENT VERIFICATION (Second Checker) may perform the verification at the same time as the First Checker and should record the AS LEFT position of each component. D. Operator Perfonning the INDEPENDENT VERIFICATION (Second Checkt .ust perform the verification independently of the First Checker, and should record me AS FOUND position of each component. Page 88 of 100 l Indian l'olnt Unit 2 Consx!! dated Edison Company ofNY REACTOR OPERATOR EXAMINATION A maintenance mechanic calls the control room and reports that a vah e he has just removed from the service water system has a black and white STOP TAG attached to the handw! -l that requires the valve to be in the OPEN position. Which ONE of the following actions should you take regarding this report? A. No action is necessary since the valve was tagged OPEN. Work protected by this tagout is not affected. B. Obtain the tagout number and report the finding to the work control center. Work protected by this tagout should cease until the discrepancy is corrected. C. Direct the maintenance anechanic to place the tag on an adjacent valve which is in the OPEN position. Work protected by this tagout is not affected. D. Obtain the tagout number and record the finding ir the SRO log. Work protected by this is not affected since the valve was tagged in the OPEN position. Page 89 of 100 I indian l'olnt Unit 2 l C:nsolidated lidison Company cfNl' REACTOR OPERATOR EXAMINATION While investigating an increase in RCS leakage, preparations are made to make a containment entry using SA0 219, Containment Entry and Egress. S A0 219 requires that the watch chemist sample the atmosphere inside the 80 ft. elevation personnel hatch (airlock) before the entry is made. Which ONE of the following statements is TRUE regarding the reason for sampling the atmosphere inside the airlock before entering the containtnent? A. Measure radioactive gas concentrations to determine if respirattrs are required. II. Ensure that opening inner airlock door will not result in an increase in Pall altborne radioactivity concentration causing ventilation to trip. C. Ensure that sufficient oxygen is available to support human life. D. Ensure that combustible levels of hydrogen do not exist in the airlock prior to persennel entry. a' ,,t' y/ s' ' ' 0 . e i< s ', ,, t ' w J 4 page 90 of 100 indian P: Int Unit 2 Consolidated Edison Company ofNl' REACTOR OPERATOR EXAMINATION During power operation the DIESEL BLDO FIRE PROT OPERATION annunciator is activated. Which ONE of the following statements is correct regarding the response of the Emergency Diesel Generator (EDO) Building Fire Protection System? A. EDO Building Fire Protection System is a dry line system and water i. pray is activated to the entire building by a deluge valve. The above alann could mean that all of the spray nozzles b:.ve actuated. B. EDO Building Fire Protection System is r. concentrated foam system and foam is activated locally by thermostats. The above alann could mean that the foam system has actuated. C. EDO Building Fire Protection System is a llALON system and llALON is activated heally by thennostats. The above alarm could mean that ilALON system has actuated. D. EDO Building Fire Protection System is a wet line system and water spray is activated kwally at each spray nozzle. The above alann could mean that one of the spray nozzles has actuated, i e Page 91 of 100 1 1 indian holzt Unit 2 Co solidated Edis:n C:mpany ofNY REACTOR OPERATOR EXAMINATION Station pohey requires that each individual follow r :actices that ensure that their personal radiation exposure is kept As Low As Reasonably Achievable (ALARA). Consider the foilowing situation:

• Maintenance must be perfonned in an area where the general area radiation levels are 5 mr/hr y and 100 mr/hr p'

• The maintenance activity is estimated to take 1 person I hour

• Most of the radiation is due to contamination of the floor in the area.

Select the ALARA practice that would result in the lowest achievable rcdiation exposure (total ManRem) for this job? A. Equip worker with a plastic face shield and protective clothing against contamination. D. Decontaminate the area (requires 2 people and 45 minutes) before commencing work. C. Cover the floor in plastic sheeting (requires 2 people 30 minutes) before commencing work. D. Cover floor with lead blankets ( requires 3 people 30 minutes) before commencing work. i l i l Page 92 of 100 indian P;l::t Unit 2 Cons:lidated lidis:n C:mpazy ofNl' REACTOR Ol!ERATOR EXAMINATION You are performing a Reactor Startup. Your watch relief from the on-coming shift has arrivedjust as you are ready to withdraw the control banks to approach criticality. Which ONE of the following statements describes an acceptable practice when conducting shift turnover during a Reactor Startup? A. Conduct shift turnover at the flight panel as you continue the reactor startup ensuring that you are not distracted from monitoring neutron flux. B. Do not begin control bank withdrawal. Conduct shift turnover after neutron flux has stabilized. C. Allow your relief to continue the reactor stanup as you relay pertinent watch turnover infonnation to him. D. Continue the startup while the rest of the crew conducts watch turnover. Page 93 of 100 1 Indian Point U::lt 2  ! Ccnstildated Edison Company ofNY REACTON OPERATOR EXAMINATION While performing a plant cookiown to 350'F using POP 3.3, Plant Cooldown, you notice a handwritten notation in the right margin of the page that contains the following infonnation TPC 96153.  ; Which ONE of the following statements is correct regarding the significance of this notation? A. The notation refers to a Temporary Procedure Change. I must refer to the Temporary Procedure Change Log Book located in the control room to perform the associated step.

11. The notation refers to a Temporary Procedure Change. The step that is affected will be lined out and the correct infonnation entered in the body of the procedure.

C. Handwritten notations are not permitted in plant procedures. I should notify the SRO and obtain a clean copy of the procedure. D. The notation refers to a Temporary Procedure Change. I must refer to the Temporary Procedure Change Request Fonn located at the beginning of the procedure to perfonn the associated step. 1 Page 94 of 100 I. Indian P: Int Unit 2 Consolidated Edisin Company cfNY REACTOR OPERATOR EXAMINATION , An event has occurred requiring notification of the NRC within I hour. Which ONE of , the following communications systems should be used to perform this notification?  ; i A, Emergency Notification System (ENS) Phone B. - Radiological Emergency Communication System (RECS)  ! C. Microwave Phone 1.ine D. State Emergency Management Office (SEMO) Radio i I i Page 95 of 100 - 1 -- r- .,,,y

• 4eyg,- g y ++*<** --- w npi ny m y v -----y-e-, y- w,y,mw wesm -e w e v e - . -e.-75.+,---em ,-n-r+-4-,-% mea w-,w-e----e m--en - pmet='e v > gase w y ym w7y -r

I:di:n Point Unit 2 Cons:lidated Edison Company ofNY REACTOR OPERATOR EXAMINATION i t Which ONE of the following is the minimum Emergency Plan Classification which would  ! require " Level 2" staffing? A. Site Area Emergency l B. General Emergency C. Alert . 1 D. Notification of Unusual Event i 4 Page 96 of 100 indian Point Unit 2 ConsolidatedEdison Company ofNY REACTOR OPERATOR EXAMINATION i Which ONE of the following statements correctly identifies the tagout protection that must be provided if a worker is going to work in a tank that is connected to any system? A. Each source of energy must be isolated by TWO CLOSED, and TAGGED isolation valves or by ONE CLOSED, LOCKED and TAGGED isolation valve. B. Each source of energy must be isolated by TWO CLOSED, LOCKED and TAGGED isolation valves. C. Each source of energy must be isolated by TWO CLOSED, LOCKED and TAGGED isolation valves or a safety person must be stationed at the tank opening to assist in emergency egress. D. Each source of energy must be isolated by at least ONE CLOSED, LOCKED and TAGGED isolation valves, or a safety person rnust be stationed at the tank openlug to assist in emergency egress. Page 97 of 100 hdian P: Int Unit 2 Consolidated Edison Company cfNl' REACTOR OPERATOR EXAMINATION Which ONE of the following gas samples would indicate that the gas space in the . associated equipment contains a potentially flammable mixture? A. Volume ControlTank 97% liydrogen,1% Oxygen,2% Nitrogen j B. 23 CVCS Holdup Tank 26% liydrogen, 0%_ Oxygen,74% Nitrogen C. 22 Large Oas Decav Tank 7% liydrogen,17% Oxygen,76% Nitrogen # D. Main Generator 94% liydrogen,1% Oxygen,5% Nitrogen i i i I Page 98 of 100 .,_..... _ ,-,.- _ ._. .. _ ..-.._.~ _ .._ _ _ _ _ _ _.. _.__.___ _ ____ _ _ ._.._. . _.. _ _ _,! 1:dian l*:l t Unit 2 C::s:lidated Edison C:mpany cfNl' REACTOR OPERATOR EXAMINATION Choose the selection that correctly completes the following statement: The Condensate Storage Tank area is a(n) , the Simulator Building is in the - and the Main Turbine is in the . A. Isolation Zone, Owner Controlled Area, Protected Area B. Vital Area, Isolation Zone, Protected Area C. Exclusion Zone, Owner Controlled Area, Protected Area D. Vital Area, Owner Controlled Area, Protected Area Page 99 of 100 L l Indian Point Unit 2 ) Cc::solidated Edis:n Company cfNY REACTOR OPERATOR EXAMINATION i Which ONE of the following indications is EQI an indication used in the Emergency ' Operating Procedures to verify that Natural Circulation Cooling now has been established? A. SO Pressure stable or decreasing B. Stable or decreasing loop AT C. Core Exit Temperatures stable or decreasing , i D. RCS Cold Leg Temperatures at saturation temperature for 50 pressure 1 l r l t i  : f Page 52 of 100 V Indian l*: Int Unit 2 Cc::s:lidated Edis:n C mpany cfNl' REACTOR OPERATOR EXAMINATION Fallowing a Manual Reactor Trip you are unable to verify that the Reactoris tripped in accordance with E 0, Reactor Trip or Safety injection, and FR S.I. Response to Nuclear Power Generation /ATWS. is implemented. Which ONE of the following actions is ,N,.QI a method directed to be used in FR S.I. Response to Nuclear power Generat;on/ATWS. for tripping the Main Twbine if the turbine trip cannot be verified? 'A. Manually trip the turbine from the CCR. B. CLOSE the MSIVs from the CCR. C. Trip the turbine locally at the " Front Standard". D. Manually mnback the turbine in the CCR and locally CLOSE the MSIVs Page 71 of 100 _. . . _ _ _ _ _ _ _ . . _ . . _ . _ . . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ . .._.._..__.m_ indian Point U;:lt 2 l Cons:lidated Edison Company ofNY REACTOR OPERATOR EXAMINATION i Which ONE of the following AUTOMATIC actions will NOT occur if a single control ) rod drops froni the fully withdrawn position to the fully inser1ed position while the reactor j is at 100ft power? b

• A. NIS Dropped Rod Stop 1 B, APPROACillNO ROD INSERTION LIMIT 12.5" and the ROD INSERTION

! LIMIT 0" alanns actuate on Panel SA. .i 4 C. Turbine Runback D. Rod Bottom Rod Stop { i ._ i t 4 e .i i l i 1 i 1 i "i i 4 1 Page 74 of 100 - Indian itixt Unit 2 Cens:lidated Edison Company ofNY REACTOR OPERATOR EXAMINATION When operating the Residual lleat Removal System (RilR) with the Reactor Coolant System (RCS) at reduced inventory care must be taken to control RCS water level such that the RilR pumps do not cavitate or become airbound. SOP 4.2.1, RiiR System Operetion, and AOl 4.2.1, Loss of Residual lleat Removal System, impose restrictions on minimum RCS water level based on certain conditions / parameters. Which ONE of the following parameters / conditions is b'QI a factor in determining the minimuni allowable RCS water level? A. Position of RilR Mini Flow Test Line Stop Valves (MOV 743/1870) 4 p, B. RilR System Flow Rste 4s' y)# C. Position of RilR Mini Flow Test Line Bypass Stop Valve (1819) [f D. Which RilR Dump is Running Page 77 cf 100 - - - - - - - - ~ ~ - ^ Indian l'ct::t Unit 2 Censolidated Edissn Company ofNY { t REACTOR OPERATOR EXAMINATION l t Which ONE of the following indications is NOT used by the Emergency Operating  ! 4 Procedure E 3, Steam Generator Tube Rupture, to identify the RUPTURED Steam  ;

Generator?

l' t I  ! A. Unexpected level rise in an) SO narrow range level. { l B. High Radiation in a Main Steam line. t

C.111gh Radiation from the Steam Jet Air Ejector Vent Radiation Monitor R45A/B.

l D. liigh Radiation from any 50 blowdown line (R-49). , I l i ' d i i t P i i ? Page 82 of 100 indian P: Int Unit 2 Ccas:lidated Edison Ccmpany cfNl' 1 . 1 REACTOR OPERATOR EXAMINATION l Following a Reactor Trip and Safety injection you are verifying 480VAC Busses l energized by offsite power using EOP E-0, Reactor Trip or Safety Injection. You note the following conditions associated with the 480VAC Busses:

• All 480VAC Bus Normal Feeder Breakers OPEN

• All Emergency Diesel Generators in Service e All 480VAC Bus Emergency Feed Breakers CLOSED ,

i Which ONE of the following actions should NOT be performed when completing this IMMEDIATE ACTION step? A. Start ONE charging pump in MANUAL at maximum speed. B. Ensure the following MCCs ENEROlZED e MCC 26A e MCC 26B '

• MCC 26C C. Reset Lighting

• D. Ensure the following MCCs ENERGlZED e MCC 24A e MCC 27A e MCC 29A e MCC 211 F

i Page 87 of 100 > . . . _ _ . _ . _ _ . - . _ . _ _ . _ ._____.______.____.m _ _ . _ _ _ _ _ _ . - Indian Pol:t U:lt 2 - Ceas:lidated Ediscn Ccmpa::y ofNY RE ACTOR OPERATOR EXAMINATION Which ONF. of the following Central Control Room log sheet entry examples would NQI require that the entry be RED circled and exphtined in the remarks section? i A. Technical Specification reading exceeds NORMAL limits specified on log sheet. ) B. Equipment is out of service. i ] C. Reading exceeds MIN / MAX limit specified on log sheet. D. Reading is taken one hour late due to startup activities. 2 i i i 1 .I 4 I i ? , 4 A 1 s A i 1 Page 100 of 100 c. a -,-- -- u v ,%,, ... , , , , ,-,,w - ,,v-.--%.---w,,97 %y,-p g<- py-._ , ,m, ..., r--y m . w. - . . - - - -,. -.-,,_ ,a t--,.,+-+*.. - e , EXAM QUESTION DATA SilEET Enllal K!M R/AJM11NG MWIY.EUhrILON I (KM tCYCS) A232 31 (1) Reactis ily Control I ETEht: Abihty to predict (a) the impacts of the following malf unctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct. control, or mitigate the consequences of those malfunctions or operationc  ! h/AIO11C lispected reactivity changes af ter vahing in a new rnised-bed demineraliter that has not been pre. borated QUEST 1QN; Consider the following initial conditions when selecting your answer: o Reactor Power 1007, o RCS T.,, 559"P o Control Rods Automatic /ilank D W 215 Steps o RCS Iloron 1070 PPM Chemistry has adsised the contiol room that #21 CVCS Mixed fled Demineraliter resin is exhausted. The Reactor Operator is directed to coordinate with the Nuclear NPO to place #22 CVCS Mised fled Demmerallier in hersice. c22 CVCS Mixed fled Demineraliser was replaced last week. The demineralit:r has remal.ied isolated since resin replacement. No operations have been performed on this elemin-taliter since that time. Which of the following statements describes the potential effect,if any, that pl"cing #22 CVCS Mixed led Demineraliter in service withoul saturating at cristing RCS boron concentration will base on the following RCS parameters:

• RCS T.,,

• Control Rod Position

• RCS Boron Concentration ANSWER C. RCS T.,, INCREASL!

Control Rod Position DECREASE RCS lloron Concentration DECREASE DIS'fRACTERS A. RCS T.,, DECREASE Control Rod Position INCREASE RCS lloron Concentration INCREASE 11, RCS T.,, NO CilANGE Control Rod Position NO CilANGE RCS Iloron Concentration NO CilANGE D. RCS T.,., DECREASE Control Rod Position DECREASE RCS Iloron Concentration INCREASE h.Qlli 0111ECI1Yb REERENCIE sop 3.5 Page i of 100 _ - = - __ _ ___ - . _- - ._. .. EXAM QUESTION DATA SilEET H111iM WM WA_IM ERO SAEli1HUNCHON 059 (htain Feed Water) A4.12 34 (4) Reactor llent Remm al ETliM: Ability to manually operate and monitor in the control room: WAIRPIC Initiation of automatic feedwater holation QL' lie 110E Which of the following statements correctly descr;tes the status of the hiain and Low flow Feedw ater Regulating Vahes after a reactor trip fror.) 1009 power and subsequent five (5) minute cooldown to $30*F due to a stuck open atmospheric steam dump , vahe? ~ ANSWER II. The hialn Feed Regulating Valves AND the Low Flow Feed Regulating Valves are Cl.OSFD DISTRACTERS A. The hiain Feed Regulating Valves are CLOSED, and the Low 1%w Feed Regulating Vahes are 75% OPEN (1009 power position). C. The hiain Feed Regulating Valves are full OPEN due to large lesel error signal caused by " shrink". The Low Ilow Feed ReFulating Valves are 759 OPEN i (l004 power position) D. The hlain Feed Regulating Vahes are full OPEN due to large level error signal

caused by " shrink" 'the low flow Feed Regulating Valve . are Cl.OSED.

EDIliE 1 RLUliG1Vli IEEliRl;NClii E0 Page 2 of 1(X) EXAM QUESTION DATA SilEET EYElliM K!%

• Kn_lMT1hD EAD11111thCllDN 001 tControl Red Drive) K4 07 3.7 (1) Reactivity Control STM1:

Anowledre of the CRDS design featuret si and/or interhieL(s) w hich prmide for the following: KM1011C Rcd Stops QL'111103 Consider the following initial conditions when selecting your answer: l

• Reactor Po,ser 100%

o RCS T ., 559'F o Control Rods Automatic / Bank D W 215 Steps 'Ihe narrow range hot leg RTI' for leop 23 f ails in9sntanecaal, high due to an open circuit. Which of the following statements correctly describes the ef fect of this f ailure on the Rod Control System? ANSWiiR C. No rod motion occurs AUTOh1ATIC rod in ertion and withdraw al are blocked. h1 ant f Al. tod insertion is available. A1 ANIf AL nid withdrawal is bhicked DISTRACTl!RS A. Rods utomatically insert to restore indicated T.,, to T,,,. Rod withdrawal is bhicked in AUTOh1A11C and h1 ant!Al IL No tod motion occurs. Rod insertion and withdraw al are bhicked in AUTOh1 ATlc l only. h1 ANUAL tod insertion and withdrawal are available. l D. Rmh automatically ir, sert to restore T.,, to T,,,. AUTOh1ATIC and h1ANUAL tod insertion and withdraw al are bhwked. ND1lii OlHlI11Yli BillBliNGE AOl 2H.1 Page 3 of 100 l EXAM QUESTION DATA SIIEET n_ DIRM RM KULRAllNG FK 'JECIl0N O W l?CCS) A1.18 4.0 (2) RCS Inventory Control SDiM: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: K/AIDMC Pressurlier lesel and pressure. l QL!bH103 Consider the following event when selecting your answer: o A small breals LOCA has occurred 1 o "Ihe reactor was tripped frorn 1009 power l

• A Safety injection was initiated due to low pressurlier level o

Emergency Operating Proecdure ES 1.2, Post LOCA Coold. .vn and Depressuritation is being used to reduce safety injcetion flow. Which of the following statements cortectly descrihes the anticipated response of presruriier level and pressure immediately after the first Safety iniection Pump is stopped? Ar45WER C. Preuurtier lesel AND prest.ure will Dt!CREASl! due to reduced iniection flow DISTRACTERS A. Prenuriier level AND prenure will INCRiiASE due to voiding in the reactor head

11. Pressuriier level will DECREASE, pressurlier pressure will INCRE ASE as water in the pressuriier flashes to st(am.

D. Prenurlier preuure will DI! CREASE and pressuriier level wnf INCRiiASE due to soiding in :he reactor head. NOIliS; DNiiCIlXL RilELIENCES; ES 1.2 Hackground Page 4 of 100 EXAM QUESTION DATA SIIEET SYSEihl K%# K/A RATING SAFETY lilNCTLOli !$; I W 039 ihiain and Rht Steam) K4.0M 33 (4) Reactor lleat . .cmoval Y SBihl: Knowledge of the h1RSS design feature (s) anNor interlock (s) w hich provide for the followine: .8/M IDElC Interlocks on h1SIVs and bypass s ahes OUESTION; Which of the following statements describes the purpose and operation of the hiain Steam isolation Valse (h1SIV) 86 Relays? ANSWER D. The htSIV 86 Relay actuates when the h1SIV daes not indicate full OPEN and initiates a Turbine Trip to pres ent a Safety injection signal from being gensted DISTRACTERS A. De htSIV 86 Relay actuates only when the LCR control switch is placed in the CLOSE position to ensure the h1SIV stays CLOSED. C. he htSIV 86 Relay actuates only w hen a hiain Steam Isolation signal occurs to ensure the htSIVs stay CLOSED. D. The h1SIV b6 Relay actuates when the htSiv indicates OPEN to seal in the solenoids 'o maintain the htSIV in the OPEN position. t NOB 12 OllHiCIlyli REFERENCEE I c Page 5 of 100 ) EXAM QUESTION DATA SHEE1 SYSTlihl h/Ap }UA RATING MECTJ tTNQ0N 001 K5.06 3.R (1) Meactivity Control EILhl: Knowledge of the following operationalimplications as they apply to the CRDS: K/A TOPIC: Effects of control rod motion on axial offset. QUESTION: Consider the following event when selecting your answer: With the plant at 1004 Reactor Power, control rod 11-8 (center of core), drops approximately 100 steps (62 inches) from the bank pceition of 215 steps. Which of the following statements correctly describes the effect, if any, that this event will have on axial

core power distribution (delta.1)?

Assume that a turbine runback DOES NOT occur. ANSWER A. Delta I will become more negative due to reduced power generation in the top of the core and increased power generation in the bottom of the core. DISTRACTERS B. Delta I will become less negative due to increased power generation at the top of the and reduced power generation at the tmttom of the core. C. Delta I will not chance since rod H.8 is at the center of the core and will affect all quadrants equally. D. De b I will initially become more negative then return to its original value when positive reactivity from the pow er ,rgegi,cgigj eturns reactor power to its original value. L ffi3... QIl((OflVE IWFERENCES: A Ol16.1.1 i P i 4 4 l Page 6 of 100 EXAM QUESTION DATA NIEET SYSTEM K/A! K/A RATING SAFETY [ UNCTION 003 K3.04 3.9 (4) Rx Heat Removal STEM

• Knowledge of the effect that a loss or malfunction of the RCPS will have on the followine:

K/A TOPIC.; Reactor Protection System QUESTION: During operation at 100% reactor power,6.9 KV Bus 1 normal supply breaker (UT-1) trips open due to a relay failure and a , reactor trip occurs. From the choices below. select the protection signal which initiated the reactor trip? ,gSWER D. Single loop loss of flow DISTRACTERS A Two loop loss of now B. 6.9 KV bus undervoltage C. Reactor Coolant Pump under frequency. NQJI.E OBJECTIVE REFERENCE 4; ARP FDF Page 7 of 1(X) EXAM QUESTION DATA SIIEET S.YSTEh1 K/A # J2A_]MTJNQ SAFETY FUNCTION 061 K4.04 3.1 (4) Reactor Heat Removal STEM Knowledge of the ARY design feature (s) and/or interkrL(s) w hich provide for the following: K/A TOPIC.;

Prevention of AFWpump runout by limiting AFW flow..

OUESTlDlt Consider the following event when selecting your answer: 1 Following a reactor trip from 100% power due to a loss of both Main Boiler Feed Pumps, the Motor. Driven Auxiliary Feedwater ' Pumps are being used to control Steam Generator (SG) Levels. Shortly af ter the AFW system is placed in service the piping downstream of the Ausiliary Feed Regulating valve to #24 SG ruptures. Which of the following statements correctly describes the response of the AFW system to this failure? ANSWER C. nc #23 SG and #24 SG auxiliary feed regulatingIalves will automatically modulate to prevent pump runout and maintain sufficient AFW discharge pressure to maintain auxiliary feed now to #23 GG. Feed now to #21 and #22 SG will not be affected. DISTRACTERS j A. De #24 SG auxiSarv feed regulating valve will automatically close. Feed How to #21. #22 and $/3 SG will not be affected. B. The #22 SG and #24 SG auxiliary feed regulating valves will automatically modulate to pr: vent pump runout and maintain sufficient AFW discharge pressure to maintain ausiliary feed flow to #22 SG. Feed flow to #21 and #23 SG will not be affected. D. Sinec #23 and #24 SGs are supplied by the same AFW pump, feed flow to both SGs will be completely lost, Feed Dow to 021 and #22 SG will not be affected. EDTES: QDJJiCTIVE

REFERENCES:

SOP 21.3 4

4 i

9 4

Page 8 of 100

. . - _ _ _ _ ~ _ _ _.-

EXAM QUESTION DATA SIIEET SYSTEM KIA_# K/A RATING S AI ETY FUNCTIO _B 015 K4.05 43 (7)lnstrumentation S_TliM:

Knowledge of NIS desien feature (s) and/or interlock (s) provide for the following:

K!A TOPIC:

Reactor trip QUESTION; Due to failure of a pre-amplifier in the circuit for NIS Power Range Channel N42, the lower detector output has failed to zero.

Total indicated power from NIS Power Range Channel N42 indicates 51% with the Reactor at 100% power. In order to repair the pre ampliner the channel must be removed from service by removing the instrument power fuses.

Which of the following statements correctly states the effect that this operation will have on the Power Range Nuclear instrumentation System and Reactor Protection System?

ANSWER B. The high flux bistables associated with N42 will trip. A minimum of one of the remaining three power range channels must sense a high flux condition to trip the reactor.

DISTRACTERS A. The Nigh fiux bistables for N42 will be inhibited from tripping. A minimum of two of the reme.ining three power range channels must sense a high Oux condition to trip the reactor.

C. The high Oux bistables associated with N42 will trip. A minimum of Iwo of the remaining three power range channels must sense a high Oux condition to trip the reactor.

D. The high Dux bistables for N42 will be inhibited from tripping. A minimum of one of

the remaining three power range channels must sense a high Hux condition to trip the reactor.

NOTES:

OBJECTIVf, BEFERENCES:

AOI 13.1.3 Page 9 of 100

EXAM QUESTION DATA SIIEET Si 5111A1 LOA.! K/A RATlNG SA[XIV FUNCTIO _.fi 017 K4.01 34 (7) Ic.trumentation STEM Knowledge ofITM system design feature (s) and4r interlock (s) w hich provide for the following:

j K/A TOPIC:

Input to subcooling monitors.

I QUES 110N:

Which of the following correctly identifies the instruments that provide the pressure signal for the determination of the wbcooling value displayed on Flight Panel (FD)?

ANSW11R D. RCS Wide Range Pressure Instruments (PT-402/403)

DISTRACTERS A. Pressuriier Pressure f Channels I and II)

, B. RCS OPS System Pressure Instruments (PT-413/433/443)

C. Pressuriter Pressure (Channels ill and IV)

NOTES, QBJECTIVE REFFRENCESj i

SOP 14.2 I

a a

6 a

i

?

Page 10 0f 100

EXAM QUESTION DATA SHEET TYSTFM R/A_# K/A RATING SAFETY FLkNCTIQN 072 A1.01 3.4 (7i instrumentation S.IliM:

Ability to predict and/or monitor changes in parameters (to pievent exceeding design limits associated with operating the ARM system controls including:

K/A TOPlQ Radiation Levels q OUESTION:

, Consider the following indications when selecting your answer:

e Reactor Power 100%

e Tm 559'F 3 e Control Bank D 215 steps / automatic s Containment Area Radiation Monitor (R2) Alanning a e Containment Area Radiation Monitor (R7) Alanning

• Containment Atmosphere Radiation Monitors (R41/42) Normal e Charging Pump Cell Area Radiation Monitor (R4) Alarming y e Sample Cell Area Radiation Monitor (R6) Alarming

• Plant Vent Radiation Monitors (R43/R44) Normal From the list below select the event which would explain the indications described above?

ANSWER

A. Reactor core fuel element failure DISTRACTERS H. RCS leak in containment C. RCS/CVCS leak in charging cell D. RCS/CVCS leak in sample cell NOTES

OBJECTIVE

REFERENCES:

Al2.1 J

r 4

i Page 11 of 100

EXAM QUESTION DATA SIIEET SYSTTiht W }UA RATING SAEETY FUNCT10N 068 A4.03 3.9 (9) Radioactive Release STEN 1 Ability to manually operate and/or monitor in the control room: I K/A TOPIC: l Stoppage of release if limits exceeded QUESTION:  !

While discharging #13 Waste Distillate Storage Tank (WDST) to the river using #14 Waste Distillate rransfer Pump (WDTP),

an alarm occurs on Panel SAF-1, "R 54 LIQUID WASTE DISTILLATE HI RAD / TROUBLE" due tu high radiation signal..

Which of the following correctly identifies the anticipated response of the following Liquid Radwaste System components:

• #13 and 14 WDTPs e #13 and 14 WDTP Discharge Valves (SOV CT 965-h1CV and SOV CT-982-htCV) e Common WDTP Discharge Vah e (SOV CT-97l-FCV)

ANSWER D, #13 WDTP STOPPED

1. 14 WDTP STOPPED

1. 13 WDTP Discharge Valve (SOV CT-965 h1CV) CLOSED

1. 14 WDTP Discharge Valve (SOV CT-982 h1CV) CLOSED Common WDTP Discharge Valve (SOV CT-971-FCV) CLOSED j DISTRACTERS A #I3 WDTP I UNNING

1. 14 WDTP STOPPED

1. 13 WDTP Discharge Valve (SOV CT 965-htCV) CLOSED

1. 14 WDTP Discharge Valve (SOV CT 982 h1CV) CLOSED Common WDTP Discharge Valve (SOV CT-971-FCV1 OPEN B. #13 WDTP STOPPED

1. 14 WDTT' STOPPED

1. 13 WDTP Discharge Valve (SOV CT-965-h1CV) CLOSED

1. 14 WDTP Discharge Valve (SOV CT-982-htCV) CLOSED Common WDTP Discharge Valve (SOV CT 971-FCV) OPEN C. #13 WDTP RUNNING

1. 14 WDTP STOPPED

1. 13 WDTP Discharge Valve (SOV CT-965 h1CV) OPEN

1. 14 WDTP Discharge Valve (SOV CT-982-h1CV) CLOSED Common WDTP Discharge Valve (SOV CT 971-FCV1 CLOSED NOTES:

OBJECTIVE BEERENCEF:

ARP SAF 1, Window 4-4 4

Page 12 of 100

EXAM QUESTION DATA SIIEET SYETEM K!M K!A_IMIING S AFETY FUNCTION l 071 A4.26 3.1 (9) Radioactive Release SIliM:

Ahility to manually operate and/or monitor in the control room:

K/A TOPIC:

Authoriied waste gas release, conducted in compliance with radioactive gas discharge permit.

QUESTION; The Senior Reactor Operator has directed you to initiate a Containment Building Pressure Relief. Approximately 15 minutes after the release has been initiated the S 133B, Meteorological Data Display on Accident Assessment Panel stops functioning.

Which of the followine actions should you take to compensate for this failure?

ANSWER C. Venfy that meteorological data is available and record meteorological data every hour throughout the remainder of the release.

DISTRACTERS A. Immediately stop the release and has e no further releases until the display is repaired B. Record the Plant Vent Radiation Monitor (R44) readmg every hour until the release is terminated D. Stop the release and prepare a new release permit using the most adverse meteorological conditions.

NO113S:

OBJECTIVE REEERENCES; SOP 5.2.4 (Rev 18)

Page 13 of 100

EXAM QUESTION DATA SIIEET SYSTEM NM M6_ RATING SAEfY FUNCHON 4

ON K613 3.1 (1) Reactivity Control ETILM:

Knowledge of the effect of a loss or malfunction on the following CVCS componente K/A TOPIC; Purpose and function of the horation/ dilution controller.

! QUESTION:

During power operation the CVCS Automatic VCT Makeup System initiates blended makeup due to the VCT level reaching the low !cvel setpoint An air leak on the supply line to Boric Acid Flow Control Valve, FCV-Il0A, causes air pressure to the valve diaphragm to decrease to O PSIO and prevents the valve from responding to the flight panel controller (FIC-110A). The air leak is small enough that it does not hase any significant effect on instrument air header pressure.

Which of the following statements correctly describes the effect,if any, that this failure will have on the boric acid concentration of the Reactor Coolant System?

ANSWER B. RCS Boric Acid Concentration will INCREASE because FCV 110A fails OPEN on a loss of air pressure, causing blended geup to have a higher than desired boric acid concentration.

DISTRACTERS A. RCS Boric Acid Concentration will DECREASE because FCV-110A fails CLOSED on a loss of air pressure, causing blended makeup to have a lower than desired boric acid concentration.

C, RCS Boric Acid concentration will NOT CHANGE because FCV-110A is only used when the Makeup Selector Switch is in the BORATE position.

D. RCS Boric Acid Concentration will NOT CilANGE because FCV Il0A is only used when the Makeup Selector Switch is in the M ANUAL pos: tion.

2 NOTES; OBJECTIVE

REFERENCES:

AOI 29.2 4

^

t J

Page la of 100

EXAM QUESTION DATA SilEET I SXEllihl K/A # K/A RATING SAFETY FUNCTION 013 A 1.01 4.0 (2) Inventory Control SlEhl:

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ESFAS controls including:

K/A TOPIC; RCS pressure and temperature.

QLESTION:

Consider the following conditions when selectinF your answer:

e A small break LOCA has occurred e RCS pressure is at 1900 PSIG e The Safety injection System has automatically actuated due to liigh Containment Pressure e ALL ESF equipment has responded as designed e RCS break How is approximately 800 GPhi e No Charging Pumps are operating Select the response which correctly describes the status of the following ESF equipment, and RCS parameters which will exist when staNe conditions are reached in the Reactor Coolant System? Assume that a cooldown HAS NOT been initiated, e St Pump Status e RilR Pump Status e RCS Pressure e RCS Break Flow ANSWER B. St Pump Status Running - Flow to RCS RHR Pump Status Running - On Recirculation RCS Pressure LESS THAN 1500 PSIG and GREATER THisN 150 PSIG RCS Break Flow LESS THAN 800 GPhi DISTRACTERS A, S1 Pump Status Running Recire to RWST RHR Pump Status Running - Flow to RCS RCS Pressure LESS THAN 1500 PSIG and GREATER THAN 150 PSIG RCS Break Flow LESS THAN 800 GPh1 C. SI Pump Status Running - Recire to RWST RHR Pump Status Running - Flow to RCS RCS Pressure LESS THAN 150 PSIG RCS Break Flow GREATER THAN 800 GPhi D. St Pump Status Running - Flow to RCS RHR Pump Status Running - Flow to RCS RCS Pressure LESS THAN 150 PSIG RCS Break Flow LESS THAN 800 GPhl NOTES:

O_JJECTIVE BEEREh'CJES; ES 1.2 Background Page 15 of 100

EXAM QUESTION DATA SIIEET SYSEiM J2Af JUA RATING SAFETY FUNCTION 056 A2.01 2.6 (4) Reactor lleat Removal SIDj:

Ability to (a) predict ths impacts of the following malfunctions or operations on the conden; ate system; and (b) based on those predictions, use procedures to correct, control. or mitiente the consequences of those malfunctions or operations:

h/A TOl'IC Loss of condensate pumps OUESTION:

Consider the following initial conditions when selecting your answer:

• Reactor Power i X 10'IR amps e T., 547"F l

• Plant startup in progress 2

During a plant startup, after the Main Steam Isolation Valves have been opened and Main Condenser Vacuum has been l established, a fault is detected on the S ation Auxiliary Transformer causing a loss of 138 KV electrical power to the station.

Assuming no operator action, which of the following describes the immediate impact that this failure will have on the operation of the following:

Ccndenstate Pumps Condenser vacuum Reactor ANSWER C. Condensate pumps Stopped

, Condenser sacuum Decreasing Reactor Tripped DISTRACTERS A. Condensate pumps Stopped Condenser vacuum Stabic Reactor Critical B. Condensate pumps Running Condenser vacuum Decreasing Reactor Tripped ,,

D. Condensate pumps Running Condenser vacuum Stable Reactor Critical NOTES, Air ejector steam supply ',alve will close and Circulatine Water pumps will trip. No automatic reactor trip results.

OBJECTIVE

REFERENCES:

AOI 27.1.1/ SOP 20.2 p. 7 of 33/ SOP 20.1 p. I Page 16 of 100

d EXAM QUESTION DATA SIIEET SYSTI31 E/M K/A RATib'G SAFETY FUECIlOS 022 ) A2.05 3.1 (5) Containment Integrity Elfdl; Ability to (a) prediet the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:

K/A TOplC:

Maior leak in CCS (service water)

DMfiTI.ON; During power operation a catastrophic failure of the service water supply line to the #23 Containment Fan Cooler Unit results in flooding of the containment building including the filling of the reactor cavity sump to the 46 ft. elevation of the containment

building. How and why could this failure impact the ability of the Emergency Safeguards System (ESP) to perform its design function af a design bases large break LOCA were to occur?

, Assume all other ESF equipment operates as designed when the accident occurs.

ANSWER D. Containment pressure and water level may exceed design limits due to reduction in containment free volume before the

, accident occurred.

DISTRACTERS A Containment pressure may exceed design limits due to loss of cooling water to the Containment Fan Cooler Units w hich are required to meet the minimum safeguards equipment requirements.

B. Accumulator isolation valves will be submerged and may fail to open when signaled resulting in failure of the accumulators to inject into the RCS.

4 C Cold leg recirculation will be impossible since the recirculation pumps will be submerged preventing the establishment of any recirculation path.

NOTES:

Q_BJECTIVE

REFERENCES:

Industry event /FR.Z hackcround/FS AR Cil.15 Page 17 of 100

l EXAM QUESTION DATA SilEET I l

SYST!ihl E/A! K/A RATING SAFETY FUNCTIO.h' l 015 K3.02 3.9 (71 Instrumentation S_Ilihi:

Knowledge of the effect that a loss or malfunction of the NIS will have on the following:

N/A TOP _LC; CRDS OUESTIOjt Consider the following initial conditions when selecting your answer:

• Reactor Power 1009

• T.,. 559'F e Contiol rods Bank D (6 215 Steps / Automatic During power operation with the above conditions Power Range Channel N41 fails to maximum detector output (1209). Which of the fcllov ing statements correctly describes the effect that this failure will have on the Rod Control System?

3 ANSWER H. Control rods will automatically insert until the rate of change of indicated reactor power vers':s turbine power decays to zero.

DISTRACTERS A. Control rods will automatically insert to reduce indicated reactor power to equal turbine power.

C. Control rod motion will be inhibited due to the overpower rod stop.

D. Control rods will automatically insert due to a turbine runback initiated by the change in indicated teactor power exceeding 5% in 5 seconds.

NOTES; 1

J OBJECTIVE REFERJINCES:

AOI 13.1.3 i

Page 18 of 100

EXAM QUESTION DATA SHEET SXh]IM K/A # K/A R_bTING SAFETY FUNCTION 071 A3.02 3.6 (9) Radioactive Release S3TM:

Ahility to monitor automatic operation of the Waste Gas Disposal System including:

E/A TOPIC Radiation monitoring system alarm and actuating signals.

OUESTION:

While releasing #24 Large Gas Decay Tank (LGDT) the Nuc) car NPO inadvertently opens RCV-014 (LGDT Release Hand Control Valve) to the full open position causing a the R 43/R 44 PLANT VENT HI RAD /FROUBLE alarm to actuate on panel SAF 1 in the control room, due to R-44, Plant Vent Gas Monitor, exceeding the High Radiation Gas alarm setpoint.

Which of the following statements describes the impact,if any, that this event will have on the following equipment:

e RCV-014

• PAB Supply Fan e PAB Exhaust Fan ANSWER A. RCV-014 Closes Pall Supply Fan No change PAB Exhaust Fan No Change DISTRACTERS B. RCV-014 Closes PAB Supply Fan Trips PAB Exhaust Fan No Change C. RCV-014 No change PAD Supply Fan Trips PAB Exhaust Fan Trips D. RCV-014 Closes PAB Supply Fan Nc Change PAB Exhaus Fan Trips N91118 QDECTIVE BfFERENCES:

ARP S AF-1 2

Page 19 0f 100

d EXAM QUESTION DATA SilEET l

. EYSIliM K/A # 196_R611NG EMETY l'UNCTIQN 001 K4.01 3.5 (1) Reactivity Control j S11iM:

Knowledge of the CRDS design feature (s) and/or interlock (s) which provide for the following:

h/A TQMC Rod position indication QUESTION:

During power operation the APPROACHING ROD INSERTION LIMIT 12.5" and the ROD INSERTION LIMIT 0" alarms are actuated on Panel SAF in the control room, investigation reveals that all control and shutdown rods are fully withdrawn and that the rod insertion limit is not being violated, indicating that the alanns have been actuated due to a malfunction.

Which on the following components,if malfunctioning, could cause the alanns to actuate?

ANSWER C. Pulse to Analog Converter (P/A Converter)

DISTRACI tiRS A. Individual Rod Position Indicator voltace decreases from 3.45 VDC to 0 VDC.

] B. Rod Bottom Rod Stop Bistable i D. Rod Bottom Rod Stop Bypass Bistable NOTTIS:

DBJECTIVE BffiE.RENCES:

Sys. Dese. RPIS I

4 4

J 9

1 1

Page 20 of 100

EXAh! QUESTION DATA SilEET SYSTEM MAj K/A lMTING SAEliTY IUNCT10_N 003 KS.01 3.3 (4) Reactor lleat Removal SlTM Knowledge of the operational implications of the fallowing concepts as they apply to the RCPS:

NA TOPIC:

We relationship between the RCP" now rate and the nuclear reactor core operating parameters (quadrant power tilt, imbalance, DNB rate, local power density, difference in loop Tue, pressure.

OUESTIONS Consider the following initial conditions w hen selecting your answer:

e Reactor Power 609

• T., 554*F e Control nxis Bank D @ 200 Steps / Automatic During surveillance testing it is discovered that the #22 Reactor Coolant Pump is not performing at design conditions. De flow rate in knop #22 is 79 less than design due to degradation of the diffuser ring in the pump casing. Flow in the remaining loops are at or abose design and total core flow is reduced by 59. Which of the following statements describes how this condition will affect the following:

o Core A - T e Departure From Nucleate Boiling Ratio (DNBR)

ANSWER D. Core 4 - T liigher than design DNBR Lower than design

, DISTRACTERS A. Core A - T Lower than design DNBR Lowcr than design B. Core A - T Lower than design DNBR liigher than design C. Core 4 - T liigher than design DNBR liigher than design NOTES:

OBJECTIVE

REFERENCES:

TS Section 2 0 bases Page 21 of 100

EXAM QUESTION DATA SIIEET EYSTJiM K/Af E!A _RATINO 1&FETY l UNCTIO.J 059 KID 4 14 (4i Reactor }icat Removal ETDJ:

Knowledge of the physical connections and/or cause effect relationship hetween the h1IW system and the following systems:

K/A TOPIC:

Steam Generator Water Level Control System Q1!TJillOli; Which of the following correctly identifies the signals that are used to provide input to the hiBFP Speed Control System for determining the following parameters:

o Actual Fred Regulating Vahe Differential Pressure (Delta P) o Power e 9 Startup S&nal ANSWER D. Delta P hiain Feed licader Pressure - hiain Steam lleader Pressure Power Total Steam Flow (sum of all controlling steam flow channels)

, 9 Startup Signal h1BFP Governor Control Switch DISTRACTERS A. Delta-P hiain Feed lieader Pressure hiain Steam licader Pressure Power Turbine First Stage Pressure (PT 412A) 9 Startup Signal h1BFP Governor Control Switch C. Delta-P hlain Feed lieader Pressure hiain Steam lleader Pressure Power Total Steam Flow (sum of all controlling steam flow channels)

% Startup Signal h1BFP Foxboro Speed Controller h1anual Settine D. ) elta-P hiain Feed 11cader Pressure - Steam Generator Pressure Power Total Steam Flow (sum of all controlling steam flow channels) 9 Startup Signal h1BFP Governar Control Switch NOTF1 QM)iC11YE REFERENCF1 SOP 21.1 Page 22 of 100

f EXAM QUESTION DATA SIIEET SlSll;M K/A # K/A_ RATING S AFETY FUNCTION 001 K(.. I I 2.9 (1) Reactivity Control STEM:

Knowledge of the effect of a loss or malfunction on the following CRDS components:

K/A TOPIC 1.ocation and operation of CRDS fault detection (trouble alarms) and reset system. including rod control ann mciator.

QES110lt While retrieving a dropped control in Control Bank D Group 1, with the unit at 80% power, a ROD CONTROL URGENT FAILURE alarm is received on panel SBF-l in the control room. Which of the following statements correctly describes the cause of this alann?

AN.CWER D. All rods in Control Bank D Group 2 are not moving when demanded.

DISTRAC t t:RS

, A. Remaining rods in Control Bank D Group 1 are not moving w hen demanded.

B. Alarm is caused by the affected rod position deviating by more than 12 steps from bank demand position.

C. Alarm is caused by control bank rods moving with no bank overlap.

NOTES:

QluliCTIVE BEFERENCES; AO!161.1 9

i i

Page 23 of 100

m EXAM QUESTION DATA SHEET SXSIliM E!A # K/A RATIND 5AFITY FUNCTION 013 A2.02 4.3 (2) Inventory Control SIEM:

Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based ability on those predictions, use procedures to correct. control. or mitieate the consequences of those malfunctions or operations.

K/A TOPIC:

4 Escess steam demand.

3 OUESTION; Consider the following event when selectinF your answer:

• Reactor Power 0%

o T., 530'F e Total Steam Flow 0.75 X 10'lbm/hr e Pressurizer Level 32%

o Pressuriier Pressure 2235 PSIG The plant is at hot shutdown due to a tube leak in #21 Steam Generator (SG). The Main Steam Isolation Valve for #21 SG is closed and the Reactor Operator is perf orming a cooldown to Cold Shutdown using the condenser rteam dumps.

While positioning the condenser steam dump vahes, a controllet failure causes ALL of the steam dumps to OPEN fully. A Safety injection Signalis received. Directly afterwards the Reactor Operator notes the following parameters:

o Reactor Power 0%

• T., 523*F e Total Steam Flow 0 lbm/hr e Pressuriier Level 25 9

• Pressuriier Pressure 2110 PSIG Which of the following ESF actuation signah initiated the Safety In_iection signal?

ANSWER B. High Steam Flow coincident with Low T,,,

DISTRACTERS A. High Containment Pressure C. Low Pressuriier Pressure D. Main Steam Line Delta P (21 SG GREATER TH AN 22.23.24 SG)

NOTES _;

QDJECTIVE

REFERENCES:

E 3/ Caution Page 24 of 100

EXAM QUESTION DATA SHEET LUJ11M E!M K/A RATING SAFETY FUNCTION 015 K4.07 3.7 (7) Instrumentation EllM Knowledge of NIS design feature (s) and/or interlock (s) that provide for the following:

K/A TOplC:

Permissives 9UESTION:

Select the answer which correctly identines the automatic reactor trip signals that are blocked when the POWER BELOW P-7 permissive is enabled

• l ANSWER _

' A. Piessuriier Low Pressare Reactor Trip Pressurizer liigh Level deactor Trip Two Loop Loss of Flow 6.9 KV Bus Undermitage DISTRACTERS B. Pressuriier Low Pressure Reactor Trip Ptessuriter liigh Level Reactor Trip Steam Generator Lo-Lo Level Trip 6.9 KV Bus Und woltage C. Presserizer Low Pressure Reactor Trip Press ariier Low Lesel Reactor Trip Two Loop Le ,of Flow 6.9 KV Bus Undervoltage D. Pressurizer liigh Pressu.e Reactor Tri!

Pressurizer tiigh Lesel Reactor Trip Steam Generator Lo-Lo 1.esel Trip 6.9 KV Bus Undervoltace RQTES:

OBJECTIVE

REFERENCES:

POP 3I Page 25 of 100 i

= . . - . _

EXAM QUESTION DATA SIIEET

SYSTIBJ K/A # K/A RATING SAFETY IUNCTIJO 002 KS. ' $ 4.2 (2) RCS Inventory SJ1i}]

Knowledge of the operational implications of the following concepts as they apply to the RCS.

K/A TOPIC:

Reasons for maintaining subcooling margin during natural circulation cooldown.

OUESTION:

While performing a natural circulation cooldown using ES-0.2, Natural Circulation Cooldown, you are directed to depressurite s

the Reactor Coolant System to 1890 PSIG after verifying that the RCS Hot Leg temperatures are less than 550*F. Which of the

' following statements correctly describes the reason for the maximum limit on hot leg temperature before depressurization can commence?

ANSWER D. Ensure that adequate subcooling exists te prevent void formation in the reactor head v hen plessure is reduced to 1890 PSIG.

! DISTRACTERS A. Ensure that the Delta-T limit hetween auxiliary spray fluid temperature and the RCS is not violated B. Ensure that wide range hot leg temperatures are approximately saturation temperature for SG pressure.

C. Ensure that RCS subemling is above the RCP Termination Criteria for the E-0 series of procedures.

BiOTES:

Q1UliCII'5 REFEEENCES:

3 ES 0.2 Background j

i a

h l

j 4

4 Page 26 of 100

EXAM QUESTION DATA SilEET S.LS311M K/A.# S/.A_JMllNQ SAEETY FUNCTION 002 K4.05 3.8 (2i RCS Inventory ElliM:

Knowledge of RCS design featurets) and/or interkick(s) which provide for the following:

K/A TOPIC:

Detection of RCS leak:.se 911Fas TlON; During power operation an increase in RCS leakage is noted during a routine RCS leak rate surveillance test. The subsequent RCS leakage safety evaluation determines that Reactor Coolant Drain Tank in-leakage has increased by the same amount that RCS leak. ige has increased. Which of the following leakage sources could be attributed to both the increase in RCS leakage and the increase in PCDT in-leakage?

ANSWER

{ B. Reactor Vessel Flange O-Ring l DISTRACTERS 4

A. CVCS Letdown Line Relief vijve leakage C. Pressuriier Power Operated Pelief Valve (PORV) Leakage D. Reactor Coolant Pump #3 Seal Leakage NOTESS; OlBJIiCI]YE REFERENCES; j AOI 1.7/ SOP 1.7 i

l 1

4 1

}

i a

4 N

i i

Page 27 of 100

EXAM QUESTION DATA SIIEET SYSTEM K/A.! K/A RATING EAFETY FUNCTlQN 011 A1.01 3.5 (2) RCS imentory STEM:

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR

1.CS controls including

K/A TOPLC;

, PZR level and pressure OUESTIOli.;

During power operation a pressuriier insurge results from a step decrease in power from 100% to 909. Which one of the following automatic control actions is desi nedF to prevent a step decrease in RCS pressure when T., is restored to program and a resultant outsurge occurs?

ANSWER C. w-Pressuri/er backup heaters automatically ENERGlZE when pressuriier level exceeds program by 59.

D'STRACTERS A. Charging pump speed increases to maintain pressuriier level stable.

B. itssuriier backup heaters automatically DE ENERGlZE when pressurizer level exceeds program by 5%.

D. l.etdown line isolation prevents rapid outsurge.

NOTES; ODJEC11VE RFFERENCES:

ARP FCF i

f 4

)

Page 28 of 100

! EXAM QUESTION DATA SIIEET 4

fd'SIEA1 ti/A # K/A RATING S AFETY FUNCTION 010 A 3.02 3.6 (3) RCS Pressure Contr<>l EIf3 Ability to monitor automatic operation of the PZR PCS including:

K/A TOPIC:

Pressuriier Pressure OUESTION:

Which one of the following signals are used to automatically ARM the RCS Overpressurization System (OPS) during a plant ecoldow n to Cold Shutdown?

+

ANSWER A. Two out of three (2/3) RCS Wide Range Cold Leg Temperature RTDs LESS THAN setpoint DISTRACTERS B. Two out of three (2/3) RCS Wide Range Hot Leg Temperature RTDs LESS THAN setpoint C. Two out of three G/3) RCS Wide Cold Leg Pressure GREATER THAN setpoint D. Two out of three (2/3) RCS Wide Cold Leg Pressures LESS THAN setpoint NOTES-l QWECILYE BEF3iRENCES; j ARPSGF 4

I Page 29 of 100

EXAM QUESTION DATA SilEET SYSTM1 W K/A RATING SAETY FUNCTION 002 Kl .11 4.1 (4) Reactor 11 eat Removal SIf41:

Knowledge of the physics! connections and/or cause-effect relat'; 4. ship between the RCS and the following systemt K/A TOPIC:

S/Gs, feedwater systems 91!fEllO3 While performing Emergency Operating Procedure FR 11-1. Loss of Secondary lleat Sink, following a Loss of Coolant Accident (LOCA), you are directed to discontinue use of the procedure if RCS Pressure is LESS TilAN non-faulted Steam Generator (SG) pressure (s). Whleh of the following statements correctly describes why it is not necessary to continue with heat sink restoration  !

if RCS ILsure is LESS TilAN non-faulted SG pressure (s)?

ANSWER C. Secondary heat sink is not required in this condition since core decay heat is removed by break flow.

DISTRACTERS A. FR H 1, will initiate RCS Bleed and Feed. A bleed path is not necessary if RCS pressure is low.

B. Since RCS pressure is less than non-faulted SG pressure, RHR flow may be established to remove core decay heat.

D, Since RCS pressure is less than non-faulted SG pressure, an inadequate Core Cooling condition probably exists, w hich is a higher priority procedure.

tiOTES; OBJfCT(VL

REFERENCES:

FR H.I Background Page 30 of 100

i l

EXAM QUESTION DATA SilEET SYSTEM }2 M K/A RATING SAFETY IUNCTION 035 K6.01 3.2 (4) Reactor lleat Removal SIliM:

Knowledge of the effect that a loss or malfunction of the following will have on the S/GS:

K/A TOPIC; htSIVs OUESTION:

I During power operation a Main Steam Line Break occurs inside containment on the steam line from #24 Steam Generator (SG).

A Safety injection si Fnal is received from the hiain Steam Line Delta-P circuit. During recovery operations the control room operators are unable to close the hiain Steam Line Isolation Valve for #24 SG.

Which of the followinF statements is correct regarding the ability of the operators to control Reactor Coolant System (RCS) temperature during this event?

ANSWER

, H. RCS temperature will continue to decrease until 24 SG has dried out. Subsequent temperature control will be performed using the remaining SGs. and auxiliary feed water flow.

DISTRACTERS A. RCS temperature will continue to decrease until ALL SGs have dried out. Subsequent temperature control will be performed by limiting Auxiliary Feedwater now.

C. RCS temperature will continue to decrease until 24 SG has dried out. Subsequent temperature control will be performed using the Atmospheric Steam Dump Valve (s) on the intact SGs since an automatic hiain Steam Line Isolation was actuated by the Ntain Steam Line Delta-P Safety in_iection signe.l.

D. RCS temperature will continue to decrease until ALL SGs have dried out. Subsequent temperature control will be performed by using RCS Bleed and Feed.

NOTLS.

Non retuen check valves (h1S 2) 03111CTIVE

REFERENCES:

System Lesson Plan Page 31 of 100

EXAM QUESTION DATA SHEET D'.SEM MM MA RATING SAFETY FUNCTION 023 K4.06 3.2 (7)lnstrumentation SlliM:

Knowledge of the RPS design feature (s) and/or interlock (s) which provide for the following:

K/A TOPIC:

Automatic or manual disable of RPS tript OUliSIlON; Which of the following statements correctly describes the purpose of the POWER ABOVE P.10 permissive?

ANSWER D. Allow operator to manually block she Power Range High Flux Low Setpoint and Intermediate Range High Flux reactor trips.

Presents re-instatement of the Source Rang-instruments.

DISTRACTERS A. Automatically blocks the Source Range High Flux and Intermediate Range High Flus reactor trips.

Prevents re instatement of the Source Range instruments.

B. Allow operator to manually block the Power Range High Flux Low Setpoint and Source Range High Flux reactor trips.

Presents re-instatement of the Source Range instruments.

l C, Automatically blocks the Power Range High Flux Low Setpoint and Intermediate Range High Flux reactor trips.

Prevents re-instatement of the Source Range instruments.

EQIFA QJ)JECTIVE ,MlF.RENCES:

System Lesson Plan l

Page 32 of 100

EXAM QUESTION DATA SilEET nhTl;M KMI KMJMTING M1E1Y1:UIC110N 012 Auvi 37 17: Instronientition SILM:

Abdity to snonitor autornatie operation of the RPS. Indudmrt R%10J'lt

'Irip lorie QVlllKh1 During power operation Channel 1 Prenurlier l'reuure Instrument Iails to matirnum output (l(KF4 ). While performing the subsequent actions of the Abnormal Operating Instruction (AOI) the Reactor Operator is directed to trip the following bistables:

l

• Preuuriter liigh Preuute Reactor Trip l

Preuuriier low Prenure Reactor Trip

!

• Prcouriier Low Prenure Safety injection e

Pressurlier law Preuure SI Unbhd.

Which of the following cor;ectsj states the espected status (illuminated / extinguished) of the anoeiated bistable proiltig larnp as each bistable trip switch is placed in the trip position?

I ANSW1R

!! Preuuttier liigh Preuute Reactor 'Irip Entinguished l'reuurlier Low Prenure Reactor Trip lliuminated Preuurlier low Preuure Safety injection illuminated 3 Prenuriier 1 ow Preuure Si linbhick listinruished DISTRACTIiHs _

A. l'rc u uttig ' ch Preuure Reactor 'Irip lis tinguished Prenurisci s,0w I'icuure Reactor Trip thlinguished i Preuurnier lew Preuure Salcty injection Ettinguhhed Preuuriier low I'reuure Si linHoek tiluminated C. Preuurtier Iliph Preuure Reactor Trip  !!!uminated Preuurlier Imw Prenure Reactor Trip listinguished Preuuriier Ixw Prenure Safely injection ExtinFuished

_ Preuuriier t.ow Prenure SI Unbhick llluminated D. Preuuriier liigh Preuure Reactor Trip listinguished Preuuriier low Prenure Reactor Trip Illuminated Prenuriier law Prenure Safety injection listinguished Preuuriier low Prenure Si linbhrL Ettinruished hnTlik DNiiCIlYli lilHilliNGiS AOI 28 5 Page 33 of 100 o

EXAM QUESTION DATA SilEET snlut Kaa K/AJunNO honanuNtnoN 01.1 A2 01/A2 02/A2 05 2 R/3 1/3.9 (1) Res.ctisity Control ,

ETEM:

ANhty to (a) predict the innpacts of the following malf unctions or operations on the RPIS: and (b) based on those predictions, use procedures to correct. control. or initicate the consequences of those malfunctions or oggtiont KalO]'lC leo of of fsite powcr/Imss of power to RPIS/ Reactor Trip R011S1101t Consider the following esent when 'slecting your answer:

While the plant was at 100% Reactor Power a loss of the 138 KV ofIsite power source occurred, followed by a Turbine Trip on low vacuurr. and subsequent reactor trip. The Reactor Operator is perforrning Step 1 of E-0, Reactor Trip or Sinfety injection.

Which of the following t,tatements correctly lists the indications that the 'Acactor Operator will use to serify that the Reactor trip has occurred?

ANSWiiR C. Reactor Trip lireaker Position Neutron Flus DISTRACTliRS A. Rod llottom 1.il' hts kcactor Trip fireaker Position D. Neutron Ilus Individual Rod Position Indication D. Reactor Trip litcaker Position llant Step Counters NQ311:

DIMECI1El; REERiiECl2 F. 0 Page 34 of 100

i i

EXAM QUESTION DATA SIIEET EDilliM MM MA_lM11NO SAllillll'halON 0.' 6 A4 05 3$ ($) Containtnent integrity SID1: 1 Abiluy to manually operate and/or monitor in the control room-WAlGMC Containment spray reset switchen.

RULIE1]DIt While performing actions directed by the timergency Operating Procedures, the Reactor Operator resets tix Containment Spray signal. After depressing the reset push buttons, the Reactor notes that the white indicating lights above the buttons illuminate and remain illuminrud af ter the buttons are released.

Which of the following statements coricelly describes the reason the lights illuminated when the reset push buttons were depretted and remained illuminated when the buttons were released?

ANSWl:.R A. 1hc lights illuminated when the buttons were depressed to indicate that the spray signal was reset.

The lights remained illuminated indicating that an automatic Containtnent Spray actuation slynal was present.

DISTRAC1TRS II. 1he lights illuminated when the buttons were depressed to indicate that the spray signal w as reset.

1he lights remained illuminated indicating that both containment spray pumps were runnine C. Th.i lights illuminated w hen the buttons w cre depressed to indicate that the spray signal could NOT be reset.

The lights remained illuminated indicating that an automatie Containment Spray actuation signal w as present.

D. The lights illuminated when the buttons were depressed to indicate that the spray signal was reset.

The lights remained illuminated indicating that both containment spray pumps were NOT running.

NRlliSi OlHliGlYli KlillilitiNCliS; System I.csson Plan Page 3$ of 100

EXAM QUESTION DATA SIIEET SYh1M1 h/A# K/A. RATING SAlfl111'liGON 029 Kl 03 36 (81I>lant Sersice hotem SID1 Knowledge of the phpical connections and/or cause-ef fect relationship between the Containment Purge Splem and the following spiemt h/A1Gl'lC Gaseous radiation release monitors i RL!Eil103 Which of the following correctly identifies the Process Radiation Monitors which are capable of AUTOMATICALLY terminating a Containment Purge if they sense a high radiation condition?

ANSWIR D. R 44, Plant Vent Oas Monitor R 41, Containment Particulate Monitor R-42 Containment Gai Monitor DISTRACTILRS A. R-44, Plant Vent Oas Monitor R-43. Plant Vent Particulate Monitor R 42 Containment Oas Monitor

11. R 43, Plant Vent Particulate Monitor R 41, Containment Particulate Monitor R 42 Containment Gu Monitor C. R-43, Plant Vent Particulate Monitor i R 44, Plant Vent Oas Monitor R-41. Containment Particulate Monitor SMlli DDJEGXfi liLERiiNC11 ARP S Al'-l Page 36 of 100 l

I m

l l

EXAM QUESTION DATA SIIEET S. Yell;M KIA* K!A_1M11NG SAEILl M G10N 075 A2 01 2.4 (0 Plant Senice Spiems hlliM:

Abihty to (a) predict the impacts of the following analfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct. control. or initigate the consequences of those rnallunctions or operatione KMLlR11C Safety features and relationship between condenser vacuum, turbine trip and steam dump.

01'11S119S; While petfonning a plant cooldown using the Condenser Steam Dumps, a 6.9 KV MOTOR TRIP (COMMON) alann is actuated. The Reactor Operator notes a step decrease in steam now at the same time the alarm occurs. No other annunciators cre actuated.

Which one of the following 6 9 KV motors is the most likely cause of the alarm?

ANSWFR

11. Circulating Wa"r Pumk DISTRACTI.RS A. Reactor Coolant Pump C. Ileater Drain Pump D. Condensate Pump NDlfik RillhGlYl3 RiiERiiNClii SOP 1H 1 Page 37 of 100

EXAM QUESTION DATA SilEET n.mM wM WAJMUNG SAum1!NGQN 086 K.i07 2.5 (H) Plant Scruce Splenu i S.TliM: I Knowledee of the desien feature (s) and/or interlockts) w hich pnnide for the following:

KM_lDflC I MT/O and T/G Protection QL'liS310h; Which of the followine lists correctly identifies the plant equipment protected by concentrated foarn fire suppreuion s: Sterm?

)

ANSWIiR 11, 11)droger. Seal Oil Unit

!! oiler feed Purnp Console '

Clean and Dirty Oil Storage Tanks Turhine Oil Resers oir _ , , ,

DISTRACTERS A. 11;dropen Seal Oil Unit Main and Auxiliary Transfonner Oils Systems Clean and Dirty Oil Storage Tanks Turbine Oil Resen oir C, flydrogen Seal Oil Unit

!! oiler Feed Pump Console Support Facility ignition Oil Tanks Turbine Oii Resen oir D. Main Generator Boiler Feed Prm; Console '

Clean and Dirty oil Storage Tanks Turbine Oil Resenoir NQlliSJ R1UlfILVli KlilliRiiNG'i; sop 29.6 l

Page 38 of 100

EXAM QUESTION DATA SilEET  !

SXElliM WM WAJATING SAIID'11M11Dh 086 A2.02 30 (M Plant Senice Splems S_ TIM Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection Sptem; and (b) based on those predictions, use proerdures to correct. control, or mitigate the consequences of those rnalfunctions or operations.

WA101'lC low I'PS header pressure RL!1iSIIDE Due a fire in the transformer yard, fire main pressure is decreasing because of heavy demand. Which of the following selections cortectly identines the sequence in u hich the standby fire pumps will automatically start to maintain Ore header pressure?

ANSWER D. Standby Prenure hiaintenance Pump Fire Main 11ooster Pump Fire Diesel Pump DISTRACTERS A. Fire Main ihxister Pumps l' ire Diesel Pump Standby Pressure Maintenance Pump II. Fire Diesel Pump Standby Pressure Maintenance Pump Fire Main !!ooster Pumps C. Fire Main thxister Pumps Standby Pressure Maintenance Pump Fire Diesel Pump SnTliS; Oll111C11E lElliRiiNEliS; sop 29 6 Page 39 of 100

i EXAM QUESTION DATA SIIEET hlSIliM K/A* N/A IRTlho SM3i1T11WCUR.N 062 K102 4.1 (6) Electrical STEM:

Knowledge of the physical connections anNor cause effect relationship between the AC distnbution system and the following systenw K/A191?lC; lid /G QUlilllOh';

Following a unit trip from $09 power, a station blackout occurs due to a fault on the 138 KV offsite power supply. Which one of the following selections correctly identifies the 480 VAC equipment that will automatically start WilEN the 480 VAC safeguards busses breakers are re energired?

NOTE: NO S1 SIGNAL. E.<lST ANSWER C. Three Essential Sersice Water Pumps 21 AND 23 Auxiliary Feed Water Pumps

21. 22. AND 23 Component Cooling Water Pumps DISTRAC11 IRS A. Two Non Essential Sersice Water Pumps Three Essential Sersice Water Pumps 21 and 23 hiotor Drisen Ausiliary Feed Water Pumpi it Two Essential Service Water Pumps 21 AND 23 Ausiliary Feed Water Pumps

21. 22. AND 23 Comp < ment Cooline Wate Pumps D. nree Essential Service Water Pumps 21 AND 23 Ausiliary Feed Water Pumps 21 AND 23 Component Cmline Water Pumps NQBil DDRiCIlYE ElilliluitiGiS; AOl 27.1 Page 40 of 100

EXAM QUESTION DATA SilEET SXSIEM Kn* hM_1MIlh0 SAEIY IUNQQN 062 A301 30 (6)lilectrical SIbM:

Ability to monitor automatic operation of the AC distribution system, including:

l KalGl'IC:

l Vital bus amperare QWSllON; When loading electrical equipment on the 180 VAC busses while recovering from a unit trip and station blackout you are directed to limit the load on Transformers 5,2,3 and 6, to less than 200 Amps. Which of the following selectioni correctly identifies the indication (parameter) that will be used to serify compliance with this direction?

ANSWi?R A. Station Ser$1cc Transformer liigh Side (6 9 KV) Ammeter (s)

DISTRACTiiRS

11. 480 VAC lius Ammeters C. 6 9 KV Station Ausiliary Ammeters D. Sum of indisidual equipment ammeters EQlliSi OlilliGXE REEEREb ES; IIS 0.1 l' ape 41 of 100 I

EXAM QUESTION DATA SilEET SXSJlliM Kaa h/A iMllh0 SAlliT111WC11DN 063 K302 3.5 (M lilectrical SIllM:

Knowledee of the effect that a low or malfunction of the DC Electrical System will have on the followine K/A 'IOPIC.;

Components usine DC control power 9UllE11Dlt Follow mg a Safety injectioa you receise indication that 125 VDC Control Power has been lost to all equipment powcred from 125 VDC bus #21. Which of the following selections describe 110W and Wily this failure WILL or WILT. NOT impact your ability to satisfy the following potential EOP requirements:

1

• RCP Ttip Criteria e SI Reduction

• Close Accumulator Isolation Yahc5 2

ANSWER

11. RCP Trip Criteria RCPs that have lost control pow er must be tripped locally SI Reduction No impact, control power automatically transferred to 23 DC !!us Accumulator holation Vahes No impact. control power supplied from individual breaker AC feed DISTRACTliRS A. RCP Trip Criteria No impact, RCPs can always be tripped from CCR 51 Reduction SI pumps must be tripped locally at 480 VAC Switchycar Accumulasar holation Valves No impact control power supplied from individual brealet AC feed C. RCP Trip Criteria RCPs that has e lost control pow er m. 5t be tripped locally St Reduction . Si pumps must be tripped kieally at 480 VAC Switchgear Accumulator Isolation Valves No impact, control power supplied from individual breater AC feed D. RCP Trip Criteria RCPs that have lost control pow er must be tripped locally St Reduction No impact, control power automatically transferred to 23 DC Ilus Accumulator holation Valves Valves that receive control power from 21 DC bus cannot be closed EDlliSJ D D E CI1Yli RiiEEREEGiSt A OI27.1.11 Page 42 of 100

EXAM QUESTION DATA SilEET SYS111M K%.# KM_BATISO sal LTY EVECil0N 064 A216 33 16) Electrical STlih1:

Ability to (a) predict the impacts of the following, malfunctions or operations on the ED/G system; and (b) based on those predictions, us,e procedures to correct. control. or mitigate the consequences of those malfunctions or operatione K%lO.11C leu of offsite power during fullload testing of EDGs R1!!iS110E With the unit at flot Shutdown (ilSD) a full load test is being perfonned on #21 Emergency Diesel Generator (EDOL #22 and 023 EDGs are operable and in AUTOMATIC. Af ter the EDO is fullloaded, a Orc in the Station Auxiliary Transfonner causes a lou of 6 9 KV power. Which of the following statements describe the espected response of the #21 EDO and 4RO VAC but $A?

ANSWER D. #21 EDG continues to run llus 5A Nonnal Feeder Breaker opens All loads except running blackout loads strip Non. running blackout loads sequence start DISTRACTERS A. #21 EDG trips and restarts when returned to AUTO

- 11us 5A Nonnal Feeder Breaker opens Allloads strip Illackout loads sequence start II. #21 EDG trips and restarts when returned to AUTO llus 5A Nonnal Feeder Breaker opens All loads except running blackout loads strip Non running blackout loads sequence start when bus is re.cnergired C. #21 EDO continues to run llus 5A Nonnal AND Emergency Feeder Breakers open Bus SA Emergency Feed 11reaker closes Illukout loa <ls sequence start N9.lliS!

DIUECIlyI; EEEliRiiSCliS; AOI27.1.1 page 43 of 100

EXAM QUESTION DATA SilEET HEIDI K%3 h/A J M 11EG SAtliT1111EclKtS (K)$ K2 01 3.0 (4) Reactor lleat Remos al SIliM:

Knowledge of bus nower supplies to the followine:

KBl011C RilR Pumps 9UliS110SJ Consider the following initial conditions when selecting your answer:

o Reactor Power IOO'7c o #21 RilR Pump 005 for last sit hours (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO action time)

While operating with the abose conditions the control room receives a report that a lubricating oil leak on the oil cooler foi #21 EDO has been discovered making #21 EDO INOPERABLE (7 day LCO action time). It is estimated that it will take 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to repair the leak, Which of the following statements correctly describes the actions that must be taken to ensure con pliance with j technical specifications?

ANSWER D. Complete repairs on #21 RilR Pump or be in llot Shutdown within seven hours.

DISTRACTI!RS A Complete repalts on both #21 EDO and #21 RilR within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> or place the plant in 1101 Shutdow n IL Complete repairs on #22EIKI within 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> or place the plant in llot Shutdown C. Verify operability of remaining safeFuards equipment and continue operation obsersing Technical Specification limits and action times fi r equipment out of service. '

MElli OluliCTIVE RElliBENClis Tech Spec. Section 3.7/3.3/3 0 Page 44 of 100

EXAM QUESTION DATA SIIEET 511IEM Ka* Ka_1LAUNO sal]i1XREcB03 NI A4.OR 30 (4) Reactor lleat Removal EIEM:

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

K!1LlOl'lC Steam dump vah es QUllTIDh3 You ha$e been directed to conduct a plant cooldown from llot Shutdown (llSD) to 350'F at 50'F/hr using the condenser steam dumps. Approximately one hour af ter positioning the steam dump valves in M ANUAl/PRf3SURI! MODI! to establish the desired cooldow n rate you notice that the cooldow n rate has decreased from 50T/hr to 20T/hr. Steam dump valve position has not been changed.

Which of the following statements correctly describes the reason that the cooldown rate has decrer. sed and the actions necessary to maintain a constant cooldown rate?

ANSWiiR A. Steam flow has decreased due to reduced steam pressure. Valves must be gradually opened as the cooldown progrenes.

DISTRACTliRS

11. 5.wm dump pressure controller will not allow Main Steam Pressure to decrease below the dial :.clting in MANUAL or AUTO. Setpoint must be gradually reduced as the cooldown progrenes C. Delta T between the steam temperature and the condenser ending water (circulating water) has decreased. Cooldow n rate ynnot be increased unten circulating water flow is increased D. Delta T between RCS temperature and feedwater temperature has decreased. I:ced flow must be increased to increase heat

$m.aal.

ND_ Ills; 01111iG1Yli RllliRliSCliS; System 1.cuon/ POP 3.3 Page 45 of 1(X)

l EXAM QUESTION DATA SilEET EY11D1 KM8 KM._IMIMO SAETELTEC110N (M5 Al 05 38 (4) Reactor lleat Remmal E1D1:

Ability to predict and/or momtor changes in parameters (to pievent exceeding design limits) associated with operating the MT/O sptem controls including:

K/A lOl'IC Espected response of primary plant parameters (temperature and pressure) following T/G trip QLES1101 Consider the followinF initial conditions w hen selecting your answer:

o Reactor Power 109 e RCS T.,, 548'F o Control Rods llank D 125 steps / MANUAL o Main Turbine Startup . approaching synchronous speed o lip Steam Dumps AUTOM ATIC/ Steam Pressure Mode While performing a plant startup an the turbine trips due to an overspeed condition. Which of the following selections is corteet regarding the cifeet that this malfunction will have on the following RCS parameters:

o RCS T.,,

o Average Lmp Delta T e Pressuriier Pressure o Pressuriier i evel ANSWER D. RCS T.,, INCREASE Average Loep Delta T DECREASE RCS Preuure INCREASE Pressuriier Level INCREASE DISTRACTERS A. RCS T,,, INCREASE Ascrape Loop Delta-T INCREASE RCS Pressure INCREASE Pressurlier Level INCREASE II. RCS T.,, DECREASE Average Loop Delta-T INCREASE RCS Pressure DECREASE Pressuriier Level DECREASE C. RCS T.,, INCREAoti Average Loop Delta-T INCREASE RCS Pressure DECREASE Pressuriter Level DECREASE M

OlUECTIV13 Bl;11REECES.;

Composite Page 46 of 100

EXAM QUESTION DATA SIIEET HElliM KM1 R/A_BATitiQ SAFETY ITNC1193 00N K401 31 (Mil'lant Serv!:e Splem SIEM:

Knowledpc of CCWS design feature (s) and/or interlock (s) which prmide for the following:

h/A 1Gl'lC Automatic start of standby pump.

QL'liS.11GN1 In Accordance with System Operating Proecdures, WilEN RCS temperature is GRiiA111R TilAN or EQUAL TO 350'F. the CCW Pump Auto Start Ley switch must be in the NORMAL position and W11EN RCS temperature is LESS TilAN 350'F, the CCW Pump Auto Start key switch is placed in the BYPASS Position.

l Which ONE of the following statements correctly describes the reason for placing the CCW Pump Auto Start key switch in the IlYPASS pmition WilEN RCS temperature is LIISS TilAN 350'F?

ANSWER D. Permit operation of the CCW system uith less than three pumps running u hen CCW is flowing through RilR llent Eschanters.

DISTRACTliRS A. Illock the CCW standby pump auto start feature to prevent water hammer when the RIIR system is in service.

11. Allow operation of the CCW system with three pumps running to meet heavy demand imposed by RilR heat load

_C. Technical Specifications allow defeating the Auto start 'cature below 350*F if three {CW pumps are OPER Allt.li.

NOIliS; OlllECILYb Rlil?EliNGS; SOr 4.1.2 Page 47 of 100 j

E EXAM QUESTION DATA SilEET tirmal MM RUaHMO sAIITn1EcJtos 4

076 K4 06 2N (4) Reactor lleat Remos al lill31:

Knowledge of the SWS design feature (s) and/or interlocks which provide for the followine:

K%20flC Senice water train separation DL'E1 HON; During power operation, the control room coordinates with the NPOs to shift the ihntial Service Water licader from the 12 3 lleader to the 4 5 6 lleader. When the necessary valving is completed the SWP Mode Control Switch on the SilF 1 panel is inadvertantly left in the 12 3 position, the service water system is operating in the lhree lleader Configuration.

Which ONii of the followint cornponents will he supplied with service water if a Safety Iniection sirnal is initiated?

ANSWiiR C. CCW llent lischanects DISTRACTI!RS A. Instrument Air Cornpressor lleat fischangers It I:.merrency Diesel Generators D. Contamment I:an Cooler Units EQHiS.:

, RIULGlyB RiiEEREEcliti; sol' 24.1 i

J Page 48 of 100

EXAM QUESTION DATA SIIEET MXilM Ka! IWLIMIlNO SAFETY I UNG10N 1(13 K4(* 3I m Conniinment Interrity GlM Knowledre of the containment splem design feature (s) and/or interhwl(s) w hich proside f l the followint' KMl9hD Containment isolation QWS110N; Which ONil of the following selections identifies ALL of the conditions w hich willinitiate a Containment Ventilation isolation

sifnal?

, ANSWl(R

!!. liigh Radiation (containment R 41.R 42) liigh radiation (plant vent, R 44)

Containment l'hase A lsolation Signal Contamment 111111 l'ressure Signal hianual Containment Spray Signal DISTRACTERS A. liigh Radiation (containment, R-41. R-42)

Containment isolation l'hase il Containment l'hase A isolation Signal Containment lli ill l>ressure Signal hianual Containment Spray Signal C, liigh Radiation (containment, R 41. R 42) liigh radiation (plant sent, R-44)

Containment l'hase A l!.olation Signal Containment 111111 Pressure Signal Station illackout D. liigh Radiation (containment, R-41. R-42) liigh radiation (plant vent, R 44)

Containment l'hase A lsolation Signal liigh Radiation (containment IM R 7) hianual Containment Spray Signal hiQlli Rlulfi1YU KliDIRiitiWS; SOP 11.1/11.2 Page 49 of 1(X)

__ _. . ~ _ _ . _ _

l EXAM QUESTION DATA SilEET

~

SXElliM KIM KM_lMTING SAUil111'NGON 103 K 3.02 38 (5) Containment interiity EID1:

Knowledge of the eficci that a lou or malfunction of the containment system will have on the following:

i KlAlDl'lC Lou of containment interrity during normal operation.

QMliSllQ1

Which ON!! of the following conditions would be comidered a loss of containment integrity during normal operation?

j ANSWliR j D. We;d channel seal to the equipment hatch is lost.

j DISTRACTliRS l A. An automatic containment isolation valve is found to be inoperable in the Cl.OSliD position.

11. Weld channel tone pressure indicates 50 psir.

4 C. Personnel hatch inner door indicates OPEN. outer door indicates Cl.OSl!D.

, NOTES:

DilECTIVE KliaiRENCES:

AOI 10 6 2 4

l t

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Page 50 of 100

l l

EXAM QUESTION DATA SilEET

.inILM Raa K%_MTINO sal m u m 31us 005 Al 02 33 (4) Reactor llent Renun al S.lliM:

Atiility to predict anNor monitor changes in parametern (to prevent eseceding design limits) auoelated v ith operating the RilRS controls including:

~ 1 K@ TOPIC; RilR flow rate QIT31192 In accordance with proerdure Residualllent Removal (RilR1 System flow rate should be AllEAST 2500 GPM per RilR Pump /RilR lleat Exchanger, Which ONE of the following statements is correct regarding the reason for thh limitation?

ANSWiiR A. Minimite turbulence downstream of the associated RilR llent lischanger and llCV-63N and llCV MO.

DISTRACTliRS II. I' resent the RllR pumps from reaching runout conditions C. l> resent s ortening at RCS Ilot 1.cp loop connection D. linsure RilR flow is evenly distributed via the S1 manifold to all RCS cold lers-N01118 91tlLCTlXU HElliBiliGE Sol' 411 i

l PaFe 51 of 100

EXAM QUlbTION 1)ATA SilEET SYElliM IVA # 12AlATING SAllii1HUNC110N (KXWIM lik 1.02 4.1 Station lilac Lout Sll}\:

Knowledge of the operationalimplications of the followiny concepts as they apply to the Station lilackout:

IVA.1011C Natural circulation coolint

)

Q(TdllGN.; I Which ON!! of the following indications is NOT an indication used in the litnergency Operating l'rocedures to verify that Natural Circulation Cooling flow has been established? l ANSWiiR 14 Stable or decreasing hop lietta T I)lS1RACT11RS __

A. 50 Pressure . stable or decreasing C. Core litit Ternperatures stable or decreasing

(). RCS Cold I cp Temperatures . at saturation ternperature for 50 prenure NLUlii DJuliC11Eli KliERiiNClii II-0 l' age 52 of 100

8 GAM QUESTION DATA SIIEET SLsitM RM MA._HKUNU~ SAlllHUNCllON (WWW174 EK 3 04 39 Inadequate Core Coolint Ell;M:

, Knowledge of the teamns for the following responses as they apply to inadequate Core Cooliny: I WEID11C l Trippint the RCPs QVLSilOM When using Einergency Operating Proerdure(EOP) 111, Loss of Reactor or Secondary Coolant, you are directed to operse Reactor Coolant Purnps (RCP)in accordance with EOP 1%C.2, Response to Degraded Core Cooling, rather than tripping the RCPS as directed by EOP E 1,if the RYLIS Dynarnie Range indication indicates less than the value obtained forin the following table:

449 .

4 RCPs 1/S 309 - 3 RCPs 1/S 209 -

2 RCPs 1/S 139 .

1 RCP l/S Which of the followint staternents is cortret retarding the reason for the ruldance described above?

ANSWER ll. Trippint RCPs under these conditions could lead to core uncoverint and an inadequate Core Cooliny Condition.

DISTRACTI:RS

A. RCI's operating under these conditions rnay seite when tripped preventing their res, tart in future recmcry actions.

C. Tripping RCPs under these conditions would result in an increase of rnass flow itorn the break due to phase separation of the fluid D. Tripping RCPs under hese condit6ons will result in a loa of RCS picssure control due to the loss of pleksuriier spray capability.

NR11E DillliCHEli IGLIRiiNCE1; E lllack round f

4 Page 53 of 100

EXAM QUESTION DATA SilEET SXsn3J K/M h/A J a n SO n lu cles (KKKWK AK 3 04 34 i perable/ Stuck Control Rtxi S11A1:

Knowledge of the reasons for the followine responses as they apply to the Inoperable / Stuck Control Ral:

K/AlDl'lC Tech spec litnits for inoperable rais QVldllONJ Which ONit of the following conditions is NOT a condition u hich would required a control rmi to be declared INOPER Allt.li9 ANSWER D. Control itrxl I 3 in Control llank A. Rrxlllottom til6 table fails to initiate a turhine runback.

DISTRAC11LRS A Control Rod 11.H in Control llank D is misaligned from the bank by .lM steps

11. Control R<wl 11 14 in Shutdown Ilank C is mechanically bound at 223 steps and will not trip C. Control Rod D-4 in Shutdow n llank A f alls to incet rmi drop time requirement of 2 4 seconds.

NOILS; OluliclYli EllliRIKlit Tech Spec. 310 l'aFe 54 of 100

EXAM QUESTION DATA SilEET SYS11LM R%# K/A RATING sat]illll'NC110E 00015 A A 1.22 4.0 RCP Malfunctions hllahl:

Ability to operate and/or monitor the followine as thes appij to the Reactor Coolant Pump Malfunctions (Imu of RC Flow r KMlQl'lC RCP seat f ailure malfunction R11S110b2 Consider the follov .'np indications when selecting your answer:

e Reactor Power 509 e RCS T.,, $$3*F e #23 RCP #2 Seal Standpipe Low Level Alarm actuated e #23 RCP Sealinjection flow C 0 GPM o #23 RCP #I Seal LeakofIflow 3.0 GPM l While operating with the abose conditions and indications the Reactor Operator reports that the RCS leakage calculation is normal howeser there has been an increase in total leakage into containment as evidenced by an increase in the pumping frequency of the containment sump.

Which ON!!of the followine failures or malfunctions would support Al.l.of the above indications and conditions?

ANSWl:R lt I:ailure of the #23 RCP #3 Seal

_DISTRACTERS A I ailure of the #23 RCP #2 Seal C. I ailure of the #23 RCP #1 Seal 11 17ailure of the #23 RCP Seal PacLace (#1.#2. and #3 Seah bX!llii Rl!IliCI1XIi EtilliRiiKliE A OI 1.3 Page 55 of 100

EXAM QUESTION DATA SilEET

~

D11111 KM * &%_K61]M) SAlliTY FUNCTION (XX124 AK 3.01 4.1 Emergency (rapidilloration Ellll:

Knowledge of the reasons for the tollowing responses as the apply to the Emergency lloration:

K%1Dl'1C When emergency boration,ls required QWiUlDNJ While the teactor is in the Refueling Condition the Reactor Operator identifies an unexplained increase in Source Range count rate and a steady positise 0.15 DPM Source Range Startup Rate indication. The Senior Reactor Operator directs the Reactor Operator to initiate boration of the RCS per A3.4, Uncontrolled Reactivity Addition.

Which one of the following boration flow paths and methods is the preferred method for completing this task in accordance with A3 4. Uncontrolled Reactisity Addition?

ANSWER C. Normal boration flowpath at musimum rate to charging pump suction DISTRACTERS A. RWST sia LCV i1211. Emergency RWST Makeup Stop l_ 11. MOV.333. Emergency lloration Stop to charging pump suction l D. Norma' boration flowpath at masimum rate to Volume Control Tank N111ESJ RRiliCIlYE RiiEURENCll't A 3.4 I

Page 56 of 100

EXAM QUESTION DATA SilEET 5YhlhM K%f h!A RA11NO SAILTY EUNC110N (XK10.'6 AK301 40 Inu of CCW ElLM:

Knowledge of the reawns for the following as they apply to the lou of Component Cooling Water:

K!A TODC Guidance contained in 1:0l% for loss of CCW/ nuclear service water RUllE110ft Which ONiiof the f iwing 4 will NOT result in the automatic start of a Component Cooling Water Pump?

ANSWiiR A. A lou of ofIsite electrical power is followed by Unit Trip AND Safety Iniection.

DISTRACTliRS

11. CCW llender Pressure decreases to 60 psig with two CCW pumps running.

C. An inads ettent Safety injection Signal is actuated due to an instrument failure, The 480 VAC busses are ALL energlied from offsite electrical power.

D. A Station Blackout Signalis actuated due to a loss of offsite power following s Unit Trip. No Safety injection signal is present.

NOTiiS:

OlHITTIVE BELERiiNG1; fiS 1.1 llackground l

Page $7 of 100

EXAM QUESTION DATA SilEET snTD1 WM Ra_RKnMa MuimVECn0N (xWo27 AA2 04 3.7 Prenutiier Prenure Control Malfunction SIliM:

Ability to determine and interpret the followine as they apply to the Prenuriier Preuure Control Malfunction-Wh10DC:

Tech Spee limits for RCS Prenure WESl10M Consider the following event when selecting your answer:

e The controlling pressurlier pressure channel (CilAh'ED,_j) has failed high with the unit at 1004 power o RCS pressure is stabilired at 2115 psig by manually closing the pressuriier spray valves l

After the plant is stabiliied the Senior Reactor Operator reviews the following Technical Specification Limit:

Rtanutfeelant System PtsstrJrinettaturc e anulow iter The folhwing DNIl related parameters pertain to four loop steady state operation at power leseth Freater than 98% of full rated power:

'e Reactor Coolant System T,,, 5: 587.2 'F o Preuuriser Pressure > 2190 psia e Reactor Coolant System Total Flow Re 'r 431.840 GPM ltem (b), pressurlier pressure,is not applicable durug either a thermal power change in e* cess of 59 of rated thermal power per minute, or a thermal power step change of 10% of rated thermal power, Under the applicable operating conditions, should reactor coolant temperature T.,,, or pressuriier pressure exceed the values gisen in items (a) and (b) the parameter shall be restored to its applicable range within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Which of the following statements correctly describes the appropriate response, if any, the Senior Reactor Operator should take to ensure compliance with this technical specification?

ANSWER D. Restore RCS preuure to GREATER TH AN 2190 psia within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce power to 5 989.

DISTRACTERS A. Specification for Pressuriier Pressure is not applicable because the transient was induced by an instrument Iailure.

11. Specification for Pressuriier Pressure is not applicable because there was no chance in reactor power.

C. Specification is not applicable because it only applies when power is 2 98% of BATED FULL POWER which is equivalent to 989 of the maximum attainable power les el of 108% (llich Flus Trip Setpoint)

NOTES:

011JECTIVE B1HiBEECliS; Tech. Spec. Section 3.1 Page 58 of 100

EXAM QUESTION DATA SIIEET

~

D1111M KIM KutJmIlNo hM1inmenoh' (XXO40 A A2 01 42 Steam Line Rupture ElliM:

Ability to determine and interpret the followine as they apply to the Steam Line Rupture:

K/AIQ11C Decurrence and hiention from a steam line rupture from pressure and now indication.  !

QL!! MIRE l A main steam line rupture has occurred with the plant at hot shutdown resulting in a hiain Steam Line AP Safety injection Signal. After analyting the following indications determine which ON!!of the following selections correctly identifies the portion of the hlain Steam System that has ruptured?

o 21 SG Prepute 780 pslF(decreasing slowly) o 22 SG Prenure 782 psig (decreasing slowly)

• 23 SG Pressure 340 psig (decreasing rapidly) e 24 50 pressure 775 psig (decreasing slowly) o ALL hiain Steam Line flow Indications 0 lbm/hr o ALL htSIVs indicate OPIIN ANSWlR

11. 23 SG hiain Steam 1.ine between SG and Ilow filement DISTRACTliRS l A. 23 SG h1ain Steam 1.ine Upstream of MSIVs outside contamment C. 23 SG Main Steam Line Downstream of MSIVs D. 23 SG Main Steam I ine between illow Illement and Containment Penetration N911is OlHIETIEli BElliBliNC1;1i:

System P&lD 4

d Page 59 of 100

EXAM QUESTION DATA SilEET SMi1EM MM WA RAllND SAElill.EUEllON (KKOSI AA202 3.9 las of Vacuum EIEM:

Abihty to determine or interpret the following as they apply to the less of Condenser Vacuum-Wh191'IC Conditior's requiring reactor and/or turbine trip RUliSIIDS Consider the followinF esent when selecting your answer:

o Reactor Power 15 %

• T.., 547'F e Turbine Startup in progren Turbine at synchronous speed e llP Steam Dumps Preuure blode AUTOhtATIC During plant startup with the above conditions, a failure of the exhaust boot (expansion joint) between #21 Low Prenure Turbine and the condenser results in a rapid lou of condenser s acuum to atmospheric prenure user a pernid of 5 minutes.

Anuming that all plant protection AND control systems function as designed, which ONE of the following statements correctly describes the resp (mse (INCREASE /DECREASIVNO CilANGE) of the following parameters as_3usiult of this event e Reactor Power e Turbine RPhi e Steam Flow

• T,,,

ANSWER C. Reactor Power DECREASE Turbine RPA1 DECREASE Steam Flow DECREASE T.,. INCREASE DIS'IRACTERS A. Reactor Power DECREASE Turbine RPhi DECREASE Steam Flow NO CilANGE T ,, NO CilANGE

13. Reactor Power DECREASE Turbine RPhi DECREASE Steam flow INCREASE T,, , INCREASE D. Reactor Power NO CilANGE Turbine RPhi DECREASE Steam Flow NO CilANGE T.,, NO CilANGE NDILSJ DillfiCIlYF, EEFERENCES; ARP FAF Page 60 of 100

EXAM QUESTION DATA SilEET SYT M E/A.! K/A RAllNQ SAFETY FUNCTION CIXX)$7 AA2.19 4.0 less of instrument Ilus SlliM; Ability to determine and interpret the following as they apply to the loss of vital AC instrument bus:

N/A TOPIC The plant automatic actions that will occur on the loss of ac vitalinstrument buues.

QUESTION:

During power operation at 1009 reactor power a loss of electrical power to #21 Instrument Ilus occurs caasing failure of PT-412A. Which ONE of the following statements describes the response of the High Pressure Steam Dump System?

Auume that no eqiipment or instrumentation was out of service before the failure.

ANSWER D. T,,, will fail to ! 47'F. High Pressure Steam Dumps will OPEN IE loss of load interlock trips.

DISTRACTERS A less of load interlock will trip anning the steam dumps.

11. T.< will fail to 547'F causine the Hich Pressure Steam Dumps to OPEN.

C. Steam dump actuation will be inhibited due to low of power to loss of load interlock.

NRTliS; OBJECTIVIi REIT.RENCES.;

A OI27.1.6 Page 61 of 100

EXAM QUESTION DATA SilEET l

SYST1BJ N M/ K/A RATING SA E]TI W CI1QS (XKWM7 AK l .01 2.9 Fire on site _,

S11511:

Knowledge of the operationalimplications of the following concepts as they apply to Fire on Site-K/A TO(1C Fire classification by type.

OlfESTION:

The Conventional NPO has reported a small electri;al fir; on the 5' elevalion of the Turbine Building. The following fire fighting equipment is available to the Fire Brigade. S.ieet the equipment which is most suitable for extinguishing a fire of this type?

ANSWER A. Portable CO., fire extinguisher DISTRACTERS IL High pressure water hose C. Water stream portable fire extinguisher D. Dry chemical fire estinguisher N9 bis!

OBJECTIVE REFERENCES; Safety Training Page 62 of 100

EXAM QUESTION DATA SIIEET S1SIliM Sfft! K/A RATIEQ SAFETY FUNCTION (X10068 AK3.12 4.1 CCR Evacuation ElliM Knowledge of the following responses as they apply to the Control Room Evacuation:

K/A TOPlQ Required sequence of actions for emergency evacuation of the control room.

QL!IiSIlOIi; A fire in the control building requires that the Central Control Room be evacuated. You have been designated as the First RO, and the SRO has directed you to trip the Reactor locally. Which of the following selection list in ORDER the locations and equipment from which you would accomplish this task, assuming that you are unsuccessfu after each attempt.

ANSWER C. Cable Spreadmg Room - Reactor Trip Breakers Cable Spreading Room Rod Drive MG Set Breakers 480V Switchgear Room - Rod Drive h10 Set Breakers 480V Switchgear koom - Bus 2A and 6A Supply Breakers 6.9 KV Switchgear Station Service Transformer 2 and 6 Supply Breakers DISTRACTERS A. Cable Spreading Room Reactor Trip Breakers Cable Spreading Room Rod Drive htG Set Breakers 480V Switchgear Room Rod Drive htG Set Breakers 480V Switchgear Room - Bus 3A and 5A Supply Breakers 6.9 KV Switchgear - Station Service Transformer 3 and 5 Supply Breakers B. 480V Switchgear Room Rod Dnve h1G Set Breakers 480V Switchgear Room Bus 2A and 6A Supply Breakers Cable Spreading Room Reactor Trip Breakers Cable Spreading Room Rod Drive htG Set Breakers 6.9 KV Switchgear - Station Service Transformer 2 and 6 Supply Breakers D. Cable Spreading Rooin - Reactor Trip Breakers Cable Spreading Room - Rod Control System Power Cabinets 480V Switchgear Room - Rod Drive htG Set Breakers 480V Switchgear Room Bus 2A and 6A Supply Breakers 6.9 KV Switchgear - Station Service Transformer 2 and 6 Supply Breakers NOTES:

OBJECTIVE REFERENCliS; A OI27.l.9 Page 63 of 100

E';AM QUESTION DATA SIIEET SYSTEM K/A # K/A iMTING S AFETY FUNCMG 0(XXM9 A A1.01 35 Loss of Containment Integrity SIliM:

Ability to operate anti /or monitor the following as they apply to the Loss of Containment Integrity:

K/A TOPIC:

Isolation valves. dampers. and electropneumatic devices.

011STION; While verifying Containment Isoaltion valves are in the correct position following a Safety In,iec ion actuation due to Ili ili Containment Pressure, you notice the following "Two is True" indication MOV 222, RCP Seal Leakoff Containment isolation Valve:

• LeIt side oflight illuminated - AMBER light e Right side oflight Extinguished Which ONE of the following selections is correct regarding the expected position of MOV-222, AND the indicated position of MOV-222 with respect to the "Two is True" indicating lights?

ANSWER D. Expected Position CLOSED Indicated Position OPEN DISTRAt:1ERS A. Expected Position OPEN Indicated Position OPEN B. Expected Position CLOSED l Indicated Position CLOSED C. Expected Position OPEN Indicated Position CLOSED bi_O.IEL OBJECTIVE

REFERENCES:

_ _E-0 t

Page 64 of 100

l EXAM QUESTION DATA SilEET SYSTEM h/A! K/A RATING SAFETY FUNCTION 000076 AK3.05 2.9 liigh RCS Activity S_IliM:

Knowledge of the reasons for the following responses as they apply to liigh Reactor Coolant Activity:

N/A TOPIC:

Corrective actions as a result of high fission product radioactivity level in the RCS.

OUESTION:

Which of the following statements identifies the conditions where Technical Specifications requires that RCS activity (for nuclides other than tritium with half lives of more than 30 minutes) be LESS TilAN 60/E har ci/ce?

ANSWER B. When reactor is critical or RCS temperature is GREATER THAN 500 F.

DISTRACTERS A. When there is fuel in the Reactor Vessel or RCS temperature is GREATER TilAN 200*F.

C. When the reactor is critical or RCS temperature is GREATER TilAN 350"I D. When the reactor is in " Power Operation Condition" or RCS temperature is GREATER TilAN 350*F.

NOTES:

Ql})ECTIVE

REFERENCES:

Tech Spec. Section 3.0 Page 65 of 100

EXAM QUESTION DATA SilEET SYSTEM h!/U }{/A RAT [NG SAFETY FUNCT[QN (XXXX)5 AK3.01 4.0 Inoperable / Stuck Rod EllM:

Knowledge of the reasons for the following responses r.s they apply to the Inoperable / Stuck Control Rod:

K/A TOPIC:

Iloration and emergency boration in the event of a stuck md during trip or normal operation.

QUESTION; Following a reactor trip you are directed by ES-0.I, Reactor Trip Response, to verify that all control nds are fully inserted.

Which ONE of the following rod position indications would meet the criteria for a control rod NOT being fully inseced?

ANSWER C. Proteus Computer rod position 13 steps DISTRACTERS A. Individual IRPI reading 5.5 inches B. Proteus Computer rod position 7 steps D. 0.05 Volts on Digital Volt Meter (DVM)

NOTES; OBJECTIVE BEFERENCES:

AOI 3.4/ES 0.1 Page 66 of 100

EXAM QUESTION DATA SIIEET

, SYSTE11 K/A # K/A RAllNQ SAFETY FUNCIJON _

I 000015 AK3 07 4.1 ' RCP Failure i STEh1 Knowledge of the reasons for the following responses as the apply to the Reactor Coolant Pump Malfunctions:

4 K/AIQPLC i Ensuring that SG levels are controlled properly for natural circulation cooldown.

. QUESTIOE; i While conducting a natural circulation cooldown using ES-0.2, Natural Circulation Cooldown, you are directed to 6ontrol SG levels. Which ONE of the following actions if performed could temporarily impede or reduce natural circulation now?

ANSWER B. Ausiliarv feed flow to #21 SG is rapidly increased from 50 GPM to 200 GPM DISTRAC IERS s

A. Steam generator #21 level is allowed to slowly increase to 559 as seen on the narrow range level indicator.

C. Steam generator #21 level is allowed to slowly decrease to 359 as seen on t' : narrow range level indicator.

, D. Ausiliary feed How to #21 SG is rapidly decreased from 200 GPM to 50 GPM NOTES:

i OBJECTIVE B.FS!RENCES;

ES 01 Background i

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EXAM QUESTION DATA SilEET SY1TliB K/A # R/A RATING li AFETY lilNCTION 000007 EK101 4.0 Reactor Trip STliM:

Knowledge of the reasons for the following as they apply to a reactor trip:

., K/A TOPIC:

Actions contained in EOP for reactor trip.

QUESTION:

While performing the immediate actions for a Reactor Trip using EOP E-0, Reactor Ttip or Safety injection, you are directed to de-energite 480VAC busses 2A and 6A to trip the reactor since a reactor trip cannot be verified by available indication. After 480 VAC bus 2A and 6A are re-energized you are directed to depress the Blackout Relay Reset 480V push button on the SC panell[ the Main Generaa,r Output Breakers are CLOSED, Which ONE of the following statements is correct regarding the reason that the Blackout Relay must be reset at this time?

ANSWER C. Resetting the Blackout Relay at this time removes the 480 VAC Bus 6A Undervoltage Signal from the blackout logic and E the Main Generator Output Breakers are CLOSED (86P and 86BU relays reset) a station blackout signal will be avoided.

DISTRACTERS A. Resetting the Blackout Relay at this time removes the 480 VAC Bus 2A Undervoltage Signal from the blackout logic and E the Main Generator Output Breakers are CLOSED (86P and 86BU relays reset) a station blackout signal will be avoided.

B. Resetting the I;lackout Relay at this time removes the 480 VAC Bus 2A AND 480 VAC Bus 6A Undervoltage Signals from the blackout logic and E the Main Generator Output Breakers are CLOSED (86P and 86BU relays reset) a station blackout signal will be avoided.

D. Main Generator Output Breakers will should already be OPEN. Depressing the Blackout R< lay Reset 480V push button will trip the 86P and 8t>BU relays - ausing the Main Generator Outpui Breakers to OPEN.

NOTES.

OBJECTIVE REFERENCES E.0 l

Page 68 of 100 1

4

EXAM QUESTION DATA SIIEET j SYS7B1 h/A! K/A RATING S AFETY FUNCTION 0000(N EK3.04 4.1 SBLOCA Sll2d:

Knowledge of the reasons for the following as they apply to the small break LOCA:

K/AT.DPIC:

, Starting additional charging pump OUESTIOli; l Consider the following event when selecting your answer:

• A small break LOCA has occurred and a Safety injection (SI) Signal has been actuated.

o All Si equipment has operated as designed.

• EOP ES-1.2. Post LOCA Cooldown and Depressurization has been implemented.

l While performing the actions required by EOP ES I.2, you are directed to establish maximum charging flow to the RCS. Which ONE of the following statements is regarding the reason for this action when performing a Post LOCA Cooldown and Depressurization?

ANSWER D. hiasimum charging flow is established in order to provide sufficient makeup so that SI pumps can be readily reduced during the Si Reduction sequence.

DISTRACTLRS A. hiaximum charging flow is established in order to provide maximum auxiliary spray flow capability in the event that Reactor Coolant Pumps are not running and normal spray is unavailable.

B. hiaximum charging flow is established to ensure that maximum boration capability exists.

C. hiaximum charging flow is establishei m attempt to achieve Si Termination Criteria, thus avoiding the tedious task of St Reduction.

NOTES:

OBJECTIVE

REFERENCES:

ES 1.2 Background i

Page 69 of 100

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EXAM QUESTION DATA SIIEET  ;

SYSTEh! }2Af K/A RAllNQ SAFETY FUNCTION 3

000011 EA2.01 4.5 LBLOCA

S_ TEA 1:

Ability to determine or interpret the following as they apply to a LBLOCA:

K/A TOPIC:

Verification of adequate core cooling.

i QLTSTION:

j Consider the following conditions when selecting your answer:

I o Large Break LOCA has occurred o Core Exit Thermocouple Temperatures 1210 F

• RVLIS Full Range Indication 30%

e RCS Subcooling 0*F l

! The Watch Engineer reports that the above conditions require entry into the Functional Restoration Procedures.

Which ONE of the following Functional Restoration Procedures must be implemented?

ANSWER t

A. FR-C.I. Response inadequate Core Cooling DISTRACTERS B. FR-C.2. Response to Degraded Core Cooling C. FR-H.l. Response to Loss of Secondary Heat Sink D. FR-1.3. Response to Voids in the Reactor Vessel NOTES:

QDJECTIVE

REFERENCES:

CFS Status Trees F-0 4

i I

)

i j Page 70 of 100 i

EXAM QUESTION DATA SIIEET SYSTEM ]2A # K/A_ RATING S_AFETY FUNCTION 000029 EK3.06 4.2 N!ws SIliM:

Knowledge of the reasons for the following responses as they apply to the ATWS:

K1A TOPIC.;

Reason for tripping the turbine (methods).

OUES110N.;

Following a Manual Reactor Trip you are unat>te to verify that the Reactor is tripped in accordance with E-0, Reactor Trip or Safety Injection, and FR-S.1, Response to Nuclear Power Generation /ATWS, is implemented.

4 Which ONE of the following actions is NOT a method directed to be used in FR S.1, Response to Nuclear Power Generation /ATWS. for tripping the Main Turbine if the turbine trip cannot be verified?

ANSWER C. Trip the turbine locally at the " Front Standard"

} DISTRACTERS A. Manually trip the turbine from the CCR.

B. CLOSE the MSIVs from the CCR.

!, D. Manually runback the turbine in the CCR and locallv CLOSE the MSIVs NOTES:

4 QBJECTI\T, REFERENCE;S FR S.I l

t a

i 1

Page 71 of 100

EXAM QUESTION DATA SIIEET 4

SiSTEM K/A # K/A RATING SAF'ETY 'TJNCTIOS 000038 EA 1.36 4.3 SGTR STEM:

Ability to operate and monitor the followug as they apply to a SGTR:

K/A TOPIC:

Cooldown of RCS to a specified temperature.

OUESTION:

Consider the following events when selecting your answer: -

o A tube rupture has occurred in #22 Steam Generator (SG) o A Safety injection (SI) signal has been actuated o EOP E 3, Steam Generatar Tube Rupture has been implemented

• #22 SG has been isolated e The SI signal has been reset The SRO has dirce'ed you to dump steam from the intact SGs at the maximum rate to establish a Core Exit Temperature of 488'F AND then stop the cooldown.

Which ONE of the following statements correctly describes the reason for reducing RCS temperature to this value?

ANSWER B. Estr.blish sufficient subcooling in the RCS so that the RCS will remain subcooled after pressure is decreased to #22 SG pressure.

DISTRACI LRS j A. Reduce RCS pressure by causing an outsurge from the pressurizer to minimi e leakage into the #22 SG.

C. Establish sufficient subcooling in the RCS so that the Reactor Coolant Pumps will not have to be tripped when the RCS l

pressure is decreased to #22 SG pressure.

D. Reduce temperature of RCS fluid leaking into #22SG to reduce #22SG preure to minimize potential of radioactive release through the atmospheric steam dump valve.

NOTES:

OBJECTIVE

REFERENCES:

E-3 Background I

l I

Page 72 of 100 i

EXAM QUESTION DATA SilEET SlSJhM E/M K/A RATitiQ S AFFTY FUNCTION Of0001 AK 1.17 3.4 Continuous Rod Withdrawal F_ TEM:

Knowledge of the operational implications of the following concepts as the apply to Continuous Rod Withdrawal:

K/A TOPlC; MTC OUESTION:

Consider the following event when selecting your answer:

e Reactor Power 809

• RCS T. , 557'F e Control Rods Control Bank D 195 steps / Manual e RCS Boron Concentration 980 ppm Beginning Of Life (BOL)

After withdrawing control rods to adjust Tm you note that when you release the In-Hold-Out switch Control Bank D rods continue to withdraw for an additional 20 steps. As a result Tm increases to 5'F above program.

Which ONE of the following statements is TRUE regarding this event?

ANSWER D. IF the same event occurred at EOL the INCREASE in flCS T.,. wot fl ave been LOWER.

a DISTRACTERS A. IF the same event occurred at EOL the INCREASE in RCS T,,, would have been GREATER.

B. IF the same event occurred at EOL the INCREASE in RCS T., would have been the SAME.

C. IF the same event occurred at EOL there would have been NO INCREASFiin RCS T.,,.

, NOTri i

OBJECTIV_F, BEH.RENCES:

Graphs Book

]

2 a

4 Page 73 of 100

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1 l

EXAM QUESTION DATA SIIEET i SYSTEM K/A # }UA RATING SAFETY FUNCTION d

00CX)03 AA2.02 2.7 4

Dropped Rod STEM' Ability to determine and interpret the followinc:

NA TOPIC:

Sirnal inputs to the rod control system.

! OUES110N:

Which ONE of the following AUTOMATIC actions will NOT occur if a single control rod drops from the full withdrawn

position to the fully inserted position while the reactor is at 100% power?

ANSWER

! B.

APPROACHING ROD INSERTION LIMIT 12.5" and the ROD INSER'llON I.lMIT 0" alarms actuate on Panel SA.

DISTRACI t.KS

^

A. NIS Dropped Rod Rod Stop j C. Turbine Runhack D. Rod Bottom Rod Stop NOTES:

i OBJECTIVE

REFERENCES:

AOI 16.1.1 i

i i

1 s

1 i

a

.i 1

4.

J I

Page 74 of 100

EXAM QUESTION DATA SIIEET Si STEM K/A # K/A RATING SABiTY IPNCrlON 000008 AA2.03 3.9 Vapor Space LOCA E M:

Knowledge of the reasons for the following responses as they apply to the Pressurlier Vapor Space Accident:

K/A TOPIC RCS pressure / temperature indicators / alarms (PORV tailpipe temperature.

QUESTION:

Consider the following event when selecting your answer:

• A Pressuriier Power Operated Relief Valve (PORV) has OPENED and failed to close.

• A Safety injection Signal has been actuated.

o RCS Pressure stable 1480 psig

• Pressuriier Relief Tank pressure 30 psig e Pressuriier relicf tank temperature 140*F e Pressurizer Level 909 e Pressuriier Relief Tank Level 829 Using the indications and conditions provided determine w hich ONE of the following temperatures would be indicated on the PORV Downstream Temperature Indicator?

ANSWER B. 275'F DISTRACTERS A. 593 F C. 250'F D. 140"F NOTJiS; OIUECTIVE REFERENCES; industry Event / Steam Tables Page 75 of 100

=_. . _. .- .-. .

EXAM QUESTION DATA SIIEET j SYSTEh] h/A.] K/A RATING SABiTY IUNCTION

, OfK)022 AK3.02 3.5 loss of RCS makeup SlhM:

Knowledge of the reasons for the following responses as they apply to the Lo : of Reactor Coolant Pump hiateup:

N/A TOPIC:

Actions contained in SOPS and EOPs for RCPs. loss of makeup. loss of charging and abnormal charging.

QWiSilON:

Abnormal Operating Instruction ( AOl) 3.1, Chemical Volume Control System (CVCS) hjalfunctions, states that when preparing to start a Charging Pump, the Charging Pump controller must be placed in manual AND set for approximately 209 before the pump is started?

Which ONE of statements is correct regarding the reason for setting the a ' roller to 209 before starting the Charging Pump?

1 ANSWER A. Ensure that a low bearing oil pressure trip does not occur during the charging pump start.

~

! DISTRACTERS B. hiinimite starting current on the charging pump motor.

C. Balance hi ANUAL signal with AUTO signal before the pump is started j D. Provided a minimum of M GPh1 seal injection now to each RCP as soon as the pump is started.

NOTES:

4 OlHECTIVE I BE[IEECM l AOI 3.1 l

1 J

J Page 76 of 100

EXAM QUESTION DATA SilEET SYSlliM h/ft# K/A RATING LAFETY ITJNCTION 000025 A A 1.02 38 Los of RilR ElliM:

Ability to operate and/or monitor the following as the apply to the Loss of Residual lleat Removal System:

K/ATOplC; RCS Inventory QUEST 10M; When operating the Residual liest Removal System (RllR) with the Reactor Coolant System (RCS) at reduced inventory care must be taken to control RCS water level such that the RHR pumps do not cavitate or become airbound. SOP 4.2.1, RiiR System Operation, and AO! 4.2.1, Loss of Residual Heat Removal System, impose restrictions on minimum RCS water level based on certain conditions / parameters.

Which ONE of the following parameters / conditions is NOT a factor in determining the minimum allowable RCS water level?

ANSWER D. Which RHR Pump is Running DISTRACTERS A. Position of RHR hiini-Flow Test Line Stop Valves (h10V 743/lN70)

B. RilR S3 stem flow Rate C. Position of RHR hiini-Flow Test Line Bypass Stop Valve (1819)

NOTES:

OBJECTIVE BEIT.RENCES:

SOP 4.2.1/AOI 4.2.1 Page 77 of 100

. EXAM QUESTION DATA SHEET SYSTEM K/A # }2AJATING SAFETY FUNCTION 000032 AK301 3.2 Loss of Source Range NIS Sllihl:

Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation:

K/A TOPIC:

Startup termination on source range loss.

QUESTIOli; Consider the following indications w hen selecting your answer:

e Reactor Startup in progress e Source Range N31 Count Rate 2X10' cps e Source Range N32 Count Rate 4X10' cps e Intermediate Range N35 Current <l X 10-" amps e Intermediate Range N36 Current <1X10 " amps o Source Range N31 Startup Rate 0.5 dpm e Source Range N32 Startup Rate 0.1 dpm e Control Rods Control Bank D .100 steps / MANUAL

Using ONLY the information provided determine which ONE of the following actions and associated reason is appropriate

, regarding the continuation of the reactor startup?

ANSWER D. Nuclear instrumentation is NOT indicating as anticipated. The approach to criticality SHALL be stopped AND no actions SH ALL be taken which could add positise reactivity until the discrepancy is resolved.

DISTRACTERS A. Source Range N32 is reading high due to a failure in the Pulse Height Discrimination circuitry and should be considered inoperable. "The startup may continue without further action.

B. Source Range N31 is reading low due to a failure in the Pulse Height Discrimination circuitry and should be considered inoperable. The startup may continue without further action.

C. Intermediate Range N35 AND N36 are not responding. Startup may continue as long as neutron Dux remains in the source range.

NOTES:

OBJECTIVF, BEEERENCES:

POP 1.2 (P. 6 of 12)

Page 78 of 100

EXAM QUESTION DATA SilEET SYSTEM K/M K/A RATING SAFETY FUNCTION 000033 AK3.02 3.6 loss of IR NIS STEM Knowledge of the reasons for the following responses as they apply to the Loss ofIntermediate Range Nuclear Instrumentation:

K/A TOPIC:

Guidance contained in EOP for loss of intennediate range instrumentation (failure of P-6).

OUESTION:

A reactor trip has occurred. Approximately 30 minutes after the reactor has tripped the Reactor Operator is perfonning actions directed by ES-0.I, Reactor Trip Response, when he notes the following indications:

• Intermediate Range N35 IX10

• amps (stable)

• Intermediate Range N36 IX10~" amps (stable)

• Source Range N31 0 cps (stable) e Source Range N32 0 eps (stable) e SOURCE RANGE LOSS OF DETECTOR VOLTAGE annunciator illuminated Which ONE of the following statements correctly describes the response, if any, that the Reactor Operator should take regardi: g these indications?

< ANSWER A. Manually re-energize the Source Range NIS by depressing the Train A and Train B Intermediate Range Permissive Override push buttons.

DISTRACTERS B. Manually re-energize the Source Range NIS by depressing the Train A and Train B Power Range Permissive Override push buttons.

C. Initiate rapid boric acid injection in accordance with. A 3.4. Uncontrolled Reactivity Addition.

, D. Monitor intennediate Range N35, and verify reinstatement of the Source Range NIS when both Intermediate Range instruments are less than IX10~'" amps.

NOTES; OBJECTIVE . REFERENCES._;

ES 0.1 (step 14)

Page 79 of 100

EXAM QUESTION DATA SHEET SYETE.M Ela_t K/A RATINQ SAlTITY IUNCTION 000037 AK 1.01 3.5 i SG Tube leak EIIM

_ Knowledge of the operational implications of the following concepts as they apply to Steam Generator Tube I cak:

K/A TOPlQ Leak Rate vs. Pres < ore drop DUESTION; Consider the following indications when selecting your answer:

e Reactor Power 1007r o T.,, 559'F e Total RCS Leakage 0.8 gpm (includes SG tube leakage) e SG Tube Leakage 0.25Fpm 4

'Ihe Senior Watch Supervisor has directed the control room operators to perform a reactor shutdown due to increasing secondary

' side activity caused by SG tube leakage. Which ONE of the following statements is correct regarding the anticipated change, if any,in total RCS leakage as a result of the plant shutdown?

ANSWER C. Total RCS leakage will decrease, due the decrease in SG tube leakage.

DIST"ACTERS

A. Total RCS leakage will increase, due to the increase in SG tube leakage.

i B. Total RCS leakage will remain the same, increase in SG leakage will be offset by decrease in other RCS leakage.

D. Total RCS leakage will remain the same, decrease in SG leakage will be of fset by increase in other RCS leakage.

NOTES

REFERENCES; OBJECTIVE Composite i

f 3

1 4

Page 80 of 100 4

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EXAM QUESTION DATA SHEET SlSJIA1 K/A # kWLIMTING SAFETY IUNCTION 00005l AK l .02 3.6 loss of Main Feed Water Sllal:

Knowledge of the operationalimplications of the followine as they apply to less of Main Feed Water:

b/ATOflC ElIcets of feedwater introduction on dry SG.

OUESTION:

While recovering from a Reactor Trip due to a loss of both main feedwater pumps, EOP FR H.5, Response to Steam Generator Low Levelis entered due to indication that water level in #23 SG has decreased to 09^ WIDE RANGE level. In accordance with ILH.S. the SRO directs you to feed #23 SG at LESS TilAN 100 GPM UNTIL water level is GREATER TilAN 109 as indicated on #23 SG WIDE RANGE lesel indication.

Which ONE of the following statements is TRUE regarding the reason for limiting feed water flow to #23 SG to 1.ESS RIIAN 100 GPM until level is GREATER Til AN 109 ?

, ANSWER II. Feed flow is limited to prevent unnecessary thermal shock to a " Hot Dry SG" which could result in SG tube failure.

DISTRACTERS i

A. Feed flow is limited to prevent a rapid RCS cooldown which could result in challenge to the INTEGRITY critical safety function.

C. Feed now is limited to prevent flashine in the SG which could result in liftine the SG Safety Valves.

D. Feed flow is lirnited to prevent runout conditions on the #23 Auxiliary Feed Water Pump.

NOTES:

QlMECTIVE REFERENCFJi; FR H.5 Ilackcround 1

i i

Page 81 of 100

EXAM QUESTION DATA SilEET S1S1111 K/A # K/A RATING SAIETY FUNCTION (XX1038 EA2.03 4.4 SGTR STJLNJ:

Ability to determine or interpret the following as they apply to a SGTR:

}68LlQPIC Which SG is ruptured.

QUESTION; Which ONE of the following indications is NOT used in the Emergency Operating Procedure E 3, Steam Generator Tube Rupture to identify the RUPTURED Steam Generator?

ANSWER C. High Radiation form the Steam Jet Air Eiector Vent Radiation Monitor R45A/B.

DISTRACHRS A. Unexpected !evel rise in any SG narrow range level.

B. High Radiation in a Main Steam line.

D. High Radiation from any SG blowdown line (R-49).

NOTES.;

Q_DJECTIVf;

REFERENCES:

E3 Page 82 of 100

EXAM QUESTION DATA SilEET SYElliM K/A! K/A RATING SAFETY FUNCTION 000fXN EK3.28 4.5 SBLOCA EEM:

i Knowledge of the reasons for the following responses as they apply to the small break LOCA: '

KIAlOllC i Manual ESF actuation.

QUEST 10N:

Following a manual Reactor Trip from 1009 power you are performing Step 3 of EOP E-0, Reactor Trip and Safety injection, Check if Si Actuated", and note the following indications:

e St Annunciator Not Lit e Si Pumps None Running e Pressurizer Pressure 2090 psig (stable)

• Steamline AP All < $0 psid (stable) e Steam Line flow All < 100,000 lbm/hr (stable) e Containment Pressure 0.75 psig (stable) e Pressuriier Level 59 (stable) e RCS Subcooling Margin 87'F (stable)

Which JNE of the following actions should you take in response to these indications. and Wily is the action required?

ANSWL' A. Manually initiate Safety In_iection due to Pressuriier Low Level.

DISTRAC1 t:RS B. Manually Initiate Safety Iniec ion due to failure of the lingh Steam Flow Si to actuate.

C. Check RCS subcooling table. If LESS THAN required. manually initiate Safety injection due to Low Subcooling.

D. Transition to ES-0.1. Reactor Trip Response. Safety iniection is NOT required.

NOTES:

QIMECTIVE

REFERENCES:

E-0 Page 83 of 100

1 EXAM QUESTION DATA SIIEET S_YSlliM h/A# K/A RATING SAHiTY RINCTION (XXX)ll EK3.13 3.8 LBLOCA STEM.

5 Knowledge of the reasons for the following responses as they a, ply to the large break LOCA:

K/A TOPIC:

Hot-leg injection / recirculation.

OUESTION:

A Large Break Loss of Coolant Accident has occurred. Approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after aligning for Cold Leg Recirculction, the Senior Reactor Operator implements EOP ES 1.4, Transfer to Hot Leg Recirculation. Which ONE of the following statements is correci regarding the reason for placing Hot Leg Recirculation in service at this time?

ANSWER D. Hot leg recirculation is implemented to prevent boron precipitation in the core.

DISTRACTERS l A. Hot leg recirculation is implemented to sweep non-condensible gasses from the reactor head region.

B. Hot leg recirculation is implemented to cool the reactor head to enable RCS depressurization without additional void

formation.

. C. Hot leg recirculation is implemented to refill the reactor vessel and preclude fuel rod damage at the top of the core.

NOTILS; 4

OJUECTIVE liEFERENCES:

ES-l.4 Background d

i 4

4 Page 84 of 100 4

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i l

EXAM QUESTION DATA SHEET 4

SYSTEM h/A # K/A BATING SAFETY ITNCTION 000028 AK3.03 3.5 PZR Level Malfunction EBiM:

Knowledge of the reason for the following as they apply to the Pressuriier I.evel Control Malfunction:

K/A TOPIC Felse indications of P7R level when the PORV or spray valve is opened and the RCS is saturated.

OUESTION:

Following a Steam Generator Tute Rupture, EOP E 3, Steam Generator Tube Rupture, has been implemented. The Reactor Operator is controlling RCS temperature using the Condenser Steam Dumps. 'lhe Senior Reactor Operator directs the Reactor Operator to depressurize the RCS to LESS THAN RUI'TURED SG pressure using the Pressurizer PORVs since the RCPs have been tripped.

The Reactor Operator CLOSES the steam dump valves and prepares for depressurization of the RCS. Before commencing depressurization the Reactor Operator notes the following indications:

• RCS Wide Range Cold Leg Temperatures 480 F(increasing slowly) e Core Exit Temperature 488'F(increasing slowly)

I e Pressuriier Level (hot calibrated) 09 e RCS Pressure 1370 psig (increasing)

• Ruptured SG Level 889 (Narrow Range . increas'ng)

• Ruptured SG Pressure 1015 psig (increasing slowly' i Just prior to closing the Pressurizer PORVs the Reactor Operator notes the following indications:

• RCS Wide Range Cold Leg Temperatures 506*F (increasing slowly) e Core Exit Temperature 510 I (stabic) 4

• Pressurizer Level (hot calibrated) 809 (increasing rapidly) l e RCS Pressure 760 psig (increasing) e Ruptured SG Level 859 (Narrow Range - decreasing slowly)

• Ruptured SG Pressure 1005 psig (decreasing slowly) e Which ONE of the following statements could explain Al L of the changes in the above indications that have occurred since the Reactor Operator commenced depressurization?

ANSWER B. RCS cold leg temperature has INCREASED due to REDUCED natural circulation flow.

Pressuriier level has INCREASE.D due to increased makeup flow AND voiding in the Reactor Head.

Ruptured SG level is DECREASING due to SG fluid backfilling the RCS.

DISTRACTERS A. RCS cold leg temperature has INCREASED due to flashing of RCS fluid in the holleg.

PZR level has INCREASED due to increased makeup Dow AND voiding in the Reactor Head.

Ruptured SG level is DECREASING due to SG fluid backfillinc the RCS.

C. RCS cold leg temperature has INCREASED due to REDUCED natural circulation flow.

Indicated Pressurizer level has INCREASED to due flashing of the fluid in the level instrument reference leg.

Ruptured SG level is DECREASING due to SG fluid backfilling the RCS.

D. RCS cold leg temperature has INCREASED due to STOPPAGE of natural circulation flow.

Pressurizer level has INCREASED due to increased makeup flow AND voiding in the Reactor Head.

Ruptured SG level is DECREASING due to steaming through the Atmospheric Steam Dump valve.

NOTES:

OBJECTIVE

REFERENCES:

E-3/ Composite Page 85 of 100

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4 EXAM QUESTION DATA SIIEET l

SYSTEM R/A.# K/A RATING SAFETY FUNCTION

(XXX)36 AK2.01 19 Fuel Handling Accident STEM Knowledge of the interrelationships hetween the Fuel Handling incidents and the following:

R/A TOPIC Fuel handling equipment.

OUESTION:

1 A refueling operator has just placed an irradiated fuel assembly in the containment side upender when the refueling SRO notes that refueling cavity water level is decreasing rapidly.

Which ONE of the following actions should be directed by the Refueling SRO?

ANSWER A. Disengage the manipulator from the fuel assembly and lower the upender to the fully lowered position, THEN evacuate i containment.

DISTRACTERS B. Withdraw the fuel assembly from the upender and move it to the reactor. THEN evacuate containment.

C. Withdraw the fuel assembly from the upender and stc.re it in the manipulator mast. THEN evacuate containment.

D. Disengage the manipulator from the fuel assembly and leave upright. THEN evacuate con,tainment.

N_QIE1 OBJECTIVE REFERENCE 1 AO! l 7.0.3 f-1 l

1 .

Page 86 of 100

l 1

i EXAM QUESTION DATA SIIEET SYSTEh! K/A # K/A RATING SMIITY FUNCTION 000056 AK3.02 4.4 loss of Offsite Power hllihl Knowledge of the easons for the following responses as they apply to the Loss of Offsite Power:

K/A TOPIC:

Actions contained la EOP for loss of offsite power.

OUESTIOJ.;

Following a Reactor Trip and Safety injection you are verifying 480VAC Busses energized by offsite power using EOP E-0, Reactor Trip or Safety injection. You not the following conditions associated with the 480VAC Busses:

• All 480VAC Bus Normal Feeder Breakers OPEN

• All Emergency Diesel Generators in Service o All 480VAC Bus Emergency Feed Breakers CLOSED Which ONE of the following actions should NOT be performed when completing this IMMEDigTF ACTION sten?

ANSWER C Reset Lighting DISTRACTERS

A. Start ONE charging pump in h1ANUAL at maximum speed.

B. Ensure the following h1CCs ENERGlZED 4

• h1CC 26A e h1CC 26B

= MCC 26C D. Ensure the following MCCs - ENERGlZED '

• MCC 24A e MCC 27A

= MCC 29A

= MCC211 NOTES:

1 OBJECTIVE

REFERENCES:

E-0 I

Page 87 of 100

EXAM QUESTION DATA SIIEET SYSTEM 12 M K!A RATING S AFETY FUNCTIOJ 294001 K101 i 3 .6 Valve Lineups SIliM:

Knowledge of conduct of operations requirements:

K/A TOPIC:

llow to conduct and verify valve lineups.

OUESTION:

You have been directed to coordinate the completion of a System Check Off List (COL) for the Safety injection System. You note that the COL requires Independent Verification.

Which ONE of the following statements is TRUE regarding acceptable practice when conducting INDEPENDENT VERIFICATION of items contained in a COL that requires independent verification?

ANSWER D. Operator Performing the INDEPENDENT VERIFICATICN (Second Checker) must perform the verification independently of the First Checker. and should record the AS FOUND position of each component.

DISTRACTERS A. Operator Performing the INDEPENDENT VERIFICATION (Second Checker) may perform the verification at the same time as the First Checker and should record the AS FOUND position of each component.

B. Operator Performing the INDEPENDENT VERIFICATION (Second Checker) must perform the verification hidependently of the First Checker, and should reposition the component IF it is not in the required position.

C.. Operator Performing the INDEPENDENT VERIFICATION (Second Checker) may perform the verification at the same time as the First Checker and should record the AS LEFT position of each component.

NOTES.

Of1ECTIVE

REFERENCES:

COL Title Pace 4

Page 88 of 100

EXAM QUESTION DATA SilEET i

SX HliM RA1 MA_RAllNG SAlliTX11WG10N 294001 K102 17 Sion Tari h1111:

Knowledre of conduct of overations:

MAIDl'!C Stop Tars QMLS.liUN; A inaintenance inec hanic calls the control room and reports that a s ahe he hasjust removed frorn the service water sptern has a black and white STOP TAG attached to the handw heel that requires the valve to be in the OPEN position.

Which ON!! of the following actions should you take terarding this report?

ANSWi!R lt Obtain the tagout nuruber and report the finding to the work control center. Work protected by this tagout should cease until the discrepancy is corrected DISTRAC1LRS A. No action is neceuary since the s alve was tarred OPl!N. Work protected by this tarout is not allected C. Direct the Inaintenance enechanic to place the tag on an adjacent valve which is in the OPEN position. Work protected by this tarout is not allected D. Obtain the tagout nuinber and record the finding in the SRO log. Work protected by this is not af fected since the vahe was tarred in the OPl!N position SQlliSJ OlilliG1YE SiRiiNGS;

. O lo!

l Page 89 of 100

isNAM QUESTION DATA Sill:ET I

SDiTD1 MM WlLRATLNQ M1]iTY RINCllDN 244(Kil Kl.14 3.3 Confined Spaces STliM:

Knowledge of conduct of operations terardinr:

NMul'1C ,

, Confined spaces RL'liSJ10M While insestigating an increase in RCS leakage, preparatfora are made to make a contal:u int entry using SAO 219, Containment Entry and Egress. The requires that the watch chem 4' sample the atmosphere inside the 80 f t. eles ation personnel hatch (airkwk) before the entry is made.

Which ON!! of the followiny statements is 11tUll reparding the reason for sampling the atmosphere inside the althwk before entering the containtnent?

ANSWiiR C linsure that sufficient osyren is available to support human life.

1)lS111ACTliRS A. Measure radioactive ras concentrations to detern'iine if respirators are required 11 Ensure that opening inner althwk door will noi result in an increase in pall airborne radioactivity concentration causing sentilation to trip.

1) linsure that combustible lesels of hydroren do not exist in the airlock prior to personnel entry.

Enllih; DiMliCIlYli KliRRlWCES; S A0 219

(

Page 90 of 100

EXAM QUESTION DATA SIIEET S.YhiliM Ksu K/A._lMIING SAllillLUNC110N 244001 K l .16 3$ Fire Equipment S.Il M Knowledre of 1 nc and Safety lauipment:

K/A I DE C Faility protection requirements including fire brirade and portable fire equipment (I!DO Ilullding Delitre System)

QUl211RS; During power opetation the DIESLL !!LDU 11Rii PROT OPERATION annunciator is activated. Which ONE of the following statements is correct regarding the response of the Emergency Diesel Generator (EDO)lluilding Fire Protection System? ___

j ANSWER '

D. EDO lluilding Fire Protection System is a welline system and water spray is activated locally at each spray norile, lhe abme alarm could mean that one of the spray noriles has actuated DISTRACTFRS A. EDO IluildmF ireFProtection System is a dry line $ystern and water spray is activated to the entire building by a deluge vdve. 'the above alarm could mean that all of the spray noriles have actuated l

i

11. EDO !!uilding Fire Protection System is a concentrated foarn systern and loam is acth ated kically by thermostats.1he above clarm could mean that the fourn system has actuated C. EDO Iluilding Fire Protection System is a llALON system and il ALON is activated kically by thermostats. The abmc alarm could mean that il Al DN system has actuated Nulli5; 01ULGlXti lELIRl;NCliS; ARPSOF Page 91 of 100

EXAM QUESTION DATA SilEET hYS111M h' U KM RK11MO SALE 111EQDS 294001 K I .(Lt 33 Al. ARA S11 M Knowledge of Radiokirical Praelices includinr:

K%3D11C i AI. ARA l Q Ulihll W J l Station policy requires that each individual follow practices that ensure that their personal radiation esposure is kept As low As Reasonably Achievable (ALARA) Consider the following situation:

• Maintenance inust be peia .1ned in an area where the general area radiation levels are $ rnr/hr 'y and 100 mr/hr p o '!he snaintenance activity is estimated to take 1 person I hour e

Most of the radiation is due to contamination of the floor in the area.

Select the Al.AR A practice that wouhl result in the lowest achievable radiation esposure (ttnd ersonRet i. for this lob?

ANSWiiR A. liquip worker with a plastic face shield and protective clothing against contamination.

DISTRACTF.RS II. Decontaminate the area f requires 2 people and 45 minutes) before commencing work.

C. Cover the lhwir in plastic shecting (requires 2 people 30 rninutes) before commencing work D. Cover thuir with lead blankets ( requires 3 people 30 minutes) before commencing work MulliS; DilEGYli KLERtEliS; Rad Safety Training Page 92 of 100

isXAM QUESTION DATA SilElsT sn1141 NM MA_lu1EO SAniI n utscu ols GliN K213 30 Shilt Turnos er i STliM:

Knowledpc of shif t turnmer practices:

N!A1Dl1C I)uring reactor startup.

RUliS110X You are pellonning a Reactor Startup. Your watch relief from the on corning shif t has arrivedjust as you are ready to withdraw the control banks to approach criticality.

Which ONE of the following statements describes an acceptable practice when conducting Shift turnover during a Reactor i Startup?

ANSWI.R

11. 1)o not herin control hank withdrawal. Conduct shif t turnmer after neutron flus has stabillied.

I)lSTRAC'll RS A. Conduct Shili turnmer at the flight panel as you conhnue the reactor startup ensuring that you are not distracted from anonitoring neutron flut C. A! low your relict to continue the reactor startup as you relay pertinent watch turnover infonnation to him

1) Continue the startup while the rest of the crew conducts watch turnover.

NulliS; OlutiCl1E RL;tliRiiNOiS; POP l.2. p N of 15 Page 9.T of 100

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EXAM QUESTION DATA SilEET EnILM WM MA_im11No EAum1!KcIloh' 294001 A l .01 33 l'ncedures SlliM:

Knowledge of conduct of operations regarding pnicedurec Wh3DUC Temporary Procedure Chanres QUliS3]Ob';

While perfonning a plant cooldow n to 350T using POP 3.3 Plant Cooldown. you notice a handwritten notation in the right margm of the page that contains the following information TPC 96-l$3.

Which ONii of the following statements in er.nect terarding the significance of this notation?

ANSWiiR D. The notation refers to a Temporary Procedure Change. I r.iust ref er to the lemporary Pnicedure Change Request I onn kicated at the beginning of the procedure to perfonn the amiciated step DIS 1RACTI.RS A. The notation refers to a Temporary Procedure Change. I must refer to the Temporary Pnvedure Change Log flook hwated in the control room to perform the amiciated step 11 The notation teless to a Ternporary Procedure Change. The step that is allected will be lined out and the concet inf onnation entered in the baly of the pnicedure.

C. llandwritten notations are not pennitted in plant procedures. I should notif y the SRO and obtain a clean copy of the proecdure.

NGTliE; DMEC11YU lilERENCES; OAD 27 Page 94 of 100

EXAM QUESTION DATA SilEET SXE11M K/M huLEATING 1 S$11?TY IUNCl]QN 294(X11 Al 04 30 Communications S11;M:

KnowleJte of plant communications systeme K/A 101'IE; limerfency phone systems DUlilllONI An event has occurred requiring notincation of the NRC within I hour. Which ONE of the following communications splems should be used to perform this notification?

ANSWl:R A. limerrency Notification System (liNSIl' hone DISTRACT 1!RS Il Radiological timetrency Cornmunication System (Rl!CS)

C. Microwns e l' hone 1.ine D State 1:merrency Management Of fice (SI!MO) Radio NQ11;E QlMLCllYli KlillRl!FClii SAO 124 Page 95 of 100

EXAM QUILSTION DATA SilEET liYMEM K%.! WA_ EATING SAELTY 17UNCTIQS 294(Nil A l .16 31 IImetrency Plan SlhM:

Knowledre of the IImergency I'lan-WA.lGl'J C limergency Claulfications Qk'liUlOli; o Which ONE of the following is the minimum I!rnergency I'lan Claulfication w hich would require

• level 2" stalfing?

ANSWiiR C. Alert DISTRACTliRS A. Site Area limergency II. General limergency D. Notification of Unusuallivent tiQlES; DlulLCTLYJi Kl!LIRiiNCliS.;

li l'lan li' l(K)3 i

Page 96 of 100

l EXAM QUESTION DATA SilEET SYS11111 IV M h/A_1M11NG tiMirrY litNC11QN 244001 Kl09 34 Safety SJ1ihl:

Know!_*dpc of safety requirements for:

K/AIQl'lC Sciety procedures related to high preuure (Tapping of equipment for personnel protection)

QL!131102 Which ONE of the following statements correctly identifies the taFout protection that mu:.t be provided if a worker is going to oork in a tank that is connected any system' ANSWER II. Itach source of energy must be ikolated by TWO Cl OSl?D, I.OCKl!D and TAGGED isolation valves.

DIS 1RACTERS A. Each source of energy must be iwlated by 1WO CLOSliD, and TAGGED isolation valves or by ONE CLOSliD LOCKED and TAGGED holation valve.

C. Each source of enerFy must be bolated by TWO CLOSED, LOCKED and TAGGED isolation vahes or a safety person must be stationed at the tank opening to aulst in emergency erreu.

D. Each source of enetgy must be holated by at least ONE CLOSED, LOCKED and TAGGED holation valves, or a safety person must be Stationed at the tank openinF to auht in emergency egress.

NQ1lifi; 9MiiC11Yli KlilHliNfEi; S AO .05 Page 97 of 100

EXAM QUESTION DATA SIIEET SHll;M NM MA_IM11NO SADiU EWIllON 294001 Kl.15 34 llydrocen h11iM:

Safety procedures related to llydrogen MlL191'lC lhminability lirnits.

Q1'liS11RE Which ONE of the following gas samples would indicate that the gas space in the associated contains a potentially llammable misture?

ANSWi!R C. 221.arre Gas Decay Tank 79 Ilydrogen.179 Oxygen. 769 Nittoren DISTRACTliRS A. Volurne Control Tank 979 voydrogen.1% Oxyren. 29 Nittoren il 23 CVCS Iloidup Tank 269 Ilydroren. 09 Oxygen. 749 Nitrogen D. Main Generator 949 Ilydrogen.19 Oxyren. 59 Nitrogen b1111;SJ DiMlIllni RllliimbIliE:

Safety Training _

page 98 of 100

EXAM QUESTION DATA SilEET s n nim h!M EIA_RAnNG SARIDUNCHOS 294001 Kl 05 4.1 Security XILM:

Knowledre of security procedures includine:

KM10DC

!)efinitions security innet QWSHON; Chm e the selection that correctly completes the following staterient:

The Condensate Storage Tank area is a(n) , the Simulator Dullding is in the and 1 the Main Turbine is in the .

l ANSWI:R 11 %tal Area. Owner Controlled Area. I'rotected Area DISTRACTERS A. Isolation 7.one Daner Controlled Area. Protected Area II . Vital Area. Isolation Zone. Protected Area C. lisclusion Zone. Owner Controlled Area. Protected Area NDEiS; DlithCHi'li RiiERUNGS; Security Training (GET)

Page 99 of 100

EXAM QUESTION DATA SIIEET n11ul Klu IwLMImg SANITY IvEnos 294(X11 A1.06 34 Conduct of Ops HliM:

Knowledge of Conduct of Operations including:

KMlDl'lC Ability to maintain accurate, clear, and concise logs. records, status boards, and reports.

RWiS110E Which ONE of the following Central Control Room log sheet entry examples would NOT require that the entry be RED circled end explained in the remarks section?

ANSWliR D. Reading is taken one hour late due to startup activities DISTRACTliRS A. Technical Specincation readmg eseceds NORM Al, limits specined on log sheet.

11. Ikluipment is out of sersice.

C. Reading eseceds MIN /M AX hmit specified on log sheet liQlliSJ Dl!IliCHYli RilliRiiNCliS; OAD3 h

I l' age 100 0f 100 l

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