ML17109A363

Draft Written Exam (Folder 2)

ML17109A363
Person / Time
Site:	Salem
Issue date:	10/12/2016
From:	D'Antonio J Operations Branch I
To:	Public Service Enterprise Group
Shared Package
ML16076A441	List: IR 05000272/2017301 ML16193A633 ML17031A294 ML17031A302 ML17031A337 ML17031A347 ML17109A351 ML17109A352 ML17109A358 ML17109A360 ML17109A361 ML17109A363 ML17109A364 NTD-16-051, Draft Outlines (Facility Letter Dtd. - 9/12/2016 (Folder 2)
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Download: ML17109A363 (118)

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

~_!'.stion Topic I l RO 1 Given the following conditions:

- Unit 1 was operating at 80% power, steady state.

- RCS Cb is 1100 ppm.

- Control rods begin withdrawing at 8 spm.

Of the followinq, which could be causing this rod withdrawal?

~ 1PT-505, Turbine Steam line Inlet Pressure transmitter, has failed low.

[!>-:1

_.:_j 1PT-505, Turbine Steam line Inlet Pressure transmitter, has failed high.

~ A VCT auto makeup occurred, and the 1CV175 Rapid Borate Stop Valve was open.

r-c:--i lcl._1 A VCT auto makeup occurred, and the boron addition rate was set to 1 gpm instead of 11 gpm.

IAr\swer 11 c I !Exam Level 1 IR I jcognitive Level 11 Application I IF~ ISalem 1 & 2 ~Date~ I 12/19/20161 jKA:ll 000001A202 ljAA2.02 --11ROValue:Jj 4.2l)SROValue:Jl 4.2ljsection:ll~!ROGroup~I 21!sROGroup:ll 2j iii D L_ m/Evolut~n Title I IContinuous Rod Withdrawal I [~1n j iKA Statement: J ! Ability to determine and interpret the following as they apply to Continuous Rod Withdrawal: I I Position of ememencv boration valve I IExplanat~on of j 55.41.(b)5 - The system characteristics of the emergency borate line are such that no flow will occur with a BAT pump running in

~_swei:_~ slow speed, which is the normal alignment of the CVCS makeup system. When an auto makeup occurs, the BAT pump running (in auto) swaps to fast speed, and would provide flow through an open CV75 to borate the RCS. The boration would cause RCS temperature to lower, causing ,__ SG pressure to lower, causing MT governor valve to open, causing temperature to further lower, and

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~ incorrect because PT-505 failing low would cause rod insertion, not withdrawal. B is incorrect because the rod withdrawal would eat maximum rate (72 som). Dis incorrect because it would cause a dilution, not a boration.

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L_ Reference Title Facility Ref~rence Numbe~ IReferenc~ Section I [ Page Ni] ~visionj ICharging, Letdown and Seal Injection 11205228-1 I II 1174 I Ieves Lesson Plan II NOS05CVCS00-16 I II 1I16 l l II I II 11 I

[Lo. Number =c=J Objectives I ABROD3E001 I I

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jMaterial Required for Examination iI II IQues_tion Source: \!New I!Question Modification Method: ~ I ~ed During Training Program :j D

~on Source Comments! I I

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

[QlleStionTop~ rR-0~2-----~--------------------~---------~---~--:-i IWith Unit 2 at power, which of the following failures would cause a SINGLE control rod to drop fully into the core?

~J Blown fuse for a Lift coil during demand for rod movement.

~ Blown fuse for a Stationary Gripper coil with no demand for rod movement.

!c. J Loss of Power to a single Power Cabinet during demand for rod movement.

[d.i Loss of Power to a single Logic Cabinet with no demand for rod movement.

!Answer i ru I Exam Level I ~ Icognitive Levell IApplication I [FM!mYl ISalem 1 & 2

~I 0000031<205 I rAK2.05 J IRo Value: I~ JSRO Value: I~ ~~I~ jRO Group:ILJ [sRO Group:! L J lsystem/EvolutionTitlel l_o_ro-'p-'-p_e_d_C_o_nt_ro_l_R_o_d_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___, ~--J iKA Statemen.t.J Knowledge of the interrelations between Dropped Control Rod and the following:

Control rod drive power supplies and looic circuits

!Explanation of I 55.41.(b )6- Rod movement loss of lift coil still has Moveable Gripper holding rod, rod just wouldn't move. Stationary gripper coil

.Answers: holds rod while not moving. Loss of power to Power Cabinet would affect multiple rods stationary gripper coils. Loss of Logic cabinet with a demand for rod movement could cause rods to drop during the movement cycle, as it would lose power to movable gripper and lift coils, and rods would stop when cycle timed out and stationary gripper coils re-engaged.

[ Reference Title ~I Facility Reference Number -I 1Reference Section ~ J Page NO:-j [Revisionl IRod Control Lesson Plan ii NOS05RODS-12 I 1121,44,55.1112 I I II I ii 1 I I I II I 11 iI I IL.O. Number Objectives I RODSOOE007

[Material Required for EXamination 11 II jQuestion Source: IINew IJauestion Modification Method: * *11 I ~sed Dl.!_ring Training Program I 0

!Question Source Comments! I I

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

~estion TopiCl l RO 3 I Given the following conditions:

- Unit 1 was operating at 100% power when a reactor trip signal was generated by the RPS, but the Reactor Trip Breakers did NOT open.

- The reactor was tripped from the control room when the RDMG sets power supply breakers were opened from 1CC3.

Because RPS did not successfully trip the reactor, which of the following actions is required to be directly performed by the control room crew?

1.;:*1 Trip the Main Turbine.

L, __

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~ Start the MDAFW pumps.

[~'

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c. Transfer 4KV Group buses.

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[] Open the Main Generator Output Breakers.

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[AnsWefl a  ! I lexam Level 11 R I [cognitive Level 11 Application I IFacility: 11 Salem 1 & 2 I iexamoate] I 12/19/20161 jKA:[I ooooo?K202 I[EK2.02 __[ JRo Valui] I 2.61 [SRO Value: 11 2.81 f SectionJ~ ~OUij I 1 IfSRO Group:i I 1I 1111 D

[System/Evoluti~n Title] l_R_e_a_ct_o_rT_n""'"*P_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___, ~~f_::_

• I KA Statement: I ~owledqe of the interrelations between Reactor Trip and the followinq:

eakers, relays and disconnects

!Explana~:j 55.41(b)7 - With the reactor trip breakers failing to open, the Main Turbine will not receive a trip signal (from the output of the RTBs). A manual turbine trip will be performed. The main turbine trip will cause the main generator output breakers to Answers:

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automatically open, so the crew will not directly open the output breakers. The group buses will transfer automatically on low voltage. The AFW pumps will automatically start on lo-lo level when the SGs shrink after the main turbine is tripped.

r--- Reference Title ---i i\

I Facility Re!erence Number I \Reference Sectioilj [ Page No. [ [Revision)

IReactor Protection System !1221051 I 11 1113 I I !I I II 11 I I II I ii 11 I

~umber~ Objectives I TRP001 E021 I I

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[Material Required for Examination.** !I II iQuestion Source: j jNew IJauestion Modification M~thod~I I [Used During Training Program.J D

[Question Source Comme~ I I

[comment J 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 jQuestio_r:i_Topic] I RO 4 _J Given the following conditions:

- Salem Unit 1 has performed a Rx trip and SI based on rapidly lowering PZR pressure and level.

- The crew has transitioned to 1-EOP-LOCA-1, Loss of Reactor Coolant.

Which of the following describes why the first several steps of LOCA-1 check for reasons OTHER than a LOCA for being in LOCA-1?

LOCA-1 actions assume ...

~ a loss of RCS inventory. If the ECCS injection flow is due to a Loss of Secondary Coolant, an unnecessary transition to FRHS-1, Loss of Secondary Heat Sink could be required.

~ a loss of RCS inventory. If the ECCS injection flow is due to a Loss of Secondary Coolant, the event could be terminated by isolating the faulted SG.

~' a SGTR is NOT causing the LOCA condition. If the ECCS injection flow is due to a SGTR, an unnecessary transition to LOCA-3, Transfer to Cold Leg Recirculation would be made. I

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a SGTR is NOT causing the LOCA condition. If the ECCS injection flow is due to a SGTR, the tube rupture must be addressed before returning to LOCA-1 to terminate ECCS flow. I iAOSWefl Ib I jExam Level i j R I ~ognitive Level 11 Memory I [FaCiiiiY:l ISalem 1 & 2 j IExamDate: JI 12/19/20161

[KM! 000009G418 Ir2.4.1 a ]~0Value:'l 3.3l~alue:lj4.0 l~ctionJl~iROGroup: j 11JsROGroup:jj 1 11 B:.fl D

\System/Evolution Ti~ l_s_m_a_ll_B_r_ea_k_Lo_c_A_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __,l ~o9 _I

[KA Stateiiilili]

Knowledoe of the specific bases for EOPs.

[Explanat'.on ofj 55.41.b(10)The Major Action Categories for LOCA-1 are : 1. Check for Subsequent Failure, 2. Monitor Plant Equipment for Optimal

,Answers. Mode of Operation, and 3. Determine optimal Method of Long-Term Plant Recovery. The first item checks that a faulted or ruptured SG is not the reason for ECCS injection, and either fixes it on the spot {faulted) or transitions to another more appropriate procedure With a faulted SG being the cause of the ECCS flow, LOCA-1 has a "do loop" which will wait until the SG has blown

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performed to do actions which will terminate the primary to secondary leakage, which is not performed in LOCA-1. Additionally, there is no transition from SGTR-1 to LOCA-1. The unnecessary transition to FRHS-1 is plausible because it happens on other procedures (FRCE,LOSC-2) when minimizino AFW flow to SGs when all are faulted.

1-------- Reference Title =1 L Facility R~ference Number-1 Referen~e Sectiolll ~~!le No; i [Revisio~

ILoss of Reactor Coolant 111-EOP-LOCA-1 I II 1130 I I II I II 11 I I II I 11 11 I

\L.O. Number Objectives I LOCA01 E009 1_ __ ,

\Material Required for Examination 11 II jQuestion Sciu~=J IFacility Exam Bank I~uestion' Modification Method: JI Direct From Source 1lused During Training Program] D

!Question s~_urce Comments! 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 J Outline Changes I

~estion Jopii; 11 RO 5 Given the following conditions:

- Due to a LBLOCA, Operators have entered 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation.

- NO CCW pumps are running.

Which of the followinq describes the procedure flowpath?

~] Continue in LOCA-3 to align BOTH RHR pumps, one SI pump, and one charging pump for recirc until RWST lo-lo level is reached.

[bJ Transition to LOCA-5, Loss of Emergency Coolant Recirculation, while attempting to restore CCW.

~ Transition to APPX-1, CCW Restoration, then establish cold leg recirculation.

[J Continue in LOCA-3 and align for single train recirculation without CCW.

~swei] Id l IExam Level JIR I ICognitive Levell IApplication I ~acility: 11 Salem 1 & 2 I !ExamOate: 11 12/19/20161 IKA:ll 000011A203 llEA2.03~\ROValue:Jl 3.7j!SROValueJI 4.2j\Section:ll~IR0Group:\I 1jlsROGroup:lj 1j iifO

~m/Evolution Ti~ l_L_ar....g'-e_B_r_ea_k_LO_C_A_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____.I ~_I1_=]

[KA Statement] Ability to determine and interpret the followino as thev apply to Laroe Break LOCA:

Consequences of manaqinq LOCA with loss of CCW

!Explanation of 55.41.b(10) At LOCA-3 step 11.2, with no CCW pumps running, operators are directed to go to step 124 and align for single train

~swer!=.__, operation. There is no provision to transition to APPX-1 or LOCA-5.

r-*-*-* Reference Title ~ i * *. Facility Reference Number J ~eference Section J [E:age No. I !Revision\

ITransfer to Cold Leg Recirculation 112-EOP-LOCA-3 II 11 1130 I I II II II II I

! II II II 11 I IL.O. Number Objectives I

LCA3U2E006 1_ ____,

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~terial Required for Examination 11 11

[Question Sourc~: 11 Facility Exam Bank I[Question Modification Methocl: . II Editorially Modified IIused During Training Program I D c__._____..__, _ _ _*--*~~-~--*~

!Question L..

Source .CommentSI' 174865 I

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

~uestion Topic] IRO 6 I Which of the following indications would be unexpected following a total failure of a RCP #1 seal?

Affected RCP .....

~ Seal leakoff D/P low console alarm.

~ Standpipe level high console alarm.

~ #1 seal outlet temperature rising.

~ I Seal flow lowering.

I [Faciiit}r:-11 Salem 1 & 2 I

~nswer 11 d I !Exam Level 11 R I !cognitive Level 11 Application I IExamDate: 11 12/19/20161

~I 000015K21 o I IAK2.1 O 1 IRO Value: I [3]j ISRO Value: i[}]"I section: 11~ IRO Group: ILJ iSRO Grou~ LJ I R'IJ D

!system/Evolution Title j IReactor Coolant Pump Malfunctions 11015 - _J iKA Statement] I Knowledoe of the interrelations between Reactor Coolant Pump Malfunctions and the following: I RCP indicators and controls I

!Explanation of I 55.41.b(3)Number 1 seal failure means minimum resistance I maximum seal flow, so seal flow will rise. More seal flow will be felt at

,Ans~~_:;...:.........__ the #2 seal, which will cause higher #2 seal D/P and higher leakoff going to the standpipe. The seal leakoff D/P alarm will come in because the failure of the seal results in no D/P across the seal.

i-----Reference I

Title I !Facility Reference Number! !Reference Section ] IPage No. I !Revision!

IRCP Lesson Plan II NOS05RCPUMP-12 I 1162 1112 I II I ii 1 I I Ii I 11 11 I li...o. Number Objectives RCPUMPE016 IMaterial Required tor Examination 11 'I jQuestion Source: j INew I!Question Modification Method: 11 I [!'sed During Training Program] D IQuestion Source CommentsJ I I

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

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]Question Topi~ RO 7

• ~----------------------------------------------'

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 ?

Char in flow will. ..

ia:l rise, and VCT auto makeup may initiate.

[i>~ lower, and flashing in the Letdown line could occur.

ic:l rise, and RCP seal injection flow could exceed Tech Spec limit total seal injection flow.

@J lower, and 2CC71 LTOWN HX CC CONT VALVE will not respond quickly enough to prevent Mixed Bed Demin isolation on high inlet temperature.

~-! !I:] l!S_xam Ley'!!J fL] ~nitive LeyetJ !Application I ~~~W Salem 1&2 l~~mDate: W 12/19/20161

~JI 000022K103 I~~ l!§o Value: ![I§J!SRo-~0"~~1~ ~~LJ SRO Grou D

~i;tem/Evolution Title j ILoss of Reactor Coolant Makeup

[KA ~tatement~ Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: I Relationship between charging flow and PZR level I

~anation of I 55.41.b(?) With a CCP in service, the failure LOW of the controlling PZR level channel will cause charging flow to RISE. The stem we rs: 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

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rise downstream ofthe letdown HX, and temps would not reach demin isolation lev~ls. The 2 rises ar~ incorrect becaus~ charging flow wouldn't rise, but the actions associated with higher flow are correct.

1- Reference Title-----.-_~ C Facility Reference Numbe~ !Reference Section I [_Page Noj [Revisionj ILoss of Charging 11 S2.0P-AB.CVC-0001 I 11 119 I I II I ii 11 I I II I lj 11 I lb:o. Number Objectives IABCVC1 E002 iMaterial Required for Examination 11 II Jauestion Source: *11 Facility Exam Bank !lauestiol'.! Modification Method: JI Direct From Source IJused During Training ProgramJ D

[Q\Jesti~n Source Comment~ 1125676 I

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RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes lciuestion Topic 11 RO 8 I I When responding to multiple stuck rods during the performance of EOP-TRIP-2, Reactor Trip Response, why are there two different boration times associated with borating the RCS for those stuck rods? I

~ As the number of stuck out rods increases, the amount of boration required to be added per rod also increases.

I

~ If a cooldown is in progress a higher boration total is required vs if the plant is stable in Hot Standby.

I rci L_:J in core life (BOL vs EOL) require different specific boration times to be used.

I

[ ] The different available borated water sources contain different ppm boron.

I

!Answer Id 1 1 IExam Level 11 R  ! !Cognitive Level ! IMemory I Facility;l ISalem 1 & 2 f I IExamDate: 11 12/19/20161 IKA:ll 000024K302 I[AK3.02 j !RO Value: JI 4.2l[sROVa1Lie:ll 4.4l lsection: II~ IRO Group:JI 21 ISRO Grou_e_:j[JJ lfm D

!System/Evolution Title I !_E_m_e_rg~e_n_c~y_B_o_ra_t_io_n_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~I 024 ~

IKA Statement:] Knowtedqe of the reasons for the followinq responses as they apply to Emerqency Boration:

Actions contained in EOP for emergency boration

!Explanation of 55.41.b(10) Step 7 of TRIP-2 requires boration for 2 or more control rods not fully inserted. It first attempts rapid boration from the Answers: Boric Acid Storage Tanks via a Boric Acid Transfer pump in high speed, for 35 minutes per rod. If this is unsuccessful, then boration from the Refueling Water Storage Tank via a charging pump running at greater than 87 gpm is performed for 120 minutes per rod. The BAST minimum Cb is 6550 ppm per Tech Specs, while the RWST minimum Cb is 2300 ppm.

Reference Title I Facility Reference Number I 1Reference Section 1

I 1: Pag;NOJ [ReVISio~

IReactor Trip Response 112-EOP-TRIP-2 II Step 7 11 1130 I IBoration Times for Multiple Stuck-out Control Rop" r-"C-MDC-1314 r- II 1115 11 1 I I II II 11 11 I

[i..o. Number Objectives I TRP002E006 j_ ___,

IMaterial Required !or Examination I I 11 jQuestion Source: I l_N_e_w_ _ _ _ _ _ _,I LIQ_u_e_st_io_n_M_o~d_if_ic~a_tio_n_M_e_th_o_d_:~*.cJl,1 1

_ _ _ _ _ _ _ _..... l !Used During Training Program I D jQuestion Source Comments: I I

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

~estion Topic 11 RO 9 I Given the following conditions:

- Unit 2 is in MODE 4 with RHR system providing the shutdown cooling.

- 2RH25 RHR Hot Leg Recirc discharge relief starts to discharge (lift).

With NO operator action, which of the followinq indications would occur if the valve continues to oass flow?

~ OHA-41 AUX ALM SYS PRINTER, for point 758, 24 RHR SUMP PUMP START.

~ OHA-41 AUX ALM SYS PRINTER, for point 710, RCDT PUMP START.

~I OHA C-2, CNTMT SUMP PUMP START.

~I PRT LEVEL HI-LO Console Alarm.

I I

!Ailswer l lc I iExam Level 11 R I [£ognltive Level 11 Application I [Facility: 11 Salem 1 & 2 1 IExamDate: 11 12/19/20161 IKA:I! 000025G446 I 2.4.46 I~~J@sRo Value: j[ill"~ction: II~ IRO Group:jLJ :SRO Group:jLJ

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[Sistem/Evolution Title  ! ILoss of Residual Heat Removal System 11025 -

• IKA Statement: I lant conditions.

Explanation Of] 55.41.b(5, 7,8) The 2RH25 is located upstream of normally shut 2RH26 which supplies Hot Leg Recirc during accident conditions. It Answers: _J is located in containment, and as such, discharges to the containment trench, which flows into the containment pocket sump. C-2

.._.____ will annunciate when the containment sump pump starts. All the other choices are radioactive waste paths which do not receive the discharge. RCDT and PRT are also located in containment and are plausible relief path collection points. RHR sump pump run Reference T-it-le_ _ _ _ _ J I Facility Reference Number 1 j Reference Section J[~.ci_{J!}§J IRevision 1 j Residual Heat Removal 11205332-1 I 1l 1 ===

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ILO. Number Objecti I RHROOOE008 1_ ____.

jMaterial Required for Examination lI 'I

!Question Source: 11 Facility Exam Bank I!Question Modification Method: II Editorially Modified I!Used During Training Program I D

[Question Source Commentsi 142333 changed from levels changing to alarms which would occur.

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[Comment I I

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

[§uestion Topicj I RO 1o I Given the following conditions:

- Unit 1 initiates a Safety Injection from 100% power due to a LOCA.

- 15 minutes after the SI is initiated, containment pressure peaks at 5 psig.

Which of the following identifies an automatic response which has occurred for Component Cooling Water System components, and why?

§J 2CC215 and 2CC113, Excess Letdown Heat Exchanger CCW isolation valves, received a close signal to ensure this non-essential containment flow path is isolated.

I

~ 21 and 22CC16, RHR HX CCW isolation valves, received an open signal to ensure that long term cooling of the RCS is in service when the swap to Cold Leg Recirc is required. I

!cl.

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2CC215 and 2CC113, Excess Letdown Heat Exchanger CCW isolation valves, received a close signal to ensure ALL CCW supply and return I from the containment is isolated.

~ 1.21 and 22CC16, RHR HX CCW isolation valves, received an open signal to ensure that RHR pumps do not overheat if the RCS remains above the shutoff head of the RHR pumps. I

[Arlswi!!J I a I \Exam Level 11 R I [Cognitive Leve!_J I Memory I !Facility: JI Salem 1 & 2 I IExamDate: JI 12/19/20161

!KA: 11 000026K302 I ~K3.02 ~I Ro Value: 11 3.6j ~RO Vatuel I 3.9j [Section: j~ !RO Group:; j 1j ISRO Group:! I 1I 1 Jll1 D fsYStem/Evolution Titlel j Loss of Component Cooling Water

~Statement: I Knowledqe of the reasons for the followinq responses as thev apply to Loss of Component Coolinq Water:

The automatic actions (alionments) within the CCWS resultinq from the actuation of the ESFAS

[Explanation of I 55.41.b(?,8) The SI signal sends a close signal to the CC215 and CC113, Excess Letdown HX isolation valves, as they are

!Answers:

J Containment Phase A isolation valves. The purpose of closing Phase A isolation valves is to ensure all non-essential containment penetrations are isolated on a SI. Bis incorrect because CC16s do not receive an open signal until the ARM PB is depressed and the RWST level is 15.2' and manual alignment is required to place ECCS in CLR. C is incorrect because ALL CCW supply and I *-.1..,.,""' ,, .. ...., -- j.,..,..,,1....,+,..,...,i ,......., ..,. Cl.-..~- -1 .1.L.- Dl""O ,...,....\"- ;...,. . .1.:11 because the CC16s do not recei~e an open signal until the ARM pg" is depressed and the RWST level iS 15.2', and the RHR pumps are cooled bv either flow throuqh the pump from RWST (LBLOCA) or recirc flow (SBLOCA until pp is S/D).

L .:_ .,.., * -1 *- a ,.., ~1-- - c *-- "'+ "'~ --~-

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L Reference Title *-~-- l ~Facility Reference Numbe!:__J §!tere11ce ~ection ' I LPage No] @Wis~

IComponent Cooling Lesson Plan II NOS05CCW000-11 I 1142 1111 I II I 11 11 I II I ii 1 I I

!Lo. Number. Objectives I CCWOOOE004

[Material Req'-'ired for Examination lI 'I

\Question Source: 11 Facility Exam Bank I~uestion Modification Meth.od: JI Direct From Source I!Used DI.Iring Training PrograrTil D

)Question Source Comments] 142744 I

\comment ">~/ __J I

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RO Skyscraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I l§!:estion :ropic I IRO 11 I I When responding to lowering PZR pressure IAW S2.0P-AB.PZR-0001, Pressurizer Pressure Malfunction, why are BOTH PZR PORV Stop valves closed when attempting to determine if the PORVs are the source of the pressure reduction? I l!'.j Because the PORV tailpipe temperature sensor is on the combined PORV discharge line.

[!>] To quickly terminate the leakage and prevent having to perform a Rx trip if the rate of pressure reduction is rapid.

~. To prevent pressurizing the PRT past the point where venting will be unavailable due to the high pressure interlock .

I Ld. l~se the PZR PORV and PZR Safety Valves discharge into a single line, and isolation of both PORVs will give indication of whether the 1

from the PORV's or Safeties. I

~~ Ia I lexam Level 11 R I [Cognitive Level 11 Memory I [FaCmty: 11 Salem 1 & 2 I IExamDate: 11 12/19/2016j

~l I00002?K3o3 LAK3~ !Ro Value: 11 I 3.71 SRO Value: 1 4.1I[Section:1 l..~£~§_j!Ro Groui]

1 I 11 lsRO Group:] I 1j 1111 D

[SYSiim/Evolution iiti!] IPressurizer Pressure Control Malfunction I ~---__J

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1KA Sta~ement: : Knowledqe of the reasons for the followinq responses as thev apply to Pressurizer Pressure Control Malfunction:

Actions contained in EOP for PZR PCS malfunction

]Explanation of I 55.41.b(?) As shown on dwg 203301, both the PZR PORVs and Safeties discharges combine into a single line going to the PRT.

Answers: The Safeties each have their own individual temperature sensor, while the PORVs have one sensor on a common line. B is incorrect but plausible because isolating both PORVs will stop the pressure reduction, but that's not why they are both shut per Basis Document, page 5, which states ... "Since there is no way to distinguish which valve is leaking, both PORV Stop valves are

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"--=*-in.o temperatures. If the temperature does lower, the user is directed to reopen one PORV Stop (2PR6). If temperature does not rise, then the procedure is exited. If it does, then that stop valve is closed and the other opened. Then the procedure is exited." C is incorrect but plausible because discharge is directed to the PRT, and PRT vent valve is prevented from opening >10 psig. D is incorrect but plausible because the all the discharges do combine into a single line, however all 3 safeties have their own sensors.

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Reference Title c-~[ Facility Reference Number! Reference Sectio_!l__J Page No. J i.~evisionj I IPressurizer Pressure Malfunction 11 S2.0P-AB.PZR-0001 I 11 1118 I I II I 11 11 I I II I ii II I ILo. Numii~-1 Objectives I ABPZR 1E002

!Material Required for Examination JI q

!Question Source: J INew IJQu~stion Mo~i!ication ~~!~od: J 1IUsed During Training Program I I D I

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[§ue$ticm Topic I l RO 12 I Analyze the following conditions:

- Unit 2 is operating at 100% power.

- l&C is performing Solid State Protection System testing.

- Reactor Trip Breaker (RTB) A is open.

- Reactor Trip Bypass Breaker A is racked in and closed.

- A valid reactor trip signal is generated in the Reactor Protection System, but the RPS system does not provide any output signal to trip the Rx.

Assuming that the RPS system WOULD generate the appropriate signal when MANUAL actions are performed during TRIP-1 immediate actions, which of the following contains both the operator action and the DIRECT method which will trip the Rx?

~ Depressing the OPEN PB for Reactor Trip Breaker B to deenergize the UV coil.

~~ Deenergizing the RDMG sets by opening the PZR heater bus supply breakers on 2CC3.

rc.i Depressing the OPEN PB for Reactor Trip Bypass Breaker A to energize the shunt coil.

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[] Turning either of the Reactor Trip pistol grip handles to the TRIP position to deenergize the UV coil for Reactor Trip Breaker B.

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!Answer J Id I [Exam Level J IR I !cognitive Level J IApplication I fF.lcilitY:i ISalem 1 & 2 i 1Examoate:] I 12/19/20161 IKA:ll 0000291<206 1~2.06 J!Ro Value: 112.9*1 [Ro Value: 113.1*I1secti0n:J I~ IRO Group: II 11 lsRo Group:JI 11 B D Isystem/Evolution Title-; IAnticipated Transient Without Scram iKA Statement: 1 Knowledge of the interrelations between Anticipated Transient Without Scram and the following: I Breakers, relays, and disconnects I

~lanation of I 55.41.B(7)he failure of the RPS output signal from the automatic trip will initially prevent the reactor from tripping. A is incorrect we rs: because PB energizes the shunt coil. B is incorrect because the Rx would already have tripped by another means prior to having to perform this step .. C is incorrect because the RT Bypass Breakers do not have an OPERATE function from the control console.

Tier/Group 2/1

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Cognitive Level 3 - Analysis

References:

NOS05RODSOO EL0-19, Logic drawing 221051 Schematic drawing 240111 I i Page No. I [ReViSi-00]

I i Reference Title II Facility Reference Number I [Reference Section IRPS Trip Signals !1221051 I 11 1113 I I II I 11 11 I I II II 11 11 I IL.O. Nombe~~ Objectives I RXPROTE027 I I,_ _ ___.

!Material Required for Examination

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~~stion S~u~:~;] IFacility Exam Bank 1l~uestlon Modificatio~_-Method: II Direct From Source I!Used [).!-!ring Training Program I D loueStio~ sour:~~ommeritsl I50753 I

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!Question Topic J I RO 13 I Given the following conditions:

- Unit 1 is operating at 100% power when Intermediate Range Nuclear Instrument 1N35 fails high.

- Operators are responding IAW S1 .OP-AB.NIS-0001, Nuclear Instrumentation System Malfunction, and are removing the channel from service IAW S1.0P-SO.RPS-0001, Nuclear Instrumentation Channel Trip I Restoration.

- With the PO in the rack area, OHA E-29, SR & IR TRIP BYP, annunciates.

Which of the following identifies the cause of this alarm?

~ 1N35 Control Power fuses have been removed.

[b.1 1N35 Instrument Power fuses have been removed.

r;:*-

~ 1N35 LEVEL TRIP switch has been placed in the bypass position.

I r(ll POWER MISMATCH BYPASS switch has been placed in BYPASS 1N35.

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~nswerl Ic I [EXam Level I rR I [Cognitive Level *11 Memory I [Facility: 11 Salem 1 & 2 I [ExamDate:] I 12/19/20161 jKA:ll 000033A102 I!AArn2__J [1£0 Value: If 3.0l iSROValueJI 3.11 ISect~I~ !RO Group:/j 21 !SRO Group:/ I 21 B D I

~ystem/Evolution Title I Loss of Intermediate Range Nuclear Instrumentation I [033 _ !

IKA Stateme!!t] Ability to operate and I or monitor the following as they apply to Loss of Intermediate Range Nuclear Instrumentation:

Level trip bypass I

!Explanation of 55.41.b(?) A is incorrect but plausible since control power fuses are removed when removing a PR NI from service using the same

~~wers: ____ procedure. B is incorrect but plausible as it is the next action to be performed after placing the Level Trip Bypass switch in Bypass. C is correct because when placing the level trip bypass switch the BYPASS position, the IR FLUX HI reactor trip and IR HI FLUX ROD WDRWL STOP are blocked, and OHA E-29 annunciates. D is incorrect because the IR does not have a power

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• Reference Title--***-----i l Facility Reference Number I!Reference Section 11 Page No, I [Revision:

INuclear Instrumentation Channel Trip I Restorati!I S1 .OP-SO.RPS-0001 Ii q 11 11 5 I I II II 11 11 I I II II 11 11 I

[Lo. Number Objective IABNIS1E001

!Material Required for Examination 11

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~estion Topic 11 RO 14 I Ioccurring Which of the radiation monitors listed below would provide the FIRST indication to the control room crew that a Steam Generator tube leak is with the unit at 100% power? I

~ R15 Condenser Air Ejector Monitor.

I l~ R19 SG Slowdown Rad Monitor.

I

~I R46 Main Steam line Monitor.

[] R53 Main Steamline Monitor.

[Answer 11 d I Iexam Level i IR I [Cognitive Levell IMemory I ~acility: J ISalem 1 & 2 1 IExamDate:-J I 12/19/2016 mr-\

IKA:ll 000037A106 IIAA1.06 j IRO Value: J§,sRO Value: l§lsection: II~ jRO Groue?JLJ~~~~LJ .. D

[5¥Stem/Evolution Title I ISteam Generator Tube Leak I [02-___J

~Statement: I Ability to operate and I or monitor the followina as they apply to Steam Generator Tube Leak:

Main steam line rad monitor meters

!Explanation of I 55.41.b(11) All the listed monitors will show some indication of a SGTL. However, the R53 are N2 monitors in the Main Steam line, Answers: which are very sensitive and would indicate prior to the others. The R15 samples the Main Condenser, so the steam would have to get there first (past the R53 monitors) The r46 are high range monitors, and while they can detect low levels of radiation, they would not provide any

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_,. __indication to control room crew via an alarm to alert them

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__ , to monitor them. The R19 blowdown monitor would I Reference Title 11 Facility Reference Number I ~Reference Section Ii Page No. I ~!~~

IRadiation Monitoring Lesson Plan II NOS05RMS000-17 I II 1117 I I II I 11 11 I I II I 11 lI I IL.O. Number I ABSG01 E001 IMaterial Required for Examination 11 I

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]Que!';tion Topic 1RO15 Given the following conditions:

- Unit 2 has experienced a tube rupture on 22 SG.

- After the reactor was tripped, the Main Turbine failed to automatically trip, and was manually tripped using the Turbine Trip handle on the control console.

- All feedwater flow was isolated to 22 SG during performance of EOP-TRIP-1, Reactor Trip or Safety Injection.

- 21, 23, and 24 SG NR levels are off-scale low.

- 22 SG NR level is 1% and rising slowly.

- Operators are currently performing actions in EOP-SGTR-1, Steam Generator Tube Rupture.

- Prior to performing the RCS cooldown to Target Temperature, operators are assessing AFW flow status.

Which of the following describes how AFW flow to 22 SG should be controlled?

l~J Leave AFW flow isolated to 22 SG for the remainder of the event regardless of 22 SG NR level.

[gJ Establish AFW flow to 22 SG until its NR level is > 9%, then isolate AFW flow to 22 SG. Maintain 22 SG NR level >9%.

11 IS Establish AFW flow to 22 SG until its NR level is >19%, then isolate AFW flow to 22 SG. Maintain 22 SG NR level >19%.

!(I] Leave AFW flow isolated to 22 SG until the cooldown to Target Temperature is complete, then establish AFW flow as required to keep the SG tubes covered.

[Ans""erj !E._J ~~'~\Cognitive level I !Application I ~~~Salem 1 &2 I 1ExamDate: n 12/19/20161 IKAl I000038A144 j [EA1.44j fROVallle:l 13.4*I \SRO Value: 11 3.4j [SeCtiOOJI~ IRO Group~ I 1!JSRO Group:! I 1I Ell D

]System/Evolution Title ] ISteam Generator Tube Rupture IKA Statement: I Ability to operate and I or monitor the followino as they apply to Steam Generator Tube Rupture:

Level operatinq limits for S/Gs

!Explanation of \ 55.41.b(4,10) Step 6 of SGTR-1 directs AFW flow be established to the ruptured SG if its NR level is <9%. A is incorrect but 1 Ans~rs:_~_J plausible if it is thought that you never feed a ruptured SG if intact SG's are available. C is incorrect but plausible since it is the correct action, but the incorrect level. Note: 19% is the new (REV. 30 of EOPs) SG NR level above which intact SGs are directed to be maintained at. D is incorrect but plausible if it is thought that the purpose of leaving ruptured SG isolated during cooldown is to m~-:_._: __ -----...1 nf - - ;nn IC: '-'~n+ Cr!~ - - - - *-"-'-'- ' - - ........ 1 * *1-: .... i.... i~ ..,-* -&-- in_.., __ 1n 1  :-:.1.:-.1.:-- rn +n T.-,-- .....

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Reference Title J Facility Reference Number ] IReference Section I ] I Page Nol /Revisiorl]

ISteam Generator Tube rupture lI2-EOP-SGTR-1 F II Sheet 2 II 1130 I q

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[Lo. Number ,]

Objectives ISGTR01 E006 I I,__ ___.

JMaterial Required for Examination 11 II

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[Question Topic 11RO16 Given the following conditions:

- Unit 2 has experienced a rupture of the 24 main steamline in the mechanical penetration area upstream of 24MS167.

- The Rx is tripped and a MSLI performed.

- Feed flow is isolated to 24 SG.

- Operators are preparing to transition out of EOP-TRIP-1.

Which of the following describes how the RCS cooldown rate will be affected over the next 30 minutes?

The RCS cooldown rate will. ..

~

I remain constant until 24 SG blows dry.

~ lower as the faulted SG pressure lowers.

~ remain constant until AFW flow is re-initiated to 24 SG, at which time it will increase.

[] lower until at least one intact SG NR level rises above 9%, then remain constant as AFW flow is lowered.

[Answer I !b I [Exam Level [ IR I [COgnitive Level JIApplication I [Facility:l ISalem 1 & 2 I !ExamDate-:J I 12/19/20161 IKA:ll 000040K102 I[AK1 .ci2j ~O Value: ii 3.21 jSRO ~I 3.6! [Section: II~ !RO Group:!! 1 j ISRO Group2j 1 I D

~ystem/EvolUtiOnTitle I j_s_te_a_m_L_in_e_R_u....p_tu_r_e_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.I ~=

[!<A Statement: i KnowledQe of the operational implications of the followinQ concepts as they apply to Steam Line Rupture:

Leak rate versus pressure chanQe j Explanation of J 55.41.b( 4,5, 14) As the faulted SG pressure lowers due to the steam break, break flow will lower, and the rate at which the RCS is 1Answers:  ; being cooled due to break will lower as a result. A is incorrect but plausible if it is thought that a static break will pass the same lbm of steam flow during the entire event. C is incorrect but plausible if it is thought that feed flow is initiated to keep the SG tubes wet, which only would be performed in LOSC-2 if ALL SGs were faulted. D is incorrect but plausible if it is thought that the lowering of fl CIA/"-*** ...i...;..-h ,,_ oln no.+ I i:;C"C" __ , .... CIA/"-***=-*- c:r- *-* oln -<<-o+ *i...- *-* ..-*=-- -*--- fl-* -<<--* nn CU'C: *-

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• Facilify Reference Number [ 1Reference Section I [Page No. J [Rivislon]

II II 11 1 I I II II Ii 1I I II II I 11 'I

[l.o. Number Objectives I MSTEAME016 jMaterial Required forExamil'.lation 11 'I

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[question Topic] IRO 17 Given the following conditions:

- Unit 2 is operating at 100% power, MOL.

- 21 SGFP trips.

- NO operator action is taken in response to the SGFP trip, and the Rx does NOT trip.

Which of the following is an UNEXPECTED alarm if it is locked in 2 minutes after 21 SGFP trips, and what procedure would be used to address the condition associated with that unexpected alarm?

~ OHA G-3, EHC SYS TRBL. S2.0P-AB.TRB-0001, Turbine Trip Below P-9.

11:)1 Console Alarm RC PRESS DEVIATION HI. S2.0P-AB.PZR-0001, Pressurizer Pressure Malfunction.

~

~ OHA G-44, COND POL TRBL. S2.0P-AB.CN-0001 Main Feedwater/Condensate System Abnormality.

@:] Console Alarm RC LOOPS TAVG-TREF DEVIATION. S2.0P-AB.ROD-0001, Immovable/Misaligned Control Rods.

!Answer! b I I iExam Levell !RI [cognitive Level 11 Application I ~acility: 11 Salem 1 & 2 1 IExamDate: 11 12/19/20161 IKA:ll 000054G445 I[2.4.45 I[Ro Value:l I 4.11 lsRo Value: 14.3 I!Section: 11~ [R[G!~ I J 11 ISRO Group: II 1j 1111 D

!System/Evolution Title I ,_L_os_s_o_f_M_a_i_n_F_e_ed_w_a_t_er_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___,I @Bu--~_-J

• KA statemenl-l I

Ability to prioritize and interpret the significance of each annunciator or alarm. I IExplanation of i 55.41.b(7, 10) The condensate polisher trouble alarm will be in alarm due to the CN108s AND the CN109 being open at the same IAn~~ers: i time. Dis incorrect because RC loops Tavg-Tref deviation will be expected as rods are driving in due to the turbine runback to 65%.

The ARP has operators place rods in manual, and if not successful at restoring conditions, going to AB.ROD-01. B is correct because the RC pressure deviation would not be expected, since the setpoint (+75 psig deviation)equates to when the spray valves Tho----*

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motion. The alarm response directs entry into AB.PZR-0001. A is incorrect because G-3 will be in alarm since it receives input from

~mn11n+ nf : ..... ......... ...J ....... ..J the EHC Control and Status computer, which has a Loss of Feed pump Runback alarm in, and the ARP directs performance of SO.TRB-2 to reduce power if a steam valve has shut. This question meets the K/A because the operator must be able to distinguish between expected and unexpected alarms for the loss of feedwater, and prioritize them in order to know which would require action in an abnormal Response Procedure.

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Reference Title --] I Facility Reference Number! 1Reference Section 1[Page No. ! IRevision I IMain Condensate Feedwater Abnormality II S2.0P-AB.CN-0001 I 11 1128 I I II I 11 11 I I II I 11 11 I IL.O. Number L Objectives I CN&FDWE016

!Material Required for Examination 11 q r~~e-~ti~~-~~~~ce] IFacility Exam Bank I~estion Modification MethOd:l Direct From Source v *-v ,,-***~--~~~-~~~~--~

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!Question Topic 11 RO 18 I Given the following conditions:

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

- 15 minutes after the power loss, operators have locally started 2B EDG.

Which of the following is an action that is REQUIRED to have been performed PRIOR to enerqizinQ 2B 4KV Vital bus, and why?

~ Shed non-essential DC loads to extend the time the Vital Instrument Inverters can power their AC loads.

~ Initiate and reset SI to prevent the auto start of a centrifugal charging pump and possible thermal shock to the RCP seals.

~ Deenergize ALL SECs and depress stop PBs for SEC actuated components to prevent overloading the 2B 4KV vital bus.

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[] Start the Station Blackout Compressor to provide air for operation of 21-24AF11, AUX FEED-SIG LEVEL CONTROL VLVS, to prevent over feeding the SGs when 22 AFW pp starts. I

[AnSwer 11 c I [exam Level 11 R I !Cognitive Leverl j Memory  ! [Facility: 11 Salem 1 & 2 I ~xamDate: JI 12/19/20161

~joooo55A203 l~A2.03 1 IRO Value: JI 3.9l 1SRO Value: Jl~.?j [Section: II~ [RC? Group:]j 1 j /SRO Gro1iJI 1 I .... ~stem!Evolutlon Title] j_s_ta_ti_o_n_B_la_c_ko_u_t_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.I lo55 .-:::J IKA Statement: I Ability to determine and interpret the followinq as they apply to Station Blackout: I Actions necessary to restore power I

)Explanation ofj 55.41.b(10) The Continuous Action Step for energizing a deenergized vital bus with an EDG comes AFTER the step to deenergize 1

Answers: all SEC's. The Bases Document states on page 15 that the reason to deenergize the SECs and depress the Stop PB for all SEC controlled safety related loads is to prevent the bus from overloading. It additionally states that a further reason is to prevent charging pump automatic start and possible thermal shock to the RCP seals. SI is initiated at Step 21 NOT to prevent a charging

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Blackout Coping Actions in Attachment 2 Part A of AB.LOOP-1. All the distracters are actions which will be taken during an extended loss of all AC power, but the correct answer is the only one that is required to be performed AND has the correct reason for doing it prior to power restoration. D will be performed, but it is NOT the correct reason, and is required within 60 minutes of Blackout. A and B will be performed, but are not required to be performed prior to power restoration.

~ Reference Title _..___._J I

l Facility Reference Number l IReference Section I! Page NQ.l [!§is~

jLoss of All AC Power j12-EOP-LOPA-1 II 1j17 1130 I I II II 11 lI I I II II 11 11 I

!LO. Number Objectives I LOPAOOEOO?

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[9:uesiionTo-Pic-l 1Ro 19 I Given the following conditions:

- Unit 1 is operating at 100% power.

- The crew is performing S1 .OP-ST.SJ-0001, lnservice Testing - 11 Safety Injection Pump.

- 11 SI pump is running.

- A loss of all off-site power occurs.

- 1B 4KV vital bus locks out on Bus Differential.

Which of the followinq describes Safety Injection pump status 2 minutes after the loss of off-site power?

~ ONLY 11 SI pump is running.

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~ ONLY 12 SI pump is running.

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[S BOTH SI pumps are running.

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@;] NEITHER SI pump is running.

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!Answer] Id I I Exam Level 11 R I !Cognitive Level I I Application I lfaCilitY:l ISalem 1 & 2 I iExamDate: 11 12/19/20161 t

~I000056A203 l~~~[Ifil°ISROValue:l[Ifil~~Jl~jROGroup:jLJ[SROGroup:jLJ

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jjAA2.03 ltl1 D Isystem/Evolution Titie.1 ILoss of Off-Site Power I !056-- *

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Explanation of; 55.51.b(8) neither SI pump will be running after a loss of off-site power with NO SI required. A is incorrect but plausible if it is

Answers: I thought that the 11 SI pump breaker would remain shut during the response, and the pump would restart. B is incorrect but

~ plausible if it is thought that only 12 SI pump would start because 11 SI pump was originally is an abnormal configuration. C is incorrect but plausible if it is thought the SECs would load both SI pumps, which it would only do if there was an accident signal

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I Reference Title [ C}§cility Reference Number ] !Reference Section 11 Page No. j IRevisionl j Loss of Offsite Power 1js1.0P-AB.LOOP-0001 j =t=3===jsa 1131 I

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ILo. Number Objectives I ABLOP1 E001 1_ ____.

IMaterial Required for Examination. ,11 II

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!Question Topi~ I RO 20 Given the following conditions:

- Unit 2 is at 100% power

- A loss of 2A 115 VAC Vital Instrument Bus occurs.

Which of the following describes the impact of the loss on the AFW system if a reactor trip with no SI occurs before the bus is recovered?

~ 21 AFW Pump will have to be started manually.

[§J Operators must dispatch an NEO to locally throttle 23AF21 and 24AF21.

le.I 21 AFW Pump may trip during the automatic start because 23AF21 and 24AF21 are failed open.

[] Operators must press the PRESSURE OVERRIDE DEFEAT pushbutton for 21 AFW Pump before gaining control of 23AF21 and 24AF21.

!Answer 11 b I \exam Level J IR I [cognitive Level 11 Application I [FaciiftY:l ISalem 1 & 2 j [ExamDate: iI 12/19/20161 i

IKA: 000057A106 IrAA1.06 _J [RoValue:1l 3.5! ISROValue:il 35! rsec~1~ :Ro Group:ll 1j [SRO Group~! 11 Ill D

[system/Evolution Tittel ILoss of Vital AC Instrument Bus IKA Statement: I Ability to operate and I or monitor the following as they apply to Loss of Vital AC Instrument Bus:

Manual control of components for which automatic control is lost I

[Explanation of 55.41.b(?) B is correct because 23AF21 and 24AF21 fail closed due pressure override as power is lost to the pressure transmitter.

~swers: *-*- An operator is dispatched at Step 13 of AB. Power is also lost to the valve controllers. A is incorrect because 21 AFW Pump would only have to be started if an SEC actuation occurred, and the stem states no SI has occurred. C is incorrect because the valves fail closed, D is incorrect because power is not available to the valve controllers.

! Reference Title _JL1- Facility Reference Number I Reference Sectio~ I Page N~

1 fRevisiOnj ILoss of 2A 115 VAC Vital Instrument Bus 1

11 S2.0P-AB.115-0001 I 114 1120 I I II I II 11 I I II I II II I l!::_O. Number Objectives I AB1151 E003 j_ __

'l!'Jlaterial Required for Examination II II jauestion Source: J lFacility Exam Bank I[auestion Modification Method: i Direct From Source IIUsed During Trainins;i Program I D jauestion Source Comments! 18031 o I

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§Uestion Topic JIRO 21 Given the following condition:

- Operators are performing actions in 2-EOP-FRCC-1 , Response to Inadequate Core Cooling.

Which of the followinq identifies how the PZR PORVs will be operated after startinq ALL available RCPs with CET temperatures above 1200°F?

l'i.J ISh"t BOTH PORV*.

~. Sh"t ONLY ONE PORV.

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~J IOpen ONLY ONE PORV.

I

~j 1Open BOTH PZR PORVs.

\Answer 11 d I [Exam Leveil IR I ~ognitive Level I 1 Memory  ! fFacilit}/: 11 Salem 1 & 2 I jExamDate: JI I

12/19/20161

)KA:JI oooo74K202 jlEK2.~IRova1u'!ll 3.9l[SROValue:Jj 4.oj/section:lj~IROGroUP:ll 2!1SROGrouP:1j 21 . . D lsY$iem/Evolution Title I j_1n_a_d_eq..:.u_a_te_c_or_e_c_o_o_lin_,g"----------------------------_....I ~74 --~

!KA Statement~ Knowtedqe of the interrelations between Inadequate Core Coolinq and the followina:

PORV I

~~:.~:::on of I circulation 55.41.b(10) RCPs will be started at step 25 if CETs are still >1200°F, to clear the water in the RCS intermediate leg and permit the of hot gases from the overheated core to circulate through the steam generators. If RCP(s) restart is not effective in decreasing CET temperature <1200°F (it is not per stem), the PZR PORVs (BOTH) will be opened at step 25.3 to help reduce RCS pressure to allow ECCS injection.

1 Reference Title ~I Facility Reference Number \ [Reference Sectio-~ ~ge ~ [Revision\

IResponse to Inadequate Core Cooling 112-EOP-FRCC-1 II Basis Doc 1138 1130 I I II II q II I I II II II 11 I

~.Number Objectives I FRCCOOE002 1--~

~erial Required for Examination

• 11 II

!Question Source: j INew I!Question Modification Method: i I [!Jse~ During Training Program I D

!Question SourceComments[ I I

[Comment I I I l I I I

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

[Question To~ IRO 22 Given the following conditions:

- Unit 2 has experienced a SBLOCA.

- Operators have transitioned from 2-EOP-TRIP-1 Reactor Trip Response, to 2-EOP-LOCA-6 LOCA Outside Containment: -

- The 21SJ49 COLD LEG ISOLATION VALVE has just been closed in an attempt to isolate the leak.

Which choice describes the response which would indicate successful leak isolation, and the next action to be performed IAW 2-EOP-LOCA-6?

~ RCS pressure stable, open 21SJ49.

I

~

~ I RCS pressure stable, stop 22 RHR pump.

I RCS pressure rising, stop 21 RHR pump.

@JI RCS pressure rising, close 21RH19 RHR DISCHARGE X-CONN.

I

[Answefl I c I !Exam Level! IR I [c-ognitive Level JI Memory I !Facility: 11Salem1 & 2 1 IExamDate: JI 12119120161 IKA:ll OOWE04A101 IfEA1.1 liROValue:Jl 4.ol[sRo~l 4.0l~.!'Jl~IRoGroup:jj 1j[SR(@roup:Jj 1j BO ILOCA Outside Containment E

lsystem/Evolution_Title]

IKA Stalem'" I to operate ood I " mooHoc the follo.;oo " they oooly to LOCA O"tsldo Cootolomoot onents, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and atic and manual features. I

!Explanation of~ 55.41.b(10) Distractor a is incorrect because after the leak is isolated by closing the 21SJ49, the procedure leaves it closed. Also Answers: stable RCS pressure is not the required RCS pressure indicating leak is isolated. C is correct because rising RCS pressure indicates leak isolation, and the next step is to stop 21 RHR pump because the RHR system is split and there is no discharge path.

Dis tractor D is incorrect because the 21RH19 would already have been closed at step 2. Distractor b is incorrect because both

-L-L.1- -----*u*o ---1 -"'---inn LL- .... - - - - nuc n .. - - - - - - - ------

l~o._j~~

-*--~--

I Reference Title :I Facility R~ference Number I [Reference Section I LOCA Outside Containment 112-EOP-LOCA-6 I 11 I j3o I I II I 11 11 I I II I II 11 I ILO. Number Objectives ILOCA06E002 1_ ____,

IMaterial Required for Examination 11 I

§uestion Source: 11 Facility Exam Bank I~~stlon Modification Metttod: ~ Direct From Source 1lused During Training Program I D

!Question Source Comments! 1127086 I

Icomment I I

I I

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

[Question Topic [ I RO 23 I Given the following conditions:

- A loss of heat sink has occurred.

- The operating crew is establishing RCS Bleed and Feed in accordance with EOP-FRHS-1, Loss Of Secondary Heat Sink.

- The RO opens one PORV. He reports that the second PORV will NOT open.

Which one of the following describes the consequences of the PORV failure?

~ ALL SGs will require depressurization to inject the alternate source of feedwater.

~ The RCS will rapidly re-pressurize when the SGs empty, resulting in a violation of a RCS Safety Limit.

~1 Bleed and Feed cooling of the RCS must be terminated and secondary depressurization to inject condensate pump flow must be immediately initiated. I ref] The RCS may not depressurize quickly enough to ensure sufficient SI flow to provide RCS heat removal, and other RCS openings may have to be established. I IA.nswerl Id l 1Exam Levell IR  ! 1'COgnitive Level !IMemory j [FacilitYl ISalem 1 & 2 I JExamDat~ I 12/19/20161 (KA:/! OOWE05K301 I[EK3.1 j!RoValu~l 3.4lfSRova1u6:ll 3.8l\SectiOrl':ll~iRoGrotp:]j 11\SROGroup:ll 11 JilD jsystem/Evolution rmej ILoss of Secondary Heat Sink I ~05 . ,

IKA Statement: I Knowledae of the reasons for the followina responses as thev apply to Loss of Secondarv Heat Sink:

Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivitv chanoes and operating limitations and reasons for these ooeratina characteristics.

[Explanation of j 55.41.b(8, 10) Distractor c is incorrect because Bleed and Feed is not terminated if only one PORV is open. Distractor b is incorrect

[An~wers: ___ because the RCS Safety limit is 2735 psig and will not occur with the PORV open. Distractor a is incorrect because action to align condensate pumps is already taken, and not as a contingency to Bleed and Feed. D is correct because FRHS Basis document describes the consequences of not having both PORV's open, and it is D.

I Reference Title J!. FacilitY. Reference NumbeQ \Reference Section~ IPage NO] \Revis~

IResponse to Loss of Secondary Heat Sink 112-EOP-FRHS-1 I !132 l I31 I I II I II 11 I I II II 11 11 I

!Lo. Number Objectives I

~

FRHSOOE009 1_ ____,

IMaterial Required for Examination II

!Question Source: 11 Facility Exam Bank I Question Modification During Training Program] O

!Question Source Comments! .-------------------------------------------~!

L . -~-~

[Comment

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes IQuestion Topic 1RO24I I Given the following conditions:

- Unit 2 has experienced a steam line break inside containment.

- Operators have entered FRTS-1, Response to Imminent Pressurized Thermal Shock.

Why will the operators be instructed to terminate SI and start RCP(s) if possible?

~ The soak required by FRTS-1 requires SI to be secured. RCPs are started to provide the ability to use spray to depressurize the primary.

lt?J The soak required by FRTS-1 requires SI to be secured. RCPs are started to provide mixing of cold SI and warm reactor coolant water.

~

• C. Safety Injection flow is a significant contributor to any cold leg temperature decrease or overpressure condition and must be terminated.

RCPs are started to minimize temperature gradient across SIG tube sheets. I

@J Safety Injection flow is a significant contributor to any cold leg temperature decrease or overpressure condition and must be terminated.

RCPs are started to provide mixing of cold SI and warm reactor coolant water. I

[Answer i Id I IExam Level i j R I [Cognitive Level 11 Memory I ~aciiiiY:l ISalem 1 & 2 I IExainoatel I 12/19/20161 l~I OOWE08K101 l[EK\1 ~§PYatue:JI 3.5!.SROValue:Jl 3.Bl!Section:il~[R{fGiouP:1I 1!ISROGroup:ll 11 -~ D l,Systl:!m/Evolution riiie] IPressurized Thermal Shock

~Statement: I Knowledoe of the operational implications of the followino concepts as they applv to Pressurized Thermal Shock:

Components, capacity, and function of emeroency systems.

!Explanation of 55.41.b.(10,S)A- incorrect - purpose for RCPs is not priority in FRTS-1, soak is not basis for SI. B - incorrect - soak not basis. C -

~~wers: _~ incorrect- SI basis correct, RCP basis not accurate. D-correct-page 12-13

[ Reference Title I* Facility Reference Number] !Reference Section I I Page No. I IRevisiOn]

IResponse to Imminent Pressurized Thermal Sh 112-EOP-FRTS-1 I 11 1130 I I II  ! II 11 I I II I 11 11 I IL.o. Number Objectives IFRTSOOE007 j_ ___.

~aterial R_equired for Examination :I II I

!Question Source: J Facility Exam Bank IIQuestion Modification Method:_ ']I Direct From Source I~ed During Training Program I D

~estion Source Coinmentsl 173425.

I

@om men!

-* --- 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

[auestion Topic I l RO 25 Given the following conditions:

- Unit 2 has tripped from 100% power due to a Loss of Off-Site Power.

- Operators are performing a cooldown IAW 2-EOP-TRIP-6 NATURAL CIRCULATION RAPID COOLDOWN WITH RVLIS.

Which choice identifies the MINIMUM RVLIS Full RanQe level required to be maintained durinq the cooldown, and its siqnificance?

raJ 70% to limit steam entering the RCS hot legs.

~ 100% to limit steam entering the RCS hot legs.

r::l 70% to ensure positive level indication of RCS.

6

~d. 100% to ensure positive level indication of RCS.

/Answell a I I [Exam Levelj IR I [COgnitive Levell IMemory I !Faciliti:l ISalem 1 & 2 I mDate: 11 12/19/20161 IKA:ll OOWE10K303 !IEK3.3_~[ROValue:ll 3.4l[SROValue~j 3.6jlsection:ll~/ROGtoup:JI 11/SROGroup:lj 11 IJll D I

\_~ystem/Evolution Title I Natural Circulation with Steam Void in Vessel with/without RVLIS M Statement: I Knowledge of the reasons for the following responses as they apply to Natural Circulation with Steam Void in Vessel with/without RVLIS:

Manipulation of controls required to obtain desired operatinq results during abnormal, and emergency situations.

IExplanation::'] 55.41.b(10)74% is minimum allowed at step 10, and get into a do loop until it is satisfied. The Bases Document states that if steam

~~ers: _ enters the hot legs, there may be some potential for it to reach the top of the SG U tubes, thereby disrupting the natural circulation flow circuit. By monitoring RVLIS and limiting the void growth to the top of the hot legs. the potential for introducing voids into the SG Utubes is minimized. Setpoint was changed from 74% to 70% current revision of EOPs.

r---

Reference Title  ! r I Facility Reference N~mbeG Reference Secti~ ~ge NoJ 1Revision 1 INatural Circulation Rapid Cooldown with RVLIS 112-EOP-TRIP-6 ll Bases Document 1122 1130 I I I  ! II 11 I I I I II 11 I

[LO. Number ==:i Objectives ITRP004E004 I I,_ _ ___.

[Material Required for Examination -11 'I I

'1Question source: 11 Previous 2 NRC Exams [Questi,oll Modification Method.: II Editorially Modified IIUsed During Training Program J D

'[Question Source Commentsj IUsed on Salem December 2014 NRC exam (2 exams ago.) Setpoint was changed from 74% to 70% current revision of EOPs in Dec 2015). I Jcomment '

'_. I I I I I I I

RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes fQuestion Topi(: I 1RO 26 Given the following conditions:

- A LBLOCA occurred on Unit 2.

- 2C 4KV vital bus locked out on Bus Differential.

- Both 22 and 24 CFCUs failed to start in low speed and cannot be started.

- After entering 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation when required, the crew has transitioned to EOP-LOCA-5, Loss of Emergency Recirculation when neither SJ44 valve opened.

- Containment pressure is 18 psig.

- The crew is performing Step 9 to determine the required number of Containment Spray pumps to be operated.

Which of the following identifies the Containment Spray pump(s) which will be in service if Step 9 is completed seven minutes after the RWST Lo Level alarm was received?

~ BOTH Containment Spray pumps.

[b:*1

~~

NEITHER Containment Spray pump.

121 Containment Spray pump ONLY.

I I

r.:111 ........

L_J I

~ ~

, .-* "*II V I

!An~ ~ ~ Lever_j IR I [C09nitive Level 11 Application I /Facm~ ISalem 1 & 2 I JExamDate: J 1___12_11_9_12_0_16_,I

~joowE11K102 I ~~J ~varue:\~~~~(ITI[SeciiOn:Jl~ ~~~LJ~~~~LJ D

~tem/Evoluti~e] ILoss of Emergency Coolant Recirculation

!KA Statement:l 1-----"----'"'"------'--------....___.....____,__._..._.________,..__....;.<._

=-**-~~****--.!

Coolant

___ Recirculation:

.,!explanation ofjl 55.41.b(10) Table C uses RWST level, Containment pressure, and the number of operating CFCUs in Low speed to determine how Answers: many containmernt spray pumps are required to be run. LOCA-3 was entered with RWST level at 15.2', and with 2C 4KV vital bus deenergized will be depleting at -13,000 gpm, with 2 RHR, 1 charging, 1 SI, and 1 CS pump running. This will require over 9 more minutes of injection before RWST level would be lower than 1'. With the stem stating that Step 9 is complete 7 minutes after 15.2' deenergized, so 21 will be run. If RWST drain do.;.,n calculation is.incorrectly pertorrned and it is assu'med RWST level is <1', then NO CS pumps would be required. The BOTH distracter is for RWST level between 1-15.2' if either containment presure is >47 si , or if no CFCUs were runnin L__ _~!erence Titl~--~----*~ i, Facility _Reference !"umb~:i:::J ;Reference Section_~[!~~~ ~~

ILoss of Emergency Recirculation 112-EOP-LOCA-5 1IBasis Doc I j 28 j I 30 I

~================-~===========~========[ If I

• ------------~*-----------'*---------'! j__. . . .

[!J?. Number . Objectives ILOCA05E006 1--~

lMaterial Required for Examinatio_l'l__J 12-EOP-LOCA-5 Sheet 1 AND S2.0P-TM.ZZ-0002 TANKS CURVE RWST II

~es~on SourceLJ INew I[Questio~ Modification ~ethod: J I l!J.sed During T~ining ProQraml D

~----~~-----~----:--i

~esti~n Source Comments; I I ICornn.-... ' J I

I I

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

[Question Topic 11RO27 I Following a SBLOCA inside containment concurrent with a faulted SG inside containment, the following conditions are noted:

- RCS subcooling is indicating 10°F.

- RCS pressure is 1380 psig.

- Containment spray actuation has occurred, but neither Containment Spray pump started.

- The crew is implementing 2-EOP-FRCE-1, Response to High Containment Pressure.

Which of the following describes why the Reactor Coolant Pumps (RCP) are tripped under these conditions?

~ To prevent RCP motor damage due to lack of cooling.

~ To minimize additional loss of reactor coolant from RCP seal leakoff.

~ To prevent subsequent thermal barrier damage due to backflow of RCS fluid.

I

[~ To minimize possible subsequent core damage due to pumping a two-phase mixture.

I

[Answer 11 a I ~Exam Level J IR I ]cognitive Level 11 Memory I IFacility: J ISalem 1 & 2 i !ExamDateJ I 12/19/20161 fKA:ll OOWE14K102 llEK1.c=J~§J 3.2liSROValue:jj 3.7j:section:ll~IROGroup:JI 2l~OGroup:JI 21 Bii D jSystemlEvolution Title I IHigh Containment Pressure I KA Statement: I Knowledqe of the operational implications of the following concepts as they apply to High Containment Pressure:

Normal, abnormal and emerqency operatinq procedures associated with (High Containment Pressure).

]explanation of I 55.41.b(10,3) CS actuation occurs at 15 psig in containment, as well as Phase B isolation, which shuts RCP CCW cooling inlet and

!Answers: I outlet valves. Per FRCE-1 basis document page 4, " ... RCPs are tripped since component cooling water to the RCP seals and

~'----*~..._.__j motors is isolated." B is incorrect but plausible if RCP seal package operation is not understood (running pump to secured pump).

C is incorrect but plausible based on having no CCW flow to thermal barrier. D is incorrect but plausible if it is thought a two phase

.-..:H.i., ................ 1lrl h.o .... _ ...................... lei .1.-. ......... ,..1: .... ,... ic : .......,n ........ ,. ..... ....i in .......... """' \

L_ Reference TitJe ]L Facility Reference Number Il'§.eference Section 1[hige No. j !Revision[

IResponse to Excessive Containment Pressure 112-EOP-FRCE-1 II ii 1130 I I II II 11 11 I I II II 11 11 I

--~-l 1 L.O. Number , ~ Objectiv I FRCEOOE006 I I

j_ ___.

[Material Required for Examination > lI II I

[Question Source:l Facility Exam Bank I !Question ModificC.'ltion Method: [I Editorially Modified I!used D1.1ring Training Program ] D I

~on Source Comments] Replaced a distracter which used to have containment spray water in it. as new rev of EOPs requires CS pumps NOT running to enter FRCE-1. I

~mment '"**,.. ,"",,

w *. ____J

\

I I

I

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

[Quest_~~~ Topic l I RO 28 I Given the following conditions:

- Unit 2 is operating at 4% power during a plant startup.

- 22 RCP trips.

Which of the following identifies the Tech Scee reauired action for the RCP trip?

~ Be in Mode 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

j

~ Be in Mode 3 within 15 minutes.

! *1 I

l~.'..J Immediately initiate corrective action to return 22 RC loop to operable status as soon as possible.

I

~ Suspend all operations involving a reduction in boron concentration of the reactor coolant system.

_ __J I

[Answer] j a I ]Exam Level ! IR I ]cognitive Level 11 Memory I fFacility~ ISalem 1 & 2 I ate:

I 12/19/20161

~I 003000G222 I~= ]RO Value: l[Ifil]sRO Value: ]EZJ'ISection: ]j~ !RO GrouJ>:lLJ[SRC[GfOUPJ L J lmJ D

]system/Evolution~ IReactor Coolant Pump System I [{)§3 - j

~Statement: J Knowledoe of limitino conditions for operations and safety limits. I IEXplanation

]Answers: * ...

j 55.41.b(3, 10) With the reactor in Mode 2, LCO 3.4.1.1 states all reactor coolant loops shall be in operation. The ACTION with less than all loops in operation is to be in HSB within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. B is incorrect but plausible as it has the correct action but wrong time. C and D are incorrect but plausible as they are actions taken for Mode 4 and Mode 3 RCP operation respectively.

I c-*-.--* .

Reference Title *----~ LJ=acility Reference Number \ !'Reference Section) [Page No.] !Revision\

j Salem Tech Specs II I 1 Section 3/3 4.1 113/44-1 1 l2a2 l I II II II 11 I I II II II 11 I

!L.O. Number Objectives I TECHSPE015

]Material Required for Examination 11 I jQuestion Source: J IFacility Exam Bank IJauestion Modification.Method: II Direct From Source I]Used During Training Program J D I Que~~~n Source Com!Jle~ I I

[CO"mment I I

I I

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

]Que-;tion Topic' I RO 29 Given the following conditions:

- Unit 1 is operating a 100% power.

- After performing a normal dilution for RCS temperature control, the START pushbutton for CVCS Makeup Mode Select was NOT depressed when returning the CVCS Makeup System to automatic.

Which of the following describes the effect this will have if VCT level were to continuously lower with NO operator action?

" 1SJ1 and 1SJ2, RWST to Charging Pump Suction valves will open automatically when VCT level reaches the lo-lo level setpoint, and 1CV40 i.Ci.:.J and 1CV41, VCT Outlet Isolation Valves, will shut automatically

[b:l

~-

1SJ1 and 1SJ2, RWST to Charging Pump Suction valves will open automatically when VCT level reaches the lo-lo level setpoint, and 1CV40 and 1CV41, VCT Outlet Isolation Valves will remain open.

~ A dilution will occur when the Primary Water pumps start at 14% VCT level with no boration flow aligned to the CVCS Blender.

~ The VCT will empty and the charging pumps will lose suction.

!Answer I a l I rEXam Level] IR I !Cognitive Level 11 Application I facility: 11 Salem 1 & 2 I~ I 12/19/20161

[KA:1 I004000K404 l[K4:04==1fROValuell 3.2l]SROValue:]j 3.1llsecti;;O:jlsYs l1ROGroup~]j 1l[SROGroup:lj 11 Ill D jsy$iem!EvOlutiOn--ruie] IChemical and Volume Control System I ~~

IKA Statemen_ii] Knowled e of Chemical and Volume Control S stem desi n feature s and or interlock s which provide for the followin :

Manual/automatic transfers of control Explanation of I 55.41.b(7) The eves START PB must be depressed to align the eves M/U system to whichever mode was selected (BORATE, I~~:___J DILUTE, ALT DILUTE, AUTO). Dis incorrect but plausible if it is thought that the SJ1 auto open interlock is based on having the CVCS M/U Mode Selector in AUTO and START. C is incorrect but plausible if it is thought that Primary Water pump operation would occur based on last position of CVCS M/U system (dilute per stem) when the auto makeup setpoint would be reached at 14%

L__- Reference Titre

- - ~l Facility Reference Number I 'Reference Section _JI Page No.] [Revision-;

I 1sJ1 Schematic 11211577 I II 113 I IChemical and Volume Control Lesson Plan ii NOS05WCS00-16 I ii 43 1116 I I II I II 11 I IL.O. Number Objectives ICVCSOOE006

!Material Required for Examination 11 'I

~estion Sourc~ [New I~estion Modification Method: II If Used During Training Program j D jQuestio~~~urce ~ommentsl 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 jQuestion Topic 11 RO 30 Given the following conditions:

- Unit 2 is operating at 100% power, steady state.

- The Charging System Master Flow Controller is in MANUAL.

- A malfunction in the H2 pressure regulator supplying cover gas to the VCT occurs, and VCT pressure rises from 25 to 35 psig.

Which of the following describes the effect this will have on CVCS when steady state conditions are established?

~ Charging flow will rise.

[ii]

I Letdown flow will rise.

I

~

L.

Charging flow will lower.

[~ ILetdown flow will lower.

I

[Answer[ a I I !Exam Levell IR I [cognitive Level J IApplication I fFacility: J ISalem 1 & 2 I fExamDate: J I 12/19/20161

~loo4oooK526 l~~ __ JIROValue:ll 3.1llSROValue:ll 3.2jfsection: 1l~IRo~I 11JsROGrouYJI 11 1111 D

[SYstem!Evolution Titlef IChemical and Volume Control System

[M Statement: i Knowledqe of the operational implications of the followinq concepts as they apply to the Chemical and Volume Control System:

Relationship between VCT pressure and NPSH for charging pumps 1Explanation of 55.41.b(5, 7) Raising VCT pressure will cause NPSH available to the charging pumps to rise. With the flow controller in manual, and lAnswe_rs: more suction head available, pump flow will rise. The letdown pressure control valve will respond in automatic and letdown flow will remain the same. Static pressure at the pump suction: An increase in surge tank level, increase in the pressure above the liquid in the surge tank, increase in booster pump output, or opening of the pump suction valve would

,_ r>-1 * .a.:----* - *-

I Reference '.fitle Facility Reference Number I jReference Section J I Page No. j IRevisioril IPWR Generic Fundamentals-Components **

I I

!I lj 44 ,, I I I I 11 11 I I I I 11 11 I Objectives I CVCSOOE004 l--~

[Material Required for examination 11 II

[Question source:. i INew I JQuestion Modifieation Method: 11 I Lused During Training Program j D

~stion Source Co~ I I 1£<>..mment .. __J I I I I I I

RO SkyScraper SRO Skyscraper RO SystemfEvolution List SRO SystemfEvolution List Outline Changes

[Question Topic i IRO 31 I Given the following conditions:

- Unit 2 is in MODE 4.

- 21 RHR loop is in service in SOC mode.

- 22 RHR loop is O/S and aligned for ECCS.

Which of the following would cause the temperature of the water flowing to the RCS from the RHR system to RISE?

§]

l 21SJ49 RHR DISCH TO COLD LEGS is shut.

~ 22CC16 RHR HX COMP CLG OUT valve is shut.

I

[Cl 2RH20 RHR HX Bypass Valve is throttled open.

I

~ l.21RH18 RHR HX FLOW CONT VALVE is throttled open.

I IA.Oswerl c I I ~xam Levefl j R I !cognitive Level[ IApplication I !Facility: 11 Salem 1 & 2 I fEXamDate~ I 12/19/20161

~loo5000A102  !~02 \\ROValue:ll 3.3!\sROValue:I! 3.41\Section:\l~IROGroup:jl 1!)SROGroui]j 1j Iii D

~tern/Evolution Tiflej IResidual Heat Removal System

[KA Statemeaj Ability to predict and/or monitor changes in parameters associated with operating the Residual Heat Removal System controls indudinq:

RHR flow rate iExplanation ~ 55.41.b(5) A is incorrect because with the RH19s open, the discharge from 21 RHR HX will flow through the normally open 22SJ49.

Answers: B is incorrect because the valve is normally shut, and with the 22RH12 shut there won't be any flow through 22 CCHX from 21 loop anyways. C is correct because bypassing flow around the in service HX will cause system temp to rise, as more warm RCS water is mixed in with the cooled RHR HX outlet water and returned to the RCS. D is incorrect because opening the 21RH18 will increase

.t'I-. .LL..-

-~ *~  :~ - ~- Ml.JO LIV *- .......... ~ *~

I j ]-Facility Reference Numbe~ 1Reference SectiO~ !Page NO.] ~evisio"-J

~~*--"

I Reference Title IRHR Simplified 11205332-SIMP I II 112 I IInitiating RHR 11 S2.0P-SO.RHR-0001 I !I 1129 I I !I ii !I 11 I ILQ.NUnliler-~ Objectives I RHROOOE004 I I

'--~

[Material Required for Examination 11 II lQuestion Source: J IFacility Exam Bank lQuestion Modification Method: II Direct From Source jL .' . . -

I!used During Training Prpgram. [ D

!Question Source Comments! 113-01 C RO 31

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

I

!comment I I

I I

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

~estion Topic 11 RO 32 Given the following conditions:

- Unit 1 was operating at 100% power when a SBLOCA occurred.

- During the performance of EOP-TRIP-1, RCS pressure has lowered to 1600 psig.

- 11 SI pump failed to start.

- Current TOTAL ECCS injection flow is 400 gpm.

Which of the followinq approximates TOTAL ECCS injection flow if 11 SI pump is successfully started?

fa:l 1400 gpm

~ 500gpm.

~ 1800gpm.

[ ] 11000 gpm IAnswe~ Ia I !exam ievel 11 R I !cognitive Level 11 Application i IFacil!fy: J ISalem 1 & 2 I /examDatej I 12/19/20161

=-- 006000A117

~! I~-____;" ,RO Value: 1@1sRo

-- I Valu~@Jsect1on: II~ ~~LJlSRO Group:ILJ & D

~!_!em/EVoiUtion Title I IEmergency Core Cooling System 11~06 ~-j iKA Statement: I Ability to predict and/or monitor changes in parameters associated with operating the Emergency Core Cooling System controls includinq:

ECCS flow rate

~Explanation Of] 55.41.b(8) Shutoff head for an SI pump is 1520 psid. Starting one at 1600 psig in RCS (with suction from 41' of RWST level,)

AnJ'wers:__~~! would provide no additional ECCS flow. The 500 gpm distracter if it is thought that shutoff head is slightly higher than 1600 psig.

The 1000 gpm distracter is if -full SI pump flow were added to ECCS flow. The 800 gpm is flow at 1082 psid.

L Reference Title J~cillty Reference Number~ ~ference Section __J [Page No. [ ~evisio~

IECCS Lesson Plan 11 NOS05ECCSOO I 1127 11 9 I I II I 11 11 I I II I 1 I I IL.o. Number I ECCSOOE008 Objectives "

I_ Material Required for Examination 11 11

!Question Source: 11 Facility Exam Bank I!Question Modification_ Method: II Concept Used I~sed During Training Program] D

~~stion Sour~e Comments! 161678 I

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

~estio~ Topic\ IRO 33 I Given the following conditions:

- Salem Unit 2 has experienced a LBLOCA.

- 21SJ44 fails to open when demanded and cannot be opened.

Upon completion of EOP-LOCA-3, Transfer to Cold Leg Recirculation, which of the following identifies how the 21 SJ44 failure affected the Cold Leg Recirc alignment compared to if the 21 SJ44 were open?

'":I la*; Containment spray header flow will be zero because 21 CS36 is interlocked to remain shut.

~ Containment spray header flow will be zero because 21 RHR pump will have been stopped.

~ Containment sump recirc flow will be directed through 22SJ45 and 22CS36 ONLY because 21 RHR pump will have been stopped.

@] Containment sump recirc flow will be directed through the 22SJ49 and 22CS36 ONLY because 21 SJ45 is interlocked to remain shut.

jAnswerJ j c I [Exam Level] j R I [cognitive Level -] IApplication I !Facility: ISalem 1 & 2 J 1 /ExamDate: 11 12/19/20161 1

t<AJI 006000K301 l!K3.01 -~/ROValue:il 4.1llSROValue:ll 4.2llSectlon:jl~[ROGrou~I 1jiSROGroup:/j 11 iii D

system!Evolution Title I IEmergency Core Cooling System rKAStatement

! Knowledqe of the effect that a loss or malfunction of the Emerqencv Core Coolinq System will have on the followinq:

RCS 1Explanation of J 55.41.b(8,7) When 21SJ44 fails to open, the procedure will direct stopping 21 RHR pump, and it will remain stopped as long as the

~nswers:

• 21SJ44 is shut (step 5.2) The 21SJ45 is interlocked with the 21SJ44 such that it cannot be opened with the 21SJ44 shut. The 21 CS36 interlock is with the RH1 and RH2 needing to be SHUT before the CS36 can be OPENED. A is incorrect because CS header flow will be provided by opening the 22CS36 since 22 RHR pump is running at step 22. B is incorrect because even though ALL circumstances if 22 RHR pump is running when the RWST lo-lo level alarm is received, the 22SJ49 isolation to cold legs is

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,....~ i... ...... -1 ...... .ii ..... a. n ie ................... ,...+a... ........................ ............... ..

~

shut. The final lineup will be with ALL containment sump recirculation flow going through the 22SJ45 to Charging/SI pump suctions, and throuqh 22CS36 to spray headers.

I Reference Title -- 11 Facility Reference Number *.. I ~eference Section

~=::;:;:;;;;;;;=::;:;:;;;;;;;=::;=::;=::;=::;=::;=::;=::;=::;=::;=::;~~:;:;;;;;;;:;:;::;::;:;;;;;;;:::::;;:;:;;;;;;;:::;;:;:;:;;;;;;;::;::;;.:;=::;=::;;:;:;;:;.~::==::;::==::;:;:;;;;;;;=::;~

1 IPage No.] ~Tsi001 j ECCS Lesson Plan JI lj I j 47,52 1I I

===~

ITransfer to Cold Leg Recirculation 112-EOP-LOCA-3 11 ll!.:::===I 130 l j_E_C_CS~S_im~p_lif_ie_d_D_ra_~_*n~g~~~~~~_.ll_20_5_3_50_-S_l_M_P~~~~~__.ll,~~~~~~__.ll lj_5_____,I

\L.o. Number Objectives I ECCSOOE016 I ECCSOOE006 IMaterial Required for Examination 11 'I IQuestiori Source:

i~- ' - *;

I !New I[auestion Modification Method: **. JI I 1used During Training Program] D IQuestion source Comment!) I I jComment . I I

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

[Question Topicl IRO 34 Which of the following describes the operation of the PRT Vent Vent, 2PR15?

The 2PR15 recieves a(n) ......

§J I

shut signal at 10 psig in PRT.

~ open signal at 10 psig in PRT.

I fcl Ishut signal at 100 psig in PRT.

I I

~ open signal at 100 psig in PRT.

I 1Answer] a I I iExami..~ j R I ~ognitive IeVei] IApplication l~Salem1&2 I JExamDate: 11 12/19/20161

[KA: i 007000A404 I ~:0-4--i JRo Value: 112.6*! !SRO Value: JI 2.6*jlSection: II SYS I RO Group:[ I 1j JSRO~ I I 11 Iii D iSystem/EvolutionTitiel r Pressurizer Relief Tank/Quench Tank System

~Statement:! Abilitv to manuallv ooerate and/or monitor in the control room: I PZR vent valve I

!Explanation ofj 55.41.b(7) PRT vent valve 2PR15 is interlocked to SHUT on a high PRT pressure of 10 psig. The PRT Rupture Diaphram will IAns~ers: --* actuate at 100 psig in the PRT.

L Reference Title IQacility Reference Number j [Reference Section J \Page No'. \ IReVisionl IControl Console CC2 I S2.0P-AR.ZZ-0012 I ii 51-52 If 39 I I I I 11 11 I I I I II 11 I

!L.O. Number Objectives PRTE006 jMaterial Required for Examination 11 II

\Question Source: 11 Facility Exam Bank I!Question Modification Method: II Concept Used I~ed During Training Progra~ D

~estion Source Comments\ I I

[comment I

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

~estion Topic 11 RO 35 I Given the following conditions:

- Unit 2 is operating normally at 65%, steady state.

- 22 and 23 CCW pumps are in service.

Subsequently 22 CCW pump trips.

Which of the following describes how the CCW system will respond or be manipulated by the control room crew?

~ 21 CCW pump will auto start when CCW header pressure lowers to 70 psig.

~ 21 CCW pump will auto start immediately since less than 2 CCW pump 4KV breakers are shut.

~ The NCO will start 21 CCW pump immediately since the 3rd CCW pump is normally aligned in MANUAL to prevent 3 CCW pumps from operating simultaneously. I idl The control room crew will enter S2.0P-AB.CC-0001, Component Cooling Abnormality, isolate the non-safeguards header, and start 21 CCW pump if CCW header pressure lowers to 80 psig. I iAnswer 11 a I ~xam Levell IR I !cognitive Levell IMemory I ['§cility: 11 Salem 1 & 2 I /ExamDatel I 12/19/20161

- m'."-'*"'1\} '1

~j 008000K401 I~~~::_] !RO Value]QJ}~RO Val~@~~!~ !RO Group:jLJJSRO Group:;LJ ~ D

~

JSystem/Evolution Titie] IComponent Cooling Water System I [008 _J JKA Stateme_!!!:J Knowfedqe of Component Coolinq Water System desiqn feature(s) and or interlock(s) which provide for the followinq:

Automatic start of standby pump iExplanatio:_::J 55.41.b(7) Normal CCW system alignment is 2 CCW pumps running in MANUAL, and one CCW pump in AUTO and stopped. A

[~swers: _ CCW pump selected to AUTO will start when either 21 or 22 CCW header pressure lowers to 70 psig. There is no interlock based on breaker positions, but other system DO have a breaker anticipatory function. AB.CCW could be entered, but no actions will be performed other than ensuring the backup pump started.

I L

Reference Title I: Facility Reference Number j [Reference SectlOrll \ Page No. \ \Revision!

I Control Console CC 1 11 S2.0P-AR.ZZ-0011 !I 1192 1160 I I II II II II I I II Ii 11 11 I

[LO. Number Objectives I CCWOOOE006 1_ ____,

JMaterial Required for Examination 11 q

!Question Source: 11 Facility Exam Bank 1lauestion Mocli~ication Method: IDirect From Source I:used During Training Progr~ D

[Question Source CommentsJ 1111918 I

IComment ]

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 11RO36 Given the following conditions:

- Unit 2 has experienced a 650 gpm tube rupture from 100% power.

- Operators are performing actions in EOP-SGTR-1, Steam Generator Tube Rupture.

- After cooling the RCS to Target Temperature, the RCS is being depressurized using PZR Spray with all RCPs running.

- During the depressurization, 2PS3 fails open.

Which of the following describes:

1) the initial effect of this failure during the depressurization

2) how the crew should proceed if actions in SGTR-1 are ineffective at stopping the pressure reduction?

The depressurization rate will ........

~ 1) rise a transition to SGTR-3, SGTR with LOCA - Subcooled Recovery will be required.

2)

~ 1) rise a transition to SGTR-5, SGTR without Pressurizer Pressure Control will be required.

2)

~ 1) remain the same

2) a transition to SGTR-3, SGTR with LOCA - Subcooled Recovery will be required.

l.:<l ! H  :- .~ I L_J 12) a transition to SGTR-5, SGTR without Pressurizer Pressure Control will be required.

IAnswerj ~ ~n_i_L~ ~ ~nitive Lev.!!] jApplication I [§C@¥] jsalem 1 & 2 I \ExamDate: \ 1_ _ _121_1_9_12_0_16_.I IKA:!lo10000A202 1A2.02 1

=:JIROValue:lj 3.9llsROVal~l 3.9l@!i:ction:Jl~ 1 Ro~I 1l[SROGrouiJI 11 D

!System/Evolution.!@!] IPressurizer Pressure Control System lKA Stat~ment~J 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 conse uences of those abnormal o eration:

Spra valve failures Explanation of I 55.41.b(10) When performing RCS depressurization in SGTR-1 with normal spray available, BOTH spray valves are opened Answers: FULLY. So the PS3 failing open will not have any initial effect on the depressurization rate. When stopping the depressurization when any condition in Table D occurs (step 19.2) both spray valves are closed. If both spray valves do NOT close, 21 and 23 RCP are tripped, and if pressure continues to drop, another RCP is stopped. The transition out will be to SGTR-3 at step 27. SGTR-5 is

Reference Title JI Facility Reference Number I !Reference Section I l§~ !Revision!

ISteam Generator Tube Rupture

~

112-EOP-SGTR-1 I ii 1130 I I II I 11 11 I I II I ii 11 I

~*Number ==:J Objectives I SGTR01 E009 I I,__ ___.

!Material Required for Examination JI II

~estion Source: Jj New I§:uestion Nfoclification Metho~1 I [~sed Duri~g Training Progr~ D

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!Question Topic ] I RO 37 I IWith normal Letdown in service, all of the following conditions would cause any open Letdown Orifice Isolation Valve 2CV3, 2CV4, or 2CV5 to automatically shut EXCEPT: I I

~ Phase A isolation signal occurs.

I

~ PZR level lowers to less than 17%.

~J 12cv2 OR 2CV277, Letdown Isolation Valve is shut from 2CC2 I

@::i I2CV7, Letdown Line Containment Isolation Valve is shut from 2CC2.

I 1Answe-r] j d I lexam Level 11 R I fCOgnitive Level IIMemory j I !Facility: I Salem 1 & 2 1 IExamDate: 11 12/19/20161

~1011000A405 IIA4.05 =!Ro va~I 3.21 [SR03ialue: II 2.91 lsection: II svs I]Ro Group:ll 21 !sRo Group:JI 21 - - D

[System/Evolution Ti~ IPressurizer Level Control System

[KA Statel'T!ent: ! Ability to manually operate and/or monitor in the control room:

Letdown flow controller

'Explanation of [ 55.41.b(7) The 2CV7 is NOT interlocked with the Orifice Isolation Valves. All the other conditions will cause letdown flow to be Answers: isolated by closing the CV3,4,5.

• • • ---.- - - - - - i

!" Reference Title .. . ,L Facility Reference Number I[Reference Section . I

! Page No:l, IRevisioOj ILogic Diagram 11224431 I q II 5 I I II I q 11 I I II  ! ii 1 I I jL.O. Number Objectives I CVCSOOE006

!Material Required for Examination iI II 1

Question Source: J jNew I!Question Modification Method: JI i lused During Training _Program I D

!_Question Source Commer}_§ I I

[~omment - -- 1 I

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RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes IQuestionTopic J 1RO 38 I IWhich of the followin~ actions chan~es the setpoint at which a Rx trip would occur while perform in~ a power ascension durin~ a plant startuE!? I

~ Placing a PRNI drawer Rate Mode switch to Reset at 4% Rx power.

I

~ Placing the High Flux at Shutdown Block switches in Block at 4% Rx power.

I

~ Depressing the Block Power Range A and B pushbuttons at 12% Rx power.

@JI- v Reset Source Range A and B pushbuttons at 12% Rx power.

IAnswefl c I I !Ex<1rn Level I IR I Icognitive Level* *11 Application I iFa~_I Salem 1 & 2 I IExamDate: J I 12/19/20161

~I 012000A101 I ~01 1IRO Valuel [IB JsRO Value: I!TI' Isection: 11~ RO Group:! LJ ~ Group:1' LJ al D I

jsystem/Evolution Titl!] j Reactor Protection System I @12-=1 IKA Statement: I Ability to predict and/or monitor changes in parameters associated with operatina the Reactor Protection System controls includina:

Trio setooint adiustment I

!Explanation of 55.41.b(?) A is incorrect but plausible if it is thought that resetting a High Flux Rate trip on a PRNI drawer would affect the Rx power Answers: trip setpoint. C is correct because it blocks the low power Rx trip (25%) when above 10% power. B is incorrect but plausible if it were to be confused with the Source Range Block switches which blocks the Source Range hi power trip. D is correct but plausible as it WOULD change the setpoint if it were performed below 10% power after the SR hi power trip was blocked, but it is prevented i .............. ........ _- . ;nn ,...r-i11.*11 ...

"~ " ~v- *- '- +~ <V<nc:i:: ,\ h, .LI-- D_-tn L

~*

Reference Title if Facility Reference Number I[Reference Section I [}>age No. I [R.eVisiOO]

IRx Protection Nuclear Instr Trip Signals 11221052 II ii 1I7 I I II u II 11 I I II II 11 11 I Objectives RXPROTE012 iMaterial Required for Examination 11 q

Question Source: J INew I!Question Modificatio~.Method: I 1lused During Training Program f D

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RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes I-Question Topic 11 RO 39 I Given the following conditions:

- Unit 2 is operating at 100% power.

- 21 and 22 CCW pumps are in service.

- 23 Charging pump is in service.

- A Safety Injection is initiated.

When performing SEC loading verification in TRIP-1, the PO reports the following equipment is not running:

- 21 Charging pump

- 21 Safety Injection pump

- #2 Emergency Air Compressor Assuming the SEC BLOCK switches on 2RP1 DO NOT WORK, which of the following identifies ONLY the power supplies which must be deenergizec in order to start the above equipment?

~

I ALL SECs must be deenergized.

1.--c

~ 2AVIB24 (2A SEC), 2CVIB9 (2C SEC).

I

~ I r:;--; I 2BVIB27 (28 SEC), 2CVIB9 (2C SEC).

riA\/l<"lriA triA ~~!"'\ ~~*11~~~ *~~ ~~!"'\

I I

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I

~ ~ ~Level 1 IR I ;cognitive level J IApplication I ~acility:! !Salem 1 &2 I~~ I 12/19/20161

~I 013000K201 IE_o:c=J IRo Value: l§lsRO ValuE'.~~~~I~ ~1.1i=JLJISRO Group:jLJ El D L~_ystemJEvolution Title I IEngineered Safety Features Actuation System I ~13 ___J I KA S_tatement: I Knowfedqe of bus power supplies to the followino: I ESFAS/safequards equipment control I

[Explanation Answers:

j 55.41.b(?) The A, B, C SEC's are powered from their respective vital instrument buses (VIB), and control the equipment powered from the respective 4KV and 460 Volt buses. 21 Charging pump is powered from B vital bus. 21 SI pump is powered from A vital bus. #2 ECAC is powered from C vital bus.

I Reference Title ,- CFacility Reference Number J [Reference Section _J [ Pag;N-0] [Revisio~

IRx Trip or Safety Injection 112-EOP-TRIP-1 ii 111 1130 I I4KV Vital buses one line 11203061 11 11 1135 I j 2c 460V vital bus one line 11601392 11 11 1124 I

[LJ.i}Jumber Objectives ISECOOOE011

'--~

!Material Required for.Examination 11 ii

[ouestion Sourc!J New I I!Question Modification Method:.* JI 1lused During Training Program I D

!Question Source Comme~ I I Icomment .,,,-:*.' 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 jauestion Topic! IRO 40 Given the following conditions:

- Unit 2 was tripped from 100% power about 16 minutes ago.

- IRNI 2N35 indicates 2.0 E-11 Amps.

- IRNI 2N36 indicates 2.0 E-10 Amps.

- Channel I SUR is -0.3 dpm

- Channel II SUR is -0.06.

Which of the following describes the condition present, and any action(s) required to be performed as a result of this condition?

§] 2N35 is over compensated. Manually energize Source Range channels.

~ 2N36 is under compensated. Manually energize Source Range channels.

I r

~ 2N35 is under compensated. Ensure Source Range channels automatically energize when 2N36 lowers to 7.0 E-11 Amps.

I

@] 2N36 is over compensated. Ensure Source Range channels automatically energize when 2N36 lowers to 7.0 E-11 Amps.

I fAOswer 11 b jKA:ll 015000A202

/system/Evolution Ti~

I [Exam Level 11 R llA2.02 I [Cognitive Level llROValue:1I INuclear Instrumentation System 11 Application  ! JFacility: IISalem 1 & 2 3.11.isROValu;!jj3.S*!lsection:ll~IROGroup:lj I /Exami>ate:j I 2llSROGroup:[j 21 I*

12/19/20161 j [015_1 D

IKA Statement: I Ability to (a) predict the impacts of the following on the Nuclear Instrumentation System and (b) based on those predictions, use procedures to correct, control, or mitioate the consequences of those abnormal operation:

Faulty or erratic operation of detectors or compensating components

~nationofJ 55.41.b(6, 10) TRIP-2 step 22 asks if both IRNI channels are reading <7E-11 A. If they are not, it asks if under compensation is ers:

-~*--,_._

preventing proper operation. If yes then the operator is instructed to manually energize SR channels. IRNI channels normally continue to lower until off scale on control console. If the reading is higher that expected, and SUR is abnormally low as indicated in stem, this provides justification for an undercompensated instrument, which allows more lower energy gammas to be seen. On I Reference Title , I Facility Reference Number IiReference Section .II Page No.I~~

IReactor Trip Response 112-EOP-TRIP-2 I ljsh 4 1130 I I II I 11 11 I I II I q 11 I

LO. Number~

Objectives I EXCOREE009 I EXCOREE010 I

Material Required for Examination 11 II jauestion Source: J IFacility Exam Bank I!Question Modification Method: * ~Direct From Source I!Used During Training Program D I

~stion sc.urce ~omm~ 1158886 I

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

!Question Topi~ IRO 41 Given the following conditions:

- Unit 1 is operating at 100% power.

- 11-14 CFCU's are in service.

Which of the following identifies the CFCU's which will be running in LOW speed if a MODE 1 SEC initiation signal occurs?

[;' ALL.

1.~:.!

~ 11-14 ONLY.

~ Primary CFCU ONLY on each Vital Bus.

~

~ Primary OR Backup CFCU ONLY on each Vital Bus based on which one was already running.

[Answe'!] Ia I !Exam Level 11 R I jcognitive Level 11 Memory I [Facili~y:l ISalem 1 & 2 - -ate: 11 p

12/19/20161 jKA: 11022000A301 i IA3.01 JIRO Value: IDJ}fsRO Value-:J@1section: II~ ~~~JLJ iSRO Group:j LJ 1111 D jsystem!Evolution Title j IContainment Cooling System IKA Statement:! IAbility to monitor automatic operations of the Containment CoolinQ System includinQ: I Initiation of safeguards mode of operation I rE:xplanauon o~ 55.41.b(7) during a MODE 1 SEC initiation, (SI with vital power) ALL CFCUs running in high speed will be stopped, then ALL 1 Ans~rs: ____ CFCUs will start in LOW speed. B is incorrect but plausible if it is thought that only the running CFCUs will swap to LOW speed. C is incorrect but plausible as other SEC controlled equipment has Primary/Secondary start function (SW pumps). D is incorrect but plausible if it is thought C could be correct but is based on what pump is already running (like SW pumps do)

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I Reference Title .1 Facility Reference Number I I Reference Sectionj [_!>_age N~ !Revisionj ISafeguards Equipment Control System 11 NOS05SEC000-07 I !114-15 11 7 I I II I 11 11 I

! II I ii 11 I

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1L.O. Number Objectives I CONTMTE007

[ __ ~

!Material Required for EXamination 11 II IQuestion Source:* IINew I!Question Modification Method: ** *ii I l!!_s~-~ [)uring Training Program D I

§u~~§~--~~urce CommentsJ I I Icomment

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[auestion__Topic 11 RO 42 I 21 CFCU is running in Low speed during its weekly exercise and flush.

Which of the following identifies how steady state service water flow through 21 CFCU would be affected when the CFCU is transferred from Low Speed to High Speed?

Service water flow will ...

~~r*

Hia..,

~ I remain the same.

I

[] I rise or lower based on initial SW header pressure.

I IAnswe'l c I I !Exam Le~ j R I [Cognitive Level 11 Memory I FacilitY:l ISalem 1 & 2 f I JExamDate: 1 I 12/19/20161

~I 022000K101 1IK1.01 ~ ~O Value: II 3.sl ISROValue: II 3.71 ISection: II~ !RO Group:il 11 ISROGroup;jl 1j 1111 D

~/Evolutioiifi!l!] j Containment Cooling System iKASt~~ Knowledge of the physical connections and/or cause-effect relationships between Containment Cooling System and the followino:

SWS/coolino svstem I

=nation of, 55.41.b(9,8) The CFCU SW flow control valve SW223 has a position limiter on it, typically 50% travel. The SW223 opens on a start we rs:

-~**-**~-______]

i signal from either low speed or high speed. With a mechanical stop employed, SW flow will be the same for high speed or low speed operation. The stem states "steady state SW flow" because the CFCU is normally stopped for 30 seconds when transferring speed. Distracter Dis plausible if it is thought that SW header pressure would change, which would affect SW system flows, but in

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L Reference Title _:J' Facility-Reference Numbe~ [Reference Section _J [Page ~<i] IRevisi_ci!i]

I1sT CFCU SW Valves 11 S2 .OP-ST .SW-0010 I 1118 1120 I I II I 11 11 I I II I 11 Ii I

[LQ. Number---~ Objective ICONTMTEOO? I 1Mate_rial Required for Examination 11 II

§uestion Source: i IPrevious 2 NRC Exams I!Question Modification Method: .fl Direct From Source I~!_~d During Training Program I D

~uestion Source ComfflentS1 I Salem Dec 2014 NRC exam I

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

[auestion Topic I IRO 43 I Given the following conditions:

- Unit 1 has experienced a LOCA.

- Control room operators have progressed to the point where the SECs have been reset.

- Containment pressure was 4.5 psig when the SECs were reset.

Which of the following describes the Containment Spray System response should a HI-HI containment pressure signal be generated at this point in the accident?

CS system valves ...

~

would realign for spray, the CS pumps would be started by the SEC.

~ would realign for spray, the CS pumps would have to be manually started.

~ must be manually realigned for spray, but the CS pumps would be started by the SEC.

@]

must be manually realigned for spray, the CS pumps would have to be manually started.

!Answer 11 b I IExam Level  ! IR I Icognitive Level 11 Application I lsalem 1 & 2 1 IExamDate: 11 12/19/20161

~:I I026000A301 1l!'~:!_l!_---= JRO Value: II 4.31 SRO Value: II 4.5j lsection: II~ IRO Group:ll 11 ISROGroup:[LJ IJll D

~System/Evolution Title] _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___,I @Zf---

j_c_o_n_ta_in_m_e_n_t_S..:..p_ra-'y_S-'y'-s_te_m I KA Statement: I Abilitv to monitor automatic operations of the Containment Sprav Svstem includinq:

Pump starts and correct MOV positioninq I IExplanation of ; 55.41.b(7,8,9)The SEC controller operates the CS pumps at 2 different point in the sequence UNTIL the SEC is reset. The SEC Answers: ONLY control the CS pumps, not the CS valves. The valve realign on the hi-hi containment pressure signal whenever it is received, but once the SEC is reset, it will not start the CS pumps since the sequencer is no longer active.

I Reference Title ..

j L Facility Reference Number I [Reference Section 11 Page No. I IRevision'.

IContainment Spray Lesson Plan II NOS05CSPRAY-06 I 1142-43 11 6 I I II I 11 I I I I II I 11 II I

[L.O. Nu_mber I CSP RAYE006 Objectives 1_ ___.

jMaterial Required for Examination 11 II

- I I

.., ~xam Bank I!Question Modification Method: II Direct From Source 1IUsed During Trainin!J Pr<_>_9-_ran:i JD

!Question Source Comments I 180567 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 TopiCl  !RO 44 I Given the following conditions:

- Unit 2 is performing actions in 2-EOP-LOCA-1 for a LOCA.

- Containment pressure peaked at 20 psig.

When performing CS Pump Stop Criteria steps in LOCA-1, the RO depresses both Reset Spray Actuation pushbuttons with Containment pressure at 15.1 psig.

Which of the following describes the effect of this action, if anything?

~

I There will be no effect.

~ Containment Spray pumps will stop.

,..-1 I

[<:~ Containment Spray Actuation signal will reset.

I fdl IContainment Spray valves 21/22CS2, 2CS14, 2CS16, and 2CS17 will shut.

I

[Answer J c I I IExam Level J IR I !cognitive Level J IApplication I !Facility: 1 ISalem 1 & 2 I [EXami:>ate: J I 12/19/20161 IIA4.05 _1 [Ro Value: 11 3.51 ISRO Value:* j 3.51 !section: 11~@? GrouP:1 I 1 j fSRO Group:! I D-lI

KA: 026000A405 1 I

~temfEvOlution Title] l_c_o_n_ta_in_m_e_n_t_S..._p_ra_.y_S.....Y.._s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~l 026 _J

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

Containment spray reset switches

Explanatio:j 55.41.b(?) Containment Spray actuation relays have retentive memory, which allows relays to be manually reset with an actuation

!Answer:-.:... signal still present. So even though containment pressure remains above the Containment Spray actuation setpoint of 15 psig, the actuation signal can be reset. The Containment Spray pumps and Spray valves CS2, CS14, CS16, and CS17 do NOT reposition to their normal positions. The CS14 is normally open with power removed. The remaining CS valves are normally shut and open

• * - - - .&.1.-.. ... f"C: ... : ___ , --...1 -----" hn ... 1....... 0'4 *-.&.:\ i.\...- - - - - * ...... .&. * .... .L:-.- -**---1 "1c- ........... .a.

I 1--- Reference Title ---:::J [-*Facility Refer~nce Number_j [Reference Section((I Page No. ! [Revisiofl!

IRPS Logics Safeguards Actuation Signals

!1221057 II II 1123 I I II II ii II I I II II 11 1 I I

[i.o. Number -----i Objectives I CSP RAYE008 I I

1_ ___.

[Material Required for Examination 11 II

Question Source:

• 11 New I!Question Modification Method: 11 I

1lused During Training Program I D

[Questi~~ 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

~e$tion Topie'] I RO 45 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 following identifies which Hydrogen Recombiners can be started when directed by procedure if required?

~ 11 ONLY.

~ 120NLY.

r:::-*1 11AND12.

~

I

[] Neither Hydrogen Recombiner is available.

I

~nswer.J Ib I !Exam Level 1IR l !cognitive Level 1 lApplication I fF.icmtY:l ISalem 1 & 2  ! IExamDate: 11 12/19/20161

!KA: 11028000K201 !jK2.01 =IROValue:Jl2.s*l[Sifova1ue:jj2.a*jls~ction:ll~[f§5f~I 2j/sROGroup:[j 2j Ill D

!System/Evolution Title i j Hydrogen Recombiner and Purge Control System

!KA Statement: I Knowledge of bus power supplies to the following:

Hvdrooen recombiners jExplanation of I 55.41.b(9). Hyd recomb are powered from 1A and 1B 460 volt vital buses, which are powered from their respective 4KV vital 1Answe.rs: buses. With 1A bus locked out on diff, 1A 460 will not have power. Only 12 is available.

I Reference Title !I Facility Reference Number I !Reference Section~ i: Page NoJ I Revision]

I1A Aux Building 460V Bus One line 11601231 II 11 1117 l j 1B Aux Building 460V Bus One line 11601232 II. II 1118 I I II II II 11 I

[L.o. Number Objectives I CONTMTE004

!Material Required for Examination 11 I

(Question Source: \ Previous 2 NRC Exams 1IQueStion Modification Method: !I Direct From Source I!Used During Training Program I D "

!Question Source Comments! I I

~mment .... _=-_]

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

[auestion Topic ] IRO 46 I I Of the following, which one describes how the Spent Fuel Pool level will be lowered if required, IAW S1 .OP-SO.SF-0001, Fill and Transfer of the Spent Fuel Pool? I

~J Gravity drained to the RWST.

r*1 L~ Pumped with Spent Fuel Pool Cooling pump to RWST.

-*1 8 Pumped with Refueling Water Purification pump to in service WHUT.

I

~ Drained via the Spent Fuel Pool Skimmer loop to the Drain Header to the in service CVCS HUT.

I

!Answer! b I I [gOxamievel 11 R I [cognitive Level 11 Memory I [Faciliti] ISalem 1 & 2 I iExamoa~ I 12/19/20161 IKA:ll 033000K105 l:KLlIB~~OValue:l!2.7*1 1 SROValue:Jl2.8*lfSection:1l~IROGroup:[j 21~~1 21 Im D

[SYSteffitEvolution T~ j Spent Fuel Pool Cooling System IKA Statement: I Knowledge of the physical connections and/or cause-effect relationships between Spent Fuel Pool Cooling System and the followin :

RWST lexPl_an.ation, 55.41(8,13,4) The RWST normal level is -41 '. The puts the water level at 141 '. The Spent Fuel Pool is -128'. There can be no

~wers: gravity drain TO the RWST, although it will work coming FROM the RWST to the SFP. Transferring water from the SFP to the

--- **-- RWST is done by manipulating valves within the SFP Cooling system, and pumping the water with the Spent Fuel Pool pumps to the RWST. The other method is to drain it to the eves HUT. The Refueling Water Purification pump cannot take a suction on the l

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Reference Title *-__JI Facility Referenc~ Nu'!.'_ber I IReference Section j [Page No'. l ~~vision 1

IFill and Transfer of the Spent Fuel Pool IIS1 .OP-SO.SF-0001 I 11 1119 I IUnit 1 Spent Fuel Cooling 11205233 I 11 1130 I I II I 11 11 I IL.o. Number Objective I SFPOOOE010 1--~

[Material Required for Examination 11 II

\auest!on Source: \IFacility Exam Bank I\Question Modification Method: II Editorially Modified I~~d During Training Progra~ D

[§Jestion S~urce Commen§ I I

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I fQUestion Topic I 1RO 47 IWhich of the following describes an interlock on the Unit 2 Spent Fuel Handling Crane? I

~ A high load weight cutoff prevents lifting and movement of loads >2200 lbs. over the spent fuel pool.

~ A high radiation signal on 2R32A, Fuel Handling Crane, will lockout ALL crane motion to prevent excessive radiation dose rate on bridge.

~ A bridge over travel limit will stop the bridge from contacting the hard stops even if the direction button is held down, to prevent damage to the bridge.

I I

[] A high radiation signal on 2R5, FHB-SFP, OR 2R9, FHB-New Fuel Storage Area prevents upward crane movement to prevent exc radiation does rate on bridge.

[Ailswefl Ia I [Exam Level I IR I [Cognitive Level 11 Memory I IFacility: 11 Salem 1 & 2 I IExamDate:-1 I 12/19/20161

[KA:[ I034000A302 I [~3.o2 IROValu~l2.5*lfSROValue:[l 3.1j[Section:ll~~OGroup:[I 2llSROGroup:ij 21 .- [SYStem/Evolution Title j IFuel Handling Equipment System I io34 IKA Statement: I Abilit to monitor automatic o erations of the Fuel Handlin Load limits IEXplanation 55.41.b(10, 12) The overload cutout is provided to prevent lifting loads >2200 lbs. If they can't be lifted, then they can't be moved I.Answers: over the spent fuel in the pool IAW TR 3.9.7, and is tested IAW TR 4.9.7. Bis incorrect because downward crane movement is still available to lower the fuel bundle. C is incorrect because on Unit 2, the proximity switches only transfer the crane to slow, but movement will continue if the direction PB is continued to be depressed. On unit one, it stops movement. The 2R5 and 2R9 are

,-- Reference Title -=1 Facility Reference Number l!§te.rence Section . 11 Page No. [ [Revision L. J ISpent Fuel Pool Manipulations 11S2.0P-IO.ZZ-0010 I Ii 1133 I ITechnical Requirements Manual l!sGS-TRM I 3.9.7 11 TRM 3/4 112 I I u I II 11 I iLO. Number Objectives IIOP010E004 1_ _____..

[Material Required for Examinati~ I II IQuestion Source: 11 Facility Exam Bank I!Question !"odification Method: *.* II Direct From Source I[used During Training Program j D Com~en§J 1111924

[Question Source I

@:omment ~-*'

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[§u;stion Topi~ I RO 48 Unit 2 is performing a plant startup to full power.

Predict the Steam Generator Narrow Range Levels for the following TURBINE power levels.

Assume a normal power ascension.

10% Turbine Power= - -% NR level.

60% Turbine Power=  % NR level.

46%.

=r5%

'b. 38.5%. 44%.

~J 133%. 46%.

@.],33%. 44%.

~swer I Ib l iExam Level 11 R I [Cognitive Levell IApplication I [FacmtY.11 Salem 1 & 2 I IExamDate: 1 I 12/19/20161

~:JI 035000A 101 l[A1.01 jlRO Value: II 3.61~Value:11 3.8! !section: 11..~ [RO Group:jj 21 ~O Group:ll 21 lt!i D jSystem/E;olution Title j !_s_te_a_m_G_e_ne_r_a_to_r_S~y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.I ~ ~~

[<A Statement:l Ability to predict and/or monitor chanQes in parameters associated with operatinQ the Steam Generator System controls includinQ: I SIG wide and narrow range level durina startup, shutdown, and normal ooerations I Answers:

j

~xplanation of 55.41 (4) SG NR level is programmed so that as Turbine power rises from 0-20% (Turbine steam line inlet pressure) it rises from 33-44%. On Unit 1 only,20-100 it rises from 44-48. A is incorrect but plausible if candidate thinks programmed level is based Turbine power but uses Unit 1 ramp from 20-100 (calculated at 60% power. Bis correct. C is incorrect but plausible if it is thought that that the ramp is from 20-100% and uses Unit 1 ramp. Dis incorrect but plausible if it is thought that that the ramp is from 20-100%.

,-- --------11 L__ Reference Title Facility Reference Number _J !Reference Section -] I Page N'OJ ~_visfOO]

IAdvanced Digital Feedwater Control System !INOS05ADFWCS-08 I 1114,17,22 I 18 I I II I 11 11 I I II I ii 11 I ru>.-Number --==:J Objectives ICN&FDWE008 I I

1_ ____,

\Material Required for Examination -. II q I

[Question Source: >[ Facility Exam Bank I!Question Modific~tiol1 Method: II Significantly Modified Iiused During ,Training Program [ D

[QUestion Source Comments] 1125789 Modified "reactor power levels" to "turbine power levels". Changed correct answer from d to b.

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RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes lauestion TqPIC]  ;======~==========~========~========~~~~~~!

_R_0_4_9_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.

1 Given the following conditions:

- Unit 1 is operating at 1x10-8 Amps.

- A small (0.1 %) steam leak occurs in containment on a single SG.

Which of the following describes how this will affect Rx power with NO operator action?

Rx power will rise ....

~ until it stabilizes after reaching the POAH.

l!>:J until the power excursion is terminated by a Rx trip on OT/DT.

~ initially, then stabilize at a power slightly above 1X10-8 Amps.

lc!:J until the power excursion is terminated by a Rx trip on high power (low range).

~~ ~ ;EXam-Levell ~ \Cognitive Level] IApplication 12/19/2016

~I 039000K508 I~= [Ro Value: l[Ifil"[SRoVaiU&:J[Ifil"isection: II~ IRo Gro~LJISRO Group: IL J Bl D lsystem/Evolutioi1}:1ti!J j Main and Reheat Steam System I ~]_9__ .J I

~Statement: ~edqe of the operational implications of the followinq concepts as they apply to the Main and Reheat Steam System: I of steam removal on reactivity I I

!Explanation of 55.41.b(1,5) At 1x10-8 Amps, the reactor is exactly critical. The steam leak will cause Rx power to rise. Before reaching the Point Answers: **-*' of Adding heat, reactor power will not have a temperature coefficient feedback. The power rise will be terminated AFTER reaching the POAH, as the negative reactivity added will offset the positive reactivity from the steam leak. The 2 reactor trips are plausible if it is thought that power will continue to rise uncontrollably, as the OT/DT trip is always active (not re-instated above 10% power as

-L'--- .f.ril"\ir"> ,., ..'"'\ -1 .1.1-. h" -'- - - * *- 11_ ...... - - - ' - .f.ri... .1r1 .__ - . .a.: .. ,,.. r-l--  ; ... ;,.. --t. t... LI. - - - .1...;...,,., '1nO/_ - - * *--

--~-- Reference Title ~I Facility Reference Number I!Reference Section 11 Page No. i [Revisio~

I I 11 11 I I  ! ii 11 i I I 11 11 I

!L.O. Number 1 Objectives IREACOEE001 IMaterial Required for Examination .. 11 11

§uestion Source: I !New I [auesticmModification Method:

• I! 1lused During Training Program ] D j

!Question Source Comments[

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~omment 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

!Question Topic 11 RO 50 Given the following conditions:

- Unit 2 is operating at 46% power steady state.

- The Main Generator trips due to a ground fault.

Which of the following describes how the Main Generator trip affects the plant with NO operator action?

r---:

_aJ A reactor trip will NOT occur. Main steam dumps will open and control rods will insert. Tavg will stabilize -547 °F.

I~ A reactor trip will NOT occur. Main steam dumps will open, and control rods will insert. Tavg will stabilize -552 °F.

~ A reactor trip will occur due to the Main Turbine tripping from the Generator Protection trip. The crew will perform TRIP-1 stabilization and diagnostics actions, then transition to TRIP-3.

[1 A reactor trip will occur due to the Main Turbine tripping from the Generator Protection trip. The crew will perform lmmedate Actions of TRIP-1 then transition to TRIP-2.

!Answer 11 b I [Exam Level J IR I :cognitive Levell IApplication I !Facility: 11 Salem 1 & 2 I IExamDate:-11 12/19/20161

~:]I 045000K301 1IK3.01 j [ROValueJI 2.al fSRO Valu~I 3.11 fsection: fI~ iRO Group~! 2! ISRO Group:jj 21 IR D 15Ystem/Ev~l~tioi!iiiliij j Main Turbine Generator System I [~45 -HJ 1

KA Statement: I Knowledqe of the effect that a loss or malfunction of the Main Turbine Generator System will have on the followinq: I Remainder of the plant I

~~:;;~:-~~] operator 55.41.b(4,5,6,7) The generator trip causes a MT trip, but at less than P-9 (49% Rx power), a turbine trip will NOT occur. With no action, rods will insert in auto to lower Tave. The steam dumps will open in Tavg Control Mode (Load Reject) as Tave rises during the load rejection, and the dumps will modulate to maintain Tave 5 deg above Tref of -563.

i Reference Title -~! Facility Reference Numb~ !Reference Sect~ I Page N-0] IRevision:

IMain Steam Dumps Lesson Plan II NOS05STDUMP-12 I 1135-36 1112 I I I I 11 11 I I I I 11 1 I I

[L~O. Number Objectiv ISTDUMPE008 1_ ___.

I Material Required for Examination 11 q lauestion Source: 11 Facility Exam Bank I!Question Modification Method:

• II Concept Used IlUsed During Trainirig Program j D fQuestion Source Commentsj 157628 (100% power question changed initial power to 46% and added MT trip as the load reject vs SGFP trip)

I Icomment ___J I

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

~uestion Topi~ IRO 51 I Given the following conditions:

- Unit 2 is at 35% power during a plant shutdown performed IAW S2.0P-IO.ZZ-0004, Power Operation.

- 21 SGFP is manually tripped as directed by S2.0P-SO.CN-0002, Steam Generator Feed Pump Operation.

- 24BF19 fails open, and 24 SG NR level rises and remains >67%.

Which of the followino describes the effect of this failure, and how will the crew respond?

~ NO AFW pumps start immediately when 22 SGFP trips. If total AFW flow is <22E4 lbm/hr when checked in TRIP-2, the crew will transition to FRHS-1 FRHS-1

~ NO AFW pumps start immediately when 22 SGFP trips. If total AFW flow is <22E4 lbm/hr when checked in TRIP-2, the crew will start AFW pumps to establish at least 22E4 lbm/hr AFW flow. Flow

~] ONLY 21 and 22 AFW Pumps automatically start when 22 SGFP trips. If total AFW flow is <22E4 lbm/hr when checked in TRIP-2, the crew will transition to FRHS-1 FRHS-1 I

~ ONLY 21 and 22 AFW Pumps automatically start when 22 SGFP trips. If total AFW flow is <22E4 lbm/hr when checked in TRIP-2, the crew will start AFW pumps to establish at least 22E4 lbm/hr AFW flow. Flow I

!Answer i b I I ['Exam Levefl IR I !cognitive Levell IApplication I [Facility: \ ISalem 1 & 2 I \Examo~I 12/19/20161 1 ~1059000A201 lf.:2.01 _JIROValue:J!3.4*lfSROValue:ll3.6*llsection:\j~\ROGroup:\I 11\SROGro~LJ Ill D 1System/Evolution T~ [_M_a_in_Fe_e_d_w_a_te_r_S...:y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___,! @59 --=:J 1KA Statement:] Ability to (a) predict the impacts of the following on the Main Feedwater System and (b) based on those predictions, use procedures to correct, control, or miti ate the consequences of those abnormal o eration:

Feedwater actuation of AFW s stem I

~Exf)lanation of 55.41.b(4) The high SG NR level in one SG initiates Feedwater Isolation. This trips both SGFPs. Normally, a trip of both SGFPs JAnswers~ would start the MDAFW pumps ONLY (21 and 22), but this does NOT occur when the SGFP signal is caused by a FW Isolation.

With power initially at 35%, SG NR levels will remain above the AFW pump auto start level of 9%, so NO AFW pumps will start IMMEDIATELY, they will ALL start when SGNR levels are <9%. If levels did not shrink low enough, then when asked in TRIP-2 I Reference Title

-* ~[ Facility Reference Number **.. I~erence Section 11 Page No.J fRevisiOn]

IFluency List II NOS05FLUNCY-09 II 1114 11 9 I IReactor Trip Response 112-EOP-TRIP-2 JI 1ISh1 l I30 I I II II 11 1 I I i.o. Number Objectives I CN&FDWE006 1,Maferial Required for Examination 11 11

\Question Source:. \ New I I\Question Modification Method: ~ I \Used During Training Program ) D

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Question Sourc:e Commentsj I

[comment ' . __cJ I

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

~estion Topic 1 Ro 52 I I IWhich of the followinl:J identifies how over-coolinl:J of the RCS is prevented on an uncomelicated manual Rx trip from 100% power? I 18.i I P-1 O aciuateo.

~ P-12 actuates.

[(:] I Feedwater Isolation.

~11 Feedwater Interlock.

I

!Answer] Id I J¥xam Level 11 R I ~ognitive Level 11 Memory I IFacility: J ISalem 1 & 2 1 IExamDate: iI 12/19/2016 I

IKA: I 059000K 105 1IK1.05- - - .JROValue:i13.1*llsROValue:ll 3.2lisection:ll~[RoGro~j 1llsROGroup~I 11  !!I] D LSystem/Evolution Tille] j_M_a_in_F_ee_d_w_a_te_r_S_.y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.I !059--

IKA Statement: I Knowledqe of the physical connections and/or cause-effect relationships between Main Feedwater System and the followinq: I RCS I

~~~~~~~-of 55.41.b(?)Feedwater interlock actuates when 3/4 RCS Tavgs <554°F and at least one Reactor Trip and associated bypass breaker open. This shuts the BF19's and BF40 Feed Reg Valves. Feedwater Isolation occurs when 2/3 SG NR levels on 1/4 SG's reaches 67% OR on a SI signal, and shuts BF19s, 40s, 13s, and trips the SGFPs. On an uncomplicated Rx trip this will not occur. P-1 O is 3/4 PRNls <10% power, and blocks the low power Rx trips. P-12 is 3/4 RCS Tavgs <543°F, and will shut the Steam Dump valves.

()n "'.:!in ............... ...,.._1: ................ ...1 4. .. : - ........................ rl 1mne l.t11'll m .... ....l.11"""'" .... fn ..................... 1 T"'.:11\' .... -=iit i:;.A70 -:anrl .&. ................ l.uill ........ 4. .............. a... CA')

i Reference Title I if=acility Reference Number _J [Reference Section_ ] !f>age No. I !Revisionj IReactor Trip Response 112-EOP-TRIP-2 II 119 1130 I I !I II II 11 I I II II 11 11 I Objectives ICN&FDWE006 I Material Required for Examination ii II

§Uestion Source:

• 11 Facility Exam Bank I~uestion Modification MethOci: !I Direct From Source Ilt:!_sed During Training Program D I IQue~tion Source Commen~ 1159006 I

icomment

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

[Question Top~ I RO 53 I j Which of the followin9 describes the desi9n pureose of 21-23AF141, AUX FDWTR PMP INLET AUTO VENT RLSE valves? I

~ To prevent AFW line voiding and water hammer upon AFW pump start.

I

~ To facilitate filling lines upon aligning alternate water sources to suction header.

I

~

IC. To facilitate draining lines upon return from alternate water source alignment to the suction header.

I

@;11.To prevent AFW line temperature gradient if backleakage to AFW from Main Feed occurs with AFW pumps secured.

I

[AOSWer] Ib I ;Exam Level ' ~ I cognitive Levell IMemory I ~a~ility: 1ISalem 1 & 2 I ~I 12/19/20161 fKA=ll 061000K505 I ~.05

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• I !RO Value: 1fEj[sRO Value:JI~J1section:Jj~ !RO Group:l[JJ ~~ Group:1LJ

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1System/Evolutlon Titl;i IAuxiliary I Emergency Feedwater System I *061 --~

IKA State'!li!iliJ Knowledge of the operational implications of the following concepts as they apply to the Auxiliary I Emergency Feedwater System: I Feed line voidina and water hammer I

!Explanation of 55.41.(b)?,8. Alternate water sources can be aligned to the AFW pump suction header through the normally jacked closed AF52 1Answers: Alternate Suction Valves from Demin water, Fresh Water I Fire Protection, and Service Water. During performance of S2.0P-

---***-~~

SO.AF-0001, Auxiliary Feedwater System Operation, Section 5.10, Alternate Sources Alignment, operators are instructed to ensure open the Auto Vent Isolation Valves 21-23AF140. This is to allow the alternate suction line to fill and vent through the auto vents,

.L..:-L.. -.o .L--- .&.I-.- lino i~ *--.&.-..J

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Reference Title

--~i I Facility Reference Numb~ I Reference Section I [Page No. I ~~isionl IAuxiliary Feedwater System Operation 11 S2.0P-SO.AF-0001 I 1165 1140 I IAuxiliary Feedwater System Lesson Plan II NOS05AFW000-15 I I 155 1115 I I II I II 11 I

[Lo. Number Objec IAFWOOOE004 I AFWOOOE013 j_ _____,

IMaterial Required for Examination 11 q

[Question Sourc=._J New I I !Question Modification Method: i I ~uring Training Program ,] D

~~~tion Source Comments] I I

!comment

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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 IRO 54 Given the following conditions:

- Unit 1 is operating at 100% power, steady state.

- The Pressure Override circuit for 11 AFW pump fails, effectively removing the Pressure Override circuit from 11 AFW pump control.

Which of the following describes the effect this failure will have on the AFW system?

~ 11AF21 and 12AF21 will remain shut until manually opened if required.

I

~ , ::, ,;-.;:: , and 14AF21 will remain shut until manually opened if required.

I

~ 21 will open to the position corresponding to current AF21 demand.

@:] ~aod 14AF21 wrn opeo to the po*moo omespoodiog to ourreot AF21 demaod.

iAnsweri Id I [Exam Level] IR I [cognitive Leve* 11 Application I IFacility: .11 Salem 1 & 2 I ~~amDate: JI 12/19/20161

~JI oa1000Kao1 llK6.01 _ ___llROValue:Jl 2.5ljSROVa1~j2.B*!~~l~IROGroup:lj 1jlSROGroup:ll 11 ffll D

~ystem/EVoiUtiOrilltle] j Auxiliary I Emergency Feedwater System I[~~-~ -

I KA Statement: I Knowledqe of the of the effect of a loss or malfunction on the followino will have on the Auxiliarv I Emeroencv Feedwater System:

Controllers and Positioners I

!Explanation of 1 55.41.b(4,7) The Pressure Override circuit provides runout protection for the AFW pumps by maintaining the AFW pump discharge Answers: ~ valves associated with each pump shut until AFW pump reaches a certain pressure, then allows them to open. So during normal standby operation, this circuit keeps the AF21 s shut even though normal demand for the valves is set at -98%. With this circuit removed from affecting the associated AFW21 valves, they will open to current demand. The 11 AFW pump feeds 13 and 14AF21 s n-" *l..,iln..&.t-- '1') /\C:\J\/n1rmn.C---.I- '1'1 ---1 '1') /\C:l")-1~

I Reference Title J[ Facility Reference Number ] !Reference Section I i Page No~ Revision!

IAFW System Lesson Plan II NOS05AFW000-15 I 1127-29 1115 I I II I 11 11 I I II I II 11 I ii.o. Number Objectives IAFWOOOE008 1_ ____.

IMaterial Required for Examination* iI II

~estion Source: 1 INew I!Question Modification Method: ] I !Used During Training Program I D IQuestion Source Comments! I I Icomment I I

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I Jauestion Topic 11 RO 55 Given the following conditions:

- Unit 2 operators are responding to a valid safety injection signal.

- Safety Injection has been RESET.

- ALL SECs have been RESET.

- ALL EDGs are running unloaded.

- 4KV Vital Susses 2A and 2C are aligned to 23 SPT.

- 4KV Vital Bus 28 is aligned to 24 SPT.

Subsequently, 24 SPT experiences a failure causing its secondary voltage to drop and stabilize at 3600 volts.

Which of the following describes the effect of this failure on 28 Vital Bus?

28 Vital Bus will ...

~ fast transfer to 23 SPT.

~ remain loaded onto 24 SPT.

I

~ be energized by its EDG in ACCIDENT loading.

i-:-c

~- be energized by its EDG in BLACKOUT loading.

I

!Answer I !Ci jExam Level 11 R I !Cognitive Level 11 Application I !Facility: 11 Salem 1 & 2 1 IExamDate: I l_ _ _ 12_11_9_12_0_16_,I IKA:/I 062000K302 I~~= IRO Value: IQJJ~RO Value: IG~~JI~ ~~LJ !SRO Group:jLJ D ISystem/Evol_LJ_!i_~ Title I ,_A_.c_.E_l_e_ct_ri_ca_l_D_is_t_rib_u_t_io_n_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.

'1 KA Statemer:!t.J Knowtedoe of the effect that a loss or malfunction of the A.C. Electrical Distribution will have on the followino: l ED/G I

~;~~;t;;:j Ctheissustained 55.41.b(7,8) - A is incorrect because 3600V is not <70%, so the vital bus transfer relay will not energize. B is incorrect because degraded voltage relay will cause a UV signal to be generated for that bus.

incorrect because after SI is RESET, the SEC will not actuate in MODE Ill Dis correct because 3600V is below the setpoint for the degraded voltage (95.%) relays. When these relays actuate then 28 SEC

..:11 -*-;n ')Q I I<'\/ h* *~ <-~m -<<-a~ -- *-- ---' ---' "-- ho,~ ;n ~~/"\r\C: II* /C;n-*- Q, ,~ I I\/\

Reference Title J [ . Facility Reference Numbe~-i Reference Section I 11 Page No. I [RevisiO~

14160 Electrical System I 1 NOS054KVRL Y-00 II 1123-28 11° I I II II 11 11 I I II II II 11 I

[LO. Number Objectives I 4KVACOE006

~erial Required for Examination 11 II IQue~t!on Source: 11 Facility Exam Bank I~estion Modification Method: JI Editorially Modified I[used During Trailiing Progra~ D IQuestion_~_?-urce Comments! 145175 I

[comment

• • - J I

I I

RO SkyScraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes IQuestfunTopicl IR056 While operating at 100% power, the 1A 28 VDC bus becomes deenergized.

Which of the following identifies an effect this failure will have?

~J I

13 AFW pump will start.

~~] OHA G-6, 11 SGFP TRBL will annunciate.

I B Control Room Emergency Lighting will light.

I

'dJ All indication on 1RP4 status panel will be lost.

L~--~

I

!Answer] Id I \Exam Levell IR I [cognitive Level 11 Memory I \Facility: \ ISalem 1 & 2 I /ExamDate:l I 12/19/20161

[!(A.~1063000K302 jlK3.~~ LRo Value: JI 3.51 ~O Value: JI 3.?j /Section: llE!'.~_jfRo Group:[! 1 j jSRO Group:[ I 1 I Ell D

[system/Evolution Title I l_o_.c_._E_le_c_tr_ic_a_ID_is_tn_*b_ut_io_n_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.I ~3 -=I

[!<A Statement:] Knowledge of the effect that a loss or malfunction of the D.C. Electrical Distribution will have on the following:

Components usina de control power I

[Explanation of 55.41.b(?) A is incorrect but plausible since loss of 125VDC control power to MS 132 steam supply valve will start cause it to fail

,Answers:

L.* _ _ _ _ _ open and start 13 AFW pump. Bis incorrect since 115 VAC power is supplied to alarm panel functions. C is incorrect because 125VDC supplies control room emergency lighting. The purpose of the 28VDC Control Power System is to provide DC electrical power under normal, transient, and accident conditions to:

~ ~ . ,.

b.Main Control Room Control Console Bezel controls c.Annunciators d.Vital Instrument Buses e.Non safety-related equipment f.Status panel RP4 in the main control room

,------- I ~-:;:--i ~--,

Reference Title * ** I Facility Reference Number ] iReference Section I ycye No. I 1Revision1 IDC Electrical System Lesson Plan iINOS05DEELEC-09 I 1110-11 11 9 I I#1 Unit 28 voe One Line Drawing 11211357 I 1jsh 1 1115 I I II I II 11 I

~~~~] Objectives I DCELECE008 I I

!Material Required for Examination 11 II

,!Question Source: ..

j INew I!Question Modification Method: *

• i I !Used Our_ing Tr~ining Program J D I

._ *-~--

~--"-----~~.~---~.~

[§uestion source comments1 I

11..0!Tu.,.,iiL .. i I I I I I I

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes lauestion To~ IRO 57 Given the following conditions:

- 2A EOG is running and develops a fuel oil leak.

- 2A EOG Fuel Oil Day Tank level lowers below the start setpoint for both the REGULAR (33") and BACKUP (27") EOG Fuel Oil Transfer Pumps.

- The leak is isolated, and 2A EDG Fuel Oil Day Tank level is recovering.

With 2A EOG Fuel Oil Day Tank level now at 35", which of the following identifies which EOG Fuel Oil Transfer Pumps will be running, if any?

~ NEITHER.

~ BACKUP ONLY.

I

~ REGULAR ONLY.

[] BOTH REGULAR AND BACKUP.

!Answer 1 d l I I Exam Level 11 R I !Cognitive Level ! IApplication I i.Facility: ] ISalem 1 & 2 I I 12/19/20161

[KA: 11064000K103 !l~~~=fiID~l 3.6ISROValue::j 4.0llsection:]j~~~~I 1llSROGroup:ll 11 1!111 D f systemfEvolutio~ej j Emergency Diesel Generators  ! ro64 IKA Statement:! Knowledge of the physical connections and/or cause-effect relationships between Emeroencv Diesel Generators and the followino: I Diesel fuel oil supply svstem I

~::;;~~~-~-:~

55.41 (7) Regular starts at 33" and Backup at 27", as stated in stem. Once started, they BOTH run until the FULL level is reached (44"). The Backup does NOT turn off when the normal level is reached (33"). It is not require to know the exact setpoint at which the pumps turn off, only that it is substantially higher that when the Regular pump turns on, and that the backup pump will turn off at the same level as the Regular pump.

i Reference Title *I c-Facillty Reference Number I [Reference Section 11 Page No. I IRevision INOS05EDG000-12 11 EMERGENCY DIESEL GENERAT I 1144 1112 I I II I 11 11 I I II I 11 11 I f L.O. Number Objectives I EDGOOOE007 iMaterial Required for Examination 11 I iQuestion I

Source: IINew I!Question Modification Method: .** II 1IUsed During Trai_fl!rl~-~rogram I D jauestion Source Comment~ I I

~omment ~-*- . J 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 11 RO 58 Given the following conditions:

- Unit 1 is operating at 100% power.

- 1A EOG starting air receiver 11A is CIT.

- 11 B EOG starting air compressor is CIT.

- A Loss of Off-Site Power occurs.

- 1A EOG trips shortly after starting.

- 11A starting air compressor will NOT run.

By design, what is the minimum number of subsequent 1A EOG start attempts available with the air start system in this configuration?

lc!J

((] 3 re'

~

4

@:J 5

[Answer 11 a I [Exam Level 11 R I [cognitive Level 11 Memory  ! [Facility:-! I Salem 1 & 2 I lexamClate: 11 12/19/20161

~j I064000K607 l[M:o? =IROValue:!l 2.71,SROValue:jl 2.9j~~lsys i!RoG!OUP?JI 11lsROGroup:JI 11 iilfEJ

'1system/EvolutionTitie] IEmergency Diesel Generators 11064 __J

[KA st~~ Knowledge of the of the effect of a loss or malfunction on the followinq will have on the Emerqency Diesel Generators:

Air receivers

!Explanation of! 55.41.b(8) Salem FSAR section 9.5.6, "Each receiver is sized to hold sufficient air for three cold diesel starts". Each EOG has 2 air Answers: receivers, but with one CIT, there will be 2 starts left after 1 attempt at starting.

I Reference Title J CJ:.acility Reference Nu'!'~ [Reference Section Page No] ll§:vislo0111 IEOG Lesson Plan II NOS05EDG000-12 I I[48-49 1112 I IDiesel Engine Auxiliaries 11205241 I 1ISh1 1142 I I !I I ii 11 I

'L.O. Number Objectives EDGOOOE014

!Material Required for Examinatio~ I II

~n Sourcej IFacility Exam Bank I!Question Modification Method: II Editorially Modified Ilt:J:;ed During Trainin_g Program I D

[Question Source Comments[ 180450 I

@"on:iment I I

I I

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

~estion Topic 11 RO 59 Given the following conditions:

- Unit 2 is performing a normal Liquid Release from 21 CVCS Monitor Tank via 22 SW header to Unit 1 CW IAW S2.0P-SO.WL-0001, Release of Radioactive Liquid Waste from 21 CVCS monitor Tank.

- The Radwaste Overboard Discharge Flow Recorder is OPERABLE.

- The 2R18 and the 2R41D rad monitors are OPERABLE.

- After commencing the release, the operator is compiling the Initial Data IAW appropriate attachment.

- When recording the 2R 18 reading, it reads 1OE6 cps.

- The 2R18 red alarm light is lit on the 104 panel.

- The 2WL51 indicates OPEN in the control room.

Which of the following describes what these indications mean, and how the operating crew should proceed?

~ The 2WL51 should have automatically closed on the valid 2R18 high radiation alarm. The NEO should shut 2WL51 locally.

l'b.i

__ j The 2WL51 should have automatically closed on the valid 2R18 high radiation alarm. The NCO in the control room should shut 2WL51.

rs ' 2R18 has failed high. Verify the flow recorder and 2R41 Dare OPERABLE, block the R-18 input to the 2WL51 on 2RP1, and continue the liquid release.

i(j:J

,,_.._..._.;i The 2R18 reading is well below the expected Hi Rad Alarm setpoint. Verify the flow recorder and 2R41 Dare OPERABLE, block the R-18 1 .f. ,f.,... .A.I-- I"\\ Lii t::::.'1 ,.,...,0'1 __ _. -- . , ,,... .a.t-- 1;,.. ,;,..i -- - -

I .- - -- - - . - ' - -- - - -.---- - -- --*

~~ ~ ~xam Level 11 R ' .. Level I 1Cogn1t1ve i 1I Appl1cat1on

. . I Salem 1 & 2 I IFacd1ty: J I LExamDate: l_ _ _12_1_19_12_0_1_,6I J

~I 068000K610 I~= !Ro Value: l@lsRo Value: J0"lsection: II~ ~~~[]lsRO Group:ILJ D

!System/Evolution Title I j_L_iq'""u_id_R_a_d_wa_s_te_S_y.._s_te_m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~

~ell!!liliJ Knowledge of the of the effect of a loss or malfunction on the followino will have on the Liouid Radwaste Svstem:

Radiation monitors I I Explanation of 55.41.b(11, 13)6.82E5 cps is the ALARM setpoint which will automatically shut the 2WL51 (S2.IC-CC.RM-0028) The current i:Answers:

_______ ,, reading is above the setpoint at which the 2WL51 should have automatically shut, and the NCO should shut the valve remotely.

Additionally, S2.0P-SO.WL-0001 for releasing the tank, Step 5.5.9, says if the 2R18 alarms, then the NEO is to inform the NCO to

[ _ Reference Title IRELEASE OF RADIOACTIVE LIQUID WASTE ii S2.0P-SO.WL-0001

-=

shut the 2WL51. There is no provision for closing the valve locally.

if;Cility Reference Num~ [Reference Section II 11 Page No. I 1Revisior1]

1113 1127 I II II 11 11 I I II u 11 11 I l!::O. Number --=1 Objectives I WASLIQE005 I I Material Required for Examination

• 11

'I

\Question Source:

1 1

I Facility Exam Bank I\Question Modification Method: ._.~Editorially Modified I [used D_!:!~ing Training P!ogram I D IQuestion Source_ Comments I 183676 I

[comment 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 11RO60 When performing a Source Check of a Process radiation monitor, which of the following describes how to initiate the Source Check, and why?

Activate the Source Check ......

§] for 30 seconds or less to prevent solenoids from overheating.

~ just until indication of rising level is detected. This will prolong source life.

~ for 30 seconds or less to prevent unwanted automatic action on high radiation level.

@:] just until indication of rising level is detected. This will ensure radiation level outside monitor remains less than detectable.

!Answer J Ia I !exam Level 11 R I !cognitive Level 11 Memory I IFacility: 11 Salem 1&2 1IExamDate: J I 12/19/20161

[KA:JI 073000G132 112.1.32 . : .RO vaiueJI 3.8!-SRO Value: 114.0 I [Seetion:JI~ IRO GrouP?JI 11 [SRO~I 11 mo

~stem/Evolution!~ IProcess Radiation Monitoring System - - * ..***-

1~..=:J IKA Statement]

Ability to explain and apply all system limits and precautions. I rexplanationj 55.41.b(11) Both the system operating and surveillance procedures contain precautions limiting check source operation to 30

~nswers: seconds or less. B is incorrect but plausible because it's a true statement , but it is not the reason. C is incorrect but plausible since there are some monitors whose alarm action functions are not automatically blocked during a source check. D is incorrect because source check operation is not intended to raise rad levels outside component.

[_ Reference Title

.* . 1[ Facility Reference NumbE!r. I [Reference Section . \ [!age No~I [Revision, IRadiation Monitors - Check Sources lIS1 .OP-ST.RM-0001 I 113 1131 I IRadiation Monitoring System Operation 11 S1 .OP-SO.RM-0001 I 114 1138 I I II I 11 11 I jL.O.Number ~ Objectives I RMSOOOE006 I I

I_ ___..

IMaterial Required for Examination 11 q IQue~tion Source:

• 11 New 1Jauestion Modification Method: JI 1lused Dur!ng Training Program I D

@luesfion Source Comme-ntSJ I I

~mment }*.;;;,::-~'

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 Jauestion Topic I 1_R_0_6_1_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~I 1

Given the following conditions:

- Unit 2 is operating at 100% power during a summer heat wave.

- 21, 24, and 26 SW pumps are in service.

- SW header pressure is 106 psig.

- The 2SW308 #2 Bay Overpressure Protection valve slowly fails open.

Which of the following describes how the plant will respond with NO operator action?

OHA's B-13 21 SW HDR PRESS LO and B-14 22 SW HDR PRESS LO will annunciate at 1) _ _ _ psig, and the standby SW pump will auto start at 2) psig.

~ 1)99.5

2) 95.5.

~ 1)99.5

2) 99.5.

~ 2) 95.5. I

- * - :==========================================================='..11 id.! 1) 105.0

• *I 2) 99.5.

[Answer] ~ [EXam!eW!] ~ !cognitive Level J IMemory I I !Facility: 1 Salem 1 &2 I ,_ - -**

1 12/19/20161 1

KA:ll 076000K402 I[Bi~2~ !RO Value: !Q2JlsRO Value:l[J]"~~I~ !RO Group:ILJISRO Group:iLJ E D

~~!!m/Evolution Title I l_s_e_rv_ic_e_W_a_t_er_S_,y:...s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __....I ~76-~-=~

IKA Statement~ Knowledae of Service Water Svstem desian feature(s) and or interlock(s) which provide for the followina: I Automatic start features associated with SWS pump controls I

~nation of I 55.41.b(?) Low pressure 99.5, auto start 95.5 wers:

' Reference Title __JI! Facility Reference Number *"l \Reference Section~ IPage No. '1 jRevlsionl IOverhead Annunciator Window B 11 S2.0P-AR.ZZ-0002 II 1128-29 I 136 I I II II 11 11 I I II II 11 11 I

[i:<fNumber Objectives ISWBAYSE009 1_ _____.

!Material Required for Examination

• 11 II

!Question Source: 11 Facility Exam Bank I!Question Modification Method: ~ Direct From Source I[!Jsed During Training Program I D

[ouesti~n Source Commeni!J 1157024 I

Icomment --

__J I

I I

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

[Question Topic 11 RO 62 I IUnit 2 ECAC is running loaded for testing IAW S2.0P-PT.CA-0001, Emergency Control Air Compressor Functional Test.

Which of the following would have an effect on Control Air Header pressure? I

~l

~

2A 4KV to 460V breaker 2A4D trips.

2C 4KV to 460V breaker 2C4D trips.

I I

~ I2E 4KV to 460V breaker 2E6D trips.

I l<!:] I2H 4KV to 460V breaker 2H5D trips.

I

~swer] !b I iExam Level 11 R I icognitive Level I!Memory I \Facility: 11 Salem 1 & 2 I \ExamDate: 11 12/19/20161 IKA:il 078000K202 1~=1[RqyaiUe:Jl3.3*l\SROValue:il3.5*l~fl:Jj~fROGroup:JI 11\SROGroup:ij 11 D

[System/EvolutiollTitle\ l_1n_s_tr_um_en_t_A_ir_S_.y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _..I ~8 ~

~temeil!J Knowledge of bus power supplies to the following:

Emeraencv air comoressor

!Explanation of j 55.41.b(S) Unit 2 ECAC is powered from 2C 460 volt bus. Station Air Compressors are powered from Group buses.

IAnswe_rs: _ i

~-

Reference Title j, Facility Reference Null")ber ] [Reference Section 11' Page NOl ~evision[

,[

ILoss of 2C 460/230V Vital Bus II S2.0P-AB.460-0003 I 117 11 7 I I II I ii 1 I I I II I 11 l

[LO. Number I CONAIRE005

~

I Objectives "

[Material Required for Examination II 'I jQuestion Source: I !New I!Question Modification Metho.d:

• l I ~ed Durin{! Training Program] D

~ion Source Coinmen~ I I Lcomment _ '

I I

I

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes l1 louestion~ _R_o_6_3_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.I Which of the following describes Station Blackout (SBO) Compressor operation and its connection to the Control Air System during a station blackout?

The SBO Compressor _(1 )__ automatically start on a loss of all station power. After starting, the SBO _(2)_ the 1A and 2A Control Air Headers.

~ 1) will must be manually aligned to 2)

[§;] 1) will 2) automatically supplies

~ 1)will NOT

2) must be manually aligned to I

@J 1)will NOT

2) automatically supplies

!Answer I~ !exam Level I~ I

@:ognitive Levei] Memory I faciiltY:l j Salem 1 & 2 I~ I 12/19/20161 IKA:ll 079000K401 1IK4.01 1 IRo Value: lf))JsRO ValueJ@!SeCtiOO:ll~ IRO Group:JLJ~o Group:ILJ IJj D

'!System/Evolution Title] 1_s_ta_f_1o_n_A_ir_S...:;y_s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.l L079 .=:J IKA Statement: J Knowfedoe of Station Air System design feature(s) and or interlock(s which orovide for the followino:

Cross-connect with IAS I

)Expla~ation of) 55.41.b(S) The SBO Compressor is required to be isolated from, and independent of plant safety related equipment except when Answers: required during operation during a blackout or other situations when there is a total loss of Control Air. The SBO is manually started and manually aligned to the CA headers 1A and 2A.

I Reference Title Ji"

• Facility Reference Number J !Reference Section 11 Page No. I [Revision!

ISBO Diesel Air Compressor !ISC.OP-SO.CA-0001 I 11 1114 I I II I 11 11 I I II I lj 11 I IL.O. Nutnber Objectives I STAAIRE013 1_ ____.

!11r1aterial Required for Examination 11 II

[auestion Source: .. ] New I IJauestionModification Method: 1: I iused During Training Program j D

§uestion Source Comments] I I

!Comment.> ,' +<*~

I I I I I I I

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

~estion ToPiCl RO 64 I Given the following conditions:

- 28 EDG is in service paralleled with the 28 4KV vital bus for a scheduled surveillance run.

- A fire occurs in 28 EDG room, and the fire detector in the room senses the fire.

Which of the following identifies how the Fire Protection system responds, and how the EDG operator should respond?

~ The EDG C02 fire suppression system must be manually activated by the operator as they exit the EDG vestibule area.

~ The EDG Halon fire suppression system must be manually activated by the operator as they exit the EDG vestibule area.

C]

L_

The EDG C02 fire suppression system will automatically discharge after a 13 second delay, the operator must immediately exit the 28 EDG area.

/d. The EDG Halon fire suppression system will automatically discharge after a 13 second delay, the operator must immediately exit the 2: ::::

area.

IAnswefl a I I IExam Level J IR I !Cognitive Level 11 Memory I IFacilitY:-1 ISalem 1 & 2 11examoatel I 12/19/20161

~1086000K504 I[K5:04===11Ro Value: II 2.91 ISRO Value:Jl3.5*1 lsection:J1..~)::'.§__j1Ro Group:JI 2j ISRO Group:lj 21 1111 D

!system/Evolution Title j j_F_ir_e_P_ro_t_e_ct_io_n_S-'y"'"s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.l §1~--:J IKA Statement:! Knowled e of the o erational im lications of the followin Hazards to personnel as a result of fire t e and methods of rotection

.;;:;~iono~ 55.41.b(?) The EDGs are protected by a C02 fire suppression system. This previously AUTOMATIC system was permanently wers: changed to a MANUAL activation system under Salem DCP 80115237. The Fire alarm will alert the operator to the fire, at which

---

time they exit the EDG room/control room to activate the fire suppression system from the vestibule area. There is a 13 second after activation before the system dumps. The Halon distracters are plausible because Salem has other vital areas wlhich have Reference Title J [ Fa!:ility Reference Number j !Reference Section ] [Paga-NO] [REMSIOn]

i==========================~~==================~;:;;::;;============-

~lE=D=G=L=es=so=n=P=la=n================il~N=O=S=05=E=D=G=00=0=-1=2========;~==========~ii 1112 I l==l==================ll~============l~=========il II I 1-------~ll l, _ _ _~11 II I

[~:_~.!!!:l~~.!:__ ____ J Objectives IFIRPROE006 I I

I IMaterial Required for Examination *dI II jQuestion Source: I I New I[auestion Modification MethQd: 11 I [used During Training Program I D

~ues~on S~urce Comment~ I I 1C .....,, .. ,, ..... ~u. **'

I I

I I

RO SkyScraper I. .

SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I la\Jestion Topic 11 RO 65 I Given the following conditions:

- Unit 2 was operating at 100% power when a steam leak upstream of 22MS167 occurred.

- The Rx was tripped and a MSLI performed successfully.

- Operators have transitioned out of EOP-TRIP-1.

- The PO is attempting to open 21-24SS94s, SG B/D Sample Valves, but they will not open.

- SGBD sample isolation bypass has been RESET.

Of the following, which identifies the reason the valves won't open?

~ 22 SG NR level is <9%. 9%.

I

~ CA330s have not been reopened.

I

~ ~ .. '.:'.:FPs have trip signals locked in.

I

~ Either Train A or Train B Phase A isolation failed to reset when its reset PB was depressed.

I IAnsw!':J d I I IExam Levell IR I ICognitive Level 11 Application j !Facility: 11 Salem 1 & 2 I ExamDate: 11 f 12/19/20161

~I 103000A404 I A4.04 I[ROYalUel§lsRO Value: i[ill"[Section: II~ !RO Group:JLJ[SRO Group:[ L J Im I

D

!system/Evolution Titre] IContainment System 11103 J IKA Statement: I Ability to manually operate and/or monitor in the control room:

Phase A and phase B resets

'Explanation of 55.41.b(?) For a trip and SI due to a single faulted SG (unisolable) the flow path will go from TRIP-1 to LOSC-1. The SI will NOT Answers: have been reset in TRIP-1, nor will it be reset in LOSC-1. The SGBD sample isolation reset will be performed in LOSC-1 (step 6.1) in order to open the SS94's. The step prior to that is RESET PHASE A. This is due to the fact that the blowdown isolation bypass only bypasses the lo-lo level input into the AFW auto start circuit, which closes the SS94's. If the Phase A hasn't been reset, the nA- ~"In --.1. ho C:C:QA ..... c-1 *--l!-..I ""lir &--....,.. l"\I 1tc-~..I ..... ........... .&.

i Reference Title l[ Facility Reference Number I !Reference Section 11 Page No. I IRevisionJ IRPS AFW Startup Logic Diagram 11221064 !I 11 1 js I ISS94 Loop Diagram ll 621216-1 II II lj 1 I I II II 11 11 I r;---~**

1L.O. Number RXPROTE019 IMaterial Required for Examination ..: 11

'I

!Question Source: J IFacility Exam Bank IJQuf;)~~ion M~dification Method: JI Direct From Source 1IU!;;ed During Training Program I D

!Question So_urce Comments,l 162124 I

',i; Icomment **,: (::i ;, ~*

,;',* I I I I

! I

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

[Question Topic 11 RO 66 I During shift turnover, the on-coming RO notices an OHA with reflash capability has 1 (one) piece of red translucent tape diagonally across its window box.

Which of the following describes what the status of this OHA is?

~ The alarm is inoperable, and will not annunciate under any circumstances.

~ The alarm is identified for heightened awareness, and has all functionality present.

~ The alarm has at least one, but not all, inputs disabled, and may not be a reliable source of information.

I

@J IThe alarms reflash capability has been defeated such that if already in alarm, a second alarm will not be annunciated.

I

[Answer 11 c I [exam Level 11 R I ~ognitive Level 11 Memory I !Facility: 11 Salem 1 & 2 I IExamoaTel I 12/19/20161 f KA:[l 194001G101 If2T1 __J /Ro Value: JI 3.81 [SRO Value: 114.2 j iSection: i PWG I!Ro Group:!I 1 j [SRO Group:rj 11

• D jsystem/Evolution Title I IKA Statement: I ..-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

Knowledge of conduct of operations reouirements. I 1

1 Expfanation,of j 55.41.b(10) The single piece of tape is placed on the OHA to alert the operator that the alarm is not a reliable source of jAnswers:

• J information. 2 pieces of red tape in a "X" signifies that the entire OHA window is INOPERABLE. Additionally, if one or more inputs to a multiple input annunciator are inoperable, then red tape should be placed diagonally across the annunciator window. A is incorrect because the window can still alarm from an operable input, since it is not ("X'd). B is incorrect because the tape signifies a.L LL- ic ...,,......,..,....f.hinn .,..,........,,.. .* ,:+h it C!n jf ...I .... ,... ... f\J()T h,....,,... f1 1fl f1 ,,...,..,f.;,.. ... ,..i; ... , r": ic ,..,... .... ,... ....... ,......, ,,...,...., ,...i......,,.,,... n :,... * ,,...,.... *h,...

reflash capability is not defeated, and as long as a second valid input comes in with one already in, the reflash capability of the I alarm will cause it to annunciate.

c= Reference Title . ~I *Facility-Reference Numb~ [Reference Section *

===========================~====::::::::;;:;;:;:;;;::::=::;;:::::::::=::::=:::=;;::::============:.

I L'§i~ ~~~~

=====:::::::;ll.~======:;l i=la==:::;l I 2 I

1 l Operator Burdens Program 11:.:::o=P=-AA==-=10=2=-1::03=-=10=0=1

~1::::::::::::::::::11~============-1!~=========1~1 l, _ _ _ _ _ _ _ ___.11 11 _ _ ____.ll_~l l_

===1~1 ___.i

! *1 L.O. Number ___.__:__J Objectives I

1 CONDOPE005 I I,_ _ ____,

iMaterial Required for Exa111ination 11 'I

!Question Source: 11 Facility Exam Bank I!Ques~ion Modification Method: *

• 11 Direct From Source I \Used Duri11g Training Program j D iQuestiC>n Source CommentS] I I

~omment I I

I I

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

[Question Topic I 1,....R_0_6_7_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.

Given the following conditions:

- Unit 2 is operating at 60% power.

- Power was reduced when 22 SGFP tripped 2 days ago, and has remained at this level for 2 days.

- Operators are preparing to start a Main Turbine power ascension using dilution and automatic rod control to maintain Tavg and AFD on program.

- Tavg-Tref deviation is 0°F.

IAW OP-AA-300, Reactivity Management, how should the power ascension be started?

Assume this is a normal power ascension with all equipment available.

lc!J The crew concurrently initiates the Main Turbine load ascension and a RCS dilution.

~ The crew initiates a RCS dilution. As soon as the dilution is in progress the Main Turbine load ascension is initiated.

~ The crew initiates a RCS dilution and waits until a RCS temperature rise is detected. Then the Main Turbine load ascension is initiated.

[ ] The crew initiates the Main Turbine load ascension and waits until a RCS temperature lowering is detected. Then the RCS dilution is ini"iti"ated. I IAoswer I fC=J !Exam Level I fR=J' lcoqniti:ye Level I !Memory I !Facility: WSalem 1&2 l~Examoa:§] _ _ _12_1_19_12_0_1~6I

~I 194001G102 I[2.1.2 I [ROVaiUe:l@fsRO Value:ifJ1section: fj~ ~~LJ~RO Group:f L J D !System/Evolution Title I IKA Statement* I Knowledqe of operator responsibilities durinq all modes of plant operation. I I IEx:planation of 55.41(5, 10) Listed under the responsibilities of the Reactor Operator, page 7, 4.6.5, "Typically, during planned load changes where Answers: j dilution or boration is required, start with the dilution or boration. The initial effects (RCS temperature change) of the reactivity change should be seen prior to initiating the load change. (PWR)" All the distracters have the correct actions in the wrong order. i I Reference Title II Facility ~eference Number J IR£1ference Section 11 Page No. f 1Revision1 IReactivity Management 11 OP-AA-300 II 111 117 I I II II 11 lI I I II II 11 11 I [LO._~umber Objecti I CONDOPE005 '--~ IMllterial Required for Exarnin(l;tion 11 II !Question Source: 11 Facility Exam Bank I jQuestion Modification Meth " - irect From Source IIUsed During Tra!O!ng Prog~I i1Question Source Comments\ 1125830 [Coron.;-;... *.  :* ...*;:. *.*.*.

                                                                                                          ..        i I                                                                                                                    I I                                                                                                                    I I                                                                                                                    I

RO Skyscraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [Ql!"esticm Topi,c J ,_R_0_6_s_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____. IWhich of the following conditions will REQUIRE the suspension of fuel movement in the Unit 2 Rx vessel?

     ~ Chemistry reports Rx Cavity boron concentration is 2499 ppm.
     ~ A NEO reports BOTH 100' elevation containment airlock doors are open.
     ~ The PO depresses Fire Outside Control Room on Unit 2 Control Area Ventilation.

rd.J

     ~

Containment Radiation Monitor 2R12A fails causing a Containment Ventilation Isolation signal. [Answef] ~[Exam Level I~ [cognitive Level J jApplication alem 1 & 2 12/19/20161 IKA:jl194001G140 112.1.40 1[ROValue:ll 2.s1\sROVa~l3.9 l!Section:llPWG llROGroup:jj 1lfSROGroup:ll 11 ml D r.--*-: [system/Evolution Title I ----------------------------------------'~NER~I rKA Statement*..I Knowledoe of refuelino administrative requirements. IExplanation 'Answers: j 55.41.b(11) C is correct because operation in the recirculation mode requires suspension of fuel movement. (SO.CAV P&L 3.6.3 When aligned to FIRE OUTSIDE CONTROL AREA (Recirculation Mode),Core Alterations and movement of irradiated fuel is NOT ~----*- permitted (TIS Bases 3/4.7.6)). Dis incorrect because Containment Radiation monitors are not required to be operable for Mode 6 or Fuel Movement or Core Alts per Tech Specs. B is incorrect because the airlock doors are only required to be CAPABLE of h - ' - - -h.,t <>--' --- ho - - - - IQ') ()D _QT ('A "-(\(\(\7 ---- P.\ f; ie * ._ ___ , *-- "-- ('()I 0 limit*-*._ ____ ie '11".lQ ! Reference Title ILFacility Reference Number 11 Reference Section **. ]IPage]OJ IRevision[ ISalem Tech Specs !I I II 11 I IRefueling Operations-Containment Closure 11 S2.0P-ST.CAN-0007 I Ila 1 j2s I IControl Area Ventilation Operation 11 S2.0P-SO.CAV-0001 I 1 ls , , 41 I r:-:: 1L.O. Number J Objectives I REFUELE012 \Material Required for ,:.Xamination : 11

                                                                                                                                                                           'I I

!Question Source;. ] Facility Exam Ba~ lfi"uestion Modific,ation Method; . *~Direct From Source j jused During TrainingProgram] D [Question Sourc_e Comments! 1110741 I ~mment .*"-.. I I I I

RO Skyscraper I SRO Skyscraper t RO System/Evolution List I . SRO System/Evolution List I Outline Changes [QuestionToPicl _R_0_6_9_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___. IWhen performing a plant startup, when is MODE 2, STARTUP, entered IAW S2.0P-IO.ZZ-0003, Hot Standby to Minimum Load? ol Bank withdrawal is imminent. When the Reactor Trip Breakers are closed. When Shutdown Bank withdrawal is imminent. Level ~!cognitive Level 11 Memory  ! :Facility: ISalem 1 & 2 1 IExamDate: 1 I 12/19/20161 IKA:]j 194001G235 ~~~'RO Value: II 3.61[sROValue:Jl4.5 I !Section: II PWG 1 IRO Group:ll 11 ISRO Group:ll 11 . . D [SYStem/Evolu-tio_n_T_it-le~J

                                 ....-------------------------------------- GENERI                                                                               -~---*-

IKA Statement* I .--~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~! Ability to determine Technical Specification Mode of Operation. Explanation of 55.41.b(6) Step 4.2.21.3 When the withdrawal of Control Bank "A" is imminent, PERFORM the following: Answers: RECORD time of Mode 2 entry in the Control Room Narrative Log. UPDATE WCM to Mode 2. RECORD Date and Time of Control Banks withdrawal in Attachment 3, Technical Specifications Pre-Criticality Surveillance Data,

                      -         l1 Reference Title .                      j[    Facility Reference Number               11 Reference Section        Ir Page No.l iRevisionl

~::=::=::=::=::=::=::=::=::=::=::=::=::=::=~~::=;;;;;;;;;;;::=o=;;;;;;;;;;;;;;;;;;;;;;;;::=::=::=::=;;;;;;;;;;;.~::=::=::=::=::=::==:'. ~IH=o=tS=ra=n=d=by=t=o=M=in=im=u=m=L=oa=d==========~il:=S=2.=0=P=-IO=.=ZZ=-=00=0=3=========11.:.============='~11=s==:=:::l~l4=3====I I II II II 1:.:1=~I l, _ _ _ _ _ _ _ ___.q _ _ _ _ ____.ll, _ _ ___,lj _ __.1 l_ ___.I jL.O. Number Objectives IIOP003E004

 !Material Required for Examination            *l I                                                                                                                        II
 !Question Source: 11 Facility Exam Bank                I!Question Modification Method:

• 11 Direct From Source I!Used During Training Program [ D IQuestion So~_rce Comments 11155638 I

IComment I I I I

RO SkyScraperJ ***SRO Skyscraper I RO System/Evolution List I SRO System/Evoluti~n Lis~ J Outline Chan~es J [ci4e~tipiWropi~1 l,_R_o_10_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___.I Given the following conditions:

    - Unit 2 is operating at 100% power.
    - 100% AFD Target is -1.5.
    - Between 0100 - 0140 today, 40 AFD penalty minutes were accumulated during a Main Turbine Valve Test which caused a rapid power reduction to be performed when a steam valve could not be re-opened initially.
    - At 2300, 22 SGFP trips.
    - At 2301, Rx power is 87% and lowering with AFD at -9.1.
    - At 2303, Rx power is 66% and the Turbine load reduction is complete, with AFD at -10.1 Assuming power and AFD remain at 66% and -10.1 respectively, which of the following describes the required power reduction the crew must take IAW Salem Tech Specs?

Power must be reduced less than 50% by ....... .

1::::

      ~      12351.
      @] 12353:* . . . . . ***********- . ............... ....                                             -- ... .. ----                  --- *-*-*                              -***** . . .                                  **-** ...

I fAJ,$1.vef I @=] IExan,J L~~el l ~ l.¢ognitive*L:eye~;d IApplication I ~~:iicill~;I ISalem 1 & 2 I !e~;'11P~tetj I 12119120161 it<A:J 1194001 G239 Ii2.2.39 I !.RP Y;:th.ie~i [ifil" t§B.9.V:~1ue: I[B [$~sti0[}: 11~ ~RO.C?rollp:j LJ t§J~Q:t;?.rpllp:I LJ Iii D tsY5tem/Evo1Utipn :t"!#e: I * - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' IGENERI !

! l<A Statemenf1 Knowledqe of less than or equal to one hour Technical Specification action statements for svstems.

[EXplanatiqh'ot* 55.41.b(5,6)LCO 3.2.1 action a (for >90% power) and action b (50-90% power) applies. The MT auto runback upon a SGFP trip [Answers: .*:

• occurs at 15% per minute. The AFD Target at 100% power (Rev. 18 of Tables) is -1.5, with a band of +6, -9. The target is adjusted for Rx power, so with power at 87%, Target is -1.3, and band of -9 = -10.3, so at 2301 AFD is still within the Target Band. With power at 66%, Target is -0.99, and band -9= -9.99. With indicated AFD at -10.1, AFD is out of band at 2303, and Penalty Minutes

                                      - - - * * -

• 11-.fo '1 ,r._ *- -*:- .+ .... ,..., ,.f,,..."..I ~-

                                                                                                                 - - o--"
                                                                                                                          \A  ~ ... ._ A/"\ ....... i .... ,

outside of Target Band at 2303+20=2323. The action for outside of band for 60 minutes of penalty minutes (but still within COLR ti . .1. _. .,....l")A

                                                                                                                                                                                                            -*--*   :11        An-:-**--

limits) is to lower power to <50% within 30 minutes. So Rx Thermal power must be <50% with 30 minutes of 2323, which is 2353. The distracters are using the 15 minute action time required for power >90%, or using the 2301 time as the time AFD is out of band (plausible if the -9 of Tarqet Band is incorrectly used as the actual AFD limit) or both. 1*.;.  ::;,::Y{/, :\: >. Reference riile"'**j1 . *\'

                   , '""""::i:;>-' *;

~;

; ;;;.~::/,

                                                                                          *1 !:+/-2 ~.t:~si1HyReferens~...~l.lri16e~1::,:;:':I !Ref~.renee'"secu<>rt:. j IPagE;::N6FI                             !R"elt,isio11I ISalem Tech Specs                                                                           11                                                                llLC03.2.1                          I13/4 2-1    11218          I

!rabies iIS2.RE-RA.ZZ-0011 II II 11248 I ISalem Core Operating Limits Report !ICOLR Salem 2 II 116 11 6 I I FLUNCYE002 Objectives

RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I IQuestionTopic J IRO 71 I IWhich Emergency Classification results in an automatic extension of annual dose (TEDE) limits for ERO personnel that have an NRC Form-4 on file, and what is the dose limit extended to? I

   ~      SITE AREA EMERGENCY, 5000 mrem.

c--*1 I

   ~ SITE AREA EMERGENCY, 4500 mrem.

I [J IALERT, 5000 mrem. I

   @;:]I ALERT, 4500 mrem.

I !Answer 11 d I !exam Level ! IR I rcognitive L~vel 11 Memory I )Facility: 11 Salem 1 & 2 i IExamDate: 1 I 12/19/20161

KA:]l 194001G304 112.3.4 ==1 [RO Value: II 3.2j ~RO Value:ll 3.7 Ilsection: i PWG I:Ro Groi:liJI 1j !SRO Group:ll 11 llJI D ISystem/Evolutio~ * - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' ~~ENERTJ

[!<A Statement* I Knowtedoe of radiation exposure limits under normal or emeroencv conditions. I IExplanation of I 55.41.b(12, 10) Upon an ALERT declaration, workers with a NRC Form 4 on file have their dose limit raised from 2,000 mrem to i~_!~r~ 4500 mrem. I Reference Title I! .Facility Reference Number~ lReference Section 11 Page No. i !Revision! IOPERATIONAL SUPPORT CENTER (OSC) RAii NC.EP-EP.ZZ-0304 II q 1116 I I II ll ii 11 I I II II 11 11 I IL.O. Number Objectives IRADCONE003 IMaterial Require~ for. Examination_J I II I [Question Source:.* [ Facility Exam Bank I[auestion Modification Method: II Direct From Source IIUsecl.During Training Prog~ D [Question Source Comments[ 1135687 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

 !Question Topic]   I RO 72                                                                                                                                           I A Containment Entry must be made in Mode 1.

When is Radiation Protection Supervisor (RPS) approval required AND what is the limit for the number of personnel that may enter containment? Assume the Shift Manager has NOT granted permission to exceed the procedural limit The RPS must qive prior approval if the entrv is made durinq ....

    ~   a power ascension at >5% I hr ONLY, and no more than 10 people may be in containment at one time.

I

    ~   any reactor power change at >5% I hr, and no more than 10 people may be in containment at one time.

II I

    ~ a power ascension at >5% I hr ONLY, and no more than 20 people may be in containment at one time.

I [] any reactor power change at >5% I hr, and no more than 20 people may be in containment at one time. I

!Answer 1 Id     I [Eiam Level J R  I           I ~ive Level 11 Memory                     I [Facility: 11Salem1 & 2             I~               I        12/19/20161 IKA:/l 194001G313          I[2.3.13 =11ROV~lue:jj              3AliSROValue:jj 3.B!lsection:llPWG liROGroup:il                    11!sROGroue:JI         11 Ill D jSystem/Evolution Title   I I                                                                                                                              j (GENERI j iKA Statement*..I Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entrv requirements, fuel handlina responsibilities, access to locked hiah-radiation areas, aligning filters, etc.                     I

[Explanation Ofj 55.41.b(12) SC.SA-ST.ZZ-0001, (Rev. 5) SALEM CONTAINMENT ENTRIES IN MODES 1 THROUGH 4, requires prior RP 1Answers: . .~ Supervisor approval to enter containment in Modes 1 or 2 during ANY power change >5% I hr. It also limit entry to 10 people per operable air lock. 2 airlocks = 20 people. See steps 2.4 & 3.2. I i Reference Title 11 Facility Reference Number * \ !Reference .Section * [~o:l. !Revision\ ISALEM CONTAINMENT ENTRIES IN MODES IISC.SA-ST.ZZ-0001 I 113,4 11 5 I I II I ii ,11, I I II 11 11 I 1L.O. Number Objectives I CONTMTE012 [Material Required tor Examination .*I I 'I

\Question Source:     1 IFacility Exam Bank           Ii.~ufi!st1011 Mocliflcation Method.:   JI Editorially Modified      11 i.Jsecl During Training Program I     D
\Question Source Comments! 1159346 MODIFIED ANSWERS FROM "5%" TO "5% I HR."

I I

                                                                      )'c jcomment                            . 'f~<'

I I I

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

  ~Question Topic*! IRO 73                                                                                                                                                                     I IWhich of the followin~ contains conditions that warrant ent!l'. to EOP-LOCA-5 "Loss of Emer~encx Coolant Recirculation"?                                                                     l
    ~    RWST LOW LEVEL alarm annunciates with <62% level in Containment Sump.

[§J The hot leg recirculation flow path is unavailable due to blockage of flow paths. [] RWST LOW-LOW LEVEL alarm annunciates requiring all ECCS pumps with suction from the RWST to be stopped.

    @;]  LOCA outside containment is indicated, and both RHR pumps were stopped IAW EOP-LOCA-6, LOCA OUTSIDE CONTAINMENT.
~swer 11 a        I IExam Level 11 R                 I Icognitive Level               11 Application          I~ !salem 1 &2                     I IExamDate: 11                    12/19/20161 jKA:lj 194001G404                I[2.4.4 -~!RO Value~j              4.51 ISRO Value:Jl 4.7 j ISection: II PWG liRO Group:ll I

11 ISRO Group:ll 11 11::.i)l! D 1system!Evolution T~

                                     !------------------------------------'[GEN~

IKA Statement: I Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operatinq procedures. I.Explanation of 55.41.b(10) EOP-LOCA-5 can be entered from 5 places. 1. LOCA-1 step 16 with no RHR pump and associated SJ44 available; 2. Answers: LOCA-3 step 2, Containment sump level less than 62%; 3. LOCA-4 step 5, No RHR pumps running; 4. LOCA-6 step 6.2, with RH1 ,2,26 shut, 21 and 22 RH19 shut, 21 and 22SJ49's shut and RCS pressure not rising; 5. Loop from LOCA-5 step 28 when RWST level is still above LO-LO (1.2') setpoint. Answer is correct because containment sump level should be rising as RWST

                                     **--!-- r- ;.....             *- . . . . . . =* :........   ..        ..-**In h - .

1- *- : .... 1- L

                                                                                                                                                                 . . n .. r;n,,... r'I *--**-

tn ---*or"" R\11/C::T '-* *-' -'*-----' as 1 CS pp continued to draw water from RWST. D is incorrect because LOCA 5 only requires RHR pumps to be available, not in service. B is incorrect because there is no provision in LOCA-4 to check if the flowpath is blocked or not. It only has the alignment performed, then return to procedure in effect. I Reference Title I Facility Reference Number Ii!§ference Section 1 I Page No.! !Revision: ILoss of Emergency Recirculation 112-EOP-LOCA-5 I 111 1130 I ITransfer to Cold Leg Recirculation 112-EOP-LOCA-3 I 111 1130 I I II I 11 11 I IL.O. Number I LCA3U1E007 Objectives 1_ ____.

!Material Required for Examination 11 I j [Question Sourg Facility Exam Bank I~estion MC>~ification Method~_j Editorially Modified I[~ed During Training Program J D [QUe-stlon Source Comm~ 142264 . Changed correct answer wording from "RWST inventory is being depleted .... " to "RWST LOW LEVEL alarm annunciates ... ". I [COmment ... . *. I I I I

RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I jQuesti<>n Topic I 1,_R_0_74_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~1 Unit 2 has entered S2.0P-AB.STM-0001 EXCESSIVE STEAM FLOW due to an extraction steam line rupture. A Reactor trip and MSLI have been performed, and the steam leak has been isolated. Safety Injection was not required and did not initiate. Which of the following describes how S2.0P-AB.STM-0001 will be used following performance of EOP-TRIP-1, Rx Trip or Safety Injection? S2.0P-AB.STM-0001 .....

    ~ must be re-entered from point it was left and completed regardless if entry conditions are still met.
    ~ must be re-entered from the beginning and completed regardless if entry conditions are still met.

uld be re-evaluated to see if entry conditions are still met. If they are, re-enter AB from the beginning. I Id be re-evaluated to see if entry conditions are still met. If they are, re-enter the procedure from where it was left at Rx trip. I [AOSWe!J ~ [Exam Level I ~ [Cognitive Level 11 Memory I ffiCmtY:l lSalem 1 & 2 I rexa~ I 12/19/20161 r@1194001G411 I§ rfROVarueJI 4.01JsROValue:Jl4.2 1isee~1PWG ljROGroup:JI 1![ROGroup:JI 11 iifEJ jsystem/Evolution Title I ---------------------------------------' ~~NERI I [KA Statement* I Knowledge of abnormal condition procedures. lexpranat~_:j 55.41.b(10) Step 4.4.1 If a reactor trip occurs while performing an AOP, either intentionally or as directed by the AOP, the AOP should be exited and the EOP entered. If the AOP calls for a trip of the reactor, the action is defined in the same manner as in 4.2.5 s: (fourth bullet, EOP-TRIP-1) for EOP action steps. There are cases where steps must be completed as directed in the AOP (i.e. TRIP affected RCPs) prior to entering the EOPs. Once this action is completed, then entry into EOPTRIP-1 is directed by procedure I Cr'\O, - --1'. - .r, __ .a.L A-Ml *'- 1ol....I h,..

                                           "~~
                             - ;,...,. .LL            LL -              -~-~&. . -1 ,.. ...                       '- - - - ...... ....,,, ..... *   *'-'- ;Li-re-evaluated to determine if performance of the AOP is stili" required,(i.e. the reactor trip did not terminate the abnormal condition). If performance is required, the AOP should be re-entered from the beoinninQ.

L ..._. Reference Title

                                    -                      II        Facility Reference Number                  I!Reference Section                            _J[Page~~~

II OP-AA-101-111-1003

                 ""~*

j Use of Procedures I If 31 11 6 I I II I 11 11 I I !I I 11 11 I ,..-~----~* il.O. Number Objectives I PROCEDE006 IMaterial Required for Examination

• 11 II

\Question Soun::e:

• 11 Facility Exam Bank I\Question Modifi~tion MethOd: ~ Editorially Modified I!Used During training Progra!!!J D fQUestion Source CommentsJ 172934 changed from following TRIP-2 to TRIP-1 based on new rev to EOPs I

jcomment

• I I

I I

RO Skyscraper I . SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I jQu~stion '(opic 11 RO 75 I Given the following conditions:

   - Unit 2 is operating at 100% power steady state.
   -  The unit has achieved "black board" status on the OHA's with NO OHA's currently in alarm.
   -  An expected OHA is NOT received when opening a panel which provides an OHA.

Which of the following describes the control room response IAW S2.0P-AB.ANN-0001, Loss of Overhead Annunciator System, if it is determined that a complete and total loss of ALL Overhead Annunciators has occurred and cannot be restored within 15 minutes?

     ~    Initiate alternate alarm check of the OHA system every 15 minutes by opening and closing a RPS or SSPS cabinet door.

lg] Initiate a controlled shutdown at <5% per minute to bring the unit off line. Line.

     ~    Initiate continuous control console walkdown.

I

     @J 'Trip the Rx and GO TO EOP-TRIP.

I (exam Level [ j R I !cognitive Level I !Memory I

!Answer 11 c                                                                                                   jsalem 1 & 2         I \ExamDate: 11         12/19/20161 IKA:[l 194001G432              ![2.4.~!~0Value:ll 3.6l/SROValue~j4.0 l[section:JIPWG                              liROGro;p!]I       1!jSROGroup:ll      1j 1111 D lsystem/Evolutio'n Titli]

fKA Statement: I Knowledoe of operator resoonse to loss of all annunciators. IExplanation of I 55.41.b(?) With a loss of all annunciators, tripping the Rx or initialing a power reduction are not warranted unless overriding plant Answers: . j status would direct it, not just from the annunciator problem. Initiating continuous walkdowns of control consoles(and P-250, etc.) is directed. With a total loss of OHA's, you don't check functionality by trying to initiate an alarm every 15 minutes, that's performed to verify status if some functionality is present. I Reference Title II Facility Reference Number **I !Reference Section I !'Page No. I [Revision! ILoss of Overhead Annunciator System JI S2.0P-AB.ANN-0001 I q14 1123 I i II I 11 11 I I II I II 11 I )LO. Number Objectives I ABANN1 E002

!Material Required for Examination            JI                                                                                                                       II I

IQuestion Source:> f Facility Exam Bank I[Question Modification t'v1eth0d: * ] Direct From Source I!Used bl.Iring Training Program ID

!Question Source C()mments) 1155633 I
~omment                                            ',_..<);
                                                                                             '1',

I I I I

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List . Outline Changes IQIJe~ti(>n,IopiWI I SRO 76 I Given the following conditions:

  .. Unit 1 is operating at 100% power.
  .. An automatic Rx trip occurs when a loose wire in Main Power transformer Phase B causes a Main Generator trip.
  .. The RO performs 1-EOP-TRIP-1 Reactor Trip or Safety Injection immediate actions with no SI required.
  .. The CRS transitions to 1-EOP-TRIP-2, Reactor Trip Response after verification of immediate actions.
  .. Only TRIP-1 Immediate Actions were performed prior to transitioning to TRIP-2 .
  .. Upon entry into EOP-TRIP-2, the PO reports:
     .. 11 AFW pump has failed to start .
     .. 12 AFW pump has started but a malfunctioning Pressure Override circuit is keeping 11AF21and12AF21 shut.
     .. 13 AFW pump tripped as it was accelerating during start.
     .. All SG NR levels are off-scale low.
     .. SPDS indicates a Heat Sink Red Path exists .

Which of the following describes the next action the CRS should take?

     ~I Direct the PO to establish Aux Feedwater flow IAW 1-EOP-TRIP-2, Reactor Trip Response.

I rhl I nirAr.t thA P(") tn "c:t"'hli,:;h ~bin r A""' JAW 1- i=nP.. TRIP-? P""'"t"' Trin - I

     ~I                                                                                                                                                                                                                        I
     ;sJ    Immediately transition to 1-EOP-FRHS-1, Loss of Heat Sink, ONLY if 3 of 4 SG WR levels are <32% to establish SG Bleed and Feed.

[ ] Immediately transition to 1-EOP-FRHS-1, Loss of Heat Sink, based on the SPDS indication of a valid Red Path, and perform a SGFP Prompt Recovery IAW S2.0P-SO.CN-0007, SGFP Prompt Recovery. ~Ansv:Jerl fQ" !el(alnt;evefl ~ !cognitlveLeveC.j IApplication I [f:acln1.Y:l lsa1em 1 &2 1l~l(?~l)~t~;J I 12119120161 ~ !000007 A202 !IEA2.02 I l'i~o Value: !@!$R(l YalueL..~.~ isectiori:J j~ !Ro Groi.lp~]LJ :SR() Group:! LJ IBj ~ jsystem/Evolutipn Jitle I IReactor T r i p i_0_07_ _ 'KA statement: I Ability to determine and interpret the following as they apply to Reactor Trip: I Proper actions to be taken if the automatic safety functions have not taken place I !Explanation ofl 55.43.b(5) CFST's become active upon exit from EOP-TRIP-1. However, from CFST procedure .... "SPDS is not designed to be Answers:

• 1 used as a primary indication, and no actions should be based upon SPDS indications without verification of the primary indications, which are the installed Control Room 1E instruments." The crew ould perform the actions at Step 3 to check total AFW flow >22E4 lbm/hr, and if not, start 11-13 AFW pumps as necessary to establish it, which would include defeating Pressure Override circuit.

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transition to another procedure) and the actions i~ that pro cedure. The FRHS-1 distracters are plausible 0 if incorrect application of CFST usage after transition out of TRIP-1 are applied. l i /;:;:,'.,' ReferenceJitle ,:.'.\:*:::f;',',  :;.; :I Facility Reference Numb.er , IIReteren*~e ~ecifor(i' 11 Pag~ !Revision[ j Critical Safety Function Status Trees II EOP-CFST-1 II 11 1131 I IReactor Trip Response ii EOP-TRIP-2 II Sh 1 ii 1130 I I II n ii 11 I !LO. N1.1m1J~r, : - *<**]' .* Objectives I TRP001 E001 I TRP002E005 1--~

RO Skyscraper I

                                        *SRO Skyscraper         I RO System/Evolution List                I SRO System/Evolution List                            I Outline Changes                                 I
         , ..... , ,., .. Pr~]

(ouEistkin'T9 ...* IsRo 77 I Given the following conditions:

        -    Salem Unit 2 is in Mode 3, at NOP, NOT.
        -    All four RCPs are in service.
        -    22 RDMG set is in operation for testing, with its motor breaker and generator breaker shut.
        -    Reactor Trip Breaker "B" is racked in and shut.
        -    Reactor Trip Breaker "A" and both Reactor Trip Bypass Breakers are open.

Which of the following identifies how Tech Specs will be applied if a RCP were to trip, and the bases for that application? I

           ~ Only ONE RCP is required to be in operation to provide decay heat removal and ensure mixing, prevent stratification, and produce gradual reactivity changes during boron concentration reductions in the RCS. A cooldown to Mode 4 is NOT required.                                                                                                                                                                                I lb:] I Only TWO reactor coolant loops are required to be operable to provide decay heat removal and to satisfy single failure criteria. A cooldown                                                                                                                                            to
           ~ Mode 4 is NOT required.                                                                                                                                                                                                                                                                        I
           '.SJ I ALL RCPs are required to be in operation to provide decay heat removal and ensure mixing, prevent stratification, and produce gradual reactivity changes during boron concentration reductions in the RCS. A cooldown to Mode 4 IS required.                                                                                                                                                                                    I fd::  IALL RCPs are required to be operable to provide decay heat removal and to satisfy single failure criteria. A cooldown to Mode 4 IS required.

I [Answer.I a I I iExam,.Level l rs  ! (Cognitive Level .* J jApplication I 1 lt.~Ci)ity:,J Salem 1 & 2 I !E:~a&ti:>ate; 1 I 12/19/20161

     ~I 000015G237                                                 i!RO Value:.: r:IfillsRO Valuej~ 1section:.l I~ !RO Group:!LJ (SfWf:3roi.lp:f LJ a ~                                                                I"'                                                                             ~

112.2.37

systern/Evoiuti6n Tltle j j Reactor Coolant Pump Malfunctions i \015 I

[KA Statemenf i Ability to determine operability and/or availability of safety related equipment. I IExprarau<>r  :<>tI 55.43.b(2) The first part of this questions require RO knowledge to determine what constitutes an "energized" rod control system. 3

     !Ans.wers: ; * .. i of the 4 RTB and RTBBs being open is considered de-energized, even with a RDMG set in operation and a single RTB shut. (TS page 3/4 4-2a). With the unit in Mode 3 (rod control deenergized means can't be in Mode 2, and at NOT can't be in Mode 4), LCO
                                    ,,3.4.1.2 requires    TWO operable RC loops including RCP, SG, and loop. It also requires ONE loop in operation when rod control is
                                                                                                                                           *L..- *-:- r.f -:l "'j...,,...J.o. C("D ...I .... -. ... - - _,........,, .: ................
                                                     \At:.i.1- C:("'\I IC oro~ /-'":!nrl ,-..,,..,~'\ - - ... --L..1.o : .... :.a.:..,fh meeting the minimum requirement for the LCO. The bases for this LCO is as stated in correct answer. Choice B is partially correct,
                                                                                                                                                                                                                                           -:::1 ... .i.: .... ,., t-. .............. ....1 l"ln ctill since the operability portion is correct and the cooldown not required part is also correct. However, the bases does not reflect the reason as being single failure criteria. Single failure criteria refers in this case to the minimum required loops IN OPERATION.

(Bases B3/4 paqe 4-1.) I '

                          * **" Refer~nce Title, : * *.                     * . II'   facllity R~ferehcei'Num,be(                                 .* .j \Referenee ~ection' ** i IPage r-Jo: j [Revisionl ISalem Tech Specs                                                           I                                                                      II LCO 3.4.1.2                                  II3/4-2,2a 1144/282 I I                                                                           I                                                                      II                                              II                                !I                                     I I                                                                          I                                                                      II                                              II                                11                                     I

[Lo. Number, Objectives I RCSOOOE009 1_ ___,

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes f<;l\jestioriT&pic<J ISRO 78 Given the following conditions:

  -   Salem Unit 2 is operating at 4% power awaiting final approval to enter Mode 1.
  -   Main steam dumps are in service maintaining Tavg at 548°F.
  -   21 SGFP is in service, 22 SGFP is latched and at idle speed.
  -   All AFW pumps are secured and aligned for normal standby operation.

Subsequently:

  -   An automatic Rx trip signal is received, but the Rx does not trip.
  - All attempts to trip the Rx from the Control Room fail.

Which of the following describes how the CRS should proceed? Initiate manual rod insertion ...... laJVhen enter 2-EOP-FRSM-1, Response to Nuclear Power Generation, and start both MDAFW pumps.

    ~I then enter 2-EOP-FRSM-1, Response to Nuclear Power Generation, verify feed flow >44E4 lbm/hr from SGFP's.

I I

    ~ then continue in 2-EOP-TRIP-1, Reactor Trip or Safety Injection, then TRIP-2, Reactor Trip Response, since FRPs are not in effect in MODE 2.

fcll Ithen continue in 2-EOP-TRIP-1, Reactor Trip or Safety Injection, then TRIP-2, Reactor Trip Response, since reactor trip is confirmed with Rx Iµuwt::r "-07o. Answer I ~ !Exam levelj ~ !Cogn1!1\l~ ~~y~Lyj I Application

                                                               ,*' ",'.-:;.""*-">.,:*                       .. 11 Salem 1 & 2 I ~f~flhty:                               I IExamDate:j !___1_21_1_91_20_1_,61

~I 000029G244 112.2.44 ItRQYatue: l@[sRo Value)~ [section: II~ tft.OG(ouf):j[JJlsRg(;~ol!i:>;ILJ ~ jsy5~e:rrrEvolutiori Title I IAnticipated Transient Without Scram rKA. I statelllenFI Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. I IExpl~nation of I 55.43.b(5) The EOP network (including FRPs) are in effect in Modes 1-3. Just because Rx power is initially <5% does not indicate Answers:. : 1 a Rx trip confirmation, which also includes negative SUR and power lowering. FRSM doesn't check any feed flow, it specifically states AFW flow, so while Main Feed flow is still occurring, operators will still start both MDAFW pumps in FRSM-1 . I **** R.etei'ellce Titi.e :* ;;\:'*:*:*;;' *..: L FacifltY R.~f~~~nce ~umber :Y *1 iRefere,llc;e.Sectioh' . ';***J l'l:ia9e.r"o:l IR.evisi0ri 1 IReactor Trip or Safety Injection 112-EOP-TRIP-1

• I ii 1131 I IResponse to Nuclear Power Generation 112-EOP-FRSM-1 I II 1130 I I II I ii ii I IL'.6. ~umbe~ .***:. *.*1 Objectives IFRSMOOE001 I TRP001 E007 1--~

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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I 1auestion-Top1cj lsRo 79 I Given the following conditions:

  - Control room operators are responding to a Steam Generator tube rupture IAW 2-EOP-SGTR-1, Steam Generator Tube Rupture.
  -  The ruptured SG has been identified and isolated.
  -  All RCPs are in service.
  -  The RCS has been cooled down to Target Temperature, and depressurized to restore inventory.
  - Before SI can be terminated, the ruptured SG goes water solid.

Which of the followinq identifies how the CRS should proceed, and why? I

    ~ Continue in SGTR-1, Steam Generator Tube Rupture, and terminate SI and establish normal charging/ letdown, to ensure primary to secondary leakage is stopped.                                                                                                                                                                                           I
    ~I Continue in SGTR-1, Steam Generator Tube Rupture, and stop all but 23 RCP to minimize heat input to the RCS and prevent having to use a PORV to subsequently depressurize the RCS.                                                                                                                                                                              I I
    ~ Enter EOP-FRHS-3, Response to SG High Level, with SG NR level >92%, to minimize spread of secondary contamination and establish SGBD to lower SG level.                                                                                                                                                                                                 I
    @;]I Enter EOP-FRHS-3, Response to SG High Level, with SG NR level >92%, and place affected SG MS10 in manual shut to prevent steam release from affected SG until the TSC has performed a SG Overfill Evaluation.                                                                                                                                          I

!Ail~ )Exam Leve}] Is I ;cognitive Level J IApplication I [FacitltY:l ISalem 1 & 2 I !Examoate:] I 12/19/20161 [KA:ll 000037A214 1*[AA2.14H~LR0Value:il 4.0ll~ROValu:!J~lsection:ll~Rc>~I 2![SROGroup:fl 21 El~ fSY$tem/Evolution Titie] ISteam Generator Tube Leak ~Statement: I Ability to determine and interoret the followino as thev apply to Steam Generator Tube Leak: Actions to be taken if S/G qoes solid and water enters steam lines ~planation of I 55.43.b(5) FRHS-3 is a Yellow Path FRP, and may be entered at the CRS's discretion, which makes the transition to FRHS-3 Answers: distracters plausible. Additionally, FRHS-3 does establish SGBD to lower affected SG level, but only after a check for SGTR has been performed, which would kick out (or back to in this case) to the SGTR series procedure in effect. (Step 10.7 of FRHS-3) FRHS-3 does not take action to minimize the spread of secondary contamination, as that is addressed in the SGTR series. FRHS-

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~**---- Re~erence Title****---.. -  ;-* Facility Reference Number .*1 [Refe_r:!nce Section ~ ~ge No. I [Re"visio~ ISteam Generator Tube Rupture 112-EOP-SGTR-1 I 11 1130 I IResponse to Steam Generator High Level j j 2-EOP-FRHS-2 I II I j3o I IPost SGTR Cooldown 112-EOP-SGTR-2  ! 11 1130 I ~:Number Objectives I SGTR01 E007 1_ ____, !Material Required for Examination _J I II [Question Source: *11 Facility Exam Bank IJauestion Modification Method:; JI Editorially Modified I[Used During Trainin~f_i>rogram] D jauestion Source Comments/ 145725 changed FRHS-3 distracters and SGTR-1 distracter, and added basis to all choices. Correct answer remains

• the same, albeit with basis added.

I

@~mment .                                                                                                                                      l I

I I

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes I lh1.iesti()!1 Jopic ! I SRO so Given the following conditions:

 - Unit 2 is in MODE 6 with core reload in progress.
 -  1O 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 following would prevent continuation of fuel movement into the reactor?

   ~     ILoss of Control Air to containment.

I I

   ~ Racking down the 22 RHR pump 4KV breaker.                                                                                                                                                                                       I I
   ~I Both 100' elevation containment airlock doors are opened.

I [cl.l IWith both SRNls operable, only ONE is capable of providing audible indication in the control room. I

Answer 1 a I I jExam .Level J Is l l.<

:ogniti\/e Level i IMemory I !Facility:J ISalem 1 & 2 I 1examDaie: 11 12/19/20161

~I 000065G142 112.1.42 j JROValue~ J@[SRO Valuej~ !Section: 11~ [RO Group:i LJ jSRO Group:I LJ §1 "'

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iSystern/Evolutiori Title I j Loss of Instrument Air 11065 I iKA Statemeht* ! .. Knowiedqe of new and spent fuel movement procedures. /Explanation of/ 55.43(7, 6) This question is SRO level because of the knowledge require for fuel handling procedures, and the ability to IP.nswers: ..*.....* i continuously apply that knowiedge when operating the fuel handling equipment. The requirement for SRNl's is BOTH operating 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 1l . Facility Refer~nc4'!* NI.Imber . I\Re{~~e11ce Section* . ll Page t"o. I [Revi~Jonl IReac Pene Area & Cont Control Air ii 205347-1,3 I 11 I I I II ii I I "q 11 11 11 I I l~.o. Number> * * *.

• Objectives IREFUELE007 IIOP009E002

'--~ [au~~tJ§r $~u?C'e: J IPrevious 2 NRC Exams I!Question Modification M~~tlQc:i:  ;~Direct From Source IiusedDuripg Training Program'I D \auestioh*soqr-90,<:;~rl11J1Eilltsj ,_1_4-_0_1_N_R_c_e_xa_m_D_e_c_2_0_1_5*_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____. 6<, x~~,, > ,'.1~v ~~<<"<\ =~=,'**'"'\ .. ,, ':**:-?'~:>=:*-* ., ' ':<"({i*::.,*,.,; I I I

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

  ~uestll)n*l'opi~ I ISRO s1                                                                                                                                                                                         I Given the following conditions:
   -  Salem Unit 2 was operating at 100% power when a LOCA occurred.
   -  Operators are performing actions in 2-EOP-LOCA-1.
   -  20 minutes after the Rx was tripped:
       - RWST level is 19 feet and lowering
       - RCS pressure is 350 psig and stable.
       - Containment pressure peaked at 16 psig, and is 13 psig and dropping slowly.
       -  Containment radiation monitors read< 1R I hr.
       - RVLIS Full Range is 100%.
   - 26 minutes after the Rx was tripped, the RWST lo-lo level alarm is received.
   - When checking containment sump level at the beginning of 2-EOP-LOCA-3, the RO reports that containment pressure is now reading 1.5 psig.

Which of the followinq identifies the HIGHEST ECG classification for these conditions?

     ~ 14 point Alert.

I [b.115 point Alert. I

     ~    l7 point Site Area Emergency.

I r--" e:J Io µu1m .:me J-\lea r::r11eryer11.;y. IA11swer I @=] [Exam Level I ~ f<;og!litiv~L~v~I 'I IApplication I jF3,~i,ltty: J ISalem 1 & 2 I jexalTID~te:i l___1_2_11_9_/2_0_16.....I [01 oooo69G441 II2.4.41 \ fRova1~e::I [I~Hs~o.va1yeJ~ [sect!On: II~ (~o Group:H:=JJ [sl'o GrotjI>:I [ ] ~

;System/Evolution Title         J ILoss of Containment Integrity IKAStatement*i . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - , .

Knowledge of the emergency action level thresholds and classifications. iE~pi;:lnatiO,n of! 55.43(1,5) A LOCA which results in RCS pressure lowering to 350 psig will result in subcooling being lost. This yields 5 points due 1 Answer5: :: .! to the loss of the Reactor Coolant System Barrier. Containment pressure rise followed by rapid, unexplained pressure drop (13 psig to 1.5 psig during a LOCA over 6 minutes) is 3 points for the loss of the containment barrier. The 4 point alert is if the loss of subcooling was not recognized and the loss of containment not recognized. The 5 point alert is if only the loss of the RCS barrier is

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[~~~lli!Si~.?l~g; ii IFacility Exam Bank I [~y!~l~2:tS:~~!~{!\!~~*~lD:,'A1~.J~il Direct From Source [c;to~iqgrs~~rc~:~~mm~~tsl 1136652

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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I *Outline Changes I rQuestiqnJopic j SRO 82 I I Given the following conditions:

   -   Unit 2 is responding to a LOCA coincident with a LOOP, and has transitioned from EOP-LOCA-1, Loss of Reactor Coolant, to EOP-LOCA-2, Post LOCA Cooldown and Depressurization.
   -   RCS cooldown has been initiated.
   -   The RCS has been depressurized as required with a PZR PORV.
   -   All RCS Thats are 450°F.
   -   When performing Step 19 RCP Status, the RO reports that RCS subcooling is NOT >0°F.

Which of the following identifies how the CRS should proceed?

      ~I Initiate SI to restore RCS subcooling and return to LOCA-1, step 1.

I

      ~I Start ECCS pumps as necessary to restore PZR level and return to LOCA-1, step 1.

I I

      ;c.: Initiate SI to restore PZR level, and ensure BIT isolation valves open if previously shut.

I [dl I Start ECCS pumps as necessary to restore RCS subcooling, and open BIT isolation valves if previously shut. I

Answer'; Id j ;examlevel 11 s I ico~l1itive Levet'?i I Application I !Facility:! I Salem 1 & 2 I tE'fa!Jlpate;J I 12119120161

[KA:jj OOWE03A201 IJEA2.1 j:Rova1ue:/l 3.4j[sRoya11.JeL_~lsectioQ:l!~tRoG~o1.n):jl 2l!SRO(;fo6p:jj 21 "Ii"~ [system/Evolution Title I j LOCA Cooldown and Depressurization 1 IE03

• KA Statement: : Ability to determine and interpret the following as thev aooly to LOCA Cooldown and Depressurization:

Facility conditions and selection of aooropriate procedures during abnormal and emerqencv ooerations. I l [fxplariatio!1 of 55.43.(b) This question is SRO level based on having to assess conditions, then select a section of procedure, in this case going .Answers: , l from Step 19.1 RNO to Step 34, bypassing charging and SI pump reduction steps based on inadequate RCS subcooling. The distracters stating to restore PZR level are incorrect because the stem states that RCS depressurization has been performed, which is only stopped when PZR level is adequate (>25%). The initiate SI distracters are incorrect because only specific required ECCS I "" h~

                                                                       ~     ., in:.:~1 C:: I ~~**                                         h.-..-. ,....,...,..., ..... ,...,,,i thinking that the next steps would include isolating the ECCS accumulators, which are not part ~f the choices and cou.ld lead to confusion in answering question.

Tho Df"C'.' Th-L 'n

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RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [gues!ton:fopi~l 1sRo83 Given the following conditions:

  -   Unit 2 was operating at 15% power prior to synchronizing the Main Generator.
  -  A Main Steam line rupture occurred that resulted in multiple Steam Generators depressurizing in containment before 2 steam generators could be isolated from the 2 faulted SGs.
  -  The 2 faulted SGs Tcolds are reading 270°F and lowering.
  -   The intact 2 SG Tcolds are 330°F and stable.
  - RCS pressure is 500 psig and slowly lowering.
  -   Containment pressure is 16 psig and slowly lowering.
  -   All SG NR levels are <9%.
  -   Total AFW flow is 24E4 lbm/hr.
  -   Source Range Nls are NOT energized.
  - Intermediate Range SUR is 0.0 DPM.

With CFST's in effect, which of the followinq identifies the procedure entry required, and actions which will be performed in that procedure?

     ~ 12-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions. Maintain AFW flow >22E4 lbm/hr until at least ONE intact                                                                                                                                                                                     II SG NR level is >15%, stop all ECCS pumps except 21 or 22 charging pump.
     ~ 12-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions Isolate any faulted SGs, depressurize RCS with ONE PORV                                                                                                                                                                                       I to within 100°/hr Cooldown Curve.
    !c. i 12-EOP-FRSM-2, Response to Loss of Core Shutdown. Energize Source Range channels and verify SR SUR is 0 or negative.
     ~ 2-EOP-FRSM-2, Response to Loss of Core Shutdown. Establish AFW flow >44E4 lbm/hr, borate RCS until JR SUR is negative.

[Ansv...er I ~ [exam Level I @=]" !Cognitive .Level .1 IApplication I !FacHity: i ISalem 1 & 2 I !~~~mDate: J I 12/19/20161 ~I oowEosG406 I 2.4.6 IJko Value: I§!sRo y~1µej_.i2J ~section; 11~ JRo.Gi:gup:i LJ !~Ro G~9pp:j LJ ~ ~ !System/EvolutiC>l'I Title<! IPressurized Thermal Shock  ! l~E_os_ _ IKA.statemeni:I . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - _ _ , . Knowled e of EOP miti ation strategies. I IExplanptj.on.pf 55.43(5) A is correct because the stem conditions result in a PURPLE path on FRTS. Actions for maintaining AFW flow (Step 3.5)

Answers: * ** *. and ECCS pump reduction (Step 12) are correct. B is incorrect because depressurizing the RCS to restore conditions within the 100°F/hr curve is performed in FRTS-2 in response to a Yellow Priority Condition. FRTS-1 is entered from either RED or PURPLE conditions, and with SG NR levels <9% (which means you're less than 15% adverse) you are directed to maintain AFW flow> 22E4

2. The FRTS is a higher priority, and is a PURPLE path. C is incorrect because it is the wiong procedure with. the correct actions of 1that procedure. D is incorrect because it is the wron procedure, and the actions are erformed in FRSM-1.

1* * . **

• Reference Title ** * .* 1[ *.* *.* FacilitfRefereric~ Number *. I[Ref~ren9esection *  !(P11g~ No. I /R~yisioni ICritical Safety Function Status Trees 112-EOP-CFST-1

!Response to Imminent Pressurized Thermal Sh 112-EOP-FRTS-1 ~=================~============== l================~~========:::=::; j Response to Loss of Core Shutdown I!_2_-E_O_P_-_F_R_S_M_-_2 _ _ _ _ _ ___, _ _ _ _ _ _ _ __, ---~ !LO I

      .. Number.********
    ..  . . . .                                           Objectives I FRTSOOE002

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RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I 1Questi6'n Topic j ISRO 84 I Given the following conditions:

   -    Unit 2 experienced a LOCA from 100% power.
   -    At step 16 of LOCA-1, Loss of Reactor Coolant, a transition to LOCA-5, Loss of Emergency Recirculation, was made with both RHR pumps stopped and unable to be started.
   -    Just prior to initiating a cooldown, the RWST LO level alarm is received.
   -    CONT SUMP CH NB LEVEL >62% lights are lit.
   -     RCS subcooling is <0°F.

Which of the following describes how the CRS should proceed? [] I Remain in LOCA-5 and initiate a RCS cooldown at maximum rate. I

      ~I Remain in LOCA-5 and initiate a RCS cooldown at less than Tech Spec limit.

I ["Zl I.Transition to LOCA-3, Transfer to Cold Leg Recirculation, perform valve realignment steps, return to procedure in effect. I fd.'l ITransition to LOCA-3, Transfer to Cold Leg Recirculation, align cont sump to RHR suction, then remain in LOCA-3 until direction is provided by the TSC. I Answer 11 b I !Exam Level J IS I !cognitive Level l IApplication l~alem1&2 I t!=xamoate: 11 12/19/20161 ~I OOWE11A202  !/EA2.2 /!RO \Jalue:.jD'.]'fSRO ValueL-~ isectiqn,:jj~ lRO Grolip:1LJ jSR,O Group:jLJ Iii ~ "" ISystem/Ev'olutlonTitle) ILoss of Emergency Coolant Recirculation I ~iE_1_1-~ 1 KA Statement:: Ability to determine and interpret the followina as they apply to Loss of Emeraencv Coolant Recirculation: l Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments. l \Explanation (* . -_,;:*"; . . of I 55.43.(b)5 The stem says the transition to LOCA-5 was at LOCA-1 step 16 so RWST level >15.2'. There is no LOCA-3 CAS in 'Answers: ... >j LOCA-5. The cooldown in LOCA-5 is done at <100°F/hr, which is the TS RCS cooldown limit to prevent an unwanted Thermal Shock condition, which would complicate matters since CFSTs are still in effect in LOCA-5. Restoring any Train of recirc is the concern with no RHR pumps, the transfer to CL recirc will occur only after a RHR pump has been restored.

       **              *. Reference Title                    *'.\!   f~cility ReferenceNu'mber , 1 IReference Section            J i: Page No. I jRe"'.ision' ILoss of Emergency Recirculation                                 112-EOP-LOCA-5                         I                           11           1130        I I                                                                II                                     I                           !I           11          I I                                                                II                                     I                           II           II          !

tLO~ Numb~r ,, I LOCA05E001

                               .. . J Objectives  I llVfatefiaLRequire~ for i:Xamrnation)                 Il l9U:es~foi:i.*~'1Hrce: . 11 Facility Exam Bank               I\Qu~~tiori Modifi~~ti.ollllA~!h/?~:. ~,ii Editorially Modified fQuestio.n so.urea qC>rt1rfientsj 1140182
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RO Skyscraper I ... SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I

  \Qli~stjonT&picj I SRO 85                                                                                                                                         I Given the following conditions:
    - Unit 2 has experienced a MSLB at the Mixing Bottle.
    -  All attempts at MSLI have failed, and 21-24MS167s remain open.
    - Operators have just completed SI termination steps in EOP-LOSC-2, Multiple Steam Generator Depressurization, and PZR level is being maintained stable.
    -  AFW flow to each SG is 1.0E4 lbm/hr.
    -  The RO reports rising pressure in 22 SG.

Which of the followinq describes how the CRS should proceed, and why? []I Transition to EOP-LOSC-1, Loss of Secondary Coolant, since one SG is now available for subsequent recovery actions. I

      ~I Transition to EOP-LOSC-1 Loss of Secondary Coolant and stop RCPs if RCS pressure is <1350 psig, since RCPs cannot be stopped in LOSC-2.                                                                                                                                                I
      !cl
      ~

IRemain in EOP-LOSC-2 since returning to EOP-LOSC-1, Loss of Secondary Coolant will require a transition to EOP-LOCA-1, Loss of Reactor Coolant, upon completion. I [] I Remain in EOP-LOSC-2 until positive control can be established over the cooldown after the remaining Steam Generators have fully depressurized, then transition to EOP-LOSC-1, Loss of Secondary Coolant. I !Answer 11 a I IExam l..evel 11 s I !cognitive Level 11 Memory 1 !Fftcility; 11 Salem 1 & 2 I \Ex~ipoa~~U I 12/19/20161 IKA:ll OOWE12A201 I jEA2.1 llROValue:\l 3.2l[~ROValuef~*section:!l~[RoGrolip:fl 11/SRCl(fo)up:ll 11 lffl ~ lsy5tefu!Evolution Title I I Uncontrolled Depressurization of all Steam Generators  !~121

/KA Statement: I Ability to determine and interpret the followinq as they apply to Uncontrolled Depressurization of all Steam Generators:

Facility conditions and selection of appropriate procedures during abnormal and emergency operations. IExplan'i!!ion ,Of :I 55.43(5) LOSC-2 CAS states that upon a pressure rise in any SG except when performing SI termination in Steps 8-20, GO TO 1Answers:,. ** .'! EOP- LOSC-1. The stem states that it is after Step 20. LOSC-1 Basis Document, page 7, states that.."Any cooldown operations that are performed as subsequent recovery actions will require at least one nonfaulted SG." 1*.

1. .  ; ReferE)nce Title ,*(

l f: I

• FacilityRefer~nce Number fv JR.efererlce Section '.j [ Pag~}fa.:\ [Revision:

IMultiple Steam Generator Depressurization 112-EOP-LOSC-2 I II 1130 I ILoss of Secondary Coolant iI2-EOP-LOSC-1 I ,,11 1130 I I II I ll l IL.o.. Numbe'r 1 ** Objectives I LOSC02E005 iMat~rial*Requir,edforEXart1i~atio(l,. 11 i<lU~sti~r;i ,$?~rce; *.JI IFacility Exam Bank I!us~~ buf'li}g Jraining'Program . \ D ioue,:>tio~ s6ur~~;.~o~.~ents: I

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141989

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I (Comment I I I

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [~uestioll Topic 11 SRO 86 Given the following conditions: I

  - Unit 2 is operating at 100% power.
  - Chemistry reports at 1000 on March 10th that a routine sample of the RCS indicates that DEl-131 is 8 uCi/gm.
  -     A second sample confirms the elevated reading.

If DEl-131 remains at this level, which of the following identifies the Tech Spec required time that the below listed action must be performed, and the bases for performing that action? Be in at least Mode 3 with Tavq <500°F bv .....

     ~       1600 on March 10th. This prevents the release of activity should a steam generator tube rupture occur since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric release.                                                                                  I
     ~       1600 on March 12th. This prevents the release of activity should a steam generator tube rupture occur since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric release.                                                                                  I le.*.;

l..__J 1600 on March 10th. This ensures the resulting 2 hour dose at the Protected Area Boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed primary-to-secondary steam generator leakaqe rate of 1 oom.

    @]       1600 on March 12th. This ensures the resulting 2 hour dose at the Protected Area Boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed primary-to-secondary steam generator leakage rate of 1 qpm.

[Answer Irti mm I !Exam Level Ifsm mml lcogriitive Level *<1 jMe~-OrYm I 1,Facility;j I Salem 1 & 2 I !Exan:ioate: 11 12/19/2016! [KA:\I 002000G225 I~i2._2._25_~-l IR9 Vallie: Ii 3.21 [sR.(JVatue)~ !section: JI~ [Ro Group:!! 21 !sRo Grotfp:! I 21 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __.l [002 [System/Evolution Title! l_R_e_a_ct_o_r_c_o_ol_a_nt_S"""'y'-s_te_m fKAStateinent; I Knowledqe of the bases in Technical Specifications for limitinq conditions for operations and safety limits_ IExplana,ticm of j 55.43.(b)Z LCO 3.4.9 states that with activity greater than 1.0 uCi/gram DEl-131 but below the line of graph 3.4-1, then operation 'Answers: .. -*-- can continue for 48 hours, then HSB with tavg <500°F in six hours. The bases for the limitation of specific activity is contained in the correct answer and distracter a. The other 2 distracters contain the bases for the LIMIT on RCS activity, but with the wrong area with regards to where the dose is limited. Protected area Boundary is wrong, Site Area Boundary is correct. f ReterenceTit1e*-. _-**I! {Fa~ili!j< R~feren~e Number5' :*j t~eferen~'section Ir ~~ge No. / fB,evlsiopj ISalem Tech Specs 11 113 I j 314 4-23, 11258 I l II II II 11 I i II II q 11 I i~.o. Nurnber I RCSOOOE009 Objectives

Materiat_ReqlJiredlor ~xain,}pat_ipn  !

                                             **,i;J SRO 86 Salem Tech Specs Figure 3.4-1 DEi vs Rated Thermal Power                                                  lj taue~tionsource:*J            jFacility Exam Bank I!au~~.u~rl'iV!p,difl~~tiC)n\\il_~t!loC!:'\      Iconcept used               I[use~,bi.ifln9'rrruhingProgra.il J o i_aues~ioll ~ouicece>&'fri~l'!~§) j Combination of 86338 (limit) and 58158 (bases)

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 l(;}ue~tipl'I topiCj ISRO 87 I Given the following conditions:

  -  Unit 2 is operating at 100% power.
  -  A LOCA occurs which results in RCS pressure lowering to 900 psig.

Which of the following describes the response of the EGGS Accumulators, and when will the CRS direct actions regarding the EGGS Accumulators? Assume conditions requiring a transition to LOCA-3, Transfer to Cold Leg Recirculation, will not be oresent durina the response to the LOCA. fal INO EGGS Accumulators will discharge. Accumulators will be isolated when RCS Thats are <375°F in LOCA-1, Loss of Reactor Coolant. I

    ~I ALL ECCS Accumulators will discharge. Accumulators will be isolated when RCS Thats are <375°F in LOCA-1, Loss of Reactor Coolant.

I I

    ~ NO EGGS Accumulators will discharge. Accumulators will be isolated when RCS Thats are <375°F in LOCA-2, Post LOCA Cooldown and Depressurization.                                                                                                                                                                             I
    @J I   ALL ECCS Accumulators will discharge. Accumulators will be isolated when RCS Thats are <375°F in LOCA-2, Post LOCA Cooldown and Depressurization                                                                                                                                                                              I

[Ans~er) j c l [Eifal11Level i Is I lcognitivel-e\lel JIMemory I [FaciHty: ! ISalem 1 & 2 I IExarrio~te~ 11 12/19/20161 IKA:ll 006000A212 1:A2.12 liROVatue:!j 4.5l!sR.C>.value}~!section:lj~:RoGrol.lp:ij 1llSROGroup:d 1j B ~ f System!Evolutiori Title j j Emergency Core Cooling System 11006 )KA Statement: l Ability to (a) predict the impacts of the following on the Emergency Core Cooling System and (b) based on those predictions, use procedures to correct, control, or mitiqate the consequences of those abnormal operation: Conditions requiring actuation of EGGS \E~plani;lti()n of 55.43(b)5. First part of question requires knowledge of EGGS Accumulator operation. Second part of question requires knowledge !'.Answers: *

• of a section of procedure, LOCA-2, where EGGS Accumulators will be isolated when RCS Thats <375 (Step 35.3. EGGS

                   - Accumulators would only be isolated in LOCA-1 AFTER Cold Leg Recirc was established at step 19.2. Stem states LOCA-3 will not be applicable.

f :<  ; Reference Title * \ I 'r , Facility Reference Nriinber **: !Referenee Section i l !.~age No. I f~evisiohl IPost LOCA Cooldown and Depressurization 112-EOP-LOCA-2 I 11 1130 I ILoss of Reactor Coolant 112-EOP-LOCA-1 I ii ii 30 l I II I 11 11 I Objectives I ECCSOOE009 I ECCSOOE013

!gJe~tion~8uDce:        qj_N_e_w_ _ _ _ _ __.l,L[Q2~""e~sc:::fi~()2nc._M"-'o:_d""}f_ic2')""'.t:_f<>"-n2,M....:*c..:e*t""*~:_*.()"-d:_r...:c****Jlil---------'l l.Vsed'.[)ori~~fTrairii~g*f'.rogranfl D
)9u~s!i()~**sqilrc~*c~ritlp~l1ts!    I                                                                                                                                                                   I
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RO Skyscraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I

  !aoestiori To'pic.J      ISRO as                                                                                                                                              I Given the following conditions:
   -  Fuel handling is in progress in the Unit 2 Spent Fuel Pool when a fuel assembly in the Spent Fuel Handling Tool is dropped.
   -  Gas bubbles are observed in the vicinity of the dropped fuel assembly.
   - 2R5, Fuel Handling Building (FHB) radiation monitor goes into alarm and stabilizes at 25 mR/hr.

Which of the following describes the effect of this event, and contains actions that will be performed IAW S2.0P-AB.FUEL-0001, Fuel Handling Incident?

     ~I ALL_ Fuel ~andling Crane motion is locked out to prevent further damage to an affected fuel assembly. Ensure all available FHB Exhaust fans are 1n service.                                                                                                                                                       I I
     ~ ALL Fuel Handling Crane motion EXCEPT downward movement is locked out to prevent raising a damaged fuel assembly. Ensure the FHB Truck Bay Roll Up Door is closed.                                                                                                                                     I
     @] I The FHB Evacuation alarm actuates to alert all personnel of the FHB high radiation condition. Evacuate ALL personnel from the FHB until Radiation Protection has performed area surveys.                                                                                                                      I
     @] I  FHB ventilation automatically swaps to place the Charcoal Filter in service to prevent a release to the environment. Ensure the FHB Watertight Door remains closed except for normal personnel passage.                                                                                                   I
 !Answer 11 d            I I exam Level 11 s              I !Cognitive, Le'vel ** d I Memory                I [raciUty; 11 Salem 1 & 2            I      ITloate:   11   12/19/20161
 ~I 034000A201                   I !A2.01            IIRoValue:jO]'l~RO\faluej~ 1Section:;ll~ [RO Group:iLJ[SRO Group:ILJ                                            m' ~

fsystemfEvolutionJitle I j Fuel Handling Equipment System I /034 I

 /KA Statement: l       Ability to (a) predict the impacts of the following on the Fuel Handling Equipment System and (b) based on those predictions, use rocedures to correct, control, or mitigate the consequences of those abnormal o eration:

Oro ed fuel element Explanation of! 55.43(7) The 2R5 in alarm (11 mR/hr alarm, 7mR/hr warning) swaps the FHB exhaust ventilation to the Charcoal Filter and starts

 ~An_s_w_e_r_s._-~    both FHB Exhaust Fans. The normal configuration for FHB ventilation is the single Supply Fan running, and BOTH Exhaust Fans running. C is incorrect because non-essential personnel are evacuated, as actions are required to be performed prior to evacuating ALL personnel. The alarm will actuate. A and B are incorrect because the FH crane only locks out as described in B with the
                  **.. Re,ference.title *.** ** **         * '*\1 I..:Facility: Reference Number                                           l
                                                                                                          ;,*/ !.~eferen~e. section* *. *.. IPage ~°! /,Revision 1

jr=:==============================~====================::;.:;'. Fuel Handling Incident !I S2.0P-AB.FUEL-0001 I Ij I j5 I IRadiation Monitoring System Lesson Plan II NOS05RMAS000-14 I Ij 1117 j i1 ~~~~~~~~~ll I ii ii I IL.C>. Nufo.ber i

• I AB FU EL01 E002 Objectives I_ __.

tQues~()ll sq.urc~: >J IFacility Exam Bank I !<:iuesti,~!1.Nlo~[fiCCl~\.~iiMe,th'o~~:< IDirect From Source I!used [}~ring'Jrail1Mg Prqgrarn l D 1.a~esuqn,,s6urc:~c;oip~eri!~l 159573 I:=::=====-=================::============-================-=~1

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RO Skyscraper I***** SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I. . Outline Changes I luuestid\'i"fopic'.l I sRo 89 I Given the following conditions:

    -  Unit 2 is operating at 84% power.
    -  All Condensate pumps are in service.
    -  NO Heater Drain pumps are in service.
    -  The Condensate Polisher is in service.
    - Power is reduced based on the Condensate I Heater Drain pump configuration Subsequently, 21 Condensate pump trips.

Which of the followinq describes the effect on the Condensate system of the pump trip, and how should the CRS respond?

      ~I The 21-23CCN108s Polisher Bypass valves automatically open when SGFP suction pressure lowers <320 psig. Reduce reactor power to 65%

or less. I [§J I The 2CN47 23/24/25 Heater Strings Bypass Valve automatically open upon the Condensate pump trip. Reduce reactor power to 30% or less. I

      §]I     21A Condenser level will rise. Open Polisher Bypass valves and reduce reactor power to 65% or less.

I

      @]I 21A Condenser level will lower. Open Heater Strings Bypass valve and reduce reactor power to 30% or less.

I [Answer 1Ic I !Exam Le"l/ei i j s I lcogilitiye Leyet IApplication J I lfaqffify: 11 Salem 1 & 2 I fE.~a.mq~te: I J 12/19/20161 q-11~A2_.0_4_~1":RQValue:Jj 2.6ltSROValue~ 2.8*jrSect19n~JI~ I 1~11 ~

KA:ij 056oooA2o4

                                                                                                                             ~

iRO Gr()UP,;d 1itSRO (3roup:Jj 1 (systenl/Evol!JtionTifle 'I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____.! lo56 j_c_o_n_de_n_s_a_te_S_y,_s_te_m lKA Statement:1 Ability to (a) predict the impacts of the following on the Condensate System and (b) based on those predictions, use procedures to correct, control, or mitiqate the consequences of those abnormal operation: Loss of condensate pumps !Explanation 6t 1 55.43(b)5. This question is SRO based on having to determine the section of AB.CN which will be performed, and the sub-section of

J\nsvvers:' * * '1 Attachment 2 knowing that with 2 Cond pumps running with no HOP running power reduction to 65% or less is required. The CN108s do NOT auto open on low suction pressure, but are directed to be opened when SGFP suction pressure lowers <320 psig, which it WILL, based on initial power level and initial pump configuration. The CN47 only auto opens on a SGFP trip, not a Cond

                                   - +.. : .... Tho ('~IA7 ;.,,      r1;---*-"'  In,,.,_-----"' 11.i::Ti::O *'-- ("'~111"\0- - - -   rl<---'-"' I n ' - - - - - - - " ' in /\.Cl ('~I
                                                                                                                                                                                          *'-- - - - - * - - nf   .1....:-t... ie expected to restore suction pressure above the point which would require CN47 opening. Hotwell level will RISE with the cond pump O/S, as condensing is still occurring, with the pump not running.
                                                                                                                                  'I [R~ference's~dion';                '] [:Page' N<i.1
           .*       *, .~Reference Title

[Main Feedwater I Condensate System Abnorma

                                                                  ;j:;.,*

JI ** Faci1iiy R.eterence Numbe{ ll S2.0P-AB.CN-0001 I

                                                                                                                                                                         ,,                   IR.evision) 1!28              I I                                                                                II                                                 I                                     ii                 II               I I                                                                                !I                                                 I                                     !I                11                l

[,L.o:,,Numbkr ; ' .: I Objectives I ABCN01 E003 I I CN&FDWE008 I [g~~str<)~ ~o~r~e§\'.j l_N_e_w_ _ _ _ _ _ __.l [~~~~ii§r{M'1-(!,iticatJo~.:M~tflO,.<i(; 'I _________.! !ysed 0U'r,1n:9:;r!~lf!in9'f>rd9ram I o fQl.l~~tfoil~~yref~Go?l~.e~~l I I 1~------------------------------------------' I I I

RO SkyScraper I SRO Skyscraper I RO System/Evolution List

• I-*

SRO System/Evolution List j Outline Changes I [Question I Topi~ SRO 90 I Given the following conditions:

  -  Unit 2 is operating at 7% power during a normal power ascension.
  -  A total loss of Control Air occurs on Unit 2 causing control air header to rapidly lower.
  -  The CRS enters S2.0P-AB.CA-0001, Loss of Control Air, and operators manually trip the reactor.

Prior to any operator action being performed other than TRIP-1 immediate actions, how is feed flow to the SGs affected by the loss of control air, and how should the CRS respond?

    ~     MDAFW pumps previously in service will cease to supply feed to the SGs as the 21-2AF21's fail closed. Dispatch operators to locally manually open 21-24AF21s to establish 22E4 lbm/hr total AFW flow.                                                                                                                                                                                                                                                                                       I
    ~     MDAFW pumps previously in service continue to supply all SGs, but now at runout flow with 21-24AF21s failed open. Dispatch operators to locally manually throttle 21-24AF21 s to prevent MDAFW pump failure.                                                                                                                                                                                                                                                                                    I
    ~     The feed flow being supplied to SGs from the SGFP in service was lost when the Feedwater Interlock actuated with the reactor tripped and Tavg low setpoint reached. Direct the PO to start 23 AFW pump since Pressure Override circuit will prevent using 21 or 22 AFW pumps as feed supply.

fril

    ~     The feed flow being supplied to SGs from the SGFP in service was lost when the Feedwater Interlock actuated with the reactor tripped and Tavg low setpoint reached. Direct a Field Operator to establish manual control of 23 AFW pump which started when 2MS132 opened to prevent 23 AFW pump runout and SG overfill.

!Answer I [ci' .. , !Exam Lever *I rs *1 lcognit!V~ L~ver I rAppliratio;:; m m m I lr~~ilftx: j ISalem 1 & 2 j [Ex~~Dat~: I ! 12/19/20161 tKA:ll 061000A207 II~A_z_.0_7_ _: IRO ya:1ue:il 3.4l [s1fo'va1ueL22! fsettiorl:. II~ IRCfGrciup:JI 111sROGroiJ)l; I 1j [system/Evolution Title I IAuxiliary I Emergency Feedwater System l ]061 IKA statement: l Ability to (a) predict the impacts of the following on the Auxiliary I Emergency Feedwater System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Air or MOV failure IExplanation of i 55.43(b)5 This question is SRO level based on the requirement to know what section of a procedure to use in AB.CA. At 7% power, !Answers:' < l Main Feed will already have been established and AFW secured. The rapid loss of CA, and the trip in AB.CA (which would have been directed when the BF19s/40s failed SHUT on loss of CA indicate that feed flow from the operating SGFP was lost. The 2MS132 would have opened on the loss of CA (fail open valve) to start 23 AFW pump, which would be operating at max speed. ITho I\ D ...J; ... ,..,....,1-,... ~n ......................... ....,'"fl"\ 1.........,...,11. ,.. ............ t...li~h ................. ...J MDAFW pump initial flow are incorrect but plausible if the operator doesn't not know at what power the !OP directs swapping AFW-to Main Feed. Choice a has wrong failure mode, but correct action if that were the correct failure mode. Choice b has correct

                                                                                                                                                                           ............ .i. ...... 1 .(,.... .. 4ot... ..... ............ ,..,.. ....... ,...;,, .......... in ... i-. .... ..................... ,......,,..,.,........ Th .....

failure mode and action. Choice c has correct status of feed flow, but wrong action

                                                                                                      *****I ; **

• f'.acility Reference Number. ; J t~efererice section < l ~ Page No..11.R.evisionl

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[  ;.:';' Reference Jitle *.. **, ILoss of Control Air II 82.0P-AB.CA-0001 I II 1121 I j Auxiliary Feedwater System Lesson Plan II NOS05AFW000-15 I ll 1115 I I II I II 11 I I 1 L.O. Number AFWOOOE006 Objectives I I ABCA01 E002 '--~

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

  ~uestion Top.ic       I ISRO 91                                                                                                                                                          I Given the following conditions:
   - Unit 2 is operating at 100% power.
   -  28 EOG is CIT for scheduled maintenance, with a scheduled return to service in 24 hours.
   -  It is discovered that 2A EOG monthly surveillance was not performed within its 31 day required periodicity.
   -  The required 2A EOG surveillance was last performed 33 days ago.

Which of the following identifies the status of 2A EOG IAW Tech Specs, and whv?

     §;]I   INOPERABLE because it has exceeded its 31 day surveillance requirement.
     ~I INOPERABLE since the 24 hour delay time past the 31 day requirement has been exceeded.
     !21  I OPERABLE because the normal surveillance interval plus 25% extension has not been exceeded.

Id* II OPERABLE because the surveillance can be performed within the 24 hour delay time which starts upon discovery of the missed surveillance. iA.riswe:rl c I I !Exam Leyel i Is  ! lco9nitj~e L~vel *d Application I !F~"!Hty: J lSalem 1 & 2 I (EFaTDate~ !j 12/19/20161 (KA:ll 064000G240 I!2.2.40 *!~OVah.ie:'!I 3.4j ;sR.oValueJ~ is~'ction: 11~ [Rb pr6up:1 I 1 j ISflO Group:! I 11 wa D lsystem/Evoluti9il litle J IEmergency Diesel Generators I 1064

 \KASfatement:\,......-------------------------------------------,.

Explanation 9t*I 55.43(2) Tech Spec 4.0.2 states ... "Each Surveillance Requirement shall be performed within the specified surveillance interval with Answers! ." **: a maximum allowable extension not to exceed 25 percent of the specified surveillance interval." Since the 25% of 31 days has not been exceeded, the EOG remains OPERABLE, since its surveillance is not required to be performed until 31+7. 75 days. A is incorrect because of the 25% time. B is incorrect because the 24 hour delay time is from time of discovery. Dis incorrect because Reference .Title  :.* * ***** j f * ** :Facility ~efer~nce Number ( i

• lRefereri,ce Secuo'.11 *.*. *. j j P~ge No. j [Revision!

ISalem Tech Specs II qSurveillance Requirem 113/4 0-3 I j279 I I 11 II lj I I l l, _ _ _ _ _ _ _ _____.11 II ii lj I l~~q. NurnberV<** 'i .*I Objectives I TECHSPE011 I ITECHSPE014 I ffilles~e:>ri§o~fc,g~f) IFacility Exam Bank

 \tiu~syo!';.sqli~r~~;cc)rr!:me~isl l . . 4 _ 1 _ 9 _ a 1 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . . , . , 1
                   ,:*\ -*~/*~,     ,,*, :~<.>: :;

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~estion Topic 11 SRO 92 Given the following conditions:
 -    Unit 2 is at 40% power performing a shutdown.
 -    4 SW Bay is isolated due to a leak on the 25SW3, 25 SW Pump Discharge Isolation Valve.
 -    Operators are performing the shutdown to comply with TSAS 3. 7.4 because difficulties arose during the leak repair of the 25SW3,
 -    2A EOG is supplying 2A 4KV vital bus for a scheduled surveillance.
 -    22 Charging pump is in service.
 -    23 Charging pump is available.
 -    21 and 23 SW pumps are in service.
 -    22 SW pump is in auto and standby.
 -    23 SW pump trips, and one minute later 2A EOG output breaker opens on 2A 4KV Vital Bus Differential Which of the following identifies:

1) The effect this will have on the SW system

2) How the CRS should respond

    ;a.J         1) 22 SW pump will auto start on low SW header pressure.

2) Enter S2.0P-AB.SW-0001, Loss of Service Water Header Pressure. Reduce loads on Service Water listed on Attachment 2, Loads Affected by a Loss of Service Water.

    ~            1) All SW flow capability is lost.

2) Enter S2.0P-AB.SW-0005, Loss of All Service Water. Trip the Rx, confirm the trip, and stop RCPs to limit the heat input to the CCW system.

1c.1 c.......c I 'J Mil <:>vv 11uvv (.;i::tjJau1111y 1::> ru::>t.

2) Enter S2.0P-AB.SW-0005. Trip the Main Turbine and reduce Rx power <5% to lower heat input to the RCS.

[d.l 1) 22 SW pump will auto start on low SW header pressure.

2) Enter S2.0P-AB.SW-0001. Place 23 Charging pump in service and remove 22 Charging pump from service.

   *v

[p.nswer J lli:J'. I tExam ,Level l '~ .. ,; i @=]" !C9gmt1ve._Lev,,el I IAppl1cat1on I lf,~~1l*!Y~ j jSalem 1 & 2

                                                                                                                         ' ~ '* "**._;*: i *;

I [E~arnD~t7: Il___1_21_1_91_20_1_,6j ~!<Jl.:lj 076000A201 IIA2.01  : IROValUe:lfililsRo va1ue)2:Z:J 1secti6r1SH~ t~o oioup:ILJ ts1w pr()lip:jLJ ~

sy5fein/EY"oll1ii~n Title I l_s_e_rv_i_ce_w_a_t_er_s_y._s_te_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ' - - - -

KAstaJ~ Ability to (a) predict the impacts of the following on the Service Water System and (b) based on those predictions. use procedures to correct. control, or miti ate the conse uences of those abnormal o eration:

Loss of SWS 1exp1;a 1 11 ati?B9t 55.43(b)5 21-26 SW pumps are powered from AA,BB,CC vital buses. With all 4 SW Bay isolated, 24,25,26 SW pumps are fAnswer~: *

• unavailable. A Bus Differential signal will open the EOG output breaker, and ALSO prevent the vital bus infeed breakers from Station Power to close. This will result in no SW pumps running. AB.SW-5 states to perform actions listed in correct answer. The other AB.SW-5 action is wrong in that tripping the MT when power is <49% will NOT trip the Rx. The 2 distracters regarding 22 SW 1 *.;**:-*. RefererJce;Title * '

.

:::;==:;==:;;;;;;;;;;;~====~==:;;;:;;:;;;;;;;;;;==::;;;;:;==;,

I!*. 1 'F:acilityRef~rence Number  ; *I fRefe~eri¢esection ***.*l f PageNo.'l l~~visioni !~L=o=ss=o=f=Al=IS=e=rv=ic=e=W=a=t=er============:::::::;!l,~s=2=.0=P=-A=B=.S=W=-=0=00=5========~1~==========::::::;q li.:14===== ILoss of Service Water Header Pressure !I S2.0P-AB.SW-0001 ll-=======:=!!j lj~1=6==== ,~========:!.Ill==========-1 ll 11 _ __. iLO. Nombe(;; ,.* *

• Objectives IABSW04E004

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  !auestiol'\ Topie 11 SRO 93                                                                                                                                                                                                                                                                                                                                              I Given the following Initial Conditions at 0800:
   -  Unit 2 is in MODE 5.
   - Containment Purge is in service.

At 0830, Containment Purge is stopped to support a maintenance visual inspection of purge ductwork inside containment. At 1015, a Containment Ventilation Isolation (CVI) signal inadvertently actuates due to a rad monitor switch mis-positioning. At 1030: The duct inspection is complete. Radiation Protection reports containment was not breached in the past 2 hours except to allow entry and exit of workers through the airlock. Chemistry reports there is operational assurance that no radiological changes to containment environment have occurred and requests containment purge be restored. IAW S2.0P-SO.WG-0006, Containment Purae to Plant Vent, is a new containment ourae release form required and why?

     ~I Yes,            because of the CVI signal.

I []I Yes, because the purge was secured. I [2J j No, because the termination was <4 hours and there is operational assurance of unchanged radiological conditions. I

     @] No, because the termination was <12 hours (one shift) and containment has not been breached except as allowed for worker ingress/egress through containment airlocks.

(AAswer1 ~ lexam'l.eve1J [I::} lcognitive,l,.eyef'j jMemory I fa~i\ity: 1 lsalem1 &2 I IE~all\qate:,'! I 12/19/20161

~ j 103000G123                                           J 12.1.23                             f iR,O VallJe: l@fSR,O Vallie)~ ISectiort: lj~ fRo,Gj-0]1p:jLJ \sfio G~oJp:j L J                                                                                                                                                                         ID ~
!system!Evoh.ifion'.:ritlEi'j                                  IContainment System                                                                                                                                                                                                                                                                      ! 103-
'KAStafement:!

Ability to perform specific system and integrated plant procedures during all modes of plant operation. IExplanationofl 55.43(4 )Procedure S2.0P-SO.WG-0006 allows for reinstatement of purge if of short duration (-4 hours) and containment rad I !Answers:* * *'* conditions have not changed ( P&L 3.3) Short duration is NOT 12 hours, and distracter does not address rad conditions. New effluent permit not always required. Can block CVI signal IAW Att 2, Temporary Termination and Reinstatement of Containment Purge. l I' .Reference Title'* ~ t'

                                                                ~'
                                                                                    \,:
                                                                                                                       ., Ir               Faci'!ty Ret,ererjce Number .                                 *IIRefereri~e. s~cuc>rl

• 1r f>ageNo.II i~evision:

IContainment Purge to Plant Vent II 52.0P-SO.WG-0006 I 1130 I I II I lI I I !L.o. Number** Objectives l! i "ll Ii I I WASGASE011

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• SRO System/Evolution List Outline Changes IQuestiollTopic !I SRO 94 Given the following conditions:

I

    -  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.
    -  A Notification has been submitted.
    - The assembly has been in the SFP for 100 months.
    - The SM has given permission, and Radiation Protection has been notified of the movement.

Which of the following describes the Operations Department requirements for this evolution IAW S2.0P-IO.ZZ-0010, SPENT FUEL POOL MANIPULATIONS? A qualified Senior Reactor Operator ... I

      ~ must directly monitor the fuel movement from the crane trolley.

I [bJl I is NOT required if SFP boron concentration is verified >2000 ppm. I [SJ Imust be present in the Fuel Handling Building during the fuel movement. I

      @] I   is NOT required to observe the fuel movement if a Qualified Reactor Engineer is present.
                                                                        .. . . Level                                 . .. .1 I
 /Answer/        ~           1t,:xam Level I ~ ;Cogmt1ve    '                             j j Memory          1 IFacriity: i I Salem 1 & 2         I j~xa~Da~e: 11         12/19120161
 /KA:/l194001G135                    112.1.35            )lROValue:,j              2.2jjSROVah.foj~(section:i1PWG                i[Ro;(3r()up:11    1li5ROGroup:/I      11 Ell~
 \systE!rnli:xotlltion Titre I         *--------------------------------~[GENER!                                                                                                       I rKA StatE!mE!nt:'l Knowledge of the fuel-handling responsibilities of SROs.

l:~~~~;on pf] 55.43(7) S2.0P-10.ZZ-0010, Precautions and Limitations, 2.2 states a Reactor Engineer OR SRO must be assigned for Spent Fuel Pool manipulations. A is incorrect because even if a SRO was assigned, they are only required to supervise from the area, not specifically on the trolley. B is incorrect because SFP boron concentration is not a pre-requisite to who is required to supervise fuel movement. C is incorrect and D is correct because a RE OR SRO is required.

                    * ... )~eferenceTi~le' *         .
                                                '.,i<-i,
                                                             ,-:..*: *. *:I [ ;: . FacilftyRefer,em;e)lµrhlJer '   ;j /RefE)rence section )LJ ,PageN~~ I !Revision!

!Spent Fuel Pool Manipulations I II S2.0P-10.ZZ-0010 I II 1133 I I !I I q II I

!                                                                          II                                       I                           ll          II         I i);.J>. f\tu rilhe'r     :    ** I           Objectives I   REFUELE012                     I I,____,

l,P:uE!~~or\~ollr5e:*j IFacility Exam Bank I [a~e~tio~* MCl~lfi~il_?~;flA~thg~: ,1** IDirect From Source I )us'e~.otirin$ Traiqing ~rogram *1 D 1~9estion;,so~rcE!commetj1sl I

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aao747 I

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  ~estionTopit 11 SRO 95                                                                                                                                                         I Given the following condition:
   -  Salem Unit 2 is in a refueling outage, with all fuel removed from the reactor vessel.

Which of the following identifies when Mode 6 will be entered during actions to refuel the reactor IAW S2.0P-10.ZZ-0009, Defueled to Refueling? Mode 6 is entered when ......

     ~'fuel movement is "imminent".

I

     ~ the Fuel Transfer Gate Valve is opened.

I I

     ~ the first fuel assembly is lowered in the reactor vessel.

I I

     @J I all Administrative Requirements are satisfied and the SM signs Attachment 1 authorizing Mode 6 entry.

I (~nswer 1 Ic I ieJ<amlevel 11 s I [cpgllitiv~'.L'ev~1 JI Memory I lfaciHty: l ISalem 1 & 2 11~~~fnl?~te~il I 12/19/20161 iKA:il 194001G136 I12.1.36  !. !Ro \falue:*1r.:I§Jfs1wva1uej~ !Secticm; 11~ :RO Group;J

                                                                  .,.,                                           **;         I*

LJ !SRO Group:)LJ Biil ~ [systemfEvolotionJitle j  !-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ (GENERI I 'KA Statement* I Knowledqe of procedures and limitations involved in core alterations.

 ,~P1~nat~<? 3 o~~I 1,Answers 55.43.b(6) IOP-9 specifically states (Step 5.1.6), that when the first fuel assembly is lowered into the reactor vessel, Mode 6 is
            **. "'      entered. A is incorrect but plausible if it is thought that you enter Mode 6 just prior to moving fuel. B is incorrect but plausible if it is thought that the action of preparing for fuel movement by interconnecting the Spent Fuel Pool with the Rx cavity is when Mode 6 is entered. Dis incorrect but plausible if it is thought that once the SM gives the final approval for fuel movement that mode 6 is
                        --~- ....... -!

I I. i"' Reference .Title

                                                       }~*
                                                           ..    :I 1** *.Fa,ci)ity Refererice ~uri,ttier. .* j !Re:te~ence Section*

• IIPageNo;.j [Revisic.ml IDefueled to Refueling II S2.0P-IO.ZZ-0009 I 114 1129 I I II I 11 11 I l II I II 1I I iLO. Number: ., . ii Objectives I IOP009E004 1--~

!Puesti~r(~~urceJ          l IFacility Exam Bank                 I \~u~s~i,?~~*s>~ifiea!iOn f'il\etJ1?d'~i:,i'I Editorially Modified      I fuse~ Rin'fng ]:raining Program aj o

======-:::::::::::::::::::::::::::::::::::::::.::::=:::=::.============-================-~1

[Q'u~~tl~n. ~0,'.ur~e,:<~?!Pme~ts: 1118980 replaced 2 distracters with 2 new distracters, correct answer remains the same. I I I I

RO Skyscraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I .Outline Changes I [Questiori ToPiCl SRO 96l Given the following conditions:

    -   Unit 2 is operating at 100% power, SOL.
    -   2A EOG is inoperable for maintenance and won't be available for 24 hours.
    -   The current time is 1200.
    -   Which of the followinQ would require the Unit to be in Hot Standby by 1900?
      ~ 128 EOG is declared inoperable due to an oil leak and Maintenance reports 2 hours later that it will be at least 12 hours to repair.

lliJ IA slow nitrogen leak causes 21 ECCS Accumulator pressure to lower to 595 psig, and initial attempts to raise nitrogen pressure are unsuccessful.

      ~!The containment 100' elevation airlock air leakage test is UNSAT \AW S2.0P-ST.CAN-0004, CONTAINMENT AIR LOCK LOCAL LEAK RATE TEST.

I

      ~ The CRS is informed that a faulty test device was used during the recent refueling outage to test and adjust the lift setpoint of all PZR Code Safety Valves, and that actual lift pressure for ALL 3 valves is 2,735 psig

!Answer 1 Id I lexarniLevef! Is I jcc)gnitive Level J IApplication I IFacility:'.11 Salem 1 & 2 11e~<i.P9ate: J I 12/19/20161 IKA:Jj 194001G222 112.2.22 i\RoValue:ll 4.opsRbValuef~(se,dion:JIPWG lfRC)Group:)I 11lsR0GrbuP,:J! 11 i1§1 2] lsystemlevolutio.ri Tit~e I [KAstaterne!ltil .--------------------------------------------=-

                        !Knowledge of limiting conditions for operations and safety limits.

IEJ(PlCi~Citi~n}?~*) 55.43(2,5) Only one PZR Code safety can be inop in MODES 1-~. and this is 3.0.3. 2 _ED~s inoperable in MODES 1-4 is 2 hours to !Answers: *. ; <* restore, then 6 to HSB. 1 Accumulator has 24 hours to restore. Airlock has 24 hour action time.

• Reference "I:itle .* .t * . *\
• it**! .fapility Reference Number * /I !Re,fereriC:e, Sectioh ' :JI Page' No. I lfievlsforij j~s=a=le~m=T~ec=h=S=pe~c~s~~~::.::;.;:::~~:::;;;:;.:~!l.~3~/4~8~-1~.8=-2=,=3/=4/=4-=6,~3=/4=/5=-1=,3=/4=6=1:;::::;~:.:::;.:;;;:==~~;.::;;;:.lj lj I

'~l _ =================li*~==========~l~====~'I _ _ _ _ _ _ ___.11 11 _ _ ____.lj 11II II tL:o;_**~umbet. **,,:, Objectives I TECHSPE015 j_ ___. [a~e$tio~ ~<iur;ce; ,.j l_N_e_w_ _ _ _ _ __.I ~ql1e~t1,p'n1 M~d,m§a~f~~-'lllW1.~d:.;ltJ __________.I fused our,~i;fTraipio9 Program 1 o l<lt!esr1<>r.i~c:i~r<:r¢#oiw~*nr~11 I

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RO SkyScraper f SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I [Question Topicj !SRO 97 Given the following conditions:

   -  Salem Unit 2 is in Mode 4.
   -  All RCS Tcolds are 280°F.
   -  21 Charging pump is in service.
   -  22 RHR loop is in service in SOC.

While hanging a tagout for 23 Charging pump, the suction valve for 21 charging pump is closed instead of the suction valve for 23 Charging pump. Which of the following identifies how the CRS should respond?

     §]I       Restore at least one ECCS subsystem to operable status or maintain RCS Tavg <350°F.

[g"] I Immediately suspend all operations which may result in a change in RCS boron concentration. [c:l l Within 1 hour action shall be initiated to place the unit in Cold Shutdown within the next 24 hours.

     @] I      Restore at least one ECCS subsystem to operable status within 1 hour or be in Cold Shutdown within the next 20 hours.

!Ansl/ier 11 d I lexa!Tl Level 11 s I [cogni~iye .Level :J IApplication I Ifa~i!ity'.: :I jSalem 1 & 2 I [~*am[)at~~*l I 12/19/20161 jt<A:l! 194001G237 112.2.37 i!RO,V,alue:j~!~RO V~luej~ 'f5ect1on:jj~ !RO Group:ILJ [SRQ ~roup:ILJ Bl ~ [s;iste~/Eyolution Title I *-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~__.!GENERll il<Astatementl . . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - , Abili to determine operabili and/or availabilit of safe related e uipment. [.E.J,CP*J*an.. atio.pof., 55.432.b(~) This question meets the KA by having to determine that with _RCS r.colds<312°F, all but~ charging pump must be l.J>.nsi.yers:* .** *' *: rendered inoperable, as well as both SI pumps, and that one ECCS loop 1s required to be operable With that 1 charging pump, and closing the suction valve of that pump makes it inoperable. Distracter a is incorrect because it is the action (for the same LCO 3.5.3) if a required operable RHR pump were to become inoperable. Distracter b is incorrect because only positive reactivity

• Referei'lceJitle: L. ' . ; *...*: */ l ~.* .:, F;acitity Reference Nunitj~f ;'. 11 (R~ferencEi sElctid~ \ +. *:l I. e~~e,)~.o/.l l~evisipni ISalem Tech Specs II I lj 314 5-7 I j 258 I I 11 I Ii 1I I I II I !I 1I I

[Ll:();' Nuh)ller\: > Objectives I ECCSOOE010 (CiU,~stipt;i~p}l,(f~:p<>r#.fue~~il I 1-~~-~~~~~~~~~~~~~~~~~~~-~~~~~~~~~~~~~~~--'l I I I l

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

• I Outline Changes I

IQi.i~stion Topi(; !ISRO 98 I Given the following conditions:

   -  Unit 1 is shutdown during a refueling outage.
   -  A normal release of 14 Waste Gas Decay Tank to the plant vent is scheduled to be performed on day shift IAW S1 .OP-SO.WG-0011, Discharge of 14 Gas Decay Tank to Plant Vent.

When reviewinq the schedule, which of the following activities is allowed to be scheduled during the period 14 WGDT is being released?

      ~I Release of 11       WGDT.

I

      ~   IInitiation of Unit 1 VCT purge.

I [I ITransfer of gas between 12 and 13 WGDTs. I

      @;] IAligning Unit 2 Vent Header to Unit 1 Waste Gas Compressor suction.

I !Answer! b I I lexarJi Levefj Is l [cognitiye'Leve1/l IMemory I 1ra~intr:'i ISalem 1 & 2 11.l:x~mpate:i I 12/19/2016! fKA:Jl 194001 G311  : /RO Value: IQ]° l~RO Valuej~ !Sectiqn:.11~ IR()Group:ILJ[SR() Group:JLJ -*, I& ~ !system/Evoiution Title I i~~~~-~-~~~-~~--~~-~------~-~--~--~~IGENE~

\Kii.statement:!

Ability to control radiation releases. /.Exeiah~t'.on,.* ot 55.43(4) A is incorrect because only a single WGDT is allowed to be released at a time (page 3). B is correct because the release Answers. *..* 1 procedure specifically allows a VCT purge to plant vent to occur during the WGDT release (page 16). C is incorrect because the release procedure specifically disallows transfer of gas between tanks when another tank is being released (page 3). D is incorrect because while waste LIQUID can be transferred from one unit to the other, waste GAS cannot. 1* Reference Title * .:Ii .:';F~cility R.efer:~nce N~mber *** I[R.efel"e~ce Sectiori :* IfPage No.'l IRevisioni IDischarge of 14 Gas Decay Tank To Plant Vent !IS1 .OP-SO.WG-0011 I II I j31 I ITransfer of Waste Gas !I s1 .OP-SO.WG-0007 I II 116 I I II I II 11 I [L.o.. t;lumber ;J I WASGASE011 Objectives '--~ j9oe.~Uon S(:)~rc~:"*J IFacility Exam Bank

!que~~iops§lu*c~,~omwerl~sl             1.-4-3-64_4_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _                                            __,I I:comment**

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RO Skyscraper I SRO Skyscraper RO System/Evolution List I SRO System/Evolution List Outline Changes [QuestiortJopicj r S - R _ 0 _ 9 _ 9 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - : 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 operator response?

     ~I Do not implement 2-EOP-FRTS-1 until 2-EOP-FRCC-2 is completed because thermal shock is a lower priority CFST.

I [] I Implement 2-EOP-FRTS-1 immediately because the potential damage done to the RPV by delaying entry into FRTS after entry conditions are met may be irreparable. I I

     ~ Do not implement 2-EOP-FRTS-1 until 2-EOP-FRCC-2 is completed because while in FRTS the core will continue to boil away injected accumulator water, and could lead to a RED path for CC.                                                                                                I

[] I 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. I [AnsyverJ jc I [f:x!iinLevetJ ls I !cognitiveJ..eve( I!Memory I [FaciUiy:1lsa1em1 &2 j jEJ(amo~1e:J I 12119120161 IKA:ll 194001G404 I1~2._4.4_~1 iROValue:!@!sRO V:alue~~ [Sec~ior:i: II~ [RO Group:jLJfSRO Group:\LJ Ell f

                                                                                                                                                                   ~

~ls-y-~t-e-.rri-/E_..v_o_iu_t_fo_n_T_i_tl-e,J IGENERI i

KA st.aterrient* I Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal ooeratino procedures.

t~.i:>~~natfo~ o~J 55.43.b(5) Stopping the depressurization to go to FRTS would cause the cooldown to be stopped, and a thermal soak to be jAnswers: ***.** ; . 1 performed. The core will continue to boil away injected accumulator water, and could lead to a RED path for CC. Step 15 Continuous action Step. r.: . Reference Title > . ... A r} I Facility Relerence Number ** * !Reference sec#ofr . H I J;~ge N9, !Revision[ I Response to Degraded Core Cooling Basis Doc 12-EOP-FRCC-2  ! II 1123 11 2 I I II II

                                                                                                          ,,                     II          11        I I                                                                       II                                                        II          11        l I FRCCOOE006 Objectives

M.aterial RequiredJor~ami\1~tiC,ff ** .! I II

[c;iue~tlqn so~~c;e:.* 11 Facility Exam Bank [~~esti(>r:i5~6rce'c§~01.eritsl 1..-5-9_3_1_5_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _..,.I, I I I

RO Skyscraper SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes [9uE!~ti6n:.J9pic 1SRO l 1oo I Given the following conditions:

  - Unit 2 is in MODE 3, performing a normal cooldown/depressurization going to Cold Shutdown IAW S2.0P-IO.ZZ-0006, Hot Standby to Cold Shutdown.
  -    RCS pressure is 870 psig.
  -    RCS Tes are 395 deg.
  -    PZR level is being maintained stable by adjusting charging flow during the cooldown.
  - Charging flow is 79 gpm.
  - The 45 gpm orifice is in service.

A step change of 10 gpm has just been performed on charging flow (from initial value of 79 gpm) to maintain PZR level stable, and is determined NOT to be associated with the cooldown in progress. Which of the following describes an acceptable procedure proqression, and the outcome of usinq the procedure(s)?

      ~I Enter S2.0P-AB.LOCA-0001, Shutdown LOCA. Isolate letdown, adjust charging flow, go to Section 4.0.

I

      ~I Enter S2.0P-AB.RC-0001, Reactor Coolant System Leak. perform an RCS inventory balance, attempt to identify source of leak.

I

     @]    I  Enter S2.0P-AB.RC-0001, Reactor Coolant System Leak, transition to AB.LOCA if leakrate is determined to be > normal charging system makeup capacity.                                                                                                                                             I r<il IEnter S2.0P-AB.LOCA-0001, Shutdown LOCA.                       Isolate letdown, re-align charging pump suction to RWST, isolate containment penetrations       I I ClllU L:UllllllUti 111 /-\0.LU\.,/-\-1.                                                                                                                       I

!Answer j ~ jExailJ Level J [CJ [Cognitive Level 11 Application I jF~~!lity: 11 Salem 1 & 2 I IE!C~mDate: i j 12/19/20161

~ 1194001 G409                        I]2.4.9          i !Ro Value: 1§!SROValuej___gJ [section: 11~ \Ro. Gr9up:! LJ ls Ro Group:] LJ R!l ~

[s~tem/Evolution Title I iGENERI I 1 !KA: ~fatement: .---------------------------------------------___,. Knowledge of low power/shutdown implications in accident e.g., loss of coolant accident or loss of residual heat removal) mitigation strate ies. Explanation of 55.43.b(5) The 79 gpm charging flow given in the stem to start,+ the 10 gpm step,= 89. The stem states that a 45 gpm orifice is in Answers: service. The first thing to do in AB.LOCA is isolate letdown. This will remove 45 gpm from the required charging flow necessary to maintain PZR level stable, so step 3.3 (asking if PZR level can be maintained stable or rising) will be YES. Adequate subcooling is available, and with charging flow less than 100 gpm at 3.13, GO TO Section 4.0. While technically acceptable to go to AB.RC-1 I*~'.!' <;:, * ** ReferenceTitle'.'t<:': .* 11. Facility ~eferenc~ ,r;iµmber * .* j !Ref~r~ilce Section. : '1' lIPage ~o. J [~evision j Shutdown LOCA !j~s=2.=0=P-=AB=.L=O=C=A-=1====="~====~1:.=I=~I 18 I I ll~=======:ll~====::=,l!=I==I ==I !_ _ _ _ _ _ ____.ll, _ _ _ _ _____.11, _ _ ____.lj _ __.l !_~I

                                                                                                                                                     !=I jL;O.,Nt1Jnber\'                                Objectives I ABLOCAE007 1_ ____.

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