ML082820496

Final - RO & SRO Written Operator Licensing Examination with Answer Key (401-5 Format) (Folder 3)

ML082820496
Person / Time
Site:	Salem
Issue date:	07/26/2008
From:	Public Service Enterprise Group
To:	Brian Haagensen Operations Branch I
Hansell S
Shared Package
ML080030005	List: ML081060248 ML082350047 ML082470469 ML082730736 ML082730775 ML082730779 ML082730784 ML082730786 ML082730790 ML082820486 ML082820496 ML082820517 ML082820534 ML082820540 ML082820562 ML082970764 ML082970767
References
U01686
Download: ML082820496 (127)

Text

ES-401 Site-Specific RO Written Examination Form ES-401-7 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Date: Q 1.t Is, F aciIitylU nit: S ale,,Y-q Region: II III IV (5 Reactor Type: W d C E L_! S K I G E D Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet 3n top of the answer sheets. To pass the examination, you must achieve a final grade 3f at least 80.00 percent. Examination papers AI be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.

Applicant Certification qII work done on this examination is my o m . I have neither given nor received aid.

Applicant's Signature Results ixamination Value - Points

\pplicant's Score Points ipplicant's Grade - Percent ES-401, Page 30 of 33

Reactor Operator Answer Key

1. c 26. c

2. c 27. d

3. b 28. b

4. d 29. b

5. a 30. b

6. d 31. a

7. c 32. b

8. a 33. a

9. d 34. c IO. d 35. a

11. c 36. a

12. b 37. b

13. a 38. c

14. d 39. b

15. b 40. b

16. a 41. c

17. a

18. a

19. a 44. a

20. b 45. c

21. a 46. b

22. b 47. a

23. c 48. d

24. b 49. c

25. d 50. d Page 1

51. b

52. d

53. c

54. d

55. d

56. a

57. d

58. b

59. c

60. b

61. a

62. c

63. c

64. c

65. a

66. a

67. d

68. a

69. d

70. c

71. b

72. b

73. d

74. d

75. b Page 2

Given the following conditions

Q\

~- Unit 2 is operating at 95% power.

1- Control Band 'ID" rods are at 215 steps.

- Rod Control is in AUTO.

IWhich of the following describes how the plant will be affected if PT-505, Turbine Steamline Inlet

___ -___ _- ~ _ _

r--

The reactor will trir, on Over TemDerature D/T.

__~--____-__ ~-

I1"D" control bank rod motion will stop at Group Demand Counter indication of 227 steps (ARO)

'from Control ~-

Grade

_ _ _Interlock

~ _ _ _ C-I _ _1, and the Rx -~~

will NOT~trip._ _ ~ _ _ ~ _ _ ~

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

[ L D ~-

control bank rod motion will stop at Group Demand Counter indication of 227 steps (ARO)

from Control Grade Interlock C-2. and the Rx will NOT trir,.

OOOOOI_K1~3

- - - _ ~ ~

_ ~-~

of the operational implications of the following concepts as they apply to Continuous Rod IPT-505 feeds the auto rod control logic circuitry as the "reference" signal. When it fails high, the rod icontrol system sees that actual temp (auct hi Tave) is much lower than reference temp (PT-505) and will rwithdraw rods automatically. The rods are almost fully withdrawn from the core initially at 215 steps, and Ithe remaining steps have very low reactivity associated with them. There will not be enough positive reactivity added from the rod withdrawl to cause any automatic Rx trip. Interlock C-I 1 blocks AUTO rod control at the cycle dependent all rods out (ARO) position, which for cycle 17 is 227 steps per S2.RE-IRA.ZZ-0011. A & B are incorrect because the Rx will not trip. D is incorrect because C-2 is the overpower rod stop at 103% power, and the rods do not add enough reactivity to reach that, plus the ARO ABROD3E001 - as aDDlled Describe the oDeration of the followinq .. to S2.OP-AB.ROD-O003(Q):

a) Reactor Control Unit b) Logic Cabinet c) IRPl d) Bank Demand e) System Alarms CMonday, September 15,2008 9:24:36AM .~

I Page IOf87 I

~ - - ~ - ~ ~

IUnK 2 isoperating normally at 100% power when the unit is manually tripped.

- ~_ Qz Compared to their pre-trip values, which of the following indications will be present one minute ifollowing the uncomplicated ~- ~ Rx trip? --

__- ~-

-- - ~ -

_ ~ _ _ _ _~ _ ~ . _ __ _ _

~~ _ _ ~

~

~ -

~~

~ -~ ~ --

IPZR level 50% and lowering. -- ~

~ _ ~ _ _ _ _ _ _ _ ~ ~~- _ -~ ~ _ ~ _ _ _ _ ~ ~

Eeal Injection

~ . ~ flow to all _RCPs

_____ _ _ ___ has-~

___ risen -1 gpm.

~ -- ~~~ ~ ~ _ _ _ _ _ _ ~ _ _

- -~~ ___ ___ _ _ _ ~ _ _ ~ - _ _

JCharging

-__ ~

System flow has lowered to 75 gpm. -~ -~ _ _ ~ _ _ _ _ _ _

_ ~ _ _ _ ~ _ _ _ _ _ _ ~ ~ _ _ ~ _ _ -~

~ __ _ _ _ ~

pC71, Letdown HX CC flow control valve demand has lowered -25%.

~- ~ ~_ ___

AppkAt ion ~ - _8/25/2008 nd Abnormal Plant Evolutions 000007A109 I 2' L_____

_ _ _ ~ _ _ _ ~

-~ ~~ ~

te and / or monitor the foll pply to Reactor Trip:

-~

~~ ~ ~ _ _ . _ _ _ _

___ ~~ ~ -~

~_ .~_ _ _

correct because the initial shrink following the trip will reduce PZR level well be is incorrect and C is correct because PZR level will still be above the programmed level derived from the auctioneered Thot, and charging flow will be lower due the charging system Master Flow Controller demand lowering, and driving charging flow lower. The CV71 valve will not have changed position, and Ithe lower total system charging flow will mean the total seal injection flow will lower proportionately. D is incorrect because normal letdown come from 23 loop cold leg. The cold leg temperature will RISE following a Rc trip as SG pressure rises. The letdown temperature will RISE, and the Letdown HX CCW

'flow will need to RISE to maintain temperature. This will cause CC71 valve demand to RISE. When iverified in the simulator. the rise in valve demand was verv small. 1%. But it will definitely NOT lower.

~ ~- ~-

CVCSOOEOI5 LOR NCT Given plant conditions, relate the Chemical and Volume Control System with the following, Pressurizer Level Control System RCS Temperature Control Main TubinelGenerator Reactor Coolant Pump seal injection flows Automatic Control Rod Control VCT Makeup Nuclear Instrumentation Emergency Core Cooling System Residual Heat Removal System Component Cooling Water System Pressurizer Pressure Control System Pressurizer including Pressure Relief Tank Waste Gas Waste Liquid Service Water 4 Kv Vital AC System 480 V Vital AC System 240 V Vital AC System 1

~

~~

_~

~

~

-~

-- ~~ -~ ~~

-:E :Monday, _September

_ _ _ _ _ _ _ - ~ ~ - _

15, 2008 9:24:36 AM 1 Page2of87

LMonday,_September I 15,~ - AM_ ~ - I ~

2008 9:24:36 ~ i-Page3of87

Unit 1 experienced a SBLOCA.

A manual Rx trip and Safety Injection were initiated.

RCS pressure is 1085 psig.

The hottest CET is 554.0 degrees.

Operators are determining whether conditions are present to allow a transition to EOP-TRIP-3, SI Termination, at Step 9, SI Flow Reduction Criteria.

Due to a concern with the indication of the Subcooling Margin Monitor, the CRS asks the RO to determine RCS subcooling using Steam Tables.

which of the following identifies current RCS subcooling, and whether the transition to TRIP-3 is lappropriate? (Assume all other conditions required to make the transition are SAT.)

I subc cooling is degrees, and the transition to TRIP-3 be made.

wbcooling is met, and the transition is warranted. If the candidate uses 1085, the subcooling would be very close to zero.

I W" F'

?&ifereflceTlth, I LOCAOIEOII A. Determine a discrete path through the EOP B ~ Determine an appropriate transitionout of

-~ ~- ~ _ _~- ~~ ~ __ ~~ __ ~ ~~~

-~ -~ __

2 flowchart. Steam Tables

~-

~

.~

Monday, September 15, 2008 9:24:36AM 1 Page4of87 1

~ ~ -- - _~ - ~ ---

balem Unit 2 has experienced a rupture of a RCS cold leg which has resulted in containment pressure peaking at 18 psig.

With all systems actuating as expected, which of the following choices identifies the containment

_,isolations

~-

which have occurred, and the reason why they have occurred?

~- _ -- - - ~ - -

---

-- -~

~

- ~~~~ ~ ~ _ _

[Phase A to ensure non-ess containment penetrations are isolated; Feedwater to isolate Phase A to ensure non-essential containment penetrations are isolated; Phase B to isolate additional potential release

.- __ - paths from __ containment. - _ _ _ _ _ ~ _ _ _ _ ~ _ _ _ _ _

4Wetgni;eTitle ,' '""

Reactor Trip or Safety Injection LOCAOlE007 Identify possible radioactivity release paths for a Loss of Coolant Accident, and describe how the actions of 2-EOP-LOCA-1 minimize the potential for a release

~

- ~_

-rce: Previous 2 NRC Exams Question source Cfili%a&

pnday,ptembefi5,2008 -. -

9:24:36 AM-- I Page5of87 1

With Unit 1 operating at 100% power, which of thefollowing will cause RCP Standpipe level(:)

os to

,Failure of a RCP #3 Seal.

I 7CV71 Seal Pressure Control Valve. fails closed.

8/25/2008

~

-- ~

rgency and AbnormaFlant Evolutions

~

lfunctions

- ~______

I(B) Level would lower; (D) 2CV71 regulates flow to seals (C) More charging into RCS not through #I

~ _ _ . ~

[=-I=---

Monday, September 15, 2008 9:24:36 AM

. _ ~ _ _ _

I Paae6of87 1

Given the following conditions:

- Unit 2 was in MODE 4 with 21 RHR loop providing shutdown cooling, and 22 RHR loop aligned for ECCS.

- 21 RHR pump began cavitating due to a valve being mispositioned during a tagging Which of the following describes __ the preferred flow rate when starting the RHR pump, and why?

~ _- _ _ _ _ -~

-~

_ . _ _ ~

~ _ - ~ _ _ ~ _

[Higher flow rate to sweep - entrained - ~_ air .-

- _ ._. ~ __- ___ _ .___- - _

[Egher

- ~ flow rate to quickly terminate the temperature rise in the RCS.

_ - - - -~ __ - ___ _

[Lower flow rate to prevent

_ ~ _ _ _ high starting current on the RHR pump.

_ _ ~

L - ~ ~

- _ - ~ _ _ _ _ - __- ___ . _ ~ _ _ -_ ~ ____ _-

ILower flow rate to limitsudden cooldown and to minimize level loss caused by collapsing voids.

1 Application Salem 1 & 2 I Emergency and Abn l000025A205 6

-_ _ _ _- ~~ __ ~~

Lty to determine and interpret _ _ the

_ _ following

~ - as they apply to Loss of Residual Heat Removal System:

~ ~ _~

- - ~ _ _ -~ -___ ___ -

1 -.-3 1*1-.3- 5*

states that it is for the reason as stated in entrained air is the method used when time does ABRHRlE004 Describe, in qeneral terms, the actions taken in S2.OP-AB.RHR-0001 and the bases for the actions in accordance with the p a y , September 15, 2008 9:24:36 AM Paqe7of87 1

~_

iGiven the following conditions:

q7 Unit 1 is operating at 100% power, steady state, with no surveillance procedures or testing in progress.

The peak outdoor temperature has exceeded 95 degrees for the past 7 days.

Service Water system problems combined with the high ambient temperature has caused Component Cooling Water system temperatures to rise.

The unit CRS is attempting to reduce CCW system flows.

_~ _

normal flow is 1,000 gpm. C is correct because Spent Fuel Pool normal flow is 3,000. RCP thermal lbarrier normal flow is-40 a m .

a) General arrangement of the Component Cooling Water system.

b) CC System loads c) Making up to the CC Surge Tank d) Safety precautions for handling/working with chromated systems.

~ _ ~ _ _

~~

~ ~ -

~

~ - _ _ _ _ _ __ _ ~_ ~_ - ~

~~~ ~- __ ~

~ ~~ ~ __ ___ I r M o d a v . SeDtember 15.2008 9:24:36 AM 1 Page8of87 1

- - - - p - - - - - - - I -

Given the following conditions:

- Unit 2 is in MODE 4.

1- RCS pressure is 350 psig.

1- All wide range cold leg temperatures are 310°F.

1- Pressurizer Overpressure Protection System is ARMED.

IPredict the plant response to RCS wide range pressure transmitter 2PT-405 failing high with NO IDirect From Source LMonday, I - 15,~2008~9:24:36_AM ~ ~ _ I _Page9of87

September

_ _ _ . _ _ ~ _

_ I

IGiven the following conditions:

- A 650 gpm tube rupture has occurred on 22 SG while operating at 100% power

- 22 SG NR level rose to 93% before SI could be terminated.

I- 22 SG NR level is currently 89% and dropping slowly.

Which of the following describes how the crew is allowed to utilize 22 SG during performance of SGTR-2, Post SGTR Cooldown?

~~ ~- ~- ~ _ _-_ -~ ____ ~- _ _ _ _ _ ~ - ~ _

[NOT allowed to steam 22 SG to prevent re-initiating primary-to-secondary leakage as SG ressure lowers.

lF-- ~ ~ ~~ _ ~ ~_ _~

~~~~ - __ _______ - _ _ ~ - ~ - - __

lallowed to steam 22 SG ONLY with TSC approval, AND an acceptable release rate calculation 1- d 1- R 7 1 Application 8/25/2008

~_ _ ~ -- _ _ -_____- ~- -~

lw2- Ability to determine

~

~ ~ ~~

_ _ ~and interpret the following as they apply to Steam Generator Tube Leak:

-~ ~ ~ - _ ~ -_ __-

_ _ ~_ - _ __

~ _ _ _ ~ -___ ~- _ _ ___ __ _~ ~ _ _ ~-~ __ _

to be taken if S/G solid and water enters steam lines

%, then the TSC is to pe overfill evaluation. It also states NOT to release steam unless directed by the TSC. Step 8 asks if the TSC recommends steaming the ruptured SG. If YES, then an acceptable calculated off-site release dose is required PRIOR to continuing to Step 25 which is the RCS cooldown at which time the ruptured SG COULD be used. A is incorrect because even if the NR level is now <92%, it was above 92%, and is considered to have water in the steam line. C is incorrect because re-initiating pri to sec flow is not the concern assocuitaed with high SG level. B is incorrect because 10CFR20 limits, while they always apply

~ - . -

Monday, Page 10of 87

_ _September

_ _ _ 15,_ 2008 _ _ _AM~

~ 9:24:36 I

~

~

I IWhen responding toalarge SGTR, which of the following identifies OperatorAction time ired by the Salem FSAR, and- the reason for it?

~

~~~

~

~~

~

~~~

~

___ ~~~

~~

shut the SG Blowdown Isolation Valves, GB4s, within 10 minutes to limit the spread o contamination 1-_ -_____._- ~~~ -~_ ~~ ___- _ _ _ ~

___ ~~~ ~ ~ ~__-

p - ~~~ _ _ ~ . ~ _ _ _ ______ ~~

reestablish letdown within 45 minutes to prevent PZR overfill and water relief through PORVS/Safeties.

________ _ _ _ _ _ _ _ ~ _ _ _ _ _ _ _ _ _ _ _ _ .____ ~- ~-_p ~ -__ -~ .~

_ _ ~ _ _ _ ~ ~ _ _ _ _ _ ~~ ~~~ ~~~ ___- - _ _ - ~ _ _ _ - -__

p

~

llsolate feedwater into and steam flow out of a ruptured SG within 10 minutes to minimize the lloss of mass from the RCS.

Lp ~~ _ _ ~ ~ __ ~~ -____ ~ ~~ ______ ~ ___-

-____ ~~~~ _____ _p ~ p -

_ _ ~ ~ ~ ~ ~ ~ -

esponses as they

~ ~~ p ~ __apply to Steam Generator Tu ure:

nventory balance, S/Gtube rupture, and plant shutdown ~ 4.2 4.5 I

is incorrect because ai correct because it is the action, time, and reason to terminate an Inadvertent SI. C is incorrect because Ithe action is correct but the reason is to prevent SG overfill. D is correct because FSAR takes credit for ,

50 minute isolation.

r Reference Title 1 Steam Generator Tube Rupture Rx Trip or Safety Injection Question IWodffication Method: I I Monday, September 15,2008 9:24:36AM I Page11 of87 1

_. . . . . ^ ..._. .

_~ ~ - ~- ~- - \

IWhich of the following describes the reason why steam dumps are blocked from opening on a Icondenser when vacuum lowers to 20" vacuum?

~

the condensers with low vacuum causes.. . ~_ -

~~ ~ ~ ~-

'substantial pitting of the condens

~~

__ _ _ _ _ _ ~ ~~~ _ ~ ~ _ _ ~ _ _ _ _ _ _

r---- - _.. _________ ~ - ~-

uneven heating and premature

~- -~

- ~~ failure of the__LP turbine exhaust hood.

_~ ~ _ _ _ _ _ -

- _ _ _~ -~ ~ _ _ _ _ _

Ivacuum to degrade further causing a reduction of NPSH to the condensate pumps.

- ~ -

~- ~ ~~ _ _ ~- - ~ - - ~~ ~ - - ~

icondensate depression to lower to zero, and flashing in the condensate system will occur.

condensate pump suction. This is why AB.COND has operators monitor cond pump suction temp I because it is expected to rise. D is incorrect because it will not lower to zero. B is incorrect because while exhaust hood boot temp may rise, it is not the

~ _ _ _ _ _ _ -~ reason for stopping steam flow

~~~ _~

~ - - ..

~-

- - ~

Technical

~

~

B&es

~ ~~

Document.

_ _ _ ~ _ _~- ~ ~~ -

~-

~~~ ~______ ~ ~

_ _ _ _ ~

STDUMPEOO6- LOR Outline the interlocks associated with thefollowing Steam Dump System components: (Licensed Operator & STA only)

_ _ _ _ _ _ . ~. ~ _

1 Monday, September 15, _ 2008 9:24:36 AM I Paae12of87 1

1 G6 en the fo I Iow ing co ndi i ons:

- Unit 1 is operating at 100% power.

- 11 AFW pp is C/T.

I- The unit trips and auto SI is actuated due to an unisolable Main Steamline rupture I on 11 SG.

- 12 AFW pump did not start when demanded, and can NOT be started.

- 13 AFW pump tripped 2 minutes after starting, and can NOT be immediately reset.

- 11 SG has blown dry.

'Which of the following describes the mitigation strategy in attempting to feed 11 SG with the icondensate system during performance of FRHS-1, Response to Loss of secondary Heat Sink?

- - - -~

~ _ _ _ _ _ - -

-_____-~

'1-1 SG may be fed @ 1E4 Ibm/hr to re-establish a heat sink when been properly aligned. Depressurize ANY other SG to establish condensate flow to assist in lrecovery.

Lp_______-- - - -

_ _ _ - ~ ~ _ _ _ - - _ p ~ ~ ~ __ _ _- _ - - p ~ ~ ~- ~ _ _ _ _ _ _

~ p ~ _____ ~ - -____

- -- ~

[TlTG-may NOT be fed due to potential tube damage from cold feedwater introduction to a dry SG. DePressurize ONLY 12 OR 14 SG to establish condensate flow to assist in recovery.

p ~ - - . pp--_. - _ _ _ p -

_ p _ _ _ _ _ _ _ - _ _--_ - -

IISG may NOT be fed due to potential tube damage from cold feedwater introduction to a dry SG. Depressurize 12, 13, or 14 SG to establish condensate flow to assist in recovery.

p _ ~ - - _ _ _ _ _ ~

_ _ _ _ _ _ ~

_ _ _ _ p ~ -

- p - - ~ - - ~ _ _ _ _ _ ~ . -~_-____ -

-- ~ ~ _ _ _ _ _ _ _ _ _- _ - - - _ _ _ _ _ ~ ~ - - - -~-____- - ~ p - ~ _ _ _ _ _ _ _ _ -

- -p~-__________

I11 SG may NOT be fed to prevent steaming the faulted SG. Depressurize 12, 13, or 14 SG to lestablish condensate flow to assist in recovery.

~- ~ . - - - p - ~ _______p-

~ ~ - _ _ _ _ _p - - - -

[Lossof Main Feedwa SG. The ERG basis states that the thermal shock of feeding a dry SG could cause a tube leak or rupture that would be unable to be isolated until the secondary boundary was restored. This is why 11 SG will NOT be fed, and A is incorrect. C and D are incorrect because the procedure states that if other SG are available, then 11 and 13 SGs should be steamed last to maximize steam supply for TDAFW pump. The stem states that the 13 TDAFW pp cannot be immediately reset, which infers that it may be able to be reset, so conserving inventory in 13 SG is correct. That leaves 12 or 14 SG to be selected to depressurize.

~~

I Monday, September 15,2008 9:24:36 AM I Page13of87

Which of thefollowing describes thebasis for why Functional RestorationProcedures (FRPs) are NOT implemented until specifically directed in EOP-LOPA-1, Loss of All AC Power?

~- - ~

~ ~-

-_____- .~

ALL FRPs are written on the pr

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

1LOPA-1 actions must be performed in sequence. Implementing FRPs interrupt the sequence land timing of steps.

- - -~ ~~~~~ - . -

ICertain diagnostic steps must bebetformed to minimize RCS leakage through the RCP seals.

lThese steps are specific to LOPA-I and are not performed in any FRP.

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

~

-- ~~ - - -

~ - ~ _ _ _-- ~ -

ILOPA-1 includes all the key actions of RED path FRPs. Performing FRPs would be redundant iand prolong the time until RCS depressurization wa med.

~

a R- Memory- ~- Salem 1& 2

- ~ _ _ _ _ _ ~~

I I Emergency - Procedures / Pla ,

I prioritizing emergen I - 2.8' 3.8

~

~~ J ption t ita1 bus is energiz provide power for controlling equipment to provide mitigating functions for the loss of power as per bases document. B and D are incorrect because LOPA-I ..."is written to implicitly monitor and maintain critical safety functions."

C is incorrect because the steps in LOPA-I are not diagnostic, they are performance. While C might be considered a true statement if the term "diagnostic" were interpreted differently, it is still an incorrect answer to what is asked for in auestion stem.

I "-

Reference Title I Learning Objectives LOPAOOEOOl For the following analyzed transientdaccidents.

A. Loss of all offsite power B. Loss of all AC power

1. Describe the analysis assumptions.

2. Describe the protective features that mitigate the event (N/A for a loss of all AC power).

3. Describe the analyzed plant response 0,describe the expected fuel failure mechanism~ ~ _ _ _

~-

~

I

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

_ _ _ _ ~ _

I Monday September 15,2008 9:24:36 AM L--L-.-. - .~

I Paqe 14 of 87 ~ 1

i Given the f o Ilowingc o nditGns :

- Salem has lost all off site power in August, and outside air temperature is 95 deg. F.

~- Unit 2 was operating at 100% power prior to the loss of off site power.

- ALL 4KV vital busses are energized from their respective EDGs.

- Off site power has not been restored 15 minutes after loss.

1- Hope Creek remains at 100% power with its off-site power supplied.

Which of the following statements is correct regarding indications available on, and the status of, the Unit 2 P-250 computer?

~p -

- -~ -~ -- __

~ p~

~

____ p~ __ ___ ___

1No indications are

~ _ _-~ ~ available

_ _ _because

- - ~-the P-250 does not have power.

- ~

~~~ ~ - p - p ~ ~~ ~~ -~ ~

___. p p -- ~ ~ ~ ~~ _ _ _ _ _ _ _ ~

'Main Steam

~ _ _ - __Dump valve position - 8%. P-250 is

~~~ ~~~ ~powered

_ _ ~

from a vital bus-

__ ~~ ~~-

from an EDG.

_ _ ~ ~

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

IContainment Average Temperature - 115 degrees. P-250 has power from Hope Creek substation.

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

~- ____ ~ - p ___ p~ __ ~p p~ ~~ ~~ ~~ - p ___ -~

es will be blocked closed because of the loss of Group bus power Ito Circulating Water pumps. RCS loop Tc's in the loops will be at saturation temperature for the SGs at I1015 psig because of natural circulation flow in the loops. P-250 is powered from Hope creek substation l#1, and will remain powered up since stem states Hope Creek remains at power, and would have tripped a The Control Room locations of the P-250 Computer terminals (N/A NEO) b The function of each P-250 Computer Control Room control and indication

1) USE all of the function buttons provided at the top of the process diagram windows 2 ) ACCESS the System Status Diagram

3) DIAGNOSE the status of the system using the functions provided by the System Status Diagram

4) EXPLAIN the significance of having a drop 254 on the highway

5) ACCESS the Base Alarm System

6) DEFINE alarm priorities and EXPLAIN how to distinguish them on the display

7) DEFINE point Quality and DETERMINE the quality of any points displayed on the screen

8) MODIFY the alarm screen to change between displaying a current alarm list, an alarm history and a list of unacknowledged alarms

9) USE the alarm filtenng capabilities to focus alarm displays on user definable parameters IO) DEFINE the differences between analog and digital point records

11) DISPLAY point information on desired points by using at least 3 methods

12) EXPLAIN all of the information being displayed in the Point Information window when in the Reduced mode of display

13) PERFORM a Point Search to find groups of points meeting user define attributes

14) EXPLAIN the purpose of the Trend package

15) BUILD a Mini Trend

16) DISPLAY a trend group

17) DESCRIBE the 5 types of trend layouts

18) BUILD and MODIFY multiple point trends, including adding shading (color)

19) BUILD and MODIFY trend groups

20) VIEW point values at various points on a trend

[ ---: _ _ _~

[

- p ~ ~ ~ - _ p _ _ _ _ - _ _ ~

~

Monday September 15, 2008 9:24:36 AM Page 15 of 87 ~

- _ _ ~~ ~ _ _ _ _ _ _ -

~~~

~~ ~~ ~~ ~

21) DlSPLAYa GbulartGnd

22) DESCRIBE the difference between Live and Histoncal trends

23) MANEUVER through the customized Salem Station process diagrams to display desired plant information c The effect each P-250 Computer control has upon P-250 Computer components and operation (NIA) d The plant conditions or permissives required for the P-250 Computer Control Room controls to perform their intended I

I -

Question S u m Comments:

~~~ __

Monday, September 15, 2008 9:24:36 AM Page 160f 87 ~

L-- -.

Given the following conditions:

~- Unit 2 is operating at 100% power.

- OHA B-I8,2C 125VDC CNTRL BUS VOLT LO, annunciates in the control room.

- Operators identify that the 2C 125VDC bus is deenergized.

IWhich of the following identifies why the ARP directs operators NOT to transfer Vital Busses and Distribution Cabinets to emergency DC

~ ~ ~

power?

-- ~

~~~~

~~ -

~~

~

~

~~

'Any undesired action has already occurred when power was lost, transferring power now would

- ~~~ lead to an additional

~

plant excursion.

- ~ - _ _ _ _ ~- _- ~~ ~~ -- __- ~ ~~

-___ ~~ -~ - ~ ~ - ___ ~~ __ - __ ~

~Asingle failure could result in cross connecting 2 vital power supplies required to be maintained 1 -_ ~ ~

separate.

~ - -

___ ~ ~~ ~_ ~- -__ ~ -_ ~~ -

-__- ~ _ _ _ -~ -~ ~ ~~ _-__ _ _ ~- -~ ~ ~- __~ _~

Overloading

-~

_ _ - ~ of the remaining

~ ~ -~ energized 125WDC busses

~ ~~ ~~

- ~ from _ _ _the _ _swapped ~ _

~

components. ~

y apply_to _ Loss of DC Power:

~ _ _ - ~ -~

~

~

~ -~

tained in EOP for loss

~ _ _ ___

_ ~~ ~- __ __ -__ 4.0 4 . 2

_~ ~ __ ~ ~

Ily in service to the DC busses comes from the 3 vital busses. If DC control power is transferred from deenergized 125VDC bus to its alternate, then the operability of the AC bus is affected, because a single failure could cause 2 AC busses to be x-connected land violate the separation requirement for Vital AC power. A is incorrect because while it may be true ithat re-energizing loads may cause component operation, it is not the reason for the precaution. C is iincorrect because the DC busses are riormally supplied from their respective chargers and have very low lloading associated with them under normal conditions. D is incorrect because the availability of DC

/Doweris not the same as eneraizina the trir, coil, which would not happen.

Overhead Annunciators Window B Leamind mjectids DCELECE014 Given a DC Electncal Svstem failure, oredlct t h e effect of the DC Electrical System fallure on the following: (Llcense Operator 1

and STA only)

Emergency Diesel Generators mponents-~ using DC control PO _

-~- ~~ __- ~

_ _ ~ _ _~ ~

1-Monday, September 15,2008 9:24:36 AM--

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

I Paae 170f 87 1

- - -~

,Why was the discovery of elevated levels of Tritium in groundwater surrounding-Salem-a concern, and where did it comefrom? ~~

~- -

Tritium is a weak beta emitter, so it can be h a k f Spent Fuel Pool had clogged drains which caused fuel pool liner leakage to escape through seismic

- ~ _ gaps _ _ _in_the ~ building.

-- ~ ~- ~ -~ ~~ _ _ _ ~ -

~-_____ - ~~~~ ~-~~ - ~- ~ ~~~ ~

'Tritium is a weak beta emitter, so it can be harmful if ingested by swallowing. The Salem Unit 12 ~-normal radioactive liquid release pathway had undetected under ground pipe leakage.

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

. - -~ ___-

tritium is a high energy gamma emitter, so it can be harmful if a person is near the area in

~ ~ ~~ ~ ~~ ~~

iwhich elevated tritium concentration is present. Salem Unit 1 Spent Fuel Pool had clogged idrains which caused fuel pool liner leakage to escape through seismic gaps in the building.

- -_____- ~~ -~ ~ _ _ _ -

~~ -~ - -___ ~~ -

Tritium is a high energy gamma emitter, so it can be harmful if a person is near the area in lwhich elevated tritium concentration is present. The Salem Unit 2 normal radioactive liquid irelease pathway had undetected under ground pipe leakage.

- ~ ~ _ _ _ _ _ ~ -

~~ - --

~ ~ _ _ -_ ___ - ~ - ~ ~~

~~~

b K L knowledge of the operation; implications of the following concepts as they apply to Accidental Liquid i Radwaste Re1eas

- - - ~ _ _ _ _ _ - -~ -- ~ -~ ~~~ -~

1 - 1 -~

lAK1.01 Types of radiation, their units of intensity and the location of the sources of radiation in a nuclear i 2.7,3.1 um groundwater contamination is a big concern in the nuclear industry. Salem has experienced tritium in the groundwater, and identified it as coming from a leak in the Unit 1 Spent Fuel Pool liner lcombined with clogged drains that allowed this water to back up and escape the building through "seismic gaps".. Tritium is a concern in the drinking water supply because it is a beta emitter, and can be ingested hv drinkina c o n t a m i n a t e d water.

Groundwater Protection - Data Collection Questionaire

~-_____--

RADCONE007 For plant or industry events associated with the Radiation Protection Program:

' A. Summarize each event B. Identify the root cause(s) of the everit C. Describe the events likelihood of occurrence at Salem Nuclear Generating Station D. Describe established or alternative actions which might prevent the events (re)occurrence at Salem Nuclear Generating Station

- ~ ~~~ ~~

_ _ _ _ _ ~ -

L n d a v . September 15.2008 9:24:36 AM Page 1 8 o f 8 7 I

\

iGiven the following conditions:

~- Unit 2 is in MODE 3, NOT and NOP I- A total loss of all Control Air occurs.

speed stop would not match the flow lost to the RCDT from the seal return relief valve lifting,

~~ - - ____

~ -

- ~ - - p - p p

_ _ _ _ -_____p p _ _ _

p

~

- _____~-.

pecific bases for t states, "Prior to commencing a cooldown, the RCS must be borated to cold shutdown conditions. Without Control Air, there is no way to reduce RCS inventory. Thus, the amount of boron that can be added before the cooldown commences will depend on the available space in the PZR. Due to the slow rise in level, this may become limiting. Therefore, the charging pump suction is '

transferred to the RWST early in the event. This ensures that any addition to the RCS is at RWST concentration. C is incorrect because the charging pump suction will auto swap (MOV's) to the RWST on lo lo level in the VCT. B is incorrect because while the recirc line does go back to the VCT, it is not the basis for the transfer. D is incorrect because seal return is not isolated on loss of air.

ABCAOlE002 Describe, in general terms, the actions taken in S2.OP-AB.CA-O001(q)and the bases for the actions in accordance with the Technical Bases Document.

-p______- --

r-I%nTay, Lppp September 15-08 9:24:36 AM

- Paae 19 of 87 '

~ ~~ ~

Which of the followina identifies the maior concern withthe primary automatic fire suppression IFlooding and subsequent loss of vital equipment.

~ ~~ ~_ _ - ~

electrocution from water contacting energized equipment.

-- ~ ~ -_ ~~ ~

~ ~ ~~~~

Long term health effects from brief exposure to extinguishing agent.

~000067K101

______~ ~ ~ _ _ _ _ _ ~____- ____- ~ _ _ _ _ ~

~ - _ _ _ _ ~p~ ~~ - _ ~ _ _ _ _ _ _ ~ _ _ ._ __ _ __~-_~

IB and C are incorrect because water is not used. D is incorrect because brief C 0 2 exposure is not associated with long term health effects, it ISthe near term death effect that is the concern.-

~ _ -

~-

_ _ ~_____

~ ~- ~ ~

FIRPROE004 Describe the function and operating charactenstics for the following Fire Protection System components:

Fire Barrier Components:

Fire Doors Fire Dampers Penetration Seals Fire Proofing Mannite Walls Energy Shields Protective Wraps and Coatings

b. Fire Detection Devices:

Ionization detector Thermal detector Smoke and Fire detectors

c. Fire Protection Subsystems:

Water Supply System Preaction Deluge System Wet-Pipe Sprinkler System Foam System Carbon Dioxide System Halon Svstem 1

Monday

-Lp Septembe7?5,2008 9:24:37 AM --

I Page20of87

IWhich of the following transients is analyzed toresu-ltin the highest containment pressure AND (greatest mass leakage out of containment?

~~ ~~ ~ ~ ~- ~~

Main Feedline break.

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

~~ ~~ ~ ~ ~~~~~ ~~

~~ ~~ ~~

f design basis Steam Line Break inside containment.

~~~ - -

~ ~ __

~ ~~

~~~

_ -~ _- - ____

lcontainment structures LOCA is the governing condition." This makes C and D incorrect. A PZR space ILOCA will be much smaller and have a much smaller effect on containment Dressure.

I ReferenceTitle

~

CONTMTEOOI Describe the purpose and design basis for the following Containment and Containment Support Systems subsystems:

Containments Containment Airlocks Containment Isolation System Containment Fan Cooler System Containment Iodine Removal System Rod Dnve Ventilation System Reactor Nozzle Support Ventilation System Reactor Shield Ventilation System Containment Pressure D Vacuum Relief System Hydrogen

~- Recombiner System -~~

~- ~-~ -

-~ ~~

~~~

I Monday, September 15,2008 9:24:37 AM Paae21 of87 1 I

Given the following conditions:

- Unit 1 has experienced a Rx trip due to a LOOP.

1- A Safety Injection occurred after the Rx trip.

1- Concurrent events have caused conditions to deteriorate to the point that a transition

~ to FRCC-1, Response to Inadequate Core Cooling has been made.

~

I IWhich of the following describes how RCS pressure will be lowered to the point where ECCS umulators will inject into the RCS?

- - - - - - ~ - - _

~- - -~ - _ - ~~ _ -

RCS pressure will be lowered

- - - ~ -- -~

using the PZR spray- valves,-

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

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

Intact SGs will be depressurized by dumping steam at maximum rate using MSIO Atmospheric IRelief valves.

-~ - - - ~ ___ -_ - - _ ~ ~ _ _ _ _ _ _ _ ~ _ _ _ _ _ ___ __

- - ____________ - - ~

~IntactSGswillbe depressurized by using the Main Steam Dumps in MS Pressure Control IMANUAL and dumping steam at maximum rate. - _ ~ -

- . _ _ _ _ _ ~ _ _ __

~ ~ ~ _ _ _ _ _

- _ ~ - -

~RCSpressure will be lowered by dumping steam using MSIO Atmospheric Relief valves while

- - ~ _ _ _ _ _ _ ~

determine and interpr --_

~ _ _ _ _ _ _ - _ _ _ _ _ _

-~

ldumps are the preferred method of dumping steam, but they are NOT available due to the loss of off-site

,power, which has caused no circulating water pumps to have power, thus removing the main condenser permissive to allow steam dump operation. The alternate path states to dump steam using MSIO atmospheric reliefs at maximum rate. While the maximum rate MAY cause a cooldown less than 100 degrees per hour, the cooldown rate is NOT ensured. That makes distracter D incorrect.

-- ~~~~~ -~ -

-_ - ~~ -

I Referenat ntle Response to Inadequate Core Cooling ~-

FRCCOOE003 Describe the plant response to actions taken in the following EOP step sequence(s).

A. 2-EOP-FRCC-1 Steps: 4, 5, 5.1, 6, 9, 13, 15, 15.3, 17, 18, 19, 25, 26, and 27

-2 Steps: 14, 16, and

--____________~ _ ~ ~~~

I

_ _ _ _ ~

p n d a y , S e p t e m b e r 15, 2008 9:24:37 AM I Paae22of87 1

I The crewis performing EOP-LOCA-2, POST LOCA COOLDOWN ANDDEPRESSURlfiTlON. 22 ICharging Pump has been stopped. Conditions are met for stopping one Safety Injection (SI) Pump.

r M i n d a y , September 15,2008 9:24:37 AM I Page23of87 1

- ~-

EOP-LOCA-6, LOCA Outside Containment directs actions to verify valve positions for only ONE Residual Heat Removal. Most likely source of leak.

[Chemical and Volume Control. Most likely source of leak.

~ _ _

LOCA06E001 Describe the EOP mitigation strategy for a LOCA OUTSIDE CONTAII

-~ -

~ ~~ ~~~~

---

-~

' Mondav. SeDtember 15. 2008 9:24:37 AM Page24of87

~-~ ~ _

Qz3 FRHS-1, Response to Lo&?of Secondary Heat Sink, Step 3 asks, "Is RCS pressure greaterthan

,ANY INTACT OR RUPTURED SG pressure" which of the following statements is correct if the operator answers NO?

~ - _

~~

-~ ~

__- -~

ilMMEDlATELY go to Step 23, Bleed and Feed Initiation, since there is no decay heat removal-ioccurring through -~

the SGs.

~ ~_ -~ __ ~- __- _ _ _ - - - ~ ~ ~ _ _ _~ _ _

~_____- ~ -_____ -~ __ _ _ ~ ~ _ _ ~

Return to Procedure in effect. Attempts to establish a secondary heat sink would be iineffective at reducing RCS L--- _. _____- -~ _ _ _ _temperature since SG pressure is higher than RCS pressure.

______ ~~~ _____ ~ --_____ -~~

~- - ~ _ _ ~ _ -__- -~ -~ ~- - ~~ ~_ __ __ ~

IReturn to Procedure in effect. The RCS has experienced a LOCA large enough such that a lsecondary heat ____ sink is NOT required because core decay heat is being removed

~~ ~~

~- ~ ~~ _ - ~~ _-by break flow. -

-~ ~ _ _ _ ~ ~~ _ ~ ~ - _ _ _ _ _ -~ ~

IIMMEDIATELY trip all RCPs to prevent further loss of reactor coolant through the LOCA, since ia LOOP later in the event could L--_- ~_ ~-

cause a more -~~

severe loss of coolant or two-phase

_~ ~~ ~-

RCS~ _ _flow.

~~

ndary Heat Sink and the foll

~ ~~

removal systems, and relations between the proper operation of these systems to the operation of the facilitv.

re i ta s to a need to nk. If pressure is below SG pressures, then a LOCA of sufficient size is present, and break flow will be removing decay heat, along with ECCs injection. Distracter A is incorrect because the criteria for going to Bleed and Feed is SG WR level. Distracter B is incorrect because a secondary heat sink could actually be established, and could reduce RCS temperature by

- _ _ _ steam from SG. Distracter D is incorrect ~-

,dumping ~- -~ - ~ - ~-

- ~ because it is the RCP trip criteria for a SBLOCA.

~~ ~~ - - ~~

-~

I Reference Title I Loss of Secondary Heat Sink ~ ~~_

FRHSOOEOIO Describe the basis for each step, caution, and note, in 2-EOP-FRHS-1 thru 5 Monday, September 15,2008 9:24:37 AM

_ _ _ _ _ ~ - -______

~

Page25of87 I

x . . _. . -..,.," . , I^ . , . . ._.,

Given the following conditions:

1- Unit 2 is in Mode 3 following a planned shutdown after a 300 day run.

- A transformer fault results in the total loss of off,-sitepower.

'15 minutes after the transformer fault, with NO operator action, the following indications are present:

- All RCS WR Thot's are 559°F and rising slowly.

- All RCS WR Tcold's are 547°F and stable.

I- All SG pressures are I010 psig and rising very slowly.

- All SG NR levels are 39% and stable.

- PZR level is 23% and rising slowly.

Which statement describes the status of the RCS and action(s), if any, that must be performed by p the control room IAW S2.0P-AB.RC-0004 NATURAL CIRCULATION?

I b R Comprehension- 812512008 OOWE09AI03--

24

~- -- ~~ -~ -

monitor the following asth

~- ~~ ~

Its during abnormal and em AB.RC-0004 step 3.6 identifies ALL the con met for natural circulation to be I occurring. With RCS Thots still rising, it is NOT occurring. Step 3.7 directs the operator to feed the SGs to maintain them within +/- 5% of programmed band. Programmed band plus 5% = 38%,so feeding in this situation is not directed. Steam dumping is directed to maintain or lower CET temps.

L---p ABRC04E002 a) Determine the appropnate abnormal procedure.

b) Describe the plant response to actions taken in the abnormal procedure.

) Q al bteritsl %qukedfw maminritlon 1 Bank Vision Q80355 p n 2 a y , September 15,2008 9:24:37 AM- I Page26of87 1

QZC Given the following conditions forunit 1:

1- A reactor trip and SI occurred at 0700 1- RHR system problems resulted in a loss of recirculation capability

- Current time is 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> prior to entering EOP-LOCA-5, Loss of Emergency Recirculation, the following conditions were present:

RCS subcooling = +1O"F All RCPs are secured 11 and 12 Charging Pumps are running with equal flow, and each pumps flow remains stable.

BIT flow - 350 gpm RVLIS full range 95%

11 SI Pump flow - 100 gpm 12 SI Pump flow - 100 gpm Containment pressure 4.9 psig Which of the following identifies the ECCS pumps that should be run following determination of minimum SI flow for decay heat removal?

d 1- R 7 1 Application 8/25/2008

~ _ _~- ~~ -~ ~ -- ~~ ~ _ _ _ _ - __

of the interrelations between Loss of Emergency Coolant RGirculation and the following:

~

1 an la --

4.3

~

removal systems, and relations between the proper operation of these systems to the operation of the facility.

-~ ~ ~~

NE centrifugal, a mps will b e ~

reduced to ONE. Starting at Step 19 of LOCA-5, with RCP's secured with 6 0 degrees subcooling, will use Figure A to determine the ECCS flow required vs. time after trip. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> equals 360 minutes, which is -225 gpm. With the stem stating that charging pump flows remain the same, you need both the 175 gpm and EITHER SI pump flow flow. Both SI pumps alone would provide 200 gpm. The 200 gpm line clearly crosses the time line past 400 minutes.

1 Rsaetance Title

~~ _.__- -_ Page27of87 1 Monday, September 15, 2008 9:24:37 AM ~

IFacilitv Exam Bank IQWstion Modification Method: IEditorially Modified

~ _ _ _ ~ _ _ _ ~

mdav.<emba5.:24:37 AM I Paae28of87 1

-~

IGGen the following conditions:

Unit 1 has initiated a MANUAL Rx trip and SI on a large Steam leak.

MSLl has succeeded in closing 14MS167 ONLY.

11-13 SG pressures are all -710 psig and dropping.

14 SG pressure is 830 psig and rising.

Total AFW flow is 24E4, with -6E4 to each SG.

ALL SG NR levels are off-scale low.

1-EOP-TRIP-1, Reactor Trip or Safety Injection, is in effect.

lOOWE12K302 26 b e a m Generators:

TOTAL AFW flow >22E4 while the 3 faulted SG's are being isolated. There is no step in LOSC-1 if student looks ahead for AFW flow requirements, but the next procedure in line, LOCA-1 states to maintain

- . >22E4 until at least one SG NR level is >9%. While the stem does not state where in TRIP 1

_~

the operators are, the CAS for maintaining AFW flow is at step 20, before any transition point.

_____- ~~ -_ -

~ -~ ~- ~~ ~~ ~ _

Reactor 15,2008 9:24:37 AM M o n d a y , September- Page29of87 1

~ - ~ - ($27 How do loop flow and core flowdiffer when operating THREE RCPs as compared to operating

'FOUR RCPs?

-- RCPUMPEOIS Chemical and Voiume Control System

- p -

Component Cooling Water System Service Water System Containment Isolation Signal Reactor Coolant Drain Tank Reactor Coolant System Reactor Coolant Pump lube oil Reactor Coolant Pump seal system p -

- - p - - - p - p - - -

- - - p p - - - - -- - - - --_-_-- p p -- - -- - -

_~________-

L-Ipp----

Monday, September 15, 2008 9:24:37 AM 1 Page 30 of 87

~ 1

IGiven the following conditions:

1- Unit 1 is in Mode 6.

/- 11 RCP Motor is uncoupled from the pump.

'- RCS Loop 11 is full.

1- Maintenance is working on the 11 RCP pump.

'Which of the following - -

describes how leakage of reactor coolant up the RCP shaft is minimized?

- - - - ~ ~~ _~

~-__

~-

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

'Seal

~-

injection flow which is maintained-during this

_ ~- -_ ~_ -~

condition.

- ~~ - -~

~ ~

~ -~ ~~

~~

~ ~ _ -

[Backseating-~ ~~ the pump shaft with

- ~

the top

~_ _ _of the~ thermal

- - barrier assembly.

~ - - - ~ - ~~ _ _ _

-- -~ ~ __ ~ ~ - ~ - ~ ~- _ ~- ~- -- ~ ~ ~ ~ - ~ ~ - -- ~ - _

~ ~

nozzle dam installation prevents RCS water from entering the RCP

-~ _ ~

- ~ -~ shaft~-

~ area. _ ~ ~~

- ~~ ~- ~- ~ _~

~ _ ~ _~_ ~ - ~- - -~ ~~ ~ -~

-~ - t back to the RCDT.

ctor Coolant Pump Sys When the RCP pump is uncoupled from th accomplished by lowering the RCP shaft -1 inch, which allows the top of the shaft to mate with the top of the Thermal Barrier. This will reduce the leakage up the shaft from 5-10 gpm to -1 gpm, which is sufficiently low enough to keep the water level below the #I seal runner. A is incorrect because seal injection is NOT in service. B is correct as described above. C is incorrect because nozzle dams are 1 mot used to isolate and drain piping. D is incorrect because the leakage up the shaft is collected and LearnindlOb/ealv?ss RCPUMPEOO4 LOR NCT Describe the function of the following components and how their normal and abnormal operation affects the Reactor Coolant Pump:

Impeller Turning Vane Diffuser Diffuser Adapter Thermal Barrier and Heat Exchanger Pump Radial Bearing Controlled Leakage Seal Assembly Lower Motor Radial Bearing Upper Motor Radial Bearing Flywheel Anti-Reverse Rotation Device Oil Lift Pump

~~ ~

~~

~~- ~ ~ -

Bank IEditonally Modified VISION Q41815 m o n d a v . September 15, 2008 9:24:37AM 1 Page31 of87 '

~ ~~

'During Unit 2 steady state MODE 1 operation, the Chemistry Technician reports that thelevel of Fluoride in the RCS is elevated at the CVCS demin outlet. A second confirmed sample places the lunit in an Action Level 1 per CY-AP-120-100, Reactor Coolant System Chemistry.

I Which of the following identifies the major concern with continued operation at this level, and what quired course of action? _ __ _

~-

~ ~~ ~ ~

~

~~ ___ ~ _ _ ~ ~ _ _ ~~~

corrosion of RCS and comoonents. IMMEDIATELY commence a load reduction at

'minute and trip the Rx when power reaches 20%.

~~ ~

~ ~ _- ~ -~ _ -~

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

corrosion of RCS and components. If concentration can NOT be reduced below the Action iLevel 1 limit within 7 days, a Technical Review shall be performed and a program for implementing corrective measures instituted.

~~ ~ ~ _~ ~~ ~_ _ ~ _ _ ~~- ~

~- ~~ ~ ___

__ __ ~_ - _ ~ -~ ~ ~- ~ _ _

___ ~ -~ __ ~_ ___ ~

,Accelerated depletion of CVCS cation demin resin. Reduce power 50% within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> icommence a load reduction at 5% / minute and trip the Rx when power reaches 20%.

~~ ~ _ ~ _ ~ ~ ~~~ ___ ~_

~ ~ ~~~ ~~ ~ _ _ -_ __

_~ _ _ _ ~~~ ~ _ _ ~ ~ ~ ~_ ~~~ ~~~ ~~ ~ _ _

_ _ ~

Accelerated depletion of CVCS cation demin resin. If concentration can NOT be reduced below the Action Level 1 limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, initiate an orderly unit shutdown and cooldown

~- ~ ___ -__

lbased on those predictions, use procedures to correct, control, or mitigate the consequences of those

.5 delineates the reason as.. .."The chemistry limits and action levels presented herein are appropriate for protecting system materials, ensuring fuel performance, and controlling radiation field buildup." In I general, there are 3 action levels associated with chemistry at Salem. Action Level 1 is a condition where ichemistry is elevated outside the norm, and may have long term adverse consequence, and only requires lmonitoring and attempts at correction. Action level 2 is declared at levels which, if allowed to continue indefinitely, would lead to increased incidence rates of corrosion, and require a unit shutdown if not corrected within a certain period of time. Action Level 3 is a condition where chemistry is well beyond the Iboundary at which accelerated corrosion will occur, and requires prompt action to shutdown and cooldown the plant below 250 degrees. A is incorrect because Action Level 1 does not require a unit ishutdown. B is correct. C and D are both incorrect because although increased levels of impurities in the RCS will dedete CVCS demins more rapidly, it is an effect of the elevated levels, not the concern with

'the elevated levels, and the actions

~_ ~ _~ ___ - ~~

are incorrect.

_~ __ ~ ~ _ ~ ~ ~ ~ ~ ~- ~

Refueling Water Storage Tank Rod Control System Pressurizer Relief Tank Chemical and Volume Control System Reactor vessel Level Indication System

~ _ _ _ _- - _ ~ _ _ _ _ _ _ _ _ _ _ ~ ~

[ I

~~

Monday, September 15, 2008 9:24:37 AM Page32of87 L

~

Emergency Core Cooling System Pressurizer Reactor Coolant Drain Tank Main and Auxiliary Feedwater Systems Nuclear Instrumentation Reactor coolant Pumps Spent Fuel Pool Purification Refueling Canal Refueling Water Purification Main TurbinelGenerator

~ _ _ _ _ _ _ _ ~ -____

~ ~ ~ _ _ ~ _ _ _ _ ~ ~~~

~~ ~~

~

-- ~~~ ~ -~~~ _ _ _ ~~~ ~ _ _

~~

kWerial Rsqdrcid for Examination 1

QuestionSoum: 1 New 1 I Question Source Comments:

r r

Monday,

~~~ -

September 15, 2008 9:24:37 AM--

-__--- Page33of87 I

Given the following conditions:

-630 I- Unit 2 is operating normally at 100% power-.

~- 2CC71, LETDOWN HEAT EXCHANGER TEMPERATURE CONTROL VALVE, fails to l the full closed position due to its temperature sensor failing low.

r flow through the Letdown HX, letdown fluid temperature will rise. The temperature sensor that controls the CC71 (2TE130A) is the same one that actuates the I

2SV496, which is what controls 2CV21 to bypass the demins. VCT temperature will rise, and available INPSH to the CVCS pumps will lower. C is incorrect because the interlock between the CV7 and CC71 is the opposite. CV7 closing will shut the CC71. D is incorrect because the same temperature sensor is used for CC71 and demin bypass valve CV22. With the sensor failing LOW, the CV21 will never divert lthe hotter letdown fluid pastthe demineralizer.

- ~ p p - ~ ~ pp -p ~ -__ p- - - - ---

etdown, and Seal Injection p p -

~ p -p~

failure on the following: (License Operator and STA only)

Automatic Rod Control Component Cooling Water System Reactor Coolant Pumps Pressurizer Level Control System Reactor Coolant System Pressurizer Pressure Control System I

@d Facilitv Exam Bank fr;

- ~-

L

- p Monday,

.p p - ~ ~

September 15, 2008 9:24:37 Ah4

~ ~ - p p -

I Paae34of87 1

IGiven the following conditions:

431

'- Unit 1 is preparing to initiate RHR in shutdown cooling mode during a late-cycle forced outage to MODE 5 two months before a scheduled refueling outage.

- The RCS has been borated to the required CSD boron concentration as shown in 1 S I .RE-RA.ZZ-0016, Curve Book.

- The RHR system was last in service during a forced outage at BOL.

'If RHR system boron concentration is NOT adjusted during sampling, which of the following describes how RCS boron concentration will be affected when the RHR system is placed in Shutdown Cooling?

RCS boron concentration..

R Comprehension Salem 1 & 2 8/25/2008

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

- ~ ~ ~~ -

E.

-~ Knowledge of the operational implications of thefollowing concepts as they apply to the Residual Heat

~- -

3 . 2 -3.4 ill ich is ill be, since it was not in service since being taken out of service following the previous shutdown. The CSD boron concentration for this shutdown 2 months before refueling is 900 ppm at a -

core burnup of -1 1,300 EFPH. (SI .RE-RAZZ-0016, Page 121, Table A) Without adjusting RHR boron Iconcentration, placing RHR in service with the previous HIGHER concentration will BORATE the RCS

'further. The candidate does NOT need to know exact numbers, so no procedure is provided. They DO need to know that required CSD boron concentration LOWERS over core life. The last distracter is Component Cooling Water System

~~~ ~~~

Pressurizer Spent Fuel Pool Cooling Chemical Volume Control System Reactor Coolant Pumps Emergency Core Cooling System Service Water System Refueling Water Storage Tank ~

~ ~~ ~~- ~ ~- ~ _ _ ~ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~

LMonday,

~

_ September

_ - l15,2008

__ _AM_ _ _ - ' ~Page35of87 9:24:37 ~L 1

~ _ _ _

LMonday I I September L . 15,2008 9:24:37 AM 1 Page36of87

Which of the following running Safety Injection pump discharge flows is conistent with the RCS pressure shown during a LOCA?

Application I &2 006000A305 006 Emergency Core Cooling System 32

~ _ _ _ __

a. /Ability to monitor i~ ~-

~ _ _ _

~~ ~

~~ ~ _ _ ~

tions of the Emergency Core Cooling System including:

~

~~~ ~ ~

A3.05 Safety lnjec

~_ ~~ ~~~ -~ __ ~ -- _ _ ~ ~~_

The pump curve for SI pumps in contained in ergency Plan document. The cand responsible for the document, but are responsible for knowing the basic pump characteristics, including the pump curve. A is incorrect because the pump shutoff head is -1,520 psig, so there should be no flow iat 1,765 psig. (1765 psig is AUTO SI initiation pressure) Distracter C is incorrect because runout flow is 1650 gpm, but should not be present at 1000 psig, as runout pressure is 650 psig. Distracter D is below runout pressure, but runout flow. B is correct because it is the only pressure flow combination that falls Lecuning Objectivss ECCS03E008 Identify and describe the Control Room controls, indications. and alarms associated with the Emergency Core Cooling System, including. (Licensed Operator & STA only)

The Control Room location of Emergency Core Cooling System control bezels and indications.

The fiinction of each Emergency Core Cooling System Control Room control and indication.

The effect each Emergency Core Cooling System control has upon Containment Spray System components and operation.

The plant conditions or permissives required for Emergency Core Cooling System Control Room controls to perform their intended

_ - ~

function ~-- ~ -

- ~~- ~_ _ --

~

~ _~ ~~

~~~

~~

I

_ Monday,

~

~ _ _ 15, September

_ _ - _ ~ 2008

_ _ 9:24:37

_ AM

Given the following conditions:

Q33 1- Unit 2 is operating at 100% power when a LBLOCA occurs.

~- When 21 RHR pump starts, the pump mechanical seal fails.

Which of the following describes the effect this will have on 21 RHR pump and its ability to perform lits ECCS function IAW Salem FSAR?

Salem 1 & 2 8/25/2008

'RHR Pump shaft seal failure that will generate leakage not to exceed 50 gpm.) B is incorrect because the seal leak is not analyzed for the injection phase because cool RWST water is flowing through the pump as it is being injected to the RCS and the pump will NOT overheat. C and D are incorrect because the FSAR analyzed leak does not affect the function of the RHR pump to perform its design function. The IRHR pumps are provided mechanical seal cooling from the CCW system. Salem FSAR, Section 16.3.2.11, page 6.3-42, discusses the seal failure during recirculation. It is postulated to be 6 0 gpm, and

,the RHR sump is sized to accommodate it for 30 minutes. Subsequent to that, the leaking seal is expected to be isolated by operator action, and is NOT expected to impact the recirculation phase of ECCS. The seal leak is not analyzed for the injection phase because cool RWST water is flowing

~ ~- ~

ECCSOOE002 Describe the desian bases of the Emerqency Core Coolinq System. (Licensed Operator & STA only)

El---_-----___--

Monday, September 15, 2008 9:24:37 AM - I Page38of87

-~ -

Monday,

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

~

September 15,2008 9:24:37AM _ _ - _

Page 39 of 87

Removing the PRT rupture disc during a refueling outage.

IDrawing a bubblein the PZR without a vacuum in the RCS.

~- -

,In service charging pump relief valve lifts and remains open.

~~ -~ ___ _ _ _ ~

&2 8/25/2008

-~

__1 007000K502 Pressurizer Relief TankfQuench Tank System --___~

34 nowledge of the operational implications of the following concepts as they apply to the Pressurizer Relief TanklQuench Tank System:

~~

-~ - _ _ ~ __

-~

~ ~~ ~

K5.02 Method of forming a steam bubble in the P flowpath would exi PRT. D is incorrect because the relief valve I S directed to the containment sump. B is incorrect because the PRT must be at atmospheric conditions to remove the rupture disc. C is correct because PZR level is raised until PRT level rises -~ ~~

LearningObjectives

- __ - .- __ - .__ _ _ -- - -.-- - - - I PZRPRTEOf2-- NCT DISCUSS the procedural requirements assGciatedwith the PressurizerandPiessurker R e G j Tank. incluing an explanation

_ _ _ ~ ~ _ _ _

r G n d a y , September 15,2008 9:24:37 AM I Page40of87 1

Q35 Unit 2 control room has been evacuated duringa Security Event.

~ ~ ~~~ ~~~ ~ ~ ~~ ~ ~ ~ ~p~ ~ -~

IReactor Coolant Pumps to allow restoration of forced flow cooling of the RCS.

~~ ~

~~ ~~

p~

Station Air Compressors to ensure positive control of AOV'sreqiredfor AppendixR safe ~

h shutdown.

L ~~ p~

~~

~~ -~ ~~ ~ p~

~~~ ~

4KV Vital Bus lnfeed Breakers to ensure EDG output breakers can be shut on a deenergized lvital bus.

p~ ~~

~-~ ~~p the CCW pumps.

I Refemce ntla w X" I

_ _ _ _ _ _ ~ - - ~

1 Monday, September 15,2008 9:24:37AM---

L- - -~ ~-

1 Page41 of87

-~ ~

,Given the followingconditions: ~

- RCS temperature - 547°F.

~~~~ -______~ - ~- ~~

K3.-- IKnowledge of the effect that a loss or malfunctionofthe Pressurizer Pressure Control System will have on because RCS pressure will rise and open a PORV, but that will cause pressure to drop rapidly, not slowly. C and D are incorrect because the action is the opposite of what will happen, I.e, spray valves will ABPZRI EO01 Describe oDeration of the Pressurizer Pressure control system as applied to S2.0P-AB PZR-0001(a).

hutr8tiOn Source: Facility Exam Bank Question WrceComments: Vlsion Q75902 Monday, September L-_____-. 15, 2008 9:24:37

___ _ _ AM_ _ ~

Page42of87 1

Given the following conditions:

Q37

- Operators are responding to an inadvertent SI from 100% power.

- Equipment malfunctions have severely slowed operator progression through the EOP network.

1- Operators are at the step to transition into TRIP-3.

1- PZR level is 93% and rising slowly.

IWhich of the following describes how FRCI-1, Response to High Pressurizer Level, should be utilized for this situation?

New Rev. 2 Supplement 1 Actual-KlA values are RO:3.4, SR YELLOW path FRP, and as such is entered at the CRS direction, and is not REQUIRED to be entered.

Candidate must recognize it is a YELLOW path to discount C and D. The condition IS met for entry with PZR level > 92%. The first step in FRCl asks if the SI pumps are running, and if so, returns to procedure in effect. Since Yellow paths are not required to be entered when entry conditions are met, it is wrong to enter it and then leave it within a step. With the SI pumps in service, normal charging and letdown cannot be established.

I 1 V" 1 Reference title 1

~~

Monday, September 15,2008 9:24:37 AM I Page43of87  ;

Given the following conditions:

~- ~~ - - Q38 1- Unit 1 is operating at 30% power.

- Rod control is in MANUAL.

- I&C is performing a Sensor Calibration on 1PT-456 PZR pressure Channel I I (two) 1- The channel has been removed from service and all applicable RPS bistables have been placed in the tripped condition IAW S I .IC-SC.RCP-0018, 1PT-456 Pressurizer 1 Pressure Protection Channel II.

I ng identifies a condition which would cause an automatic Rx trip to occur? ~~

~~ _~ - -~ -~ ~ ~ - ~- - - - _ _ -

-~ RCS ~ loop 11 Tc fails LOW.

-__ ____ ~_ ~ ~ - - - ~ ~ ~- -

r - ~ ~ ~- -___ _ _ _ _ - ~ -~ - ~ -~ --- ~ ~- - -- ~ -- __ ~ __

lp-- ressurizer Pressure

_- - ~ - -~

Channel IV (four)

-~ ~ - - ~

fails LOW

-- ~ -L -- -~ ~ ~~~ ~ __

- _ _ ~ __ -~ -- ~ - - ~ -~ ~ ~

The front door of SSPS Protection Racks #I 2 and # I 3 for CHANNEL Ill (three) is opened by Imista L- - ke -:~- ~ -__ __ ~ ~ -~ -- ~- ~_ ~ __ -- ~ ~~ ---

Salem 1 & 2 8/25/2008 channel is taken out of service for testing that all RPS system bistable are placed in their tripped lcondition, which is stated in stem. A is incorrect because a single RCP tripping <P-8 (36%) will not icause a trip, the coincidence is 2/4. B is incorrect because Tc failing low (510 deg) would cause Loop 11 Tavg to lower, and this would put this channel further from a trip setpoint. enough to trip the bistable for OT/DT or OP/DT. C is correct because the channel IV of PZR pressure failing low would satisfy the 2/4 PZR low pressure Rx trip. Also, with rod control in manual as stated in the stem, autioneered high Tavg rising would NOT cause inward rod movement, which it would if rods were in auto, and eventually cause a Rx trip because rods would never stop inserting. D is incorrect because: SSPS rack door iannunciates an OHA in control room. Placing 2 trains of SSPS in test at the same time causes a General Warning on both trains, which causes an automatic Rx trip. When testing a single component, the entire train is not placed in test, but knowledge of how the 2 trains interface makes the door opening a plausible distracter.

I Y'W" ReferisntxTltle ' ' 7 Solid State Protection System Train A Functional Test

~ _- ~ ~ -

RXPROTEOIJ

- -~

Describe the differentiate between the followinq - of RPS testing: (Licensed Operator and STA Only)

- types a) Analog testing M o n d a y , September 15, 2008 9:24:37 AM I Page44of87 1

tion on 1PT-456 PZR pressure Channel I1 (two) service and all applicable RPS bistables have IAW S I .IC-SC.RCP-0018, 1PT-456 Pressurizer C R Memory '\ Salem 1 & 2 System

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

K4. Knowledge of Reactor F'rotection System desig ) which provide for the tripping <P-8 (36%) will not w (510 deg) would cause Loop

SSPS rack door mponent, the entire opening a plausible Solid State Protection System Train A Functional Test

.- I Monday, September 15,

_ _ _ _2008

~ _ _ _ - 9:24:37 AM i Page45of87- 1

Which of the following Rx trips is designed to prevent exceeding local-power density limits for ;he Over Power Delta Temperature. (OPIDT)

IRCS LOSSof Flow >P-8 o the Reactor Protection control system actuations:

a) OT DT Reactor Tnp, Rod Block, and Turbine Runback b) OP DT Reactor Tnp, Rod Block and Turbine Runback c) P-12 d) Feedwater Interlock e) High Steamline Flow Safety Injection

- ~~

~

~ ~~~ ~~~ ~~~

~-

RCTEMPEOIO Is, ala&<, andlndications associated with the Reactor Cool Instrumentation System, including:

~~~~~

a) The Control Room location of Reactor Coolant Temperature Instrumentation System control bezels, alarms, and indlcations b) The function of each Reactor Coolant Temperature Instrumentation System Control Room control and indication tpoints associated with the Reactor Coo I ~ -~

Monday, September 15, 2008 9:24:37 AM I Page46of87 ,

IGiven the following condition:

1- Unit Iis operating at 100% power.

1- A spurious SI signal is generated in the RPS system.

~- All systems function as expected for this condition.

1- AFW is reduced to 22E4 Ibm/hr right after the Immediate actions of TRIP-I are complete.

~~ ~_ ~ ~ ~ ~ _~ - -~ ~_ ~ ~_~ ~~~ ~ - _

nitor changes in parameters associated with operating the Engineered Safety ystem controls including:

mount of cold RWST water being pumped into the RCS will have a very small effect on temperature, since the RCPs will still be running and decay heat being generated. 603 degrees is the 100% power Thot. Tavg of 547 degrees is the setpoint of the Plant Trip controller for steam dumps.

There will be a delta between Th and Tc, so Thot won't be right at 547. There will be very little power generation 5 minutes after the trip. 530 would be a representative temperature for Th if a large injection 1 of RWST water were being injected with the RCS at a much lower pressure. 555 is representative of

-1 0% power generation.

~ ~ ~~ _-

Rx Trip o

~- ~

RCTEMPEOOS

~ -~ _ _ Identify and describe the local controls and indications associated with the Reactor Coolant Temperature Instrumentation, including:

a) Not applicable to this lesson Monday, September 15, 2008 9:24:37 AM Page47of87 ~

Q4 I IGiven-the following conditions:

- Reactor Power is 75%

- A failure of control rods to move in AUTO or MANUAL has occurred.

~ and Position Indication Systems:

Rod Cluster Control Assembly (RCCA)

Control Rod Dnve Mechanism (CRDM)

Rod Drive MG Sets Reactor Tnp and Tnp Bypass breakers Reactor Control Unit Power Cabinets Logic Cabinet components Pulser Master Cycler Slave Cyclers Bank Overlap Unit h DC Hold Cabinet i Rod Position Indicator (RPI) Coils j Signal Conditioning Modules

k. PulsehtoOAnalog (P to A) Converters I. Rod Bottom Bistables m Rod Insertion Limit Comparator Mawial Requlriidfor Examination 1 Question hwax I Faulitv Exam Bank ifie 1 - I QuestionSourn Comments:

6ndazbx5.2008:24:37

~

AM Paqe48of87 1

Given the following conditions:

'- Unit 2 is operating at 100.0% indicated NI power, and 3459 MWth indicated calorimetric power.

1- Tave-Tref deviation is zero.

- A high level is detected in the 25A feedwater heater.

I

'Assuming all expected automatic actions occur, which of the following indications will be present in lthe Control Room 2 minutes after the high level is detected? ~~~

- ~-~ ~ --

~ ~ ~~ ~

-~ -~

k Lo Lvl alarm wi- - -~ ~~ ~

-~

~ - ~ P ~ ~ ~ ~ ~~ ~~

~~ ~~~ ~ _ _ ~ ~

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

IRC Loops Tave-Tref L- ~ P P P P ~ - P -

deviation alarm

~- ~ ~~ --

will be locked in on 2CC2. -

~ ~~ ~ P~ ~ ~ ~ ~ ~ ~ -_ -~ P~

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

NI power will read lower than actual power due to colder feedwater entering the steam lgenerators.

~- ~~ -~ ~~ ~ ~ ~~ - ~ ~~ ~ ~ ~ ~ -~ -~ ~ P - ~ -~ -~ ~ ~-

~ - -~ ~- ~ -~ - ~ ~ ~P~ ~ - ~- - ~- ~~ ~~ ~~ ~- -

IN1 power will read greater than calorimetric power due to the lower FW temp used in Icalorimetric

-~ ~- - calculation.

~~ ~ ~~ - -~ --

~ -~ - ~

~~ ~

I Instrumentation System controls including:

ita allow less thermal neutrons to leak from the core and be seen by the Nl's. B is incorrect because with the deviation at 0.0 prior to the event, the change in temperature will not be great enough to cause the I alarm tu cume in at 3 degrees deviation. A is incorrect because the HDT level will rise, not lower, and Learning Objectives EXCOREEz9 ldentifv anddescribethe Control Room controls, indications, andalarms associated wlth the Excore Nuclear lnstrumentatlon System, including:

The Control Room location of Excore Nuclear Instrumentation System control bezels and indications.

The function of each Excore Nuclear Instrumentation System Control Room control and indication.

The effect each Excore Nuclear Instrumentation System control has upon Excore Nuclear Instrumentation System components and operation.

The plant conditions or permissives required for Excore Nuclear Instrumentation System Control Room controls to perform their intended function.

The setpoints associated with the Excore Nuclear lnstrumentatlon System control room alarms. ~

~

EXCOREEOOI Describe the DurDose of the Excore Nuclear Instrumentation System.

1 Paoe49of87

043 c=:-. Monday, September -~

15, 2008 9:24:37 AM - I Page50of87 1

Given the following conditions: -

- Unit 1 initiated a manual trip and safety injection coincident with a loss of off-site power.

- I C EDG failed to start.

1- Prior to the event, 11-14 CFCUs were running in HIGH speed.

8/25/2008 K201

~

45

~ ~ -~ _~ __ ~~ - -~ ~~ ~~ - - ~ _- ~~ - ~ ~- -~ ~

ies to the following:

-_ ~ -- _~ ~ ~~

he 11-15 CFCUs are powered from A,B,C,B,C 460 volt vital busses respectively. AL LOW start signal upon a MODE Ill (SI plus Blackout) SEC initiation. With C bus deenergized , only 1,12, and 14 CFCUs will be running. With 5 CFCUs to choose from, there had to be 2 CFCUs that were ,

in 3 of the choices, as opposed to the remaining 3 CFCUs, ~ - ~ which were only selected in 2-~

~- -~ ~~ ~- ~ of the choices. '

~~ - ~ -

CONTMTE004 State the power supply to the following Containment and Containment Support Systems components, including voltage level and 1EINon 1E containment Fan Cooling Units, including breaker alignment for Fast and Slow speed.

Containment Iodine Removal Fans (Licensed Operator & STA only)

Control Rod Drive Ventilation Fans (Licensed Operator & STA only)

Reactor Nozzle Support Ventilation Fans (Licensed Operator & STA only)

Reactor Shield Ventilation Fans (Licensed Operator & STA only)

Hydrogen Recombiners ed Operator & STA only) _

~

~~~~

~~

_ _ ~ __ -

~ - - -~

1 Concept Used

~- ~_

L

~

_- ~ _ _ -

Monday, September 15, 2008 9:24:37 AM--

__  : , Page51 of87

~

given the following conditions:

- Unit 2 has experienced a Large Break LOCA.

1-- RWST level has reached the semi-automatic swapover setpoint.

21 RH4 does not shut after 21SJ44 opens.

with the RWST at 8, which of the following identifies how this failure will affect Containment Spray Iflow as compared to pre-swapover flow, and how will the failure be addressed in EOP-LOCA-3, I

Transfer to Cold Leg Recirculation?

~~~~ __ ~~ ~~ __ ~~ - ~~ ~ ~ ~ -~ _ _ ~ ~ -

lower due to direct flow from the RWST to the containment sump. Operators will manually close 21RH4 at the valve.

~~~ ~~~~ ~~~ ~ ~~ ~~ ~~ ~~ ~ ~~ ~ -__ ~

~~~ ~~~~~ ~ ~~ -- -~ ~~ ~~ ~~ ~~~ ~~ __ -

stay essentially the same. Operators will shut the 2SJ69 to isolate the flowpath from the RWST to the RHR pump suction.

L - ~ _ _ ~~~~ ~~~

~ ~~ -~

~

~

__ ~~~~ ~- ~~~ __ -~ ~ ~~ ~ - ~~~ ~~~ _ _ ~

rise or lower depending on containment pressure. Operators will shut the 2SJ69 to isolate the flowpath from the RWST to the RHR

~~~~ _____ ~ __ pump suction.

~~ - __ ~~ ~~ ~~ ~- ~~ __ _____-

- ~~ ~ ~- ~

~~ ~ ~ ~~ ~

~~ ~~~ ~~

A2. Ability to (a) predict the impacts ofthe following on the Containment Spray System and (b) based on those predictions, use procedures to correct, control, or mitigate the conse nces of those abnormal operation:

-~ ~~~ __ ~~~

will ithout connection through the SJ69 supply to the RHR pumps. C is correct because CS flow will not be impacted by the valve malfunction. Step 5 of LOCA-3 directs that when the SJ44s are open Ithe SJ69 is shut. reaardless of RH4 position. D is incorrect because the auto transfer failure will not 80 95 nent operation for each step in 2-EOP-LOCA-3 (Material Required for Examination 1 Mondav. Seotember 15, 2008 9124137AM I Paae52of87 1

~- ~

IWhich of the following identifiesthe component(s) used for gaseous iodine removal from containment atmosphere?

~ -- ~

llodine Removal Units during accident conditions and during normal conditions.

-~ ~~

Containment Spray during accident conditions, and Iodine Removal Units during normal conditions

~ ~~

IContainment Spray and Iodine Removal Units during accident conditions, and neither during inormal conditions. ~ ~ - ~ ~ ~- ~- - ~

Containment Spray and Iodine Removal Units during accident conditions, and Iodine Removal Units during normal conditions.

~ ~~ ~~ -

&2 8/25/2008 l027000K101 Containment Iodine Removal System 47

~~ ~- __ ~~ ~~ ~~ ~ --- ~ ~

.1 K IKnowledge of the physical connections andlor cause-effect relationships between C stem and the following:

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

~- ____ ____ ~-

~~ -~ ____

• 3.7*

Containment Spray system op containment atmosphere. There is no direction in the work to operate Iodine Removal Units.

IRUs would be DlaCed in service in non-accident conditions at direction of Radiation Protection upon Containment Ventilatio n Containment Support Systems:

Containment Fan Cooler System Containment Iodine Removal System Rod Drive Ventilation System Reactor Nozzle Support Ventilation System Reactor Shield Ventilation System containment Pressure Q Vacuum Relief System

-~ --_ ~ -~

~~- - -~ - ~-~ -

QuestionSource Commnti:

~ - - ~

Qusstion'ModiRcationMethod:

Vision Q29202 Slightly changed stem from "mechanisms" to "component(s)"

~ ~-

' Eztorially Modified-

~ ~- ~~

M o n d a y~ , September

_ - _ 15, _ 2008 9:24:37 AM Page53of87 1

'Following a severe earthquake, Salem's Electric Plant status is as follows:

Q47 NO off-site power available.

Unit 3 is providing power to the 13KV ring bus sections 3, 4, and 5.

The SPTs supplied from these ring bus sections are energized and providing power to their respective 4KV busses.

2A DG is supplying 2A 4KV vital bus.

2B and 2C 4KV vital busses are deenergized.

028000K201 48 and 1EINon 1E.

Containment Fan Cooling Units, including breaker alignment for Fast and Slow speed.

Containment Iodine Removal Fans (Licensed Operator & STA only)

Control Rod Dnve Ventilation Fans (Licensed Qperatar & STA only)

Reactor Nozzle Support Ventilation Fans (Licensed Operator & STA only)

Reactor Shield Ventilation Fans (Licensed Operator & STA Only)

~ ~- ~ -~

~ ~ ~ ~ _ _~~ _ _

[ Page54of87

During Spent Fuel movement in the Spent Fuel Pool prior to a refueling outage, the Fuel Ha;dling

,Crane area radiation monitor (2R32A) fails HIGH when the fuel handling tool and attached spent fuel assembly are being raised. The crane hoist has NOT yet been fully raised.

IUnder these conditions, which of the following correctly describes restrictions concerning tached to the crane? - -

ovement of the fuel assembly must be terminated until an HP adiation Monitonng System components:

RIB, Control Room Inlet Duct Monitor R5, FHB 0 SFP Area Radiation Monitor R7, In-core Seal Table Area Radiation Monitor R9, FHB 0 New Fuel Storage Area Radiation Monitor RIOA, Personnel Hatch 0 Containment Elev I O O E Area Monitor RIOB, Personnel Hatch 0 Containment Elev 130A Area Monitor R1I A , RIZA, R12B, Containment Particulate, Noble Gas, and Iodine Monitor R13A, B, C D & E CFCU Service Water Monitors R17A and B, Component Cooling Liquid Monitor R18, Liquid Waste Disposal R19A, B, C, & D, Steam Generator Blowdown Liquid Monitors R32A, Fuel Handling Crane Area Radiation Monitor R36, Evaporator and Feed Preheaters Condensate Monitor R41D, Plant Vent Radiation Monitor uid PASS Room Area Radiation Monitor I -~ - ~ _ _

Monday, September 15, 2008 9:24:37 AM

~

I Page55of87

Given the following conditions:

Salem Unit 1 is operating at 73% power.

Main Generator output is 888 MW.

A grid disturbance causes load to drop to 798 MW.

The Rx does not trip.

The following indications are present:

- The "Block Cooldown" split bezel light on the Steam Dump control bezel is illuminated.

- The "Block Non-Cooldown" split bezel light on the Steam Dump control bezel is illuminated.

Which of the following identifies the status of the Main Steam Dumps?

-~ ~ _ _ ~- -- ~~ ~-~ --- __ ~~ ~- -~ __ - - ~

'armed,

-~

but the valves ~ are blocked

_ --from opening.

- _ ~- ~ --

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

~ _ _~ -~ - -- -~ ~ ~ ~ -~ -- ~~ - - ~ ~-

'NOT armed, and SHOULD NOT be armed.

~~- ~~ ~ ~~ -~~ - ~ ~- - - ~ ~ __ ~ ~ ~~ -- ~ _ _ ~~ ~

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

- __ ~ _ _

8/25/2008 123 MWe) will arm the steam dumps instantaneously. In this case, ile this is >IO% of current load, it is not > I O % of full load, and will not arm the dumps. Therefore, the split bezel indication is correct, and the steam dumps are not armed, and Steam Dump Control Logic Drawing

~ _ ~ _ _ _ _ ~ _ _ _

Mondav. SeDtember 15. 2008 9:24:37 AM

~ - _ .

I Page56of87

Given the following conditions:

- Unit 1 has just entered MODE 1.

1- Rx power is 8%.

- Power is being raised slowly in preparation for rolling the Main Turbine.

- 11 SGFP is in service supplying FW to SGs.

- ALL AFW pumps are aligned for normal standby operation.

- A spurious MSLl actuates.

which of the following describes the effect this will have on the AFW pumps with NO operator 1 Page57of87 I

QS I

'Which of the following indications would be present ONE MlNUTEfollowing a normal manual Rx all BF19s and BF40s shut due to the Feedwater Interlock signal. SGFPs remain in service.

~ ~~ ~~

~- ~~

All BF19s and BF40s have modulated closed and the SGFPShave tripped.

~~ ~ ~~ ~~

All BF19s and BF40s have modulated closed and SGFPs remain in service.

CN&FDWE008 LOR Identify and describe the Control Room controls, indications, and alarms associa System, including:

The Control Room location of Condensate and Feedwater System control bezels and indications. (Licensed Operator & STA only)

The function of each Condensate and Feedwater System Control Room control and indication (Licensed Operator & STA only)

The effect each Condensate and Feedwater System control has upon Condensate and Feedwater System components and operation. (Licensed Operator & STA only)

The plant conditions or permissives required for Condensate and Feedwater System Control Room controls to perform their intended function (Licensed Operator & STA only) d Operator & STA on1

~

~

The setpointsassociated

~-

~~ - ~~ ~~ with the ~

~

~ ~ ~~ ~

~~ ~~ ~~~

IMatwkI Requiredfor ExamDnatim 1 1 7 1 New ~ ~

Question Source Comments:

[Monday, September 15, 2008 9:24:37 AM I Page58of87

Which of the following describes the effect of 21 AFW pump failing to auto start on anormal k x trip

~~

Overcooling of the RCS during the initial 5 should NOT be reduced, ~- ~~~

and overfeeding of 23 and 24 SGs will occur.

~~ - ~ -

~~ ~ ~

~ __ ~~~- ~~ ~ _ _ ~ ~~~ ~

__ ~~ ~~

~~ ~~ __ ~~~ ~~~ ~~

~ ~~ ~~ ~~ ~

IOveKooling of the RCS during the initial 5 minutes following the trip. 23 AFW pump speed Ishould NOT be reduced, and overfeeding of 21 and 22 SGs will occur.

- ~~ ~~~~ ~~~ ~~ ~~ ~~ ~ -~ ~~~ ~~ ~~ ~~~~ _____

-~

~~~ ~~~~~~

- _ _ ~ ~ _ _ -_____

- ~-

-~ ~

~ ~~ ~~ ~~~

loperator action to throttle the 21-24AF1 'I, S/G LEVEL CONTROL VALVES, will be required to lprevent overfeeding the SGs, since 23 AFW pump will NOT be secured unless BOTH AFW lpumps are running in EOP-TRIP-1 Rx Trip or Safety Injection.

~~

_____ ~~~ ~ - _ _ _ _ _ ~~ ~ -~ ~~~ ~~- ~~

~~~ ~~ ~~ __ ~~~

~~

~~

_ _ ~ ~~ ~~ -~ ~-~~ ~ ~~ ~~~ ~

Operator action to throttle the 21-24AF11, S/G LEVEL CONTROL VALVES, will be required to prevent overfeeding the SGs, since 23 AFW pump speed will NOT be lowered to minimum the AFI 1s to balance flow to each of the SGs and maintain levels and pressures approximate. A and B lare incorrect because overfeeding will NOT occur since operators are directed to lower AFW flow. C is 004 NCT Describe the function of the following components and how their normal and abnormal operation a Feedwater System.

Motor-driven Auxiliary Feedwater Pumps Turbine-driven Auxiliary Feedwater Pump Turbine-driven Auxiliary Feedwater Pump Start-Stop Valve (MS132)

Turbine-driven Auxiliary Feedwater Pump Trip Valve (MS52)

Turbine-driven Auxiliary Feedwater Pump Speed Control Valve (GOV) (MS53)

AFW Pump Alternate Suction Header Supply Valves (AF52s)

Motor-driven AFW Pump Recirculation Flow Control Valves (AF140)

Motor-driven AFW Pump Discharge Flow Control Valves (AF21)

Turbine-driven AFW P

~~ ~~

- ~ - - ~

~

I Mondav. SeDtember 15, 2008 9:24:37 AM Page59of87 1

Given the following conditions:

Q53

- 2A 4KV Vital bus experienced a loss of bus voltage.

1- 2A EDG energized the 2A 4KV bus.

I- The SEC sequenced loads in accordance with MODE II*.

I- The normal source to the bus is now available.

lWhich of the following describes the method for restoration of the normal power supply to the 2A 4KV Vital Bus in accordance with S2.OP-SO.DG-0001, 2A DIESEL GENERATOR OPERATION?

The EDG is ...

with t

'Distribution and the fo

-- ~ -~

A and B are incorrect because the ED mode. C is correct because it is in accordance with the SO section 5.12. D is incorrect because there is no ability to parallel across a 4KV vital bus feeder breaker.

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

-~

Generator, including:

The Control Room location of Emergency Diesel Generator control bezels and indications. (Licensed Operator & STA only)

The function of each Emergency Diesel Generator Control Room control and indication. (Licensed Operator & STA only)

The effect each Emergency Diesel Generator control has upon Emergency Diesel Generator components and operation.

(Licensed Operator & STA only)

The plant conditions or permissives required for Emergency Diesel Generator Control Room controls to perform their intended function Diesel Generator control room alarms Facility Exam Ban

_ Monday,

_ ~ _ _September

~ _

15,_

2008 9:24:37 AM

_ _ _ . ~

[

~

Page60of87 1

Given the following conditions:

I

- Unit 2 is operating at 100% power.

- 2C Vital 4KV Bus is aligned to 24SPT (breaker 24CSD closed).

- Power is lost to 2C Vital 125 VDC Bus.

'- Prior to restoring power to the 2C DC Bus, 24 SPT is deenergized.

I IWhich of the following describes the status of 2C 4KV Vital Bus for these conditions?

-~

~ -

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

i- ergized

-- from the 2C EDG. ~~ ~ ~ - ~ _ _

~~~~~ ~ ~- ~

R 1A.C. Electrical Distribution

- ~- - - ~~-

K1. Knowledgeof the physical connections and/or cause-effect relationships between A.C. Electrical

Distribution and the following

~

.o ts for rs will remain "as is". The EDG breaker can not close onto the bus even though it is ldeenergized

- -~ ~~

because

- -~

one

-~

of the interlocks to -shut the EDG output

~~ ~~

~ ~ - -~ breaker is both infeed breakers

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

-~

Learning Objectives

.L DCELECEOl3 NCT Given plant conditions, relate the DC Electncal System with the following:

AC Electncal System Battery charger and battery

-~

Battery ventilation s Question Modification Methd: Direct From Source Questlon Source Comments: Vision (267359

-~ ~ _ ~ _ _ _ _ _ _ _ _ _ _ ~ ~ _ _

I Monday, September 15, 2008 9:24:37AM 1 Page61 of87

Which of the following choices identifies an adverse effect of a ground on a 125VDC buslbattery, and the method in which operators perform ground isolation IAW S2.OP-S0.125-0004, 125VDC

'Ground Detection?

iA ground.. . ~

~

~-

~

~

~~

s causes a higher level of current to flow in the system.

transfer to the backup battery charger to determine if the I/S charger is the cause of the

_ _ ~ -~ ~ ___ -~ ~ ~ ~~ ~

__ ~~ _ _ ~ ~~~ ~~

Ionthebattery associated with the bus causes voltage reading on the bus to become unreliable ldue to the excessive current flow. Individually deenergize each load on the bus, then re-renergize

~~

~- if that load

~~~ is not the source of the ground.

-~~~ ~ ~~ ~ ~~

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

~~~ ~ ~ ~- ~ ~~ ~~ ~ ~~ ~~~ __ ~

on the bus causes voltage reading on the bus to become unreliable due to the excessive current flow. Transfer to the backup battery charger to determine if the I/S charger is the cause -__ ~

of the ~ ground.

~~ ~~~ - ~ ~~ ~~~ ~ ~ ~~~ ~~ ~~~ ~~ ~ _ _

~~~ - _ _ ~ ~~~ ~~ ~ __ ~~ ~ ~~ __ ~~ ~- ~~~ ___

the buscauses a higher level of current to flow in the system. Individually deenergize each

'load on the bus, then re-energize

~~~~~ ~~ ~~- ~ _

if that load

_ ~~~ -

is not the source of the ground.~-

~ ~~~ ~ ~~ ~~~~ _ _ ~ -

has operators transfer to the standby battery charger if bus voltage is low, and battery current is present, so these are plausible distracters. The bus voltage is higher than the battery voltage, so a ground on the battery would not cause bus current to rise.

teaming Dbjectivles #

DCELECE008 ldentifv and describe the Control Room controls. indications, and alarms associated with the DC Electrical System, including:

The Cbntrol Room location of DC Electrical System control bezels and indications. (Licensed Operator & STA only)

The function of each DC Electrical System Control Room control and indication. (Licensed Operator & STA only)

The effect each DC Electrical System control has upon DC Electrical System components and operation. (Licensed Operator &

STA only)

The plant conditions or permissives required for DC Electrical System Control Room controls to perform their intended function (Licensed Operator & STA only)

~~ ~~ ~~

~ ~~ ~ ~

~

~~ ~

- - ~

- ~-

rMonday, September 15,2008 9:24:37AM Page62of87 J

IGiven the following conditions:

- - ~~

Q5c:

1- Operations Management has received information from a Vendor about certain components associated with the 28 VDC Electrical Distribution system.

-' The Vendor is performing testing at their manufacturing facility to prove compliance with applicable standards. This test data is expected to be received in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by Salem.

I- As a precaution, the Operations Manager wants the 28VDC system monitored much more closely than normally required for the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

'Which of the following identifies the method in which this additional monitoring should be made known to the shift operators IAW OP-AA-102-104, Pertinent Information Program?

I 1OPERATlONS MEMOS, ETC."

short-term, written philosophy and instructions from Operations Management to Shift Crews providing information which includes, but is not limited to:

1- Special plant operations,

- Operating administrative requirements,

- Priorities,

- Manpower requirements and availability,

- Equipment deficiencies,

- Housekeeping,

- Special data taking,

- Orders to cover backshifts and weekends, Pertinent Information Program

' Page63of87  ;

1

~

Monday, September 15, 2008 9:24:37 AM / Page64of87

-~ ___________ ____-

'Given the followingconditions:

Q57 I

I- A total loss of all AC power occurred at Salem.

I- Operators were successful in restoring power to a single vital bus with its respective EDG.

- The EDG output breaker just tripped on bus differential.

which of the following describes how the EDG will operate following its output breaker trip?

I

, r ~ ~ ~~ ~ -- ~~ ~-~- -~ ~~ -

'tripped

~~ -

when the EDG local MCC lost power since the Fuel Oil pumps-~

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

-~ ~~ ~~ ~ lose power.

~- ~~ ~ ~

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

will continue to run until its Fuel Oil Day Tank empties due to the loss of power.

~~ ~~ ~ ~~~ ____ -~

~- ~-

____ ~~

A is incorrect because the Fuel 0 are powered from theEDGlo and D is correct because EDGs have shaft driven fuel oil pump specifically so they do not need external

~~

- Diesel Generators:

Lubricating Oil System Jacket Cooling Water System Fuel Oil System Starting Air System Turbo-charger Turbo Boost Air System Exciter 0 Regulator Governor/Speed Control

_ _ ~ ~ -

'Monday, I September

_ __ _ ~ _ 15,2008

_~ _ _ 9:24:37

~ _ AM -~

Page65of87 1

Given the following conditions:

- Preparations are in progress to perform a release of 21 Waste Gas Decay Tank IAW

~ S2.OP-SO.WG-0008, Discharge of 21 Gas Decay Tank to Plant Vent.

1- The 2R41A rad monitor is declared INOPERABLE due to an intermittent power supply failure.

IWhich of the following describes the effect this will have on releasing 21 GDT?

~~~ ~~~ ~- ~~ ~~~ ~~ ~ -~ ~ -~ ~- ~ _ _ -~

~A2. Ability to (a) predict the impacts of the following onthe WasteGas Disposal System and (b) based on

,those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal ithout the R41A or D if the double sampling and analysis, and double release rate calculations are performed IAW Section 3.0 of Attachment 2. A is incorrect because the 2R41D does not have to be OPERABLE. C is incorrect because there does not have to BE a low range noble gas monitor as long as B above is performed. D is incorrect because while it is true that the monitor will be blocked. it does not Dreclude startina the release.

I Reference ntle of 21 Gas Decay Tank to Plant Vent.

~ - ~ --

Learning Objectives WASGASEO11 Identify the differences between Unit 1 and Unit 2 Radioactive Waste Gas System components, parameters, and operation.

a) Not applicable to this lesson

~- ~~

LMonday, Septembe?57008 9:24:37 AM I Page66of87 1

- -~ -

rform a release of 21 Waste Gas Decay Tank IAW f 21 Gas Decay Tank to Plant Vent.

INOPERABLE due to an intermittent power supply failure.

ffect this will have on releasing 21 GDT?

~ ___ - ___ __ _ ~. ~~

~ ~ -- ~~~ ~ ~ ____ -~

,CAN be started after double Sam , and double release rate kalculations are performed.

- ~ - _____-

r-

~~~

- ~ - -

- 7

~~

-~ ~ ~~

~ -~ ~ - - - ~ - ~~

can NOT be started since there i onitor available to automatically terminate the release on high radiation.

~~ ~~~~ ~~ _____ ____ ~ ~- ~ --- ~~ ~ - ~ -

-~ ~~ ~- ____- ~

~~ ____ --____ ____ ~~ -~ ~~ --- -~ ~~ ~ ~ -- _____ ____

can NOT be started since the 2ND17572-ZW aste Gas Decay Tank Block Switch is rate calculations are performed IAW Section 3.0 of Attachment not have to be OPERABLE. C is incorrect because there does BE a low range noble gas true that the monitor will be

- - -~ ~~~~

~ - - ~- -- ____ ~ ~~ ~ - - -~ ~- ~ ~~ ~

'Materisl Requiredfor ExaRBination I QuestionSource: New Question Source Commentas: "

\

[-Monday,September 15, 2008 9:24:37AM

- - ~ - - _ _ _ _ _

~ ~ -

Paae67of87 1

While removing a source, R P personnel drop it o n the floor 10 ft. from an area monitor. If this are; 2 Whr, what is th

- dose rate 1 ft. from the dropped source?

- ~

~ __ ~

~ ~-

40 Whr.

~400Whr.

- ~ _ _

RADCONEOO6--

~~ ~

Perform the following, using the principles of time, distance, and shielding:

a) Describe the methods used to shield personnel from alpha, beta, gamma, and neutron radiation.

b) State how time, distance, and shielding are used for dose reduction.

c) Define stay time and perform calculations to determine stay time or dose.

d) Calculate the dose rate at a distance from a gamma point or line source.

shielding problems

~~~~ ~

Quesrtldn Source: Other Facility

~ _ _ _

Lfvbnday, September 15,2008 9:24:37 AM

~ . _ _ _

I Page68of87 1

Given the following conditions:

Q60 I

- Due to Main Condenser problems, a rapid power reduction on Unit 2 has been performed from 100% at 15% per minute, IAW S2.OP-AB.LOAD-0001.

1- The power reduction is stopped at 22% when the problem is corrected.

- The Reactor Operator reports that RCS temperature has just dropped below 541O F and continues to slowly lower.

stem.

Reference W e 'I e actions taken in S2.OP-AB.LOAD-0001(q)and the bases for the actions.

~

I-IIP.

Monday September

~ 15,2008 _ _ _ AM

~ _ _ 9:24:37 - -

Paae69of87 I

Given the following conditions: ~

- Unit 1 is operating at 100% power.

- BOTH SW header pressures are 118 psig.

~- 11,I 3, and 15 SW pumps are in service.

LearningObjectives SWBAYSEOO9 State the setooints. coincidence. blocks and Derrnissives for automatic actuations associated with the Service Water System -

Intake Bays.

1 Monday, September 15,2008 9:24:37 AM , Page70of87

iWhich of the following identifies how the Emergency Control Air Compressors (ECAC) respond to I Air header pressure on both the " Aand "B" headers?

~ - -

~~~~~ ~

The Unit 1 ECAC will start at 85 psig on If Control Air he e pontinues to degrade and reaches 80 psrg, the Unit 2 ECAC will then start.

~ ~- ~ ~ ~~~ ~ ~- - ~~ ~~- -~~ _ _ _ _ ~

~ ~- - ~ ~ ~-

~~--- ~ - ____ -~ ~ ~ ~~ ~~ ~ ~~

The Unit 2 ECAC will start at 85 psig on (CA header "B". If Control Air header "A" pressure lcontinues to degrade and reaches 80 psig, the Unit 1 ECAC will then start.

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

~ ~~ -~ ~~ ~~ ~~

~ -~ ~~

~

~~ ~ -___ ~ -~ ~~ ~ ~- ~ -~ -- ~- ~-~ ~ ~ ~ - -

BOTH ECACs will start at 85 psig on their respective CA header.

-~ - ~- ~- ~

IBOTH ECACs will start at 80 psig on their respective CA header.

~- ~ -~~ ~

Mondav. SeDternber 15, 2008 9:24:37AM I Page71 of87 1

IGiven the following conditions:

1- Unit 2 is operating at 100% power.

- OHA A-7, FIRE PROT FIRE, actuates.

- 2RP5 indicates the following:

- Zone 59 - Air and Water Deluge, Containment El. 100 Panel 335 is lit.

- Zone 74 - Smoke and Fire Detector, Containment El. 100 Panel 335 is lit.

~ - ~~ ~ ~

, Fire Protection Manual Isolation Valve, in the 2RP5. FP147 is not an automatic valve.

& STA only)

The Dlant conditions or permissives required for Fire Protection System Control Room controls to perform their intended function. (Licensed Operator & STA only) the Fire Protection System control room alarms. (Llcensed Operator & STA only)

-~ -~ -

~ ~~-

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

[Material Required for Examination I Editorially Modified QUeStiStion %WCe COSlWtd6: Wsion Q82748

~ _ _ _ _ _ _

Monday, _

September

. ~

15,2008 9:24:37 AM -- I Page72of87 1

,Unit 2 is in Mode 6 and core alterations are in progress.

I

'Which ONE of the following conditions, taken by itself, will result in a loss of the ability to establish pontainment closure withinthe 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time lirriit if required?

-~ - - - - -

~ ~ - p

'Temporary hoses have b

- - - p p - p run through the 100' airlock lAW applicable procedure.

- - p - - ~ - ~

- ~-

- p - - ~

- - - - p - - - p _ - - - p - - - - -- - ~ --

The containment 100' Airlock Inside Door becomes cocked and can NOT be closed.

- p - - p p - - - - - - - - - - - -

- p - - p - - - -

~~ - p - ~ p - - - - ~ ~

21~SG secondary - - - manways removed with 21 MSI 67 open and its operator

- ~ p ~ - - disconnected.~- ~

Only 6 bolts can be found for securing the Containment Outage Equipment Hatch (OEH) iinside door.

8/25/2008 03000K303 1

-~ -66

'being closed. The OEH only requires 4 bolts for closure. Having a path from inside containment into the secondary side of the SG and through the MSLl valve outside containment, and no way to close the MSlV rwould result in containment closure being unable to be established in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Learning Objectfves IPrevious 2 NRC Exams 12006RO exam Q18

-~

Monday, September 15,2008 9:24:37AM-

~. ~ ~~ --

, Page73of87 1

ICharging Systems SI systems flow meter indication has been lost.

- ~-

IRCS pressure is 170Opsig and going down. ~ ~~~~

I IRHR Dump flow is 1700 gpm and steady.

1940016117 I

- 67

~

2.1 Conduct of Operations Ability to make accurate, clear and co 3.51

__ ~ ~ - -- -

Procedure states in section 4. tion of indicator reading should be provided in th format of PARAMETER-VALUE-TREND (with rate when appropriate)." It also states in 4.4.1

'that..."ENSURE all communications are clear, concise, and free of ambiguity." Using the word "lost" idoes not define the indication, it does not give status of indication. A is correct because is gives specific I

'trainkomponent, current status, and trend. E3 is incorrect because slang is used (lost). C is incorrect because going down is not a correct identification of rate of trend. D is incorrect because it does not provide train.

I Reference Title I Learning Objectives

_ _ ~ _ _ _ _ _ ~ - - __

I Mondav. Seotember 15. 2008 9:24:37 AM Paue74of87 1

se of the Containment Spray System?

sig following a Loss of ICoolant Accident (LOCA). ~ ~ ~ ~~~ ~

~~ ~~

'Maintain containment pressure less than the test pressure of 54 psig following a Main Steam Line Break (MSLB) inside containment.

~~ ~

~~ ~

llnject a mixture of borated water and Sodium Chloride (NACI) into the containment atmosphere following a LOCA to minimize exposure to the public following a LOCA.

~~ ~ ~ ~~ ~ ~~ ~ ~ ~~ ~~

Inject a mixture of borated water and Sodium Hydroxide (NAOH) into the containment iatmosphere following

~ ~~

~~ ~ ~~

~~~ ~~~

Salem UFSAR, s Learning Objectives

~

M o n d a y T G c m b e r 15,2008 9:24:37 AM -

~~~

Page75of87 1

Given the following conditions:

Q47

- Unit 2 is operating at 100% power.

- 21 and 22 AFPs were declared inoperable at 0405 on April 1st due to being sprayed with leaking SW and failing a subsequent motor megger.

1- A unit shutdown is commenced at 0615 to comply with the associated Tech Spec Action Statement.

Which of the following identifies a situation where the shutdown would be stopped prior to taking the unit off line?

Reference provided. ~ --

~ - ~ ~~ ~- ~

- ~~

~ --

~-~~~

rMaintenance reports that they are certain at least one of the AFW pump motors can be dried and restored to OPERABLE status -~~ by noon of that day. ~ - - ~~- ~- ~ -~ -~ ~ -~ -- -~

_ _ - ~ ~ -~ - ~~ ~ -~ ~ ~ -~ _ ~ ~ ~ -- ~~ - - -

The Electric System Operator (ESO)calls the Control Room and orders the unit shutdown be iplaced

~- - ~~

on hold to maximize

~ - - -~

generation.

~ - ~ ~ ~ ~ ~- ~ - ~ ~ _ _ ~-~ -

~ ~ ~ - -- -~ - ~

- ~ - ~ _~ ~_ ~ ~ ~ ~ ~- ~ ~ ~ - ~- -_ - ~~ - - ~-

More

-~ -_ than one

~- ~ -

Control -

~ ~

Bank D rod

~ _ - is identified as being

~-~ ~ ~~ -~ ~~ -

stuck with-~the - Rx >50% power. -~

~~ - _ ~- ~

~ ~~ ~ -~ - ~ -

,%-AFtrip

- ~ linkage is discovered disengaged and bent,

~ _~ ~ ~ ~ ~~ - and can NOT be reset.

~- ~ - _

-~

Interlock) C is incorrect because the AB for stuck rods states that if a power reduction is required, which lit is in this situation, to borate as necessary to reduce turbine load. The Action Time for being in HSB for this TSAS is 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (1005), so the report that the AFW pps could be restored outside this 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> limit will not stop the shutdown. If the candidate uses the 0615 commencement of the shutdown as the start of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> clock, thev mav think the shutdown could be stopped. B is incorrect because the ESO does not have the authori<y to &der anything related to Rx operation. ~

Specification action. (License Operatdr and STA only)

~

NCT State the Technical Specification assoclated with the component, parameters and operation of the Auxiliary Feedwater s) (Non-licensed Operator)

~ ~ - _ ~-

- ~ ~

_ _ ~-

_ _ ~

- ~~

~~

-~ - ~~

/ Page76of87 ]

MondzSepternber 15,2008 9:24:37 AM Page 77 of 87 Given the fo Ilowingc o nditions:

i- Unit 2 is in MODE 6.

- Containment Closure is established IAW S2.OP-ST.CAN-0007, Containment Closure.

Support Systems, including:

The Control Room location of Containment and containment Support Systems control bezels and indications. (Licensed Operator & STA only)

The function of each Containment and Containment Support Systems Control Room control and indication. (Licensed Operator

& STA only)

The effect each Containment and Containment Support Systems control has upon Containment and Containment Support Systems components and operation (Licensed Operator & STA only)

The plant conditions or permissives required for Containment and Containment Support Systems Control Room controls to perform their intended function. (Licensed Operator & STA only)

The setpoints associated with the Containment and Containment Support Systems control room alarms. (Licensed Operator &

-~ ~ ~

~- - -

~

~

~~ ~~ ~~

~~

~~ ~~ ~

-~~

~

I Paae78of87 ~

Given the following conditions:

Q64

~- Unit 2 is operating at 100% power, with no active Tech Spec LCOs in effect.

- 2C EDG is being C/T for a scheduled 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> maintenance window.

I IIn addition to the TSAS for the INOPERABLE EDG, 3.8.1 . I .b, which of the following identifies a Comprehension 008 194001G234

_~ __

71

~~

r determining the inter rnal effects on core re _ _ ~

- ~ - ~ _ _

~~~~~~ ~

~- ~_ ~~ ~~ ~

plement 1 K / A K 2 . lyze the effect of maintenance activities, s wer sources, on the status of limiting conditions for operations. When addressing an INOPERABLE EDG in TSAS 3.8.1 .I .b, Action b.2 states "Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, declare required systems or components supported by the inoperable diesel inoperable WHEN A REQUIRED REDUNDANT SYSTEM OR COMPONENT IS INOPERABLE.. .." (My emphasis) This precludes the necessity for "cascading" tech specs to the components who have their emergency power supplied from the inoperable IEDG, but have all redundant systems and components remaining operable. In this case, no other indication is provided in the stem that would lead to the conclusion that any required systems were not Salem Tech Specs Material Required for &mination Questkm Source: New Qu6lstion Source Comments:

I Page79of87 1

-~ a70 lWhich of the following statements is TRUE regarding personnel exposure limits during an ALERT at Salem-~ -IAW NC.EP-EP.ZZ-0304,

~_~ - ~ -

___ ~ -~ ~ OSC Radiation Protection -~ Response?

The EDOs authorization for an Emergency Exposure can be delegated to the Radiological Assessment Coordinator

~

~ - __

(RAC) ONLY.

~ _~ -~ - p -~ - ~ ~- - ~ ~ ~ - p - - _ _ _ _ ~

r_p -~ ~ - _

p ~ _ - ~~- - - ~ ~- -~ ~_ - __ ~ -~- ~ --_ ~~ -

an operators yearly dose limit is automatically raised to 4500 mrem regardless of their NRC

~ _ _ --~ ~ ~ _ _ p_ _ ~ _ ~ - _ ~_~ ~- ~ ~ p~ ~~ ~

IAn operators yearly dose limit is automatically raised to 4500 mrem ONLY if their completed INRC~-

L _ p Form- 4 is on record.

~-_~ __ ~ ~- ~_~ ~ _-_ _-_ ~

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

r- ~

~~~ ~ --- ~ ~~~ - p ~ ~ _ p~~ ~~ ~ ~ p~ ~ ~p -- ~- ~

Emergency exposure can be authorized for someone who has previously received an

,Emergency Exposure, as long as the previous exposure resulted in less than 10 Rem in the current year 1- ~ --* ~~ ~~ ~ __ ~ ~~ -- ~ ~~ ~- - - ~ -~ -~ - ~~ ~ ~

because a current NRC Form 4 is required to be on file. C is correct because of B above. D is incorrect because the rule for NOT considering an Emergency Exposure as an Emergency Exposure is if the resultant exposure for the year is less than or equal to 4.5R after finishing the exposure.

Reference Title I OSC Radiation ,Protection Leamlng Objectives

._ I _.. . - __ I I . __ -_ _ I __ ---- - - -- . . - I_

RADCONE003 D k c u z t h e emergency exposure limits in accordance with applicable station procedures, pertaining to

~_ ~_

A. Automaiic dose extensions during an emergency.

B. Position responsible for authonzing emergency exposure.

C Guidelines for personnel receiving emergency exposure.

D. Emergency Exposure Limit for life saving Monday, September 15,2008 9:24:37AM-- I Page80of87

'During an outage, the replacement of a CVCS valve will require 3 person-hours of work without a

~ 7 1 respirator, or 5 person-hours with a respirator. External dose rate for the area is 75 mRem/hr, and Ithe airborne activity is 15 DAC. (2000 DAC-hrs represents 1 annual limit of intake and is equivalent to 5 Rem CEDE.)

8/25/2008 194001G304 excess OF those authorized. ___

rds that may arise during normal, abnormal, or emergency conditions. The first choices is external dose without respirator.

a) Describe the methods used to shield personnel from alpha, beta, gamma, and neutron radiation.

b) State how time, distance, and shielding are used for dose reduction.

c) Define stay time and perform calculations to determine stay time or dose.

d) Calculate the dose rate at a distance from a gamma point or line source.

~ ~ ~-

~~ _~

- ~~ -~ ~ _~ ~_ ~ ~- -

~-

r M o g a_

y ,_

September 15, 2008 9:24:37AM -

'The Unit 2 CRS has directed a Unit shutdown, based on RCS activity exceeding TS 3.4.9 limits.

m2-Which of the following Tech Spec required actions performed after the Rx is shutdown is designed iMSlVs are closed.

~- ~~ ____ ~ ~- -- ~ ~~ ~~~ ~ ~~ ~~ ~ ~- ~ ~~

~ ~ ~ .~ -~ ~- -~ ~ ~~ ~ ~ ~ ~~ -~ - - ~ ~- ~ ~ ~ ~~

RCS is cooled down __ below 500F.~~-

, ~~ ~~

~~~

~~

~

~~

~

~

~

~~ ~~

~- ~

~

~~ ~

~ ~~

~

~-

-~

-~ ~-

~-

~~ ~~

~

Main steam dump- valve-~~setpoints are raised.

~ ~~

~ ~ ~~

n a timely manner. B is I correct because the Bases for TS 3.4.9 states as such. C is incorrect because the ruptured SG atmospheric relief setpoint would be raised, not all the steam dump settings. D is incorrect because ,

Condensate polishing would help clean the secondary plant but not an action performed in accordance 1 lwith the actions of TS 3.4.9 Question Yodifrdon MfHhoel: Editonally Modified Monday, L- -~

September

-~~ -

15, 2008 9:24.37 AM

~

~- -

i Page82of87 1

IGiven the following conditions:

Q73 I

' Unit 1 is responding to a LOCA, and due to a Core Cooling PURPLE path is taking action IAW FRCC-2, Response to Degraded Core Cooling.

1- While performing actions IAW FRCC-2, the STA identifies a Containment Environment l PURPLE path is present due to containment pressure, and been verified by control conso le i ndication.

- No other RED or PURPLE paths are present.

Which of the following describes how the crew should proceed IAW OP-AA-101-111-1003, Use of Memo 8/25/2008 IGENERIC

-p_ -- ~- - _ -~ - ~ -~ _ _ - _ ~p Procedures

/ Plan - -

~-

~ _

p - - p p p - ~ p - - ~ ~ p

- p

~~p p~ ~- - -p e of general -

pp p ~

guidelines for E

_ p _ _ - p

-AA-101-111-1003, Use of Procedu E FRP is entered, the ating crew remains in the FRP until either the FRP directs a transition to another procedure (which lmay be the procedure in effect), or a higher priority challenge is detected, in which case the current FRP is suspended and the operating crew transitions to the FRP directed by the higher-priority challenge." A is incorrect because FRCE is not a higher order PURPLE path. B is incorrect because the transition is not made if the condition is not currently PURPLE. C is incorrect

'because once entered, a FRP is performed to completion OR when a higher order FRP entry condition is identified.

I Reference Title 1 Use of Procedures A,-Immediate Actions B -Continuous Action Summaries C.-Communications D -Log Keeping E.-Application of Notes and Cautions F.-Transitions

-- - p pp

- - p

~

L-l__-p_-----

Monday, September 15, 2008 9:24:37 AM p --

Page83of87 ,

,Given the following conditions: ~

Q74 I

,- Unit 1 Control Room has been evacuated clue to a fire in the Relay Room IAW l S I .OP-AB.CR-0002, Control Room Evacuation Due to Fire in the Control Room, l Relay Room, 460/230 Switchgear Room, or 4KV Switchgear Room.

1- The Reactor Operator assigned to perform Attachment 4 places the EDG Fire 1 Emergency Bypass Switches 69/1, 6912, and 69/3 in BYPASS while the EDG is running unloaded.

Which of the following choices identifies how this will affect 2C EDG operation?

The EDG... -~

~_ ~~ ~ ~ ~ ~

~

lvoltage regulator will swap into lsochr regardless of its previo IEDG can NOT be operated in parallel with the grid when the control room has been ~ ~ ~ ~~~~~ -~evacuated.

~

~ ~ ~ - ~ - ~ - - ~~_ _ ~ - ~~ ~ __ ~ ~ _~ -~ -

will trip due to the break-before-make switch characteristic, and must be manually restarted if needed

~ ~-~~ for power. -~~ ~ ~ ~ ~~~ ~ ~ ~ ~ _~~

- ~~ -_ ~~

~~-~ ~- ~~ ~ _ ~ _ ~ - ~_~ ~ ~ ~ _ ~ ~~ - ~~

room ventilation will automatically shutdown due to insertion of a "fire detected" signal into its loperating circuitry.-

~ ~~ ~ ~ ~~

~~ ~ ~ ~ - ~ -

- _ ~ ~ ~ ~~ ~-~~~-~ - ~ -- ~ ~ ~ _ -

will continue to~run,~and- all ~non-SEC _

~~~~~~~~~~ _trips ~ will - be reinstated. - ~ -

id R Memlory Salem I& 2 v.-n n...., t i l -..-...

ntinn I ~~ ~~~~ ~

Reference W e I Control Room Evacuation Due to Fire in the Control Room, Relay Room, 460/230 Switc

~ ~~

including:

The location of Emergency Diesel Generator local controls and indications. (Licensed Operator & Non-licensed Operator only)

The function of Emergency Diesel Generator local controls and Indications. (Licensed Operator & Non-licensed Operator only)

The plant and conditions or permissives required for Emergency Diesel Generator local controls to perform their intended function. (Licensed Operator & Non-licensed etpoints associated with the Emergenc

_~ ~

~- -- ~ ~- ~~~

_~________

I L Monday,

_ _September

_ _ - _ 15,2008 9:24:37 AM 1 Page85of87 1

Given the following conditions: ~

q7s

- Unit 2 is operating at 80% power.

1- Unit 1 and Hope Creek are operating at 100% power.

I- Unit 2 Main Generator is operating normally, with output of 1000 Mwe, 325 MVAR OUT.

I- Unit 2 Main Generator H2 gas pressure is '75 psig.

- The Electric System Operator (ESO) calls the Unit 2 control room and reports a grid disturbance has tripped several non nuclear units off line.

I- All transmission lines remain in service.

I- The ESO requests Salem Unit 2 to adjust its MVAR loading to the maximum allowed per IEngineering Evaluation A-5-500-EEE-1686.

Which of the following identifies how the change will be accomplished, and identifies the MVAR loading CLOSEST to what the maximum MVAR loading will be?

~- ~~~ ~- - _ ~

Transfer the Voltage Regulator to MANUAL IAW S2.OP-SO.TRB-0001, Turbine Generator Istartup Operations. Utilize the MANUAL ADJUSTER to obtain 400 MVAR OUT.

L - _ ~ ~ --

- ~ - ~~ - - ~ - -

~ ~ _ -

Transfer the Voltage Regulator to MANUAL IAW S2.OP-SO.TRB-0001, Turbine Generator Startup Operations. Utilize the MANUAL ADJUSTER to obtain 650 MVAR OUT.

-~ ~ ~~~ ~_ - - ~

lb Applic 008 en&c Knowledge and Abilities _ ~-

C 77

- ~- - -~ ~

~- -~~ _ _ _ ~ _ _ -~~ - - _ _

2.4 Emergency - ~

Procedures

/-Plan- ~ _ _ - ~

tpoints and oper ~

esP 3.3 New Rev. 2 Supplement 1 KIA is 077 Generator Voltage and Electric Grid Disturbances 2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response lmanual. Using the A-5-500-EEE-1686-1686 Rev. 8, the operator should either go ldirectly to Curve 3S2, page 258, or can use the Tables section to identify which curve to use. Table 13 is for 3 unit operation, Power System Stabilizer in service, which is the correct mode of operation for PSS.

Curve 3S2 is for 3 unit operation, with all transmission lines in service. The stem also states generator gas pressure is 75 psig. Using 1000 MW and the 75 psig line the intersection for maximum MVAR

'loading is 650 MVAR OUT. To realch this, the operator adjusts the Auto Volts Adjust PB to RAISE Iwhich raises excitation to the Main Generator and causes MORE MVAR to be sent to the grid. The distracters are either to use the manual volts adjust, or the limit for 100% electrical loading, not the t.eamirlg Objectives GENC02E012 - ldentifvand describe the Controlk6om Gntrols, indications and alarms associated withthe Unit 2 Main Generator and Hydrogen Cooling System, including:

~

a) The Control Room location of Unit 2 Main Generator and Hydrogen Cooling System control bezels and indications b) The function of each Unit 2 Main Generator and Hydrogen Cooling System Control Room control and indication c) The effect each Unit 2 Main Generator and Hydrogen Cooling System control has upon Unit 2 Main Generator and Hydrogen Cooling System components and operation

__ --___ ~ _ ___ _ -

~-~ - - ~ -

I c~ -

Monday, September 15, 2008 9:24:37 AM l Page86of87 I---

d) The plant conditions or permissives required for Unit 2 Main Generator and Hydrogen Cooling System Control Room controls to perform their intended function

- ~ ~ ~

e ) The setpgngassogated with ~-~

~~~ - p~ p -- p _ _ -~~~~

the Unit 2 Mal

~

' Paae87of87 1

U S . Nuclear Regulatory Commission Site-S pecific Written Examination Applicant Information

-~ ~~ ~

Name: Region: I Facility: Salem 1 & 2

- ~~~~~ - ~ ~~ ~

License Level: SRO Reactor Type: W

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

Start Time: Finish Time:

~~

~-

~ ~~- ~- -~ ~

Instructions

~ -~ ~ -

Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected SIX hours after the examination starts.

-~

_~

~ ~~ ~ _ _ ~ ~

~ ~

Applicant Certification

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

~--

All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature

~~-~ ~~~ ~~ ~

ResuIts

~

~ ~~ ~

Examination Value Points

- - ~ - ~ ~

~ ~- ~

Applicants Score Points Applicants Grade Percent 1I

1. c

2. b

3. d

4. a

5. a

6. b

7. b

8. c

9. c IO. d

11. d

12. b

13. a

14. a 15, c

16. b

17. b

18. a

19. a

20. c

21. a

22. c

23. a

24. d

25. d Page 1

IGiven the following conditions:

Unit 2 is operating at 100% power.

A total loss of off-site power occurs, and the Rx automatically trips.

ALL Control Bank "A" control rods fail to trip and remain fully withdrawn.

ONLY 2C vital bus energizes from its EDG, 2A and 2B vital busses remain deenergized.

As 21 MSIO starts to open as expected for RCS temp control, it fails full open.

The following indications are presents during the second pass through the Immediate Actions of EOP-TRIP-I, Rx Trip or Safety Injection:

- ALL PR Nl's indicate 0% power.

- IR NI indication is 3x10-6 A and stable

- IR SUR is 0.0 dpm on BOTH channels;.

Which of the following identifies the current status of the Rx , and how operators should respond?

- - _ _ _ - _ _ _ ~ ~ ~

~~ ~~ ~- ~ _ _ _ _ _ _ _ _ ~ -

ctor trip is NOT confirmed, ao to LOPA-I , Loss of All AC Power.

~~ ~- - ~ ~ ~ - _ _ _ - ~ ~~- - _ _ _ _ _ _ _ ~ - - - -

The reactor trip is confirmed, continue in TRIP-I, after Immediate Actions are completed.

- - - ~ ~ _ _- .

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

~ ~ _ _ _~- - ~ ~-

'The I-- reactortrip is NOT _- confirmed,

- go to FRSM-I, Response to Nuclear Power Generation.

-~ ~~ _______- --- - ~ _ _ _ _ _ _ _ ~ ~

- - - ~ _ _ _ ~ -

the reactor trip is confirmed, transition to TRIP-2, Rx Trip Response after verifying SI not iactuated or demanded.

L---____ ~ ~ - ~ _ _ _ _ _ _ _ - ~ ~ - - - ~ -~ --- - - -

The criteria for trip are 3/4 power range NI <5%, IR NI level dropping, IR SUR NEGATIVE. With a SUR of ZERO, Rx Itrip is NOT confirmed. Transition is to FRSM. A is incorrect because a single Vital Bus is energized and ithe transition to LOPA-1 is made if NO vital busses are energized. B is incorrect because the Rx trip is lnot confirmed. D is incorrect because the Rx triD is not confirmed.

Concept used, numbers given in stem different. Not really a "new" question.

Thursday, August 21, 2008 10:37:44 AM

- ~-

Page 1 of35

~

Given the following conditions:

Unit 1 is operating at 100% power.

Charging system problems result in NO Unit 1 charging pumps running OR being available to be run.

The decision is made to align and run 23 charging pump (Unit 2) from Unit 2 RWST to Unit 1 CVCS system.

The proper lineup is completed, and 23 charging pump is now supplying the Unit 1 charging system.

Unit 1 PZR level is rising very slowly.

Which of the followi dentifies the next action to be performed, and why?

of Unit 2 RWST level.

~ _ _ _ _ _ ~ ~~~~ ~~ ~ ~~ -~ ~~

_ _ _ _ ~ ~~~ ~_____

-~ ~ ~ ~ _ _ _ ~ _ _ ~ ~~~ - ~ -

~~~ ~ ~~ _ _ _ _ ~ _ _ _ _ _ - ~ ~~

ICommence a Unit 1 shutdown IAW S I .OP-IO.ZZ-O004, Power Operation, due to the boration of the RCS from Unit 2 RWST.

~~~ ~ ~~ ~

~~ ~~~ ~~~ ~ - _ _ _ _ _ ~ ~- ~ ~ ~~

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

Reestablish normal letdown on Unit 1 IAW S I .OP-S0.CVC-0001 , Charging, Letdown, and Seal Injection to control PZR level.

~~ ~ ~~~~ _ _ _ _ ~ ~~~ ~- ~~

~- ~~- ~

~~~~~ _____ ~~ ~~~ ~~ ~ ~~ -_____ ~

Initiate makeup to Unit 2 RWST IAW S2.OP-S0.CVC-0006, Boron Concentration Control, to ensure minimum Tech Spec required level in Unit 2 is maintained. _____ ~~ ~-~ ~~ ~~~

____ _____~ ~

~~ --

hey apply to Loss of Reactor Coolant

~~ ~~ ~~ _____

~~ ~~~ ~~

charging pump will be evaluated or other contingency actions implemented (AB.CVC-1 Step 3.49 NOTE) ,

IC is incorrect because normal letdown can NOT be restored since the letdown isolation valves are iinterlocked with the charging pump breakers such that at least ONE charging pump breaker has to be Ishut to open letdown isolation valves. Excess letdown may be placed in service to control PZR level. D lis incorrect because while it will probably be done, it is not the next step. It will be done after the shutdown of Unit one has started.

LearningWaotrves

. .- . . " . -. .. ~. . .. .. . . . .. -.. .. _. ..- ....-..-..... - - __II ABCVClE003 Given a set of initial plant conditions:

Determine the appropriate abnormal procedure.

Describe the plant response to actions taken in the abnormal procedure.

blished by the abnormal procedure.

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

1 New 1

Question Source Comments:

1 Page2of35 I

~~ ~~~~~

Thursdav. Auaust 21. 2008 10:37:44 AM Page3of35 1

~~ ~

IUsing the following list of components and assuming them to be in service, select the choice which contains ONLY those locations of a CCW leak that could cause Radiation Monitors 2R17A and 12R17Bindications to rise. There is no RCS primary to secondary leakage.

~~~~ ~

to Loss of Component Cool

~~~ ~

~~ ~ ~ ~ ~ _ _ _ _ _ ~ _ _ _ _ _ ~

~~~~

~ ~~~~~~ ~ ~

ing through them at higher age would not be radioactive.

The RCP Seal Water HX seal water is at higher pressure than CCW system. The SFP HX is at a lower

,pressure so leakage would be out of CCW system. The R19A SG Blowdown Rad monitor, would not cause counts to rise, since normal ops assumed in stem does not infer SG tube leakage. The distracters containina I t and V are incorrect.

Component Cooling Abnormality

~~~~ ~~

, ABCCOI E006 For the following analyzed transientslaccidents:

a) Determine the exDected alarms and indications.

bj Describe the anaiysis assumptions.

c) Describe the protective features that mitigate the event.

~pd) Describe the expected plant response.

~p

~~ _____

~ ~ ~ ~ ~ _ _ _ _ _ _ _

_ _ ~ ~ ~ ~ _ _ _ p Thursdav. Auaust 21.2008 10:37,44 AMpp Page 4 of 35

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

'Given the following conditions:

1- Unit 2 was operating at 100% power, steady state.

'- The following indications have changed as indicated over the past 2 minutes.

l - PZR Master Pressure Controller output is rising slowly.

1 - Charging flow has lowered from 86.4 gpm to 84 gpm and continues lowering slowly.

- Letdown flow has lowered from 80 gpm to 79 gpm and continues lowering slowly.

- PZR level has risen from 57.3% to 57.8% and continues rising slowly.

Which of the following identifies the problem occurring, and how will the malfunction be addressed?

~ ~~

~ -

psig. Enter AB.PZR-0001, Pressurizer Pressure Malfunction.

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

-. - ~

the controlling PZR pressure channel has failed at 2232 psig.

~~~~~~ ~ ~

Enter S2.OP-AB.PZR-0001, ipressurizer Pressure Malfunction.

~~~~ - - ~ ~ ~ -

~ ~~~ -~ ~ --

~~~~~~ -~ ~-~- ~ ~ ~ _ ~~~~ _ ~~ _ ~~~~ ~

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

'The controlling PZR level channel has failed at 55.5%. Enter S2.0P-AB.CVC-0001, Loss of G 2 . 0 7 Makeup flow indication

~~~ - ~~~

~ ~~~ - - ~ ~ _ _ ~ _ - ~ ~~

I 3.1 3.1 55.43(5)PZR pressure indication is purposely NOT given in stem as it would make question too easy.

Operator is required to use diverse indications to diagnose problem. With the MPC input from controlling channel failed slightly above actual pressure at beginning of transient, the MPC output will rise, since it lthinks pressure is too high, and will act to deenergize heaters and open spray valves. As actual pressure ldrops due to the rising spray flow, the RCS will expand into the PZR, causing PZR level to rise. This will cause charging flow to lower as PZR level rises above program. Letdown flow will lower as RCS Pressurizer Pressure Malfunction ABPZRI E003 a) Determine the appropriate abnormal procedure.

b) Describe the olant resDonse to acitons taken in the abnormal Drocedure

- ~~~

~~

cj ~

Describe the h a 1 plant conditio

~~~~~~ - --

~ ~~

~~ ~

ABPZRl EO01 Describe oDeration of the Pressurizer Pressure control svstern as aDDkd to S2.OP-AB.PZR-O001(Qb

~ ~ ~ ~~ ~ ~

Thursday, August 21,2008 10:37:44 AM Page 5 of 35

'Given the following conditions:

- Unit 1 is performing a Rx startup.

I- Rx power is 2E4 cps.

I- SUR is .2 dpm.

1- During the refueling outage, BOTH IR NI detectors were replaced.

~- IR NI indication for both N35 and N36 is flashing at 1x10-I I A .

'Which of the following describes the status of the IR instrumentation, and the required action(s) that will be performed?

BOTH IR Nl's.. . -

- ~ ~ ~

Nl's INOPERABLE and enter TS 3.0.3.

~ _ _ ~ ~~ ~ ~~ - -~ ~ - ~- ~ -- -~ ~ -

&e under compensated. Stabilize power, block SR Hi Flux trip, and correct compensating L

voltage problem for BOTH IR Nl's prior to exceeding 5%-~~~

- ~ - ~ ~

Rx power~ ~ ~ -

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

are overcompensated. Stabilize power ~cP-6and correct compensating voltage problem for at Ileast -~ ONE -~ IR NI within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or open Rx Trip Breakers within

~ - - ~ - ~ -~- the next~- 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. -~

~

~~

Instrumentation:

20, 000 counts, the decade of overlap should already be present. With the Hi Flux Trip at 100,000 counts, there can't be proper overlap. With no other information in the stem to provide inference of any other problems with the Nl's EXCEPT that both IR detectors were lreplaced, the IR Nl's should be declared INOPERABLE. There is only an action in 3.3.1.I for ONE INOPERABLE IR NI, with BOTH INOF'ERABLE TS 3.0.3 is entered. C is correct because even if the candidate though they were reading low due to overcompensation, the P-6 block would not be manually performed without the power above P-6 interlock to allow P-6 to be blocked. Under compensation would cause a hiaher than exDected readina.

4 "

I RefwenceTitle I b m i n q Objectives

. ..". - . . . . ..- . . __ ___" . __ . I .__I. ".. _- .. ..- . .-. .. .

~ EXCOREEOOS

~~- ~

Identify and describe the Control Room controls, indications, and alarms associated with the Excore Nuclear Instrumentation System, including The Control Room location of Excore Nuclear Instrumentation System control bezels and indications The function of each Excore Nuclear lnstrumentatlon System Control Room control and indication.

The effect each Excore Nuclear Instrumentation System control has upon Excore Nuclear Instrumentation System components and operation.

The plant conditions or permissives required lor Excore Nuclear Instrumentation System Control Room controls to perform their intended function.

The - associated

-setpoints -- ~ with

~-the Excore Nuclear Instrumentation System control room alarms.

~- - - -~~

~ -~ ~

~~ --

I I- Thursday,

~~

~-

~

~-

August 21, 2008 10:37.zAM- -

~~

~ ~

1 Page6of35

-~

~

~~~~~

Thursday, August 21, 2008 10:37:44 AM

~~ ~

Page7of35 1

lGiven the following conditions:

- Unit 1 is operating at 15% power.

1- Main Steam Dumps are in MS Pressure Control-AUTO.

- PT-507 Steamline Pressure Transmitter, fails high, and an automatic Safety Injection initiates 1

on Hi Steam Flow and Lo Steam Pressure after all the steam dumps open fully.

- The 24MS167 valve status light on 2RP4 is flashing.

or action,

-~~

which of the following would -

identify that the-24MS167

~

~ -

has failed to shut?

~ -~ - ~ ~ ~

- - SG NR level would shrink out low I'24 2 1 --Iconduct

- Of Operations ~ - - -~ ~~~~-

- ~ -~ - - - -

~ ~- -- --

~ ~

- - -~~ ~~~- - ~

main& primary and seconda

~- - ~ _ -- - ~ -

-2.3

~

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

ual KA is 0040 Steamline Rupt 1.45 Ability to identify and interpret diverse%dications to validate response of another indication. SRO Value 4.3 55.43(5) The auto SI on high stm flow coincident with lo steamline pressure will initiate a MSLI. There is some period of time in seconds in which ONLY the failed open 24MS167 will continue to pass flow through to the steam dumps before they automatically shut on lo lo tave at 543 degrees. This will cause NR level, WR level, and steam pressure for 24 SG to all be lower than the other 3 SGs. The NR level IS a good distracter if the operator thinks it lwill shrink low out of the indicating band, which it won't because of the low initial power level. The High Steam flow bistable does not "lock in" its indication on 2RP4, and will clear when the high flow condition ,

Overhead Window F

~

-~ - ~~ ~ -~ ~ ~-

-- MSTEAME003 NCT Draw a one-line diaqram of the Main Steam Svstem which indicates the following:

Major Component Main steam System Componenis

~~ ~

Steam Generators Steam Generator Flow Restrictors Safety Valves Atmosphenc Steam Dump Main Steam Warm-up Valves Main Steam Drain Valves Main Steam Isolation Valves Main Steam Mixing Bottle Main Steam Stop and Governor Valves Turbine Bypass Valves

b. Major Main Steam System Flowpaths Main Steam Flow Auxiliary Feedwater Pump Steam Flow Turbine Bypass Steam Flow Main

~

Steam line Drains- ~-

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

r ~~

Thursdav. A u o u s ~ 2 l 2008

. 10:37:44 AM

IGiven the following condition:

- Unit 2 lost off-site power while in MODE Ei

- When attempting to reset the SEC IAW S2.OP-AB.LOOP-0001, Loss of Off-Site Power, 2A SEC would not reset.

I- An operator deenergized the 2A SEC cabinet by opening its power supply breaker from 1 the Vital Instrument Bus.

~- Off site power was subsequently restored, and AB.LOOP-0001 has been completed and exited.

IWhich of the following describes how this breakers status will be tracked until it is restored to Normal config uration? -~~ ~

~ _ _

P-AA-108-101-100 L-~Tagged/Not

- - ~ - ~ Tagged Discrepancy sheet.

_ __ _ _ _ _ _ ~ ___ ~

~ - - - ~ _ _ ~- ~~ ___

- ~ _ _ _ _ _ ~

~~ ~ ~~ ~ _ _~- -____ ~- - ~ - - ~~ ~ - __-

'When S2.0P-AB.LOOP-0001 was completed, the Work Clearance Module (WCM) was updated IAW OP-AA-108-101-1002.

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

-~ - ~ - - -~ __ ~- ~

~ - - _ _ ~ - ~ -- ~ -~ - - -~ ~ ~ ~ ~~ -

IAW OP-AA-108-101, Control Of Equipment and System Status, Attachment 1, Abnormal Component Position Sheet.

~ -- ~ - _ _ ~ - - -- ~~ ~~ -~ ~ -~

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

OP-AA-108-101 directed updating the breaker position in the WCM during performance of 0000566214 re incorrect becau has been terminated (applicable sections completed). C is incorrect because it is a condition directly lprohibited to use an ACPS IAW OP-AA-101 Section 4.1 . I .

MISCAP007 LOR NCT Given a set of plant conditions and SH.OP-AP.

conditions and summarize the actions necessary to properly maintain the status of systems and components, IAW SH.OP-(#laterialRequired fiir bcamilwttion 1

' ~

Thursday, August 21,2008 10:37.44 AM Page 10 of 35 L----

Which of the following describes a situation which is considered an unmonitored radioactive liquid l reIease?

,Waste from 21 CVCS Monitor Tank, and the 21WR25, 21 CVCS Monitor Tank drain valve is lopened.- ~ ~ ~~~ ~

~~~~~ p

~ ~ _

p

~

p _ ~~

~ ~ ~~~ ~ ~-~

p~~ ~~~ ~ ~~ ~ ~~~ p ~ p ~ ~ ~ ~~

iA Radioactive Liquid release is performed IAW S2.OP-S0.WL-0002, Release of Radioactive ILiquid Waste from 22 CVCS Monitor Tank with the knowledge that 2R18, Liquid Radwaste

[Effluent Line Monitor, is INOPERABLE.

p p ~ ~ ~~ - ~ ~ ~ p ~

~p ~

p ~ . ~

p ~

p p

~

p~~

p - ~

~~ ~ ~~p~ ~~~~~ ~p~

~~~~~ ~ p - p ~ p~~

with the 2FR1064, Radwaste Overboard Discharge Flow Monitor INOPERABLE, sample flow iis lost to 2R18 during an authorized radioactive liquid release from 2 WMHUT IAW S2.0P-

~SO.WL-0003,Release of Radioactive Liquid Waste from 2 Waste Monitor Holdup Tank.

pppppppL ~

_ _ ~

~ p ~- _ ~_

~~~

~

_ _ ~ ~~~~~ ~ _ _ ~

~ p -

p ~ _ _ ~ ~ p p ~

I ASpent Fuel Pool tell-tale line is dripping 20 drops per minute. __

AGj

~p

~~ ~~~~~p S Salem 1 & 2 25

~p~ ~ p ~ ~ p ~p~~ ~~~ ~~~~ ~p ~~ ~~

12.4 Fmergency Procedures / Plan ~ p p p ~~ ~ ~ ~ p _ _ ~ ~ p p p 2.4.21 Kno ecl to assess the status of safety functions including: 1. i 3.7 4.3 Reactivity control 2. Core cooling and heat removal 3. Reactor coolant system integrity 4.

i ped to the drain header 0 sh. I ) , and will be monitored. The R18 is allowed to be INOPERABLE as long as double lsampling/analysis is performed and the overboard flow monitor is OPERABLE, as this provides a method iof monitoring the release based on how hot the tank is, the dilution flow, and the rate at which it is ldischarged. The Spent Fuel Pool Liner Leak Detection system consists of channels under the Spent Fuel lpool, which lead to 17 tell tale drains. 'The small (20 dpm) amount of leakage will be contained in the channels, and the tell tales collect in a common sump, which is then pumped to in service Waste Monitoring hold Up Tank. If the 2R18 is OPERABLE during the pre-release preparations, then losing sample flow to it during a release will require termination of release since it is unmonitored.

Release of Radioactive Liquid Waste from 21 CVCS Monitor Tank ~~

~~~~ ~ ~ ~~~~ - p~

appropriate Technical Specifkition action. (License Operator and STA only)

NCT State the Technical Specification associaited with the component, parameters and operation of the Radioactive Liquid Waste System including the Limiting Condition for Operation(s) (LCO) and the applicability of the LCO(s) (Non-licensed

~

Operator) ~~~~ ~~

~p - p ~ ~ _ p _ ~ ~ ~ ~ p

~~p ~

WASLIQE012 LOR NCT Discuss the procedural requirem d with the R a d i o a z e Waste System, includin L

of major precaution and limitations in the R

~~

~~ ~ ~ p p - p ~ ~~

~ ~

i

~- p ~ p 1 Thursday, August 21,2008 10:37:44 AM- Page 11 of35

~ ~p~ ~-

'Given the following conditions:

~- Unit 2 is operating at 100% power.

i- A containment pressure relief in in progress IAW S2.OP-SO.CBV-0002, Containment i Pressure-Vacuum Relief System Operation.

- 2VC5 and 2VC6, Containment Isolation Valves, are open.

,- An RCS leak occurs, leading to a F b trip and Safety Injection.

I IWhich of the following identifies how containrnent integrity will be restored after the SI is initiated, and what action is required if the 2VC5 and 2VC6 do not shut, and can not be shut, when ldemanded?

Reference provided. -~

-~~

utomatically

~~ p ~ _ _ _ ~ ~ _ _ _ - ~ -~ ~ p ~~ _ _ ~~ ___- -___- ~ ~ p

'The 2VC5 and 2VC6 will automatically shut directly from the CVI signal. Declare a Site Area

'Emergency.

L-p _ _ ~

p ~ _ _ ~~ __ -~ ~~-

812512008 iate Phase A isolation occurred based on an RCS leak, so the leak has to be bigger than what a centrifugal charging pump can handle to maintain PZR level > 17%. This would be at least 3 and could possibly be 4 points on ECG 3.2

'RCS barrier. Additionallv. the oDen containment as stated in the stem would be 2 points on ECG 3.3 IContainment barrier, so {he point total would be5 or 6, which is a SAE.

~ --- - ~ p p p ~p~ p~~

including: (Licensed Operator and STA Only) a) The Control Room location of Reactor Protection System control bezels and indications b) The function of each Reactor Protection System Control Room control and indication c) The effect each Reactor Protection System control has upon Reactor Protection System components and operation d) The plant conditions or permissives required for Reactor Protection System Control Room controls to perform their intended function control room alarms

- ~ p ~ p -~

~ - p p p ~- ~~~

Question Awortiwcation Method:

___ ~~~ - p ~ ~

Thursday, August 21, 2008 10:37.44 AM Page 12 of 35

- _ _ _ ~ ~ _ _ p p~

~

Thursday3gust 21, 2008 10:37:44 AM - Page 13 of 35

[ __ ~ _ _ _ ~

~

Given the following conditions:

~- Unit 1 is operating at 100% power when it receives an inadvertent Safety Injection signal from the RPS system.

- The control room crew has progressed through the EOP network.

,- Operators are preparing to exit EOP-TRIP-3, Safety Injection Termination.

t which of the following conditions, if present when determining the proper procedure transition, iwould require the transition out of TRIP-3 be made to S I .OP-I0.ZZ-0006, Hot Standby to Cold Shutdown? ~~ - ~ ~~~ ~ - ~~

NO RCPs are operating.

~~~~ ~ _____- ~~~ ~~~ -~ ~~ - - ~- ~~~~ ~ ~ _ _ _

~~~

~~ ~~ ___ ___ -~ -~~ _ _ _ _ _ ~-

~-

NO - _ control room air - conditioning chilled water pumps are OPERABLE.

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

~

d S Comprehension Salem 1 & 2 00WE02G222

~~

~~

ired it would be to IOP-8 Maintaining Hot Standby. B is incorrect because the EDGs are secured in TRIP-3 if off-site power is powering the vital busses. D is correct because TSAS 3.7.10 requires 2 chilled water pumps, and there is no action for 2 INOPERABLE pumps, which is TS 3.0.3. With the unrt already in Hot Standby following the trip, the unit must be placed in Hot Shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. C IS incorrect because it is onlya 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> TS and would not TRP003E008 A. Determine a discrete path through the EOF.

rmi P I Requiredfor Exirmination I Question Modification dll@hed: I Page 140f 35 I

Given the fo Ilowing co ndiions:

1- Unit 2 was operating at 100% power when 21 SGFP tripped.

~- The Main Turbine ran back as expected.

~- Control Rods did not respond in AUTO, and the RO is inserting them in manual to restore Tavg

'to program.

- PZR pressure is 2238 psig.

- OHA E-20, PZR HTR ON LVL HI is in alarm.

- BOTH PZR Backup Heater groups are OFF.

Which of the following explains how the heaters could be OFF in this situation, and the action that should be performed?

The Backup Heaters are OFF because.. - -~

~-

. -~ - ~~ ~ _

-~

~-

_ ~~ ~- -

r---- ____ ~~~

ithe PZR Master Pressure Controller (MF'C) output is demanding ALL PZR heaters off in iresponse to the PZR pressure > 2235 psig. Ensure Charging System Master Flow Controller lis responding to reduce charging flow IAW S2.OP-S0.CVC-0001 , Charging, Letdown, and Seal Injection. -____ - - ~~ ~- - ~ ~~~ - ~-~ -~ ~ -

~ - _ _ _ _ - ~ - ~ - ~~~ ~-~ -____ _____- ~

the PZR MPC is a PID type controller, and the signal to turn on the heaters has not been ldeveloped. Cycle MPC from AUTO to MANUAL to remove the reset windup, and return to lAUTO IAW S2.OP-SO.PZR-0005, Pressurizer Pressure Control System operation.

i -- ~ - ~ _ _ _ _ _ - ~- - - ~ __ - ~~ --__ -~ -

~ ~ _ _ _ ~________ -~ ~ - -- - ~- ~~- ~~~ ~~ ~

lthe operating charging pump has tripped, causing a Letdownlsolation, which blocks the heaters from energizing. Start an available charging pump IAW S2.OP-AB.CVC-0001, Loss of harging. ~~~

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

~ ~~ ____ ~ --

their respective AUTO/MAN controls &e in MANUAL. P k c e BOTH sets of heaters in ON and will be to turn OFF the heaters. Use of the term reset windup refers to the way of removing saturation conditions in the controller and was used in the past on Salem's MSIO controllers. C is incorrect because there is no interlock between charging Ipumps and heaters, the interlock is between charging pumps and lletdown isolation. D is correct because MANUAL mode of operation overrides any other automatic signal to operate heaters.

ReferenceTitle I Overhead Annunciators Window E.

~~ ~~

PZRP&LEOOG Outline the interlocks associated with the following Pressurizer Pressure and Level Control System components.

P-7 and the Pressurizer Low-Pressure Reactor Trip P-I 1 and the Pressurizer Low-Pressure Safety Injection Block P-7 and the High Level Reactor Tnp

-_____- ~ ~~~

_ _ _ _ ~ - ~

~- -

Thursday, August 21, 2008 1037:44 AM

_____~- - - _ _ _ _ ~

Low-temperature interlock with PORVs Isolation of Letdown Orifices Low-level Pzr Heater Cutout Material RequlredfofExaminatton 1 Question80w: New Question ModincationMethod:

Question Souree Comments: 1 I Thursday, August 21, 2008 10:37:44 AM

~

Page 16 of 35

given the following conditions on Unit 2:

1- A LBLOCA has occurred.

i- Operators are performing 2-EOP-LOCA-5, LOSS OF EMERGENCY RECIRCULATION.

i- Containment pressure is 15.1 psig and is rising slowly.

Which of the following describes how the Containment Spray system will be operated, and why?

2.4 IEmergency ~Procedures

~ _/ Plan_ __

_ _ ~ - ~~

~~~

~~

~~~

~

~~

~~ -~

~

&e of _ the_ bases

~ - _ _

for prioritizing

-~ ~

s

~~ ~~~~

~~~ ~

55.43(5) Upon entering FRCE-1, s are to be operated IAW LOCA-5. The basis document states that this is because in FRCE, maximum available heat removal system operability is warranted to reduce containment pressure, whereas in LOCA-5 a less restrictive criteria permits reduced spray pump operation depending on RWST level, containment pressure, and # of CFCU's operating. The less restrictive criteria in LOCA-5 is used because recirculation flow to the RCS is not available, and it is very important to conserve RWST water, if possible, by stopping containment spray pumps. So while the operator WILL enter FRCE-1 due to IPURPLE path of containment pressure > 15 p i g , the containment spray pumps will be operated IAW ILOCA-5.

~ -~

FRCEOOE006 Describe the basis for each step, caution, and note in 2-EOP-FRCE-1 thru

/Yaterkl Required for Exmlnation I Question Souroe: Facility Exam Bank QuestionSotme Cwnmentii:

Page 17 of 35 ,

unit 1 is in Mode 5 with a containment purge in progress.

-~ ____ - ~ -

~~ ~ ~~ -- ___ ~. ~

Failure of the Plant Vent Flow Monitor with all Auxiliary Building Exhaust Fans operating.

~~~ - ~~

~FailGeofRMS Channel R l 1A, Containment Particulate, with Channel R41D, Plant Vent Noble IGas Release Rate Composite, operable

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

'Failure of RMS Channel R12A, Containment Noble Gas, with Channel R41D, Plant Vent Noble Gas Release Rate Composite, operable

~~

~- ~- ~ _ _

,at a negative pressure, the potential exists for an unmonitored release from the Aux Building. P&L 2.10 istates that if the plant vent flow monitor is unavailable, then ALL ABV fans must be in service. As long as

-the R41D remains OPERABLE, the 11A or 12A are not required.

~ - - - ~~ ~ -~ ~ _ _ _ ~

Containment Purge to Plant Vent

~--

1 IFacility Exam Bank Direct

_ _ _ _ ~ -____~

Thursday, August 21, 2008 10:37:44 AM Page 18 of 35

- ~~~ - ~~

Given the following conditions:

lI-- Salem Unit 2 is operating at 73% power.

An electrical panel in the CW building has been exposed to the rainy outdoors I environment when the roof panel above it is removed.

~- Operators perform a power reduction to 40% power in response to Circulator malfunctions associated with the now wet panel.

I-Operators trip the Main Turbine, and continue lowering power in response to lowering condenser vacuum.

I- 23 SG NR level detectors (2 of 3) see a Bad Quality input, and the 23 loop ADFACS swaps to manual. Actual NR level is 30% and rising when the swap occurs.

Which of the following describes how the 23E3F19 and 23BF40 will be operated?

~~ ~~ ~~ ~~

~~~~ ~

'While continuiniin S2.0P- Tur elow P-9, BOTH 2 i23BF40 will re&ire operator control to prevent a high level in 23 SG.

~~ ~~~~~~ ~ ~~~ ~~ ~~ ~~ ~~ ~ ~~~ ~~~ ~~~ ~~~ ~ ~

~ ~~~~~~~ ~~~~~ ~~~~ ~

~ ~~~~ ~~ ~~~~~ ~~~

Using S2.OP-AR.ZZ-0007, Overhead Annunciators Window G for OHA G-7, ADFCS Swap to

'Manual, BOTH 23BF19 and 23BF40 will require operator control to prevent a low level in 23

~SG. ~~ ~~~ ~~~~~ ~ ~~ ~ ~~ ~~ ~~~ ~~~~ ~~~ -

~~~ ~~~ ~~~ ~~ ~~

~~~~~ ~ ~~~~ ~ ~~~ ~~ ~~ ~~~

IAfier entering S2.OP-AB.CN-0001 , Main FeedwaterKondensate System Abnormality, ONLY ~

'the 23BF40 will require operator control to prevent a low level in 23 SG since 23BF19 closed ion Feedwater Interlock.

~~ ~

~~~ ~ ~ ~~ ~~~~ ~~~ ~~ ~ ~~~ ~

~ ~~~~ ~~~~ ~~~~ ~ ~~ ~ ~~ ~~~~

~~ ~ ~ ~~~~

Using S2.OP-AR.ZZ-0007, Overhead Annunciators Window G for OHA G-I 5, ADFCS Trouble, ONLY the 23BF19 will require operator control to prevent a high level in 23 SG since 23BF40 flow is small compared to 23BF19 flow.

~~~~~ ~ ~~~ ~ ~~~ ~ ~~ ~~~ ~ ~ ~~ ~~~

A2. ~ Ability to (a) predict the impacts of the following on theM2n Feedwater System and (b) based &those trol, or mitigate the consequences of those abnormal operation:

ue to the problems associated with controlling SG level after the turbine was tripped and power reduction continued. The stem states that the power reduction continues after the turbine was tripped.

AB.TRB is required to be entered for the turbine trip, and operators stay in AB.TRB to perform the power reduction since vacuum is still lowering (Steps 3.11-3.12) With actual SG level at 30%, below

,programmed level, the demand on the BF19s and 40s will be higher than that required to maintain normal llevel. This will cause NR level to rise without operator action to correct it. All of the distracters contain lincorrect actions, but their procedures are applicable to the conditions.

~ ~~

Technical Bases Document. ~~ ~ ~~

~~ ~

ABTURBE005 a) D e K k i n e the appropriate b) Describe the plant response to actions taken in the abnormal procedure.

~ ~~~~~

De I plant that is ed b nor ure.

~~ ~~

Page19of35 ,

Page 20 of 35 Given the following conditions:

- Unit 2 is operating at 100% power.

1- 2B EDG is running in parallel on the 2B 4K:V Vital bus.

'The 500 KV ring bus loses all off-site power hot already running. Enter EOP-TRIP-1 Rx Trip or Safety Injection.

~ ~_____ -~______ ~ -~ ~- ~ ~~ ~ ~ ~ _~ ~- _ ~~ _____~

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

~ _ _ _ _ _ _ ~

~

~ ~ ~~~

2B EDG will remain connected to 2 9 vital bus. NO Blackout loads will start since 2B SEC still senses voltage on 2B 4KV Vital bus. Enter S2.OP-AB.LOOP-0001, Loss of Off-Site Power.

- - ~ ~ ~______ _____ ~-

~ - ~ - ~

_____ _~ ~ - _ ~~ ~~

~~

~~~ ~ ~ ~~

-~ -~ ~~

~

,2B EDG output breaker will trip, 2B bus will strip, the EDG output breaker will close, and IBlackout Loads will sequence on the bus. Enter EOP-TRIP-1 Rx Trip or Safety Injection.

i~~ ~~ ~~~

_ - ~ ~ _ - ~-~

~ ~ -_- ~ ~

~-

_ ~ - -

\zl

~- ~~ ~~ ~ -~ -____- ~- ~~

~~ ~_ _____~ ~~ - ~ -

~2BEDG output breaker will trip, 2B bus strip, the EDG output breaker will close, and e bus.Enter S2.OP-AB.LOOP-0001, Loss of Off-Site Power.

~- ~ -~ ~- ~ ~~

_~ ~~

lity to (a) predict the impac erators and (b) based on--

those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal e busses if energized by opening EDG output

, and sequence loads on due to start current

~ ~~~~ ~- _

-____ ~~

~~ - - ~ _ ~~ ~ - ~~ -~ ~~~~~

~~

~~

2A Generator Endurance Run EDGOOOE008 LOR NCT Identify and describe the Control Room contro Generator, including:

The Control Room location of Emergency Diesel Generator control bezels and indications. (Licensed Operator & STA only)

The function of each Emergency Diesel Generator Control Room control and indication. (Licensed Operator & STA only)

The effect each Emergency Diesel Generator control has upon Emergency Diesel Generator components and operation (Licensed Operator & STA only)

The plant conditions or permissives required for Emergency Diesel Generator Control Room controls to perform their intended function.

The setpoints associated with the Erne Page21 of35 I Thursday August 21 2008=37:44 I l p ( L -~ AM

Given the following I- 1 WMHUT is in recirc, a sample has been drawn and is in the process of being analyzed.

~- The RWO mistakenly places 1 WMHUT in service.

- One hour later, the RWO recognizes his error when the high level alarm for 1WMHUT lannunciates at the Unit I104 Panel.

1- The RWO immediately returns 1 WMHUT to recirc.

IWhat effect, if any, will this have on the release preparations for 1 WMHUT IAW S I .OP-S0.WL-

~0003?

The CRS will direct the.. .

I Reference provided. -~ - ___ __

~ ~~

~~

~ _ _ - _ _ __ ~~ ~~

~~

~~

~~

~ -~ __

~

~

~- ~~

~- - -~

~

~ -_____ -

RWO to verify the additional volume a tank does not exceed 1YOof total tank volume

[068000A202 -~

~ ~-

level from the tank curve, and Ithen will yield a recirc time IAW Att 2 of WL-0003, of 306.67 minutes prior to sampling. This is 5.1 1 Ihours. Any extrapolation error would be in the lower tank level based on the way the rectangle is placed on the graph for the high level alarm. The 4.33 hour3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> distracter is if the candidate uses the wrong tank (CVCS Monitor tank)

I -~ - ~- -~ ~

Reie+c&Title I Release of Radioactive Liquid Waste from #I WMHUT Leaminqobjectives __ . -_

__ l_l - , _I - _ - -

• I-- I WASLIQEO-12 LOR NCT DISCUSS the proceduya: ;equirementsassociated with the Radioactive Gquid Waste System, including an explanation

_ _ ~~- ~~

2008 10:37:44AM Page22of35 1- Thursday, August 21, __

~ _ _ ~~ ~

l

, Page23of35

,Which of the following describes why FRCE-3, Response to High Containment Radiation, is a

'YELLOW path FRP performed at the discretion of the CRS, instead of a RED or PURPLE path lFRP, which is REQUIRED to be performed upon meeting its entry requirements?

~~

~~

~

~

undo actions t re ntered Pe 1completion. ~- _ _ _ _ ~

~ _ _ _ _

_____ ~

~

~~

~~~~

~~

~

~~

~ - ~

~

~

~-

~-

~~~~

~~~ - ~~

~ _ _

~

i The actions performed in FRCE-3 are redundant to actions which are expected to have

'occurred

~ _ automatically._ ~~ ~ ~~~ ~- ~~ ~~ ~~ ~~~~ ~- ~~ ____- ~~~ -_____~

~~ ~- ~ _ _ _ _ ~- ~~ ~~~ ~ ~~~ ~~~ ~- ~~~ ~~~~ ~

IConditions requiring entry into FRCE-3 would not result in the release of radiation outside the lcontainment building.

~~~ ~ _ _ _~ ~_ _ ____~ ~~

~- ~~ ~___ ~~~ ~ ____ ~~~~

~ ~~~ ____~ ~~~ ~~~ ~~~~ ~~~ ~~~ ~~~ ~- ~~ -~ ~ ~~~ _____

h conditions have degraded to where the entry conditions are met for FRCE-3, then higher priority

_ RED _ or ~PURPLE

~_ _ _ ~ _ path

_~_FRPs ~ ~~~ will be present,

~~ ~- and must

~ __ be performed first. ~~ -~~~

~- -

~~ ~

ergency operations.

~~

rrect because there are only 2 "actions" taken in FRCE-3, to ensure the containment pressure/vacuum relief path is isolated through VC5 and 6, and starting CFCUs in low speed to ensure the filters are realigned. D is incorrect because ithe entry conditions for FRCE-3 is containment radiation >2R/hr. This does not always mean that other IFRPS will be in effect. A is incorrect because of B above, and the CRS should direct entry into a YELLOW Dath onlv when its Derformance will not imDede other actions critical to plant recovew.

Reference Title Learning Objectives TRPOOI E007 Describe the Critical Safetv Function Status Tree hierarchv-in accordance with SC.OP-AP.Z2-0102(c1)andEOP-CFST-1 Page24of35 1

Given the- fol lowing co nditions:

I

- Unit 2 is operating at 100% power.

- 21 Control Air Dryer is in service and operating in the AMLOC mode.

~- When automatically switching from the leRt to right desiccant chamber, the CAI 884, Purge Exhaust Valve for the left dryer opens as expected.

Which of the following describes the consequences of this valve failing to close when the desiccant

~~~

t action(s), if any, will be required to be performed?

~

~~

~ ~

-~

~ _ _ _ _

~

- ~~

~ _ _ _ _

purge valve on thedesiccant chamber, even as it iremains out of service, will lower control air header pressure to the point of starting the Unit 2

[ECAC. Enter S2.OP-AB.CA-0001, Loss of Control Air.

____ ~ _ _ _ _ ~~ _ _ _ _ ~ ~~ ~~ ~~ ~~~

-~

~ _ _ _ _ _ _ _ ~ ~~~ _ _ _ _ ~ ____ ~ ~~ ~~~ ~- ____ ~- ____ ~~~ ~~

IThe affected desiccant chamber will still be automatically be placed in service when the timing lsequence demands it, and a rapid lowering of control air header will occur. Enter S2.0P-lAB.CA-0001 Loss of Control

_ _ _ _ _ _ _ ~

I

_ _ _ _ ~ ~~-

Air.

~ _ _ _ _ ~~~ ~~ ~ - ~ ~ _ _ _~ ___ _ _ ~~ -~ ~

~ ______~ ____ _ _ _ _ _ ~ -~ - ~~~ ~~~ ~- ~~ ~~~ ~~~ ____ _____

The affected desiccant chamber will be interlocked from going in service, and the 21 Control Air Dryer will be removed from service IAW S2.OP-SO.CA-0004, 21 Control Air Dryer Operation. ____ ~ _ _ _ _ ~- ~ ~~ ~ ~~ ____ ~-

~~~~ -____ ~~~ ~~ ~- -~ ~~ ~~ ~ ~ ~- ~-

The continued loss of air through the purge valve will lower control air slightly. Use S2.0P-

~SO.CA-0001,Control Air System Operations to remove 21 Control Air Dryer.

-~ -____ ~~~ ~~~ ~~~ ~~

____ ~~~~ ~ _ _ _ _ - ~~

~~ -

~~- ~ ~~~

____ ~~ ____-~ ~ - - ~ ~

--A .2 Ability to (a) predicttheimpass of thefollowingon thelnstrument Air System and (b) based on those ate the consequences of those abnormal operation:

above. C is correct because of B above, and the local panel alarm will alert operators to the fact that the sequence has not completed. The SO provides direction for removing a dryer from service as the redundant dwer(22) _ . .is normally in service also. D is incorrect because the procedure referenced will not direct CA dryer operation.

Reference Title I barrpinq Objactivea

._l__lll - _-_I I _ .. I-. . . .- . . . . .. - . . .-

- CONAlRE004 NCT Describe thefunction of the following coinponents and how their normal and abnormai operation affects the Control Air

- ~ -

System.

Control Air Dryers Control Air Receivers Emergency Control Air Compressors (ECACs)

Emergency Control Air Dryers Excess Flow Check Valves (EFCVs)

Station Blackout Compressor CA Containment Isolation Valves PORV Control Air Accumulators

~

~~ ~

-~ ~ ~

~~ ~

~~~~~

~

Thursday, August 21 2008 10:=44

~ _ _ _ _ ____

AM ~

~

1 Page25of35 1

IMaterial Required for Examination 1 QuestionModification Method:

-~ ~~ ~~~ ~~

I Thursday, August 21, 2008 10:37:44 AM- 1 Page26of35

(Chemistryreports that the lithium concentration in the RCS is approaching its upper limit- What is

'the preferred action that should be taken to restore the lithium concentration to its proper level?

maximized to accelerate the cleanup using the CVCS mixed beds.

not provide lithium control. Placing the cation bed in service per chemistry direction will remove lithium from the RCS. B is incorrect because the cation bed is placed in service at normal flow (75) gpm, and the mixed bed does not lower lithium. C is incorrect because it would take a long time to lower lithium this way and create a large amount of liquid waste. D is incorrect

~~ ~ ~~ - ~ because the cation bed is placed in service.

~~ ~~ ~- ~

CVCSOOE004 LOR NCT Describe the function of the following components and how their normal and abnormal operation affects the Chemical and Volume Control System:

LetdownKharging Letdown lsolaiton Valves, CV2, CV277 Regenerative Heat Exchanger Letdown Orifices Letdown Orifice Isolation Valves, CV3, CV4, CV5 Letdown Releif Valve, CV6 Letdown Line Containment Isolation Valve, CLV RHR Flow Control Valve, CV8 Letdown Heat Exchanger Low Pressure Letdown Control Valve, CV18 Temperature Control Valve, CV21 Demineralizers (Mixed Bed, Cation, and Deborating Inlet Valve to Deborating Demin, CV27 Reactor Coolant Filter Diversion Valve, CV35 CVCS Holdup Tanks Volume Control Tank VCT isolation Valves, CV40, CV41 Chemical Mixing Tank Charging Pumps (Centrifugal and PD)

Miniflow Recirc. Valves, CV139, CV140 Seal pressure Control Valve, C W 1 Chg. Line Containment IsoI. Valves, CV68. CV69 Charging to Loop 3 Valve, CW7, Loop 4 Valve, CV79 PZR Auxiliary Spray Valve, CW5 CCP Flow Control Valve, CV55

b. RCP Seal Water Seal Water Injection Filters Seal Bypass Flow Valve, CV114 Seal Water Return Isolation Valve, CV104 Seal Water Return Relief Valve, CV115 Seal Return Cont. IsoI. Valves, CV116, CV284 Seal Return Filter Seal Water Heat Exchanger

c. Excess letdown Excess Letdown Isolation Valves, CV278, CV131 Excess Letdown Heat Exchanger Excess letdown Flow Cotrol Valve, CV132 Excess Letdown Diversion Valve, CV134

d. Makeup Primary Water Storage Tank Primary Water Makeup Pumps Boric Acid Batch Tank Boric Acid Tanks Boric Acid Transfer Pumps Boric ,Acid Filter Boric Acid Blender Primary Water Flow Control Valve, CV179 Boric Acid Flow Control Valve, CV172 Charging Pump Suction Valve, CV185 VCT Makeup Isolation Valve, CV181

-~ - - ~- -

~~

Thursday, August 21,2008 10:37.44 AM Page28of35

- _ ~

- -~ ~ ~~

Given the following conditions:

1- Unit 1 was operating at 100% power when 11 Condensate Pump tripped.

I- One minute later, 11 SGFP pump tripped.

iWhich of the followina actions performed by the RO without CRS direction is NOT consistent with

- - ~ _ - - ~ ~ - ~ _ _ ~ ~ ~ - ~ _ ~ - - ~ _ P _ _ ~ - - ~ - ~

,Tripping the reactor based-~

-P ~ ~ on- SG NR - levels at~-

~~ P~ 18% and lowering.

~- - ~ - ~ - -

P ~~ - ~ ~

- - ~- ~ _ _ -~ ~ ~ - ~ ~~

'Tripping L ~ - the Main Turbine

~- based

- - -SG on - NR levels at 18% and lowering.

- ~ P - ~

~

~

~

P

~P P~~~ -- ~ - ~ ~ ~P~~ -

- ~~ ~ ~ _ _ ~

Conduct of Operations

~ -

what is inserting in auto following a load reduction is conservative. B is not correct because the operator is expected to take action prior to reaching an automatic trip setpoint. C is correct because tripping the

/turbine in this situation is forcing an automatic Rx trip on the turbine trip, and is not an expected action when power is >P-9 (49%) power. D is incorrect because reducing load is required for a SGFP trip, and L m f n g Objectives ZZ-O005(Q), Station Operating Practic utious and conservative decisions.

~~ P- - ~

~ - - ~ P Page 29 of 35

iDuring a refueling outage,who is responsible for the FINAL authorization priorto entry into a IPLANNEDSafe Status of Orange or Red? p - p p - p p p p p p p p - p p- - p - p p Shift Outage Manager.

p - - p p p - p -- - -

'Shutdown Safety Manager.

- ~ - - -~ -

director of Work Management.

~ does not provide final approval.

~- -~

- p I

Question Source Comments:

I Page30of35 i

~~ ~

Given the following conditions:

1- A male radiation worker at Salem Station returned 3 weeks ago from outage support at Limerick Station.

His Total Effective Dose Equivalent (TEDE) received at Limerick was 75 mrem.

- He received an exposure of 50 mrem to his right hand during an inspection while wearing special

'dosimetry.

I- The worker's current TEDE from Salem for this year is 75 mrem.

- The worker had an chest x-ray one week ago estimated at 25 mrern exposure.

excess of those authorized. I rn RO 3.7 55.43(4) Federal Limit is 5000 mr. Total TEDE is from all sites worked at.

_ _ _xray IMedical ~ not

- added to dose. Extremity dose not added- to dose.. ~ _ _ ~ ~~~ ~

Exposure Control and Authorization Lesmring OUectlvss

- -_ -_ - - _ _ - ____._I__ - ._ - - - -. - - - __ -

RxDCONEO&

~- ~-~

- List the following external radiation exposure limits, in accordance with Station Procedures. iOCFR20, and Reg Guide 8.13 A. 10CFR20 dose limits for external, internal, and total whole body, skin, extremities, and eyes, as well as extension limits and requirements

6. Administrative dose control levels for Category 1 and 2 Workers, as well as extension limits and requirements C. Reg. Guide 8.13 limits and administrative dose control levels for Declared Pregnant Women D. 10CFR20 and Administrative limits for members of the general public and minors E. Category 1 Radiation Worker

~ -_ ~ _ - -~~ ~ _~ _-- _~_ ~

~ ~

~

~

~

Q~&b?l l#OeliRcat/gn M&h&jd: Significantly Modified Q68091 Changed stem to result in different answer. Changed distracters.

- ~ ~ -

- ~~ ~ ~ ~ ~~~

l _ - ~ ~ ~ ~- ~~ ~-

Thursday, August 21, 2008 10:37:44 AM Page 31 of 35

Given the following conditions:

Unit 2 is operating at 100% power.

- 23 & 25 CFCUs have been C/T for emergent corrective maintenance for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

'- A crew of 5 people entered containment at 1415 to investigate a rise in the RCS leakrate, with a IHeat Stress stay time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

I- At 1416 22 CFCU breaker trips.

170minutes after the crew entered containment, the NCO reports that 2R12A CONTAINMENT GAS

,EFFLUENT is reading double what it was when the crew entered containment, and containment

'average air temperature is rising slowly.

Which of the following describes the effect, if any, on the personnel in containment IAW SC.SA-

~ST.ZZ-0001SALEM

~ ~~ - CONTAINMENT ENTRIES IN MODES ~ ~ -- 1 THROUGH

~ - _ - - -- 4? ~

~ -- -- -- - -- ~ __ - ~ ~~

- ~

~

~

~

The control room will contact the crew in containment by flashing the containment lights, and idirect them to -~ exit the containment.

~ - _ _ - ~ - ~ - ~ -

~~

_~

'Since the 2R12A is expected to rise with an RCS leak, the crew may remain in containment until their Heat Stress stay time is complete.

_ _ _ - ~ - - -

ANY increase in radiation levels in containment while it is occupied REQUIRES dispatching a

'Radiation

_ -~ _-

Protection technician

- ~ - ~~ ~ ~_ - ~- -

into--containment

~

~- to evacuate containment.

~~ ~ _ ~ ~ ~ - - ~ - ~ ~- - -~ -_

Personnel in containment may continue their inspection while monitoring for any continuing rise in radiation level. If radiation levels on 2R12A increase by a factor of 4 from original level,

_use the page system to direct

~ ~ - - personnel in containment

_ _ - _ _ to exit. -

~-

~ - - - - _ -

se to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to high radiation areas, aligning filters,etc. SRO value 3.8 155.43(4) SCRP-TI.ZZ-1102 states that the SRPT, upon discovering a 50% rise in RMS data, shall prohibit any subsequent entries to containment and DIRECT the control room to contact any work parties

,and have them exit containment. IAW SC.SA-ST.ZZ-0001, 3.2.1 The containment lighting, when

....'I flashed, is the preferred method the Control Room uses for requesting communications with the work Containment Entries at Power Salem containment Entries in modes 1-4 - -- _ -~~~~

_ - _ ~ ~ ~ ~ _

~_~

-_ - ~- - ~ - ~ -

- -~~

- precaution

- - - and_ limitations

~ in the Radiation Monitoring System procedures ~ ~ - ~ _

~

Materid'Requtredfor &mtnation'*

1 QwtsU~k~ biource: Previous 2 NRC Exams IDirect From Source Thursday, August 21,2008 10:37:44 AM

_ ~

Page32of35 ,

Page33of35 1

,Giventhe following conditions:

- Salem 1 and 2 are operating at 100% power.

- Hope Creek is operating at 100% power.

- Fire Brigade manning consists of 6 qualified personnel, which includes one Fire

' Brigade Leader.

~- The site ambulance is involved in an accident during a training exercise, and two Fire Protection Operators require off-site transportation to Salem Memorial Hospital.

I Which of the following describes the status of the Fire Brigade with four members per the Salem FSAR, and action(s), if any, which are required to be performed IAW the appropriate Fire artment Procedures? _ _

- ~ - - ~ _ _ _ _ ~ ~ ~ _~

The Fire Brigade remains adequately staffed. Only four members are required

~FSAR.

~~ -~~

No-compensatory

_____ - ~

measures

~~ - ~~

are required. -

~ _~ _~ - ~ - ~ ~ ~ ~- ~ - ~ ~ ~

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

The Fire Brigade remains adequately staffed. Notify the Superintendent- Fire Protection Operations if-~ callout is

~ ~ initiated - _ for _

~~~

any ~Fire Protection-~ ~

Operators..

~ - ~ - _ ~~ ~ -~ ~ - ~ -

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

The Fire Brigade staffing is inadequate. Initiate call-out of qualified personnel to ensure manning is restored to six members within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise submit a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report to the INRC.-~ -~

~~ ~ ~ ~ ~_ ~ _ ~ ~ - ~ -_ ~ - _ ~ -~ ~

- ____ ~ ~ ~~ _ ~ ~ -~ ~~ - - -

The Fire Brigade staffing is inadequate. Initiate call-out of qualified personnel to ensure manning is restored to five members within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise initiate a Notification and review

'for licensing

~ _- ~~-

commitment

~ - ~

violation.

____ ~~- - ~ - - - ~-~ - ~ - _~~~ _~~~~ - ~_ ~

2.4- IEmergency Procedures / Plan Knowledge of facility pro ing fire briga tate ra 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> tors requiring transportation off site due to accident staffing is INADEQUATE. NC.FP-AP.ZZ-0001, 5.2.5, states that if brigade manning falls below FIVE for more than

,2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, initiate a notification and review for licensing commitment violation.

_ ~ ~~~

~ -~

'Which-of the following Security Events reported to the Shift Manager by the proper authority IREQUIRESfull staffing of the Emergency Response Organization IAW the Salem ECG?

-~ ~~~ _ - ~ ___

~- - ~_ ~- -~ -~ ~ -- ~- ~- ~ _ ~ _ _ _ - _ __ ~-

'The discovery of a pipe bomb in a car being searched at the Site Access Road Security

~ ~~ ~ ~ - - ~ __ ~ ~- ~- ~-

_ _ ~

e~-

of procedures relatin

~~~ _

~ ~

explosive attack, airliner impact, or other hostile action is occurring or has occurred within the Owner Controlled Area." The candidate must also know that the full ERO is not required to be activated until the ALERT level.

1