ML101520347

Salem - Final Written Examination with Answer Key (401-5 Format)

ML101520347
Person / Time
Site:	Salem
Issue date:	04/20/2010
From:	Public Service Enterprise Group
To:	Silk D M Operations Branch I
Hansell S
Shared Package
ML092470038	List: IR 05000272/2010301 ML101130322 ML101130328 ML101130338 ML101130402 ML101520331 ML101520347 ML101520364 ML101520414 ML101880243 ML102520088 ML111730058
References
TAC U01764, 50-272/10-301, 50-311/10-301 50-272/10-301, 50-311/10-301
Download: ML101520347 (117)

Text

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RQ I Unit 1 completed a rapid do=w=n=po=w=e=r:::;fr=0=m=1 0;::0:;0;'=0

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I With Rod Control in automatic, which one of the following conditions would require ent into S1.0P-AB.ROD-0003, Continuous Rod Motion? Control rods step OUT 2 steps at 8 spm with Tave-Tref deviation of 0.0 Control rods begin stepping IN at 18 spm with a Tave-Tref deviation of plus 1.0 Control rods begin stepping OUT at 8 spm with a Tave-Tref deviation of Control rods step IN at 40 spm with a Tave-Tref deviation of +4.0 degrees following a small load I Continuous Rod ,Explanation of 55.41.b(5.10

} B is correct because control rods are not expected to begin automatically stepping until Tave-Tref reaches 1.5 iAnswers:

degrees, either + or -. Additionally, following a downpower, Xenon will be building in and rod motion would be expected to be in the L,.___, ..... __ ___ outward direction.

C is incorrect because of B explanation above. C is incorrect because steady state rod stepping of 3 steps or less is not entry condition to ROD-3. D is incorrect because following.

a load rejection:

Tavg is expected to rise. The rod speed ,-_...._--.......Reference Section iRevision' L.............J .L.a: Number I ABROD3E001

'Material Required for Examination

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§T Given the following conditions:

-Unit 2 is holding Rx power at 1x10-SAmps during a Rx -A single shutdown bank control rod drops fully into the core .-IRNI SUR becomes negative, and IR power is I Which of the following describes why ALL Control Rods will be manually inserted lAW S2.0P-AB.ROD-0002, Dropped Rod? RCS Tavg will lower below the pOint at which the Rx is allowed to remain critical.

Attempting to recover the dropped rod would constitute an approach to criticality.

A Tech Spec shutdown is required per 3.1.3.1, Movable Control Assemblies Group Height. Insufficient time is available to stop the power reduction prior to the automatic I'e-energization the Source Range NI's and :>UUI:>tRjUtlfll

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...1 Knowledge of the reasons for the following responses as they apply to Dropped Control Rod: Actions contained in EOP for dropped control rod Explanation of i 55.41.b(5,6, 10) S2.0P-AB.ROD-0002, Step 3.4 asks if the Rx is subcritical as a result of the dropped rod. The conditions in the stem :Answers:

• indicate it is. The procedure then directs insertion of all control rods. The Bases Document states that the reason rods are inserted is that an attempt to recover the dropped rod in AB.ROD-2 would constitute an approach to criticality.

In addition to the above reason, the distracters are incorrect because: A is incorrect because temperature would not drop to 541 on the insertion of a single i control rod at 1X10-S Amos. Cis'

• because a sinale control rod being inoperable doe:! DQt reQuire a unit sQutdoWD.

0 is i incorrect because the Rx would not trip when the source range energized since the SR counts would be below the trip selpoint.

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-,.,L.O. Number i I lABROD2E002

IRO 3 Ii Given the following conditions:

-Unit 1 is at 74% reactor -Control Bank 0 rod 101 was at 178 -Remaining Control Bank '0' rods are at 198 1-Control Room crew has entered, S1.0P-AB.ROD-0001(Q)

Control Rod, and actions to realign the rod have been

-During the realignment of the rod, the PIA converter was NOT set to actual position as required in the

-Subsequently, a plant transient occurs and rods begin to drive in until OHA Rod Insertion Limit LO, Relative to the ACTUAL Rod Insertion Limit, at what rod height will the alarm actuate? Would this be conservative or non-conservative?

WhX1 I Alarm will actuate 10 steps BELOW Rod Insertion Limit. This would be nonconservative since adequate Shutdown Margin will NOT be assured." ! I lb. i lAlarm will actuate 10 steps BELOW Rod Insertion limit. This would be nonconservative since the ejected rod analysis assumes all control rods I l..-' !are pOSitioned within 12 steps of the Group Demand Counter <85% power. I Itlarm will actuate 30 steps ABOVE Rod Insertion Limit. This would be conservative since the alarm would still provide ear1y warning of inadvertent dilution, I Alarm will actuate 30 steps ABOVE Rod Insertion Limit. This would be conservative since acceptable axial/radial flux profiles will be maintained, and radial p eakin g factors will be NOT be excessive.

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!Inoperable/Stuck Control Rod responses as they apply to Inoperable/Stuck Control Rod: Rod insertion limits on 55.41.b(6)

A is correct because all CBD rods except 101 would be disconnected from the bank and 1 D1 withdrawn 20 steps. If Answers: ' PIA converter is not reset, anything fed from the PIA converter, such as the RIL computer, will see the Bank at 218 steps (198 begin with plus an additional 20 steps to align 101). Thus the total error is 20 steps in a non-conservative direction.

The LOW is set to alarm at 10 steps above the Rod Insertion LIMIT. Thus, if the alarm is 20 steps non-conservative, the alarm will actuate 10 . , . accident reactivity effect is part of the bases for the RIL, but FSAR Section 15.4.7 for Rod Control Cluster Assembly Ejection refers I I to entire control bank position relative to the RIL or fully inserted for reactivity associated with the ejected rod. Additionally, operators would have reduced power to at least 75% in AB.ROD-01, and the requirement is =/-18 steps <85% power. C and Dare incorrect because if it is assumed that the 20 step disagreement should be added to the 10 step alarm, 30 steps above the is I plausible.

An inadvertent dilution could cause rods to step in too far and thus cause this alarm. Rod position impact on Axial Flux is plausible, since having the rods inserted too far will cause AFD issues. I". .

, I Immovable/Misaligned Control Rod II S1.0P-AB.ROD-0001 11 Bases Document /15 of 6 /1 6 I I, I Salem UFSAR II II Section 15 Accident An , 115.4-72,7116,23,18 I,Rod Control and Position Indications Systems II NOS05RODSOO-10 J I 11 50-52 /1 10 iMaterlal Required for Examination ! I II IQuestionSource:'

i f Facility Exam Bank !

1 Editorially Modified J 0 I "'-.....on c.,.;;m.nts 11080307, mo,mod to on"", moo> th," 1 ro"opt m"t"1li, 'eo"' to roIT"", a""",."",,", ....,... ..i 'UY'UVIIO' Ri:lf",o",-<, Material, 81,OP-AR ""7 1\" OHA Window E, Rev. 16 pages 11-12 I I I I

tQuestio_n RO 4 ! I Which one of the followina actions does NOT occur within the first minute followin9 a Unit 2 Rx trie from 40% eower? !Main Power Transformers Group 1 cooler tums off. I I 1-9 and 9-10 Generator Output 500 KV breakers open. I I RCP power supply busses transfer to their altemate source of power. .14KV Vital Busses power supplies transfer to their altemate (not emergency) source of power. !Answer' I d I L Exam Level . I R I Level I Memory I fF:CilitY:

i I Salem 1 & 2 I Ifxanioate:

'! I 5117120101 ,-, ....... '-, '. w , If .1 I .'

• 1 .1

[J 000007K202 " I

__i and the followin Breakers, rela s and iExplanation Of;!55A1.b(10)

For the Rx to be operating at 40% power, the Main Turbine must be online and the Main Generator outputting to ,Answers:

'I' grid. When the Rx trips. the Main Generator output breakers always trip. (While a main turbine/generator trip does not '--'--' -.----cause a Rx trip. a Rx trip ALWAYS causes a Main Turbine/Generator trip. The 40% indicated in the stem is below the power)setpoint at which a Main Turbine trip does NOT cause a Rx trip. The Main Generator Exciter Field Breaker trips when I I ,Main Turbine trios. The EFB is jnterloQ]sfld.with the Group 1 coolfl[ . . stop, The RCPs are powered from the 4KV Group Busses, which are powered form the output of the Main Generator operating.

When the Main Generator trips on the Rx trip. the Group busses automatically swap to off-site power. The 4KV vital busses are powered from off-site power and do not swap power sueelies on a Rx trie/Main Turbine Generator trip ,.-------------

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No. , I Main Power Transformer Cooling II S2.0P-SOAkv-0004 11 II 11 6 I I Reactor Protection System RX Trip Signals I 221051 11 'I , II I 113 , I I I II !I I LO. Number ;, ObjilctiyeS

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ITRP001E021 Required fOr Examination . : I ue Ion urea ommen s:

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I RO S I Given the following conditions;

-Unit 1 is operating at 100% power. -A very small leak develops on 2PR3, PZR Code Safety Valve. -PZR tailpipe temperature has stabilized at 228 degrees. -PRT pressure is S.3 psig. Which of the following identifies a condition which will result in PZR taileiee temeerature rising? I IThe PRT is partially drained. I I I PZR pressur!;J drifts 1 psig higher. I IThe PRT rupture disk develops a leak. I I I Additional nitrogen cover gas is added to the PRT. I I Leve! J I Comprehension I !FacllitY:l1 Salem 1 & 2 I :ExamDate:!

i I S/17I201°1

_ I d I I R Ii, L. _ ' ,.' ,. -,'. '1 r-::-::7 , ----t %,

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I Pressurizer Vapor Space Accident I

' Know!ed e of the 0 erational im lications of the followin conce ts as the a I to Pressurizer Va or S ace Thermod amics and flow characteristics of 0 en or leakin IExplanation 0flISS.41.b(3,S)The pressure drop across the Safety Valve will cause the downstream pressure to equal that in the PRT. Enthalpy will i I Answers: i be constant.

The fluid in the tailpipe will be at saturation temperature for the temperature in the PRT. The tailpipe temp is raised by

..--"-raising PRT pressure.

A is incorrect because any draining in the PRT would tend to allow the gas in the PRT to expand into the vacated area and cause pressure to go down. B is because th: pressure seen in the upstream side does. not control the ;Reference Section 1

!SRO RO Ust 6 , 'Given the following conditions:

-Unit 1 was operating at 100% power EOL when a 70 gpm leak developed in the -Operators tripped the Rx and initiated a Safety

-When the Main Turbine automatically tripped following the Rx trip, all Off-Site deenergized.

Which of the following describes how SG pressure will be controlled?

ISG Atmospheric Relief Valves11-14MS10s will modulate as required to maintain SG pressure at 1015 psig. Main Steam Dumps will modulate in the Main Steam Pressure Control Mode as required to maintain SG pressure at 1005 psig. Dump Valves AND the 'leXPlanationOfl 55.41.b(4)Steam Generator pressure control after a normal Rx trip can be from Main Steam Dumps or SG Atmospheric Answers: i valves. Normally, the Main steam dumps will ARM from the P-4 signal, and the steam dumps will modulate as required to maintain -.*

RCS Tavg at 547 deg, with RCS loop Tc lower than Th due to the residual power still being made in the Rx. This Tc temperature will dictate what the SG pressure is. However, when off-site power became deenergized, power was lost to all the Unit 1 . . . . iLO. Number I LOCA01 E003 I LOCA01 E008 Material Required for Examination 11 :Question Source: . .. ;New Ilo"....tinn **.v ...".......VII ***"'......... , !\ .. 1......

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I RO 7 I C __ ___ ____ ___ --.-.-JI, ; I Given the following conditions:

I-Unit 2 is in Mode -A cooldown to Mode 5 is in progress due to RCS back leakage into 23 SI

-RCS Tavg is -RCS pressure is 2200 -23SJ54 Accumulator Isolation Valve is selected to VALVE OPERABLE on -23SJ54 was shut from the control room and power remains available to the Which of the following describes the reseonse of the 23SJ54 if a Desisn Basis LOCA occurs on 22 Looe Cold Leg? I 23SJ54 opens ''-<lily. I23SJ54 remains closed. The RO must depress the 23SJ54 OPEN pushbutton 2CC1. I 23SJ54 remains closed. The RO must rotate 23SJ54 RECIRC-OVERIDE-OPEN switch to OPEN on 2RP4. :ii! I23SJ54 remains closed. The 230V control power breaker must be cycled to permit automatic opening. '--'I I I a I lE:x am Level !R I

'J I Memory I[FaCility:]

I Salem 1 & 2 i iExamDate:1 I 5/17/20101

000011A113 I iEA1,13 " IRO Value: i 4.1*IISRO Value: :1 4.2!SectionG

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-2J If 5 0 EXPlanatiOnofil'55.41.b(8,7)

As shown on marked up dwg 217127, the 23SJ54 will always open when it receives an OPEN Signal REGARDLESS of /' IAnswers:

the position of the Power Lockout SWltch( Valve operable or Lockout).The lockout portion of the circuit applies to the CLOSING " I circuit only. B is plausible if operator doesn't understand that as long as power is available, the SJ54 will open on a SI Signal. C is I

.." Title J ,. " ":'" ',_/.:f ",'" i.o. Number ',"','1, " ',r" 0,::, .,nc","1 ',1" i Objectiv+s

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r RO 8 Which of the following describes the effect of a total failure I the #1 seal would be indicated by a low #1 sealleakoff flow condition with a corresponding rise in #2 seal leakoff flow. f the #2 seal would be indicated by a high #2 sealleakoffflow condition with a corresponding rise in #1 sealleakoff flow. I the #3 seal would be indicated by an rise in the makeup requirement to the RCP standpipe.

fd.1 ANY RCP seal will result in rising radiation levels in containment. [KAq 000015K207 I

I Reactor Coolant Pump Malfunctions

'Knowled e of the interrelations between Reactor Coolant Pum Malfunctions and the RCP I I I 1 55.41.b(3, 10} A is incorrect because when the number 1 seal fails, it means that there is no resistance to flow across that seal !Answers:

i I anymore, so seal leakoff flow will rise because more flow is getting to. the downstream side of #1 seal. #2 seal leakoff flow would rise. but there is no direct indication of it, and the first part of distracter is wrong anyways. B is incorrect because the #2 seal failure would not affect the #1 sealleakoff.

which is based on #1 seal supply pressure.

C is correct because failure of the #3 seal means I JellS backprellSyre on the #2. seal. so less water would be going to standpipe hecayse mQIll water would be Dowina th[Quao 'I failed #3 seal. D is incorrect because only a failure of the #3 seal would affect containment radiation levels.

since the remaining 2 I seals flow to closed systems, VCT for #1 leakoff and Standpipe to liquid waste collection system. ---..J

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!iMaterial._.Required . for Examination ., ., Ii!r' ..' -.... -

Source: ' I Facility Exam Bank I Modification J Editorially Modified ,lUsed 0

I Q111143, replaced second #2 seal failure with ANY Rep distracter.

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.. -.""... w,lQuestion ToP1CJ RO 9 I Given the following conditions:

-Unit 2 is operating at 80% -Normal Letdown is secured, and Excess Letdown is in

-PZR level is on program and I Which of the followin9, if left uncorrected, would result in cavitation of the oeerating charging eume? 12CV132, EX L TDWN FCV, fails shut. I 12CV40 OR 2CV41, VCT OUTLET STOP VALVES, are shut. I 12CV35, VCT 3 WAY INLET V, fails to the FLOW TO HUT position.

I 1.2cV134, EXC LTDWN 3 WAY VALVE, placed in FLOW TO RCDT position. jAnswer._

I b .Le..,el.1 R I I Comprehension I

• I Salem 1 & 2 I !ExamDate:

1 I 5/17120101 I 000022A202 I

3.21 j022---Abili to determine and inte ret the following as they aeely to Loss of Reactor Coolant Makeup: I Char in um roblems I [expianation of! 55.41.b(5)

With the stem condition that PZR level is stable, charging and letdown must be equal. Excess letdown comes from the Answers: I RCS to the charging pump suction via the seal return to the VCT, upstream of the CV40 and 41. D is incorrect because changing L_.. _ .._,-_. i the CV134 to flow to RCDT will not cause cavitation, since excess letdown design flow is only 25 gpm, Ihe makeup system will be . able 10 maintain VCT level, and that is what the charging pumps will be drawing the!r suction from. A is incorrect because tile loss of is isolated, and the Excess Letdown flowpath is not in the normal letdown line. B is correct because if the CV40 or 41 (in series valves) shuts, then the suction flowpath to the charging pumps is isolated.

The auto swap to the RWST for charging pump suction on 10 10 VCT level will NOT occur because excess letdown is stili going into the VCT, causing level to rise with no outlet path to the Reference

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1 56 ,66.40 I I I I i.ri. Number I CVCSOOE006 I CVCSOOE008 I CVCSOOE015

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'rio J'SRO SvstemlE;o,utiOni.istl Cbanges , ... Given the following

-Unit 2 is in MODE -21 RHR loop is supplying shutdown

-22 RHR loop is aligned for -21 RHR pump During the crew's response operators start 22 RHR Which ONE of the following is the MINIMUM required stable 22 RHR pump flow established by S2.0P-AB.RHR-0001, Loss of IExamDate:

!

I I Loss of Residual Heat Removal System LPI um switches, ammeter, dischar e ressure au e, flow meter, and indicators iExPlanationofi i 55.41.b.(10,8).

The minimum required stable pump flow is 1800 gpm, regardless of other conditions which may be required in 'Answers:

dAB. The range of flow required by AB.RHR-1 is 1800-3000 gpm, and the AB f10wpath will terminate at step 3.68, where flow at 1800 3000 gpm. There are several additional conditions which have to be met to exit the AB, which are

!stable or dropping, RCS level >97.5', and at least one RHR pump running. A is plausible because it is the flow which goes through ahe automatically oPEloing RHR minimum flow valves 22BH29. B . . . . . I is plausible because it is the upper end of the required flow band. ____"____

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I Loss of "S2.0P-AB.RHR-0001 J I II 1117 I I Initiating 11 S2.0P-SO.RHR-0001 Ii 11 6 II2=4=:::' II Jl ,,!1----' !L.O. Number I "

I ABRHR1E005 Material Required for Examination ....._ .. II " : I Other Facility PJietho.d: ) Concept Used ![Used During Training 1[J I Palisades 8/2003 NRC I , , '" .I......'*...* " f*'" I I I t ROSkVScraperf liSt' 1 list I"OutlineCha.;'ges 1*

fO 11 If !Given the following conditions:

-Unit 2 is operating at 100% -21 CC pump is -21 CC HX SW flow centrol is in manual for a Waste Liquid

-Operators receive 2CC1 Console Alarm 21(22) CC HDR PRESSURE If the cause of the low pressure cendition is a CCW leak on the discharge flange of 23 CCW pump, which of the following centrol room would also be present within 5 minutes after the alarm is Assume the CCW Eume(sl in service centinue to run, and NO oeerator action is taken. lin service CVCS HUT level rising. I I The standby CCW pump will have auto started. 121/22 RHR HX CCW FLOW LO alarms have annunciated . *I L d.!! 2CC71, L CO NT VALVE, demand has risen. .' . . .. I I tns!,er..1 d I Level' I R I Level .._.1 I Comprehension I/Facility:

i I Salem 1 & 2 J IExamDate:.:

I 5/17/20101 000026G145 I__

i 4.31 j 4.3 IGroup3 r 11 Group:, I 11 0 1--_---'-___--"'________________________

..... ____J :KA Statement: , ,--_____________. _______________________________

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IAbilit to identi and inte ret diverse to validate the res onse of another indication.

'exPlanation Ofi I 55.41.b(8)

A is incerrect because floor drain cellection in the Aux Building is directed to the WHUT for cellection, not the CVCS I iAnswers:.. . " HUTs. B is incerrect because with the unit at power, there will normally be 2 CCW pumps running. The 3rd pump is normally in ,___. -_..... , standby, but in the stem is tagged out. C Is incerrect because the alarms will already be locked in because the CC16s are shut at I power. It will not annunciate.

D is cerrect because the 2CC71 is an air fail closed valve. Its valve demand will rise, calling

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r I: Outline'Changes t Ii IRO I I Given the following conditions: .-Unit 2 operators have just entered 2-EOP-FRSM-1, Response to Power

-Step 6 concerning SI Component alignment is contained in a double bordered I inside a shaded rectangular I Which of the following describes the significance of the layout of this step? I I Step 6 is ... Ian Action Step. Operators will answer a yes/no question and branch to the right or left based on the answer. a Continuous Action Step. Its action applies from the time the procedure is entered until the end of the procedure.

I I a Continuous Action Step. Its action applies from the first time the step is reached through the end of the procedure.

I I an Action Step. It is performed once when encountered.

It must be initiated, but is not required to be completed prior to moving on step.IE=J I

I I Salem 1 & 2 fi{A: !I000029G414 I

!!tIe:: I Anticipated Transient Without Scram ,

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Knowled e of eneral uidelines for EOP usa I EXplanation of ' 55.41.b.(10)

A is incorrect because the description is of a Decision step. B is incorrect because it is a correct description of an II Answers: ' action contained in the Continuous Action Summary, of which there IS NOT one In FRSM. D is incorrect because it is the

.. _nn _u _n _ n II Reference Title I Facility Reference Number Reference Section . Page No. IRevision.

0 11

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I Response to Nuclear Power Generation lI2-EOP-FRSM-1

!II I :::12:::4=:::.::::; II II II 11---, !,Material Required for Examination

..; I 0I New' !!Question Method: I iUsed During Training II I I I'0 u. :, ". " il. 1 !I. I I I Given the following conditions:

-Unit 1 is performing a Rx startup by control rods lAW S1.0P-IO.ZZ-0003.

Hot to Minimum -Both IRNI 1 N35 and 1 N36 have just come on scale at 1X1 0-11 A failure in the Permissive P-10 circuitry causes the "Power Above P*10" logic to satisfied . Which of the following identifies the status of the SRNI indication?

BOTH Source Ran e Nl's read on 1CC2. and actual Source Ran e counts in the Rx are Explanation of Answers: . rRevisiorll

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/0/' Obje<:tivea fUsed During Training Program 1 '-__

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i I Which of the following describes an action that will occur or be directed, based on Area Radiation Monitors, for a Fuel Handling Incident regardless of I ! I whether it occurs in Containment or in the Fuel Handling Building?

!Automatic isolation of affected areas ventilation systems. I Automatic re-alignment of areas ventilation systems to maximize charcoal 1 Rlhr in the affected area requires ALL personnel are to exit the area. 'Explanation 55.41.b(11)

A and S are incorrect because FHB ventilation will re-align through charcoal filters and start all exhaust fans, containment will not realign automatically. C is correct because CAS of AS.Fuel states such. D is incorrect because upon a fuel handling incident, all non-essential personnel are directed to leave area at beginning of procedure, no radiation level is Checked, it is always performed. .........Reference Title iL.o. Number lV-'1 IABFUEL01EOO2 . I'''''

... ; Required for Examination I II, ![Question Source: ..... , Iu ..."c: . !iQuestil ..."'............J jUsed During Training .

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." " I I J RO List' 'SRO Svstem1Evoiution " Outline Changes .1 !Question Topic j i Given the following conditions:

Unit 2 has experienced a 10 gpm tube leak on 24 SG while operating at 45% -A unit SID is to be performed per management Which of the following identifies how the CVCS system will be operated lAW S2.0P-AEl.SG-0001, Steam Generator Tube Leak, during the shutdown, and why? Leave the Charging System Master Flow Controller in AUTO. This will maintain PZR level on program. System Master Flow Controller in MANUAL and maintain PZR level 35-70%. This will prevent letdown isolation when the charging flow remains >65 gpm in Manual or Auto control. This will prevent flashing in the letdown line when the in I II I Ensure charging flow remains >87 gpm in Manual or Auto control until the reactor is tripped. This ensures the RIL is not violated while initiating a continuous boration via the makeup system.

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4.21 114.51 11 1/

I Knowledge of the reasons for the following responses as they apply to Steam Generator Tube

!Actions contained in EOP for RCS water invento!}!.

balance S/G tube rupture and plant shutdown I ,Explanation of 155.41.B(10)

Step 3.26A directs charging flow to be placed in manual and PZR level maintained 35-70%. This ensures that sufficient

!iAnswers:

!I inventory exists to prevent PZR level from shrinking below 17% when the reactor is tripped and RCS cooldown is initiated.

Keeping I l flow in AUTO is incorrect.

Flashing in the letdown line is always a concern, but flow will not be controlled for this reason. RIL are . always a concern, and a continuous boralion may be performed while performing the SID, but 87 gpm is not a limit for charging flow _, !19aHR tl:!ill

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.. * . Reference : Facility Reference Number 'Reference Section .I Page No. i"v c,._. __ __ __ .. ...__________**_.,. ___ ,_ _____ ". ___ . _____, __L ,'_ _____ , __ ______ , __ ,___*_____, ____, _, ___

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• _____ _____ _____ -k.w-. __ ** ____._I Steam Generator Tube II S2.0P-AB.SG-0001 j I Step 3.26A and Bases II !127 I il ]1 il II I I 11 JI tJ !I I ,L.O. Number I i

" 1 I RO 16 ! Given the following conditions:

-Unit 1 is operating at 1<:)0% power. -A large steamline break occurs at the Mixing Bottle. Which of the following identifies why the Bleed Steam Check Valves shut following the Main Turbine trip? , ITO ensure that Gland Seal & Heating Steam System Header does not become over pressurized.

I ITO prevent steam back:f1ow from the Feedwater Heaters to the Main Turbine preventing turbine overspeed

! I I To prevent Main Condenser vacuum degradation to the point where Main Steam Dumps would become unavailable.

I ITO ensure any steam remaining in the Main Steamline after MSLI is directed to the Main Condenser for secondary inventory control. I iA.. "..... :lfxam Level .-:.........i Level I I Memory I j I Salem 1 &2 I I 5117120101 KA:1 0OOO40K303 I iAK3.03

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I Vision QL 1420. Modified distracters for continuity, replaced C and D with new distracters.

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.*". RO SvstelriiEvolution list I List Z;OutiineChanges r

II Given the following condition:

J 11-Unit 2 was operating at 100% power when a total loss of all AC power occurred. ,

of the following is an action that is REQUIRED to have been performed PRIOR to energizing 2B 4KV Vital bus, and why? loads to extend the the Vital Instrument can power their AC loads. and depress stop PBs for SEC actuated components to prevent overloading the 2B 4KV vital bus. Explanation 55.41.b(10)

The Continuous Action Step for energizing a denenergized vital bus with an EDG comes AFTER the step to deenergize all SEC's. The Bases Document states on page 15 that the reason to deenergize the SECs and depress the Stop PB for all controlled safely related loads is to prevent the bus from overloading

.. It additionally states that a further reason is to charging pump automatic start and possible thermal shock to the RCP seals. SI is initiated at Step 21 NOT to prevent a Dump from run nina. but rather to Drelleot 1tl .... . restored.

Non essential DC loads are shed at Step 35 to extend the batteries power capability.

The SBO is started as Blackout Coping Actions in Attachment 2 Part A of AB.LOOP-1.

All the are actions which will be taken extended loss of all AC power, but the correct answer is the only one that is required to be performed AND has the for it to power restoration.

D will be but it is NOT the correct reason, and is required within

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I Ability to determine and interpret the following as they apply to Loss of Off-Site Power: Reactor tri alarm --L Explanaticln ot*'55.41

,8(7) A is incorrect because the setpoint is <90% design flow with power >P-8 (36% power). The RCPs have a huge IAnswers:

'" specifically designed to maintain pump speed as high as possible dU'"ing coastdown, so there will not be an instantaneous drop '-...*.* _. __.*_."*.J loop flow. 8 is incorrect because the Rx trip causes the turbine trip, not the other way around <P-9. C is incorrect because the I i:.

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......'" for Examination II I Question Source: !New I Ion ! I During Training Program I :J i'"'u.....uun Source C"'......"'...... I I ......... , " ."'......... '. >, . ..... I I I I I

t" Li$t'J Which of the following describes the bases for why certain Air Operated Valves (AOVs) have redundant air supplies which automatically swap to the unaffected Control Air (CAl header on a loss of the Primary air supply? I Prevents the erratic operation of Feedwater Reg valves which occurs when control air supply pressure lowers to -90psig. Prevents exceeding a Safety Limit when the normal air supply to the PZR PORVs is isolated on a Phase A containment isolation signal. Ensures that a ruptured SG secondary side can be fully isolated within the time required to prevent a release in excess of allowable limits. Ensures that a loss of an individual CA header does not result in a loss of CA to instruments andlor air-operated devices required and controlled shutdown.

IExamDate:;

! rExptanation of i 5S.41.b(7)

C is incorrect because all the valves on the secondary side of a SG required to be shut to isolate a during a SGTR IAnswers: . per SGTR-1 are fail closed on loss of air valves (MS7, MS1 D, MS18, GB4, and MS167 (because the MS169 and 171 fail open) Page l ....__....37 of AB shows MS fail closed, and SGTR sheet 1 Ruptured SG Isolation identifies valves required to be shut for isolation.

B is incorrect because the PZR PORVs have a separate backup air supply system (Accumulator).

However, the PORVs are not required . . . .

Valves, (UFSAR 5.5.13) The PZR PORy§ are de . limit PZR pressure to a value below the high pressure reactor trip setpoint.

A is incorrect because the Rx is required to be tripped control air header pressure falls below 80 psig in part because the Feed reg valves operation does become erratic below that cC ryes -. 'liS"" IABCA01EOO1 I ! I l!'.'terial Exa.ll1 ination "'1 ! !I [Question Source:! 1 New I Question Modification Method: I I

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____________________________________________________________________________Given the following conditkms:

1'1 -Unit 2 Control Room has been evacuated due to a Security -The TSC has identified 4 control rods did not insert on the Rx trip, and remain I Which of the following actbns required to establish Rapid Boration is the ONL Y action that can be PERFORMED by the Plant Operator at the Hot ;Shutdown Panel 213 lAW S2.0P-AB.CR-0001, Control Room Evacuation?

Start a Boric Acid pump in FAST speed. Adjust charging flow to > 75 gpm above seal injection flow. Shut the 21CV160 and 22CV160, BA XFR RECIRC CONTROL VL VS. I 000068K20 1 Ex.. ** planation of .155.41.b(10,7)

All of the choices are actions that are performed to establish Rapid Boration when required.

The only action that can Answers: be performed from the Hot Shutdown Panel is operation of the Boric acid pump. The CV175 is locally manually opened. The CV55 I IL -is adjusted locally by adjusting air supply t: 2CV55 at panel 216-2.

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i I Given the following conditions:

-Umt 2 IS operating at 100% power. -There is no primary-to-secondary leakage. -Excess letdown is in service due to a problem with the 2CV18, Letdown Pressure Control Valve, which is currently shut. -A fuel pin failure occurs, releasing a large amount of fission products into the RCS. Of the following radiation monitors, which would FIRST show a change because of the failed fuel? Room Area Monitor. i I ,AnY 2R19, Steam Generator Slowdown Monitor. -.--J IEJ I

!Salem 1 &2 I I 5/17/20101 I

0 EXplanation of'155.41.b(11

,5)With the CV18 shut, normal letdown will be out of service, and if out of service for an extended period of time will have Answers: ,1 Excess letdown placed in service. Excess letdown does NOT pass through the 2R31 process monitor. The RC filter also will not ,,____*___**_..** __ T ______,_**_____have flow from the discharge of the mixed bed demins since normal letdown is secured. The stem states that there is no pri to sec leakage, so the Ri9s should be unaffected.

The excess letdown line flowpath goes to the suction of the charging pumps where it 'Co.

-'.

IABRC02E001 I RMSOOOE007 I I RMSOOOE014

...... ....,",. I I RO 22 I I 'Given the ,,,lIvVH",,, C',ond,tions:

I -Unit 2 is operating at 55% power. 600 -Operators are currently in S2.0P-AB.GRID-0001, Abnormal Grid. due to grid

-Main Generator MVARS OUT are at the maximum allowable per Artificial Island Operating Guide and Documentation, per Load Dispatcher request due 500KV grid

-Main Generator voltage regulator is in !The control room receives OHA H-13, GEN FLD OVRVOL T. If the alarm does not clear within 10 seconds, which of the following identifies what will occur FIRST as a DIRECT result of this condition with i operator IMain Turbil 'will trip. I IiiJ I Exciter Field Breaker will I Main Turbine runback will occur. I utput breakers will open I I I.__ .*. , -_.. ,..: " c.*.*._.__.I = = , i ., -r"nswer Level :

LeveLJ I ApplicatiOn i

Salem 1 & 2 I

• ExamDate.

J1 5/17/20101 KA:J I 000077K102

[RO

'System/Evolution rTtie:] I Generator Voltage and Electric Grid Disturbances Statement:

dge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Explanation of II' 55.41.b(1

0) The ARP for OHA states that the alarm annunciates at1 05% of rated field Voltage. The J1 K relay attempts to iAnswers:

i voltage to the 100% value, and after 5 seconds if voltage is not reduced to <: 105% it swaps the regulator to manual. If after , seconds the voltage is still >105%, with the Main Generator output breakers shut, it will trip the Main Turbine, which in tum will the Main Generator output breakers 30 seconds later. D is incorrect because the main gen output breakers are NOT affected as a .., . . . ..

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__________________________________________

---l Given the following Unit 2 conditions: -A manual safety injection was initiated when RCS leakage exceeded charging capability.

-The crew has progressed through the EOP's and is now in TRIP-3, Safety 21 Charging Pump is

-Both SI Pumps and both RHR Pumps were just stopped. Which of the followin arameters will be evaluated at the SI Reinitiation Criteria ste to determine the need for reinitiatin ECCS flow? :c.

• subcooling and PZR level. d.. RCS subcooling and SG Pressure.

IEx'planation of' 5S.41.b(10}

Step 10 asks for ReS subcoofing

> 0, which is obtained from the Subcooiing Margin Monitor (SMM) on RP4. It also Ii Answers:

• asks for PZR level >11 %. A is incorrect because CET's are not checked, but PZR level is. B is incorrect because while this i! -_******._-j combination is what gives operators their input into the SMM, there is no trend involved.

C is correct. D is incorrect because while II SG pressure will be correlated to RCS loop Tc, it is not checked.

__

____...._.__--_..______________fLO. Number ITRP003E005 I TRP003E003 i , ...... "'......".'1 .... "' for Examination

! I I I Source: I Facility Exam Bank IIOIIAlIttinn

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Topic I 180 24 II Given the following condtions:

--Unit 2 is operating at 25% power during a power ascension following a refueling -21 charging pump is in A RCS leak is identified, and operators enter S2.0P-AB.RC-0001.

Reactor System Leak. I -After reducing letdown flow to minimum, and raising charging flow to maximum. the I following indications are present: 1 I . PZR 1"",,1 ;, 27% aodloweri", 0.1 % "'.'Y 45 "",,"d,.i -22 RHR sump pump run alarm is locked in. I ,Which of the following describes the proper course of action for these conditions, and why? Trip the Main Turbine and enter S2.0P-AB.TRB-0001, Turbine Trip Below P-9. Safety Injection will only be required ifVCT level can NOT be I 1 L--maintained above 4%. ,.......-...I b. iTriP the Rx and initiate Safety Injection.

The leak rate exceeds the CVCS system make-up capacity.

Action will be taken in LOCA-6. LOCA Outside Containment to isolate the leak from the RHR system. I Isolate letdown to establish rising PZR level and place the CVCS Make-Up system in manual to raise VCT level. This will ensure PZR level II can be maintained above 11 % and preclude having to initiate Safety Injection.

II[J Place a second centrifugal charging pump in service. Continue in AB.RC-0001 to perform leak identification and isolation steps. Initiate a unit shutdown to comply with the actions of TSAS 3.4.7.2 Reactor Coolant System Operational Leakage. I ! i ...r*****.. Explanation of ' iAnswers:

Number Ob.;ectivesl Question Source: 'j Previous 2 NRC Exams Method: I Direct From Source Question jsalem 12/2006 NRC RO Exam il"' ._u..--I I I I

__ I

[RO 25 I I Given the following conditions:

I -Unit 2 is operating at 100% power. -22 charging pump is -A technician working in the control room rack area inadvertently causes a momentary P-14 signal to be actuated.

-Shortly after the Rx trip, 23 SG Safety Valve 23MS15 lifted and remains open. -An automatic Safety Injection occurs on Main Steamline DIP. -2A 4KV vital bus is locked out on bus differential.

-NO AFW flow can be established, and operators transition to 2-EOP-FRHS-1, Response to Loss of Secondary Heat Sink. Which of the following identifies how these conditions will be mitigated in 2-EOP-FRHS-1?

=-, I Operators will initiate Bleed and Feed because no charging pumps are available.

I I Steam Generators will be fed with a SGFP since SI reset action will allow prompt recovery of a SGFP. i Steam Generators will be fed using the Condensate system because the Safety Injection signal isolated Service Water to the SGFPs. I I I Steam Generators will be fed using the Condensate system because the in P-14 signal is maintaining a trip signal to the SGFPs. .-=. . -" i r--I .

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IAppllcatlon 1 & 2 5/17120101 SystemlEvoll.ltlonTHIe:'

L.____________

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I Knowledge of the interrelations between Loss of Secondary Heat Sink and the following:

I I Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and I ! automatic and manual features.

lr.Ex.. PI2.. natiO...n of .lI55.41.b.(10,4,7)

A is incorrect because the remaining centrifugal charging pump required to be operating (21) to prevent the II Answers: ' transition to Bleed and Feed from step 4 IS available, its power supply is from 2B 4KV vital bus. 23 charging pump (positive displ I , ----,---pump) is supplied from A vital bus. B is incorrect because step 12,1 asks if Safety Injection has been initiated, and if yes, bypasses ! the SGFP recovery action and goes straight to condensate system recovery.

C is correct and D is incorrect because the SGFPs are I I I unavailable due to the 51 signal which caysed the to clQ§e the SW to building isolallim the triP si!JDal from j ' P-14. P-14 is not a "lock in" relay, and the stem states a momentary signal is the cause of the P-14 actuation, which tripped the I /., SGFPs and Main Turbine and isolated FW yObjeotlvQsl

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... J IALL RCPs should be stopped because the drop in the #1 seal DfP during depressurization may result in pump damage. lb. " but one RCP should be stopped because loss of forced circulation flow would require more time to reach equilibrium conditions.

I ALL RCPs should be stopped to avoid mechanical damage that can occur due to cavitation when the RCS reaches saturation conditions.

I should be stopped because stagnating during depressurization could cause a bubble to form in the reactor ,Cognitive Level I '.

I A is correct because the Basis Document on page 35 states that the reason the RCPs are stopped is due to the iAnswers:

i I' anticipated loss of Number 1 seal requirement.

Continued operation may result in damage to the RCPs, B is incorrect because , "". while it will take longer on natural eirc when RCPs are stopped, it's not the reason why, and one RCP is not left running. C is I incorrect because cavitation should not be expected, since the procedure will be cooling down the RCS to <375 Thot. D is incorrect I l Will Q9LR§, lef trtlD nioQl........u._.."" ......""" ...... ,_,......_"".._,..",_..""._.',,,... ,,,,,,,,,, __, ',i" Objectives)

I27 I-A LBLOCA has I ! -Operators are performing 2-EOP-LOCA-5, LOSS OF EMERGENCY I 1-A PURPLE path on Containment Environment Status Tree has just occurred i containment pressure reached 15 Which of the following describes how the Containment Spray system will be operated, and why? The Containment Spral: System is operated as directed in ... ia.' !2-EOP-FRCE-1, Response to Excessive Containment Pressure, since restoration of the critical safety function takes precedence.

I 2-EOP-FRCE-1 because actions ooncerning Containment Spray operation are more restrictive.

2-EOP-LOCA-5 since FRPs are NOT implemented during the performance of LOCA-5.

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i 1., II Given the following

-Unit 1 is operating at 100% -1 E 460 volt bus is deenergized following a trip of its feed

-Tagging is in progress to allow troubleshooting of 1E 460 volt The operator mistakenly opens the 1 F 460 volt bus feed breaker, deenergizing the 1 F 460 volt Which of the following describes a consequence, if any, of this Unit 1 Reactor will trip due to the loss of BOTH Rod Drive Motor Generators.

will trip due to the loss of a SINGLE Rod Drive Motor Generator.

ill NOT trip because BOTH Rod Drive Motor Generators are still in service. will NOT trip because ONE Rod Drive Motor Generator is suffiCient to maintain power to the Rod Control system. 001 000K201 :System/Evolution Title::

....

1 Knowledge of bus power supplies to the following:

lOne-line diawam of power supply to MIG sets. i .ExPlanation of 55.41.b{7,6) 11 and 12 Rod Drive Motor Generators are powered from 1 E and 1 G 460 volt busses. A single RDMG set is sufficient I Answers: ; to power Rod Control. Since the unit is still at 100% power subsequent to the loss of the 11RDMG , the loss of 1F 460 volt bus will '....

not trip both Rod Drive MG sets. A is incorrect because 1F 460 bus does NOT power 11 RDMG set. B is incorrect because a single RDMG set can power Rod Control. C is incorrect because only 1 RDMG set is still in service. D is correct because 1 RDMG set will I*Dower rod control The KIA is met RpMG sets are POwElrnQ rnauirns knowledae of the one-line !drawing of the power supplies.

I I Material Required for Examination I I I I Facility Ex;:=am=B=an::=k=:..:.li

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,---'--..Statement:

e of the effect that a loss or malfunction of the Reactor Coolant S stem will have on the followin : ,Explanation 55.41.b(9)

A is incorrect because CFCU leak detection will rise as the moisture in containment is removed from the air by B is correct because the containment recirc sump level is used to determine if sufficient level is in containment following a LBLOCA to allow operation on Cold Leg Recirc. There is a smaller containment pocket but it does not have an indication on 2CC2. and is not called the "recirc" sump. C is incorrect because containment dew Reference IL.O. Number IQuestion Source: _________--1 ...

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SR?'SiWscBPer,J T iQuestion Topic I IRO 30

______________________________________________________________________Back-up seal cooling flow is provided to ALL RCPs by opening the CV114. Seal Leakoff Bypass valve. Back-up seal cooling flow is provided to individual RCPs by opening the CV75 PZR Auxiliary Spray Isolation valve and closing the

..r.tIV'" CV104. RCP Seal LeakoffValve.

'I Reactor coolant flows up the RCP shaft through the Thermal Barrier where it is cooled by the CCW in the Thermal Barrier Heat Exchanger.

and then flow continues along the shaft to the radial bearing and RCP Seal package radial bearing and into the Therma' Barrier, where the heat from the RCS and the bearing are dissipated rrier. This cool water then flows through the #1 seal.

I I Reactor Coolant Pump System :KA Knowledge of Reactor Coolant Pump System design feature(s) and or interlock(s) which provide for the

! eals ! 'Explanation of ". 55.41.b.3)

B is incorrect (but plausible) because Prereq 2.3.2 of SO.RC-0001, RCP Operatior has Aux Spray put in operation for lAnswers:

maintaining PZR level while preparing to start RCP. A is incorrect because CV114 is only opened at low RCS pressure when seal leakoff lowers due to low DIP. D is incorrect because hot RCS is cooled first by Thermal barrier before reaching Radial bearing and Iseals. ""..... ,."",,. .. Number ..;;..

Page No.

!I RCP Lesson Plan II NOS05kl,..;r'UMP-09 I

! 11,________

J RCPUMPE004 I Material Required for Examination , _ __ -.

'Question Source: i I Facility Exam Bank I iQuestion Modification Method: II Editorially Modified L -:Question sOurceCommentS:

I Vision 058727 Replaced 3 distracters with more plausible ' i

... : i . .c i I I I I II, I I II: I I II I J II I 'I !DU!ing Traifllng i 0 I I Given the following cond:tions:

-Unit 2 is operating at '.2% -24 RCP With NO operator action, which of the following describes how indicated loop flow and loop OfT in RCS loops 21-23 will be it<A?I 003000K505

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....[Question TOil!S I RO 32 I Given the following conditions:

-Unit 2 is at 100% -Core Burnup is 8,000 -Rod Control is in 1-Main Turbine Governor Valves are FULLY 1-A normal AUTO makeup initiates to the -The boron addition rate is set 5 gpm higher than required for present RCS i (SSUmin g NO operator action, what will be the effect on Rx power and RCS Tave 15 minutes after the auto makeup is complete?

Reactor Power will be , and RCS Tave will be . :!J I lower, I I I lower, higher. I lower. I I higher, I higher, higher.I I I I IAnswer-ra-\Exam t.evel lR""I!Cognitive Level I I Comprehension I

Salem 1 & 2 5/17/20101

!KAjI 004000A110 I

iRO l!<A

for Exa l1l ina!ion" -.J ! II I Facility Exam Bank !euest!on MOdification Method: JlSignificantly Modified !iUsed During TrainingPrO!!ram!

U l'1uestion Source Comments I I Vision Q78012 Modified stem to include governor valves full open, which changes correct answer. Removed part of ..-..

..

I answer for Main generator electrical output status. I'I .tnt I I ,I I II I I

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_____--11 I Residual Heat Removal

_________________________________________________Il:Xplanal]On of 55.41.b(8) 11 and 12 RHR pumps are powered from "N and "9" 4KV vital busses respectively.

pumps, (11 and 12 11 and 12 CS) are powered from A and C. Unit 2 SW pumps are powered in reverse order, 21/22 from C, and 25/26 from considering plausible distracters.

Charging pumps 21 and 22 are powered from 9 and C busses, again when ,_________________________________________________________

I RHROOOE005 I ,Material Require<!

for Examination i I I Source: I I Other ModlfieaiionMethod:

] Concept Used I[Used During

-CJ ljFermi 7/7/03 NRC Exam J ...... L .' .' ". " 'I I I I

ko t'Sfio siWScraPtirJ _ flO Ust ,

'IOutlinEl dlli,I1Qesl 3:1 ........ ... ......... ' ___...__* __._.__.** __J Given the following conditions:

-Unit 2 was operating at 100% power when a LOCA occurred.

I 1-RCS pressure is currently 500 psig and stable. I I -Containment pressure is 9 psig and rising slowly. I-RWST level is 23' and lowering.

I Which of the following is the CLOSEST to the total amount of ECCS injection flow?

I I Emergency Core Cooling System J Ability to manually operate and/or monitor in the control room: I I Pumps I :Explanation of ,155.41.b(8) 2,500 is correct, because runout flow for each of 2 charging pumps is 560 gpm at 550 psid, and runout flow for each of Answers: 'I the 2 SI pumps is 675 gpm @675 psid. Operating at runout flow would yield a maximum flow of 2470. 7670 is if containment spray __***.* .* .J I flow (5200) was added to SI and CVCS flow, but no CS flow would be expected at 9 psig in containment.

and is not considered WCCS injection flow. 11,470 is if the.2 RHR pumps run.out flow is added (4,500 @ 130 psid x 2 = 9000 gpm+ 2470 gpm) 16,700 is SRO SvstemlEvolution List" R035 II -

Using the below list, which of the following choices contains ONLY the interlocks whiGh must be met to open 12SJ45 RHR TO CHG SI PMPS STOP MOV? I. Both 1RH1 and 1RH2 RHR COMMON SUCT MOV must be shut. II. Either 1RH1 or 1RH2 RHR COMMON SUCT MOV must be shut. I ililo Both 11SJ44 and 12SJ44 CONT SUMP SUCT VALVE MOV must be open. IV. Either 11SJ44 or 12SJ44 CONT SUMP SUCT VALVE MOV must be open. V. 12SJ44 CONT SUMP SUCT VALVE MOV must be open. VI. Both 1SJ67 and 1SJ68 SJ PUMP MIN FLOW VALVE must be shut. VII. Either 1SJ67 or 1SJ68 SJ PUMP MIN FLOW VALVE must be shut. '11' III. VII ! III, V, VI. r:! ' E:.J I"' V, VII. I i :d. f IV, VI. IIAnsw... 1iCognltlvelevel

! I Memory I &2 I iExamDate:

I, 5/17/20101 ,KA:JI006000K416 I K4.16 v.!Ilue=l!

3.21 [SRO II l'!0 11 [sRo.Group:H3 Etl 0 [System/Evol Title: I L" ' l,j"""'y Core Cooling System 1 006 I Knowledge of Emergency Core Cooling System design feature(s}

and or interlock(s) which provide for the following:

I I Interlocks between RHR valves and RCS explanation oil i 55.41.b(8, 10) Using Logic Diagram 224421, it shows that to open 12SJ45 you must have: 1. Either 1 SJ67 or 1 SJ68 shut (VII); 2. I I RH1 " 1 RH2 'h"1 (II); 3. 12SJ44 op'" (V). Th, 'I,tract", ,II 0001"0 ,I 00' of Ih"hol= whl,h I, 001'0""'

to Reference Title ! Facility Number !"""'""'....... Section ! Page No. CVe: System No. 12SJ45 Chg Suct 11224421 I II 113 I I I II ! I I I I II i I I Required for Examination

._-""".......---------_

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-Unit 2 was operating at 100% reactor power when a Reactor Trip and Safety Injection were initiated due to lowering Pressurizer pressure.

-Five minutes after the SI actuation.

containment humidity and pressure have just begun rising. Assuming NO operator actions were taken, which of the following would result in these conditions?

I I Pressurizer Relief Tank/Quench Tank System a loss or malfunction of the Pressurizer Relief Tank/Quench Tank System will have on the following: !------------------

__________

.____--J!______*__________________

____

II I Other Facility I:Question MethOd:--]

Editorially Modified J [Question Source Gomr,nents I I Point Beach 1/20/2006 NRC exam, modified PORV failure to Safety failure for realism. Remove PRT pressure and !,

..

____ i temp ind in stem to raise level of difficulty.

Removed cont sump pump start because it would not happen immediately . II I il, II __________________________________

---J 1 I RO Skvsc.,I:SROSkvsc,,:per' 1"SRO

"OutlilleChanQe;!

.-. , Given the following VVIIUlUVII<>.

-Unit 2 is operating at 100% -Operators receive OHAA-6, RMS HI RAD OR TRBL, and determine the 2R17A I 2R17B, Component Cooling Header 21 and 22 Rad Monitors are both in I ,-CC Surge Tank level is determined to be rising very slowly. I I Which of the following describes the impact of the radiation alarms. and what action(s) will the control room crew be directed to perform lAW I AB.CC-0001, Component Cooling Abnormality?

12CC149. CC Surge Tank Vent valve automatically shuts to prevent contaminating the in-service WHUT. Drain the CC Surge tank to maintain I. I level < 58%. 'b.; 12CC149. CC Surge Tank Vent valve automatically shuts to prevent contaminating 22 Aux Building Exhaust Filter Unit. Ensure Surge Tank I --Makeup valve is shut. -.-.J lc. i 12CC131, RCP Thermal Barrier Isolation valve automatically shuts to prevent a RCP Thermal Barrier leak from contaminating the CCW I ..system. Shift the Aux Building ventilation Normal Areas exhaust to 22 HEPA PLUS CHARCOAL.

---

.-J !2CC131, RCP Thermal Barrier Isolation valve automatically shuts to prevent a RCP Thermal Barrier leak from contaminating the CCW system. Shift the Aux Building ventilation Emergency Areas exhaust to 22 HEPA PLUS CHARCOAL.

I

[CJ [Exam Level. [C]

I Memory I facility:

1 I Salem 1 & 2 i 5/17/20101 I

I!S C I Component Cooling Water System r ,uuo ! :KA Statement::

Ability to (a) predict the impacts of the following on the Component Cooling Water System and (b) based on those predictions,

---" i procedures to correct, control. or mitiQate the conseguences of those abnormal PRMS alarm I rEx.*-Planat.ion Of.:. A is incorrect because draining the Surge tank is NOT appropriate here, because the procedure directs it to be drained to a 55 gallon I I:Answers:

i drum. The Basis Document page 3 specifically states, "The CCW Surge tank is drained to maintain level <58% for cases where ,---,--

2R17NB are NOT in warning or alarm (Step 3.10)". B is correct because the CC149 vent valve and the CC147 tank relief valve I combine into a common header which goes both to the ABV ventilation system and the Contaminated Floor Drains system. With ! Surge Tank leyel rising. any potential source of inleakage is isolated.

regardless of whether a high rad conditioo exists. ThA r:r: surge tank M/U is isolated at step 3.4, Just prior to checking R17AlB status at step 3.5. C and 0 are Incorrect because while high I CC retum flow from the RCP thermal barriers will isolate the CC131, it would result in a rapid rise in CC surge tank level. Additionally, the ABV filter would not be selected since the source of the radiation in the CC surge tank has been isolated by the " automatic closure of the CC149. I Component Cooling Abnormality II I! 11 2-8 11 13 1________----'11 --11 II 11_---'

I J ! I....g,

..

! I 'e;;. ,<<: ..: I Source:. j I New J I I.'"","'"".",,,,,, ....... i . I ,Used

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I[Question Source Comments J J ....

.. J J J

• " ROSVStem/Evolutlon list I.

Ust*1 , ...."..

iQuf!stlon TopiC J r RO 38 I Given the following conditions:

-Unit 2 is operating at 1 :)0% -Both PZR spray valves have been placed in manual control to erratic

-PZR Pressure Channel IV has failed low. NO action has been taken to place it in tripped condition i PZR Master Pressure Controller is selected to Channel I for control. I --PZR Backup heaters are in Auto and OFF. -A main turbine control failure results in a rapid load reduction.

-RCS temperature, PZR level, and PZR pressure rise rapidly. i-As RCS pressure rises past 2335 psig, the NCO observes that PZR level is 2PR1 is open, 2PR2 is closed, and PZR backup heaters are Which of the following how the PZR pressure control system is I C"[JVII .... :a.J I Malfunctioning, the PZR backup heaters should be deenergized.

I ! J Operating as expected for given I Malfunctioning, 2PR1 should be closed. I d-:-I Malfllodiooioc 2!:B2 sbould be olleo I

[0" I Comprehension I I Salem 1 &2 i [Examrilite':]

I 5/17/20 10 1 [KA:I 010000K1 03 i Value: DISRO Group:iD mmm KA Statement:

• IlEXplanation !55.41.b(7)

With PZR pressure channel IV failed LOW, and the bistables NOT tripped per stem conditions, 2PR2 is inoperable and I jAnswers:

I will NOT automatically open on high PZR pressure of 2335 psig since it needs 2/2 channels (PZR pressure channels 1/ & IV) to ! ..............open. The PZR heaters are ON since actual PZR level is 5% > clipped program level of -59%. (PZR programmed level cannot be >60%, and the PZR BtU heaters energize when actual PZR level is 5% above programmed leveL) Normally, PZR BIU heaters would . . P R > . . .. v' h nn I fr I service, which has not been done as per stem conditions. " Reference

__.-

Numiler_"

[pageNo.'

I Pressurizer Pressure Malfunction II S2.0P-AB.PZR-0001 II 1122 11 18 I I Overhead Annunciators Window E II S2.0P-AR.ZZ-0005 J I 11 29-30 11 18 I II II II 11=1*..*.,.,/.*0..*.'.'i>?i...

..... ' .'. ""'.], . OhJectiv.es I ,Material Required for Examination

• i ,L ,,' '"

..I II I I Facility Exam Bank I IQuestion Modification Modified I During Training Program l 0 IVision Q67201 Added BIU heater status and PZR Master Pressure Controller Control channel to stem --.J \ ......** , .' I I I

'.m dirtiir;;Change;l

[QueStlOnTopTcl

/§9 1 I Which of the following would cause a loss of position indication for 1 PR2, PZR PORV, on 1 CC2? Loss of 115 Volt AC bistable power to bistable power to Loss of 28 Volt DC control power to 1PR2 control circuitry.

Loss of 230 Volt AC motive power to 1 PR2 valve operator.

iEXplanation of 'I' 55.41.b(7)

A and B are incorrect because bistable power would affect 1RP4 indication if not present, not a loss of console indication.

D is incorrect because the PORV is an air operated valve. The PORV Block valve is powered from 230 VAC. C is correct because the console OPEN I CLOSE PBs receive their power from 28VDC as does all the control PBs on the control consoles.

A and B are plausible because of the between the PORV operation circuitry and which pressure controller

.. :

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i I RO I Given the following conditions: Unit 1 is operating at 100% power. -PZR level Channell (one) is selected for CONTROL. -PZR level Channell! (tVilO) is selected for ALARM. -PZR level Channell! (two) fails LOW. Which of the following describes the effect that this will have on PZR level control, and the action(s}, if any, that will be performed lAW AB.CVC-0001, Charging Malfunction? .IOHA E-36, PZR HTR OFF L VL LO will annunciate, but charging and letdown flow will remain constant since it is the alarm channel and actual I PZR level is on program. I :1):': !1CV2, LETDOWN LINE ISOL and 1CV277, LETDOWN LINE ISOL will shut Channell (one) must be selected for Alarm to clear the interlock L--J and allow opening of the 1CV2 and 1CV277. IPZR level will rise. Charging System Master Flow Controller must be placed in manual. Charging flow will be raised to 85-90 gpm prior to I UfJt::ICllIIIy the 1CV3, 1CV4, or 1CV5 LETDOWN ORIFICE ISOL valves.

level will lower. Charging System Master Flow Controller must be placed in manual. Charging flow will be loweredto 85-90 gpm prior to I operating the 1CV3, 1CV4, or 1CV5 LETDOWN ORIFICE ISOL valves.

' I c I level :

L!vel= I Application i

I Salem 1 & 2 I I 5/17/20101 , ,.... l; 0* I r KA.II 011000A211 I A2.11 !System/Evolution Title:i _..__...

KA Statem!.'!.!:s Ability to (a) predict the impacts of the following on the Pressurizer Level Control System and (b) based on those predictions, rocedures to correct, co f those abnormal 0 Failure of PZR level instrument

-!Explanation of 55.41.b(7,10)

D is incorrect because PZR level will rise due to the loss of letdown and the continued charging flow. A is incorrect

[Answers:

because while the OHA will annunciate, either of the Control OR alarm channel will shut the orifice isolation valves CV3,4,5, along their respective CV2 or 277. B is incorrect because only 1CV277 will shut on an Alarm channel failure, not 1CV2 also. Am"'lrm;olll\l the unaffected available channel will be selected for not channel I. C is correct because PZR level will rise

',sRD SkvsCnlP8i\J Outline Changes, I

[RO 41 !Given the following conditions:

-Unit 2 is operating at -Containment Pressure Channell (one) failed high and was removed from service 1 I S2.0P-SO.RPS-0005, Placing Containment Pressure Channel in Tripped -A technician troubleshooting the failed channel inadvertently de-energizes the power Containment Pressure Channel III

-The Rx does NOT I Which of the following identifies what has happened, and what actions, if any, that must be taken? isolation signals are expected for this condition.

Enter Tech Spec 3.0.3. An ATWT has occurred since the Rx should was met. Trip the Rx and to EOP-TRIP-1, A Containment Phase B isolation Signal has been generated, and the Rx is NOT expected to have tripped. If the Phase B signal can NOT be reset within 5 minutes, trip the Rx and go to EOP-TRIP-1 Rx Trip or Safety Injection.

An ATWT has occurred since the Rx should have tripped on the Containment Spray initiation signal generated when 2/4 coincidence for Hi-Hi Containment pressure was met. Trip the Rx and go to EOP-TRIP-1, Rx Trip or Safety Injection.

5/17/20101 I

ISystem/Evolution Title: 1 ..Explanation of . 55.41.b(7)

Containment pressure channell is dedicated to Containment Hi-Hi pressure and Containment Spray/MSLI, and does not Answers: ' feed into the Safety Injection coincidence of 2/3. The stem states that the technician deenergizes the power to channel III, which

.. would prevent energizing its associated bistable, so an automatic SI will not occur. The containment hi-hi bistables are energized to I actuate, and are not tripped during channel removal. Since there is no SI or Phase B signal, no actuations will occur. The loss of a I second channel of containment pressure will cause Tech Spec 3 0 3 to be entered. because there ir;; only an action for 1 of 4 cont I pressure channels bein.l:L0/S under ESFAS Tech Spec 3.3.2.1 I==::=::;

i 266 i________________________________________________

. ROSkv&;'oer1

_SRbSkYscraPiif I ";ROSvstemlEvo!utlon Ust ... 42 Given the following conditions:

-Unit 2 is at 75% -Console alarm "LOSS OF TRIPPING CAPABILITY" Alarm is received for "An Trip Breaker !i Which of the following describes the effect on RTB "A" from this condition?

I A Manual Rx Trip signal will ONL Y energize RTB "N UV trip coil. Depressing the RTB "A" OPEN pushbutton on 2CC2 will cause a Rx !An Automatic Rx trip signal will ONLY energize the RTB "A" Shunt trip coil. :(i' I A Manual Safety Injection initiation will NOT directly cause RTB "A" breaker to open. I-'I ----.J I

Id I

!R IApPlication I

Isalem 1 &2 I I 5/17/20101 I

..

..

h sical connections andlor cause-effect relationships between Reactor Protection System and the following:

I:El'phlnationOt]

55.41.b(7)

The loss of tripping capability alarm indicates that power to energize the shunt trip of the RTB is not present. The shunt 1 Answers: trip is energized from ANY auto or manual trip, and ANY auto or manual SI. The UV coils, which are deenergize to actuate only i get trip signals from ANY auto or manual rx trip, and any AUTO SI. Additionally, the OPEN PB on the control console ONLY I energizes the shunt trip coii. For the above conditions

...... A is incorrect because a manual Rx trip will only . " fB pnl'! eae[gjzeQ shunt trip coil which has no power to enerajze, C is irlQQlIect ,I because there is no power to the shunt coil AND the UV coil would deenergize.

D is correct because a manual safety injection signal ONLY acts to energize the shunt which has no power. The word "directly" is part of the correct answer because the Rx will tri from the redundant train, and the when its UV coil is de-ener ized '

for Examination

.I II I.

I Facility Exam Bank I M_()dificationMethod:q

] Significantly Modified I flJsed Pio£p:all1:

1 C j I Vision Q77831 Kept stem conditions, changed to new correct answer. Changed 2 of 3 distracters.

I Ii I

.. 1 mmm ........._mm !QuestionTopTcl I RO j the f:lloWing completes the description of the ECCS design basis single failure criteria for the INJECTION phase of an i i The ECCS is designed to withstand an sin Ie failure and still erform its intended safe function.

I I Engineered Safety Features System Knowledge of the operational implications of the following concepts as they apply to the Engineered Safety Features S , EXPlanation-of] Functions, 6.1.1.4 states ...I I !Answers:

i one or one passive failure, but not in to a single failure in the injection phase" UFSAR Section 6.3.1 i

..

states.....The system is effective

... and is tolerant of failures of any single component or instrument channel to respond actively in the I system." Reference Number Reference Section fRe"TiIOl1l r"-;;;;;;::..c.;;;;"';

I ESFOOOE010 1___--1 L......_____ ____ __..._>> ___

RO iff IR044____________________________________________________________________________

.____of the following describes the automatic action(s) which will occur, if any, if 14 SG Steam Flow Channell! (two) were to fail high with NO A Safety Injection will actuate, causing a Rx trip. The Rx will trip, but a Safety Injection will NOT initiate.

A Main Steamline Isolation will occur, and a Rx trip will occur on High PZR pressure following I

I Engineered Safety Features owledge of the of the effect of a loss or malfunction on the following will have on the Engineered Safety Features ,.. ....'CUalllUII stem:_______________________________________________________________________________________

-4 and detectors 1 iExplanation of i 55.41.b(7)

There is no Hi Steam Flow Rx Trip. There is a High Steam Flow SI, but it requires both 1/2 high steam flows on 2/4 steam ! ,Answers: . generator COINCIDENT with either 10 steam generator pressure or 10-10 Tave. When a SG Steam Flow channel is removed from I it only trips the High Steam Flow signal to SI. When the second SG gets it 1/2 high steam flow, there is still no coincidence with steam pressure or Tavg to initiate a SI. There will just be 2 SCi's that have a Hi Stm Flow signal associated with them. Steam,Generatpr water level cpntrol will npt be affected because Digital Feedwater wi!! remove the failed channel fr.Pm the contro . . The Main Steam line isolation Signal is developed just as the SI signal is developed, which requires the 10 steam pressure or 10-10!Tavg. I iMaterial Required for Examination i! !I I New i iQuestlon !

Program :::J 1...! rQuestionSource commentS]

I I ;'.,{>'., ' ;>. ': ...*.. ;..:,: ........*.

........,.::.. 1 J I !

I RO I Which of the following identifies how Power Range Nuclear Instrument indication on the control console would change from BOL to EOL with II compensatory 1 At 100% power at EOl, the Power Range Nls would indicate __than actual power in the core. This occurs II Reference I I LOWER. Over core life Axial Flux will shift lower in the core, and PRNls are positioned to monitor the upper 2/3 region of the core. I Over core life fuel burnup produces a negative reactivity effect, so The location of the PRNls makes them more sensitive to power produced in the upper 1/2 of the core, and power production shifts over core life. This will be seen as a lower total power, as the I

Till!:l , in the

.a. ..nllir..'; for ......u .._ .. II I SUYI<,;U; INew i In.... ., .."" M;"difh..oltion ,"",..",\.1, II I iUsed During I rammg * ,"111......!

[J QUestion sOUrce CommentSl I______

___J I .. .1\,;"''''''''

J I Use this ONLY I I I I SROSvste;JfvolGilollUst OutlilleChanges*

Jm. I Which of the foJlowing conditions would result in Containment Fan Coil Units swapping from High Speed operation to Low Speed operation?

Containment Spray pumps just received a valid automatic start signal from SSPS. ALL off-site power becomes deenergized while operating at 100% power. A Rx trip Signal is generated in SSPS Train A ONL Y. A SBLOCA results in an automatic Safety Injection.

KAJ 022000A301

System/Evolution Title:! L... ...___....,KAStatement:-

Ability to monitor automatic operations of the Containment Cooling System Initiation of safeguards mode of

!ExPlanation of i 55.41.b(8}

A is incorrect because Containment Spray pumps automatically start at 15 psig in containment.

The CFCUs would have il Answers: shifted operation when containment pressure of 4 psig initiated a Safety Injection.

B is incorrect because the CFCU's are not I loaded on the EDGs upon a MODE II (Blackout)

SEC initiation.

C is incorrect because a Rx trip signal does not feed into the SEC !. I logic. D is correct because a Safety Injection results in the SEC initiating in MODE I (or 3 if a blackout occurs) which swaps running : CFCUs from hioh speed to lo.w speed. Note: Djstracters Band Care diffelllll1.

thouob each would result in a RlS: to!:;' since B I requires knowledge that a MODE 2 SEC initiation does not start CFCUs, as well as knowledge that a Rx trip does not calise a SEC initiation

.. Reference Number ***. .,. "'for""***.g.

II ........" Source:

• I New I Method: il Il{jsed During lTi;unmg j [J. *!Ol".",tlt'ln Source ...............

I I ....., iCI".m....m:*(Si . "); .... :.;/; ;.'..1' "j .; I I I I

" .... ' ....-. *m.,* ... 1 II Given the following conditions:

I -Unit 2 has experienced a double ended rupture of 21 RCS loop cold leg at 100% power. -Containment Spray automatically initiates, but neither 2CS16 or 2CS17, SPRAY ADD TK MOT OP DISCH Valves open. -Only 21 Containment Spray Pump starts. Which of the following describes the effect of this failure, and the EARLIEST time the EOP network will direct openinll of these valves? II Containment pressure may rise above the design pressure of 47 psig. 2-EOP-LOCA-1, Loss of Reactor Coolant, will direct 2CS16/17 after performing SI reset actions. , Containment Sump pH may lower <7.0 and provide inadequate absorption of iodine in the sump liquid. 2-EOP-LOCA-1 will direct positioning I I of 2CS16/17 after performing SI reset actions. I Containment pressure may rise above the design pressure of 47 psig. 2-EOP-TRIP-1, Rx Trip or Safety Injection, will direct positioning of 2CS16/17 when performing Containment Spray Actuation verification step. f Containment Sump pH may lower <7.0 and provide inadequate absorption of iodine in the sump liquid. 2-EOP-TRIP-1 will direct positioning of I I--=CS16/17 when performing Spray verification step. ., ....

jd I I

1 I Memory I I Salem 1 &2 i 5/17/20101 fKA: 11 026000AZ05 I AZ.05 I;R<?value:

1 3.7I[SRO Value:jl IIROGrour:>J 11 11 S*i.

[J 'SystemlEvolutionTitie?

I Containment S ra S '026i P Y Y KA Statement:; , Ability to (a) predict the impacts of the following on the Containment Spray System and (b) based on those predictions, use

....

• procedures to correct, control, or mitigate the consequences of those abnormal operation:

Failure of chemical addition tanks to iniect ;Explanation of;'I 55.41.b(9,2)

A is incorrect because keeping containment pressure below design pressure is accomplished with CS flow, not additive iAnswers:

tank flow, and will be addressed in TRIP-1. B is incorrect because it will occur in TRIP-1. C is incorrect because keeping .. containment pressure below design pressure is accomplished with CS flow, not additive tank flow. D is correct because FSAR i I section 6.2 states that the spray add tank, along with borated water provided during injection, assures sump pH >7.0 to ensure I I adeouate absorotion of iodine in the sumo liouid i I .-..... ---.....

i Reference Title :, Facility "',.",, .. Number Reference Page No.

ISalem II 116.2 !16.2-21 11 33 I II I I II I II J I II I iL.O.........,... I CSPRA YE004 i I I I I for ....... " i L I ,Question Source: c i INew L. :Question SourCe commentSl I ...

..-'....J iQuestion Modification Method: ..*. 1 IIUsed During !rammg r '1:1' ..... i 0 I "' ...... ,. , , .). ..... i.i**, " i.' , ". "'* i I I I

w SRO SVslemlEvolUflOri'UstJ*

OutllneChahges

'I I RO 48 Which of the following describes operation of the Containment Iodine Removal Units? The Containment IRUs are manually started from the Control Room ... or Hot Shutdown Network. on request from Radiation Protection.

in the SEC SI actuation modes.

I Containment Iodine Removal System lAbility to manually operate and/or monitor in the control room: ! iCIRS fans I 'ExplanationofH 55.41.b(9)

A is incorrect because there are no IRU controls on HSD panel. B is incorrect because IRU operation is not coptained in I I ,An,we,,, i I ,n, EOP. C;, oon," poe not, in 51.0P-S0.CBV-0001.

Contalnm,,"'

V,",iI,tlon Op'oo, thot """ whoo IRU, '" to '" nm. D L.__*.__ *.is incorrect because the SECs do not start IRUs in any Mode.

i R.".", <011,,<0 Title Facility '>V'w'w"Vw Number Reference Section .* Page No. 'F" .y. 11 i C0",a,,,,, ' Vt::IIU1dllUli uperation I[S1 :()F'-SO.CBV:OOg1 J !19 1125 I I I I I I I j J I I I 1 !

.SROSvstemlevokitiOn tist I I

Which of the following sources of hydrogen in containment during a DBA would MOST likely produce hydrogen in quantities which would require placing a Hydrogen Recombiner in service durin!:! performance of EOP-LOCA-1, Loss of Reactor Coolant? Breakdown of reactor coolant chemicals (ammonia, lithium,_

hydrogen peroxide) when exposed to oxygen. Zirconium-water reaction involving the zirconium fuel cladding and the reactor coolant. Radiolytic decomposition of reactor and containment sump water _____->, ___

I Hydrogen Recombiner and Purge Control 028 Knowledge of the operational implications of the following concepts as they apply to the Hydrogen Recombiner and Purge Control S stem: Sources of h dro en within containment I [Explanation OfH 55.41.b(9) "When inadequate core cooling has occurred, the containment hydrogen may be as much as 10-12 volume percent, III ,Answers:

_ "I depending on the amount of metal-water reaction(to produce hydrogen) that has occurred in the core." The hydrogen recombiners 1 _________* are put in service if required during performance of LOCA-1 at step 24, between CL Recirc and HL Recirc, less than 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> since : SI actuation.

I, I j' :RevisiOn!

ITMOOOE006 I CONTMTE001 I LOCA01 E008 1 IMaterial Required for Examination -. II ...... I I I I

'RO skvsciioorl I lit>

I.1;t'l SRO List) ..

Cha(igeS:

.. 1 .. ***********************,1

I RO I Which of the following would be the FIRST indication that a 5 gpm leak in the Unit 2 Spent Fuel Pool/liner has KA:!I033000A302 Title: ..................

....

______________________

---

4 sump pump normally aligned in auto, and has a rating greater than 5 gpm, Answers: , therefore high level alarm would never annunciate.

C is incorrect because while the tell-tale leakage can be seen, it does not ......--.. -. have a flow alarm. D is incorrect because all the leakage would be collected under the SFP, then routed through the teU-tales and collected in the sump. The 2R5 and 2R9 are located on a different level and far from where the would be and Reference Section ILO, Number ** **

... _U ,Objectives:;

i tir I ;:'rt"uuur:::OO7 .... Required for Examination .....-,.. II I ...\AI............Question Source: !New lIOu_inn'" ... ... I I!Used During Training I :.J Question Source' CommentSl I I__..--,-.....

I I.... ... , Ii . " I I I

• .. "

RO 51 I I Which of the followins identifies the effect of 2R32A, Fuel Handling Crane Area Monitor, failing HIGH? I I ITransfers Fuel Handling Building ventilation to HEPA plus Charcoal filters ONLY. I I I Locks out ALL Fuel Handling Crane motion other than downward motion of load. I!Locks out lateral movement of the Fuel Handling Crane trolley unless the hoist is in the Full Up position. !I, , ,1.Transfers Fuel Handling Building ventilation to HEPA plus Charcoal filters, and starts ALL FHB ventilation Exhaust Fans. !

.[Cf [Cognitive Level': LMemory I [FaCility:

i I Salem 1 & 2 I[ExamDate:

_: I 5117/20101 i

,__

!alue] 3.31 SYS I 21 SRO 21 0 r5ystem/Evolution TitlS:1 I Fuel Handling Equipment System J J:!<A I KnowledQe of the of the effect of a loss 2!" malfunction on the followins will have on the Fuel Handlins Eguiement I I Radiation monitorinq systems I I!Explanation of A and D are incorrect because the R32A does not affect FHB ventilation, the 2R5 and 2R9 Area Monitors do. Cis ,

__*1 incorrect because that is a normal crane interlock.

B is correct bec:ause the R32A interlock only permits lowering a suspended load. i i .*

v, ,,,.. Title Facility ,,,,'"v.........v Number v ...." Section ' Page No.

• I"UIIUI IIell n II S2.0P-A.8 RAn-0001 ! I I 127 I I Unit 2 Fuel ,

Crane 11242785 I I I I II I I J I '*_/;,/,.-1 It..O. Numner ,

1 I ABRAD1 E001 1 Required!or I II QUestion Source

,I Facility Exam Bank I iQuestion Modification Method: Concept Used I O'uring Training Programl ul L _____

I 'Question Source Comment;)

I Vision Q48703 L&,____,,--_.____,...J d ,...." I J I II list* I" SRO svStemiEvoluUori Changes ;QuestionTOpic-1 i RO L_,_______,_, I I With Unit 1 operating at12% power. which one of the following conditions will cause an AUTOMATIC Rx trip? I' rEXamDate:!

____.___* 035000K112 iKA StatemenD h sica/ connections and/or cause-effect relationshi s between Steam Generator S stem and the following: __

__

Explanation of 55.41.b(7)

A is incorrect because with power> P7 (10% power) but <P-8 (36%),2 RCP trips are required for Rx trip. B is iAnswers:

because PZR hi level trip is at 90%. 70% is the High Level alarm. C is correct. 0 is incorrect because Turbine trip does not cause Rx trip < P-9 (49%)

R.eference Section LO.

..---.--'I STMGENE008 ISTMGENE014

................

Required for ................." Jl I *n,,"'ti

....n .......u. ...... 1/IJ*lIN .*.,,"'...........................

"'. ,II j ,Used During Irammgvl:I

...."i 0 i *nll.........n Source Comments i I J .,' j 1-......._ *** " , ,.' .' I I I

'.

• .*1 iQuestion Topic i I SRO 1 ;-----'---

...........Given the following conditions:

-Unit 2 is operating at 75% power, EOL. -RCS boron concentr3tion is 100 ppm. -Control Bank D is at 144 steps. Which of the following C'..ontains both a condition which would cause control rods to withdraw continuously, and the interlock which would Rapid Borate Stop valve opens. Control Grade Interlock C-11. [Explanation of' 55.43(6) 2CV175 opening will cause boric acid flow to the charging pumps suction and into the RCS. Control rods will start stepping Answers: out and will continue to withdraw in response to the lowering temperature caused by the negative reactivity insertion.

Power will _____ . . remain fairly constant since steam demand is not changing, so RC loop delta T wiU remain constant.

Control rods will withdraw until (C-11 is reached, which is the "all rods out" interlock and stops auto rod withdrawal, The RC loop temp detector failing low will NOT I cause rods \.Q step out due \.Q the inout in " .' . I will LOWER RCS boron concentration as it removes boron to saturate itself. This would cause temp to RISE, and rods would move in. PT-505 failing high would initially cause rods to rapidly withdraw, but the power mismatch and the high level dip ofthe 1st stage , pressure signal act to stop rod withdrawal prior to all rods out or any Delta T signal. Power peaks around 85%, and rods will actually I begin to insert due to the power mismatch portion of rod control. The C-2, C-3 and C-4 interlocks all act to block rod withdrawal, and are 103% power, within 3% of OT/DT limit and 3% of OP/DT limit respectively. [Revision!

! I

• J:!'IlIir-1 for ......., ......... " .. I i?".."'tinn Source: I New ....::..*" Method: IIUsed During I I ...""'111......"" i 'lIf*..... !Question SOurce Comni'entS! "

......g ... .... , '," , J RO SkvScraperl'SRO Ro SRO "1

I e iSRO 2 Given the following conditions:

-Unit 1 is operating at 100% -Rx Engineering has Gonfinned a leaking fuel pin this -A Containment Pressure Relief is in service lAW S1.0P-SO.CBV-0002, Pressure Vacuum Relief, and 1VC5 and 1VC6 CONT PRESSNAC RELIEF are -The pressure relief was started with BOTH 1R12A. Containment Noble Gas 1R41D Plant Vent Noble Gas INOPERABLE as allowed by the 1-1 R12A is INOPERABLE due to the Containment APD Sample Pump being Which of the following describes how 1VC5 and 1VC6 will be shut if a PZR steam space leak were to occur and cause containment radiation levels to rise? I Assume a Rx Injection are NOT required. 11R45D, Plant Vent Filter, will automatically shut 1VC5 and 1VC6 upon receipt of a Hi Radiation A1ann signal. I 11R11A, Containment Particulate, will automatically shut 1VC5 and 1VC6 upon receipt of a Hi Radiation A1ann signal. I IKA

'..

..-., 'I I'Explanation of 55.43(4) 1R11A choice is incorrect because it, along with the 1 R12A and 1 R12B are the Containment APD. If the sample pump is II jAnswers:

OIS, then no air flow from containment is being pumped through the unit, and none of the 3 detectors will see the increas.e in .....radiation, NOR will they perform their cont isolation function.

The pressure release is specifically allowed to start with the 1R12A II' I and 1 R41 0 lnop, as long as requirements for double sample/analysis is perfonned.

The R45D does not perfonn any isolation I functjons.

The control room will be rectlJjred to shut the VC5 and 6, since no auto jsolaljQn would occur. The R41A low ranae I monitor would still be available to provide the indications in the stem. The initiation of a CVI signal would send a close signal to I 11VC1,4,5 and 6, since valves VC2 and VC3 were removed during a DCP. . I c-*----**** -..

....

..:

.....--..

,"""--" r,--"', j' ... ..... Refer!nce Title .. . . Facility REiference

Reference Section I Containment Pressure-Vacuum Relief !I S1.0P-SO.CBV-0002

.-J I Off Site Dose Calculation Manual I Salem aDCM .-J I !1 24-26 ___________

---

'IS1.0P-AB.RAD-0001 II 11 15-18 iUsed During Training Program L_"____.___.._______________

__________

___.___ ,

RO SkyScraper SRO Skyscraper I RO System/Evolution list I SRO System/Evolution List I Outline Changes I 3 I Given the following conditions:

-Unit 2 was at 100% power when a 500 gpm SBLOCA

-Operators are performing actions in EOP-TRIP-1, Rx Trip or Safety Which of the following would prevent charging pump ECCS flow from being present, and when is the EARLIEST that action will be directed establish EJ I.NEITHER 2SJ12 nor 2SJ13, BIT OUTLET VALVES repositions; during Safeguards Valve Alignment in TRIP-1. I INEITHER 2SJ4 nor 2SJ5, BIT INLET VALVES repositions; after SI reset in LOCA-1, Loss of Reactor Coolant. I IVCT OUTLET STOP VALVES 2CV40 shuts and 2CV41 remains open; during Safeguards Valve Alignment in TRIP-1. I id.ll RWST TO CHG PMPS STOP VALVES 2SJ1 opens and 2SJ2 remains shut; after SI reset in LOCA-1, Loss of Reactor Coolant. I I II Level [CJ [Cognitive Level : I Application I

I Salem 1 & 2 I lExamDate:

r= 5/17/20101

'-" I I KA: 1000009A213

__

0 Title:' I Small Break J Ability to determine and interpret the following as they apply to Small Break CharQinQ pump flow jExplanation of 55.43(5) B is incorrect because the BIT inlet valves would be positioned correctly in TRIP-1. (Right condition, wrong procedure)

A 'Answers:

correct because the 2 parallel valves are normally shut, and would be positioned correctly in TRIP-1. C is incorrect because these ..... valves are in series, and only one has to open. (Wrong condition, right procedure)

D is incorrect because these 2 valves are parallel, and only one has to open.(Wrong condition, wrong i" Facility Reference-Nul1liier-

--PageNo-:-:

I Reactor Trip or Safety Injection ITRIP-1 II 1127 I I II-II II I II-II II 11_--,1 IL.O. Number II Objectives I L TRP001 E021 I Material Required for Examination 1 II [Question Source: : [Bew IIQuestion Modification I I Used During Trainiflg 0 L_ L--. . II Question Source CommerrtSl I .... ..

I __--j I I I I I I

....

I SRO 4 I ,

i Given the following conditions:

-Unit 1 is responding to a -When the crew is pe10rming SI reset actions in 1-EOP-LOCA-1, Loss of Reactor ONL Y Train A SI will reset on the Safeguards -Operators continue in LOCA-1, and transition to 1-EOP-LOCA-3, Transfer to Cold Recirculation, when Which of the followin9 identifies how the Control Room Su!!ervisor would direct the ECCS to be realiened to Cold Recirc?

pumps will be unable to be stopped when realigning suction to containment sump. Go to 1-EOP-LOCA-5, Loss of Emergency ecirculation.

I, I A loss of off-site power will cause the SECs to load in MODE", and Safeguards loads will have to be manually restarted while continuing in LOCA-3. 111SJ44 and 12SJ44 CONT SUMP SUCT VALVES will not be able to be opened after 11 RH4 and 12RH4 RHR PUMP SUCT MOV Go to I e: i 111 CC16 and 12CC16, RHR HX COMP CLG OUT valves would not open when operators armed them at 15.2' RWST level. Continue in LOCA-I 3 while dispatching operators to locally open the CC16 valves. ..

!Application I Salem 1 &2 I c= 5/171201011 iKA.:;1000011A202 ....

afety injection EXplanatIOn-Of i 55.43(5) When performing LOCA-1, SI reset actions occur early in the procedure.

When SI on train B does not reset. the SI input !Answers:

signal to the SECs will remain. The procedure then BLOCKS this signal when the SEC's can not be reset This BLOCK remains in I .--..

place, so when the loss of off-site power occurs after SEes have been reset. the SEes will NOT have a Sf input signal. and will load in MODE II (Blackout).

The Step 3 CAUTION in LOCA-3 prior to Safeguards Reset states that if a loss of off-site power occurs after 5! reset then saftlQ.lJards Igads wi!! ball!:l to be manually restarteg.

6. is incorrect because tbe SECs WILL be I the RHR pumps CAN be stopped, C is incorrect because the SJ44 and RH4 valves are able to be positioned, D is incorrect I because there is no auto arming capability on UNIT ONE, I

____

__

____I LOCA01 E006 I ECCSOOE008 I SECOOOE005

, I"""'".... Required for I I II I'n.....,tinn lUI. """" ,..,

Source: I New ..... _ .."" !I I'Used During I Htlnlng ::1'

D [Question Source Comments II '-...--'..."---........ _.I .I...v ...........

' ...... ,.,." ..,.. ,..*.....,.... , r . I I I !I I II, I SRO SvstemlEvokrt1on List* '.( **Outline Changes !!QueStionTOPic J I SRO 5 !I Given the following conditions: .I -Unit 1 is operating at 100% power. Charging system problems result in NO Unit 1 charging pumps running OR being I -The decision is made to align and run 23 charging pump (Unit 2) from Unit 2 RWST to the to be i Unit 1 CVCS system lAW S1.0P-AB.CVC-0001, Loss of I -The proper lineup is completed.

and 23 charging pump is now supplying the Unit I charging I 1-Whoh Unit 1 PZR level is rising very slowly. 0"" follow;,.

'oo@" I'" ,,,.301;00 to be ee"",,,, od*.0' -, Commence a Unit 2 shutdown lAW S2.0P-IO.ZZ-0004, Power Operation.

due to the loss of Unit 2 RWST level. I I I b. i I Commence a Unit 1 shutdown lAW S1.0P-IO.ZZ-0004, Power Operation, due to the boration of the RCS from Unit 2 RWST. I I Reestablish normal letdown on Unit 1 lAW S1.0P-SO.CVC-0001.

Charging.

Letdown, and Seal Injection, to control PZR , I IAdjust 11-14CV98 RCP SEAL WTR FLOWADJ VALVES, lAW S1.0P-SO.CVC-0001.

Charging, Letdown. and Seal Injection.

to ensure 6-12 gpm flow to each RCP.I !:Answer '. I b /Exam Level fS'! 'Cognitive Level I!ApplicatiOn i

Salem 1 &2

• ..5/17/20101 i

I Loss of Reactor Coolant Makeup to determine and interpret the following as they apply to Loss of Reactor Coolant Makeup: I Charging pump problems I I ,Explanation ofl 55.41(5) B is correct because the RWST boron concentration is always going to be higher than the RCS boron at power, and will I i Answers: ! result in a continuous boration of Unit 1. (AB.CVC*1 Step 3.49) A is incorrect because RWST level is monitored.

and if it approaches ,

minimum level, continued operation of 23 charging pump will be evaluated or other contingency actions implemented (AB.CVC-1 II' I Step 3.49 NOTE) C is incorrect because normal letdown can NOT be restored since the letdown isolation valves are interlocked with , the chargina pump breakers such that at least ONE charging PIJrnR ha§ to be shut to open letdown isolation Excess letdown may be placed in service to control PZR level. 0 IS incorrect because while seal injection flow adjustment might have to be I made, it is done by throttling 1CV71CHG HDR PCV, not adjusting manual valves in the field. , "

I Previous 2 NRC Exams I[Question MOdJficationM&thoci:-li EditOrially Modified I["""",0" I,;';!:C SRO "'pi".. '"",d" db";' "0,,""" pro"'",, 10 do", "''"..,,, ;,. I[ . .'.. , ;;,!-"'....,....... , I , "'" '" "" " "" " I I I II

SRO List I OoiiineChanges I

r SRO 6 !

II Given the following conditions:

I !-Unit 2 is operating at 100% power, MOL. -21 SGFP trips. I I i ,-NO operator action is taken in response to the SGFP trip. 11 Which of the following is an UNEXPECTED alarm if it is locked in 2 minutes after 21 SGFP trips, and what procedure would be used to address the I condition associated with that unexpected alarm? [OHA G-3, EHC SYS TRBL. S2.0P-AB.TRB-0001, Turbine Trip Below p-g. !Console Alarm RC PRESS DEVIATION HI. S2.0P-AB.PZR-0001, Pressurizer Pressure Malfunction. I OHA G-44, COND POL TRBL. S2.0P-AB.CN-0001 Main Feedwater/Condensate System Abnormality.

I [console Alarm RC LOOPS TAVG-TREF DEVIATION.

S2.0P-AB.ROD-0001, Immovable/Misaligned Control Rods. I

[:J L!ve.' . ! IApplication j

I Salem 1 & 2 I I 5/17/2010 1/ I 000054G446 r-2.4.46 '

4.2 I 11 I 11

SystemlEvoh.rtion Title;

L_._.__. __. ....___ _..

that the alarms are consistent with the plant conditions.

Explanation of. 55.43(5) The condensate polisher trouble alarm will be in alarm due to the CN108s AND the CN109 being open at the same time. i Answers: .D is incorrect because RC loops Tavg-Tref deviation will be expected as rods are driving in due to the turbine runback to 65%. The !

ARP has operators place rods in manual, and if not successful at restoring conditions, going to AB.ROD-01.

B is correct because I the RC pressure deviation would not be expected, since the setpoint (+ 75 deviation)equates to when the valves are.full r....

....--* .. -..-....

....-" ...--.....-'... -.....-.-...... .....

10'--*-..... ...

...___

...

...

...

_____I Control Console 2CC2 IIS2.0P.AR.ZZ-0012 iI 1136 11 35 I I Overhead Annunciators Window G II I!68-69, 8 1144 I I 11__ II II II I I Material Required for Examination I Source:-'" i New I i?u:stiO...Modification Method: J 'iUsed During Training I I Ad',"OM' 'of,mnre S2.0P-AB.CN-0001, Maio F,adwata,/Coo'",.." S",om

'.24 pa" 6

--. ---..l n I I I !

.i ""RO list) SRi> SvstemJEvoiutlonlist J Outline ChangeS

!Question Topic I ISRO 7 Given the following conditions:

-Unit 2 is operating at 100% -The performance of the daily RCS leak rate calculation determines that Leakage has risen from 0.07 gpm to 0.50 -Shortly after the RCS leak rate is performed, Chemistry reports that a routine sample the RCS indicates that DEI-131 is 8 uCi/gm, and a second sample confirms the 1

-2R31, Letdown Line Radiation Monitor is NOT in Which of the following is the FIRST regulatory notification required if Rx power and RCS activity levels remain at these levels? :t<A11_00_O_O_76_G_4_41_---11

syst!m/EVoluti()n I High Reactor Coolant Activity J __.. __._ action level thresholds and classifications.

I iExphlnation'oij 55.43(5) Tech Spec 3.4.9 for RCS activity will allow operation at this power level with this activity level for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before a unit I' I Answers: ! shutdown is initiated.

The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report distracter is from EAL 11.1 Technical Specifications, that requires the report upon initiation ,of a unit shutdown directed by Tech Specs. which would happen after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of operation at this RCS activity level OR if the RCS I 'leak rate was> ONE gpm unidentified leakage, However, EAL 1.1.1.a RCS Activity states that operation at this level for >48 hours 'I'j I reouires UE declaration, whjch would occur before the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification.

Ihe 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> djstracters are fouOd tt[Ql,!ahout " the Reportable Action Level section of the ECG. Additionally, if candidate goes to IS 3.7.1.4 for specific activity of SECONDARY

.1 coolant by mistake, they will think the level is exceeded and the unit shutdown will have to be done within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, ,

iRevlsion'

_..--..

I 1258/262 i 1-------______11------------'1-------'1---" I I I;Material Required for Examination

• I SRO 7 ECG with CFST Section removed, and Unit 2 Tech Specs 3.4.9 AND 3.7.1.4 .J o' ,Question Source: ; I..'=acilily Exam Bank I iQUeStiOn Modification Method:]!Significantly Modified I iUsed During'TralningJ'rogram 0

...

.._*__.__...

I Source I Changed from required TS action to first E-plan notification required.

Added RCS ULR rise for added complexity.

I ..f> ..)';; *.**

• • • .* j; 'c. ..... '. .. ... .. i I I I

• RO skvscra,;rilsRO List ] S . SRO SVsten1JEvolutk,;,Ii.Ist

,;:

TOpiCll SRO 8 , Given the following conditions:

-Unit 2 was operating at 100% power when a LOCA occurred.

1-The crew is now in 2*EOP-LOCA-1, and the STA observes the following conditions:

-All rods are fully inserted.

-No RCP's are operating. -A total of 10 CETs are reading between 725-750 deg. F. -Remaining CET's are reading between 660-690 deg. F. -Containment pressure is 24 psig. -Containment sump level is 52%. -RWST level is 17 ft. -All loop Tc's are between 275-290 deg. F. RVLIS Full Range indicates 43%. -RCS pressure is 265 psig. -Only 2A 4KV Vital Bus is energized.

At this time, which of the following identifies the procedure which must be implemented FIRST? Reference provided.

fa. !1.2-EOP-FRCC-2, Response to Degraded Core Cooling. Ii II !2-EOP-FRCC-1, Response to Inadequate Core Cooling. . . !2-EOP-FRCE-1, Response to Excessive Pressure.

I EOP-FRTS*2, Response to Anticipated Pressurized Thermal Shock EJ §LeveI: Is I I Application I

i-------'

I I :!"aterial for Examination

' !2-EOP-CFST

-1 Figure 4A Thermal Shock Umit A Curve (page 71 of II I!Qi.iestion Modification Editorially Modified I[USed During Training Programl 1I3cility Exam Bank L, ,,' ' --iQuestionSource Comments I Q77770 replaced LOPA-1 and LOCA-5 as distracters with FRTS-2 and FRCC-1, and modified stem to have II ___,__,____,_c Purple path present. Answer remains the same II '!'1" ..."'". ; I I j I

. .. .. '............. J'.'.:. ' ...**7 ',r" RO SkyScraper

SRO Skyscraper

.. 1:'

.. 1... Given the following conditions:

I -While operating at 100% power, Unit 2 experiences a large, unisolable steamline affecting ALL Stearr.

-The control room crew is responding to the event in the EOP's, and the rupture

-RCS pressure is 1349 psig and

-ALL SG WR levels are 20% and

!l' '1' Which of the following describes an action, and the bases for that action, which will be performed to mitigate the potential adverse consequences of the event? I Running RHR pumps will be stopped in LOSC-2, Multiple Steam Generator Depressurization, if RCS pressure is greater than 300 (420 adverse) psig and stable or rising. to prevent potential damage to the RHR pumps due to heat up of the recirculated fluid. ibJ IAFW flow will be reduced to no less than 1.0E4 Ibmlhr in LOSC-1. Loss of Secondary Coolant, to ensure steam generator tubes are kept wet ii' L and minimize the possibility of SGTR when the steam generators have depressurized. . i RCPs will be tripped in LOSC-2 after verifying ECCS flow is established so that CET temperatures do not become superheated if forced II .. .t I circulation in the RCS stops. I.

1 23 AFW will be stopped if not needed for AFW flow in LOSC-1, to extend the before all steam generators completely blow down. It j fQ" L!vel I I Memory I I Salem 1 &2 c= 5/17/201011

/KA:lI00WE12G418

_',

21 I Uncontrolled Depressurization of all Steam Generators

-.J ...KA Statement:

i .....-.-....

Knowledge of the specific bases for EOPs. i ,Explanation of 155.43(5)

C is incorrect because the RCPs are not stopped if a cooldown is in progress.

RCP trip criteria specifically asks if a '\ 'Answers: . cooldown is in progress, and if it is, does NOT stop the RCPs at step 6 of LOSC-2. A is correct because LOSC-2 Basis Document.

,_.._, page 24 states..."On the RHR system where the pump recirculates on a small volume circuit, there is concem for pump and motor overheating.

The RHR HX cooling isolation valve CC16 does not automatically open, so the fluid being recirculated in the RHR ,system is not beina cooled If it weee iIJiectjng, tbe cool water flowing thwugh the pump would be providing coolina, 13 if c.orrect I because it will be done in LOSC-2, not LOSC-1, with the right reason. D is incorrect because while there is a step in LOSC-1 for !stopping 23 AFW pump. it comes AFTER the transition out of LOSC-1 to LOSC-2 occurs.

Rc:qulred for Examination

_, i I ... ........ I I I I II

";110 t tiRo

[I .RO List "1 Outline ChamlesJ 10 Given the following conditions:

-Unit 2 has experienced a LBLOCA coincident with a loss of off site -2C 4KV vital bus locked out on bus

-2B SEC did not Assuming one train of ECCS equipment is operating, which of the following identifies the FIRST action which will restore the minimum complement of equipment to assure containment integrity is maintained lAW Salem FSAR? I Resetting 2C SEC. ;b.. , Depressing START PB for 21 CFCU. ,

.....es.....s.....

_________________________________r

__Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and I instrument inte retation. . Explanation oflI55.43(5, 1) FSAR Section 6 and 15 both state that the minimum complement of Containment Spray Pump/CFCUs require,d to ensure II !Answers: . ., containment integrity along with a train of ECCS in operation is 1 CS pump and 3 CFCUs. With the conditions in the stern. only 21 L._______; CS pump will be running on A bus. C bus will be deenergized because a bus differential signal locks out all power to the 'Additionally.

the power supplies to the are A,B.C,B,C for CFCUs, so only 2.1 CrCU will in operation; A is incorrect lthat will restore the 3 CFCUs needed. D is incorrect because C bus has no power. [.0.

L ___.___ICSPRAYE002 I I CONTMTE002 1___---1 for EA..... " ..dion Ii I ......." ........iO-;u...tiM Source: I New I I I Used During murllng Pivl:j.G.., [J In..""fi,," Source .......I . I Iv................

','.,,,' ...........,..... ...* ... " ...."... II I , II il I SRO

'1 " .. ...

J§O 11 I Given the following conditions:

Unit 1 is performing a Tech Spec required shutdown due to 11 and 12 charging pumps being

-The unit is tripped from 20% -Upon entering EOP-TRIP-2, Reactor Trip Response, the following conditions

-3 control rods remain full -PZR level is 8% and

-NO AFW pumps are running and the lowest SG NR level is 1CC131, RCP Thermal Barrier Retum Valve has shut. I Which of the following identifies the reguired actions for these conditions?

c!J 1.lnitiate rapid boration for 105 minutes while continuing in TRIP-2. I I Initiate Safety Injection and go to EOP-TRIP-1, Rx Trip or Safety Injection.

I Either start at least ONE AFW pump or immediately go to FRHS-1, Loss of Secondary Heat Sink. I Stop ALL RCPs and transition to EOP-TRI P-4, Natural Circulation Cooldown, when directed to perform cool down required per Tech Specs. I

.. j . j I =. : I , . -IApplicatlOn I faCility-1 &2 5/17120101 I Emergency Core Cooling System ..-..--

__--:--:------:::-:-:--:---:----:-:-:----

..

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

Conditions requiring actuation of ECCS l iexPlanation of 55.43(5) A is incorrect because while the rapid boration for 105 minutes is directed by TRIP-2, the higher priority condition of PZR Answers: ..' level <11 % requiring SI initiation and transition to TRIP-1 is present. B is correct because with PZR level <11 % and unable to be ..._ .... _--.-maintained

>11% (which will be the case with only 23 charging pump available, plus letdown is isolated), then initiation of SI and I transition to TRIP-1 is required per TRIP-2 CAS. C is incorrect because transition to FRHS-1 is not performed until SG NR level <9% with no AFW OLmQS running, D incorrect because CC131 closing does oot isolate all cooling to RCP's. it only isolates I thermal Barner return flowpath, and as does not reqUire stoppiog all RCP's as per CAS of TRI P-2 if all RCP cooling Yfas

__ I ECCSOOE015 I TRP002E005 I.M.M., Req",..d "" Ex.m'n.tion

'--I InIlA.dl",n Source: i lli ew IIQuestion Metho<: I ,l,Ised DUri,ng Trai,,!ing D. ! [Question' i I I ;;,,': ;;/';';.1

.;' ;_w......w ..1 ; ; .. ';"";;?:

.. ;........ 1 I 1 I

"""

TopicJ §O 12 I r Given the following conditions:

I -Unit 2 is operating at 100% -PZR Pressure Channel I (one) is selected for

-PZR Pressure Channel IV (four) is selected for i-A channel calibration of Loop 21 RCS Narrow Range Th and Tc channels is being I performed lAW S2.IC-CC.RCP-0001, 2TE-411A-B

1. 21 RX COOLANT LOOP DELTA T-TAVG PROTECTION CHANNEL -All required bistables have been placed in their test Which of the following describes the result if PZR Pressure Channel IV (four) instrument transmitter loses power, and what action, if any, is required to be [!erformed?

E: I A Rx trip will occur on Over Power 1 Delta T. Operator action will be required to shut PZR Spray valves to prevent Lo PZR Pressurel Safety Injection in EOP-TRIP-2, Rx Trip Response.

I IA Main Turbine Runback will occur on Over Temperature!

Delta T. Depress GO to backup runback signal on Digital EHC consollAW ... AR.ZZ-0012.

Control Console CC2. I A Rx trip will occur on Over Temperature

/ Delta T. Perform immediate actions of EOP-TRIP-1.

no extra mitigative actions will be required.

I I I.NO plant response other than annunciation of alarms will occur. Verify alarms are consistent with actual plant condition lAW ARPs. II I llAnswer EJ !examLevel' rs==r iCoQnrnveLeVel'

!APPlication I

1 & 2 5/17/201011 I

lSystemiEvOlution Title:i I Non-Nuclear Instrumentation System

-.'------.. ' I . I 1=2=TE=-=41=1=A-=B=#=2=15=5=::;I RPS PZR Press and Lvi Control 1/221060= J I i 17 FI

____

, I RCTEMPE007 I RCTEMPE008 I___-.i I I *** , .."':..... n,.."t..........

!.?rExamination d euestion Source: I INew IIQuestion Modification

....... i IUsed During """""---_....._-I D euestion I I '.I",on.."':" IL J ! I I .....uu"." [Que-suOil-Topici i SRO 13 J_,._---'---

_______ _____ __________

J IWhile operating at 100% power during the winter. 21 CFCU high speed breakers trip on over Which of the following identifies the effect this will have. and what action will be taken in response to the CFCU I I Containment humidity will remain constant.

Open 12SVDC control power to the high speed AND low speed breakers to ensure 21 CFCU does I I not start on a valid auto start signal lAW Tech Spec 3.6.2.3. Containment Cooling System. .1 SW flow to the remaining in service CFCUs will rise noticeably.

Start a backup CFCU in High Speed lAW S2.0P-SO.CBV-0001, Containment Ventilation Operation to restore normal system configuration. \ :ell SW flow to the remaining in service CFCU's will remain constant.

Start a backup CFCU in Low Speed lAW S2.0P-SO.CBV-0001 to prevent L-l rise in containment tempernture I I icill Containment humidity will rise noticeably.

Open 12SVDC control power to the high speed breakers to restore operability of 21 CFCU lAW i I Tech Spec 3.6.2.3 ' I Answer._ I b I Level ., S I Level '! I Application J

I Salem 1 & 2 I iExamDate:

i '-S/17/20101 , -',.__.,'--"-. KA:l1 022000A20 1 I,A2.01 :RC?y'alue:J!

2.71.section:

ISYS I Gr'>.up:" 1 , Gr,<lup:'

I 1!

.. 1 022---'

Tltle:J ICoru" ....... ,,"" Cooling ......Ii(A Statement:

' I Ability to (a) predict the Impacts of the followmg on the Containment Cooling System and (b) based on those predictions.

use

..-i procedures to correct, control. or mitigate the consequences of those abnormal operation:

i Fan motor over-current Explanation of ,IS5.43(5)

SW flow will rise due to the removal of 1950 gpm from the system when 21 CFCU trips. The flow to the other CFCUs will Answers: .1 rise due to increased SW header pressure.

The O/S CFCU will be started to restore normal operating configuration of 4 CFCUs in l,____...-C! high speed. Salem removes the 125VDC from the high speed breakers.

but that alone does not make the CFCU operaole.

since the cause of the breaker trip is not kno:n. Containment humidity will rise. but a very small amount. and not noticeably.

!

Reference Number_. IReference ipage NO:-iRevlsion*

I r** -.....

II Control Console 2CC1 II

-.J I i \S6 ! II -.J! 1/ II !

I CONTMTE007 I

I\J_ew______-'1 in. .." MvdifiwtivlI

....... I I jUsed During Training Program I 11 .. laueStlOnsourceCommentsi I ...... .'. ! I I I II RO SRO.Skvscrauerl

§014 Which of the following identifies how post accident radiation releases from Containment following a lBlOCA are maintained less than that assumed II, in the Salem UFSAR? . ;a'I limiting Rx Thermal Power to 3459 MWth to limit amount of fission product buildup. Containment Spray system operation to maintain containment pressure < Test Pressure 54 psig. system operation to maintain containment pressure < Design Pressure 47 psig. allowable fuel defects, and replacement of leaking fuel bundles to maintain the effective ratio of leaking to intact fuel of <10%. 'SystemlEvolution Title?

____

i r-----------------------.---------------------------,.

Explanation of

• 55.43(4) Salem UFSAR, section 6 states that reactor containment ensures that post-accident leakage is to a rate Answers: ' 0.1 % of the free containment volume per day at the design pressure of 47 psig. Containment spray operation is required to keep ..-----.." containment pressure <47 psig. A is incorrect because limiting thermal power ensures that fuel peaking and hot channel factors are maintained

.. B is incorrect because while the containment is tested at 54 psig test pressure, the FSAR states it is the design . . SlOt leakage D is incorrect because the as:?umptjon in peellon lS foe a DBA is i:}8t 1% I, i failed fuel is assumed to be present before onset of accident.

i___________

__-______________

R_.e_f_er_e_n_c_e_ii_u_m_be_F

______________________IL:O:Numher' . ...

!Material Required for Examination , I--II :Ouestion Source: .. ' \:§w !iQuestion'Modificatlon Method: "1 1 I[used During ) CJ c __ccc _._

• __"c ** _

..

..

L..... 'Question sc,-urce'CommentS I I**___**__**_______M....____; I .'. . ..*.. I* 1....... * ....."*.. 1 I I I I J iRb SkyScraper

...

I" Outline Change$. 1.

Given the following conditions:

-Unit 2 is in MODE -OHA C-35 SFP LO -The NEO dispatched to investigate reports SFP level just below the 10 alarm setpoint, appears to be -No leak identification action has been I Which of the following describes the actions reguired for this condition? I Occasional SFP low level alarms are to be expected due to the leak on the SFP liner, refill the SFP using preferred source CVCS HUT water if I L_. available to maintain boron concentration as high as possible lAW S2.0P-SO.SF-0001.

,'Direct the operator to investigate source of possible leak, and refill the SFP using preferred source demineralized water lAW 0001, FILL AND TRANSFER OF THE SPENT FUEL POOL. I IMMEDIATELY GO TO S2.0P-AB.FUEL-0002,LOSS OF REFUELING CAVITY OR SPENT FUEL LEVEL, to isolate the SFP cooling pumps individually to isolate the most likely source of leakage. II I Monitor 2R5 and 2R32 SFP Area Radiation Monitors, which will initiate 22 HEPA PLUS CHAR mode of FHV lAW ':'L..ur'-,",o.

Number

.......SFPOOOE006 rla.'... Required for Examination

.. II ,....

• .1 l.rrevious 2 NRC Exams I Editorially Modified Comments']

112/2006 NRC SRO exam Q16 I "'"......""

I RO SkvsCriDef ISSRoskvscralltlrl';;;Ro sVStemtE;olutlOn List I l 0'Utline 'Changes I .......

TOPiC] §o 16 While performing actions in 2-EOP-FRCE-1, Response to Excessive Containment Pressure, due a steam rupture in containment, the Step 6 CAUTION states, "AT LEAST ONE SG MUST BE MAINTAINED AVAILABLE FOR RCS COOLDOWN." Which of the follow!n!:!

describes how that caution will be applied to the subseguent steps for SG Identification and isolation if ALL SGs are faulted? Choose ONE SG to remain available for RCS cooldown and feed it at a rate which maintains RCS cool down <100 deg.lhr., while isolating all sources of feedwater to the other 3 SG's. Choose ONE SG to remain available for RCS cooldown and feed it between 1.0-5.0E4 Ibm/hr, while isolating all sources of feedwater to the other 3 5G's. 1.0E4 Ibm/hr total COMBINED _____

___]

4.31 'Sectio'!D I 11 11 r.-.---.....

iKA Statement: ... --..

II'"'K-n-o-w-:"'e-d':""g-e-o-=-f-op-e-r-at-=-io-n-a-II:-'m-p-:-li-ca-t-io-n-s-o":"'f

!ExPIanatlon of 55.43(5) D is correct because the CAS following the CAUTION speCifically says if ALL SGs are faulted then minimize feed to ,Answers:

SG to NL T 1.0 E4 Ibmfhr. This is NOT step memorization, because the operator should know that with ALL SGs faulted, ALL ..-'--.....need to be fed at the minimum verifiable feed flow to keep the tubes wet. This prevents a situation where SG' are allowed to dry out, then subsequent feed is initiated and creates significant thermal stresses on the SG components.

A and B are incorrect to keAO flow . to All SGs. ThAV . .I steps dealing with ALL SGs being faulted. The 100 deg/hr cooldown rate limit is found in numerous places in the EOP network, and is the entry condition for FRTS-1. The 1-5E4 Ib/hr is found in FRHS-1 when feed has been restored and certain WR level is established.

C is incorrect because 1.0 E4 is the minimum verifiable feed flow to each SG.

t ,

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Facility ""'",,,,,.....,, ,.............

""'''''''''

Section i Page No. ...--.--' I R"'JJv, ' to ,.... Corllall" ""' Pressure I!

r:RCE-1 I I 1122 I I J I I II I I I ! I II I IMaterial Required for Examination i l... . . -. ... "' , .. II Source:., I New ii

[J.--...

In......ti....n Source Comments i I ......" ... >>:i',. *. ii'** .IF ....... ' ,.! I I 1 ROSkVSCraPflrlm RO SvstemlEvoluflOnUst

[QUestion Topic", ,'sRo 17 '-___._. _. , ____ _

Given the following conditions:

-2C EDG is completing a loaded run lAW S2.0P-ST.DG-0003, 2C EDG Surveillance -The NEO performing the unloading has reduced electrical load to 500 KW, 250 KVAR -The NEO announces the next action he is performing is lowering load to 0 KW and the EDG output

-Which of the following identifies any potential concerns associated with this action, and the Field Supervisor, how should you Reactive loading must also be reduced to 0 KVAR or the large arc across the breaker contacts when it is opened leads to reduced breaker life. Ensure NEO also this reactive load reduction.

The EDG voltage regulator will automatically respond to raise voltage as speed is reduced. Ensure NEO lowers voltage to prevent I excitation trip of EDG. The EDG may overspeed when the output breaker is tripped. Immediately stop the NEO from reducing load. Immediately stop the NEO from reducing load.

System/Evolution Title: :KA Statement:

' Reference Section J ..... m ..__...........

_._..m i!::..C>.:

__ ,12 IFI I I *!,.!C.... , lQ., Required for Examination

! I I ...tjon Source: I New I IIUsed During 'raining y.a...

D Question Source Comments I I I I I RO SvsteIDIEvolutionUst

.. ... " ... ,.*... l[Question*Toplcl 18 I i ! Given the following conditions:

-Unit 1 is shutdown curing a refueling outage. -A normal release of 14 Waste Gas Decay Tank to the plant vent is scheduled to performed on day shift lAW S1.OP-SOWG-0011, Discharge of 14 Gas Decay to Plant When reviewln9 the schedule, which of the followins activities is allowed to be scheduled during the period 14 WGDT is being released?

I I Release of 11 WGDT. !.Inltiation of Unit 1 VCT purge. Transfer of gas between 12 and 13 WGDTs.I I I d*. , Aligning Unit 2 Vent Header to Unit 1 Waste Gas Compressor suction. I I b I [EXa ll1 Levee I S I 1 Memory I I Salem 1 & 2 I c= 5/17/20101 i KA:'1071000G218

!

_J Va'lI e j I 2.61 I SYS I 21 fSR9 0

IWaste Gas Disposal System

__________________________________________________________-. Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, atlon, etc.

,---" """""_w lQuestion Topic I IsRO 19 L .. , _, I I The control room operators are currently responding to an accident.

which has required implementation of the EOPs. They have recently made a REQUIRED transition to a Purple Path Functional Restoration Procedure (FRP). While performing steps of this procedure, a condition arises which is covered by an action contained in the Continuous Action Summary (CAS) of the EOP that was in effect. Which of the following identifies the correct course of action to be taken lAW OP-AA-101-111-1003, Use of Procedures?

1 3 ...* 1 Continue actions in the FRP until it is exited. Do NOT perform EOP CAS action until EOP is the procedure in effect. i I Perform the EOP CAS action as soon as it is identified ONLY if it does not conflict with or reverse any action being taken in the FRP. I Continue actions in the FRP until it is exited. ONLY perform the EOP CAS action if it is also contained in the CAS section of the FRP. fd.'1 Perform the EOP CAS action as soon as it is identified since CAS actions remain in effect from the point the EOP is entered until the EOP has I been completed, L .. "3 r-----;-r---1 ..

.. llMemOry !

1 & 2 ---.J ..

r= 5/17/201011 Title:J 1---------------------------------------'

GENERil-...

rl<A Statement:

! .....-........""..

Explanation of A is correct because once a FRP is entered, it is performed until completion OR transition to a higher priority Cis Answers: inrr)rrpM because FRPs do not have CAS's. Band D are incorrect because CAS actions from EOPs are NOT in effect in FRPs ...-....

it is a YELLOW priority FRP. 'L.O. Number L-..-___, _______.. __ _ ITRP001E007 ITRP001E010 Material Rio",.." ""u for Examination L I puestion Source: ,INew IQuesilon Source CommenW I

...

Ii' u ....,4;, u. II l'Used During I rammg .-. ul:I, ..... D I ,C"'nn.",... .,. i J I I I

• • I?sRlloskl'6CraPtfI*

1 20 Given the following

-Unit 2 is performing a cooldown lAW S2.0P-IO.ZZ-0006.

Hot Standby to Cold Shutdown.

-The unit has entered MODE 5. and the cooldown rate has been lowered to 45 deg/hr. -PZR temperature is 400 deg F. The board NCO reports that PZR hot calibrated level on all 3 channels indicates 40% and is stable. Which of the following identifies what the Cold Calibrated channel will be reading. and how charging flow should operated?

Reference provided. . !Cf:J 1 35%. Maintain charging flow stable.

[Q"

[CJ IAPPlication I

ISalem 1 &2 I c= 5/17/20101 194001G125 I

...

___________________________________________

.______________________________ret reference materials, Explanation of ; 55,43(5)Using Exhibit 1 of IOP-6, page 2. shows that with hot cal level at 40% at 400 deg., the ACTUAL level in PZR is between Answers: 35%. Using Page 1 and the ACTUAL PZR level of 35% and going across to the 400 deg line, cold cal level will be 3C%. After the cooldown rate has been reduced as per stem, IOP-6 has operators raise charging flow to establish 80% cold cal level. The distracters either have the incorrect level or maintain stable charging flow. Salem experienced a PZR drain-down event in 2008 due . . . . Referlenc:e Section IIOP006E009 I PZRP&LE008

.. ROSkiSCrapercl'SRO SkYScraper RO:svstefnlEvolutlonUSt J;JSROSystemlEVOlutionUst J' 'Outline Changes i SRO_21__________. Given the following Unit 2 is perfonning a reactor startup by control rods lAW S2.0P-IO.ZZ-0003, Hot Standby to Minimum Estimated Critical Conditions -Cb= 500 , -Control Bank D = 121 , -Xe -12,000 When the ICRR value reaches 0.125, the Predicted Critical Rod Height is 67 What action{s), if any, are required to be taken in response to this Predicted Critical Rod References Continue the reactor startup, and evaluate the post startup data for Initiate rapid boration, insert Control Rod Banks, and recalculate the ECC. Insert the Control Rod Banks and recalculate the ECC prior to withdrawing Control Rods. 20[lroV'21 prior to continuing with the startup.

1_______--' fRo. ".alue:'I[ROcjrouP:.1 J 11 'System/Evolution .. __ __...

... , ...

_________________________________________________________________rReViSiOnl 1--------___________________

ii_________________________.,_________

'c';", .,..:::::-n I Material Required for Examination ISRO 21 S2.RE-RA.ZZ-0016 Rev. 4 Salem Unit 2 Cycle 18 COLR Rev. 2 S2.0P-IO.ZZ-0003 Rev. 31

--." .. " : Question Source' 'I Facility Exam Bank I MethOd:.J Significantly Modified I During Training Program I Source book rev and format Replaced one distracter.

,.. iComment I I I (Quest/onToPlc:

SR022 *. ... --... IIWhiCh of the followin9 describes the bases for maintaining an OPERABLE Auxilia!y Feedwater in Modes 1-3 lAW Tech Seec 3.7.1.2? I I Ensures the capability to cooldown and maintain the RCS at <500 OF for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a SGTR assuming failed fuel. I I Remove decay heat and maintain the RCS at HSB conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 'UIIUVVII'l:j a complete loss of off-site power. I IEnsures that the RCS can be cooled down to <350 deg F from normal conditions following a complete loss of off-site power. I IProvide the RCS heat removal capability necessary to prevent a challenge to the pressurizer safety valves during a full power ATWT. I I I c I Level *1 S I ;Cognitive Level J I Memory I

!Salem 1 & 2

I 5/17/20101

....

11 ...----------------------------------------;-!__________________________________________________________________-J System/Evolution Title:: _____.

I Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. I 155.43(2)

All of the distracters are incorrect because the Tech Spec Bases document does not contain those reasons. Additionally:

B II Answers: .! is plausible because the operability of the AFWST is for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> maintaining HSB with steam discharge to the atmosphere.

A is *..-...

plausible because the RCS cooldown during a SGTR to <500 degrees is to limit the potential off-site dose. C is correct per D is plausible because the ATWT bases document for AFW flow is for decay heat removal. I , .

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.i.*** "-**' -.... ! Title I

..Facility Reference';lurrlber-*

Reference Section ! Page No. IRevlsion'

I Salem Tech Specs Bases I J 13.7.1.2 !iB3/4 7-2 11258 I I I i I I I I I I I I I I L--__ ________* AFWOOOE011 II

..*u_r.ce ....c I_Fa_Cl_'I_itY_Exam Bank I Modified I

.... ...... Questlon-So-urceCommentsl -I

. t LiSt l'OlitlineChangesJ ISR_O_2_3__________________

• ______________________________________________________Mr. Doe is a Hope Creek Station employee and has received 1900 mrem routine TEDE for the current calendar year. ALL dose recorded at Creek Station. He is expected to receive an additional dose of 450 mrem on his current job assignment AT His lifetime exposure is 5500 mrem. lAW RP-AA-203, Exposure Control and Authorization, and prior to performing the job, written approval for increasing his dose limit to 3000 TEDE for the calendar year must be received from the .._____________________

--' Station Manager and Site Vice President.

______

System/Evolutlon Title

:

.*__

ffii!owledge of radiation exposure limits under normal or emer\ilency conditions.

I I",';'.n.llon of 155.43(4)

D;, ",, boo",,, 'he ,p,,,,,;3, "",',""" "e' Up" 3,000 m'6ffl-RP M,,,,.,, 'p In 4,000 m'.m* RP M,""., ,"d I I An!w,,""_ --S','oo Moo,,,",, >4,000 Sit. "'" P',,"ool

.""n ",me fo"" of """b, 'f MoO "f the p",""", I RADCONE002 rused

__..__

Q65481, changed a distracter from Station Manager ONLY to Site Vice President only to make more plausible.

I I .*.

.\,

LiSt .SROSvSteinlEvolutlOn LiSt "' Outline Changes I SRO 24 An explosion and fire at the RAP tank area has resulted in a possible large spill of radioactive water in the area. An Alert has been declared and all required facilities are activated and staffed. The Fire Department has determined that off-site assistance from the local fire department is needed. lAW S2.0P-AB.FIRE-0001, Control Room Fire Response, which choice identifies who must authorize requesting off-site fire department assistance?

Security Duty Supervisor.

I Nuclear Fire Protection Supervisor.

I Radiological Assessment Coordinator (RAC). Shift Manager / Emergency Duty Officer (EDO). I !Explanation of*I' 55.43(5) CAS ATT. 1, Fire Dept. Support, Caution prior to step 3.0 in AB.FIRE-1 states, "In the event of a radiological emergency, :Answers:

the Nuclear Fire Protection Supervisor should obtain permission from the EDO/SM prior to calling for off-site assistance." A is plausible because security is required to be notified whenever off-site assistance is requested (CAS 2.0) C is plausible because they will be leading the fire brigade and will be the person to request the off-site assistance through the EDO. 0 is plausible because . . . *........

LO.Number

__,_.___,_---.1 IFIRPROE012

.., J sR()

IOl.rtline Changes).

1 I SRO 25 !'With Salem Unit 2 operating at 100% power, which of the following conditions will require operator performance of the designated Immediate Action i steps lAW the associated procedure?

',An auto Safety Injection signal does NOT trip the Rx. Initiate SI lAW 2-EOP-FRSM-1, Response to Nuclear Power Generation. i IA PZR Spray valve fails open. Place the affected spray valve in manual S2.0P-AB.PZR-0001, Pressurizer Pressure J 2A Vital Instrument Bus becomes deenergized.

Place Rod Control in manual S2.0P-AB.115-0001, Loss of 2A 115V Vital Instrument 1 id.llThe Electric System Operator calls the control room and reports a valid SMD of K-6 and ESOC EXCESS MVAR ALARM. Depress SMD #1 RUNBACK lAW ! I S2.0P-AB.GRID-0001, Abnormal Grid. II i :Ainswe r . E.J iExam I PI'I!!nitive 1 altAI !Memory !

Salem 1 & 2 5/17/2010lJ ii<iCStatement:

L. __* __

I controls Reference Section 1__________________________________.__________________--__--__

__

__iAB1151E002 I J I Required for Exa'!'lnati<:m

• I II . I New IIO...."tinn Modification Method: !iUsed During Training Program 0 IQuestion Source Comments' I references TRIP*1 Sh 1 and FRSM-1 Sh 1

_____..-J I I !Comment L __

I I I I I____________________________________