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Final Written Examination with Answer Key (401-5 Format) (Folder 3)
	ML16050A331
Person / Time
Site:	Salem
Issue date:	12/04/2015
From:	; Public Service Enterprise Group
To:	Todd Fish; Operations Branch I
Shared Package
ML15135A303	List: ML15159B204; ML15342A448; ML16050A330; ML16050A331; ML16081A285; ML16081A304; ML16081A334; ML16081A353; ML16081A364; ML16081A397; ML16081A406; ML16145A208;
References
	TAC U01910
	Download: ML16050A331 (120)
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{{#Wiki_filter:U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: Region: I Date: 12/21/2015 Facility: Salem 1 & 2 - License Level: RO Reactor Type: W Start Time: Finish Time: Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected SIX hours after the examination starts. Applicant Certification All work done on this examination is my own. I have neither given nor received aid. Applicant's Signature

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Results Examination Value Points Applicant's Score Points Applicant's Grade Percent

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I fouestion Topic I IRO 1 I Unit 2 is operating at 60% power with rod control in auto when Turbine Steamline Inlet Pressure transmitter 2-PT-505 fails high over 10 seconds. Which of the following describes the effect on the rod control system from this failure? Control rods will move in the ___ direction, and the Power Mismatch signal will initially cause rods to move_ _ rapidly than if the circuit was not part of rod control.

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   ~    Iinward, more.

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   ~     outward, less.

I []I outward, more. I !Answer 11 d I !Exam Leve'i] I R I cognitive Level i IApplication I !Facility: 11 Salem 1 & 2 I IExam Date: 11 12/21/20151 IKA:ll 000001K105 j IAK1.05 I !Ro Vcillle:JI 3.51 lsRO Vallie: i 3.81!Section:11~ [ROGroup:!j 21 !sRo Group:JI 21 lil~ D !System/Evolution Title] IContinuous Rod Withdrawal 11001 IKA Statement: I~e of the operational implications of the following concepts as they apply to Continuous Rod Withdrawal: I turbine-reactor power mismatch on rod control I !Explanation of 55.41.b(6) The power mismatch section of the Reactor Control Unit compares the difference in the rate of change of turbine power [Arisl/Vers: *.* and rx power, and adds a signal to the temperature error circuit. With turbine power (PT-505) rising rapidly over 10 seconds and rx power not changing at all, the power mismatch signal will cause control rods to move out at a higher rate than what would be expeected from the Terr signal alone, in an attempt to raise Rx power at a rate similar to what turbine power was doing. I Reference Title ii Facility Reference Number IIReferef)ce Section i IPage No. I rRevision! !Rod Control Lesson Plan 11 NOS05RODSOO I 1j2a 1112 I I II I II ,II, I I ii I 11 I fL.O. Number i Objective I RODSOOE006 1_ _____, !Material Required for EXarf\fr1ation ' 11 II IQueition source: i l_N_e_w_ _ _ _ _ _~I Llo~u~e~s_ti_o_n_M_o_d_if_ic_a_t_io_n~M_et_h_o_d~=~*.Jil1, __________.I !used Dui'ing Training Program ID loue~tic:msollrce Comments: I I

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RO Skyscraper SRO Skyscraper I .-..R..o...s..y...st_e_m_IE...vo...1..ut..io...n_L_is_t- _s_R_o_s_y_s... te..m..1E_v_o... 1u... tio..n..L...is_t_ Outline Changes I I jOuestion TOpic j RO 2 I IWhich of the followin~ describes the reason for the sequence of dia~nostic steps in EOP-TRIP-1 Reactor Trip or Safety Injection? I I

    ~ SGTR diagnosis has the higher priority because of the highest probability of radioactive release.

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    ~I SGTR diagnosis has the higher priority to minimize the potential for component failure due to water in the Main Steam lines.

I I ic.: Main Steam Line break diagnosis has the higher priority because a high energy steam break could potentially mask other failures. I l

    ~] Main ~team Line break diagnosis has the higher priority because AFW must be isolated to remain within accident analysis assumptions for containment pressure.                                                                                                                                                                  I

!Answer: j c I !Exam Level ! IR I :cognitive Level I 1 Memory I iFacHity: ! ISalem 1 & 2 I jExamDate: i l

                                                                                                                                                                    .      .. I 12/21/2015!

[KA:[! 000007K301 \ IEK3.01 1IRova1ue:!j 4.0!iSROValue:1! 4.6IJSection:ll~iROGroup:I 1!:sROGroup:I 1! -~~ 0 jsystem/EvolutiollTitl~J 1_ R _ e _ a _ c _ t o _ r _ T _ r i ' - p - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' ~[o_o7_ _ IKA Statement: j Knowledge of the reasons for the followinq responses as they apply to Reactor Trip: I Actions contained in EOP for reactor trip l !Explanation of' 55.41.b( 10) SG pressure is checked prior to tube rupture criteria in TRI P-1 to see if any SG is faulted, as that can mask other Answers:

accidents.

Reference Title** p . Facility Reference Number I!Reference Section If Page No. / !Revision/

[* IRx Trip or Safety Injection ii EOP-TRIP-1 I II ! j2s I ITRIP-1 Lesson Plan 11 NOS05TRP001 I I159 l I7 I I II I 11 ii ! [LO.Number I TRP001 E017 Objectives I ,__ ___.

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Outline Changes I IQllestioriTopic 11 RO 3 I I Which of the following is the basis for establishing I maintaining SG Narrow Range level between 9-33% (non-adverse containment) for small or intermediate LOCA's IAW 2-EOP-LOCA-1, Loss of Reactor Coolant? I I

      ~ Ensures SG available if a RCP has to be started later in the event.

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      ~ Maintains a static head of water to reduce any existing SG tube leakage.

I 81 Ensures adequate feed flow or SG inventory to ensure a secondary heat sink. I

      @J  IMaintain the water level above the top of the U-tubes to prevent depressurizing SG.

I iAnswer 11 c I [Exam Level 11 R I iCognitivelevel l I Memory I !Facility: 11 Salem 1 & 2 I [ExaITIDate: .11 12/21/20151 [KA:ll 0000091<203 IiEK2.03 *!Ro Value: II 3.0! )SRO Value: II 3.3! jsection: ii~ iRo Grollp:jj 11 lsRO Grour>:ll 1I D

systernlEvolution Title I !_s_m_a_ll_B_re_a_k_L_o_c_A_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.I f009

!KA Statement: i Knowledqe of the interrelations between Small Break LOCA and the followinq: I S/Gs I

;Explanation of! 55.41.b(10) Distractor B is incorrect because it is the basis for S/G level I feed flow for a LARGE break LOCA. Distractors A and D

,I.An>swers:. ... *.*** are incorrect because the Basis document states that the reason for maintaining adequate feed flow I SIG level is to ensure a secondary heat sink. C is correct because EOP-LOCA-1 Basis Document states the purpose of establishing 9% level is .. "To ensure adequate feed flow or S/G inventory to ensure a secondary heat sink for small or intermediate size LOCAs and secondary break

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FaciHty Reference Number* I:Reference Section 11 Page No. j !Revision' j Loss of Reactor Coolant 112-EOP-LOCA-1 II Bases Doc 1112 1128 I I II n ii 11 I I ii II II 11 I fL.0.Number I LOCA01 E009 Objectives

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Outline Changes I louestio~ Jopic 11 RO 4 Given the following conditions:

  -  Unit 2 is in Mode 3 following a shutdown after a 200 day run.
  -  Core burnup is 18,000 MWD/MTU.
  -  All RCP's are in operation.
  -  Main steam dumps are in MS PRESSURE CONTROL-AUTO@ 1005 psig.
  -  A transformer fault results in the total loss of off-site power.

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

  - All RCS WR Thot's are 559°F and rising slowly.
  -  All RCS WR Tcold's are 547°F and stable.
  -  All SG pressures are 1015 psig and stable.
  -  All SG NR levels are 39% and stable.
  -  PZR level is 23% and rising slowly.

Which of the following identifies:

1) The status of natural circulation

2) The action(s) that will be performed by the control room IAW S2.0P-AB.RC-0004 Natural Circulation based on that determination of natural circulation status

    ~      1) Natural circulation NOT established                                                                                                                       I

2) Raise the steam dumping rate. I

    ~      1) Natural circulation NOT established

2) Raise the feeding rate to all steam generators.

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1) Natural circulation IS established

2) Start AFW pumps to maintain steam generator narrow range level in the normal band.

    @]     1) Natural circulation IS established

2) Feed steam generators to maintain steam generator narrow range level in the normal band and adjust MS-1 O's to enhance heat removal.

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I [Ansv.:er J ~ I.Exam Level.! [CJ Cognitive Level . I IApplication I iFacHity: I j Salem 1 & 2 I JExamDf!te: J l___1_2_12_1_12_0_1__,5j ~loooo15K101 llAK1.01 1!ROValue:JG]"ISROValue:]~:section:ll~fROGroup:lf3JsROGroup:ILJ D ~ystem!Evolution Title I IReactor Coolant Pump Malfunctions !KA Statement: I ! Knowledqe of the operational implications of the followinq concepts as they apply to Reactor Coolant Pump Malfunctions: I Natural circulation in a nuclear reactor power plant I

                    !j IEJ(planation of 55.41.b(5) The question meets the KIA based on the Loss of RC Flow portion of the KIA. AB.RC-0004 step 3.6 identifies ALL the Answers:

I 11 conditions that must be met for natural circulation to be occurring. With RCS Thats still rising, it is NOT occurring. Step 3.7 directs the operator to feed the SGs to maintain them within+/- 5% of programmed band. Programmed band plus 5% = 38%, so feeding in this situation is not directed. Steam dumping is directed to maintain or lower CET temps. l Reference Title H Facility Reference N_umber .*** * [ [Referel)C~ Section 11 Page No.[ f~evision'. INatural Circulation 11 S2.0P-AB.RC-0004 I !11 118 I I !I I II 11 I I II I 11 ii I jLO. Number > Objectives ! ABRC04E001

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                                           ! RO 4 Steam Tables lflo~s,f~n ~oti[:c;:~"co[hi:i1~n!;>! 180355, removed some window dressing in stem. Modified format to two part answer I

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[Question TopiC J I RO 5 Given the following conditions:

   -  Unit 2 is operating at 100% power.
   -  21 charging pump is in service.
   -  22 charging pump is CIT.
   -  Subsequently, 21 charging pump trips.
   -  23 charging pump is placed in service after its suction is aligned to the RWST.

Which of the following identifies a subsequent action which will be directed by S2.0P-AB.CVC-0001, Loss of Charging, and why?

     ~I Shutdown Unit 2 due to boration of the RCS from the RWST.

I jb.: Restore normal letdown to establish adequate VCT level for normal 23 charging pump operation. l=J IPlace Excess Letdown in service to establish adequate VCT level for normal 23 charging pump operation.

     @]I Open 2CV464 Charging Cross-Tie Isolation valve to prevent requiring a Unit 2 shutdown due to boration of the RCS from the RWST.

IAnswer j a l IExam Level R I !cognitive Level IMemory J 11 J I !Facility: i ISalem 1 & 2 I jExainDat 12/21/2015

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~I 000022K302 I:AK3.02 i"1RO Value:!~ ;SRO ' Value: I~jsect1on: *I:I~ IRO Group:iLJjSROGroup:jLJ ~\.:~ D 1 I jsystem/Evolution Title J Loss of Reactor Coolant Makeup j 1022 I J<A Statement: I Knowledqe of the reasons for the following responses as they apply to Loss of Reactor Coolant Makeup: Actions contained in SOPs and EOPs for RCPs, loss of makeuo, loss of charqino, and abnormal charoinq i.Explanatioriof I 55.41.b(10) When the inservice charging pump tripped, letdown automatically isolated with all charging pump breakers open. If 23 I !Answers: I charging pump is placed in service from the RWST, it means it was not able to be placed in service with VCT as suction source at step 3.7. The note prior to placing 23 charging pump in service from Unit 2 RWST states that a unit S/D will be required because of borating the RCS from the RWST.

1. Reference Title /i ***Facility Reference Number

i !Reference Section dfPage No'. j IReyi~ion:

ILoss of Charging !IS2.0P-AB.CVC-0001 II Bases Doc 11 3 1I9 I I II II II II I I II II II II I jL.o. Number ..* Objectives I ABCVC1 E003 ,_ _ ___. IMateflal Required for E~afoinaUon ,;j I 11

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 !Question Topic J 1RO 6                                                                                                                                                                    I Given the following conditions:
  - Unit 2 is in MODE 4.
  -  21 RHR loop is in service providing SDC.
  -  22 RHR loop is aligned for ECCS.
  - Subsequently, 21 RHR pump trips on overcurrent.

Which of the following describes how SDC flow will be established IAW S2.0P-AB.RHR-0001, Loss of RHR? 22 RHR pump supplying flow .....

    ~I through 21 RHR HX and ONLY 21SJ49 RHR Disch to Cold Legs.

I IE] Ithrough 22 RHR HX and ONLY 22SJ49 RHR Disch to Cold Legs. I

    ~I through 21 RHR HX and BOTH 21SJ49 and 22SJ49 RHR Disch to Cold Legs.

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    ~I through 22 RHR HX and BOTH 21SJ49 and 22SJ49 RHR Disch to Cold Legs.

I !Answer 11 d I IExani Level 11 R I !cognitive Level 11 Memory I !Facility: ! ISalem 1 & 2 I JExamDate: 1 I 12/21/20151 iKA:ll 000025K201 I AK2.01 1IROValue:jj 2.9jlsROValue:ll 2.9ilsection:ll~fROGroup::j 11lsROGroup:!I 11 !Sl~I D [System/Evolution Title i j Loss of Residual Heat Removal System ! KA Statement: I Knowledqe of the interrelations between Loss of Residual Heat Removal Svstem and the followinq: RHR heat exchanqers I jExpfanation of! 55.41.b{10,8) With a single loop of RHR in service for SDC, the loop cross-tie valves 21/22RH19 are open. They would only have .Answers: I been shut if BOTH loops were in SDC mode, to split out the RHR loops to verify minimum flow requirements are being maintained (S2.0P-SO.RHR-0001, page 36). There is no direction to close either cross tie valve in Attachment 2 of AB.RHR when swapping the ECCS loop to SDC mode. There is also no direction to shut either SJ49 RHR to RCS valves, since cooling flow to all 4 RCS

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IiReference Section .* j IPage No. !Revision! j ILoss of RHR !I S2.0P-AB.RHR-0001 I ii 1118 I IInitiating RHR !I S2.0P-SO.RHR-0001 I 11 ! l2s I IRHR Simplified drawing ii 205332-SIMP I 11 112 I fL.(). Number**..

IABRHR1E004 ,____. iMaterialRequlred for EJ(Cll:Jlination **** 1 I louesti.oh_Soufoei I l_N_e_w_ _ _ _ _ __..l louestionModification lll!eth?d: *. J I \Osed During Traini11g Pr~gram"! D fqu~sti.on Source Comments! I I

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RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List I .Outline Changes touestion Topic I IRO 7 I Given the following conditions:

  -  Unit 2 is operating normally at 100% power steady state.
  -  Console alarm RC PRESS DEVIATION HI, annunciates.
  - Subsequently, the board operator then observes the following indications:
     - PZR pressure master controller output dropping slowly
     - Both PZR spray valves closed
     - All PZR heaters energized
     - PZR pressure channel I (selected for control) dropping slowly
     - PZR pressure channels II, Ill, & IV rising slowly Which of the following procedures should be used for these conditions?

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    ~ S2.0P-AB.STM-0001, Excessive Steam Flow.

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    ~ S2.0P-AB.ROD-0003, Continuous Rod Motion.

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    ~ S2.0P-AB.RC-0001, Reactor Coolant System Leak.

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    @]I  S2.0P-AB.PZR-0001, Pressurizer Pressure Malfunction.

I !.AA.svver 11~ !Exam Level 11~ :cognitive Level I Application I I [F(ldlity: !]Salem 1 & 2 I IExam Date: fl ___1_2_12_1_12_0_15_.I ~I 000027G107 112.1. 7 i :Ro Value:j G] lsRo Value:j~ !Section: !I~ !Ro Group:jLJ iSF{O Group:l LJ !;~! D !System/Evolution Title i IPressurizer Pressure Control Malfunction I W~ IKA Statement: I Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. I fExptapation pf,I 55.41.b(5, 10) The three distracters would cause all the bulleted items except the last one to occur. The 3 unaffected channels rising Answers::. * :> would discount the 3 items (steam leak, RCS leak, or rod insertion) that could be causing the event. I Reference Title I! Facility Reference Number j jReference Section j I.Page NC?'.: jRe\(ision! j Pressurizer Pressure Malfunction 11 S2.0P-AB.PZR-0001 II 11 1118 I ! II II 11 11 I I II II 1 I I LO. Number IABPZR1 E001 Objectives " [Material Required for Examination iI I IQuestion Source:*, ,j Facility Exam Bank i 1ftluestion Modificatio,n !\l[~thod:. ; Direct From Source l lusedDOri~g.iraiO.in9 Program j D louestiorisource.somirientsl 1.--42_6_8_2_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _...,.1 1 !comment I I I I I '

RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I lauestion Topic 11 RO s I Which of the following describes the effect if a PZR level instrument reference leg were to develop a leak with the Master Flow Controller in auto? Indicated PZR level would initially ..... [al Irise, then actual PZR level would rise. I

    ~I rise, then actual PZR level would lower.
    !b.:

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    ~       lower, then actual PZR level would rise.

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    !ct. j Ilower, then actual PZR level would lower.

I !Ansv./er ! b I I !Exam Level i j R I !cognitive Level 11 Application j jFacility: 11 Salem 1 & 2 1 IExamDate: J I 12/21/20151 IKA:il 000028K101 1IAK1.01 'iRO Value: :j2.s*1 isRO Value: !I 3.1 Ifsection:!j~ :Ro Group:jj 1 2j SRO Group:lj 21 ~ltJ D !system/Evolution Title I IPressurizer Level Control Malfunction :028 IKA Statement: i Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunction: PZR reference leak abnormalities I !Explanation of I 55.41.b(7) As level in the reference leg lowers, the pressure of the actual wight of water in the PZR would appear to be greater than Answers: it was, as it is pushing against a lower reference pressure from the reference leg. This would result in indicated PZR level rising. The rising level would cause an automatic response of the Master Flow Controller to lower charging flow, which would cause actual PZR level to lower. I> Reference Title I l~ Facility Reference Number **. i jRefererice Section *I IPage No. f IRevision; IGeneral Physics Lesson Plan Components I 1 PC071 r4 Sensors I 11 11 4 I I II I ii 11 I i II I 11 11 I IL.o. Nllrnber IPZRP&LE015 1_ ___, !Material Requfr~dfo~ Examination**. i I ia':.i~stion Source;. 1 IFacility Exam Bank I ldllestion Modific:ation Method; .**.. ,j Concept Used I fused During Training Program I D ~9~e,stior1Source Cornmentsf 159797

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!Question Topic      l I RO 9                                                                                                                                   I Which of the following describes the difference in effect of a rapid boration performed during an ATWS at BOL or EOL?

At 1) , the rapid boration will insert more reactivity due to the 2) differential boron worth . I

  ~I      1) EOL
   . *. . 2) HIGHER I

[gJ 11) EOL

2) LOWER I le.! 11) BOL

  ,_ 2) HIGHER I

[ ) 11) BOL

2) LOWER I

/Answer 11 a 1 IExalTI Levell IR I !cognitive Level 11 Memory I !Facility: I j Salem 1 & 2 I [Exam[)a~e: 11 12/21/20151 IKA:il 000029K103 j IAK1.03 liROValue:1j 3.ellsRova1ue::l 3.sl!sectic)n:il~iROGroup:ll 11isR6Group:ll 11 t~~J! D lsysterri/EvolutiOn Title I j Anticipated Transient Without Scram 11029 iKA Statement: i ! Know!edqe of the operational implications of the follow!nq concepts as they apply to Anticipated Transient Without Scram: Effects of boron on reactivity IExpl~nat,i?p ofj 55.41.b(5) Differential boron worth rises over core life, and will insert more reactivity during an ATWT at EOL. .Answers: * .*

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Facility geterepce Number I!Reteren~eseCtion * ** * *I ! Page ~o. I IRevisionj ICurvebook 11 S2.RE-RA.ZZ-0016 I 11 1I7 I I I I II 11 I I I I ii 11 I

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 !Question Topic 11 RO 1o Given the following conditions:
  -  Unit 2 experiences a Rx trip from 100% power when a Main Generator trip causes a Main Turbine trip.
  - Upon transition to TRIP-2, Reactor Trip Response, the PO reports OHA A-6 RMS HI RAD has annunciated, and Radiation Monitor 2R19C - 23SG Slowdown, is in alarm with rising counts.

Which of the following describes how the crew should proceed?

    ~I Initiate SI and go to TRIP-1, Reactor Trip or Safety Injection.

ibl IEnter S2.0P-AB.SG-0001, Steam Generator Tube Leak, while continuing in TRIP-2.

    ~I Do NOT initiate any procedures based solely on a 2R19 alarm immediately following a Rx trip.

I idl IEnter S2.0P-AB.RAD-0001, Abnormal Radiation, to verify validity of alarm while continuing in TRIP-2. I !Answer 11 d I /Exam Level 11 R I !Cognitive Level 11 Application I !Facility: 11 Salem 1 & 2 I [ExamDate:'l I 12/21/20151 iKA: 11 000037A201 l1AA2.01 I IRO Value: JI 3.olfsRo Value: II 3.41 tsection:.11~ !Ro Grciup:Jj 21 iSRO Group:: I 2j l!J D ISystem/Evolutiori Title ! ISteam Generator Tube Leak f j037 [KA Statement:! Ability to determine and interpret the following as they apply to Steam Generator Tube Leak: Unusual readinqs of the monitors; steos needed to verify readinqs !Explanation of I 55.41.b(10) AB.SG says in note under entry conditions that R19s are not accurate immediately following a unit trip and should not !Answers: ***.*. * .1 be used as the sole basis for entering it. However, AB.RAD should be entered and the alarm verified as directed at step 3.2. The SI is not warranted without corroborating indications

l Reference Title It Facility Reference Number

IIReference Seetion ** *Ii Page r-z,o. I [Revision 1! .. *. . , .

ISteam Generator Tube Leak II S2.0P-AB.SG-0001 II 112 1129 I IAbnormal Radiation 1IS2.0P-AB.RAD-0001 II 111 1130 I I II II 11 II I !Lo. Number<** I ABSG01 E003 Objectives 1_ ____, !.Material Requfred for Examin(ition * *! I iQUestion .s9urc~: t l_N_e_w_ _ _ _ _ _ __.I ;auestion Mo,dificatic:in Method: *** __________,I tused During Tr~iriih9 Prografrl I D 1aue5:tion Source Cofrlrr,~nts) I I

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RO Skyscraper i SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I jQuestiori Topic j Ro 11 I Given the following conditions:

   - Unit 2 is performing a controlled shutdown and cooldown from 100% power due to a 5 gpm tube leak on 22 SG IAW 82.0P-AB.SG-0001, Steam Generator Tube Leak.
   -  After completing the Immediate Actions of EOP-TRIP-1, Reactor Trip or Safety Injection, following the Rx trip from 20% power, the RO reports that control rod 202 is stuck in the fully withdrawn position.

Which of the followinq identifies the action, if any, the crew will perform in response to the stuck rod? I la. i Initiate a rapid boration for 35 minutes during performance of EOP-TRIP-2.

     ~I Initiate a rapid boration for 35 minutes in 82.0P-AB.SG-0001 after exiting the TRIP series procedures.

[] I No actions are required for a single stuck rod because SOM for the cooldown to 503 degrees is adequate. [] I No actions are required for a single stuck rod until the Auto SI Block is performed during RCS depressuri~ation to 1900 psig. [Answer 11 b I !Exam Level 11 R I [cognitive Level 11 Application I 1 !Fa!=ility:.1 Salem 1 & 2 I IExamDate: 11 12/21/20151 ~I000038G416 lfROValue:l@[sROValue:j[Dfsection:Jl~:goGroup:JLJfSROGroup:ILJ f:in

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I 112.4.16 D fsy§tem/Evoluti<?nTitle I ISteam Generator Tube Rupture 11038 I ll<A statement' I Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. IExplarya~ion of I 55.41.b(10) Question meets KIA because knowledge is required of the relationship between AB performance in the initial response [Answers: * .* I to the tube rupture and EOP performance which addresses stuck rods. Boration for a single stuck rod is not performed in the EOP series, but IS performed in the AB for tube rupture to meet SOM requirements for the initial cooldown to 500°F. Step 3.26.H in AB.SG states to trip the turbine, then trip the Rx at 20% power during the shutdown. The 5 gpm size of the leak will allow for a

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cause a transition to SGTR-1. AB.SG would be re-entered at step 3.27 following exit of TRIP-2, and 3.28 directs rapid boration for

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each stuck rod for 35 minutes. The rapid boration will be initiated before any depressurization starts in 3.29, so the distracter reqardinq depressurization is incorrect. re \

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                                     ** Reference Title .                                                             ll         Facility Refere.nce. N.umber

i!Reference Sec~ion** . *l I Page* No. I tRevision:

J Steam Generator Tube Leak II S2.0P-AB.SG-0001 l 11 1129 I j Reactor Trip Response 112-EOP-TRIP-2 ! ii 1I I I II I ii 11 I fL.o:Nutnber * *** Objectives SG01E005

J94~~tion~p~S~~/ 11Previous2 NRC Exams I tq\'~;;,H~H.~P~!!l~'\?~~it,§.~~=.fil Direct From Source io~~~~Iq]l.~p~f;f~.fR~nientsf 1_1_1_-0_1_(_1_2_12_0_1_2_e_x_am_>_R_o_a_9_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~

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Outline Changes I iauestionTopic l IRO 12 I How does the INITIAL Tavg response for a large steam line break (SLB) compare to the Tavg response for a large feed line break (FLB)? (Initial Tavg response is before automatic protective actions occur)

    ~    I
    ~I Rises for a SLB and lowers for a FLB.

Lowers for a SLB and rises for a FLB. I I I

    ~ Rises for both a SLB and a FLB, rises more for a SLB.

I I

    ~ Lowers for both a SLB and a FLB, lowers more for a FLB.

I

Answer J I b I !Exam .Level ! IR I !cognitive Level 11 Application I !Facility: 11 Salem 1 & 2 1 IEX:amoate: ii 12/21/20151

~I 000054K101 I AK1 .01 I!RO Value: i [DJ jSRO Value: l@tsection: iI~ [RO Group:i LJ [SRO Group:! LJ I,~ D r~,,,,._,,,_,T"~ ls~~~w~~~~j_L_o_~_cl_M_~_n_F_e_e_~_a_t_e_r_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _I!~ fKA Statement:! ! Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater: I MFW line break depressurizes the S/G (similar to a steam line break) [Explanation of j 55.41.b(5) A steam line break will cause MORE steam to be drawn from the SG, causing Tc to go down, causing Tavg to go down.

Answers: * . *...* J A Feedline break will send LESS cold water into the SG so it's Tc will rise, and Tavg will rise. KIA match because the operational implication of the SG depressurization (similar to a steam line break) is that an operator would have to use diverse and alternate indication to discern what is acrually happening to the SG.

I Reference Title li Facility Reference Number .. ![Reference Section * ,j t Page No. ! fRevision! I I I 11 11 I I I I 11 iI I I I I 11 11 I iLb.Number Objecti ICN&FDWE016 ,__ __, IMaterial Required for Ex:aminatiorl >i I )Question Sol.lrce:: 11 Facility Exam Bank I f OuestiOhModifidtion Method: I Direct From Source )~~~nS~~efo~~~11~8-0_7_8_9 _____________________________________ ~, [comment *.* *'. I I I i I

RO Skyscraper SRO Skyscraper. I RO System/Evolution List I SRO System/Evolution List Outline Changes lauestionTopicj I RO 13 Given the following conditions:

  -   Salem Unit 1 is operating at 100% power.
  -   Salem Unit 2 is in Mode 4 with 21 loop of RHR in SOC mode.
  -   Subsequently, a loss of all off site power occurs.
      -   ALL Unit 2 EOGs fail to start or trip shortly after starting, and ALL Unit 2 4KV vital buses are de-energized.
      -   The RO reports 21RH18 RHR HX FLOW CONT VALVE indicates 6% open and 2RH20 RHR HX BYP VALVE indicates 15% open, and questions their actual position.

Which of the following describes the operation of the 21 RH18 and 2RH20? I

    ~ BOTH the 21RH18 and 2RH20 fail AS IS on a loss of air and the indication is correct.

I

    ~ The 21RH18 fails open and the 2RH20 fails shut, and the board indication shows last known position before the loss of power.

I I

    ~ BOTH the 21RH18 and 2RH20 fail SHUT on a loss of air, and the board indication shows how far they shut before losing all air pressure for operation.                                                                                                                                                                                  I
    '.dl i 80Ild  tbQ 2:'.J b?bl:l 8 ~od 2l2h:l20 f:;Jil 012Et\I oo       lcso:s;;:: of ~ic :aod tbg bo:;ii:d iodic~tioc ~bcuus;:
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brnu f:;Jr: tbQu ooQoQd bQfocg losio,.. .... 11 .. I I

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L:.cJ for operation. I !Answer I ~ !Exam Level: ~ !Cog11itive L:evel. I IMemory ! !FaciHty: j' ISalem 1 & 2 I jExamDate: Il___12_12_1_12_0_1_,5I ~I 000055A201 1IEA2.01 ! !Ro Value: IfTI"isRO Value: I[IZJl~ection: 11~ [RO Group:j LJ iSRO Group:! LJ D [System/Evolution Title.I l_s_ta_t_io_n_B_la_c_k_o_ut_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___, ~-~ IKA Statement:! Ability to determine and interpret the followino as they apply to Station Blackout: Existinq valve positioninq on a loss of instrument air system !Explanation of I 55.41.b(?) The 21RH18 and the 2RH20 fail AS IS on a loss of control air, and both are supplied air exclusively from the "A" header. !Answers: . . . . ! Unit 2 ECAC supplies the "A" air header. With the Unit 2 ECAC also failing to start, control air pressure will bleed away fairly rapidly. The console indication will indicate actual valve position since 115VIB power should be available from the inverters for at least 2 hours following a LOPA. The valve positions given in the stem are what would be expected with a loop of RHR in SOC mode. I

        \"/"       i Reference Title                            Ir       Facility Reference Number                ******I iReference Section . ;1 f Page)lo~ I !Revision; ILoss of Control Air                                             11 S2.0P-AB.CA-0001                                     II                           1136      1I18                I I                                                                II                                                      n                            II        11                  I I                                                                Ii                                                      n                            II        11                  I Objectives I LOPAOOE010

,_ _ ___. [Material RE\q~fr~d tor exarTilrfaiion /I I !I la4e~tion ,source: ** 11 New 1lauestiqn .~.odification Meth~~.: .*.* ~--------~I !used Ouring Training Program I D lauestion Source,Coinl"l1,e~tsj I I

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RO System/Evolution List I SRO System/Evolution List I Outline Changes I .. louestion Topic i I RO 14 I Given the following conditions:

  -  The 2B Vital Instrument Bus (VIB) Uninterruptible Power Supply (UPS) Static Switch has been placed in the Bypass to Alternate position IAW 2B VITAL INSTRUMENT BUS UPS SYSTEM OPERATION.
  -  ALL power supplies to the 2B VIB UPS are available.
  -  Subsequently, the breaker from 2B 230VAC bus to the 2B VIB AC Line Regulator fails and opens.

What will be the effect on 2B VIB? The 2B VIB wilL . ..... - - . - --* I

                                                                                                                       ..
    ~ NOT deenergize because the Static Switch will automatically swap to the Preferred Source.

I

    ~I NOT deenergize because the inverter will automatically power the 2B VIB through the inverter auctioneering circuit from its DC source.

I [c. f I deenergize until manual operator action is taken to re-energize the 2B VIB by placing the Static Switch in Normal and placing the Test Transfer switch to N (Normal). I

    @:]I deenergize until manual operator action is taken to re-energize the 2B VIB by placing the Static Switch in Isolate (Alternate) and placing the Test Transfer Switch to N (Normal).                                                                                                                                                                                                                        I fAnswer I      !c          IExam Level ! R       !                    lcogl'litive            L~vel              !

I Agglication I IFacility: WSalem 1 & 2 l~ExamD?te: W 12/21/20151 i it{A: 000057A101 IIAA1.01 i iROValue: 1!3.7*1 :sRo Value: ii 3.71 [section: !I~ tRo Group:JI 11 ISRO Group:JI 1 I D

system/Evolution Title I j Loss of Vital AC Instrument Bus 1w-:

I . ' iKA Statement: I Ability to operate and I or monitor the followino as they apply to Loss of Vital AC Instrument Bus: I Manual inverter swappinq I 1£xplanation of ! 55.41.b(7) The VIB UPS static switch is transferred from norm to alt by placing the test transfer switch to ALT IAW Section 5.4 of Answers: I S0.115-0012. Then the Manual Bypass Switch is placed in Bypass to Alternate to physically position contacts B1, B2, and B4 (closed) and B3, B5 open. The VIB will deenrgize when power is lost to the AC line regulator, which is the Alternate source. Placing the static switch in Normal and test transfer switch to N (from alternate) is directed by S0-115-0012, Section 5.7.3 and

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could be powered from another source. A is incorrect because automatic transfer is unavailable with the staic switch not in the

                                                                                                                                                                                                                                                .I.... ... ,...

Normal posiiton. B is incorrect because the static switch is aligned to alternate, and while the DC power is supplying the inverter, the inverter output cannot flow throuoh static switch with B3 and B5 contacts open. I * .. Reference Title *II,; >Facility Reference Number *. J 1Reference Section 11 Page No. ) !Revision: I2B VITAL INSTRUMENT Bus UPS SYSTEM o II s2.0P-so.115-0012 ll 5.4 1119 11 6 I !2B VITAL INSTRUMENT BUS UPS SYSTEM o II s2.0P-so.115-0012 !IExhibit 1 (Static Switch I156 11 6 I I !I II II 11 I !Lo. Number

I 115VACE014

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I Outline Changes I

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[Question Topic 11RO15 Given the following conditions:

  -  Unit 2 tripped due to a transformer problem.
  -  2A and 2C Vital Busses are energized in SEC Mode 2.
  -  2B Vital Bus is de-energized because 2B Diesel Generator failed to start.
  -  OHA B-10, 2B 125VDC CNTRL BUS VOLT LO, just actuated.

In accordance with S2.0P-AR.ZZ-0002, Overhead Annunciators Window B, which of the following describes the required action? -- la. i I .. . Place the 2B2 Battery Charger in service.

I

    ~I Transfer the 2B1 Charger to its alternate power supply.

[c.l ITransfer 2B 125 VDC bus loads to their alternate source. [] I Ensure the 2B2 Battery Charger has automatically energized.

Answe_r ! I a I !Exam Level i IR I !Cognitive Level 11 Application I !Facility:J ISalem 1 & 2 I !ExamDate:*11 12/21/20151

~1000058A101 llROValue:jjilljsROValue:IQ]'[sec;tion:,fl~ [ROGroup:ILJiSROGroup:ILJ l[~;1i;I D

                                                                                                                                                       ,* ~ ,,. IE~::F llAA1.01 ISysterryEvolutionTitle            I  ILoss of DC Power                                                                                                            I ~lo_ss__

IKA State merit: I Ability to operate and I or monitor the followinq as they apply to Loss of DC Power: Cross-tie of the affected de bus with the alternate supply I [Explanation of I 55.41.b(S) The 2B1 battery charger is normally in service. The use of 2B2 Battery charger is limited to 7 days per Tech Specs and !Answers: is not normally in service. There is no auto swap. Transferring loads is only done if the backup battery charger cannot be placed in service. I < Reference Title * .. >it Facility Reference Number-,-*.*. IIReference Section *_ ... j I Page No. I 1~evisionl IOverhead Annunciators Window B !I S2.0P-AR.ZZ-0002 I 1121-22 1136 I I I I ii 11 I ! I I 11 11 I [L*.O.-Number:;i_"* I DCELECE005 I_ _____. IMater'ial ~equired for e.Xai:nfoation' 'I I ll I [ouesfioll;s#&rce:,\ Facility Exam Bank I 1lauestion Modific~tia,11 Met~~d:

Editorially Modified 1llJsed oll;!"illg Training program I D

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I

  !auestionTopic 11RO16                                                                                                                                                         I Which of the following identifies when the 21/22SW122, CC HX SW INLET VALVES automatically reposition, and why?

The 21/22SW122 valves ... I

    ~ open on ANY SI signal to ensure cool CCW is supplied to the RHR HXs and pump seals.                                                                                       I I
    ~I close on ANY SI signal to ensure SW pump runout does not occur with all CFCUs running.

I I

    ~ open on a SI signal coincident with a LOOP to ensure cool CCW is supplied to the RHR HXs and pump seals.

I l:!J Iclose on a SI signal coincident with a LOOP to ensure SW pump runout does not occur with all CFCUs running. I !"Answer Id i I JExarl1 Level 11 R I !cognitive Level I

                                                                           ! Memory                ! l_acilfty: / ISalem 1 & 2         I /Examoat~: 11          12/21/20151 J. *"":

tKA:JI 000062K302 2.91fSection:f!~IROG~oi.lp:lj ~11 D

                                                                                                                                                               },"!'1, liAK3.02           /iROValue:Jl 3.6ifSROValue:Jl                                                           1ljSROGroup:!I       11 Is ystem/Evo I uti on Title !
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j Loss of Nuclear Service Water I 1062 IKA Statement: I Knowiedge of the reasons for the following responses as thev apply to Loss of Nuclear Service Water: The automatic actions !aliqnments) within the nuclear service water resultinq from the actuation of the ESFAS !Explanation of 55.41(7,8) The SW122 valve are normally open and operate in conjunction with the CCW HX SW outlet valves SW127. During a SI .Answers:* coincident with a Loss of Offsite Power, the SW122 valves are automatically repositioned from open to shut. SEC Mode Ill closes SW122's due to SW Pump runout (and potential loss of all SW) concerns with all CFCU's l/S if only 2 SW pumps are operating, due to EOG/SEC failure. Also, the SW122's must stroke closed in less than 30 seconds to ensure the CFCU's are operational within l

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I Reference Title 11 Facility Reference Number ./I :Reference Section, Page No~ I !Revision:

                                                                                                                               * .. 11 IService Water System - Nuclear                               11 NOS05SWONUC                             II                          !134-35 ! I13          I I                                                             II                                         II                          II         11          I I                                                             !I                                         n                           II         11          I

!Lo. Number i Objectives I SWONUCE006

j_ ___. f Material. R.eqyired fo~ Examiriatio1,,1.:~;;j I [Question ~ource~ 11 Facility Exam Bank 1lauestion ModificationMetrod: .; ~Direct From Source 1lus~d Duri11g<Trainjrig Programj D

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  !Question Topic
  ',.  .       ,  *'  ,

I IRO 17 I I Both Salem Units are operating normally at 100% power when Unit 2 recieves OHA A-7 FIRE PROT FIRE. Which of the following is a correct response when assessinq the affected Fire Zone(s) on 2RP5, and why? I []I If ANY rows "Fire" light is illuminated ensure at least one Fire Pump is running to supply fire protection water by verifying OHA A-15 FIRE PUMP 1/2 RUN. I

      ~I If a fire is indicated in the Relay Room select Fire Outside Control Area in both control rooms to prevent smoke from entering control rooms.

I [c::

IIf ANY C02 I Halon Discharge lights are lit then ensure EDG control room ventilation has automatically stopped to prevent egress of gas to adjoining areas. I [] I. If fire indic~tion for BOTH zo.nes 59 and 74 are received, then open 2FP147 Fire Protection Containment Isolation to provide normally isolated fire protection water to containment. I !Answer i Id I !exam Level 11 R I !cognitive Level* 11 Memory 1 !~acility: i ISalem 1 & 2 I IExan:iOate: I I 12121120151 [l<A:JI oooo67K302 IIAK3.02 i!Rova1Ue::j 2.51[sRova1ue:q 3.3li8edion:ll~IRoGroup::I 21fsRpGrou!>:JI 21 all D '

system/Evqlution Title! !_P_la_n_t_F_ir_e_o_n_S_it_e_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _...... I067 i: KA Statement:! Knowledqe of the reasons for the followlng responses as they apply to Plant Fire on Site:

Steps called out in the site fire protection plan, FPS manual, and fire zone manual IExplanati<m()f j 55.41.b(10,4) A is incorrect because the FIRE light can illuminate if a manual fire pull box is activated, which only gives indication

                        *I tAnswers: . * ** *.*. and does not initiate fire protection water fiow. C is incorrect because Halon supplied to relay rooms would not indicate stopping EDG supply ventilation. B is incorrect because for a fire in Relay Room (physically located outside CR but using same AC system)

Fire Outside Control Area would not be selected. D is correct because the fire protection line to containment is normally isolated

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i Reference .Title> :[

                                                                             '     Facility ~eference Number                      I !Reference Section

I[Page No. I 1Revision 1 IControl Room Fire Response !I S2.0P-AB.FIRE-0001 I 112 11 9 I IOverhead Annunciators - Window A I

II S2.0P-AR.ZZ-0001 I 1123-24 l l56 I I II I 11 11 I [LO.Number

I FIRPROE008 Objectives I

!Material Requinid             for ExafTJin~tion

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\auesticmSoU,rce:'              j j_N_e_w_ _ _ _ _ _~I [aue$tionM9dification Metho~: **. --------~I !Used During ;r{airiing Program I D lQuesti()nSourc~Comrnents\ !---------------------------------------~'

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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [ouestioh Topic 11 RO 1's I Given the following conditions:

  -  Operators are performing actions in 2-EOP-FRCC-1, Response to Inadequate Core Cooling.
  -  Containment pressure is 2 psig and stable.

Which of the following describes the SG level and AFW flow requirements PRIOR to initiating the SG Depressurization to Inject ECCS Accumulators? SG NR level in at least one SG must be greater than ___

    ~ 19% OR total AFW flow must be >22E4 lbm/hr.

I

    ~ 115% OR total AFW flow must be >22E4 lbm/hr.

I

    ~ 19% AND total AFW flow available of >22E4 lbm/hr.

I [ ] 115% AND total AFW flow available of >22E4 lbm/hr. I JAnswerJ I a I iExamLe>/el J IR I !Cognitive Level ! IMemory I !Facility: 11 Salem 1 & 2 I !ExamDate: i I 12/21/20151 Group:ILJ !SRO Group:j LJ r ~I 000074K203 IiEK2.03 IiRO Value:!~ ' ,SRO Value: I~ lsect1on: 11~ :RO ' '.1

                                                                                                                                                                 !!II    D lsystel-ri{Evolution Title i j 1nadequate Core Cooling                                                                                                            11074      I

KA Statement: I Knowledqe of the interrelations between Inadequate Core Coolinq and the followinq: I AFW pump I fExplanat!on of I 55.41.b(10) Step 13 asks if any SG NR level is >9% (15% adverse). Since containment pressure is stated as 2 psig, it is below the

!Answers...*. 1 4 psig at which adverse numbers would be used. If SG NR level is >9% in at least one SG, the step for asking if >22E4 lbm/hr AFW flow is present is bypassed. If SG NR level is <9%, then it requires 22E4 lbm/hr. So you need at least 9% OR actual total AFW flow >22E4 lbm/hr. The availability is plausible as it is used in other Emerg procedures, but in this case a secondary heat sink

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I Reference Title !~ Facility Reference .Number . I!Reference Section J[ I Page No. [Revision' IResponse to Inadequate Core Cooling !12-EOP-FRCC-1 I ii 2 1122 I I II I II 1 l I I II I II II I iL(). N1,1mber Objectives I FRCCOOE005 j_ ___. (Material Required for Exam_ination 11 11 [Que~ti().I} S()~rc,~\ ] j_N_e_w_ _ _ _ _ _ __.l IQl.lestion Modification Method: , :jl I !used Dliring.TraiOingPr69r<!mJ D fOue~t!!'~,~~urce(:omments/ I I

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(auestion T?Pic ! IRO 19 Unit 2 is at 80% power when RMS channel 2R31, Letdown Line-Failed Fuel Process Rad Monitor starts a rising trend. When responding with S2.0P-AB.RC-0002, High Activity in Reactor Coolant System, how will operators differentiate between a crud burst and failed fuel as the cause of the rising 2R31 indication?

    ~I By monitoring 2R31. Fuel damage will cause the indication to rise at a higher rate than a crud burst.

I

    ~ By raising letdown flow rate to 120 gpm. The 2R31 readings will lower if crud burst caused the rising trend but will NOT lower if failed fuel caused the rising trend.

I

    ~ By requesting Radiation Protection to survey the letdown pipe area in the Auxiliary building. Radiation levels will be higher due to failed fuel than from a crud burst.                                                                                                                                        I
    @JI By requesting a Shift Chemistry Technician perform a radiological analysis of the RCS. A crud burst will show different concentrations of certain radionuclides than will failed fuel.                                                                                                                   I

!Answer 11 d I !Exam Level i I R I !cognitive Level 11 Memory I IFa~Hity: 11 Salem 1 & 2 I [Exarnbate: 11 12/21/20151 [KA:f I000076A203 I!AA2.03 i !Ro vaiue: I 2.5j jSRO Value: JI 3.olfseciion: II~ iRO Gi:oup:JI 21 lsRo Groi.i?:I! 2j 1~1 D fsystem/Evolution Title I IHigh Reactor Coolant Activity 11076 [KA Statement: ! Ability to determine and interpret the following as they apply to High Reactor Coolant Activity: RCS radioactivitv level meter jExpla~ati()p of I 55.41.b(5, 10) D is correct.This is the method as directed by procedure to determine if there is failed fuel. A is incorrect because

Answers:

' . , . . .

I there is no procedural guidance for the operator to use to determine source of elevated readings by how fast indications are rising.

C is incorrect because Radiation Protection is sent to perform surveys to repost areas as necessary for personnel protection, not to determine source of activity.. Bis incorrect because Letdown is maximized to expedite RCS cleanup for valid elevated activity, not tn -'~*~--;no~~ .. ~~ nf tho ~~.: .. :+.

i Reference Title

                                                            -1, q     FaC:ility Reference Number        I[~eference Section       J ! Page No. I !Revision:

IHigh Activity in Reactor Coolant System ll S2.0P-AB.RC-0002 II II 1 Is I I Ii II II 11 I I II II Ii 11 I 'Lo: Number J Objectives I ABRC02E001 I ABRC02E003 ,_ _ ___.. [.Material Requirecl f()r Examination j I 11

\OuesHon     ~olll"~~;      J j Facility Exam Bank         1jau~~fi?rtM()dificaVon       M~th()~:  J  Direct From Source       Iiused Di.iring Training P~dgrarn I D fQyesti()"1-S()urc~Co)'Jlrpents! l , . . . 7 _ 1 _ 4 _ 8 _ 3 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . . . . , . l i I                                                                                                                    I I                                                                                                                    I I                                                                                                                    I

RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic 11RO20 Given the following conditions:

  - Salem Unit 1 is offiine.
  -   Salem Unit 2 operating at 95% power, 1150 Mwe, with its Power System Stabilizer (PSS) out of service.
  - Unit 2 Main Generator gas pressure is 75 psig.
  -   Hope Creek is operating at 100% power, with its PSS out of service.
  -   The Hope Creek 5-6 breaker is out of service.
  -   Subsequently, a 500KV grid disturbance results in lower than normal grid voltage.

If a power reduction was required, which of the following identifies Main Generator loading which is outside the allowable for Salem Unit 2 IAW A-5-500-EEE-1686, Artificial Island Operating Guide? Trip-A-Unit is NOT armed. Salem Unit 2 operating at Mwe with MVAR loading out.

     ~ 11000, 150.

nl

     ~
     ~   !1000, 575.

I

     ~ 11100, 575.

IA11swer '. ~ IExam Level I ~ ':Cognitive - - Level j IApplication I \,Fac;ility:i ISalem 1 & 2 I !Exa(TID,ate: I I 12/21 /20151 ~joooo77A201 liM2.01 f!Rova!Ue:J@[sROValue:l~lsecticm:Jj~[RO.Group:)LJ!s~oGroup:JLJ ID D IS:ystem/Evolution Title I j Generator Voltage and Electric Grid Disturbances If~0_77_~

l<A Statement: ! Ability to determine and interpret the following as they applv to Generator Voltage and Electric Grid Disturbances: I Operatinq point on the oenerator capability curve I IExplan~tion of/ 55.41.b(4) With Unit 1 O/S and the HC 5-6 breaker O/S, the correct curve is 2S2H-5-6 on page 291. With both Units PSS O/S,. the Answers: 1 red dashed line will be used for allowable generator excitation. A is incorrect because the PSS is O/S. If either units PSS was IN service, then it would be correct. The 2 distracters with higher MVARS are both within the limit. Since there are two different Mwe loading conditions, and the choices for each are high/low, the answer cannot be obtained by ruling out 2 of the choices because

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I Reference Titl~ .' * . **.IL > facility Reference Numb_er j [Reference Sedioi:i . 11: Page No. j !~evisionl IArtificial Island Operating Guide 11 A-5-500-EEE-1686 I 11291 1111 I I II ii 11 I I II I II 11 I ]LO~ Num!Jer * * *. \ Objectives IGEN002E016 I I GEN002E017 I '--~

I IMat~rialfiequireq for El!:C1!T!in.9tJoi;l.:;\iJ'.! RO 20 A-5-500-EEE-1686, pages 123,291,308 lj l<it~~~ti~J?.*so.Yrce: / IPrevious 2 NRC Exams I/9uesf!-6il M.~dlfis~ti~~ M~W§Ci: ;' ~ Significantly Modified 1lµs,~d olidng Tf#iningJ~:r<>9ram',\ D

1 fQ~~g;t!gfl s~.!Jrce9()qifi1entsi 113-01 RO 021 all 4 choices changed to different combination of numbers. No original numbers from 13-01 exam

  *          * **    * *       **   used. Answer position changed.

I I I I I I !

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes

 !Question Topic        i I RO 21 Given the following conditions:
  -   A SBLOCA has occurred.
  -   The crew is performing the actions of EOP-LOCA-2, POST LOCA COOLDOWN AND DEPRESSURIZATION.
  - SI pumps have been stopped.
  -   Normal Charging is aligned.
  -   The crew is depressurizing the RCS using normal spray.

Which of the following describes the strategy for controlling RCS subcooling during the depressurization?

    §]I Subcoolino will be ...

minimized to reduce RCS break flow. I I

    ~ maximized to ensure continued RCP operation.

I

    ~ maximized to prevent a challenge to the core cooling critical safety function.

I l.d. I

        ! minimized to ensure pressurizer level remains above the lower limit to allow heater operation during the RCS cooldown.

I IAnswer I

            .!al         iExain Level   I fRl      1Coal1itlve Level ] I Memorv                  ! !Facility: q  Salem 1 & 2             l~ExamDate: W          12/21/20151 IKA:ll OOWE03A103                IiEA1.3 _jiRO Value: i           3.7l lsROValue:Jl 4.1!     [secti6n:Jl~IRO Group:JI                     211s~o Grollp:JI    21         D

[System/Evolution Title j j LOCA Cooldown and Depressurization

KA Statement: I ll~.plal'lat\o. h of I 55.41.b(10) Strategy of step 31 is to depressurize and attempt to minimize subcooling so that break flow is reduced, due to the

.Answers:** .*. I minimal makeup provided by charging pumps. B is incorrect because RCP operation is not required for this event, although desired. C is incorrect because core cooling should not be challenged on loss of subcooling at these temps and pressures(this point in the cooldown) D is incorrect because PZR heater operation may be required to reduce the rate of increase in pressurizer

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                  <.Reference Title                       I!                                         i Facility Reference Number * \Reference Section                  .* i I Page No.\ !Revision!

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[Lo. Number J I LOCA02E001 Objectives '--~ !Mat~ri!ll Requiredfor EXamination?*. ; I lj [o':'estie~~ourc~: **J IFacility Exam Bank l i(l~e~tion llJ!odifi~~tion<ftii~thocf,: ;+~Direct From Source I !u~ed During Training Ptogralll I D

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Outline Changes I . . t<lue~tion Topic ! IRO 22 I Given the following conditions:

  - Unit 2 is attempting to identify and isolate a LOCA outside containment.
  -  2-EOP-LOCA-6, LOCA Outside Containment, has just been entered.
  -  The source of the water is backleakage from the 23 cold leg injection line.

Assuming that any valves required to be operated during LOCA-6 operate correctly, which of the following leak locations would NOT be isolated while using 2-EOP-LOCA-6?

    ~I On the valve inlet flange on 22SJ49, RHR DISCH TO COLD LEGS.

I

    ~I On the valve outlet flange on 21SJ49, RHR DISCH TO COLD LEGS.

I rz1 IBetween the 2RH20, RHR HX BYP VALVE and the 2RH26, HOT LEG ISOL VALVE. I

    @]I  Between the 2RH2, RHR COMMON SUCT VALVE, and 22RH4, RHR PMP SUCT VALVE.

I \Answer 1 I b l IExa!llieveti IR I !cognitive Level : 11 Application I (Facility: ! ISalem 1 & 2 I IExamDate:' 11 12/21/20151

..

~I OOWE04K202 i~!SRO V~ilue: ir.I§j" !Section: \I~ \Ro Group:\ LJ )SRO Group:\ LJ Ill;, li~ I l 11 EK2.2 ! !RO Value: D isyster:n/Evo!Ution Title I ILOCA Outside Containment I ~J iKAStatement:i KnowledQe of the interrelations between LOCA Outside Containment and the followinQ: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility. fExplanation of 55.41(3,8) 2-EOP-LOCA-6 closes/checks closed the following valves: 2RH1 OR 2RH2, 21 and 22RH19s, 2RH26, 21 and

Answers: 22SJ49s. Using drawing 205332-SIMP, it shows that any leak between the RH1/2 and the SJ49s will be isolated with the above valves closed. The only location which wouldn't be affected by those valve being closed in the downstream/outlet side of the SJ49 valves. The stem statement of proper valve operation was inserted to preclude a candidate from assuming a leaking valve may not I

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Reference Title 11

                                                              *'          Fac:ility Reference Number,                i !Reference Section                       i      11 page No. I IRevision[

ILOCA Outside Containment !12-EOP-LOCA-6 I II 1121 I I ii 205332-SIMP I ,,11 11 I I II I 1 I I /LO,, Number// Objectives I LOCA06E002 I_ ___. !Material Required tc>r.ej(a!lliriation .\I I la1;1e~tion Soup::e: xl l Facility Exam Bank IJouesti.j>n llJlo~ificati~n Met~od:'.: I Direct From Source I!Used Durlllg Training Pr69r<1m l o !Question Source Comrri'~ritsl 1133792. Used 3 NRC exams ago Sept 2011 RO exam I icomment I I I I l I

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes

  !Questi<?n Topic      i IRO 23 Given the following conditions:
   -   The crew is in 2-EOP-FRHS-1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, and the criteria for initiating RCS bleed and feed has been met.
   -    Prior to the actual procedure steps there is a CAUTION statement that reads:

TO ESTABLISH RCS HEAT REMOVAL BY RCS BLEED AND FEED, STEPS 24 THRU 29 MUST BE PERFORMED QUICKLY AND WITHOUT INTERRUPTION Which of the following describes the basis for that statement? I

      ~ Stopping RCP's is the first step in the process. This terminates all RCS heat removal until bleed and feed is initiated.
      ~I Expeditious performance of the steps allows time for other compensatory actions if bleed and feed actions are unsuccessful.

l:J IDelay allows core cooling to degrade further. RCS pressure rises such that ECCS flow is lower when bleed and feed is initiated. I [] I Establishing SI flow and then delaying opening the PZR PORVs may lead to damage to the PORVs and Code Safety Valves when they pass water. I !Answer 1 c I I jexam Level I lR I !cognitive .Level 11 Memory I I !Facility: 1 Salem 1 & 2 I jExamDate: 11 12/21/20151 ~joowEo5G420 l1RoVatue:1(2]'tsROValue::@1section:;!~IROGroup:[LJ~~LJ !:;i~I D J.~~~ I ' I ! ! *I I 112.4.20 ' iSystem/Evolutioii Title i ILoss of Secondary Heat Sink I E05 1 I IKA Statement* I Knowledge of operational implications of EOP warnings, cautions, and notes. 1i:xplanati?"'Pt,.j 55.41.b(1 O)Per FRHS Basis document states that delay allows further degradation of cooling, followed by RCS pressure rise, and ,Answers: * .**

I core uncovery may be greater because ECCS flow is limited by RCS pressure. A is incorrect because there is still some cooling after the RCP's are stopped since there is water left in the SG's. B is incorrect because there are no alternative steps. D is incorrect because ECCS flow is actually reduced by the delay - the PZR will not go solid.

f* Reference Title ****!I .. Facility Reference Number ** . 1[Refere~ce. Section.* *;I tPage No.! IRevision 1 ILoss of Secondary Heat !12-EOP-FRHS-1 II Basis Doc 11 7 1124 I l II II 11 11 I I II II 11 11 I il.O. . Number. I FRHSOOE009 Objecti '-~---'far lMateri~I ReqUirea Examination. *. II 11 tou~~tio~ sou,rc.ff l IFacility Exam Bank I [Question ¥odificatioll;o/)etl}<;>e: JDirect From source I(used pµring Training Prograi:rt I D fpu,~stiorfs~i.lfoe corrimeritsj 180840

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RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes

 !Question Topic 11 RO 24 Given the following conditions for Unit 2:
   - A LOCA has occurred.
   - While performing 2-EOP-LOCA 1 "LOSS OF REACTOR COOLANT" 22 RHR pump motor seizes and power is lost to 21 RHR pump.
   - The crew enters 2-EOP-LOCA-5, "LOSS OF EMERGENCY RECIRCULATION".
   -   A cooldown has been initiated as directed in 2-EOP-LOCA-5.
   -   During the cooldown, the crew restores power to 21 RHR pump.
   -   RWST Level is 30' Based on current plant conditions, which of the following describes the mitigation strategy?

g) IContinue with the cooldown and start 21 RHR pump when directed in 2-EOP-LOCA-5. I

      ~ Return to 2-EOP-LOCA-1 and continue recovery actions with the step previously in effect.

I I

      ~ Start 21 RHR pump and continue actions of 2-EOP-LOCA-5 until the RWST LO Level alarm actuates.                                                                 I
      @] I Start 21 RHR pump and transition to 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation, to verify recirculation flowpath.

I !'Answer 1 I b I !Exam Level 11 R I jcognitive Level

f 1 Memory I !Facility: ! ISalem 1 & 2 I IExarnDate: ! I 12/21/20151

[KA:ll OOWE11A103 !1EA1.3 : jRo'Vatue: ll 3.7l[SRO Value: :j 4.2lfSection: !I~ iRO Group:JI 1![sR.o Group~!I 1j IJ~ D !System/EvolUtlon Title l ILoss of Emergency Coolant Recirculation I 1E11 IKA Statement:! Abilit to operate and I or monitor the followin as the a pl to Loss of Erner enc Coolant Recirculation: Desired operatin results durin abnormal and emer enc situations. IEXPl~nation of 55.41.b(10) B is correct because Continuous action step states that if any train of emergency recirculation capability is restored then iAriswers: the crew should return to the procedure and step in effect. This is consistent with the organization of the EOPs. C is incorrect because Continuous action Step 6.1 directs return to evaluate train status and a return to the procedure in effect; A is incorrect because continuation of the cooldown in LOCA-5 is not required. The purpose of the procedure is mitigation and recovery of

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Iquestion is not considered SRO level based on only requiring general mitigative actions j.;'*, Reference Title*.* I[

Facility Reference Number * '[ReferencE! Section* / [ f Page Nod [Revision[

ILoss of Emergency Coolant Recirculation 112-EOP-LOCA-5 I ii 1125 I I II I 11 ii I I !! I II 11 I fG.c): ~umber** *** *. ..... *[ Objectives I LOCA05E008 I il\llaterial Required t,or~J<amiriatiofj . ;:j I ll 1aue~ti.~n Source;. IIFacility Exam Bank I 1 laHist~<<:in Modifip~~eri:,,(IA~tho~! *** Direct From Source I ~Used During Training Pr(:>9raJli j D

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I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I.. .. louestioh Jopic 11 RO 25 I Given the following conditions:

  - Unit 1 has experienced a MSLB at the mixing bottle.
  - MSLI has failed to close ANY MS167.
  -  Operators have transitioned out of 1-EOP-LOSC-1, Loss of Secondary Coolant.
  - RCS pressure is 1345 psig and dropping slowly.
  -  RCS Tes are dropping.

Which of the following describes Reactor Coolant Pump strategy, and why? RCPs should ... I

    ~ be tripped to minimize the heat input into the RCS.

I

    ~I continue to be run, because RCP pressure dependent trip criteria are only applicable during a LBLOCA.

I [] I continue to be run, since RCP pressure dependent trip criteria is not used when a cooldown is in progress. I [) I be tripped so that 2 phase flow doesn't develop if they were to be tripped later in the event, leading to higher peak clad temperatures. I IAnswer 11 c I [Exam Level I l R I !cognitive Level. ! I Application I IFac~lity: 11 Salem 1 & 2 I !ExamDate: 11 12/21/20151 I lKA: 1 OOWE 12A202 llEA2.2 !IROValue:!j 3.4l[SROValue:jl 3.9llsection:ll~IROGroup:ll 1l[sR0(3roup:ll 11 !II D iSystem/Evolutioh Title i IUncontrolled Depressurization of all Steam Generators I IE12 1 KAStatement: I~ determine and interpret the following as thev apply to Uncontrolled Depressurization of all Steam Generators: rence to appropriate procedures and operation within the limitations in the facilitv's license and amendments. I [F'<Pl~nati?~. ?1.1 55.41.b(10,3)The RCP trip criteria with regards to pressure in a LOSC condition is for pump protection only. The Generic issues ,Answers:** J segment of the ERG executive volume describes the SBLOCA scenarios where pumping coolant out the break then stopping RCP's leads to peak clad temps in excess of 2200 degrees. That is not applicable here. What is more important in the LOSC is forced flow. LOSC-2 basis document page 12. I " Reference Title'* Ii Facility Reference N,umber * .I [Reference Section > l IPage No'. I !Revision* IMultiple Steam Generator Depressurization 111-EOP-LOSC-2 I q12 1122 I I I I 11 11 I ! I I II 11 I Objectives I __ LOSC02E003 , IMatefialReq1;1fr~dfor Exarhi11ation **l j lj IS:~~stion so.urce: *I IFacility Exam Bank I [9u,e~tion Modificatiorl:ivfetho,c!M~* ~Direct From Source I!used During "fiail1ing Prpgram ID

~ollestipJ1 sourc~*com~entsJ 163166

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[ouestionTopic I l RO 26 I Given the following conditions:

  - Unit 2 has experienced an event which has resulted in 24 SG pressure rising to 1115 psig.
  -  A MSLI has been performed.

How many total SG Safety Valves will be open if 24 SG pressure remains at 1115 psig and all Safety Valves operate when expected?

    ~12                                                                                                                                                        I

[)13 I

    !c.f 14 I
    @]Is                                                                                                                                                       I

!Answer 11 b I !exam Level 11 R I \Cognitivl! Level 1 IMemory I !Facility: ISalem 1 & 2 1 I [Ex~mDate: !I 12/21/20151

~jooWE13K202 "iiROValue:J~ISROValue:J@1section:;;l~:ROGroup:JLJ1SROGroup::LJ ~ D I *.* ., I ':"'] l,EK2.2 iSystem/Evolution Title i jSteam Generator Overpressure 11E13 I ,KA Statement: I Knowledge of the interrelations between Steam Generator Overpressure and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility. i [Explanation of 55.41.b(4) Each SG has five safeties, with lift setpoints of 1070, 1100, 1110, 1120, 1125 psig 1Answers: * . *. I

j iReference Section

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I ' Page No. I [Revision: I

                 *. Reference Ti.tie ...                It   Facility Reference Num1:>er:

IUnit 2 Main Steam System 11205303 Sheet 2 I lj61 I I II I 11 11 I I II I II 11 I !Lo. Nu.!Tlber I STMGENE008 Objectives j_ ____. [Material Required torExaminatiori. !I lj IQuestion s?urce: ... 11 Facility Exam Bank I [puestion IVlodification Method:,r l Direct From Source I [usec!During Trajning Program ;l D jouestr~n Sou.rce Comm7,nts\ 1152360 used 3 NRC exams ago Sept 2011.

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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes

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!Question Topic      J 1RO 27                                                                                                                                 I IOf the following choices, which is the only automatic action expected to occur as containment pressure rises from 12 psig to 18 ~sig during a LOCA? I nl 1a*:

I

    ~    I Ei    I Feedwater Isolation.

I l__J

    ~    I Main Steam line Isolation.

Containment Ventilation Isolation I I I !Answer 1 c I !Exam Level ! ~ !cognitive L.evel 11 Memory ! !Facility: 11 Salem 1 & 2 I \ExamDate:;i I 12/21/20151 IQ2j [SRO Value: IQ2j lSection: 11~ :RO Group:! LI !SRO Group:[ LI l~~i; D

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I !KA:ll OOWE14A101 ! !RO Value: ls~~~~~~ooTWej l-H~ig~h_C_o_n_t_ai_n_m_e_~_P_r_e_u_u_r_e_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~llE14 'KA Statement:: Ability to operate and I or monitor the followinq as thev aoplv to Hioh Containment Pressure: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. I i !Explanation of 55.41.b(7) Phase A isolation occurs on any SI signal, which at the very least would have occurred at 4 psig in containment if not Answers:. l sooner. Feedwater Isolation occurs on any SI or SG NR level >67%. Main steamline Isolation occurs at 15 psig. Containment Ven isolation occurs any SI, RMS alarm of associated monitors, Phase B. The SI signal would have already actuated. I Reference Title j; Facility Reference Number *I IReference Sec.tion j IPage No. j jRevision: ILicensed Operator Fluency List ii NOS05FLUNCY I 1113-16 11 9 I I II I q 11 I I II I II !I I IL'.O'. Number **

Objectives FLUNCYE002

\M(lterial Required for i:xamination : II [auestion So1.frce:. J IFacility Exam Bank I !<lu~sticin.Modification Method: j Concept Used I!used During Training Progretm I O 1auestfon Source Cofl}'!le~ts\ I

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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I ......... ..... louestionTuPicJ IRO 28 I Which of the following identifies when will OHA E-48 ROD BOTTOM will clear during a reactor startup? When 1) rods are withdrawn past 2) steps.

     ~ 11) Control Bank A 2)20                                                                                                                                                              I
     ~ 11) Control Bank A 2)35                                                                                                                                                              I
     ~     1) Control Bank D

2) 20 I

[] 1) Control Bank D 2)35 I !A.nswer 11 a I !exam Level ! IR I !Cognitive Level . JI Memory I !Facility: 11 Salem 1 & 2 1 IExamDate: ! j 12/21/2015! ~I 001 OOOA302 *IRO Value: IfJ2i !SRO Value:!~ *'Section: 1I~ jRO Group:! LJ !SRO Group:] LJ lit~&~ D

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\system/Evolution Title j j Control Rod Drive System 11001 I !KA Statement: i Abilit to monitor automatic operations of the Control Rod Drive S stem includin hei ht !Explanation of I 55.41.b(6) There are 3 Rod Bottom Bypass Bistable Modules, for Control Banks B, C, and D only !Answers: . i 1)Blocks OHA E-48, ROD BOTTOM for its own bank when entire bank is <35 steps

2)When all banks are on bottom, OHA E-48, ROD BOTTOM alarm is illuminated.

3)When Control Bank A is >20 steps alarm is cleared (Control Banks B, C, D, bypassed)

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!>:* Reference Title.: ** II . Facility Reference Numbe.r . j ;Ref.erence Se~tion . 11 Page ~o. I IRE!yisionl IRod Control and Position Indications Systems L !INOS05RODSOO I lj 51-52 1112 I I II I 11 II I I II I ii 11 I LO. Number.* IRODSOOE008 Objectives I ,_ _ ___. itii1aterial Reqliireci tor e~amination 11 II [ouestion Source: 11 Facility Exam Bank 11auestion Modificati~.n fv1etho~: : ~Concept Used 1lused otii"ing Trainil"lg P/ogramdl D l<:luestion source Comments! 157724 =======-==::::==============~===========-================--!

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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes [Question Topic,J 1_R_0_2_9_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___. Given the following conditions:

  - Unit 2 is in MODE 6.
  - The Reactor vessel head is removed and on its storage stand.
  - The Refueling Cavity is being filled from the RWST IAW S2.0P-SO.SF-0003, FILLING THE REFUELING CAVITY. (Assume RWST level was at 40.5' when the unit was shutdown)
  - Refueling cavity water level is 11 O' and rising.

Which choice identifies the indications which will be present on the Control Room Console? PZR Cold Calibrated level is ...

    ~ 13%; RWST level is 27'.
    ~ 180%; RWST level is 20'.

rel Ioff-scale low; RWST level is 40'.

    @] I off-scale high RWST level is 1O'.

!Answer I [L:J 'Exam L~vel : !&:J !Cognitive Level I !Application _l_JFacUity: wSalem 1 & 2 l~ExarTlpat~: w___1_2_12_1_12_0_15_,I ~ j 002000A111 I A 1.11 1 I!Ro Valut}: !@[SRO Value: l@fsection: ii~ [Ro Group:j[}j ls Ro Group:j LJ D [System/Evolution Title> I l_R_e_a_ct_o_rC_oo_l_an_t_S....:y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __, ~--~ !KA Statement: 11 Ability to predict and/or monitor chanqes in parameters associated with operatinq the Reactor Coolant System controls including: I I Relative level indications in the RWST, the refueling cavity, the PZR and the reactor vessel durinq preparation for refueling I fExplanation of : 55.41.b(3,)The refueling cavity at 11 O' will hold -112,000 gallons. The RWST normal operating level is a minimum of 40.5 feet prior jAnswers: ; * ** j to refueling (364,500 gal). Subtract and the total left in RWST will be 253,000 gallons, which will be -27.5'. The corresponding PZR cold cal level will be > the 0% level at the 108' 11" elevation in containment. This question does not require memorization of tank levels, but rather an understanding of the physical connections and relative elevations of the systems. ! .Reference Title il Facility Reference Numbt}r:. .. !\Reference Section J i Page .No. I !RevisionJ IDraining the Refueling Cavity II S2.0P-SO.SF-0004 11 1129-30 1117 I I II II q 11 I I 11 II 11 11 I [LO.Number Objectives I RCSOOOE006 1_ ___. !Material RequiredfofExarriinatiori <I I lj [Question Spurce:. J IFacility Exam Bank I !Question .Mo,dificati9n Method: . :I Direct From Source Iiused During'rralning Prpgrain I D

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 !Question Topic 11 RO 30                                                                                                                                                    I IPrior to opening 1CV114, RCP Seal Bypass Valve IAW S1 .OP-SO.RC-0001, RCP Operation, all of the following conditions must be met EXCEPT:                                   I
   ~    I RCS pressure < 100 psig.

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   ~I Any RCP seal leakoff flow< 1 gpm.

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   /cl  I 11-14CV104, SEAL LEAKOFF valves open.

I [] I Seal water flow to each RCP is at least 6 gpm. I [Answer 11 a I iExain Level I I R I !cognitive Level 1 IMemory I !Facility: 11 Salem 1 & 2 I texamoate: !I 12/21/2015!

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~I 003000A407 @ET 'SRO Value: 10]" '~Section: 1!~1RO Group:JLJ !SRO Group:I L J ~ i;>~. ~ I

I IA4.07 1 IRO Value: I D I

!S~tem/Evolution Title I Reactor Coolant Pump System I ~0_03_~ IKA. Statement: I Ability to manually operate and/or monitor in the control room: RCP seal bypass IExplap~tton of ! 55.41.b(3) RCS pressure must be between 100-1,000 psig. If pressure is less than 100 psig, the CV114 is required to be shut. All Answers: : distracters are listed in SO.RC-1 step 5.2.1.

           . *****.** .. /*Reference iitte                      II  **Facility Reference Number'>    i lReference Section .*     j i Page N()i! iR,eyi~ionl I Reactor Coolant Pump Operation                                  I 1 S1 .OP-SO.RC-0001                  I                           1113         1134         I I                                                                II                                   I                           q            II           I I                                                                II                                   I                           II           11           I

\L.o. Number I RCPUMPE013 {s Objectives I 1_ _____, [Materi13t Required for ExamiQation . II 11 fQuestton Source.: j IFacility Exam Bank 1 IQues~ion l'Aodification M~thod: J Concept Used I [Used During "fraining Program J D i~~~~~~~l~4_3_0_~-----------------------------------~I jcomment** I I I

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SRO Skyscraper I...... RO System/Evolution List I SRO System/Evolution List I Outline Changes I lauestion Topic 11RO31 I Operators are preparing to start 21 RCP IAW S2.0P-SO.RC-0001, Reactor Coolant Pump Operation. Which of the following would prevent the RCP from starting when its START PB is depressed? 21 RCP ...

     ~   I#1 seal D/P < 200 psig.

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     ~I   4KV breaker trip springs not charged.

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     ~I STANDPIPE LEVEL LO alarm                 locked in.
     @JI Oil Lift Pump discharge pressure <500 psig.

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                                                          ... Level . !IMemory                          .. .1 ISalem 1 & 2 I

["O" 1Exam Level ! ~ !Cogmbve i I 1Fac11lty. I ,ExamDate.: :11I

! *

)Answer) 1 1 12/21/2015h l~:l j 003000K101 IIK1 .01 [I Ro Vallie: I[]]" :5Ro Value: I12] !section: ii~ IRo Group:j LJ lsRo Group:I LJ !!'~ D IS:Ystem/Evoiution Title I j Reactor Coolant Pump System 11003 rKA Statement: l KnowledQe of the physical connections and/or cause-effect relationships between Reactor Coolant Pump System and the followinQ: I RCP lube oil I iExplanation !. . of iI 55.41.b(3) A is incorrect because it is a manaul RCP trip setpoint stated at Step 3.2.9 third bullet. B is incorrect because the trip

Answers: i springs are charged when the breaker closes, it woiuld be the closiong springs not charged which would prevent the 4KV braker form closing. C is incorrect because the standpipe level low has no interlock to prevent pump starting. D is correct as shown on drawings.

I Reference Title I Facility Reference Number I~Reference Section* j !Page No. I !Revision! IReactor Coolant Pump Instrumentation 11220424 I ii ii 05 I IRCPs and RCPs Lift Oil Pumps 11224405 I ii 114 I I II 11 II ii I iLd. Number I RCPUMPE006 Objectives J 1_ __ . . )l\llaterial Required for Examination 11 II 1auestioh Source; 11 Facility Exam Bank I[auestiC>ri Me>dificatie>l"l ~et~o~: JI Concept Used I tused During Traiiling Prograr:n J D fOU esHe> n Sq~r~e corri~eri.t~!

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RO Sky Scrape~ SRO Skyscraper i __R_o_s_y_st_e_m_1E_v_o1_ut_io_n_L_is_t..... __sR_o_sy_s_te_m_1_Ev_o_1u_ti_on_u_st...... Outline Changes

 \Question Topic I !RO 32                                                                                                                                                            I I With normal charging and letdown in service, which of the following valves failing closed is interlocked to cause 1CV4, Letdown Orifice Isolation Valve, to automatically shut?                                                                                                                                                      I
   ~ 11CV18, Letdown Pressure Control valve.
   ~ 11 CV??,

I 23 Loop Chg Line Stop valve. ic. i 11 CV?, Letdown HX Inlet Isolation valve. I

   ~I 1CV2, Letdown Isolation valve.

I !Answer J d I I IExam Level i IR l [cognitive Level 11 Memory j jF~cility: 11 Salem 1 & 2 1 IExamDate: 11 I 12/21/20151 IKA:JI 004000K413 IiK4.13 I!Rd Value: lj3.2*1 iSROValue:1I 3.51 [section: II~ ~RO Group: I 1 11 lsROGroup:ll 11 1 1~'~ D lsystern/Evolution Title J IChemical and Volume Control System 11004 IKA Statement: J I Knowledqe of Chemical and Volume Control Svstem desiqn feature(s) and or interlock(s) which provide for the followinq: Interlock between letdown isolation valve and flow control valve !Explanaticm pf I 55.41.b(?) The 1CV2 being closed would cause any of the 3 open orifice isolation valves to shut. 2CV4 is stated in the stem to be iAnswers: *.*.' the open isolation valve. i I.

Reference Title **I:. Facility Reference Number
I/Reference Section * !

11 Page No. !Revision/ I1CV4 Logic Diagram ii 224430 I ii l l5 I ! II I II 11 I I II I ii 11 I Lo. Number

I Objectives I CVCSOOE006 j_ ___.

[Material Required for Examination II II IQuestion Sciurce: .*.11 Facility Exam Bank I fGW.e~tion M9clification Method:/ . JConcept Used 1lused outing TrainJng Program. I D

),Questici~ SourceCofnllJents/ 161622
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[Question Topic I IRO 33 I Which of the following methods is used prior to cooldown to remove hydrogen from the RCS after a unit shutdown to enter a refueling outage, and why is hydrogen removed? deqassification is used to remove hydrooen from the RCS to prevent I

   ~ Chemical, an inadvertent crud burst during RCS cooldown.

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   ~ Mechanical, an inadvertent crud burst during RCS cooldown.

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   !:§] IChemical, having an explosive concentration present when oxygen is introduced to system.

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   ~I Mechanical, having an explosive concentration present when oxygen is introduced to system.

I IAnswer 1 Id I !ExarhLevel 11 R l !cognitive Level i IMemory I IFacilitY:'J ISalem 1 & 2 I \Ex~mDate: \ I 12/21/20151 ~I 004000K549 I[K5.49 I \RO Value: llJ2itsRO Value: l@lsection: II~ \Ro Groop:jLJISRO Group:\LJ i55l31 P-illl~,t~ ltliJ1~!6Iill D [system/Evolution Title I IChemical and Volume Control System 11004 I

KA Statement:! Knowledge of the operational implications of the following concepts as they aooly to the Chemical and Volume Control System:

Puroose and method of hydrooen removal from RCS before openina svstem: explosion hazard, nitrogen puroe [Expl~r;iat'.on of I 55.41.b(3,5) Both chemical (addition) and mechanical (replacement of H2 with N2) degasses are performed during a 1 Answers. . .* I shutdown/cooldown. Mechanical is most effective at NOP/NOT (see P&L 2.1.5) A is incorrect because chemical degass is performed when RCS temp is less than 250 (see pre-req 3.1.7). B is incorrect because of wrong reason for mechanical degas. Cis incorrect because chenical degass is performed to initiate crud burst (addition of H202). i '* Reference Title . I' : *.Facility Reference Number '* I!Reference Section mment I I I l I I I

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes lauestion Topic 11 RO 34 I Given the following conditions:

- Unit 2 has experienced a Large Break LOCA. - 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation, is complete with NO abnormalities encountered. - Operators are currently at step 26, "Preparation for Hot Leg Recirc", of 2-EOP-LOCA-1, Loss of Reactor Coolant. - Off-site power is supplying all 4KV Vital busses.
If BOTH RHR pumps are operating, what would be the effect if 22 RHR Pp were to trip?
~ 121 and 22 SI pumps would begin to cavitate.
I
~ 122 Containment Spray Pump would lose NPSH.
I lS 121 and 22 Charging pumps would begin to cavitate. I
@] I Flow to the Containment Spray header would be lost.
I IAnswer ! d I I [Ex~m Level i IR I !cognitive Level i IApplication I if'acility: j ISalem 1 & 2 I [ExamDat:: J I 12/21/20151 IKA:JI 005000K305 j IK3.05 11Ro Value:!l3.7*1 iSRO Value:i13.B*l lsection: ljsYS l[Ro Gfoup:!I 11 ISROGroup:il 11 II! D [System/Evolutic:m Title i IResidual Heat Removal System I ~I0_0_5_~ IKA Statement:! Knowied e of the effect that a loss or malfunction of the Residual Heat Removal S stem will have on the followin ECCS lExplanati<m of j 55.41.b(5) Distracter A is incorrect because the closure of the RH19's is done to prevent RHR pump runout if only a single RHR Answers: i pump is operating, so the SI pumps will not lose suction. D is the correct answer. LOCA-3 explicitly states that if BOTH RHR pumps are operating, then 22CS36 is opened to supply containment spray from 22 RHR pp discharge. Distracter C is incorrect charging pumps, as well as the SI pumps, will not lose suction.
. 11se all containments. .
I. Reference Title *! I
Fa(::ility Reference Number :I IReferenceSection P~ge No. I [Revision' l~L=o=ss==of=R=e=a=c=m=r=C=o=o=la=nt==============~ll~E=O=P=-=L=O=C=A=-1==============~~==============~~====::::;.!.::=::::::::=:::.
!Lo. Nlllllber J I ECCSOOE016 Objectives I RHROOOE016 j_ ____, !Material Requir~<J for Examinafigp **I I I \a!JE!stion Source: *J Facility Exam Bank I ISp~5-ticm Mc@ficaHon.Method: . {I Editorially Modified 1 lusedollrlng TrliiJlin~(Progt,am I D [auestiorlSolirce Gomme~tsj
I Modified correct answer terminology from Containment Spray flow would be lost to Flow to the Containment Spray header would be lost.
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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question Topic 11 RO 35 I Given the following conditions:
- Unit 2 is operating at 100% power with 22 charging pump in seNice. - 23 charging pump is out of seNice and operable.
Which of the following Tech Spec LCO's would be applicable if 22 charging pump were to trip? [ ] 13.5.4 - Seal Injection Flow.
~ , 3.5.2 -ECCS Subsystems >350°F.
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§] ,3.1.2.4 -Charging Pumps - Operating.
[ ] , 3.1.2.2 - Boration Flowpaths - Operating. I I IAnswer 11 b I [EX:a~ Liwe1 IR I t :cognitive Level *! IApplication I !Facility: 11 Salem 1 & 2 I !ExarnOate: 11 12/21/20151 I~i2_.2_.4_2_~1 IRO Value:!~ [SRO Value:j~ ~Section: II~ IRO Group::LJ :sRO Group:JLJ Ila~ D
>X"'*~-;F<'"c~"~
I IKA: 11 006000G242 ~[s-y-st_e_m_/_E_v_o_lu-ti_o_n_T-it-le~I IEmergency Core Cooling System 1 i006 I rKA Statement*! Ability to recoqnize indications for svstem operatinq parameters which are entN-level conditions for technical specifications. IE~plctnation of i 55.41.b(6, 10) RO candidates are responsible for "Above the Line" knowledge of Tech Spec LCO's. A is incorrect because the LCO IArtsi.vers: ***. *.. .
I is for maximum seal injection flow which is set by adjusting manual seal injection throttle valves. Plausible because many procedures direct establishing 6-12 gpm seal injection flow per pump not to exceed 40 gpm total. Bis correct because the LCO states 2 complete trains of ECCS are required, and 22 charging pump (hi head ECCS) is required for the B train of ECCS. C is
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pumps were inoperabie. D is incorrect 'because 2 boration flow paths from RWST through charging pumps remain available: as does the Boric Acid tank flowpath with one BAT pump and one charqinq pump.
.*.*.Reference Title ;>*/'
II .. Facility Reference Number. *I!Reference section *** :J f Page No: I !Revision! j Salem Tech Specs II I II 11 I ! II 11 II 11 I I II I 11 II I Objectives tlVl?tJiria!ReqliiredfoiE.xamination ; 11 !j fq~~~ti?ll Source.:
11 New 1 louestion. Moclification ~?t.Mct: g,*1--------~I !used Dt:iringJrainirig Program:J D IQue,stJor,i Soprc~c?rnrn~~tsi I*~~~~~--~~~~~~-~~-----~~~~~~~~~~~~~--~~~--'
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RO SkyScraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes IQuestion Topi~ IRO 36 Given the following conditions:
- 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.
Assuminq NO operator actions were taken, which of the followino would result in those conditions?
~ IRCP #1 Seal failure.
I [g'.] IPressurizer Spray Valve failed open. I I
~ Pressurizer Safety Valve failed open.
[] I Steam Generator Slowdown piping failure. I !Answer 11 c I !Exam Level I lR I !cognitive Level JIComprehension I :Facility: i ISalem 1 & 2 I IExarriDate: !I 12/21/20151 ~!007000K101 l!K1.01 ltROValue:i~!SROValue:j[ITI[section:il~IROGroup::[]:sROGroup:!LJ BJ D
*J'.'F!l
!System/Evolution Jitle I l Pressurizer Relief Tank/Quench Tank System 1 l~o_0_7_ _ iKA Statement: I Knowledge of the physical connections and/or cause-effect relationships between Pressurizer Relief Tank/Quench Tank System and the following: Containment system i fEicplariatio"n of 55.41.b(9)The failure of a RCP #1 seal will not be seen in containment outside of closed systems. The excess seal leakoff flow past ,Answers: .* *. , l the #2 seal will be seen as a rise in RCDT level. The PZR safety failing open will cause the PRT to pressurize and the rupture disk to rupture, causing the saturated steam in the PRT to continuously be vented to containment. The Spray valve failing open would cause the lowering PZR pressure, bit would not cause changing containment conditions. Reference Title
I: .Facility Reference Number*. I Reference ,Section j [ Page No. I !Revision:
INo. 2 Unit Reactor Coolant !1205301-1 I II 1 lsg I II I 11 11 I II I ii 11 I IL.a. Nurriber .. *. I PZRPRTE008 Objectives I ,__ ____, [Material Required for Examiilation * ;:;J I 11 I l<:l~estio~. Source~*--' J Facility Exam Bank !lauestiori Modificatio~.11A7thbd: ** _** 11 Direct From Source I iuse~Quring'TrainingPrograrri'* I D IQues,tion s.o~~cl! Comrrieritsi ISalem May 2010 RO NRC exam, created from Point Beach 1/20/2006 NRC exam I
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!Question Topic 11 RO 37 I Given the following conditions: - Unit 2 is operating normally at 100% power. - Excess Letdown is in service due to a problem with an orifice isolation valve. - The 2CC215, EXCESS LETDOWN HX CC INLET Vair supply line breaks, and all air is vented from valve.
Which of the following describes the effect this failure will have on Excess Letdown temperature, and how will operators respond? Excess Letdown temperature will. .. la. I I rise and the 2CC215 bypass will be throttled open to restore normal letdown temperature. I I
~ lower and the 2CC215 bypass will be throttled open to restore normal letdown temperature.
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~I rise and Excess Letdown flow will be secured to prevent lifting 2CV115 eve RCP SEAL WTR INJECTION RETURN RELIEF VLV.
I I Id. j lower and Excess Letdown flow will be secured to prevent VCT temperature from lowering to the point where 23 charging pump must be secured. I I.Answer I j c I !Exam .Level I lR I !Cognitive Level*. I IApplication I !Facility: 11 Salem 1 & 2 1~1 12/21/20151 ~ j 008000A205 I1RO Value: i@:;!JfsRO Value: IQ}i !Section: 11~ \RO Group:] LJ :SRO Group:J LJ I?~:' D
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1IA2.05 !System/Evolution Title 1 j Component Cooling Water System I ~lo_08_ _ !KA Statement: I Ability to (a) predict the impacts of the following on the Component Cooling Water System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Effect of loss of instrument and control air on the position of the CCW valves that are air operated !Explanation oft 55.41.b(?) The Excess Letdown HX CC isolation valves 2CC113 (outlet) and 2CC215 (inlet) are fail close (air and power) isolation Answers: ! valves. With excess letdown in service and no cooling flow, temperature and pressure will both rise. Excess letdown flows into the seal return header and rising pressure causes seal return line relief valve to open. Knowledge of both the effect (valve failure position) and operator action (isolate excess letdown) required. SO.CVC-3 P&L 3.3 states to maintain excess letdown pressure
;'1t::.n ,....,_;,... +,.... ........ .... i:.c.1.:-- .............. 1 . ... , ....... - .1:-.c '
1* Reference Title. r Facility Reference Number * ** \ lReference Section
Ij Page No. I l~evisionl IExcess Letdown II S2.0P-SO.CVC-0003 II ii 116 I Ieves drawing 11205228-2 II I 184 I I JI Ii
"!I 11 I ILO.~umber Objectives I CCWOOOE012
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[Qm~stion ~ou~c;~/ 1j_N_e_w_ _ _ _ _ ___,l jCluestiori ~()C'imcatiop Method: .. ~ _ _ _ _ _ _ _ __.I Jused[>,uril)g Training Program I D
;~uestion Source Comm~.~t::i.11 I !~---------------------------------------'
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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I .. louestion Topic i IRO 38 I Given the following conditions:
- Salem Unit 2 is operating normally at 100% power. - The controlling PZR level channel fails low. - As the operators respond IAW S2.0P-AB.CVC-0001, Loss of Charging, and before an operable channel is selected for control, the RO reports that PZR backup heaters in auto are energized with the controlling PZR level channel indicating 0%.
Which of the following describes the operation of the PZR B/U heaters? PZR heaters in auto 1) be energized because 2) [ ] 11) should
2) PZR level has risen 5% above program.
[ ] 11) should
2) PZR pressure has lowered to the auto on setpoint of 2210 psig.
~ 11) should NOT
2) PZR low level cutoff at 17% should be keeping heaters OFF.
[ ] 11) should NOT
2) PZR pressure has remained above the auto on setpoint of 2218 psig.
!Answer 1 c I I !exam Level J IR I I !Cognitive Level .* J Application I [Facility: 11 Salem 1 & 2 I JExamDate: i I 12/21/20151 iKA:JI 010000A402 j:A4.02 : [Ro Value: JI 3.61 !sRO Value: II 3.41 lsectiori: II~ 'RO Group:!! 11 lsRo Group:JI 11 Ill D !System/Evolution Title I j Pressurizer Pressure Control System I ,010 IKA Statement: I~ manually operate and/or monitor in the control room: ters I [Explanation ' , .' . .-*. of lI 55.41.b(?,5) The low level on the control channel will cause an automatic letdown isolation. Charging flow will continue and raise
Answers: *
i PZR level. The backup heaters are designed to energize at 5% above program to ensure stauration conditions are maintained in the PZR. However, either an alarm or control channel failing low deenergizes all PZR heaters. Nothing has occurred which would cause a PZR pressure change except for the rise in PZR level. Pressure will not lower. PZR B/U heaters are designed to energize
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I Reference Title d Facility Reference Number }I tR~ference Section Ii PageNc). j jRevision' ILoss of Charging 11 S2.0P-AB.CVC-0001 II 11 119 I
- rl Annunciator Window E I 1 S2.0P-AR.ZZ-0005 ilOHA E-20 1129 1120 I I II II 11 11 I
!LO. Number Objectives I PZRP&LE006 1_ ____. M?terial Required for Examination/ II I fauestion Sou.rce: J l_N_e_w_ _ _ _ _ __.I faliestion Mo~ificaifon Method~, l_________.1 lusecl Duririg Training Progra01 IO iauestiori Somce Comments! I I
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!Question TOpic\ I RO 39 I IWhich of the following describes how the PZR Spra:I'. Nozzle is prevented from being thermallx shocked during normal operation? I ~I Initiating spray flow when the Spray Nozzle delta T exceeds 320°F.
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~ PZR Backup heater groups operating in auto forces continuous spray flow.
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~ PZR Control heater group firing to maintain PZR pressure at setpoint forces continuous spray flow.
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@]I A small amount of spray flow is bypassed around the PS1 and PS3 Spray Valves to keep the spray line continuously warm.
I !Answer ! j d I [Exam Level !IR I !cognitive Level 11 Memory I !Facility: 11 Salem 1 & 2 I IExamDate: 1 I 12/21/20151 IKA:ll 010000K401 !M.01 I!Ro Value: I 2.7i;SROValue:H 2.9! !Section; II~ fRO Group:!! 11 lsRO Group:JI 1j t!l'I D !System/Evolution Title.I IPressurizer Pressure Control System 11010 !KA Statement: I Knowledge of Pressurizer Pressure Control System desiqn feature(s) and or interlock(s) which provide for the follo\Ning: Spray valve warm-up I IExplanation~f I 55.41.b(5,7) PZR spray bypass flow is set during unit startup while@ NOT/NOP to ensure spray line temp is >500°F INith both ~Answers: I spray valves shut. Salem runs with one PZR B/U heater group in MANUAL which forces the spray valves open a small amount to provide continuous spray for boron mixing also, but that is not a choice. Spray flow is NOT initiated if spray line delta T exceeds 320°F. Control groups are SCR controlled and normally fire to maintain pressure ON program, and do not force spray fiow. ' f*;< Reference Title II *.*Facility Reference Number. .! !Reference Section *. l IPage No. / jRe\lision/ I Setting Pressurizer Spray Bypass Flow 11 S2.0P-SO.PZR-0008 II 11 114 I I II II q 11 I I II n II 11 l !Lg. Number. Objectives I PZRP&LE012 ,_ _ ___. l
!Questicm SCii.lice: .* l_N_e_w_ _ _ _ _ _ __.I L[Q_u:c__e_s-'tiolc'n.:..****..:..M_o_d_ifi~rc~a~ti..:..o_n..:..M..:..*e_t..:..~..:..C>_Ci~:.-J1 _ _ _ _ _ _ _ __,l /useif During Training Pr9gramd.j D
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RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes IOuestion Topic'! I RO 40 I Given the following conditions:
- Unit 2 was operating at 90% power with 21 charging pump in service when the controlling PZR Level Channel I failed low. - The Charging Master Flow Controller was placed in Manual when directed by S2.0P-AB.CVC-0001, Loss of Charging. - The alternate PZR level channel has been selected as the controlling channel. - Letdown has been restored.
Which of the following identifies a consequence of returning the Master Flow Controller to auto PRIOR to returning PZR level to program as directed in S2.0P-AB.CVC-0001 ? Charging flow will ...
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rise, and VCT auto makeup may initiate.
~ lower, and flashing in the Letdown line could occur.
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~ rise, and RCP seal injection flow could exceed Tech Spec limit total seal injection flow.
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@]I lower, and 2CC71 LTDWN HX CC CONT VALVE will not respond quickly enough to prevent Mixed Bed Demin isolation on high inlet temperature.
I !Answer\ I~ !Exam Level I I~ !Cognitive Level 1 lAppl1cat1on -I !Facility: i j Salem 1 & 2 1lExamDate:11 _ _ _ 12_12_1_12_0_1_,5! I
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I !KA: i 011000A404 1 IA4.04 i iRO Value: I i"EJ fSRO Value: :[.Ifil [section: 11~ fRO Group::LJ !SRO Group:j L J D [SYStem/Evolution Title I IPressurizer Level Control System
KA Statement:! o manually operate and/or monitor in the control room:
Transfer of PZR LCS from automatic to manual control I ! . * .. rExplanation of I 55.41.b(7)With a CCP in service, the failure LOW of the controlling PZR level channel will cause charging flow to RISE. The stem !Answers: * ** 1 stated that MFC was taken to manual when directed IAW AB, so there was sufficient time for actual charging flow to rise substantially. With actual level higher than programmed level, if the MFC were placed in auto it would force charging flow to lower. If charging flow lowered to <-60 gpm, inadequate cooling of letdown flow would occur in the regenerative heat exchanger, and I, ....................... lino fl..,,c-hioa l~lCI 1ld CCCI IC IbP r'r-'71 it- nf"'l-........ ***ilh --'* -1 no/_ l"\l""IOn ............ L-.-- ................. ,..,f ............ _ +,.... ,.., ............ if*- .......... **- "---,n ............. +r. rise downstream ofthe letdown HX, and temps would not reach demin isoiation levels. The 2 rises are incorrect because charging flow wouldn't rise, but the actions associated with higher flow are correct. I Reference Title :1 Facility Reference Number .. I!Reference Section II Page No. I !Revision' ILoss of Charging 11 S2.0P-AB.CVC-0001 I ii 119 I I II I II II I I II l q 11 I \Lo. Numbe¥ *** Objectives I PZRP&LE015 ,_ ____,
I [g~~~ticm sou~g~n 11 Facility Exam Bank 11 Used Duriri~irr~fni~g Progran( I D
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!9ij~,~!~ersouf~~~s9 ~m,e,ntsj 1125676 J
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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I
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lauestionTopic 11RO41 I Which of the following describes the interface between the Reactor Trip Handles on 2CC2 and the Reactor Protection System? Turning either Reactor Trip Handle to the trip position is designed to operate I
~ the UV trip ONLY for Reactor Trip AND Reactor Trip Bypass breakers.
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~ the shunt trip ONLY for Reactor Trip AND Reactor Trip Bypass breakers.
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~ BOTH the shunt trip and UV trip for Reactor Trip AND Reactor Trip Bypass breakers.
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@J IBOTH the shunt trip and UV trip for Reactor Trip breakers, and the shunt trip ONLY for the Reactor Trip Bypass breakers ..
I [Answer 11 c I IExarn Level j j R I !cognitive Level i IMemory I !Fa'sility: l ISalem 1 & 2 I (Exam bate:' 11 12/21/20151 IKA:ll 012000A401 1IM.01 !jRoyalue:1j 4.5jlsROValue:ll 4.5jlsection:ll~1RoGroup:jj 1ll5R6Group:/j 11 1~~1j D jSY5tern/Evohition Title j j_R_e_a_ct_o_r_P_ro_te_c_ti_o_n_S..:..y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __, ,_0_12_~ IKA Staternent: I Ability to manually operate and/or monitor in the control room: I Manual trip button I f Explan~tion of j Salem has 2 Reactor Trip Handles/Switches. Either switch operates BOTH the UV and shunt trips for BOTH the reactor trip ,Answers: . 1 breakers and bypass breakers. The distracters are plausibler because: 1) an automatic reactor trip ONLY actuates the UV trip; 2) manually tripping the reactor trip breakers from the control console ONLY actuates the shunt trip. I Reference Title >j;,, Facility Reference Number.. 1~Reference Section
I 1 Page No. i !Revision!
IReactor Protection System Reactor Trip Signals !I221051 II 11 1113 I I II ll II 1 I I I II !I II II I iL.0. Number
Objectives IRXPROTE010 IRXPROTE007 IJVlaterial Required for EJ<amin.ation II lj 1auestion ~6Jrce:.oj j_N_e_w_ _ _ _ _ _ __.l lauesti()!l Modificati()n.M~thog: **** 1--------~I !Used [)uring Training Program j D
!,Question ~o.4fce Comments\ I I
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RO Skyscraper SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes f louestion Topic !R042 I IWhat would be the effect on the Reactor Protection S}'.stem if the 2B Vital Instrument Bus were to become deener~ized with the unit at 100% power? I I
~ SSPS Train B slave relays would not actuate on a Safety Injection signal.
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~ OHA A-34 SSPS TRNA TRBL in alarm due to loss of 1 of 2 45VDC power supplies to Train A logic cabinet.
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~I Logic coincidence for Containment Spray actuation would go from 2/4 to 1/3 due to channel II bistable tripped.
I []I 2RP4 bistable lights flashing for all channel II indications due to train disagreement between SSPS Trains A and B. I !Answer 11 a I IExam Levell I R I :cognitive Level 11 Application I !Facility: 11 Salem 1 & 2 1 IExamDate: 11 12/21/20151 IKA:!I 012000K201 iIK2.01 I !Ro Value: JI 3.3lfSRO Value: !I 3.71 lsection: II~ (RO Group:H 1 j ISRO Group:ll 11 l!B D !System/Evolution Title 1 l_R_e_a_ct_o_r_P_ro_te_c_ti_o_n_S"'"y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~l lo12 IKA Statement: I Knowledge of bus power supplies to the following: RPS channels, components. and interconnections I
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[Expl<1ryauon of 55.41.b(?) B is incorrect because Train A 45 VDC power comes from A and D vital power. C is incorrect because CS bistables are
Answers: 1 energize to actuate. D is incorrec because there is no disagreement since none of the slave relays have energized with the other train ot having energized.
Reference Title'. * . 1.1* Facility Reference Number
i!Reference Section******* .
,. .. . . .* .. Jf Page No.j IRevisionj IOverhead Annunciators Window A "S2.0P-AR.ZZ-0001 I 11 I 156 I ISolid State Reactor Prat Train A AC Power Distr 11 Drawing 240136 I 11 114 I I Ii i 11 11 I
!Lo. Number ** 1.*,. Objectives I RXPROTE011 I RXPROTE020 1_ ____. !Material Required for Examination J I Ij IQuesti?n Scnfrce:.>11 Facility Exam Bank 1IQue;;tion~.9~ificatl?nMethod: i Direct From Source I !Used During Trainirig Program Dl iQuesti~11;>oufce ComT~nts[ I
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Used on Salem June 2004 NRC exam fcOmment * **.. *.1 I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I
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Outline Changes I f Ouestion Topic ! IRO 43 I Given the following conditions:
- Unit 2 is operating at 100% power when a turbine trip event and subsequent switchyard disturbance result in an undervoltage condition (<70%) on the 2A & 2C 4160 V vital buses. - 2 minutes later, a Safety Injection signal is generated.
Which of the following describes the plant response when the SI occurs? I
~ EDG output breakers remain shut and safeguards loads sequence on each vital bus.
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~ EDG output breakers remain shut, blackout loads are stripped, then safeguards loads are sequenced on each vital bus.
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~ EDG output breakers open, blackout loads are stripped, EDG output breaker shuts, then only safeguards loads are sequenced on each vital bus. I Id. IIEDG output breakers open. blackout loads are stripped, EDG output breaker shuts, then safeguards and blackout loads are sequenced on ~ each vital bus. I
!Answer i c I I tExam Level 11 R I ;Cognitive Level *. I IApplication I iFacility: 1ISalem 1 & 2 I 'Exa.mDate:.1 j 12/21/20151
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~j013000K112 llK1.12 liROValue:ILlJ}!SROValue::Glsect1on:!j~fROGroup:jLJ:SROGroup:,LJ '1!1~ D !system/Evolution Title I I Engineered Safety Features Actuation System I i013 I IKA Statement: I Knowledge of the physical connections and/or cause-effect relationships between Engineered Safety Features Actuation System and the following: ED/G Explanation of I 55.41.b(?) 2/3 4KV vital bus UV causes ALL 3 4KV vital buses to load in SEC MODE II BLACKOUT. This mode starts ALL EDGs Answers:
J and sequences BLACKOUT loads onto ALL vital buses. When the SI occurs, the SEC initiates a MODE Ill, which opens any running EDG breaker, strips whatever loads are energized, then sequences SAFEGUARDS loads onto ALL buses. Distracters are plausible based on determining the 2 buses 2A and 2C load individually in BLACKOUT based on the UV on those 2 buses only.
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Reference Title I .* . Facility Ref~rence Numbef i !Reference Section I1Page No:l IRevisiOnl ISafeguards Equipment Controller Lesson Plan 11 NOS05SECOOO II 1119 l I6 I
!I u II 11 I II n ii !I I
[L~O. Number Objectives IESFOOOE005 I SECOOOE010 1--~ IMater,iaL~equired for ~iirni~.a~ion
1 I Ij t9uestlon Source: j IFacility Exam Bank I 'I
[<luestion N,l~~ification Method: . Concept Used I !:us~d During Training Program I D [aui:;stion source Comtp~ll.ts!
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Outline Changes I iauestion Topic ! I RO 44 I A loss of off-site power resulted in a reactor trip. The operating crew is attempting to confirm natural circulation flow but neither SPDS NOR the Plant Computer is operating. You are the 3rd NCO and have been assigned to monitor and log CET's at the CET Control Panel. Which of the following describes the "ALL" Mode at the Train A GET Control Panel? It displays ......
~I Train A CET temperatures in sequential order.
I [bl ITrain A CET temperatures from lowest to highest reading. I
~I any Train A CET >700°F, then remainder of Train A CETs from lowest to highest.
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@] Ithe two highest reading Train A CET's in each quadrant then sequentially display all Train A CETs.
I [Answer l Id I tExam Level 11 R I !cognitive Level I !Memory I IFacllity: / ISalem 1 & 2 I IExamDate: J j 12/21/2015j ~I 017000K402 r--:-i IiK4.02 I!RO Value:: [DJ !SRO Value: IG]" :Section: ll~ fRO Group:; i LI 'SRO Group:I LI w; iii~ D
'-l"-l

system/Evolution Title! !in-Core Temperature Monitor System II~0_17_ _

KA Statement: I Knowledoe of In-Core Temperature Monitor Svstem desion feature(s and or interlock(s) which provide for the following:
Sensinq and determination of location core hot spots i \Expl;:mation of 55.41.b(?) Table C of CFST-1 states that in ALL Mode the display will progress through the first and second highest CETS in each Answers: *' quadrant, then sequentially display all cETs assigned to that channel. A is incorrect because does not display in sequential order without first displaying the 2 highest in each quadrant. B is incorrect because it doesn't display lowest to highest. C is incorrect because the 700°F noted in choice is criteria for purple path CFST for Core Cooling. l .. Reference Title .
\' Facility Reference Number *. ! \Reference Section. *** 11 Page No.1 IRevisionl f Critical Safety Function Status Trees 112-EOP-CFST-1 I q 15 1125 I I II I 11 11 I I II I 11 11 I li.:~o. Number**
Objective~ I INCOREE016 1_ __ , IMaterial Required for EJ<ainination ' I I II (Question S~iJri::e: 11 Facility Exam Bank I I[ouestion Modifi~ation Method:.: Editorially Modified I:lised During Training Pfo9ram j D
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3 - 1 6 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' '
lc<>inment
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RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes iouestioh Topic I IRO 45 Given the following conditions:
- Unit 2 is operating at 100% power. - A Main Steam line break inside containment occurs. - The Rx was tripped, but the MSLI failed to shut ANY MSIV. - Safety injection was initiated successfully. - Containment pressure has risen steadily and is currently 11 psig. - RCS pressure is 1700 psig and lowering slowly. - Safeguards reset actions have just been completed in 2-EOP-LOSC-2, Multiple Steam Generator Depressurization.
Which of the following describes how the Containment Cooling system will respond if containment pressure were to rise to >15 psig? I
~ Both Containment Spray pumps will start and Containment Spray valves will reposition. 'b.l INeither Containment Spray pump will start, and Containment Spray valves will reposition.
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~ Both Containment Spray pumps will start, and Containment Spray valves will NOT resposition.
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~ Neither Containment Spray pump will start, and Containment Spray valves will NOT resposition.
[Answer i IE__J !Exam Level: I~ 1cogmt1ve Level.
.. 11 Application 'llW' I lfac1hty:n ...
Salem 1 & 2 llExamDate: Il___12_12_1_12_0_1__,5i ~I 022000A301 I[A3.01 ; !RO Valµe; 1 [ITI 'sRO Value: I@[section: 11~ jRO Group:! LJ !SRO Group:I LJ D \system/Evolution Title I IContainment Cooling System ;022 iKA Statement: I Ability to monitor automatic operations of the Containment Cooling System including: Initiation of safeguards mode of operation fExplan~ti?n ofl 55.41.b(9,7) Containment Spray Pump Sequencing Answers: 1)If the SSPS Containment HI-HI Pressure signal is not present when the SEC initially tries to start the Spray Pumps, the SEC contact will re-open 2)0nce the SEC has completed the last step of its loading sequence, the CS Pump start contact is re-closed
-"'l-IU-l" r,., ~. **- **~
r"C' 0 ........... ,... .... n1 ...... .J..... - . ..., ...... Ib)lf the SEC has been reset the CS Pumps will NOT respond to a HI-HI Containment Pressure until the SEC is again actuated H f::".acility Reference Number* *** I!Reference Section
I[PageNo.\ (Revision[
I i . Reference Title ISafeguards Equipment Controller Lesson Plan 1INOS05SEC000-06 I 1117 11 6 I ! !I I II lI I I II I 11 11 I il.O. Number > * **.
i Objectives I CSP RAYE009
'--~ l~aterial Require~ for,Exami.gation). J I II [ouestior't sollrce,: . f l_N_e_w_ _ _ _ _ _ _..I f otiestif>n Modifit:~tiori ~ethp9: >] _________.! iusei:I Dµririg Training Prograrn I o iollesticmSource co~nierit~! I I
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I RO System/Evolution List I SRO System/Evolution List I Outline Changes I tOuesticm Topic I lRO 46 I Given the following conditions:
- Operators are responding to a LBLOCA IAW 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation. - RWST level is 8.9' and lowering as expected.
Which of the the following describes the effect if 21CS2 Containment Spray pump discharge valve experienced a short and motored closed? Containment Spray Header flow will. ..
~I lower to O gpm.
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~I lower but remain> 0 gpm.
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~ be unaffected since 21 CS36 RHR CS STOP VALVE is open supplying all spray flow I
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~I be unaffected since 22CS36 RHR CS STOP VALVE is open supplying all spray flow.
I !Answer j a I I !.Exam Level ; IR I !Cognitive Level J IApplication I IF:acility: 11 Salem 1 & 2 I !ExamDate: 11 12/21/20151 [KA:l I026000K302 I!K3.02 i IRb Valu~: 114.2*1 :SRO Value::! 4.31 lse,ction: II~ IRO Grc>up:ij 11 !sROGrolli:>:ll 11 !lf~ D lsysteffi!Evolution.Title I j_c_o_n_ta_in_m_e_n_t_S..:..p_ra"""y_S-'y'-s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __,11026 [KA Statement: I Knowiedoe of the effect that a loss or malfunction of the Containment Sprav Svstem will have on the followino: Recirculation sprav svstem !Explanation of I 55.41.b(S,10) When RWST level reaches 15.2', operators will beging alignment to Cold Leg Recirc IAW LOCA-3. 22 Containment Answers: : Spray pump is stopped first if both Containment Spray pumps are operating. With the stem condition of current RWST level, 21 CS pump will still be running. When RSWT level reaches lo-lo setpoint, the remaining CS pump (21) will be stopped, and 21CS36 opened to supply recirculation spray flow. I I Reference Title ll Facility Reference Number. I[Reference Section .* I!Page No.j jReyisiC>ni I Transfer to Cold Leg Recirc 112-EOP-LOCA-3 I q 1129 I I II I 11 11 I I 11 I ii 11 I Objectives ILCA3U1E006 ,_ ____, jMatetial Required for Examination
I I 11 f(lliestion Source: 1j_N_e_w_ _ _ _ _ ___.l l<ltte,stiop.1V19pificatiC>niVtethod: .*** ~ _ _ _ _ _ _ _ __.I fused During TrafriingPrograrn.i D l(;lue,s.tion sl)urcfComments 11 I
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RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution Li~ Outline Changes iauestidnJopic 11RO47 I Given the following conditions:
- Salem Unit 1 was operating at 100% power when a LOCA occurred. - A manual reactor trip and manual SI were initiated. - When the Main Generator output breakers opened, a loss of off-site power occurred. - 1A vital bus locked out on bus differential.
Which of the followinq identifies which Containment Iodine Removal Units (IRUs) can be started if required? 1.:11
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I f!U 112 IRU ONLY. I
~ 111 or121RUs.
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.d.11 NEITHER IRU is available.
I !Answer i Id I !Ex,am t:evel 11 R I !cognitive Level 11 Memory I IFaciHtY: 11 Salem 1 & 2 1 IExa!TIDate: 1 I 12/21/20151
!R.O Value: I[Dj' ,SRO Value: III£]' !section: II~ /RO Group:j LJ [SR,O Group:j LJ <
~I 0270001<201 I'K2.01 i ~ .' D ISYstem/Evoiution Title I I Containment Iodine Removal System ! 1,027 I IKAStafoment: J Knowledqe of bus power supplies to the followinq: Fans I IExplanatio~of 1 55.41.b(9) Containment IRUs are powered from G and E non-vital 460VAC. With the loss of off-site power, none of the non vital Answers: *.* 1 busses are energized. The distracters are based on the operator knowing that the loading of a vital bus in Mode IV doesn't have any bearing on !RU operation. ! \ ;"/ .**.. Reference Title 11: ** FacilityReference Number
I!Reference Section. 11 Page No. j !Revision\
!1E1 Aux Building 460-230V One line 11207916 I 11 1126 I !1E1 Aux Building 460-230V One line !1207919 I q q23 I I II I II 11 I !LJ>. Number* Objectives I CONTMTE003 ,_____, !Material Required for Examination
J I toues~i~n Source: 11Previous2 NRC Exams I 1o~estio~Modif!?atioriJV1ethod; \1 Direct From Source I !used During Traii:iing Program l D
!Question Sourc~;Somm~n~sj 1_1_1-_0_1_R_o_N_R_c_a_4_7_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ [comment.* II
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SRO Skyscraper I RO System/Evolution List f SRO System/Evolution List I Outline Changes 1ouestionTopicJ ,_R_0_4_8 ~---------------------------------------------~! _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~ Given the following conditions:
- Unit 2 has experienced a Large Break Loss of Coolant Accident (LOCA). - Containment H2 concentration has risen to 2%. - 21 H2 Recombiner has been placed in service with containment pressure at 4.1 psig. - 24 hours later, containment H2 concentration has remained at 2%, and containment pressure has risen to 5.1 psig.
Which of the following describes the effect the higher containment pressure has on H2 Recombiner operation, and how should the crew proceed? H2 Recombiners are _ _ effective as containment pressure rises with a constant power setting. The power setting must be ___ IAW S2.0P-SO.CAN-0001, H drogen Recombiner Operation. ----------------------------- la'.J less, raised.
~ more, raised. ~ less, lowered.
fd.: more, lowered L:...___J !Answer I r&:J rExam evel i rL.J' !cognitive Level I! Application 1 IF~cility: HSalem 1 & 2 l~ExalllDate:H _ _ _12_1_2_11_20_1_5~I IKA:!j 028000A201 !IA2.01 ilROValue:i[ill[sROValue:l§isection: ii~ IRO (3roup:1LJISRO Group:ILJ D [System/Evolution Title I IHydrogen Recombiner and Purge Control System [KA Statement: I Ability to (a) predict the impacts of the following on the Hydrogen Recombiner and Purge Control System and (b) based on those predictions, use procedures to correct, control, or miti ate the conse uences of those abnormal o eration: fExplanation of j 55.41.b(?,8, 10) H2 re~ombiners are P.laced i~ service when directed in the EOPs with containme~t H2 concentration between 2-~%. 1 AnsWers:
_- ** : The hydrogen recomb1ners use electric heating elements to elevate the temperature of the containment atmosphere. As shown 1n Attachment 2, RECOMBINER POWER CORRECTION FACTOR CURVE, a higher containment pressure would result in a higher power correction factor, which would cause recombiner power setpoint to rise. So initially the higher pressure would cause the
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I Reference Title Ii ' Facility Reference Number> i !Reference s~ction JI Page No: I iReylsion' IHydrogen Recombiner Operation II S2.0P-SO.CAN-0001 I 114,7,8 1I9 I I II I 1I I I II I "II 11 I !LO.Number Objectives I CONTMTE012 ,_ ____, !MaterialR'.~qufrea fc)r¢can}in'ation , l IRO 48 S2.0P-SO.CAN-0001 (because KJA says so!) II !Oue~tion Source: 11 Facility Exam Bank I Jl.ised outing "Tr'.a'inin9 Pr,ogram l D [a~_estion Sourc,e Comments! 180552 I
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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question Topic! IRO 49 Given the following conditions:
- Salem Unit 2 is 7 days into a refueling outage. - The core is partially offloaded with 7 bundles remaining in the Rx. - Fuel movement is in progress, and S2.0P-IO.ZZ-0010 SPENT FUEL POOL MANIPULATIONS is in effect. - SFP temperature is 120 deg. F. - 21 SFP becomes air bound, trips on motor OL, and can NOT be restarted. - 22 SFP pump will NOT start. - SFP hi level alarm is in alarm. - SFP heatup rate is 12 deg F/ hr.
If SFP cooling can NOT be restored, which of the following choices describes an adverse consequence of this Joss of Spent Fuel Pool cooling?
§:] IRx cavity overflow due to rising SFP level if the Gate Valve remains open.
[]I Increased production of radioactive waste liquid as the tell-tale drains flow rises to the FHB sump. I
~ Increased radiation levels at the FHB charcoal filter due to Spent Fuel off-gassing at temps> 150 deg. F.
I [d. ! Inability to place a raised Spent Fuel bundle into any location in the pool due to rising radiation level on 2R32 Fuel Handling Crane Area L...:..J Monitor. B I [Answer: I~ :Exam Level !~ 1cognit1ve Level 11 Memory .. r ISalem 1 & 2 I !Facll1ty: I IExamDate: ! ,___12_12_1_12_0_1_..51 ~I 033000K303 ! \K3.03 \ IRO Value: i 12QJ:sRO Value: !@~section: q~ !Ro Group:\ [JJ\sRO Group:! LJ ~~11'. D fsystem/Evolution Title I I Spent Fuel Pool Cooling System I !033 I IKA Statement: I Knowiedge of the effect that a loss or malfunction of the Spent Fuel Pool Cooling System will have on the following: Spent fuel temperature I !Explanation of I 55.41.b(13, 10) Distractor b is incorrect because the pool would overflow into the ventilation system, not come over the physical wall [Answers: of the pool. C is correct because rising radiation will be seen as fuel off-gassing and is expected to occur as temp increase to 150 deg. Distractor a is incorrect because any overflow will go out the ventilation openings in the SFP. Distractor d is incorrect because it is always possible to lower a SF bundle. I
Reference Title 1: Facility Reference Number I!Reference Section Ii Page No: I !Revision:
ILoss of Spent Fuel Cooling !Is2.0P-AB.SF-0001 II 11 1112 I II II II 11 I I! !I !I 11 I fL~O. Number Objectives !SFPOOOE007 '--~ IMc;iterial Reqllired for Exa~i!l,ation Ii 11
!Question Source: 11 Facility Exam Bank l IQuest\~n Modifiqatio,n Method:. ;::I Editorially Modified I iUsed Puring Training Program .1 D !Question Source Comme~ts[ I Used on Salem June 2004 NRG RO exam (6 exams ago)
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1aues~ion Topic] I RO 50 I Given the following conditions:
- Unit 1 is operating at 95% power performing a 1% per minute load reduction due to high radio frequency in the Main Generator. - 11 SG Narrow Range Level channel II is undergoing a channel calibration IAW S1.IC-CC.RCP-0035, 1LT519#11 STEAM GENERATOR LEVEL PROTECTION CHANNEL II, and all associated bistables are tripped.
Which of the following identifies the consequence if a second channel of Narrow Range level on 11 SG were to fail to 30% with NO operator action? 11 SG level will become ... la. c__j I I higher than program because 11BF19 and 11 BF40 ONLY swapped to manual and will be over feeding 11 SG. I I
,b. i lower than program because 11BF19 and 11 BF40 ONLY swapped to manual and will be under feeding 11 SG.
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'c.
L_j I l higher than program because both SGFPs and 11BF19 and 11 BF40 swapped to manual and will be over feeding 11 SG. I I [d. ! lower than program because both SGFPs and 11BF19 and 11BF40 swapped to manual and will be under feeding 11 SG. I
Answer i I a I !Exam Level 11 R I !cognitive Level 11 Application I wacillty: i ISalem 1 & 2 I !Examoate:'! I 12/21/20151
\KA: I035000K603 1 i IK6.03 jROValue:il 2.6jlSROValue:jj 3.oj[sectiol'l:lj~tROGroup:il 2psROGroup:lj 2j L~~~ D jsystern/Evolution Title j j_s_te_a_m_G_e_ne_r_a_to_r_S..;.y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.I ~I fKA Statement: I Knowiedge of the of the effect of a loss or malfunction on the followinq will have on the Steam Generator Svstem: SIG level detector I !Explanation of I 55.41.b(?) SG NR level is programmed from 33-44% up to 100% power. As the downpower continues, the BF19 will be closing in [Answers: I response to the lower steam flow requiring less feed. The 11BF19 and 11 BF40 (expected to be shut at this power level) swap to manual upon the second NR level channel failure. The SGFPs do not. When the 11BF19 swaps to manual, it will have a certain demand on it. As power (and steam flow) continues to lower, the demand will be higher than required, and SG NR level will l... ............ """'o. i....~ ..... i....,... .. .i.a... ......... n'"'"'"1r"-:arn 1,..,,,,....1
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I Reference title * . . . I . Facility Reference Number ItReference Section . I* Page No. j [Revision! IOverhead Annunciators Window G !IS1 .OP-AR.ZZ-0007 I 1111 1142 I I I I 11 11 I I 11 I 11 11 I [1.,.:0*. Number< I CN&FDWE004 Objectives 1_ ___. fMaterial Required for Exarninatiort' II II 1auestion source:> 1l_N_e_w_ _ _ _ _ _~I '-i.a~u~e~s_,_ti__:occ.n*__:M~o~d:.c:..if_ic...ia~t~io:::.n:::.**M-'--.-..::et-..::}l__:o_d__c:-..::->Jjl1_________.l lus~C1'6uring :i:~aining Progsarii Io iauestion Source <:;or.rtmentsi I
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Outline Changes I .. jauestionTopic I I RO 51 I Salem Unit 2 is operating at 95% when the PO reports steam flow is rising on all SGs with no readily apparent reason. Which of the following describes how the steam flow will affect the plant, and how operators should proceed if steam flow continues to rise uncontrollably IAW S2.0P-AB.STM-0001, Excessive Steam Flow? Reactor power will rise at 1) rate as the increased steam flow. The crew will 2)
~ 11) a higher
2) initiate a MSLI to determine if a safety injection is required. I
[gJ 11) the same
2) Trip the reactor and confirm the trip, then initiate a MSLI to determine if a Safety Injection is required. I fc.! 11) a higher
~ 2) Initiate a power reduction to ensure reactor power remains <100% while attempting to identify and isolate the leak. I fci.l 11) the same t-=.:.J 2) Initiate a power reduction to ensure reactor power remains <100% while attempting to identify and isolate the leak. I jAnsW'er 11 b I iExam Level ! IR I !Cognitive L._evel ** i I Comprehension I [Fa~Uity; 11Salem1 & 2 I~ I 1212112015!
jKA:il 039000A205 jjA2.05 I!Ro Value: II 3.31 lsRo Value:ll 3.61 [section: II~ )Ro Group:ll 11 jsRo Group:;! 1j Ill D !system/Evolution Title I j Main and Reheat Steam System I row=-i IKA Statement: l Ability to (a) predict the impacts of the following on the Main and Reheat Steam System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: lncreasinq steam demand, its relationship to increases in reactor oower fExpl<imat'.On>~! 1 55.41.b(1,5, 10) AB.STM does direct a power reduction at Step 3.8 if the EHC system is not causing the turbine to be the source of 1Answers. *' I the rising steam flow. The stem states that there is no readily apparent reason for the rising steam flow, and EHC malfunction would be apparent. Also, the Continuous Action Summary is in effect at Step 1. CAS Step 1.1 states that at any time if reactor power is rising uncontrollable, trip, confirm, initiate MSLI. If souce of steam leak is isolated, then go to TRIP-1. If not, initiate SI and
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, i * ** j Page No, I [Revisloni IExcessive Steam Flow ii S2.0P-AB.STM-0001 I 11 11 9 I I II I 11 11 I I II I II 1I I IL.o:fllutnber< . I Objectiv I MSTEAME008 I IABSTM1E001 I
, __ ___. IMaterial ~equired fofExamfriation II II jOuestio.~s~urce: ..*J l_N_e_w_ _ _ _ _ _~I Lia_u_e_s_:ti_:o21}.~M-'o_d~If_:ic_:aLt~io~n_M_** _etc::*~L<>L9L:*_:.* .'-'i,1 ________~1 luse<(buring Trainjng Program .j O
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I SRO System/Evolution List I .... Outline Changes I (auestion Topic 11 RO 52 I Given the following conditions:
- Unit 2 has experienced a loss of all feedwater flow initiated by a Feedwater Isolation signal (P-14) condition on 23 SG. - To mitigate the event after the P-14 has cleared an NCO has been directed to start 21 SGFP in accordance with S2.0P-SO.CN-0007, Rapid SGFP Recovery. - The NCO successfully relatches the 21 SGFP but the speed of the pump does not rise automatically to minimum speed as he anticipated.
Which of the following is the cause of this response?
~ IThe P-14 signal "seal-in" feature.
I
~ 121 SGFP PUMP SPEED CONTROL is in AUTO. ~ 121 SGFP ENABLE/DISABLE switch is in the DISABLE position.
I I [ ] 121 SGFP speed was >160 rpm when the latch push button was depressed. I [Answer J Id ! iExam Level 11 R I !cognitive Level 11 Memory ! IFacility:*l ISalem 1 & 2 1 IExanioate:' 11 12/21/20151 !KA: iI059000A107 !IA1.07 llRo Value:ll2.5:J !s~o vciiue::j2.6*l lsectfon: II~ !RO Group:!! 11 lsRq.Group:!I 11 falil D j.Systeni/Evolution Title l j_M_a_in_F_ee_d_w_a_te_r_S-'y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.l 1059 !KA Statement:! Ability to predict and/or monitor chanqes in parameters associated with operatinq the Main Feedwater System controls including: Feed Pump speed, includinq normal control speed for ICS !Explanation of I 55.41.b(4,10). A is incorrect because the P-14 signal automatically clears when the SG level lowers <setpoint, there is no seal in ,Answers: ..*.*. ** j circuit. Bis incorrect because auto speed control prevents latching of the SGFP. C is incorrect because the Enable/Disable switch in the disable position only removes the ADFWCS from controlling SGFP speed, the switch is placed in Disable when starting the SGFP. Dis correct because the SGFP may be latched with speed <160 rpm, but it will not automatcally raise speed to minimum 111nn - - - ;~1~ -- --'\
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.* Reference Title Ii *.. f=acility Reference Number J [Reference Sectio(l. ]IPage No. [ !Revision[
ISGFP Prompt Recovery 11 S2.0P-SO.CN-0007 I ii 6 114 I I II I !I 11 I I II I ll 11 I !Lo. Number
I SGFPLOE006 Objectives I
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/Material Required forExamination)*i l lj jauesfion Sol.ire~: 11 Facility Exam Bank l ldu'estion Modific(itiqn>Metpod: .. Editorially Modified I I[used DuringT~ciining ~rograrri I D I -;::======-=========::::::::::::::::::::::::::::::::::::::~===========-================-~t rauestion .source Commehtsj ~4112. _Removed second part of question which asked what you had to do to get speed to raise. Replaced 1mplaus1ble distracter.
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I SRO System/Evolution List I Outline Changes J.... fQuestionTopic I IRO 53 I I With both 11 and 12 AFW pumps in service providing 6E4 lbm/hr flow to each SG, what would be the response if both 13AF21 and 14AF21 were shut! fully?
~ 111 AFW pump automatic recirc valve would open.
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~I The single automatic recirc valve for the MDAFW pumps would open.
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\c. ! IThe Pressure Override Circuit would actuate to prevent dead heading 11 AFW pump.
I
@] I Normal MDAFW pump recirc flow would rise through the orificed continuous recirculation line.
I JAnswer ! ja I lexam Lev~rj j R I !cognitive Level .11 Application I lf=aci!jty:' 11 Salem 1 & 2 I !Exarnbate: 11 12/21/20151 y !KA:lj 061000K503 1IK5.03 I:RO Value: II 2.6j !SR,O Value: :12.9*1 ISection:*.!j~ lRO Group:ll 1I !SRO Group:lj 1 I l!E ' D I
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system/Evolutio*n Title. i Auxiliary I Emergency Feedwater System I i061 I

KA Statement: Knowled e of the o erational implications of the followin Feedwater S stem:
Pump head effects when control valve is shut iExplanation ofj 55.41.b(7,8) The MDAFW pumps each has its own dedicated recirc line and associated automatic recirc valve, which opens to
.Answers

** *
1 maintain aFW pump flow >180 gpm. With the 6E4 lbm/hr in stem for each of the 2 SGs being supplied from 11 AFW pump that =
240 gpm, the valve would initially be closed and open when flow lowers <180 to prevent the increased pump head from causing pump damage from overheating. B is incorrect but plausible if it is thought there is a common recirc line. C is incorrect because
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i. ,* **, '* *.Reference Title ,'. :h,' . Facility Reference Number IAuxiliary Feedwater System Lesson Plan II NOS05AFW000-14 I ! j 19-20 1 I I II I 11 1I I II I II 1I l lL.o; Number* Objectives I AFWOOOE008 1_ ____,
[llllaterial Requir~d for Examif1atror1;i
II 11 i<:llle~tion S9uice:.A l_N_e_w_ _ _ _ _ __,l [Q~~S,~ion "19clification~etlloCI:: J_________.I [used During Training Program ID fal1estioo,sourceCo.mrn~nts] ' - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -.....!
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!Question Topic j RO 54 I Given the following conditions: - Unit 2 is performing a normal shutdown at 20% per hour IAW the IOP's. - 22 SGFP is shutdown. - Current Rx power is 24%. - 21 SGFP trips.
With NO operator action, what will be the status of the Auxiliary Feed Pumps?
~I ONLY the MDAFW pumps start immediately upon the trip of 21 SGFP.
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~I ONLY the MDAFW pumps start when SG NR level in 1/4 S/G's lowers to 14%.
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~ The MDAFW pumps AND the TDAFW pump start immediately upon the trip of 21 SGFP.
I [] I The MDAFW pumps AND the TDAFW pump start when NR level in 1/4 S/G's lowers to 14%. I IAnsw~r 11 a I !exam Level 11 R I ;cognitive Level J IApplication j jFaciliW: 11 Salem 1 & 2 I lExamDate: 11
'. ..* >,, '
12/21/2015!
Q]" !s~o Value: [Ej !Section: II~ [RO Group:! LJ !SRO Group:jLJ g
;w. ~I 061 OOOK602 " '~ , .~
11 K6.02 ' I ,Ro Value: I
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D [System/Evolution Title J IAuxiliary I Emergency Feedwater System j 1051 i IKA Statement: l KnowledQe of the of the effect of a loss or malfunction on the following will have on the Auxiliary I Emergency Feedwater System: Pumps fExplaf!~tion of j 55.41.b (4) MDAFW pumps auto start when both SGFPs are tripped as shoen on logic drawing 221064. The TDAFW pump does Answers: .*. not. The MDAFW pumps also auto start on 2/3 NR level channels in one SG lowers to 14%, but in this case they will already be running. The TDAFW pump starts on 2/3 NR level channels on 2/4 SGs.
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*F,acility Reference Number\ , /~eference Section*. J! P~ge No../ IRevision!
IRPS AFW pumps startup 11221064 I II lj a I IAFW System LP II NOS05AFW000-14 I lj 33 1114 I I II I ii II I
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Objectives AFWOOOE006 IMaterfalRequired td'r Examination. j I II
!9u~~tion sou~c~: 11 Facility Exam Bank 1JQue~tiof!Modification Metho.CI:\ I Editorially Modified I IUsed Du.ring Training PrograrilJ D
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Outline Changes I.... [Question Tbpic I IRO 55 I Given the following conditions:
- 2C Emergency Diesel Generator (EOG) is operating in parallel with the 500KV grid for a 24 hour endurance run IAW S2.0P-ST.DG-0014, 2C DIESEL GENERATOR ENDURANCE RUN, following a complete overhaul. - Cumulative run times for all individual EOG load limits are less than 10% of rated. - While operating at 2525 KW three hours into the test, the operator mistakenly adjusts 2C EOG speed control resulting in MW loading rising to 2610 KW.
What are the consequences, if any, of continued EOG operation at this KW load? l Operation for the remainino 21 hours of the test ...
~ will not have any adverse effect on 2C EOG.
I I
~ will result in exceeding the 30 minute load limitation for 2C EOG.
I rs I will result in exceeding the 2 hour load limitation for 2C EOG. I
@] Iwill result in exceeding the 24 hour load limitation for 2C EOG.
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.la! 'Exam Level .
I !RI !Coanitivel,.vel I IAnnlication I IFaci}ity: I Salem 1 & 2 IJ~~mOate.:*W 12/21/20151 !KA: Ij 062000A 101 I~jA_1._01_ _1JRO Value: II 3.41 Jsifo Value: 11 3.81 iSectiori: 11~ [RO Group:! I 11 JsRO Group:/ I 1I D [System/Evolution Title I l_A_.c_._E_l_ec_t_ric_a_l_D_is_tr_ib_u_ti_o_n_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.j J062 !KA. Statement: I Ability to predict and/or monitor chanc::ies in parameters associated with operatinq the A.C. Electrical Distribution controls includinq: I Significance of DIG load limits I I i~planationgt 55.41.b{8) The EOG load limitations are maximum of: 2600KW continuous, 2600-2750KW for 2000 hours, 2750-2860KW for 2 tAnswers: * * .. hours, and 2860-3100KW for 30 minutes. With the EOG operating at 2610KW for 21 hours, the EOG will not exceed any limits, operation between 2600 (cont) and 2750(2000 hours) KW, since the stem stated that the cumulative run time for ALL EOG load limits was <10%, which would be 210 hours for this limit. t Reference' Title *
11 *Facility Reference Number *** J[Reference ~ection

I 1Page No:'.1 IRevisioni j 2c DIESEL GENERATOR SURVEILLANCE TE s2.0P-ST.DG-0003 II II P&L 3.5 II l j52 I I !I II I[ 11 I I II II 11 11 I
[Lo. Nµrnber I EDGOOOE012 Objectives I 1_ ____. iMafr~rialRequired for Exalllinatlon;* '. j I Ij I [ouesti9:'.n Source:
j Facility Exam Bank 1lOue~tionModification Meth~d:' J Editorially Modified I JUsed DufingTl-aiiling Progrl:iin.j D
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fOuestio~ s(')IJrce Commenti'ii 163989 replaced 2000 hr distracter (implausible) with 24 hour distracter I fcommenf'
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lauestionTopic 11 RO 56 I IIAW Salem FSAR Section 8.3.3.2, Station Batteries, on a total loss of all AC power the station vital 125 voe batteries are designed to supply vital station loads for a minimum of hours. I nl
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[d.jl8 I [Answer 11 b I IExam Level J IR I !cognitive Level 11 Memory I l~acility: 11 Salem 1 & 2 I !ExamDate: 11 12/21/20151 !KA: 11 062000K303 IiK3.03 liROValue:lj 3.7jlsROValue:ll 3.9llsectioll:lj~i~OGroup:1I 11Js~0Gr,oup:il 1j ~lilJ D !System/Evolution Title] j_A_.c_._E_l_ec_t_ric_a_l_D_is_tr_ib_u_ti_o_n_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.l ID62 !KA Statement: I Knowledge of the effect that a loss or malfunction of the AC. Electrical Distribution 'Nill have on the follo'Nin~i: DC svstem IExplanatio~ pf'. 55.41.b(?,8) IAW Salem FSAR Section 8.3.3.2, Station Batteries, states ... "The batteries are sized for 2 hours of operation after a Answers: . * . ; ! loss of ac power, based upon the required operation of the de emergency equipment." 1 i .Y **
Reference Title I'\ .*.**Facility Reference Number**** *. !!Reference Section 11 Page No. I IRevisionl
!Salem FSAR I 118.3.2.2 118.3-18 1114 I I I Ii 11 11 I I l II ii !I I I.Co; Nurrib~r. *...*. Objectives I DCELECE002 '--~ llVlaterial Required for Examination *~* I I 11 fou~~ti~ll Source: 11 Facility Exam Bank I (ques,ti?n Mpdifi~~tion Meth4ct: <iJ Direct From Source I1.Used During Training Pr,ograriij D fauestfon ~6}-lrce Solll!llerts) L091 o I;:::::====:::..'.::::::'.::::=:'.::::=:'.::::=:'.::::=:::::::'.::::=:::::='.===========-==============-=~1 I
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IC2oesUonTopic ! IRO 57 Given the following conditions:
- The control room receives Auxiliary Typewriter alarm point 0676, 2G 4KV GROUP BUS LOSS 125 voe CONTROL PWR. - An operator investigates and reports 125 VDC breaker 2BDC1AX12, 2G 4KV Bus Control Power Supply (Reg) tripped.
Which of the following identifies the effect, if any, this will have on 24 RCP 4KV breaker? If 24 RCP is running it will. .. [] I trip immediately. I lb.JI continue to run but will not trip if required. I
~ trip if a RPS trip signal is subsequently developed, but would not be able to be re-started if directed.
I
@] I continue to run and be unaffected as emergency control power from the alternate control power supply will automatically be provided.
I 1Answerj b I I !Ex<lm Level 11 R l [Cogniti.Je Level 11 Application I !Facility: i ISalem 1 & 2 1 IExamDate: 11 12/21/20151 IKA:il 063000K401 IIK4.01 1lr{ci Value:JI 2.7j ISROXalue:Jl 3.0*l lsection: H~ IRO Group:11 I 1 lsROGroup:il 11 mi D iSystem/EyollltionJitle i j_o_.c_.E_l_e_ctr_i_ca_l_D_is_t_rib_u_t_io_n_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-'I :o63
KA Statement: i Knowiedae of D.C. Electrical Distribution design feature(s) and or interlock(s) which provide for the following:
Manual/automatic transfers of control !Explanation of 55.41.b(S) 4KV Group bus control power is supplied from the DC electrical system. The alternate DC supply does not automatically 'Answers: transfer to supply DC control power to the bus, it must be manually transferred when the normal supply is lost. 4 KV breakers cannot be tripped remotely without 125VDC available to energize trip coil. A is incorrect because there is no power to energize the trip coil, plus no trip signal would be present. Bis correct. C is incorrect because it would not trip if required (different from A is
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Facility Reference Number * " j [~efer~nce Section 11 Page NoJ [Revisi011[
~- 24 Reactor Coolant Pump ii 211538-2 I 11 1122 I II .
ii 11 I II : ii 11 I I DCELECE007 1_ ____. lnni:lterial Required {gr i;iallli£1a~ion .. J I 11 IQuestion so'Urce: :j IFacility Exam Bank 11au~stiOI) Modi!iCJ!ti~n Method: ~11 Editorially Modified 1lused DuringTrainin$f Ptog~am lD 1a~e,~tioll sollrce.pomm~r,t~~I
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iOuestion Topic I IRO 58 I Given the following conditions:
- 21 Diesel Fuel Oil Storage Tank indicated level is 96.0". - 22 Diesel Fuel Oil Storage Tank indicated level is 96.1". - 21 Diesel Fuel Oil Transfer Pump is selected to REGULAR. - 22 Diesel Fuel Oil Transfer Pump is selected to BACKUP. - The Diesel Fuel Oil Storage Tank area C02 Tank fails, and a projectile opens a hole in the bottom of 21 DFOST which completely empties.
Assuming no other damage from the C02 Tank failure has occurred, which of the following identifies how makeup flow to the EOG Day Tanks will be provided, if at all, if the 2C EOG starts automatically to provide power to 2C Vital bus on a single bus UV condition? Makeup flow to the EOG Day tanks will be provided by ......
~I BOTH DFO Transfer pumps because DFOSTs are normally cross connected but check valves will prevent 22 DFOST from draining out the rupture. I
[g] 121 DFO Transfer pump ONLY because the REGULAR pump is aligned to the highest storage tank level during normal surveillance testing. I l.<::J INEITHER DFO Transfer pump because DFOSTs are normally cross connected and both will be empty. I
@] 122 DFO Transfer pump ONLY because 21 DFO Transfer pump is aligned to an empty tank.
I !Answer] I~ !Exam .Level 11~ 1cognitiye Level .*J IAppl1cat1on I !Facility: i jSalem 1 & 2 1iE:!CamDate:i j___12_12_1_12_0_1_,5I ~I 064000K608 I!K6.08 1 !RO Value: <@is Ro Value:l@!secti9~::J I~ [RO Group:! LJ !SRO qrol:if>:I LJ B; D !system/Evolution Title j IEmergency Diesel Generators 11064 !KA statement: l Knowledqe of the of the effect of a loss or malfunction on the followinq will have on the Emerqencv Diesel Generators: Fuel oil storaqe tanks IExplatiatfon cit: 55.41.b(8) DFOSTs are normal isolated from each other on the outlet side by the closed 2DF35, 21/22 DFO STOR TANK X-CONN
)
!Answers: VALVE. Each tank is supplied by its respective transfer pump. Return (overflow from DFO Day tanks) is directed to the tank which has its DFO transfer pump selected to lead, so that overflow won't be directed to the storage tank from which suction is not being taken. With an empty 21 DFOST, 21 transfer pump will still receive a start signal (at 33"), but has no fuel to pump. As Day Tank I-**-' ---"-* 'o~ tn '-* **-* ')') 11:1--1 * .,n\ n * - - *'" -*~..t I~+ ')7"\ ~n.-1 -*-**'.-lo"-* <--- ')') -*-*--- *--
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II Facility Reference NI.Imber '>I [Reference s~ction

Ij Page No: i !Revisionl IFuel Oil ll 205249-3 I 11 1130 I IEOG Lesson Plan 11 NOS05EDG000-11 l 1144 1111 I
! !I I II II I jLo. Number. Objectives I FUEOILE004 j_ ___. iMatericil ~~qufredfor EXalliination *** *1 I [a~~stion Source:. J j_N_e_w_ _ _ _ _ _ __,l l91Jestion ~odi~cati,cin. ~~~h9~: **** J________~l Jlised During}~aJnif!Q f>'rogram I D fQuestionSou?c~ Comr!ie~~sl I I
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. . \auestion Topic J I RO 59 An explosive mixture is prevented from being present in the Waste Gas Holdup Sysytem JAW Salem Tech Spec LCO 3.11.2.5, Explosive Gas Mixture by limiting the 1) concentration to 2) or less. §] 11 ) oxygen 2)2% ~ 11) oxygen 2)4% ~ 11) hydrogen 2)2%
I [ ] 11) hydrogen !Answer 11 a I !Exanflevel i IR I !cognitive Level
JIMemory I !Facility{j ISalem 1 & 2 1 !ExamD~~e: 11 2)4%
I 12/21/20151 i IKA: 071000K504 liK5.04 ilR0Value;!j 2.5!tSROValue:!j 3.1llsection:Jl~[ROGroup:Jj 21!sROGr6up:ll 2j lllJ D /syst~m/Evolution Title I IWaste Gas Disposal System I [071 [KA Statement:! Knowledge of the operational implications of the following concepts as they apply to the Waste Gas Disposal System: Relationship of hvdrooen/oxvoen concentrations to flammability I [E)(plan~tiC>n ofi 55.41.b(13) Salem TS LCO 3.11.2.5 staes that oxygen concetration in Waste Gas Holdup System shall be maintained less than or
.Answers:
i equal to 2% . 4% is not correct all of the time. Hydrogen concentration is monitored but not address in Tech Specs I*
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Facility Reference Number **l\Reference Section .* 1\Page No:\ \Revi~ioril j Salem Tech Specs I lj 3.11.2.5 113/411-1511282 I I I II II II I I I II -
ii 11 -- I - Objectives I WASGASE009 ,____. !Material_ Requfred ft;>r EJ(all'linatioh.
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[tfoe~tion squrct,J j_N_e_w_ _ _ _ _ __.I !8tjesti~R'!'Aodiycation ~~thod: 1_________,l [usedD~rirtgJrainirigProgr,am I D 1a.uestio~ so'Urc~ Com~(!ntSI I I
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RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes I ************** *-* ......... . iouestiori Topic I I RO 60 I Given the following conditions:
- Unit 1 is operating at 100% power. - Control room operators are preparing to perform a Containment Pressure Relief IAW S1 .OP-SO.CBV-0002, CONTAINMENT PRESSURE-VACUUM RELIEF SYSTEM OPERATION. - Containment radiation levels are NORMAL for 100% power operation with no failed fuel.
After opening the 1VC5 and 1VC6 CONT PRESS/VAC RELIEF ISOL valves to initiate the pressure relief, which choice describes how the respective radiation monitors indication will respond?
1R12A - Containment Gas Effluent

1R41 B - Plant Vent Noble Gas Intermediate Range

1R41 D - Plant Vent Noble Gas Release Rate
~ 11 R12A rises; 1R41 B rises; 1R41 D rises.
I [b'.111R12A rises; 1R41 B constant; 1R41 D constant. I
\cl I 1R12A constant; 1R41B constant; 1R41D rises.
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@;] I1R 12A constant; 1R41 B rises; 1R41 D constant. I I I .. . ...
Answer 'I~ !Exam
! Level.'~ I I 1Cogmt1ve Level .. 11 Application J !facility. i ISalem 1 & 2 1 jExamDate.1 l 12/21/20151 ~lo73000A101 j1A1.01 !1Rova1ue:1~JsRova1ue:IO]"l,sectii:m:ll~lRbGroup:\LJ1s1mGroup::1 LJ lz~! iSystem/EvolutionTitle j IProcess Radiation Monitoring System I -:0-73~~ [KA Statement: I Ability to predict and/or monitor changes in parameters associated with operating the Process Radiation Monitoring System controls includin : Radiation levels [Explanation of I 55.41.b(11 )1R12A is sampling containment atmosphere, so it will NOT rise when the pressure relief is started. 1R41 B is an !Answers: * : intermediate range monitor that normally does not have sample flow through it. It's sample flow will start when the lo range 1R41 A
~ monitor nears its high end of monitoring range. It's indication will not change during a pressure relief with NORMAL containment radiation levels. The R41D provides the gaseous effluent release rate (uCi/sec) by combining (product of) the on-range R41A . . . .
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I[Reference section*** \~'I li:>a9E! Nq.] iRevisionl IContainment Ventilation System Operation iIS1 .OP-SO.CBV-0001 I ll 1 j2s I IAbnormal Radiation 11 S1 .OP-AB.RAD-0001 I 1132 I I
IL(). Number. / Objectives II I
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11 11 I I RMSOOOE008 ,_ _ ___.
lj l<:l.l!'.~~tion,.$~ur£t j IFacility Exam Bank I[9iJesti~!JM§~if!ca1,%1 Mel~(,d: ?' IDirect From Source i9u7,~tii?:n.~6urcii*c&faiJ'\en!sJ 150514, used on June 2004 Salem NRC RO exam (7 exams ago) I
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!Question Topic I RO 61 I I Given the following conditions: - Unit 1 is operating at 25% power. - 1B EDG is running in parallel with station power on 1B 4KV Vital Bus. - 13 and 16 SW pumps are in service, 11 SW pump is in AUTO. - 1A 4KV Vital bus becomes deenergized due to a Bus Differential signal.
1 minute after the 1A 4KV Vital bus deenergizes, with NO operator action, which of the following contains ALL the SW pumps which will be running? ral I I jb:1111, 15. I rc:i113,16. I fd.l j 15, 16. I 1Answer 1 a I I !Exarri Level 11 R I JcoghitiveLevel i IApplication I !Facility: i ISalem 1 & 2 I [Exa?lDate: 11 12/21/20151 ~I 076000K201 V~lue:1@2:J;sROValue:i[}]'lsection: 1j~ IRO Group:ILJ1SRO GrQup:1LJ l;:!J l"~
.;,,\. ! I I 1IK2.01 i ~' ] iRO .. *' * [" ,(
D !system/EvolutiorlTitre ! lService Water System 11076 I IKAstateme!lt: I Knowledge of bus power supplies to the following: Service water jt::~planath:m Answers: ofl
55.41(7) A bus powers 15, and 16 SW pumps. On a single bus UV as described in the stem, only that bus would load in blackout loading. A bus is locked out on Bus Differential (deenergized), and the loss of 16 SW pump would cause header pressure to lower to where the auto pump (11) would start. Only one SW pump is aligned for AUTO which is the normal at power configuration for the SW pumps, one in auto, and the rest in manual. 12 SW pump would never start unless 11 pump did not on a SEC initiation, that is "h' ;t ;c, --* .,_, _ _, ;n --* nf'h- -h-:--- 111 C::\A/ n* *mn **-.,Jn --* -*~.+ -:--~ D "* "" --**~* *~-~- --* *-* . .,h:~h ;c **h, ;t :--" '
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any of the choices. There can be confusion about the running EDG and the loss of A vital bus' causing a MODE 1i (Blackout), which would strip busses and load the primary SW pump on each bus. The unit 1 SW pump power supplies are reversed from unit 2 (21/22 pumps A bus, 25,26 pumps C bus) i f Page No. I IRevision1
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lI Facility Reference Num.ber /j /Reference Set;tion IService Water Pump Operation II S1 .OP-SO.SW-0001 I II,, 1127 I IUnit 1 4KV Vital Buses One line 11203002 I 1134 I I II I II 11 I tLo. Nutnber .* *. Objectives I SWBAYSE005
,_ _ __, IMat~rialRequiredfor EJ<aminatiol"l 11 I ['auestion Source: 11 Facility Exam Bank I[ouestlC>n fJl'?dific~tiC>n'M~thp~: IDirect From Source IIOsed l)ifr.ing Training Program I o [auestion s9urce <;~~111entsi 1152970, used on Sept 2011 NRG RO exam (3 exams ago) I iComrnent I I I
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Outline Changes I i(luestiori Topic J 1RO 62 I Given the following conditions:
- Unit 2 is in MODE 3, NOT, NOP. - # 2 ECAC was manually placed in operation to collect bearing vibration readings JAW S2.0P-SO.CA-0001, Control Air System Operation. - PRIOR to starting the# 2 ECAC, BOTH Control Air headers were at 96 psig. - 2C 4KV vital bus senses an UV condition, and the bus loads in MODE II*. - 2A and 2B 4KV vital buses remain powered from SPT's. - 5 minutes after the UV signal, Unit 2 ECAC oil pressure indicates O psig and has been 0 psig for> 1 minute.
With NO operator action, which of the following describes the effect this will have on the Control Air System? Assume the ECAC Motor Overload has not actuated at any time.
# 2 ECAC is I , and Control Air header(s) is/are
[J I NOT running, "A", lower than "B" header.
!b.
running, "A", higher than "B" header.
I I i_§_{ NOT running, BOTH, 96 psig. I I I I ldl running, BOTH, 96 psig. [Answer.I @=] !Exam Level I ~!cognitive Level 11 Application I [Facility: 11 Salem 1 & 2 1 IExamDat~:J I 12/21/20151
~I 078000A301 I1A3.01 I!Ro value:J[IijfsROValue:J@jSection: JI~ fRO Group:JLJ!~Ro Gro~i>:ILJ Ill D !System/Evolution Title: !instrument Air System I *~*0_78_~ *KA Statement: j Abilit to monitor automatic operations of the Instrument Air S stem includin Air ressure lExplanation .of/ 55.41.b(?) The ONLY trip which remains active for the ECAC after ANY SEC start is motor overload. The ECAC operating !Answers: .* * ***
1 characteristics are such that at 95 psig and above header pressure, the ECAC will NOT be supplying the CA header. With the stem conditions of 96 psig prior to and after the ECAC was originally started, the subsequent (SEC) stop and restart of the ECAC will have NO effect on CA header, since the Station air headers supplying the CA header were not affected by the II* loading of 2C vital
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i' Reference Title :r Facility Reference Nurribe~ i[Reference Section * ' j fPage No,i jRevisionJ IControl Air System Operation II S2.0P-SO.CA-0001 I !I 1114 I IControl Air Lesson Plan ii NOS05CONAIR-12 I 1130-31 11 I I II I ii ii I [LO. Number
N Objectives I CONAIRE008 1_ ___.
[q~,esfil),p sp~rce~'i1 IFacility Exam Bank I jQtie~ti?r(~.~~}f!g~tiop.~eitj~q; 'I Direct From Source I [used [)urin9:i:tairirlg' Bf9.Qram; I D r:ou~~if?n ~ourcil CoriiiJi~r,t~I ;::.======-================:::===========-==============--,.1 187627 I
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I SRO System/Evolution List I Outline Changes I [Question Te>?ic I I RO 63 I I In addition to providing a source of water for Fire Protection, Fresh Water and Fire Protection Storage Tank water can be aligned to which one of the following systems? I g) I Service Water. I [gJ I [] I Main Condensate. I
@J I Auxiliary Feedwater.
Spent Fuel Pool Cooling. I IMemory I
Answer 1 j c I !Exam Level [ IR I !cognitive Level J I !Facility: 11 Salem 1 & 2 I IEXaml)ate~ 11 12/21/20151 KA:ll 086000K103 IiK1.03 I !Rb Value: II 3.4*lfSRO Value: 113.5*l[Section: n~ 1RO G~C>up:q 21 !SRO Group:! I 21 ~}"IJ D

system/Evolution Title I l_F_ir_e_P_ro_t_e_ct_io_n_S'""'y'""s_te_m
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____,l 1086 lt<A State111ent: I Knowledqe of the physical connections and/or cause-effect relationships between Fire Protection System and the following: AFW System '.Explanation of I 55.41.b( 4) Fire protection water can be aligned to the AFW system through a normally disconnected spool piece. !AnsWer~: : ! Reference Title <' ii . Facility Reference Number, . !!Reference Section*** I!Page No. j !Revisionj INo. 1 & 2 Units Fire Protection 11205222-4 I 1163 I I I II II I I
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I iLO. Numbe'r** .* * rExainination. :j I II [question.S,oul'ce: 11 Facility Exam Bank 1Jou~s!i()n"Mod\fication M~thod: {J Direct From Source I !used burir:1g"Trail)ing Program JD
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RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes (Question Topic'.! I RO 65 I IWhich of the followin~ identifies a condition in which the Unit 2 LCO for Containment lnte~rity, 3.6.1.1 Modes 1-4, would NOT be met? ! I
~ A manual valve or Blind Flange outside containment required to be to be closed during accident conditions cannot be visually verified in correct position due to its location in a High Radiation Area. I I ~ The containment 100' elevation airlock doors are operated by procedure to allow entry into containment for Rad Pro to take radiation surveys.
I 81 A CVCS Letdown Orifice Isolation Valve fails to fully close on a failure of the controlling PZR level channel LOW. I [:] IA SW Accumulator nitrogen cover gas pressure falls below the minimum required. I jAnswer 11 d I !Exam .Level !IR I !cognitive Level 11 Application I !Facility: 11 Salem 1 & 2 ! IEx1(:111}Dat~: JI 12/21/20151 lf<A: 111030008240 I 2.2.4?___j !Ro Value: II 3.41 jSRO Value: l14. 7 J !Section: 1I~ [RO Group:II 1 I fSROGroup:l j 1I B D jsystem/Evolution ntre I j_c_o_n_ta_in_m_e_n_t_S_,_y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.I ! 103 [KA S t a t e m e n t : ! . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . . , . Abilit to a I Technical Specifications for as stem. Explanation of 55.421.b(9, 10). A is incorrect because 4.6.1.1.a specifically states that a manual valve or blind flange in a high radiation area may ,Answers~ ** ** 1 be verified using Admin controls. B is incorrect because while containment airlocks are required to be operable IAW 4.6.1.1.b (per spec 3.6.1.3 airlock), the doors are allowed to be opened for normal transit entry and exit. C is incorrect because CIV have their own TS 3.6.3 which is less restrictive that containment integrity and is not included in surv requirements for 3.6.1.1. D is correct j Accumulator l~vel pressure and temp
** Reference Title *j r: *,* 1 Facility Reference Number>'. I!Reference Section
j \Page No. I !Revision:
ISalem Tech Specs I 113.6.1.1 II 11 I I I lj 3.6.2.3 II 11 I I I 113.6.1.3 II 11 I !Lo. Number Objectives I I CONTMTE010 .M(:1feriai .~equi~~d fqfB@Tliriatipn :. II i Que~~io~ ~g~tc~: dl l_N_e_w_ _ _ _ _ __.l !auestion Modification Method: *. jauestion source Cornmentsl I I
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RO Skyscraper SRO Skyscraper RO System/Evolution List I SRO System/Evolution List Outline Changes [auestionTopic j RO 66I I Given the following conditions for Unit 2:
- Rx is in Mode 3, NOT, NOP. - RWST concentration - 2450 ppm - 21 BAT concentration - 6650 ppm - 22 BAT concentration - 6650 ppm - 22 BAT level - 43%
Which of the following describes the LOWEST level for 21 BAT that meets or exceeds the operability requirements for the BAT? [JI I
~154%.
I [ ] 192%. I [ ] 196%. I [Aflswer,i b I I !Exam Levell IR I 1cC>gnitive Level I
! Application I [Facility: J ISalem 1 & 2 1 IExatppate:J I 12/21/20151 fJ']" !SRO Value: IfID !section: II~ !RO Group:ILJ !SRO Group:I LJ lf'fij ~ .. T'J
[KA:jl 194001G125 112.1.25 1IRO Value: i [systell'liEvolution Title I * - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' iGENERI J IKA Statement*!., Ability to interpret reference materials, such as oraohs, curves, tables, etc. [E.xpfanaii.onpf.,! 55.41.b(6,7) TS 3.1.2.6 requires a certain amount of borated water available. The BASTs are normally cross connected so the fAnswers: * **[ TOTAL volume of the 2 tanks is what is required to be above the limit. With RWST boron concentration of 2450, and both BAT tanks at 6650, the intersection is - 93.5%. If 1 tank is at 43 % the other tank must be at 50.5%. To preclude picking the wrong answer because of interpolation, the correct answer of 54% is 3.5% higher than required. !' ' Reference Title IL Facility Reference Number.
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j f Page No. J !RevisionI j Saqlem Tech Specs II I ii 1 I I I !I I 11 11 I I II I ii ii I jLO.Number Objectives ICVCSOOE010
'--~ lMaterial Requiredfor Exall'lination. I ITS Figure 3.1-2 Boric Acid Tank Contents 1a~~stibii s?µrc~; 11 Facility Exam Bank 11a~7~ti.ol"l llJlodi~cationMethod:.*.J Editorially Modified I jus~d oµring'frainingPrqgram'J D /Cl.1,1es~ic)n SourcE:J ~c:>rnrn~nt~J ;::::======-.:::============:::::::::===========-==============-=~1 j 135342 replaced a distracted and slightly modified correct answer. j I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List . I SRO System/Evolution List Outline Changes I [auestion Topic ! I RO 67 I jWith Salem Unit 2 in Mode 6 on November 20th, which of the followin9 conditions would prevent Core Alterations from bein9 commenced? I
~ I One Source Range NI is inoperable.
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~ The reactor has been subcritical for 100 hours.
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~ Any one of the Containment Airlock doors is open. ,
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@] I Only one loop of RHR is in service in Shutdown Cooling mode.
I lf.nswer: 11 a I !Exam Level j IR I !Cognitive Level 11 Memory I iFacilitY: [ ISalem 1 & 2 I [ExamDate:j I 12/21/20151 ~ [KA: j j 194001G136 Il2.1.36 i IRO ,Value: i[Ifil ;SRO Value:IEJ] Section: Ii~ IRO Group:iLJ !SRO Group:ILJ I I mD
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lsY5tem/Evoiution J:itle I i-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~__.[GENERI[ IKAstatement:! Knowledge of procedures and limitations involved in core alterations. I 1E~pl~nation of 55.41.b(10) A is correct because LCO 3.9.2 states 2 SR Ni's must be operable. B is incorrect because between Oct 15-May 15th Answers: .. :". ** only 80 hours of subcriticality is required per LCO 3.9.3.a. C is incorrect because one ONE airlock door (per airlock)has to be capable of being closed per LCO 3.9.4.b. Dis incorrect because only one RHR loop is required to be in service per LCO 3.9.8.1 I
Reference Title * ******.11 Facility ~.efer:ence Nllrhber .* .;j [Reference *section

lk~age No. j !Revision:
j Salem Terch Specs I I II 11 I l I I 11 1I I I I I II ,, I IL.a: Number I IOP009E004 Objectives I ,_ ____, lilJ!aterfal Required for Examiriati 0 n :) I II lai.f~s!fon sol.lrce: J l_N_e_w_ _ _ _ _ _ __.l lauestion llJlod,ificat!on MeJ~()d: *. 1--------~l Ju~edDl.lri~g TrainingProgralnJ D j'a!-!~stip~ source Commen\~J j I
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iouestion Topic l RO 68 I I Given the following conditions:
- Both units are operating at 100% power. - Reactor Engineering has determined that a single fuel assembly in the Spent Fuel Pool must be moved to a new storage location in the Spent Fuel Pool. - All administrative requirements are complete to allow the movement of the Spent Fuel.
When the field operator arrives at the Spent Fuel Building, he notices that while a Qualified Reactor Engineer is present on elevation 130', a Licensed SRO is not. Which of the following describes the Operations Department requirements for this evolution IAW S2.0P-IO.ZZ-0010, Spent Fuel Pool Manipulations?
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The fuel movement 1) occur because a SRO 2)
~11)CANNOT direct the fuel movement from the crane trolley.
2) shall I
[_] 11) CAN only required to be "in the area" for spent fuel moves.
2) is I
~ 11) CANNOT is required to provide oversight of the Reactor Engineer directing the fuel move.
J1<\l"'A.i 2)
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L..,.j Inot required to obseNe the fuel movement since a Qualified Reactor Engineer is present iAnswer I @=] \Exam Lever j ~ !cognitive Level \ IMemory I [Facmtf [ j Salem 1 & 2 I JE)(~inDate: j j___1_21_2_11_20_1~5i ~! 194001G142 i 12.1.42 ijROValue:l[°3]\sROValue:i['D[section: ll~fROGroup:iLJlsROGrollp:iLJ D lsysternlf:volution Title I iKA Statement' I Knowledoe of new and spent fuel movement procedures. I \Explanation ()f \ 55.41.b(10) Precaution and Limitation 2.2 of S2.0P-IO.ZZ-0010 states ... "IF ANY spent fuel manipulation(s) being performed in the
Answers

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i Spent Fuel Pool, THEN ASSIGN Reactor SeNices, qualified SRO, or Reactor Engineer to supeNise spent fuel manipulation(s)."
Since the RE is present, fuel movement can occur without a SRO. I*.*.*. Reference .Title H Facility R~ference Numb.er >I [Reference Section /I! Page No. J iRevision ISpent Fuel Pool Manipulations II S2.0P-IO.ZZ-0010 II lj 2 I 133 I I II II II 11 I I II ii 11 11 I (L,o.Nl.lmber.'
Objectives I IOP010E005 l
louestlon Sour~~: ! Previous 2 NRC Exams [.ciuestiori M9dific<:ition tv1ethod:) I :I I iusedOuririgff3inirigProgram .j D fQll~~tion ~ol.lr~e c()il)meiitsf I!-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~--! I [Corrimeot./ ... > ' I I I I I I
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RO System/Evolution List I SRO System/Evolution List J Outline Changes ( IOuestion Topic 11 RO 69 Given the following conditions:
- Unit 1 is at 100% power. - 11, 13, and 16 SW pumps are in service, with 15 SW pump selected to Auto. - Subsequently, a loss of the 500 KV Switchyard occurs. - 1A 4KV Vital Bus has de-energized due to a Bus Differential relay actuation. - Unit 1 has initiated a MANUAL Safety Injection (SI).
Which of the following identifies the Service Water Pumps which will be running 2 minutes after the SI has been initiated? [ ] 111 and 14. [gJ 112 and 13.
~ 113 and 15. @J 114 and 16.
1AnswE!r 1 I a I !Exam Level 11 R I !cognitive Level 11 Application I 1 IFatiiity: 1 Salem 1 & 2 I iExamDate:[ I 12/21/2015! !KA:jl 194001G203 jj2.2.3 ! I [Ro Value: 1 3.81 rsRo Vah.ie:ll3.9 I[section: 11 PWG I[Ro Gi-oup:l I 11 lsRo Group:il 1j 5~1~ D [system/Evolution Title II I :GENERI I KA Statement*;.. (multi-unit license) Knowledoe of the desion, procedural, and operational differences between units. I [Explanat~~n o*f
Answers

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l 55.41.b(?) This question meets the K/A because SW pumps are powered from opposite vital buses when Unit 1 is compared to Unit
2. U1 pumps are C,C,B,B,A,A, whereas U2 are A,A,B,B,C,C. The Lead pump on B bus is always 14 unless 14 is not available.
With 1A bus deenergized, 15 and 16 SW pumps have no powerThe SW pump selected to auto (15) will not start on low pressure (nor will it have power) as it will be locked out by SEC initiation. [ : ....*.. *ReferenceTitle .* ****-*([* <Facility Reference Number*.* I!Reference Section' < \l Page NC>/ fRevision; 1 1 I1B 4KV vital bus one line diagram 11203002 II 11 1134 I I II II 11 11 I I II II 11 1 I I IL.o*. ~_urnber ; Objectives ISWBAYSE005 \ __ __, I.Material Requir~d for Examination* II Ij \ouestiC>ll Source: ;11 Facility Exam Bank I \ouesti2n M,0 difi,6~tior(l\t1~1~~d: I
** Editorially Modified 1l~~ed Durin~iTl'ailling Pfogram I D
[ou.estion sourcE!£ommentsj l , . . . . 4 _ 8 _ 9 - 9 2 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - = - 11
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!Question Topic I IRO 70 I An Independent Verification (IV) of valve position is required in an area with a 75 mrem/hr dose rate.
For this job, which of the followino is the longest time allowed for the IV before "hands-on" verification may be waived?
~ 15 minutes.
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~ 17 minutes.
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~ 19 minutes.
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@] 111 minutes I !Answer l b I I !Exam Level 11 R I !cogri'itive Level 11 Application I !Facili~y: 11 Salem 1 & 2 I !Exanibate: 11 12/21/20151 r~~~""'*:m i IRO Value:I ~ ISRO Vallie:! [TI} ~Section: :j~ 1RO Group:JLJ [SRO Group:: L J li~~~ ~ ,.
jKA:ll 194001G214 i I 112.2.14 !System/Evolution Title ! * - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' IGENERI I II KA statement* I Knowledoe of the process for control lino equipment configuration or status. iExpla;;iation !.Answers: ofl 55.41.b(10) 10 mrem is the dose above which an IV is not required to be performed. A= 6.25 mrem. B=8.75 c=11.25 mrem
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.~*, : *:,,,>. *Reference Title 'I Facility Refe.rerice Number I ,j [Reference Section 11 Page No. j !Revision!
IComponent Configuration Control ll OP-AA-108-101-1002 II Att 11 Step 1.5.1 1165 11 7 I I I I II I I lLO'. Number **.
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11 11 I IMISCAPOO? !Material Reqllired torExamination*;*J I IQuesti?pSo.llrc~:>J IFacility Exam Bank I !8'Lies!i?.~ 1vf?~itii:ation Method:.] Significantly Modified I!usedptlrlng"frainfr1g Program I o [Questi<:>n ~olirc~ <;cjf!lmentsi 160955 changed dose rate which changes correct naswer to a previous distracter. I
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tOuestiof! Topic 11 RO 71 Given the following conditions:
- Salem Unit 2 is operating at 100% power. - 21 SW pump is err. - 22, 25, and 26 SW pump are in service. - Subsequently, 22 SW pump discharge strainer clogs, 22 SW pump is stopped, and 22 SW pump is declared inoperable. - 2 hours after being secured, maintenance discovers a crack in the strainer drum which will take 1 day to repair.
Of the following, which is the only method of tracking SW pump status which is NOT performed IAW OP-SA-108-115-1001, Operability Assessment and Equipment Control Program? Updating the is NOT required.
~I Operational Status Board.
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!b* I Control Room Narrative Log.
I [Cl I Tech Spec Action Statement Status Board. f";Jl IIgcbcic~I StH~cifk:~tioc Ac:tioo st~tQ I -- I
~, ~
Answer I ~ !Exam Level i ~ !Cognitive Level >11 Memory I 1 IF;acility: ! Salem 1 & 2 I /Exa111D~te: j !___12_12_1_12_0_1~51
~I 194001G223 Ij2.2.23 *'Ro Value: l[TI"lsRo Value:l~!Section:H~ !Ro Group:jLJisRO Group:ILJ D (system/Evolution Title I IKA Statement: I Abilitv to track Technical Specification limiting conditions for operations.
Explanation of 55.41.b(10) Initially, actions wil be attempted to clear the strainer clog, so it won't be readily apparent that the repair would take iAnswers: longer than one shift. (Sect 5.2.4) When it becomes apparent that it will take longer than one shift, Section 5.2.5 will be performed also. The Operational Staus Board is used during emergencies and is located in the control room area.
I *.*Reference Title
**.H ,Fac;ility Reference. Number [Reference section **I IP,age No. IIRe,yision: *1 j Operability Assessment and Equipment Control II OP-SA-108-115-1001 I 11 1I7 I I II I 11 11 I I II I II 11 I
[LO. Number ., Objectives I TECHSPE015 1_ ____. lauestion Source: ;:J l_N_e_w_ _ _ _ _ ___.I !auestlo!l ~odificcitio.n Meth()d: ij_________.I !used pliring'.'.fraliiirig Program. l D [guestio(l ~.?urce Comme~~~l I I
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RO System/Evolution List I SRO System/Evolution List I Outline Changes I
!auestion Topic ! IRO 12 I Given the following conditions: - A Unit 2 shutdown is in progress. - A containment entry is required to inspect an INOPERABLE component on elevation 78' outside the bioshield. - The Unit is being shutdown at 10% per hour.
Which of the followinq additional approvals is required to authorize a containment entry other than the SM/CRS under these conditions?
~ I Operations Director - Salem.
I
~I Station Vice President- Salem.
c.l jWork Control Center (WCC) Supervisor.
I
@J I Radiation Protection Supervisor (RPS).
I I !Answer J Id I [Exarri Level* 1 I R I ;*cognitive Lev,el
I I Memory I iFacilify: IISalem 1 & 2 I jeX:ainOate:j I 12/21/20151 IKA:ll 194001G312 112.3.12 I !Ro Value: JI 3.21tSROValue:Jl 3.7 l lsecti6f"l:jj PWG lfRb Gro(Jp:[I 11 \SRO Group:il 11 if!ll D fSystemlEvolutionTitleJ
[t<A Statement* I Knowiedge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, alioning filters, etc. jE,xplaQationo~ [ Normal Containment entries at power are governed by SC.SA-ST.ZZ-0001 & RP-SA-102. Authorization to access containment is 1~nswers: . 1 provided by the SM/CRS. However, to access containment when power is being changed >5%/hr, the Radiation Protection Supervisor's approval is required. (Pre-req 2.4) [ ******Reference Title.. * ,:,*,, ll Facility Reference Number . <1 [Reference Section <!IPage No. j }Revision: ISALEM CONTAINMENT ENTRIES IN MODES II SC.SA-ST.ZZ-0001 I q I 15 I I \I I ii 11 I I II I II 1 I I ILo. Number Objectives I RADCONE004 ,_ _ __. /Material. Required fc)i: Examinatipn, *l . II jQuesHon S:?~rce: 1 lFacility Exam Bank I jaue~ticm 'M?dific~~ion.Metho.a:: JEditorially Modified Itus~d riurtng TraillJ~g Program.I o lQllestlon Sollrce CoriJrJ1ents] 178014 added abbreviation to each title because its found that way in the procedure. I
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RO Skyscraper
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I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I .Outline Changes I [Question Topic 11 RO 73 I IWhich setpoint? of the following Area Radiation Monitors (ARM) will cause a ventilation system alignment change when it reaches its High Radiation Alarm I
~12R44A, Containment High Range.
I
~12R32A, Fuel Handling Crane.
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~12R9, id. j I New Fuel Storage.
2R52, Liquid PASS Room. I I I.Answer! c I I lexam Level i IR I jcognitive l..evel I j Memory I !Facility: j ISalem 1 & 2 I !ExarnDate:*.11 12/21/20151
,.. !ID !section: iI~ iRO Grou,p:J 0 0 5om*7 ~II D '~
IIRO Value; I' [IE ISRO v.alue: 1 I >:, [KA:ll 194001G315 112.3.15 :SRO Group:! !system/Evolution Title I !~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-'
!GENERI; IKA statement: l edge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel orin e ui men!, etc.
!Explanation of 55.41.b(11) C is correct because it realigns FHB ventilation through the charcoal filters and starts both FHB Exhaust fans. B is iAnsw~rs: incorrect but plausible because its auto function is to prevent Fuel Crane motion except in downward direction. D is incorrect since ~~~-~~ it only has alarm light outside the PASS room which activates, but plausible because of the high radiation levels which would be expected in that area of the aux building following an accident. A is incorrect since it has no automatic function, but is plausible
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,~/ Reference Title *** *
II Facility Reference NulJ1ber .. *I [.~efere11ce :Section I [Page No: I iReviSionl IS2.0P-AB.RAD-0001 11 Abnormal Radiation II Attachment 5 RMS ch 1114-16 1130 I I II I 11 11 I I II i 11 11 I
!Lo'.. Number RMSOOOE005 [Material Requir~d :for Exami1JCl,tior1? II I fQuestion So~rce: >i Facility Exam Bank I /al.l~sti~h ~~di!!C:Ci~i<;mMethod:, IDirect From Source I[Qse'd oUriiig Jrclrnillg Program l D fauesti~ns.~u{~e.<;orryme~t~/ 1125827 used on 9/2011 Salem NRC RO exam. 3 exams ago. I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes lauestion Topic 11RO74 I Given the following conditions:
- Salem Unit 1 is operating at 100% power. - 1A EOG is CfT for maintenance. - A 500KV switch yard malfunction causes a loss of offsite power, and 1B and 1C EDGs do not load due to 1Band 1C 4KV vital buses locking out on Bus Differential.
Of the following, which one states a procedure entry and its correct flowpath allowed by the rules of procedure !AW OP-AA-101-111-1003 Use of Procedures?
~I Enter EOP-LOPA-1 directly and perform immediate actions, which states to trip the Rx then trip the Turbine. ~I Enter EOP-LOPA-1 directly and perform immediate actions, which states to trip the Rx and confirm the Rx trip, then trip the Turbine. ~I Enter EOP-TRIP-1 and perform immediate actions, which states to trip the Rx and confirm the Rx trip, trip the Turbine, initiate SI, then transition to EOP-LOPA-1 based on no vital buses energized. @]I Enter EOP-TRIP-1 and perform immediate actions, which states to trip the Rx and confirm the Rx trip, trip the Turbine, initiate SI ONLY if conditions warrant, then transition to EOP-LOPA-1 based on no vital buses energized. I
!Answer 11 a I IExam Level 11 R I I
!Cognitive Level ! Application I jFacility: i ISalem 1 & 2 I ;examDate: 11 12/21/20151 IKA:ll 194001G401 I_;2._4._1_~! IRO Value: l~ISRO Value: I~ 1Sect1on: II~ )RO (3roup:)LJ :SRO Group:j L J t
BJ ,~ ;~ D l~s~ys-t-~m-./-E-v-ol~ut-io_n_T-it-le~j IGENER! I iKA Statement: 1 Knowledoe of EOP entry conditions and immediate action steps.
ekpfanation of I 55.41.b(10) Either TRIP-1 ot LOPA-1 can be entered upon a total loss of all AC power. If entered, the flowpath for TRIP-1 does not
!Answers:* . reach the SI evalution step before the kickout to LOPA-1 on no 4kv vital buses energized, so both TRIP-1 distracters are incorrect. LOPA-1 does not confirm the Rx trip (since there is no power to do anything about it anyways). ...'
Reference Title It Facility Reference Number . *I !Referenc*e Section* IIPage No~ I !Revision:
I Rx Trip or Safety Injection 111-EOP-TRIP-1 II 11 1127 I ILoss of All AC Power 111-EOP-LOPA-1 11 ii 1125 I IUse of Procedures II OP-AA-101-111-1003 11 1110 11 6 I I LOPAOOE009 Objectives 1_ _____. [auestio,? S()~rce:] l_N_e_w_ _ _ _ _ _ __.l la~~stioll MMificati~n,M,~tllo,g: v _________.1 lused DuringTrarning ~rc)gram I D
!Oue~!iof1 Source C()mm~ntsJ I I *~-----------------------------------------' !Comment .**
I I I I I I
RO SkyScraper I SRO Skyscraper
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I... RO System/Evolution List I ... SRO System/Evolution List I
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Outline Changes I louestiqn Topic 11 RO 75 I Given the following conditions:
- An Alert has been declared at Salem, and all required notifications have been made by the Primary Communicator. - Conditions degrade to the point where a Site Area Emergency is declared.
Which of the following identifies the PRIMARY method which the Primary Communicator will use to make notifications to the States of Delaware and New Jersey, and how long from the SAE declaration do they have to make those notifications !AW Attachment 6, Primary Communicator Log of the Salem ECG?
~I NETS phones within §]I 15 minutes.
I fc.', I NETS phones within 60 minutes. I L___J []I ESSX phones within 15 minutes. I ESSX phones within 60 minutes. I i*Answer.l j a I !Exam Level 11 R I jco911itive Level 11 Memory I !Facility: !ISalem 1 & 2 I fi='xamDate: 11 12/21/20151 IKA:Jl 194001G429 I~:2_.4_.2_9-~! iROValue: :OJ}fsRO Value:J!ITilsect1on: JI~ iRO Group:ILJ1SRO Group:jLJ *--<:".<,
' ' I.
E D ~ls-y-st_e_m_/E_v_o_lu-t-io_n_T_it-le~'l IGENERI I IKASfafomenfl Knowledqe of the emerqency plan. [Explanafi~n of j 55.41(10) Salem ECG, lists the communications systems in order of preference. The NETS (Nuclear Emergency iAnswers: *
Telecommunications System) is the primary closed circuit communication system for off-site notifications. The ESSX is also a closed circuit system, which is used as a backup for NETS. The notifications to the States must be made within 15 minutes of the declaration of an Emergency, even if a lower classification emergency is already in progress. The 60 minutes is plausible if the
---..J:.J-.1.- .&.L-~-l*e '-'--.a.:-- ie , ___ ' *- _: ___ ~,.., ".:I, ................ -": ........... ' ,,*r *** : Facility Reference Number
iIRe~erence Section If Page ~o. I IRevision; I .Reference.Title*..* :'. 1 r~
IPrimary Communicator Log 11 EP-SA-111-F6 II II 1112 I j Emergency Preparedness Training Communicat 11 NEPCOMMDTYSC II II ljo5 I ! II II 11 1 I I GENISSE013 Objectives I !Question so1:1rce: 11 Facility Exam Bank 1 lq~~stioll'~odifi9~t.toll:M~th()d: *. IDirect From Source I[Used f?uring Traihi~g Program I D
\otiesti~11~PLn'ce c;o~1!1ef1t~l 1Used on Salem Sept 2011 NRC exam (3 exams ago)
I fcoriunent ***.* ** \ *. I I I
U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: IR . I eg1on: I i I Date: 12/21/2015 I Facility: Salem 1 & 2 I License Level: SRO I Reactor Type: W I I I Start Time: i Finish Time: I i Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours to complete the combined examination, and 3 hours if you are only taking the SRO portion. App11cant Certmcation All work done on this examination is my own. I have neither given nor received aid. Applicant's Signature Results RO/SRO-Only/Total Examination Values --I --I -- Points Applicant's Score --I --I -- Points Applicant's Grade --I --I -- Percent
Senior Reactor Operator Answer Sheet Circle the correct answer. If an answer is changed write it in the blank. NAME:
1. a b c d

2. a b c d

3. a b c d

4. a b c d

5. a b c d

6. a b c d

7. a b c d

8. a b c d

9. a b c d

10. a b c d 11 . a b c d

12. a b c d

13. a b c d

14. a b c d

15. a b c d

16. a b c d

17. a b c d

18. a b c d

19. a b c d

20. a b c d

21. a b c d

22. a b c d

23. a b c d

24. a b c d

25. a b c d Page 1
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [auestion Topic J ISRO 1 Given the following conditions:
- Unit 1 is performing a Rx startup IAW S1 .OP-IO.ZZ-0003, Hot Standby to Minimum Load. - Power is stable at 1x10-8 Amps for Low Power Physics Testing. - A Shutdown Bank "A" rod drops fully into the core. - The Rx does not trip.
Which of the following identifies:
1) How the CRS should proceed

2) Why rod withdraw! is not allowed?
Direct the RO to ...
~ 1) trip the Rx.
2) A dropped rod recovery would constitute an approach to criticality.
~ 1) fully insert all Control Bank and Shutdown Bank rods.
2) A dropped rod recovery would constitute an approach to criticality.
~ 1) trip the Rx.
2) The depressed power distibution profile in the area around the dropped rod may cause power production in other parts of the core to exceed Tech Soec limits.
@:] 1) fully insert all Control Bank and Shutdown Bank rods.
2) The depressed power distibution profile in the area around the dropped rod may cause power production in other parts of the core to I VAVVVU I VVl I ""tJ'VV lllllll>=I.
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!Ans""°~ ~ ~m Level 1 IS I [C09flitive L~~U I Application I *~ I Salem 1 & 2 I iExa~ 1_ _ _12_12_1_12_0_1_,sl Bl ooooo3G409 I',?-4.9 ____1~~~a~~§~e>_~ls~~~i~f!JI~ ~~-G~LJ !SRO Grou!):;[] ~ System1E1folution Titl~J l_D_ro~p_p_e_d_C_o_n_tr_o_IR_o_d------------------------------~ ____ IKA Statement=- ~- . ,. -~-~ Knowledge of low power/shutdown implications in accident e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. I IExplanation !Answers: 01 55.43.b(6,5) This question is SRO level based on the knowledge of internal effects on core reactivity of a dropped rod while exactly critical, and the procedural direction to insert all rods based on becoming subcritical. The bases for that insertion is that performing - - * *.. **-**-- a dropped rod recovery when the problem has been fixed, which would normally be done at power, would NOT be done if the Rx were subcritical, because withdrawing that rod would constitute an approach to criticality and is required to be performed IAW the
... 4-..,rf11n ............. ,.,.....1.1ro '

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L_ Reference Title
-*. -- --=11 Facility Reference Number _J 1Reference Section 11 Page No. J [Revision j Dropped Rod II S1 .OP-AB.ROD-0002 I 11 ! !10 I l II I 11 II I I II I 11 11 I
!LO. Number ~ Objectives ABROD2E002 I I I
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[Material Required for Examination iI 'I jOuesti~~ Source: I I New I!Oues~~~ Modific~ion Method:__J I [t1sed During Training Program 1 D [Ques~~n source Comments1 I I I:_.... *-* .. l I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic I ISRO 2 Given the following conditions:
- Unit 2 is operating at 100% power. - 21 charging pump is in service. - 22 and 23 charging pumps are operable. - Subsequently, 21 charging pump trips due to a breaker malfunction. - The CRS enters S2.0P-AB.CVC-0001 and places 23 charging pump in service.
Which of the following:
1) Identifies ALL the Tech Spec LCO(s) which will be entered

2) What the required action(s) is/are if 21 charging pump remains inoperable for the next 4 days?
The CRS will enter .....
~ 1) LCO 3.1.2.2.b for not having required boration flowpaths, LCO 3.1.2.4 for not having required charging pumps, and LCO 3.5.2.a for not having required ECCS subsystems available.
2) Be in Mode 3 and borated to a SOM equivalent to at least 1% delta k/k within 78 hours of pump trio.
~ 1) LCO 3.1.2.2.b for not having required boration flowpaths, LCO 3.1.2.4 for not having required charging pumps, and LCO 3.5.2.a for not having required ECCS subsystems available.
2) Be in Mode 4 within 84 hours of pump trip.
~
lS 1) LCO 3.5.2.a ONLY for not having required ECCS subsystems available.
2) Be in Mode 3 and borated to a SDM equivalent to at least 1% delta k/k within 78 hours of pump trip.
rc.v l____J
1) LCO 3.5.2.a ONLY for not having required ECCS subsystems available.

2) Be in Mode 4 within 84 hours of pump trip.
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~~~ @=] ~ri:i_L~~ js I~~ 1_ _ _ 12_12_1_12_0_15~1 r 1 lcognitiveLeveIJ !Memory I §nty:j !Salem 1 &2 ~I 000022G222 I ~_j [RC>:Vai~e~ [Ifil~rSeCtion:ll~ Ro Group:]LJ!sRo Group:[LJ 1 ~ ]system/Evolutionii!&] ILoss of Reactor Coolant Makeup 1 KA siaieme.m:::l
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Knowledge of limiting conditions for operations and safety limits. ~lan.ation of i 55.42.b(2) This question is SRO level based on NUREG-1021, ES401, Attachment 2, page 17, 11.B, 1st bullet for application of Ans!'fE!.':!;:__j required actions. lnoperability of a single charging pump in Modes 1-3 only results in entry into the ECCS LCO. The 2 other LCOs would be entered upon inoperability of the second charging pump. The required action is to restore within 72 hours or be in Hot Shutdown within next 12 hours. The incorrect action in the distracters is for when 2 charging pumps are inoperable.
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I Reference Title ~FacilitY__~eference Numb~ 1Reference Section ][iiage~ ]Revision! j Salem Tech Specs I l j 3.5.2, 3.1.2.2, 3.1.2.4 II 11 I I I II Ii 11 I I I II ii 1 I I IL.o. Number Objectives I CVCSOOE010 ,____.
!Mate~ial Required for E_xamination ;I ii f Que~tion Source: JINew I[9~~stion Modification Method~- ...I 11 I jUsed O~ing Training ~~gram I D [Question ~~~rce Comments] I I ,;:_******-Hl I I I I I ! I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [§uestion Topic 11 SRO 3 I Given the following conditions:
- Unit 2 is in Mode 5. - Rx vessel level is 97.4'. - 22 RHR loop is in service. - RHR HX inlet temperature is 125°F. - 21 RHR loop is available. - 21,24, and 26 SW pumps are in service. - 21 charging pump is in service. - Subsequently, a complete loss of Service Water occurs.
Which of the following identifies a procedure the CRS will enter, and the FIRST action operators will perform in that procedure? I
~ S2.0P-AB.SW-0005, Loss of All Service Water. Start 23 charging pump.
I I
~ S2.0P-AB.SW-0005, Loss of All Service Water. Stop 21 charging pump.
I I
~ S2.0P-AB.RHR-0002, Loss of RHR at Reduced Inventory. Start 21 RHR pump.
I
@;]I S2.0P-AB.RHR-0002, Loss of RHR at Reduced Inventory. ..
Stop 22 RHR pump. I
@=] ~~~~~ IS I ~~~~ 1_ _ _1_2_12_11_20_1~51 *~~~--~ ~"-,--~---
[Answer I I [cognitive Level 11 Application I !Facility:: I Salem 1 & 2 ~I 000025G120 12:1~2_?_=~] i~O Value: J~[SRo ValueJ~ !section: II~ [ROGroup~LJlsRO Group:ILJ ~ 'system/Evolution Title! ILoss of Residual Heat Removal System IKA Statement: I Ability to interpret and execute procedure steps. I !Explanation of i 55.43.b(5) This question is SRO level based on having to assess facility conditions, select the correct operating procedure, and IAnswers: ! know specific actions taken in that procedure. BOTH procedures may be entered, however, AB.SW-5 has a CAS that states if RHR is in service GO TO AB.RHR-1 or RHR-2 depending on RPV level. AB.SW-5 distracters are plausible because procedure states that if 23 charging pump is IMMEDIATELY available, then place it in service based on its being cooled by CCW, and will extend the
..:........ +n ... 11 ........ nl--=-- .........................,I"'\ in---*=-- +,..... -1: ...................... - - . ... . . ..... -. ..... 1: ..... - ... fl"\ i+ LI:-1.... ....... ----:.a.: ...... 1........... ... ... .
___ ,.: ....... -1 J: .. - : ..... :.1.;...,1 conditions that ALL SI and charging pumps EXCEPT one are required to be cleared and tagged with RCS temp< 312°F, and that there will not be another charging pump to go to. 21 charging pump WILL be stopped, but not before letdown is isolated. AB.RHR checks a RHR pump in service, so starting a second RHR pump will not be performed. The in service RHR pump will be stopped to preclude damage to the pump with RPV level <97.5' I L.. Reference
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Title 11 Facility Reference Number J[Ref_erence Section______l IPageNQ.l iRevision ILoss of All Service Water !IS2.0P-AB.SW-0005 II 11 114 I ILoss of RHR at Reduced Inventory II S2.0P-AB.RHR-0002 II 11 1114 I I II II 11 11 I IL.O. Number Objectives IABRHR1E004 1--~
!Material Required for Examination 11 lj !Questi~~~ou~ IFacility Exam Bank I[~uestion Modification Method: jSignificantly Modified It!Jsed During Traini!1~-~~ogram 1 iD ~on source co~ments] ,_
~-- I v AB.RHR actions to which procedure and which action.
I !Cornn.-... I I I I I I I
RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes f [Question Topic 11 SRO 4 Given the following conditions:
- Operators are responding to a 650 gpm tube rupture on 22 SG which occurred while operating at 100% power, IAW EOP-SGTR-1, Steam Generator Tube Rupture. - All off-site power was lost when the Main Generator breakers opened. - The RCS was cooled down to Target Temperature, and then depressurized to stop primary-to-secondary leakage. - Subsequently, off-site power is restored, and all 4KV Group Buses are now energized from off-site power. - The crew is evaluating RCP re-start. - RVLIS Upper Range indication is 98%.
Which of the following contains:
1) ONLY the criteria in addition to RCS Subcooling which are checked to determine if RCP re-start will be performed

2) How the re-start of the RCP be accomplished if allowed The crew will check ....
~- 1 j ~~ ~:_~level and the PZR saturated. ~, a RCP directly at SGTR-1 Step 49 based on RCP support conditions being desired but not required to start the RCP. ~ 11) RVLIS Full Range indication and SG NR level. *--- 2) Start a RCP directly at SGTR-1 Step 49 based on RCP support conditions being desired but not required to start the RCP. ~ 11) PZR level and the PZR saturated.
2) Start a RCP using S2.0P-SO.RC-0001, Reactor Coolant Pump Operation to ensure all support systems and P&Ls for starting a RCP are met.
[ ] 1) RVLIS Full Range indication and SG NR level.
2) Start a RCP using S2.0P-SO.RC-0001, Reactor Coolant Pump Operation to ensure all support systems and P&Ls for sta v : ~~,.,
met. !Answer I ~ :exam Level' ~ !Cognitive Level 1 I Memory I !Facility: 11 Salem 1 & 2 I IExamDate:J I 12/21/20151 ~:ijoo0038A217 l[~~~*~ __J[f!~~l~_!:H 3.8jl~_ROValue)~:section:il~IROGroup:ll 1jlSROGroup:ll 11 Im~ 's~~l§~ol1.1_ti<:>'"1!i-1!~J lSteam Generator Tube Rupture 11038 [KA StatE!_~nt~ Abilit to determine and inter ret the followin as the ap I to Steam Generator Tube Ru ture: Explanation of 55.43.b(5) This question meets SRO only criteria listed in NUREG-1021, ES-401, Attachment 2, 11.E. Figure 2, 1st bullet for 5th Answers: Block for assessment and implementation of a procedure or section of procedure, namely SO.RC-0001 vs direction strictly in the EOP. SGTR-1 does not contain a step which directly starts a RCP, rather it directs RCP start IAW SO.RCP-0001, which will require support conditions satisfied to start the RCP. There are other places in EOP network where RCPs are started without regard to
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Reference Title -------1 c--.=acilitYRefilrencetiiJffiiier-- *1 [Ret&encesecuorll ~~ [ReviSIOrll ISteam Generator Tube Rupture lI2-EOP-SGTR-1 ILo. Number Objectives 01E009
!Material Required for Examination i! II I jQuestion s~-u~c~_:J New I [Ouestio~ ~o~ification Method~! I [!Js_E!~_c.>uring Training Program D I ~~stion Source Commentsj I I 1:_,.,.,,_,,. I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I I Question Topic 11 SRO 5 Given the following conditions:
- Unit 1 is operating at 100% power. - 21, 24 and 26 SW pumps are in service. - 21 and 22 SW header pressures are 108 psig. - The following OHAs annunciate sequentially in this order: - B-13, 21 SW HOR PRESS LO - B-14, 22 SW HOR PRESS LO - B-15, TURB AREA SW HOR PRESS LO. - B-48, SW VL V RM FLOODED.
The standby SW pump starts automatically, and OHAs B-13, B-14, and B-15 clear. Which of the following describes where a SW leak could be located which would produce these alarms, and the procedurally directed actions which would mitiQate the event? I
~ On a CFCU supply piping in the 78' Mechanical Penetration Area. S2.0P-AB.SW-0001, Loss of Service Water Header Pressure, will direct isolating multiple CFCUs IAW Attachment 5 CFCU leaks, until header pressure stabilizes and sump pump runs stop.
I
~I Upstream of the 2ST901, TURB LO CLR ST RET VLV. S2.0P-AB.SW-0002, Loss of Service Water-Turbine Header, will direct operators to adjust 2ST901 TURB LO CLR ST RET V, and 2ST1 TG AREA SW PRESS CONT VLV, to compensate for the SW leak.
I
~ On a CFCU supply piping in the 78' Mechanical Penetration Area. S2.0P-AB.SW-0001, Loss of Service Water Header Pressure, will direct isolating a single CFCU which would be readily identifiable from the control room.
rd.. IUpstream of the 2ST901, TURB LO CLR ST RET VLV. S2.0P-AB.SW-0002, Loss of Service Water-Turbine Header, will direct removing the 11v1C:1111 1uru111e110111 :;erv11,;e 111 µreC:1JJIC:111urr 101 r:;u1C:111r1y 111e 11.:>/-\ r1eC:1ue1. i§!\Y~ ~ [fxam-Level ! @=] [cognitive LevelJ lApplication I ~iliti] ISalem 1 & 2 I jExamDate: 1 !___12_12_1_12_0_1_.sl IKA:]j 000062G445 112.4.45 I ~~[DJ SRO Valll_~j~ !section: JI~ !Ro Group~LJjSRO Group:ILJ ~
gystem/Evolutio~~!itltij ILoss of Nuclear Service Water 11062 -~

-~J IKA Statement-I
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to prioritize and interpret the siQnificance of each annunciator or alarm. IExplanation of I 55.43.b(S}This question meets SRO only criteria listed in NUREG-1021, ES-401, Attachment 2, 11.E. Figure 2, 1st bullet for 5th Answers: ---~ Block for assessment and implementation of a procedure or section of procedure in knowing where the leak could be (assessment), then correct procedure implementation and steps taken in the peocedure. The leak location could be in the TGA with the conditions in the stem except for the SW valve room flooding. Knowledge of where the SW valve room and what piping is there
--1 +,... --- ........ ,.,1 ...... .a.:........ Tho ')C:TQf'\'1 ...... ilrl ........................ - .... Tr:!./\ --* - - - ' ~-..... 1 . . . - - ' *- - - .... -.Lr.,..:- ,...,..., ,1r1 ........ *- .... . nf header pressures. If it did not operators would be directed to ake manual control of 2ST901 and 2ST1. If it is thought that the TGA header must be isolated to stop the leak, then removing the MT from service would have to occur. The multiple cFCU isolations directed in attachment 5 is for leaks of undetermined CFCUs in containment, and refers to containment sump pump runs and trying to isolate the leak by stopping a bunch of CFCUs. Step 3.11 states if a single component can be isolated, and it can, to isolate it.
Additionally, the SW indication in control room would identify that a single CFCU is affected. The sump pump runs referred to are containment sump pump runs, and are if the SW leak is in containment. - Reference Title --] ~C:ility Reference Nu01ber ![Reference Section J[Page No. I IRevision: L__ -- ILoss of Service Water Header Pressure 11 S2.0P-AB.SW-0001 II ii ! I16 I IOverhead Annunciators Window B ll S2.0P-AR.ZZ-0002 II ii 81 1136 I I II II 11 11 I IL.O. Number Objectives I ABSW01 E004 ,_ _ __.
]Material Required for Examination JI II [~uestion Sou~<:~ IFacility Exam Bank I~e~~on Modification Metho~: JI Significantly Modified IIUsed P_ti_ring Training Program D J IQuestion~~iource Commen!sJ 1153928 used on Salem SRO NRC exam more than 2 exams ago. Modified to add actions required. I iCorr.... y'" I ! I I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question Topic-I ISRO 6 Given the following condition: Both Unit 1 and Unit 2 control rooms have been evacuated due to a toxic gas release on site. When performing local actions to stabilize the plant:
1) Where will PZR level be determined

2) How will PZR level be controlled ra,: 11) At the Hot Shutdown Panel 213.
~ 2) By ensuring the Charging System Master Flow Controller is controlling PZR level on program. ~ 11) At the Charging Pumps Flow and Pressure Panel 216-2.
2) By ensuring the Charging System Master Flow Controller is controlling PZR level on program.
le. i 11) At the Hot Shutdown Panel 213.
~* 2) By establishing local control of the CV55 CHARGING FLOW CONTROL VLV to maintain PZR level 22%-77%.
[dl 11) At the Charging Pumps Flow and Pressure Panel 216-2. L~. 2) By establishing local control of the CV55 CHARGING FLOW CONTROL VLV to maintain PZR level 22%-77%. iAnswe~ Ic I 1exam Level i Is I @_09niiive Level 11 Memory I [Fa~ility: [ j Salem 1 & 2 I ~amDate: i I 12/21/20151 I [KA: 1 oooo6aA2o7 I~AA2.07 JIRova1ue:-il 4.1![Rova1ue) 4.3l[S!CtiO.nJl~IRoGroup:[I 21[SFio~I 21 !IE ~ [SYStemTE:volution Title_j j_c_o_n_tr_o_IR_o_o_m_E_v_ac_u_a_ti_on_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___..1 1~-- ] ~~tement:: Abilit to determine and inter ret the followin as the a pl to Control Room Evacuation: PZR level ]Explanation of\ 55.43.b(S) This question m_eets SRO o~ly criteria listed in NURE?-1021, ES-401, ~ttachm_ent 2, _11.E. Figure 2, 1~t bullet for 5th 1Answers: ___J Block for assessment and 1mplementat1on of a procedure or section of procedure 1n knowing which procedure will be used to
- evacuate the control room, and how the RO will maintain PZR level, and where it will be indicated. AB.CR-1 contains field actions which will be directed by the CRS. AB.CR-3 for toxic gas directs the control room evacuation to occur using AB.CR-1, but does not . . .
I ---- . Reference Title 1 Facility ~eference Number
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i [Reference Section [i Pagl:!~ !Revisioni IControl Room Evacuation 11 S2.0P-AB.CR-0001 I II 1122 I IControl Room Habitability 11 SC.OP-AB.CR-0003 I ii ii 6 I I II I 11 II I !Lo. Number __J Objectives IABCR01 E003 j_ ____. [Material Required for Examination 11 q !Question Source] New I I[Question Modification Method: JI I [Used During Training Program DI 1 Question__~C>.~ice Comment~ I I IC011 .... ~*** I I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes IQuestion Topic] ISRO 7 I Given the following conditions:
- A SBLOCA has occurred on Unit 2 outside containment. - Actions of EOP-LOCA-6, LOCA OUTSIDE CONTAINMENT, have failed to isolate the break. - RCS pressure is 1440 psig and continues to lower.
Which of the following identifies the procedure that will be used upon transition from EOP-LOCA-6 and the actions that will be directed in that procedure after the transition?
~ EOP-LOCA-1 Loss of Reactor Coolant. Add makeup to RWST, initiate a cooldown, and minimize injection flow.
I
~ EOP-LOCA-5 Loss of Emergency Coolant Recirculation. Add makeup to RWST, initiate a cooldown, and minimize injection flow.
Fl -- _ -1, Loss of Reactor Coolant. Check for subsequent failure and conserve makeup inventory. id i EOP-LOCA-5, Loss of Emergency Coolant Recirculation. Check for subsequent failure and conserve makeup inventory. L_"_ ~nswe_r 11 b l ~am Leven I s I rcognitive Level *IApplication I :_F~cilit~: 11 Salem 1 & 2 I fE:xamDate: 11 12/21/20151 iKA:ll OOWE04A201 I:EA2:-1=~=J IRo Value: II 3.4!:SRO Value)~ 'Section: 1_§.~.§_ _JRO Group:il 11 SRO Group:il 1 1 11 li:~!,i ~ IKA Statement:* !Explanation of j 1Answe~__1 =====--~~e~~e_!i_!l_E!______________ Ji Facility Reference Numbe!:::J Reference Section _ J [~~!!_':_~~ ~~ ILOSS OF EMERGENCY COOLANT RECIRCU 112-EOP-LOCA-5 I I !::::12=5== ILOCA Outside Containment Ij 2-EOP-LOCA-6 I ! I21
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IL.0. Number 008 !Material Required for Examination 11
'I I
!Question Source: 11 Previous 2 NRC Exams !Question Modification Method: i Direct From Source 1lused During Training Program I D
;Question Source Comm~ 113-01 NRC SRO exam (Dec. 2014) --~~~*~-*~~~~~*--
I 1Comment I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [auestion Top~ ISRO 8 Given the following conditions:
- Unit 1 has experienced a Main Steam line Break (MSLB) inside containment from 100% power. - Main Steam Line Isolation has failed to close any MS167. - Safety Injection was manually initiated, with all components operating as expected. - 11 AFP is err. - 12 and 13 AFP's tripped after starting. - Reactor Coolant System pressure is 1100 psig. - All Reactor Coolant Pumps have been tripped. - Containment pressure is 16 psig and rising. - All Wide Range (WR) Steam Generator (SG) levels are 35% and dropping. - All Steam Generator pressures are 465 psig and dropping. - Reactor Coolant System Tc's have dropped from 540 to 438°F in 40 minutes.
If these condition were present when transitioning out of 1-EOP-TRIP-1, Reactor Trip Response, which procedure must be entered and what action must be taken upon that transition?
~ 11-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions. Reset Safeguards Actuation and restore normal charging and letdown. ~b~j , 1-EOP-FRHS-1, Response to Loss of Secondary Heat Sink. Initiate feed and bleed ONLY when 3/4 SG WR levels have dropped <32%.
ic.: 11-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions. Shut all MS10's and steam dump valves .
.... ....... ............ ... ...
[ ] 11-EOP-FRHS-1, Response to Loss of Secondary Heat Sink. Initiate feed and bleed immediately. ]An~~lll/~~ ~ ~ll'l L~~I_ '~ I !Cognitive Lev!l_J IApplication I iFacmtYJ ISalem 1 & 2 I 'ExamDa~i:_:J I
~
IS 12/21/20151 ~JI oowEo5A202 I[EA2.2 - -1 IRo value: J[EJ[sRo va1u~~~H~ [RociroUP:lLJ[Sifo-G"roup:] LJ Ill ~ ~-Y~!em/Evolution 1:i~1eJ ILoss of Secondary Heat Sink I [Eo§"~ [KA Statement: 11 Ability to determine and interoret the followina as they apply to Loss of Secondary Heat Sink: I * ..
*- - procedures and operation within the limitations in the facilitv's license and amendments.
1Explanation of 155.43(5) 55.43.b(5) This question is SRO level based on having to asses the conditions given in stem and determine both what IAnswer_s: procedure will be entered and the action(s) required in that procedure, Dis correct because the conditions given in stem would transition to FRHS-1 due to a RED path of no AFW flow and <9% NR level. The Bleed and Feed initiation criteria are when SIG WR levels are <36%(adverse), NOT 32% (normal) as in distracter B. Distracters A & Care incorrect because it is a lower priority oc:n .......&.t... ... ._ _ _ ._ if 1 C" --"=-- ir ______ ... .C-- ... .__ -----...J,,,.n
' .. ~**-
Reference Title *1 [ __Facility Reference-f>,lllmber I [Refere_n~e section ! [Page No:_l JRevisionj IResponse to Loss of Secondary Heat Sink 111-EOP-FRHS-1 II ii lj 21 I IResponse to Imminent Pressurized Thermal Sh 111-EOP-FRTS-1 II II 1122 i ICritical Safety Function Status Trees 111-EOP-CFST-1 II q 1122 I l':.:0. Number Objectives I FRHSOOE005 I FRHSOOE013 j_ ___,
IMateria!_Required for Examination 11 II §uestion So~rc~LJ lFacility Exam Bank I[Question Modification ~_:thod: JDirect From Source I!Used D_u.r:~~ Training Program ! D ~E)stion Source C~~-~ 143189, Used on Salem SRO NRC exam more than 2 exams ago. I I Comment I I I I I I I
RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
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l§:i.les.E?~!~~J 1 _s_R_o~9~------------------------------------------~ Given the following conditions:
- Unit 2 is responding to a degraded core cooling condition in accordance with 2-EOP-FRCC-2, "Response to Degraded Core Cooling." - After depressurizing to inject the accumulators, the STA reports a RED priority on the Thermal Shock Status Tree, and recommends transitioning to 2-EOP-FRTS-1, "Response to Imminent Pressurized Thermal Shock."
Which actions describe the o erator response?
~ Do NOT implement 2-EOP-FRTS-1 until 2-EOP-FRCC-2 is completed because thermal shock is a lower priority CFST.
P-FRTS-1 immediately because the potential damage done to the RPV by delaying entry into FRTS after 1ent:rv 1cor1dition1s <:ire I arable.
~ Do NOT implement 2-EOP-FRTS-1 until 2-EOP-FRCC-2 is completed because while in FRTS the core will continue to boil a*wa*v in1iectedl I accumulator water, and could lead to a RED path for Core Cooling.
f d:! Implement 2-EOP-FRTS-1 immediately because it is a higher priority CFST and rules of usage stipulate the transition to a higher priority procedure always takes precedence over notes and cautions. [Answer I~ rexam*LeVel Is I [Cognitive Level __ IMemory I [Facility_:_i ISalem 1 & 2 I ;~x~J I 12/21/20151 ~I oowE06G423 I[2~4-:-23 *u -~~-~!!~.!:J [ ] ] i s Ro Valuej~ [seciiOO:l I~ RO Group:[[] ~~~_!>3 [ ] [hstem/Evolution Title j j_D_e""'g_ra_d_e_d_C_o_re_c_o_ol_in..;;:g'--------------------------------' Knowled e of the bases for rioritizin ,Explanation of j 55.43.b(5) This question is SRO level based on knowledge of diagnostic steps and decision points in EOP that involve transition !Answers: , points to event specific emergency contingency procedures. Knowledge of the bases for when NOT to implement a RED priority CFST even when indications are present is just as importanct as knowing when to implement. in this case to preserve and inventory in the reactor pressure vessel. Stopping the depressurization to go to FRTS would cause the cooldown to be stopped, and a
. . . . -- -.
Eventually, CETs and/or RVLIS level values could exist which would require transition to FRCC-1 via a RED path CFST. The
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Isto pin of the cooldown could lead to a de raded core coolin condition to deteriorate to an inadequate core coolin condition. Reference Title i[ Faci_litx_Re!e~~~ce-Number I ;Reference Section J [f~J ~~ IResponse to Degraded Core Cooling
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=

=

=

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I21
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112-EOP-FRCC-2
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l~-----------------'1----------------' --------~ ------'1---~ 11..:0. Number~ Objectives I FRCCOOE006 '--~ !Material Required for Examinat~.'_"! _ _j I !Question Source: i IFacility Exam Bank I!Question Modificati~n Method: sed During Training ~~~!-~rrl__J D [auestion source Comments:
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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question_jii>ICJ !SRO 10 I Given the following conditions:
- Salem Unit 2 tripped from 100% when a loss of off-site power occurred. - All systems respond as expected, and operators progress normally through the EOP's. - It has been determined a rapid natural circulation cooldown will be performed.
Which of the following describes procedures/actions which are REQUIRED to be performed PRIOR to transitioning to either 2-EOP-TRIP-5, Natural Circulation Rapid Cooldown Without RVLIS, or 2-EOP-TRIP-6, Natural Circulation Rapid Cooldown With RVLIS, and the bases for them?
~ 12-EOP-TRIP-2, Rx Trip Response, must be performed to ensure adequate SDM and natural circulation have been established, containment
cooling remains in service, and equipment not needed for the cooldown has been secured. I
~ 12-EOP-TRIP-4, Natural Circulation Cooldown, must be performed to ensure adequate SDM and upper head cooling have been established, SI signals have been blocked, and initial cooldown/depressurization have been performed. I ~ 12-EOP-TRIP-4, SI Initiation Criteria step must be performed to ensure SI will not be required prior to blocking SI before a cooldown can be initiated.
[ ] 12-EOP-TRIP-2, SI Initiation Criteria step must be performed to ensure SI will not be required prior to blocking SI before a cooldown initiated. ~~swe-r] Ib I rexam Level 1 Is I !cognitive Level 11 Memory I ~ciiityl ISalem 1 & 2 l~xa~~I 12/21/20151
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~I oowE10A202 I~~-__]~ Vatue]~!SRo Valuef~ [sectio!l:JI~ IRo Group::LJ~~LJ ~§'I' ~ lsystem/EvolUtiOr!Title-] INatural Circulation with Steam Void in Vessel with/without RVLIS I i-E:1°==J [1-_<A Statem-ent:] o determine and inter ret the followin as the a I to Natural Circulation with Steam Void in Vessel with/without RVLIS: Explanation of :
Answers: i
_ ~~ference Title --~] ~ittty-Refen~~ce Numb~~] [Retere~ce ~~ction 1 [F>_~~ [R.evi5i0n1 INatural Circulation Rapid Cooldown with RVLIS 112-EOP-TRIP-6 !=======:::::::::!i.==== 2=3==:::. i.::j INatural Circulation Rapid Cooldown without RVLll 2-EOP-TRIP-5 I23 !~~~~~~~~~~~~~~~~~--'!-~~~~~~~~~~~--' ~~~~~~~~~ -~~--'~, ::::::::::::::: [L..o. Num~~r ~ Objectives TRP004E005 I I I iMaterial Required for Examinati~ I II e~:stion Source: J IFacility Exam Bank I[Question Modification Method: _ ~ Direct From Source I!Used Duri~g Training Program I D !Question S~~~ce Comments! 1153885 I I !--******-*" -- I I I I I I I
RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
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1auestion Topic I _s_R_o_1_1_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___. Given the following conditions:
- Unit 2 is operating at 75% power. - 21 Waste Holdup Tank (WHUT) is being processed via the Portable Liquid Radwaste Processing System. - Console alarm SURGE TANK LEVEL HI-LO for the CC system alarms. - Surge tank levels are 59.3% and 59.2% on Channels A and B, and rising slowly. - No other alarms are present.
Which of the following describes the effect if Surge Tank Level were to continue to rise, and what action the control room will take in response to the risin level? 2CC149 CC SURGE TANK VENT VLV will auto close to prevent overflow. Stop processing 21 WHUT as directed in the S2.0P-AR.ZZ-0011, Control Console 2CC1. Overflow of the Surge Tank will contaminate the Waste Holdup System with chromates. Close 2CC149 CC SURGE TANK VENT VLV 2CC2 as directed in the S2.0P-AR.ZZ-0011, Control Console 2CC1. 9 CC SURGE TANK VENT VLV will auto close to prevent overflow. Open the CC Surge Tank Drain Valve from 2CC1 asnec:es:sary tc1 I evel <100% as directed in S2.0P-AB.CC-0001, Component Cooling Abnormality. Overflow of the Surge Tank will contaminate the Waste Holdup System with chromates. Direct a NEO to locally drain the Surge Tank to 55 gallon drum as necessary to maintain Surge Tank Level <100% as directed in S2.0P-AB.CC-0001, Component Cooling Abnormality. [A~ ~ iEXM1 Level ] @=:] cognitive Level~: IApplication I I [~C:ili~~:_J ISalem 1 & 2 I IExamDatej I ~I 008000A202 IiA2.02 I [Rova!Ue:]@~Ro vaiuej~ !section: ii~ [Ri:[Grc>~P:1LJ ~~!~~LJ
system/Evolution Title I I Component Cooling Water System
KA Statement: I Ability to (a) predict the impacts of the following on the Component Cooling Water System and (b) based on those predictions, use es to correct, control, or mitigate the conseauences of those abnormal operation:
Hiah/low surae tank level !Explanation of I 55.43.b(5) This question is SRO level based on having to asses the conditions given in stem and determine which action in which Answers: procedure will be taken. The 2CC149 auto closes on hi radiation in the CCW systen, nnot high pressure, to prevent radiation release from the CCW system. The Surge Tank will be locally drained. The overflow from the tank will go to the Waste Holdup system and contaminate the WHUT. WHUT processing will be stopped in both the AB.CC and ARP procedures. Locally draining
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*---*1 [*--Facility Reference_ Number *1 [Pa9;NOJ iRevi~~~ -----~---------
Reference Title rn I lReference Section I u-*-~-*---* Component Cooling Abnormality 11 S2.0P-AB.CC-0001 II 11 1114 I I Control Console 2CC1 11 S2.0P-AR.ZZ-0011 11 11 1160 I 1205331 11205331 Sheet 1 II 11 I 154 I [Lo. Number Objectives CCWOOOE008 IMaterial Required for Examination
~- JI II ~~~stion Source: [ j Facility Exam Bank I~uestion Modification Method: II Concept Used I[Used During Training Program I ~ ~~~~~~~Source Comments[ 182785 modified to make SRO level and fit KA.
I 1: . . . ,.......... \ I I I I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I
@luestion Topic I ISRO 12 I Given the following conditions: - Unit 2 is operating at 100% power. - 2PR1 fails open and remains open.
Which of the following identifies how this affects the PZR Master Pressure Controller (MPC) response, and what consequences, if any, are associated with the actions performed by the crew IAW S2.0P-AB.PZR-0001, Pressurizer Pressure Malfunction?
~I MPC output will RISE. A unit shutdown will be required if 2PR1 cannot be restored to operable status within 72 hours.
I
~ MPC output will LOWER. A unit shutdown will be required if 2PR1 cannot be restored to operable status within 72 hours.
I
~ MPC output will RISE. The unit may continue to operate indefinitely after the initial mitigative actions are completed.
I I
~ MPC output will LOWER. The unit may continue to operate indefinitely after the initial mitigative actions are completed.
I ~nswer J Ib I IExam Level 11 S I 1cognitive Level JIApplication I 11 Facility: ; Salem 1 & 2 I IExam Date: 11 12/21/20151 [~I 01 OOOOA203 ~--------*1
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[~ystem/Evolution Title 1.;..;..:...:..:..:;;,,;.;:.:.;_;_;_;.;;,,;,.;;.;..;_..:..;,...;.;..:;.......:,.;::..:...;.;,;.;.;__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___J 1010 i - .--------------, ....------------------------------------------------: KA Statement: I Ability to (a) predict the impacts of the following on the Pressurizer Pressure Control System and (b) based on those predictions, use rocedures to correct, control, or miti ate the consequences of those abnormal operation: PORV failures Explanation of 55.43(2) This question is SRO level because of the Tech Spec knowledge required, and what actions TS directs for different PORV 1 Answers: 1 malfunctions. Additionally, while the question doesn't specifically ask what procedure to use (too easy for AB.PZR), it does require knowledge of the actions IN that procedure. The MPC raises output when actual pressure rises, and lowers as actual pressure lowers. As pressure lowers due to the open PORV, the output will lower to turn on heaters and close spray valves. When the PORV
. . . . . .
is required. A PORV isolation that DOESN'T require shutdown if not fixed is aleaking PORV, which is isolated by its Block Valve I with ower maintained to the Block valve. [L.~o. Number Objectives I ~1E002 LE010 I Material Required for Examinatio_n ___J I II !Question Source: I I Facility Exam Bank 1IOuestion Modification Method: II Direct From Source 1lused During Training Program I D !Question Source Comment~ 1125725 used on Salem SRO NRC exam more than 2 exams ago I !Comment l I I I
RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I r*--------- ----- -* ISRO 13
;Question Topic; I Given the following conditions: - Unit 2 is conducting a rapid plant shutdown due to a loss of condenser vacuum. - Two IRPls in control bank D go dark, and their P-250 readings are O steps. - Maintenance Controls Dept. will be unable to repair the IRPls until after shutdown. - Rx Engineering has made no specific recommendations outside of procedural direction regarding boration.
Which action is required when the plant is placed in Hot Standby?
~I Borate to Cold Shutdown SOM to prevent a Yellow Path on FRSM from occurring during the shutdown.
I
~I Borate an additional 540 ppm to prevent a Yellow Path on FRSM from occurring during the shutdown.
I I
~ Borate an additional 540 ppm since it is assumed that the reactivity associated with the affected rods is unavailable for shutdown. @] IBorate to Cold Shutdown SOM since it is assumed that the reactivity associated with the affected rods is unavailable for shutdown. I !
~nswe~ Ic I I Exam Level 11 s I IC?9_ni~f/eiev~ I Application I fFacilitY:l ISalem 1 & 2 I [Exa_rnl)ate:] I 12/21/201511 ~I 014000A202 I[A2m--Hl ~~!~~J [TI" is Ro Valuej~ 'section: 11~ IRO Group:1 LI jsRO Group:I LI I& li'l !system/Evolution Title] IRod Position Indication System I r014- ---] l~ S~~mef'l!:J ~ity to (a) predict the impacts of the following on the Rod Position Indication System and (b) based on those predictions, use cedures to correct, control, or mitigate the consequences of those abnormal operation: Loss of oower to the RPIS lExplanation of j 55.43.b(6,5) This question is SRO level based on having to determine the effect on core reactivity based on the 2 control rods which !~~~ers:___ _J have to be assumed to remain fully withdrawn in the absence of IRPI indication. Additionally, the sRO must select the portion of AB.ROD-4 (CAS action 2.0) which requires an additional 270 ppm boration for each failed IRPI if a shutdown is performed before
!
the IRPI is declared operable. Boration to cold shutdown conditions is not required
. ' :+h ~u~~*~'"' -~~t~\ :~1~-~ +h~-~h* * ,__ for a shutdown to hot standby. It assumes the ,; ,_ ~~-~ I i Reference Title - __J L Facility Reference Nu.mber _ ~eference ~~(;~i?':_l J --~,c=~~~~()J §.evi~()_l"lj IRod Position Indication- Failure LO**-*-*- .,,._, *-
11 S2.0P-AB.ROD-0004 I ii lj 10 I I II I 11 11 I I II I 11 11 I !Lo. Number .......... "*-*-*-****-*--*-*----- Objectives ! ABROD4E002 ,_ _ ___. !Material Required for Examination :I II iQuestion Source: 1 IFacility Exam Bank I!Question Modification Method: i 1 lused During Training Program I D Question Sourc~~~~m4:_~~ 1140829 1 I IC"'"'**v'" I I I I I I I
!Material Required for Examination I I I I IQuest~?~ Source: i New IlCl~~estion Modifi~~tion Method: i I lUsed During Training Program D I I
~
[Ques~!?n source comme~§ I i.--*<111**-llL l I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic 11 SRO 16 I Given the following conditions:
- Unit 1 was operating at 100% power when a catastrophic failure of 12 SG steam piping occurred inside containment. - The Rx tripped and SI initiated. - 12 SG rapidly blew down to containment, and containment pressure peaked at 16 psig prior to transition out of EOP-TRIP-1 Reactor Trip or Safety Injection to EOP-LOSC-1, Loss of Secondary Coolant. - Operators have just transitioned out of LOSC-1 to EOP-LOCA-1, Loss of Reactor Coolant, with the following conditions: - RCS pressure is 1780 psig and rising slowly. - RCS subcooling is 100°F. - PZR level is 22% and stable. - 11, 13, and 14 SG levels are 17% and rising slowly.
Which of the following describes containment cooling operation during subsequent EOP performance? Containment Sorav will be secured in ......
!a.J i --~' 4 , Post LOCA Cooldown and Depressurization, when containment pressure is < 4 psig. Containment Spray Additive Tank will be ~- * *- ... based on normal radiation levels in containment.
[b:1 I EOP-TRIP-3, SI Termination, when containment pressure is< 4 psig. Containment Spray Additive Tank will be isolated based on normal radiation levels in containment.
~ I EOP-TRIP-3, SI Termination if containment pressure is less than 13 psig. CFCUs will continue to operate in Low Speed.
[11 LOCA-2, Post LOCA Cooldown and Depressurization. CFCUs will continue to operate in Low Speed.
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I iFaciiitY:l ISalem 1 & 2
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[Ans~~ I@=] ~ognitive Level I !Application I JExamDate: 11 _ _ _1_21_2_11_2_0_15_,I
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rexam Level ~1026000A208 I A2.08 _i~~~0]"~-~~-~-22l~!~~~]~~<?~~~~LJ~~LJ ~ [fyStiffiiEvolutioi:i_Title 1 l_c_o_n_ta_in_m_e_n_t_S_p_ra~y_S_y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~ IKA Stateme_1_1t_:J 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 mitii:iate the consequences of those abnormal operation: Safe securing of containment spray when it can be done) !Explanation of i 55.43.b(5) This question is SRO level based on having to asses the conditions given in stem and determine both what procedure ,Answ~!l>~ will be entered and how containment cooling/spray equipment will be operated. With conditions in stem, the transition out of LOCA-1 to TRIP-3 will be made at step 9 based on adequate subcooling, SG NR level status, and PZR level, otherwise the transition would be made to LOCA-2 at Step 18 with RCS pressure >420 psig. Both procedures have CS terminated. Neither procedure I
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[Lo. Num~er Objectives I CSPRAYE012 I TRP003E005 ,__ ___,
iMaterial Required for Examination ~1 'I /Question Source: 11 New
~~., ... " *-*~**J I [Question Modification Method:
L--~~~--- **-~~~--~*--** JI I [used ~ming Training_j:>~ogram ] D I-Question ~~urce Commen~~ I I
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RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
!Question Topic 11SRO17 I Given the following conditions: - Unit 2 is in Mode 5 entering a refueling outage. - 2C EOG was CIT yesterday for scheduled outage window. - The normal 31 day surveillance test of 2A EOG was completed SAT 2 days ago while in Mode 4. - Subsequently, the Unit 2 CRS notices that the 18 month Hot Restart surveillance for 2A EOG, was NOT performed as scheduled after the normal 2A EOG 31 day run surveillance, and is now outside its required periodicity including any allowable grace time.
Which of the following describes: I
1) The status of 2A EOG

2) When the associated Hot Restart test must be oerformed I
~ Operable. The Hot Restart test must be performed prior to entering Mode 4.
I
~I Operable. The Hot Restart test must be performed prior to entering Mode 6.
I I
~ Inoperable. The Hot Restart test must be performed prior to entering Mode 4.
I [dl I Inoperable. The Hot Restart test must be performed prior to entering Mode 6. I I *-- I IAAswer I' ~ LExam Level @=] ~ve Level l j Application j I [Facility: 11 Salem 1 & 2 I ~xamDate: J l___1_2_12_1_12_0_15_.I jKA:il 064000G120 I ~~-o-~I !Ro Value]~~~@~ [section: 11~ 'ROi3fOOP:10JsRo Group~O ~ [SYstem/Evolution Titl~ IEmergency Diesel Generators [KA Statement: I Abilitv to interoret and execute procedure steps. I \Explanation of I 55.43.b(2) This question is SRO level based on being able to apply TSAS 3.8.1.2 for electrical power in Modes 5 and 6, and the
Answer~ action required based on the conditions in the stem. RO knowledge would be "above the LCO line", and SRO level for actions below the line. LCO 3.8.1.2 requires 2 operable EDGs, and the stem states 2C is already tagged out. The surveillance requirements of 4.8.1.2 specifically state that certain surveillances are NOT required to maintain EOG operability. The bases for
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performance of the surveillance requirement, and to preclude de~energizing a required ESF bus or disconnecting a required offsite circuit during performance of surveillance requirements ...... It is the intent that these surveillance requirements must still be capable of being met, but actual performance is not required during periods when the DG and the offsite circuit are required to be operable. During startup, prior to entering Mode 4, the surveillance requirements are required to be completed if the surveillance frequency has been exceeded .... " The provided reference does not give answer, but allows the operator to interpret the surveillance requirement, and in any case does not contain the action required which is located in the bases. c:=-- ._. Reference Title
- ==i: Facility Referen_ce Number _J jReference--Section -11 Page No~ IRevi~!C>~
ISalem Tech Specs I lj 3.8.1.2 I13/4 8-7a i 1245 I ISalem Tech Specs I 11 Bases 11 B3/4 8-3 l I282 I I I n 11 II i iL.O. Number Objectives I EDGOOOE011 1-~
~terial Required for Examinatio.n j j SRO 017 Tech Spec 3/4.8.1 Electrical Power Systems, A.C Sources II l~uestion S_<>_urceJ INew I~estion Mod_ificationMethod: *1 I !Used During _Training Program I D ~ ~--~------ntS] stion Source _c~mments I I 1r~~---* ---111***-** .. I I I I
RO SkyScraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I lauestion Topic] ISRO 18 Given the following conditions:
- Unit 2 is operating at 60% power, steady state. - 21A Circulator is CIT. - The Condensate Polisher is O/S and in standby. - Subsequently, the 13 KV South ringbus breaker D-E opens causing a loss of 23 SPT.
Which of the following describes:
1) The plant status 15 minutes after the loss of 23 SPT

2) How the operators will respond to this event
~ 1) The reactor remains at power.
2) Operators will be performing S2.0P-AB.CW-0001, Circulating Water System Malfunction, and establishing Low Pressure Turbine Hood Sprav.
~ 1) The reactor was manually tripped.
2) Operators will be utilizing S2.0P-AB.COND-0001, Loss of Condenser Vacuum, in conjunction with the TRIP series of EOP's, to break condenser vacuum.
~] 1) The reactor remains at power.
2) Operators will be performing S2.0P-AB.CHEM-0001, Abnormal Secondary Chemistry, and placing the Condensate Polisher in service for the expected rise in Dissolved 02.
Jd.
1) The reactor was manually tripped.
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~ !exam Levei"I @=]" lcognitiveievel 11 Application I !Facility-:-] ISalem 1 & 2 I jExamDate: i I 12/21/20151 IKA:jlo7soooA202 l~~~~-*[Ro-vaiue:jl 2.s1[SRO-va1ue)-22JlsectfoniJl~[R§"Group:]I 21:sR0Group~I 21 l~~tJJ ~
[~ystem/Evolution Title I j_c_ir_c_u_la_tin.....;g:;.._W_a_te_r_S-'y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.I !075 iKA ~tatement: I Ability to (a) predict the impacts of the following on the Circulating Water System and (b) based on those predictions, use gcedures to correct, control, or mitiqate the consequences of those abnormal operation: ss of circulatinq water pumps lExplanation of] 55.43.b(5) This question is SRO level based on knowing the appropriate procedure or section of procedure used in applying the Answers: conditions which will occur based on the initial conditions in the stem. 23 SPT powers 23 CW bus, and 24 SPT powers 24 CW bus. --~-~-.-- 23 CW bus powers the "A" circulatros and 24 SPT powers the "B" circulators. The loss of 23 SPT will cause a loss of 2 additional circulators in addition to 21A which is CIT. The operator must know how many circulators are now not running, 3 or 4. If it is
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thought condenser vacuum would degrade or DO increased based on losing hotwell levels and its effect on condensate pump seals. Manually bypassing the Turbine Hood Spray is allowed (and directed) in AB.CW-1, even though normally it remains secured
> 15% power. (AB.CW-1 paqe 5)
C___ Reference Title I L Facility Reference _Number i ~eference section ] !f>-a9e No.1 ~~J ICirculating Water System Malfunction 11 S2.0P-AB.CW-0001 II qs I 135 I I II II ii II I I II II 11 11 I IL.O. Number Objectives I ABCW01 E004 1--~
lMaterial Required for Examination 11 II I IQues~()n Source: I New 1IQuesti~n Modifica!i~n Method: J I !Used During Training Program_! D [Qu~~tion source CommeniSl I I !c_...... _... . I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic 1 ISRO 19 Given the following conditions:
- Unit 2 is performing a reactor startup by control rods IAW S2.0P-IO.ZZ-0003, Hot Standby to Minimum Load. - Estimated Critical Conditions are: - Cb= 500 ppm - Control Bank D = 77 steps - Xe free - 11,756 EFPH When the ICRR value reaches 0.125, the Predicted Critical Rod Height is 122 steps on Control Bank D.
What action(s), if any, is/are required to be taken in response to this Predicted Critical Rod Height? I
~ Continue with no special actions are required.
I
~ Continue the reactor startup, and evaluate the post startup data for trend.
I [~ Initiate rapid boration, insert Control Rod Banks, and recalculate the ECC.
@] I Insert the Control Rod Banks and recalculate the ECC prior to withdrawing Control Rods. .. ..,
I ~~~ ISalem 1 & 2 !___12_12_1_12_0_15_.I
.
[Answ__~ ~ IExam Level J [I:J !Cognitive Level r 11 Application I !Exam Date: J ~I 194001G123 I E~...=:J ~vaiU-e]@sRo value)~ [See!ion:]j~ IRo Group:ILJ~~LJ ~ !system/Evolution Title I ~emelili] ----*-*- Ability to perform specific system and inteqrated plant procedures during all modes of plant operation. Iexplanation of: 55.43.b(6) Table 1-8 77 steps is 1079.4 pcm. 122 steps is 877.0 difference is 202.4 pcm. With <300 pcm difference between ECC jAnswers: ___ __/ and predicted at the eightfold position, there is no action, and the startup will continue with no additional action required. [ Reference "fitl~ --*----
Facility Reference Number I !Reference Section 11 Page NoJ §:eviSioi1:
ICurve Book 11 S2.RE-RA.ZZ-0016 I 11 11 8 I IHot Standby to Minimum Load I 1 S2.0P-IO.ZZ-0003 I 11 11 I I II I ii 11 I Objecti j_ ___, ]Material Required for Examination 11 SRO 19 S2.RE-RA.ZZ-0016 Rev. 8 II I !Question SourcE!:l Facility Exam Bank I[Question Modification Method: II Concept Used 1lused During Training Program I D [auestion Source Comments[ 1151642 Used 3 NRC exams ago Sept 2011. Different numbers based on different revision of Curve Book. I Icomment I I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic J 1SRO 20 Choose the answer which contains ONLY actions directed to be performed IAW S2.0P-IO.ZZ-0007, Cold Shutdown to Refueling, BEFORE Rx Vessel Head detensioning would be initiated for a refueling outage starting in November. I. The Rx shall be subcritical for at least 168 hours. II. Direct. continuous communication between the control room and refuel floor is established. Ill. The RCS is drained to <104' elevation. IV. Unit CRS AND SM approval.
~1 ~1111. IV. ~,I.Ill. @Jiii, iV.
iAnswer 11 b I [Exam Level i Is I [cognitive Level *11 Memory I ~~ci~!ty: J ISalem 1 & 2 I ~ll"ID~te: ~ I 12/21/20151 iKA:[l 194001G141 !f2.1.41 I !Ro Value: q 2.81 iSRO Value) I
3. 7 !section: 11 PWG I iROCfrou~ I 11 lsRo Group: II 11 lrfl D lsystem1Evotutionrit1e] I I 1GENERI I IKA Statement: I Knowledge of the refueling process. I
]Explanation of 1 55.43(6,2) This question is SRO level based on knowing the requirements which must be met before alterations affecting core Answers: configuration will be started entering a refueling outage. The reactor does not have to be subcritical for 168 hours (Oct 15-May 15th) prior to moving fuel in the reactor (TSAS 3.9.3) Direct communications is required during CORE ALTS, and detensioning the head is NOT core alts. I '----
*--*--*-**----*-
Reference Title ] ~_Faciiity Reference Number . J[R"it;;~;nce section 11 Page No.J !R.eviSion! ICold Shutdown to Refueling 11 S2.0P-IO.ZZ-0007 I II lj 11 I I II I 11 11 I I II I 11 II I 1L.O. Nu111ber Objectives IIOP007E002 '--~ Material Required for Examination l I II I Question Source: 11 Facility Exam Bank I!Question Modification Method: II Editorially Modified I!Used During Training Program I D
!Question-~~!~.: Commentsj I I -.-~
[comment .. I I I I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I
~estion Topicl I SRO 21 I Given the following conditions: - Unit 2 is in MODE 6 with core reload in progress. - 10 fuel assemblies have been moved into the Rx. - Rx cavity level is 26' above the RPV flange. - 21 RHR loop is in service in Shutdown Cooling. - 22 RHR loop is O/S and available.
Which one of the followinq would prevent continuation of fuel movement into the reactor?
~ I Loss of Control Air to containment.
I I
~ Racking down the 22 RHR pump 4KV breaker.
I
;c:f I Both 100' elevation containment airlock doors are opened. ~I With both SRNls operable, only ONE is capable of providing audible indication in the control room.
~~-~w~ Ia I !Exam Level 11 s I ~?gnitive Level I i Application I Facility: I J 1Salem 1 & 2 I rexamDate: i I 12/21/20151
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!KA:Ji 194001G201 I!_~-~-- ROValue:il
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4.5ltSRO~~~iSectiorliJIPWG liROGroup~j 1jjSROGroup:JI 11 1111 ~ jSystem/Evolution Title '.
......-----------------------------------------,. --*-------
1 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' IGENERI I ~temerlt] *-*-
lily to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that uld affect reactivity.
rplanation of Answers: l 55.43(7, 6) This question is SRO level because of the knowledge require for fuel handling procedures, and the ability to continuously apply that knowledge when operating the fuel handling equipment. The requirement for SRNl's is BOTH operable and providing VISUAL indication in the Control room, with ONE providing AUDIBLE indication in the control room. The manipulator crane is air powered for gripping, so the loss of air to containment would preclude being able to perform core alts. Only ONE RHR
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.~--* ---- Reference Title --1[ Facility Refe~ence-Number J IRefere~ce Section 1 lPage No] ~<:>-'!
----*--**
IRefueling Operations
c. -
11 S2.0P-SO.SF-0009 I 11 1118 I IReac Pene Area & Cont Control Air 11205347-1,3 I 11 1142,36 I I II l 11 11 I 1 L.o. Number Objectives IREFUELE007 I IOP009E002 '--~ Material L......__.. . Required for- Examination iI I II [Question I Sourc~ Facility Exam Bank I!Question Modification Method: JI Editorially Modified !tUsed During Training Progra.tll D fQuistion S-~-~rce Comments! 1125721 changed to reload from defuel. Used on Salem SRO NRC exam more than 2 exams ago. I IC-..i****-* I I I I
RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
!Question Topic I ,_s_R_o_2_2_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~
IWhich of the following Salem events requires a 15 minute notification?
~ A unit shutdown is initiated to comply with Technical Specifications. ~ An oil discharge directly into the Delaware River with a visible sheen. ~ An endangered species of sturgeon is found deceased during circ water trash rack cleaning. @J Rx power is determined to be greater than 3459 MWth after removing a Feedwater Heater from service.
12/21/20151 ~I 194001G238 I[2.2.38 I !Ro Value: !0"[S[ovaiUeL.~.~ Section: lsystem/Evolution::::+/-l!'!J ~tatement* _ _ _ _ _ _ _'.'._Ji Knowledge of conditions and limitations in the facility license. IExplanation of [ 55.43.b(1) This question is SRO level based on the knowledge required of Conditions and Limitation required in the facility license, 1Answers: 1 and meets the criteria in ES-401, Attachment 2, page 17, II.A, 4th bullet. Knowledge of events which require 15 minute notifications is critical for SRO to know from memory to faciliate timely and correct notifications. The question matches the KIA because Salem Unit 2 Facility Operating License, Renewed License No. DPR-75, Section C.2, Technical Specifications and Environmental
- '"' - -*-*-- --' -- *--'"" ..,;u ha ---*-*--' I" \Al . *'-- - - '""-- <>C> ---*-*--..< ;n A .. . 0 Appendix B, Section 4.1, Unusual or Important Environmental Events, staes that the NRC must be notified of ... "Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation ... ". RAL 11.5.2.b contains conditions for which an oil spill is reportable in 15 minutes. The distracters are all reportable, A is 11.1.1.a (4 hour report) C is 11.5.2.c (4 or 24 hour report) and Dis at least a one hour report. Question is balanced with 2 environmental choices and 2 Tech Spec choices. - - - - * - * - - - - * [-----
I Reference Title _ _ j F~cility Refert:!!'ce_ Number 11 Reference Section [I Page No. [ [Revision! lsalem ECG 11 ii RAL 11.5.2.b II 11 4 I lsalem ECG II ljAttachment 16 114-5 11 4 I ISalem Operating License II II ii II I [f.o. Number Objectives I EL0_11.d , __ __,
!Material Required for Examination 11 II [Questi~_ll_S.~U.~~~~J INew I;Question
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Modification Method: i I !Used During Training Program I D I
~
[Question Source Comments: I I 1=-*llll*v**" I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I
!Question Topic ] SRO 23I I Given the following conditions: - Unit 2 is operating at 100% power. - The operating crew entered S2.0P-AB.RC-0002, High Activity in Reactor Coolant System, when RMS channel 2R31, Letdown Line Monitor, went into WARNING.
Which of the following is required to be performed IAW S2.0P-AB.RC-0002, and why? The CRS will. .. I
~ direct Radiation Protection Technician to take surveys to determine if radiation levels may have changed access requirements.
I I
~ direct Radiation Protection Technician to survey the letdown line in the vicinity of 2R31 to confirm the suspected rise in RCS activity.
I I
~ direct Chemistry Technician to sample hourly for isotopic analysis to determine predominant radiation hazard (gamma, neutron, beta, alpha).
I l<1J I direct Chemistry Technician to initiate confirmatory sample analysis beca .. - *- .. - - - .. level changes. I !Answer 1 a I I iExam Lev~ IS I LCog~itive Level_ J IMemory I [Facilfty;**11 Salem 1 & 2 I IExamDate: t I 12/21/20151 ~I 194001G314 I 2:~-~4 --,,Ro~l 3.4l[SROValueJ 3.aj[Section:JIPWG IRq_~!~~JI 11lsROGroup:il 1j IJ~i ~
! - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' l<3.~~~~Q rc;-*-----~---1 LSystem/Evolution Title IKA Statement]
Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emeraencv conditions or activities. !Explanation o~ 55.43(4) This question is SRO level based on the "why" the action will have to be taken, as knowledge of radiation hazards which is ,Answers: occurring during an abnormal situation. A is correct as described in S2.0P-AB.RC-002 basis, so that prompt identification and subsequent notification of plant personnel is ensured. B is incorrect because chemistry sampling confirms 2R31 readings, not survey results. D is incorrect because rising counts does indicate dose level changes. C is incorrect because the hourly isotopic
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Reference Title i I Facility Reference Number . J[~!!~i

..e!1£..!..~..£..~~ [PaQe-No. i :Revision 1 I High Activity in Reactor Coolant System 11 S2.0P-AB.RC-0002 II 11 I 1 a I I II II 11 iI I
! II II 11 11 I Lo. Number___ 1 Objectives ABRC02E003 I I I !,Material Required for Examination II II [auestion source: 1 IFacility Exam Bank I!Question Modification Method: II Direct From Source I[iiSed During Training Program ] D I
~estion Source Commentsj Used on Salem SRO NRC exam more than 2 exams ago.
I I <111*t-lll I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
!Question Topi~J ISRO 24 I Given the following conditions: - At time T-0, while operating at 100% power, the Auxilary Typewriter prints an alarm without OHA A-41, AUX ALM SYS PRINT annunciating. Operators immediately enter S2.0P-AB.ANN-0001, Loss of Overhead Annunciator System and begin assessing the OHA system functionality. - A T+5 minutes, and prior to verifying if either SER is in command, the Rx trips. - At T +14 minutes, the PO reports that neither SER is in command.
Which of the following identifies:
1) The correct ECG classification

2) The time at which it should have been declared?
~ 11) Alert
_, . 2) T +5 minutes. I
~ 11) Alert . 2) T+14 minutes.
I jc., 11) Unusual Event
~ 2) T +5 minutes.
[ ] 11) Unusual Event
-- 2)T+14minutes.
IA"!::~*-"'!' I r;:==J rs=::::I 10e?~!t!*!e ' ::*:e! I I _\nn!i<;;'>ti0R L____~------~
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! iFacilit : Salem 1 & 2 ...tExamDate: 12/21/2015 I
lKA:jj 194001G432 112.4.32
~ystem/Evolution Title]
IKA statement: I . . . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . , . e of operator res onse to loss of all annunciators. Planationo~ 55.43.b(1) This question is SRO level based on the requirement to declare an emergency. The requirement to declare an wers: emergency under any of the loss of annunciators (S5) EALs is the loss for greater than or equal to 15 minutes. Declaring at T +5
--------
minutes with 10 minutes of assessment time left is not indicated even though the significant transient (Rx trip) has occurred, because the AB.ANN takes actions in an attempt to restore functionality of OHA system, and it is not a long drawn out procedure.
. . . .
I Reference Title 11 Facility Reference Number I :Reference Section 11 Page No. I [Revision] lsalem ECG I I 1 S5 - Instrumentation II 11 I I I II 11 11 I I I II ii 11 I jL.o. Number___ ! Objectives IABANN1E002 I I,_ _ ___.
!Material Required for Examination : ISRO 024 ECG Section S5 I !Question Source~ INew I!Question Modification Method: ~ 1lused During Training Program I D lQuestion Source Comments] I I ,-------- ---
jComment I I I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question Topic ! SRO 25 I I IWhich of the following identifies a condition during a declared Emergency which REQUIRES a Protective Action Recommendation, either initial or upgrade, when the TSC/EOF have NOT been activated yet? I I
~J ANY General Emergency initial declaration.
I l§J I . - - **- esults in a radiological release to the environment. I I
~ ANY time the wind shifts after the initial 15 minute notifications have been made during a General Emergency.
I IC!~ ANY event which in the judgement of the Emergency Coordinator could result in exceeding 1OCFR Part 100 limits. !Answer J Ia I Exam Levell j S I ~ognitive Level JI Memory I !Facility: [ ISalem 1 & 2 I ;ExamDate: 11 12/21/2015 [KA:!l 194001G444 I.2.4.44 . .,IROValue;*Jl 2.4!,SR~~[Section:]jPwG I ROGroup:ll 1j[SRO-Group:JI 11 Bia D !system/Evolution Titl~] ,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____, IGENERI I I KA Statement-:; Knowledqe of emerqencv plan protective action recommendations. [Explanation of I 55.43.b(5, 1) This question is SRO level based on the knowledge of how to implement the associated section of the ECG, EP-SA-
~~vyer~~ 111-F4, Attachment 4, General Emergency, and the conditions which require making a PAR or PAR upgrade. The ICMF for a GE requires a PAR, see Appendix 1. There is a PAR for a Rapidly Progressing Severe Accident, a PAR for Hostile Action, and a defualt PAR, one of which must be made. Bis incorrect because a radiological release is defined as "Any release above normal,
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for a wind shift (see Appendix 1). Dis incorrect because there is no "judgement" directed in the PAR based-on exceeding radiation limits, the PAR is determined based on the the GE that directed its implementation .
...-.. . ---****~-.,
Reference Title . 11 Facility Reference Number J :~i:,terence Section I; Pa9."'..~~ iRevisionl IGeneral Emergency ~ .-.~
-- ,.,u 11 EP-SA-111-F4 II 11 1102 I I II II 11 11 I I II II 11 11 I ILo. Number Objectives I EL0_24.a
!Material Required for Examination 11 II [Question Source: Ii New I I!Question Modifica~i-~n Method: II I fused Duri_n!;J Training Program.1 D 1 !Question Source Commentsj I I 1c._., ... ,...,.otL
I I
I I
U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: Region: I Date: 12/21/2015 Facility: Salem 1 & 2 - License Level: RO Reactor Type: W Start Time: Finish Time: Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected SIX hours after the examination starts. Applicant Certification All work done on this examination is my own. I have neither given nor received aid. Applicant's Signature
..
Results Examination Value Points Applicant's Score Points Applicant's Grade Percent
RO SkyScraper
... . .
I..... SRO Skyscraper I
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RO System/Evolution List I . .. SRO System/Evolution List I Outline Changes
....
I fouestion Topic I IRO 1 I Unit 2 is operating at 60% power with rod control in auto when Turbine Steamline Inlet Pressure transmitter 2-PT-505 fails high over 10 seconds. Which of the following describes the effect on the rod control system from this failure? Control rods will move in the ___ direction, and the Power Mismatch signal will initially cause rods to move_ _ rapidly than if the circuit was not part of rod control.
~1 I ~ Iinward, more.
I QI
~ outward, less.
I []I outward, more. I !Answer 11 d I !Exam Leve'i] I R I cognitive Level i IApplication I !Facility: 11 Salem 1 & 2 I IExam Date: 11 12/21/20151 IKA:ll 000001K105 j IAK1.05 I !Ro Vcillle:JI 3.51 lsRO Vallie: i 3.81!Section:11~ [ROGroup:!j 21 !sRo Group:JI 21 lil~ D !System/Evolution Title] IContinuous Rod Withdrawal 11001 IKA Statement: I~e of the operational implications of the following concepts as they apply to Continuous Rod Withdrawal: I turbine-reactor power mismatch on rod control I !Explanation of 55.41.b(6) The power mismatch section of the Reactor Control Unit compares the difference in the rate of change of turbine power [Arisl/Vers: *.* and rx power, and adds a signal to the temperature error circuit. With turbine power (PT-505) rising rapidly over 10 seconds and rx power not changing at all, the power mismatch signal will cause control rods to move out at a higher rate than what would be expeected from the Terr signal alone, in an attempt to raise Rx power at a rate similar to what turbine power was doing. I Reference Title ii Facility Reference Number IIReferef)ce Section i IPage No. I rRevision! !Rod Control Lesson Plan 11 NOS05RODSOO I 1j2a 1112 I I II I II ,II, I I ii I 11 I fL.O. Number i Objective I RODSOOE006 1_ _____, !Material Required for EXarf\fr1ation ' 11 II IQueition source: i l_N_e_w_ _ _ _ _ _~I Llo~u~e~s_ti_o_n_M_o_d_if_ic_a_t_io_n~M_et_h_o_d~=~*.Jil1, __________.I !used Dui'ing Training Program ID loue~tic:msollrce Comments: I I
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RO Skyscraper SRO Skyscraper I .-..R..o...s..y...st_e_m_IE...vo...1..ut..io...n_L_is_t- _s_R_o_s_y_s... te..m..1E_v_o... 1u... tio..n..L...is_t_ Outline Changes I I jOuestion TOpic j RO 2 I IWhich of the followin~ describes the reason for the sequence of dia~nostic steps in EOP-TRIP-1 Reactor Trip or Safety Injection? I I
~ SGTR diagnosis has the higher priority because of the highest probability of radioactive release.
I
~I SGTR diagnosis has the higher priority to minimize the potential for component failure due to water in the Main Steam lines.
I I ic.: Main Steam Line break diagnosis has the higher priority because a high energy steam break could potentially mask other failures. I l
~] Main ~team Line break diagnosis has the higher priority because AFW must be isolated to remain within accident analysis assumptions for containment pressure. I
!Answer: j c I !Exam Level ! IR I :cognitive Level I 1 Memory I iFacHity: ! ISalem 1 & 2 I jExamDate: i l
. .. I 12/21/2015!
[KA:[! 000007K301 \ IEK3.01 1IRova1ue:!j 4.0!iSROValue:1! 4.6IJSection:ll~iROGroup:I 1!:sROGroup:I 1! -~~ 0 jsystem/EvolutiollTitl~J 1_ R _ e _ a _ c _ t o _ r _ T _ r i ' - p - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' ~[o_o7_ _ IKA Statement: j Knowledge of the reasons for the followinq responses as they apply to Reactor Trip: I Actions contained in EOP for reactor trip l !Explanation of' 55.41.b( 10) SG pressure is checked prior to tube rupture criteria in TRI P-1 to see if any SG is faulted, as that can mask other Answers:
accidents.
Reference Title** p . Facility Reference Number I!Reference Section If Page No. / !Revision/
[* IRx Trip or Safety Injection ii EOP-TRIP-1 I II ! j2s I ITRIP-1 Lesson Plan 11 NOS05TRP001 I I159 l I7 I I II I 11 ii ! [LO.Number I TRP001 E017 Objectives I ,__ ___.
\11i1aterial Required for exalllinati<>11" 11 11 I !'Question Sourte: ) Facility Exam Bank I /,au~stion<Modification Method; * ~Direct From Source If Used Duiing Training Program ! D
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lau,es.ticm S()urce Com,m~ntsl 144816 j
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Outline Changes I IQllestioriTopic 11 RO 3 I I Which of the following is the basis for establishing I maintaining SG Narrow Range level between 9-33% (non-adverse containment) for small or intermediate LOCA's IAW 2-EOP-LOCA-1, Loss of Reactor Coolant? I I
~ Ensures SG available if a RCP has to be started later in the event.
l I
~ Maintains a static head of water to reduce any existing SG tube leakage.
I 81 Ensures adequate feed flow or SG inventory to ensure a secondary heat sink. I
@J IMaintain the water level above the top of the U-tubes to prevent depressurizing SG.
I iAnswer 11 c I [Exam Level 11 R I iCognitivelevel l I Memory I !Facility: 11 Salem 1 & 2 I [ExaITIDate: .11 12/21/20151 [KA:ll 0000091<203 IiEK2.03 *!Ro Value: II 3.0! )SRO Value: II 3.3! jsection: ii~ iRo Grollp:jj 11 lsRO Grour>:ll 1I D
systernlEvolution Title I !_s_m_a_ll_B_re_a_k_L_o_c_A_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.I f009
!KA Statement: i Knowledqe of the interrelations between Small Break LOCA and the followinq: I S/Gs I
;Explanation of! 55.41.b(10) Distractor B is incorrect because it is the basis for S/G level I feed flow for a LARGE break LOCA. Distractors A and D
,I.An>swers:. ... *.*** are incorrect because the Basis document states that the reason for maintaining adequate feed flow I SIG level is to ensure a secondary heat sink. C is correct because EOP-LOCA-1 Basis Document states the purpose of establishing 9% level is .. "To ensure adequate feed flow or S/G inventory to ensure a secondary heat sink for small or intermediate size LOCAs and secondary break
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!* Reference Title ;(
FaciHty Reference Number* I:Reference Section 11 Page No. j !Revision' j Loss of Reactor Coolant 112-EOP-LOCA-1 II Bases Doc 1112 1128 I I II n ii 11 I I ii II II 11 I fL.0.Number I LOCA01 E009 Objectives
,_____, [Ma,terial ~equired for Exa111ination *** II foue~tion S,ourc~: 11 Facility Exam Bank I [ouestiOn Modification Method:.**.* *~Direct From Source I /!Jsed During t~ainlng Program ~l D
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Outline Changes I louestio~ Jopic 11 RO 4 Given the following conditions:
- Unit 2 is in Mode 3 following a shutdown after a 200 day run. - Core burnup is 18,000 MWD/MTU. - All RCP's are in operation. - Main steam dumps are in MS PRESSURE CONTROL-AUTO@ 1005 psig. - A transformer fault results in the total loss of off-site power.
15 minutes after the transformer fault, with NO operator action, the following indications are present:
- All RCS WR Thot's are 559°F and rising slowly. - All RCS WR Tcold's are 547°F and stable. - All SG pressures are 1015 psig and stable. - All SG NR levels are 39% and stable. - PZR level is 23% and rising slowly.
Which of the following identifies:
1) The status of natural circulation

2) The action(s) that will be performed by the control room IAW S2.0P-AB.RC-0004 Natural Circulation based on that determination of natural circulation status
~ 1) Natural circulation NOT established I
2) Raise the steam dumping rate. I
~ 1) Natural circulation NOT established
2) Raise the feeding rate to all steam generators.
!c.i L....J
1) Natural circulation IS established

2) Start AFW pumps to maintain steam generator narrow range level in the normal band.
@] 1) Natural circulation IS established
2) Feed steam generators to maintain steam generator narrow range level in the normal band and adjust MS-1 O's to enhance heat removal.
..
I [Ansv.:er J ~ I.Exam Level.! [CJ Cognitive Level . I IApplication I iFacHity: I j Salem 1 & 2 I JExamDf!te: J l___1_2_12_1_12_0_1__,5j ~loooo15K101 llAK1.01 1!ROValue:JG]"ISROValue:]~:section:ll~fROGroup:lf3JsROGroup:ILJ D ~ystem!Evolution Title I IReactor Coolant Pump Malfunctions !KA Statement: I ! Knowledqe of the operational implications of the followinq concepts as they apply to Reactor Coolant Pump Malfunctions: I Natural circulation in a nuclear reactor power plant I
!j IEJ(planation of 55.41.b(5) The question meets the KIA based on the Loss of RC Flow portion of the KIA. AB.RC-0004 step 3.6 identifies ALL the Answers:
I 11 conditions that must be met for natural circulation to be occurring. With RCS Thats still rising, it is NOT occurring. Step 3.7 directs the operator to feed the SGs to maintain them within+/- 5% of programmed band. Programmed band plus 5% = 38%, so feeding in this situation is not directed. Steam dumping is directed to maintain or lower CET temps. l Reference Title H Facility Reference N_umber .*** * [ [Referel)C~ Section 11 Page No.[ f~evision'. INatural Circulation 11 S2.0P-AB.RC-0004 I !11 118 I I !I I II 11 I I II I 11 ii I jLO. Number > Objectives ! ABRC04E001
IMatehaL~~c)9ired f,(;,r;;J=X:<1m!nati9ri. I
! RO 4 Steam Tables lflo~s,f~n ~oti[:c;:~"co[hi:i1~n!;>! 180355, removed some window dressing in stem. Modified format to two part answer I
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[Question TopiC J I RO 5 Given the following conditions:
- Unit 2 is operating at 100% power. - 21 charging pump is in service. - 22 charging pump is CIT. - Subsequently, 21 charging pump trips. - 23 charging pump is placed in service after its suction is aligned to the RWST.
Which of the following identifies a subsequent action which will be directed by S2.0P-AB.CVC-0001, Loss of Charging, and why?
~I Shutdown Unit 2 due to boration of the RCS from the RWST.
I jb.: Restore normal letdown to establish adequate VCT level for normal 23 charging pump operation. l=J IPlace Excess Letdown in service to establish adequate VCT level for normal 23 charging pump operation.
@]I Open 2CV464 Charging Cross-Tie Isolation valve to prevent requiring a Unit 2 shutdown due to boration of the RCS from the RWST.
IAnswer j a l IExam Level R I !cognitive Level IMemory J 11 J I !Facility: i ISalem 1 & 2 I jExainDat 12/21/2015
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~I 000022K302 I:AK3.02 i"1RO Value:!~ ;SRO ' Value: I~jsect1on: *I:I~ IRO Group:iLJjSROGroup:jLJ ~\.:~ D 1 I jsystem/Evolution Title J Loss of Reactor Coolant Makeup j 1022 I J<A Statement: I Knowledqe of the reasons for the following responses as they apply to Loss of Reactor Coolant Makeup: Actions contained in SOPs and EOPs for RCPs, loss of makeuo, loss of charqino, and abnormal charoinq i.Explanatioriof I 55.41.b(10) When the inservice charging pump tripped, letdown automatically isolated with all charging pump breakers open. If 23 I !Answers: I charging pump is placed in service from the RWST, it means it was not able to be placed in service with VCT as suction source at step 3.7. The note prior to placing 23 charging pump in service from Unit 2 RWST states that a unit S/D will be required because of borating the RCS from the RWST.
1. Reference Title /i ***Facility Reference Number
i !Reference Section dfPage No'. j IReyi~ion:
ILoss of Charging !IS2.0P-AB.CVC-0001 II Bases Doc 11 3 1I9 I I II II II II I I II II II II I jL.o. Number ..* Objectives I ABCVC1 E003 ,_ _ ___. IMateflal Required for E~afoinaUon ,;j I 11
\Question Sour,2e: Il_N_e_w_ _ _ _ _ __.l [Questi<>:n )'d6dification Method: ....~ _ _ _ _ _ _ _ __.I Jusedbui'ing Training Program. I O )Que~tio11~purceComme~ts/ I j !~--------------------------------------~
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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I. . SRO System/Evolution List I Outline Changes I
..... !Question Topic J 1RO 6 I Given the following conditions: - Unit 2 is in MODE 4. - 21 RHR loop is in service providing SDC. - 22 RHR loop is aligned for ECCS. - Subsequently, 21 RHR pump trips on overcurrent.
Which of the following describes how SDC flow will be established IAW S2.0P-AB.RHR-0001, Loss of RHR? 22 RHR pump supplying flow .....
~I through 21 RHR HX and ONLY 21SJ49 RHR Disch to Cold Legs.
I IE] Ithrough 22 RHR HX and ONLY 22SJ49 RHR Disch to Cold Legs. I
~I through 21 RHR HX and BOTH 21SJ49 and 22SJ49 RHR Disch to Cold Legs.
I
~I through 22 RHR HX and BOTH 21SJ49 and 22SJ49 RHR Disch to Cold Legs.
I !Answer 11 d I IExani Level 11 R I !cognitive Level 11 Memory I !Facility: ! ISalem 1 & 2 I JExamDate: 1 I 12/21/20151 iKA:ll 000025K201 I AK2.01 1IROValue:jj 2.9jlsROValue:ll 2.9ilsection:ll~fROGroup::j 11lsROGroup:!I 11 !Sl~I D [System/Evolution Title i j Loss of Residual Heat Removal System ! KA Statement: I Knowledqe of the interrelations between Loss of Residual Heat Removal Svstem and the followinq: RHR heat exchanqers I jExpfanation of! 55.41.b{10,8) With a single loop of RHR in service for SDC, the loop cross-tie valves 21/22RH19 are open. They would only have .Answers: I been shut if BOTH loops were in SDC mode, to split out the RHR loops to verify minimum flow requirements are being maintained (S2.0P-SO.RHR-0001, page 36). There is no direction to close either cross tie valve in Attachment 2 of AB.RHR when swapping the ECCS loop to SDC mode. There is also no direction to shut either SJ49 RHR to RCS valves, since cooling flow to all 4 RCS
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IiReference Section .* j IPage No. !Revision! j ILoss of RHR !I S2.0P-AB.RHR-0001 I ii 1118 I IInitiating RHR !I S2.0P-SO.RHR-0001 I 11 ! l2s I IRHR Simplified drawing ii 205332-SIMP I 11 112 I fL.(). Number**..
IABRHR1E004 ,____. iMaterialRequlred for EJ(Cll:Jlination **** 1 I louesti.oh_Soufoei I l_N_e_w_ _ _ _ _ __..l louestionModification lll!eth?d: *. J I \Osed During Traini11g Pr~gram"! D fqu~sti.on Source Comments! I I
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RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List I .Outline Changes touestion Topic I IRO 7 I Given the following conditions:
- Unit 2 is operating normally at 100% power steady state. - Console alarm RC PRESS DEVIATION HI, annunciates. - Subsequently, the board operator then observes the following indications: - PZR pressure master controller output dropping slowly - Both PZR spray valves closed - All PZR heaters energized - PZR pressure channel I (selected for control) dropping slowly - PZR pressure channels II, Ill, & IV rising slowly Which of the following procedures should be used for these conditions?
I
~ S2.0P-AB.STM-0001, Excessive Steam Flow.
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~ S2.0P-AB.ROD-0003, Continuous Rod Motion.
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~ S2.0P-AB.RC-0001, Reactor Coolant System Leak.
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@]I S2.0P-AB.PZR-0001, Pressurizer Pressure Malfunction.
I !.AA.svver 11~ !Exam Level 11~ :cognitive Level I Application I I [F(ldlity: !]Salem 1 & 2 I IExam Date: fl ___1_2_12_1_12_0_15_.I ~I 000027G107 112.1. 7 i :Ro Value:j G] lsRo Value:j~ !Section: !I~ !Ro Group:jLJ iSF{O Group:l LJ !;~! D !System/Evolution Title i IPressurizer Pressure Control Malfunction I W~ IKA Statement: I Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. I fExptapation pf,I 55.41.b(5, 10) The three distracters would cause all the bulleted items except the last one to occur. The 3 unaffected channels rising Answers::. * :> would discount the 3 items (steam leak, RCS leak, or rod insertion) that could be causing the event. I Reference Title I! Facility Reference Number j jReference Section j I.Page NC?'.: jRe\(ision! j Pressurizer Pressure Malfunction 11 S2.0P-AB.PZR-0001 II 11 1118 I ! II II 11 11 I I II II 1 I I LO. Number IABPZR1 E001 Objectives " [Material Required for Examination iI I IQuestion Source:*, ,j Facility Exam Bank i 1ftluestion Modificatio,n !\l[~thod:. ; Direct From Source l lusedDOri~g.iraiO.in9 Program j D louestiorisource.somirientsl 1.--42_6_8_2_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _...,.1 1 !comment I I I I I '
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I lauestion Topic 11 RO s I Which of the following describes the effect if a PZR level instrument reference leg were to develop a leak with the Master Flow Controller in auto? Indicated PZR level would initially ..... [al Irise, then actual PZR level would rise. I
~I rise, then actual PZR level would lower. !b.:
I ill
~ lower, then actual PZR level would rise.
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!ct. j Ilower, then actual PZR level would lower.
I !Ansv./er ! b I I !Exam Level i j R I !cognitive Level 11 Application j jFacility: 11 Salem 1 & 2 1 IExamDate: J I 12/21/20151 IKA:il 000028K101 1IAK1.01 'iRO Value: :j2.s*1 isRO Value: !I 3.1 Ifsection:!j~ :Ro Group:jj 1 2j SRO Group:lj 21 ~ltJ D !system/Evolution Title I IPressurizer Level Control Malfunction :028 IKA Statement: i Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunction: PZR reference leak abnormalities I !Explanation of I 55.41.b(7) As level in the reference leg lowers, the pressure of the actual wight of water in the PZR would appear to be greater than Answers: it was, as it is pushing against a lower reference pressure from the reference leg. This would result in indicated PZR level rising. The rising level would cause an automatic response of the Master Flow Controller to lower charging flow, which would cause actual PZR level to lower. I> Reference Title I l~ Facility Reference Number **. i jRefererice Section *I IPage No. f IRevision; IGeneral Physics Lesson Plan Components I 1 PC071 r4 Sensors I 11 11 4 I I II I ii 11 I i II I 11 11 I IL.o. Nllrnber IPZRP&LE015 1_ ___, !Material Requfr~dfo~ Examination**. i I ia':.i~stion Source;. 1 IFacility Exam Bank I ldllestion Modific:ation Method; .**.. ,j Concept Used I fused During Training Program I D ~9~e,stior1Source Cornmentsf 159797
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!Question Topic l I RO 9 I Which of the following describes the difference in effect of a rapid boration performed during an ATWS at BOL or EOL?
At 1) , the rapid boration will insert more reactivity due to the 2) differential boron worth . I
~I 1) EOL . *. . 2) HIGHER I
[gJ 11) EOL
2) LOWER I le.! 11) BOL
,_ 2) HIGHER I
[ ) 11) BOL
2) LOWER I
/Answer 11 a 1 IExalTI Levell IR I !cognitive Level 11 Memory I !Facility: I j Salem 1 & 2 I [Exam[)a~e: 11 12/21/20151 IKA:il 000029K103 j IAK1.03 liROValue:1j 3.ellsRova1ue::l 3.sl!sectic)n:il~iROGroup:ll 11isR6Group:ll 11 t~~J! D lsysterri/EvolutiOn Title I j Anticipated Transient Without Scram 11029 iKA Statement: i ! Know!edqe of the operational implications of the follow!nq concepts as they apply to Anticipated Transient Without Scram: Effects of boron on reactivity IExpl~nat,i?p ofj 55.41.b(5) Differential boron worth rises over core life, and will insert more reactivity during an ATWT at EOL. .Answers: * .*
I I Reference Title .**. lf;

Facility geterepce Number I!Reteren~eseCtion * ** * *I ! Page ~o. I IRevisionj ICurvebook 11 S2.RE-RA.ZZ-0016 I 11 1I7 I I I I II 11 I I I I ii 11 I
\L.o *.Number Objectives OOE004 !Material Required forExani,in.ation *.. 1 l II !.Question sour~fi:/i l_N_e_w_ _ _ _ _ __.I !<:iu,e~ti~p Modification Method;< l_________.l luse9 ,OurintJ Training Pr,ogram l D [auestiori ~c)u~~e ~()mmentsl I I
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!Question Topic 11 RO 1o Given the following conditions: - Unit 2 experiences a Rx trip from 100% power when a Main Generator trip causes a Main Turbine trip. - Upon transition to TRIP-2, Reactor Trip Response, the PO reports OHA A-6 RMS HI RAD has annunciated, and Radiation Monitor 2R19C - 23SG Slowdown, is in alarm with rising counts.
Which of the following describes how the crew should proceed?
~I Initiate SI and go to TRIP-1, Reactor Trip or Safety Injection.
ibl IEnter S2.0P-AB.SG-0001, Steam Generator Tube Leak, while continuing in TRIP-2.
~I Do NOT initiate any procedures based solely on a 2R19 alarm immediately following a Rx trip.
I idl IEnter S2.0P-AB.RAD-0001, Abnormal Radiation, to verify validity of alarm while continuing in TRIP-2. I !Answer 11 d I /Exam Level 11 R I !Cognitive Level 11 Application I !Facility: 11 Salem 1 & 2 I [ExamDate:'l I 12/21/20151 iKA: 11 000037A201 l1AA2.01 I IRO Value: JI 3.olfsRo Value: II 3.41 tsection:.11~ !Ro Grciup:Jj 21 iSRO Group:: I 2j l!J D ISystem/Evolutiori Title ! ISteam Generator Tube Leak f j037 [KA Statement:! Ability to determine and interpret the following as they apply to Steam Generator Tube Leak: Unusual readinqs of the monitors; steos needed to verify readinqs !Explanation of I 55.41.b(10) AB.SG says in note under entry conditions that R19s are not accurate immediately following a unit trip and should not !Answers: ***.*. * .1 be used as the sole basis for entering it. However, AB.RAD should be entered and the alarm verified as directed at step 3.2. The SI is not warranted without corroborating indications

l Reference Title It Facility Reference Number
IIReference Seetion ** *Ii Page r-z,o. I [Revision 1! .. *. . , .
ISteam Generator Tube Leak II S2.0P-AB.SG-0001 II 112 1129 I IAbnormal Radiation 1IS2.0P-AB.RAD-0001 II 111 1130 I I II II 11 II I !Lo. Number<** I ABSG01 E003 Objectives 1_ ____, !.Material Requfred for Examin(ition * *! I iQUestion .s9urc~: t l_N_e_w_ _ _ _ _ _ __.I ;auestion Mo,dificatic:in Method: *** __________,I tused During Tr~iriih9 Prografrl I D 1aue5:tion Source Cofrlrr,~nts) I I
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RO Skyscraper i SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I jQuestiori Topic j Ro 11 I Given the following conditions:
- Unit 2 is performing a controlled shutdown and cooldown from 100% power due to a 5 gpm tube leak on 22 SG IAW 82.0P-AB.SG-0001, Steam Generator Tube Leak. - After completing the Immediate Actions of EOP-TRIP-1, Reactor Trip or Safety Injection, following the Rx trip from 20% power, the RO reports that control rod 202 is stuck in the fully withdrawn position.
Which of the followinq identifies the action, if any, the crew will perform in response to the stuck rod? I la. i Initiate a rapid boration for 35 minutes during performance of EOP-TRIP-2.
~I Initiate a rapid boration for 35 minutes in 82.0P-AB.SG-0001 after exiting the TRIP series procedures.
[] I No actions are required for a single stuck rod because SOM for the cooldown to 503 degrees is adequate. [] I No actions are required for a single stuck rod until the Auto SI Block is performed during RCS depressuri~ation to 1900 psig. [Answer 11 b I !Exam Level 11 R I [cognitive Level 11 Application I 1 !Fa!=ility:.1 Salem 1 & 2 I IExamDate: 11 12/21/20151 ~I000038G416 lfROValue:l@[sROValue:j[Dfsection:Jl~:goGroup:JLJfSROGroup:ILJ f:in
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I 112.4.16 D fsy§tem/Evoluti<?nTitle I ISteam Generator Tube Rupture 11038 I ll<A statement' I Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. IExplarya~ion of I 55.41.b(10) Question meets KIA because knowledge is required of the relationship between AB performance in the initial response [Answers: * .* I to the tube rupture and EOP performance which addresses stuck rods. Boration for a single stuck rod is not performed in the EOP series, but IS performed in the AB for tube rupture to meet SOM requirements for the initial cooldown to 500°F. Step 3.26.H in AB.SG states to trip the turbine, then trip the Rx at 20% power during the shutdown. The 5 gpm size of the leak will allow for a
.. ............... +....... q.o.rl c-h1 .+.-.1 ...... ,.... "'.:lnrl .~II -::::ilen -::ill ........ +i.......... ............. +/"\
cause a transition to SGTR-1. AB.SG would be re-entered at step 3.27 following exit of TRIP-2, and 3.28 directs rapid boration for
+~ TOID." Th-*--*- -
n~ ~r.:To ~;-----*;~ -*--- ;n TOID_'l "*-
. **- .i~
each stuck rod for 35 minutes. The rapid boration will be initiated before any depressurization starts in 3.29, so the distracter reqardinq depressurization is incorrect. re \
' *,;._ .. :~ .':,'. : ., ** Reference Title . ll Facility Refere.nce. N.umber
i!Reference Sec~ion** . *l I Page* No. I tRevision:
J Steam Generator Tube Leak II S2.0P-AB.SG-0001 l 11 1129 I j Reactor Trip Response 112-EOP-TRIP-2 ! ii 1I I I II I ii 11 I fL.o:Nutnber * *** Objectives SG01E005
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Outline Changes I iauestionTopic l IRO 12 I How does the INITIAL Tavg response for a large steam line break (SLB) compare to the Tavg response for a large feed line break (FLB)? (Initial Tavg response is before automatic protective actions occur)
~ I ~I Rises for a SLB and lowers for a FLB.
Lowers for a SLB and rises for a FLB. I I I
~ Rises for both a SLB and a FLB, rises more for a SLB.
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~ Lowers for both a SLB and a FLB, lowers more for a FLB.
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Answer J I b I !Exam .Level ! IR I !cognitive Level 11 Application I !Facility: 11 Salem 1 & 2 1 IEX:amoate: ii 12/21/20151
~I 000054K101 I AK1 .01 I!RO Value: i [DJ jSRO Value: l@tsection: iI~ [RO Group:i LJ [SRO Group:! LJ I,~ D r~,,,,._,,,_,T"~ ls~~~w~~~~j_L_o_~_cl_M_~_n_F_e_e_~_a_t_e_r_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _I!~ fKA Statement:! ! Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater: I MFW line break depressurizes the S/G (similar to a steam line break) [Explanation of j 55.41.b(5) A steam line break will cause MORE steam to be drawn from the SG, causing Tc to go down, causing Tavg to go down.
Answers: * . *...* J A Feedline break will send LESS cold water into the SG so it's Tc will rise, and Tavg will rise. KIA match because the operational implication of the SG depressurization (similar to a steam line break) is that an operator would have to use diverse and alternate indication to discern what is acrually happening to the SG.
I Reference Title li Facility Reference Number .. ![Reference Section * ,j t Page No. ! fRevision! I I I 11 11 I I I I 11 iI I I I I 11 11 I iLb.Number Objecti ICN&FDWE016 ,__ __, IMaterial Required for Ex:aminatiorl >i I )Question Sol.lrce:: 11 Facility Exam Bank I f OuestiOhModifidtion Method: I Direct From Source )~~~nS~~efo~~~11~8-0_7_8_9 _____________________________________ ~, [comment *.* *'. I I I i I
RO Skyscraper SRO Skyscraper. I RO System/Evolution List I SRO System/Evolution List Outline Changes lauestionTopicj I RO 13 Given the following conditions:
- Salem Unit 1 is operating at 100% power. - Salem Unit 2 is in Mode 4 with 21 loop of RHR in SOC mode. - Subsequently, a loss of all off site power occurs. - ALL Unit 2 EOGs fail to start or trip shortly after starting, and ALL Unit 2 4KV vital buses are de-energized. - The RO reports 21RH18 RHR HX FLOW CONT VALVE indicates 6% open and 2RH20 RHR HX BYP VALVE indicates 15% open, and questions their actual position.
Which of the following describes the operation of the 21 RH18 and 2RH20? I
~ BOTH the 21RH18 and 2RH20 fail AS IS on a loss of air and the indication is correct.
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~ The 21RH18 fails open and the 2RH20 fails shut, and the board indication shows last known position before the loss of power.
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~ BOTH the 21RH18 and 2RH20 fail SHUT on a loss of air, and the board indication shows how far they shut before losing all air pressure for operation. I '.dl i 80Ild tbQ 2:'.J b?bl:l 8 ~od 2l2h:l20 f:;Jil 012Et\I oo lcso:s;;:: of ~ic :aod tbg bo:;ii:d iodic~tioc ~bcuus;: -
brnu f:;Jr: tbQu ooQoQd bQfocg losio,.. .... 11 .. I I
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L:.cJ for operation. I !Answer I ~ !Exam Level: ~ !Cog11itive L:evel. I IMemory ! !FaciHty: j' ISalem 1 & 2 I jExamDate: Il___12_12_1_12_0_1_,5I ~I 000055A201 1IEA2.01 ! !Ro Value: IfTI"isRO Value: I[IZJl~ection: 11~ [RO Group:j LJ iSRO Group:! LJ D [System/Evolution Title.I l_s_ta_t_io_n_B_la_c_k_o_ut_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___, ~-~ IKA Statement:! Ability to determine and interpret the followino as they apply to Station Blackout: Existinq valve positioninq on a loss of instrument air system !Explanation of I 55.41.b(?) The 21RH18 and the 2RH20 fail AS IS on a loss of control air, and both are supplied air exclusively from the "A" header. !Answers: . . . . ! Unit 2 ECAC supplies the "A" air header. With the Unit 2 ECAC also failing to start, control air pressure will bleed away fairly rapidly. The console indication will indicate actual valve position since 115VIB power should be available from the inverters for at least 2 hours following a LOPA. The valve positions given in the stem are what would be expected with a loop of RHR in SOC mode. I
\"/" i Reference Title Ir Facility Reference Number ******I iReference Section . ;1 f Page)lo~ I !Revision; ILoss of Control Air 11 S2.0P-AB.CA-0001 II 1136 1I18 I I II n II 11 I I Ii n II 11 I Objectives I LOPAOOE010
,_ _ ___. [Material RE\q~fr~d tor exarTilrfaiion /I I !I la4e~tion ,source: ** 11 New 1lauestiqn .~.odification Meth~~.: .*.* ~--------~I !used Ouring Training Program I D lauestion Source,Coinl"l1,e~tsj I I
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RO System/Evolution List I SRO System/Evolution List I Outline Changes I .. louestion Topic i I RO 14 I Given the following conditions:
- The 2B Vital Instrument Bus (VIB) Uninterruptible Power Supply (UPS) Static Switch has been placed in the Bypass to Alternate position IAW 2B VITAL INSTRUMENT BUS UPS SYSTEM OPERATION. - ALL power supplies to the 2B VIB UPS are available. - Subsequently, the breaker from 2B 230VAC bus to the 2B VIB AC Line Regulator fails and opens.
What will be the effect on 2B VIB? The 2B VIB wilL . ..... - - . - --* I
.. ~ NOT deenergize because the Static Switch will automatically swap to the Preferred Source.
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~I NOT deenergize because the inverter will automatically power the 2B VIB through the inverter auctioneering circuit from its DC source.
I [c. f I deenergize until manual operator action is taken to re-energize the 2B VIB by placing the Static Switch in Normal and placing the Test Transfer switch to N (Normal). I
@:]I deenergize until manual operator action is taken to re-energize the 2B VIB by placing the Static Switch in Isolate (Alternate) and placing the Test Transfer Switch to N (Normal). I fAnswer I !c IExam Level ! R ! lcogl'litive L~vel !
I Agglication I IFacility: WSalem 1 & 2 l~ExamD?te: W 12/21/20151 i it{A: 000057A101 IIAA1.01 i iROValue: 1!3.7*1 :sRo Value: ii 3.71 [section: !I~ tRo Group:JI 11 ISRO Group:JI 1 I D
system/Evolution Title I j Loss of Vital AC Instrument Bus 1w-:
I . ' iKA Statement: I Ability to operate and I or monitor the followino as they apply to Loss of Vital AC Instrument Bus: I Manual inverter swappinq I 1£xplanation of ! 55.41.b(7) The VIB UPS static switch is transferred from norm to alt by placing the test transfer switch to ALT IAW Section 5.4 of Answers: I S0.115-0012. Then the Manual Bypass Switch is placed in Bypass to Alternate to physically position contacts B1, B2, and B4 (closed) and B3, B5 open. The VIB will deenrgize when power is lost to the AC line regulator, which is the Alternate source. Placing the static switch in Normal and test transfer switch to N (from alternate) is directed by S0-115-0012, Section 5.7.3 and
~ A n* n ;,... ..... 1.... ,,...;1... - if i+ ;.,.. ......... ....i .......... __ ,_*-.LI... .,..,... *--- .C--- "'-- _.......:_ -* .:.1. i+
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- ..... h ... LI... * *'- **- rl..1 hi"\
could be powered from another source. A is incorrect because automatic transfer is unavailable with the staic switch not in the
.I.... ... ,...
Normal posiiton. B is incorrect because the static switch is aligned to alternate, and while the DC power is supplying the inverter, the inverter output cannot flow throuoh static switch with B3 and B5 contacts open. I * .. Reference Title *II,; >Facility Reference Number *. J 1Reference Section 11 Page No. ) !Revision: I2B VITAL INSTRUMENT Bus UPS SYSTEM o II s2.0P-so.115-0012 ll 5.4 1119 11 6 I !2B VITAL INSTRUMENT BUS UPS SYSTEM o II s2.0P-so.115-0012 !IExhibit 1 (Static Switch I156 11 6 I I !I II II 11 I !Lo. Number
I 115VACE014
touesti§ff~o!Jrce: : 1l_N_e_w________,I Jgtiestion Modifi§,t.t~9;M~t~pct: ** [c:tu~sti~~:~oUrce co~rfl~rit~'I
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[Question Topic 11RO15 Given the following conditions:
- Unit 2 tripped due to a transformer problem. - 2A and 2C Vital Busses are energized in SEC Mode 2. - 2B Vital Bus is de-energized because 2B Diesel Generator failed to start. - OHA B-10, 2B 125VDC CNTRL BUS VOLT LO, just actuated.
In accordance with S2.0P-AR.ZZ-0002, Overhead Annunciators Window B, which of the following describes the required action? -- la. i I .. . Place the 2B2 Battery Charger in service.
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~I Transfer the 2B1 Charger to its alternate power supply.
[c.l ITransfer 2B 125 VDC bus loads to their alternate source. [] I Ensure the 2B2 Battery Charger has automatically energized.
Answe_r ! I a I !Exam Level i IR I !Cognitive Level 11 Application I !Facility:J ISalem 1 & 2 I !ExamDate:*11 12/21/20151
~1000058A101 llROValue:jjilljsROValue:IQ]'[sec;tion:,fl~ [ROGroup:ILJiSROGroup:ILJ l[~;1i;I D
,* ~ ,,. IE~::F llAA1.01 ISysterryEvolutionTitle I ILoss of DC Power I ~lo_ss__
IKA State merit: I Ability to operate and I or monitor the followinq as they apply to Loss of DC Power: Cross-tie of the affected de bus with the alternate supply I [Explanation of I 55.41.b(S) The 2B1 battery charger is normally in service. The use of 2B2 Battery charger is limited to 7 days per Tech Specs and !Answers: is not normally in service. There is no auto swap. Transferring loads is only done if the backup battery charger cannot be placed in service. I < Reference Title * .. >it Facility Reference Number-,-*.*. IIReference Section *_ ... j I Page No. I 1~evisionl IOverhead Annunciators Window B !I S2.0P-AR.ZZ-0002 I 1121-22 1136 I I I I ii 11 I ! I I 11 11 I [L*.O.-Number:;i_"* I DCELECE005 I_ _____. IMater'ial ~equired for e.Xai:nfoation' 'I I ll I [ouesfioll;s#&rce:,\ Facility Exam Bank I 1lauestion Modific~tia,11 Met~~d:
Editorially Modified 1llJsed oll;!"illg Training program I D
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!auestionTopic 11RO16 I Which of the following identifies when the 21/22SW122, CC HX SW INLET VALVES automatically reposition, and why?
The 21/22SW122 valves ... I
~ open on ANY SI signal to ensure cool CCW is supplied to the RHR HXs and pump seals. I I ~I close on ANY SI signal to ensure SW pump runout does not occur with all CFCUs running.
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~ open on a SI signal coincident with a LOOP to ensure cool CCW is supplied to the RHR HXs and pump seals.
I l:!J Iclose on a SI signal coincident with a LOOP to ensure SW pump runout does not occur with all CFCUs running. I !"Answer Id i I JExarl1 Level 11 R I !cognitive Level I
! Memory ! l_acilfty: / ISalem 1 & 2 I /Examoat~: 11 12/21/20151 J. *"":
tKA:JI 000062K302 2.91fSection:f!~IROG~oi.lp:lj ~11 D
},"!'1, liAK3.02 /iROValue:Jl 3.6ifSROValue:Jl 1ljSROGroup:!I 11 Is ystem/Evo I uti on Title ! ---;=:::::::::::::::::..::======-===-=======:..:.::=...:=====::====-=======...:.:::==-===========-
j Loss of Nuclear Service Water I 1062 IKA Statement: I Knowiedge of the reasons for the following responses as thev apply to Loss of Nuclear Service Water: The automatic actions !aliqnments) within the nuclear service water resultinq from the actuation of the ESFAS !Explanation of 55.41(7,8) The SW122 valve are normally open and operate in conjunction with the CCW HX SW outlet valves SW127. During a SI .Answers:* coincident with a Loss of Offsite Power, the SW122 valves are automatically repositioned from open to shut. SEC Mode Ill closes SW122's due to SW Pump runout (and potential loss of all SW) concerns with all CFCU's l/S if only 2 SW pumps are operating, due to EOG/SEC failure. Also, the SW122's must stroke closed in less than 30 seconds to ensure the CFCU's are operational within l
~n fr.II~***'-~ **~--lo Ill :-:+:~.:~- ITc~h c-~~ "l/A "'.! '> T. ., "'-'"
I Reference Title 11 Facility Reference Number ./I :Reference Section, Page No~ I !Revision:
* .. 11 IService Water System - Nuclear 11 NOS05SWONUC II !134-35 ! I13 I I II II II 11 I I !I n II 11 I
!Lo. Number i Objectives I SWONUCE006
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!auesti<>,n sour~e_cC>mf11e11ts1 1_1_25_6_7_5_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.I jcomment I I I I I !
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!Question Topic ',. . , *' ,
I IRO 17 I I Both Salem Units are operating normally at 100% power when Unit 2 recieves OHA A-7 FIRE PROT FIRE. Which of the following is a correct response when assessinq the affected Fire Zone(s) on 2RP5, and why? I []I If ANY rows "Fire" light is illuminated ensure at least one Fire Pump is running to supply fire protection water by verifying OHA A-15 FIRE PUMP 1/2 RUN. I
~I If a fire is indicated in the Relay Room select Fire Outside Control Area in both control rooms to prevent smoke from entering control rooms.
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IIf ANY C02 I Halon Discharge lights are lit then ensure EDG control room ventilation has automatically stopped to prevent egress of gas to adjoining areas. I [] I. If fire indic~tion for BOTH zo.nes 59 and 74 are received, then open 2FP147 Fire Protection Containment Isolation to provide normally isolated fire protection water to containment. I !Answer i Id I !exam Level 11 R I !cognitive Level* 11 Memory 1 !~acility: i ISalem 1 & 2 I IExan:iOate: I I 12121120151 [l<A:JI oooo67K302 IIAK3.02 i!Rova1Ue::j 2.51[sRova1ue:q 3.3li8edion:ll~IRoGroup::I 21fsRpGrou!>:JI 21 all D '
system/Evqlution Title! !_P_la_n_t_F_ir_e_o_n_S_it_e_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _...... I067 i

KA Statement:! Knowledqe of the reasons for the followlng responses as they apply to Plant Fire on Site:
Steps called out in the site fire protection plan, FPS manual, and fire zone manual IExplanati<m()f j 55.41.b(10,4) A is incorrect because the FIRE light can illuminate if a manual fire pull box is activated, which only gives indication
*I tAnswers: . * ** *.*. and does not initiate fire protection water fiow. C is incorrect because Halon supplied to relay rooms would not indicate stopping EDG supply ventilation. B is incorrect because for a fire in Relay Room (physically located outside CR but using same AC system)
Fire Outside Control Area would not be selected. D is correct because the fire protection line to containment is normally isolated
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i Reference .Title> :[
' Facility ~eference Number I !Reference Section
I[Page No. I 1Revision 1 IControl Room Fire Response !I S2.0P-AB.FIRE-0001 I 112 11 9 I IOverhead Annunciators - Window A I
II S2.0P-AR.ZZ-0001 I 1123-24 l l56 I I II I 11 11 I [LO.Number
I FIRPROE008 Objectives I
!Material Requinid for ExafTJin~tion
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\auesticmSoU,rce:' j j_N_e_w_ _ _ _ _ _~I [aue$tionM9dification Metho~: **. --------~I !Used During ;r{airiing Program I D lQuesti()nSourc~Comrnents\ !---------------------------------------~'
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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [ouestioh Topic 11 RO 1's I Given the following conditions:
- Operators are performing actions in 2-EOP-FRCC-1, Response to Inadequate Core Cooling. - Containment pressure is 2 psig and stable.
Which of the following describes the SG level and AFW flow requirements PRIOR to initiating the SG Depressurization to Inject ECCS Accumulators? SG NR level in at least one SG must be greater than ___
~ 19% OR total AFW flow must be >22E4 lbm/hr.
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~ 115% OR total AFW flow must be >22E4 lbm/hr.
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~ 19% AND total AFW flow available of >22E4 lbm/hr.
I [ ] 115% AND total AFW flow available of >22E4 lbm/hr. I JAnswerJ I a I iExamLe>/el J IR I !Cognitive Level ! IMemory I !Facility: 11 Salem 1 & 2 I !ExamDate: i I 12/21/20151 Group:ILJ !SRO Group:j LJ r ~I 000074K203 IiEK2.03 IiRO Value:!~ ' ,SRO Value: I~ lsect1on: 11~ :RO ' '.1
!!II D lsystel-ri{Evolution Title i j 1nadequate Core Cooling 11074 I
KA Statement: I Knowledqe of the interrelations between Inadequate Core Coolinq and the followinq: I AFW pump I fExplanat!on of I 55.41.b(10) Step 13 asks if any SG NR level is >9% (15% adverse). Since containment pressure is stated as 2 psig, it is below the
!Answers...*. 1 4 psig at which adverse numbers would be used. If SG NR level is >9% in at least one SG, the step for asking if >22E4 lbm/hr AFW flow is present is bypassed. If SG NR level is <9%, then it requires 22E4 lbm/hr. So you need at least 9% OR actual total AFW flow >22E4 lbm/hr. The availability is plausible as it is used in other Emerg procedures, but in this case a secondary heat sink
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I Reference Title !~ Facility Reference .Number . I!Reference Section J[ I Page No. [Revision' IResponse to Inadequate Core Cooling !12-EOP-FRCC-1 I ii 2 1122 I I II I II 1 l I I II I II II I iL(). N1,1mber Objectives I FRCCOOE005 j_ ___. (Material Required for Exam_ination 11 11 [Que~ti().I} S()~rc,~\ ] j_N_e_w_ _ _ _ _ _ __.l IQl.lestion Modification Method: , :jl I !used Dliring.TraiOingPr69r<!mJ D fOue~t!!'~,~~urce(:omments/ I I
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(auestion T?Pic ! IRO 19 Unit 2 is at 80% power when RMS channel 2R31, Letdown Line-Failed Fuel Process Rad Monitor starts a rising trend. When responding with S2.0P-AB.RC-0002, High Activity in Reactor Coolant System, how will operators differentiate between a crud burst and failed fuel as the cause of the rising 2R31 indication?
~I By monitoring 2R31. Fuel damage will cause the indication to rise at a higher rate than a crud burst.
I
~ By raising letdown flow rate to 120 gpm. The 2R31 readings will lower if crud burst caused the rising trend but will NOT lower if failed fuel caused the rising trend.
I
~ By requesting Radiation Protection to survey the letdown pipe area in the Auxiliary building. Radiation levels will be higher due to failed fuel than from a crud burst. I @JI By requesting a Shift Chemistry Technician perform a radiological analysis of the RCS. A crud burst will show different concentrations of certain radionuclides than will failed fuel. I
!Answer 11 d I !Exam Level i I R I !cognitive Level 11 Memory I IFa~Hity: 11 Salem 1 & 2 I [Exarnbate: 11 12/21/20151 [KA:f I000076A203 I!AA2.03 i !Ro vaiue: I 2.5j jSRO Value: JI 3.olfseciion: II~ iRO Gi:oup:JI 21 lsRo Groi.i?:I! 2j 1~1 D fsystem/Evolution Title I IHigh Reactor Coolant Activity 11076 [KA Statement: ! Ability to determine and interpret the following as they apply to High Reactor Coolant Activity: RCS radioactivitv level meter jExpla~ati()p of I 55.41.b(5, 10) D is correct.This is the method as directed by procedure to determine if there is failed fuel. A is incorrect because
Answers:
' . , . . .
I there is no procedural guidance for the operator to use to determine source of elevated readings by how fast indications are rising.
C is incorrect because Radiation Protection is sent to perform surveys to repost areas as necessary for personnel protection, not to determine source of activity.. Bis incorrect because Letdown is maximized to expedite RCS cleanup for valid elevated activity, not tn -'~*~--;no~~ .. ~~ nf tho ~~.: .. :+.
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-1, q FaC:ility Reference Number I[~eference Section J ! Page No. I !Revision:
IHigh Activity in Reactor Coolant System ll S2.0P-AB.RC-0002 II II 1 Is I I Ii II II 11 I I II II Ii 11 I 'Lo: Number J Objectives I ABRC02E001 I ABRC02E003 ,_ _ ___.. [.Material Requirecl f()r Examination j I 11
\OuesHon ~olll"~~; J j Facility Exam Bank 1jau~~fi?rtM()dificaVon M~th()~: J Direct From Source Iiused Di.iring Training P~dgrarn I D fQyesti()"1-S()urc~Co)'Jlrpents! l , . . . 7 _ 1 _ 4 _ 8 _ 3 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . . . . , . l i I I I I I I
RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic 11RO20 Given the following conditions:
- Salem Unit 1 is offiine. - Salem Unit 2 operating at 95% power, 1150 Mwe, with its Power System Stabilizer (PSS) out of service. - Unit 2 Main Generator gas pressure is 75 psig. - Hope Creek is operating at 100% power, with its PSS out of service. - The Hope Creek 5-6 breaker is out of service. - Subsequently, a 500KV grid disturbance results in lower than normal grid voltage.
If a power reduction was required, which of the following identifies Main Generator loading which is outside the allowable for Salem Unit 2 IAW A-5-500-EEE-1686, Artificial Island Operating Guide? Trip-A-Unit is NOT armed. Salem Unit 2 operating at Mwe with MVAR loading out.
~ 11000, 150.
nl
~ ~ !1000, 575.
I
~ 11100, 575.
IA11swer '. ~ IExam Level I ~ ':Cognitive - - Level j IApplication I \,Fac;ility:i ISalem 1 & 2 I !Exa(TID,ate: I I 12/21 /20151 ~joooo77A201 liM2.01 f!Rova!Ue:J@[sROValue:l~lsecticm:Jj~[RO.Group:)LJ!s~oGroup:JLJ ID D IS:ystem/Evolution Title I j Generator Voltage and Electric Grid Disturbances If~0_77_~
l<A Statement

! Ability to determine and interpret the following as they applv to Generator Voltage and Electric Grid Disturbances: I Operatinq point on the oenerator capability curve I IExplan~tion of/ 55.41.b(4) With Unit 1 O/S and the HC 5-6 breaker O/S, the correct curve is 2S2H-5-6 on page 291. With both Units PSS O/S,. the Answers: 1 red dashed line will be used for allowable generator excitation. A is incorrect because the PSS is O/S. If either units PSS was IN service, then it would be correct. The 2 distracters with higher MVARS are both within the limit. Since there are two different Mwe loading conditions, and the choices for each are high/low, the answer cannot be obtained by ruling out 2 of the choices because
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I Reference Titl~ .' * . **.IL > facility Reference Numb_er j [Reference Sedioi:i . 11: Page No. j !~evisionl IArtificial Island Operating Guide 11 A-5-500-EEE-1686 I 11291 1111 I I II ii 11 I I II I II 11 I ]LO~ Num!Jer * * *. \ Objectives IGEN002E016 I I GEN002E017 I '--~
I IMat~rialfiequireq for El!:C1!T!in.9tJoi;l.:;\iJ'.! RO 20 A-5-500-EEE-1686, pages 123,291,308 lj l<it~~~ti~J?.*so.Yrce: / IPrevious 2 NRC Exams I/9uesf!-6il M.~dlfis~ti~~ M~W§Ci: ;' ~ Significantly Modified 1lµs,~d olidng Tf#iningJ~:r<>9ram',\ D
1 fQ~~g;t!gfl s~.!Jrce9()qifi1entsi 113-01 RO 021 all 4 choices changed to different combination of numbers. No original numbers from 13-01 exam
* * ** * * ** used. Answer position changed.
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!Question Topic i I RO 21 Given the following conditions: - A SBLOCA has occurred. - The crew is performing the actions of EOP-LOCA-2, POST LOCA COOLDOWN AND DEPRESSURIZATION. - SI pumps have been stopped. - Normal Charging is aligned. - The crew is depressurizing the RCS using normal spray.
Which of the following describes the strategy for controlling RCS subcooling during the depressurization?
§]I Subcoolino will be ...
minimized to reduce RCS break flow. I I
~ maximized to ensure continued RCP operation.
I
~ maximized to prevent a challenge to the core cooling critical safety function.
I l.d. I
! minimized to ensure pressurizer level remains above the lower limit to allow heater operation during the RCS cooldown.
I IAnswer I
.!al iExain Level I fRl 1Coal1itlve Level ] I Memorv ! !Facility: q Salem 1 & 2 l~ExamDate: W 12/21/20151 IKA:ll OOWE03A103 IiEA1.3 _jiRO Value: i 3.7l lsROValue:Jl 4.1! [secti6n:Jl~IRO Group:JI 211s~o Grollp:JI 21 D
[System/Evolution Title j j LOCA Cooldown and Depressurization
KA Statement: I ll~.plal'lat\o. h of I 55.41.b(10) Strategy of step 31 is to depressurize and attempt to minimize subcooling so that break flow is reduced, due to the
.Answers:** .*. I minimal makeup provided by charging pumps. B is incorrect because RCP operation is not required for this event, although desired. C is incorrect because core cooling should not be challenged on loss of subcooling at these temps and pressures(this point in the cooldown) D is incorrect because PZR heater operation may be required to reduce the rate of increase in pressurizer
. . . <.Reference Title I! i Facility Reference Number * \Reference Section .* i I Page No.\ !Revision!
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[Lo. Number J I LOCA02E001 Objectives '--~ !Mat~ri!ll Requiredfor EXamination?*. ; I lj [o':'estie~~ourc~: **J IFacility Exam Bank l i(l~e~tion llJ!odifi~~tion<ftii~thocf,: ;+~Direct From Source I !u~ed During Training Ptogralll I D
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Outline Changes I . . t<lue~tion Topic ! IRO 22 I Given the following conditions:
- Unit 2 is attempting to identify and isolate a LOCA outside containment. - 2-EOP-LOCA-6, LOCA Outside Containment, has just been entered. - The source of the water is backleakage from the 23 cold leg injection line.
Assuming that any valves required to be operated during LOCA-6 operate correctly, which of the following leak locations would NOT be isolated while using 2-EOP-LOCA-6?
~I On the valve inlet flange on 22SJ49, RHR DISCH TO COLD LEGS.
I
~I On the valve outlet flange on 21SJ49, RHR DISCH TO COLD LEGS.
I rz1 IBetween the 2RH20, RHR HX BYP VALVE and the 2RH26, HOT LEG ISOL VALVE. I
@]I Between the 2RH2, RHR COMMON SUCT VALVE, and 22RH4, RHR PMP SUCT VALVE.
I \Answer 1 I b l IExa!llieveti IR I !cognitive Level : 11 Application I (Facility: ! ISalem 1 & 2 I IExamDate:' 11 12/21/20151
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~I OOWE04K202 i~!SRO V~ilue: ir.I§j" !Section: \I~ \Ro Group:\ LJ )SRO Group:\ LJ Ill;, li~ I l 11 EK2.2 ! !RO Value: D isyster:n/Evo!Ution Title I ILOCA Outside Containment I ~J iKAStatement:i KnowledQe of the interrelations between LOCA Outside Containment and the followinQ: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility. fExplanation of 55.41(3,8) 2-EOP-LOCA-6 closes/checks closed the following valves: 2RH1 OR 2RH2, 21 and 22RH19s, 2RH26, 21 and
Answers: 22SJ49s. Using drawing 205332-SIMP, it shows that any leak between the RH1/2 and the SJ49s will be isolated with the above valves closed. The only location which wouldn't be affected by those valve being closed in the downstream/outlet side of the SJ49 valves. The stem statement of proper valve operation was inserted to preclude a candidate from assuming a leaking valve may not I
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Reference Title 11
*' Fac:ility Reference Number, i !Reference Section i 11 page No. I IRevision[
ILOCA Outside Containment !12-EOP-LOCA-6 I II 1121 I I ii 205332-SIMP I ,,11 11 I I II I 1 I I /LO,, Number// Objectives I LOCA06E002 I_ ___. !Material Required tc>r.ej(a!lliriation .\I I la1;1e~tion Soup::e: xl l Facility Exam Bank IJouesti.j>n llJlo~ificati~n Met~od:'.: I Direct From Source I!Used Durlllg Training Pr69r<1m l o !Question Source Comrri'~ritsl 1133792. Used 3 NRC exams ago Sept 2011 RO exam I icomment I I I I l I
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!Questi<?n Topic i IRO 23 Given the following conditions: - The crew is in 2-EOP-FRHS-1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, and the criteria for initiating RCS bleed and feed has been met. - Prior to the actual procedure steps there is a CAUTION statement that reads:
TO ESTABLISH RCS HEAT REMOVAL BY RCS BLEED AND FEED, STEPS 24 THRU 29 MUST BE PERFORMED QUICKLY AND WITHOUT INTERRUPTION Which of the following describes the basis for that statement? I
~ Stopping RCP's is the first step in the process. This terminates all RCS heat removal until bleed and feed is initiated. ~I Expeditious performance of the steps allows time for other compensatory actions if bleed and feed actions are unsuccessful.
l:J IDelay allows core cooling to degrade further. RCS pressure rises such that ECCS flow is lower when bleed and feed is initiated. I [] I Establishing SI flow and then delaying opening the PZR PORVs may lead to damage to the PORVs and Code Safety Valves when they pass water. I !Answer 1 c I I jexam Level I lR I !cognitive .Level 11 Memory I I !Facility: 1 Salem 1 & 2 I jExamDate: 11 12/21/20151 ~joowEo5G420 l1RoVatue:1(2]'tsROValue::@1section:;!~IROGroup:[LJ~~LJ !:;i~I D J.~~~ I ' I ! ! *I I 112.4.20 ' iSystem/Evolutioii Title i ILoss of Secondary Heat Sink I E05 1 I IKA Statement* I Knowledge of operational implications of EOP warnings, cautions, and notes. 1i:xplanati?"'Pt,.j 55.41.b(1 O)Per FRHS Basis document states that delay allows further degradation of cooling, followed by RCS pressure rise, and ,Answers: * .**
I core uncovery may be greater because ECCS flow is limited by RCS pressure. A is incorrect because there is still some cooling after the RCP's are stopped since there is water left in the SG's. B is incorrect because there are no alternative steps. D is incorrect because ECCS flow is actually reduced by the delay - the PZR will not go solid.
f* Reference Title ****!I .. Facility Reference Number ** . 1[Refere~ce. Section.* *;I tPage No.! IRevision 1 ILoss of Secondary Heat !12-EOP-FRHS-1 II Basis Doc 11 7 1124 I l II II 11 11 I I II II 11 11 I il.O. . Number. I FRHSOOE009 Objecti '-~---'far lMateri~I ReqUirea Examination. *. II 11 tou~~tio~ sou,rc.ff l IFacility Exam Bank I [Question ¥odificatioll;o/)etl}<;>e: JDirect From source I(used pµring Training Prograi:rt I D fpu,~stiorfs~i.lfoe corrimeritsj 180840
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!Question Topic 11 RO 24 Given the following conditions for Unit 2: - A LOCA has occurred. - While performing 2-EOP-LOCA 1 "LOSS OF REACTOR COOLANT" 22 RHR pump motor seizes and power is lost to 21 RHR pump. - The crew enters 2-EOP-LOCA-5, "LOSS OF EMERGENCY RECIRCULATION". - A cooldown has been initiated as directed in 2-EOP-LOCA-5. - During the cooldown, the crew restores power to 21 RHR pump. - RWST Level is 30' Based on current plant conditions, which of the following describes the mitigation strategy?
g) IContinue with the cooldown and start 21 RHR pump when directed in 2-EOP-LOCA-5. I
~ Return to 2-EOP-LOCA-1 and continue recovery actions with the step previously in effect.
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~ Start 21 RHR pump and continue actions of 2-EOP-LOCA-5 until the RWST LO Level alarm actuates. I @] I Start 21 RHR pump and transition to 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation, to verify recirculation flowpath.
I !'Answer 1 I b I !Exam Level 11 R I jcognitive Level
f 1 Memory I !Facility: ! ISalem 1 & 2 I IExarnDate: ! I 12/21/20151
[KA:ll OOWE11A103 !1EA1.3 : jRo'Vatue: ll 3.7l[SRO Value: :j 4.2lfSection: !I~ iRO Group:JI 1![sR.o Group~!I 1j IJ~ D !System/EvolUtlon Title l ILoss of Emergency Coolant Recirculation I 1E11 IKA Statement:! Abilit to operate and I or monitor the followin as the a pl to Loss of Erner enc Coolant Recirculation: Desired operatin results durin abnormal and emer enc situations. IEXPl~nation of 55.41.b(10) B is correct because Continuous action step states that if any train of emergency recirculation capability is restored then iAriswers: the crew should return to the procedure and step in effect. This is consistent with the organization of the EOPs. C is incorrect because Continuous action Step 6.1 directs return to evaluate train status and a return to the procedure in effect; A is incorrect because continuation of the cooldown in LOCA-5 is not required. The purpose of the procedure is mitigation and recovery of
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Iquestion is not considered SRO level based on only requiring general mitigative actions j.;'*, Reference Title*.* I[
Facility Reference Number * '[ReferencE! Section* / [ f Page Nod [Revision[
ILoss of Emergency Coolant Recirculation 112-EOP-LOCA-5 I ii 1125 I I II I 11 ii I I !! I II 11 I fG.c): ~umber** *** *. ..... *[ Objectives I LOCA05E008 I il\llaterial Required t,or~J<amiriatiofj . ;:j I ll 1aue~ti.~n Source;. IIFacility Exam Bank I 1 laHist~<<:in Modifip~~eri:,,(IA~tho~! *** Direct From Source I ~Used During Training Pr(:>9raJli j D
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I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I.. .. louestioh Jopic 11 RO 25 I Given the following conditions:
- Unit 1 has experienced a MSLB at the mixing bottle. - MSLI has failed to close ANY MS167. - Operators have transitioned out of 1-EOP-LOSC-1, Loss of Secondary Coolant. - RCS pressure is 1345 psig and dropping slowly. - RCS Tes are dropping.
Which of the following describes Reactor Coolant Pump strategy, and why? RCPs should ... I
~ be tripped to minimize the heat input into the RCS.
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~I continue to be run, because RCP pressure dependent trip criteria are only applicable during a LBLOCA.
I [] I continue to be run, since RCP pressure dependent trip criteria is not used when a cooldown is in progress. I [) I be tripped so that 2 phase flow doesn't develop if they were to be tripped later in the event, leading to higher peak clad temperatures. I IAnswer 11 c I [Exam Level I l R I !cognitive Level. ! I Application I IFac~lity: 11 Salem 1 & 2 I !ExamDate: 11 12/21/20151 I lKA: 1 OOWE 12A202 llEA2.2 !IROValue:!j 3.4l[SROValue:jl 3.9llsection:ll~IROGroup:ll 1l[sR0(3roup:ll 11 !II D iSystem/Evolutioh Title i IUncontrolled Depressurization of all Steam Generators I IE12 1 KAStatement: I~ determine and interpret the following as thev apply to Uncontrolled Depressurization of all Steam Generators: rence to appropriate procedures and operation within the limitations in the facilitv's license and amendments. I [F'<Pl~nati?~. ?1.1 55.41.b(10,3)The RCP trip criteria with regards to pressure in a LOSC condition is for pump protection only. The Generic issues ,Answers:** J segment of the ERG executive volume describes the SBLOCA scenarios where pumping coolant out the break then stopping RCP's leads to peak clad temps in excess of 2200 degrees. That is not applicable here. What is more important in the LOSC is forced flow. LOSC-2 basis document page 12. I " Reference Title'* Ii Facility Reference N,umber * .I [Reference Section > l IPage No'. I !Revision* IMultiple Steam Generator Depressurization 111-EOP-LOSC-2 I q12 1122 I I I I 11 11 I ! I I II 11 I Objectives I __ LOSC02E003 , IMatefialReq1;1fr~dfor Exarhi11ation **l j lj IS:~~stion so.urce: *I IFacility Exam Bank I [9u,e~tion Modificatiorl:ivfetho,c!M~* ~Direct From Source I!used During "fiail1ing Prpgram ID
~ollestipJ1 sourc~*com~entsJ 163166
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[ouestionTopic I l RO 26 I Given the following conditions:
- Unit 2 has experienced an event which has resulted in 24 SG pressure rising to 1115 psig. - A MSLI has been performed.
How many total SG Safety Valves will be open if 24 SG pressure remains at 1115 psig and all Safety Valves operate when expected?
~12 I
[)13 I
!c.f 14 I @]Is I
!Answer 11 b I !exam Level 11 R I \Cognitivl! Level 1 IMemory I !Facility: ISalem 1 & 2 1 I [Ex~mDate: !I 12/21/20151
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~jooWE13K202 "iiROValue:J~ISROValue:J@1section:;;l~:ROGroup:JLJ1SROGroup::LJ ~ D I *.* ., I ':"'] l,EK2.2 iSystem/Evolution Title i jSteam Generator Overpressure 11E13 I ,KA Statement: I Knowledge of the interrelations between Steam Generator Overpressure and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility. i [Explanation of 55.41.b(4) Each SG has five safeties, with lift setpoints of 1070, 1100, 1110, 1120, 1125 psig 1Answers: * . *. I
j iReference Section
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*. Reference Ti.tie ... It Facility Reference Num1:>er:
IUnit 2 Main Steam System 11205303 Sheet 2 I lj61 I I II I 11 11 I I II I II 11 I !Lo. Nu.!Tlber I STMGENE008 Objectives j_ ____. [Material Required torExaminatiori. !I lj IQuestion s?urce: ... 11 Facility Exam Bank I [puestion IVlodification Method:,r l Direct From Source I [usec!During Trajning Program ;l D jouestr~n Sou.rce Comm7,nts\ 1152360 used 3 NRC exams ago Sept 2011.
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!Question Topic J 1RO 27 I IOf the following choices, which is the only automatic action expected to occur as containment pressure rises from 12 psig to 18 ~sig during a LOCA? I nl 1a*:
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~ I Ei I Feedwater Isolation.
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~ I Main Steam line Isolation.
Containment Ventilation Isolation I I I !Answer 1 c I !Exam Level ! ~ !cognitive L.evel 11 Memory ! !Facility: 11 Salem 1 & 2 I \ExamDate:;i I 12/21/20151 IQ2j [SRO Value: IQ2j lSection: 11~ :RO Group:! LI !SRO Group:[ LI l~~i; D
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I !KA:ll OOWE14A101 ! !RO Value: ls~~~~~~ooTWej l-H~ig~h_C_o_n_t_ai_n_m_e_~_P_r_e_u_u_r_e_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~llE14 'KA Statement:: Ability to operate and I or monitor the followinq as thev aoplv to Hioh Containment Pressure: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. I i !Explanation of 55.41.b(7) Phase A isolation occurs on any SI signal, which at the very least would have occurred at 4 psig in containment if not Answers:. l sooner. Feedwater Isolation occurs on any SI or SG NR level >67%. Main steamline Isolation occurs at 15 psig. Containment Ven isolation occurs any SI, RMS alarm of associated monitors, Phase B. The SI signal would have already actuated. I Reference Title j; Facility Reference Number *I IReference Sec.tion j IPage No. j jRevision: ILicensed Operator Fluency List ii NOS05FLUNCY I 1113-16 11 9 I I II I q 11 I I II I II !I I IL'.O'. Number **
Objectives FLUNCYE002
\M(lterial Required for i:xamination : II [auestion So1.frce:. J IFacility Exam Bank I !<lu~sticin.Modification Method: j Concept Used I!used During Training Progretm I O 1auestfon Source Cofl}'!le~ts\ I
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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I ......... ..... louestionTuPicJ IRO 28 I Which of the following identifies when will OHA E-48 ROD BOTTOM will clear during a reactor startup? When 1) rods are withdrawn past 2) steps.
~ 11) Control Bank A 2)20 I ~ 11) Control Bank A 2)35 I ~ 1) Control Bank D
2) 20 I
[] 1) Control Bank D 2)35 I !A.nswer 11 a I !exam Level ! IR I !Cognitive Level . JI Memory I !Facility: 11 Salem 1 & 2 1 IExamDate: ! j 12/21/2015! ~I 001 OOOA302 *IRO Value: IfJ2i !SRO Value:!~ *'Section: 1I~ jRO Group:! LJ !SRO Group:] LJ lit~&~ D
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\system/Evolution Title j j Control Rod Drive System 11001 I !KA Statement: i Abilit to monitor automatic operations of the Control Rod Drive S stem includin hei ht !Explanation of I 55.41.b(6) There are 3 Rod Bottom Bypass Bistable Modules, for Control Banks B, C, and D only !Answers: . i 1)Blocks OHA E-48, ROD BOTTOM for its own bank when entire bank is <35 steps
2)When all banks are on bottom, OHA E-48, ROD BOTTOM alarm is illuminated.
3)When Control Bank A is >20 steps alarm is cleared (Control Banks B, C, D, bypassed)
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!>:* Reference Title.: ** II . Facility Reference Numbe.r . j ;Ref.erence Se~tion . 11 Page ~o. I IRE!yisionl IRod Control and Position Indications Systems L !INOS05RODSOO I lj 51-52 1112 I I II I 11 II I I II I ii 11 I LO. Number.* IRODSOOE008 Objectives I ,_ _ ___. itii1aterial Reqliireci tor e~amination 11 II [ouestion Source: 11 Facility Exam Bank 11auestion Modificati~.n fv1etho~: : ~Concept Used 1lused otii"ing Trainil"lg P/ogramdl D l<:luestion source Comments! 157724 =======-==::::==============~===========-================--!
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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes [Question Topic,J 1_R_0_2_9_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___. Given the following conditions:
- Unit 2 is in MODE 6. - The Reactor vessel head is removed and on its storage stand. - The Refueling Cavity is being filled from the RWST IAW S2.0P-SO.SF-0003, FILLING THE REFUELING CAVITY. (Assume RWST level was at 40.5' when the unit was shutdown) - Refueling cavity water level is 11 O' and rising.
Which choice identifies the indications which will be present on the Control Room Console? PZR Cold Calibrated level is ...
~ 13%; RWST level is 27'. ~ 180%; RWST level is 20'.
rel Ioff-scale low; RWST level is 40'.
@] I off-scale high RWST level is 1O'.
!Answer I [L:J 'Exam L~vel : !&:J !Cognitive Level I !Application _l_JFacUity: wSalem 1 & 2 l~ExarTlpat~: w___1_2_12_1_12_0_15_,I ~ j 002000A111 I A 1.11 1 I!Ro Valut}: !@[SRO Value: l@fsection: ii~ [Ro Group:j[}j ls Ro Group:j LJ D [System/Evolution Title> I l_R_e_a_ct_o_rC_oo_l_an_t_S....:y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __, ~--~ !KA Statement: 11 Ability to predict and/or monitor chanqes in parameters associated with operatinq the Reactor Coolant System controls including: I I Relative level indications in the RWST, the refueling cavity, the PZR and the reactor vessel durinq preparation for refueling I fExplanation of : 55.41.b(3,)The refueling cavity at 11 O' will hold -112,000 gallons. The RWST normal operating level is a minimum of 40.5 feet prior jAnswers: ; * ** j to refueling (364,500 gal). Subtract and the total left in RWST will be 253,000 gallons, which will be -27.5'. The corresponding PZR cold cal level will be > the 0% level at the 108' 11" elevation in containment. This question does not require memorization of tank levels, but rather an understanding of the physical connections and relative elevations of the systems. ! .Reference Title il Facility Reference Numbt}r:. .. !\Reference Section J i Page .No. I !RevisionJ IDraining the Refueling Cavity II S2.0P-SO.SF-0004 11 1129-30 1117 I I II II q 11 I I 11 II 11 11 I [LO.Number Objectives I RCSOOOE006 1_ ___. !Material RequiredfofExarriinatiori <I I lj [Question Spurce:. J IFacility Exam Bank I !Question .Mo,dificati9n Method: . :I Direct From Source Iiused During'rralning Prpgrain I D
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!Question Topic 11 RO 30 I IPrior to opening 1CV114, RCP Seal Bypass Valve IAW S1 .OP-SO.RC-0001, RCP Operation, all of the following conditions must be met EXCEPT: I ~ I RCS pressure < 100 psig.
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~I Any RCP seal leakoff flow< 1 gpm.
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/cl I 11-14CV104, SEAL LEAKOFF valves open.
I [] I Seal water flow to each RCP is at least 6 gpm. I [Answer 11 a I iExain Level I I R I !cognitive Level 1 IMemory I !Facility: 11 Salem 1 & 2 I texamoate: !I 12/21/2015!
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~I 003000A407 @ET 'SRO Value: 10]" '~Section: 1!~1RO Group:JLJ !SRO Group:I L J ~ i;>~. ~ I
I IA4.07 1 IRO Value: I D I
!S~tem/Evolution Title I Reactor Coolant Pump System I ~0_03_~ IKA. Statement: I Ability to manually operate and/or monitor in the control room: RCP seal bypass IExplap~tton of ! 55.41.b(3) RCS pressure must be between 100-1,000 psig. If pressure is less than 100 psig, the CV114 is required to be shut. All Answers: : distracters are listed in SO.RC-1 step 5.2.1.
. *****.** .. /*Reference iitte II **Facility Reference Number'> i lReference Section .* j i Page N()i! iR,eyi~ionl I Reactor Coolant Pump Operation I 1 S1 .OP-SO.RC-0001 I 1113 1134 I I II I q II I I II I II 11 I
\L.o. Number I RCPUMPE013 {s Objectives I 1_ _____, [Materi13t Required for ExamiQation . II 11 fQuestton Source.: j IFacility Exam Bank 1 IQues~ion l'Aodification M~thod: J Concept Used I [Used During "fraining Program J D i~~~~~~~l~4_3_0_~-----------------------------------~I jcomment** I I I
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SRO Skyscraper I...... RO System/Evolution List I SRO System/Evolution List I Outline Changes I lauestion Topic 11RO31 I Operators are preparing to start 21 RCP IAW S2.0P-SO.RC-0001, Reactor Coolant Pump Operation. Which of the following would prevent the RCP from starting when its START PB is depressed? 21 RCP ...
~ I#1 seal D/P < 200 psig.
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~I 4KV breaker trip springs not charged.
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~I STANDPIPE LEVEL LO alarm locked in. @JI Oil Lift Pump discharge pressure <500 psig.
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... Level . !IMemory .. .1 ISalem 1 & 2 I
["O" 1Exam Level ! ~ !Cogmbve i I 1Fac11lty. I ,ExamDate.: :11I
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)Answer) 1 1 12/21/2015h l~:l j 003000K101 IIK1 .01 [I Ro Vallie: I[]]" :5Ro Value: I12] !section: ii~ IRo Group:j LJ lsRo Group:I LJ !!'~ D IS:Ystem/Evoiution Title I j Reactor Coolant Pump System 11003 rKA Statement: l KnowledQe of the physical connections and/or cause-effect relationships between Reactor Coolant Pump System and the followinQ: I RCP lube oil I iExplanation !. . of iI 55.41.b(3) A is incorrect because it is a manaul RCP trip setpoint stated at Step 3.2.9 third bullet. B is incorrect because the trip
Answers: i springs are charged when the breaker closes, it woiuld be the closiong springs not charged which would prevent the 4KV braker form closing. C is incorrect because the standpipe level low has no interlock to prevent pump starting. D is correct as shown on drawings.
I Reference Title I Facility Reference Number I~Reference Section* j !Page No. I !Revision! IReactor Coolant Pump Instrumentation 11220424 I ii ii 05 I IRCPs and RCPs Lift Oil Pumps 11224405 I ii 114 I I II 11 II ii I iLd. Number I RCPUMPE006 Objectives J 1_ __ . . )l\llaterial Required for Examination 11 II 1auestioh Source; 11 Facility Exam Bank I[auestiC>ri Me>dificatie>l"l ~et~o~: JI Concept Used I tused During Traiiling Prograr:n J D fOU esHe> n Sq~r~e corri~eri.t~!
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\Question Topic I !RO 32 I I With normal charging and letdown in service, which of the following valves failing closed is interlocked to cause 1CV4, Letdown Orifice Isolation Valve, to automatically shut? I ~ 11CV18, Letdown Pressure Control valve. ~ 11 CV??,
I 23 Loop Chg Line Stop valve. ic. i 11 CV?, Letdown HX Inlet Isolation valve. I
~I 1CV2, Letdown Isolation valve.
I !Answer J d I I IExam Level i IR l [cognitive Level 11 Memory j jF~cility: 11 Salem 1 & 2 1 IExamDate: 11 I 12/21/20151 IKA:JI 004000K413 IiK4.13 I!Rd Value: lj3.2*1 iSROValue:1I 3.51 [section: II~ ~RO Group: I 1 11 lsROGroup:ll 11 1 1~'~ D lsystern/Evolution Title J IChemical and Volume Control System 11004 IKA Statement: J I Knowledqe of Chemical and Volume Control Svstem desiqn feature(s) and or interlock(s) which provide for the followinq: Interlock between letdown isolation valve and flow control valve !Explanaticm pf I 55.41.b(?) The 1CV2 being closed would cause any of the 3 open orifice isolation valves to shut. 2CV4 is stated in the stem to be iAnswers: *.*.' the open isolation valve. i I.
Reference Title **I:. Facility Reference Number

I/Reference Section * !
11 Page No. !Revision/ I1CV4 Logic Diagram ii 224430 I ii l l5 I ! II I II 11 I I II I ii 11 I Lo. Number
I Objectives I CVCSOOE006 j_ ___.
[Material Required for Examination II II IQuestion Sciurce: .*.11 Facility Exam Bank I fGW.e~tion M9clification Method:/ . JConcept Used 1lused outing TrainJng Program. I D
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[Question Topic I IRO 33 I Which of the following methods is used prior to cooldown to remove hydrogen from the RCS after a unit shutdown to enter a refueling outage, and why is hydrogen removed? deqassification is used to remove hydrooen from the RCS to prevent I
~ Chemical, an inadvertent crud burst during RCS cooldown.
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~ Mechanical, an inadvertent crud burst during RCS cooldown.
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!:§] IChemical, having an explosive concentration present when oxygen is introduced to system.
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~I Mechanical, having an explosive concentration present when oxygen is introduced to system.
I IAnswer 1 Id I !ExarhLevel 11 R l !cognitive Level i IMemory I IFacilitY:'J ISalem 1 & 2 I \Ex~mDate: \ I 12/21/20151 ~I 004000K549 I[K5.49 I \RO Value: llJ2itsRO Value: l@lsection: II~ \Ro Groop:jLJISRO Group:\LJ i55l31 P-illl~,t~ ltliJ1~!6Iill D [system/Evolution Title I IChemical and Volume Control System 11004 I
KA Statement:! Knowledge of the operational implications of the following concepts as they aooly to the Chemical and Volume Control System:
Puroose and method of hydrooen removal from RCS before openina svstem: explosion hazard, nitrogen puroe [Expl~r;iat'.on of I 55.41.b(3,5) Both chemical (addition) and mechanical (replacement of H2 with N2) degasses are performed during a 1 Answers. . .* I shutdown/cooldown. Mechanical is most effective at NOP/NOT (see P&L 2.1.5) A is incorrect because chemical degass is performed when RCS temp is less than 250 (see pre-req 3.1.7). B is incorrect because of wrong reason for mechanical degas. Cis incorrect because chenical degass is performed to initiate crud burst (addition of H202). i '* Reference Title . I' : *.Facility Reference Number '* I!Reference Section mment I I I l I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes lauestion Topic 11 RO 34 I Given the following conditions:
- Unit 2 has experienced a Large Break LOCA. - 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation, is complete with NO abnormalities encountered. - Operators are currently at step 26, "Preparation for Hot Leg Recirc", of 2-EOP-LOCA-1, Loss of Reactor Coolant. - Off-site power is supplying all 4KV Vital busses.
If BOTH RHR pumps are operating, what would be the effect if 22 RHR Pp were to trip?
~ 121 and 22 SI pumps would begin to cavitate.
I
~ 122 Containment Spray Pump would lose NPSH.
I lS 121 and 22 Charging pumps would begin to cavitate. I
@] I Flow to the Containment Spray header would be lost.
I IAnswer ! d I I [Ex~m Level i IR I !cognitive Level i IApplication I if'acility: j ISalem 1 & 2 I [ExamDat:: J I 12/21/20151 IKA:JI 005000K305 j IK3.05 11Ro Value:!l3.7*1 iSRO Value:i13.B*l lsection: ljsYS l[Ro Gfoup:!I 11 ISROGroup:il 11 II! D [System/Evolutic:m Title i IResidual Heat Removal System I ~I0_0_5_~ IKA Statement:! Knowied e of the effect that a loss or malfunction of the Residual Heat Removal S stem will have on the followin ECCS lExplanati<m of j 55.41.b(5) Distracter A is incorrect because the closure of the RH19's is done to prevent RHR pump runout if only a single RHR Answers: i pump is operating, so the SI pumps will not lose suction. D is the correct answer. LOCA-3 explicitly states that if BOTH RHR pumps are operating, then 22CS36 is opened to supply containment spray from 22 RHR pp discharge. Distracter C is incorrect charging pumps, as well as the SI pumps, will not lose suction.
. 11se all containments. .
I. Reference Title *! I
Fa(::ility Reference Number :I IReferenceSection P~ge No. I [Revision' l~L=o=ss==of=R=e=a=c=m=r=C=o=o=la=nt==============~ll~E=O=P=-=L=O=C=A=-1==============~~==============~~====::::;.!.::=::::::::=:::.
!Lo. Nlllllber J I ECCSOOE016 Objectives I RHROOOE016 j_ ____, !Material Requir~<J for Examinafigp **I I I \a!JE!stion Source: *J Facility Exam Bank I ISp~5-ticm Mc@ficaHon.Method: . {I Editorially Modified 1 lusedollrlng TrliiJlin~(Progt,am I D [auestiorlSolirce Gomme~tsj
I Modified correct answer terminology from Containment Spray flow would be lost to Flow to the Containment Spray header would be lost.
I !Comment***** I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question Topic 11 RO 35 I Given the following conditions:
- Unit 2 is operating at 100% power with 22 charging pump in seNice. - 23 charging pump is out of seNice and operable.
Which of the following Tech Spec LCO's would be applicable if 22 charging pump were to trip? [ ] 13.5.4 - Seal Injection Flow.
~ , 3.5.2 -ECCS Subsystems >350°F.
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§] ,3.1.2.4 -Charging Pumps - Operating.
[ ] , 3.1.2.2 - Boration Flowpaths - Operating. I I IAnswer 11 b I [EX:a~ Liwe1 IR I t :cognitive Level *! IApplication I !Facility: 11 Salem 1 & 2 I !ExarnOate: 11 12/21/20151 I~i2_.2_.4_2_~1 IRO Value:!~ [SRO Value:j~ ~Section: II~ IRO Group::LJ :sRO Group:JLJ Ila~ D
>X"'*~-;F<'"c~"~
I IKA: 11 006000G242 ~[s-y-st_e_m_/_E_v_o_lu-ti_o_n_T-it-le~I IEmergency Core Cooling System 1 i006 I rKA Statement*! Ability to recoqnize indications for svstem operatinq parameters which are entN-level conditions for technical specifications. IE~plctnation of i 55.41.b(6, 10) RO candidates are responsible for "Above the Line" knowledge of Tech Spec LCO's. A is incorrect because the LCO IArtsi.vers: ***. *.. .
I is for maximum seal injection flow which is set by adjusting manual seal injection throttle valves. Plausible because many procedures direct establishing 6-12 gpm seal injection flow per pump not to exceed 40 gpm total. Bis correct because the LCO states 2 complete trains of ECCS are required, and 22 charging pump (hi head ECCS) is required for the B train of ECCS. C is
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pumps were inoperabie. D is incorrect 'because 2 boration flow paths from RWST through charging pumps remain available: as does the Boric Acid tank flowpath with one BAT pump and one charqinq pump.
.*.*.Reference Title ;>*/'
II .. Facility Reference Number. *I!Reference section *** :J f Page No: I !Revision! j Salem Tech Specs II I II 11 I ! II 11 II 11 I I II I 11 II I Objectives tlVl?tJiria!ReqliiredfoiE.xamination ; 11 !j fq~~~ti?ll Source.:
11 New 1 louestion. Moclification ~?t.Mct: g,*1--------~I !used Dt:iringJrainirig Program:J D IQue,stJor,i Soprc~c?rnrn~~tsi I*~~~~~--~~~~~~-~~-----~~~~~~~~~~~~~--~~~--'
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RO SkyScraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes IQuestion Topi~ IRO 36 Given the following conditions:
- 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.
Assuminq NO operator actions were taken, which of the followino would result in those conditions?
~ IRCP #1 Seal failure.
I [g'.] IPressurizer Spray Valve failed open. I I
~ Pressurizer Safety Valve failed open.
[] I Steam Generator Slowdown piping failure. I !Answer 11 c I !Exam Level I lR I !cognitive Level JIComprehension I :Facility: i ISalem 1 & 2 I IExarriDate: !I 12/21/20151 ~!007000K101 l!K1.01 ltROValue:i~!SROValue:j[ITI[section:il~IROGroup::[]:sROGroup:!LJ BJ D
*J'.'F!l
!System/Evolution Jitle I l Pressurizer Relief Tank/Quench Tank System 1 l~o_0_7_ _ iKA Statement: I Knowledge of the physical connections and/or cause-effect relationships between Pressurizer Relief Tank/Quench Tank System and the following: Containment system i fEicplariatio"n of 55.41.b(9)The failure of a RCP #1 seal will not be seen in containment outside of closed systems. The excess seal leakoff flow past ,Answers: .* *. , l the #2 seal will be seen as a rise in RCDT level. The PZR safety failing open will cause the PRT to pressurize and the rupture disk to rupture, causing the saturated steam in the PRT to continuously be vented to containment. The Spray valve failing open would cause the lowering PZR pressure, bit would not cause changing containment conditions. Reference Title
I: .Facility Reference Number*. I Reference ,Section j [ Page No. I !Revision:
INo. 2 Unit Reactor Coolant !1205301-1 I II 1 lsg I II I 11 11 I II I ii 11 I IL.a. Nurriber .. *. I PZRPRTE008 Objectives I ,__ ____, [Material Required for Examiilation * ;:;J I 11 I l<:l~estio~. Source~*--' J Facility Exam Bank !lauestiori Modificatio~.11A7thbd: ** _** 11 Direct From Source I iuse~Quring'TrainingPrograrri'* I D IQues,tion s.o~~cl! Comrrieritsi ISalem May 2010 RO NRC exam, created from Point Beach 1/20/2006 NRC exam I
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!Question Topic 11 RO 37 I Given the following conditions: - Unit 2 is operating normally at 100% power. - Excess Letdown is in service due to a problem with an orifice isolation valve. - The 2CC215, EXCESS LETDOWN HX CC INLET Vair supply line breaks, and all air is vented from valve.
Which of the following describes the effect this failure will have on Excess Letdown temperature, and how will operators respond? Excess Letdown temperature will. .. la. I I rise and the 2CC215 bypass will be throttled open to restore normal letdown temperature. I I
~ lower and the 2CC215 bypass will be throttled open to restore normal letdown temperature.
I
~I rise and Excess Letdown flow will be secured to prevent lifting 2CV115 eve RCP SEAL WTR INJECTION RETURN RELIEF VLV.
I I Id. j lower and Excess Letdown flow will be secured to prevent VCT temperature from lowering to the point where 23 charging pump must be secured. I I.Answer I j c I !Exam .Level I lR I !Cognitive Level*. I IApplication I !Facility: 11 Salem 1 & 2 1~1 12/21/20151 ~ j 008000A205 I1RO Value: i@:;!JfsRO Value: IQ}i !Section: 11~ \RO Group:] LJ :SRO Group:J LJ I?~:' D
~W,'!.~;
1IA2.05 !System/Evolution Title 1 j Component Cooling Water System I ~lo_08_ _ !KA Statement: I Ability to (a) predict the impacts of the following on the Component Cooling Water System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Effect of loss of instrument and control air on the position of the CCW valves that are air operated !Explanation oft 55.41.b(?) The Excess Letdown HX CC isolation valves 2CC113 (outlet) and 2CC215 (inlet) are fail close (air and power) isolation Answers: ! valves. With excess letdown in service and no cooling flow, temperature and pressure will both rise. Excess letdown flows into the seal return header and rising pressure causes seal return line relief valve to open. Knowledge of both the effect (valve failure position) and operator action (isolate excess letdown) required. SO.CVC-3 P&L 3.3 states to maintain excess letdown pressure
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1* Reference Title. r Facility Reference Number * ** \ lReference Section
Ij Page No. I l~evisionl IExcess Letdown II S2.0P-SO.CVC-0003 II ii 116 I Ieves drawing 11205228-2 II I 184 I I JI Ii
"!I 11 I ILO.~umber Objectives I CCWOOOE012
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!Material Reqllifodfor Examina~ion
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[Qm~stion ~ou~c;~/ 1j_N_e_w_ _ _ _ _ ___,l jCluestiori ~()C'imcatiop Method: .. ~ _ _ _ _ _ _ _ __.I Jused[>,uril)g Training Program I D
;~uestion Source Comm~.~t::i.11 I !~---------------------------------------'
fcomlllent I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I .. louestion Topic i IRO 38 I Given the following conditions:
- Salem Unit 2 is operating normally at 100% power. - The controlling PZR level channel fails low. - As the operators respond IAW S2.0P-AB.CVC-0001, Loss of Charging, and before an operable channel is selected for control, the RO reports that PZR backup heaters in auto are energized with the controlling PZR level channel indicating 0%.
Which of the following describes the operation of the PZR B/U heaters? PZR heaters in auto 1) be energized because 2) [ ] 11) should
2) PZR level has risen 5% above program.
[ ] 11) should
2) PZR pressure has lowered to the auto on setpoint of 2210 psig.
~ 11) should NOT
2) PZR low level cutoff at 17% should be keeping heaters OFF.
[ ] 11) should NOT
2) PZR pressure has remained above the auto on setpoint of 2218 psig.
!Answer 1 c I I !exam Level J IR I I !Cognitive Level .* J Application I [Facility: 11 Salem 1 & 2 I JExamDate: i I 12/21/20151 iKA:JI 010000A402 j:A4.02 : [Ro Value: JI 3.61 !sRO Value: II 3.41 lsectiori: II~ 'RO Group:!! 11 lsRo Group:JI 11 Ill D !System/Evolution Title I j Pressurizer Pressure Control System I ,010 IKA Statement: I~ manually operate and/or monitor in the control room: ters I [Explanation ' , .' . .-*. of lI 55.41.b(?,5) The low level on the control channel will cause an automatic letdown isolation. Charging flow will continue and raise
Answers: *
i PZR level. The backup heaters are designed to energize at 5% above program to ensure stauration conditions are maintained in the PZR. However, either an alarm or control channel failing low deenergizes all PZR heaters. Nothing has occurred which would cause a PZR pressure change except for the rise in PZR level. Pressure will not lower. PZR B/U heaters are designed to energize
-.+ 'JIJ'1n --=--I- ' - ~nrl +. ,.n - U ~+ 'l'l1 Q --;~ ' - - - - - - * - - ' -*
I Reference Title d Facility Reference Number }I tR~ference Section Ii PageNc). j jRevision' ILoss of Charging 11 S2.0P-AB.CVC-0001 II 11 119 I
- rl Annunciator Window E I 1 S2.0P-AR.ZZ-0005 ilOHA E-20 1129 1120 I I II II 11 11 I
!LO. Number Objectives I PZRP&LE006 1_ ____. M?terial Required for Examination/ II I fauestion Sou.rce: J l_N_e_w_ _ _ _ _ __.I faliestion Mo~ificaifon Method~, l_________.1 lusecl Duririg Training Progra01 IO iauestiori Somce Comments! I I
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!Question TOpic\ I RO 39 I IWhich of the following describes how the PZR Spra:I'. Nozzle is prevented from being thermallx shocked during normal operation? I ~I Initiating spray flow when the Spray Nozzle delta T exceeds 320°F.
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~ PZR Backup heater groups operating in auto forces continuous spray flow.
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~ PZR Control heater group firing to maintain PZR pressure at setpoint forces continuous spray flow.
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@]I A small amount of spray flow is bypassed around the PS1 and PS3 Spray Valves to keep the spray line continuously warm.
I !Answer ! j d I [Exam Level !IR I !cognitive Level 11 Memory I !Facility: 11 Salem 1 & 2 I IExamDate: 1 I 12/21/20151 IKA:ll 010000K401 !M.01 I!Ro Value: I 2.7i;SROValue:H 2.9! !Section; II~ fRO Group:!! 11 lsRO Group:JI 1j t!l'I D !System/Evolution Title.I IPressurizer Pressure Control System 11010 !KA Statement: I Knowledge of Pressurizer Pressure Control System desiqn feature(s) and or interlock(s) which provide for the follo\Ning: Spray valve warm-up I IExplanation~f I 55.41.b(5,7) PZR spray bypass flow is set during unit startup while@ NOT/NOP to ensure spray line temp is >500°F INith both ~Answers: I spray valves shut. Salem runs with one PZR B/U heater group in MANUAL which forces the spray valves open a small amount to provide continuous spray for boron mixing also, but that is not a choice. Spray flow is NOT initiated if spray line delta T exceeds 320°F. Control groups are SCR controlled and normally fire to maintain pressure ON program, and do not force spray fiow. ' f*;< Reference Title II *.*Facility Reference Number. .! !Reference Section *. l IPage No. / jRe\lision/ I Setting Pressurizer Spray Bypass Flow 11 S2.0P-SO.PZR-0008 II 11 114 I I II II q 11 I I II n II 11 l !Lg. Number. Objectives I PZRP&LE012 ,_ _ ___. l
!Questicm SCii.lice: .* l_N_e_w_ _ _ _ _ _ __.I L[Q_u:c__e_s-'tiolc'n.:..****..:..M_o_d_ifi~rc~a~ti..:..o_n..:..M..:..*e_t..:..~..:..C>_Ci~:.-J1 _ _ _ _ _ _ _ __,l /useif During Training Pr9gramd.j D
,[Question S()!-l~~EiC()ml,Tientsl I I
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RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes IOuestion Topic'! I RO 40 I Given the following conditions:
- Unit 2 was operating at 90% power with 21 charging pump in service when the controlling PZR Level Channel I failed low. - The Charging Master Flow Controller was placed in Manual when directed by S2.0P-AB.CVC-0001, Loss of Charging. - The alternate PZR level channel has been selected as the controlling channel. - Letdown has been restored.
Which of the following identifies a consequence of returning the Master Flow Controller to auto PRIOR to returning PZR level to program as directed in S2.0P-AB.CVC-0001 ? Charging flow will ...
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rise, and VCT auto makeup may initiate.
~ lower, and flashing in the Letdown line could occur.
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~ rise, and RCP seal injection flow could exceed Tech Spec limit total seal injection flow.
I
@]I lower, and 2CC71 LTDWN HX CC CONT VALVE will not respond quickly enough to prevent Mixed Bed Demin isolation on high inlet temperature.
I !Answer\ I~ !Exam Level I I~ !Cognitive Level 1 lAppl1cat1on -I !Facility: i j Salem 1 & 2 1lExamDate:11 _ _ _ 12_12_1_12_0_1_,5! I
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I !KA: i 011000A404 1 IA4.04 i iRO Value: I i"EJ fSRO Value: :[.Ifil [section: 11~ fRO Group::LJ !SRO Group:j L J D [SYStem/Evolution Title I IPressurizer Level Control System
KA Statement:! o manually operate and/or monitor in the control room:
Transfer of PZR LCS from automatic to manual control I ! . * .. rExplanation of I 55.41.b(7)With a CCP in service, the failure LOW of the controlling PZR level channel will cause charging flow to RISE. The stem !Answers: * ** 1 stated that MFC was taken to manual when directed IAW AB, so there was sufficient time for actual charging flow to rise substantially. With actual level higher than programmed level, if the MFC were placed in auto it would force charging flow to lower. If charging flow lowered to <-60 gpm, inadequate cooling of letdown flow would occur in the regenerative heat exchanger, and I, ....................... lino fl..,,c-hioa l~lCI 1ld CCCI IC IbP r'r-'71 it- nf"'l-........ ***ilh --'* -1 no/_ l"\l""IOn ............ L-.-- ................. ,..,f ............ _ +,.... ,.., ............ if*- .......... **- "---,n ............. +r. rise downstream ofthe letdown HX, and temps would not reach demin isoiation levels. The 2 rises are incorrect because charging flow wouldn't rise, but the actions associated with higher flow are correct. I Reference Title :1 Facility Reference Number .. I!Reference Section II Page No. I !Revision' ILoss of Charging 11 S2.0P-AB.CVC-0001 I ii 119 I I II I II II I I II l q 11 I \Lo. Numbe¥ *** Objectives I PZRP&LE015 ,_ ____,
I [g~~~ticm sou~g~n 11 Facility Exam Bank 11 Used Duriri~irr~fni~g Progran( I D
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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I
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lauestionTopic 11RO41 I Which of the following describes the interface between the Reactor Trip Handles on 2CC2 and the Reactor Protection System? Turning either Reactor Trip Handle to the trip position is designed to operate I
~ the UV trip ONLY for Reactor Trip AND Reactor Trip Bypass breakers.
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~ the shunt trip ONLY for Reactor Trip AND Reactor Trip Bypass breakers.
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~ BOTH the shunt trip and UV trip for Reactor Trip AND Reactor Trip Bypass breakers.
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@J IBOTH the shunt trip and UV trip for Reactor Trip breakers, and the shunt trip ONLY for the Reactor Trip Bypass breakers ..
I [Answer 11 c I IExarn Level j j R I !cognitive Level i IMemory I !Fa'sility: l ISalem 1 & 2 I (Exam bate:' 11 12/21/20151 IKA:ll 012000A401 1IM.01 !jRoyalue:1j 4.5jlsROValue:ll 4.5jlsection:ll~1RoGroup:jj 1ll5R6Group:/j 11 1~~1j D jSY5tern/Evohition Title j j_R_e_a_ct_o_r_P_ro_te_c_ti_o_n_S..:..y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __, ,_0_12_~ IKA Staternent: I Ability to manually operate and/or monitor in the control room: I Manual trip button I f Explan~tion of j Salem has 2 Reactor Trip Handles/Switches. Either switch operates BOTH the UV and shunt trips for BOTH the reactor trip ,Answers: . 1 breakers and bypass breakers. The distracters are plausibler because: 1) an automatic reactor trip ONLY actuates the UV trip; 2) manually tripping the reactor trip breakers from the control console ONLY actuates the shunt trip. I Reference Title >j;,, Facility Reference Number.. 1~Reference Section
I 1 Page No. i !Revision!
IReactor Protection System Reactor Trip Signals !I221051 II 11 1113 I I II ll II 1 I I I II !I II II I iL.0. Number
Objectives IRXPROTE010 IRXPROTE007 IJVlaterial Required for EJ<amin.ation II lj 1auestion ~6Jrce:.oj j_N_e_w_ _ _ _ _ _ __.l lauesti()!l Modificati()n.M~thog: **** 1--------~I !Used [)uring Training Program j D
!,Question ~o.4fce Comments\ I I
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RO Skyscraper SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes f louestion Topic !R042 I IWhat would be the effect on the Reactor Protection S}'.stem if the 2B Vital Instrument Bus were to become deener~ized with the unit at 100% power? I I
~ SSPS Train B slave relays would not actuate on a Safety Injection signal.
I I
~ OHA A-34 SSPS TRNA TRBL in alarm due to loss of 1 of 2 45VDC power supplies to Train A logic cabinet.
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~I Logic coincidence for Containment Spray actuation would go from 2/4 to 1/3 due to channel II bistable tripped.
I []I 2RP4 bistable lights flashing for all channel II indications due to train disagreement between SSPS Trains A and B. I !Answer 11 a I IExam Levell I R I :cognitive Level 11 Application I !Facility: 11 Salem 1 & 2 1 IExamDate: 11 12/21/20151 IKA:!I 012000K201 iIK2.01 I !Ro Value: JI 3.3lfSRO Value: !I 3.71 lsection: II~ (RO Group:H 1 j ISRO Group:ll 11 l!B D !System/Evolution Title 1 l_R_e_a_ct_o_r_P_ro_te_c_ti_o_n_S"'"y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~l lo12 IKA Statement: I Knowledge of bus power supplies to the following: RPS channels, components. and interconnections I
!
[Expl<1ryauon of 55.41.b(?) B is incorrect because Train A 45 VDC power comes from A and D vital power. C is incorrect because CS bistables are
Answers: 1 energize to actuate. D is incorrec because there is no disagreement since none of the slave relays have energized with the other train ot having energized.
Reference Title'. * . 1.1* Facility Reference Number
i!Reference Section******* .
,. .. . . .* .. Jf Page No.j IRevisionj IOverhead Annunciators Window A "S2.0P-AR.ZZ-0001 I 11 I 156 I ISolid State Reactor Prat Train A AC Power Distr 11 Drawing 240136 I 11 114 I I Ii i 11 11 I
!Lo. Number ** 1.*,. Objectives I RXPROTE011 I RXPROTE020 1_ ____. !Material Required for Examination J I Ij IQuesti?n Scnfrce:.>11 Facility Exam Bank 1IQue;;tion~.9~ificatl?nMethod: i Direct From Source I !Used During Trainirig Program Dl iQuesti~11;>oufce ComT~nts[ I
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Used on Salem June 2004 NRC exam fcOmment * **.. *.1 I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I
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Outline Changes I f Ouestion Topic ! IRO 43 I Given the following conditions:
- Unit 2 is operating at 100% power when a turbine trip event and subsequent switchyard disturbance result in an undervoltage condition (<70%) on the 2A & 2C 4160 V vital buses. - 2 minutes later, a Safety Injection signal is generated.
Which of the following describes the plant response when the SI occurs? I
~ EDG output breakers remain shut and safeguards loads sequence on each vital bus.
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~ EDG output breakers remain shut, blackout loads are stripped, then safeguards loads are sequenced on each vital bus.
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~ EDG output breakers open, blackout loads are stripped, EDG output breaker shuts, then only safeguards loads are sequenced on each vital bus. I Id. IIEDG output breakers open. blackout loads are stripped, EDG output breaker shuts, then safeguards and blackout loads are sequenced on ~ each vital bus. I
!Answer i c I I tExam Level 11 R I ;Cognitive Level *. I IApplication I iFacility: 1ISalem 1 & 2 I 'Exa.mDate:.1 j 12/21/20151
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~j013000K112 llK1.12 liROValue:ILlJ}!SROValue::Glsect1on:!j~fROGroup:jLJ:SROGroup:,LJ '1!1~ D !system/Evolution Title I I Engineered Safety Features Actuation System I i013 I IKA Statement: I Knowledge of the physical connections and/or cause-effect relationships between Engineered Safety Features Actuation System and the following: ED/G Explanation of I 55.41.b(?) 2/3 4KV vital bus UV causes ALL 3 4KV vital buses to load in SEC MODE II BLACKOUT. This mode starts ALL EDGs Answers:
J and sequences BLACKOUT loads onto ALL vital buses. When the SI occurs, the SEC initiates a MODE Ill, which opens any running EDG breaker, strips whatever loads are energized, then sequences SAFEGUARDS loads onto ALL buses. Distracters are plausible based on determining the 2 buses 2A and 2C load individually in BLACKOUT based on the UV on those 2 buses only.
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Reference Title I .* . Facility Ref~rence Numbef i !Reference Section I1Page No:l IRevisiOnl ISafeguards Equipment Controller Lesson Plan 11 NOS05SECOOO II 1119 l I6 I
!I u II 11 I II n ii !I I
[L~O. Number Objectives IESFOOOE005 I SECOOOE010 1--~ IMater,iaL~equired for ~iirni~.a~ion
1 I Ij t9uestlon Source: j IFacility Exam Bank I 'I
[<luestion N,l~~ification Method: . Concept Used I !:us~d During Training Program I D [aui:;stion source Comtp~ll.ts!
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Outline Changes I iauestion Topic ! I RO 44 I A loss of off-site power resulted in a reactor trip. The operating crew is attempting to confirm natural circulation flow but neither SPDS NOR the Plant Computer is operating. You are the 3rd NCO and have been assigned to monitor and log CET's at the CET Control Panel. Which of the following describes the "ALL" Mode at the Train A GET Control Panel? It displays ......
~I Train A CET temperatures in sequential order.
I [bl ITrain A CET temperatures from lowest to highest reading. I
~I any Train A CET >700°F, then remainder of Train A CETs from lowest to highest.
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@] Ithe two highest reading Train A CET's in each quadrant then sequentially display all Train A CETs.
I [Answer l Id I tExam Level 11 R I !cognitive Level I !Memory I IFacllity: / ISalem 1 & 2 I IExamDate: J j 12/21/2015j ~I 017000K402 r--:-i IiK4.02 I!RO Value:: [DJ !SRO Value: IG]" :Section: ll~ fRO Group:; i LI 'SRO Group:I LI w; iii~ D
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system/Evolution Title! !in-Core Temperature Monitor System II~0_17_ _

KA Statement: I Knowledoe of In-Core Temperature Monitor Svstem desion feature(s and or interlock(s) which provide for the following:
Sensinq and determination of location core hot spots i \Expl;:mation of 55.41.b(?) Table C of CFST-1 states that in ALL Mode the display will progress through the first and second highest CETS in each Answers: *' quadrant, then sequentially display all cETs assigned to that channel. A is incorrect because does not display in sequential order without first displaying the 2 highest in each quadrant. B is incorrect because it doesn't display lowest to highest. C is incorrect because the 700°F noted in choice is criteria for purple path CFST for Core Cooling. l .. Reference Title .
\' Facility Reference Number *. ! \Reference Section. *** 11 Page No.1 IRevisionl f Critical Safety Function Status Trees 112-EOP-CFST-1 I q 15 1125 I I II I 11 11 I I II I 11 11 I li.:~o. Number**
Objective~ I INCOREE016 1_ __ , IMaterial Required for EJ<ainination ' I I II (Question S~iJri::e: 11 Facility Exam Bank I I[ouestion Modifi~ation Method:.: Editorially Modified I:lised During Training Pfo9ram j D
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3 - 1 6 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' '
lc<>inment
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RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes iouestioh Topic I IRO 45 Given the following conditions:
- Unit 2 is operating at 100% power. - A Main Steam line break inside containment occurs. - The Rx was tripped, but the MSLI failed to shut ANY MSIV. - Safety injection was initiated successfully. - Containment pressure has risen steadily and is currently 11 psig. - RCS pressure is 1700 psig and lowering slowly. - Safeguards reset actions have just been completed in 2-EOP-LOSC-2, Multiple Steam Generator Depressurization.
Which of the following describes how the Containment Cooling system will respond if containment pressure were to rise to >15 psig? I
~ Both Containment Spray pumps will start and Containment Spray valves will reposition. 'b.l INeither Containment Spray pump will start, and Containment Spray valves will reposition.
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~ Both Containment Spray pumps will start, and Containment Spray valves will NOT resposition.
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~ Neither Containment Spray pump will start, and Containment Spray valves will NOT resposition.
[Answer i IE__J !Exam Level: I~ 1cogmt1ve Level.
.. 11 Application 'llW' I lfac1hty:n ...
Salem 1 & 2 llExamDate: Il___12_12_1_12_0_1__,5i ~I 022000A301 I[A3.01 ; !RO Valµe; 1 [ITI 'sRO Value: I@[section: 11~ jRO Group:! LJ !SRO Group:I LJ D \system/Evolution Title I IContainment Cooling System ;022 iKA Statement: I Ability to monitor automatic operations of the Containment Cooling System including: Initiation of safeguards mode of operation fExplan~ti?n ofl 55.41.b(9,7) Containment Spray Pump Sequencing Answers: 1)If the SSPS Containment HI-HI Pressure signal is not present when the SEC initially tries to start the Spray Pumps, the SEC contact will re-open 2)0nce the SEC has completed the last step of its loading sequence, the CS Pump start contact is re-closed
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r"C' 0 ........... ,... .... n1 ...... .J..... - . ..., ...... Ib)lf the SEC has been reset the CS Pumps will NOT respond to a HI-HI Containment Pressure until the SEC is again actuated H f::".acility Reference Number* *** I!Reference Section
I[PageNo.\ (Revision[
I i . Reference Title ISafeguards Equipment Controller Lesson Plan 1INOS05SEC000-06 I 1117 11 6 I ! !I I II lI I I II I 11 11 I il.O. Number > * **.
i Objectives I CSP RAYE009
'--~ l~aterial Require~ for,Exami.gation). J I II [ouestior't sollrce,: . f l_N_e_w_ _ _ _ _ _ _..I f otiestif>n Modifit:~tiori ~ethp9: >] _________.! iusei:I Dµririg Training Prograrn I o iollesticmSource co~nierit~! I I
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I RO System/Evolution List I SRO System/Evolution List I Outline Changes I tOuesticm Topic I lRO 46 I Given the following conditions:
- Operators are responding to a LBLOCA IAW 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation. - RWST level is 8.9' and lowering as expected.
Which of the the following describes the effect if 21CS2 Containment Spray pump discharge valve experienced a short and motored closed? Containment Spray Header flow will. ..
~I lower to O gpm.
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~I lower but remain> 0 gpm.
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~ be unaffected since 21 CS36 RHR CS STOP VALVE is open supplying all spray flow I
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~I be unaffected since 22CS36 RHR CS STOP VALVE is open supplying all spray flow.
I !Answer j a I I !.Exam Level ; IR I !Cognitive Level J IApplication I IF:acility: 11 Salem 1 & 2 I !ExamDate: 11 12/21/20151 [KA:l I026000K302 I!K3.02 i IRb Valu~: 114.2*1 :SRO Value::! 4.31 lse,ction: II~ IRO Grc>up:ij 11 !sROGrolli:>:ll 11 !lf~ D lsysteffi!Evolution.Title I j_c_o_n_ta_in_m_e_n_t_S..:..p_ra"""y_S-'y'-s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __,11026 [KA Statement: I Knowiedoe of the effect that a loss or malfunction of the Containment Sprav Svstem will have on the followino: Recirculation sprav svstem !Explanation of I 55.41.b(S,10) When RWST level reaches 15.2', operators will beging alignment to Cold Leg Recirc IAW LOCA-3. 22 Containment Answers: : Spray pump is stopped first if both Containment Spray pumps are operating. With the stem condition of current RWST level, 21 CS pump will still be running. When RSWT level reaches lo-lo setpoint, the remaining CS pump (21) will be stopped, and 21CS36 opened to supply recirculation spray flow. I I Reference Title ll Facility Reference Number. I[Reference Section .* I!Page No.j jReyisiC>ni I Transfer to Cold Leg Recirc 112-EOP-LOCA-3 I q 1129 I I II I 11 11 I I 11 I ii 11 I Objectives ILCA3U1E006 ,_ ____, jMatetial Required for Examination
I I 11 f(lliestion Source: 1j_N_e_w_ _ _ _ _ ___.l l<ltte,stiop.1V19pificatiC>niVtethod: .*** ~ _ _ _ _ _ _ _ __.I fused During TrafriingPrograrn.i D l(;lue,s.tion sl)urcfComments 11 I
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RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution Li~ Outline Changes iauestidnJopic 11RO47 I Given the following conditions:
- Salem Unit 1 was operating at 100% power when a LOCA occurred. - A manual reactor trip and manual SI were initiated. - When the Main Generator output breakers opened, a loss of off-site power occurred. - 1A vital bus locked out on bus differential.
Which of the followinq identifies which Containment Iodine Removal Units (IRUs) can be started if required? 1.:11
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I f!U 112 IRU ONLY. I
~ 111 or121RUs.
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.d.11 NEITHER IRU is available.
I !Answer i Id I !Ex,am t:evel 11 R I !cognitive Level 11 Memory I IFaciHtY: 11 Salem 1 & 2 1 IExa!TIDate: 1 I 12/21/20151
!R.O Value: I[Dj' ,SRO Value: III£]' !section: II~ /RO Group:j LJ [SR,O Group:j LJ <
~I 0270001<201 I'K2.01 i ~ .' D ISYstem/Evoiution Title I I Containment Iodine Removal System ! 1,027 I IKAStafoment: J Knowledqe of bus power supplies to the followinq: Fans I IExplanatio~of 1 55.41.b(9) Containment IRUs are powered from G and E non-vital 460VAC. With the loss of off-site power, none of the non vital Answers: *.* 1 busses are energized. The distracters are based on the operator knowing that the loading of a vital bus in Mode IV doesn't have any bearing on !RU operation. ! \ ;"/ .**.. Reference Title 11: ** FacilityReference Number
I!Reference Section. 11 Page No. j !Revision\
!1E1 Aux Building 460-230V One line 11207916 I 11 1126 I !1E1 Aux Building 460-230V One line !1207919 I q q23 I I II I II 11 I !LJ>. Number* Objectives I CONTMTE003 ,_____, !Material Required for Examination
J I toues~i~n Source: 11Previous2 NRC Exams I 1o~estio~Modif!?atioriJV1ethod; \1 Direct From Source I !used During Traii:iing Program l D
!Question Sourc~;Somm~n~sj 1_1_1-_0_1_R_o_N_R_c_a_4_7_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ [comment.* II
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SRO Skyscraper I RO System/Evolution List f SRO System/Evolution List I Outline Changes 1ouestionTopicJ ,_R_0_4_8 ~---------------------------------------------~! _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~ Given the following conditions:
- Unit 2 has experienced a Large Break Loss of Coolant Accident (LOCA). - Containment H2 concentration has risen to 2%. - 21 H2 Recombiner has been placed in service with containment pressure at 4.1 psig. - 24 hours later, containment H2 concentration has remained at 2%, and containment pressure has risen to 5.1 psig.
Which of the following describes the effect the higher containment pressure has on H2 Recombiner operation, and how should the crew proceed? H2 Recombiners are _ _ effective as containment pressure rises with a constant power setting. The power setting must be ___ IAW S2.0P-SO.CAN-0001, H drogen Recombiner Operation. ----------------------------- la'.J less, raised.
~ more, raised. ~ less, lowered.
fd.: more, lowered L:...___J !Answer I r&:J rExam evel i rL.J' !cognitive Level I! Application 1 IF~cility: HSalem 1 & 2 l~ExalllDate:H _ _ _12_1_2_11_20_1_5~I IKA:!j 028000A201 !IA2.01 ilROValue:i[ill[sROValue:l§isection: ii~ IRO (3roup:1LJISRO Group:ILJ D [System/Evolution Title I IHydrogen Recombiner and Purge Control System [KA Statement: I Ability to (a) predict the impacts of the following on the Hydrogen Recombiner and Purge Control System and (b) based on those predictions, use procedures to correct, control, or miti ate the conse uences of those abnormal o eration: fExplanation of j 55.41.b(?,8, 10) H2 re~ombiners are P.laced i~ service when directed in the EOPs with containme~t H2 concentration between 2-~%. 1 AnsWers:
_- ** : The hydrogen recomb1ners use electric heating elements to elevate the temperature of the containment atmosphere. As shown 1n Attachment 2, RECOMBINER POWER CORRECTION FACTOR CURVE, a higher containment pressure would result in a higher power correction factor, which would cause recombiner power setpoint to rise. So initially the higher pressure would cause the
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I Reference Title Ii ' Facility Reference Number> i !Reference s~ction JI Page No: I iReylsion' IHydrogen Recombiner Operation II S2.0P-SO.CAN-0001 I 114,7,8 1I9 I I II I 1I I I II I "II 11 I !LO.Number Objectives I CONTMTE012 ,_ ____, !MaterialR'.~qufrea fc)r¢can}in'ation , l IRO 48 S2.0P-SO.CAN-0001 (because KJA says so!) II !Oue~tion Source: 11 Facility Exam Bank I Jl.ised outing "Tr'.a'inin9 Pr,ogram l D [a~_estion Sourc,e Comments! 180552 I
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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question Topic! IRO 49 Given the following conditions:
- Salem Unit 2 is 7 days into a refueling outage. - The core is partially offloaded with 7 bundles remaining in the Rx. - Fuel movement is in progress, and S2.0P-IO.ZZ-0010 SPENT FUEL POOL MANIPULATIONS is in effect. - SFP temperature is 120 deg. F. - 21 SFP becomes air bound, trips on motor OL, and can NOT be restarted. - 22 SFP pump will NOT start. - SFP hi level alarm is in alarm. - SFP heatup rate is 12 deg F/ hr.
If SFP cooling can NOT be restored, which of the following choices describes an adverse consequence of this Joss of Spent Fuel Pool cooling?
§:] IRx cavity overflow due to rising SFP level if the Gate Valve remains open.
[]I Increased production of radioactive waste liquid as the tell-tale drains flow rises to the FHB sump. I
~ Increased radiation levels at the FHB charcoal filter due to Spent Fuel off-gassing at temps> 150 deg. F.
I [d. ! Inability to place a raised Spent Fuel bundle into any location in the pool due to rising radiation level on 2R32 Fuel Handling Crane Area L...:..J Monitor. B I [Answer: I~ :Exam Level !~ 1cognit1ve Level 11 Memory .. r ISalem 1 & 2 I !Facll1ty: I IExamDate: ! ,___12_12_1_12_0_1_..51 ~I 033000K303 ! \K3.03 \ IRO Value: i 12QJ:sRO Value: !@~section: q~ !Ro Group:\ [JJ\sRO Group:! LJ ~~11'. D fsystem/Evolution Title I I Spent Fuel Pool Cooling System I !033 I IKA Statement: I Knowiedge of the effect that a loss or malfunction of the Spent Fuel Pool Cooling System will have on the following: Spent fuel temperature I !Explanation of I 55.41.b(13, 10) Distractor b is incorrect because the pool would overflow into the ventilation system, not come over the physical wall [Answers: of the pool. C is correct because rising radiation will be seen as fuel off-gassing and is expected to occur as temp increase to 150 deg. Distractor a is incorrect because any overflow will go out the ventilation openings in the SFP. Distractor d is incorrect because it is always possible to lower a SF bundle. I
Reference Title 1: Facility Reference Number I!Reference Section Ii Page No: I !Revision:
ILoss of Spent Fuel Cooling !Is2.0P-AB.SF-0001 II 11 1112 I II II II 11 I I! !I !I 11 I fL~O. Number Objectives !SFPOOOE007 '--~ IMc;iterial Reqllired for Exa~i!l,ation Ii 11
!Question Source: 11 Facility Exam Bank l IQuest\~n Modifiqatio,n Method:. ;::I Editorially Modified I iUsed Puring Training Program .1 D !Question Source Comme~ts[ I Used on Salem June 2004 NRG RO exam (6 exams ago)
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1aues~ion Topic] I RO 50 I Given the following conditions:
- Unit 1 is operating at 95% power performing a 1% per minute load reduction due to high radio frequency in the Main Generator. - 11 SG Narrow Range Level channel II is undergoing a channel calibration IAW S1.IC-CC.RCP-0035, 1LT519#11 STEAM GENERATOR LEVEL PROTECTION CHANNEL II, and all associated bistables are tripped.
Which of the following identifies the consequence if a second channel of Narrow Range level on 11 SG were to fail to 30% with NO operator action? 11 SG level will become ... la. c__j I I higher than program because 11BF19 and 11 BF40 ONLY swapped to manual and will be over feeding 11 SG. I I
,b. i lower than program because 11BF19 and 11 BF40 ONLY swapped to manual and will be under feeding 11 SG.
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'c.
L_j I l higher than program because both SGFPs and 11BF19 and 11 BF40 swapped to manual and will be over feeding 11 SG. I I [d. ! lower than program because both SGFPs and 11BF19 and 11BF40 swapped to manual and will be under feeding 11 SG. I
Answer i I a I !Exam Level 11 R I !cognitive Level 11 Application I wacillty: i ISalem 1 & 2 I !Examoate:'! I 12/21/20151
\KA: I035000K603 1 i IK6.03 jROValue:il 2.6jlSROValue:jj 3.oj[sectiol'l:lj~tROGroup:il 2psROGroup:lj 2j L~~~ D jsystern/Evolution Title j j_s_te_a_m_G_e_ne_r_a_to_r_S..;.y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.I ~I fKA Statement: I Knowiedge of the of the effect of a loss or malfunction on the followinq will have on the Steam Generator Svstem: SIG level detector I !Explanation of I 55.41.b(?) SG NR level is programmed from 33-44% up to 100% power. As the downpower continues, the BF19 will be closing in [Answers: I response to the lower steam flow requiring less feed. The 11BF19 and 11 BF40 (expected to be shut at this power level) swap to manual upon the second NR level channel failure. The SGFPs do not. When the 11BF19 swaps to manual, it will have a certain demand on it. As power (and steam flow) continues to lower, the demand will be higher than required, and SG NR level will l... ............ """'o. i....~ ..... i....,... .. .i.a... ......... n'"'"'"1r"-:arn 1,..,,,,....1
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I Reference title * . . . I . Facility Reference Number ItReference Section . I* Page No. j [Revision! IOverhead Annunciators Window G !IS1 .OP-AR.ZZ-0007 I 1111 1142 I I I I 11 11 I I 11 I 11 11 I [1.,.:0*. Number< I CN&FDWE004 Objectives 1_ ___. fMaterial Required for Exarninatiort' II II 1auestion source:> 1l_N_e_w_ _ _ _ _ _~I '-i.a~u~e~s_,_ti__:occ.n*__:M~o~d:.c:..if_ic...ia~t~io:::.n:::.**M-'--.-..::et-..::}l__:o_d__c:-..::->Jjl1_________.l lus~C1'6uring :i:~aining Progsarii Io iauestion Source <:;or.rtmentsi I
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Outline Changes I .. jauestionTopic I I RO 51 I Salem Unit 2 is operating at 95% when the PO reports steam flow is rising on all SGs with no readily apparent reason. Which of the following describes how the steam flow will affect the plant, and how operators should proceed if steam flow continues to rise uncontrollably IAW S2.0P-AB.STM-0001, Excessive Steam Flow? Reactor power will rise at 1) rate as the increased steam flow. The crew will 2)
~ 11) a higher
2) initiate a MSLI to determine if a safety injection is required. I
[gJ 11) the same
2) Trip the reactor and confirm the trip, then initiate a MSLI to determine if a Safety Injection is required. I fc.! 11) a higher
~ 2) Initiate a power reduction to ensure reactor power remains <100% while attempting to identify and isolate the leak. I fci.l 11) the same t-=.:.J 2) Initiate a power reduction to ensure reactor power remains <100% while attempting to identify and isolate the leak. I jAnsW'er 11 b I iExam Level ! IR I !Cognitive L._evel ** i I Comprehension I [Fa~Uity; 11Salem1 & 2 I~ I 1212112015!
jKA:il 039000A205 jjA2.05 I!Ro Value: II 3.31 lsRo Value:ll 3.61 [section: II~ )Ro Group:ll 11 jsRo Group:;! 1j Ill D !system/Evolution Title I j Main and Reheat Steam System I row=-i IKA Statement: l Ability to (a) predict the impacts of the following on the Main and Reheat Steam System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: lncreasinq steam demand, its relationship to increases in reactor oower fExpl<imat'.On>~! 1 55.41.b(1,5, 10) AB.STM does direct a power reduction at Step 3.8 if the EHC system is not causing the turbine to be the source of 1Answers. *' I the rising steam flow. The stem states that there is no readily apparent reason for the rising steam flow, and EHC malfunction would be apparent. Also, the Continuous Action Summary is in effect at Step 1. CAS Step 1.1 states that at any time if reactor power is rising uncontrollable, trip, confirm, initiate MSLI. If souce of steam leak is isolated, then go to TRIP-1. If not, initiate SI and
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tReference
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, i * ** j Page No, I [Revisloni IExcessive Steam Flow ii S2.0P-AB.STM-0001 I 11 11 9 I I II I 11 11 I I II I II 1I I IL.o:fllutnber< . I Objectiv I MSTEAME008 I IABSTM1E001 I
, __ ___. IMaterial ~equired fofExamfriation II II jOuestio.~s~urce: ..*J l_N_e_w_ _ _ _ _ _~I Lia_u_e_s_:ti_:o21}.~M-'o_d~If_:ic_:aLt~io~n_M_** _etc::*~L<>L9L:*_:.* .'-'i,1 ________~1 luse<(buring Trainjng Program .j O
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- Unit 2 has experienced a loss of all feedwater flow initiated by a Feedwater Isolation signal (P-14) condition on 23 SG. - To mitigate the event after the P-14 has cleared an NCO has been directed to start 21 SGFP in accordance with S2.0P-SO.CN-0007, Rapid SGFP Recovery. - The NCO successfully relatches the 21 SGFP but the speed of the pump does not rise automatically to minimum speed as he anticipated.
Which of the following is the cause of this response?
~ IThe P-14 signal "seal-in" feature.
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~ 121 SGFP PUMP SPEED CONTROL is in AUTO. ~ 121 SGFP ENABLE/DISABLE switch is in the DISABLE position.
I I [ ] 121 SGFP speed was >160 rpm when the latch push button was depressed. I [Answer J Id ! iExam Level 11 R I !cognitive Level 11 Memory ! IFacility:*l ISalem 1 & 2 1 IExanioate:' 11 12/21/20151 !KA: iI059000A107 !IA1.07 llRo Value:ll2.5:J !s~o vciiue::j2.6*l lsectfon: II~ !RO Group:!! 11 lsRq.Group:!I 11 falil D j.Systeni/Evolution Title l j_M_a_in_F_ee_d_w_a_te_r_S-'y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.l 1059 !KA Statement:! Ability to predict and/or monitor chanqes in parameters associated with operatinq the Main Feedwater System controls including: Feed Pump speed, includinq normal control speed for ICS !Explanation of I 55.41.b(4,10). A is incorrect because the P-14 signal automatically clears when the SG level lowers <setpoint, there is no seal in ,Answers: ..*.*. ** j circuit. Bis incorrect because auto speed control prevents latching of the SGFP. C is incorrect because the Enable/Disable switch in the disable position only removes the ADFWCS from controlling SGFP speed, the switch is placed in Disable when starting the SGFP. Dis correct because the SGFP may be latched with speed <160 rpm, but it will not automatcally raise speed to minimum 111nn - - - ;~1~ -- --'\
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ISGFP Prompt Recovery 11 S2.0P-SO.CN-0007 I ii 6 114 I I II I !I 11 I I II I ll 11 I !Lo. Number
I SGFPLOE006 Objectives I
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/Material Required forExamination)*i l lj jauesfion Sol.ire~: 11 Facility Exam Bank l ldu'estion Modific(itiqn>Metpod: .. Editorially Modified I I[used DuringT~ciining ~rograrri I D I -;::======-=========::::::::::::::::::::::::::::::::::::::~===========-================-~t rauestion .source Commehtsj ~4112. _Removed second part of question which asked what you had to do to get speed to raise. Replaced 1mplaus1ble distracter.
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I SRO System/Evolution List I Outline Changes J.... fQuestionTopic I IRO 53 I I With both 11 and 12 AFW pumps in service providing 6E4 lbm/hr flow to each SG, what would be the response if both 13AF21 and 14AF21 were shut! fully?
~ 111 AFW pump automatic recirc valve would open.
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~I The single automatic recirc valve for the MDAFW pumps would open.
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\c. ! IThe Pressure Override Circuit would actuate to prevent dead heading 11 AFW pump.
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@] I Normal MDAFW pump recirc flow would rise through the orificed continuous recirculation line.
I JAnswer ! ja I lexam Lev~rj j R I !cognitive Level .11 Application I lf=aci!jty:' 11 Salem 1 & 2 I !Exarnbate: 11 12/21/20151 y !KA:lj 061000K503 1IK5.03 I:RO Value: II 2.6j !SR,O Value: :12.9*1 ISection:*.!j~ lRO Group:ll 1I !SRO Group:lj 1 I l!E ' D I
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system/Evolutio*n Title. i Auxiliary I Emergency Feedwater System I i061 I

KA Statement: Knowled e of the o erational implications of the followin Feedwater S stem:
Pump head effects when control valve is shut iExplanation ofj 55.41.b(7,8) The MDAFW pumps each has its own dedicated recirc line and associated automatic recirc valve, which opens to
.Answers

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1 maintain aFW pump flow >180 gpm. With the 6E4 lbm/hr in stem for each of the 2 SGs being supplied from 11 AFW pump that =
240 gpm, the valve would initially be closed and open when flow lowers <180 to prevent the increased pump head from causing pump damage from overheating. B is incorrect but plausible if it is thought there is a common recirc line. C is incorrect because
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i. ,* **, '* *.Reference Title ,'. :h,' . Facility Reference Number IAuxiliary Feedwater System Lesson Plan II NOS05AFW000-14 I ! j 19-20 1 I I II I 11 1I I II I II 1I l lL.o; Number* Objectives I AFWOOOE008 1_ ____,
[llllaterial Requir~d for Examif1atror1;i
II 11 i<:llle~tion S9uice:.A l_N_e_w_ _ _ _ _ __,l [Q~~S,~ion "19clification~etlloCI:: J_________.I [used During Training Program ID fal1estioo,sourceCo.mrn~nts] ' - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -.....!
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!Question Topic j RO 54 I Given the following conditions: - Unit 2 is performing a normal shutdown at 20% per hour IAW the IOP's. - 22 SGFP is shutdown. - Current Rx power is 24%. - 21 SGFP trips.
With NO operator action, what will be the status of the Auxiliary Feed Pumps?
~I ONLY the MDAFW pumps start immediately upon the trip of 21 SGFP.
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~I ONLY the MDAFW pumps start when SG NR level in 1/4 S/G's lowers to 14%.
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~ The MDAFW pumps AND the TDAFW pump start immediately upon the trip of 21 SGFP.
I [] I The MDAFW pumps AND the TDAFW pump start when NR level in 1/4 S/G's lowers to 14%. I IAnsw~r 11 a I !exam Level 11 R I ;cognitive Level J IApplication j jFaciliW: 11 Salem 1 & 2 I lExamDate: 11
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Q]" !s~o Value: [Ej !Section: II~ [RO Group:! LJ !SRO Group:jLJ g
;w. ~I 061 OOOK602 " '~ , .~
11 K6.02 ' I ,Ro Value: I
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D [System/Evolution Title J IAuxiliary I Emergency Feedwater System j 1051 i IKA Statement: l KnowledQe of the of the effect of a loss or malfunction on the following will have on the Auxiliary I Emergency Feedwater System: Pumps fExplaf!~tion of j 55.41.b (4) MDAFW pumps auto start when both SGFPs are tripped as shoen on logic drawing 221064. The TDAFW pump does Answers: .*. not. The MDAFW pumps also auto start on 2/3 NR level channels in one SG lowers to 14%, but in this case they will already be running. The TDAFW pump starts on 2/3 NR level channels on 2/4 SGs.
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'i Reference Title 11 I
*F,acility Reference Number\ , /~eference Section*. J! P~ge No../ IRevision!
IRPS AFW pumps startup 11221064 I II lj a I IAFW System LP II NOS05AFW000-14 I lj 33 1114 I I II I ii II I
.L.o. Number.
Objectives AFWOOOE006 IMaterfalRequired td'r Examination. j I II
!9u~~tion sou~c~: 11 Facility Exam Bank 1JQue~tiof!Modification Metho.CI:\ I Editorially Modified I IUsed Du.ring Training PrograrilJ D
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Outline Changes I.... [Question Tbpic I IRO 55 I Given the following conditions:
- 2C Emergency Diesel Generator (EOG) is operating in parallel with the 500KV grid for a 24 hour endurance run IAW S2.0P-ST.DG-0014, 2C DIESEL GENERATOR ENDURANCE RUN, following a complete overhaul. - Cumulative run times for all individual EOG load limits are less than 10% of rated. - While operating at 2525 KW three hours into the test, the operator mistakenly adjusts 2C EOG speed control resulting in MW loading rising to 2610 KW.
What are the consequences, if any, of continued EOG operation at this KW load? l Operation for the remainino 21 hours of the test ...
~ will not have any adverse effect on 2C EOG.
I I
~ will result in exceeding the 30 minute load limitation for 2C EOG.
I rs I will result in exceeding the 2 hour load limitation for 2C EOG. I
@] Iwill result in exceeding the 24 hour load limitation for 2C EOG.
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.la! 'Exam Level .
I !RI !Coanitivel,.vel I IAnnlication I IFaci}ity: I Salem 1 & 2 IJ~~mOate.:*W 12/21/20151 !KA: Ij 062000A 101 I~jA_1._01_ _1JRO Value: II 3.41 Jsifo Value: 11 3.81 iSectiori: 11~ [RO Group:! I 11 JsRO Group:/ I 1I D [System/Evolution Title I l_A_.c_._E_l_ec_t_ric_a_l_D_is_tr_ib_u_ti_o_n_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.j J062 !KA. Statement: I Ability to predict and/or monitor chanc::ies in parameters associated with operatinq the A.C. Electrical Distribution controls includinq: I Significance of DIG load limits I I i~planationgt 55.41.b{8) The EOG load limitations are maximum of: 2600KW continuous, 2600-2750KW for 2000 hours, 2750-2860KW for 2 tAnswers: * * .. hours, and 2860-3100KW for 30 minutes. With the EOG operating at 2610KW for 21 hours, the EOG will not exceed any limits, operation between 2600 (cont) and 2750(2000 hours) KW, since the stem stated that the cumulative run time for ALL EOG load limits was <10%, which would be 210 hours for this limit. t Reference' Title *
11 *Facility Reference Number *** J[Reference ~ection

I 1Page No:'.1 IRevisioni j 2c DIESEL GENERATOR SURVEILLANCE TE s2.0P-ST.DG-0003 II II P&L 3.5 II l j52 I I !I II I[ 11 I I II II 11 11 I
[Lo. Nµrnber I EDGOOOE012 Objectives I 1_ ____. iMafr~rialRequired for Exalllinatlon;* '. j I Ij I [ouesti9:'.n Source:
j Facility Exam Bank 1lOue~tionModification Meth~d:' J Editorially Modified I JUsed DufingTl-aiiling Progrl:iin.j D
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fOuestio~ s(')IJrce Commenti'ii 163989 replaced 2000 hr distracter (implausible) with 24 hour distracter I fcommenf'
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lauestionTopic 11 RO 56 I IIAW Salem FSAR Section 8.3.3.2, Station Batteries, on a total loss of all AC power the station vital 125 voe batteries are designed to supply vital station loads for a minimum of hours. I nl
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[d.jl8 I [Answer 11 b I IExam Level J IR I !cognitive Level 11 Memory I l~acility: 11 Salem 1 & 2 I !ExamDate: 11 12/21/20151 !KA: 11 062000K303 IiK3.03 liROValue:lj 3.7jlsROValue:ll 3.9llsectioll:lj~i~OGroup:1I 11Js~0Gr,oup:il 1j ~lilJ D !System/Evolution Title] j_A_.c_._E_l_ec_t_ric_a_l_D_is_tr_ib_u_ti_o_n_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.l ID62 !KA Statement: I Knowledge of the effect that a loss or malfunction of the AC. Electrical Distribution 'Nill have on the follo'Nin~i: DC svstem IExplanatio~ pf'. 55.41.b(?,8) IAW Salem FSAR Section 8.3.3.2, Station Batteries, states ... "The batteries are sized for 2 hours of operation after a Answers: . * . ; ! loss of ac power, based upon the required operation of the de emergency equipment." 1 i .Y **
Reference Title I'\ .*.**Facility Reference Number**** *. !!Reference Section 11 Page No. I IRevisionl
!Salem FSAR I 118.3.2.2 118.3-18 1114 I I I Ii 11 11 I I l II ii !I I I.Co; Nurrib~r. *...*. Objectives I DCELECE002 '--~ llVlaterial Required for Examination *~* I I 11 fou~~ti~ll Source: 11 Facility Exam Bank I (ques,ti?n Mpdifi~~tion Meth4ct: <iJ Direct From Source I1.Used During Training Pr,ograriij D fauestfon ~6}-lrce Solll!llerts) L091 o I;:::::====:::..'.::::::'.::::=:'.::::=:'.::::=:'.::::=:::::::'.::::=:::::='.===========-==============-=~1 I
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IC2oesUonTopic ! IRO 57 Given the following conditions:
- The control room receives Auxiliary Typewriter alarm point 0676, 2G 4KV GROUP BUS LOSS 125 voe CONTROL PWR. - An operator investigates and reports 125 VDC breaker 2BDC1AX12, 2G 4KV Bus Control Power Supply (Reg) tripped.
Which of the following identifies the effect, if any, this will have on 24 RCP 4KV breaker? If 24 RCP is running it will. .. [] I trip immediately. I lb.JI continue to run but will not trip if required. I
~ trip if a RPS trip signal is subsequently developed, but would not be able to be re-started if directed.
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@] I continue to run and be unaffected as emergency control power from the alternate control power supply will automatically be provided.
I 1Answerj b I I !Ex<lm Level 11 R l [Cogniti.Je Level 11 Application I !Facility: i ISalem 1 & 2 1 IExamDate: 11 12/21/20151 IKA:il 063000K401 IIK4.01 1lr{ci Value:JI 2.7j ISROXalue:Jl 3.0*l lsection: H~ IRO Group:11 I 1 lsROGroup:il 11 mi D iSystem/EyollltionJitle i j_o_.c_.E_l_e_ctr_i_ca_l_D_is_t_rib_u_t_io_n_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-'I :o63
KA Statement: i Knowiedae of D.C. Electrical Distribution design feature(s) and or interlock(s) which provide for the following:
Manual/automatic transfers of control !Explanation of 55.41.b(S) 4KV Group bus control power is supplied from the DC electrical system. The alternate DC supply does not automatically 'Answers: transfer to supply DC control power to the bus, it must be manually transferred when the normal supply is lost. 4 KV breakers cannot be tripped remotely without 125VDC available to energize trip coil. A is incorrect because there is no power to energize the trip coil, plus no trip signal would be present. Bis correct. C is incorrect because it would not trip if required (different from A is
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m~~**~"' I I ** > Reference Title Ii
Facility Reference Number * " j [~efer~nce Section 11 Page NoJ [Revisi011[
~- 24 Reactor Coolant Pump ii 211538-2 I 11 1122 I II .
ii 11 I II : ii 11 I I DCELECE007 1_ ____. lnni:lterial Required {gr i;iallli£1a~ion .. J I 11 IQuestion so'Urce: :j IFacility Exam Bank 11au~stiOI) Modi!iCJ!ti~n Method: ~11 Editorially Modified 1lused DuringTrainin$f Ptog~am lD 1a~e,~tioll sollrce.pomm~r,t~~I
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iOuestion Topic I IRO 58 I Given the following conditions:
- 21 Diesel Fuel Oil Storage Tank indicated level is 96.0". - 22 Diesel Fuel Oil Storage Tank indicated level is 96.1". - 21 Diesel Fuel Oil Transfer Pump is selected to REGULAR. - 22 Diesel Fuel Oil Transfer Pump is selected to BACKUP. - The Diesel Fuel Oil Storage Tank area C02 Tank fails, and a projectile opens a hole in the bottom of 21 DFOST which completely empties.
Assuming no other damage from the C02 Tank failure has occurred, which of the following identifies how makeup flow to the EOG Day Tanks will be provided, if at all, if the 2C EOG starts automatically to provide power to 2C Vital bus on a single bus UV condition? Makeup flow to the EOG Day tanks will be provided by ......
~I BOTH DFO Transfer pumps because DFOSTs are normally cross connected but check valves will prevent 22 DFOST from draining out the rupture. I
[g] 121 DFO Transfer pump ONLY because the REGULAR pump is aligned to the highest storage tank level during normal surveillance testing. I l.<::J INEITHER DFO Transfer pump because DFOSTs are normally cross connected and both will be empty. I
@] 122 DFO Transfer pump ONLY because 21 DFO Transfer pump is aligned to an empty tank.
I !Answer] I~ !Exam .Level 11~ 1cognitiye Level .*J IAppl1cat1on I !Facility: i jSalem 1 & 2 1iE:!CamDate:i j___12_12_1_12_0_1_,5I ~I 064000K608 I!K6.08 1 !RO Value: <@is Ro Value:l@!secti9~::J I~ [RO Group:! LJ !SRO qrol:if>:I LJ B; D !system/Evolution Title j IEmergency Diesel Generators 11064 !KA statement: l Knowledqe of the of the effect of a loss or malfunction on the followinq will have on the Emerqencv Diesel Generators: Fuel oil storaqe tanks IExplatiatfon cit: 55.41.b(8) DFOSTs are normal isolated from each other on the outlet side by the closed 2DF35, 21/22 DFO STOR TANK X-CONN
)
!Answers: VALVE. Each tank is supplied by its respective transfer pump. Return (overflow from DFO Day tanks) is directed to the tank which has its DFO transfer pump selected to lead, so that overflow won't be directed to the storage tank from which suction is not being taken. With an empty 21 DFOST, 21 transfer pump will still receive a start signal (at 33"), but has no fuel to pump. As Day Tank I-**-' ---"-* 'o~ tn '-* **-* ')') 11:1--1 * .,n\ n * - - *'" -*~..t I~+ ')7"\ ~n.-1 -*-**'.-lo"-* <--- ')') -*-*--- *--
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Reference Title*

II Facility Reference NI.Imber '>I [Reference s~ction

Ij Page No: i !Revisionl IFuel Oil ll 205249-3 I 11 1130 I IEOG Lesson Plan 11 NOS05EDG000-11 l 1144 1111 I
! !I I II II I jLo. Number. Objectives I FUEOILE004 j_ ___. iMatericil ~~qufredfor EXalliination *** *1 I [a~~stion Source:. J j_N_e_w_ _ _ _ _ _ __,l l91Jestion ~odi~cati,cin. ~~~h9~: **** J________~l Jlised During}~aJnif!Q f>'rogram I D fQuestionSou?c~ Comr!ie~~sl I I
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. . \auestion Topic J I RO 59 An explosive mixture is prevented from being present in the Waste Gas Holdup Sysytem JAW Salem Tech Spec LCO 3.11.2.5, Explosive Gas Mixture by limiting the 1) concentration to 2) or less. §] 11 ) oxygen 2)2% ~ 11) oxygen 2)4% ~ 11) hydrogen 2)2%
I [ ] 11) hydrogen !Answer 11 a I !Exanflevel i IR I !cognitive Level
JIMemory I !Facility{j ISalem 1 & 2 1 !ExamD~~e: 11 2)4%
I 12/21/20151 i IKA: 071000K504 liK5.04 ilR0Value;!j 2.5!tSROValue:!j 3.1llsection:Jl~[ROGroup:Jj 21!sROGr6up:ll 2j lllJ D /syst~m/Evolution Title I IWaste Gas Disposal System I [071 [KA Statement:! Knowledge of the operational implications of the following concepts as they apply to the Waste Gas Disposal System: Relationship of hvdrooen/oxvoen concentrations to flammability I [E)(plan~tiC>n ofi 55.41.b(13) Salem TS LCO 3.11.2.5 staes that oxygen concetration in Waste Gas Holdup System shall be maintained less than or
.Answers:
i equal to 2% . 4% is not correct all of the time. Hydrogen concentration is monitored but not address in Tech Specs I*
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Facility Reference Number **l\Reference Section .* 1\Page No:\ \Revi~ioril j Salem Tech Specs I lj 3.11.2.5 113/411-1511282 I I I II II II I I I II -
ii 11 -- I - Objectives I WASGASE009 ,____. !Material_ Requfred ft;>r EJ(all'linatioh.
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[tfoe~tion squrct,J j_N_e_w_ _ _ _ _ __.I !8tjesti~R'!'Aodiycation ~~thod: 1_________,l [usedD~rirtgJrainirigProgr,am I D 1a.uestio~ so'Urc~ Com~(!ntSI I I
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!coiilment I I I I
RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes I ************** *-* ......... . iouestiori Topic I I RO 60 I Given the following conditions:
- Unit 1 is operating at 100% power. - Control room operators are preparing to perform a Containment Pressure Relief IAW S1 .OP-SO.CBV-0002, CONTAINMENT PRESSURE-VACUUM RELIEF SYSTEM OPERATION. - Containment radiation levels are NORMAL for 100% power operation with no failed fuel.
After opening the 1VC5 and 1VC6 CONT PRESS/VAC RELIEF ISOL valves to initiate the pressure relief, which choice describes how the respective radiation monitors indication will respond?
1R12A - Containment Gas Effluent

1R41 B - Plant Vent Noble Gas Intermediate Range

1R41 D - Plant Vent Noble Gas Release Rate
~ 11 R12A rises; 1R41 B rises; 1R41 D rises.
I [b'.111R12A rises; 1R41 B constant; 1R41 D constant. I
\cl I 1R12A constant; 1R41B constant; 1R41D rises.
I
@;] I1R 12A constant; 1R41 B rises; 1R41 D constant. I I I .. . ...
Answer 'I~ !Exam
! Level.'~ I I 1Cogmt1ve Level .. 11 Application J !facility. i ISalem 1 & 2 1 jExamDate.1 l 12/21/20151 ~lo73000A101 j1A1.01 !1Rova1ue:1~JsRova1ue:IO]"l,sectii:m:ll~lRbGroup:\LJ1s1mGroup::1 LJ lz~! iSystem/EvolutionTitle j IProcess Radiation Monitoring System I -:0-73~~ [KA Statement: I Ability to predict and/or monitor changes in parameters associated with operating the Process Radiation Monitoring System controls includin : Radiation levels [Explanation of I 55.41.b(11 )1R12A is sampling containment atmosphere, so it will NOT rise when the pressure relief is started. 1R41 B is an !Answers: * : intermediate range monitor that normally does not have sample flow through it. It's sample flow will start when the lo range 1R41 A
~ monitor nears its high end of monitoring range. It's indication will not change during a pressure relief with NORMAL containment radiation levels. The R41D provides the gaseous effluent release rate (uCi/sec) by combining (product of) the on-range R41A . . . .
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I[Reference section*** \~'I li:>a9E! Nq.] iRevisionl IContainment Ventilation System Operation iIS1 .OP-SO.CBV-0001 I ll 1 j2s I IAbnormal Radiation 11 S1 .OP-AB.RAD-0001 I 1132 I I
IL(). Number. / Objectives II I
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lj l<:l.l!'.~~tion,.$~ur£t j IFacility Exam Bank I[9iJesti~!JM§~if!ca1,%1 Mel~(,d: ?' IDirect From Source i9u7,~tii?:n.~6urcii*c&faiJ'\en!sJ 150514, used on June 2004 Salem NRC RO exam (7 exams ago) I
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!Question Topic I RO 61 I I Given the following conditions: - Unit 1 is operating at 25% power. - 1B EDG is running in parallel with station power on 1B 4KV Vital Bus. - 13 and 16 SW pumps are in service, 11 SW pump is in AUTO. - 1A 4KV Vital bus becomes deenergized due to a Bus Differential signal.
1 minute after the 1A 4KV Vital bus deenergizes, with NO operator action, which of the following contains ALL the SW pumps which will be running? ral I I jb:1111, 15. I rc:i113,16. I fd.l j 15, 16. I 1Answer 1 a I I !Exarri Level 11 R I JcoghitiveLevel i IApplication I !Facility: i ISalem 1 & 2 I [Exa?lDate: 11 12/21/20151 ~I 076000K201 V~lue:1@2:J;sROValue:i[}]'lsection: 1j~ IRO Group:ILJ1SRO GrQup:1LJ l;:!J l"~
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D !system/EvolutiorlTitre ! lService Water System 11076 I IKAstateme!lt: I Knowledge of bus power supplies to the following: Service water jt::~planath:m Answers: ofl
55.41(7) A bus powers 15, and 16 SW pumps. On a single bus UV as described in the stem, only that bus would load in blackout loading. A bus is locked out on Bus Differential (deenergized), and the loss of 16 SW pump would cause header pressure to lower to where the auto pump (11) would start. Only one SW pump is aligned for AUTO which is the normal at power configuration for the SW pumps, one in auto, and the rest in manual. 12 SW pump would never start unless 11 pump did not on a SEC initiation, that is "h' ;t ;c, --* .,_, _ _, ;n --* nf'h- -h-:--- 111 C::\A/ n* *mn **-.,Jn --* -*~.+ -:--~ D "* "" --**~* *~-~- --* *-* . .,h:~h ;c **h, ;t :--" '
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any of the choices. There can be confusion about the running EDG and the loss of A vital bus' causing a MODE 1i (Blackout), which would strip busses and load the primary SW pump on each bus. The unit 1 SW pump power supplies are reversed from unit 2 (21/22 pumps A bus, 25,26 pumps C bus) i f Page No. I IRevision1
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lI Facility Reference Num.ber /j /Reference Set;tion IService Water Pump Operation II S1 .OP-SO.SW-0001 I II,, 1127 I IUnit 1 4KV Vital Buses One line 11203002 I 1134 I I II I II 11 I tLo. Nutnber .* *. Objectives I SWBAYSE005
,_ _ __, IMat~rialRequiredfor EJ<aminatiol"l 11 I ['auestion Source: 11 Facility Exam Bank I[ouestlC>n fJl'?dific~tiC>n'M~thp~: IDirect From Source IIOsed l)ifr.ing Training Program I o [auestion s9urce <;~~111entsi 1152970, used on Sept 2011 NRG RO exam (3 exams ago) I iComrnent I I I
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Outline Changes I i(luestiori Topic J 1RO 62 I Given the following conditions:
- Unit 2 is in MODE 3, NOT, NOP. - # 2 ECAC was manually placed in operation to collect bearing vibration readings JAW S2.0P-SO.CA-0001, Control Air System Operation. - PRIOR to starting the# 2 ECAC, BOTH Control Air headers were at 96 psig. - 2C 4KV vital bus senses an UV condition, and the bus loads in MODE II*. - 2A and 2B 4KV vital buses remain powered from SPT's. - 5 minutes after the UV signal, Unit 2 ECAC oil pressure indicates O psig and has been 0 psig for> 1 minute.
With NO operator action, which of the following describes the effect this will have on the Control Air System? Assume the ECAC Motor Overload has not actuated at any time.
# 2 ECAC is I , and Control Air header(s) is/are
[J I NOT running, "A", lower than "B" header.
!b.
running, "A", higher than "B" header.
I I i_§_{ NOT running, BOTH, 96 psig. I I I I ldl running, BOTH, 96 psig. [Answer.I @=] !Exam Level I ~!cognitive Level 11 Application I [Facility: 11 Salem 1 & 2 1 IExamDat~:J I 12/21/20151
~I 078000A301 I1A3.01 I!Ro value:J[IijfsROValue:J@jSection: JI~ fRO Group:JLJ!~Ro Gro~i>:ILJ Ill D !System/Evolution Title: !instrument Air System I *~*0_78_~ *KA Statement: j Abilit to monitor automatic operations of the Instrument Air S stem includin Air ressure lExplanation .of/ 55.41.b(?) The ONLY trip which remains active for the ECAC after ANY SEC start is motor overload. The ECAC operating !Answers: .* * ***
1 characteristics are such that at 95 psig and above header pressure, the ECAC will NOT be supplying the CA header. With the stem conditions of 96 psig prior to and after the ECAC was originally started, the subsequent (SEC) stop and restart of the ECAC will have NO effect on CA header, since the Station air headers supplying the CA header were not affected by the II* loading of 2C vital
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i' Reference Title :r Facility Reference Nurribe~ i[Reference Section * ' j fPage No,i jRevisionJ IControl Air System Operation II S2.0P-SO.CA-0001 I !I 1114 I IControl Air Lesson Plan ii NOS05CONAIR-12 I 1130-31 11 I I II I ii ii I [LO. Number
N Objectives I CONAIRE008 1_ ___.
[q~,esfil),p sp~rce~'i1 IFacility Exam Bank I jQtie~ti?r(~.~~}f!g~tiop.~eitj~q; 'I Direct From Source I [used [)urin9:i:tairirlg' Bf9.Qram; I D r:ou~~if?n ~ourcil CoriiiJi~r,t~I ;::.======-================:::===========-==============--,.1 187627 I
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I SRO System/Evolution List I Outline Changes I [Question Te>?ic I I RO 63 I I In addition to providing a source of water for Fire Protection, Fresh Water and Fire Protection Storage Tank water can be aligned to which one of the following systems? I g) I Service Water. I [gJ I [] I Main Condensate. I
@J I Auxiliary Feedwater.
Spent Fuel Pool Cooling. I IMemory I
Answer 1 j c I !Exam Level [ IR I !cognitive Level J I !Facility: 11 Salem 1 & 2 I IEXaml)ate~ 11 12/21/20151 KA:ll 086000K103 IiK1.03 I !Rb Value: II 3.4*lfSRO Value: 113.5*l[Section: n~ 1RO G~C>up:q 21 !SRO Group:! I 21 ~}"IJ D

system/Evolution Title I l_F_ir_e_P_ro_t_e_ct_io_n_S'""'y'""s_te_m
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____,l 1086 lt<A State111ent: I Knowledqe of the physical connections and/or cause-effect relationships between Fire Protection System and the following: AFW System '.Explanation of I 55.41.b( 4) Fire protection water can be aligned to the AFW system through a normally disconnected spool piece. !AnsWer~: : ! Reference Title <' ii . Facility Reference Number, . !!Reference Section*** I!Page No. j !Revisionj INo. 1 & 2 Units Fire Protection 11205222-4 I 1163 I I I II II I I
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I iLO. Numbe'r** .* * rExainination. :j I II [question.S,oul'ce: 11 Facility Exam Bank 1Jou~s!i()n"Mod\fication M~thod: {J Direct From Source I !used burir:1g"Trail)ing Program JD
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RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes (Question Topic'.! I RO 65 I IWhich of the followin~ identifies a condition in which the Unit 2 LCO for Containment lnte~rity, 3.6.1.1 Modes 1-4, would NOT be met? ! I
~ A manual valve or Blind Flange outside containment required to be to be closed during accident conditions cannot be visually verified in correct position due to its location in a High Radiation Area. I I ~ The containment 100' elevation airlock doors are operated by procedure to allow entry into containment for Rad Pro to take radiation surveys.
I 81 A CVCS Letdown Orifice Isolation Valve fails to fully close on a failure of the controlling PZR level channel LOW. I [:] IA SW Accumulator nitrogen cover gas pressure falls below the minimum required. I jAnswer 11 d I !Exam .Level !IR I !cognitive Level 11 Application I !Facility: 11 Salem 1 & 2 ! IEx1(:111}Dat~: JI 12/21/20151 lf<A: 111030008240 I 2.2.4?___j !Ro Value: II 3.41 jSRO Value: l14. 7 J !Section: 1I~ [RO Group:II 1 I fSROGroup:l j 1I B D jsystem/Evolution ntre I j_c_o_n_ta_in_m_e_n_t_S_,_y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.I ! 103 [KA S t a t e m e n t : ! . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . . , . Abilit to a I Technical Specifications for as stem. Explanation of 55.421.b(9, 10). A is incorrect because 4.6.1.1.a specifically states that a manual valve or blind flange in a high radiation area may ,Answers~ ** ** 1 be verified using Admin controls. B is incorrect because while containment airlocks are required to be operable IAW 4.6.1.1.b (per spec 3.6.1.3 airlock), the doors are allowed to be opened for normal transit entry and exit. C is incorrect because CIV have their own TS 3.6.3 which is less restrictive that containment integrity and is not included in surv requirements for 3.6.1.1. D is correct j Accumulator l~vel pressure and temp
** Reference Title *j r: *,* 1 Facility Reference Number>'. I!Reference Section
j \Page No. I !Revision:
ISalem Tech Specs I 113.6.1.1 II 11 I I I lj 3.6.2.3 II 11 I I I 113.6.1.3 II 11 I !Lo. Number Objectives I I CONTMTE010 .M(:1feriai .~equi~~d fqfB@Tliriatipn :. II i Que~~io~ ~g~tc~: dl l_N_e_w_ _ _ _ _ __.l !auestion Modification Method: *. jauestion source Cornmentsl I I
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RO Skyscraper SRO Skyscraper RO System/Evolution List I SRO System/Evolution List Outline Changes [auestionTopic j RO 66I I Given the following conditions for Unit 2:
- Rx is in Mode 3, NOT, NOP. - RWST concentration - 2450 ppm - 21 BAT concentration - 6650 ppm - 22 BAT concentration - 6650 ppm - 22 BAT level - 43%
Which of the following describes the LOWEST level for 21 BAT that meets or exceeds the operability requirements for the BAT? [JI I
~154%.
I [ ] 192%. I [ ] 196%. I [Aflswer,i b I I !Exam Levell IR I 1cC>gnitive Level I
! Application I [Facility: J ISalem 1 & 2 1 IExatppate:J I 12/21/20151 fJ']" !SRO Value: IfID !section: II~ !RO Group:ILJ !SRO Group:I LJ lf'fij ~ .. T'J
[KA:jl 194001G125 112.1.25 1IRO Value: i [systell'liEvolution Title I * - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' iGENERI J IKA Statement*!., Ability to interpret reference materials, such as oraohs, curves, tables, etc. [E.xpfanaii.onpf.,! 55.41.b(6,7) TS 3.1.2.6 requires a certain amount of borated water available. The BASTs are normally cross connected so the fAnswers: * **[ TOTAL volume of the 2 tanks is what is required to be above the limit. With RWST boron concentration of 2450, and both BAT tanks at 6650, the intersection is - 93.5%. If 1 tank is at 43 % the other tank must be at 50.5%. To preclude picking the wrong answer because of interpolation, the correct answer of 54% is 3.5% higher than required. !' ' Reference Title IL Facility Reference Number.
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j f Page No. J !RevisionI j Saqlem Tech Specs II I ii 1 I I I !I I 11 11 I I II I ii ii I jLO.Number Objectives ICVCSOOE010
'--~ lMaterial Requiredfor Exall'lination. I ITS Figure 3.1-2 Boric Acid Tank Contents 1a~~stibii s?µrc~; 11 Facility Exam Bank 11a~7~ti.ol"l llJlodi~cationMethod:.*.J Editorially Modified I jus~d oµring'frainingPrqgram'J D /Cl.1,1es~ic)n SourcE:J ~c:>rnrn~nt~J ;::::======-.:::============:::::::::===========-==============-=~1 j 135342 replaced a distracted and slightly modified correct answer. j I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List . I SRO System/Evolution List Outline Changes I [auestion Topic ! I RO 67 I jWith Salem Unit 2 in Mode 6 on November 20th, which of the followin9 conditions would prevent Core Alterations from bein9 commenced? I
~ I One Source Range NI is inoperable.
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~ The reactor has been subcritical for 100 hours.
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~ Any one of the Containment Airlock doors is open. ,
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@] I Only one loop of RHR is in service in Shutdown Cooling mode.
I lf.nswer: 11 a I !Exam Level j IR I !Cognitive Level 11 Memory I iFacilitY: [ ISalem 1 & 2 I [ExamDate:j I 12/21/20151 ~ [KA: j j 194001G136 Il2.1.36 i IRO ,Value: i[Ifil ;SRO Value:IEJ] Section: Ii~ IRO Group:iLJ !SRO Group:ILJ I I mD
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lsY5tem/Evoiution J:itle I i-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~__.[GENERI[ IKAstatement:! Knowledge of procedures and limitations involved in core alterations. I 1E~pl~nation of 55.41.b(10) A is correct because LCO 3.9.2 states 2 SR Ni's must be operable. B is incorrect because between Oct 15-May 15th Answers: .. :". ** only 80 hours of subcriticality is required per LCO 3.9.3.a. C is incorrect because one ONE airlock door (per airlock)has to be capable of being closed per LCO 3.9.4.b. Dis incorrect because only one RHR loop is required to be in service per LCO 3.9.8.1 I
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lk~age No. j !Revision:
j Salem Terch Specs I I II 11 I l I I 11 1I I I I I II ,, I IL.a: Number I IOP009E004 Objectives I ,_ ____, lilJ!aterfal Required for Examiriati 0 n :) I II lai.f~s!fon sol.lrce: J l_N_e_w_ _ _ _ _ _ __.l lauestion llJlod,ificat!on MeJ~()d: *. 1--------~l Ju~edDl.lri~g TrainingProgralnJ D j'a!-!~stip~ source Commen\~J j I
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iouestion Topic l RO 68 I I Given the following conditions:
- Both units are operating at 100% power. - Reactor Engineering has determined that a single fuel assembly in the Spent Fuel Pool must be moved to a new storage location in the Spent Fuel Pool. - All administrative requirements are complete to allow the movement of the Spent Fuel.
When the field operator arrives at the Spent Fuel Building, he notices that while a Qualified Reactor Engineer is present on elevation 130', a Licensed SRO is not. Which of the following describes the Operations Department requirements for this evolution IAW S2.0P-IO.ZZ-0010, Spent Fuel Pool Manipulations?
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The fuel movement 1) occur because a SRO 2)
~11)CANNOT direct the fuel movement from the crane trolley.
2) shall I
[_] 11) CAN only required to be "in the area" for spent fuel moves.
2) is I
~ 11) CANNOT is required to provide oversight of the Reactor Engineer directing the fuel move.
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L..,.j Inot required to obseNe the fuel movement since a Qualified Reactor Engineer is present iAnswer I @=] \Exam Lever j ~ !cognitive Level \ IMemory I [Facmtf [ j Salem 1 & 2 I JE)(~inDate: j j___1_21_2_11_20_1~5i ~! 194001G142 i 12.1.42 ijROValue:l[°3]\sROValue:i['D[section: ll~fROGroup:iLJlsROGrollp:iLJ D lsysternlf:volution Title I iKA Statement' I Knowledoe of new and spent fuel movement procedures. I \Explanation ()f \ 55.41.b(10) Precaution and Limitation 2.2 of S2.0P-IO.ZZ-0010 states ... "IF ANY spent fuel manipulation(s) being performed in the
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i Spent Fuel Pool, THEN ASSIGN Reactor SeNices, qualified SRO, or Reactor Engineer to supeNise spent fuel manipulation(s)."
Since the RE is present, fuel movement can occur without a SRO. I*.*.*. Reference .Title H Facility R~ference Numb.er >I [Reference Section /I! Page No. J iRevision ISpent Fuel Pool Manipulations II S2.0P-IO.ZZ-0010 II lj 2 I 133 I I II II II 11 I I II ii 11 11 I (L,o.Nl.lmber.'
Objectives I IOP010E005 l
louestlon Sour~~: ! Previous 2 NRC Exams [.ciuestiori M9dific<:ition tv1ethod:) I :I I iusedOuririgff3inirigProgram .j D fQll~~tion ~ol.lr~e c()il)meiitsf I!-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~--! I [Corrimeot./ ... > ' I I I I I I
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RO System/Evolution List I SRO System/Evolution List J Outline Changes ( IOuestion Topic 11 RO 69 Given the following conditions:
- Unit 1 is at 100% power. - 11, 13, and 16 SW pumps are in service, with 15 SW pump selected to Auto. - Subsequently, a loss of the 500 KV Switchyard occurs. - 1A 4KV Vital Bus has de-energized due to a Bus Differential relay actuation. - Unit 1 has initiated a MANUAL Safety Injection (SI).
Which of the following identifies the Service Water Pumps which will be running 2 minutes after the SI has been initiated? [ ] 111 and 14. [gJ 112 and 13.
~ 113 and 15. @J 114 and 16.
1AnswE!r 1 I a I !Exam Level 11 R I !cognitive Level 11 Application I 1 IFatiiity: 1 Salem 1 & 2 I iExamDate:[ I 12/21/2015! !KA:jl 194001G203 jj2.2.3 ! I [Ro Value: 1 3.81 rsRo Vah.ie:ll3.9 I[section: 11 PWG I[Ro Gi-oup:l I 11 lsRo Group:il 1j 5~1~ D [system/Evolution Title II I :GENERI I KA Statement*;.. (multi-unit license) Knowledoe of the desion, procedural, and operational differences between units. I [Explanat~~n o*f
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l 55.41.b(?) This question meets the K/A because SW pumps are powered from opposite vital buses when Unit 1 is compared to Unit
2. U1 pumps are C,C,B,B,A,A, whereas U2 are A,A,B,B,C,C. The Lead pump on B bus is always 14 unless 14 is not available.
With 1A bus deenergized, 15 and 16 SW pumps have no powerThe SW pump selected to auto (15) will not start on low pressure (nor will it have power) as it will be locked out by SEC initiation. [ : ....*.. *ReferenceTitle .* ****-*([* <Facility Reference Number*.* I!Reference Section' < \l Page NC>/ fRevision; 1 1 I1B 4KV vital bus one line diagram 11203002 II 11 1134 I I II II 11 11 I I II II 11 1 I I IL.o*. ~_urnber ; Objectives ISWBAYSE005 \ __ __, I.Material Requir~d for Examination* II Ij \ouestiC>ll Source: ;11 Facility Exam Bank I \ouesti2n M,0 difi,6~tior(l\t1~1~~d: I
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[ou.estion sourcE!£ommentsj l , . . . . 4 _ 8 _ 9 - 9 2 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - = - 11
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!Question Topic I IRO 70 I An Independent Verification (IV) of valve position is required in an area with a 75 mrem/hr dose rate.
For this job, which of the followino is the longest time allowed for the IV before "hands-on" verification may be waived?
~ 15 minutes.
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~ 17 minutes.
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~ 19 minutes.
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@] 111 minutes I !Answer l b I I !Exam Level 11 R I !cogri'itive Level 11 Application I !Facili~y: 11 Salem 1 & 2 I !Exanibate: 11 12/21/20151 r~~~""'*:m i IRO Value:I ~ ISRO Vallie:! [TI} ~Section: :j~ 1RO Group:JLJ [SRO Group:: L J li~~~ ~ ,.
jKA:ll 194001G214 i I 112.2.14 !System/Evolution Title ! * - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' IGENERI I II KA statement* I Knowledoe of the process for control lino equipment configuration or status. iExpla;;iation !.Answers: ofl 55.41.b(10) 10 mrem is the dose above which an IV is not required to be performed. A= 6.25 mrem. B=8.75 c=11.25 mrem
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IComponent Configuration Control ll OP-AA-108-101-1002 II Att 11 Step 1.5.1 1165 11 7 I I I I II I I lLO'. Number **.
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11 11 I IMISCAPOO? !Material Reqllired torExamination*;*J I IQuesti?pSo.llrc~:>J IFacility Exam Bank I !8'Lies!i?.~ 1vf?~itii:ation Method:.] Significantly Modified I!usedptlrlng"frainfr1g Program I o [Questi<:>n ~olirc~ <;cjf!lmentsi 160955 changed dose rate which changes correct naswer to a previous distracter. I
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tOuestiof! Topic 11 RO 71 Given the following conditions:
- Salem Unit 2 is operating at 100% power. - 21 SW pump is err. - 22, 25, and 26 SW pump are in service. - Subsequently, 22 SW pump discharge strainer clogs, 22 SW pump is stopped, and 22 SW pump is declared inoperable. - 2 hours after being secured, maintenance discovers a crack in the strainer drum which will take 1 day to repair.
Of the following, which is the only method of tracking SW pump status which is NOT performed IAW OP-SA-108-115-1001, Operability Assessment and Equipment Control Program? Updating the is NOT required.
~I Operational Status Board.
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!b* I Control Room Narrative Log.
I [Cl I Tech Spec Action Statement Status Board. f";Jl IIgcbcic~I StH~cifk:~tioc Ac:tioo st~tQ I -- I
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Answer I ~ !Exam Level i ~ !Cognitive Level >11 Memory I 1 IF;acility: ! Salem 1 & 2 I /Exa111D~te: j !___12_12_1_12_0_1~51
~I 194001G223 Ij2.2.23 *'Ro Value: l[TI"lsRo Value:l~!Section:H~ !Ro Group:jLJisRO Group:ILJ D (system/Evolution Title I IKA Statement: I Abilitv to track Technical Specification limiting conditions for operations.
Explanation of 55.41.b(10) Initially, actions wil be attempted to clear the strainer clog, so it won't be readily apparent that the repair would take iAnswers: longer than one shift. (Sect 5.2.4) When it becomes apparent that it will take longer than one shift, Section 5.2.5 will be performed also. The Operational Staus Board is used during emergencies and is located in the control room area.
I *.*Reference Title
**.H ,Fac;ility Reference. Number [Reference section **I IP,age No. IIRe,yision: *1 j Operability Assessment and Equipment Control II OP-SA-108-115-1001 I 11 1I7 I I II I 11 11 I I II I II 11 I
[LO. Number ., Objectives I TECHSPE015 1_ ____. lauestion Source: ;:J l_N_e_w_ _ _ _ _ ___.I !auestlo!l ~odificcitio.n Meth()d: ij_________.I !used pliring'.'.fraliiirig Program. l D [guestio(l ~.?urce Comme~~~l I I
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!auestion Topic ! IRO 12 I Given the following conditions: - A Unit 2 shutdown is in progress. - A containment entry is required to inspect an INOPERABLE component on elevation 78' outside the bioshield. - The Unit is being shutdown at 10% per hour.
Which of the followinq additional approvals is required to authorize a containment entry other than the SM/CRS under these conditions?
~ I Operations Director - Salem.
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~I Station Vice President- Salem.
c.l jWork Control Center (WCC) Supervisor.
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@J I Radiation Protection Supervisor (RPS).
I I !Answer J Id I [Exarri Level* 1 I R I ;*cognitive Lev,el
I I Memory I iFacilify: IISalem 1 & 2 I jeX:ainOate:j I 12/21/20151 IKA:ll 194001G312 112.3.12 I !Ro Value: JI 3.21tSROValue:Jl 3.7 l lsecti6f"l:jj PWG lfRb Gro(Jp:[I 11 \SRO Group:il 11 if!ll D fSystemlEvolutionTitleJ
[t<A Statement* I Knowiedge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, alioning filters, etc. jE,xplaQationo~ [ Normal Containment entries at power are governed by SC.SA-ST.ZZ-0001 & RP-SA-102. Authorization to access containment is 1~nswers: . 1 provided by the SM/CRS. However, to access containment when power is being changed >5%/hr, the Radiation Protection Supervisor's approval is required. (Pre-req 2.4) [ ******Reference Title.. * ,:,*,, ll Facility Reference Number . <1 [Reference Section <!IPage No. j }Revision: ISALEM CONTAINMENT ENTRIES IN MODES II SC.SA-ST.ZZ-0001 I q I 15 I I \I I ii 11 I I II I II 1 I I ILo. Number Objectives I RADCONE004 ,_ _ __. /Material. Required fc)i: Examinatipn, *l . II jQuesHon S:?~rce: 1 lFacility Exam Bank I jaue~ticm 'M?dific~~ion.Metho.a:: JEditorially Modified Itus~d riurtng TraillJ~g Program.I o lQllestlon Sollrce CoriJrJ1ents] 178014 added abbreviation to each title because its found that way in the procedure. I
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I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I .Outline Changes I [Question Topic 11 RO 73 I IWhich setpoint? of the following Area Radiation Monitors (ARM) will cause a ventilation system alignment change when it reaches its High Radiation Alarm I
~12R44A, Containment High Range.
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~12R32A, Fuel Handling Crane.
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~12R9, id. j I New Fuel Storage.
2R52, Liquid PASS Room. I I I.Answer! c I I lexam Level i IR I jcognitive l..evel I j Memory I !Facility: j ISalem 1 & 2 I !ExarnDate:*.11 12/21/20151
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IIRO Value; I' [IE ISRO v.alue: 1 I >:, [KA:ll 194001G315 112.3.15 :SRO Group:! !system/Evolution Title I !~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-'
!GENERI; IKA statement: l edge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel orin e ui men!, etc.
!Explanation of 55.41.b(11) C is correct because it realigns FHB ventilation through the charcoal filters and starts both FHB Exhaust fans. B is iAnsw~rs: incorrect but plausible because its auto function is to prevent Fuel Crane motion except in downward direction. D is incorrect since ~~~-~~ it only has alarm light outside the PASS room which activates, but plausible because of the high radiation levels which would be expected in that area of the aux building following an accident. A is incorrect since it has no automatic function, but is plausible
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II Facility Reference NulJ1ber .. *I [.~efere11ce :Section I [Page No: I iReviSionl IS2.0P-AB.RAD-0001 11 Abnormal Radiation II Attachment 5 RMS ch 1114-16 1130 I I II I 11 11 I I II i 11 11 I
!Lo'.. Number RMSOOOE005 [Material Requir~d :for Exami1JCl,tior1? II I fQuestion So~rce: >i Facility Exam Bank I /al.l~sti~h ~~di!!C:Ci~i<;mMethod:, IDirect From Source I[Qse'd oUriiig Jrclrnillg Program l D fauesti~ns.~u{~e.<;orryme~t~/ 1125827 used on 9/2011 Salem NRC RO exam. 3 exams ago. I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes lauestion Topic 11RO74 I Given the following conditions:
- Salem Unit 1 is operating at 100% power. - 1A EOG is CfT for maintenance. - A 500KV switch yard malfunction causes a loss of offsite power, and 1B and 1C EDGs do not load due to 1Band 1C 4KV vital buses locking out on Bus Differential.
Of the following, which one states a procedure entry and its correct flowpath allowed by the rules of procedure !AW OP-AA-101-111-1003 Use of Procedures?
~I Enter EOP-LOPA-1 directly and perform immediate actions, which states to trip the Rx then trip the Turbine. ~I Enter EOP-LOPA-1 directly and perform immediate actions, which states to trip the Rx and confirm the Rx trip, then trip the Turbine. ~I Enter EOP-TRIP-1 and perform immediate actions, which states to trip the Rx and confirm the Rx trip, trip the Turbine, initiate SI, then transition to EOP-LOPA-1 based on no vital buses energized. @]I Enter EOP-TRIP-1 and perform immediate actions, which states to trip the Rx and confirm the Rx trip, trip the Turbine, initiate SI ONLY if conditions warrant, then transition to EOP-LOPA-1 based on no vital buses energized. I
!Answer 11 a I IExam Level 11 R I I
!Cognitive Level ! Application I jFacility: i ISalem 1 & 2 I ;examDate: 11 12/21/20151 IKA:ll 194001G401 I_;2._4._1_~! IRO Value: l~ISRO Value: I~ 1Sect1on: II~ )RO (3roup:)LJ :SRO Group:j L J t
BJ ,~ ;~ D l~s~ys-t-~m-./-E-v-ol~ut-io_n_T-it-le~j IGENER! I iKA Statement: 1 Knowledoe of EOP entry conditions and immediate action steps.
ekpfanation of I 55.41.b(10) Either TRIP-1 ot LOPA-1 can be entered upon a total loss of all AC power. If entered, the flowpath for TRIP-1 does not
!Answers:* . reach the SI evalution step before the kickout to LOPA-1 on no 4kv vital buses energized, so both TRIP-1 distracters are incorrect. LOPA-1 does not confirm the Rx trip (since there is no power to do anything about it anyways). ...'
Reference Title It Facility Reference Number . *I !Referenc*e Section* IIPage No~ I !Revision:
I Rx Trip or Safety Injection 111-EOP-TRIP-1 II 11 1127 I ILoss of All AC Power 111-EOP-LOPA-1 11 ii 1125 I IUse of Procedures II OP-AA-101-111-1003 11 1110 11 6 I I LOPAOOE009 Objectives 1_ _____. [auestio,? S()~rce:] l_N_e_w_ _ _ _ _ _ __.l la~~stioll MMificati~n,M,~tllo,g: v _________.1 lused DuringTrarning ~rc)gram I D
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Outline Changes I louestiqn Topic 11 RO 75 I Given the following conditions:
- An Alert has been declared at Salem, and all required notifications have been made by the Primary Communicator. - Conditions degrade to the point where a Site Area Emergency is declared.
Which of the following identifies the PRIMARY method which the Primary Communicator will use to make notifications to the States of Delaware and New Jersey, and how long from the SAE declaration do they have to make those notifications !AW Attachment 6, Primary Communicator Log of the Salem ECG?
~I NETS phones within §]I 15 minutes.
I fc.', I NETS phones within 60 minutes. I L___J []I ESSX phones within 15 minutes. I ESSX phones within 60 minutes. I i*Answer.l j a I !Exam Level 11 R I jco911itive Level 11 Memory I !Facility: !ISalem 1 & 2 I fi='xamDate: 11 12/21/20151 IKA:Jl 194001G429 I~:2_.4_.2_9-~! iROValue: :OJ}fsRO Value:J!ITilsect1on: JI~ iRO Group:ILJ1SRO Group:jLJ *--<:".<,
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E D ~ls-y-st_e_m_/E_v_o_lu-t-io_n_T_it-le~'l IGENERI I IKASfafomenfl Knowledqe of the emerqency plan. [Explanafi~n of j 55.41(10) Salem ECG, lists the communications systems in order of preference. The NETS (Nuclear Emergency iAnswers: *
Telecommunications System) is the primary closed circuit communication system for off-site notifications. The ESSX is also a closed circuit system, which is used as a backup for NETS. The notifications to the States must be made within 15 minutes of the declaration of an Emergency, even if a lower classification emergency is already in progress. The 60 minutes is plausible if the
---..J:.J-.1.- .&.L-~-l*e '-'--.a.:-- ie , ___ ' *- _: ___ ~,.., ".:I, ................ -": ........... ' ,,*r *** : Facility Reference Number
iIRe~erence Section If Page ~o. I IRevision; I .Reference.Title*..* :'. 1 r~
IPrimary Communicator Log 11 EP-SA-111-F6 II II 1112 I j Emergency Preparedness Training Communicat 11 NEPCOMMDTYSC II II ljo5 I ! II II 11 1 I I GENISSE013 Objectives I !Question so1:1rce: 11 Facility Exam Bank 1 lq~~stioll'~odifi9~t.toll:M~th()d: *. IDirect From Source I[Used f?uring Traihi~g Program I D
\otiesti~11~PLn'ce c;o~1!1ef1t~l 1Used on Salem Sept 2011 NRC exam (3 exams ago)
I fcoriunent ***.* ** \ *. I I I
U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: IR . I eg1on: I i I Date: 12/21/2015 I Facility: Salem 1 & 2 I License Level: SRO I Reactor Type: W I I I Start Time: i Finish Time: I i Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours to complete the combined examination, and 3 hours if you are only taking the SRO portion. App11cant Certmcation All work done on this examination is my own. I have neither given nor received aid. Applicant's Signature Results RO/SRO-Only/Total Examination Values --I --I -- Points Applicant's Score --I --I -- Points Applicant's Grade --I --I -- Percent
Senior Reactor Operator Answer Sheet Circle the correct answer. If an answer is changed write it in the blank. NAME:
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2. a b c d

3. a b c d

4. a b c d

5. a b c d

6. a b c d

7. a b c d

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9. a b c d

10. a b c d 11 . a b c d

12. a b c d

13. a b c d

14. a b c d

15. a b c d

16. a b c d

17. a b c d

18. a b c d

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20. a b c d

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22. a b c d

23. a b c d

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25. a b c d Page 1
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [auestion Topic J ISRO 1 Given the following conditions:
- Unit 1 is performing a Rx startup IAW S1 .OP-IO.ZZ-0003, Hot Standby to Minimum Load. - Power is stable at 1x10-8 Amps for Low Power Physics Testing. - A Shutdown Bank "A" rod drops fully into the core. - The Rx does not trip.
Which of the following identifies:
1) How the CRS should proceed

2) Why rod withdraw! is not allowed?
Direct the RO to ...
~ 1) trip the Rx.
2) A dropped rod recovery would constitute an approach to criticality.
~ 1) fully insert all Control Bank and Shutdown Bank rods.
2) A dropped rod recovery would constitute an approach to criticality.
~ 1) trip the Rx.
2) The depressed power distibution profile in the area around the dropped rod may cause power production in other parts of the core to exceed Tech Soec limits.
@:] 1) fully insert all Control Bank and Shutdown Bank rods.
2) The depressed power distibution profile in the area around the dropped rod may cause power production in other parts of the core to I VAVVVU I VVl I ""tJ'VV lllllll>=I.
---* --* .-- -
!Ans""°~ ~ ~m Level 1 IS I [C09flitive L~~U I Application I *~ I Salem 1 & 2 I iExa~ 1_ _ _12_12_1_12_0_1_,sl Bl ooooo3G409 I',?-4.9 ____1~~~a~~§~e>_~ls~~~i~f!JI~ ~~-G~LJ !SRO Grou!):;[] ~ System1E1folution Titl~J l_D_ro~p_p_e_d_C_o_n_tr_o_IR_o_d------------------------------~ ____ IKA Statement=- ~- . ,. -~-~ Knowledge of low power/shutdown implications in accident e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. I IExplanation !Answers: 01 55.43.b(6,5) This question is SRO level based on the knowledge of internal effects on core reactivity of a dropped rod while exactly critical, and the procedural direction to insert all rods based on becoming subcritical. The bases for that insertion is that performing - - * *.. **-**-- a dropped rod recovery when the problem has been fixed, which would normally be done at power, would NOT be done if the Rx were subcritical, because withdrawing that rod would constitute an approach to criticality and is required to be performed IAW the
... 4-..,rf11n ............. ,.,.....1.1ro '

-*--

L_ Reference Title
-*. -- --=11 Facility Reference Number _J 1Reference Section 11 Page No. J [Revision j Dropped Rod II S1 .OP-AB.ROD-0002 I 11 ! !10 I l II I 11 II I I II I 11 11 I
!LO. Number ~ Objectives ABROD2E002 I I I
,,
[Material Required for Examination iI 'I jOuesti~~ Source: I I New I!Oues~~~ Modific~ion Method:__J I [t1sed During Training Program 1 D [Ques~~n source Comments1 I I I:_.... *-* .. l I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic I ISRO 2 Given the following conditions:
- Unit 2 is operating at 100% power. - 21 charging pump is in service. - 22 and 23 charging pumps are operable. - Subsequently, 21 charging pump trips due to a breaker malfunction. - The CRS enters S2.0P-AB.CVC-0001 and places 23 charging pump in service.
Which of the following:
1) Identifies ALL the Tech Spec LCO(s) which will be entered

2) What the required action(s) is/are if 21 charging pump remains inoperable for the next 4 days?
The CRS will enter .....
~ 1) LCO 3.1.2.2.b for not having required boration flowpaths, LCO 3.1.2.4 for not having required charging pumps, and LCO 3.5.2.a for not having required ECCS subsystems available.
2) Be in Mode 3 and borated to a SOM equivalent to at least 1% delta k/k within 78 hours of pump trio.
~ 1) LCO 3.1.2.2.b for not having required boration flowpaths, LCO 3.1.2.4 for not having required charging pumps, and LCO 3.5.2.a for not having required ECCS subsystems available.
2) Be in Mode 4 within 84 hours of pump trip.
~
lS 1) LCO 3.5.2.a ONLY for not having required ECCS subsystems available.
2) Be in Mode 3 and borated to a SDM equivalent to at least 1% delta k/k within 78 hours of pump trip.
rc.v l____J
1) LCO 3.5.2.a ONLY for not having required ECCS subsystems available.

2) Be in Mode 4 within 84 hours of pump trip.
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~~~ @=] ~ri:i_L~~ js I~~ 1_ _ _ 12_12_1_12_0_15~1 r 1 lcognitiveLeveIJ !Memory I §nty:j !Salem 1 &2 ~I 000022G222 I ~_j [RC>:Vai~e~ [Ifil~rSeCtion:ll~ Ro Group:]LJ!sRo Group:[LJ 1 ~ ]system/Evolutionii!&] ILoss of Reactor Coolant Makeup 1 KA siaieme.m:::l
*-
Knowledge of limiting conditions for operations and safety limits. ~lan.ation of i 55.42.b(2) This question is SRO level based on NUREG-1021, ES401, Attachment 2, page 17, 11.B, 1st bullet for application of Ans!'fE!.':!;:__j required actions. lnoperability of a single charging pump in Modes 1-3 only results in entry into the ECCS LCO. The 2 other LCOs would be entered upon inoperability of the second charging pump. The required action is to restore within 72 hours or be in Hot Shutdown within next 12 hours. The incorrect action in the distracters is for when 2 charging pumps are inoperable.
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I Reference Title ~FacilitY__~eference Numb~ 1Reference Section ][iiage~ ]Revision! j Salem Tech Specs I l j 3.5.2, 3.1.2.2, 3.1.2.4 II 11 I I I II Ii 11 I I I II ii 1 I I IL.o. Number Objectives I CVCSOOE010 ,____.
!Mate~ial Required for E_xamination ;I ii f Que~tion Source: JINew I[9~~stion Modification Method~- ...I 11 I jUsed O~ing Training ~~gram I D [Question ~~~rce Comments] I I ,;:_******-Hl I I I I I ! I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [§uestion Topic 11 SRO 3 I Given the following conditions:
- Unit 2 is in Mode 5. - Rx vessel level is 97.4'. - 22 RHR loop is in service. - RHR HX inlet temperature is 125°F. - 21 RHR loop is available. - 21,24, and 26 SW pumps are in service. - 21 charging pump is in service. - Subsequently, a complete loss of Service Water occurs.
Which of the following identifies a procedure the CRS will enter, and the FIRST action operators will perform in that procedure? I
~ S2.0P-AB.SW-0005, Loss of All Service Water. Start 23 charging pump.
I I
~ S2.0P-AB.SW-0005, Loss of All Service Water. Stop 21 charging pump.
I I
~ S2.0P-AB.RHR-0002, Loss of RHR at Reduced Inventory. Start 21 RHR pump.
I
@;]I S2.0P-AB.RHR-0002, Loss of RHR at Reduced Inventory. ..
Stop 22 RHR pump. I
@=] ~~~~~ IS I ~~~~ 1_ _ _1_2_12_11_20_1~51 *~~~--~ ~"-,--~---
[Answer I I [cognitive Level 11 Application I !Facility:: I Salem 1 & 2 ~I 000025G120 12:1~2_?_=~] i~O Value: J~[SRo ValueJ~ !section: II~ [ROGroup~LJlsRO Group:ILJ ~ 'system/Evolution Title! ILoss of Residual Heat Removal System IKA Statement: I Ability to interpret and execute procedure steps. I !Explanation of i 55.43.b(5) This question is SRO level based on having to assess facility conditions, select the correct operating procedure, and IAnswers: ! know specific actions taken in that procedure. BOTH procedures may be entered, however, AB.SW-5 has a CAS that states if RHR is in service GO TO AB.RHR-1 or RHR-2 depending on RPV level. AB.SW-5 distracters are plausible because procedure states that if 23 charging pump is IMMEDIATELY available, then place it in service based on its being cooled by CCW, and will extend the
..:........ +n ... 11 ........ nl--=-- .........................,I"'\ in---*=-- +,..... -1: ...................... - - . ... . . ..... -. ..... 1: ..... - ... fl"\ i+ LI:-1.... ....... ----:.a.: ...... 1........... ... ... .
___ ,.: ....... -1 J: .. - : ..... :.1.;...,1 conditions that ALL SI and charging pumps EXCEPT one are required to be cleared and tagged with RCS temp< 312°F, and that there will not be another charging pump to go to. 21 charging pump WILL be stopped, but not before letdown is isolated. AB.RHR checks a RHR pump in service, so starting a second RHR pump will not be performed. The in service RHR pump will be stopped to preclude damage to the pump with RPV level <97.5' I L.. Reference
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Title 11 Facility Reference Number J[Ref_erence Section______l IPageNQ.l iRevision ILoss of All Service Water !IS2.0P-AB.SW-0005 II 11 114 I ILoss of RHR at Reduced Inventory II S2.0P-AB.RHR-0002 II 11 1114 I I II II 11 11 I IL.O. Number Objectives IABRHR1E004 1--~
!Material Required for Examination 11 lj !Questi~~~ou~ IFacility Exam Bank I[~uestion Modification Method: jSignificantly Modified It!Jsed During Traini!1~-~~ogram 1 iD ~on source co~ments] ,_
~-- I v AB.RHR actions to which procedure and which action.
I !Cornn.-... I I I I I I I
RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes f [Question Topic 11 SRO 4 Given the following conditions:
- Operators are responding to a 650 gpm tube rupture on 22 SG which occurred while operating at 100% power, IAW EOP-SGTR-1, Steam Generator Tube Rupture. - All off-site power was lost when the Main Generator breakers opened. - The RCS was cooled down to Target Temperature, and then depressurized to stop primary-to-secondary leakage. - Subsequently, off-site power is restored, and all 4KV Group Buses are now energized from off-site power. - The crew is evaluating RCP re-start. - RVLIS Upper Range indication is 98%.
Which of the following contains:
1) ONLY the criteria in addition to RCS Subcooling which are checked to determine if RCP re-start will be performed

2) How the re-start of the RCP be accomplished if allowed The crew will check ....
~- 1 j ~~ ~:_~level and the PZR saturated. ~, a RCP directly at SGTR-1 Step 49 based on RCP support conditions being desired but not required to start the RCP. ~ 11) RVLIS Full Range indication and SG NR level. *--- 2) Start a RCP directly at SGTR-1 Step 49 based on RCP support conditions being desired but not required to start the RCP. ~ 11) PZR level and the PZR saturated.
2) Start a RCP using S2.0P-SO.RC-0001, Reactor Coolant Pump Operation to ensure all support systems and P&Ls for starting a RCP are met.
[ ] 1) RVLIS Full Range indication and SG NR level.
2) Start a RCP using S2.0P-SO.RC-0001, Reactor Coolant Pump Operation to ensure all support systems and P&Ls for sta v : ~~,.,
met. !Answer I ~ :exam Level' ~ !Cognitive Level 1 I Memory I !Facility: 11 Salem 1 & 2 I IExamDate:J I 12/21/20151 ~:ijoo0038A217 l[~~~*~ __J[f!~~l~_!:H 3.8jl~_ROValue)~:section:il~IROGroup:ll 1jlSROGroup:ll 11 Im~ 's~~l§~ol1.1_ti<:>'"1!i-1!~J lSteam Generator Tube Rupture 11038 [KA StatE!_~nt~ Abilit to determine and inter ret the followin as the ap I to Steam Generator Tube Ru ture: Explanation of 55.43.b(5) This question meets SRO only criteria listed in NUREG-1021, ES-401, Attachment 2, 11.E. Figure 2, 1st bullet for 5th Answers: Block for assessment and implementation of a procedure or section of procedure, namely SO.RC-0001 vs direction strictly in the EOP. SGTR-1 does not contain a step which directly starts a RCP, rather it directs RCP start IAW SO.RCP-0001, which will require support conditions satisfied to start the RCP. There are other places in EOP network where RCPs are started without regard to
. . . 0 . . . .
Reference Title -------1 c--.=acilitYRefilrencetiiJffiiier-- *1 [Ret&encesecuorll ~~ [ReviSIOrll ISteam Generator Tube Rupture lI2-EOP-SGTR-1 ILo. Number Objectives 01E009
!Material Required for Examination i! II I jQuestion s~-u~c~_:J New I [Ouestio~ ~o~ification Method~! I [!Js_E!~_c.>uring Training Program D I ~~stion Source Commentsj I I 1:_,.,.,,_,,. I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I I Question Topic 11 SRO 5 Given the following conditions:
- Unit 1 is operating at 100% power. - 21, 24 and 26 SW pumps are in service. - 21 and 22 SW header pressures are 108 psig. - The following OHAs annunciate sequentially in this order: - B-13, 21 SW HOR PRESS LO - B-14, 22 SW HOR PRESS LO - B-15, TURB AREA SW HOR PRESS LO. - B-48, SW VL V RM FLOODED.
The standby SW pump starts automatically, and OHAs B-13, B-14, and B-15 clear. Which of the following describes where a SW leak could be located which would produce these alarms, and the procedurally directed actions which would mitiQate the event? I
~ On a CFCU supply piping in the 78' Mechanical Penetration Area. S2.0P-AB.SW-0001, Loss of Service Water Header Pressure, will direct isolating multiple CFCUs IAW Attachment 5 CFCU leaks, until header pressure stabilizes and sump pump runs stop.
I
~I Upstream of the 2ST901, TURB LO CLR ST RET VLV. S2.0P-AB.SW-0002, Loss of Service Water-Turbine Header, will direct operators to adjust 2ST901 TURB LO CLR ST RET V, and 2ST1 TG AREA SW PRESS CONT VLV, to compensate for the SW leak.
I
~ On a CFCU supply piping in the 78' Mechanical Penetration Area. S2.0P-AB.SW-0001, Loss of Service Water Header Pressure, will direct isolating a single CFCU which would be readily identifiable from the control room.
rd.. IUpstream of the 2ST901, TURB LO CLR ST RET VLV. S2.0P-AB.SW-0002, Loss of Service Water-Turbine Header, will direct removing the 11v1C:1111 1uru111e110111 :;erv11,;e 111 µreC:1JJIC:111urr 101 r:;u1C:111r1y 111e 11.:>/-\ r1eC:1ue1. i§!\Y~ ~ [fxam-Level ! @=] [cognitive LevelJ lApplication I ~iliti] ISalem 1 & 2 I jExamDate: 1 !___12_12_1_12_0_1_.sl IKA:]j 000062G445 112.4.45 I ~~[DJ SRO Valll_~j~ !section: JI~ !Ro Group~LJjSRO Group:ILJ ~
gystem/Evolutio~~!itltij ILoss of Nuclear Service Water 11062 -~

-~J IKA Statement-I
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to prioritize and interpret the siQnificance of each annunciator or alarm. IExplanation of I 55.43.b(S}This question meets SRO only criteria listed in NUREG-1021, ES-401, Attachment 2, 11.E. Figure 2, 1st bullet for 5th Answers: ---~ Block for assessment and implementation of a procedure or section of procedure in knowing where the leak could be (assessment), then correct procedure implementation and steps taken in the peocedure. The leak location could be in the TGA with the conditions in the stem except for the SW valve room flooding. Knowledge of where the SW valve room and what piping is there
--1 +,... --- ........ ,.,1 ...... .a.:........ Tho ')C:TQf'\'1 ...... ilrl ........................ - .... Tr:!./\ --* - - - ' ~-..... 1 . . . - - ' *- - - .... -.Lr.,..:- ,...,..., ,1r1 ........ *- .... . nf header pressures. If it did not operators would be directed to ake manual control of 2ST901 and 2ST1. If it is thought that the TGA header must be isolated to stop the leak, then removing the MT from service would have to occur. The multiple cFCU isolations directed in attachment 5 is for leaks of undetermined CFCUs in containment, and refers to containment sump pump runs and trying to isolate the leak by stopping a bunch of CFCUs. Step 3.11 states if a single component can be isolated, and it can, to isolate it.
Additionally, the SW indication in control room would identify that a single CFCU is affected. The sump pump runs referred to are containment sump pump runs, and are if the SW leak is in containment. - Reference Title --] ~C:ility Reference Nu01ber ![Reference Section J[Page No. I IRevision: L__ -- ILoss of Service Water Header Pressure 11 S2.0P-AB.SW-0001 II ii ! I16 I IOverhead Annunciators Window B ll S2.0P-AR.ZZ-0002 II ii 81 1136 I I II II 11 11 I IL.O. Number Objectives I ABSW01 E004 ,_ _ __.
]Material Required for Examination JI II [~uestion Sou~<:~ IFacility Exam Bank I~e~~on Modification Metho~: JI Significantly Modified IIUsed P_ti_ring Training Program D J IQuestion~~iource Commen!sJ 1153928 used on Salem SRO NRC exam more than 2 exams ago. Modified to add actions required. I iCorr.... y'" I ! I I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question Topic-I ISRO 6 Given the following condition: Both Unit 1 and Unit 2 control rooms have been evacuated due to a toxic gas release on site. When performing local actions to stabilize the plant:
1) Where will PZR level be determined

2) How will PZR level be controlled ra,: 11) At the Hot Shutdown Panel 213.
~ 2) By ensuring the Charging System Master Flow Controller is controlling PZR level on program. ~ 11) At the Charging Pumps Flow and Pressure Panel 216-2.
2) By ensuring the Charging System Master Flow Controller is controlling PZR level on program.
le. i 11) At the Hot Shutdown Panel 213.
~* 2) By establishing local control of the CV55 CHARGING FLOW CONTROL VLV to maintain PZR level 22%-77%.
[dl 11) At the Charging Pumps Flow and Pressure Panel 216-2. L~. 2) By establishing local control of the CV55 CHARGING FLOW CONTROL VLV to maintain PZR level 22%-77%. iAnswe~ Ic I 1exam Level i Is I @_09niiive Level 11 Memory I [Fa~ility: [ j Salem 1 & 2 I ~amDate: i I 12/21/20151 I [KA: 1 oooo6aA2o7 I~AA2.07 JIRova1ue:-il 4.1![Rova1ue) 4.3l[S!CtiO.nJl~IRoGroup:[I 21[SFio~I 21 !IE ~ [SYStemTE:volution Title_j j_c_o_n_tr_o_IR_o_o_m_E_v_ac_u_a_ti_on_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___..1 1~-- ] ~~tement:: Abilit to determine and inter ret the followin as the a pl to Control Room Evacuation: PZR level ]Explanation of\ 55.43.b(S) This question m_eets SRO o~ly criteria listed in NURE?-1021, ES-401, ~ttachm_ent 2, _11.E. Figure 2, 1~t bullet for 5th 1Answers: ___J Block for assessment and 1mplementat1on of a procedure or section of procedure 1n knowing which procedure will be used to
- evacuate the control room, and how the RO will maintain PZR level, and where it will be indicated. AB.CR-1 contains field actions which will be directed by the CRS. AB.CR-3 for toxic gas directs the control room evacuation to occur using AB.CR-1, but does not . . .
I ---- . Reference Title 1 Facility ~eference Number
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i [Reference Section [i Pagl:!~ !Revisioni IControl Room Evacuation 11 S2.0P-AB.CR-0001 I II 1122 I IControl Room Habitability 11 SC.OP-AB.CR-0003 I ii ii 6 I I II I 11 II I !Lo. Number __J Objectives IABCR01 E003 j_ ____. [Material Required for Examination 11 q !Question Source] New I I[Question Modification Method: JI I [Used During Training Program DI 1 Question__~C>.~ice Comment~ I I IC011 .... ~*** I I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes IQuestion Topic] ISRO 7 I Given the following conditions:
- A SBLOCA has occurred on Unit 2 outside containment. - Actions of EOP-LOCA-6, LOCA OUTSIDE CONTAINMENT, have failed to isolate the break. - RCS pressure is 1440 psig and continues to lower.
Which of the following identifies the procedure that will be used upon transition from EOP-LOCA-6 and the actions that will be directed in that procedure after the transition?
~ EOP-LOCA-1 Loss of Reactor Coolant. Add makeup to RWST, initiate a cooldown, and minimize injection flow.
I
~ EOP-LOCA-5 Loss of Emergency Coolant Recirculation. Add makeup to RWST, initiate a cooldown, and minimize injection flow.
Fl -- _ -1, Loss of Reactor Coolant. Check for subsequent failure and conserve makeup inventory. id i EOP-LOCA-5, Loss of Emergency Coolant Recirculation. Check for subsequent failure and conserve makeup inventory. L_"_ ~nswe_r 11 b l ~am Leven I s I rcognitive Level *IApplication I :_F~cilit~: 11 Salem 1 & 2 I fE:xamDate: 11 12/21/20151 iKA:ll OOWE04A201 I:EA2:-1=~=J IRo Value: II 3.4!:SRO Value)~ 'Section: 1_§.~.§_ _JRO Group:il 11 SRO Group:il 1 1 11 li:~!,i ~ IKA Statement:* !Explanation of j 1Answe~__1 =====--~~e~~e_!i_!l_E!______________ Ji Facility Reference Numbe!:::J Reference Section _ J [~~!!_':_~~ ~~ ILOSS OF EMERGENCY COOLANT RECIRCU 112-EOP-LOCA-5 I I !::::12=5== ILOCA Outside Containment Ij 2-EOP-LOCA-6 I ! I21
~~~~~~~--'*~~~~~~~~~__..! 1~1===
IL.0. Number 008 !Material Required for Examination 11
'I I
!Question Source: 11 Previous 2 NRC Exams !Question Modification Method: i Direct From Source 1lused During Training Program I D
;Question Source Comm~ 113-01 NRC SRO exam (Dec. 2014) --~~~*~-*~~~~~*--
I 1Comment I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [auestion Top~ ISRO 8 Given the following conditions:
- Unit 1 has experienced a Main Steam line Break (MSLB) inside containment from 100% power. - Main Steam Line Isolation has failed to close any MS167. - Safety Injection was manually initiated, with all components operating as expected. - 11 AFP is err. - 12 and 13 AFP's tripped after starting. - Reactor Coolant System pressure is 1100 psig. - All Reactor Coolant Pumps have been tripped. - Containment pressure is 16 psig and rising. - All Wide Range (WR) Steam Generator (SG) levels are 35% and dropping. - All Steam Generator pressures are 465 psig and dropping. - Reactor Coolant System Tc's have dropped from 540 to 438°F in 40 minutes.
If these condition were present when transitioning out of 1-EOP-TRIP-1, Reactor Trip Response, which procedure must be entered and what action must be taken upon that transition?
~ 11-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions. Reset Safeguards Actuation and restore normal charging and letdown. ~b~j , 1-EOP-FRHS-1, Response to Loss of Secondary Heat Sink. Initiate feed and bleed ONLY when 3/4 SG WR levels have dropped <32%.
ic.: 11-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions. Shut all MS10's and steam dump valves .
.... ....... ............ ... ...
[ ] 11-EOP-FRHS-1, Response to Loss of Secondary Heat Sink. Initiate feed and bleed immediately. ]An~~lll/~~ ~ ~ll'l L~~I_ '~ I !Cognitive Lev!l_J IApplication I iFacmtYJ ISalem 1 & 2 I 'ExamDa~i:_:J I
~
IS 12/21/20151 ~JI oowEo5A202 I[EA2.2 - -1 IRo value: J[EJ[sRo va1u~~~H~ [RociroUP:lLJ[Sifo-G"roup:] LJ Ill ~ ~-Y~!em/Evolution 1:i~1eJ ILoss of Secondary Heat Sink I [Eo§"~ [KA Statement: 11 Ability to determine and interoret the followina as they apply to Loss of Secondary Heat Sink: I * ..
*- - procedures and operation within the limitations in the facilitv's license and amendments.
1Explanation of 155.43(5) 55.43.b(5) This question is SRO level based on having to asses the conditions given in stem and determine both what IAnswer_s: procedure will be entered and the action(s) required in that procedure, Dis correct because the conditions given in stem would transition to FRHS-1 due to a RED path of no AFW flow and <9% NR level. The Bleed and Feed initiation criteria are when SIG WR levels are <36%(adverse), NOT 32% (normal) as in distracter B. Distracters A & Care incorrect because it is a lower priority oc:n .......&.t... ... ._ _ _ ._ if 1 C" --"=-- ir ______ ... .C-- ... .__ -----...J,,,.n
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Reference Title *1 [ __Facility Reference-f>,lllmber I [Refere_n~e section ! [Page No:_l JRevisionj IResponse to Loss of Secondary Heat Sink 111-EOP-FRHS-1 II ii lj 21 I IResponse to Imminent Pressurized Thermal Sh 111-EOP-FRTS-1 II II 1122 i ICritical Safety Function Status Trees 111-EOP-CFST-1 II q 1122 I l':.:0. Number Objectives I FRHSOOE005 I FRHSOOE013 j_ ___,
IMateria!_Required for Examination 11 II §uestion So~rc~LJ lFacility Exam Bank I[Question Modification ~_:thod: JDirect From Source I!Used D_u.r:~~ Training Program ! D ~E)stion Source C~~-~ 143189, Used on Salem SRO NRC exam more than 2 exams ago. I I Comment I I I I I I I
RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
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l§:i.les.E?~!~~J 1 _s_R_o~9~------------------------------------------~ Given the following conditions:
- Unit 2 is responding to a degraded core cooling condition in accordance with 2-EOP-FRCC-2, "Response to Degraded Core Cooling." - After depressurizing to inject the accumulators, the STA reports a RED priority on the Thermal Shock Status Tree, and recommends transitioning to 2-EOP-FRTS-1, "Response to Imminent Pressurized Thermal Shock."
Which actions describe the o erator response?
~ Do NOT implement 2-EOP-FRTS-1 until 2-EOP-FRCC-2 is completed because thermal shock is a lower priority CFST.
P-FRTS-1 immediately because the potential damage done to the RPV by delaying entry into FRTS after 1ent:rv 1cor1dition1s <:ire I arable.
~ Do NOT implement 2-EOP-FRTS-1 until 2-EOP-FRCC-2 is completed because while in FRTS the core will continue to boil a*wa*v in1iectedl I accumulator water, and could lead to a RED path for Core Cooling.
f d:! Implement 2-EOP-FRTS-1 immediately because it is a higher priority CFST and rules of usage stipulate the transition to a higher priority procedure always takes precedence over notes and cautions. [Answer I~ rexam*LeVel Is I [Cognitive Level __ IMemory I [Facility_:_i ISalem 1 & 2 I ;~x~J I 12/21/20151 ~I oowE06G423 I[2~4-:-23 *u -~~-~!!~.!:J [ ] ] i s Ro Valuej~ [seciiOO:l I~ RO Group:[[] ~~~_!>3 [ ] [hstem/Evolution Title j j_D_e""'g_ra_d_e_d_C_o_re_c_o_ol_in..;;:g'--------------------------------' Knowled e of the bases for rioritizin ,Explanation of j 55.43.b(5) This question is SRO level based on knowledge of diagnostic steps and decision points in EOP that involve transition !Answers: , points to event specific emergency contingency procedures. Knowledge of the bases for when NOT to implement a RED priority CFST even when indications are present is just as importanct as knowing when to implement. in this case to preserve and inventory in the reactor pressure vessel. Stopping the depressurization to go to FRTS would cause the cooldown to be stopped, and a
. . . . -- -.
Eventually, CETs and/or RVLIS level values could exist which would require transition to FRCC-1 via a RED path CFST. The
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Isto pin of the cooldown could lead to a de raded core coolin condition to deteriorate to an inadequate core coolin condition. Reference Title i[ Faci_litx_Re!e~~~ce-Number I ;Reference Section J [f~J ~~ IResponse to Degraded Core Cooling
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112-EOP-FRCC-2
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l~-----------------'1----------------' --------~ ------'1---~ 11..:0. Number~ Objectives I FRCCOOE006 '--~ !Material Required for Examinat~.'_"! _ _j I !Question Source: i IFacility Exam Bank I!Question Modificati~n Method: sed During Training ~~~!-~rrl__J D [auestion source Comments:
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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question_jii>ICJ !SRO 10 I Given the following conditions:
- Salem Unit 2 tripped from 100% when a loss of off-site power occurred. - All systems respond as expected, and operators progress normally through the EOP's. - It has been determined a rapid natural circulation cooldown will be performed.
Which of the following describes procedures/actions which are REQUIRED to be performed PRIOR to transitioning to either 2-EOP-TRIP-5, Natural Circulation Rapid Cooldown Without RVLIS, or 2-EOP-TRIP-6, Natural Circulation Rapid Cooldown With RVLIS, and the bases for them?
~ 12-EOP-TRIP-2, Rx Trip Response, must be performed to ensure adequate SDM and natural circulation have been established, containment
cooling remains in service, and equipment not needed for the cooldown has been secured. I
~ 12-EOP-TRIP-4, Natural Circulation Cooldown, must be performed to ensure adequate SDM and upper head cooling have been established, SI signals have been blocked, and initial cooldown/depressurization have been performed. I ~ 12-EOP-TRIP-4, SI Initiation Criteria step must be performed to ensure SI will not be required prior to blocking SI before a cooldown can be initiated.
[ ] 12-EOP-TRIP-2, SI Initiation Criteria step must be performed to ensure SI will not be required prior to blocking SI before a cooldown initiated. ~~swe-r] Ib I rexam Level 1 Is I !cognitive Level 11 Memory I ~ciiityl ISalem 1 & 2 l~xa~~I 12/21/20151
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~I oowE10A202 I~~-__]~ Vatue]~!SRo Valuef~ [sectio!l:JI~ IRo Group::LJ~~LJ ~§'I' ~ lsystem/EvolUtiOr!Title-] INatural Circulation with Steam Void in Vessel with/without RVLIS I i-E:1°==J [1-_<A Statem-ent:] o determine and inter ret the followin as the a I to Natural Circulation with Steam Void in Vessel with/without RVLIS: Explanation of :
Answers: i
_ ~~ference Title --~] ~ittty-Refen~~ce Numb~~] [Retere~ce ~~ction 1 [F>_~~ [R.evi5i0n1 INatural Circulation Rapid Cooldown with RVLIS 112-EOP-TRIP-6 !=======:::::::::!i.==== 2=3==:::. i.::j INatural Circulation Rapid Cooldown without RVLll 2-EOP-TRIP-5 I23 !~~~~~~~~~~~~~~~~~--'!-~~~~~~~~~~~--' ~~~~~~~~~ -~~--'~, ::::::::::::::: [L..o. Num~~r ~ Objectives TRP004E005 I I I iMaterial Required for Examinati~ I II e~:stion Source: J IFacility Exam Bank I[Question Modification Method: _ ~ Direct From Source I!Used Duri~g Training Program I D !Question S~~~ce Comments! 1153885 I I !--******-*" -- I I I I I I I
RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
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1auestion Topic I _s_R_o_1_1_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___. Given the following conditions:
- Unit 2 is operating at 75% power. - 21 Waste Holdup Tank (WHUT) is being processed via the Portable Liquid Radwaste Processing System. - Console alarm SURGE TANK LEVEL HI-LO for the CC system alarms. - Surge tank levels are 59.3% and 59.2% on Channels A and B, and rising slowly. - No other alarms are present.
Which of the following describes the effect if Surge Tank Level were to continue to rise, and what action the control room will take in response to the risin level? 2CC149 CC SURGE TANK VENT VLV will auto close to prevent overflow. Stop processing 21 WHUT as directed in the S2.0P-AR.ZZ-0011, Control Console 2CC1. Overflow of the Surge Tank will contaminate the Waste Holdup System with chromates. Close 2CC149 CC SURGE TANK VENT VLV 2CC2 as directed in the S2.0P-AR.ZZ-0011, Control Console 2CC1. 9 CC SURGE TANK VENT VLV will auto close to prevent overflow. Open the CC Surge Tank Drain Valve from 2CC1 asnec:es:sary tc1 I evel <100% as directed in S2.0P-AB.CC-0001, Component Cooling Abnormality. Overflow of the Surge Tank will contaminate the Waste Holdup System with chromates. Direct a NEO to locally drain the Surge Tank to 55 gallon drum as necessary to maintain Surge Tank Level <100% as directed in S2.0P-AB.CC-0001, Component Cooling Abnormality. [A~ ~ iEXM1 Level ] @=:] cognitive Level~: IApplication I I [~C:ili~~:_J ISalem 1 & 2 I IExamDatej I ~I 008000A202 IiA2.02 I [Rova!Ue:]@~Ro vaiuej~ !section: ii~ [Ri:[Grc>~P:1LJ ~~!~~LJ
system/Evolution Title I I Component Cooling Water System
KA Statement: I Ability to (a) predict the impacts of the following on the Component Cooling Water System and (b) based on those predictions, use es to correct, control, or mitigate the conseauences of those abnormal operation:
Hiah/low surae tank level !Explanation of I 55.43.b(5) This question is SRO level based on having to asses the conditions given in stem and determine which action in which Answers: procedure will be taken. The 2CC149 auto closes on hi radiation in the CCW systen, nnot high pressure, to prevent radiation release from the CCW system. The Surge Tank will be locally drained. The overflow from the tank will go to the Waste Holdup system and contaminate the WHUT. WHUT processing will be stopped in both the AB.CC and ARP procedures. Locally draining
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Reference Title rn I lReference Section I u-*-~-*---* Component Cooling Abnormality 11 S2.0P-AB.CC-0001 II 11 1114 I I Control Console 2CC1 11 S2.0P-AR.ZZ-0011 11 11 1160 I 1205331 11205331 Sheet 1 II 11 I 154 I [Lo. Number Objectives CCWOOOE008 IMaterial Required for Examination
~- JI II ~~~stion Source: [ j Facility Exam Bank I~uestion Modification Method: II Concept Used I[Used During Training Program I ~ ~~~~~~~Source Comments[ 182785 modified to make SRO level and fit KA.
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RO Skyscraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I
@luestion Topic I ISRO 12 I Given the following conditions: - Unit 2 is operating at 100% power. - 2PR1 fails open and remains open.
Which of the following identifies how this affects the PZR Master Pressure Controller (MPC) response, and what consequences, if any, are associated with the actions performed by the crew IAW S2.0P-AB.PZR-0001, Pressurizer Pressure Malfunction?
~I MPC output will RISE. A unit shutdown will be required if 2PR1 cannot be restored to operable status within 72 hours.
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~ MPC output will LOWER. A unit shutdown will be required if 2PR1 cannot be restored to operable status within 72 hours.
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~ MPC output will RISE. The unit may continue to operate indefinitely after the initial mitigative actions are completed.
I I
~ MPC output will LOWER. The unit may continue to operate indefinitely after the initial mitigative actions are completed.
I ~nswer J Ib I IExam Level 11 S I 1cognitive Level JIApplication I 11 Facility: ; Salem 1 & 2 I IExam Date: 11 12/21/20151 [~I 01 OOOOA203 ~--------*1
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[~ystem/Evolution Title 1.;..;..:...:..:..:;;,,;.;:.:.;_;_;_;.;;,,;,.;;.;..;_..:..;,...;.;..:;.......:,.;::..:...;.;,;.;.;__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___J 1010 i - .--------------, ....------------------------------------------------: KA Statement: I Ability to (a) predict the impacts of the following on the Pressurizer Pressure Control System and (b) based on those predictions, use rocedures to correct, control, or miti ate the consequences of those abnormal operation: PORV failures Explanation of 55.43(2) This question is SRO level because of the Tech Spec knowledge required, and what actions TS directs for different PORV 1 Answers: 1 malfunctions. Additionally, while the question doesn't specifically ask what procedure to use (too easy for AB.PZR), it does require knowledge of the actions IN that procedure. The MPC raises output when actual pressure rises, and lowers as actual pressure lowers. As pressure lowers due to the open PORV, the output will lower to turn on heaters and close spray valves. When the PORV
. . . . . .
is required. A PORV isolation that DOESN'T require shutdown if not fixed is aleaking PORV, which is isolated by its Block Valve I with ower maintained to the Block valve. [L.~o. Number Objectives I ~1E002 LE010 I Material Required for Examinatio_n ___J I II !Question Source: I I Facility Exam Bank 1IOuestion Modification Method: II Direct From Source 1lused During Training Program I D !Question Source Comment~ 1125725 used on Salem SRO NRC exam more than 2 exams ago I !Comment l I I I
RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I r*--------- ----- -* ISRO 13
;Question Topic; I Given the following conditions: - Unit 2 is conducting a rapid plant shutdown due to a loss of condenser vacuum. - Two IRPls in control bank D go dark, and their P-250 readings are O steps. - Maintenance Controls Dept. will be unable to repair the IRPls until after shutdown. - Rx Engineering has made no specific recommendations outside of procedural direction regarding boration.
Which action is required when the plant is placed in Hot Standby?
~I Borate to Cold Shutdown SOM to prevent a Yellow Path on FRSM from occurring during the shutdown.
I
~I Borate an additional 540 ppm to prevent a Yellow Path on FRSM from occurring during the shutdown.
I I
~ Borate an additional 540 ppm since it is assumed that the reactivity associated with the affected rods is unavailable for shutdown. @] IBorate to Cold Shutdown SOM since it is assumed that the reactivity associated with the affected rods is unavailable for shutdown. I !
~nswe~ Ic I I Exam Level 11 s I IC?9_ni~f/eiev~ I Application I fFacilitY:l ISalem 1 & 2 I [Exa_rnl)ate:] I 12/21/201511 ~I 014000A202 I[A2m--Hl ~~!~~J [TI" is Ro Valuej~ 'section: 11~ IRO Group:1 LI jsRO Group:I LI I& li'l !system/Evolution Title] IRod Position Indication System I r014- ---] l~ S~~mef'l!:J ~ity to (a) predict the impacts of the following on the Rod Position Indication System and (b) based on those predictions, use cedures to correct, control, or mitigate the consequences of those abnormal operation: Loss of oower to the RPIS lExplanation of j 55.43.b(6,5) This question is SRO level based on having to determine the effect on core reactivity based on the 2 control rods which !~~~ers:___ _J have to be assumed to remain fully withdrawn in the absence of IRPI indication. Additionally, the sRO must select the portion of AB.ROD-4 (CAS action 2.0) which requires an additional 270 ppm boration for each failed IRPI if a shutdown is performed before
!
the IRPI is declared operable. Boration to cold shutdown conditions is not required
. ' :+h ~u~~*~'"' -~~t~\ :~1~-~ +h~-~h* * ,__ for a shutdown to hot standby. It assumes the ,; ,_ ~~-~ I i Reference Title - __J L Facility Reference Nu.mber _ ~eference ~~(;~i?':_l J --~,c=~~~~()J §.evi~()_l"lj IRod Position Indication- Failure LO**-*-*- .,,._, *-
11 S2.0P-AB.ROD-0004 I ii lj 10 I I II I 11 11 I I II I 11 11 I !Lo. Number .......... "*-*-*-****-*--*-*----- Objectives ! ABROD4E002 ,_ _ ___. !Material Required for Examination :I II iQuestion Source: 1 IFacility Exam Bank I!Question Modification Method: i 1 lused During Training Program I D Question Sourc~~~~m4:_~~ 1140829 1 I IC"'"'**v'" I I I I I I I
!Material Required for Examination I I I I IQuest~?~ Source: i New IlCl~~estion Modifi~~tion Method: i I lUsed During Training Program D I I
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[Ques~!?n source comme~§ I i.--*<111**-llL l I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic 11 SRO 16 I Given the following conditions:
- Unit 1 was operating at 100% power when a catastrophic failure of 12 SG steam piping occurred inside containment. - The Rx tripped and SI initiated. - 12 SG rapidly blew down to containment, and containment pressure peaked at 16 psig prior to transition out of EOP-TRIP-1 Reactor Trip or Safety Injection to EOP-LOSC-1, Loss of Secondary Coolant. - Operators have just transitioned out of LOSC-1 to EOP-LOCA-1, Loss of Reactor Coolant, with the following conditions: - RCS pressure is 1780 psig and rising slowly. - RCS subcooling is 100°F. - PZR level is 22% and stable. - 11, 13, and 14 SG levels are 17% and rising slowly.
Which of the following describes containment cooling operation during subsequent EOP performance? Containment Sorav will be secured in ......
!a.J i --~' 4 , Post LOCA Cooldown and Depressurization, when containment pressure is < 4 psig. Containment Spray Additive Tank will be ~- * *- ... based on normal radiation levels in containment.
[b:1 I EOP-TRIP-3, SI Termination, when containment pressure is< 4 psig. Containment Spray Additive Tank will be isolated based on normal radiation levels in containment.
~ I EOP-TRIP-3, SI Termination if containment pressure is less than 13 psig. CFCUs will continue to operate in Low Speed.
[11 LOCA-2, Post LOCA Cooldown and Depressurization. CFCUs will continue to operate in Low Speed.
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I iFaciiitY:l ISalem 1 & 2
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[Ans~~ I@=] ~ognitive Level I !Application I JExamDate: 11 _ _ _1_21_2_11_2_0_15_,I
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rexam Level ~1026000A208 I A2.08 _i~~~0]"~-~~-~-22l~!~~~]~~<?~~~~LJ~~LJ ~ [fyStiffiiEvolutioi:i_Title 1 l_c_o_n_ta_in_m_e_n_t_S_p_ra~y_S_y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~ IKA Stateme_1_1t_:J 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 mitii:iate the consequences of those abnormal operation: Safe securing of containment spray when it can be done) !Explanation of i 55.43.b(5) This question is SRO level based on having to asses the conditions given in stem and determine both what procedure ,Answ~!l>~ will be entered and how containment cooling/spray equipment will be operated. With conditions in stem, the transition out of LOCA-1 to TRIP-3 will be made at step 9 based on adequate subcooling, SG NR level status, and PZR level, otherwise the transition would be made to LOCA-2 at Step 18 with RCS pressure >420 psig. Both procedures have CS terminated. Neither procedure I
..... :+h ........ f"C::: C" .......... , A ....1....1:+;, ,,... .a.,...,...1, ~ ....... 1.... .&.: ....... .* i...~ .... i.... ie nl~ ..... :i....1 .... ,...;,, ........ +h ... 4. 4.L.. .... ............. .&. ic- <:::1o .... .a. ... -.m 1 ...... 1 1e ~ I r'\('A nr f""t:'f"I I operation, which would. be. goverened by the System operati.ng Procedure when EOP network weas exited to the Integrated operating Procedure (IOP) c Reference Title -- =:Ji Facility Reference Numb~ iReferenc~ Sectio0 ~..?J !Revisionj I SI Termination II EOP-TRIP-3 I II 1125 I j Post LOCA Cooldown and Depressurization II I 11 1125 I I II I 11 11 I
[Lo. Num~er Objectives I CSPRAYE012 I TRP003E005 ,__ ___,
iMaterial Required for Examination ~1 'I /Question Source: 11 New
~~., ... " *-*~**J I [Question Modification Method:
L--~~~--- **-~~~--~*--** JI I [used ~ming Training_j:>~ogram ] D I-Question ~~urce Commen~~ I I
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l'-OJT,..,wo* -i I I I I ! I
RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
!Question Topic 11SRO17 I Given the following conditions: - Unit 2 is in Mode 5 entering a refueling outage. - 2C EOG was CIT yesterday for scheduled outage window. - The normal 31 day surveillance test of 2A EOG was completed SAT 2 days ago while in Mode 4. - Subsequently, the Unit 2 CRS notices that the 18 month Hot Restart surveillance for 2A EOG, was NOT performed as scheduled after the normal 2A EOG 31 day run surveillance, and is now outside its required periodicity including any allowable grace time.
Which of the following describes: I
1) The status of 2A EOG

2) When the associated Hot Restart test must be oerformed I
~ Operable. The Hot Restart test must be performed prior to entering Mode 4.
I
~I Operable. The Hot Restart test must be performed prior to entering Mode 6.
I I
~ Inoperable. The Hot Restart test must be performed prior to entering Mode 4.
I [dl I Inoperable. The Hot Restart test must be performed prior to entering Mode 6. I I *-- I IAAswer I' ~ LExam Level @=] ~ve Level l j Application j I [Facility: 11 Salem 1 & 2 I ~xamDate: J l___1_2_12_1_12_0_15_.I jKA:il 064000G120 I ~~-o-~I !Ro Value]~~~@~ [section: 11~ 'ROi3fOOP:10JsRo Group~O ~ [SYstem/Evolution Titl~ IEmergency Diesel Generators [KA Statement: I Abilitv to interoret and execute procedure steps. I \Explanation of I 55.43.b(2) This question is SRO level based on being able to apply TSAS 3.8.1.2 for electrical power in Modes 5 and 6, and the
Answer~ action required based on the conditions in the stem. RO knowledge would be "above the LCO line", and SRO level for actions below the line. LCO 3.8.1.2 requires 2 operable EDGs, and the stem states 2C is already tagged out. The surveillance requirements of 4.8.1.2 specifically state that certain surveillances are NOT required to maintain EOG operability. The bases for
.i.i...~ .... ic th".lt ...... ....1 ........... *f. ............... tn .a.i....,.. c:nr::? n ............... 1-11orl . .:.i.i... ... L..-. .... u .... :.&. .... - - ................. "'- ,,__ , l"\r "- ......... * ,______ ... ,_ ,.., *-'--
performance of the surveillance requirement, and to preclude de~energizing a required ESF bus or disconnecting a required offsite circuit during performance of surveillance requirements ...... It is the intent that these surveillance requirements must still be capable of being met, but actual performance is not required during periods when the DG and the offsite circuit are required to be operable. During startup, prior to entering Mode 4, the surveillance requirements are required to be completed if the surveillance frequency has been exceeded .... " The provided reference does not give answer, but allows the operator to interpret the surveillance requirement, and in any case does not contain the action required which is located in the bases. c:=-- ._. Reference Title
- ==i: Facility Referen_ce Number _J jReference--Section -11 Page No~ IRevi~!C>~
ISalem Tech Specs I lj 3.8.1.2 I13/4 8-7a i 1245 I ISalem Tech Specs I 11 Bases 11 B3/4 8-3 l I282 I I I n 11 II i iL.O. Number Objectives I EDGOOOE011 1-~
~terial Required for Examinatio.n j j SRO 017 Tech Spec 3/4.8.1 Electrical Power Systems, A.C Sources II l~uestion S_<>_urceJ INew I~estion Mod_ificationMethod: *1 I !Used During _Training Program I D ~ ~--~------ntS] stion Source _c~mments I I 1r~~---* ---111***-** .. I I I I
RO SkyScraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I lauestion Topic] ISRO 18 Given the following conditions:
- Unit 2 is operating at 60% power, steady state. - 21A Circulator is CIT. - The Condensate Polisher is O/S and in standby. - Subsequently, the 13 KV South ringbus breaker D-E opens causing a loss of 23 SPT.
Which of the following describes:
1) The plant status 15 minutes after the loss of 23 SPT

2) How the operators will respond to this event
~ 1) The reactor remains at power.
2) Operators will be performing S2.0P-AB.CW-0001, Circulating Water System Malfunction, and establishing Low Pressure Turbine Hood Sprav.
~ 1) The reactor was manually tripped.
2) Operators will be utilizing S2.0P-AB.COND-0001, Loss of Condenser Vacuum, in conjunction with the TRIP series of EOP's, to break condenser vacuum.
~] 1) The reactor remains at power.
2) Operators will be performing S2.0P-AB.CHEM-0001, Abnormal Secondary Chemistry, and placing the Condensate Polisher in service for the expected rise in Dissolved 02.
Jd.
1) The reactor was manually tripped.
~**--J - ,,... - ""'" * .
I
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Ill
-* ..... - * .... .... . - ~ - .. - -
lAnswer 1
~ !exam Levei"I @=]" lcognitiveievel 11 Application I !Facility-:-] ISalem 1 & 2 I jExamDate: i I 12/21/20151 IKA:jlo7soooA202 l~~~~-*[Ro-vaiue:jl 2.s1[SRO-va1ue)-22JlsectfoniJl~[R§"Group:]I 21:sR0Group~I 21 l~~tJJ ~
[~ystem/Evolution Title I j_c_ir_c_u_la_tin.....;g:;.._W_a_te_r_S-'y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.I !075 iKA ~tatement: I Ability to (a) predict the impacts of the following on the Circulating Water System and (b) based on those predictions, use gcedures to correct, control, or mitiqate the consequences of those abnormal operation: ss of circulatinq water pumps lExplanation of] 55.43.b(5) This question is SRO level based on knowing the appropriate procedure or section of procedure used in applying the Answers: conditions which will occur based on the initial conditions in the stem. 23 SPT powers 23 CW bus, and 24 SPT powers 24 CW bus. --~-~-.-- 23 CW bus powers the "A" circulatros and 24 SPT powers the "B" circulators. The loss of 23 SPT will cause a loss of 2 additional circulators in addition to 21A which is CIT. The operator must know how many circulators are now not running, 3 or 4. If it is
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thought condenser vacuum would degrade or DO increased based on losing hotwell levels and its effect on condensate pump seals. Manually bypassing the Turbine Hood Spray is allowed (and directed) in AB.CW-1, even though normally it remains secured
> 15% power. (AB.CW-1 paqe 5)
C___ Reference Title I L Facility Reference _Number i ~eference section ] !f>-a9e No.1 ~~J ICirculating Water System Malfunction 11 S2.0P-AB.CW-0001 II qs I 135 I I II II ii II I I II II 11 11 I IL.O. Number Objectives I ABCW01 E004 1--~
lMaterial Required for Examination 11 II I IQues~()n Source: I New 1IQuesti~n Modifica!i~n Method: J I !Used During Training Program_! D [Qu~~tion source CommeniSl I I !c_...... _... . I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic 1 ISRO 19 Given the following conditions:
- Unit 2 is performing a reactor startup by control rods IAW S2.0P-IO.ZZ-0003, Hot Standby to Minimum Load. - Estimated Critical Conditions are: - Cb= 500 ppm - Control Bank D = 77 steps - Xe free - 11,756 EFPH When the ICRR value reaches 0.125, the Predicted Critical Rod Height is 122 steps on Control Bank D.
What action(s), if any, is/are required to be taken in response to this Predicted Critical Rod Height? I
~ Continue with no special actions are required.
I
~ Continue the reactor startup, and evaluate the post startup data for trend.
I [~ Initiate rapid boration, insert Control Rod Banks, and recalculate the ECC.
@] I Insert the Control Rod Banks and recalculate the ECC prior to withdrawing Control Rods. .. ..,
I ~~~ ISalem 1 & 2 !___12_12_1_12_0_15_.I
.
[Answ__~ ~ IExam Level J [I:J !Cognitive Level r 11 Application I !Exam Date: J ~I 194001G123 I E~...=:J ~vaiU-e]@sRo value)~ [See!ion:]j~ IRo Group:ILJ~~LJ ~ !system/Evolution Title I ~emelili] ----*-*- Ability to perform specific system and inteqrated plant procedures during all modes of plant operation. Iexplanation of: 55.43.b(6) Table 1-8 77 steps is 1079.4 pcm. 122 steps is 877.0 difference is 202.4 pcm. With <300 pcm difference between ECC jAnswers: ___ __/ and predicted at the eightfold position, there is no action, and the startup will continue with no additional action required. [ Reference "fitl~ --*----
Facility Reference Number I !Reference Section 11 Page NoJ §:eviSioi1:
ICurve Book 11 S2.RE-RA.ZZ-0016 I 11 11 8 I IHot Standby to Minimum Load I 1 S2.0P-IO.ZZ-0003 I 11 11 I I II I ii 11 I Objecti j_ ___, ]Material Required for Examination 11 SRO 19 S2.RE-RA.ZZ-0016 Rev. 8 II I !Question SourcE!:l Facility Exam Bank I[Question Modification Method: II Concept Used 1lused During Training Program I D [auestion Source Comments[ 1151642 Used 3 NRC exams ago Sept 2011. Different numbers based on different revision of Curve Book. I Icomment I I I I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Question Topic J 1SRO 20 Choose the answer which contains ONLY actions directed to be performed IAW S2.0P-IO.ZZ-0007, Cold Shutdown to Refueling, BEFORE Rx Vessel Head detensioning would be initiated for a refueling outage starting in November. I. The Rx shall be subcritical for at least 168 hours. II. Direct. continuous communication between the control room and refuel floor is established. Ill. The RCS is drained to <104' elevation. IV. Unit CRS AND SM approval.
~1 ~1111. IV. ~,I.Ill. @Jiii, iV.
iAnswer 11 b I [Exam Level i Is I [cognitive Level *11 Memory I ~~ci~!ty: J ISalem 1 & 2 I ~ll"ID~te: ~ I 12/21/20151 iKA:[l 194001G141 !f2.1.41 I !Ro Value: q 2.81 iSRO Value) I
3. 7 !section: 11 PWG I iROCfrou~ I 11 lsRo Group: II 11 lrfl D lsystem1Evotutionrit1e] I I 1GENERI I IKA Statement: I Knowledge of the refueling process. I
]Explanation of 1 55.43(6,2) This question is SRO level based on knowing the requirements which must be met before alterations affecting core Answers: configuration will be started entering a refueling outage. The reactor does not have to be subcritical for 168 hours (Oct 15-May 15th) prior to moving fuel in the reactor (TSAS 3.9.3) Direct communications is required during CORE ALTS, and detensioning the head is NOT core alts. I '----
*--*--*-**----*-
Reference Title ] ~_Faciiity Reference Number . J[R"it;;~;nce section 11 Page No.J !R.eviSion! ICold Shutdown to Refueling 11 S2.0P-IO.ZZ-0007 I II lj 11 I I II I 11 11 I I II I 11 II I 1L.O. Nu111ber Objectives IIOP007E002 '--~ Material Required for Examination l I II I Question Source: 11 Facility Exam Bank I!Question Modification Method: II Editorially Modified I!Used During Training Program I D
!Question-~~!~.: Commentsj I I -.-~
[comment .. I I I I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I
~estion Topicl I SRO 21 I Given the following conditions: - Unit 2 is in MODE 6 with core reload in progress. - 10 fuel assemblies have been moved into the Rx. - Rx cavity level is 26' above the RPV flange. - 21 RHR loop is in service in Shutdown Cooling. - 22 RHR loop is O/S and available.
Which one of the followinq would prevent continuation of fuel movement into the reactor?
~ I Loss of Control Air to containment.
I I
~ Racking down the 22 RHR pump 4KV breaker.
I
;c:f I Both 100' elevation containment airlock doors are opened. ~I With both SRNls operable, only ONE is capable of providing audible indication in the control room.
~~-~w~ Ia I !Exam Level 11 s I ~?gnitive Level I i Application I Facility: I J 1Salem 1 & 2 I rexamDate: i I 12/21/20151
*-- -- ..
!KA:Ji 194001G201 I!_~-~-- ROValue:il
--*"-
4.5ltSRO~~~iSectiorliJIPWG liROGroup~j 1jjSROGroup:JI 11 1111 ~ jSystem/Evolution Title '.
......-----------------------------------------,. --*-------
1 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' IGENERI I ~temerlt] *-*-
lily to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that uld affect reactivity.
rplanation of Answers: l 55.43(7, 6) This question is SRO level because of the knowledge require for fuel handling procedures, and the ability to continuously apply that knowledge when operating the fuel handling equipment. The requirement for SRNl's is BOTH operable and providing VISUAL indication in the Control room, with ONE providing AUDIBLE indication in the control room. The manipulator crane is air powered for gripping, so the loss of air to containment would preclude being able to perform core alts. Only ONE RHR
!--- iC' ""--* oi--..J f..-. 1-- in - '-- ;n ~Al""\r"lt:: " ') C _on , ____ - * - * - - *'*--' +n ho f"\Ct::C A r">I t:: **"'--- <>'>.' 1...... 1 -'--**-*'--fl---- '
.~--* ---- Reference Title --1[ Facility Refe~ence-Number J IRefere~ce Section 1 lPage No] ~<:>-'!
----*--**
IRefueling Operations
c. -
11 S2.0P-SO.SF-0009 I 11 1118 I IReac Pene Area & Cont Control Air 11205347-1,3 I 11 1142,36 I I II l 11 11 I 1 L.o. Number Objectives IREFUELE007 I IOP009E002 '--~ Material L......__.. . Required for- Examination iI I II [Question I Sourc~ Facility Exam Bank I!Question Modification Method: JI Editorially Modified !tUsed During Training Progra.tll D fQuistion S-~-~rce Comments! 1125721 changed to reload from defuel. Used on Salem SRO NRC exam more than 2 exams ago. I IC-..i****-* I I I I
RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
!Question Topic I ,_s_R_o_2_2_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~
IWhich of the following Salem events requires a 15 minute notification?
~ A unit shutdown is initiated to comply with Technical Specifications. ~ An oil discharge directly into the Delaware River with a visible sheen. ~ An endangered species of sturgeon is found deceased during circ water trash rack cleaning. @J Rx power is determined to be greater than 3459 MWth after removing a Feedwater Heater from service.
12/21/20151 ~I 194001G238 I[2.2.38 I !Ro Value: !0"[S[ovaiUeL.~.~ Section: lsystem/Evolution::::+/-l!'!J ~tatement* _ _ _ _ _ _ _'.'._Ji Knowledge of conditions and limitations in the facility license. IExplanation of [ 55.43.b(1) This question is SRO level based on the knowledge required of Conditions and Limitation required in the facility license, 1Answers: 1 and meets the criteria in ES-401, Attachment 2, page 17, II.A, 4th bullet. Knowledge of events which require 15 minute notifications is critical for SRO to know from memory to faciliate timely and correct notifications. The question matches the KIA because Salem Unit 2 Facility Operating License, Renewed License No. DPR-75, Section C.2, Technical Specifications and Environmental
- '"' - -*-*-- --' -- *--'"" ..,;u ha ---*-*--' I" \Al . *'-- - - '""-- <>C> ---*-*--..< ;n A .. . 0 Appendix B, Section 4.1, Unusual or Important Environmental Events, staes that the NRC must be notified of ... "Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation ... ". RAL 11.5.2.b contains conditions for which an oil spill is reportable in 15 minutes. The distracters are all reportable, A is 11.1.1.a (4 hour report) C is 11.5.2.c (4 or 24 hour report) and Dis at least a one hour report. Question is balanced with 2 environmental choices and 2 Tech Spec choices. - - - - * - * - - - - * [-----
I Reference Title _ _ j F~cility Refert:!!'ce_ Number 11 Reference Section [I Page No. [ [Revision! lsalem ECG 11 ii RAL 11.5.2.b II 11 4 I lsalem ECG II ljAttachment 16 114-5 11 4 I ISalem Operating License II II ii II I [f.o. Number Objectives I EL0_11.d , __ __,
!Material Required for Examination 11 II [Questi~_ll_S.~U.~~~~J INew I;Question
. _,,_,. __,
Modification Method: i I !Used During Training Program I D I
~
[Question Source Comments: I I 1=-*llll*v**" I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I
!Question Topic ] SRO 23I I Given the following conditions: - Unit 2 is operating at 100% power. - The operating crew entered S2.0P-AB.RC-0002, High Activity in Reactor Coolant System, when RMS channel 2R31, Letdown Line Monitor, went into WARNING.
Which of the following is required to be performed IAW S2.0P-AB.RC-0002, and why? The CRS will. .. I
~ direct Radiation Protection Technician to take surveys to determine if radiation levels may have changed access requirements.
I I
~ direct Radiation Protection Technician to survey the letdown line in the vicinity of 2R31 to confirm the suspected rise in RCS activity.
I I
~ direct Chemistry Technician to sample hourly for isotopic analysis to determine predominant radiation hazard (gamma, neutron, beta, alpha).
I l<1J I direct Chemistry Technician to initiate confirmatory sample analysis beca .. - *- .. - - - .. level changes. I !Answer 1 a I I iExam Lev~ IS I LCog~itive Level_ J IMemory I [Facilfty;**11 Salem 1 & 2 I IExamDate: t I 12/21/20151 ~I 194001G314 I 2:~-~4 --,,Ro~l 3.4l[SROValueJ 3.aj[Section:JIPWG IRq_~!~~JI 11lsROGroup:il 1j IJ~i ~
! - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' l<3.~~~~Q rc;-*-----~---1 LSystem/Evolution Title IKA Statement]
Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emeraencv conditions or activities. !Explanation o~ 55.43(4) This question is SRO level based on the "why" the action will have to be taken, as knowledge of radiation hazards which is ,Answers: occurring during an abnormal situation. A is correct as described in S2.0P-AB.RC-002 basis, so that prompt identification and subsequent notification of plant personnel is ensured. B is incorrect because chemistry sampling confirms 2R31 readings, not survey results. D is incorrect because rising counts does indicate dose level changes. C is incorrect because the hourly isotopic
~. , ......... : .... L._ .................................. ~~,
IC*~ '> ~O\ ___ ,,
'
Reference Title i I Facility Reference Number . J[~!!~i

..e!1£..!..~..£..~~ [PaQe-No. i :Revision 1 I High Activity in Reactor Coolant System 11 S2.0P-AB.RC-0002 II 11 I 1 a I I II II 11 iI I
! II II 11 11 I Lo. Number___ 1 Objectives ABRC02E003 I I I !,Material Required for Examination II II [auestion source: 1 IFacility Exam Bank I!Question Modification Method: II Direct From Source I[iiSed During Training Program ] D I
~estion Source Commentsj Used on Salem SRO NRC exam more than 2 exams ago.
I I <111*t-lll I I I I
RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes
!Question Topi~J ISRO 24 I Given the following conditions: - At time T-0, while operating at 100% power, the Auxilary Typewriter prints an alarm without OHA A-41, AUX ALM SYS PRINT annunciating. Operators immediately enter S2.0P-AB.ANN-0001, Loss of Overhead Annunciator System and begin assessing the OHA system functionality. - A T+5 minutes, and prior to verifying if either SER is in command, the Rx trips. - At T +14 minutes, the PO reports that neither SER is in command.
Which of the following identifies:
1) The correct ECG classification

2) The time at which it should have been declared?
~ 11) Alert
_, . 2) T +5 minutes. I
~ 11) Alert . 2) T+14 minutes.
I jc., 11) Unusual Event
~ 2) T +5 minutes.
[ ] 11) Unusual Event
-- 2)T+14minutes.
IA"!::~*-"'!' I r;:==J rs=::::I 10e?~!t!*!e ' ::*:e! I I _\nn!i<;;'>ti0R L____~------~
~------
fcX"!!! ;:::_;;jl 1
! iFacilit : Salem 1 & 2 ...tExamDate: 12/21/2015 I
lKA:jj 194001G432 112.4.32
~ystem/Evolution Title]
IKA statement: I . . . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . , . e of operator res onse to loss of all annunciators. Planationo~ 55.43.b(1) This question is SRO level based on the requirement to declare an emergency. The requirement to declare an wers: emergency under any of the loss of annunciators (S5) EALs is the loss for greater than or equal to 15 minutes. Declaring at T +5
--------
minutes with 10 minutes of assessment time left is not indicated even though the significant transient (Rx trip) has occurred, because the AB.ANN takes actions in an attempt to restore functionality of OHA system, and it is not a long drawn out procedure.
. . . .
I Reference Title 11 Facility Reference Number I :Reference Section 11 Page No. I [Revision] lsalem ECG I I 1 S5 - Instrumentation II 11 I I I II 11 11 I I I II ii 11 I jL.o. Number___ ! Objectives IABANN1E002 I I,_ _ ___.
!Material Required for Examination : ISRO 024 ECG Section S5 I !Question Source~ INew I!Question Modification Method: ~ 1lused During Training Program I D lQuestion Source Comments] I I ,-------- ---
jComment I I I I I I
RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question Topic ! SRO 25 I I IWhich of the following identifies a condition during a declared Emergency which REQUIRES a Protective Action Recommendation, either initial or upgrade, when the TSC/EOF have NOT been activated yet? I I
~J ANY General Emergency initial declaration.
I l§J I . - - **- esults in a radiological release to the environment. I I
~ ANY time the wind shifts after the initial 15 minute notifications have been made during a General Emergency.
I IC!~ ANY event which in the judgement of the Emergency Coordinator could result in exceeding 1OCFR Part 100 limits. !Answer J Ia I Exam Levell j S I ~ognitive Level JI Memory I !Facility: [ ISalem 1 & 2 I ;ExamDate: 11 12/21/2015 [KA:!l 194001G444 I.2.4.44 . .,IROValue;*Jl 2.4!,SR~~[Section:]jPwG I ROGroup:ll 1j[SRO-Group:JI 11 Bia D !system/Evolution Titl~] ,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____, IGENERI I I KA Statement-:; Knowledqe of emerqencv plan protective action recommendations. [Explanation of I 55.43.b(5, 1) This question is SRO level based on the knowledge of how to implement the associated section of the ECG, EP-SA-
~~vyer~~ 111-F4, Attachment 4, General Emergency, and the conditions which require making a PAR or PAR upgrade. The ICMF for a GE requires a PAR, see Appendix 1. There is a PAR for a Rapidly Progressing Severe Accident, a PAR for Hostile Action, and a defualt PAR, one of which must be made. Bis incorrect because a radiological release is defined as "Any release above normal,
_ :... -- I
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for a wind shift (see Appendix 1). Dis incorrect because there is no "judgement" directed in the PAR based-on exceeding radiation limits, the PAR is determined based on the the GE that directed its implementation .
...-.. . ---****~-.,
Reference Title . 11 Facility Reference Number J :~i:,terence Section I; Pa9."'..~~ iRevisionl IGeneral Emergency ~ .-.~
-- ,.,u 11 EP-SA-111-F4 II 11 1102 I I II II 11 11 I I II II 11 11 I ILo. Number Objectives I EL0_24.a
!Material Required for Examination 11 II [Question Source: Ii New I I!Question Modifica~i-~n Method: II I fused Duri_n!;J Training Program.1 D 1 !Question Source Commentsj I I 1c._., ... ,...,.otL
I I
I I}}

v t e Letter Sequence Other
TAC:U01910, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML16050A330, ML16050A331, ML16081A397, ML16081A406, ML16145A208
MONTHYEAR2015-12-04 [Table View]

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