ML15009A286

Final Written Examination with Answer Key (401-5 Format) (Folder 2)

ML15009A286
Person / Time
Site:	Salem
Issue date:	12/03/2014
From:	David Silk Operations Branch I
To:	Gauding G Public Service Enterprise Group
Shared Package
ML14174B408	List: IR 05000272/2014301 ML14191A180 ML14329A180 ML14337A574 ML14356A583 ML14356A584 ML14356A585 ML14356A586 ML15009A252 ML15009A280 ML15009A286 ML15009A311 ML15009A340 ML15020A701 ML15156A443 ML15156A635
References
TAC U01894
Download: ML15009A286 (119)

Text

U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: Region: I Date: 12/15/2014 Facility: Salem 1 & 2 License Level: RO  ! Reactor Type: W

- - - - - -t- - - - - - - - - - - -

Start Time: Finish Time:

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

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

Applicant's Signature Results Examination Value - - - - - - Points Applicant's Score - - - - - - Points Applicant's Grade

- - - - - - Percent

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-t~~irr~'&t1fum \R01 - .. ---- ----------* - -*-- - --- ...-.--- - - -- .. - ---------* ---- _. _______

Given the following conditions:

- Unit 2 is operating at 100% power.

-Aoegraaeostarorwat~llng-system*-conditlon-ca~ses*a Maln*!3enerator-Stetor----------------------- - - - - - -

Water runback to occur.

The runback terminates at 80% power when the initiating signal clears.

Durtng the runback, the RO reported 2 control bank D rods have stopped moving at 215 steps. -

The CRS entered S2.0P-AB.ROD-0001, Immovable I Misaligned Control Rods, and rod contrails placed In manual.

Control Bank D Group Demand Is 185 steps.

A Rx ~ip Is not required.

Which of the following ldentiftes the minimum required action for how the crew w!ll proceed?

Assume the 2 Inoperable-control rods w!ll f\IOT-be restored to operable status, and the remaining controL rods.wiii.NOT be realigned to the Inoperable rods.

!Jj Place the unit In Hot Standby.

~ I Place the unit In Hot Shutdown.

~?ill Reduce power to <50%, power operation at up to 50% may continue.

m; IReduce power to less than 75%, power operation at up to 75% may continue.

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

~ue-Stion Topic.: i RO 2 Given the following conditions:

- Unit 2 was responding to a SGFP trip from 100% power at EOL.

- 22 SG NR level reached 16% and continued to drop.

- lAW S2.0P-AB.CN-0001, Main Feedwater I Condensate System Abnormality, the CRS directed the RO to trip the Rx.

- The RO turned the Rx Trip Handle on 2CC2 and performed the immediate actions of EOP-TRIP-1.

When reporting his review of the OHA's prior to the first shift brief in the EOP's, the RO reports the following "F" Window alarms are locked in:

F-3 21 SG LVL LO-LO F-11 22 SG LVL LO-LO F-19 23 SG LVL LO-LO F-27 24 SG LVL LO-LO F-36 TRB TRIP & P-9 F-44 MAN RX TRIP INITIATED The F-11 OHA is red, while all the others are white.

Which of the following describes the information provided by the "F" OHA Window Boxes?

~r IIC III;:)L f'.A LllfJ .,. s LO-LO SG NR level. An ATWT has occurred. *-*1 I

~ The first Rx trip signal was the manual Rx trip. The F-36 window indicates the Main Turb failed to automatically trip.

'C.' ~The first Rx trip signal was the manual Rx trip. The red box only indicates the first automatically generated Reactor Protection System trip.

~The first Rx trip signal was LO-LO SG NR level. Only a review of the Sequence of Events Recorder can determine whether or not an ATWT has occurred.

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_____ - - - - Titlel IReactor Trip

,lsystemlEvol~tion i.KA St~t;;men~~. ~ ility to determine and interpret the followingas they apply to Reactor Trip:

_~or trip first-out indication Iexplanation of 1 55.41.b(7) The "F" windows have dual backlights, red and white. The first signal to be generated to trip the Rx is locked in RED,

'Answers: and can only be reset with a keyswitch and SM permission. In the above condition, the time it took to order and carry out the

--* - -

• manual Rx trip was sufficient to allow SG NR level to lower past the auto trip setpoint of 14%. Since a manual trip was ordered but an auto trip occurred, the SER must be reviewed quickly to determine which signal was sent to the RPS system first; a manual trip

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an auto trip occurred before the manual trip, while the F-36 does indicate the turbine tripped before the Rx. C is incorrect because of B above AND because the RED box is the first TRIP signal, not the first AUTO TRIP signal. occurred as the Rx was tripped though the second part would be correct if it was thought that a manual trip occurred first. A is incorrect because an ATWT may or may not have occurred, and Dis more correct because the SER must be reviewed to determine if the manual trip was initiated before the auto trip occurred.

- ----- -- ---- - - ------ ---**--- ----- ----1 ---- --------r-;::;:- -

-*Reference Section

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Reference Title - - -1 1

Facility Reference Number

--- --- -* 1 Page No. . - Revision' IOverhead Annunciator Window F II S2.0P-AR.ZZ-0006

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

I lj I 16 I IOverhead Annunciator System II NOS05ANN00-06 I J141 j 6 I II -*-- Tl I l[_ __j I L.O. Number l I OHAOOOE008 I _j

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

~uesticm Topic 1Ro 3 I Given the following conditions:

- Unit 2 is operating normally at 100% power when one PZR safety valve fails full open.

- All plant systems respond as designed.

Which of the following identifies how PZR level will respond after the Rx is tripped?

PZR level will lower initially, then ...

liJ Irise rapidly until the PZR becomes water solid.

I I

'b.: rise very slowly until the PZR becomes water solid.

I rc.J I lower rapidly until the PZR empties and remains empty.

I 611ower very slowly until the PZR empties and remains empty.

I Ans~er II a I Exam Level: IR I Cognitive-Level II Application I IFacility: Isalem 1 & 2 I Exa111oate~ll 12/15/20141

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• KA Statement:

'!Explanation of 55.41.b(5,7, 14) A characteristic of a vapor space accident is that pressure and level will initially lower, then as the RPV begins to

,Answers: i void, level will rapidly rise in the PZR, whereas a LOCA would lose pressure and level.

Reference Title Objectives l!Jtatl;)!_i~I-~I;)(:JUir_ed for Examinati()n :I II

~estion S~urce_:_j IFacility Exam Bank I Question Modificati~~ Method:

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[ci~estionSo~~ce Comme~ts-~ 057432 changed from why level rises rapidly to what does level do I

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

-auesti~~ -Topic i 1Ro 4 I Complete the following statement:

EOP-LOCA-4, Transfer to Hot Leg Recirculation, is performed when directed during the response to a Large Break LOCA to ...

a. Iensure enough boron remains in the RPV to provide adequate SDM as long as vessel level remains > 39% RVLIS.

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b. prevent thermal gradients across the upper vessel from becoming fissures which could divert recirculation flow from the core.

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'c. ensure boron does not concentrate in the reactor vessel (due to boil off) to the point of solidification and blockage of coolant channels.

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• ~

prevent thermal stratification of the fluid in the core which would add to the assumed 1% fuel damage which has already occurred and is accounted for in the accident analysis. I

!Answer j c

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Explanati~;, of I

~llswer~:_______j

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Re-re_r_e_n-ce-T-itl_e_ --------- :I J Facility_ R_~ference Nu~ber _ Referenc-esecti~r;- JiPage No. i Re~~~~~~

r===~-=-====~========~~~~======~~

ILoss of Reactor Coolant 112-EOP-LOCA-1 II Basis Document !I 51 1[28 I

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- -***' Ob"Jecf1ves _I

! LOCA01E010 Material Required ---****-

for Examination

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I qu~~!ion Topic IRO 5 I IWhich of the following is an unexpected control room indication I alarm if a RCP thermal barrier rupture occurs? l

-a.l Cooling surge tank level lowering.

I

~ 12CC131 Thermal Barrier Return Valve, indicates closed in AUTO.

rc~~ Rising activity or alarm on 2R17A or 2R17B, Component Cooling Radiation Monitor.

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!ct] RCP Thermal Barrier DISCHARGE FLOW HI console alarm will annunciate then clear shortly afterward.

I

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I 12/15/20141

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~1~!e_ll'l~~volution Title IReactor Coolant Pump Malfunctions I i015 -~

edge of the interrelations between Reactor Coolant Pump Malfunctions and the following: I I

'Explanation of j 55.41.b(3, 7) The high flow alarm would come in as reactor coolant flows into the thermal barrier CCW system. This increased flow Answers: 1 would cause a momentary hi flow alarm, then the CC131 return valve would auto shut on high flow. The RCS flow into the CCW

• -*- ** -- system would be seen on the CCW surge tank rad monitors R17A and B. CC surge tank would RISE, not lower.

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____:::_:::.:::. ~-] Fac~l_ity Refere-nc-_~--N-u-mber  ! ~re~ce section I ~~ ~o:J ,R~

j Reactor Coolant Pump Abnormality II S2.0P-AB.RCP-0001 I II 1121 I I Component Cooling System Simplified 11205331-SIMP I II II 0 I II ll _ _ _ _ ____.l II II I

'Lo. Number Objectives I~----*-***

ABRCP1 E005

Material Required for Examination

• • • -* - II !I

Question Source: II Facility Exam Bank I'Ques~io~ n,todification Method:_j Editorially Modified IrlJ;ed During Training Program I []

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I Question Source Comments

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RO SkyScraper I SRO Skyscraper RO System/Evolution List SRO System/Evolution List Outline Changes Euestion Topic'  !---------------------------------------------'

.-------------------------------------------------------------------------------------------~!

Given the following condition:

- Unit 2 was operating at 100% power when 23 Charging Pump tripped.

Which of the following identifies an action which must be performed lAW S2.0P-AB.CVC-0001, Loss of Charging, prior to starting a charging pump, and wh ?

~ Check RCP seal inlet temperature <225°F to prevent damage to seals when CVCS flow is restored.

'c] Shut 2CV55, Charging Flow Control Valve, to prevent water hammer on the Regenerative Heat Exchanger.

If] Open 2CV71 Charging Header Pressure Control Valve, to prevent seal injection flow from being re-established> Tech Spec limit of 40 gpm total to all RCPs.

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'system/Ev:o~tion T_itl~ ILoss of Reactor Coolant Makeup i p22 __ - I

~ ~tatem~nt] I Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup:

Consequences of thermal shock to RCP seals

• ..  ::1 Explanation of i 55.41.b(1 0) AB.CVC-1 states to check RCP seal inlet temp <225 OR seal injection isolated. Seal isolation is not one of the

'Answers: available choices. The bases for AB.CVC-1 says this is done using LOPA-1 as guidance. LOPA-1 bases doc says that seals are

~--***--***

isolated (because in LOPA you have additionally lost all CCW flow and seals HAVE heated up) to protect RCPs from seal and shaft damage that may occur when a centrifugal charging pump is started. While VCT is the source of NPSH to the eves pumps, it is I

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to prevent e~cessive flow, and the CV55 is normally full open at power. The cV71 is not adjusted until after the c"vcs pump is started, but the reason is correct.

I I Reference Title _ j :- Facility Reference Nu~ber - ! Reference Section __j i Page No. I rRevision-ILoss of Charging II S2.0P-AB.CVC-0001

---

---- - - - - - - - -- *---- ---- --- - - - c ___ - - - ~ --- ---- i i II Bases Doc 11 2 11 9 I j Loss of all AC power ~~~-EOP-LOPA-1  : ll Bases Doc i134 1127 I I II II II ]I I r~-O~be_r__ J I ABCVC1 E002 I

,Material Required for Examination , I II I

I

method, If any, of establishing Rapid Boration If the 2CV175, Rapid Stop Valve will not open

~~ I Knowledqe of the interrelations between Emerqencv Boration and the followinq:

!Valves

~~ 55.41.b(6) Using the 2CV175 to establish rapid boratlon Is the most direct way. However, with closed, there are 3 other ways lAW 1-~ S2.0P-SO.CVC-0006 in which Rapid Boratlon can be established. B is incorrect because the 2CV40 or 41 VCT outlet valves are required to be shut, otherwise the RWST water will not have enough head to be sucked into the charging pump suction, with 30 psig in the VCT. Cis correct because the flowpath from th:;, BAT pumps !~rough the 2CV172?~~~;~e 17~?~~~~~~ablish flow. Dis

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List J SRO System/Evolution List I Outline Changes I l§uestion !()P~ IRO 8 I Given the following conditions:

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

- 21 RHR pump began cavitating due to a valve being mispositioned during a tagging release.

- The CRS entered S2.0P-AB.RHR-0001, Loss of RHR, and stopped 21 RHR pump.

- Plant conditions allowed time for normal restoration and local venting of the RHR System.

Which of the following describes the preferred flow rate when starting the RHR pump, and why?

'~-=I Higher flow rate to sweep entrained air from system.

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b~ Lower flow rate to prevent high starting current on the RHR pump.

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

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I d; Lower flow rate to limit initial sudden cooldown and to minimize level loss caused by collapsing voids.

I

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~!i~em/Evo!uti~! l Loss of Residual Heat Removal System I 9_~--

IKA statement::

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[Explanation of I 55.41.b(1 0) The preferred rate is a lower flow. The CAUTION on page 10 states that it is for the reason as stated in the correct Answers: ' choice above. Higher flow rate to sweep entrained air is the method used when time does NOT allow a normal venting as described in the stem. (CAUTION PAGE 14)

Objectives IABRHR1E004 I I

_r,naterial REjquired f()l' Examinatiorl_j I II

!auestion Source:

j IFacility Exam Bank _j [Question Modification Method: JDirect From Source Iiused During-Training Program ; D Que~~~~-~~u~ce_C_oiTI_ITients I0120590 (used on 8/2008 RO NRC exam, 4 NRC exams ago.)

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

~uE!!>lionToPic I I.R09 I Given the following conditions:

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

- A Component Cooling Water leak results in entry into S2.0P-AB.CC-0001, Component Cooling Abnormality.

- Make-up can maintain CC surge tank level > 38%

- The crew has implemented ATTACHMENT 4, Leak Isolation Method.

- When make-up is stopped, surge tank level lowers with either CC header in service.

Of the following, which is the only component that could be leaking to cause these indications?

~-, 22 CCW HX.

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b_.i I I fc:ll23 Charging pump mechanical seal HX.

1 I

d~l~ Boric Acid Evaporator Distillate Cooler HX.

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.Answer' j b I ~xam Leveilj R I [Cog.,itiv~ Le~elj IApplication  !!Facility: 11 Salem 1 & 2 I fExamDate:J j 12/15/20141

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I_KA:!!oo0026A102 '

~~teffiTE~uticm Title 1

j Loss of Component Cooling Water I 'o26- l

~ St(:ltem~nt~ Ability to operate and I or monitor the following as they apply to Loss of Component Cooling Water:

Loads on the CCWS in the control room

!Explanation of 55.41.b(4) With the stem stating that the leak continues with EITHER CC header in service, that means the leak must be on the

• ~An~sw_e~rs:___ _j Non-Safeguards header, which is supplied from both CC headers. Of the 4 choices, 2 are on the safeguards header, and 2 are not. Of the 2 possible answers, the Boric Acid Distillate Cooler HX is not normally in service. That leaves the SFP HX, and SF cooling pressure is< CCW pressure, meaning the leak would be out of the CCW system. This question is different from Question 5 in th~* i~ ,.,, oadi"n '"'"' oira~ "f ~""'"'""' """""" ~' orinro >n,-o "f tho IIR t" I"""'"' tho la<>v uhora<>~ rlo oocti"n &; ic Ia Thermal Barrier question j

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Numb~-~~ 'Referer1ce S(lcti~rl__-1 ~~_e NOJ ~evisio~

I Component Cooling System Abnormality II S2.0P-AB.CC-0001 I il 1114 I I Component Cooling System Simplified 11205331-SIMP I II 11° I I II II il iI l Objectiv l111Jate_rial ~equired for Ex~minatioll I I II

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~esti~n~~~-e_Co~m~m~en~t~ jri0--57--73--2------------------------------------------------------------------------~~

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I lciuesti()-n Tof1ic l 1R01 o Given the following conditions:

- Unit 2 is operating at 60% power.

- There is a power ascension in progress at a rate of 10%/hr.

- PZR Pressure Channel Ill, PT-457 is selected for CONTROL.

Which of the following describes RCS pressure response if PZR Pressure Channel Ill fails low with no operator action?

RCS pressure will rise until. ..

~~~ONE PZR PORV I

~ BOTH PZR PORV's

[(] I the PZR Spray Valves open.

l<!J I a PZR Code Safety Valve opens Allswer1l a I rExam Level II R I ~gnitive Lev~ IApplication I [Facility:~ ISalem 1 & 2 I ~xan1Date: i I 12/15/20141

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Title:

= Pressurizer Pressure Control Malfunction 1 1027 ~-

~_stater11en!=J Knowledge of the interrelations between Pressurizer Pressure Control Malfunction and the following:

Controllers and Positioners

Explanation of 55.41.b(7) The failure of the controlling PZR Pressure channel causes the Master Pressure Controller to sense a low pressure 1Answers: condition, and its output will go to zero. A 0% demand will cause all PZR heaters in auto to energize, and PZR Spray valves to shut. RCS pressure rises slowly but spray valves will not open because the MPC still sees a low pressure condition from the failed low PZR pressure channel. The PZR PORVs 2PR1 and 2PR2 are 2/2 coincidence required to open, from PZR pressure channels 1/':l <>nrl ?{,1

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psi g.

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i____ _____ Reference !itle ****--..**---~ L . -~cility Re!ere__nce Num~~':_j Referemce S~ction .. ~_ll_age "!(;)~J 8_~yisioni IPressurizer Pressure Control Malfunction II S2.0P-AB.PZR-0001 1

I I 1118 I l RPS PZR Pressure and Level Control 11221060 I I

,, 117 I 1PZR PORV Valves 11"2~1357

~ :I 1115 i

~C>~~'!'~------- I Objectives IABPZR1E001 I I

.IV!aterial Reqllired for Exa!Uinati_o~ I J

@~:stion So~rce: Jl Facility Exam Bank J [QUestion Modification Method:- ---[~ditorially Modified l[i.lsed During Training-Program]

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oue&io~ToeicJ ~----------------------------------------------------------------------------------------4 Given the following conditions:

- Unit 1 has experienced a RCS leak while operating at 40% power.

- The crew is responding lAW S1.OP-AB.RC-0001, Reactor Coolant System Leak.

- As conditions continue to degrade without an automatic or manual Rx trip, which of the following identifies a condition where an ATWT is present and a manual Rx trip is required?

lf-1

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,system!Evollltion-Tit~ IAnticipated Transient Without Scram

!KA~ta~meriOr-----------------------------------------------------------------------------------~

Ability to verify system alarm setpoints and operate controls identified in the alarm response manuaL Explanation ofl 55.41 .b(7) A is incorrect because it is the normal value for loop DIT at 40% power. B is incorrect because with Rx power <P-9 I~nsw~r~ --' (49%), a turbine trip does not initiate a Rx trip. Cis incorrect because the 17% threshold for PZR level is heater isolation, not Rx trip. D is correct because the auto trip setpoint for PZR pressure is 1865 psig.

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_;'io1~kySer~~e'r;l ;;'S[{~j~Y~'t}P?r::J ;"'~ ~ctsy5~fu!~~lt!~9AiJ~~- j- ~- sRil;si.~t~llit~>v9Jo~?1f}'~t~J i.:crJII~nlcti~ncie~l

!9'~1io!M I Ro 12 J Given the following conditions:

Unit 2 is performing a Rx startup. _

- Power Is 1.0E3 cps.

- Source Range Nuclear Instrument (SRNI) Channell (2N31) fails LOW.

Which of the followinq identifies why power must be maintained less than P-67

~ SR/IR overlap at 3.0-5.0 E3 cps cannot be verified with only one SRNI.

~ Permissive P-6 will not energize when required with only a single SRNI channel.

~ The ability to monitor Rx power on anything other than a one dimensional plane is lost. I

~ A single SR channel cannot be considered reliable with no other Rx power Indication to verify It against.

I

~ ID ~ r.o [Goirtt!tl>1li'EJ IMemory I ~I Salem 1 & 2 I (!!~nf'll~ I 12/15/20141 tlE;]Ioooo32K3o1 IJAK3.o1 J~@tmi.Y\ituiiL~~!m((~[JJtl'IR~\!~OJ Ill~

~~ ILoss of Source Range Nuclear Instrumentation I ~l a I to Loss of Source Ran e Nuclear Instrumentation:

Below P-6, the SR and IR Nis may not be overlapped. This in actuality reduces Rx power Indication to a single channel, and while adequate for shutdown monitoring, cannot be relied upon to provide Rx power indication when performing a startup. Tech Spec bases for 3.3.1.1 for Rx trip Instrumentation generalizes about all Rx trip and ESF Instrumentation, but states that the maintaining operability Is to. *:"2.) the specified ccincldenc;> logic and sufficient redundancy is maintained to permit a channel to be out of service

~~

IEXCOREE012 I I I I I II

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List f SRO System/Evolution List I Outline Changes I rcluestionTopiCl IRO 13 I During movement of irradiated fuel in the Spent Fuel Pit with the Rx in Mode 1, a Spent Fuel Assembly is not fully withdrawn from its rack before the Spent Fuel Crane is moved. The Spent Fuel Assembly is visibly damaged when the crane moves.

Which of the following conditions would require ALL personnel to evacuate the Fuel Handling Building lAW S2.0P-AB.FUEL-0001, Fuel Handling Incidents?

[£ I Radiation level in the FHB reaches 1 R/hr.

I

,.-,I

~ Bubbles coming from the damaged fuel assembly.

I I

lc_j Fuel Handling Crane motion locked out upon reaching Radiation Monitor 2R32A alarm setpoint.

~-*I Automatic re-alignment of the Fuel Handling Building exhaust filter train to HEPA plus Charcoal has occurred.

LAn~wefll a I ~m ~e~elj IR l ~ognlthte Level II Memory !IFacilit~ ISalem 1 & 2 1 ~amDate:11 12/15/2014!

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Titlel IFuel Handling Incidents I 036--

fKA Statement:

'------*----*~~

1 Knowledge of the operational implications of the following concepts as they apply to Fuel Handling Incidents:

Radiation exposure hazards fexplanation Ofl The R32A distracter is plausible if it is thought that because the crane can't be moved everyone should evacuate. AB.FUEL-1 CAS 1 Answe~~-*-l 1.0 states to evacuate at 1R/hr. Not LOD 1 because AB.FUEL-2 (Loss of Refueling Cavity or Spent Fuel Pool Level) don't evacuate FHB until 2R /hr. Bubbles coming from fuel assembly may be present, but are not cause for evacuation until the radiation from them, or any other cause, reaches 1R/hr. Ventilation Realignment is expected to occur either on high local rad level or manual

• <>rfoo<>finn

~-- -

Reference Title

_j  ! .Facility Reference .Number-= !Reference Secti~_j Page No. ! IRevisiOnl I Fuel , *a* ou*n '\:1 Incident I ~? OP-AR.FUEL-0001 I I I 5 I I I I I I I l I I I I I IL().Niiffibe;:- -- l Objectives I ABFUE2E002 I

L~aterial Required for Examina~ I :U IQuestion Source: j IFacility Exam Bank I ~estion Modification Method: -,, Concept Used j[used During Training Programl D I

L_____.__ - *- * -

• ________j

~-Question Source-Comments' 078657

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• _____ __j I

((Comment *---=]

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I

RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I ra~estion Tol>? IRo 14 I Which of the following describes when rising radiation levels on 2R19A, STM GEN SLOWDOWN RAD MONITOR, will automatically close the 21GB4, SG BID OUTLET !SOL VALVE, and why?

2R19A in will close the 21GB4 raJ r;;;-ning, to prevent the spread of contamination from a Steam Generator Tube Rupture (SGTR) on 21 Steam Generator to secondary terns. I

-~ I Alarm, to prevent the spread of contamination from a Steam Generator Tube Rupture (SGTR) on 21 Steam Generator to secondary systems.

I rc.f,_warning, to prevent backfeeding contamination from 21 Steam Generator to any other Steam Generator through the unaffected Steam

~ Generators blowdown lines. I J

~- Alarm, to prevent backfeeding contamination from 21 Steam Generator to any other Steam Generator through the unaffected Steam

• - Generators blowdown lines. I I:A.n~w~ Ib I iExamLevel ~ IR I 'cognitive Levell! Memory I [i=aCilitY:i ISalem 1 & 2 I ,Exam Date: J I 12/15/2014!

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~=!I 000038K303 I'

fS_ystem/Evolution T!~ j Steam Generator Tube Rupture iKA stateme~t:] edge of the reasons for the following responses as they apply to Steam Generator Tube Rupture:

atic actions associated with high radioactivity in S/G sample lines

~---

[Explanation of I 55.41.b(11) B is correct because isolating the blowdown path from the S/G to the condenser will prevent the spread of Answers:***---** contamination, and also will prevent any type of release from the main condenser to atmosphere. A is incorrect because the auto

- **-~-*

closure occurs upon an Alarm signal, not warning. C and D are incorrect because the S/Gs each have its own blowdown line, so backfeeding contamination is not possible through the blowdown lines.

L_

• • • Refere~ce Title=:=-= C Faci~ity Reference Number ].Reference Section J! Page~ ~~~isio~

IRadiation Monitoring Systems Operation !I S2.0P-SO.RM-0001 j 1j2o j j38 j

!I I _j II II I II I I II II I Objectiv

!_Material Required for Examination.

~----**---*----*-

, I II eu-~~tion Source: j I I ilused D_uri~g Tr_ainl~!!.fl_I"Ogr~_i []

• * --** -~* --~----~----,

Question Source Comments I**~

Facility Exam Bank

• • ---****--*****-~

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-;::;::;::::;::=::_::=======:~==--=::::::::.:=:=========--===-===---==-===-~1

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

[Oues!i~n To~ IR015 During an event in which a steam line rupture cause ALL SG's to blow down completely, why is AFW flow and steam release used to prevent the RCS from heating up?

RCS heatup will. ..

a Ithe result in a larger Delta-T between the AFW injection flow and the internal temperature of the SG J-tubes, which can cause water hammer in feed ring when incoming AFW flashes to steam.

[§:. I f cause PZR level to rise, which will repressurize the RCS. The severe cooling of the RPV downcomer combined with the pressure rise can

• cause a flaw in the vessel to propagate threatening the integrity of the vessel.

I

~~ result in a larger Delta-T between the ECCS injection water from the RWST and the RCS cold leg injection points, which can cause a flaw at the ECCS to cold leg piping weld to propagate threatening the integrity of the RCS.

ldJ I cause PZR level to rise which will repressurize the RCS. The thermal stress on the SG secondary side components from excessive heat transfer during the blowdown, combined with the pressure rise across the SG tubes can cause tubesheet deformation and leakage.

1Ans~erlj b I [Exam-Level II R I ~nitive_Levelj IMemory I 11 Facility: ' Salem 1 & 2 I LExamDateJ I 12/15/20141

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~joooo40K101 -------

~y~te~/Evoiution Titiel !Steam Line Rupture I 'o4o-- I lKA ~taterf1El_nt:

• Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture:

Consequences of PTS Explanation Of 55.41.b(3,4) RCS heatup after a rapid cooldown/depressurization can result in a Pressurized Thermal Shock condition, as described IAnswers: on page 2 of FRTS-1 Basis Document. Distracter a is incorrect while it might be what happens, its not why the RCS is prevented

~ from heating up, and the reason is loosely based on the Indian Point Feed Line water hammer event. Distracter cis incorrect

-~--

because it is RPV failure that is a concern, not cold leg piping. Distracter D is incorrect because the pressure delta across the 11h 1hA<: i<: <::r, *no rl. * 'th ~ r. m * ,,;, ,f nrim!>n.* <:inA nnt th to *h

• nrnhl"m I

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Reference Title L-........... *..* _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - -

-~ 'l -Facillty _Refe-rence Number _j ~eren~e Sec!ion~ r Page No.-I fflViSi()ll, IResponse to Imminent Pressurized Thermal Sh 112-EOP-FRTS-1 l1 Basis Doc 11 2 i 125 I I II Jl II II I I II II II II J r--: ~ *-- -~]

~~er_~--- Objectives 1 FRTSOOE002 I

~terial ~equired for Examination JI II

~Cluestion So~rce:J j New I fOuestio~ Modification Method: ] IluSed During Training Progra~ D r:-**------*-----1, Question Source Comments

• • ---------~*- *--~-- --

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Comment

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I I

RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I l(;lue~ii_o_n_ Topl~]l RO 16 I Given the following conditions:

- Operators are recovering from a total loss of all AC power.

- While in 2-EOP-LOPA-1, Loss of All AC Power, 2B 4KV vital bus has been energized from off-site power.

- The crew has transitioned out of LOPA-1 to LOPA-2, Loss of All AC Power Recovery I Sl Not Required.

- Safety injection was initiated as directed in LOPA-1, but is not required.

Which of the following describes how the listed equipment has, or will be, operated?

~a],22 CCW pump was started as soon as a SW pump was started in LOPA-1.

J

~b.]l21 Charging pump will be started after RCP seal return valve 2CV116 is shut.

I rc. 121 Charging pump was started as soon as a SW pump was started in LOPA-1.

I fd~, 122 CCW pump will be started after Thermal Barrier return valve 2CC131 is shut.

I r.Answe? d I I!exam Level I i R J Cogniti~e Leveill Memory I ,Facility: II Salem 1 & 2 I ~- .~* te: ll 12115120141

~ **-* r-**

~:II 000055A 107 iSystEmt/Evo~tion Titlil IStation Blackout I o55 I--- I I~ statem_e_rttil Ability to operate and I or monitor the following as they apply to Station Blackout: I Restoration of power from offsite I 1

!Explanation of 55.41.b(1 0) A is incorrect due to not starting CCW pump until Thermal Barrier return is isolated. B is incorrect because seal return

Answers: isolation is not the concern, seal injection to a hot RCP seal is. CVCS pump not started until RCP seal inlet is isolated.

--~** **-*-*---*'* '

Reference Title

~J __Facility Referer~ce N~lll~er _] ~eference_Section_ __:.= ~ge ~ ~~~ision I

Loss of All AC Power 112-EOP-LOPA-1 I II,, 1127 I I Loss of All AC Power RecoveryiSI Not Required !12-EOP-LOPA-2 I 1122 I II Jl I II II I

~NUJ!I~. _ _j Objective ILOPAOOE013 I I

I IMaterial Required for Examination I I II Question S()u~ce:  ! IFacility Exam Bank  !]Question Mod.ificati()n. Metho£!1 Editorially Modified I~Used During Training Program J []

~esti()_~ ~~rc~_~()mments 1 I 0148063 modified to include procedure transition in stem.

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

[oll_eSt~o~TopicJ 1~<)_ 17. I IWhich of the following describes how a power reduction would be performed after a loss of the indicated Unit 2 115VAe Vital bus? I I

~'!-J A loss of 2A 115VAe Vital bus would require only the use of boration due to the loss of input to control rod speed and direction.

I EJ IA loss of 28 115VAe Vital bus would require only the use of manual rod insertion due to the loss of eves totalizer function.

I

~~: lA loss of 2e 115VAe Vital bus would require only the use of manual rod insertion due to the loss of eves totalizer function.

I

[d.j 'A loss of 20 115VAe Vital bus would require only the use of boration due to the loss of input to control rod speed and direction.

I

!Answer 1 d I I ~~xam Le~ IR i lcognitive Lev~IJ IMemory I rFac~lity;lj Salem 1 & 2 1 ~am nate: ~ I 12/15/20141

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1!oooo57K301 lrAK3.o1 l

lsystemlE'folutio~!~ Loss of Vital Ae Instrument Bus I ~~~----=

~statementi

---

*----~

!Explanation of]

~l;wers: .. ___j e vital loads is not required, knowledge of how rod control is affected by each of the 4 155VAe vital instrument buses is not a.

C ___-~itt~---=-~~ L:"F~~~~~ ~cesectioo:=J ~Page N~ ~e~

ILoss of 2A(B,e,D) 115 VAe Vital Instrument Bu II S2.0P-AB.115-0001 (2,3,4)

' ---------*-~

LE-_Number _ J I AB1151 E003

[ I l I Material Required for Examination ..J I II II New  : i iOuestion- Modification Method:

[9uestion Source:

~****--*---*--*

. - - -.... *--*** -----, ~**-*--~w*----*****-----**--oo*

11 1[\&ed During Training Program

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]Question Source Comments]

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I I

I

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

[9!:'estion Topic: IRO 18 Given the following conditions:

- Unit 2 is in MODE 3, NOP, NOT.

- The control room receives OHA B-18 2C 125VDC CNTRL BUS VOLT LO

- Upon further investigation, the NCO reports that 2C 125VDC bus voltage is at 126 volts, and no current is indicated on 2RP9.

Describe the condition which is present, and the actions required to be taken?

2C 125VDC bus is ...

l

~J within the normal operating band, direct maintenance to raise the charger float voltage.

I

[~l~ experiencing a minor short-to-ground, initiate S2.0P-S0.125-0004 125VDC GROUND DETECTION.

I I

~~ below the Tech Spec minimum setpoint, secure the operating battery charger and place the standby battery charger in service.

I

~~~I above the Tech Spec minimum setpoint, ONLY continued monitoring for any indication of further voltage degradation is required.

.._ I I

[AnslllfE!I'_j a I "Exam LeveiJ IR I l~_ognitive L~V:~JJ IApplication I ~~1ht~j ISalem 1 & 2 I [ExamDat~: i I 12/15/201411

.~=Jioooo58G446 ~~~~~@-s~~~~~~~~~~~U~~JU

[SystemfEvOilliionliiiel

[KAsta!~~e~t~ir---------------------------------------------------------------------------------------~

Ability to verify that the alarms are consistent with the plant conditions. I

'Explanation of [ 55.41.b(8) A is the correct answer because the control band as specified in the NCOs logs is 125-139.8V. Voltage is in the normal jAnswers:_ _____ band, and the AR states to have maint adjust the float voltage. Distractor b is incorrect because there is no indication of a ground.

Distracter cis incorrect because action IS required lAW ARP. Distracter dis incorrect because voltage is above the TS limit Reference Title ~ --F_a_c_il-it-y Ret-e-re-n-ce.N_u_m-be-r IRetere~e-se~ti(IQ ~~~ ~~~o~

IOverhead Annunciator Window B jj S2.0P-ARZZ-0002 II ll-:::13=5=::::::::;j 1.:::::13=5=::::;1 Flc=o=nt=ro=IR=o=om=Lo=g=s=M=od=es=1-4=====;ii~=S2=.0=P=-=DL=.Z=Z=-0=0=03====~~~FA=tt=.1====::::::::;1148 1IF9=7=::::;1 11 _ _ _ _ _ _ _ _ _____.111 _ _ _ _ _ ___.11 II 11 _ __.1

~~~_____j Objectives IDCELECE008 I j

I Material Required forJ:xaminati~ll _j I II 1

Question Source: *11 Facility Exam Bank llauestion Modification Metho~: 1Direct From Source 1

ILused During Training Program I D

Question 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

[Qu~siion Topis l Ro 19 I Given the following conditions:

- Unit 2 is operating at 7% power.

- Unit 1 and Unit 2 Operators receive Console Alarm CONTROL AIR PRESSURE LO.

- Station Air header pressure is 110 psig and steady.

- Control Air header "A" pressure is 78 psig and dropping slowly.

- Control Air header "B" pressure is 93 psig and steady.

Which choice describes the actions required to be performed by the Unit 2 operators?

~I Immediately trip the reactor and GO TO 2-EOP-TRIP-1 REACTOR TRIP OR SAFETY INJECTION.

I

~I GO TO S2.0P-AB.CA-0001 LOSS OF CONTROL AIR, and trip the reactor due to loss of the "A" control air header pressure.

J iJIInsert control rods to lower power to <5%, and start 21 and 22 AFW pumps lAW S2.0P-AR.ZZ-0011 CONTROL CONSOLE 2CC1.

I I

~ GO TO S2.0P-AB.CA-0001 LOSS OF CONTROL AIR, and verify redundant air panels have swapped to the "B" control air header.

I IAnswer:J d I I !Exam Leyelll R I lcogJ1itive Lev~llj Application I racility~ lSalem 1 & 2 I!Examoate:ll 12/15/20141

- -- - ]

lKAl f000065G1 07

[System/l::volut~o_ll Title !_;__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___,

2_65_ J

[KA statement*

--** --*---- ~_j 1

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. I

explanation oil 55.41.b(7)A is incorrect because a reactor trip is not required since the given condition in the stem does not indicate a loss of IAnswers: _ _ j Control Air on both headers. B is incorrect because a reactor trip is not required. D is correct because the ARP for the alarms directs the operators to go to AB.CA, and they will verify swap of panels after reading NOTE at step 55 or 63. Distracter C is incorrect because there is no direction to lower power and start AFW pps.

!L -Refe~~ce Title ---~---:-_::J ~Facility ReferenceNumber -- -l ~eference Section-:; ~CI~ No.J [~evisjOnl

' Loss of Control Air j- S? OP-AR.CA-0001 I I 1118 I I I I II I I I I II i

[!:~~-~--___; Objectives I ABCA01 E003 I l_Po'1aterial Required for Examin(ltion '1 I II

~onSource:j ~acility Exam I Bank 1 @it.estion Modification Method: II Direct From Source 1fUsed During f~ainillg Program ; D

Question So_ul'_c_e ~~rn~en~ I075662 I

-- -~

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I I

I

Given the following conditions:

- Unit 21s operating at 100% power.

- ---Th6re~1Sil0pilmary-to-seconaary-lea:Kage:-- - - - - --- - - - - --- - - -

Excess letdown Is In service due to a problem with the 2CV18, Letdown Pressure Control Valve, which Is currently shut.

- A fuel pin failure occurs, releasing a large amount of fission products into the RCS.

Of the following radiation monitors, which would show a change because of the failed fuel BEFORE the others?

jji 12R26, Reactor Coolant Filter Monitor.

~ 12R31, Letdown Heat Exchanger Monitor.

~ 12R4, Charging Pump Room Area Monitor.

rrJjAny 2R19, Steam Generator Slowdown Monitor.

~ r~ 51~ D I~MifWJ~!l!j !Application I~ !salem 1 &2 I ~~r,;41 12/15/20141

~loooo7eA104 1§1.o4 _j~M'l!D@~!!L~~(ili:J~Oi~~[JJ IIJI D l~l !High Reactor Coolant Activity I~

[@:]l!iilflii~IDTI Ability to operate and I or monitor the following as they apply to High Reactor Coolant Activity:

Failed fuel-monitoring equipment I r_.~~~~-~~~ 55.41.b(11,5)With the CV18 shut, normal letdown will be out of service, and if out of service for an extended period of time, will

~~l!~ have Excess letdown placed in service. Excess letdown does NOT pass through the 2R31 process monitor. The RC filter also will not have flow from the discharge of the mixed bed demins since normal letdown is secured. The stem states that there Is no pri to 1~~~~~~k~'g'e~"~ t~~ ~~;s sh~u},~ ~~n~~~ffected. -~h~ excess letdown ~~e flowpat~~~~e;, t~ ~~nen suction of the ch~~?~~g'~'um,~~ ~.ere 0

definition of failed fuel monitoring equipment, but rather the method of monitoring for failed fuel under other than no~al condlti~ns using installed plant equipment. *

~~-J!I!/"~"!4~~~mtfr!JtfJtillTh!L~ IJMJ,'i~&i'i;tijon~ ~!;~~

IRadiation System Monitoring II S2.0P-SO.RM-0001 I ll___j j3a I Ieves System II 205228 I Jl In I IHigh Activity In the Reactor Coolant System II SZ.OP-AB.RC-0002 I II IL:J

~9-~l:il I ABRC02E001 I I I I I

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

'O~estion Topii: LRo 21 Given the following conditions:

- Salem Unit 1 is offline.

- Salem Unit 2 operating at 95% power, 1150 Mwe, with its Power System Stabilizer (PSS) out of service.

- Unit 2 Main Generator gas pressure is 75 psig.

- Hope Creek is operating at 100% power, with its PSS out of service.

- The Hope Creek 5-6 breaker is out of service.

- A 500KV grid disturbance results in lower than normal grid voltage.

Which of the following identifies Main Generator loading which is outside the allowable for Salem Unit 2 lAW A-5-500-EEE-1686, Artificial Island Operating Guide?

Trip-A-Unit is NOT armed.

Salem Unit 2 operating at Mwe with MVAR loading out.

~ 11100,225.

Lt>~, 11100, 525.

@:] 1150, 525.

~Ans~-:.r_ Ic I l§t_rnLev!'._ IR I ~gnitive Lev.aiJ IApplication I FaciliiY:ll Salem 1 & 2 I ~xamD~~ I 12/15/20141

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i --- ~---

1 KA: 1 000077A202 system/Evolutio!l_ Titl~] IGenerator Voltage and Electric Grid Disturbances fKAstatemeBt] Abilityto determine and interpret the followinQ as they apply to Generator VoltaQe and Electric Grid Disturbances:

Voltage outside the generator capabilitycurve Explanation ()i] 55.41.b( 4) With Unit 1 0/S and the HC 5-6 breaker 0/S, the correct curve is 2S2H-5-6 on page 291. With both Units PSS 0/S, the

!An~~e_!!:__ __ ' red dashed line will be used for allowable generator excitation. A is incorrect because the PSS is 0/S. If either units PSS was IN 1

service, then it would be correct. The 2 distracters with higher MVARS are both within the limit. Since there are two different Mwe loading conditions, and the choices for each are high/low, the answer cannot be obtained by ruling out 2 of the choices because thoro '"' lrl h<>HO 1,-, ho ? ,-.,-,rror-1 <>nc,Mor<: fr.r !hom t,-, ho ,-.,-,rror-1 1\1,-,* <> rlirorl ln,-,ko on hor<>o oco co,;or<>l rlifforonl l=ino oroc <>ro niHon

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

Reference Title _

---

~[__*--*-

_ _ __j

---**-~--

~acility Re_ference Number

- [RefurE - - - -...

j Reference Section 1----

, Page No.

~-------I

• Revision IArtificial Island Operating Guide II A-5-500-EEE-1686

!I 11291 1 11 I I II II II i I I II n II I I

--* - - - - --------1

.L.O. Number - Objective~

I

• • ---~*-*

GEN002E016 I GEN002E017

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

[aue~tic:>n_TolliCj 1Ro 22 I Given the following conditions:

- A Small Break LOCA has occurred.

- The crew is performing the actions of EOP-LOCA-2, POST LOCA COOLDOWN AND DEPRESSURIZATION.

- All Sl pumps are running.

- All RCPs are running.

- RCS cooldown via Main Steam Dumps is ongoing.

- RCS Tave is 510°F and lowering at a rate of 90°F/Hr.

- PZR level indicates 26% and rising.

- The RCS depressurization has been secured and RCS pressure is 1310 psig and stable.

Which of the following describes the next major action to be implemented in the EOP to mitigate the current conditions?

I fa:~ Stop ALL RCPs due to pressure< 1350 psig and ECCS flow established.

I I

[b~ Stop the cooldown. Energize all PZR heaters to collapse voids and stabilize PZR level.

I I

~S Stop all but one RCP and begin the Sl flow reduction process by stopping ECCS pumps.

I

['I Recommence the RCS depressurization using normal spray to collapse voids and refill the PZR.

I

~s~ ~ ~-Level

~** ., '

1 IR I ~()gnitive_ Lev':'_J IComprehension I lfacili~ ISalem 1 & 2 I ~xamD~ j___12_1_15;.;../2_0_1......J41

~KAjoowEo3K102 I~K1I_~~o~~~~~alu_!_L~ 1 s~o:;si~L~~U~~r?~OJ;55!ifi, :1t.

D

~ysterri/Evoluti~m Title1 ILOCA Cool down and Depressurization

~-

I ECJ3--

~~ S!Cilement:l Knowledge of the operational implications of the following concepts as they apply to LOCA Cool down and Depressurization:

Normal, abnormal and emergency operating procedures associated with (LOCA Cooldown and Depressurization). I

[Explanation of I 55.41.B(1 O)The idea for depressurization is to refill the pressurizer. Since the pressurizer is already filled (>25%), go directly to flow

,Answers: reduction. There will be no voids if RCPs are running, and there is no CAS transition to TRIP-3.

L...--*******~~ .. ~

~----*

Reference Title --~ c-.=a~ility Reference Numb~ I""'"'"""" Section -~ 1Pa9e No~1 !Revisi~

ILoca r., and uepressurization jj2-EOP~I nr.A-? I I 1125 I I II I I !I,, I I II p I I

~L~~llm~ ___ _1 Objectives I LOCA02E001

~aterial ~equir~d for Examination II !I

~uestion Sourcs IFacility Exam Bank J ~stion -Modification Method: ~ Editorially Modified ll!Jsed Duri~g Trai~ing Program l D IO:uestion Source CommentS] 1.074666

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I LvVIIIIIICIIL

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~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

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

- Unit 2 is operating at 100% power.

- 2PR2 is leaking, and 2PR7 is shut to comply with TSAS 3.4.5. action a.

- Both SGFPs trip.

- When the Main Generator breakers opened, 2B 4KV vital bus deenergized and remains deenergized.

- Only 23 AFW pump started, and it tripped 2 minutes after the Rx was tripped.

- No AFW pumps are in service or can be started.

- Operators have transitioned out of EOP-TRIP-1.

Which of the following identifies how Bleed and Feed of the RCS will be accomplished lAW 2-EOP-FRHS-1, Response to Loss of Secondary Heat Sink?

~I Sl pump injection and bleed flow from 2PR1 only.

I I

~ Charging pump injection and bleed flow from both PORVs.

I S ISl pump injection and bleed flow from the reactor head vent valves.

I

@J ICharging pump injection and bleed flow from 2PR1 and the reactor head vent valves.

I JAnsw~~ II:J ~Ill ~--*-

level , j R II Cognitive L~~ IApplication I 1 Facility~ ISalem 1 & 2 I E_~amDa!e:_ 1 j___1_21_1_51_2_0_14_,1 jKA:!IoowEosK1o2 IE~-~~o~~J.i~J~~~~~~~~~o~[J}[R_~o~[J} ~~.~~ D

---

~--------------------------------~----------------~~~05~

jEXplanation of ' 55.41.b(1 0) The 2PR7 would be opened if it had power in this case to allow 2PR2 to be used as part of the Bleed path. With 2PR7

Answ~~~- J shut and B bus deenergized, both PORV block valves cannot be opened per step 26.1 of FRHS-1. Rx head vents are the next step. A single Charging pump will be supplying the feed portion, as step 25.1 asks if EITHER charging pump is running, and 22 will be after Sl initiation at step 24.

Reference Title LO. Number- - -

~--*------- Objectives 1FRHSOOE006 I I M~t~rial Required for Examinatio_rt__j I  :: 11

9~estioll Sourc:~ IFacility Exam Bank  !!Question Modification Method:

• -----*--------****~--

~Direct From Source I ~sed DlJI"in~l"~inin~_~l"_~ra~J D comment~

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• --******--**-****--- .... _., ___ 10122559 I

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

!au~SiiOn T~pic-11 Ro 24 I Given the following conditions:

- Unit 2 is responding to a Saturated Core Cooling condition lAW 2-EOP-FRCC-3, due to a loss of subcooling following a Reactor Trip and Safety Injection.

- RCS pressure is 1600 psig and stable.

- Containment pressure is 2 psig.

Which of the following would be an indication that ECCS flow is injecting into the RCS lAW FRCC-3?

Ia~ 1 ~1 s1 'flow 110 J

~I All Sl Accumulator pressures dropping slowly.

I I

~] RHR pump discharge flow reads 500 gpm on 21 SJ49 flow meter.

I

@.J ICharging flow reads 290 gpm on Sl systems charging flow meter.

I

~s1111Ell"j Id I [E~a~ Le~ei 1IR I l~ognitive Level ] I Comprehension l ;Facility:

L_

jsalem 1 & 2 I.Exam~JI 12/15/20141

'I KA: OOWE07K202

~

System/Evolu~_on Title

• IKA. St~te_menT:l Knowledge of the interrelations between Saturated Core Cooling and the following:

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

'Explanation ofl 55.41.b(8) Basis document for 2-EOP-FRCC-3 identifies the minimum charging flow of 100 gpm on Sl systems charging flow meter t-nsllll~!>:__ , as indicating injection to RCS. Sl pump flow of >100 gpm also indicates injection to RCS, but procedure asks if RCS pressure is

~ less than 1540, which it is not, and then skips the step to check Sl flow since it is not expected to be present above the shutoff head of the Sl pumps. RHR pumps shutoff head of 210 psid (with suction from the RWST at -30 psig) would not allow injection until

! QI""'C:: nr. *c-o ora '" m* or-h oar th 1 <=;<=;() ~~ *n rli *~* I"'* *~* /1,-., 'ol~tno d ora h<=tndi<::.Rnfl-65.0 n<:in <:n Ithey would not be able to inject until RCS preisure was below that of the accumulators so distracter B is wrong

-- ---- *---- .. --- ---- ....., ~-* ---*-*---- ---- -~ ~---**----------

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Reference Title _, LFacility Reference ~umber__. ~ference Section .. _I Page N~ ~vision; I

-~ ----* ---* --*** ---.

Response to Saturated Core Cooling 112-EOP-FRCC-3 II II 1120 I II II II Jl iI I II II II II iI I

~L.O. Number

--* ---- *-- Objecti 1FRCCOOE005 I ;Material Require~ for Examillation ' ! II

,Question Source:ll Facility Exam Bank

~-,******-----***--~

ILici~estion Modification Method: 'I Editorially Modified I~~l)u_!ing Training Prog!aiTJ_) D II rauestiOn Source Comments

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078017 changed RCS pressure in stem to ensure >1540, and changed charging flow from 315 to 290 to make question look different.

I

• • • • • ~~

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

[Que~tion ToP§] 1Ro 25 I Given the following conditions:

- Operators are performing actions in FRTS-1, Response to Imminent Pressurized Thermal Shock.

- When performing the Sl Termination Criteria step, the following conditions are present:

- RCS subcooling is 20°F.

- All RCPs are stopped.

- RVLIS Full Range is 99%.

Which of the following describes the RCP start strategy, and why?

!aJ IDo NOT start a RCP because subcooling is not adequate.

I

~~~Do NOT start a RCP because the Reactor Pressure Vessel contains voids.

I I

~C._, Start a single RCP in ANY loop regardless of SG NR level to prevent thermal creep failure of SG U-tubes.

I I

~J Start a single RCP ONLY in a loop which has SG NR level >9% to mix cold incoming ECCS water and the warm reactor coolant water.

I r~

!Ansi/Ver. d I I rEJ(am Le,;eo IR I !Cognitive Le~ IMemory I [i=acilitylj Salem 1 & 2 I ~xamDateJ I 12/15/20141

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'* I 1

System/Evolution Title IPressurized Thermal Shock

.J I lEOS I L_

~*-*--------

'KA Statement~1~1ity to determine and interpret the following as they apply to Pressurized Thermal Shock:

_ _ __ herence to appropriate procedures and operation within the limitations in the facility's license and amendments.

,Explanation of 1 55.41.b(10) FRTS Step 9 looks at subcooling of the RCS along with adequate vessel level to determine if a RCP is required, and if

!Ans_ll'/~rs:. ~ one can be started. The initial criteria are >50°F subcooling, and adequate vessel level as indicated by RVLIS. With less than 50°F subcooling, go straight to RCP start step, which requires all RCPs stopped and subcooling is >0°F. Then start the RCP lAW SO.RC-1, which has additional starting restrictions, one of which is 9% SG NR level. As per the bases document, the reason for cbrt;nr. " RrP 'onrlor thoco * * ;., t, mi ,..,..,lrl J=('('C:: fl"" .,;th u<>rm ~::>rc:: "~tor I

--=-== '

Refe~ence Title-Response to Imminent Pressurized Thermal Sh

• * ---=-J [-Facility Reference f.hunber

!I EOP-FRTS-1 j 'Ref!rence_§e-ctk.m-II Bases Doc

~ ~age _r-.Jo. ~visi~

1112-13 1

1125 i II Jl. II II I[ I II II II II Jl i

[L.o. N~mber- ~

--*-----**- _____ _j Objective IFRTSOOE002 I I I I I

~aterial Required for Examination I I ::  : II jCluestion Source: j INew  ![Question Modification Method: __ j

] I ;use~ During Training Progra~ D I

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~--*---~---

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,Question Source Comments,

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~llestioll-'!i_pi~i IRO 26 Given the following conditions:

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

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

Which choice identifies the MINIMUM RVLIS Full Range level required to be maintained during the cooldown, and its significance?

I~ 174% to ensure positive level indication of RCS.

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

8174% to prevent steam from entering the RCS hot legs.

I

'd.] 1100% to prevent steam from entering the RCS hot legs.

Ii

• j\n$;e:r:J Ic _ 1 'Exam Levell R I I :C09niti~e Le~el J IMemory I ~ility:ll Salem 1&2 IIExamDate:-11 12/15/2015 i.I<A_j!OOWE10A103 1~_12_=-_]@:>~uE!_j(B"~-~~IS~i~l~~o~e=;[J]~~G~: '

, *- -1 [J} ~ D jsystemiE~ofuti-on!~ INatural Circulation with Steam Void in Vessel with/without RVLIS I 'E1~~ _I IKA Statement:*

---

~-----~*---

Ability to operate and I or monitor the following as they apply to Natural Circulation with Steam Void in Vessel with/without RVLIS:

Desired operating results during abnormal and emergency situations.

'Explanation of I 55.41.b(1 0)74% is minimum allowed at step 10, and get into a do loop until it is satisfied. The Bases Document states that if steam

~swer::;: __ _. _; 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.

~-~--**-**

Reference Title JL .. Facility Refere~ce_ N"lunber __j §:_e_fere~ce Se_cticm 1

[_Page-No.-! lRevisio~

I II

~-*~-------* --**-------***~----*

Natural Circulation Rapid Cooldown with RVLIS 112-EOP-TRIP-6 Bases Document 1122 I 123 I I

II II H I I II Jl II I r:-*- ----** - - - * - -

c!:::O~u~ __ j Objectives

[i-RP004E004 I I I I I 1Mate_rial ReCiuiredfor E~amination J 1:  :  ::  ::  :  : II i_Ouest~on ~ource: JIFacility Exam Bank I!auestion Modification Method: Jl Direct From Source IfUSed Du~ing Training Program I D

'Question Source Comments

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I la~~~s!ie>~rOI)i£J L13Q_27 I Given the following conditions on Unit 2:

- A LBLOCA has occurred.

- Operators are performing 2-EOP-LOCA-5, Loss of Emergency Recirculation.

- Containment pressure is 15.1 psig and is rising slowly.

Which of the following describes how the Containment Spray system will be operated, and why?

The Containment Spray System is operated as directed in ...

io.l ~-FRCE*1. R"poooe to E>oe"'" Cootalomeot Pceo'"ce. "oe ceotocotioo of the e<moat ,.rety f"ootioo "'" pceoedeooe.

~-' -5 because it establishes minimum required containment spray flow and conserves RWST inventory.

• ~* ~2-EOP-FRCE-1 because actions concerning Containment Spray operation are more restrictive.

!d".-1 ILOCA-5 since FRPs are NOT implemented during the performance of LOCA-5.

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I EA2.1 ~::~ ~

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~~OOWE11A201

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SRO Group]

System/Evolution T~ ILoss of Emergency Coolant Recirculation iKJ\§t~tement: l~o determine and interpret the following as they apply to Loss of Emergency Coolant Recirculation:

conditions and selection of appropriate procedures during abnormal and emergency operations. I

,Explanation of

[Answers:

j 55.41.b(1 O)Upon entering FRCE-1, step 3.1 asks if LOCA-5 is in effect. The yes path states that CS pumps are to be operated lAW LOCA-5. The basis document states that this is because in FRCE, maximum available heat removal system operability is

• -* warranted to reduce containment pressure, whereas in LOCA-5 a less restrictive criteria permits reduced spray pump operation depending on RWST level, containment pressure, and# of CFCU's operating. The less restrictive criteria in LOCA-5 is used I rt>rirr11l:>tinn flnw In lht> R('C::. i<: nnt ::>\/::>il:ohlt> :ond it i<: """' imnnrt:ont In . RWC::T w::>lt>r if nn<:<:ihlt> h* * <:tnnninn containment spray pumps. So while the operator WILL enter FRCE-1 due to PURPLE path of containment pressure;. 15 *p~ig~ the containment spray pumps will be operated lAW LOCA-5.

I Reference Title ------: lf:=ac.ility Reference Number~ .Reference Sectio~ I Page_No.j [fevision]

ILoss of Emergency Coolant Recirculation 112-EOP-LOCA-5 I II J 125 I j Response to Excessive Containment Pressure 112-EOP-FRCE-1 I II 1122 I II II I II II I

__=-._] Objectives I LOCA05E005 I

I Materi_al Required for Examination , I lj rc:tu_~stion Source: II Facility Exam Bank l[~uestion Modificat_ion Method: J Direct From Source I fUsed During Training ~rogram 1 D

~~t0nsou~ecommen~~ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

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

[auestion Top~ 1R028

---

*-------- i Given the following conditions:

- Unit 1 is operating at 85% power steady state, MOL.

- Rod control is in AUTO.

- Control Bank Dis at 185 steps.

- Control rods begin withdrawing with no demand signal present.

- Operators place rod control in MANUAL and rod motion stops.

- A Rx trip is not generated, nor is one required by plant conditions.

- Operators determine that Control Bank D rods have withdrawn a total of 10 steps.

Which of the following identifies the effect of the rod motion?

/a. I Overpower Delta Temperature trip (OPDT) setpoint has risen.

I

[b._: I Axial Flux Difference (AFD) has become less negative.

I I

/~;J Quadrant Power Tilt Ratio (QPTR) has risen.

I II Shutdown margin has lowered.

rd.

~INE!i] Ib I [Exam _Level-, R I jCogniti~e LE~vel] IApplication I ~~cility: -~~Salem 1 & 2 1 [Examoate:-JI I

12/15/20141

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~

iSyst~mfEvolution T!t~ IControl Rod Drive System I [001 ~

li<A ~c:ttement:J d e of the operational implications of the followin concepts as the appl to the Control Rod DriveS stem:

trol rod motion on axial offset 55.41.b( 1) Salem normally operates with a negative AFD except for very late in core life. As rods move out, more power will be produced in the upper half of the core, and indicated AFD will become less negative. A is incorrect because the OP/DT setpoint is not dependent on rod position. QPTR should be unaffected because the change in power will be seen on all planes equally. Dis incorrect because SDM is not affected by rod position, since the rods are still trippable.

rl------------

il..~!'ll!~ ______ ] Objectives I RXOPERE019 l I Materia,! Required for Exami~ation jI II i~uestion ~ou~ IFacility Exam Bank I~uestion Modification Method~~ Editorially Modified I~ed During "f"rainingPr~_g_ram1 D

!Question Source Co-mmentsJ,_083980

~-*-** * - * - - *...* * * - - - * - - - - *

• I Comment

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

~estionl"_epi~~ IR029 I Given the following conditions:

- Unit 1 is operating at 100% power.

- Reactor Coolant flow measurements determine that 11 RC loop has experienced a 5%

reduction in flow from its expected 100% flow.

Which of the following identifies how this 5% flow reduction in 11 loop has affected the primary plant in relation to the previous 100% flow conditions?

Assume the 3 other loop flows remain the same.

~ IDelta T in the 11 RCS loop will be lower.

I

[§] I

~ ~ ~

pr~ssure in fJ I Steam 11 SG will be higher.

The reactor core will be operating closer to DNB.

l I

d. IDemand on the Pressurizer variable heaters at 2235 psig will be lower.

I

~wer 1 Ic  !!Exam Levell R I I ~gnitive Level -II Comprehension I ~Hity: -~~ Salem 1 & 2 lrexarnoate~l I 12/15/20141 Y:")("m0\

I l~ 002000A303

~st~m~E~ollltion Titlel IReactor Coolant System

~Statement:] Ability to monitor automatic operations of the Reactor Coolant System including:

![002 Pressure, temperatures, and flows

!Explanation of , 55.41.b(2,3)A lower single loop flow will cause total flow through the core to lower. Using Q=mc(D/T) if mass flow rate lowers, then

~}:\nsw~rs: _ _j the D/T has to go up if power remains the same, which it will due to MT gov valve reaction. This will cause the core to be operating closer to DNB.

- [R;****-- -----..., ,- ______,_

~-----*-R~terence-iitie-----~~-- Facility Reference Number- ..1 Reference Section  : I Page No.] ~~visionj II I II II I II J Jl II I II I II II I

~~~-=-_] Objectives I RCSOOOE006 I I RCSOOOE013 I

!Material Required for Examination

- -- -* -  ! I =  :  ::::: II

~stion Sour~E!] IFacility Exam Bank I[Question Modification Method:

J Direct From Source  ![!!sed-During Training Program-~ D

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tuestion sourcec~mrnents 10111914

___ .. _ , ---*********~---~ **--**

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j\.OmiiiCIIL

• . - - - J J

I J

RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I f§u!Stion TopicJ 1Ro 3o Given the following conditions:

- Unit 2 is operating at 75% power.

- 21 CCW pump is CIT for maintenance.

- 21 charging pump is in service.

Which of the following would ALWAYS require entry into S2.0P-AB.RCP-0001, Reactor Coolant Pump Abnormality?

~~~CCW Surge Tank level begins rising.

[t>J l2A 4KV vital bus locks out on Bus Differential.

~~: l2C 4KV vital bus locks out on Bus Differential.

. riJAny RCP Shaft vibr:tion indicates 4 mils on RP3. .

I j ~nswe~ Ic I ~xam Level_j IR !ICognitive Leve~ IApplication ll Salem 1 & 2 lrExamDate:~lj. 12/15/20141

~.c: """"<2";

1 KAStatemen~ Knowledge of bus power supplies to the following:

CCW pumps

!Explanation of 1 55.41.b(7,8) A is incorrect because there are reasons other than Thermal Barrier rupture that can cause CCW surge tank rise. D is

!Answers: _ _ 1 incorrect because normal shaft vibration is - 4 mils, but plausible because flange vibration >3 mils is entry condition. 2A supplies 23 charging pump, not 21. 2C bus supplies 23 CCW pump, and with 21 CCW pump CIT would cause OHA D20-23 to annunciate on low bearing water flow, which requires entry into AB.RCP.

, -.. * - * - - - - - * - - * - - - * - *-*--*-r -~*--**-* ..__..._. _ _,_,_.. - ...- . rc::. _._._._, *--- r*--- __ .., r**~-- ---

I ____ -~eference Titl~- _ _ _ __; i_Facility Refere~~~~umbea:__j "Reference ~ection_j J'age No .. ,Revision!

jReactor Coolant Pu~p Abnormality !IS2.0P-AB.RCP-0001 II liz 1 121 I l II II :  : ~ ~~ ][ f I

'I II II II J I 11..-:-o:Num-be-r-

'-** ----------------- Objectives 1RCPUMPE005 I ABRCP 1E004 jMaterial Required for Examination

. - I I=:  :

=  :  : II IQuestion Source: *11 New ] I1Used During Trai~ing Program j 0

-- ---------- , I

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llauestion Modification Method:

/Question Source Comments

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I lauestion Topic 1 IR<)31 I Given the following conditions:

- Unit 2 is operating at 100% power.

- 2CC190, RCP THERM BAR CC OUTLET V, fails shut.

Which one of the following describes the effect on RCP temperatures, if any, as a result of this failure?

ALL RCP ...

~J lower motor bearing temperatures will rise.

I

~ bearing temperatures will remain the same.

[(;' ratures will rise.

~

[d.l motor winding temperatures will rise.

i.A:ns~er I Ib I [exalll_Levelj IR 1 'Cogniti~elevel I

' Application 1~Salem 1 & 2 1 Exam Date: II 12/15/20141 I

[KA: 003000K604 I K6.04 ..

j~lue:il2.al[sRoval~eJ 3.11~e<:tion:ilsvs .IIRoGroup:ll 1li~Ro~

IReactor Coolant Pump System ISystem/Evoluti()~_!itle 003

[t<A Statement: 1 will have on the Reactor Coolant Pump S stem:

1 Explanation of 55.41 (3) The CCW line supplying the RCPs is a single line supplying both bearing cooling and Thermal Barrier cooling. Once the Answers:

1 line inside containment splits, the CCW from the Thermal Barriers has its own, separate return line, which is isolated by the 2CC190 (inside containment) and 2CC131 (outside containment.) The Thermal Barrier CCW flow acts to cool reactor coolant flowing upwards through the thermal barrier upon a loss of seal injection flow. With normal seal injection, the loss of CCW to the thermal II used During Training PrograiT;' D

[(;()mment

RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I lo~estio~TopiCj 1 RC>32 I Unit 1 is operating at 100% power when the 1CC71, Letdown HX CC Control Valve fails to 50% open and remains 50% open.

Which of the following describes the impact of this failure?

ta~ RCS temperature will rise. Perform a boration of the RCS if required to restore Tavg to program.

J

[~ RCS temperature will rise. Remove CVCS Demineralizers from service, then place Excess Letdown in service to restore demineralization

~

capability. I

~~e CVCS Letdown Demineralizers will be bypassed when letdown temp reaches 136°F. Lithium addition required to control RCS pH will be

~

~

gher than normal. I Id.

L_

The CVCS Letdown Demineralizers will be bypassed when letdown temp reaches 136°F. Lithium addition required to control RCS pH will be lower than normal.

I r*Facility: II Salem 1 & 2 I

rAnswe~, Ia  !,Exam ~e~ IR I ~(,gnitive~ Level] IApplication I iExamoaie:] I 12/15/20141 c~j 004000A230 ~~~30 *~ IRO-~~Iue: II 3.31 ~~-0 Va~u~]~ ~tion: *1~ ~~Group;! I 11 ~R~ Gro~p::l 11 ~5,j~! ~

1 s_ystem!_E~olution-Title 1

j'C_h_e_m-ica-1a_n_d_V_o_l_u_m_e_C_o_n_t_ro_I_S_y_st_e_m----------------------------:- 004~*-_ _l

~ ~tatem~n-~ ~~ility to (a) predict the impacts of the following on the Chemical and Volume Control System and (b) based on those predictions, se procedures to correct, control, or mitigate the consequences of those abnormal operation:

Reduction of boron concentration in the letdown flow; its effects on reactor operation Explanation of I 55.41.b(5) Boron affinity of resin bed is affected by temperature of coolant passed through bed. a. At lower temperatures, borate ion 1

1~nsw~rs._*~--* bonding to exchange site contains three boron atoms. b. At higher temperatures, borate ion contains only one boron atom c. Result of this characteristic is that at lower temperatures resins are more efficient at removing boron from coolant than at higher temperatures. B is incorrect but plausible if it is thought that the Excess Letdown line contains demins that would restore boron I . "hon ni<>I"Orl in coruii"O r <>nrl n <>ro. lrO "ill ho If)* "orinn in lolrl()IAin lino n()' ric inn hoi"<IIICO Lithium control would be affected during normal daily chemical additions c- -~-- ------* *--

Reference Title

• ---~

- __ Facility Refe~nce _N_umbe~~ -R~ference sec-tion - .I Page N_Q.l [Revision.

IGeneral Physics LP- Demineralizers and ion Ex II

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I !131 II I I II J II II I I JL I Jl II I

-LO. Number - - ;

Objectiv I

I___ -~-"**------'

CVCSOOE015 Material Required for Examination 1

I II I

II I

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

'aliesti~n ro!">ic]IRo 33 I IWhich of the following identifies the vital 4KV power supplies to the 11 and 12 RHR pumps, respectively? I

~~Ab1 B bus.

I

~0~ A bus; C bus.

J I

[c. 1 B bus; C bu1 J

I

,d. 1 B bus; A bus.

J

~s.;.;eill a I Exam Level IR ! ~<?ognitive Levei-n' IMemory I 1 1 [Facility: : Isalem 1 & 2 I ~xamDat:: I ! I

~~~005000K201 1~01 'ROValue:I[I§}'SROValue~~~Section:JI~~OGro~[J}.sROGroup:[J} ~3:, 0

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"sy~~emlEvoh.Jti~ Titie] IResidual Heat Removal System I roo5- I

~ siatemer~t~l Knowledge of bus power supplies to the following: I RHR pumps I Explanation of I 55.41.b(8) 11 and 12 RHR pumps are powered from "A" and "B" 4KV vital busses respectively. Other ECCS pumps, (11 and 12 Sl,

.Answers: and 11 and 12 CS) are powered from A and C. Unit 2 SW pumps are powered in reverse order, 21/22 from C, and 25/26 from A, L..............~ *****--****'

when considering plausible distracters. Charging pumps 21 and 22 are powered from Band C busses, again when considering plausible distracters.

L

--Reference Title ____ ----, ~- Facility Reference Number Jl Reference section! PageNo.J -Revisio~

INo. 1 Unit 4160V Vital Busses One Line 1 2o3oo2 II I 1134 I I I II I II I II I n I Jl J ll.CfNumber--___ -,

IRHROOOE005

[_____.~

/Material Required for Examination II 'I Question Source:

IFacility Exam Bank I Q~estlon Modi_fication Method: !I Direct From Source I:usec:J_During TrainingProgra~J 0 Questi~~Source ~~~ments/j Used on Salem 5/2010 NRC RO exam (3 exams ago,) developed from 2003 Fermi NRC Exam I

. LComment I

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

[~~**------*-,

~_sti()nTopi:_l RO 34 I J Given the following conditions:

- Unit 1 is operating at 100% power.

- 11 CFCU is C!T.

- 1C EDG is paralleled to 1C 4KV Vital Bus for monthly run.

- A 1" line connected to RCS loop 11 shears off.

- Operators initiate a Rx trip and Sl.

- 1A 4KV Vital bus locks out on Bus Differential.

Which of the following describes the difference in containment pressure response between this LOCA, and one with the same conditions except 1C EDG was initially aligned for normal standby operation, and why?

I

[a.] the same since all required pumps will be running.

I I

~ higher since only ONE Containment Spray pump will be operating.

J I

~ higher since NEITHER Containment Spray pump will be operating.

_I 1~ I lower because ECCS will inject faster due to the EDG already being running.

I

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[t<A-stateiTie;rt:l ~ty to manually operate and/or monitor in the control room:

mps

' ~ - :l Explanation of* 55 A 1.b(3, 7) A is correct since a 1" break will be within the capability of the two high pressure injection pumps (11 and 12 eves 1 pps) to prevent a major lowering of PZR pressure. The loss of 1A vital bus would affect 11 RHR, 11 CS, and 13 Charging pp, (and 1

Answers:

' - - - * * -~ .. - .. , _ _ J 11 CFCU which is C!T) none of which would be injecting for ECCS or for containment pressure control, since the small size of the leak would not cause RCS pressure to drop to their shutoff heads, or containment pressure to rise for CS requirement.

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• Revision ISalem UFSAR I II Section 15.3 I I 25 J I I I I I I

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~--~~-~~

~stiol'l Topic I Given the following conditions:

- Unit 2 experienced a LOCA while operating at 100% power.

- A Rx trip and Sl initiation were successful.

During the response to the LOCA in the EOP network, which of the following overhead alarms would be UNEXPECTED if it were to occur?

Assume containment pressure peaks at 10 psig during the event.

raJ I D-41, BIT DISCH PRESS HI.

I Jb] IC-~2, 22 CFCU ~RFLO TRBL.

~ ~~:  :~: :::  : I 1

I c] C-10, CNTMT SUMP OVERFLO I

id~l I D-48, SUBCLG CH B MARGIN LO I

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lsyst~m/Evolution fitl;: IEmergency Core Cooling System 11oo6_____j

'KA Statement**

Ability to veri~ that the alarms are consistent with the plant conditions. I IExplanation of' 55.41.b.7) OHA C-12 is expected whenever the CFCU is in slow speed, which it would be for Sl initiation. OHA D-41 would NOT be

~An~wer~:__ J expected, setpoint is 2610 psig, and charging pump discharge pressure would be much less than that, above the RCS pressure which would be lower due to the LOCA. C-10 would be expected as the containment sump would fill after the Phase A isolated containment and the leak filled up the sump. With a peak cont press of 10 psig, the LOCA, will definitely lose subcooling, and alarm ic !:>I 1 nolO' m!:>rnin In C!:>h or!Oiinn

• • -* ***-* --- *~-**~-~--~-*~, ~--**~----~ . -. ,--~. ---------~ rc::--* __ ....___ ..,

~*-**--*--** Reference Tit~ _ _ _ _:__j _Facility Reference_ Number__j iRefE!rence Section--~~ Page No-J ~evision I Overhead Annunciator Window C I j S2.0P-AR.ZZ-0003 J II IIJ 7 I IOverhead Annunciator Window D !I S2.0P-AR.ZZ-0004 I II 1126 I II II I II II I

~~~~-==:J Objectives IECCSOOEOOS I I I

~ateriCII R':quire~ forExalllination _ l j II rco-n,--

I I

I

RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List f Outline Changes I iol!e~tionr<>~ 1Ro 36 I With Unit 1 operating at 100% power. 1 PR1 opens in automatic with no demand to open, and cannot be shut.

Which of the following describes the effect of this failure, and how the actions the crew should perform lAW S1.0P-AB.PZR-0001, Pressurizer Pressure Malfunction will affect this event?

~-~.The PRT rupture disk will rupture when pressure reaches 10 psig if the 1PR6 Block Valve is not shut.

I fb_.~ The PRT rupture disk will rupture when pressure reaches 100 psig if the 1PR6 Block Valve is not shut.

I I

I

[c~ If PZR heaters cannot restore pressure, the Rx will be manually tripped before an auto trip is generated on OT/DT at 2100 psig.

J rd., l.lf PZR heaters cannot restore pressure, the Rx will be manually tripped before an auto trip is generated on low PZR pressure at 1985 psig.

I

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Sys!em1Evoluti_~~ IPressurizer Relief Tank/Quench Tank System I i~97-n r--------~

!<A Statement:_j Ability to (a) predict the impacts of the following on the Pressurizer Relief Tank/Quench Tank System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Stuck-open PORV or code safety iEXptanation otl 55.41.b(5, 7,1 0) AB.PZR directs closing the PORV block valve if the PORV cannot be shut. If it is not shut the PRT rupture disk will An_!>_wers: _j rupture at 100 psig. 10 psig is the high pressure alarm setpoint. There are steps in AB.PZR for operating PZR heaters, but for a PORV failure the heaters will be unable to maintain PZR pressure. The 2 trip setpoints are incorrect. The OT/DT trip setpoint is not a psig value, but its equivalent value is- 2,000 psig. (Actual Salem data on actual Rx trip) The low PZR pressure Rx trip is at 1865

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==~~~~-

j Pressurizer Pressure Malfunction

== [~~~~~ii~}i~~b~_! ~!~t;~c;ti~

i1 S1.0P-AB.PZR-0001 II II

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~_Nu~~er__ .....J Obj I PZRPRTE009 I ---

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I.Material Required

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for Examination

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

~~esti<m Top~ 1Ro 37 I Given the following conditions:

- Unit 2 is in MODE 3@ NOT, NOP_

- 21 and 22 CCW pumps are in service,

- 23 CCW pump is in MANUAL,

- 2C 4KV Vital Bus senses an undervoltage condition, and loads in SEC MODE II*_

Which of the following identifies the Tech Spec consequence of this event on the CCW system?

TSAS 3,7.3, Component CoolinQ System is ...

~:-11 entered due to not having 2 loops of CCW operable.

J I

[b~ NOT entered because ALL CCW pumps remain operable.

I

~<;.11 NOT entered because 2 of the 3 CCW pumps remain operable.

I I

d~ entered due to high system flow from 3 CCW pumps in service through 2 heat CCW HX's.

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' 5!431 l~yst~m/Evolution Title] IComponent Cooling Water System 1 0_()_8-- -

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~StMemen~.---------------------------------------------------------------------------------------~

Abilit to determine operabilit and/or availabilit of safet related equipment.

c - - * - ------,

,Explanation of I 55.41 .b.(7) CCW TSAS 3.7.3 requires 2 independent loops of CCW. The bases for that states that in order to have 2 operable IAnswers:

• --~---* -U**-*'

loops, ALL 3 CCW pumps must be operable along with HX's and valves, etc. When the 2C SEC senses the undervoltage condition, it will open the 2C bus infeed breakers, start the EDG, strip loads, close the EDG output breaker, the sequence on BLACKOUT loads. 23 CCW pump is a blackout load, but not an ACCIDENT load. Additionally, the SEC locks out AUTO/MAN function of the c - . - -.. - ..---1

,LO. Number L __ _ _ _ _ _ _ _ __

Objective I CCWOOOE010 I I c!-'lateriCII Required for E)(aminatiO'.':_j I II IQuestion Source: *11 New 1!Question Modification Method: il I fused During Trainin~ Pr()gramJ D

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,auestio~"J"o_pi~ IRO 38 I I With both PZR Spray Valves 2PS1 and 2PS3 in AUTO, which of the following describes the effect, if any, of 2PS3 PZR Spray Valve demand failing to 50% demand? I La*__ , No effect as the 2PS1 would close and transfer normal spray capability to 2PS3.

I I

~~j All PZR Backup heaters in auto will energize when PZR pressure lowers to 2210 psig.

I li_l, All PZR Backup heaters in auto will energize when PZR pressure lowers to 2218 psig.

I I

icC~ PZR pressure will initially lower, and the Control Group heaters will fire full time to restore pressure w/o auto B/U heaters required.

I

[~nswer~ Ib I ~Exam Levelj IR I ~Cognitive Level ; IApplication I ~aciHty:ll Salem 1 & 2 I :ExamDate:J I 12/15/20141 r~: 1 [o1ooooK3o1 J~~~01___ 1 1 R_~~~~~ ~~~"~~~~ ls_t:_cti_o~~~~ ~~~~~u~~~~u_£:JU ~111 o I l isyStem/EvolutiOnTitle-1 IPressurizer Pressure Control System  ! ,()1 0- e

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!~ State_rr~_ent:_; Knowledge of the effect that a loss or malfunction of the Pressurizer Pressure Control System will have on the following: I RCS I 1

Explanation of 1 55.41.b(5) Normal PZR spray demand is -13% on each PZR spray valve, as Salem runs with one set of B/U heaters in MANUAL 1

Answers:

!.': ____ . ~ **--- - ~*

! ON. The failure to 50% effectively doubles the actual spray flow. The 2PS1 WILL shut, but more spray than needed is now present especially since 2PS3 is the dominant spray flow. PZR B/U heaters in auto will energize at 2210 psig, they turn off at 2218 psig.

The control group heaters are for fine pressure control and do not have the capability to maintain pressure with 50% spray demand.

I L- - ~ --- -- ~Refen~nce Title-~ ~--l 1 Facility Reference Number :~eference * - - * - - - - - - - - ---*- .---*--- - - l r * - - - - 1 Section e II Pag~No.j LRevisio~

I I

---

~- *-------~--

Pressurizer Pressure Malfunction j S2.0P-AB.PZR-0001 II il j 118 I I PZR Pressure and Levei~C~trol LP __jl NOS05PZRP&L-09 Jl _jl 119 j I II II II Jl J

--~~***- ~---*---1

~~N~rnl'~ __ Objectives I

_j PZRP&LE008 lMaterial Requi~edfor Examination I I  :  :  : :  :  : n

~u~tion Source] INew I[Question Modification Method: J IJUsed During Training Program II D

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I j~uestion T~pi~  !RO 39 I Given the following conditions:

- PZR safety valve PR3 is stuck slightly open.

- Charging pumps are maintaining RCS pressure at 1910 psi g.

- PZR vapor space temperature is 630°F.

- The PRT level and pressure are 75% and 5 psig respectively.

- A NCO notes that the tail pipe temperature for PR3 indicates 310°F. He states that he believes that there is a problem with the indication since it is not reading as he expects for the current conditions.

What should the indication read?

'a.: 1162°F.

I

~~. I

~I I 0°1 I

[d.~ 1630°F.

I

~swer:J [U Exam Leve~ ~ !cognitive Level_j !APEiiCCI!ion J ffacmty:J1~IE)Ill_ 1 & 2 I [ExamDat~ I 1_2/15£2014)

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[<A~ 01 OOOOK502 lsystemiEvoiution Titlel

~------

IPressurizer Pressure Control System ll010 _j

~-Statemenl:l Knowledge of the operational implications of the following concepts as they apply to the Pressurizer Pressure Control System:

Constant enthalpy expansion through a valve iExplanation ofl 55.41.b.(5)With the PZR at 1910 and 630, the liquid is saturated (page 13 of steam tables). Since throttling is a constant enthalpy

~~wers: _ _I process, the downstream must be saturated for the pressure in the PRT. With 5 psig (20 psia), steam tables show a sat temp of 227.918 oF for 20 psia. (page 11 ). The 162 distracter is if psia is uded for PRT press vs psig. 310 is value given in stem, and 630 is what PZR is.

Reference Title -j I -!~i~eterence Number j 1Reference Sectioo~ !~age N~ [Re~ision'

! Steam Tables II II II lj _ _ j II II II ll__j II II II II _j

[!:~N~b~=-] Objectives I PZRPRTE008 j IR~: 39 Steam Tables ll\llaterial Required for Examination

:  : !I 1 auesti~n Sourc~:_j IFacility Exam Bank  ![Question Modification Method: !I Significantly Modified !lused During Training Program JD

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~estiC>_n r~pic] IR04o I Given the following conditions:

- Unit 2 is operating at 100% power.

- Console alarms SEAL WATER FLOW LO annunciate for ALL 4 RCPs.

Which of the following failures has led to these alarms?

~a=j12CV71 CHG HDR PCV has failed shut.

I

',I PZR level program setpoint has failed high.

ib*

I

c:~ 12CV115, Seal Return Relief valve has lifted and failed to reseat.

I

.~CII Charging System Master Flow Controller demand has failed to 20%.

I

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System/E~olution Title IPressurizer Level Control System I ro11 ---

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.KA Statem~11t:

• Knowledge of the of the effect of a loss or malfunction on the following will have on the Pressurizer Level Control System: I Correlation of demand signal indication on charging pump flow valve controller to the valve position I 1

Explanation of' 55.41.b(5,6,7,8) The 2CV71 PCV is located on the charging line upstream of the tap to go to seal injection. Its closure would cause

,~swers__: _ _ i full pressure/flow to go to the RCP seals. PZR level program signal failing high would cause charging flow to remain the same (program is clipped at -100% programmed level) or go up slightly, and also seal injection flow. The CV115 lifting on the return line should have no effect, or if any, it would cause seal inj flow to rise if it lowered seal return header pressure enough. The charging

'<O\/d<>m m<>d<>r flnw rnntrnll<>r rl<>m<>nrl ;., nnrm<>lll. -.100/, "" ?00/, rl<>m<>nrl ""' drl ni\/<> -1/? nnrm<>l rh<>rninn fln1M t\lnrm<>l rh<>rninn flow is -90 gpm. This matches the intent of the KA, as there is no indication of CV55 flow control valve position (which controls charging flow when a centrifugal charging pump is in service) other than open/shut/ or indeterminate. The Master Flow controller controls the PDP charging pump speed, and hence its flow, when its in service, and controls the charging FCV CV-55 when centrifugal pump in service.

~-=--=____Reference Title ==~ J L_ Facility Reference N~mber J ~erence s~c_tion ___; 1Page No.] ~evisi_on 1 IS2.0P-SO.CV~-0002 II Charging pump operation J II 1140 I I II I II II I II Jl l II II  : ~ I Ll~~m~ ___ l Objectives IPZRP&LE008 I I I

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~~IR041 Given the following conditions:

Unit 1 Is returning from .a refueling o.utage*. _ .

RCS heatup and pressurization Is In progress lAW S1.0P-IO.ZZ-0002, Cold Shutdown to Hot Standby RCS pressure is 1B50 psi g.

RCS Tave Is 51 O'F.

A 2,000 gpm RCS leak occurs In containment.

Which of the following Identifies how the Reactor Protection System will respond with NO operator action?

~ An automatic Safety Injection will occur at 4 psig in containment.

IEJ An automatic Safety Injection will occur when PZR pressure lowers to 1765 psig. I NO automatic Safety Injection will occur because the RPS System Auto Sl Block has not been Unblocked yet.

NO automatic Safety Injection will occur because the 2 running centrifugal charging pumps will respond In auto to the lowering PZR level.

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~J\lj'lj!~,ij~ j Reactor Protection System I @12=:1 ifWSiat~iiientrl r.K 7n_o_wl-:-e~d:-ge-of~l:-ow-p-ow_e_r-:-/s-:-h:-ut:-:d-own-:-im-p-::ll-ca-:t:-io-ns-i:-n-a-cc-:id 7e-n-:-t-e.-g-.,-:-lo_s_s_of:-c-o-o:-la-nt:-a-c~cl:-:de-n7t_o_rl:-o-ss-o-:f:-r-es7Jd 7 u-a-:-lh:-e-a-:-t-re_m_o-va-:1):-m-l:::tlg-a-::tl:-on-;

strate les.

55.41.b(7) (Note: The choices say Safety lnjction "will occur" vs signal generated" to preclude anyone from saying that the Lo PZR Pressure Sl signal WILL occur, Its just blocked.) During the unit return to service, the Auto Sl Block (from ANY auto Sl signal) is UNBLOCKED at step 5.2.21 of IOP-2. At that point, the unit Is preparing to enter MODE 4, (>200'-<350'F.) The Lo PZXR PRESSURE Sl remain BLOCKED until after the R?,Sis pressurized >1915 at step 5.3.23. So with 1650 pslg in stem, It >1111 still be complement of ECCS equipment were' available. This leak size will cause containment pressure to rise well in excess of 4 psig, wlhich is where the AUTO Sl will occur. DIs incorrect both because only a single centrifugal charging pump is allowed to be In Iservice, and the runout flow of 550 gpm is insufficient to keep RCS pressure from degrading, and In any case containment pressure would still rise re ardless.

~!Ei&\iiiiOi+:;J IRXPROTE012 I I

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

!Question Topiel 1RO 42 Given the following conditions:

- Unit 1 is operating at 80% power.

- A large quantity of river grass starts building up on the Circ water traveling screens and condenser waterboxes.

- A rapid power reduction is initiated lAW S1.0P-AB.LOAD-0001 RAPID LOAD REDUCTION, to maintain condenser backpressure.

- During the power reduction, the NCO places rod control in MANUAL and continues to drive rods in.

- The turbine is put on hold at 20%, with condenser backpressure at 4.8" Hg and stable.

- Reactor power and temperature continue to lower due to an excess amount of negative reactivity inserted with control rods and boration, and reactor power reaches 7% before stabilizing.

- The NCO starts to withdraw control rods in manual to restore RCS Tave which has dropped to 545°F.

As the NCO continues to withdraw control rods continuously, which of the followinq will terminate the power rise, and whv, lAW Salem FSAR?

I

,a~ Rod Block at 20% power equivalent amps on 1/2 IR Nl's to protect against DNB.

[1:>:] I High power reactor trip (low range) at 25% on 2/4 PR Nl's to protect against DNB.

rc.' IRod Block at 20% power equivalent amps on 1/2 IR Nl's to ensure hiqh RCS pressure will not result PZR Safetv valve opening.

I

[cjJ I High power reactor trip (low range) at 25% on 2/4 PR Nl's to ensure high RCS pressure will not result PZR Safety valve opening.

1 Ans~, ~ ~X<!_Ill_LeveiJI R

~--,

![Cognitive LeveiJ Memory I I [a~~~ ISalem 1 & 2 I E_x_amD_at:J I 12/15/20141 rKA:Jio12oooK4o2 !l~o2==:J~~@~~~u!j~[s~~:]~~~uB[]Is~(;~:][] ~~ D

~-yst~ITt/Evolut~on Tit'!J IReactor Protection System ~  :~ ] ~12-

~ Statement: 11 Knowledqe of Reactor Protection System design feature(s) and or interlock(s) which provide for the following:

IAutomatic reactor trip when RPS setpoints are exceeded for each RPS function; basis for each I_'Expl*a* na.ti~m of_; 15.2.2.1. Uncontrolled rod withdrawal at power. Unless terminated by manual or automatic action, the power mismatch and Answers: I resultant coolant temperature rise would eventually result in DNB. The high neutron flux, high pressurizer pressure, and I. -. - ~ -~ overtemperature DT trip channels provide adequate protection over the entire range of possible reactivity insertion rates, i.e., the minimum value of DNBR is always larger than the limit value. While the rod block signal may be generated, it will not act quickly

! <>nf'lllnh IC:.<><> C:.<>l<>m !'.nril Iensu;e DNB is avoided 7fh ,n 1,;

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No. '.~evision 1

!Salem FSAR I 1115 I II J

! II I !I I I II I II i

~~~ Objectives PROTE004

i: RO Sky~cfaper I ; SR?~lll'~c~aper -I '. RO sys!eii11Evolullon List_: I ;' S~8 ~¥~lem/EVplutl!{n}'s\:J :g?qtli_l'le Changes 'I

~~'Eft1fi1~ I RO 43 IWhich of the following identifies a 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, Iwhich maynot be met if the ESF Actuation System fails to Initiate ECCS components during a LOCA? I

~ The peak fuel element cladding temperature may not be maintained below 1800' F.

I

~ Reactor vessel water level may Initially lower below the top of the active fueL I

~ Cladding oxidation may exceed 17% of the total clad thickness at any location in the core.

I

~ The hydrogen generated from the Zlrc-water reaction may exceed 10% of the hydrogen generated if all of the zirconium surrounding the fuel reacted.

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~)M._ii!t§IJJt[~ jEngineered Safety Features Actuation System I~

((J(~~mJI I Knowledge of the effect that a loss or malfunction of the Engineered Safety Features Actuation System will have on the following:

(Fuel 55.41.b(7)ECCS is a system which ESFAS actuates. 10CFR50.46 paragraph b delineates the Acceptance Criteria for a LOCA.

Also Salem FSAR contains the same criteria copied from 10CFR46 In section 15.4.1.1. Choice A is incorrect because the fuel cladding temperature criteria is 2200'F. B Is not correct because the Accident Analysis for the blowdown phase of a LBLOCA states that the top half of core is uncovered for lengthy period

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~*~~ ~-- -- ~* - -~

~

Question Topic LR()44 I IUnder which of the following conditions would ALL outward rod motion be blocked? I ra IRods are at ARO position.

~ PRNI Channel 2N43 fails high.

I Df"'C

[Cl ?1 OT/DT is 64. rF with an OT/DT trip setpoint of 69°F.

~ PT-505 Turbine Steam line Inlet Pressure Transmitter fails to 0 psig.

I Answer J[0" 1 Exam Level II R I [<::ognitive Level I

! Application I Facility:-~~ Salem 1 & 2 I ExalllD~te: Jl 12/15/20141

~~015000K402 J~IJe]@~~~~[ROGrou~[]~oc:;rou:P:J[] i8*~. D I K4.02 System/Evolution Title]

- ...~

INuclear Instrumentation System l[o15

-~-~_t(:lt(:lrnel'lt:[ ~owledge of Nuclear Instrumentation System design feature(s) and or interlock(s) which provide for the following: I Rod motion inhibits I ixplanation of] 55.41.b(6, 7) A is incorrect because at All Rods Out position (Control Grade Interlock C-11 ), all AUTO outward rod motion is Answers: blocked. This position is set for each fuel cycle, meaning ARO is a number, not a physical stop in the core. B is correct because 1/4 PR Nl >103% is C-2 and blocks ALL outward rod movement. Cis incorrect because the control grade interlock C-3 is actuated within 3% of the OT/DT Rx trip setpoint. 64.7/69=93.8%. Dis incorrect because with steam line inlet pressure< 15%, (Permissive P-

'?\ ntthAI<>rrl <>ttln rnrl mmt.:>m.:>nt i<: hlnrk.:>rl m<>ntt<>l <:till \lnrk<:

~-- **-***--- *-

Reference Title Facility Reference Number 1 o .. ~.

Section Page f'l(): j ~visiol'l, I Licensed Operator Fluency List I NOS05FLUNCY-09 I I 11 119 I I I I II I I I I II I j FLUNCYE002 I

I I

lM_at(:lriai_Required for Examination '

I II

!Question Sourc~ INew I[Ciue~tion Modification Meth~d: -- ll I ru~ed Du~i~g Training Program D I I

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Comment~

!Question Source

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[ouestion Topicj 1RO 45 I IWhich of the following indications is the ONLY one allowed to be used as part of the steps performed to verify natural circulation is occurring if the Core Exit Thermocouple Processing sx.stem becomes de-energized lAW 2-EOP-CFST-1 Critical Safety Function Status Trees? I ra. IPlant Col lputer readings.

I

.ib.l,_subcoolil Margin Monitor , c:au11 'loP*

I ic:. ~ IInstalled Control Room Class 1E readings.

I I

~J Safety Parameter Display System readings.

I

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~ysterO!EW>IutiooTitlel lin-Core Temperature Monitor System I :o17 lt<A Statemimt-~ Knowledge of the effect that a loss or malfunction of the In-Core Temperature Monitor System will have on the following:

Natural circulation indications

[fxplancltlon ofl 55.4 1.b(7) GET's are the primary indication of RCS temperature. The CET Processing system takes the input from all GET's,

,Answers:

1~***-***--~**- ~

I coverts it to a digital signal, and sends them to various places, including the Subcooling Margin Monitor, SPDS, and the eP-250 Computer. With the system (2 trains) deenergized, there is no CET indication to send anywhere. The only remaining indications for RCS temperature are provided via the Class 1E control console indications. The CET temperature indications normally provided r.n ~Dn~ <>nrl Dl<>nt r, "'"' lht> r.nt> . '" thr.c:p svsteoos beinn IISPd as prim<>n* inr!ir<>lir.nc: <>nrl "'"' t>Vt>mn** fmm 11=

Irequirements *

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1_ __ _ _Befer~ce Title______ j i~Facilitr_Refe~nc~umb~__j !Reference Sec~ion~ _; [!age N~j ~evisionJ ICritical Safety Function Status Trees 112-EOP-CFST-1 II 112 1125 I lin Core Nuclear Instrumentation Lesson Plan !I NOS051NCORE-04 il 1128 i14 I il II  : i[  :~ ~II iC I

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'~J?~~~r:_ __ I Objec IINCOREE007 I iiNCOREE011 J

I 1

Material Required

. .. *.. for.

Examination

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~u__estion s~~rce~__l INew  : r[Question Modification Method:

~

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auestion source comments[ 1

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I Question Topic: IRO 46 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-ia. rise.

[b.- lower.

~c.

remain the same.

I,_d._j: rise or lower based on initial SW header pressure.

1 Answer II c I I Exam_~_e~ ~ [cognitive Level !Memory I [Ficiiity: J ISalem 1 & 2 1 'Examoate] 1 12/15/20141 li<A_:il 022000A104 I A1.04...

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[system/Evolution TitJel co-~-

IContainment Cooling System --------

1 022

[KA Statement: I Ability to predict and/or monitor changes in parameters associated with operating the Containment Cooling System controls including:

Cooling water flow Explanation of I 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

'Answers: ' 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 D is plausible if it is thought that SW header pressure would change, which would affect SW system flows, but in l<:l<><>rivc:t"t"'tn c:t<>"rl'*"bl<>

• ilwn*lirl nnl

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-----** ----l Facility Reference Number Referen~esection--1 Page No. * 'Revision:

I ""'"'""""' Tit!_e ...

1 1ST CFCU SW Valves II S2.0P-ST.SW-0010 I I! 1a 1120 I I I I II I I I I II I 1- ---- ----- .....

1L.O. Number Objectives I CONTMTE007 Material Required for Examination I I II Question Source: ]I New I[QuestionModiflcation Method: I I [used ou~ing Training Pro{Jra.lt j 0 Q~esti~n_Sour~e_Comments]l I

, 'C:onlment

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

~est~~-!()picJ Given the following conditions:

- Salem Unit 2 is operating at 100% power.

- An electrical fault causes 22SW20, Nuclear Header Supply valve to shut, and it cannot be opened.

Which of the followin identifies how man CFCU's will lose all coolin water availabilit , and which action would restore coolin water to all CFCU's?

l<l~: 2, open 21 SW23 and 22SW23- Nuclear Header X-over valves.

l~l 3, open 21 SW17 and 22SW17- SW Discharge Header X-over valves.

l":_!

1\!!~'lleiJ Ia I ~xaml~velll R l[~ognitive Levell! Application l[~~cility: .I Salem 1 & 2 ll~~~mDate:  ! I 12/15/20141 it<A:II 022000A204 I1\2.04-= [Ro Valuel~ SROV(lll1eL].3J[section: il~ [Rc>~O ~oup:~ e~~t~~ ~

jsystem/Ev()l~~i()l') !"itle ] IContainment Cooling System 11~~2-  :

I:KA Statement: Ability to (a) predict the impacts of the following on the Containment Cooling System and (b) based on those predictions, use procedures to correct, control, or miti ate the consequences of those abnormal operation:

rexplanation of 1 55.41.b(4,7,9) The CFCUs are supplied cooling water from the Nuclear headers, with 21 and 22 supplies from 21 nuc header, and

!Answers:  ! 24 and 25 supplied from 22 nuc header. 23 CFCU is supplied from BOTH nuc headers via a check valve arrangement, so the loss

~"~-----

of flow to 21 nuc header will not affect cooling to 23,24, or 25 CFCUs. The SW Bay x-connect valves are normally open, and even if closed would not restore SW flow, since it could not flow past the shut 22SW20. The SW23s are located downstream of the nuc I SWONUCE016 I i

I IMaterial r=-*- - ------** --

Required_!()~ _i::X(lmil1_(lti()_l1 _I I II

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[auestion So_urce Comments1j 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-To-piC_] IRO 48 I IWhich of the following contains ONLY the action(s) and/or conditions REQUIRED to electricall~ reset an AUTOMATIC Containment S[!ra~ initiation? i I

/aJ Depress BOTH Reset Phase B PBs, then depress BOTH Reset Spray Actuation PBs at ANY containment pressure.

I

[b. I Containment pressure <15 psig. Depress BOTH Reset Sl PBs, depress BOTH Reset Spray Actuation PBs.

I LC:._I Containment pressure <15 psig. Depress BOTH Reset Spray Actuation PBs.

I lCIJ I Depress BOTH Reset Spray Actuation PBs at ANY containment pressure.

I

~~nswe~ Id I ~m Level

• IR I @C)gnitive Level] IMemory I'FaciHty:&lem _1 & 2 1 li:~amoate:ll 12/15/20141 I[A~___ ~~"_ai~I:TIJ'~~~~[J] l~~~_:j~~ [ff~u_e_;i[J} ~()~£=] [J} qa~) o lKA~j 026000A405 '

~st!l;n/~~ol~tiC)n~ IContainment Spray System 1 o26 lKA statement?

Exp.Ja. natio*n** ofj' 55.41(9,7)Containment Spray actuation relays have retentive memory, which allows relays to be manually reset with an actuation

!Answers: signal still present. For this reason, B and C are incorrect because containment pressure is not required to be less than 15 psig.

-- ---- --- A is incorrect because Phase B is not required to be reset to reset Cont Spray. D is correct because BOTH trains of CS have to be reset, and electrically can be reset regardless of cont pressure.

Reference Title

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[L..~Urn_i:l!!:_ -~- J Objectives I CSPRAYE008 I Material Required !e>r Exalllinatio~__j I  : II

§:uell_t~_n S~urce: .J IFacility Exam Bank I !Question L_

Modification Method: . lj Direct From Source It!!_sed Durin!l Training ProgramJ D l:O~-=~t!()n*s~ur~:com~entsi 10147704 I

Comment

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[Qu!stiOn Topicl 1Ro 49 I Given the following conditions:

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

- A manual reactor trip and manual Sl 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?

1~, 1.11 ONLY I

~J 112 ONLY.

l

~-~ 111 AND 12.

I t

~-~ Neither Hydrogen Recombiner is available.

I

[AilswerJI b I iExam Lev~ IR I !cognitive Level 1 IApplication i ~acilityj ISalem 1 & 2 I [i:xamDateJ I 12/15/20141 ff"\

I_KA:J I~28000K201 n*-B, 1

system/EvOlutlonTitie[

I_ _ --*--*-*~--*

j Hydrogen Recombiner and Purge Control System l o2a 1

[KA §_t~~111_ent~ IKnowledge of bus power supplies to the following:

I Hydroqen recombiners

explanation of] 155.41.b(9). Hyd recomb are powered from 1A and 1B 460 volt vital buses, which are powered from their respective 4KV vital l~li~e_r

_s_:__ _ _j buses. With 1A bus locked out on diff, 1A 460 will not have power. Only 12 is available.

I 1 *-- *--- --**------*-*-*---, ~--*-*-*-*-*---- ~-***--*-**--*-~*l

________ _Beference Title______ __j '---__!acility Reference Nu~er

,*--*~,

_j ~eference~ection___j LPage !'!_()-.j r***-**--

Revisio~

11A Aux Building 460V Bus One line 11601231 II II 1116 I 11 B Aux Building 460V Bus One line 11601232 il II 1118 I I JI Jl II II I rf_:"~~~~~~-.J Objectives I CONTMTE004 i I I 1-----1

~ate rial Required for Examination

1:  :  : : ::  ::  :  : : II

!.9\J:stion Source:~ jFacility Exam Bank I[Question Moclification Method: ~ Significantly Modified !lused During Traini~ProgramJ 0 1§~iioo S~rce_~~mm-;~ts: I

~~--~

0125691 and 0147705 combined l

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[Ciuest~-()n T_opic -1 Which of the following addresses items 1 below:

1. A reason for maintaining Unit 2 Cycle 21 SFP boron concentration at 2127 ppm or greater in Mode 6 lAW Salem Unit 2 COLR

2. The preferred water source to establish or restore that boron concentration level if less than required lAW S2.0P-SO.SF-0001, Fill and Transfer of the Spent Fuel Pool.

Ensures SFP boron concentration is always > RCS and Refueling Cavity boron concentration. eves Holdup Tanks.

ron concentration is always> RCS and Refueling Cavity boron concentration. Refueling Water Storage Tank c.

'd. Ensures Keff of 0.95 or less at All Rods In, Cold Zero Power conditions with a 1% delta k I k uncertainty added. Refueling Water Stora\Je 11 Tank.

f)i.~ ~ ~~- [CJ @()~~ c_::___ . amDate: I 12/15/20141 l~:JI o33oooA2o1 1 ~2:or=~ ~~~Ci~u-~~~~ue]~ ~DI~ ~~JDJ~i3r~0 '$111 ~  :,,,_"~ ~,,,~WA

~yst_e_mlE_If()l~tion Titi~J j Spent Fuel Pool Cooling System 11033_~-=

~---- , __ ---:--1 KA Statement::

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

Inadequate SDM

!Explanation of I 55.51.b(13,4) Salem COLR, page 9 . Section 2.6 states the 3 criteria for maintaining 2127 ppm, which ensures the most restrictive Answers: of those 3 is met. One of those is the Keff<0.95 as listed in choices c and d. A and B are incorrect because while the boron concentration is MAY be higher than that in RCS, it is not the reason for the 21271imit. D is incorrect but plausible because Demineralized water is the preferred makeup source to the SFP under normal circumstances. but would NOT be used to raise I hnrnn

• hot'"'"'" it h"" nn hnrnn in it

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Reference Title

• • -.. -~--I - - - - * - - - - - * - * - - - ***-*

__ ~cility ReferElnce NuiTI_b~ -~E)fE)rence Section_

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IPage f\l_o~

~.C>I'll ICore operating Limits Report for Salem Unit 2 C II COLR Salem 2

~.... ~.. *-~ ~****

I II Jl5 I IFill and Transfer of the Spent Fuel Pool. il S2.0P-SO.SF-0001 I II 1120 I I II I I ll I

~te~ial Require~ for Exa111ination I II

~~uestion sourc!J INew I Question Modification Method: t JUsed Duri_~gTraining Progr~~-] []

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'OiJ:_siion ~~~ce co111ments-j I

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'clu~stion-Topic II RO 51 I IWhich of the followin~ would cause an automatic Main Steam line Isolation si~nal to occur with NO operator action? i

[a:-11 An automatic Safety Injection signal occurs on Steam line D/P.

I I

~J All Main Steam Dumps fail full open while operating at 20% power.

I 1':.: I NR level on a single SG rises above 67% with the Unit operating at 75% power.

I

• I A Phase B Isolation signal is generated during SSPS testing with the Unit operating at 100% power.

[d.

I Answ~ Ib I [Exam_~e;el :I R I fCogniti;e Levell IApplication I 'Facility: II Salem 1 & 2 1 1 ExamDat~ I 12/15/20141 r?"N

"'1\ll

~~~ 039000A302

~stem/Evol~tlon TitlE!] IMain and Reheat Steam System

~ ~taternllnt: ~lily to monitor automatic operations of the Main and Reheat Steam System including:

lation of the MRSS

~lanation of I Steam dumps will pass 52% total steam flow. From 0-20% power, the steamflow setpoint is 40%. It also requires Tavg <543°F or

• Answers: steam pressure <600 psig. With 50% load, Tavg would rapidly lower from where it was at 15-18% power (where we normally synch

~--*-** * * * - -

gen) to< 543. The steam dumps will turn off at 543°, but not before generating the isolation signal.

~- * - - *--******---*

Reference Title ~ 1* Facility Reference Number J~* ~- S~ction  ! ~geNo.J '"Revisio;ll Licensed Operamr Fluency List I NOSO'iFIIJNCY-09 I I 119 I I I I II I J I I II I r;-r********- ---*-*-**-,

l._~u-~____ j Objectives I MSTEAME015 I I I I I 1 Materi~l Required for Examination 1 I

~stion Source:_ .I Facility Exam Bank I[QlrestionModitic:;ation M~thod: JConcept Used  !/used During Training Progranl] D rau:~tion~o~rce C?mme~t~ 1040425 made into conditions, not just setpoints 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

!ouestionT~pic 1 IRO 52 Given the following conditions:

- A Unit 2 plant startup is in progress.

- Reactor power is stable at 18%.

- The Main Generator is rolling unloaded at 1800 rpm.

- Main Steam Dumps are controlling in AUTO in MS Pressure control.

MS Dump Pressure setpoint is raised 5 psig.

With no other operator action, several minutes later you will notice:

[a.,l is l~:..~ Reactor power is> 18%.

I I

~~~! Control rods have stepped in.

I id.

I 'I Control rods have stepped out. I

[Ansv./er i Ia I ~x~~~ Level IR I !Cognitive Level 1 IApplication II Facility:_! ISalem 1 & 2 llExamDate:_. I '

12/15/20141

' ~ '-.., ' \ ' . "

]~ Vai~O]" SRO Value)~ Sect1on: Jl~ ~oup: U SR()G~U !itfo.~ D

~- ...

[~:Jj 039oooKso8 JK5.08

~~l§_v_c>IIJ!i<'n TitleJ IMain and Reheat Steam System 039 1- . ----------------------------------;

1 KA Sta!E!n1e11!:J ..;..o_ns.;.....;;;.of;...t;;;..h;..;;e..:.fo.:..l.;..;lo'-w_in-><-.;;;.;;;.;.;.:..;.,;;..;.;c..;...;....:..-'-'-..;:.;,;;=<.;..;..;....;...;...;.._...;__.;;;.;..;..;;;...;..;;..;;c...:...;..;.;..;.;..;.:..;;..;..;.;_;;;..,c.:..:.~'------+

'Explanation of 55.41.b(5)With steam dumps open in MS pressure control auto, raising the setpoint will cause steam dumps to shut to increase 1

[!...!l.s;11\1ers: ... ~: steam header pressure to setpoint pressure This will cause a lower steam flow and higher temperature, which causes lower Rx power. Control rods are not placed in auto until >P-2, which is 15% Turbine power, which is not online yet, so rods will be in manual and no operator action stated in stem.

Reference Title Reference Number

~L~O. Number IRXOPERE021

__j I

Objectives I

I I

Material _Required for Examin_~tio~ I II Questi~n Sou!ce: II Previous 2 NRC Exams I[Question Modification Method: -~Significantly Modified IIused During Training Program D I

I Que~ti~nSourceCo~rn~~ Changed stem from lowered 5 psig to raised 5 psig which changes answer from >18% to <18%.

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!comment J

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[§~estion Top~ 1Ro 53 I Which of the following parameters will change in the direction indicated following a Main Turbine trip from 90% power, once new steady state conditions have been reached?

~ ITAC system D/P will be lower.

I

-~J ITAC system sup pi~ temp will be lower.~

\c_] I I TGA SW header pressure will be higher.

~*-~Main Condenser Hotwelllevels will be higher.

I 1

.A.ns_wer 1

Id I 1 Ex~m Le~ IR  ! 'C()gnitive Level 1 IApplication I [Facility:lj Salem 1 & 2 1fE:xamDate: .J I 12/15/20141 I ~6 _ _ 1~~~e.=J@~~~Iu~~e_~~~~ ~~ue=_DJ ~?~OJ !~ 0 Y~'*"~1

[!<A]! 045000A106

-*'T I"

LSystEmllEvolUtlon Title~ IMain Turbine Generator System 1 1 045-

[KA siatemE!~t:

• Ability to predict and/or monitor changes in parameters associated with operating the Main Turbine Generator System controls including:

Expected response of secondary plant parameters following T/G trip

!eXplanation of! Turbine aux cooling system D/P is maintained at pre-set setpoint, and automatic valve will operate to maintain it, so actual D/P will

!Answers: J not change based on heat load or flow. TGA SW header pressure is regulated by ST1, which maintains downstream pressure of 80

• • *-** --- psig, so it will modulate to maintain pressure stable. TAC system HX outlet is controlled at setpoint automatically. Hotwelllevels will rise as the turb trip initiates a Rx trip >P-9, and the BF19s and 40's will shut on FW interlock. There will be no "goes out" from the

• h,-,twollc hoot it ;11 dill ho rorobting "goes in" from tho C::to::.m rlo om~ c~tdom from tho M<>in do<>m c~tctom. <>nrl th;, <::r.:c \Mill ho forl Ifrom the AFW pumps Reference Title ITGA SW system Lesson Plan

• = II I~acility Reference Nurrlber NOS05SWTURB-04 1

!I 1

Reference Section :J ~age NoJ [l§visio~

II j lo4 I II II __jj il II I II II II II II I Objectives I MNTURBE013 I I

j_ ____,

[Materi~l Required for Examination__j I Questi~~ Sour~~ INew __j ~estion ~odifi:ation~eth~d: I I :used During Tr(lining Pro~ran]- l 0

-** - - --- - - - - - r lQue~tion S~urce Comments I 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 auestion Topic- !RO 541 1

Given the following conditions:

- Unit 1 is operating at 85% power.

- All Condensate and Heater Drain Tank pumps are in service.

- The Condensate Polisher is in service.

- 11 Condensate Pump trips.

Which of the following describes the effect that this pump trip has on the Main Feedwater system?

~ ISGFP suction pressure will lower.

~~J !The 11-13CN108s, Polisher Bypass valves will open.

§~ 11CN47, 13/14/15 Heater Strings Bypass valve will open.

'cl.~JMain Feedwater temperature entering the SGs will lower.

I

!Answer~ Ia j :~xam Lev_Ellj IR I ~gnitive Leve_l_j IApplication I ;facility: Jl Salem 1 & 2 I ~xamDate: i I 12/15/20141 f-'.,.-'"' u,o

~t KA~I 056000K103 I1K1.03

[System/Evolufion-!itleJ I Condensate System !lo56-J I.L-*----****--*--,

,KA Statement: 1 Knowiedge of the physical connections and/or cause-effect relationships between Condensate System and the following: I MFW. I 1

1 Explanation of 55.41.b(4} B is incorrect because the CN108s open on a SGFP trip, not a condensate pump trip. Cis incorrect because the 1CN47 1Answers: I auto opens at 265 psig, which won't be reached. A is correct because the loss of flow from the condensate pump will cause SGFP

~---- ----*- ....

suction pressure to lower. D is incorrect because the reduced feed flow initially would cause feed temp to rise, not lower based on O=m.i..T

,..---*--**--*-**----*-----*---* ----***-- - - - - - * * - * - * - - * - r = - - **-**----*-11Ji _ _ _ -------,

,_ _____ Reference Title**-----~ [_Facility ~eference Number __j: Reference ~ecti~, Page NCIJ ~ision; I Main Feedwater I Condensate System abnormal!! S1.0P-AB.CN-0001 II il 1120 I I Jl Jl : ~ il iI J l II II II II I Objectives

[ __ ___.

[_Material Required for Examination i I  ::  :: ===::  :: Jl

~e_s_tion s_o_urce: _]j_N_e_w_ _ _ _ _ __,l[ou_es_ti~ Modif_ic_ati_on_M_e_th_o_d:_]l! _ _ _ _ _ _ _ __.l )Used Du~in~_!rai~ing_f'rogra~ D

~Question sourcecomments 1 I

1 L - * - * - - * - * ._ ...... _

!Comment 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

[ouesfi~n r~pic" IRO 55 I Given the following conditions:

- Unit 1 is in MODE 1.

- Rx power is 8.1 %.

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

- 11 SGFP is in service supplying FW to SGs.

- ALL AFW pumps are aligned for normal standby operation.

- A spurious MSLI actuates.

Which of the following describes the effect this will have on feed to the SGs with NO operator action?

l'l* pumps and the TDAFW pump will start when SG level(s) drop(s) to the lo lo level setpoint.

I ib,' ~ MDAFW pomp* will *tart wheo 11 SGFP trip,. The TDAFW pomp "" *tart wheo SG le""l' ,h,lok 1ollowiog the R' trip.

I

'c;:-1 ALL AFW pumps will remain in standby.

-- SGFP.

Sufficient steam will be supplied through the 11-14MS18s, MS STOP BYP VALVES to supply 11 I

'Ci:!

I ALL AFW pumps will remain in standby. 11 SGFP will remain in service since at this power level it is being supplied with steam from the Heating Steam System. I Answer

Ia I [Exam Level ll R I .Cog-~itive ieyeii IApplication I ,.Facility:] ISalem 1 & 2 1 [e~amoail] I 12/15/20141

--* F ,.":'~'~'\

I

~: i 059000A403

~-

1 L059 I KA Statement: Ability to manually operate and/or monitor in the control room:

Feedwater control during power increase and decrease

'EXplanation of 55.43.b(5,4)D is incorrect because the operating SGFP(s) will be placed on Main steam supply prior to exceeding 5% power (IOP-3, IAnswers: step 5.4.10), and will lose their steam supply when the MSLI signal closes the MSIVs AND the MS18 bypass valves. A is correct

- . .- - - * * " ____ __J because the MDAFW pumps and TDAFW will start on lo lo level in SGs as the SGFP coasts down after losing its steam supply.. C is incorrect because the MS18s shut on the MSLI also. B is incorrect because the SGFP will not trip. KA is applicable since

~

nf hm ~b;n c.

• <::Hdom ;., ocorl rlo or;nn nl,nt cbrlo on

. uhon ;tc do"m co onnh.* ;.,

"nrl tho offorl nf tho MSLI as it applies to that steam supply. Additionally, this question when used in requal typically results in an -25% miss rate, on an open book exam, based on when certain actions are taken in the lOP, the status of the MS valves, etc. 2008 Annual was classified as high miss (>30% miss) 2009 Seg. 1 (25%), 2010 Annual (25%), 2011 Seg 1 (8.3%)

c----=-_--§~~----=-:_j ~~~~~~- ~nc_~_]~a~ ~~n1 I Hot Standby to Minimum load i1 S1.0P-IO.ZZ-0003 jl lj37 1132 j II II II II i[ J I II II II Jl I Objectives I I II Material Required for Examination *

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

II IFacility Exam Bank I[Question Modificationl'v'lethod: _I Direct From Source lQuestion Source: I[u~ed Durin~ "]"rainin_~t_J)rogram D I

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!Question Ti>~icj RO 56I Given the following conditions:

- Unit 2 has experienced a reactor trip from 100% power.

- 22 AFP Pressure Override Protection circuit has malfunctioned, causing the AF21's (Auxiliary Feedwater Isolation Valve) supplied from this pump to remain shut.

With NO operator action, choose the indications which would be present 2 minutes after the reactor trip.

I

~J AFW flow indication reading 0 gpm for 21 and 22 SGs.

I Lb. IAFW flow indication reading 0 gpm for 23 and 24 SGs.

I

~c.-,21 and 22 SG levels rising slower than 23 and 24 SG levels.

J

':c!* 123 and 24 SG levels rising slower than 21 and 22 SG levels.

I

'Answer' c I I 'Exam Level 1 IR 1 !cognitive LE)vel IIApplication I jlFacility:: Salem 1 & 2 I [E~am o~te] I 12115120141

~:JI 061000K602 I ~d-1 ~-y~:I[TIJ ~<?_~~~:Q:B ~c~J~~_j ~<:)~P~[JJ~()~_:J=:Ij. ~3 D

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lsystem/E:volution Titlej j Auxiliary 1 Emergency Feedwater System I jo61-rKA Stat~ment:1 Knowled_g_e of the of the effect of a loss or malfunction on the following will have on the Auxiliary I Emergency Feedwater System: I

~-****---*---

Pumps I

,Explanation of-~ 55.41.b(4)With the 22 AFP Pressure override protection controlling 21 and 22 AF21 s, c would be the correct answer because 23

'An~INEl_~_:__ ___; AFP would still be supplying AFW to 21 and 22 SGs through the AF11s, but there would be more flow going to 23 and 24 SGs since they are being supplied flow from 21 MD AFW pp plus the TD AFW pump. Distracters a and bare incorrect because TOTAL AFW flow (from MDAFW pps and TDAFW pps combined) is indicated on 2CC2. Distracter d is incorrect because 23 and 24 WR levels r;dn,., f<>ctor lhom ?1 <>nrl 'J'J ~r:: 111/R loHolc hor<>ooco nf tho rnmh;n<>Hnn nf ~An <>nrl Tn llr::\11/ nne coonniH;n,., foorl In lhnco Igenerators.

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l___ Reference Title

"*--'""---*u***--**-*----****-*--***

_ _ I_ ~~ility Refere11ce Number_j -~efere11ce Secti~*-' Pag!~ ~evisiol'l; 1

! AFW Simplified drawing 11205336-SIMP I II 111 I I AFW System LP II NOS05AFWOOO I 1128-29 1113 I I II II

=

II II i

L.O. Number~--- *-

~---***---**-*-*

1 Objectives AFWOOOE008 .I I

I iMateriCII Reqllired fo~ Exami_nation -~ I II eu~stionS_ource:_!l Facility Exam Bank I e_~estion_ ModificationMethod-:~1 Editorially Modified i~sed Dur~llg Training Program D 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~Top~ IHO 57 I Given the following conditions:

- Unit 2 is operating at 100% power.

- B 4KV Vital bus Undervoltage (UV) testing is in progress.

- A problem results in the "B" SEC loading 2B 4KV bus in the II* Mode.

With NO operator action, choose the conditions below which will be observed 3 minutes following the SEC actuation.

I. Rx power >100%

II. 22 RHR Pp running Ill. Steam dump demand signal of 100%

IV. 22 CC Pp running

v. 22 CS Pp running VI MTLO outlet temp rising VII. SGBD flows = 0 gpm VIII. 22 CCHX outlet temp rising

~~~III,III.VIII I

[b:lll IVVI VII I

[Cl,ll, V, VI, VIII

[j I"'* IV, V, VII

[Answer 11 b I [Exan1 L!~ IR I ~o_gnitiv!LeveiJ IApplication I [~aciliti:J ISalem 1 & 2 I e.a~Date: I 12/15/20141

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~~em/l§:voluUon Title~ IA. C. Electrical Distribution 1 o62 1

}SA StatemEl_n_tll Knowledge of the effect that a loss or malfunction of the A. C. Electrical Distribution will have on the following:

Major system loads I Explanation of ' 55.41.b(7,8)Reactor power will rise due the auto start of 22 AFP on SEC mode II*. 22 CC pump is powered off 2B vital and starts on

!Answer~:_____ . II*. SGBD isolation occurs on auto AFW pp start. 22SW122 does not close on blackout . 2SW26 closes, causing cooling water to MTLO cooler to lower to none.

c*.** -~- -~- ~

Reference Title

--~-

--1 I

1

-Facility Reference Number

-*-*** -~-~---~

l~eference Section

____-:] ~- *~-- *-*-~"'

• Page No .. ,Revi!~

I Safeguards Equipment Cabinet Lesson Plan II NOS05SEC000-06 i 11]2,13,17116 I 12RP4 Status Panel 11218489 I II 1126 i II II I q II I

,---* ~---- ***-*--

~:~.:!'l._lJ_mb_e..__ ... J Objectives 14KVACOE008

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

~estio~ l"opic : 1Ro sa J

'-When in Modes 1-4, which of the following describes a condition which would ALWAYS require entering TSAS 3.8.2.3, 125-Volt DC Distribution-Operating, for less than the three required 125 VDC Bus Trains being OPERABLE?

_j

~~-~~Battery current for any 125 VDC Bus is 0 amps.

I I

II>~ Placing ANY of the backup 125 VDC battery chargers in service.

[c: 11 ANY of the six 125 VDC battery chargers loses its power supply.

@:'I Discovery during operator rounds that any battery room temperature is 90°F.

I

'P.Oswe_rJ b I I[Ex(lm. Level j jR I [cognitive ~Ewel II Memory  ! ~iitY:ll Salem 1 & 2  ! Lexamoate~ I I 12/15/20141

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Syste;.;:;/Evolution Titl~ ID.C. Electrical Distribution I o-63

=-* ---* -

li<A St_aternent~~dge of the physical connections and/or cause-effect relationships between D.C. Electrical Distribution and the following:

charger and battery

'ExplanationOfl 55.41.b(8) A is incorrect because during normal operation, the Battery Charger is supplying normal system loads, not the battery, Answers:

-~"

1

• - - * , .... ___1

, and battery cutrrent is expected to be zero. B is correct as described in LCO 3.8.2.3. C is incorrect because loss of power to a B/U charger does NOT cause entry, and choice says ANY. Dis incorrect but plausible based on operators knowing normal room temperature, and not necessarily memorizing battery electrolyte max temp (110 and 120 for 2 different types of batteries.)

---

---- *---*-*-----**--- --! !Reference section-~ IP(lge No. 'Revis}o_r:JJ ll Reference Title Facility Reference Number i j125 VDC electrical One Line 11203007 I II J j3o I

I I I II II I I I I II II I Objectives IDCELECE003 j_ ____,

[auestion Source: , New I I jQuestion Modification Method: 11 I [!Jsed During Training Program L ]

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

[Ques!i_<>n TopiC' 1RO 59 I Given the following conditions:

- 2B EDG is paralleled to 2B 4KV vital bus for a normal surveillance run lAW S2.0P-ST.DG-0002, 2B Diesel Generator Surveillance Test.

- 2B EDG is operating with 2525 KW load.

Which of the following identifies the consequence, if any, if the operator attempts to place the 28-DF-GCP-1, 2B Diesel Gen Loading Switch in AUTO (ISOCR)?

The 2B EDG will. ..

~JI unafft::~.,;,.,u.

I

,-:-~~speed

-~~ , and trip uv"'"IJ""u.

I rc.1 Islow down and stall when speed is < 800 rpm.

I

[dJltrip on either reverse power or output breaker over-current.

I

[!\nswe!J d I 1 Exam _!..evel" R 1 1 I I I [Cognitive Level] Application II Facility;] ISalem 1 & 2 IJExamoate: 1 12/15/20141

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I~A3._05 1 iROValue_:_JI~J[SROV_~~on: ~~ ROGroup:[J}~oGrouiJ[J} f:*q D

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[KA:Jjo64000A305

'!system/Evolution "fiill!l lEmergency Diesel Generators I [964

--~**--***-

[.<A, Stat~ment: 1 Ability to monitor automatic operations of the Emergency Diesel Generators including:

Operation of the governor control of frequency and voltage control in parallel operation

[Explanation of I If the Generator Loading switch is in the Auto Mode in parallel operation the generator will attempt to pickup large +/- VAR loading IAnswers: because it is attempting to control grid voltage. There is no SPT to EDG control interlock. The EDG will not speed up.

L___,, ...- - . - - * * * * - -

• II Reference Title I

Facility Reference Number~ l~::~. :;,.~:section-; Page No. IRevisiOrJl

! EDG Lesson Plan I NOS051'"'""""*11 I 1168 1111 I I J Jl II I I I II II I 1

L.O. Number Objecti EDGOOOE004

[Material Requir~d for Examinatio~_J I II

't'-'01Tl'"'"ln__ ---- . - * - ---.

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

auesiiOil To!lic-j 1Ro 6o I Unit 2 is releasing 22 eves Monitor Tank through the cross-connect line to Unit 1SW, then to Unit 2 ew system.

If a high radiation condition occurs, how will the release will be terminated?

l~ 11 WL 115 Waste Discharge Hdr x-conn valve will be manually shut.

I

[b:~l 12WL 115 Waste Discharge Hdr x-conn valve will automatically shut.

I

[iJ 11WL51 Liquid Radwaste Overboard Stop Valve will automatically shut.

I

~cl_] 12WL51 Liquid Radwaste Overboard Stop Valve will automatically shut.

I

~swer: Id I 'Exam~ IR I lCognftive Levell IMemory I lt=acm§IJ ISalem 1 & 2 ilExamoat.~ll 12/15/20141

• - .. ~"'--.,~. F" '

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fKA::!o6soooA4o4 fsysl:em/Ev!JIUtion Title I..::L;.;:iq~u;.;;id...:R..;.:a;.;;d..;.;w..:;a.:.;st:.:.e...:S.!..y.:.;st:.:.e..;.;m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _--ll 1668--_j J

,KA siatem~nt] Ability to manually operate and/or monitor in the control room:

Automatic isolation

!Explanation Of] 55.41.b(11, 12) The unit initiating the release will have the flow through its own R18 radiation monitor, and it will auto close on high 1

AnsV\/l!rs: __ __j radiation. Use of the cross connect line does not put flow through the opposite units R18 rad monitor, nor will its isolation valve 1WL51 be opened or in the flowpath. The 2WL 115 is a remotely operated valve but does not have an auto close function. The 1WL 115 is a normal locked shut manual valve.

I

-~-= ~=-~e_fer~~ce Title Units 1 and 2 Radioactive Liquid Waste

• =-=' 2acmt~

Reference:

-Number j[205239-SIMP 1

I

~erence_:section

,,11-Pag~ No.[lRevisio~

11 2 I I I I II II I I I I II Jl I Objectives I WASLIQE005

! WASLIQE007 I

IMaterial Re<Juired for Examination j I II

~stionSource_:j INew

1 I:Question Modification Method: il I fused During Tiaining Program J D I.

Question Source Comments

'* ***-*-- ****--~--------~

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List J SRO System/Evolution List I Outline Changes I 1 at.~esiionTopi~ 1Ro 61 Given the following conditions:

- Unit 2 is operating at 100% power with an identified small fuel pin leak. II\ v A/

- A 5 gpm tube leak occurs on 22 SG.

Of the followinfij, which is the onli radiation monitor that will NOT show a chan~e from this tube leak? \

'~j 12R19B, 22 SG Slowdown.

\\~JJ/

v /

fj).1 12R46A- 22 Main Steam Line.

/ I c~ I NOOU<OOO"<OO Ejector.

I

~~~~~2R41 0-Piant Vent Release Rate.

I Ans~er] Ib I ~mlev~ IR I 'cognitive ~eve!J IApplication 1 fi:acmty: j Salem 1 & 2 1

I ~xainDate: J I 12/15/20141

!LA~1 __ Jf~v~eJ@L~~~-~~~n~~~~o~o~U~<?~o~Jo ~-~~ D

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'KA:l!o73000A101 1

L~ste~/Ev~iutio_rl}i~ j Process Radiation Monitoring System I ~3- =

_KA Statement;"l Ability to predict and/or monitor changes in parameters associated with operating the Process Radiation Monitoring System controls including:

Radiation levels 1Explanation o~ 55.41.b(4) The R46A-D monitors provide continuous monitoring of high-level, post-accident releases of radioactive noble gases via

'Anlillllers: ___ 1 the atmospheric steam relief and I or safety valves. R19 blowdown monitors and R15 condenser air ejector monitors will respond first, then R41D as it makes its way into plant vent.

j Radiation Monitoring Systems ~= lj S2.0P-S~.RM-0001 }I IF!=:::='::1138  !

~~R=a=dr=*at=io=n=M=o=n=im=ri=ng==Le=s=so=n=p=la=n=========;n~=N=o=s=o5=R=M=s=o=o=o-=1=6=========;l}l==============;ll19 11~1=6==~1 I_:__: --~_:_:__: __,jj __::_:_:------~il_._ _::-----~11  : i I_:----~1

~o.~ulll~e.r___~-*" Objecti I RMSOOOE007 I I

I

!Material Require~ for Examination j! II Question Source:J New L__ _ _ _ _ _ _ _ _ _ _

I 1 1 auestion Modification Method:

l ,. . 'I I .. - , "" -*

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[Question source comments,

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commerlt

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@!e!J"mfl:rm@ IRO 62 I fWhlle removing a source, RP personnel drop It on the floor 10ft. from a process monitor. If this process monitor is reading 2Rihr, what Is the I approximate dose rate 1 ft from the dropped source?

~ 20 Rfhr.

rml 40 Rfhr.

~ 200 Rfhr.

Lf] 400 Rlhr.

~ D ~~ijilf~yii;c!II:J !g£imfil1%'i::{ifiD31Application I~ !salem 1 &2 I ;!,.;~~,-;,~,....1 1 12/15/20141

~jo73000K502 1~.02 I~VIll1!dl}]~~~!m}21§:.Jeii[JJ!fflo.GrouiiiJOJ. C

~!i,'i'il1J;:\18fl'llflm";jjtll'l j Process Radiation Monitoring System I~

~ IKnowledge of the operational implications of the followingconcepts as thev apply to the Process Radiation Monitorln9. System:

I Radiation Intensity ohanQes with source distance lll!~J!~~!M'!~t; 5~.41.b(12) The formula for a point source Is Dose Rate 1 =Dose Rate 2 times the product of distance 2 squared divided by If.6;if§W{m i"""'~ distance 1 squared. DR1= 2R/h times 100/1 DR1 = 200 Rlhr.

~!!i~:,::.~~'a"riii¥!EeZ:ti :s;a~a~~~~ lfia'.il$10i11 I Equation Useful for Radiation Safety j http://energy.gov/sltes/prod!files/2 I I II I II I _jj i 1-----------'1--------'1-------'1 I m!M!§3iJJ I RADCONE006 I I _jj I

RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I Jauestiorlropis 1Ro 63 I Given the following conditions:

- Unit 2 is operating at 100% power.

- A large earthquake 5 miles from the site causes a loss of off-site power.

- The reactor trips, and a MANUAL Safety Injection is initiated.

- 28 EDG output breaker does NOT close.

With NO other operator action, which choice contains the system lineup for the Service Water System 5 minutes after the Sl?

i!: 1 12SW26-TURB AREA SW MOV STOP VLV OPEN, 22SW122-CC HX sw INLET VALVE SHUT, 23SW223-CV FANS SW OUTLET v OPEN. I

~-~ 12SW26 -TURB AREA SW MOV STOP VLV OPEN, 21SW122-CC HX SW INLET VALVE OPEN, 22SW223-CV FANS SW OUTLET V SHUT. J ic~ 12SW26 -TURB AREA SW MOV STOP VLV SHUT, 22SW122-CC HX SW INLET VALVE SHUT, 25SW223-CV FANS SW OUTLET V OPEN.,

rd.~ 12SW26 -TURB AREA sw MOV STOP VLV SHUT, 21sw122-cc HX sw INLET VALVE SHUT, 24SW223-CV FANS sw OUTLET v SHUT. I l§swer I! a j ~am Levell! R I ~_gnitive Leve1ij Application  ! iFacility: I ISalem 1 & 2 I fExamDate:J I 12/15/20141

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.I o76oooK4o1 - -

r* l I IService Water System

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~-*-*-

~s_!_em/Evo~_tion.!itle . !1676 I fi<A state"meri!~ ~wledge of Service Water System design feature(s) and or interlock(s) which provide for the following:

ditions initiating automatic closure of closed cooling water auxilia!Y building header sueply and return valves I

[Explanation of 55.41.b(7) Losing power to B 4KV, 460, and 230V bus will result in loss of power to 2SW26. 2SW26 is always open at 100%

,~swers: ***-*' power, so loss of power to it will prevent it from closing. 21 and 22SW 122s are AOVs whose control circuits will not lose power as long as 115VAC is available. They will CLOSE on a MODE 3. CFCU 223 valves will open on MODE 3. 22 and 24 CFCU's will not start in slow speed due to loss of 460 volt power. D is the only answer that has the correct combination of valves.

T;or/r.:rn 'n 'J/1

~-==-==-R~erence Title--===-] L Facility R~-ferenceNumbe_r:_~ ~ierence~ection _j I Page Noj 'Re_visionl IService Water System !I NOS05SWONUC-12 II il 1112 I ISafeguards Equipment Controller II NOS05SEC000-06 II 1113,17 11 6 I I II II II II I

~§.~~~~r_-~-:LJ Objectives I SWONUCE006 I I 1 f I t_PJ!aterial Required for Examination

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes I l§u~~sti~nTopicJ IRO 64 I Given the following conditions:

- Both Salem Units are operating at 100% power.

- The operating Station Air Compressor (SAC) trips, and none of the remaining SAC's can be started.

- Unit 2 Emergency Control Air Compressor trips immediately after starting.

Which of the following describes how Rx operation will be affected, if at all, during performance of S1/S2.0P-AB.CA-0001, Loss of Control Air?

I

§.~ Neither Rx will be required to be tripped.

1 I

~ i BOTH Rx's will be tripped based on impending BF19 closures.

I l

~_-i ONLY Unit 1 Rx will be tripped based on impending BF19 closures.

I

@., IONLY Unit 2 Rx will be tripped based on impending BF19 closures.

I IAns~erJ Ib ! !.Exam.i..e~el~ IR j 'cognitive Level J IApplication I [cicility: 11 Salem 1 & 2 I fi:xamDate~ I 12/15/20141

~~I 078000K102 I'~1Jl~ _ J L~O_'.'_a~eJ§~~o~~~~~~!'?~,[~ ~~~PJU Ls~~<li.IP~U :ff*fl* []

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I

!systemiEvolutlo~Title j !Instrument Air System I Cl7S I

1 LKA Statement:J Knowledge of the physical connections and/or cause-effect relationships between Instrument Air System and the following:

Service air

~xplanation of~ 55.41.b(7, 10) A CAS action in both units AB.CA procedure states that if BOTH CA header pressure are <80 psig, then trip the Rx.

l~n!;_~~~te~: - _i If all station air is lost, then the BF19s (Feed Reg Valves) will go shut when their air runs out. The FRY's are NOT supplied backup air from the ECAC's there is a check valve which prevents control air from going to the BF19s. Unit 1 ECAC feeds the 1B and 2B headers, Unit 2 ECAC feeds the 1A and 2A header. Page 38 of AB.CA-2 has note showing SA only supplies BF19s.

,-~-****- --*-**-*-*-*--*-*---*-**~., r* - - -**-*-~-**--**-**- ,rc *-***-*-*-*-**--*l*.. -~ --***-*. -**-**-*-

l..- **-** Facility Reference Number '~""' " " " " Section ~~~ No. 1Revisi().!!J Reference

- - -

• Title **=-**=-*=-i i 1 Loss of Control Air II S1/S2.0P-AR C:A-0001 I I 1118/20 i I I I II I

! J I II l r.-

J

• -~**--

LO. Number Objectives I

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ABCA01 E001 1Material Required for Examination II  :  :  : II fQuestion Sou~ce: J IFacility Exam Bank IL(luestion Modification Method: 1Concept Used 1 !I Used Du~ing Training Progra.:O 1 []

io~estionSoorce Comments/ IQ83875. Changed info in stem to remove CA header pressures. Changed all choices

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[§IJ_estion T~pic IRO 65 I jWhich of the following events will result in ALL containment eenetrations not su~porting ECCS functions being isolated? I I

~ Failure of a RCP #1 seal.

I

~.__I Main Steam line rupture in containment with failure of MSLI.

I rc. I I PZR PORV fails open with its Block Valve unable to be shut.

I d./ IR11A Containment Particulate Monitor fails high with the Unit at 100% power.

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'Explanation of, 55.4 1.b(9) Phase A containment isolation signal occurs at 4 psig containment pressure, and isolates non-essential containment iAI'l_swers: ___ penetrations. Phase B containment isolation signal isolates all remaining non-ECCS penetrations. A is incorrect because the seal j

failure leakage would be directed to the seal return system, and its size would not cause a Sl. The R11A failing high will cause a Containment Vent Isolation, which is part of the non-essential isolation, but not all of it The MS rupture with MSLI failure will cause Reference Title I FLUNCYE002 Objectiv ICONTMTE007 Material Required_for Examination .... 1 j e_~es~ion sourc~ New l 1~tion M-~dificatioll Method: ____ l

[!lues~on :ource Commentsj I I

[Comment I

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! During a Unit 2 plant shutdown for refueling, which of the following Is \he EARLIEST point at which only ONE NCO is required to be In \he Unit 2 control room lAW Salem Tech Specs Table 6.2-1, Minimum Shift CrewComoosition, Salem Unit 2?

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~~Mode 3.

{II Mo.de '1. _

ti IMode 5.

~~Mode6.

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-~M ~~*iii*~~~~Salem Tech Specs, Section 6.0, Admin Controls, page 6-5, Table 6.2-1. The requirements are the same regardless of

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.~iili!l;iili~--i I TECHSPE017 I I I I

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~~~~~ feed flow I

r~>~~~

J

~c.J lRCS I

--~~~JSG tube di:~t~al pressure . ..

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-~ystemlJ:volution Title~

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[KAstatement:] -----*--

I Knowledge of system purpose and/or function. J IExplanotion ol '155.41.b(7) '"'"" plan, page 11 An!wers: __ I I

l--=

IAMSAC Lesson Plan

--=---= Refe~~nce Ti!le -~---=-_J [_ Facili_!Y Refere~ce _Number_ ~eferEmce Sec~i()~

• J Page Nol [Re~isi()n

!I NOS05AMSAC0-02 II ,r11 11 2 I

!salem FSAR II !17.8 II 1125 I il II Jl II II I IL":O.Number -- ,

Objectives I l C *** --*-*-----~

AMSACOE001 MC!terial Required !or Examination II II

[Question Source: II Facility Exam Bank I :Question Modification Method:

L__________._- * *-** .*

1Editorially Modified I~e~ Durin~ Training Program j D

~ue~tionSource Comme~ts./j 041886, replaced implausible distracter with choice b above J

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[Que_~ti~n_!~picJ IRO 68 I I Which of the following events would require the transfer of spent fuel elements to the Spent Fuel Pool to be suspended during MODE 6 refueling I

,<I:_ 121 Spent Fuel Pool Cooling pump is discovered to have no oil in its bearing oil reservoir with 22 Spent Fuel Pool Cooling Pump in service.

1 I

1~:: IAn SRO over-seeing Spent Fuel Pool manipulations leaves the area under supervision of a qualified Reactor Engineer.

I I

~J Only one FHB Supply Fan and 2 FHB Exhaust Fans are running.

I

~-~I Fuel Handling Area Rad monitor 2R5 fails low.

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!SystellllEvoluti~~ -Title~ @EN ~RI-j r~Jitatemen!=J .Knowledge ...--------------------------------------------------:

of new and spent fuel movement procedures.

55.41.b(1 O)Distracter Dis incorrect because only one of the two FHB area rad monitors are required to be OPERABLE lAW TSAS 3.3.1.1, Table 3.3-6. Distracter Cis the complement of fans required to be running to have an OPERABLE FHB ventilation system. Distracter B is incorrect because the requirement for supervision of loads in the Spent Fuel Pool is a SRO OR a Qualified RE. A is correct because in S2.0P-SO.SF-0009, REFUELING OPERATIONS, P&L 3.13 specifically requires suspension of I tr<:>ncfor nf fo ool !ntn tho C::I=P "hon o!thor ?1 nr ?? C::i=P no omn hornm<>" I(")DI=Rt.P.I I= Tho In" nf -,II nil jn !hp 011000 h1 1hh >r Irenders the pump INOPERABLE. * * *

• Reference Title

[LO. f'l_umber Objective REFUELE012 Material Required for ExaminatiC>I"l_j I II ro~e:tic>_n s~~r::= ~ IFacility Exam Bank i[Questio_n Modification Method: IDirect From Source I~sed-D':I~ingTrainin_g Program __ 1 D 1 Qu:stioJ1_s_()_u~ce Comment~

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RO SkyScraper I SRO Skyscraper I RO System/Evolution List f SRO System/Evolution List I Outline Changes I lcioosti0nropic L------*---_::___j 1 1Ro 69 I

IOf the following, which is the ONLY choice which identifies when a Bezel Red Blocking Tag (RBTJ mai be used as the sole isolation point? I

[~' !When the tagged position of the component is in the fail-safe position.

I i~: !.When the component being tagged is inside the boundary of another tagging request.

I

[cJ 'When the location of the component being tagged ensures it won't be operated locally.

I

'd.l,When no other means to isolate are practical to establish a Test Boundary, as long as a hazard to personnel or equipment does not exist due L_ to energized sources. I fAns~e_.-j Id I [exam Levelj IR I ~nitive Level

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~~em/Evoluti~Titi!J fGENERI:

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Knowledge of taqqingand clearance procedures.

(Explanation of I 55.41.b(10)The specific conditions under which a bezel RBT may be used as the sole isolation point are listed in procedure A, B,

~nsw~~-__j and Care not allowed. C is incorrect because, for example, a component which would be considered inaccessible, i.e.,nn overhead enviornment, while normally inaccessible, could still be accessed.

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,_ _ ----~~terence Tit~ _____ __j l_~cilityReferenc~_~umbe.':___j ~eference_Sectio~__j ~~~N~J ~evisio_llj

[Safety Tagging Operations II OP-AA-109-115 Jl 11~ 2 Jl7 :I I J! II II II I I II II II II I l!:-~~~~!1:.- _ _j Objectives I NA0015E006 I I I I I

!Material Required for Examination II  ::  ::: ::

§_uestion Source:. 11 Facility Exam Bank I ~uestion Modification Method: lj Editorially Modified I!used During Trainit.;g Program] 0 I

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~~s~n Sour~ Com men~ 0127101, replaced poor distracter C with new rnore realistic choice. Correct answer remains the same.

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- Unit 2 has experienced an inadvertent Safety Injection, which resulted in a Reactor Trip.

- Following the trip, the Pressurizer becomes water solid, numerous failures occur, and RCS pressure rises.

Which of the following identifies the LOWEST RCS pressure at which a Tech Spec Safety Limit is exceeded, and the required time allowed for Ipressure to be reduced below that value lAW Salem Tech Specs?

1~1 12440 psig, 5 1inutes.

[13:1 1 c.: 12735 psig, 5 minutes.

-~~

~--' ~735 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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55.41.b(5, 10) Tech Spec 2.1, Safety limits, for Mode 3 (which is the mode in after a Rx trip) 2.1.2, states RCS pressure shall not exceed 2, 735 psig, and allows 5 minutes to restore. 2440 psig is top line of figure 2.1-1, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time frame is if in Modes 1 or 2.

~~~~-er--=-_:_~J Objectives rTECHSPE006 liVJaterial Required for Examination

!Que!>_tion~ourc~ IFacility Exam Bank II  ::

II Question Modification Methocl:__ 'I Significantly Modified II Used During Training Pro~l'am _j D

!I e.~'=~on Source Commen~ 1075215 modified from just the time to what pressure exceeds the safety limit, and how long to restore below.

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[§tj~stio~Top~ ' IRO 71 I I Which of the following Unit 2 situations has the LEAST amount of time to respond to the Tech Spec LCO before a power reduction or action to move to lower MODE would be REQUIRED by the associated action of the Tech Spec?

I

~: II MODI 1 75% AFD -26%.

I

,b~ !.MODE 2, a single Reactor Coolant Pump trips.

l____j I

[(0 I MODE 2, RCS Tavg lowered to 542.SOF, and has been at that temperature for 15 minutes.

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System~_Evolution Title_'

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......! EENERil LKAst~ement~r---------------------------------------------------------------------------------------~

!Expla-nation Answers:

ofl 55.41.b(10,5) A is correct because with the parameters given, the AFD is outside the "doghouse" shown in COLR Figure 2. This requires implementation of TSAS 3.2.1, Action 2.a.2, which states that power operation may continue if the indicated AFD is within

-- the limits as specified in the COLR. They are not, and the action further states that if not then reduce thermal power to <50% within the next 30 minutes. B is incorrect because the action time for 3.4.1.1 not having all 4 RCPs running in MODE 2 is one hour. C is c----==-~~~==----=--~~~~~-~~~-~~~~~ ~:

j Salem Tech Specs II II !13/4 2-1 II I ISalem COLR jICOLR Salem 2 II 1112 J!s I I II II Jl ll I

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,L.O. Number I L_ ----*

EXCOREE012

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*---- Objectives IMaterial Required for Examination_j IRO 71 Core Operating Limits Report Unit 2 rev 5 II

!Question Source: IFacility Exam Bank j louestion Modification Method:---- JEditorially Modified I[sed Durinjl_Training Pro9ra.Tl=! D I

~stio~Source~Cornrnentsj 01_18999 replaced RCS pressure distracter (~in~e ~sed on previ?us question o_n this ~xa~) with RCS Tavg < 543,

- wh1ch does not meet the m1n temp for cnt1cahty hm1t of <541, wh1ch would requ1re acllon 1n 15 mmutes I

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1 All of the following are items found on a "Radiation Worker Pocket RWP Data Sheet" as shown in RP-AA-4000, Personnel Conduct in Radiolo,..~.-,,, 1 Controlled Areas EXCEPT:

I~ I lcJ

~-] Electronic Dosimeter Dose Rate Alarm setpoint.

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Ability to comply with radiation work permit requirements during normal or abnormal conditions. I rE~planat~~n of I 55.41.b(12) All of the choices are found on RP-AA-4000 except c. Meeting the KIA for this question requires the ability to do Answers.-*--~--***-___________J

, something, so without giving a RWP and asking what you would do, it requires knowledge of how the RWP requirements are checked prior to entry into the RCS, that is, in order to fill out the pocket RWP dtat sheet, you have to know what is on it, then get info from RWP to ensure you comply with its requirements.

~===--=--~~~!itl~=--=:~~~il~~~~~- ~~-.n.-=C!i~ ~@

I Personnel Conduct in Radiological Controlled Ar II RP-AA-4000 I 111 0 114 I 1

1 II. I II II I II H J II II I

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............_ ___ . -Number Objecti 1RADCONE005 I I I I

[Materiai_Requiredfor Examin~t~ I II Question Source:

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~ Shift Radiation Protection Technician (SRPT) AND Radiation Protection Manager Plant Manager./ Shift Manager. __

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Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment en! re uirements, fuel handlln res onsibilitles, access to locked hi h-radiatlon areas, ali nin filters, etc.

55.41.b(12) Attachment 1, High Radiation Area Key Approval Authority, states that BOTH the RPM and Plant Manager/Shift Manager must approve Issuing the key.

r" :::t:~~n§!t!)tfe'!':

IHigh Radiation Area Key Controls

[~~]

IRADCONE002 I I I I I

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

'a~estionTopic 11 RO 74 Given the following:

- Unit 2 has tripped and experienced a loss of all AFW

- The crew has just transitioned from step 20.1 of EOP-TRIP-1 to EOP-FRHS-1, Response to Loss of Secondary Heat Sink.

- The STA has just completed his first pass through the CFST's and reports the following Status Tree conditions:

Shutdown Margin - GREEN Core Cooling - GREEN Heat Sink- RED Thermal Shock- GREEN Containment Environment- GREEN Coolant Inventory- GREEN Assuming conditions do not change, what is the required monitoring frequency for the CFST's lAW OP-AA-101-111-1003, Use of Procedures?

CFST monitoring ...

a. *I L__j be continuous.

~I is required every 30 minutes.

I 81 is required every 10-20 minutes.

d. I may be suspended with CRS concurrence until FHRS-1 is completed.

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1I[SR0Group lt<A statemel"li:l r---------------------------------------------___,.

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Explanation of :

!Answers: I An example would be a condition which results in no Sl but Heat Sink RED path occurs shortly after Reference Title

• L£a~~ity R~_erence Number ~-] ~rere-~~~ [PagE!N~ ~~

IUSE OF PROCEDURES II OP-AA-101-111-1 003

[L.O. Number -~

Objectives PROCEDE003

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

[9uestionTopic 1Ro 75 I A problem has occurred at Salem Unit 1 which results in the declaration of an ALERT.

Which of the following actions is required to be performed by the Secondary Communicator at Salem?

Assume the ALERT is the first emergency classification made.

[a.JI.Complete the Operational Status Board (OSB) Form, and update it every 60 minutes.

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[!).j !.Complete the Major Equipment and Electrical Status (MEES) Form and update it every 15 minutes.

I Ell The Emergency Response Data System (EROS) must be activated within 15 minutes of the Alert declaration.

I

~I The Emergency Response Data System (EROS) must be activated within 60 minutes of the Alert declaration.

I

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~~~~~~tion Titlel j_GENE~

~Stareme~t=~--~~~----------~------------------------------------------------------------------~

Knowledge of the emergency plan.

jExplanation Ofl 55.41.b(10) Cis incorrect and Dis correct because EROS must be activated within 60 minutes of ALERT declaration. B is incorrect

[!\nswers:. ~- because it is updated after significant plant change or classification change. A is incorrect because the OSB is updated every 15 J

minutes if requested by the TSC.

c--=-=~~---§~~_i!i~--=--==L.=~~~c~~~'J~Tr~~~ll__j~~ :~~

ISecondary Communicator Log  ![EP-SA-111-FS !I Attachment 8 II 2 112 l I ll. II II II I

'I II II II II I L!::_.~umbe."._.~_J Objective I GENISSE013 I I

!!'ate~ial Required for Examination JI II

Question Source:

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. IFacility Exam Bank Ijauestion Modification Method: il Direct From Source llUsed During Trai~ing *Program JD 1

auestionsourcecomment~ 1a6oo33

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U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: Region: I Date: 1211512014 Facility: Salem 1 & 2 License Level: SRO Reactor Type: W Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

~pplicantCertification ________

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

Applicant's Signature Results ROISRO-Only/Total Examination Values I I Points Applicant's Score I I Points Applicant's Grade I I Percent

Senior Reactor Operator Answer Sheet Circle the correct answer. If an answer is changed write it in the blank. NAME:

1. a b c d

2. a b c d

3. a b c d

4. a b c d

5. a b c d

6. a b c d

7. a b c d

8. a b c d

9. a b c d

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

12. a b c d

13. a b c d

14. a b c d

15. a b c d

16. a b c d

17. a b c d

18. a b c d

19. a b c d

20. a b c d 21 . a b c d

22. a b c d

23. a b c d

24. a b c d

25. a b c d Page 1

RO SkyScraper f SRO Skyscraper f RO System/Evolution List I SRO System/Evolution List I Outline Changes I IC}~estio~-Topic_ 1 1sRo 1 I Given the following conditions:

- Unit 1 is operating at 85% power.

- Control Bank D rods are in auto at 180 steps withdrawn.

- AFD is -1.0 with a Target AFD of -1.5.

- A power reduction from 100% to 85% was completed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ago when 21 Condensate Pump tripped.

Rods begin stepping out at 8 spm.

Tavg is below program and rising slowly.

PZR spray demand has lowered and is now rising slowly.

Of the following, which correctly identifies what is causing these indications and how should the crew respond?

• a]

IA small steam leak has developed. Enter S1.0P-AB.STM-0001 Excessive Steam Flow, place rods in manual to limit the power rise, initiate actions to locate the leak. I

~-~A Xenon transient is in occurring. Leave rods in auto to control the oscillation, while manually diluting to dampen oscillation lAW S1.0P-IO.ZZ-~

• 0004, Attachment 1, Dampening Xenon Oscillations.

I c~~ The 1CV185, Makeup to Charging Pump Suction Valve has opened. Enter S1.0P-AB.ROD-0003, Continuous Rod Motion, place control rods

- Iin manual. and terminate the boration.

I

~ 1N44 has failed low. Enter S1.0P-AB.NIS-0001, Nuclear Instrumentation Malfunction, place control rods in manual, and remove the channel from service. I I 1 Cogni~ve Lev!!_j IApplication I Facilitrj ISalem 1 & 2 I _E:xamDate~ I___12...;./.:....15:.:../2:...;0...;.1-.J41

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LsisieiTI/Evojution Title I IContinuous Rod Withdrawal 1 1 oo1

,-g.A.~_ia!erl,ejrt-11 Ability to determine and interpret the following as they apply to Continuous Rod Withdrawal*

!Uncontrolled rod withdrawal, from available indications

[Explanation of ..l55.43.b.(5). The indications given in the stem reflect outward rod movement caused by an unwanted boration of the RCS. The

'Answers: 1

! boration would initially cause a lowering of RCS temp/pressure, which would be corrected as rods withdrew, hence the rising

• * --- -- -* i demand on PZR spray as pressure rises and Tavg being below program and rising. A steam leak would also cause these I indications, but control rods would not be placed in manual. A Xenon transient large enough to cause outward rod movement would

, he indicated by a large change in AFO from noanaL..a" "h"'"n r.n lnP_LI. 1 111/ith rr.nditir.n" in dom r.f nnrm<>l 11.1=0 tho xenon oscillation is not occurring, though the action is correct if it were. The CV185 opening would cause a boration of the RCS to II occur. The PRNI failing low would cause a short duration rod withdrawal signal (during the time it is failing), but the OverPower Rod

, Block at 109% on 1/4 PRNI>109% power would prevent rod withdrawal, and the indications in stem also don't support something in which only outward rod motion has occurred.

LO. Number BROD3E002

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liH&~~ I SRO 2 Given the following conditions:

- Unit 2 was operating at 100% power when a single Control Bank 8 Group 1 control rod

• - -llrowea-into-tne-core.and-ortA-c.-48;-RoD-BOTToM<!11llunclatecd-.- - * - - - - - - - - - - - - - - - - - - - -

. The reactor did not trip, and operators responded lAW S2.0P-AB.ROD-0002, Dropped Rod .

- Rx Engineering has determined the dropped rod can be recovered,

- The condition which caused the dropped rod has been identlfled and corrected.

While preparing to recover the rod, the RO reports that both the iRPiindication and Plant Computer indication for the dropped rod show 15 steps withdrawn.

Which of the following Identifies how the CRS should proceed and the bases for that decision?

~I Remain in S2.0P*AB.ROD*0002, Dropped Rod, because the rod meets the definition of fully inserted.

I

~I Remain In 8_2.0f'0AB.80Q,OQ02, DropRed Rod, beg§Use_ the rod rec;overy SteJlS specifically addre,;s whethertheaff_8{Jtedro~ is partiaJiy or_

fully dropped. I I

~ Enter S2.0P-AB.ROD-0001, Immovable/Misaligned Control Rod, because the affected control rod Is known to be untrippable since it Indicates not fully inserted.

I tlrJ IEnter S2.0P-AB.ROD-0001, Immovable/Misaligned Control Rod, because with the affected control rod not fully inserted, the steps for Group I Step Counter manipulations for maintaining proper rod group stepping is significantly different for a partially Inserted rod vs a fully inserted rod.

~mmJ [D' ~~ Is I @!!jllt!t.fiff;'ljvlil~~ll Application I mm ISalem 1 & 2 1n:,w.IWf,;*ii>c, 1 12/15/20141

~joooo03A201 I\IQ~il)tre}!i§~g~]!JJ(J::£HI!lmr::l~t!m~t<>llPM[JJ~[]'- ill~

,.,-~,."~i ,.~

IIAA2.01

[B\iflg~t\tl'lfj.r] jDropped Control Rod I~

~~ Ability to determine and Interpret the following as they apply to Dropped Control Rod:

Rod position indication to actual rodposition

~~ 55.43.b(5) UNIT TWO, the control rods can be considered fully inserted If they Indicate 10 steps withdrawn lAW the evaluation results published In Nuclear Fuels Engineering Letter,NFS99-098, Aprll13, 1999, as stated in OP-AA-101-111-1003. At step 3.23, AB.ROD-2 asks if the affected rod Is fully Inserted. As per above, neither rod position indication available Indicates 10 steps or

,:~;!* ~~~e procedure then d~~~cts ~0 ~?h A.~~OD-1. The, ~~~ses ~o~~~,'~n~;~~~~~;hat .... "th~ ~~~~c.edure for rod realignment differs 1

Igroup stepping logic Is significantly different. The Rod Bottom' E-48 alarni comes In at 20 steps withdrawn with Group"Demand >35 steps.

~~~~:2m~tTim1!fit'ifl~' -~~£fll!rn~l~~"hfl\i~ii~MJ~~nili~4di!~~ ~

J Dropped Rod II S2.0P-AB.ROD-0002 I II j1o I Jlmmovable/Misallgned Control Rod IIS2.0P-AB.ROD-0001 I IJ Js I JUse of Procedures  : II OP-AA-101-111-1003 I II iE=J fwro.il~llli\e'~~;;tii~ ;Object(v~~J ABROD2E002 lj

~~~-*-**Jl.,

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I I

I I

RO SkyScraper I SRO Skyscraper I RO System/Evolution List j SRO System/Evolution List I Outline Changes I 3Jw~~Stion !opic 1sRo 3 1 I

Given the following conditions:

- Unit 2 is performing a normal shutdown to enter a refueling outage.

- Power was reduced from 80% to 20% and the Rx was tripped at 20% power as planned to enter the outage.

All of the following will be directly informed over the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that Salem Unit 2 is now off-line EXCEPT:

~~--~Nuclear Regulatory Commission (NRC).

I

,~I PSEG Energy Resources & Trade (ER&T).

~-~Electric System Operations Center (ESOC).

:~ : I I

'!

~ 'North American Electric Reliability Corporation. (NERC).

I

/Answer~ Id I [Exam L1well j s I L(;ognitive Level 1 IMemory i Facility: ! ISalem 1 & 2 1

1 'Examoaie:', I 12/15/20141 i',W'

~~! 000007G430

' c::-1

_systemiEvol~tionTitle]

~KAS~teme~r---------------------------------------------------------------------------------------~

Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission s stem o erator.

'Expl~anation of 55.43.b(1) Following the planned Rx trip to enter the outage, the NRC will be notified via the morning phone call if not sooner.

1

~nswers: __ PSEG ERT will be informed, as PSEG Nuclear informs them of plant status 2 times per day. OP-AA-108-107-1001, step 1

• 2.3.2.2. ESOC will be informed OP-AA-108-1 07-1001 Step 2.2.1. NERC is not informed, but is plausible based on Salem ECG, RAL 11.11, NERC Reporting. Question is not considered LOD 1 based on Annual Integrity Training which company employees are Reference Title

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

13 I I

I

__ j IELO 11.b l -

I I I

'Material Required for Examination I I JI a~estion ~~rc:_:_*l New if§_u_e~~onllll_odi_!i:a~ ~h~ J, __________,!lus~~ oll_rin_g Training ~~()9r~f!l ~ 0

~~stionSource c~~mentsj I I

'Comment

===================]==========~

I I

I

RO SkyScraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I a~estion Topic L______. - -* -- J 1sRo 4 I Which of the following sets of conditions would result in the highest (most severe) ECG classification?

Use onl~ the conditions current!~ present in each choice. I

~---

leak in the SFP liner which results in lowering SFP level with all available SFP makeup in service.

a.

IL' I

L~  : f:o1il,:; to trip when demanded, does not trip from the control room, and is tripped locally by opening the RTBs.

I r--~

!C. I activates the Hope Creek Seismic Switch and is felt in the Control Room, and causes a trip of 2 of the 3 operating SW pumps. I I

• d.J The Main Turbine has oversped during a functional test causing turbine blade disintegration, with reports of visible impact damage to the containment outside wall. I tnswer Ib.... I !Exam Levell! s I lCognit!ve Level *I Memory 1 Facilityll salem 1 &2 I iExamDate:JI 12/15/20141

"~

~

112~ _j~(lluE!J[3]"~v~Gl)rs~~Oil_=ll~i~u~[J]:s~~roliE=J

~-

,I{A: loooo29G441

\55.43 FA~,,

f~stelll/E~Iution~Title . IAnticipated Transient Without Scram I 'o2~

KA Statement:

'Explanation of 1 55.43.b(5) Lowering SFP level with all available makeup will result in an ALERT under RA2.2, which can only be escalated by Answers: means of a different RAL, probably radiation levels rising. RA2.2 says "lowering SFP level that will result in irradiated fuel becoming uncovered." An ATWT in which all attempts to trip the Rx from the control room fail is a SAE (SS3.1 ). An ATWT in which the Rx can be tripped from the control room is an ALERT. (SA3.1) An earthquake as described above is an ALERT (HA 1.1) The turbine

---

+~~lli*~~-~ . . . .

Ideclared: even though at the time it was only aUE. Salem operators should be familiar with.this.

Reference Title Reference Section LO. Number Objectives I FRSMOOE009 I r=--- I I I Material Required for Examination_j I JJ

Q~estion source:

--I New Irciuestion Modificatio~Method: J J Used During Training Program D l~u_~;ti~~ s~~ice Com~ent~' I I

Comment

"-- -*-~** --*

I I II I II. I

RO SkyScraper I SRO Skyscraper I RO System/Evolution List I SRO System/Evolution List I Outline Changes j 1 q_~~i()_~,:()llic' lsRo 5 I Given the following conditions:

- Unit 2 is in Mode 6, with refueling actively in progress.

- Emergency core position P-10 is full.

- The fuel transfer cart is at the Spent Fuel Pool.

- The manipulator crane has an irradiated fuel assembly in the mast, and is near the lower internals stand, moving toward the core.

- Gas bubbles are observed in the vicinity of the fuel assembly last placed in the core, which had experienced difficulties getting placed into the vessel.

- The refueling SRO orders all non-essential personnel to evacuate containment.

Which of the following describes how the CRS should respond lAW S2.0P-AB.FUEL-0001, Fuel Handling Incident?

Place the fuel assembly in the mast into ...

/_a. Ithe first available position, send the Fuel Transfer Cart to containment, shut the fuel transfer canal gate valve.

I

~l>:. lthe first available position, shut the fuel transfer canal gate valve, start at least one Iodine Removal Unit.

I

'c.' lits designated core position, shut the fuel transfer canal gate valve, start at least one Iodine Removal Unit.

J I

d~l its designated core position, send the Fuel Transfer Cart to containment, shut the fuel transfer canal gate valve. I I

1!-nsw~!, fCJ :Ex_a_m Leyell ~ [Cognitilfe Level i IMemory I [~<:il~ !Salem 1 & 2 1 ExamDate:

1

• -*~**-*--****-

I 12/15/2014J 036 e of abnormal condition rocedures.

55.43.b(7) With a fuel assembly in the mast tube, it is placed into its position in the core. WHEN AVAILABLE, it could be placed in

, emergency position P-10, but the stem states that position P-10 is not available.(because it may have already been filled during

- reload) Fuel loading considerations preclude putting it willy-nilly into any available position. The Fuel transfer cart must be sent to containment before the gate valve is shut. The IRU is not started unless specifically requested by Rad Pro when iodine is present in implausible, since "Fuel Handling Incident" is the .title of the procedure, and a Fuel Handiing Incident is what has occurred, with very Ilittle leewa to think it would be an hin else.

~================~~============~F=======~~~'~5==~

F===============~F===========~F========;~==~~F=~

1------------------------~------------------~1-----------~1----~1----~

L.O. Number Objecti I~---** --

ABFUEL01 E002

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

'allestion fo~ 1sRo 6 I Given the following conditions:

J ~ 0"'-'ll

)J..

- Unit 1 is operating at 100% power.

- A breaker fault occurs on the 2-6 500 KV breaker.

a,ver

-r ~. v;.rr

- The 2-6 500 KV breaker does NOT trip, but should have.

- 15 seconds after the breaker failure, Unit 1 has NOT tripped.

(AO Which of the following identifies how the Unit 1 CRS should proceed? ~

(1<7

"'"'""~'

a.

• Direct the RO to manually trip the reactor and go to EOP-TRIP-1, Reactor Trip or Safety Injection. Concurrently with EOP implementation, initiate S1.0P-AB.LOOP-0001, Loss of Offsite Power, and perform Attachment 2, Loss of Group Buses, Part A, Loss of 1E and 1H 4KV Group Buses.

- - I b.i

___j Direct the RO to manually trip the reactor and go to EOP-TRIP-1. Concurrently with EOP implementation, initiate S1.0P-AB.LOOP-0001, and perform Attachment 2, Loss of Group Buses, Part B, Loss of 1 F and 1G 4KV Group Buses.

r:::*l C.*

Enter S1.0P-AB.LOOP-0003, Partial Loss of Off-Site Power, then enter S1.0P-AB.CW-0001 Circulating Water System Malfunction, and perform a power reduction to 83% power or less.

id.

~

Enter S1.0P-AB.LOOP-0003, then enter S1.0P-AB.CW-0001 and open the Hood Spray Bypass valves 11-13MC62s.

rc:=::r ..

I

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Factllty.1Salem 1 &2 I-,Exall1Da~1---1_21_1_51_2_0_14_,1

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r::;-::;T r - : : ; - : ; 1 _"'. 5 * "4"3--.

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~¥stem/E~olotionTitle I_L_os_s_oo:...f_O_ff_-_S....;ite,.;__P_o_w_e_r-------------------------------l!lo5_EJ -l KA Statement:

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

!Explanation Ofl The breaker failure of 2-6 will cause the 5-6, 2-8, and 2-10 500 KV breakers to open, causing a loss of 500 KV bus section 2. The Answers: . failure will not cause a loss of RCPs since the Group Buses will be powered from generator output through the Aux Power

---

-- - - Transformer vs station power transformer, so there will not be a demand for a rx trip. There will be a loss of 3 circulators when 13KV ring bus south Section A loses power, but there will be at least one circulator running on each waterbox so the power rQrlllrt;nn ;., nnt '"'"' ;,.,rl tn h"' '" Th"' hnnrl cnr"" hHnO>cc HO>IHoc O>ro nnonorl nn ,;Ho....torl uh;.-h .,,-,,[rl ho .,11 nf Ithem AB-LOOP-1. is for a tot~lloss of offsite power and ,;..ill-riot be entered I

- - - -- --- ------ --_, - - -- ----- -------,.-- ----rl__]-- ---- -

Reference Title

- --- i ~cility~eferenc! Number _j Reference Se_ction ~Page No. J Revis~_n IPartial Loss of Offsite Power ll S1.0P-AB.LOOP-0003 I II i51 I ILoss of Offsite Power II S1.0P-AB.LOOP-0001 j II 1129 I IS1.0P-AB.CW-0001 ll Circulating Water System Malfunc!J II 1137 I L.O. Number I ABLOP3E002 I

I Material Re9_lli_red for Examination J I. I Question Source:

INew I 0uestion Modification *** ....

1 il II Used DuringTraining Program D

--*~*******-**-~~- ' ***-*-*-**-**-*-*

I Question source comrnenlS, 1

---~---. ------- --------'

I Comment

~"'"

I I

I

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

~u_;sti~ro~>lc_~ 1sRo 7 I Unit 1 is releasing 11 Gas Decay Tank (GOT) lAW S1.0P-SO.WG-0008, Discharge of 11 Gas Decay Tank to Plant Vent.

Which of the following identifies a condition which would require termination of the release, and why? I I

I~J The Auxiliary Building pressure turns positive. This could result in an unmonitored release of radiation from the Aux Building atmosphere.

I

~: 1 IA transfer of waste gas from 13 GOT to 14 GOT is performed. This would result in the actual release rate exceeding the 32 scfm maximum allowed . I I

.c~ A valve lineup to place 12 GOT in holdup is performed. This would result in an unmonitored release of radiation from the tank being put in holdup into the normal tank release path.

I I

~d} The pressure downstream of 1WG38, Gas Decay Tank Vent Pressure Control Valve, is 7.5 psig once the gas release has stabilized following

_, the initial start. This could result in the actual release rate exceeding the 32 scfm maximum allowed.

I iAils~e~ Ia I 'Exam lev~ Is j LCognitive Level] IMemory j 1 Facillty: I ISalem 1 & 2 I 'ExamDate~ *I

........ I 12/15/20141

- I ~ -- I

""i

*

>mt I '

~temi.fy~tion 'fit!~ IAccidental Gaseous Radwaste Release '060

!!<A S~!e~nJ:J r------------------------------------------------------:

Ability to explain and appl all system limits and precautions.

iExplanation of 55.43.b.(4) A is correct per P&L 3.9, and TSAS 3.7.7 action e, which requires negative aux building pressure or suspend all

~ll~WeEs: __

1 operations involving radioactive gaseous release via the aux building immediately. The procedure further identifies that "radioactive gas releases" is defined as Waste Gas Decay Tank releases. B is incorrect because the reason is wrong. Transfer of Waste Gas i-------~!-1*~neaJ:IlJne.r.:it:wi::J..tP,:r.Er~a.s~~~J.J[,::su!t in an unapproved _rele.ase, as the transfer path puts ga~ in the release header: The WG8 would react to isolates the tank from in.put and output, and would not put any of its contents into the release path. D is incom~-ct because the

• I downstream ressure <8.0 psi maintains release rate < max allowed of 32 scfm.

• -- l ll.(). _!!llm~er_ _ _ Objectives I WASGASE011

~Material Required for Examillation _j I II

~uesti~~ s~~ur~e:_j IFacility Exam Bank_j ~~estion Modification Method: ~I Concept Used JrUs~d ouiing '!raining Program j []

?ues~n~~u_r_c:e_~om~ellt~ I 080529- original question contained list of evolutions and asked which of them could be performed during release.

I

[ ... . -* - ----,

Comment

-.-.-c--" c*- *-**--'

I I

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-~

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

"§uestio-J'l~fopicj !S~08 I Given the following conditions:

- Unit 2 is operating at 100% power.

- Operators receive the following alarms:

- OHA B-13 21 SW HDR PRESS LO

- OHA B-14 22 SW HDR PRESS LO

- SW header pressure indication in the control room reads 98 psig for both headers.

No other OHA's have annunciated.

Which of the following describes both the possible location of a Service Water System leak which would cause these indication, and how the CRS should respond?

Assume each of the leaks is large enough to cause the indications present in the control room.

ia.' 14 Service Water Bay. Split SW Bays by closing 21SW17 and 22SW17 lAW S2.0P-AB.SW-0003, Service Water Bay Leak.

I

'!)_:_;I Nuclear header x-over line between the 21SW23 and the 22SW23. Shut EITHER SW23 using Attachment 6, Service Water Valve Malfunctions, of S2.0P-AB.SW-0001. J

~-~2B EDG Lube Oil Cooler. Isolate BOTH SW supply header isolation valves and BOTH return header isolation valves to 2B EDG lAW S2.0P-AB.SW-0001 Loss of Service Water Header Pressure. I

'C:t. 1 j21 CCW HX end bell. Ensure 22 CC HX controller set lower than 21 CCW HX, and isolate 21 Service Water Header using Attachment 4, i .:>ervrt;e vv<:uer neauer r::;urauurr, ur .:>L.ur- -1"\0 *.:::>VV vvv 1.

- I!

,-Ans-;er ~ !~~LeveiJ Is I icognitlve Le~- !Application ****1 ~acilityJ ISalem 1 & 2 I ~~mDate_:_ j___12_1_15_12_0_1__,4j

~~~joooo62A2o2 1 1 ~o~ ~va~~:~~-;~"[J]~ez~~~~l~~~P::DJ~~;.~:~DJ s~~- ~

,5¥:~!em/Evolutl0rllitle ILoss of Nuclear Service Water 11062

~ta~ent: 1 Ability to determine and interpret the following as they apply to Loss of Nuclear Service Water: I The cause of possible SWS loss I Explanation of1 55.43.b(5) The TGA SW header pressure is maintained at 80 psig by 2ST1. The low pressure alarm (which is not present) is 7 ; ]

1Ans""_ers:__ __' psig. In order for a leak to be large enough to cause both SW Nuc headers to lower to 99.5%, the TGA low pressure alarm would be in also. A leak in the Service Water Bay would also cause the bay sump high level alarm, as well as the TGA lo pressure alarm. The cross over line is normally isolated by the shut SW23's. 21 ccw hx CAN BE MANUALLY ISOLATED WITH MANUAL iVAL\ll=~ ANnnni=*::: fiJnT Rl=niiiRI= l~nl ATII\IG THE \1\/Hnl F 1\IIIC HEADER The. FOG's eacb ba11e 2 srrnolv isolation vail Ione form each header and 2 return header isolation valves going to 11 or 12 discharge path ** * . I r- __--_--__y~~~--;TJ!I_--e- ---=-~-~ =--F~!iRe~en~N~b~ I 'Reference Section- I :Pag.!:_No.j ~e_-vi-s!Ortj ILoss of Service Water Header Pressure lj S2.0P-AB.SW-0001 I I[ J j16 ]

IService Water Simplified. II ~q5342-S1Mf SheetS:1 a~d 2 I II I!  !

II II lj _ _ ____.ll II I

-I L.O. Number

~ - --- --- Objectives

[ ABSW01 E005

A SBLOCA has occurred on Unit 2 outside containment.

- Actions 6fEOP-LOCA-6;LOTI\ ODTSIDEC:Ul'JTAII\IMEI\IT, navefalled toisolate tile breaK:----

- RCS pressure is 1440 psig and continues to lower.

identifies the procedure that will be used upon transition from EOP-LOCA-6 and the actions that will be directed in that I .

EOP-LOCA-5, Loss of Emergency Coolant Recirculation. Check for subsequent failure and conserve makeup inventory.

55.43.b(5) Dis correc~ because an unisolablebreak will transition out to LOCA-5. LOCA-1 Is incorrect because It would be the appropriate transition if the break were isolated. 43.5 because the SRO Is required to understand the actions of the LOCA-6 procedure to Identify the appropriate procedure to transition to when It is complete, and actions performed In that procedure.

~~'ij~;,;~,~ij!W!Jii~'!!;;~c;;~~;a*itll<6'~lll~l!!I~ietJ9n:.!J bfiJ~ l!i¥1~

J LOSS OF EMERGENCY cooLANT RECIRCU 112-EOP-LOCA-5 I I lis I ILOCA Outside Containment 112-EOP-LOCA-6 I I 1j21 I 1--------------~~----------~~-------~1 I

~~r;;;;;;;;J I LOCA06E008 J I I I I BtwW.C§ifll!!~i~'a@p.l!!IRI 11

~~~I!!JIIIjJli:J IFacility Exam Bank l(~i~~ Editorially Modified I[trtJD1!lj1J;ir~ining l'r§9ta11 0 W!@,W'~!l:~fu~ilM) 1060930. Did not ~nsider it significantly modified because correct answer stayed correct, just made Into a 2 and 2.

changed order of 1tems In c,d to match order In procedure.

I

k. B.?~SkyS_craper _1 - s~QJ~yscr~~er~t iii'"Rb sysielli/Evoltlllo(l Ltst,;J ;.;:g~o §y$1em/~llol~fipn ~~~l ,, '; Ou~lltii! (;h~ng~~ I imJm~6~1!J /SRO 10 Given the following conditions:

___ .:_ ljnlt 2 has exQerienced a MSLB at the Main Turbine Inlet steam piping. -------------------------

All attempts at Main Steamline Isolation have failed.

Operators have just transitloned out of EOP-TRIP-1.

22 and 23 AFW pump are in service.

RCS cooldown rate Is 105'/hr.

RCS pressure is 1300 psig and dropping.

Charging system Sl flowmeter Indicates 290 gpm.

The RCS cooldown is NOT being controlled.

Which choice identifies a subsequent action that must be performed, and why?

~ Isolate AFW In LOSC-2, Multiple Steam Generator Depressurization, to any SG with NR level >9% to minimize cooldown while still ensuring the SG tubes remain wet.

Reset SGBD Sample Isolation in LOSC-2, Multiple Steam Generator Depressurization, to allow sampling of SGs and transition to SGTR series if required.

Trip all RCP's in LOSC-1, Loss of Secondary Coolant, to prevent seal package damage if RCS pressure continues to lower.

~ Trip 23 AFW pump in LOSC-1, Loss of Secondary Coolant, as part of isolating all steam flow paths from faulted SGs.

~~oowE12A201 I@AZTjf!f!!iJ~~@~~~If!:I~U~DJ ~ ~

~~~bfL~*'

IUncontrolled Depressurization of all Steam Generators I [Hz-- J

/KK!lf't!!ltiit!'l'J'!W Ability to determine and Interpret the following as they apply to Uncontrolled Depressurization of all Steam Generators:

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

~~

55.43.(b)5 A is incorrect because LOSC-2 step 5 states to maintain no less than 1.0E4 lbm/hr to each SG. The CAS 3 which talks about SGs with NR levels >9% does not apply to a CD> 100'/hr. That CAS would allow Isolating AFW to any SG is the CD rate was less than 1OO'F/hr, and all SG NR levels were >33%. B is correct because it Is performed at step 10, and if a SGTR were i~,e~tified h~~~,' the CRS ~o~l_d ~~6'~tlon to SGTR-1 and not stay In LOSC-2. C is incorrect because RCPs are tripped less than

~lmlt:~e~i-i~-- -~~~~~~~~~-~~;~~rl!m;~;JiiB:ii~iljit;J ~

ILoss of Secondary Coolant !12-EOP-LOSC-1 I I 123 I IMultiple Steam Generator Depressurization j12-EOP-LOSG-2 i I /126 I II I I I

~itirt.L~ ,JIN~E--Yi~~~~

LOSC02E002 Dl\!5,~ Facility Exam Bank I B\fsf\~-&6\lt'llii!~enfilj 1059885 used stem. Changed to add which procedure used. Replaced 2 dlstracters. Modified previous correct answer to make wrong. Made one of the new choices the correct answer.

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

~ouestion~iopicl 1sRo 11

.. -- -- - -- - - __j I

Given the following conditions:

- Operators are performing S2.0P-IO.ZZ-0002, Cold Shutdown to Hot Standby in preparation for returning the unit to service following a forced outage.

- With the unit otherwise ready to enter Mode 4, the STA reports that a RCS Water Inventory Balance has not been performed with the last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and is required in Modes 1-4 lAW Tech Spec Surveillance 4.4.7.2.1.d.

Which of the following identifies how the CRS should proceed?

fa IContinue to Mode 4 without performing the RCS Water Inventory Balance. The provisions of Tech Spec 4.0.4 are not applicable.

I

b. I, Continue to Mode 4 without performing the RCS Water Inventory Balance.

'- within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Apply Tech Spec 4.0.3 and ensure the RCS leak rate is performed I I

'cJ Direct the RO to perform S2.0P-ST.RC-0008, Reactor Coolant System Water Inventory Balance, for the normal 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration and obtain

~-- acceptable results prior to proceeding to Mode 4. I 1

I

~d. Direct the RO to perform S2.0P-ST.RC-0008, Reactor Coolant System Water Inventory Balance, for an abbreviated duration as detemined by the SMICRS and obtain acceptable results prior to proceeding to Mode 4.

I Ans~er II a I lExam Level  ! Is I ,Cognitive Le~el 1 IMemory I,Facilityll Salem 1 & 2 I 'ExamDate:: I 1211512014!

~JI oo2oooG12o i KA Statement:_]

-*~~--

Ability to interpret and execute procedure steps.

,Explanation of! 55.43.b(2) IOP-2 P&L 3.8 states .... "Performance of S2.0P-ST.RC-0008(Q), RCS Water Inventory Balance, required by TIS

'Answers:

-- -~-- ****-

_i 4.4. 7.2.1.d, is NOT required to enter Mode 4. The surveillance is NOT required to be completed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation." Additionally, TS Surveillance 4.4. 7.2.1.d specifically says the provisions of Tech Spec 4.0.4 are not applicable for entry into MODE 4. B is incorrect because 4.0.3 is applied for missed or overdue surveillances, which is not the case I

! h-'>r-'> Th-'> (' ,,:,,; n . <>r-'> . . hAf'<lll"e the R('C:: IA<>kr::>te \AJnn't he . I mtil ::>fter 1 ') h,-,,,.., ,,

operation, but they are plausible because of the following procedural direction: (P&L 3.15) A routine RCS Water Inventory Balance that is performed to satisfy TIS Surveillance Requirement 4.4. 7.2.1.d should normally be performed over a two hour duration. (P&L 3.16) IF an RCS Water Inventory Balance is to be performed for reasons other than to satisfy the requirements of TIS Surveillance Requirement 4.4.7.2.1.d, THEN the time interval of the RCS Water Inventory Balance may be determined at the discretion of the SMICRS.

-- -- ----- - - - - - -- -- - - ----, - -- - -- - -- -- ---- --: , - - - - ---- - - .- --- ------, r - - -

Reference Title *i Facility

~

Reference Number . 'Reference Section I~Page No. Revision I

__j- -- -- --- -- -

II

- - - -- - __ _ j - - - - -- --- ---- - - _ _j Cold Shutdown to Hot Standby S2.0P-IO.ZZ-0002 I 113 1159 I I Reactor Coolant System W~ter Inventory Bala~ I S2.0P-ST.RC-0008 I !17 1137 i

,I II II _jl II I L.O. Number -I Objectiv I

RCSOOOE008 fRCSOOOE011

! RCSOOOE001

'r<oJI\is~r~~~"ij_l ~;~

~~s~)i~6~B~ill_! ;~~:~~5~.!)11~¥~!~\i~p;g?y~;~( )i::~~g ilysieiilJ~~f~~~~fJ(s\;~_~f.~~~~f~e ~tia~~?fJ _

t91ffi~i[riil.&UI ISRO 12 Given the following conditions:

Unit 1 Is In MODE 5 during a plant startup.

11-RHR-Ioop-ls In-service;---~--~------~-~----------- --*-~~------* -- ------------ -----

12 RHR loop is aligned for ECCS.

i 3 RCP Is In service.

11 Charging pump Is in service.

RCS Tavg Is 175'.

RCS pressure is 310 pslg.

PZR level is 60%.

When placing the second RCP in service, RCS pressure momentarily rises to 390 pslg.

Which of the following describes the RHR system response and how the CRS should proceed?

l!J_ T_h_e_!_Rij~t~f:I_R. ~f RU=_~~y 19 _Q.Qt:UAL/\1Mt;NT_i2l!Mt:_opens. _t;otec~2..0.P-6B.PZB.:OOOJ, .E'ZRP.ress.ure.Malf.unctlon, _and ensurethaL any PZR PORV that opened In response to the RCS pressure has shut.

~ The iRH2, RHR COMMON SUCT MOV automatically shuts. Enter S2.0P-AB.PZR-0001 and ensure that any PZR PORV that opened in response to the RCS pressure has shut.

[ll The i RH3 RHR SAF RLF VLV TO CONTAINMENT SUMP opens. Enter S2.0P-AB.LOCA-0001, Shutdown LOCA, and isolate letdown to minimize RCS Inventory loss. - *

[lli The 1RH2 RHR COMMON SUCT MOV automatically shuts. Enter S2.0P-AB.LOCA-0001 and isolate letdown to minimize RCS inventory loss.

14Er.;;,;r=o=~~;;/f;;;;i'!!o;;m;;,~;;;;ll}Jiii.I!LJ 7.LJ=::r;l;;;;§~;;!'\J; ; ;,~; ; :IJ.; ~ *; :ey,; ;: v;rer; ; ?:lr=lA:=:p=:pu=ca'7:'tlo=n===r;l!; ; -~; ; ;-); [!'1Y; ; ;Y't; Jmj;:;:sa=:=le=m: : 1=: =&=:=2= 1~1!; ,.:,*"':~h'; :;J:>; :_~; J; -hl rl=:::::;;12;:;:;,1:;::;5,2;:;;o;::;i-14\

!ml oo5oooA2oz 1/A2.oz i!W)1.!Ti!J~[§:~s;~~~a.~LW ~§[Jt~1fcttl>\f@[TI~[] 8 ~

r§:iiTI!1:@11fy,1ti)11ltJl!lJ J Residual Heat Removal System I~

~mJlit.mJ Ability to (a) predict the Impacts of the following on the Residual Heat Removal System and (b) based on those predictions, use procedures to correct, control, or mltiqate the consequences of those abnormal operation:

I Pressure transient protection durinq cold shutdown I

~ 55.43(5) With RHR in service both the PZR PORVs and the 1RH3 will open at their 375 psig setpoints. The RH3 opening will not

~ be apparent to the control room, but the PORVopenlng will. AB.PZR, Attachment 3, will ensure the PORV has shut The 1RH2 has an OPENING interlock that requires RCS pressure to be <375 psig, then a keyswitch opens the valve. There Is no automatlc

~~~~~~~';, ~~~~~a,~~~t~ ~.Is ~~alv~ o~~ hl?~h~r~;;ur~;,n~~-~?~~ l!n~!~~in MODE 3 and MODE 4 with the accumulators Isolated, and I

-~9.1\t§~~

IRHROOOE004 I I

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

[Questionropic~! 1sRo 13 Given the following conditions:

- Unit 2 is performing a cooldown from NOP/NOT lAW EOP-TRIP-4 Natural Circulation Cooldown.

- The RCS has been depressurized to 1245 psig.

- PZR level is 16% and stable.

- RCS cooldown rate is 22°F/hr.

- CETs are 445°F.

- 21 and 22 CRDM vent fans are running.

Which of the following describes how the CRS should proceed, and why?

[a.l !.Stop cooldown and depressurization and perform an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> soak to prevent upper head void formation while performing additional shutdown

~ _activities as directed in S2.0P-IO.ZZ-0006, Hot Standby to Cold Shutdown.

'b.-;IContinue depressurization in TRIP-4 until <1000 psig, then close 21-24 SJ54, Accumulator Isolation Valves, to prevent injecting accumulator

'--* contents into the RCS.

• ,c. I Initiate Sl and go to EOP-TRIP-1, Reactor Trip or Safety Injection, due to the loss of subcooling.

I .1: I Start ECCS P"m" ao oeoe'"'>' to "e PZR le,el aod go to TRI P-1.

'I:Ajs~e'"J

~:]I 006000G406 IU ~~m__!:-_~~ ~ ~~g~i~~v~J !Application

'sy~llliE~Iu_~~TitiE!_~ IEmergency Core Cooling System

. !==~~~~~~~~~~~~

! 1=acility.:J !salem 1 &2 1

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I ~ExarnDate:-lj 12/15/20141 1 ro()_6 -

Knowl~f EOP mitigation strategies.

!Explanation of J 55.43.b(5) A is incorrect because the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> soak is only required if RCS pressure is< 1250 and< 2 CRDM vent fans are running.

! ~'!:iv.tllr~ __  : IOP-6 actions have already been directed to be performed earlier in the procedure so could be continued here. B is correct I

I

""" because the soak is not required. Continue RCS depress until< 1,000 psig and isolate ECCS Accumulators per step 25, and correct reason. Cis incorrect because the CAS action in TRIP-4 to initiate Sl is if subcooling cannot be maintained >0, and using lri=C:TT,hlo A itc: 1n7°1= n ic:. hof"'!:>tiC:A P7R leuelisJlp.J.!:u::k.i:nainbinorl ">110/, n<:>rr'aC: <lf"'tinn <>nrl <lf"'Hnn """lrl ho fn Iinitiate Sl. not just start ECCS eumps, since full Sl actuation would be wanted vs just starting ECCS eumps. _j

• -~-~***-*--~---*- ----- * - * - * * - - - - - - *

• l - * - * - - -~

1'Reference Section 1~.0_._Number TRP004E001 J

LMateriaiRequired for Examination I

!SRO 12 CFST Table A Subcooling Table Normal Containment II

[91J:st~n~o~c~ I

_ Facility Exam Bank !IQuestion ModificationMethod: ~ditorially Modified J!Used ~u!i~ T.':<li~ng ProgrCI_I11 I

[J I

--~----~-~--~-~------**-*~----

auestiOn-Source Cornmentsi

- ~ ~ - ..J 050353. Modified to include what procedure to use in all choices. Replaced 2 distracters with the ECCS and Sl distracters .

I

!Comment

- .. - -- - - 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 i9l1~t~o~TopicJ IS_Ro __11_4_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____,1 Given the following conditions:

- A LOCA has occurred on Unit 2.

- Operators are performing 2-EOP-LOCA-3, Transfer to Cold Leg Recirculation with all equipment and off site power available.

- The crew has isolated the charging pump and Sl pump suction from the RWST.

- RWST level is 3.5' and lowering slowly when debris in containment clogs the containment sump, causing all pumps taking suction from it to begin cavitating severely.

What affect will this have on Containment Spray flow, and how will the CRS proceed?

Containment Spray flow will ....

Ia. 1 !remain above 0 gpm due to one CS pump still taking suction from the RWST. GO TO EOP-LOCA-5, Loss of Emergency Coolant

-- Recirculation, add makeup to the RWST, stop any safeguards pump which has lost its suction source.

~~' !tower to 0 gpm due to CS flow being supplied solely from RHR pump discharge. GO TO EOP-LOCA-5 and add makeup to the RWST, stop

'- 1.any safeguards pump which has lost its suction source.

c:t! !lower to 0 gpm due to CS flow being supplied solely from RHR pump discharge. GO TO EOP-APPX-7 and stop all operating Charging, Sl,

• --~ RHR, and CS pumps.

I

~~.. *-~*~~ [E} ~~ '2'!"!_ ~ ,c?2!'!!!!'!'2 1

I '2'.'"!_j l".p; 1 i'3"t;on. i 'Facility: d Salem 1 & 2 h_ExamDat:_: 4 12/15/201411

KAll 026000A207 J[_A2._~_

~--- *---~-- ----- *-

,~ Sta!eme.J1!: i Ability to (a) predict the impacts of the following on the Containment Spray System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding), or sump level below cutoff (interlock) limit Explanation of ' 55.43.b(5) The correct transition from LOCA-3 when containment sump blockage is causing a loss of ECCS flow is to APPX-7.

!Answers:

---

* - --~

I There is a CAS item in LOCA-3 to go to LOCA-5 if emergency recirculation cannot be established or maintained ONLY if the reason is OTHER than containment sump blockage. With RWST level above 1.2' as stated in stem, 22 CS pump will have been stopped at step 8. the stem states that charging and Sl pump suction has been isolated, which occurs at step 14. The crew will be waiting at

. ':>1 mtil R\NST IPVAI ln\AIAr<:: tn1 ?':>t which limP lrl 1:>ininn r.Sn11mn 1?1 \

--- - -----***-- -- *- - - - - * ----------- --- * - - - - - - * * * * - ._.............. , - - - - - . - - ***--*----- -~  :- -------~ --- - - - - - *** 1 Reference Title

. . ._I :_F<JCility~!E!renc_e Number ..J ;R:ferenceSectio_ll__ * ~ge No~ ~vision 1 I

--* ***-*-****-* "**--~*' ***--*w***-*oo Containment Sump Blockage Guideline 112-EOP-APPX-l II I[ I lo J ITransfer to Cold Leg Recirculation 112-EOP-LOCA-3 II II 1129 I II !I. II .....II JI: i L.O. Number

~----------- Objectives .I

~~SPRAYE012 Material Required for Examination I I II cQ~esticmSourc~ I_N._:.e_w_ _ _ _ _ _....~I[Q_u_e_st~o_n_M_~~~~~t_io_n fv1_et_hod__:_)'l.----~---......1 i_tJ_sedDuring *Trai~~g Pr()~ram__l []

9ue~t~on s_ource_Commentsj I J Comment I I

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

~~&io~To~c l1 ~s~;IR~01~5----------------------------------------------------~-----------~1 Given the following conditions:

- Unit 2 is 7 days into a refueling outage on June 15th.

- The core is partially offloaded with 7 bundles remaining in the Rx.

- Spent Fuel Pool (SFP) temperature is 124°F.

- SFP level is 5" above normal.

- 21 SFP trips on motor OL, and can NOT be restarted.

- 22 SFP pump is placed in service with SFP temperature at 138°F.

Which of the following contains both an expected plant response to this failure, and a condition which must be met prior to transferring the remaining fuel bundles into the spent fuel pool?

The SFP ...

[a.J I High level OHA annunciates. 21 SFP cooling pump must be restore to operable status. I J>] I High Level OHA annunciates. The 22 SFP must have normal and emergency power supply availability verified. l

' - neralizer automatically bypassed flow when 22 SFP was placed in service and demin inlet temp reached 136°F. 21 SFP cooling pump be restore to operable status.

I

~~I 033000A202 System!E~oluti()n Title- ISpent Fuel Pool Cooling System 033

...______ - r"'""------------------------------------------------:

~ StatE!111ent: Ability to (a) predict the impacts of the following on the Spent Fuel Pool Cooling System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Loss of SFPCS

,ExPlanation of.i 55.43(5) IOP-10, Spent Fuel Pool Manipulations, states (P&L 3.4) that transfer of spent fuel into the SFP is to be suspended until

'Answers:

BOTH SFP pumps are OPERABLE. If in service, the demin was removed from service lAW CAS 2.0. OHA C-27 SFP Lvl hi will occur at 6" above normal, and would be expected to occur with a pool heatup of 14 degrees, as noted in CAUTION in AB.SF on top of page 2. The SFP demin does not auto bypass, but is plausible because the CVCS demin does, at 136°F. The SFP demin would

~ *~lh. f, '~

• 1/\. lA* ARS.E r.A C:: ;t, ') n * <:;_!=P ll'>mn i<: 1"l.n°!= Thi<: niiP<Otinn es the requirement for SRO only based on required knowledge of the section of the procedure being used to allow fuel handling commence. This is the first bullet of the last block of the Clarification Guidance for SRO-Only Questions.

-~~~-~ --=-~efere~ce r]t~ -------=-_j ~_!acility Reference Number- j ~efere_n_c_eS_e_ct_ion __j ~g_e_No. :Re;isi~

ISpent Fuel Pool Manipulations !I S2.0P-IO.ZZ-001 0 II II 1132 l j Loss of Spent Fuel Pool Cooling II ~2.0P-AB.SF-0001 II lj ] 111 J IOverhead Window C II S2.0P-AR.ZZ-0003 II 1132 1117 I LO. Number Objecti ISFPOOOE013

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aue5ti011101Jk: ~ lsRo 16 Given the following conditions:

- Unit 2 is performing a Rx startup.

- Rx power is currently stable at 6%.

- 22 SGFP is supplying Main Feed to all S/G's.

- Steam dumps are controlling Tave in MS Pressure control- Manual set at 980 psig.

- All MS10's are closed in AUTO at 1015 psig.

- 23TB40 fails 50% open.

- Auctioneered high RCS Tavg is 540.9°F and slowly lowering.

- PZR pressure is 1984 psig and slowly lowering.

Which of the following describes how the CRS should apply Tech Specs, and why?

, Assume no auto Rx trip setpoints are reached.

I

~-J Restore RCS Tave to at least 541 oF within 15 minutes, or Rx trip breakers must be opened within the next 15 minutes because adequate SDM cannot be assured. I I

!i:)~ Restore RCS Tave to at least 541 °F within 15 minutes, or Rx trip breakers must be opened within the next 15 minutes because 1- vuoCtb,v instrumentation is not within its normal Uf-1<::1 """ 1\:J range. I

;

-, I

"--- J Restore PZR pressure to at least 1985 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or lower Rx thermal power to <5% rated thermal power in the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> beca1 initial FSAR analysis for minimum DNBR is not met. I

! I

'd~! Restore PZR pressure to at least 1985 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or lower Rx thermal power to <5% rated thermal power in the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> because  !

lure rne:uyur LU LII!J 1ur Ul\10 ler<:neu !JIULe<.;uve a<.;uurr~ r~ ueruw ure llllllllllUIII a~~unreu varue.  !

~~sii\I!!'J [ : ] ~~~."!_L._e~~,js 1 I l '_(;_o~~~!_~~I_,!Applicaiion ' :_j ~~~~:~ j~a~em 1 &2 j ~~~~D~t_El:_~ , _ _ _1_21_1_51_2_0_14_.1

~~:jj 041oooAzoz I:A_3.()2 - ~ ~~~ifieJ~ ~~~oj~l~e_[J] ~~_!!~~~~ ~~~§-~IJ_()~[JJ ~~_(;_r_o_i~[JJ ~4.l ~

"systen1ie~oiuti0n-Titie

-*--*---~----*- -

~ I Steam Dump System a~d Tu~bine

- Bypass Control

~---- --~--~ -,

~~S.!C'I!~_E!_n__t:' Ability to (a) predict the impacts of the following on the Steam Dump System and Turbine Bypass Control and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Steam valve stuck open Btplanation of ~ 55.43.b(2) RCS Tavg is required to be maintained at 541 oF or greater in Modes 1 and 2 lAW TSAS 3.1.1.4. If not restored within 15

'~'!_S_'IVE!_r!>: ___ 1 minutes, the plant must be in Hot Standby in the next 15 minutes, which would be accomplished by opening the RTB's. The bases for this temp is to ensure 5 different things, one of which is that protective instrumentation within its normal range. A is incorrect because SDM margin is not one of the 5 listed bases for minimum temp for criticality. C and D are both incorrect because of the It;,.,.,, ro,-,, ,;rornont ~,h;,--h ;., ') hr" "" 1,-, ro<>,--h -IC:R <>rlrl;t;,-,n,fl,, n ;" f, >rthor ;n,--,-,rro,--1 h.,,..., '""' ;t., h"'""'" ;c ln,--,-,rro,--1 I

---~-~----- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - 1 ____________ l _______ - - - - - -

,' _ _ _ _ _ Reference Title ~ 1 _ ~ac;ili_t}'_ Reference Number ,,Reference Section _1i ~~e No.; ~evi!;i~n~

I Salem Tech Specs H Jl II I! I II II II J!,, I II II II j

---

--I

~~-~-_!-liJ_~b!r ____ 1 RCSOOOE009 I

I

__ ::~?~%~\i~9~~!l'f:J !:~~ii§:~~~t?_P~;*,J. ~~S}~~Itnile~b!\lt~~nt~si:~t~:-~~o*s!sl~.m!Ev_61~:~~~-~~~~_r;~:~$:~~J_~~-;ch~ft~~~l!}]

((@!!'!J;m§l JSRO 17 Given the following conditions; Unit 21s operating at 100% power EOL.

1\filileaCurrenfTransform~alnGenerator mell6rlng c1rcwtcauses8Maln _______ _

Generator trip, causing a Main Turbine trip, resulting in an automatic Rx trip.

The Feedwater Interlock (FWI) falls to actuate when expected, and has not actuated, Which of the following Identifies a consequence of the FWI failure, and how will It be addressed?

I!J Excessive cooldown of the SGs, Initiate Slln TRIP*1 based on degrading board parameters.

1!1 Cavitation of the Condensate Pumps, Initiate Slln TRIP-1 based on degrading board parameters. __

~ Excessive cooldown of the SGs. Trip both SGFP's In TRIP-2 after verifying adequate AFW ftow.

l[fJ Cavitation of the Condensate Pumps. Trip both SGFP's In TRIP-2 after verifying adequate AFW ftow.

~ 0 ~il fO ~~tt£.l:~Jk.l:\'1.1)1~ !Application I~ !salem 1 &2 IIJ¥~~~hpiJ~::~ J 12/15/20141 fml 059000A203 jfA2.03 l~@mJI!ti!!l.2.!:l ~!mrit9"1>1!m!UI!\l~:lillt'iffi!IO Ia ~

fsystei]iJ~f<iW!6ttl.W!%!l 1[osg 1 l\it~(~I'JP'!i'ib'e&ii I CN&FDWE009 I I TRP002E002 I ITRP001E013 I

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

'_§u~stjQr\)~ic(  !SRO 18

,,__._ff_

Given the following conditions:

- Unit 2 is in Mode 3, NOP, NOT.

- 2A EDG is being run for a monthly surveillance, and is currently loaded to 2550 KW.

- 21 CCW pump trips.

- 5 minutes after the CCW pump trip, the field NEO reports the breaker for 21 Diesel Fuel Oil Transfer Pump is tripped.

I Which of the following describes the Tech Spec entry required and its Bases? I ja.! Enter Tech Spec 3.8.1.1.b.2 for EDG Fuel Oil Modes 1-4 since BOTH Fuel Oil Transfer pumps are required to be operable. This ensures that

'I)~

sufficient power would be available for the safe shutdown of the plant and for the mitigation and control of accident conditions.

I Enter Tech Spec 3.0.3 because all redundant equipment in the CCW system is not available, making both loops of CCW inoperable. This ensures that sufficient safety related CCW would be available for the safe shutdown of the plant and for the mitigation and control of accident conditions.

c. Enter Tech Spec 3.8.1.1.b.2 for EDG Fuel Oil Modes 1-4 since BOTH Fuel Oil Transfer pumps are required to be operable. This is to ensure a timely unit shutdown is performed when the plant operation cannot be maintained within the limits of safe operation defined by the Limiting Conditions for Operation and its action requirements. I

'd., nter Tech Spec 3.0.3 because all redundant equipment in the CCW system is not available, making both loops of CCW inoperable. This is to !

nsure a timely unit shutdown is performed when the plant operation cannot be maintained within the limits of safe operation defined by the I

L-11111111 \J II Ill 11,:, lUI v' Cl ll II II ll I Ill II Ill II

~s..v_erJ ~ '§c.a~ !:_e\f:l 1 Is i'C_<>~~tive LEwe_l_ -~ I Application j 1Facili~: ISalem 1 & 2 I 'Exa-m~:>_a~:~ j___1_21_1_51_2_0_14_.1

,_~:~jo64oo?G222 l:f212_~ = .~R§_~~~~[I§j~~§_~ru~(ID~cjl_o~jj~*-Fi~~~iJOJ~~§_~o~~[] '55t4if ~

~~~e@~"'u§nTi~ I Emergency Diesel Generators 064

,KJ\Siate"me~:

'-~--- ~-----*--

1 Knowledge of limiting conditions for operations and safetx: limits. I

'Explanation of 1 55.43.b(2) A is correct because it is gleaned from TS bases section for elec power sources. The bases in choice b is from the ac

'Answers:

• -* ______ j* power sources. C is incorrect because even thought TS is correct, the reason is the 3.0.3 bases. D is incorrect because both the TS and bases are incorrect.

I

---

-- ---- ---*---*--*----.---.---*-**------ ----- --l--Fa-Cility________ ----------, ----1,-----

Reference Title *-

__ _j I Reference* Number 1 Reference Section__ 1 -~~e_ No.

-*--*--'----*----*- --~

'Revision.

_________ _j ISalem Tech Specs I I I ,3/~/8~1 II I Salem Tech Specs Bases I I I B 3/4 8-2 ,,

i J

I I I II I rl:O. Number*--*-

Objectives l

EDGOOOE011 Comment

- - - - - - - c - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - _j I

I

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

~§_uestfon T~13ic i jSRO 19 Given the following conditions:

- Unit 1 is performing a Rx startup.

- Rx power is 3E4 cps.

- SUR is .2 dpm.

- During the refueling outage, BOTH IR Nl detectors were replaced.

- IR Nl indication for both N35 and N36 is flashing at 1x1 0-11 A.

Which of the following describes the status of theIR instrumentation, and the required action(s) that will be performed?

Both IR Nl's ...

I i~ should be reading -1 x1 0-1 OA. Declare BOTH Nl's INOPERABLE and enter TS 3.0.3.

I

,~~ should be reading -1 x1 0-10A. Declare BOTH Nl's INOPERABLE and enter TSAS 3.3.1.1.

ic~"

!; under compensated. Stabilize power, block SR Hi Flux trip, and correct compensating voltage problem for BOTH IR Nl's prior to ceeding 5% Rx power.

'd-:-: !.are overcompensated. Stabilize power <P-6 and correct compensating voltage problem for at least ONE IR Nl within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or open

- Rx Trip Breakers within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

'Answer i j a I :Exam Level' Is I 'cognitive Level' IApplication I ,FacmtY:i ISalem 1 & 2 I fExallloate: I I 12/15/20141

[KA~~194001G120 11_2.2-30_ __l'!ID-vatueJI4.6L!R~\,/alu~J 4.6 i~ection~liPwG 1:_~-- _l[J},.,_~ ~.->-li!>~:[J] if1~: ~

l<A State-ffient:

---

IAbility to interpret and execute procedure steps

Explanation of 1 55.43(2)There is at least a one decade overlap required between the SR and IR Nl's when raising power. (Step 5.2.33) With the SR Answers

1 indication at 30,000 counts, the decade of overlap should already be present. With the Hi Flux Trip at 100,000 counts, there can't

~*-* -- --- --- be proper overlap. With no other information in the stem to provide inference of any other problems with the Nl*s EXCEPT that both IR detectors were replaced, theIR Nl's should be declared INOPERABLE. A is correct, and b incorrect, because there is only an i ,;,rtinn in"\':\ 1 1 fnr ()1\IC lf"\DCDIIDI I= ID . R()T~ 11\f()PI=RARI I= TC:: ~ n ~. n ic.  :<-

candidate thought they were reading low due to overcompensation, the P-6 block would not be manually performed without the power above P-6 interlock to allow P-6 to be blocked. Cis incorrect because under compensation would cause a higher than readinQ.

______ Reference Title _ .J f___ _!_at;_!l_itL~~_!!r~c~~~ _I ~e.':!n~ S~t~ _ 1i Page N_c:>:_! 1 ~vi~~

=~H=o=tS=ta=n=d~by:to~=M~ini~m=um~=L=oa:d::::=====~:~~~=S1=.0=P=*=IO=.Z=Z=-0=0=03=========;11~============4!~~~3=2==~1

[salem:TechSpecs II = : =: := I[ ll _ _j~j~j 11, _ _ _ _ _ _ _ _____,11,~_ _ _ ____.11 II !l _ ___.l LO. Number I

--*-' Objective I--* ---~--

EXCOREE009

)Material Required for Examination I I ~,. .. : :  :  ::  :  : II I

'c:luestion Source: J Facility Exam Bank

~ I Question Modification Method:

1 jDirect From Source I~Used During Training Pr<>gra_rll] 0 I

- - L So~~c:_~o~ilt_ent~

pue_sti?n Used on Salem 8/2008 NRC SRO Written exam (4 NRC exams ago)

I iComment

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

[QuestionioPic~ 1sRo 20 Given the following conditions:

- Unit 2 is performing a cooldown lAW S2.0P-IO.ZZ-0006, Hot Standby to Cold Shutdown.

- Mode 5 has just been entered.

- PZR temperature is 400° F.

- The board NCO reports that PZR hot calibrated level on all 3 channels indicates 95% and is stable.

Which of the following identifies what the Cold Calibrated channel will be reading, and how charging flow should operated?

!a. 157%. Raise charging flow.

1

~**

I b.~_57%. Lower charging flow.

'~~,81%. Raise charging flow.

'd. i 181%. Lower chargir flow.

'An~ ~ 'Exam Lev~l. ~ Cog~itive Level I I J\ ,,;, ,,;,

I [FacilitY:] ISalem 1 & 2 I ExarriiJatell 12/15/20141

~JI194oo1G12s 12.1.~~-~~o~,IJ~:il3.9l SRC?_\f.alue)~[s~ction:]IPwG liR()Gro~p:l 1j~oq!()llll][JJ '55.43 ~

'-:-/2;:~-

'System/Evolution Title

~. ------**- - l - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - l ~N_ERI i KA Statement*

L..C. -----~

Ability to interpret reference materials, such as graphs, curves, tables, etc.

cExplanation of 55.43(5)Using Exhibit 1 of IOP-6, page 2, shows that with hot cal level at 95% at 400 deg., the ACTUAL level in PZR is -66%.

Answers:

--. -*~

Using Page 1 and the ACTUAL PZR level of 66% and going across to the 400 deg line, cold cal level will be 57%. After the cooldown rate has been reduced as per stem, IOP-6 has operators raise charging flow to establish 80% cold cal level. Prior to the 80% requirement, the procedure requires 25-53% (5.1.5 and 5.1.16) The distracters either have the incorrect level or to lower I cbacgiog fi!JlA! Salem expecieoced a E!Ze dtaio-dnwn ouont in ?flflil rlrro tn '""" th::m ::~rl"':"t"' ~rnr<>rl'""' rlir

• n~<>r.:~rr>r understanding of the hot cal/ cold cal correlation in PZR level, and heightened emphasis on this relationship and better procedure direction has been applied. This is not direct lookup due to the changes in required Cooldown rate, PZR level, and the transition between hot calibrated and cold calibrated level.

,-~----- *----~------ *-- -l-~---~-**** . *--------... * - * -.--**-*

Reference Title r Facility Reference Numbef'___J :Reference Section ~~ge _('l()~ ~V:'<:>'vJ IHot Standby to Cold Shutdown I S2.0P-I077-nnnn I I I 44 I l 1: I I I I il li I I I I 1

L.o:N"urnber ,

I Objectives

'IOP006E009 I PZRP&LE008 I I

[Material Required for Examination ...

I ISRO 20 S2.0P-IO.ZZ-0006 Rev. 44 .... _u Question Source: Jl Facility Exam Bank I!Question-Modification Method:

ISignificantly Modified I c .. .. ..

!tJ~_(i_[)lJ.ri_llgTraining Pr_<>~ram []

Sou~ce Com~~en~~ j 08-01 r?'uestion NRC SRO (3 exams ago)

I

[Comment

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A Temporary Configuration Change Package (TCCP} will be converted to a Permanent Configuration Change lAW CC-M-112 Temporary Configuration Changes, and CC-M-103 Configuration Change Control for Permanent Physical Plant Changes.

Implemented permanent configuration change must be accepted by Operations prior to closing the TCCP.

permanem configuration change Ops .l'.cceptan"edate

• as the TCCP removal date. _

55.43.b(3} When implementing a Design Change Package (DCP} to make a TCCP permanent, the removal work order for the TCCP must be completed, even If the only task is the removal of the TCCP tags, update of Operations TCC Tracking Log, and completion of referenced administrative controls and OWDs. This may be done in the DCP work order and the completion of the removal work order would reference the DCP work order but the removal work order for the TCCP should not be cancelled. B Is

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ITemporaryConfigurationChanges llcc-M-112 I 11~114 I II 11--.JI 11

_ _ _ _ _ _ _ _--111 I li__jj I

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I MISCAPE001 I

.:;R_o ~~y§£riii~txl E*SJ:l<t~~ys~~apet*l ~~o*systeml=~of~tion Usi: *1 ;* ,'s~trs%teinl~volut\o~\f~~J ,;*q~~-ine~qf~~~~~il LcJlOl!!fioruiroPl~ I SRO 22

!Which of the following describes the bases for maintaining an OPERABLE Auxiliary Feedwater System in Modes 1-3 lAW Tech Spec 3.7.1.2?

~ Eures the capability to cooldown and maintain the RCS at <500 *f for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a SGTR assuming failed fuel.

I I

~ Remove decay heat and_malntaln the RCS at HS13 conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> jollowi[1g a <:COfT1pJete los~ of_ ofj,site power.

I

~ Ensures that the RCS can be cooled down to <350 deg F from normal conditions following a complete loss of off-site power. I IK:j Provide the RCS heat removal capability necessary to prevent a challenge to the pressurizer safely valves during a full power ATWT.

I

~D!!~ILJ~~-~Memory ~~~~Salem1&2 lffiff~l 12/15/2014!

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m_~!li~ll!l],~lJli!!J (GENERI]

!Ki\'sta~~

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. I e!!! 55.43(2) All of the dlslraclers are incorrect because the Tech Spec Bases document does not contain those reasons. Additionally:

B is plausible because the operability of the AFWST is for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> maintaining HSB with steam discharge to the atmosphere. A Is plausible because the RCS cooldown during a SGTR to <500 degrees is to limit the potenllal off-site dose. Cis correct per Bases.

D Is plausible because the ATWT bases document for AFW flow Is for decay heat removal.

e::.:::~f*fJ'ji'~~~;f:eoliltyJ~~!l!DJ;f~-!l~~~ ~

ISalem Tech Specs Bases I ,,3.7.1.2 iiB3/47*21~

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AFWOOOE011

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B¥l[s~~~~~'fii.ID Vision 060263. Last used 3 NRC exams ago at Salem.

RO SkyScraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes I l:c:t~~stion Topic~ ISRO 23 Given the following conditions:

- Unit 2 is performing a Tech Spec required shutdown from 100% power lAW Tech Spec 3.0.3., at 20% I hr.

- Current Rx power is 85%.

- A containment entry is being planned in an effort correct the condition which has caused 3.0.3 entry.

- The personnel entering containment will be going ONLY to the 78' elevation outside the bioshield.

Which of the following identifies the Rx power limitation, if any, lAW SC.SA-ST.ZZ-0001, Salem Containment Entries in Modes 1-4, which must be met for this containment entry, AND a Salem position who is REQUIRED to authorize this containment entry?

I

~~~; No power limitation. Radiation Protection Supervisor.

~~ I Rx power <50%. Radiation Protection Supervisor.

I

~~ No power limitation. Operations Manager.

~~~ Rx power <50%. Operations Manager.

,.l;;x~m. l 1.c~oi1iua IM~oo::;.,; ,Facilit~H Salem 1 & 2 H~am~~t::_ rl 1

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Isystem/EvolutionTitle ' 'GE-NERI-1 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handlinq responsibilities, access to locked high-radiation areas, aligningfilters, etc.

r -- 1 Explanation of ' 55.43(4) Normal Containment entries at power are governed by both of the below listed procedures. Authorization to access

'Answers: i containment is provided by the SM/CRS/Designee. However, when power is being changed >5%/hr, the Radiation Protection Supervisor (RPS) approval is required. Since the SM/CRS/Designee is not one of the available choices, the only correct answer is the RPS. Since the stem asked whose specific approval is required, the Operations Manager is not correct, even though he/she

~;,.,n~ tho ~nn *~I ,f ~II ,.,, '~ ""~arlo'"'~ l>rlrl;t;, d . lhArA

• n, n. >r *,~;,,,,;,  ;~tw*~ *Hh 'nt~; ~ ont." Th" RPM

!approves a"ccess to certain areas of containment, b~t the stem states they are ~9oing into those.

- - * - - - - - ---*- -----------*--- ----,-----~ -- -~~-----~~~ ~ *-- ---*--*~~---, .-~-~---.

'-------~!~e'!_n~!i!l~. ____ ~ -~ ~-F~il_i!y~e_~r~c_!N_I:I_~~- ~e_!Elre_~cEl_!)e__ct~n_ .J !!'_a~~*J ~R~vi:>_io~

j Salem Containment Entries in Modes 1-4 II SC.SA-ST.ZZ-0001 I lj li5 j IContainment Entries at Power II RP-SA-1 02 I II I !0 I II II II __j I I I

,_ -~-~----* *---. Objectives J E012

.Material Required for Examination

• -* -* **- . _j I II puestio~S_c>_~rce:~ j ~acility Exam Bank J ,Questio~ M_od~ic~i~ Method: _ _] Significantly Modified J!used During 'J"rai~n_{J~r~gr_a_m_~ D I

Question Source Comments

- ---- *- ~

Corrected from RPM to RPS.

~ ~ ~:  : __j

,,_ 1 I

J I

RO SkyScraper I SRO Skyscraper I RO System/Evolution List SRO System/Evolution List I Outline Changes f L<luesti<>n_TOiJ~<:_; 1sRo 24  !

'Which of the following identifies a condition which will ALWAYS require suspending any Functional Restoration Procedure (FRP) in use prior to completion, and why? . I

'a.

IA new CFST condition occurs on a status tree different from the one which directed the current procedure implementation. This ensures the most recent plant conditions are used when assessing critical safety functions. I

'-b.ll the A Continuous Action Summary item directs transition. This ensures a transition out to the proper procedure is m~de regardless of what step operator is in the procedure. J I

~ 1 RWST level lowers to the low level setpoint Establishing Cold Leg Recirculation ensures both long term cooling of the core and ECCS Acceptance Criteria are maintained. I

d~ lA loss of all three vital buses occurs. FRPs assume at least one 4KV vital bus is available for mitigative actions.

J

~Anwe~ Id I Exam Leve_l i Is 1

I ~Cognitive Level 11 Memory I :Facility: 1 ISalem 1 & 2 I ~ExamDate~ I 12/15/20141

- ., ~:--"'SfP" i~~~194001G418 112~~--~~R_S>~~~~@~~~"'!I~~~~C!'o_n_:J~~J~~_c:;~u~j3~~-~r~u~f:J} ~~,l~ ~

I l

ISysteffile~C>IutionTitie

~~---- --- _________ J

~ GENERI-1 L ----- ---

iKA Statement* J 1

Knowledge of the specific bases for EOPs. I

!Explanation of 55.43.b(5) dis correct because EOP-LOPA-1 Bases document says ... "this EOP is also entered anytime, from anywhere, on the

'Answers: --*-I symptom of a loss of all AC power." A is incorrect because only a HIGHER RED or PURPLE path than the one directing current

~***-*~-*--*

procedure entry would require suspending. B is incorrect because FRPs don't have Continuous Action Summary like EOPs do. c is incorrect because RWST lo level would only go to LOCA-3 when directed in whatever FRP you were in. Automatically going to nr t._ "l r.n ' " R\1\/C::T '"' rlr.as NOI a liMa\££ aCCIIC ao matter wbere ia any proce.d1~<e liD! Lare

---*-*-- -- ---- -- --- *- **-*-- -------*-*--*-*------------- -* -- *--- -------*-*----,: ***- *-- - *- .. ---------*1

'c- ~cili~Reference _l!um~l'__ J ,~f~_ence SectiOJl__j ~CI!I~<>:_! ~Ellfision

• ----~-------*- ~--*-- ~-

I

_____ --~~erence Title _ _ _ _ _ J j Loss of All A~ Po~er ___JI2-EOP-LOPA-1 II Bases Document IIi : ] [~7 : I I II II II II l I !I II II II I

~--- -* -~ .. ___ ,_ ~ *-*-

1 LO. Number  !

Objectives I

LOPAOOE014

~ -------~-~

I I

_Material Required forExamination ..

I I r I

II

~

Question Source: i Facility Exam Bank I[auestion Modification Method: ,, Significantly Modified 1[!.~sed During Training Program

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

1

,9~~~~o~ ~o~~~5:<>IT!:0en~ 1.057948 modified to remove "none of the above" choice, and added bases to meet KIA.

,comment

.~-.-* -*-- -~

- -*- -~

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

auestio~n Topic ' SRO 25 I I I

Given the following conditions:

- Unit 2 is at 100% power.

- An Unusual Event is declared due to a loss of all overhead annunciators for greater than 15 minutes.

- Prior to the Primary Communicator making any of the 15 minute notifications for an Unusual Event (UE), the problem is corrected and the annunciators are restored to operable.

_yvhich of the following identifies how the EmergencyCoordinator should proceed?

~~-=I Terminate making the 15 minute notifications and ensure the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> After-the-Fact report is made.

I

'!>~'!Direct the Primary Communicator to make all required 15 minute notifications, then retract the UE lAW the proper attachments.

'<:~ IDirect the Primary Communicator to make all required 15 minute notifications, then terminate the UE lAW the proper attachments.

d-j_l.Terminate making the 15 minute notifications and ensure the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report for Major Loss of Emergency Assessment Capability is made.

I

~Aris;,.,eJ"_J Ic I IExamLev_el i Is I ~ognitive Le~el I

, Memory lrFacllity:

1 ISalem 1 & 2 liExamDate: ij 12/15/20141 c* .,. ,.,.

~~~:JIJJJ ~R~~Y<l~e~!~c~n:J~ ~0 ~o~JU ~~r~: [J] '55,~4i ~

r I l

,KA: (1I4oo1G429

- ~-~* -- --~ -1

System/Evolution Title GENERI!

1----------------------------------------' L. -~*

KA Statement*

~ ~-* -- '

Knowledge of the emergency plan.

I * - -:1

'Explanation of 1 55.43.b(1 )The following discusses Short Duration Events which are corrected before declaration. This requires declaration then Answers: termination, so the actual declaration in question above should be no less restrictive. "For some events, the condition may be

~

~--*-~

corrected before a declaration has been made. The key consideration in this situation is to determine whether or not further plant damage occurred while the corrective actions were being taken. In some situations, this can be readily determined, in other l

I c:ilile>linnc: fllrthPr (p tl r'l"\1"\bnl re>~inrh"mic:ll"\, c:e>mnlinn me>*. hP rle>c:"in. th" PI/Pnl "'".

terminate the emerge~cy on'ce-assessment shows that-there 'were no consequences" from the event and other termination criteria are met."

. e>n~

c L

Refere~ce ritie~- *- *- ~ !~ ~Cl<:iiitrReference Number~ ~eference se-ction ~ ~ge No._! ~e,.~on_

Event CI::J: -"'* *"* Guide Introduction & Usa I EP-SA-111-101 I 1 13 II o I I I I II I I I I II I Objectives I

I

_Material Required for Examination J I II

,Question Source:

--* --**-~ --*

IFacility Exam Bank i!au:stio~ Modification -Met~od: Jl Editorially Modified I~~ed ouring::J"..:ain_~$1. Pr<:)~ram D iauestiC>r\sourcecoffiments 1

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Comment

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