ML12353A456
| ML12353A456 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 11/20/2012 |
| From: | Operations Branch I |
| To: | Public Service Enterprise Group |
| Jackson D | |
| References | |
| TAC U01859 | |
| Download: ML12353A456 (117) | |
Text
u.s. Nuclear Regulatory Commission Site-Specific
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Written Examination
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Applicant Information
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Name: Region: I Date: 12/3/2012 Facility: Salem 1 & 2
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License Level: RO Reactor Type: W Start Time: Finish Time:
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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
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Examination Value Points 0--*- ...
Applicant's Score Points
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Applicant's Grade Percent
, Gh"," th' foll<>Mn, oond;tio",
I
- Unit 2 is at 30% power.
- Control Bank 0 rod 203 drops fully into the core.
The reactor does not trip.
The crew is performing actions to recover the rod lAW S2.0P-AB.ROD-0002, Dropped Rod.
Which of the following islare EXPECTED to alarm durin the control rod recovery, and when?
IOHA's E-8 RIL LO and E-16 RIL LO LO. These occur when the PIA Converter is reset to zero prior to withdrawing the dropped rod.
OHA's E-8 RIL LO and E-16 RIL LO LO. These occur when the STARTUP PB is depressed to reset the Rod Step Counter for Control Bank o. I I
OHA E-40 ROD BANK URGENT FAIL This occurs when the Rod Bank Selector Switch is moved past the AUTO position into the Individual Bank Select position for Control Bank D.
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,~~:i~i~~t~~~kBh~~:b~:~;~:n:;IL. ~~iS:)~en withdrawing the dropp~~ ro~ after the CRDM lift C~~~itches f~;r all the unaffected, I
~_; [CJ ;ExamLe\l~ ~ ~.!tl'l'!'!:'~l:i~ IApplication I ;Facllity: i 1Salem 1 &2 I tExam,Date:; I 12/3/201211
~j 000003K205 IIAK2.0~Rov~~.gRovaiOe::r:IE[sectIOnJl~ iRO Group;U[SROGroup:'U lEi) []
. . IDropped Control Rod I
'KAStatementl
- - - - - - _ . Knowledge of the interrelations between Drooped Control Rod and the following: I Control rod drive Dower supplies and logic circuits I Explanation 5'f, 55.41 .b(6) A is incorrect because the PIA converter is only reset (and would cause the alarms) for a GROUP 1 rod recovery, and An~werl>_:_ _ . the affected rod is 2D3 (Group 2). B is incorrect because the STARTUP PB is not depressed during a rod recovery, only when performing a startup. C is incorrect because contro! rod movement when passing through auto is a concern, and the alarm would not occur then. D is correct becaus," a Power Cabinet Regu!ation failure occurs when the rod movement demand signal is sent out i I tn r"ntml R",nir n ,<11'1", <lnr1 >"nl\, 1'1"" r"r1 ,lith m('\\/<>m<>nt I
1---*';-'.*.* .- - Reference Title -~
-~---,-.~
i .. Facility Reference Number .'1 ;Reference Section M.-----l ' _. I : p~geNo. i lRev'i~ion
- IDropped' Rod II
- ____ M. ____
S2.0P-AB.ROD-0002 II II 11 10 I IRod Control Lesson Plan I[EOS05RODSOO-11 II II 1111 I I Ie II II II I LO.Number
'--.~.~-
IABROD2E003
- MatElrial Requi red for f:xamin atronI I
_ _ _ _ _ _ _ _...JIUsed During:Tra~riingprogram i 0 I
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I 1
ROSkV~cm~r ;l~:rSl1:iskvs~~geP I SRO*S~ili~..rilE~of~tio~r~flt~ ~,.PfJt!irl~"cIDl5~~!.1
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.~uestionTop!:J Given the following conditions:
- Unit 1 has performed a manual Rx trip from 100% power.
- No 51 was required.
- 11 SG NR level is 9.1 % and rising slowly
- 12-14 5G NR levels are 8.9% and rising slowly.
- Total AFW flow is 24E4lbm/hr.
Which of what AFW flow is removal?
At least 22E4 Ibm/hr must be maintained until ALL SG NR levels are at least 9%.
At least 22E4lbm/hr must be maintained until at least one SG NR level is at least 15%.
AFW flow may be reduced to less than 22E4 ItJm/hr to isolate the SG tube leak indicated on 11 SG.
AFW flow may be reduced to less than 22E4 Ibm/hr because the secondary heat sink requirement is met ExamLevel ii icognitiveJ:.:.vel<.!
~I 000007K106 IEK1.06m~Rovalue:i[~ ~~Blsection: II~ ~~U~O Group:IU System/Ev?lljtion TItle! I_R_e_a_ct_o_rT_r..:ip_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _...1 ' -_ _---'
j<ASt~t~-;:;:;entJ IKnowledge of the operational implications of the following concepts as they apply to Reactor Trip:
I Relationship of emergency feedwater flow to S/G and decay heat removal following reactor trip I: REi!E!rence -T-itJ-e-~m----. .- -Fa~Reference Number mmlReferE!rl~e Secti~: l-~(I~ ~~
II Reactor Trip Response I[EOP-TRIP-2 ---.J Isases document 1111 11 28 I II II II 11 II I iI II 1\ 11 11 I LO. Number ...~
ITRP002E006 IMaterial Requ!r,ed for Examin Clti <:l11
- I II ra~estionSource] I_N_e_w______.....I'qliestionModification Method: _ _ _ _ _ _ _ _......1 jUsed D-~ringTraining Prograrn l 0
iIQuestion Source CVII"'I~"'~" I I I-------------------------------------------------------------------------~
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Given the following conditions:
- Unit 2 is operating at 100% power when a catastrophic failure of RCS loop 21 cold leg piping occurs.
- RCS pressure rapidly dropped to 35 psig, 22 RHR pump failed to start, and remains stopped.
- Initial RWST level was 41.1 feet.
Of the following, which is CLOSEST to the time available from the failure until the swap to Cold Le recirc will be required?
~ 124 minutes.
\33 minutes .
- ~swer .1 b 'Ex~Il1Date;11 IlExam Level! I R ICognn:ive Levelll Application I'Facilit s I Salem 1 & 2 12/3/20121 IIJD
! :011 KAStatement~ Knowledge of the interrelations between Large Break LOCA and the following:
Pumps
-explanation Oii55.41.b(8) Question stem describes LBLOCA with power. With the RCS at 35 psig, all ECCS pumps will be injecting at their
,Answers: . ' maximum rate. The flow rates used are: Charging pumps 2x560= 1120gpm(page 23); 81 pumps 2x675= 1350gpm (page 26);
-~ RHR 1x4500= 4500 (page 34); and Containment Spray pump flow of 2x2600=5200. (page 17) So, 1120+1350+4500+5200=
12,170 gpm total. With the initial RWST level of 41.1' equating to 370,000 gallons, and 15,2' level of 150,000, you need to pump in 1')')(\ ('Iii(\ ,~ Th~" 1 Q (\Q ~. ,,',,~. ., f~' ,,4 01 10 '" ** A. "V . II i" +h * . AI'"
Ipump in the entire RWST volume. Qistracter C is the time if CS pump flow is' not included.
Ii ReferenceTi~lt~le---C--C~J i Facility R~ference Numb~r:Reference Section j l_~~~ :R~
II~T~nk Capacity D~ta II S2.0P-TM.ZZ-0002 II II 11 8 i~E==c=cs===Le==ss=o=n=PI==an========,I!i~N~o:::;=50~5~EC~C~s;OO;-0;7;;;;;===~:::::;=:;.;!lf~~~~~~~~~~~~~1123,26,34 11f=7=:::;
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,LO.Number.
ILOCA01E003
- MateriaIRequire(! for Examination . . I ! RO 3 S2.0F'-TM,ZZ*0002, Tank Capacity Data. Page 28 of 34, RWST Tank Capacity
'I 19ues~i911 Source:. I !Facility Exam Bank I 'Que~~()-nnnOdificatl~~M~thOd:'1 Editorially Modified IIUsed Ourl!"!;! Traihirlgpro~rariJ' i0 Question Source Comments!
~~~~~~~~~~========~============~.
Vision Q48704. Modified correct answer from 18 to 19 minutes to reduce probability of candidate not choosing correct answer because it was LESS than calculated time. Also modified stem to say wilat is the CLOSEST time.
Given the following conditions:
- Unit 2 is operating at 30% power, steady state.
- OHA 0-29, 22 RCP BKR OPEN/FLO LO is received.
- All 22 loop RC flows are 85% and dropping.
- The red START bezel for 22 RCP is illuminated.
- The reactor has NOT tripped.
Which of the following identifies what has occurred?
122 RCP shaft has sheared.
The 22RC9, RC FLOW common low press tap isolation valve has developed a leak.
I[~nswer: I I ~m~ ~ ~~U I c Comprehension ! FaCiiitYll Salem 1 & 2 I [ExamDate: II 12/3/201211
~ I000015K210 ~~§[~ROValue: ,1J]'~~JliELJ IRQ Group::D ~RQ GrollP~U
[§~em/E;'oluti()n Tit!;] IReactor Coolant Pump Malfunctions c:..=..~,:==:':::"::':C'..J
- Knowled e of the interrelations between Reactor Coolant Pum Malfunctions and the followin :
RCP indicators and controls I fExplanationoji 55.41.b(3,7} With a RCP shaft shea~ there is no event that would cause the RCP breaker to open. For this reason, that is why the Answers: I START bezel will still be illuminated, even though loop flows are all dropping, Distracter A is incorrect because between 10%(P-1 0)
,-~~ and 36%(P-8), 1/4 RCS loop 10 flow will NOT cause a Rx trip, the coincidence is 214. Oistracter D is incorrect because there are 3 low pressure flow taps, and 1 common high pressure flow tap. Distracter B is incorrect because a seized RCP shaft would cause its Iwci01d cause that loop flow to drop, even while the bezel indication showed the breaker is still closed.
~--""-~R~e-f-erence Title ""', ! --FaCility Reference Nun:lber! ~rence Section ,1 ~age No.1 m Revision!
F=~======~==========~
I Overhead Annunciators Window D II S2.0P-AR.ZZ-0004 II II 11 23
~I==============~iI~==========~I.F!======~II I~I~
II II II II II-~
I RCPUMPE008 I IRCPUMPE009 I I I
!Matei"ia[R~qui~~d for Examination J I
[ II if" I Source:,j Facility Exam Bank I~~estiOI1 M()dificatio~' Methoa: ] Direct From Source IfUsed During Trainilig Program !D Question Sour5~Commen,~ IVision Q70312 last NRC Exam usage 4 exams ago (1212006)
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Given the following conditions:
- Unit 1 is operating at 100% power.
- 13 Charging pump is in service.
- Normal letdown is in service.
- The 1A 4KV to 460V bus feeder breaker opens, deenergizing the 1A 460/230V bus.
With NO operator action, which of the followin identifies a conse uence, if an ,of this event?
VCT level will be rising at - 1% per minute.
PZR level will be lowering at - 1% per minute.
VCT level will be lowering at - 4% per minute.
~ I000022K1 03 I~I 1;=s=ec=t=io=n::::I,::!E=P=E==-I
',-;:A=K=1.=03==::::'.;.::,R=o=v:::a:::!u=e=::,::H=3=.0:;:..1=S=R=o=v=a=Ii=Je=: =3=.4:.. ,:::f(=()=.*c;=r::ciu=p=:I~1=1:..1i=s=f(=()=G=r=ou::p:::.:',::1=1:..1-.;
iLoss of Reactor Coolant 1\..;;1a..;.;k..;.eu.;;.:p=--_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _--l ',--,0_22_ _
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~KA~:Cl~"~~~.:c:. Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup:
Relationship between charging flow and PZR level
- Explal1atfcm of I 5S.41.b(7) Normal power operation has the Positive Displacement charging pump in service (13), which is powered from 1A 460 volt
'Answers: I bus. The two centrifugal charging pumps, (11 and 12) are powered from Band C 4KV buses respectively. The thumbrule for PZR level at NOT is 75 gallons per % of level.(Page 15 of Lesson Plan) When the 1A 460 volt bus is deenergized, the breaker for the 13 charging pump does NOT trip, it does NOT have a UV trip. Therefore, the interlock for automatically closing the 3 letdown orifice I 1""I"llrm "",I,,,,,,, i<:: ni"\t "",tid;o,; {"II ,
- _.no omn i"\non
, ,\ I ol,;n' <In rom",i,..,,, In <::on/i"o ",7" ... nm ,!hl"h Ie "Arm,,1 "t power letdown flow. This will cause PZR level to lower at 1% per minute. PZR level would remain if it is thought that a charging pump remains in operation because of not knowing correct power supplies. The VCT rule of thumb is 20 gallons per %
level. With letdown flow still enterill~l the VCT at 75 gpm and no charging pump taking suction and pumping from VCT, VCT level will be RISING at -4% per minute, not lowering. The 1% VCI distracter is if there is confusion about which (VCT or PZR) will be changing at 1% per minute.
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" ". Reference Title* '" . , FaCintyReference Number iReferenceSection I,Paae N():If(e~
ILoss of Charging 1[§IOP-AB.CVC-0001 11~=====::::;.Il=1====,IIl==:::;1 I Pressurizer and PRT Lesson Plan 1§S05PZRPRT06 ~j II 11 6 1
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1 - - - - - - -__'1 _ _ _ _ _ _----lIJ II 11_---'1
[i...O:-N-u-m-ber!' ." i lABCVC1E001 1-
Given the following conditions:
- J
- UnlV'initiates a Safety Injection from 100% power due to a LOCA.
- 15 minutes after the SI is initiated, containment pressure peaks at 5 psig.
Which of the following identifies an automatic response which has occurred for Com 2CC215 and 2CC113, Excess Letdown Heat Exchanger CCW isolation valves, received a close signal to ensure this non-essential containment flow path is isolated.
21 and 22CC16, RHR HX CCW isolation valves, received an open signal to ensure that long term cooling of the RCS is in service when the swap to Cold Leg Recirc is required.
2CC215 and 2CC113, Excess Letdown Heat Exchanger CCW isolation valves, received a close signal to ensure ALL CCW supply and return from the containment is isolated.
21 and 22CC16, RHR HX CCW isolation valves, received an open signal to ensure that RHR pumps do not overheat if the RCS remains above the shutoff head of the RHR pumps.
IMemory -I Facility: II Salem 1 & 2 I !ExamDate: j
~ValueJ I 3.6] SRO vai~e:-I 3.91 [Section:] I~ -RO Group; I 11 isROGroup: II 1J 55.41.b(7,8) The SI signal sends a close signal to the CC215 and CC113, Excess Letdown HX isolation valves, as they are Containment Phase A isolation valves. The purpose of closing Phase A isolation valves is to ensure all non-essential containment
' - - - - - - - - - ' 1 penetrations are isolated on a SI. B is incorrect because CC16s do not receive an open signal until the ARM PB is depressed and the RWST level is 15.2' and manual alignment is required to place ECCS in CLR. C is incorrect because ALL CCW supply and
~.'- ---~!!r:e~~~-= ______ ~_u. ~ . ::~J '_nn.Facility Referen~e-NUril~ Reference-SeCtiOrli I Page No. .
IComponent COOling Lesson Plan II NOS05CCWOOO-07 II I1=14::::0==; ~=~
I IC II II ~~
I IC I] 11_--,
I Used During Trai
Salem Unit 1 is operating at 100% power when the PZR Master Pressure Controller demand fails high.
How will this failure affect PZR pressure control components in AUTO?
PZR B/U heaters will be BOTH PZR PORV's will be I~JI 000027A101 !~1~J !ROvalue:II~[~~~[~ I~ l!32~~llIi[JJ~OGrOuP:J U
_SyStemfi:vcilutionTItlel
.. ~
IPressurizer Pressure Control Malfunction I 027 i KA Statement: : Ability to operate and lor monitor the following as they apply to Pressurizer Pressure Control Malfunction: I PZR heaters, sprays, and PORVs I Explal1ation of. 55.41.b(7) Master Pressure controller scale runs from 0-100% demand. With MPC demand failing high, the output of the MPC calls
-AnswerS: i for maximum spray, and no backup heaters. The PZR PORVs are controlled independently of the MPC demand from actual PZR
~._ _............ _ _J pressure channels 1-4. Since actual PZR pressure is not high, PORVs have no open demand signal.
_ _ _~._ *** ~ _ _ m . _ _ _ . _ , _ _ _ _ _ **
Reference Title . . . .~ ~--FaCility Reference~.':I.~~~~=:J [Reference' ~.e.~ti0ll~-*~ I PageNo:j i Re~isio~l IPressurizer Pressure and Level Control 11221060 II II 117 I IPressurizer Power Relief Valves 11231357 II II 1114 I
)1 I[ ~I II II I
....~.-I
'L.O. Number .J
'ABPZR1 E001 I
Material Required for Examination I II IQuestion Source:
_J IFacility Exam Bank I[Question Modification Method: Significantly Modified IIU~El~ DuringTnliningProgrclll1]
'11 D
~e$tion Source~omrnents"' IVision Q50570. Changed from MPC failing low to failing high, which changes correct answer to one of the
---................ ' distracters.
I i ICorln"!:'lIt
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Given the following conditions:
- Unit 1 experienced an ATWT in MODE 1 where both Train "A" and "B" Reactor Trip Breakers (RTBs) failed 10 open.
The reactor was tripped when the pressurizer heater buses were deenergized.
- The RTBs remain shut.
An momentary inadvertent safety injection signal was generated and has cleared.
Which of the following describes the impact of depressing the Train A and Train B RESET SI pushbuttons on 2CC1 when performing Safeguards Reset Actions?
reset, and Auto SI will NOT be blocked.
I NOT reset, and Auto SI will be blocked.
NOT reset, and Auto 51 will NOT be blocked.
~1000029K206
[$~teOl~E;OIutiOl-lTItl~ IAnticipated Transient Without Scram
=-.. . . . . --.
,'KAS .. - - : I 1 Knowi ed1Qe 0 f the 'InterreIatlons
. tatement:
~
' betw een A"ntlcleated T ranslent W"Ithout 5 cram and teo h f II OWIOS:
'~-'.'-'~
Breakers, relays, and disconnects I iExplanation of I 55.41.b(7) The SI signal can be reset as shown on 221057 grid F-2 AND box and downstream LATCH-RESET. This shows the 51
.Answers: can be reset if 2 conditions are present. 1. Manually pushing the reset pb, ( MANUAL SI RESET AND BLOCK), and the 1-2 minute
~--~-~-~.-
TD has timed out after the SI signal was generated. 2. The LATCH-RESET button is reset. Right next to that AND box is another AND box, whose purpose is to block a second SI after the Rx has been tripped. Since the Rx has not tripped, there is no output f.,,~ ,10.;". A",n h~ "-nrl
. ..11.
",,... d~n*
- 10. * ,n,," ,', 'h. ,""0 h. 'h rl~h' ,f' Tho ( 1 '
- 'hi h III n ,rI,
. l' out of this box, which is one of 2 inputs to the AND box to its right. This AND box needs 2 signals to produce an output (Auto SI).
The second input into this AND box is a safety injection from any of the 4 auto SI signals above it. Hi Steamline Flow with 10 steamline pressure or 10-10 Tavg, High Sleamline Differential pressure, PZR low pressure, or Containment hi pressure.
--~.===~~':~!':l1ceT~I.j!.. ~~' ...*....*.,:=J., .-~aci!ity-~efere:"-c~~~Ol'be~_=:J ~~fere~ce Se~tiol1_ ...~ !PCl.~~~.c'J ~~~i~i~.nl I RPS Safeguards Actuation Signals 11221057 II
Ii II ! 122 I I I[ .-11 II II i!
I Ie .-11 II II I Number
-'-'~" ...' - -
IRXPROTE027 1-----'
II antly. Modified Vision Q44139. Modified inadvertent SI signal from remaining present to being clear This changes correct answer to one of the distracters.
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!QuestionTopic.
IRO 9 Given the following conditions:
- Unit 2 will be performing a controlled shutdown and cooldown from 100% power due to a 5 gpm tube leak on 22 SG lAW S2.0P-AB.SG-0001, Steam Generator Tube Leak.
- After completing the immediate Actions of EOP-TRIP-1, Reactor Trip or Safety Injection, following the Rx trip, the RO reports that control rod 2D2 is stuck in the fully withdrawn position.
.""",.,*..,,.. identifies the r"nt"lr' to the stuck rod?
Initiate a rapid boration for 35 minutes during performance of EOP-TRIP-2.
Initiate a rapid boration for 35 minutes in S2.0P-AB.SG-0001 after exiting the TRIP series procedures.
No actions are required for a single stuck rod I:>ecause SDM for the cooldown to 503 degrees is adequate.
No actions are required for a single stuck rod until the Auto SI Block is performed during ReS depressurization to 1900 psig.
IKA:11000038G408 1~~,~~~[}]"~o~[I8!~~ii~I~~~u~§~U
~~tem/Ev?lu:!i9_n Tl!I~ ISteam Generator Tube Rupture I ~3~_._i KA Statement: 1 .--~-~-'
Knowledge of how abnormal operating procedures are used in conjunction with EOPs. I "exPlanation I .Answers:
j 5S.41.b(10) Step 3.26.H in AB.SG states to trip the turbine, then trip the Rx at 20% power during the shutdown. The 5 gpm size of the leak will allow for a controlled shutdown, and will also allow the crew to transition to TRIP-2. There are no SGTR diagnostic I - - ---~ steps in TRIP-2 that would cause a transition to SGTR-1. AB.SG would be re-entered at step 3.27 following exit of TRIP-2, and 3.28 directs rapid boralion for each stuck rod for 35 minutes. The rapid boralion will be initiated before any depressurization starts i.., 'l 'JQ 'A tho";i .~,..tor ' , - l , *..,,,,,,,,, .. h.,ti"n* .
I ' "( :"
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Reference Title '-""':=J
.,.,"" , !---.-..,Facility Reference Number "'."'.
~. ........ _ .. " """",,' ......
i
[Reference Section ",f Page No.' !Revi.s~o";
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1Steam Generator Tube Leak 1@:2,OP-AB,SG-0001 II 1122 11 28 1 IReactor Trip Response 112-EOP-TRIP-2 II II 11 28 I I II II II II I
~M=a~t~e~=*_="",=",~~""~q~U~!~~ed==fo~r=E=x=a=m=in=a=t=i~='~=::"='~l==~====~==~~==========~================~==~~==~~========~==U I_Q_ue_""s_t_iO_"c;SOU_fc_e:J IFacility Exam Bank I~~_tl_Qn_'*_M_Od_Ifi_,c_-a~io_n_M_e_~_~1_E.;...d_ito_n.;...*a_IIY:...-M.;...O.;...d_jfi.;...e.;...d_--l1
- . ~se~Duiing T~ajrtlrlg progranlJ Questlo~ sou~ii1~_ent~i I Vision Q85037
~~$ti~fo~ i----------------------.--------------------------------------------------------------~
Given the following conditions:
- Unit 2 is operating at 100% power, MOL.
- The Condensate Polisher is in service -full flow.
- 21 SGFP trips.
- The CRS enters S2.0P-AB.CN-0001, Main Feedwater I Condensate System Abnormality.
- The Rx does NOT trip.
WhiCh of the followi is an UNEXPECTED alarm if it is locked in 2 minutes after 21 SGFP ?
OHA G-44, COND POL TRBL.
Console Alarm RC PRESS DEVIATION HI.
Console Alarm RC LOOPS TAVG-TREF DEVIATION.
tCognitive Level
~-*~~**-~
~""".-------.~---.----
[!<AJ I000054G446 ISyst~m/EVoIU~ion .!~I.!. I_L_o_Ss_o_f_M_ai_n_F_e_ed_wat_e_r_______________________________________
.... __ ~I05~==n*l r
iKA -- " - '.-.---,
~tatemE!!ll:j .--_______________._______________________________________________-...,.
AbUi to verif that the alarms are consistent with the plant conditions.
Explanation of.' 55.41.b(5)B is incorrect because the condensate polisher trouble alarm will be in due to the CN108s (auto open on a SGFP trip)
[Answers: i AND the CN109 being open at the same time (polisher in service). D is incorrect because RC loops Tavg-Tref deviation will be L.....- ***_ _ _ ~. __ expected as rods are driving in due to the turbine run back to 65%. C is correct because the RC pressure deviation would not be expected, since the setpoint (+75 psig deviation)equates to when the spray valves are full open. The spray valves should be shut alarm since it receives input from the EHC Control and Status computer, which will have a Loss of Feed pump Runback alarm in ..
F===================~F=:============~F=======~F===~1~26==~ 9 1
I_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~I ________________~I------------IF===~I--~
IABCN01 E005 .---J I I I~Mate.~ial Requir.El~ for Examination ~ I II iQueS!iO~ source:] IFacility Exam Bank i [Que.siio.n Modification Method: . II Editorially Modified I[Used During Training Program I 0
~==~~=~~~~~============~==~~~1 i~estiorisou;'ce commeniSllvision Q113267. Removed procedure transition part of question to make RO level (alarm not expected)vs. SRO
---.---.---~-~ level (unexpected alarm AND what procedure addresses unexpected alarm.)
I
,Cornl...'I1< ....; " . , !
I I i I I I
Given the following condition:
- Unit 2 was operating at 100% power when a loss of all AC power occurred.
- 15 minutes after the power loss, operators have locally started 2B EDG.
Which of the following is an action that is REQUIRED to have been erforrned PRIOR to ener izin 2B 4KV Vital bus, and wh ?
Shed non-essential DC loads to extend the time the Vital Instrument Inverters can power their AC loads.
Initiate and reset SI to prevent the auto start of a centrifugal charging pump and possible thermal shock to the RCP seals.
Deenergize ALL SECs and depress stop PBs for SEC actuated components to prevent overloading the 2B 4KV vital bus.
Start the Station Blackout Compressor to provide air for operation of 21*24AF11, AUX FEED-S/G LEVEL CONTROL VLVS, to prevent over the SGs when 22 AFW starts.
~l 000055A203 I~-= ~~~~~~G3[~~I'lJI~RO GrouP:.U[~~(iGrOuP:1U
~~t~t~ IStation Blackout I 1055
~ Statement:] ~~~~~~~~~~~~~~~~~~~~~~~~~--------------------------------~
55.41.b( 1O} The Continuous Action Step for energizing a den energized vital bus with an EDG comes AFTER the step to deenergize all SEC*s. The Bases Document states on page 15 that the reason to deenergize the SECs and depress the Stop PB for all SEC controlled safety related loads is to prevent the bus from overloading. It additionally states that a further reason is to prevent charging pump automatic start and possible thermal shock to the RCP seals. SI is initiated at Step 21 NOT to prevent a charging restored. Non essential DC loads are shed at Step 35 to extend the batteries power capability. The SBO is started as part of Blackout Coping Actions in AttachmEmt 2 Part A of AS.LOOP-i. All the distracters are actions which will be taken during an extended loss of all AC power, but the correct answer is the only one that is required to be performed AND has the correct reason for doing it prior to power restoration. D will be performed, but it is NOT the correct reason, and is required within 60 minutes of Blackout. A and S will be performed, but are not required to be erformed rior to power restoration.
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jQi'festion Topic I RO 12 I Given the following conditions:
- Unit 2 is in MODE 4.
- RCS pressure is 290 psig.
- RHR HX inlet temperature is 270 0 F.
- 21 RHR pump is in service in shutdown cooling.
- 22 RHR loop is aligned for ECCS.
- A loss of all off-site power occurs.
Which of the following identifies why S2.OP-AB.LOOP-0001, Loss of Off-Site Power directs operators to initiate S2.0P-AB.RHR-0001, Loss of RHR?
~~.I The 2A SEC trips 21 RHR pump and does not restart it when 2A EDG connects to 2A vital bus.
I The SEC's trip all running CCW pumps, and they do not restart when the EDGs connect to their respective vital buses.
S2.0P-AB.LOOP-0001 does not consider the initial plant conditions, and always directs initiation of S2.0P-AB.RHR-0001 regardless of Iwhether or not RHR is in operation.
~ IThe 22RH18, RHR HX Flow Control Valve, fails shut, and the 2RH20 RHR HX Bypass Flow Control Valve fails open. Action is contained in S2.OP-AB.RHR-0001 to re-establish positive control of RHR HX flow.
Answer i ~ rexami.evel] ~ ~~nitive._~evE!1j IComprehension I :FacilitY.j I Salem 1 & 2 I :ExamDate: II 12/3/20121 jKA:!1000056K302 liAK3.02~~~aluel~SR- ........:J!~ ~O~ IU 111* l]
- systemiEVohi~~1 1Loss of Off-Site Power 1056
!KA~!~tE!rnentJ Knowledge of the reasons for the following responses as they apply to Loss of Off-Site Power:
Actions contained in EOP for loss of offsite power IExplanationOfl 55.41.b(7,8) The loss of off-site power with NO Safety Injection signal is a SEC MODE" actuation, Blackout. All 3 EDG's will start, IAnswers; J the SEC will strip all loads of it's vital bus, shut the EDG output breaker and sequence on BLACKOUT loads. The RHR pumps are NOT blackout loads and will not be started. AB.LOOP-1 asks, at step 3.8, if a RHR pump was running in SOC mode. If the answer is yes, it directs initiation of AB.RHR since the LOOP will result as described above. The CCW pumps WILL be started by their
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- L.O. Number I
~- - ................... -~----
ABLOP1 E002
~aterial Required for Examinati0l'!.j I II I~~c!~~~~m~ource; --II New IIQuestion~(:)~i!i~ation Metho<tl I!used During Tr(iining Program ~ D Que~~~o~.Source Comments I I
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Given the following conditions:
Salem Unit 1 is in MODE 2 performing a startup by control rods lAW S1.0P-IO.ZZ-0003, Minimum Load to Hot Standby.
Rx power is stable at 4%.
- Vital Instrument Bus 1D inverter output breaker trips and deenergizes 1D 115VAC Vital Instrument Bus.
One minute after the loss of 1D VIB, which of the following contains the indication(s) that will be illuminated on Reactor Status Panel 1RP4, with NO I"In<',::oI,nr action?
Blue Over Power Rod Stop Manual Bypass for CH IV lamp.
Yellow High Flux PRNI CH IV for BOTH High Power and Low Power.
- Syst~Oiuti~e= ILoss of Vital AC Instrument Bus
~ Statell!.~~~ Abilit to determine and inter ret the following as they apply to Loss of Vital AC Instrument Bus:
RPS anel alarm annunciators and trip indicators
'EX.planauon of] 55.41.b(7) A is incorrect because the reactor has no trip demand from the loss of D VIB. B is incorrect because the CH IV nswers: J indication is associated with "G" 4KV RCP group bus. 4 group buses are H,E,F,G from 11, 12, 13, and 14 RCPs. C is incorrect
.-...~~ because the over power block bypass must be manually aligned. D is correct because the CH IV for BOTH the High power hi flux and low power high flux will be illuminated.
L______~!!'!~~le~._ _ ~._~ ~~~iity Referenc:eNUnl~- iReferenceSectioll.... 1i Page~~~ ~~n~
I Loss of 1D 115 Vital Instrument bus IfEOP-AB.115-0004 .---JI II 1112 I~=======::::;1[- .---JI II I ~I~
I_ _ _ _ _ _ _---'IC II II 11_---'
!=:~~~ . ~_i IAB1151E001 IMateri~1 Required for Examination j I II
~sti~n50urce(J I Facility Exam Bank I~stiO~Modificati~nMethod: ] Direct From Source IlUsed Dui'ing Training Pr?gram- [J
§'uestion Source Comments IVision Q133676
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I- Salem Unit 2 is operating at 100% power.
- All Station Air Compressors trip and none can be restarted.
i-The Unit 2 ECAC does not start, and cannot be started.
- The Unit 1 ECAC starts and trips after 5 minutes.
Which of the following identifies an action taken on Salem Unit 2 lAW S2.0P-AB.CA-0001, Loss of Control Air, and why?
~ IAn operator is dispatched to locally shut the 2DR6, AFWST M/U Valve. This is to prevent over flowing the AFWST and causing a spill of 1hydrogen peroxide to the storm drain system. ,
.1 The Rx is tripped when EITHER Control Air header lowers to <80 psig. This is to prevent an automatic trip on 10-10 SG NR level when the lBF19s associated with that header start to drift shut.
All Radwaste releases in progress are terminated. This ensures that during a gradual depressurization of the Control Air system a release is
\ not in progress when the dilution medium flowrate may be changing.
~ IAn operator is dispatched to manually control 23 AFW pump speed which is running at the high speed stop. This ensures the pump does not
~ become steam bound due to the 21-24AF11 valves failing shut with only limited recirc flow provided.
'An~er' ~ [Ex~m Level ~c()g~iii~e leve'-] IMemory -.J ~acility:ll Salem 1 & 2 1 rE~amDate: : I 12/3/20121
[~I 000065G314 I~m_._~ ~~~y~l2:.:J~~!J~~~?ti~~I~ [~to~'D~~c;r~D IK. C
@;Y!l!erri/EVoiu~rTill!J ILoss of Instrument Air 1065=-~-
KA Statement*"
I~_.~
i Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. J
[ExPlanation oil 55.41.b(10) A is incorrect because while the 2DR6 will be operated locally, the concern is the overflow of water with hydrazine in it, AnsVt.'!r.s_:___ not hydrogen peroxide. C is correct because the bases document says that on page 8 of 12. B is incorrect because BOTH CA header pressures have to be below 80 psig before the rx is directed to be tripped, but the reason is correct. D is incorrect because
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the action is correct, but the reason is wrong. The AF11s fail open, and pump runout is a concern with the speed failed at the high
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Reference Title ~---...- ..' L-FaCil!ty~eference N~mber=:J iR~ference se~~ionJ iPaSJe No*IIR&v'iSiQill ILoss of Control Air 1@*OP-AB.CA-0001 II II r 117 I I II .---JI II II 1 I II .---JI iIII II I
!L.O. Number IABCA01 E002
~aterial Requinld for Examillation II II l§~estio~Sour~~ IFacility Exam Bank IiQuestio~ Mociificatio~Method:il Editorially Modified IIUsed During Tr~inlng Program CI
~estiofl Source-comments] IVision ~41379 modified from the release valves a~e shut bec~use ... to v.:hat do you ha--:e to do (termi~ate any
.... --~~- ..... --~--....... release In progress) and why. The why was the ongrnal questions 4 chOices. Also modified per tech meal review I I~ ..........""... . ......
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- >.ROSkvscraJler,I.SROSkv$ciap~r,1 ." >RO svSt!l~voluiiOnU$t iQuestion TO~ IRO 15 Given the following conditions
- Operators are performing actions in 2-EOP-FRCC-1, Response to Inadequate Core Cooling.
- With no other RCPs in service, 23 RCP has been started lAW direction in FRCC-1 and Rx core temperature is lowering.
- 23 RCP was the only RCP able to be started.
Which of the following identifies why 23 RCP would b..:.e..:;s..:.;to;.,:pc=ed.:...;.;.IA..;..W;...;.....;F..;..R..;..C;;...C;;...-..;..1..;..?_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _---'
~ RVLlS level has risen to >57% which shows that the fuel is covered and injection flow is present.
[§J ALL SG NR levels have lowered <9% which indicates insufficient heat transfer will be available in any RCS loop.
~ At least two RCS Thots have lowered to <35()CF which indicates the core is cool and RCP forced circulation is no longer required.
~ 23 RCP #1 seal DIP has lowered to less than 250 psid which is less than the minimum required to prevent mechanical damage to RCP.
-Answer I ~ I Exam Levell ~ ~tive ~ IMemory ~ tacility: ISalem 1 & 2 I I ExamDate: ' I 12/3/20121 iKA:11000074K304 1 ~~~~__ ~RO Value: *[~ ~RO Value: I @ [Section: II~ I RO Group: U :SRO Group:1 U D iSyst~r:':l!EV~~ll~~~Titi~ IInadequate Core Cooling ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ---1
.KA Statement: I Knowledge of the reasons for the followin£1 responses as the~ appll to Inadeguate Core Cooling: I Tripping RCPs I
- Explanation o~ 55.41.b(10)The step to isolate the ECCS accumulators is 28.1, just prior to stopping RCPs at step 29. The ECCS accumulators are iAnswers
- isolated after intermittent RHR flow has been verified, since this means they have discharged based on RHR discharge pressure
- _ _ _ .J capacity. However, SI or Charging system flow is not checked until AFTER the running RCP is stopped in step 29 when RCS Thots (at least 2) are <350 c F. There is no concem for RCP damage based on seal DIP. As discussed at Step 23, normal conditions for I RrD . <>ro rlo,,;rorl hIlt n"tJeCllited Ibe OOll' tbioc th<>t ,,;11 nro"ont RrD c+<>r+ ;" n" c::r:: I\IR 10' ,01 1<00/_\ h<>"orl "n n"tont;<>1 1creep failure of the high temperature SG tubes, but LOSS of SGNR level does not require stopping the RepS. 'RVLlS leve'l must be
>57% at step 31and be combined with at least 2 RCS Thots <350 and all RCP's already stopped to exit FRCC-1 to LOCA-1.
I Reference Title i [- Facility Reier~nc~ Number---] ~~!~nce-Section ---] [§~e-~o.~ .Re~iS.~~
IInadequate Core Cooling I[EOP-FRCC-1 il II 1122 1 I IL II II II I I IL II II II I
[Lo:NUmbe~-_=:--=~
I FRCCOOE006 IFRCCOOE002 1-----'
~aterial Required for Examination I IJ l~uestion Source: II New 1 ~estion Modification Meth~1 1!Used During Training Prograr.!ll D fCiuestionsource Comments1 I I lComment I I I I I I I
fR(fiIJ~fi~t,I *. SR()S~r1If~~1 5ROS~te!ljEvpluti~nL~f ****I_tp;SVstemlEyOlutioriI~f*
.IRO 16 With the Unit 2 Rx operating at 100% power, which of the following radiation monitors would be the FIRST to provide indication that a nuclear fuel rod had developed a substantial leak?
130' Elevation Area.
tCoglli~,:,~..!-:ve'll Comprehension 1 iFaciHty: i ISalem 1 & 2 ExamDate: : I 12/3/20121
~jJI 000076A104 I[~1.04] [ROva~:1[~:SRO va'~~JQ]~?~I~~~t:l~O~o~rou~p:l0
~temlEvOiUtion'Titie~] IHigh Reactor Coolant Activity 1,076!
IKA'" m. Ability to o:Jerate and / or monitor the following as they apply to High Reactor Coolant Activity:
Failed fuel-monitoring equipment lEXplanatio-nof i 55.41 (11) With the plant operating at 100% power, letdown will be in service. Failed fuel would immediately release fission products iAnswers: ! into the RCS. The letdown line would transport these radionuclides and be detected quickly by the 2R31. The 2R2 would respond
'--.------- only if there were a leak in the RCS, since its an area monitor and area radiation levels would take a long time to rise. The R53s detect N 16 gammas in the steam line, which would only be present with primary to secondary leakage. The Charging pumps area
, monitor lAIOllld rise eventl lall\l based on the acthrjl\l risina in Ib" "h",r!'lin" fl! lid hilt it "AI lid ht:> rlil! II",d h" Ih", \lAIi 1m" ",I
- I VCT, and the letdown line ~uld see it first and much more significantlY.
L Reference Title .......J ~ Facility Reference NU~ber= [Reference Section Jl!~Jl.e_~()~J !RevisiOnj IF;R:::a:::di:::at:::io:::n:::M:::on:::it:::or:::in:::9:::s:::ys:::te:::m=====:::::;11 S2.0P-SO.RM-0001 IF;I= = = = = = ; 1 1 I FI3:::7=::::;
ICharging, Letdown, and Seal Injection I§S328-1 II Grid E-5 II I Fls:::6=::::;
1________----111 II II 11_---1 IiMaterial Required for Examination .! I I II IQue$tio:~~ou~!J 1Facility Exam Bank I~fiQn Modification Method: 'I Direct From Source Ilused During Traini ll Program" 9 ID
[Que;iion sourceCommen~11 Vision \,,/1,00 I:;'
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Given the following conditions:
- Unit 2 is operating at 100% power when the operators receive several alarms related to the 500KV grid.
- The Electric System Operator calls Unit 2 and direGts them to perform a rapid load reduction to 875 MW due to grid instability issues.
Which of the following describes how the load reduction will be performed lAW S2.0P-AB.GRID-0001, Abnormal Grid?
At the EHC Console the PO will depress ...
Ithe GO pushputton, and ensure the runback automatically stops at -66% turbine power.
I I
~ SMD #2 RUNBACK and GO PBs, then depress HOLD when Main Generator load lowers < 875 MW.
I ISMD #2 RUNBACK and GO PBs, and ensure the load reduction stops automatically at -66% turbine power. I I
- ~:J EITHER the GO pushbutton OR SMD #2 RUNBACK and GO PBs, then depress HOLD when Main Generator load lowers < 875 MW. I
~~ ~ ~ain Level i ~ [CognItive L.iV~1 JIMemory 'IFacility: ! ISalem 1 & 2 I [Exam'[)ate: , r 12/3/20121 L~ I000077A102 I~1:~,j L~<?_!~r~r~ l~~~~i~[}3~!C~~nl~ ~~~~U~~~f~~ U ~.. 0 IsySt(;;JlIEvOii.it.io;;Titie- IGenerator Voltage and Electric Grid Disturbances 1.~!2 i
~_~!~tement:J Ability to operate andlor monitor the following as they apply to Generator Voltage and Electric Grid Disturbances: I Turbine I generator controls I
,EXPlana.tion".orl 55.41.b(10,7) AB.GRID directs the load reduction directed by the ESO due to grid instability be performed lAW Att 4, which says to IAnswers: J push SMD #2 if the required end point is >765MW AND <942 MW. It says to press HOLD when the MW value is less than or equal
.- to that directed by the ESO. While dipressing the GO PB would work, the procedure says to do it a certain way to ensure consistency amongst the crews (Note on Att 4), so it would be wrong. The MT is normally set up to do a 15% per minute runback to I AAOL h ..hino ,..,rI f_Q1 {'\ ~~,,,o \ "" "hilo if ",,,lrI "0+ ,.,rI ..,hn,,1 "h,,.o* I" "nn. ,,,. ,rI I, Ih n" .rI**", rI, ,,,,.,n'+ ,11, f~~, . +h~
Iway. Candidate needs to know how to initiate load reduction, and how it is directed to be siopped.
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[?uestIOn S9Urce Comments. Vision Q120134. Modified one distracter from SMD 2 or SMD 3 to SMD 2 or GO PB since SMD 3 only appeared in
. -~....... ---......_ _. one choice, and GO only appeared in one choice. I
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~uestionT9pEB I~~~__________________,__________________________________________________________________~
Given the following:
" Unit 1 experienced a Rx trip and Safety Injection from full power due to a RCS leak.
" The control room crew is currently performing 1-EOP-TRIP-3, Safety Injection Termination.
- 11 Charging pump is in service and 12 Charging pump has been secured.
- Charging pump flow through the BIT has been isolated and 1CV68 and 1CV69, Charging Discharge Valves, have been opened.
- The RO fully opens 1CV55, Charging Flow Control Valve, and reports current PZR level is 45% and lowering slowly.
Which of the following describes the action the control room crew should take lAW 1-EOP-TRIP-3?
Re-establish charging flow through the BIT Initiate Safety Injection and return to 1EOP- Tf~IP-1, Rx Trip or Safety Injection, Step 1.
Re-start 12 charging pump to establish PZR level stable or rising, and continue in TRIP-3.
Allow 5-10 minutes for system conditions to stabilize while monitoring PZR level.
~'"IJ ~ ~~C ~ ~1~] iFacility: I ISalem 1 &2 1 fEiam0ate:-1 12/3/20121
~I OOWE02K201 '[~~:1_~i ~JD~ jSRO value~J~I~~~I~ !~O Gr()uii:I[JJ~~~ Grouii:l[]
~temiEvolutiOI'lTitie] lSI Termination IIE02 I iKAStatement:l Knowledge of the interrelations between SI Termination and the following: I Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
I Explanation of i 55.41 (7,8) After stopping one of two centrifugal charging pumps at step 4, charging flow is re-directed from BIT to normal charging Answers:
_ _ _~_ _ ~ *** m~~
! line. If this flowpath cannot maintai'1 stable or rising PZR level, the operator will re-establish BIT flow and go to LOCA-2, Post LOCA Cooldown and depressurization since control of RCS inventory is greater than the capacity of normal charging. The basis document specifically says not to re-start the idled CVCS pump, because that would restore subcooling/PZR level and you would end up back 1:>1 d.,,,if \1"" I')r.A.1 +~ TO,O_,> r. j", ' 11= TRIO~ , . 1'l /"'\//"'c MIGHT be started lAW CAS to start ECCS pumps as necessary since stem is non-speCific about actual PZR level, but continuing in TRIP-3 is not true. Step 7 has operators reestablish charging flow through the BIT and go to LOCA-2. There is no CAS action to initiate SI and go back to TRIP-1 .. There is no direction (nor reason) to allow plant conditions to stabilize, maximum charging flow should act on PZR level in seconds, not minutes to change level.
I Ir:et~lnjeCtl: ~::::~:~~:Title -. fl~~~~~:':ference ~~~~iI~!:~c~~i ~l=~=.~=~=~-=*~=o=:.*I l=~=:=Vi=Si=on=:.i Illl========~IC=======:::=:;I~1===~I I IC._____---'II____ :=1====:.1 ~I !,=I
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II Given the following conditions:
- Unit 2 was operating at 100% power when the RCS developed a SBLOCA.
- 45 minutes after the trip, 2B 4KV vital bus locked out on bus differential.
- Containment pressure is 2.4 psig and lowering very slowiy.
- Operators are now performing actions in 2-EOP-LOCA-2, Post LOCA Cooldown and Depressurization.
Which of the followin contains an alarm, which if received durin performance of LOCA-2, would r uire the associated res onse?
PZR Low Level alarm at 17%. Start ECCS pumps as necessary.
RWST Lo Level console alarm at 15.2 feet. Transfer RCS to Cold Leg Recirculation.
OHA A-6 RMS HI RAD OR TRBL associated with 2R53A, 21 MS Line Rad Monitor. Align SGBD to the Waste Header.
21 SG Program Deviation Setpoint Actual console alarm at 28% NR level. Open 21AF21 SG Level Control Valve to raise level in 21 SG.
[Answer] /U ,'exam Levell m==:J 'cOg;:JitiVe~~ IComprehension.J . ISalem 1 & 2 I .Exa
[~JOOWE03K103 !~~~~I~~m~G]"~I~~~u~~U
~m/Evoluti()n
Title:
I LOCA Cooldown and Depressurization iKA Statem;;rt~l Knowled e of the 0 erational im lications of the following concepts as they apply to LOCA Cool down and Depressurization:
Annunciators and conditions indica~lg signals, and remedial actions associated with the (LOCA Cooldown and Depressurization).
IiExplanatjon of 55.41.b(10,11) RWST 10 level at 15.:~' indicates the need to transfer RCS cooling to cold leg recirculation, as identified by the CAS IlAnswen;;:~ action in LOCA-2. The SG program deviation setpoint is +/-5'Vo, so the alam at 28% would be valid. However, 22 AFW pump has no power, so opening the 21AF21 would have no effect. The R53s are N2 monitors in the Main Steam Lines, and after the Rx is I shutdown do not provide indication of any use. The action to align SGBD is perfonned in SGTR-1. The PZR level at which ECCS
. 0 0 . ,
- Iievel' alarm comes in at 5% 'below program, which would be -22% with the lo~ Tavg expected during a SBLOCA.
_~~~ ~!!:!1~e~~~==== L~~(;ill~~~~~~1 ~~ell~ti~~.=J ~~ [i§'~
IPost LOCA Cooldown and Depressurization 112-EOP-LOCA-2 II II Ili=2=5=~
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I Ii II II II_--J LO. Numbf[==:=J I LOCA02E005 I I I I I
[Material Required for Examillatl<m I II l<:t~estion S~urce~i[ IFacility Exam Bank I§U~$t1011 NlodificaJic::m MethOd:'] Editorially Modified I[gsed DUrlngTraining Progr;;lm'] 0
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I Quest:i~n Source Comments' Vision 0127166. Removed procedure from LOCA outside containment, removed response from opening AF21
....,.~~ " , - , ..) valve as part of distracter.
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Given the following conditions:
- Unit 2 is attempting to identify and isolate a 400 gpm LOCA into the RHR system which occurred while operating at 75% power.
EOP-LOCA-6, LOCA Outside Containment, was entered from 2-EOP-TRIP-1, Rx Trip or Safety Injection.
- The source of the water is back leakage from the 23 cold leg injection line.
- A large leak in the RHR system is located on the piping between 21 and 22RH19s, RHR HX DISCH X-CONN VALVES.
Which of the foliowin com onents, if it failed to respond when directed b LOCA-6, would prevent isolation of the RCS leak outside containment?
L~J I 22SJ49, RHR DISCH TO COLD LEGS.
j22RH19, RHR HX DISCH X-CONN VALVE.
!A~~~ j d J !Ex~~ i~~~J IR I ~ognitive~L~~~ IApplication I [acUity: II Salem 1 & 2 I 'ExamDate: :1
~1 OOWE04A101 ~~[I2i~~~i~I~T~~I~ ;ROGroup:][]~~[]
~/Evolut~_nl1tle -I
~--.~~: ..
- ~ StCiteme.,!
- j Ability to operate and I or monitor the following as they apply to LOCA Outside Containment:
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
~Ianation o~ 55.41.b(7,8, 10) The leakage from 23 Cold leg flows back through the 21SJ49, then through the 21RH19 cross connect to reach the I.~swers~._~ leak. Leak isolation is attempted first by ensuring closed the RCS-RHR suction isolation valves RH1 and RH2. Then BOTH the RH19s are shut. Since the leakage from the RCS has to flow through the 21 RH19 to reach the leak, and it is a normally open valve.
Its failure to reposition when directeKl would prevent leak isolation. The SJ69 is closed after the leak is isolated as above. The o~o t,,,in (""nrH~"to "I",,, h"", ,,, "n, ",hi,..h o~o t,,,in foo~'" "hi,..h ,"" I~ I",.,,,, "hon ""t "''''
1 Iconnected ..
??<:: lAO i", ,..." tho ,..."""",ao 21 RHR feeds 21 and 23 cold legs, while 22 RHR train feeds 22 and 24
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HS-1, Response to Loss of Secondary Heat Sink, Step 3 asks, "Is RCS pressure greater than ANY intact or ruptured SG pressure".
Which of the f"IIf"lIAJinn statements is correct if the answers NO?
IMMEDIATELY go to Step 23, Bleed and Feed Initiation, since there is no decay heat removal occurring through the SGs.
Return to Procedure in effect. Attempts to establish a secondary heat sink would be ineffective at reducing RCS temperature since SG pressure is higher than RCS pressure.
The RCS has experienced a LOCA large enough that a secondary heat sink is NOT required because decay flow.
loss of reactor coolant through the LOCA, since a LOOP later in the event could cause a more flow.
1.KA:IOOWEOSK302 ! I~~*2== :Rova'Uel@fSRo-val;;T1[ITI[~I~ ~~?~[JJ~~:j[JJ
~nlIEVOluUo~ ILoss of Secondary Heat Sink IEOS
- KA Statement~ Knowledge of the reasons for the f<2!lowing responses as they apply to Loss of Secondary Heat Sink:
Normal, abnormal and emergency operating procedures associated with (Loss of Secondary Heat Sink).
IExpliilllation ofl 55,41.b(10, 14) The reason for checking RCS pressure> intact or ruptured SG pressure is to check if there is a need to be worried IAnswers: : about a secondary heat sink. If RCS pressure is below SG pressures, then a LOCA of sufficient size is present, and break flow will
~--....----~ be removing decay heat, along with ECCS injection. Distracter A is incorrect because the criteria for going to bleed and feed is SG WR level. Distracter B is Incorrect beciiuse a secondary heat sink could actually be established, and could reduce RCS f"...,,,, *~h"o h. rf .. ",,,I,,,, "f"""",, f.",..., fh, <:r-,,' . n 1<, . If I", """ "" ('A<: "f ~Oll<: h ..f I", ..""""",,, f". f.I ..."I"" 0(,0", I" ITRIP-10rLOCA-1.
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Given the following conditions:
- Unit 2 has experienced a steam line break inside containment.
- Operators have entered FRTS-1, Response to Imminent Pressurized Thermal Shock.
The soak required by FRTS-1 requires SI to bEl secured. RCPs are started to provide mixing of cold SI and warm reactor coolant water.
Safety Injection flow is a significant contributor to any cold leg temperature decrease or overpressure condition and must be terminated.
RCPs are started to minimize temperature gradient across 5/G tube sheets.
cil1,niti,,..,,,nt contributor to any cold leg temperature decrease or overpressure condition and must be terminated.
of cold 51 and warm reactor coolant water.
- KA:.100WE08K102 1~KT2-~~~~I2+/-i~IJ.~[I§j[~ll~RO~Drs-R~D r$YStemiEvolutio.nntle r IPressurized Thermal Shock
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~KA.§!.atement~ Knowledge of the operational implications of the following concepts as they apply to Pressurized Thermal Shock:
Normal, abnormal and emergency. £p'erating procedures associated with (Pressurized Thermal Shock).
- Explanation of I 55.41.b.(10,8)
Answers: ' A- incorrect - purpose for RCPs is not priority in FRTS-1, soak is not basis for SI. B - incorrect - soak not basis. C - incorrect - SI
~-.--- -- basis correct, RCP basis not accurate. D-correct -page 11-- Reflnence Titre-~.------~i~-~FaCilityReterenCeNUlllber.J 'Reference Section -] ~ge.!'~! [ReV[siOnl 11 Response to Imminent Pressurized Thermal Sh 1[E'OP-FRTS-1 ~18asis Document j 12,13 1125 !
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iMaterial Required for EXa!Dination
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~~~~~~~_.i IFacility Exam Bank I~odi~on Meiho~:_ JDirect From Source Question ~~1".rce§~mmentsi IVision Q73425. Originally on Seabrook 2003 NRC exam.
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Given the following conditions:
- Operators are performing a natural circulation rapid cooldown on Unit 1 lAW 1-EOP-TRIP-5. Natural Circulation Rapid Cooldown Without RVLlS.
- NO RCPs are running or can be started.
- The control room crew has completed the initial RCS cooldown I depressurization to 500°F 11600 pSig.
- The current time is 1300.
Of the following. which one identifies the EARLIEST time RCS Thot temperatures could be reduced below 450°?
Assume the cooldown will start at 1300 and instantaneousl be at the maximum rate allowed.
1401 1 .
1501 1 .
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- *. amD~terJ'H 1213/20121.
~JI 00WE10A102 I~----:J 'ROVaI~e:lr1}irSR~~rET~6~1 EPE I~Qijroup:I[1]~GrouElU S~temlEvoluticjil~ INatural Circulation with Steam Void in Vessel with/without RVLlS I
~ S~atemenr.j Ability to operate and I or monitor the following as they apply to Natural Circulation with Steam Void in Vessel with/without RVLlS:
IOperatinq behavior characteristics of the facili!i:.
IExplanatio~()f ! 55.41.b(10) RCS temp is at 500 degrees per the stem. The next temp reduction will be to 450° starting at step 9. and the cooldown Answers: __ ' _J is directed to be performed at <100"F per hour. This means 30 minutes of cooldown is required. The 1401 distracter is if the 50 0/hr
~ate of step 7 (initial cooldown to 500°F) is used. The 1316 distracter is if the 2000lhr PZR cooldown limit per T8 3.4.10.2.b is used.
The 1501 distracter is for both continuity of the choices. and if the 25°F per hour rate is used for cooldown.
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......:1 INatural Circulation Rapid Cooldown Without RV 1!1-EOP-TRIP.5 ~I rI 1122 I I II II II II I i
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co"mment~ IVision 0116968. Modified to include procedure name. Added that the C/O to 50QoF has been performed. Added 1 minute to each choice since the stem asks when can be below 450, and 30 minutes of cooldown would only get to It' u.
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J .. 1 i Given the following conditions for Unit 1:
- A reactor trip and SI occurred at 0700 due to a 1500 gpm RCS LOCA.
- RHR system problems have resulted in a loss of recirculation capability.
- Current time is 1300 hours54.167 days <br />7.738 weeks <br />1.781 months <br />.
Conditions present when transitioning to 1-EOP-LOCA-5, Loss of Emergency Recirculation, from 1-EOP-LOCA-1, Loss of Reactor Coolant, due to the loss of recirc capability are:
- RCS subcooling is 10°F.
- All RCPs are secured
- 11 and 12 Charging Pumps are running
- BIT flow - 350 gpm
- RVLlS full range 95%
11 SI Pump flow - 250 gpm 12 SI Pump flow 250 gpm
- Containment pressure 4.1 psig Which of the following identifies the ECCS pumps that should be run following determination of Minimum SI Flow for Decay Heat Removal?
Assume:
Equal flow from each Charging Pump, and each pump will supply half the original total flow if the other is secured.
- Each SI pump flow remains constant if the other 51 pump is secured.
,hf"'f"I,..linl"1 remains between 10°F - 45°F for the lONE Charging pump and ONE SI pump.
lONE Charging pump only.
@:J )ONE SI pump only.
fAlISWe'~l fD" 'EXamiJ~VeJlI R I ~f~v~U IApplication Facility: II Salem 1 & 2 \ Exa~~~§J I 12/3/2012\
KA: II 00WE11A202 I ~A2.2-- .RO Va.lue:.* [8J[SRO value:J@@ection:lI~ L~~~D ~~I 0
'KA Statt'!!"enE! Abilit to determine and interpret the followin as the a pi to Loss of Emer enc Coolant Recirculation:
Adherence to appropriate rocedures and operation within the limitations in the facility's license and amendments.
lEx. Planat.ion Of] 55.41.b(10,8) A 1,500 gpm RCS LOCA will deplete the RWST in 2.3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br />. The transfer to CL Recirc will already have been IAnswers: performed. During step 14 of LOCA-5, charging pumps will be reduced to ONE centrifugal, and SI pumps will be reduced to ONE .
., Starting at Step 19 of LOCA-5, with Fi:CP's secured with <50 degrees subcooling, will use Figure A to determine the ECCS flow required vs. time after trip. 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> equals 360 minutes, which is -225 gpm, but definitely LESS THAN 250 gpm. With the stem lSI pump, however, supplYing 250 gprn will supply sufficient flow.- _.
I*****. . . . . ... . . Reference Title =:=J ~:. ,=aciiity Refereft~e NU~ :Refen;;,;eeSe~i.0'.l :=J~....§J ~~~
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o l~~l3tion Topic~ I.:..R:..;.o::...::.25::...-_ _ _ _ _ _ _ _,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _......J Given the following conditions:
1- Unit 1 has experienced a MSLB at the Main Turbine inlet steam piping.
- All attempts at Main Steamline Isolation have failed.
- Operators have transitioned out of 1-EOP-TRIP-1, Reactor Trip or Safety Injection.
- RCS cooldown rate is 120o/hr.
- RCS pressure is 1300 psig and dropping.
1- Charging system SI flowmeter indicates 290 gpm.
- The RCS cooldown is NOT being controlled.
Which choice identifies an action that must be performed lAW 1-EOP-LOSC-2, Multiple Steam Generator Depressurization, and why?
ITri P all RCP's to reduce heat transfer to the SGs.
I IReduce AFW to minimize cooldown while still keeping the SG tubes wet.
I I
ISend operators to close all BF19's, BF40's, and BF22's to re-establish a secondary pressure boundary in any SG. I iAriswir' ~ [Exam Levell [C] ~gnitive Level "'*1 Application i iFaCiiity:! I Salem 1 & 2 I IExamDate: II 12/3/20121 1~100WE12K101 I,EK~~:,L,_~I~L~~y~!§~~I~I~ll~01~~~~O ~~~~, 0
- SYstem/ElfOlUtion Title! !Uncontrolled Depressurization of all Steam Generators kA Stat!l11entJ Knowledge of the operational implications of the following concepts as they apply to Uncontrolled Depressurization of all Steam Generators:
Components:, capacity, and function of emergency systems. I 1Explaml.tion Of-I 55.41.b{10A)Once out of TRIP-1, no actions other than attempting to close MSLI valve are taken in LOSC-1 prior to going to LOSC
!Answers: , 2. Maintaining >1 E4 Ibm/hr to each S/G keeps tubes from drying out, among other things. Do not trip RCP's because pressure is
. ...,-....-~ dropping due to cooldown, and the reason is wrong. Doesn't matter if it's uncontrolled or not. Distracter 0 is incorrect because we don't close BF22's, but the reason is right. Don't stop RHR pumps because pressure is still dropping, reason is right.
IMultiple Steam Generator Depressurization 111-EOP-LOSC-2 I II 1122 I I I[=-=~~========~I~I======~I~I==~I~I~I l_______--JIC II II 11_-11 fh-*~-~E~-m-b-~- . **-.
I LOSC02E004 1_ _ _ -1 LMaterial Req':llre~ for Examination. I iQuestion Source::1 I Facility Exam Bank I~stion MOdifiCationMetho(t:1 Direct From Source 19lJestiOn~~~_c:.~e6mtl1e:nts! IVision Q59885. Used on "H" RO NRC Exam (2004, 5 NRC exams ago.)
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Qu~stion~f~~ I_R~O~26 _ _________________________________________________________________________________~
Given the following conditions:
- Unit 1 has experienced a LBLOCA.
- The crew is responding lAW the EOP network.
- 3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br /> after the transfer to Cold Leg Recirc has been accomplished, the STA reports a Purple Path exists for containment Environment due to containment sump level being 80%.
Which of the following would assist in validating that a high containment sump level actual! exists?
3 SW pumps in service with SW header pressure 150 psig.
'd.'i5 CFCUs running in low speed with SW flow ;;j: - 1000 gpm each.
~ns\l\l~ Id I ~x~mievelJ IR I !co~n'itiVe~l:~vel-C:; IApplication -I !FacilitY-II Salem 1 & 2 1~ExarnDate~~ I 1213/20121
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~-Statement:j I' -.~. . :
Ability to identify and interpret diverse indications to validate the response of another indication. I IExplanation of 55.41b.(7,9)FRCE-2 checks for possible sources of the excessive sump level in containment. They are: CFCU SW flow, FP to
~swers:~_j containment isolation valve position, CCW Surge Tank level, Demin Water Storage Tank Level, Primary Water Storage Tank Level. PWST of 200,000 gallons is not enough to have raised containment sump level that much. Tank Capacity completely full is
-240,000 gallons.
I ~\A, h ,~rl, n, ,f 1t:::n 'k Hh th", ~\AI,,)ac -hilt frnm +h ~r::f" .. 1\ ~\A' Irl h in identifying SW leak in containment.
- : Ie",
Each CFCU SW flow is normally -1600 gpm, and indicates a 3,000 gpm leak into containment is possible. Fire protection Storage tank level chan!;le would be inadeguate to raise sume level that much .
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~Qu-:stion-TopiCll RO 27 I IWhich of the following identifies the minimum radiation monitor(s) that must be sensing high radiation conditions for either channel of the Subcooling Margin Monitor to automaticall:r: shift to the ADVERSE Mode?
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~ EITHER R44A OR R44B, Containment High Range.
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~ BOTH R44A AND R44B. Containment High Range.
I
~ IEITHER R2, Containment 130', OR R7 In-Core Seal Table.
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~nswer, ~Exam Level i
';;0; 10-:' S"~'bI' Cognitive Lev~ 1Memory I 1 rr'=aCility: i Salem 1 & 2 1 IExam Date: ]1 1213/20121 I
- ~: IOOWE 16A202 li EA2 .2 ~[3.01~RO"alue:]13.3Il~~ction:JI~!RO-GrOup:I[JjsROGrOup: I 21 ~~.43' [J
~ystem/Evolution Ti~ IHigh Containment Radiation lKAstatement:'
iate procedures and operation within the limitations in the facilitis license and amendments.
[EXplanation o.il 55.41.b(11) Either of the Containment high Range monitors reaching 1E5 Rlhrwill automatically place the SMM in ADVERSE lAnswers: ___ J Mode. The other area monitors in containment listed do not input into the SMM.
!Revision!
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!Questlon Topic , I.;..R;,.:O;...2.;.,.8::...-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ----I Given the following conditions:
- Unit 2 is operating at 70% power and stable after a load reduction was completed 10 minutes ago.
- Rod Control is in MANUAL control.
- The highest actual Tave-Tref deviation is 4.0°F.
Which choice identifies the rod speed that would be present initially if the Rod Control Selector Switch were laced in AUTO?
I IS6 spm.
\rAnswerl ~ ~~ ~ [C~JI~!_'::~IIAppliCation l-Facimy: ISalem 1 &2 1 ~~11 12/3/20121
~1001000A101 I rROYaiUe:l[i~isRova'ue:I@[~-~~I~~~~UlsRoGrouplU
~mfEvoh.lti-onTiile I Control Rod Drive System llSAS!a!il!l~~~ Ability to predict and/or monitor changes in parameters associated with operating the Control Rod Drive System controls including: I T-ave. and no-load T-ave I i IExplanation of I 55.41.b(6,7)Rod speed is determined in AUTO by Auct High Tavg vs. Tref (PT-SOS, Turbine steamline Inlet Pressure) .. The AUTO Answers: rod speed program is 8 spm from 1.S-3.0°F deviation. From 3.0-5,0 it ramps up linearly from 8 spm to 72 spm. This correlates to 16 spm per 1/2°F temp change. 8 spm (@3.0) + 32 spm (from 3.0 to 4.0)= 40 spm.
The 56 spm distracter is based upon using an incorrect linear ramp from 1.5 - 5.0 degrees was used.
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Ithrough over S time constants and its effect on rod speed will be zero ten minutes after the load reduction has been stopped. -
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~uesiionTop~ IRO 29 Given the following conditions:
- Unit 1 is operating at 30% power returning from a mid-cycle outage.
- Rod Control is in Manual.
- 14 RCP trips.
Which of the following describes how 14 RC Loop Tavg will be affected 5 minutes later with NO operator action when compared to pre-event Tavg in 14 RC Loop?
14 RC LoopTavQ will be ...
I
~J Lower because backflowfrom the unaffected loops will cause Tc to rise and equal Thot.
IHigher because backflow from the unaffected loops will cause Tc to rise and equal That.
I Lower because less heat transfer will occur in 14 SG due to the loss of forced flow in that loop.
- (i."JHigher because less heat transfer WIll occur in 14 SG due to the loss of forced flow in that loop.
llA.~swer* ~ ~Exa~ Level! ~ ifo~nitive level I Comprehension .JFacility:! I Salem 1 & 2 1 [EXimoat.e:l1 12/3/20121 1003000K503 I ~~:2L~~m_J ~~~~r.2:.!lISRC)VaIUE;: i@~ec:ti()~I~ [ROGroup:IU~6G~D .. :=J iSystem/Evolution i Title IReactor Coolant Pump S';;tem I --
IKA Sfutel!l~nt:J Knowiedge of the operational implications of the followins conceets as thel: aeell: to the Reactor Coolant Pume S:r;:stem: I Effects of RCP shutdown on T-ave., includin!ij the reason for the unreliabilit:r;: of T-ave. in the shutdoWln looe I
.1~XPlanation o~ 55.41.b(3) With Rx power <36%, the reactor will not trip upon a loss of a single RCP. 14 RCP will coast down, and flow will reverse Answers: ........1' in that loop. A differential pressure will exist across a loop in which a RCP is not running due to the head of other RCPs applied to the cold leg side of the vessel. This high pressure (cold leg side) will induce flow in an idle loop in the reverse direction. This means reactor coolant from the cold leg will flow back through the idle RCP and SG. This will lower Tavg in the idle loop. Th and Tc in tho <lffc"tc,; I""n 111 rC'It'"' <lin'; rh" n,inin<lll T" "ill ho nrc"t"r th<ln Th Thc? . "ith tho ThfT" ",n"-::Ilinn <>rc .
'h..,,-"I"""\ Tc will rise >Th in that loop .
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I Reactor Coolant System Lesson Plan II NOS05RCSOOO-09 II I !25 11 09 I IC ~I I II I I II ~I II II I
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I RCPUMPE016 I I I I I III/Ima..~erial Required for Examination I II I""
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~~estion Sou."~=--~o~ments! I I I
I I 1 1 I I
Given the following conditions:
- Unit 2 is operating at 100% power.
- 21 Charging pump is crr 23 Charging pump trips.
- 22 Charging pump cannot be immediately started.
Whi Flow?
VCT level will lower due to the lower sealleakoff flow, and auto makeup to the VCT will be initiated.
be restored within 5 minutes, operators will trip the Rx lAW CAS action in S2.0P-AB.RCP-0001, Reactor Coolant Barrier hElat exchanger will maintain RCP seal temperature and allow plant operation to continue while lAW S2.0P-AB.CVC-0001, Loss of Charging. This will require a the Unit 1 RWST.
~I 003000K602 , i~~~ 'if~~IJ3[~~J[TIj~s.~~I~ '~~~[J]~~1Ip:][J]
iSyStem/EvOlutlonTitiel
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IReactor Coolant Pump System ..
--.--~ ..--
iKAState,rnellt.:j Knowiedge of the of the effect of a loss or malfunction on the followinf;l will have on the Reactor Coolant Pum2 S:r:stem: I RCP seals and seal water su~ I
!Explanation of ' 55.41.b(7) A is incorrect because when all 3 charging pump breakers are open, letdown orifice isolation valves automatically shut, IAn~~~~S~.J so letdown flow is zero. Charging pumps are using no VCT capacity. Seal return will still be going to VCT, so VCT level will be rising. B is incorrect because as AB.RCP directs Rx trip if BOTH seal injection and Thermal Barrier flows are lost, not just one or the other. C is correct because it describes the f10wpath of RCP seal cooling flow when normal seal injection flow has been lost. D I I I" . her",,"'" t. R ('\1('_1 r!lre,* 1;"I"n "" I Inl+ 1 prop /1 '<I It un"lr! ror" .ire", I Inl+ ') ",h"trl"",n ""t" I Ina 1 ",h"trl,.,,,,,, h",,,,,,rI ,.,n higher borated water supplied from Unit 1 R"WST to Unit 2 Res. Trainee may recognize that TS 3.0.3 is present when no charging pumps are available, but correct answer is worded to allow time to attempt to restore charging flow.
[-~~ ~--Retereflce Titl~~-~--c--:J i--- F~cilitYRefer~nce~Number'l fReier!~ce Section '*. --] !Page-NO:! ~ViSi~
ILoss of Charging I§OP-AB.CVC-0001 II II 11 9 I IReactor Coolant Pump Abnormality 1@:2.0P-AB.RCP-0001 II II 1/21 I
,I II II II II I
~~~=~=J IRCPUMPE008 1-----'
iMaterial Required for Examination
- . II
, I?uest!~n so~::~: 11 Facility Exam Bank I[Questlon.MOdi!icatlon Pi1ethod: ] Concept Used I:used_,?uringTrainillg pr()l1r~ D
[Question Source commen~ IQ41806 concept Of_RCS flowing up shaft to supply seals. Added what is the effect part to match KIA.
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- I RO 31 Given the following conditions:
The on-coming shift assumes the watch with Unit 1 operating at 100% power steady state.
- 13 Charging Pump is in service with the Master Flow Controller in Manual.
- RCP seal injection flow is 8 gpm per pump.
- RCP seal leakoff flow is 2.0 gpm per pump.
Halfway through their shift, operators receive console alarm VCT Level Hi-Lo.
A CVCS makeup system failure has caused VCT level to rise to 87.2%.
- VCT pressure has risen to 35 psig.
- CVCS makeup is now isolated.
Which of the following identifies how the current VCT level and pressure has affected the CVCS system when compared to conditions at the beginning of the shift, and how are the operators directed to res ond lAW S1.0P-AR.ZZ-00012, Control Console 1CC2?
has risen. Ensure 1CV35 VCT 3 WAY INLET V is directed to the CVCS HUT.
Letdown flow has risen. Ensure 1CV35 VCT 3 WAY INLET V is Seal Leakoff flows have lowered. Open 1CV243 VCT VENT ISOL Seal Injection flow has lowered. Open 1CV243 VCT VENT rSOL VALVE.
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!Sys~em/Evolutio"nTitie' IChemical and Volume Control System KAS....",...",.. ,:! Ability to (a) predict the impacts of the following on the Chemical and Volume Control System and (b) based on those predictions, use procedures to correct, control, or mitiqate the consequences of those abnormal operation:
High VCT level
'. Explanation I.Answers:
of.l increased 55.41.b{7,10) Stem conditions indicate that something has occurred which has caused VCT level and pressure to rise. The pressure will cause charging flow to rise since the MFC is in manual. D is incorrect because seal injection flow will have risen because of the increased NPSH to the charging pump and resultant higher discharge pressure. Opening the VCT vent is not directed. VCT high pressure alarm occurs at 50 psig, at which time the vent would be open. A is correct because the ARP for high I"" .",1 """" t.... "'n". or, th", r\I?t=.* ';;"0, t",rI tro r\lr'<:; IT it "h", ,lrI h""", trinno,; t" f, ,II rli"",rt .,t A70 h",,,,,, ""'" nf
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same reason as A above. B is incorrect because letdown flow would not rise, it would lower due to the increase backpressure from I the VCT, but the action is correct.
....Reference Title ] L Facility ~tl!tl~tlI'lCEli~iu~m-ber'TI 'Reference Section ~OJ ~~~
~Ic=o=nt=ro=1c::::o=ns::::o=le=1=CC::::2=========,11 S1.0P-AR.ZZ-0012 II II 11 34 I
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@luestion TOP~ I RO 32 I I2CV21 , Letdown Demin Bypass Valve, will automatically reposition to bypass tile CVCS Mixed Bed Demineralizers at the Letdown HX Outlet temperature of .... I
~ I120°F.
I
~ 1 12rF I
i~:l11300F .
I l~ I136°F.
I
~swelll d I ,Exam Levell IR I ICognitive Level : I Memory I Facility: II Salem 1 & 2 I
I :E~amDate:J I 12/3/20121
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~: II 004000K416 I ~4.16 IIRO Value: I f :~ rSRO Value: II 3.01 ,~~D 0
[§tstem/Evolution Title] IChemical and Volume Control System 11 004
~'Statement:J Knowledge of Chemical and Volume Control S:!stem desifiln feature~s) and or interlock(s) which provide for the followinfil: I Temperature at which the temperatul"e control valve automatically diverts flow from the demineralizer to the VCT; reason for this diversion I
!Explanation OJ 55.41.b(5) The 2CV21 will reposition to divert flow from the CVCS demineralizers at 136°F. 130°F is nominal charging temp
'Answers: entering Regen HX. Letdown HX outlet Hi Temp Alarm at 120°F in Control Room (computer point T0145A; TE-130B). Letdown HX Design Cooldown 380°F to 12rF with 95°F CCW inlet at 1000 gpm flow rate for CCW.
I Reference Title__.__............. ___ ~I __._F.~c.i!i!y Re!~r~!'~~.~.I'1l~.r...J IReference Section II Page No.11 Revision!
IPWR Components Lesson Plan Ie II il II I Icvcs Demineralizers-Normal Operations I[I2.0P-SO.CVC-0012 ~I II 11 32 I I Overhead Annunciators Window E II S2.0P-AR.ZZ-0005 II E-41 1155 11 19 I I'f~~~-===
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- Objeetives' ICVCSOOE004 LMaterial Required for Examinati~ I II IQuestion Source: II Facility Exam Bank I~stion Modification Metho~ Concept Used IiUsed During Training Program! 0 e~,=.!!.on Sour~e c~_~men_~ IVision Q39286 Used concept of letdown flow divert at 136°F, made 2 and 2 and also added why it diverts.
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~~~~ti()~~Top}£] IRO 33 I Given the following conditions:
- Unit 2 is in MODE 4.
- RCS Cooldown is in progress.
- 21 RHR Pump and Heat Exchanger are in service to provide shutdown cooling.
- 22 RHR loop is aligned for ECCS.
- CRS directs the RCS cooldown rate be REDUCED.
Of the followin~, which describes how RHR slstem flow will be adjusted lAW S2.0P-SO.RHR-0001, Initiating RHR, to lower the cooldown rate?
ra.-IIThrOttie closed on 21 RH18, RHR Heat Exchanger Flow Control valve, while throttling closed on 2RH20, RHR Heat Exchanger Bypass valve to
- maintain total RHR flow constant.
~ I Throttle open on 21 RH18, RHR Heat Exchanger Flow Control valve, while throttling closed on 2RH20, RHR Heat Exchanger Bypass valve to
- lower total RHR flow.
~j IThrottie open on 21RH18, while throttling open on 2RH20 to raise RHR Heat Exchanger bypass flow.
I
[ ] I Throttle closed on 21 RH18, while throttling open on 2RH20 to maintain total RHR flow constant.
~sw~-...:Il d I I-Exam Level II R I I [Cognitive Leve' "1 Application I ~acility: II Salem 1 & 2 I Exam Date: II 12/3/20121 IKA: !I005000A402 11A4*02 iROValue: II~~SRO Value: lJ 3.11 !Section: il.~.~.~.~ iRO Group:11 111sRo Group: II 11~~:~~1 D
-~~-~.--.-.~ ..
iSystem/Evolutlon Tltlel IResidual Heat Removal System ~--~--
I 1005 J
KA Statem~"r:!!J Ability to manually operate and/or monitor in the control room: I Heat exchanger bypass flow control I Explanation of i 55.41.b(5,8)Throttling closed on the RHR HX outlet valve while throttling the bypass valve open will pass less water through the Answers: .. ~
. RHR heat exchanger, therefore reducing the cooldown rate while maintaining stable total RHR system flow. A is incorrect because throttling closed both the RH18 and RH20 will not maintain flow constant. B & C are incorrect because it would raise the cooldown rate by passing more flow through RHR HX.
~-.-~-
Reference Title j [ Facility Reference Num~ ~rence Section=cJ i"Page~No:l [Revision]
I Initiating RHR I ~2.0P-SO.RHR-0001 ~I II 1127 I I RHR Simplified Drawing I§S332-SIMP ~I II r 12 I I I[ ~I II II I I'-nt:.~---
~~~"~!!:. ____J IRHROOOE004 I I I I ~
~Material Required for Examination II II
~~~!i~-" Source: II Facility Exam Bank I ~stion Modification Meth~d~1 Editorially Modified IlUsed During Trainin.g Pr()gram lD e~~.:'tion So~rce comments] IVision Q77960 co I
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[Question ToPic] IRO 34 I Given the following conditions:
- Unit 2 is in MODE 5.
- BOTH loops of RHR are in service for Shutdown Cooling.
- RCS temperature is 190°F and stable.
- Each loop is supplying 1800 gpm flow.
- 2RH20 is 10% open.
- Conditions to transition to MODE 4 are NOT met.
- The air line supplying the 21 RH18, RHR HX Outlet FCV breaks, and air is lost to 21 RH18.
Which of the following describes the initial effect this air line failure will have?
~ IThere will be no effect on the RHR system.
I
~ 12RH20 will have to be throttled in the open direction to prevent aRCS cooldown.
I
~ 1.22RH18 will have to be throttled in the closed direction to prevent aRCS cooldown.
I
@J 122RH18 will have to be throttled opened to ensure RCS temperature is maintained <200°F.
I
- a:nswer II a I Exam Levell IR I ICognitive Level ! IMemory ~ IFacility: ! ISalem 1 & 2 I Exam Date: iI 12/3/20121 I I KA: : 005000K41 0
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IK4.1O IIROvalue::1 ~~'SRovalue:113.1@ectionJI~iROGrOuP::1 111sROGroUP:11 11 !~. D rsYStem/Evolutlon TltleJ IReSidual Heat Removal System 11 005 IKA Statement:J Knowledge of Residual Heat Removal System design feature(s) and or interlock(s) which provide for the following: I Control of RHR heat exchanger outlet flow I IEXPlanation ofl 55.41.b(7,8) A is correct because the RH18 valves are fail as-is valve. Losing the air supply will not affect the stable system
,Answers: J conditions as described in the stem. B is incorrect but plausible if it is thought the 21 RH18 fails open. C is incorrect but plausible if it is thought the 21 RH 18 fails open. (J is incorrect but plausible if it is thought the 21 RH 18 fails shut.
I ReferenceTltle-------l: Facility Reference Number i [Reference Section i I Page N~ ~~
ILoss of Control Air I§.OP-AB.CA-0001 ==,II~======:::.I ~13=6==;1 ~11=7=~I IRHR Simplified Drawing 11205:332-SIMP II II 112 I I Ie ~I II I~I~I
!L.O. Number ==:J IRHROOOE004 I I I I I
~aterial Required for Examination II II iQuestion source:] New I I IQuestion Modification Method: ]\ I ~s~~ Dur~,"!Q T!ai!ling Program 1 D
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-- I Salem Unit 1 is operating at 100% power when a loss of ALL AC power occurs.
Which of the followin~ identifies a conseguence if ALL AC power remains deener~ized for at least one da:C I
.,.containment temperature rise to near saturation conditions as RCS inventory is released which will complicate recovery when AC power is restored. J
~: !Containment degradation will result from sustained pressure above 15 psig after RCDT reliefs lift and remain open.
I
~] ILOSS of ECCS pumped injection capability coupled with RCP seal leakage will result in core uncovery.
I IPZR PORVs cycling would ultimately result in S8LOCA.
I
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,......,..---~-- .. ----~~ "~-~~'--l (System/Evolution
'-- ........... __.. ._...... Title i ii Emergency Core Cooling System .. j
!KA ~!C!te.'!!.en.!:.; Knowledge of the effect that a loss or malfunction of the Emergency Core Cooling System will have on the following: I RCS I
'Explanation of ! Containment pressure is expected to rise to - 3 psig and temperature by 40°F as the RCS drains through the RCP seals. Time to
'Answers: : core uncovery as shown on Figure 40 in lesson plan, best case, is <20 hours0.833 days <br />0.119 weeks <br />0.0274 months <br />. RCS inventory will be released to containment,
~
- - however, the containment is designed for a'LBLOCA in which all the mass in the RCS is released to the containment and long term recovery is not affected.
I D ..<'.
,~"."'." ......" Title I Facility Reference Number ! ,~""'" .... Sectfon--- Page No** [ReViSion:
- Loss of All AC Power Lesson Plan II Nn!:::n"l nPAon-OA I I I 4 I 1 I I I I I I I I I
'Material R!9lJired for Examin~ion II II IIQuestion Modification
~-
~est§:nS~~~<:e Comments, DC Cook 2002 NRC Exam, modified to Salem conditions and replaced poor distracters.
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Given the following condition:
- Unit 1 has initiated a Safety Injection while in MODE 3 in response to a LBLOCA.
Choose the set of valves which would did NOT 11 SJ49 AND 12SJ49, RHR Discharge to Cold Leg.
[ ] 111 SJ44 AND 12SJ44, Containment Sump Isolation.
IAilsw~ *EX8fr1 Level] ~~hltive~.':.veIJ
!i<A:1I 006000K610 I~~:~_J ~~r.~~~I.u.!:.]r:TIr~I~ 'ROCi~lDJi$Ro~DJ rsystemiEvolution~Title-: IEmergency Core Cooling System It<AStatement:] Knowledge of the of the effect of a loss or malfunction on the followin!:l will have on the Emer!:lenc~ Core Cooling S::stem: I Valves I EXplanation o~ 55.41.b(8) A is correct because BIT inlet (SJ4/5) and outlet valves (SJ12/13) are normally shut and receive an open signal from iAn_~!,ers:__ SSPS on a SI. B is incorrect because while the CV40 and 41 receive a close signal, their failure to position will not affect charging pump suction, since it automaticall~f realigns with the SJ1 and SJ2 opening to provide suction from the RWST. C is incorrect because SJ49 valves are normally open at power, and do not re-position during a LOCA, but would expect to have ECCS injection I fin' ,f",,.,... 01 lO ". '" fl, ,nn", RI '(11 A n , , ,
- h",..." '"' <::i IAA ,.,h,." "ro ",,,ono,, ,n n("A. "I fl, "in", +r.,ndor f" ("I ro"ir,.. "nrl I do not provide any ECCS infection flow.
l-~}acilitY Referenc; N~mber . *. JRefere~ce Section ILPage No.-! I~evisioll;
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Reference Title :
IECCS Simplified Drawing
'~
11205250-SIMP II II II I IPreparation of the Safety Injection system for 0 I@] .OP-SO.SJ-0001 II 11 28 1114 I I I[ II II II I 1'.-- -'--~~..-.....
~~~-"--"'.
I ECCSOOE016
!Matertal Required for Examination iI
~~~~'?~ INew IIQuesiion r¥1Q(lffication Me~.~od:
iQues~~~__~_~rce~mm~n~ I I
- 1............"'...
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,;fRO~SkvScraner01 !i~~Rosijkril~lf~:~l . . RQ ~~nltEvotutiQn list I ....S~9 Svstem/Evolution List I Out,Uti:1Cllanae:S: :I
!Question Topic:j I RO 37 I Which of the following describes Pressurizer Relief Tank (PRT) response when a bubble is being drawn in the PZR after a vacuum refill of the Reactor Coolant System lAW S2.0P-SO.RCS-0002, Vacuum Refill of the RCS?
PRT ....
~ Ipressure will rise slowly as operators vent air and non-condensibles by opening the Pressurizer PORVs.
I
~ 1.level will rise rapidly as the Pressurizer PORVs cycle automatically in response to the solid PZR expanding.
I t;J Ipressure will rise rapidly as the Pressurizer PORVs cycle automatically in response to the solid PZR expanding I
@J Ilevel will rise slowly as operators maintain Pressurizer pressure during RCP bumps by opening the Pressurizer PORVs.
I
[!.nswe~ Ia J [Exam Level] I R I ~gnitive level II Memory I I [Facility:] Salem 1 & 2 IIE,,_amoateJ I 12/3/20121
~: II 007000K502 IK5.02 IIRoval_ue:][I.1J ~OValue: II 3.4I~ection: II~ fRo Grou~UI5ROGroup.~1 11 !~'1~ D
_-0 ISystem/Evolution L__ Title.JI IPressurizer Relief Tank/Quench Tank System ri<A Statement:j Knowledge of the operational implications of the following concepts as they apply to the Pressurizer Relief Tank/Quench Tank System:
Method of forming a steam bubble in the PZR I
!Explanation of 55.41.b(3) A is correct because operators will perform a 10-15 minute vent of the PZR while drawing a bubble (Step 5.3.28), with IAnswers: i PZR level 40-60% (Step 5.3.5). There will be minimal liquid carryover, but venting will slowly raise PRT pressure. 8 and Care incorrect because the PORVs are controlled in manual. 0 is incorrect because the RCP bumps are performed prior to a vacuum being used in the RCS, and PORVs are in auto during bumps, but will be opened after the RCP is secured for venting.
! Reference Title i L Facility Reference Numb~ !Reference Section II Page No*1! Revision, IVacuum Refill of the RCS 1[}2.0P-SO.RC-0002 II II 11 31 I I IC ~I il iI I I IC II il II I
~ _ _ _ _ _ O~
L.C. Number I IPZRPRTE012 I Material Required f?r .Examination JI II IQuestion Source: Oil New L...-
I iQuestion Modification Method:
II I[Used Ouring Training Program] D
'Question Source co~
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B!)_~~II~R~~~~.i1i1 With Unit 2 in MODE 4, which one of the following describes the normal and loss-of-air positions of the Component Cooling Water Surge Tank Vent Valve 2CC-149?
2CC149 is normal and fails a total loss of its air
- open.
rl~===================================================================================;1 l~:.J
!~I 00SOOOK40S 11I<4.02-~; [*~~~[ROVaiUe:]ITI@!~~I~1 ~]!~[]SROGroup:j[]
"SYSteintEVOIt.itiOn~TitieJ 1Component Cooling Water System i Knowledge of Component Cooling Water System design feature(s) and or interlock(s) which provide for the following*
Operation of the surge tank, includin<~ the associated valves and controls I rExpl~;:;ation of ! 55.41.b(7). 2CC149 is a normally open vent valve which automatically shuts on high radiation in the Surge tank from 2R17A, and
!Answers:.... J fails shut on loss of air (and loss of control power.). The valve is AU (automatic operation) in ALL Modes of operation.
_ .."" ...... ~-----"-----~
Reference Title ll~Fac@yRefere;:;ceNumbe~ lRef~~!l1ce Secti~ !Page N"O:] .e. .:_t.
INo.2 Unit Component Cooling 11205331-1 II Grid F-2 II i 53 1 1 IComponent Cooling Lesson Plan II NQ.S05CCWOOO-07 II 11 20 117 I II I[ II II II I I
._ L.O._ Number ICCWOOOE006 i
Material Required for~mination I I
!n. .. :
Source: .11 Facility Exam Bank I [QueStion Moijlfication Method: ] Editorially Modified I.used Duril1~.!rainl.n~pr~!Jr~ [J
!QlIestionSourceCommentsl .::..... *.;..,_;.__:::____"..... ....J ,
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I Vision Q39302 editorially modified to remove window dressing and common components found in all choices.
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-QueSii~~)"op~ I_R_O_3_9_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _---t Following a loss of offsite power, 2A 4KV Vital Bus fails to reenergize.
Which of the following describes the PZR heater group(s) which are available, or will be made available, to maintain PZR pressure while responding lAW TRIP series EOPs?
up heater group 21 only.
IBoth backup heater groups only, All backup and central heater groups.
la I ~am'level' IR I I Application I [Facility: I Salem 1 &2 1 'E~amDatell 12/3/20121
~c?Ylllue: 1[3.01IsRovalue: 'I 3.41 ~-;;;tiOn~: I~RO Group:11 11 ~O GrouP:11 11 rsYsterl'l!~,,:()lutionTitie' IPressurizer Pressure Cont:ol System 55.41.b(7) Control Group heaters are, powered from 2G non vital bus, and does not have an emergency power supply. 21 Backup Heater Group is normally powered from 2G non vital bus. but has an emergency power supply from the 2C vital bus. 22 Backup Heater Group is normally powered from 2E non vital bus, but has an emergency power supply from the 2A vital bus.
Reference NllmtlAt [Revisionl
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F===================~F===============~F========~F===~~115===~
1========================:::::l-F===========::.~=====l_\==:::::;.1~12===::::;
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l~_a~eriaIRequired for ExaminationJ I II Source: IFacility Exam Bank IIQuestltm MQdift¢~t;ionMetho~1 Significantly Modified ![Used D':'.rin9..Trainirl~.~~OgrarllJ [J
!QuestionSourceCommentsi IVision Q58200. Changed loss of 2B bus to loss of 2A bus which changes cerrect answer from both backup heater L... _~_~ .. groups to only 22 backup heater group. I ii
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~on TopiC! RO 40 I I Given the following conditions:
- Unit 2 is operating at 100% power.
- A power reduction from 100% to 20% Rx power will be performed at 1% per minute lAW S2.0P-AB,LOAD-0001, Rapid Load Reduction,
. Prior to initiating the down power, the Charging Master Flow Controller fails as is .
Which of the foliowinfJ is CLOSEST to what actual PZR level will be when the down power is completed and RCS Tavg is exactly on program?
[8.11 22%,
I
~128%,
I 1
51 59 I
1 %.
I IAns...,er, b I I 'EXamievel 11 R I [Cognitive Level 11 Comprehension IIFacility: 1Salem 1 & 2 I IExamoate:J I 12/3/20121 IKA:1011000K604 IIK6,04 l~va'ue:113.11~Rova'ue:113.1![SectionqSYS 1:~c:lc:;rouP~lI 2j"SROGroup:1 21 !Y~~~' D ISystem/EvoltJ!~on!J!,!~ IPressurizer Level Control System 1011 KA Statement] Knowledge of the of the effect of a loss or malfunction on the following will have on the Pressurizer Level Control System: I Operation of PZR level controllers I i~~~;;:~on Of] 55.41.b(7). Program PZR level is clipped at 59%. As the down power occurs, there will be an outsurge from the PZR as the RCS contracts due to lowering Tavg, At 20% power, RCS Tavg exactly on program Is 551.6*, (AB.ROD-3 Attachment 1) which would give a program level of 28.3%. (AB,ROD-3 Attachment 2). A is incorrect but plausible since it is the no load PZR program level. 0 is incorrect but plausible if the candidate thinks that with charging in manual PZR level remains at its current level. C is the PZR II"""" f"r RnO/_ ~~ "n' ""r (<;{:I;Q <;o~\
_~~."'."'.. Title II Facility Reference Number in..... Section ! Page No. * :Revisi()~
IContinuous Rod Motion II S2.0P*AB.R()n.ooo~ II I 1121 1 I II II I II 1 I II II I II I 1'!"lif:!riaIReqUired for Examim\~ion !
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I
- QUesti~I'lTop~
With Unit 2 at 100% power, Containment Pressure Cl1annell (one) indication became erratic and the channel was removed from service lAW S2.0P SO.RPS-0005, Placing Containment Pressure Channl,1 in Tripped Condition.
What is the ro::."""""., if Containment No response other than channel related alarms.
An AUTO Safety Injection actuation on 2/3 channels tripped, Safety Injection, Containment Spray, Main Steamline Isolation and Phase B Isolation all actuate.
Main Steamline Isolation and Phase B Isolation. Containment Spray valves reposition but the pumps do not start.
~I 012000A301 I[A3.01--:J ~,,~~~OValue:![TIJ~ectiOrl:l~ ~~DJ ISRO Group:]
~~/E:vOlutionT'itle= IReactor Protection System iKA Statement:] Abilit to monitor automatic operations of the Reactor Protection System including:
Individual channel IExplanation of j 55.41.b(7) Cont press Channell only feeds Cont Hi-Hi (Spray act) it does not feed the Cont Hi (SI) circuits. Containment Spray I ~~wers:J s~stem bistables are energ,ized t? actuate, so whe~ the failed channel is removed from service, its Spray actuation bistable is NOT tripped, it is removed from Inputting to Spray coincidence to prevent one of the remaining channels from actuating cont spray If It fails, This leaves the SI circuitry still 213 on channels I. II, and III, and the containment spray actuation goes to 2/3 of the remaining I
~ ___~eference Title ___ J '-'-Facility ~El!ElrElnce Number ~1'l~~S~~tlc:>rl~ I Page No:] ~~
IRPS Safeguards Actuation System 11E1;=05=7========~F=======::::;11 11~2=2=~
1Placing a Containment Pressure Channel in the iI S2.0P-SO.RPS-0005 I 112 I I[ II II~===='
[L.o.Num;ber-*..- -
IRXPROTE012 IMaterial ~El9ull"$d for Examination II I I tC1~~stion Modification Method: j iaUeSlio'n'source:j Facility Exam Bank
,Question Sou~:~ C0I'l'lIl1~nts IVision Q134992 Direct From Source I IUsed During!~il'liniProgram . C j I".............* .,.'
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,Question TOf:li.C ~ RO 42 I Given the following conditions:
. Unit 2 is in MODE 2 during a startup .
- A LOCA causes containment pressure to exceed 1~5 psig.
Which of the following valves indicating OPEN in the control room means that it has failed to reposition properly?
121SW122, CC HX SW INLET VALVE.
I I
~ 22CC3, 21*23 HEADER X*OVER VALVE.
I 1238F22, SG FW STOP CHECK VALVE.
I
~~4~. ,MAIN STEAM~OLATION VAL'IE. ] .
Answ.er ~ 't::xam level I ~ Cognitive l~v!1 IMemory J liiacmtYll Salem 1 & 2 I :Examllaie:] I 12/3/20121
~J!013000A403 1 A4 *03 . ,!t.O VaitieJ Q:~ Is~C) Value: II 4.71 fSectiol1: I.~~_JROGrooi):{JJ ,SR() Group£ I 11 ItfJ C
!~~tElrniE;oiution TitleJ IEngineered Safety Features Actuation System I !01 3mm
]
KA Statement:] Ability to manually operate and/or monitor in the control room: I ESFAS initiation I
'Explanation ot' 55.41.b(7) 21SW122 is a normally open valve that recieves a CLOSE signal upon a MODE III SEC initiation, SI plus Blackout. It's Answers: ... --~."
status is displayed on 2RP4. 22CC3 is a normally open valve which does not have any automatic action. It is plausible because other CCW system valves reposition on SI and/or Phase B, and the SJ 113 valves also have "X-Over" designators, and they reposition on RWST 10 10 level. 23BF22 does not receive a shut signal from the MSU signal that occurs at 15 psig in containment.
I .
lui ~AC:1 ~7 \1,,1\,,:.., ''''''''''' C:~IIT "".-.n,,1 "" Ina ni.n; ..."nbinrncnl "",<",,,ro 11" """..,\
I Reference Title I Facility Reference Number I,"""'" "'....'" Section Page No.: ._,
I "",f"''IIrds A ... t""tion Signals 11221057 I I 1122 I I I I I I I I I I I
!l.O. Number~==-:J IESFOOOE021
- LMaterial Required for Examination JI II
[Questions~~~:':J I=N=e=w=~====:I~[Qu;~ue::'s::t::iO::'~::~M: O=d=i: fiC=at.*=*.=iQ="::M=e=!=l'lo=*tCJ=:::***::::i~========~II::u::se::d::;~:I):u=ri::ng=_=T::ra:::in:::l:::ng=p=l'o=g=ra::*m:::--=]~~D
~!~ti~n.~~~~ce Comments l
I II I I I II----------------------------.------~I
".."""'---"'~--------
Unit 2 was tripped from 100% power about 16 minutes ago.
- IRNI 2N35 indicates 2.0 E-11 Amps.
- IRNI 2N36 indicates 2.0 E-10 Amps.
- Channell SUR is -0.3 dpm Channel II SUR is -0.06.
Which of the following describes the condition present, and any action(s) required to be performed as a result of this condition?
2N35 is under compensated. Ensure Source Range channels automatically energize when 2N36 lowers to 7.0 E-11 Amps.
2N36 is over compensated. Ensure Source Range channels automatically energize when 2N36 lowers to 7.0 E-11 Amps.
IComprehension I 'FaciiiiY.11 Salem 1 & 2 I .Exam~~te: j I
[System/Evolution TItle' ! Nuclear Instrumentation System
- ~ I
~ ~ij.tement~ Ability to (a) predict the impacts of the follOwing on the Nuclear Instrumentation System and (b) based on those predictions, use
- procedures to correct, control, or mitigate the consequences of those abnormal operation:
Faulty or erratic operation of detectors or compensating components l.~;:~:~~_~:~ preventing 55.41.b(6,1 0) TRIP-2 step 38 asks if both I RNI channels are reading <7E-11 A. If they are not, it asks if under compensation is proper operation. If yes then the operator is instructed to manually energize SR channels. IRNI channels normally continue to lower until off scale on I:ontrol console. If the reading is higher that expected, and SUR is abnormally low as indicated in stem, this provides justification for an undercompensated instrument, which allows more lower energy gammas to be seen. on I
in<:trllm"nt _
I Reference Title ] ~~Fac.!lit¥..Reference Number : !Reference Section II Page~No:' ~Si~"J l .. "_. -
IReactor Trip Response 112-EOP-TRIP-2 ---.-JI 115 11 25 I I II ---.-JI II II I I II II II II I
- ~*b~' ,
- ........
iLO..........."'.
~,.~~ ~EE009 I
~; Objectives J
EXCOREE010 J
I iMat~!ia.1 ~!q\Jlred for Examination II II SQ"~~ IFacility Exam Bank I[Question f.'odification Method: il Significantly Modified IfUsed During Trai~~ P;ogram] r IQ\JestionSource comQ'1~ntsll Vision Q61948. Added SUR for both channel as this would be available in CR and gives additional indication of
-~-.-."--- compensation problem. Replaced 2 distracters with additional over/under compensated choices with wrong actions. I c' :',"
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I I !RO 44 IOueSHonTopic I Given the following conditions:
- A LOCA is in progress.
- Operators have transitioned out of EOP-TRIP-1, Reactor Trip or Safety Injection.
Which of the following indicates a sueerheat condition exists in the core, and what CFST is applicable?
15 or more CETs > 1200°F. RED path for Core Cooling.
I I 15 or more CETs > 1200"F. PURPLE path for Core Cooling.
I
~ IALLCETS 650°F with RVLlS Full Range 35%. RED path for Core Cooling.
I IALL CETs 650°F with RVLlS Full Range 74%. PURPLE path for Core Cooling.
IAnsWeIJ Ia I :~arJl~evel
- IR I ~~gnitiv~_L~~el c: IApplication ~ LFaCiiitiJ ISalem 1 & 2 I ExamDatell 1213/20121 I
1 1i<A: II 017000K503 IK5.03 "lROValue:il3.71:~~OValue:'14.11[Section: .1~.ROGrouP:1 21~f(<?Group::1 21 .~~ D
""~~,,---,,~" ~--~,
Isystem!~vOlu~"(m
Title:
lin-core Temperature MOnitor System I 017 I~KA~"'~~~ml!lt~::J' Knowledge of the operational implications of the following concepts as they apply to the In-Core Temperature Monitor System: I
, _______ " m _________ ,_ *** _______ ,
! Indication of superheating I Explanation of, 55.41.b.(5,10)With CETs indicating >1200°F ... ."this temperature indicates that most liquid inventory has already been removed Arlswers; J from the RCS and that core decay heat is superheating steam in the core."(page 6) This is a FRCC RED path entry to FRCC-1
'""" ---"._-- Response to Inadequate Core Cooling. B is incorrect because its priority (PURPLE) is wrong. The 2 distracters with CETs at 650 OF are wrong first because CFST Basis Doc, page 10 states that at least 5 CETs have to be >700°F to indicate superheat is present at tho ~". \I\/ith -7r\r\ 0 t::: th,
- ill~, ,ntn ;nt~ r~" r".~lIn. 011001 t::: ,th If 0\11 I~ t:::,,11 0" ,f~' ..-"lQO r" U because of the lack of superheat and RVLlS level indicates PURPLE Path .. iJ is incorrect becauseit is not indicated that superheat I I exists, and a YELLOW path not PURPLE path.
~.~ Referen~e Title J. F~ciiity Reference Num.~.!! -. ~eierence Section _.=:l ~ ~~~
ICritical Safety Functions Status Trees Basis Do 112-EOP-CFST-1 ~ I i16, 10, Fig 11 25 I I IC ~I II II I I IC ~I II II I
~umber~--.J I FRCCOOE001 I I I I ~
Material Required for Examination
- I I
~i~~.":~!3J INew I 'Question MOdificatfon MethOd::1
~~~stion sourCecomfTIen§
_ " .. " **mt ./ ......... ************ I I
I I
Given the following conditions:
- Unit 2 has experienced a LBLOCA coincident with a loss of off site power.
- 2C 4KV vital bus locked out on bus differential.
- 2B SEC did not actuate.
Assuming one train of ECCS equipment is operating, which of the following identifies the FIRST action which will restore the minimum complement of equipment to assure containment integrity is maintained lAW Salem FSAR?
[~ IResetting 2C SEC.
~ IDepressing START PB for 21 CFCU.
Depressing START PB for 22 AND 24 CFCUs.
Rotating key switch to ON for 21 Containment Spray pump.
~~] I022000A 102 iKA Statement:] Ability to predict and/or monitor changes in parameters associated with operating the Containment Cooling System controls includin :
Containment ressure
- . EXPlanation oil 5541 ,b.(8)FSAR Section 6 and 15 both state that the minimum complement of Containment Spray Pump/CFCUs required to ensure Answers: J containment integrity along with a train of ECCS in operation is 1 CS pump and 3 CFCUs. With the conditions in the stem, only 21
... CS pump will be running on A bus, C bus will be deenergized because a bus differential signal locks out all power to the bus.
Additionally, the power supplies to the CFCUs are A,B,C,B,C for 21-25 CFCUs, so only 21 CFCU will be in operation, A is incorrect l _ _ _ _ _ _-f-D.QC.:3J.J.SCliUIo.a..=I..,;,./;:a.o:x..s.IaIJ:..aJ:1¥-=ac!s "ntil the hilS bas Oo\Me..
Ithat will restore the 3 CFCUs needed, D is incorrect because 21 CS pump will already be running:.
~N~~_
I CSPRA YE002
~
IMaterial R,equired for Examination .J I II IQuestion source:j Facility Exam Bank I Ii~uestion M()~i~~iitionfv1~t~?d;. 11 Editorially Modified IIUsed,DUrillgTraining programJ 0 Comments I IVision Q113269.
Question Source
.---~---
Modified Distracter D from 22 to 21 CS pumps since 2 of the answers (A and D) had equipment that would not be available due to loss of 2C bus . I
.... ..,. . . . . . ~ . . L
...**...... .~
I I
I
Given the following conditions:
- A loss of reactor coolant has occurred which results in containment pressurE: rapidly rising to 18 psig.
down the control boards 25 minutes later to prepare for a crew brief, which of the following locked in Overhead alarms would be ADD TK lVllO.
CFCU WTRFLO TRBL.
IB-7, TURB AREA SW HDR PRESS HI.
lj\nswer II b I ~x.am LeiiiiJ IR I r£()~~i!i"EJ~evel i IApplication .J !Facility: ~ ISalem 1 & 2 ExamDate: i I 12/3/201211
](A:] ~~@~~O Value: 1@J~e~~n:Jl~ ~rc;~1 I 11 p~:a3' * *
"':"':"'~--':""~~:':"':"':":":"";""___________________---ll ~===
. KA Statement~ r-------------------------------------------:
~----
rocedures.
Explanation of I 55.41.b(8) Spray eductor flow is nominally 75 gpm. Normal level 75%. Administratively max level 90% (3900 gal as shown in I Answe'.!:......... J SC.CH-AD.CS-0415) Alarm occurs at 67%. Normal level 3,400 gallons. Alarm at 3,050 gallons. 0-43 will occur - 5 minutes into event. Transition out ofTRIP 15 minutes. E-5 is not expected to be in alarm 25 minutes after the trip because it is indication of the SR instruments being energized when they should not be with respect to turbine power above or below 15%. Source range
_ _ _ _ _ _~.Lt.t:l!:rua.iti.<.::aJJ._~c.e.t:g_i.:z.:e.J::i.E!llIll.e.6i.1J..jL5.-..1 8 min!!tes {alloMOng a trip . ,
alignment problem with the CFCU running. It is plausible because the CFCU Airflow Trouble alarm Will be in alarm, as it occurs when the damper alignment is not correct for running in HIGH speed, and the CFCU will be running in LO speed following an accident. 8-7 would not be in alarm as SW to the TGA is automaticall isolated b the SECs.
Section ;Revision!
E============~~:::::::========::========~~=====';~~==4135,17,261
~~~============~~~~~~==~~======~~==~18 I 1 - - - - - - - - - - - - ' 1 - - - - - - - - - ' 1 - - - - - - - '1-----' I I IL.O. " .........,"
ICSPRAYE008 I
I I
- Material Required!or Examination ; I I IOliestion Source: I!New I Question M<:)~i!i~ation Metho~11 I ;U~dDLIring Training pro"T~rrI' 0
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[§uestion-ropi£J IRO 47 Given the following conditions:
- Salem Unit 1 was operating at 100% power when a LOCA occurred.
- A manual reactor trip and manual SI were initiated.
- When the Main Generator output breakers opened, a loss of off-site power occurred.
- 1A vital bus locked out on bus differential.
Which of the following identifies which Containment 10ljine Removal Units (IRUs) can be started if required?
~ 1111RU ONLY. I
~112IRUONLY. I
~ 111 or 12 IRUs. I
@J INEITHER IRU is available. I
[Artswer'l rD" 'Exam~ ~ ICognitiveLevel II Memory J 'FacilitYlI Salem 1 & 2 1 IE"x~m~ate:'J I 12/3/20121 IKA:11027000K201 1 K2.01 ':'Rovalue:I[~SRovalue:I[£j!section:!I~~OGrouP:'U~~Group:[JJ E:\ D 11 027 _0 IKA Statement: 1 Knowledqe of bus power supplies to the following: 1 Fans I
=nation ofl 55.41.b(9) Containment IRUs are powered from G and E non-vital 460VAC. With the loss of off-site power, none of the non vital Answers: I busses are energized. The distracters are based on the operator knowing that the loading of a vital bus in Mode IV doesn't have any bearing on IRU operation.
Reference Title ,.
I Ciacility RefEl.':.El nce Numb~~-.J [Reference Section ] :Page 'r;io~l rRevisiOn]
11 E1 Aux Building 460-230V One line 1§7916 ~I rl 11 26 I 11G1 Aux Building 460-230V One line 1§7919 II II 11 23 1 I Ie II [I II I ICONTMTE004
- Material L-,
Required_
for Examination i I II l<iuestionsou~ce: -II New IiQuestion Modification Method~1 ]IUsed During Training 'Progra~ D IQuestion Source Comments]
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I iComment ~-~ ~~.~,",,~~-.-,. -- .~
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IQuesti0!l Topic: IR048 I II Which of the following describes how uncovering of the fuel in the Spent Fuel Pool (SFP) is prevented if a leak were to develop on the in-service Spent Fuel Pool CoolinQ Pump? I IAutomatic trip of the in-service SFP Cooling pump on low level in the SFP.
I L~ !Automatic makeup to the SFP combined with the 10 level alarm to alert the Control Room.
I ILocating the SFP Cooling pump suction line close to the surface of the pool, and an anti-siphon hole in the return line to the SFP.
r I
~] locating the SFP Cooling pump discharge line close to the surface of the pool and an anti-siphon hole in the suction line to the SFP Cooling pumps. I
~swer' c I j [EXam Level . !R I l~Ognitiyei.e"-ElO IMemory I ~Cility: 'I Salem 1 & 2 I IExam~at~:1 [ 12/3/20121
...- _." r 1~-::g_3_J I~!~ lEi ~~'lli ~~.il~ ~()_~~~i OJ _S~R~J>:l OJ r;:
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IKA:, 033000K403
~ii;liv§volutI~~!J ISpent Fuel Pool Cooling System I
~Statement:ll Knowiedge of Spent Fuel Pool Cooling System design feature{s) and or interlock(s) which provide for the following' Anti-siphon devices I
- ExplanationOtl 55.41.b(4) There are neither auto SF? pump trips nor auto M1U to the SFP. The suction line is located 4 ft below the surface to IAnswe rs : i minimize level lost if a leak were to develop below the level of the pool in the SFP Cooling system, while the return to the SFP is
- - - .. ~ located 6' above the fuel. The return line has a 1/2' hole drilled in it to prevent it from siphoning the level back to the SFP Cooling I pumps on a leak.
c=:= . _~~~?~~~== ,,---!~~fer~~~~ ~rencesectiOn~*'1 pag(FN6..l~~
ISpent Fuel Pool Cooling 11205.233 --.J I II 11 26 I I IC" II II I I IC 11 II II I iL,O:Number
~---~.~.-----~
ISFPOOOE013
![Material Required for Examination
- I II IQ~estiol1 Scitlr~e: i I=F~ac~i=lity~E~xa:::m=B:a:n:k==I.::~~. u~e=s=ti=O=.~=-M=O~d~i=fi=ca~t~io=n=*. *=M=.et=*. h=o=d=:=:=l::c:o:::n:c:::e:Pt:::u:::S:::ed:::====I':iu::s::e=d:::D::U=r:ln:****9:-=T=ra=in:j=n.:~=p=r=o.:gr=a=m=I~C~1 I
l~uestion Source C()mme~~j Vision Q60120 regarding when in service pump would lose suction I iCotJUfh:"n I I I I I I I
"f ~~ ~h" iii"' m:'Iiii"l u! .1rkk:. '~,"\"H;'" ' "" ''' ** C<< ~'1 ~ " "** ",,7/,,"'"
I ~uestionTOPiC
- IRO 49 I During a steam leak, the CRS directs the PO to FAST close 11MS167 from the 1CC3 bezel. The PO depresses the NORMAL dose PB on 1CC2 instead, and the 11MS167 starts closing hydraulically,
":1.
Which choice ,what will happen if the vl'",'a,v, then pushes the FAST close PB for 11MS167?
I
~ The vent valves11 MS169 and 11 MS171 immediately open, allowing hydraulic pressure to close valve against only atmospheric pressure" 1b.'1 The MSIV hydraulic pump immediately stops, depressurizing the hydraulic header, and allows main steam pressure to dose the valve.
fc.1 The hydraulic sequence will continue until the Valve Fully Closed (33CVO) contact is closed. All other operation of the valve is locked out until
~ the hydraulic pump is deenergized.
[ci:ll A solenoid valve immediately opens, equalizin~l hydraulic pressure on both sides of the operating piston, and vent valve
" 11 MS171 open to allow main steam pressure to close the valve.
~AnsweEj 1d I Exam Levell] R I~v~.;;!~.v LeVell I Ar'll')lication 1 If'acWty
- ISalem 1 & 2 1 ' ...... u . . . . . . . ." . iI 12/3/20121 i~JI 039000G128 112.1.28 [~o Value: ![UJSRO Value: 114.1 ! f~ection: JI~ [tio GroupTI 1 11 ~O Group:l! 11" 0 IMain and Reheat Steam System
,-~~ '~'~'_~.~~~v_ . . . _~
~~y~tem/EvolutJon Titl~_,
,~ Stare~~~~j~ __~~~__________~__~~__~______________~____~______________________________~
Knowiedge of the purpose and function of major system components and controls.
Explanation of 1 55.41.b(4) Using Logic drawings 239916 and 239917: The Emergency Trip signal is generated from the (Fast) CLOSE PB, and acts i~~~!~_J just the same as ~ S~feguards Train MSLI or ~i~h Stm Lin~ Flow SI signal. SV-1 closes (was open to direct hydraulic pressure to bottom of hydraulic piston.) SV-3 opens, equaliZing hydrauliC pressure on both Sides of the hydraulic piston, and allows the hydraulic fluid to act as buffer to prevent 11 MS167 from slamming closed. The solenoids for 11 MS169 and 11 MS 171 open, venting
",;r ",nn In", "",I"",,, "non I" ",II"" ~n<::: nr",,,,,,,,,,, f'\n In.. n"lt",n r,f In-.. I"",or nn",r",l;nN n;d"n I" nr;"o Iho n;"" "n o>"",;n<>l .
Ipressure. The hydraulic pump does immediately stop running.
l~,__------~erence Title ,.--] [~FaclIjtYReference Number ; iReferenceSec~- ~~__N~ IRevisionl IMain Steam System Stop Valves Vent Valves 11239916 F========~IF====~F===!-~===;
II II r 17
! Main Steam System 11MS167 Stop Valves Hyd 11l=2=39=9=17========i1~~1= = = = = = : ; 1 ~I==~I!12 I II_______---'I!_ _ _---JII_~I :=1~
IMaterial Req,:!ired for Examination j1 II
~~ue~t~~_s_~~~!J ! Facility Exam Bank IIQue$d~o Modification Me~hOd: IDirect From Source I~dDtiring Training ...,ro{,Jr~_m c:
L~~~~ti?n_~~~rce CQmmen!~ IVision Q80671 I
- ~~ ...... ...
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- 1 I
I I
.. ............. -------------------------------------------------------I
!Question To-pi~ IRO 50 IiI Given the following conditions: I
- A Unit 2 plant startup is in progress.
- Reactor power is stable at 18%.
- The Main Generator is rolling unloaded at 1800 rpm.
- Main Stearn Dumps are controlling in AUTO in MS Pressure control.
MS Dump Pressure setpoint is lowered 5 psig.
With no other operator action, several minutes later you will notice:
I !I::11 Reactor Power is <18%.
I~:i I Reactor power is > 18%.
I I
[(J IControl rods have stepped in.
[~J IControl rods have stepped out.
I I
~sv..er: Ib I !Exam Levell IR I !Cognitive !-evel II Application I rFacllft;.:j ISalem 1 & 2 !IExamDate: .1 12/3/20121
[KA: II 039000K508 1~5.08 IRO Value: II 3.61 ~SRO Value: ICINSection:1 SYS iRo Group:! I 1ISROGroup:! j 11 &\11 D
-~_~r'"'T'7J7'7' rSystem/Evolution Title. IMain and Reheat Stearn System ----
t~~t~~~f!l~Ilt: Knowiedge of the operational implications of the following . as they apply to the Main and Reheat Steam System: I Effect of steam removal on reactivity I IExplanation of .. 55.41.b(5)With steam dumps open in MS pressure control auto, lowering the setpoint will cause steam dumps to open to reduce Answers: ,............
steam header pressure to setpoint pressure This will cause a higher steam flow and lower temperature. which causes higher 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 stern.
- Reference Title !
- Facility.,,,.,,,,........ Number ,I !,..."'."'....." Section 1:---'
I Page No, i l~EI"isi~
I II I II I I II I II I I II I II I l!::.c:l:I\I':IIl'l~~ __.
IRXOPERE021 iMaterial Required for Examination II Ii l~~~~~i~~~~~r~=-J IFacility Exam Bank IlQuestion MOdi~onMettlo~ Direct From Source I[sed Dur1rm Trainin~Programm D
~==~~~~~~~======~==~====~~I IQue~i~n~~~r=-e~~~~~I11~.:nts IVision Q67278 modified from rod pull to lowering MS dump setpoint.
'_,I.,~ ~,_, ~_ 1I6t!~. J ...,.......
Given the following conditions:
- Unit 1 is operating at 100% power when a Main Turbine trip causes a reactor trip.
The Main Steam Dumps do NOT arm.
Which of the f"II,,,Mir,,, describes the effect this failure to arm will have on the Reactor Coolant over the next 5 minutes?
PZR PORVs will open to restore RCS pressure to 2235 psig.
rise and PZR spray will keep pressure below 2335 RCS Tavg will stabilize at a temperature below 547°F based on reset curve of SG Safety Valves which have lifted.
will stabilize at a temperature above 547°F based on 11-14MS10 operation in conjunction with at least one SG Safety Valve having I~~:~ __ ] IRO Value:G~sRo-VaTue:TI~L~~I~ IRO Group: I lSySiem/Evolution======~ Title. iSteam Dump System and Turbine Bypass Control I . _.. - ~ !
iKAStateme~i:~J Knowledge of the effect that a loss or malfunction of the Steam Dump System and Turbine Bypass Control will have on the following:
RCS IExplanation of 55.41.b(4,5,) With a reactor trip, core heat production will lower rapidly. The Steam Generator Atmospheric Reliefs, MS10s, will
- Answers:
,~ ~ .. .. ........ .....
~ ~~ ~
l open to establish RCS temp -551-552 OF. The RCS pressure will not rise enough to open the PORV's, much less the PZR
~ Safeties, which are meant to relieve a loss of load with the Rx still at power. PZR Spray valves will open rapidly and fully as required to prevent PORV operation. The Tavg distracters will not occur as the Safeties will not open, and would reset well before lowering In ,~, ora .-1 ('I('I&; n~in In" I"",rl t"""n fr., &;A7 0 C:'
~ , ,
I~
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"~-,-"--,
PZR and PRT Lesson Plan Reference Title
==r, I~OS05PZRPRT-06
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~acility Reference~lJl'\'lber *iReference S~ction! I Page No.
II II 116 I I II II III i II I I Ie ~I ,' II II I
.L.O. Number ISTDUMPE011 IMaterial Required !~~.~.amination _._-" I II Source: i INew I~estion Modificath:ln ~~tJ'l<>d: 1 I Used D~ring TrainingpI'()SJr~§'] 0 Source ,",V..11""""j I, I
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~l.1flstion TOpic! IRO 52 Of the following, choose the choice which contains ONLY actions that automatically occur on a Unit 2 Main Turbine trip from 100% power, with NO operator action.
I. Running EHC pumps trip II. 500KV breakers 1-9 and 9-10 open.
III. Emergency Bearing Oil pumps start.
IV. 4KV Vital buses swap power supplies.
V. 4KV Group buses swap power supplies Exciter .
~:! I045000A311 Il~11~~-- ~rill~~~~§~]I~ROGroup~DIrSRO Group.~DI
- sys:t;mlEvOiu~n Title~ !Main Turbine Generator System
~. Stat~;';'ent:J Generator trip
[Explanation ot]55.41.b(4) Running EHC pumps do not auto stop, but plausible because F-32 DEHC trip occurs on turbine trip. 1-9 and 9-10 are Answers: the Unit 2 Main Generator output breakers, and they open automatically on every turbine trip. Emergency bearing oil pumps do not
~----.-.---~ start but plausible because aux bearing oil pump will start. 4 KV group buses are powered from APT when Main Generator is operating, an automatically swap to Station Power Transformers powered from off site power upon when the output breakers open.
~ Reference Title--;-'~ =~FaciiityRe!~r.~nce Numb;;--ll~!~~rence section:] ~ ~f.~
IOverhead Annunciators Windows F,G,H \I S2.0P-ARZZ-0006,7,8 II II 11 15.48, 19 1 I Generator Voltage regulator Exciter Field Break 11601 037 ~I II I 16 I I Ie ~I II II I IMNTURBE006
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[~~i~~;;:r~~p~] I.;..R;.;;0....;5:...;3~_ _ _ _ _ _ _ _-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ---'1 Given the following conditions:
- Unit 1 is operating at 45% steady state power.
- All Heater Drain Pumps are O/S.
- All Steam Flows and Feed Flows are 40% and stable
- Nl's indicate 45% on each channel and stable.
- RCS TavgfTref deviation is O.O"F.
Which of the following is CLOSEST to the programmed value of SG Feed Delta-P lAW S2.0P-SO.CN-0002, Steam Generator Feed Pump Operation?
'~~'150 ,id.
~ 160 psid.
Ie: 180 psid.
[Ci~, 1150 psid. I IAnswer I ~EXam[evell ~ I
- Cognitive!:evel-j Application I IFacility: : ISalem 1 & 2 1 IExamDate: ' I 1213/20121
- KA:I059000A107 IA1.07 *Rova'ue::~i~~<:>~r.~~:nsectiCm:JI~Ro~TI)UP~DI$ROGroUP~D
- D sYsteiriJEv~i'~tion Title IMain Feedwater System i!<A ~taJ~ment: 1 Ability to predict and/or monitor changes in parameters associated with operating the Main Feedwater System controls inclu Feed Pump speed, including normal Gontrol speed for ICS I iExplanation ofl 55.41.b(4) SG Feed DIP (delta between feed pressure and SG pressure) is controlled by adjusting SGFP speed, and is
- Answers:...---,---. I programmed based on total °/" Steam Flow. Actual SGFP speed in rpm is only a result, not a controlled parameter. The KiA intent
~-----.---.-,
is met by asking how the feed pressure DIP is controlled, which itself controls SGFP speed. 50 is the minimum DIP from 0-15%.
150 psig is the 100% DIP. 60 psld is if candidate used linear scale from 0%-100% steam flow.
I a"". " **,," Title
.no"""" ******* Facility Reference Number ' IR..f ... " .....'" Section 'J I Page No. i I Revision I Steam Gen"'.dlV. Feed Pump u!-'",.citliJII I S2.0P-SO.CN-0002 1 Exhibit 1 1188 11 33 I I I I II I I t I II I fiJiNUrllberI ICN&FDWE008
~aterial Required for Exami.!'atio~ I II
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""'" """ 'Tm "lj "'~i ~m 1 ' :~ ','",,-, _P<-__"""I"'" -~.) ,','" " " , ,~~~, . "" '"97 IOIl..",tin" Topic i IRO 54 I IWhich of the followine identifies how over-cooling of the RCS is erevented on an uncomelicated manual Rx trip from 100% power? I IP-1 0 actuates.
I fb~11 P-12 actuates.
I E:_j Feedwater Isolation.
IFeedwater Interlock.
I I
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1KA:! 059000K105 I[K1.05 e 1\ Group: --.!J l_'_M'_
System/~v~I~~ion Title ; IMain Feedwater System - i 1059 I iKA Statement: 1 Knowledge of the physical connections and/or cause-effect relationships between Main Feedwater System and the following: I RCS II
- Explanation of" 55.41.b(7)Feedwater interlock actuates when 3/4 RCS Tavgs <554°F and at least one Reactor Trip and associated bypass breaker J
~s""ers_:__ open. This shuts the BF19's and BF40 Feed Reg Valves. Feedwater Isolation occurs when 2/3 SG NR levels on 1/4 SG's reaches 67% OR on a SI signal, and shuts BF19s, 40s, 13s, and trips the SGFPs. On an uncomplicated Rx trip this will not occur. P-10 is 3/4 PRNls <10% power, and blocks the low power Rx trips. P-12 is 3/4 RCS Tavgs <543°F, and will shut the Steam Dump valves, I
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""." ., " Section I Page No. Ii Revision:
1Reactor Trip "'(""tN' """ 112-EOP-TRIP-2 I' Basis Document I8 11 28 I I I I II I I I I II I IL.O. Number ! Objecti~;$~~t I CN&FDWt:.uuo I
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Material Required for Examination ! I II
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[~lIest!OnTOPiC i,.;..:..::....::-=-_________,_ _______________________________- - - l Given the following conditions:
- Salem Unit 2 is operating at 100% power.
- 21 AFW pump is CIT for pump oil bubbler repair.
- A 400 gpm tube rupture occurs on 23 SG.
- The Rx is tripped and a SI initiated successfully.
- 2A 4KV vital bus locks out on Bus Differential.
Which of the f"II"'AJinn describes how 23 AFW should be utilized lAW 2-EOP-SGTR-1 Steam Generator Tube Lower 23 AFW pump speed to minimum and trip 23 AFW pump regardless of MDAFW pump status to terminate the unmonitored radioactive release from its steam exhaust.
Lower 23 AFW pump speed to minimum and trip 23 AFW pump. Do not restart until 23MS45 23 5G TO AF PUMP TURB STOP VALVE is shut.
Continue running 23 AFW pump since 22 AFW pump will only be supplying feed to a single SG.
Continue running 23 AFW pum p because it is the only source of feed flow to the SGs.
ISystem/Evol~iion
- - -.. ~-
nil:!] IAuxiliary I Emergency Feedwater System I 061
.!<A Statement] IKnowledge of the physical connections and/or cause-effect relationships between Auxiliary I Emergency Feedwater System and the ifollowing:
Main steam system I IExplanation 'o~ 55.41.b(4, 10, 13) 23 Turbine Driven AFW pump is supplied from 21 and 23 Main steam lines. Each steam line has a tap upstream tAnswers: of the M5IV. Each line has its own isolation valve (21 MS45 and 23MS45) and then its own check valve before the 2 lines combine into one line which feeds the TDAFW pump. During a SGTR. the TDAFW pump remains in service ONLY if it is the SOLE source of feed flow to the SG's. In the stem. 22 AFW pump will be running. since it is powered off of 2B 4KV vital bus, and it has received I ::m ::" .t.... d",rt . <:r.:: 1,,\1<,1 il i" . hIlt nl"".c::ihl" h""::>**",,, th"r" 1<: ::>n . ",I"",,,,, fmm ?<l. ilr=w nllmn "in"" it exhausts directly to atmosphere, but is only secured if there is another source of feed. B is correct per Steps 4.4,4.5. and 4.7 of SGTR-1. which state to lower speed and trip if it is not the sole source of feed, and to not restart until 23MS45 has been shut. C is incorrect because a single SG being fed is sufficient for heat sink status and 23 AFW pump will not continue to be run. D is incorrect because 22 AFW pump is powered from 2B 4KV vit.1 bus and it remains energized. 21 AFW pump is powered from 2A 4KV vital bus which was lost.
'~=*-c--~-R;ference*Titi;~'~"**~~J
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Facilit¥.Reference~umber i
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Refe!ence SE!~ion
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Ii Page Noj Revision' ISteam Generator Tube Rupture I[310P-SGTR-1 -.-JI II 1127 I IUnit 2 Main Steam I~303 Sheet 1 II Grid G-3 and C-3 II 11 67 I I II II II II I
]L.O. Number' I SGTR01 E009 I I ~
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.* RO s~rl *f,~~~J ;~RbsYfll!Jliution list . Outline ChanA~~~1 QUe~iOn*TOp~
Given the following conditions:
- Unit 2 has tripped from 100% power from a faulty SSPS relay on the unaffected train during SSPS testing.
- No Safety Injection has occurred or is required.
tnllnw,nn describes the fails to start?
Operator action to throttle the 21-24AF11, S/G LEVEL CONTROL VALVES, will be to prevent overfeeding the SGs, since 23 AFW pump speed will NOT be lowered to minimum speed unless BOTH AFW pumps in EOP-TRIP-2, Rx Trip Response.
action to throttle the 21-24AF11, S/G LEVEL CONTROL VALVES, will be directed to prevent overfeeding the SGs, since 23 AFW NOT be secured unless BOTH AFW pumps are running in EOP-TRIP-1 Rx Trip or Safety Injection.
ill receive more AFW flow than 23 and 24 SGs. With NO operator action, a S!eamline DIP Safety Injection will actua ill receive more AFW flow than 21 and 22 SGs. With NO operator action, a Steamline DIP Safety Injection will actuate 23/
1061000K302
'SystemJEvolutlCin Title] IAuxiliary / Emergency Feedwater System
- .KA Statement: II Knowledge of the effect that a loss or malfunction of the Auxilia~ I Emergencz: Feedwater Sz:stem will have on the followinr::r I S/G I
- Explanation ofl 55.41.b{4) 21 MDAFW pump supplies AFW flow to 23 and 24 SG. 23 TDAFW pp supplies all 4 SGs. Following a Rx trip, operators IAnSWE~.rS: ~~.J will transition to TRIP-2 after the Immediate Actions of TRIP-1 are performed and a SI is not required. After stopping the SGFPs in TRIP-2, step 3, 23 AFW pp speed is lowered to minimum or 22E4 Ibm/hr. Since there would be no flow to 23 and 24 SGs if speed was lowered to minimum, operators will throttle the AF11s to balance flow to each of the SGs and maintain levels and pressures
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......._.. _.. _-_. ..., ;--- -.-.-..... m Reference Title Fac!Iil¥.Reference Number .IRefere.,~e Section Ilpage No.1 IRevisioni I
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Reactor Trip Response 112-EOP-TRIP-2 II II 11 28 I 1 Ie II II II I I II II II i1 I I
,Material Required for Examination II II 10m.",.I..." Source: II Facility Exam Bank I§~~ion Modification Method: ,[ Direct From Source llUsed During TrainingProgri'lnl
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Which of the following describes how control room instrumentation will be affected if the 1C Vital Instrument Bus (VIB) Inverter were to experience a latched transfer?
Instrumentation powered from 1C VIS ...
is restored to its iC*Ogni1ive}.eve,11 Memory I 'Facility: ISalem 1 & 2 I iExamDate:] I
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KA Statement:
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............... -~
- Ability to predict and/or monitor changes in parameters associated with operating the A.C. Electrical Distribution controls including: I Effect on instrumentation and contr~s of switching power supplies I iEXPlanation~o1 55.41.b(7) B is incorrect because ttle VIS, and the instrumentation powered from it, remain OPERABLE as long as the inverter is
!Answe!!l~=--_~. ....' powering the Vital Bus. (P&L 3.5) The transfer of a VIB inverter takes 2/3 of 1 cycle, which is 11.1 milli seconds, which will not give enough time for the lights to respond on the instrumentation. D is incorrect because indication won't be lost, and 1D VIS is powered from 1B bus .
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Reference Title ~~ . ! Facility RI;l~I;l~ence Number i Reference Section 11C Vital Instrument Bus UPS System Operation rfS1.0p.SO,115-0013 --..JI II 119 J IOverhead Annunciator Window B II S1 ,OP-ARZZ-0002 I II 11 28 I I 11 --..JI II II I
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Question TOPiCI! RO 58 IChoose the one component below which does NOT receive power from the Unit 1 Vital 125 VDC system.
11 D Vital Instrument Bus Inverter. I 11H 4KV Group Bus control power. I IGland Sealing Steam Annunciator Panel Alarms. I ISupervisory Control and Data Acquisition (SCADA) System. I
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~] I063000K201 I K2.01 , *. r:t<:l. .~ ~ ~r:t0 v.iiue] [Ij ~... *liiiJ !RQ Grou~ U ;SRO Group:! U 1~~5L c:::
reT~W' ..tiM Title I D.C. Electrical Distribution I ~==j
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KA_S.tat~.!!I:~l1t:_ Knowledge of bus power supplies to the following:
Major dc loads I
'Explanation oil 55.41.b(4) The SCADA system is powered from the Cire Water 125 VDC system, not the vital 125VDC system. All the distracters
!Answers: I are powered from the 125 VDC vital buses. 1D Vital Instrument Bus Inverter receives power from 1B 125 VDC bus. The 1H 4KV
~...- - - - . - Group Bus receives its control power from 1A 125 VDC bus (reg) and 18125 VDC (em erg), The Gland Sealing Steam panel recieves power from the 1ADC Distribution Cabinet.
Reference Title i Facility Reference Number I i~';:'" ....,,'" Section , Page No. ! ~;:;ision!
INo.1 Unit 125 VDC One Line I 2(3007 I I 117 I IDC ::::""""va, "'1''''' Lesson Plan IN OELEC-07 I I 1/ 7 I I I I II I I-----------------------------------------------------------------------------------------------~
IL.O. Number
- DCELECEnl)7 I I
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Which of the following describes the Design Basis for the capacity of the EDG Diesel Fuel Oil Storage Tanks (DFOSTs) following a LOCA coincident with a LOOP?
DFOST(s) are I is designed to supply EDGs continuously for 4.5 da s.
IThe COMBINED volume of BOTH; TWO The COMBINED volume of BOTH; THREE.
Ic I iExam Level .1 R I ~ognitiveLevel IMemory i !FaCility] ISalem 1 & 2 I iExamDat~ I 12/3/20121
,03 [~~~ISRovalue:jG]~!~I~ ~~o.~U!SRO Group:IU Diesel fuel oil su pi s stem EX,',P,i, anatio",n O,fJI 55.41.b(8) Salem FSAR states .... The combined volume of both 30,000 gallon fuel oil storage tanks contains sufficient fuel oil at the Answers: Technical Specification minimum volume to supply two diesel generators, operating at the most limiting accident mitigation profile
,-- " - - - for LOCA with loss of offsite power, for approximately 4.5 days:
ISalem UFSAR II:=======:~ ISection 9.5.4 119.5-40 1\16 FI==========:;'1\ II II~===,.I ~I==::::::,
1------------'11 11--------'11 Ii----'
~
!L.O. Number I EDGOOOEOO4 I
I i Ef)GnnnFnO?
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IMaterial Required for Examination L..............~ ~.L........... . . :HHm ***
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[KA:lI068000K107 !~1~_J~~.r.Iil~~~~~~I~~~0I~~~OI
[SY$tem/E~I':ltion11tleJ ILiquid Radwaste System
~ StcI~r:rtent:J s between Liquid Radwaste S stem and the followin :
[Explanation ofl 55.41.b(13) RHR pump rooms each have a sump for receiving drains and leakage in the room. Each RHR pump room sump is
!AAswers: pumped to the inservice Waste Hold up Tank. A is incorrect because the Aux Building sump tank collects floor drains from locations above it, and it is located on 64'. e is incorrect because the eves HUT system receives influent which can be processed and recovered as eves quality water, not floor drains. D is incorrect because Laundry and Hot Shower drain collection
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L.O. Number~
IWASLlQE004 I
! I I
Material Required for Examination JI II
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_~uestion s~~~e... (;ommen~~1 I
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Given the following conditions:
- Salem Unit 1 is operating at 100% power, with no equipment out of service.
- Salem Unit 2 is in MODE 5, during a return from a refueling outage.
- Unit 2 is performing a normal Liquid Release from 21 CVCS Monitor Tank via 22 SW header to Unit 1 CW lAW S2.0P-SO.WL-0001, RE~lease of Radioactive Liquid Waste from 21 CVCS monitor Tank.
- The Radwaste Overboard Discharge Flow Recorder 2FR-1064 is OPERABLE.
- The 2R18 Liquid Waste Disposal and the 2R41D Plant Vent Release Rate monitors are OPERABLE.
- After commencing the release, the field operator is recording the Initial Release Data.
- When recording the 2R18 reading, it reads 10 E6 cps.
- The 2R18 red alarm light is lit on the 104 panel.
- The 2WL51 Liquid Release Stop Valve indicates OPEN in the control room.
Which of the following describes what these indications mean, and how the 0 eratin crew should proceed?
r Lo:t: The 2R18 is reading below the actual alarm setpoint. Continue with the liquid release and reset the alarm light at the 104 Panel.
~ With the 2FR-1 064 and 2R41 D OPERABLE, block the 2R18 input to the 2WL51 on 2RP1 prior to the time delay timing out and continue the liquid release.
6 The 2WL51 should have immediately automatically closed on the 2R18 high radiation alarm. The NCO in the control room should shut 2WL51.
e 2WUi1 should have immediately automatically closed on the 2R18 high radiation alarm. The NEO should shut 2WL51 locally.
~_..~~._l_e_v~e~1 ~ 'Cognit~}..ev!!J IApplication j ,FaCmtyl ,Exam Date:
!KA: :'10-73-0-0-0K-3-0-1-----:- iRQ Vaiu~Q]" ~~~@~~~~I~ ~~~~~CTI@;ROGroup:1 CTI
!SysteI1llEvolut~0n..!~!!ei IProcess Radiation Monitoring System I
- KA ~t~_l!!ent:J Knowiedge of the effect that a loss or malfunction of the Process Radiation Monitoring System will have on the following:
Radioactive effluent releases I iExplanationo'-1 55.41.b(13) 6.82E5 cps is the ALARM setpoint which will automatically shut the 2WL51 (S2.1C-CC.RM-0028). The current reading Answers: ' is above the setpoint at which the 2WL51 should have automatically shut, and the NCO should shut the valve remotely. The red alarm light for high radiation indicates the 2WL51 should have shut, memorization of alarm setpoint is not necessary. Additionally, S2.0P-SO'wL-0001 for releasing the tank, Step 5.5.9, says if the 2R18 alarms, then the NEO is to inform the NCO to shut the
')\1\/1 &;1 Tho" ;c. n, ". f~, ,,,,,,,., tho ,,,I,,,, ,,,11,, Th"., , n,' .4, f" ')'All &;1 n' .. ,fnr Ireleasing a hot tank or is there an expected spike J*Page-N~:~ [~E!~
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I _ Reference Title ... ~ r--~Faciiity Reference NUIl1~~_r . !Reference Section .*****..
IRelease of Radioactive Liquid Waste I§.OP-SO,WL-0001 .-JI II 11 25 I 12R18 Liquid Waste Disposal Process Radiation I§.IC-CC.RM-0028 .-JI II 1114 I I Ie II II II I
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I Given the following conditions:
- Unit 2 is operating at 100% power steady state.
- A field operator reports a SW leak in 2C EDG room, just upstream of 23SW:39, 2C DIESEL CLGSWVLV.
The RO reports Service Water pressure on both 21 and 22 headers has low3red from 112 to 101 psig and continues to lower.
Which of the following describes the expected SW system response, and how the operating crew will respond lAW S2.0P-AB.SW-0001, Loss of Service Water Header Pressure?
Assume SW header pressure can be restored.
The standb SW um will start when SW header pressure lowers to ...
95.5 psig. Lock out 2C EDG and declare it INOPERABLE, shut 21 SW21 AND 22SW21, DIESEL CLG SW INLET VALVES, to isolate the leak.
Lock oul2C EDG and declare it INOPERABLE, shut 21SW21 AND 22SW21, DIESEL CLG SW INLET VALVES, to isolate the leak.
95.5 psig. Lock out 2C EDG and declare it INOPERABLE, isolate the leak by shutting 21SW37 AND 22SW37, 2C DIESEL CLG SW INLET VALVES.
~dJ 199.5 psig. Lock out 2C EDG and declare it INOPERABLE, isolate the leak by shutting 21SW37 AND 22SW37, 2C DIESEL CLG SW INLET I VALVt:'>.
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System/Evolution Title. Ii Service Water System
-~-
- " _ _ _ r, IKAStatemei;!;
~--- ..........- . Ability to (a) predict the impacts of the following on the Service Water System and (b) based on those predictions, use procedures to correct, control. or mitigate the consequences of those abnormal operation:
Service water header pressure 1
!Explanatlon of 55.41.b(7, 10)Lock out the EDG(s) that will be affected and isolate the leak. Step 3.11 has operators isolate the leak. EDG must be iAnswers: . . . 1 locked out to prevent starting with no SW available. The only way to isolate the leak is to isolate both supplies from both SW L...... ~
headers by closing both SW37's. Cannot isolate both SW21s per Att 4, Steps 4.0 Band C because it would render ALL EDGs inoperable. OHA for SW header pressure low states auto start for standby SW pumps is 95.5 psig.
r=-=-__" Reference-*ntl~_~;~. *~**_: [)<I~mtY Refer~l1Ce N-~~~r .' 'j ~eferenceSe~!on J[Page No:'ll~ev~~o~
ILoss of Service Water Header Pressure I!li0P-AB.SW-0001 II II 11 16 I IOverhead Annunciators Window B II S2.0P-ARZZ-0002 II 11 28 11 35 I I II ~l II 1/ I
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I ABSW01 E004 1------'
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.§Ollrc7 ~llvision 077578. Changed from what to do to isolate leak (4 choices) to what pressure auto pump will start and how to I' isolate (made into a "2 and 2" question.
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IRO In".. .,ti""" Topic j 63 1 IWhiCh of the following identifies normal Control Air header pressure, and the pressure at which the Emergency Air Compressor will automatically start? I 1100 psig; 1
1110 psig 90 psig.
I 1100 psig; 90 psig.
I
[ii 1110 psig; 85 psig.
I IAnswer I a 1 ieXam Le;eJ] IR I i'""'lIt....' Level II Memory I I FacUity,j Salem 1 & 2 I IExamDate: 11 12/3/20121
, I
~~=::=,I.II=07::8=00::0=A3::0=1==:::1 fA3 .01 I ,~' ,
iExplanation 55.41.b(7) AB.CA Basis Document states ... "When supplied from SA through the dryers, CA pressure runs approximately 5 psig Answers: below SA pressure. CA pressure cycles along with the SA cycle. Thus, CA pressures normally run between 95 and 105 psig.The
"'~,.--,.,*.- .. ,--~-' Emergency Control Air Compressor (ECAC). should auto start if CA pressure drops to 85 psig.
Reference Title I-(AfArAnr.A Number
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Topic [R064 I Given the following conditions:
- Unit 2 is operating at 100% power when the following occurs:
- OHA A-7, FIRE PROT FIRE.
- 2RPS Fire Protection Panel, Zone 148 Work Control Ops Ready Room lamp illuminates, as does the FIRE lamp for that row of alarms.
1- The audible coded fire alarm is broadcast over the plant PA system.
For these conditions, which of the following identifies:
- 1. What these indications mean.
- 2. How the 2RPS Fire Protection Panel is reset when the condition has cleared. I II An active fire suppression system (water/C02/Halon) has activated. Reset from the Control Room.
I fire suppression system (water/C02/Halon) has activated. Reset from the Relay Room.
I IA Fire alarm (smoke or heat) has activated, Reset from the Control Room.
ld.:-IA Fire alarm (smoke or heat) has activated. Reset - from the Relay Room.
I IAnswer lid i Exam Ltwel : I R I I .C"1dllitrv.. Level II Application ~ :FaCilitY:] ISalem 1 & 2 I~I 12/3/20121
~JI086000A402 IA4.02 ~<?Y~lue:1 3.sj ,sRovilUe:1 3.SI ~ectiorKJl~ IRO Group:1 21 [SROGroup:11 21 I~[!~:
IFire Protection System
~,~---~ ................-....
iSystem/Evolution Title
~~tatement:~ Ability to manually operate and/or monitor in the control room:
Fire detection panels I
~XPlanatiOnOfl SS.41.b( 4) 1)Early warning detectors provide the Control Room with early indication of fire, but do not cause a suppression system iAnswers:
t:.::.::....._ _ _ _ _._ to actuate. Early warning Smoke Detc~ctors and Fire Detectors installed in the plant are arranged to alarm to the Control Room by zone.a) If a detector on a zone actuates, the zone indicating light on Panel RP5 will illuminate along with the appropriate group "FIRE" light, and the coded fire alarm assigned to the zone will be broadcast over the station PA system.
rln,..", <> .,.Ano ",,,t,.,,,t,,,,, tn", ",I",.m r"'"'aies iillUIlieated Iialil tbe .,"n<> ;" "'<:In, ."II~ r",,,,,,! fr"", tn" for" noon",l" in I Room.
r"-~---*'-~
I Reference Title -**.:"~**~I QacilitY~Il!llrence ~~..:nbllr __ . ~ierence Section I !I;)~ NO] §IlViSi_~~
I Fire Protection System Lesson Plan II NOS05FIREPRO-07 II 11 31 -32 117 I I IC il II II I I Ie II rI II I
~NU~b~--~i IFIRPROE008 I I I I I Material Required for Examination i I II OJlAdlnnSoufce:- INew IIQuesiion:~~~ification Method: I I[Used During Trainiri~iProgran1J 0
~~!rtion Source Comments I I I
.... >;;;;:1 II I II I II I
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~,!uestion Topi~.!
Given the following conditions:
- Operators are performing a Unit 2 Containment Vacuum Relief lAW S2.0P-SO.CBV-0002, Containment Pressure* Vacuum Relief System Operation.
- 2VC5 and 2VC6 CONT VENT ISO DAMPERS are open.
- The Vacuum Relief Damper is open.
A valid Containment Ventilation Isolation signal from the RMS system occurs.
Which of the fol identifies how this will affect the Containment Vacuum Relief in ?
ONLY the Vacuum Relief Damper will shut.
2VC5, 2VC6, and the Vacuum Relief Damper will shut 2VC5, 2VC6, and the Vacuum Relief Damper remain open (CVI blocked during Vacuum Relief).
F-'--~*-~---~"-----******l
~gnitive~evel J
' ,~.- ..
IContainment System
- KA§t<ttemj!n!:J Ability to manually operate and/or monitor in the control room: I Containment vacuum system I
!Explanation of! 55.41.b(11) RMS initiated CVI signals close the VC1,4,5.6 and the Pressure Relief and Vacuum Relief Dampers. The CVI can be
!Answers:
- - -....... ~-,-- .. ~--------
! blocked if present to allow the commencement of the relief, but is NOT present in initial conditions in stem until it occurs after vacuum relief has been started
!! Page NollRevision
~.-
Faciiity R!ference NumbElr . '1 ~eferenC!Sectiori Reference Title
- - - - - - - - -......- - - - - -.. .L I ~ ..... _ _ _ ,._~ .... JJ IContainment Pressure-Vacuum Relief System II S1.0P-SO.CBV-0002 il 99-101 i 121
~~.----- -~--
--.JI !
ICONTAINMENT AND CONTAINMENT SUPPO II NOS05CONTMT*11 II II i 111 !
I Ie --.JI II II I
!L.e. Number --~,
I k ******* _ _ _ _ ~ **** _
CONTMTE007 l"'a~E!rial Required forExamination~j I IlljsedDuring Tral,!ing P~o~ramJ 0 I
I
~~'")'!:1_f:A~_~\I!i!Q,I'_ A .i ....
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[§uestionToP!C' 1RO 66 I IThe eureose of the Reactor Coolant Pume ~RCP) Thermal Barrier is to ... I I
~ limit loss of reactor coolant up the RCP shaft if the seal package fails.
I I ~ols:al injection flow to prevent overheating of the #1 seal.
I Iprotect the radial bearing and seals from the heat of the RCS, I
Iprotect the thrust bearing from damage due to~~xcessive heat.
II I Answer~ Ic I !Eli(l;t.;i~"~1 IR Ilc~~ ..... ", Level i IMemory I Faci!jty: 1Salem 1 & 2 1 IExamDate: II 12/3/20121 i~1194001G128 I_2:1.2r=JIROVa~14.11 ~~~c:lValue: '1 4 .1 I'section:lI PWG IRO Group:JI 1!iSRO Group:! 1 11 !~~I! 0
['------------""'
~}'$temJEvolution Title. I 1 GENERI
- IKAStatement:] r:--~-:-:----~-::-~-':""""'"--------:-:--~---------------:'
Knowledge of the purpose and function of major system components and controls.
IEXpllmation ofl 55.41.b.(3) The RCP thermal barrier prevents hot reactor coolant from flowing up the shaft to the radial bearing and RCP seal IAnswers: package when normal seal injection flow is lost. The normal flow of cool seal injection water past the radial barrier into the RCS
~- ~ normally performs that function. In the event of a loss of seal injection flow, reactor coolant would flow up across the tubes of the heat exchanger(cooled by CCW) and into the pump to provide the coolant and lubricant for the radial bearing and seals. A is in,...,.,,,,,,,...t h",,...,,, "' tho th",,,,,, "I h" .... o. i" n,.,t * .,,, ., fl. i",it",. R i" in,...""",..' h, ,,....,, '0 th",. Th",.", R.,,,;, ,-l, '" ,. "'1"11"11 Iseal injection flow. D is incorrect because the thrust bearing is not in the controlled leakage seal package
~============~======~
Reference Title j C F:acility Reference Numb~ lR~!erenc~~~~tion iPageNo.- ~~
R=e=ac=to=f=C=oo=la=n=t
- =1 p=u=m=p=L=es=s=on=p=la=n===~II=N=o=s:=05=R=C=P=UM=P=-=10====::;1.:=1======~II17 I FI1=O=~
~I==============~IC=~========~lil~====~11 1_ _ _ _ _ _ _ _ _---'[1 1[--------111 I~I~
II_---l
.L~*c~_~~~=:J IRCPUMPE004 I l~aterial R~9,:,ired for Examination I II IQuestion Source:
- ..... --~-----
1 Facility Exam Bank I[§e~~on Modific:~~ion Method:ll Editorially Modified I Used During Training Program.j 0 L~~estion Source Comments: !Vision Q43989 made psychometrically .... a'<'" 'Y.
I
- iComment
.... .. 0 I I
I I
ROSkvScraperisRO Skvseraper 1 ",~ -' --,>,'
[Qlfe~t~on!op§] IRo67 I IWhich of the following conditions will REQUIRE the suspension of fuel movement in the Unit 2 Rx vessel? I I
~ Chemistry reports Rx Cavity boron concentration is 2499 ppm.
J
!b:J I.A NEO reports BOTH 100' elevation containment airlock doors are open.
I fe. IThe PO depresses Fire Outside Control Room on Unit 2 Control Area Ventilation.
I I
~_'-' Containment Radiation Monitor 2R12A fails causing a Containment Ventilation Isolation signal.
I l§s~~ Ic I !Exam Level: R I I 'Cognitive Level II Application I racil!ty: II Salem 1 & 2 II ExamDat~ I 12/3/20121
!~jI194001G140 112.1.40 J ~alue: i~ lSRO Value: :[ill" ,SectIon: il~ iRO GrouP:'USRO Group:IU IiIi I ~
f 1;"~"~,;"">';
D
~tem/Evolution Title J _ _ _ _ _ _ _ ,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _---'IIGEN@
~ Statement*]
Knowledge of refueling administrative requirements. I
!Explanation OfJ 55.41.b(11) C is correct because operation in the recirculation mode requires suspension of fuel movement. (SO.CAV P&L 3.6.3
'~sw~!= __ When aligned to FIRE OUTSIDE CONTROL AREA (Recirculation Mode),Core Alterations and movement of irradiated fuel is NOT permitted (TIS Bases 3/4.7.6)). D is incorrect because Containment Radiation monitors are not required to be operable for Mode 6 or Fuel Movement or Core Alts per Tech Specs. B is incorre'ct because the airlock doors are only required to be CAPABLE of h"inn chI It <>nri r<>n h" nn"n 1<::) nD_<::.T rai\l_nnn7 n<>n" A\, a ic . , h"r<>llc" th" r n l R li",it fn.-hnrnn . ic ')1 "lQ I ' I r I
I Reference Title I ! Facility ReferenceN~mb~ [Reference SectiOn! ~ge NoJ ! Revi~
ISalem Tech Specs II ~I II II I I Refueling Operations-Containment Closure II S2.0P-ST.CAN-0007 ~I II 1127 I IControl Area Ventilation Operation II S2.0P-SO.CAV-0001 ~I II II I 1,"-----
&0. Number i I REFUELE012 1------1 IMaterial Required for Examination .J I II
[QUestion Source: I [ Facility Exam Bank I!Question Modification Method: I Direct From Source I iUsed During Training-Program i
~ ..
I D
~u~~~i~~~Source Commentsi I Vision Q110741 I
~mment J I I I I I I
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IQuesti()n~~~i~J I RO 68 I When performing a Surveillance Procedure, the operator encounters a step which has a "$" under the line where the operator initals completion of that step, What is the significance of the "$" sign?
la. : 1.lt identifies an item required to meet Salem UFSAR acceptance criteria, which if not satisfactorily completed should be brought to the attention of the SM/CRS upon completion of the surveillance. I
~ lit identifies an item required to meet Technical Specification acceptance criteria, which if not satisfactorily completed should be brought to the immediate attention of the SM/CRS.
I
§J tit identifies a step which requires Independent Verification of its completion PRIOR to continuing to the next step.
I
[ ] lit identifies a step which requires direct oversight by an assigned Reactivity Management SRO.
I
[Answer i Ib I Exam Levell IR I [cog~iti~eie;el-'ll Memory ~ ~acilitY:-ll Salem 1&2 IIExamDate~ I 12/3/20121
[KA: 11194001 G212 I' "
1~*2.12 n :RO ValUe] I :~ iSROValue: :14.1J lSection] PWG IIRO Group:'[JlsRc:> G~0l:l~ I 11 H~~ D
[System/Evolution r~
.-~~-~---~
_______ ,_ ____________________----'1 ~f\J~~
--~-.-
'KA Statement"l .~
Knowledge of surveillance procedures, I
!E~planation oil 55.41 ,b(10) Each surveillance proceclure which contains $ signs contains a Precaution and Limitation which states, .. "Steps c_,____.._,.. ,__,: identified with a dollar sign ($) are those items required to meet Technical Specification acceptance criteria. Such steps, if not IAnswers:
satisfactorily completed, may have rEiportability requirements and should be brought to the immediate attention of the SM/CRS."
The procedure writers guide also states" Step 4.5.4 DESIGNATE Technical Specification and UFSAR acceptance criteria as II fAII",,,,,'_~' . c>, ,~ . ""0 a dollat sioe ($l Ot Iba I,,,,.hn;,.,,,1 .. ,H"n n, ,mho. ",I Iho "Ion "ho.o Iho """ '0;" ro;"on ".
where verification for' that value is checked. - UFSAR, use a cent symboi (¢) or the UFSAR number at 'the step where the value is given or where verification for that value is checked.
.. ~ Reference Title I((((-Facility Reference Nu~~~~~J §~ferenc~_~~Ctio~=J [}t!~_-~.?J [~~~is,~~~l I IMPLEMENTING PROCEDURE WRITERS GLiI II AD-AA-101-1003
~I 11 29 112 I I Ie ~I II II I I IC ~I II II I l'::-<2._~_~~_, __ ,_~
I SURVOOE006 ~
I I I I IMaterial Required for Examination II II
[Question Source: II New I ~ues~ion Modification Method~1 I [Used During Trahling Progranil D
[§uestion Source co~
.~-~.----~-
I I
!Comment ._- I I I I I
[ I
[Question TOP~
Given the following conditions:
- Unit 2 is operating at 100% power.
- Excess Letdown is in service due to a problem with the control circuit for 2CV35, VCT 3 WAY INLET V.
- I&C troubleshooting is in progress on 2CV35.
- The RO reports that 2CV35 has just swapped to thE! Flow to HUT position.
- The RO also reports that during the pre-job brief for the 2CV35 troubleshooting, it was stated that 2CV35 actual position would NOT be affected during the troubleshooting.
Enter S2.0P-AB.CVC-0001, Loss of Charging to address the unanticipated CVCS system lineup.
Contact the WCC to initiate a tagout for 2CV35 since the troubleshooting needs to have the valve deactivated.
Contact the I&C Supervisor and stop work on 2CV35 based on being outside of Procedures, Parameters or Processes (OOPS) lAW HU-AA 101, Human Error Prevention.
to maintain status control since the Component Off Normal and Off Normal Tagged report System/Evolution Title KA Statement"l Knowiedge of the process for managing troubleshooting activities. I
~XPlanation Of] 55.41.b(10) Excess letdown does not flow through 2CV35, so its movement will not affect RCS letdown, and AB.CVC-1 will not address the valve movement. While a tagout may need to be initiated, it would not happen without stopping the job in progress and Answers:
finding out in more detail from I&C 110W the valve could be affected and wilat the blocking paints needed to be. OOPS requires stopping the job because they system realignment was stated to NOT going to occur during troubleshooting. Status control does I n"t ",.... "iro ,.,., ""0,.,., on' "f I . . (
i",,' ." ""i...."
x . .
ith tho nTf fl.1 ",.,.,'" th, riff fl.1.... ".,..,.,1 " ...."iti.... " """Irl ho ."rl",torl i" <::AO I,m,m
, Reference Title Jl¥.~~ility Reference Number I'Reference Section ....Jit Page No. ! ,"~*~*~"I ...
IHUMAN PERFORMANCE TOOLS AND VERIFII§-AA-101 II II 11 8 I I [I - II II II I Ii I 11= II II II I IM;;rterial Requir~<ff9J"i~~in~!9n II Ii Sourc~f~l IFacility Exam Bank I ~~uestionmmMOdiiication Method: II Direct From Source lillli!!1 o.u.r!,I1,~ ~..:aining Program,_! C
[QueStion SQurc~~?~~~!~ IV_i_i~sion_'Q_tI2_570_7_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _--1J,
!Ci........."'... i I
I I
Or q t? .. t The Unit 2 control room receives a call from the Rad Waste Operator, who states that an isolated Gas Decay Tank in hold-up has lowered in pressure from 90 psig to 40 psig over the last 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br />, and continues to lower slowly_
Which of the following would provide confirmation that tank ressure has lowered vs instrument failure, and wh is confirmation im ortant?
Display the 2R41 D trend reading on 2RP1. An unapproved release Display the 2R41D trend reading on 2RP1. An unmonitored release may be in progress.
Direct Rad Pro to locally retrieve trend data for the Area Monitor closest to the GDT area. An unapproved release may be in progress.
Direct Rad Pro to locally retrieve trend data for the Area Monitor closest to the GDT area. An unmonitored release may be in progress.
IApplication ;ExamDaieJ I 12/3/20121 i'
Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. I ExpJanationof~ 55.41.b(11) R41 (plant vent) monitors in control room have trend function which can display historical data. If a release is in
!~rtS~!~:,~~J progress it's being monitored, but is unapproved. Area Monitors do not have trend functions locally, but some of them are trended on P-250 computer (R4 and R34)in control room.
~ Reference
._. Title ~-F.acilit¥ft~ference Num~ IRefe':en£~ Secti~~~-; I Page No. ' IRevisio~l IRadiation Monitoring System Operation II S2.0P-SO.RM-0001 II II 11 37 I I II 1\ II II I I II II II II I
~LO.............."'.
- RM~nn()I=;007 I
I I
i Sv.... , w . I_N_e_w_ _ _ _ _ _-'llQuestion,Modiflcad9nMetho~--_ _ _ _ _--'llJ$ed During Training Pri,,,-i~rn ! D
-QI.Je~tlon Source Comments I I
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1 I
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1~1J!~ti(m TOplCll RO 71 I Given the following conditions:
- Unit 2 is operating at 100% power.
- 2R19A, STM GEN BLOWDOWN RAD MONITOR, rises to the ALARM setpoint and continues to rise.
Which of the following describes how the 21-24GB4s, SG BID OUTLET ISOL VALVES respond, and why?
IONLY 21 GB4 shuts to minimize the spread of contamination from a Steam Generator Tube Rupture (SGTR) on 21 Steam Generator to secondary systems. I 121-24GB4S shut to minimize the spread of contamination from a Steam Generator Tube Rupture (SGTR) on 21 Steam Generator to secondary systems. I IONL Y 21 GB4 shuts to prevent cross contamination from 21 Steam Generator to any other Steam Generator through the unaffected Generators blowdown lines.
121-24GB4S shut to prevent cross contamination from 21 Steam Generator to any other Steam Generator through the unaffected St Generators blowdown lines.
fAnswer II a I !Exam Level i IR I 'Cog-nitive Lev~J II Memory I Facility: , ISalem 1 & 2 I :EXam~at~:ll 12/3/20121
~I194001G314 1 2 .3 .14 R9~13.4I~o-va:itleJ 3.81'Section:IIPWG IIROG!OU~~I 1IISRoGroup:1! 11 1,11 0 i,...*
,~.,~.~ ....",.... -" .*" Title I I [GENERL1
-KAStatement* ' rKnowledge
"""'-"-'---~-- ---- ---------------------------------------:
of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.
IExplanation of I 55.41.b(11) B is incorrect because only the affected GB4 is automatically shut on the Hi Rad Alarm, it is plausible because ALL iAnswers: J GB10s, GB185s, and 2GB50 shut on Hi Rad WARNING. Isolating the blowdown path from the S/G to the condenser will prevent the
, spread of contamination, and also wi,] prevent any type of release from the main condenser to atmosphere. Each S/G has its own blowdown line, so backfeeding contamination is not possible through the blowdown lines.
! Reference Title ----~==J ~-FacmtYReferenceN~-;ber--: ~erenc::e Se~if:l~ : [page No.1 ,Revision~
IRadiation Monitoring System Operation II S2.0P-SO.RM-0001 II II 11 37 I I II II II II I I Ie 11 _ _ _----111 II I
[L~9~~~~~==
IABSG01E003 I I I I I IMl:\t~ri~IRequired for Examination ml I
II
.QiieSt!onSource: :IFacility Exam Bank 11§~es~icm M()~ification Metho~ Significantly Modified 1 Used I:l.uring Training Program 10 Q,::~ion Source tommen§ll Q78166 modified to make a 2 and 2 vs just a why do we isolate blowdown I
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,Question T()pic" IRo72 I IWhich of the following Area Radiation Monitors is one that is checked in 2-EOP-LOCA-1, Loss of Reactor Coolant, when determining if a LOCA outside Unit 2 containment Is occurrlnll? I 12R10A, """"<>VII""" Hatch vUrtlCllnrrlent 100',
I 12R34, Mechanical Penetration 100'.
I 12R47, Electrical Penetration.
I
[J 12R52 , Liquid PASS Room.
I IAns,,!~'"J Ib J i.Exam Level II R I ~c:;()!;Jnitiv!__~ev~J 11 Memory 1 IFacility: II Salem 1 & 2 I lexamDate: II 12/3/20121
[i<A:'IJ 194001G315 1'2.3.15 mm] fRo'value:! j ~~SRO VC1I,:,~j 3.1 IrsecliQn:lI PWG 1[RO G~~p:1 OJ [SROGroup: I 11 Ell!
,~*****************-***~-~~N"'''''
l~r:>te'!!'Evolllti()l1litle , I "GENERI j KA Statement*
Knowiedge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment. etc.
!Explanation of ' 55A1.b(11) LOCA-1, step 16, checks for radiation outside containment by looking at 2R4, charging pump area, 2R41 0 plant vent Answers:
~' --- -.~---~-......:
process, 2R34, 1R3 Radio Chem lab area, 1R6A Sampling room, 1R20B counting room.
i[Page NO] [Revision'
-~
Reference Title * * !.,-----~~~ . Facility ~.;:;:,. ~ ..~~ _""lj!:"b.er * ,Reference Section
- .- ..... ----~- ~----
ILoss of Reactor Coolant 112-EOP-LOCA-1 I I 11 28 I IS2.0P-SO.RM-0001 II Radiation Monitoring System Oper I Attachment 1 132 -34 11 37 I I iI II II II I
~O~~IJ~_-==:J 1RMSOOOE006 ILOCA01 E007 ILOCA06E003
[Material Re5luiredforExaminati(;)'!_J I II
'Question Source: "" I Facility Exam Bank I,Que~t!on Modification Method: I Editorially Modified IIUsedD'lI~il1~Training Program-j C
_Qu~~ti~~~_o.U!:e comment~ ~~~~~e~125697, replaced I distracter with 2R10A (used to assist in determining LOCA is occurring IN cont, not I
[comment
-- ~
I I I I I I
SRO Svs~volution E~t~1 Outlim~ chanQtlGI Given the following:
- The unit has been tripped and Safety Injection initiated due to a LOCA.
- The STA observes a PURPLE path displayed by SPDS for the CORE COOLING Status Tree. with no other RED or PURPLE paths on SPDS.
- The SMM is blinking dashes on both channels.
- SPDS is displaying question marks for all CET's.
- Plant Computer CET indication shows all CET's between 50-60 e F.
Local CET Display is unavailable.
Which of the following identifies how RCS saturation tl3mperature will be determined lAW 2-EOP-CFST-1, Critical Safety Function Status Trees?
RCS pressure (PI-403 or PI-405) will be used in conjunction with CFST Subcooling Tables.
PZR pressure channels (PI-455A, 456. 457 or 474A) will be used in conjunction with Steam Tables.
PZR pressure channels (PI-455A. 456. 457 or 474A) will be used in conjunction with CFST Subcooling Tables.
[~Clm Le~~ ~ Level ! Memory Ifaciiiti! ~~
,-------, 'RO-val~ [12] [SROvalue: '!3.9 I rs~ctiOrl:11 PWG I ~O ~roup::! 1! ISRO Grou-i):ll 11
'MStatement: i , - -_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _--.,.
Ability to identify post-accident instrumentation. I Explanation ofl 55.41.b(7.10) EOP-CFST Basis Document. page 2 of 18. "The SMM should be used to determine RCS subcooling, If the SMM is Answers: j inoperable, then calculate and log RCS subcooling on Table D. The value of T-Sat is obtained by using Table A for Normal Containment or Table B for Adverse Containment." Table D footnote states ..."*RCS temperature- Use CET's (WR Thot RTD's if CET's are not available)."
r======R;;;;e;;;;fe:re:;;;;n~=c~=e~-.=TI~=lcctl:e::::~:*:::::=:i ;:I=-;; ;'-=Fa;; ;C' ;;;;i;;;;lit~{~!!~!.!"-~~~~ ~rence SeC!ion .J i PClg~ :Revisionl
~IC=rit=ica=I=S=af=ety==Fu=n=ct=io=n=St=at=us=T=r=ee=s~~~1~12=-E=O=P=-C=F=ST=-=1============~I~I============~I~!====~I~I==~l
~I==============~ilF==========~~I!~======~I~!==~IFI~I 1----------'11 --.-J[_ _ _-.lll_--JII_--Il
'QP~~ti~es
"""""""""'~'"' ,
!'Material Required for Examination !I 1 IQuestio~:C:>lJrce:11 Facility Exam Bank IIQue~ic:>~Modific~tion Method: !Significantly Modified Ilused ~~Traifling Progralllj C I IVision Q811 04.
[Question Source Comments I.~....--.~ saturation temp.
Used stem and expanded answer required from generic "what to do next" to how to determine RCS I
-.... I e"........... c I Similar to DC Cook 4/29/2004 NRC Exam questior I I I I I I
!QUestion Top§] IRO 74 I I When operating in the EOP network, which of the following describes conditional operator actions, including interprocedure transitions, that are apelicable at all times while the erocedure is being implemented? I I
l~:.j Conditional Action Step.
I
~j 1Continuous Action Step.
I lc;] IConditional Action Summary.
I
[] IContinuous Action Summary.
I IAns';;erll d J [!:xam Levell IR I 'Cognltive L.evel] IMemory II Facility:] ISalem 1 & 2 I ~xamDate:ll 12/3/20121
~11194001G419 1[2.4.19 J IRO Value: I~ ISRO Value:. ED r::~ectiOnTll~ ~O Group:1[J]lsRO Group:! [ J ]
I I
,;;5;;~3. 0
[~~!.~'.!!~,,~~ion "!i..t~~J 1_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-'1 '9~N-_E-R-~J
~tatement_J Knowledge of EOP layout, symbols, and icons. I
[exPlanation o~ 55.41.b(10)4. EOPs have a Continuous Action Summary (CAS) in the upper left corner of each flowchart sheet. The continuous
,Answers:
~ _____ ._""'." __.... _ .* _,.,J
~ action summary contains conditional operator actions, including interprocedure transitions, that are applicable at all times while the procedure is being implemented.
! Reference Title I Facility Reference Numb~ I Reference Section : [Page No. ! ! Revision!
IEMERGENCY/ABNORMAL OPERATING PRO I~-SA-108-101-2000 ~I 11 28 11 0 I IREACTOR TRIP OR SAFETY INJECTION AND II NOS05TRP001-007 II 1117 117 I I II ~I II II I
.0. N;';-mber~
l_____________J L.,
Objectives ITRP001 E001 1 I ~
I I iMaterial Required for Examination t I II
~uest~on~ource: II New I~uestion Modificati~n Meth'~d~ . JI I~ed During Training Progr~ 0
[~~est~~~~~lI.~~~~~~me.':'_!~1 I I
~mment I 1 I I I I I
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!*Clue!~~~~~~~l I~~~________________________________________________________________________________~
In which one of the following Unit 2 procedures are the Emergency Diesel Generator FIRE EMERGENCY BYPASS Keylock switches directed to be S2.0P-AB.FIRE-0001, Control Room Fire Response.
S2.0P-AB.FIRE-0002, Fire Damage Mitigation.
S2.0P-AB.CR-0001, Control Room Evacuation S2.0P-AB.CR-0002, Control Room Evacuation Due to Fire in the Control Room, Relay Room, 460/230V Switchgear Room, or 4KV Switchgear Room.
~ I'D' ~~}~ ~ Level IMemory I :FaCiiity:: ISalem 1 & 2 I 'ExamDat~ I 12/3/20121 lKA: II 194001G427 I~~:_=J ~l!:~IB[SROva~rm~~~I~ fRo GioupjD~o Group;1D
['_.m~~-.-,
LSystem/Evolution Titl~J fKA' Statement: .--~j Knowiedge of fire in the plant" procedures. I Explanation 'Ofl 55.41.b(10) AB.CR-2, Attachment 4, Reactor Operator, pages 15, 19, and 22 place the keylock switches in bypass. The 3
-Answers:
-... --~
distracters do not contain steps to perform this action. The bypass switches remove the SEC control from the EDG control, and are only operated when the control room has been evacuated due to a fire and SEC operation may be aberrant.
Reference Tljle I Facility Reference.NlJm~<~ :Referi~ce Section _ : [i:"!;Ie No** [RevisiO~
II 1Control Room Evacuation Due to Fire in the Con S2.0P-AB.CR-0002 II Attachment 4 i\15,19,22 1127 I I II II II II I I Ie II II II I
[LO. Numbf!.r~ ___ "_ mml IABCR02E003 I I I I I l.Mat~rial Required for Examination iI
~~-~(:)IJ-"~ I Facility Exam Bank Iiti~estion Modificationlll\pthod:* :1 Direct From Source 1~~~gD~ring Traillirig.£>~ogram !
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I I
I
u.s. Nuclear Regulatory Commission Site-Specific Written Examination I~~
/
~
D Applicant Information
..-- .. ~- .. .. - .. ~~-.- .. ..~.-.---~.
Name: ! Region: I i
Date: 12/3/2012 I Facility: Salem 1 & 2 License Level: SRO Reactor Type: W I
1 - - - - - . - - -..- ..- .. ._-_ .. ..~- ..- . -..
Start Time: I Finish Time:
i Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> to complete the combined examination, and 3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br /> if you are only taking the SRO portion .
-~ .. -~- .. -~.-~- ..- ..- ... ..- - ..- - -.. ----.~--.-------
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results i
RO/SRO-OnlylTotal Examination Values
- - Points Applicant's Score
-- I --1 Points r---" .. -------~-.--------
Applicant's Grade - - 1 Percent
i'Rq:~kfsiilp~fl ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~+/-Lc..=======';;;;;;:;;;j 1~~~g ooodilio",
[ - Unit 2 has performed a rapid downpower from 60% to 8% during a period of high Circ Water grassing.
Control rods remain in automatic.
- The Main Turbine remains online.
- Control rods begin to withdraw in automatic.
- Rod speed is 16 spm 1Nith Tavg at 545°F.
Which of the following describes how should the CRS res ond?
Enter S2.0P-AB.ROD-0003, Continuous Rod Motion, because control rod speed is higher than expected. Place control rods in manual and adjust Main Turbine load as necessary to return Tavg to 1Nithin 1's"F of program.
Enter S2.0P-AB.ROD-0003, because control rods should not be 1Nithdrawing. Place control rods in manual and adjust as necessary to return Tavg to within 1.5°F of program.
Lower turbine load to restore Tavg to within 1.0°F of program lAW S2.0P-AB.LOAD-0001, Rapid Load Reduction.
Trip the Main Turbine lAW S2.0P-AB.LOAD-0001.
!s ! iCogn[tive~L~vei.i IComprehension.J i~acmty:i Isalem 1 & 2 I iExambate~ I 12/3/20121 1000001A203 IROyalue:i@l~~L~~edion:!I~~~[JJ~i§'OGroGP:]Q 21
[System/Evoruti~~~ !Continuous Rod Withdrawal 001 I
'KAstatement:J Ability to determine and interpret the following as they apply to Continuous Rod Withdrawal: I
, Proper actions to be taken if automatic safety functions have not taken place !
Explanatiorlof i 55.43(2) Rod control would normally be placed in manual upon reaching 20% power in AS. LOAD. The stem states that is has nOT
!Arlswers:~ been placed in manual. ~ermissive P-2 actuates at <15% turbine laod, and prevents automatic rod withdrawal. Control rods should NOT b.e WithdraWing. A IS Incorrect. Even If rod speed higher than expected (expected should be zero and its not). the action IS incorrect because ROD-3 says to adjust control rods to reston, Tavg, not adjust turbine load. B is correct. C and D are incorrect
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I reduction procedure. Additionally, taking actions to adjust or trip trip the turbine do not address the rod failure.
_-..===._.,,-...=**=-=-. ~-R=e_fe_r~-n-C_~e:..T-=it...*-le~--==~ . . ---"C",_J r:;-Y~i~I'l~~rn.E~ ~rence Section I~ageNQ.l i_R_'~_V~iS_i~
IFC""o""n""ti:;;:nu=o""U:;;:S:;;:R:;;:O:;;:d=W""i:;;:th=d:;;:ra""w""a""l======;;;;;:'11 S2.0P-AB.ROD-0003 III! , 121 l====;
II=R=ao:::.id=L=0=ad=R=e=d=uc=tio=n========:;.lls2.0P-AB.LOAD-0001 =::.:;.iI~======.::;.q ! 11=1=7=:::;
1_ _ _ _ _ _ _ _---111 --.JI II 11_--,
r~==:J IABROD3E002
[ RXPROTE013
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!Que$tion Topici
-' - -'-'-~, .... -~.,-'-'-'
I SRO 2 J Given the following conditions:
- Unit 1 was operating at 100% power, EOL.
- A 70 gpm RCS leak in containment was identified.
1- Operators tripped the Rx and initiated a Safety Injection lAW S1.0P-AB.RC-0001. Reactor Coolant System Leak.
- While performing EOP*TRiP*1, Rx Trip or Safety Injection. Step 23, PZR PORV STATUS.
the RO reports that PZR PORV 2PR1 indicates open.
. The RO reports 2PR1 will not manually shut.
- The RO reports that 2PR6 PZR PORV Block Valve will not shut.
I Which of the following identifies how the CRS should proceed?
~ IContinue in TRIP-i. The combined size of the PZR PORV and RCS leaks will allow a transition to TRIP-3. Safety Injection Termination.
I I Continue in TRIP-i. A transition to LOCA-1, Loss of Reactor Coolant will be made based on rising containment radiation monitors.
I I
ITranSitiOn to LOCA-1. A transition to TRIP-3 will be made after PZR PORV status is checked again in LOCA-1.
I I
I I!,':--
Transition to LOCA-1. SI Termination criteria cannot be met with either PORV not fully shut.
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lKA:] I000007G120 ]2_1:20
~========:
~ystemlEyorution Title! I_R_e_a_c_to_r_T_ri.:.,.p_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _...... _~~.,,-'
,-.-. *-**-*~~-l KA Statement:J
~---'-
i Ability to interpret and execute procedure steps.
I rExplanation of: 55.43.(5) A 70 gpm leak would require SI as stated in the stem, but charging pump flow would be able to maintain both PZR level
'Answers: 1 and ReS pressure. If a PORV is op13n in TRlp*1 and it or its block valve cannot be shut, a transition is made to LOCA-1 because I
I," I the EOP doesn't know how big the PORV leak is. Once in LOCA*1. there are steps redundant to those performed in TRIP-1 to attempt to dose the PORV or its block valve, then it continues(there is no transition to another procedure based on PORV/block
,ft, 0('\0\ <::,
. . ("v"' IL1 <::1 "", 0 ".... , , * ., rh",,..k,,rl All *. '" "h...."lrl h", II met which are: Subcooling >O°F, AFW flow/Adequate SG NR level. RCS pressure stable or rising, PZR level >11%. A is incorrect because while a transition WILL be made to TRIP-3, it is not made in TRIP-1, it is made in LOCA-1. B is incorrect because while a transition to LOCA-1 might be made based on rising containment radiation levels, the transition must be made now. 0 is incorrect because the small size of the RCS leak combined with the small size of the PORV opening(as shown by normal RCS pressure). SI termination criteria can be met.
1 J L~age No-'1 [Revision[
~
--~ ~.
I: ' '. Reference TiUe'2 Facility Reference Number I iReference Section
!Rx Trip or Safety Injection 1l1-EOP-TRIP-1 II II 1127 I ILoss of Reactor Coolant 1!1-EOP-LOCA-1
-.JI II 11 25 I I il I! II II I
[L:O.'Nurnber -===:J
!TRP001 E009 I I
'" ~ ,"-<
T:ri"=~tiA~ TO~iC;lrSR03 I Given the following conditions:
- Unit 1 was tripped from 100% power when a SBLOCA occurred.
- While performing EOP-TRIP-l, 11 SG was identified as being faulted.
- All actions have been completed to isolate 11 SG.
- After transitioning to EOP-LOCA-1, the crew is performing the Faulted SG Evaluation steps.
- 11 SG pressure is 740 psig and lowering.
- 12-14 SG pressures are all 960 psig and lowering very slowly.
Which of the followin£l identifies how the CRS should respond, and why?
Iother accidents (or their severity) in nrogn'l<:;<:;
~ Transition first to EOP-LOSC-1, then LOSC-2 because all SGs are now faulted. Faulted SGs require isolation because they may be masking I
Ionly Continue in LOCA-1 since the ECCS injection is cooling the RCS and causing the unisolated SG pressures to lower. Going to LOSC-1 would perform steps which have already been performed. I I Continue in LOCA-1 since the ECCS injection is cooling the RCS and causing the unisolated SG pressures to lower. The additional subcooling provided will allow an earlier transition to TRIP-3 during SI Flow Reduction steps. I I Transition directly to EOP-LOSC-2 because all SGs are now faulted. Steps to determine if SI termination can be performed will be adversely affected due to the lowering RCS pressure from the SG fault(s), and cause unnecessary procedure performance . I
- [Answer II b I [exam LeyelJ Is IC:Ognlti>.ie.l:!vel IApplication I iFac:Wiy:j ISalem 1 & 2 I 'ExamD~te:i I 12/3/20121 iKA: ,I 000009G244 1 2 .2 .44 *IRO Value:! I ~SRO ValUe] 4.4iSectiqn: jI~RO Group: 1 11 ~OGroup~ U :il~~; D ISystemJEvolution Title L
I ISmall Break LOCA 11 009 IKAS~tement~r- ______________________________________________________________________________________~
Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect lant and s stem conditions.
Expla~ationo~ 55.43(5) The procedure transition to LOSC-1 is made before LOCA-1 when in TRIP-1. The stem says all actions have been AnSwers,: '. ,.1 completed to isolated the faulted SG. These actions would have to be performed in LOSC-1. The transition out of LOSC-1 is
~... SGTR-1 if there is a rupture or to LOCA-1. The first step in LOCA-1 is to check for faulted SG's that have not been isolated. Cis I incorrect because RCS pressure :'ill still be lowering (not stable ~r rising) baS~d o~ the faulted 5G. A is in~orrect ~ecause the other I incorrect because there is no direct t~ansition to LOSC-2, you first ha~e to enter LOSC-1. . . I Reference Title " ',-Facilit¥' ReterenceNurnb-;~-'-"Reierence Section I. PCl~ [Re~isiOrl R=e:::::a:::ct:::or=T=ri::p:::o=rs:::a=fe:::ty=ln:::ie:::ct::;io=n======.ll1-EOP-TRIP-1
- .=1 II II I ;.=12=8==::;.
L=OS=S=O=fR=e=a=ct=or=c=o=ol=an=t======:::::!,1!1-EOP-LOCA-1 =:::::!,Hl=======,11 IF12=5=::::!,
---11
- .
- :i I_L_OS_S_o_f_S_ec_o_n_da_ry-'--C_O_ol_an_t_ _ _ _ _ _~If=EoP-LOSC-1 II 11_2_2_---'
l!::O:Number I LOCA01E009 ILOSC01 E004 i
1-----'
IiMat~rial Requiredfo~ExaminationJ I I[Question Source: ! INew ,I rquestlon'Modification
~
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Question TOP~ ISRO 4 I Given the following conditions:
- Unit 2 is in MODE 5 after refueling the Rx.
- The Rx was shutdown 20 days ago.
- 21 RHR loop is providing shutdown cooling.
- 22 RHR loop is aligned for ECCS .
1- 21 RHR HX inlet temperature is 105°F and stable.
- RCS pressure is 205 psig.
- 21 RHR pump begins cavitating due to a valve being mispositioned during a tagging release.
- The CRS enters S2.0P-AB.RHR-0001, Loss of RHR, and stops 21 RHR pump.
- During performance of S2.0P-AB.RHR-0001, the crew has time for normal restoration and local venting of the RH R system.
- 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the initial loss of RHR, 22 RHR pump is started.
- The RO reports RHR flow is oscillating between 1,500-3,000 gpm.
- The highest CET temperature is 184°F.
Which of the following describes how the CRS should proceed?
I c~:..: Start 21 RHR pump and initiate Attachment 7, Hot Leg Injection.
~ IStart 21 RHR pump and initiate Attachment 8, Cold Leg Injection.
- C.i IStop 22 RHR pump and initiate Attachment 7, Hot Leg Injection.
I
@] IStop 22 RHR pump and initiate Attachment 8, Cold Leg Injection.
~s~ Id I IEXam Level i Is I Cognitive L.evel IApplication I II I
I Facility:.; Salem 1 & 2 II.ExamDate: 12/3/20121
~ _ft?~§SRO valuel~ ~~~.~I~ iRO Group: U SROGroup:l
~~tem/Evolution
Title:
ILoss of Residual Heat Removal System KA Statement: I
!Explanation of!
[Answers: I Cold Leg injection. Attachment 7 Hot leg injection is not preferred due to RCS being intact and <200°F. With the procedure stating Ito stop all RHR umps, startin 21 is not directed.
Revision1 F===================~.~============~~~========~r==~I~18==~
~================~~============~~======~F==~FI==~
J------------------------~I----------------.--~I-----------~I----~I----~
iLO. Number IABRHR1E004 I I I
Concept Used ~
Vision Q80328 concept used, expanded to make SRO level. Changed 2 distracters to start idle RHR pump_
~sti?nTO~ I~S_R_O_ 5 _________________________________________________________________________________~
Given the following conditions: I
- Unit 2 is operating at 100% power.
i
- .
- ..-~T=he~R~o~r=e~po~rt~s~th~a~tt~h~e=p~Z~R~C~OI~d~C~a=1~le~v~el~c~h~a~nn~E;.:..31~(L=I~-4~6~2~)=is~i~nd~i~ca~t~in_g_O_o/_O.
.:tJhich of the following describes how the CRS should respond, and why?
____________________________________________~i Enter TSAS 3.3.1.1 Reactor Trip System Instrumentation. This Tech Spec is based on being able to provide the overall reliability, redundancy, and assumed available in the facility design for the protection and mitigation of accident and transient conditions. Place Enter TSAS 3.3.3.7 Accident Monitoring Instrumentation. This Tech Spec is based on ensuring sufficient information is available on selected plant parameters to monitor and assess thes':,.variables following an accident. Place an INFO sticker on LI*462.
No Tech Spec entry will be made since TSAS :3.3.1.1 and TSAS 3.3.3.7 are applicable to this instrument ONLY in MODES 4-6 and during movement of irradiated fuel. Place a single pie,~e of red translucent tape across LI-462.
No Tech Spec entry will be made since TSAS 3.3.1.1 and TSAS 3.3.3.7 are not applicable to this instrument Place an INFO sticker on LI 462.
! [~salem 1 & 2 1-~~:;:;;0at;~'-l-------1-2,-'3-/2-0-1......
21
~!.<A:JI 000028G243 I~-=:J IROVaiUe:l[~j[SROValuel~ :rmGroup:11 21 iSRO Group:] ~
~======:
- Sysiem/E"olution Title I IPressurizer Level Control Malfunction KAStatement* !
-_.. :J r----------------------------------------------:
Knowledge of the process used track inoperable alarms. I EXPla.nation oil 55.43(2)(5) The PZR cold cal channel is not included in the Rx Trip Instrumentation or the Accident Monitoring Tech Specs. It is
'Answers: I used winen RCS temperature is <200°F as directed in S2.0P-O.ZZ-0006, Hot Standby to Cold Shutdown. A is incorrect because TS 3.3.1.1 is not entered, although the Bases is correct. Also, the red translucent tape is used to identify an inoperable alarm.
I;~---.~.-,~--~R-*el\-e-re-n-cc-e-T-i-tl~e~-:---~---,J ~-Facmty Reference Number * ';1 iReferel1ce Section ,!p~ [Revision ISalem Tech Specs II 11 3 .3 .1.1 , 3.3.3. 7 11l======,II~=:::;1 1Control Room Instrumentation and Alarms II sC.OP-OL.ZZ-001 0 II II 11 10 I I_H_o_tS_t_an_db...:.y_tO_C_O_ld_S_h_ut_do_wn ____ ----<11 S2.0P-IO,ZZ-0006 ~ 1--------'11 I I_ _.....JI
- L.O~Number .~
!PZRP&LE010 I I I I I iMaterial RequtrEld for Examination' I! II iQtiesiionSour~~ I_N_e_w_ _ _ _ _ _ _......1iQues~i<>n Modification Meth0t3_______________.....J1 ;Us,ed Di.ll;ngTrairii rl 9 P[ogramJ D I' SourceC?rUJ.'.l"'IJ~! I I I---------------------------------------------~
.................... "~,;' '. " , ' ; ' .....*....
,I I I
i I
Given the following conditions:
- Unit 1 is operating at 45% power.
- 12 SG NR Channel I has been removed from service while undergoing a Channel Calibration lAW 51.IC-CC.RCP-0045, 1LT-529 #12 Steam Generator Level Protection Channel I.
- 12 SG NR Channel IV fails high.
Which of the followin describes how the CRS should respond?
Enter S2.0P-AB.TRS-0001, Turbine Trip <P-9. Place control rods in manual and insert to lower Rx power <5%.
Enter S2.0P-AB.LOAD-0001, Rapid Load Reduction. Initiate load reduction to less than 50 Mwe.
Enter S2.0P-AB.CN-0001, Main Feedwater I Condensate System Abnormality. Trip the Rx.
Enter 2-EOP-TRIP-1, Reactor Trip or Safety Injection. Verify the Rx has tripped,
- AnsvieilI c i ~~am~evef,j Is I[Cogni~)le_Level i IApplication I[Facility: II Salem 1 & 2 I[ExaIllD~te; i I 12/3/20121
~l 000054A205 ) i~2,05 *!R9Yalu4: 1@~~~L.1!J ~~l~ [ROGrou~Dl~_~D
!System/Evolution TItiel iLoss of Main Feedwater 11 054 fKA~l11ent:J Ability to determine and interpret the following as they apply to Loss of Main Feedwater: I Status of MFW pumps, regulating and stop valves I Explal'1ationofl A P-14 signal is generated by 2/3 NR level channels on 12 58 being :>67%, The P-14 signal trips the Main Turbine, trips the Main Ailswers: ! Feed pumps, and shuts the BF19s and 40s and 13s. (FW Isolation signal) The reactor does NOT trip on the Main Turbine trip <P-9
""~'-' (49% power) A is incorrect because AB.TRS does not direct a Rx trip is both SGFPs are tripped, only states to lower power <5%.
This is an incorrect action because it would force an automatic Rx on 10 10 5G NR level. B is incorrect because the turbine is i alcead\l triooed C is correct becallsP A 8 CN directs a Rx trio' . ,."" ,D_1 n 11 no ""',,., "'" rof h,,'h <::",I::D, h ,,... "" '"
I which it has from the P-14 (2/3 SG NR level channels on 1 SG ; 67%). D is incorrect because below P-9 the Rx has not tripped.
IIt',
,-.'--""--~".
. ',~eference Title
. ~ ~-'-"-,--.. "-,'-.. ,. - "---"] ,--~.~,
'_.'_f'aEil!!y~~~!,,~~er, -,~E~~~~J lf~.9.* :Revisionl jIMain Feedwater I Condensate System Abnorma II=S=2.=O=P=-A=s=.=C=N-=O=OO=1======I~I~======:=:;.11 ! 126 1\=========:::::I~I=r==-======~~Il=1 I 11 ______-'11___ ====:::::::.11\= ~~~~I ~I~~
'11_---'11_.. . . .
iL.O. Number IABCN01 E004 1-----1
~aterial Required for Examination 'i I !I
,! I II~========================~I
!I,--------------------------------------------~I
!-~---------------I
Given the following conditions:
- Unit 2 is at 40% power performing a shutdown.
- 4 SW Bay is isolated due to a leak on the 25SW3, 25 SW Pump Discharge Isolation Valve.
- Operators are performing the shutdown to comply with TSAS 3.7.4 because difficulties arose during the leak repair of the 25SW3.
- 2A EDG is supplying 2A 4KV vital bus for a scheduled surveillance.
- 21 and 23 SW pumps are in service.
Which of the following describes how the control room crew will respond if 2A EDG output breaker trips on Bus Differential, and 23 SW pump trips 1 minute later on over current when supplying SW flow at runout conditions?
Trip the Main Turbine to reduce Rx power and heat input to the RCS, and enter S2.0P-AB.TRB-0001, Turbine Trip <P-9.
Enter S2.0P-AB.SW-0001, Loss of Service Water Header Pressure. Trip the Main Turbine, and reduce Rx power <5% in order to place AFW in service.
Enter S2.0P-AB.SW-0005, Loss of All Service Water. Trip the Rx, confirm the trip, and stop RCPs to limit the heat input to the CCW system, and preserve RCP seal packages.
Trip the Rx and go to TRIP-1, Reactor Trip or Safety Injection. After exiting TRIP-2, Reactor Trip Response, enter S2.0P-AB.SW-0005 to perform compensatory actions for no service water pumps operating.
IAnswer: c I I iEx~1ll Level] Is J :£§gnltive ~.evei-ll Application .J IFacility: II Salem 1 & 2 1 IExamDate: II 12/3/20121
~~tiOn Ti!!.il ILoss of Nuclear Service "Vater 062 J<A.Stateme_n_*t_~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _--:"_ _ _ _ _ _~
I Knowledge of EOP mitigation strategies
\ !Explanatkmof I 55.43(5) A is incorrect because ev~~ if the actions are (some of those) performed in AB.SW-5, the next procedure entry to AB.TRB I
!Answers:1 is incorrect since the Rx is tripped in AB.SW-5. B is incorrect because AB.SW-1 doesn't perform those actions. C is correct. Dis
. -' incorrect because AB.SW-5 should be entered before exiting the TRIP series.
I"*** Reference Title Facility Reference Numqer J.* ~ference Section~ [""Page No.!
...-~~~
Revision
.~., -----------'
1 j Loss of All Service Water II S2.0P-AB.SW-0005 II II 114 I ILoss of Service Water Header Pressure II S2.0P-AB.SW-0001 II IIF===~I ~11=6=~
I________-----'H ______---Ill ____---ll! 11_--,
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!QuestionTopic
- Given the following conditions:
- Salem Unit 2 was operating at 100% power when a LOCA occurred.
- Operators are performing actions in 2-EOP-LOCA-1.
- 20 minutes after the Rx was tripped:
- RWST level is 19 feet and lowering
- RCS pressure is 350 psig and stable.
- Containment pressure peaked at 16 psig, and is now 13 psig and dropping slowly.
- Containment radiation monitors read < 1R / hr.
- RVLlS Full Range is 100%.
- 26 minutes after the Rx was tripped, the RWST 10-10 level alarm is received.
- When checking containment sump level at the beginning of 2-EOP*LOCA*3, the RO reports that containment pressure is now reading '1.5 psig.
identifies the HIGHEST ECG classification for these conditions?
'Answer' EJ '--~_~._. [Cognitive Level 'j !Application.J ISalem 1 & 2
~ ;Rova[ue:!@SROV~'uel~~~JI~ROGrOuP:;O;SROGrot.ip:IU
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iSystemlEvolution Title! [ Loss of Containment Integrity KA Stat~rrlent~ Abilit to determine and interpret the following as they apply to Loss of Containment Integrity:
Loss of containment inte ri
,Explanation of!'55.43(1 ,5) A LOCA which results ir, RCS pressure lowering to 350 psig will result in subcooling being lost. This yields 5 points due Answers: , to the loss of the reactor Coolant System Barrier. Containment pressure rise followed by rapid, unexplained pressure drop (13 psig
- - . - - - - - to 1.5 psig during a LOCA over 6 minutes) is 3 points for the loss of the containment barrier. The 4 point alert is if the loss of subcooling was not recognized and the loss of containment not recognized. The 5 point alert is if only the loss of the RCS barrier is
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IQuestion Top~ ISRO 9 1 Given the following conditions:
- Unit 2 has experienced a MSLB at the Mixing Bottle.
- All attempts at MSLI have failed, and 21-24MS167s remain open.
- Operators have just completed SI termination steps in EOP-LOSC-2, Multiple Steam Generator Depressurization, and PZR level is bein~l maintained stable.
- AFW flow to each SG is 1.0E4 Ibm/hr.
- The RO reports rising pressure in 22 SG.
Which of the followin~ describes how the CRS should proceed, and why?
~ Transition to EOP-LOSC-1 and stop RCPs if RCS pressure is <1350 pSlg, since RCPs cannot be stopped in LOSC-2.
I
~ ITransition to EOP-LOSC-1, Loss of Secondary Coolant, since one SG is now available for subsequent recovery actions.
I
,c.II.Remain in EOP-LOSC-2 since returning to EOI::l-LOSC-1 will require a transition to EOP-LOCA-1 upon completion.
I fdJ IRemain in EOP-LOSC-2 until positive control can be established over trle cooldown after the remaining Steam Generators have fully
'-.-J depressurized, then transition to EOP-LOSC-1. I
'Answer II b I I Exam Levell S I IICognitive Level I IMemory .-J lFacility: II Salem 1 & 2 1 [Exam Date: !j 12/3/20121
- KA: 11 00WE12A201 jEA2.1 IIRO Value: II...:~ ,SRO valueL_~~rSection: II~ RO Group:! I 11 SRO Group:11 11 ~!~ ~
,System/EvolutIon Title I [Uncontrolled Depressurization of all Steam Generators IIE12 IKA Statementl Ability to determine and interpret the following as they appll': to Uncontrolled Depressurization of all Steam Generators: I Facility conditions and selection of appropriate procedures during abnormal and emer~encl': operations. I
- Explanation of I 55.43(5) LOSC-2 CAS states that upon a pressure rise in any SG except when performing SI termination in Steps 8-20, GO TO
'Answers: -.I EOP- LOSC-1. The stem states that it is after Step 20. LOSC-1 Basis Document, page 7, states that .. "Any cooldown operations that are performed as subsequent recovery actions will require at least one nonfaulted SG."
Reference Title . iCFacility Reference Num~ 'Reference Section I! Page No*1 Revision!
IMultiple Steam Generator Depressurization 1!2-EOP-LOSC-2 ---.JI II 11 26 I ILoss of Secondary Coolant I~OP-LOSC-1 ---.JI II [ 1 23 I I I[ ---.JI II II I IL.O. Number ~
ILOSC02E005 ---.J I [
I I IMaterial Required for Examination II II IQuestion Source: II Facility Exam Bank I[Question Modification Method:. II Concept Used I,Used During Training Program 10 1Question Source Comments[ IVision Q57956 concept used, and added the "why" to question.
I iComment I I
I I I I I I
Given the following conditions:
- Salem Unit 2 was performing a Rx shutdown due to indications of failed fuel after chemistry reported reactor coolant activity to be 500 uCi/gm dose equivalent 1-131.
- During the shutdown, the RO reports lowering PZR pressure and level.
. With 21 charging pump in service, PZR level continues to rapidly lower, and the RO reports containment pressure is also rising.
- The RO trips the Rx and initiates a Safety Injection, and all equipment responds as expected for the SI, RCS pressure continues to lower rapidly to 35 psig.
- The SM declares a Site Area Emergency, and all notifications associated with the SAE have been made.
Which of the following identifies a condition which would require a subsequent notification of the NRC, and the correct time for that notification?
The wind direction shifts from 0° to 180", 15 minutes, Containment radiation level exceeds 2,000 R 1 Containment sump level indicated on 2CC1 has remained stable at 46%. 15 minutes, fd.l The control room must be evacuated and operators cannot access the Auxiliary Building due to high radiation. 60 minutes.
I !
Answer I Ib I jExam level I ~ IApplication .J [FacmtY:ll Salem 1 & 2 I iexamDate:'l !___ 1,..;;2,..;;/3,..;;/2,..;;0,..;;1.,;;.;21
=KA:='i========, 2.4.30 I:R()vahj~:-i@sRovarueJ 4.1!Section: 11~'ROGrOuP:{2J!SROGrCfuP:1U ~
IHigh Containment Radiation KA Statement' Knowledge of events related to system operationlstatus that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.
:"'....';'"....... " of 55.43(1.2) The 15 minute times in the question are from the 15 minute notifications required during emergencies to the states. The
.An",wers. NRC is not required to be notified for 60 minutes. The seconci knowledge part of the question is what would cause a notification to be required, i.e. a more sever E plan classification is made, The radiation levels> 2,000 Rlhr adds 2 points from the containment barrier. The stem states that the SM declared a SAE, which would have been under FB4.L and RB2.L each of which is 5 points, I Th", ? . . n"ink H'" ,In n, ,t th", I ,nit in '" r:;::
Th", lAIinn ",hift ,mil" in " "A;:: Ilf'II ,In nrlt "' liro " n,..,tifrt'"ti,m hot'", 1<," n" PAR would have been made for the SAE, The containment sump level is NOT expected. with containment pressure at 35 psig, the entire contents of the RCS are on the floor and level would have risen, as is seen for LBLOCAs. The 15 minute time for this condition is wrong, however, The CR evac and inability to establish control of the plant in 15 minutes is a SAE, and the plant is already in a SAE.
II Page No. : Revision:
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I--.Facility Reference Nunlber !iRefererlc~~ection Reference Title ,,"
Isalem ECG Ie I! II II 1 I II - ~! II 11 I I[ ~I II rI I
Question Topic j Given the following condition:
- Unit 1 is in MODE 5 during a plant startup.
- 11 RHR loop is in service.
- 12 RHR loop is aligned for ECCS.
- 13 RCP is in service.
- 11 charging pump is in service.
- RCS Tavg is 175°.
- RCS pressure is 310 psig.
- PZR level is 60%.
When placing the second RCP in service, RCS pressure momentarily rises to 390 psig.
Which of the following describes the RHR system response, and how the CRS should proceed?
8.* The 1RH3, RHR SAF RlF VLV TO CONTAINMENT SUMP opens. Enter S2.0P-AB.PZR-0001, PZR Pressure Malfunction, and ensure that any PZR PORV that opened in response to the RCS pressure has shut.
The 1RH2, RHR COMMON SUCT MOV automatically shuts. Enter S2.0P-AB.PZR-0001 and ensure that any PZR PORV that opened in to the RCS has shut.
The 1RH3 opens. Enter S2.0P-AB.LOCA-0001, Shutdown LOCA, and isolate letdown to minimize RCS inventory loss.
The 1RH2 automatically shuts. Enter S2.0P-AB.LOCA-0001 and isolate letdown to minimize RCS inventory loss.
'Answer! EJ '--___-' L--~ __ ~~_ ......., !Application ~ lFacili~11 Salem 1&2 KA-=11 005000A202 I,A2 02 RO Value: il I
3.51 fSRO value]3section: . II~ iRO Group:'1 11 :SRO Group:l 11 System/Evolution Title I IResidual Heat q",,,,vv,,,, System IOOs' I KA Statement: i 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 mitigate the consequences of those abnormal operation:
Pressure transient protection during cold shutdown i Explanationof 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
.AnsV'lers: apparent to the control room, but the PORV opening will. AB.PZR, Attachment 3, will ensure the PORV has shut. The 1RH2 has an OPENING interlock that requires ReS pressure to be <375 psig, then a keyswitch opens the valve. There is no automatic closure associated with this valve oro high pressure. AB.LOCA is used in MODE 3 and MODE 4 with the accumulators isolated, and j
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'. .* Reference "_Title
.. . IrFacility Reference Number : iReference Section I'
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IPZR Pressure Malfunction 1@*oP-AB.PZR-0001 II II 1 16 I IShutdown LOCA l@IoP-AB.LOCA-0001 .--JI II 11 6 1 IResidual Heat Removal I§i232 Sheets 1 and 2 .--JI II I 43.40
! 1
..c...:-J Given the following conditions:
II
- Unit 2 is operating at 15% power, performing a power ascension lAW S2.0P-IO.ZZ-0003, Hot Standby to Minimum Load, prior to rolling the Main Turbine.
- The RO reports the Bistable for Intermediate Range Channell on 2RP4 has just illuminated.
1- The reactor remains at power.
Which of the following describes how this bistables current condition will affect the power ascension?
I la. j This bistable illumination is NOT expected at this point in the power ascension. Power ascension may continue while investigating cause of
~ ! faulty bistable V!-O<lU. I
,I An ATWT has occurred, attempt to trip the rea,:tor manually. If the reactor does NOT trip, verify the turbine is tripped and initiate rod insertion, then go to FRSM-1. I' IThis bistable illumination is expected during a power ascension. BLOCK both intermediate range channels by depressing the BLOCK INTERMEDIATE RANGE A and B PBs lAW OHA F-17 ARP. Continue the power ascension. I l:!J IA failure of the IR high flux trip block has occurred. Lower Rx power to less than 5% and depress BLOCK INTERMEDIATE RANGE B I pushbutton to block Train B lAW S2.0P-IO.ZZ*0004 POWER OPERATION. I
!At)sv~el! Ia I IExam Levell !S I ICognitive Level] IApplication I [F;cility:! ISalem 1 & 2 IIExa~1 12/3/2012 1 KA: II 01 ?nnnA201 I!A2.01 IROValue:!1 3.11!SRO-,,:~r\J~L2~ES~CtiC>I1J1~RO~E()~1 1liSROGroup:11 11 l~~ ~
System/Evolution Tit~
IReactor Protection System ; - .
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,KAStaterl1.El!:!!.:J Ability to (a) predict the impacts of the following on the Reactor Protection System and (b) based on those predictions, use I erocedures to correct, control, or mi!igate the consequences of those abnormal operation:
Faulty bistable operation
.Explanation of I 55.43(5)A is correct because the alarm setpoint is 25%. The action is correct because the IR hi flux trip was blocked when Rx power
!Answer~ was >P-10. TSAS 3.3.1.1, Functional Unit 5, Action 3.c states above 5% power power operation may continue. B is incorrect because the IR Hi Flux trip is blocked and a Rx trip is not expected. The actions are correct for an ATWT lAW EOP-TRIP-1.
Distracter C is incorrect because it is not expected to occur at this power level. While the light WILL light at 25% power, its output is I
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IHot Standby to Minimum Load II S2.0P-IO.ZZ-0003 ~I 11 37 11 37 I
! II ~! II II I I Ie ~I II II I
'CO.'Number I RXPROTE012 IIOP003E004 IIOP003E005
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Given the following conditions:
- Unit 1 is performing a Reactor startup lAW S1.0P-IO.zZ-0003 Hot Standby to Minimum Load.
All Shutdown Bank control rods have been fully withdrawn,
- Control Bank A is fully withdrawn.
- As Control Bank B is withdrawn past 20 steps, the RO reports OHA E-48, ROD BOTTOM has just alarmed and remains locked in.
- No other alarms are received, Which of the following describes how the system is operating, and how the CRS should roceed?
The Rod Bottom Bistable causes OHA E-48 to alarm as each control bank is withdrawn past 20 steps and is expected. The CRS should direct the reset of the Failure to reset the then continue the startup.
Bottom Bistable causes OHA E-48 to alarm as each control bank is withdrawn past 20 steps and is expected. The CRS should RO to depress the STARTUP pushbutton on 1CC2 to reset the alarm, and continue the startup.
The Rod Bottom Bistable cleared when Contra! Bank A was withdrawn past 20 steps and OHA E-48 is unexpected at this time. The CRS should enter S1.0P-AB.ROD-0002 Dropped Rod and direct the opening of the Reactor Trip Breakers to terminate the Rx startup.
The Rod Bottom Bistable cleared when Control Bank A was withdrawn past 20 steps and OHA E-48 is unexpected at this time. The CRS should place the startup on hold and initiate S1.0P-AB.ROD-0002, Dropped Rod, or S1.0P-AB.ROD-0004, Rod Position Indication Failure, to I. _._~._.d~g,:::::~:~:~ti§==:r~~~!:z.:i! !Gern;&~e!'leR£ieRJ Facility:H Salem 1 & 2 HExamDate: H 12/3/20121 IRoVa'ue:lr ~lsRova'ue-@T!~~~I~IROGrouP:11 21ISROGroup;l! 21 System/Evolution
Title:
INon-Nuclear Instrumentation System Ability to determine operability and/or availability of safety related equipment.
The Rod Bottom alarm CLEARS when CB A is withdrawn past 20 steps. This is because a Rod Bottom Bistable Bypass for each of 1 the other three control banks B,C,D bypass the alarm for their respective bank when all rods in that group are below 35 steps. This ,
means that the alarm was CLEAR when it alarmed, it did not refJash. Since no other alarms occurred, the CRS should place the II startup on hold and enter AB.ROD*2 (which will direct entry 10:0 AB.ROD-4) or enter AB.ROD-4 directly to investigate the failure.
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!Rod Control and Position Indicating Systems La II NOS05RODSOO*11 --.-J! II 1111 1 I Dropped Rod 1 § .OP-AB.ROD-0002 --.-J 1 II II I IRod Position Indication Failure II S1.0P-AB.ROD-0004 --.-J I II II I
Given the following conditions:
- Unit 2 is operating at 100% power.
- Technicians are performing a sensor calibration of Containment Pressure Channel IV lAW S2.1C-SC.RCP-0066, 2PT-948A CONTAINMENT PRESSURE PROTECTION CHANNEL IV.
All bistables and test switches are in their proper alignment for the calibration.
- While I&C Technicians are performing the calibration, the control room receives OHA C-16, PHASE B CNTMT ISOL ACT.
Which of the following identifies what will occur, and how the CRS should respond to this alarm?
Phase B isolation valves will shut ONLY. The isolation will not be able tiJ be reset until I&C returns their bistables and test switches to normal. Trip the Rx, stop all RCPs, and GO TO EOP-TRIP-1.
Phase B isolation valves will shut and Containment Spray valves will open. Attempt to reset and open the Phase B isolation valves lAW OHA C-16 ARP. If unable to reset Phase S, GO TO S2.0P-AS.RCP-0001, RCP Abnormality.
Phase B isolation valves will shut and Containment Spray valves will open. The isolation will not be able to be reset until I&C returns their bistables and test switches to normal. Trip the Rx, stop all RCPs, and GO TO EOP-TRIP-1.
Phase B isolation valves will shut ONLY. Attempt to reset and open the Phase B isolation valves lAW OHA C-16 ARP. If unable to reset Phase B, GO TO S2.0P-AB.RCP-0001, RCP Abnormality.
Il~~ r:o- ~~ ~~~~~ I Application "1 ~YJ ISalem 1 &2 l-e~te:ll 12/3/20121 J<kl1026000G450 12.4.50 I ~~@;SRQ valuel~ .~~I~ IRQ Group:U ~ Group::U
~~em/Evorution Till;: I...:Co:..:.;.;n.:.:.ta...:in...:m;,,;,e.:.;n...:t...;s~p...:ra;;.::y...:s.:.;y~s;,,;,te.:.;m....:-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-l_._ _ _- '
IKA Statement-I Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. I
[Explanation of I 55.43(5) This alarm is not expected to occur during the sensor cal, as it requires 2/4 containment pressure channels to see 15 psig.
IAnswers: . J The ARP says if cont pressure is <15 psig, attempt to reset and open Phase B isolation valves, and if unsuccessful go to AB.RCP based on the loss of CCW to the RCPs. The CS valves will reposition .
I Reference Title' FacilityRefere~ce Number~Il !Reference Section .'.. Ilf':lge No.:l ~ViSion!
IOverhead Annunciators Window C II S2.0P-ARZZ-0003 II 11 19 1117 I I II ~I II II I I Ie II [I II I l.I::9- Number
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Question Topic.i !...:s...:R...:O...:...:15.:.-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _....l Given the following conditions:
- Fuel handling is in progress in the Unit 2 Spent Fuel Pool when a fuel assembly in the Spent Fuel Handling Tool is dropped.
- Gas bubbles are observed in the vicinity of the dropped fuel assembly.
- 2R5, Fuel Handling Building (FHB) radiation monitor goes into alarm and stabilizes at 25 mRlhr.
Which of the following describes the effect of this event, and contains actions that will be performed lAW S2.0P-AB.FUEL-0001, Fuel Handling Incident?
ALL Fuel Handling Crane motion is locked out to prevent further damage to an affected fuel assembly. Ensure all available FHB Exhaust fans Crane motion EXCEPT downward movement is locked out to prevent raising a damaged fuel assembly. Ensure the FHB Door is closed.
of the FHB high radiation condition. Evacuate ALL personnel from the FHB until Iter in service to prevent a release to the environment. Ensure the FHB nPT*"f'I.,n,,1 passage.
~*****IRO value::1 3.61 :SROvaluel~ rsection~!~RO Group:! I 21 'SROGroup:'1 21 SystemJEvr.lttH...." TItle! iFuel Handling Equipment System
'KA Statement:' Ability to (a) predict the impacts of the following on the Fuel Handling Equipment System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Dropped fuel element I. .
!Explanation of i 55.43(7) The 2R5 in alarm (11 mRlhr alarm, 7mRlhr warning) swaps the FHB exhaust ventilation to the Charcoal Filter and starts iAnswers: __ both FHB Exhaust Fans. The normal configuration for FHB ventilation is the single Supply Fan running, and BOTH Exhaust Fans I
running. C is incorrect because non*essential personnel are evacuated, as actions are required to be performed priorto evacuating ALL personnel. The alarm will actuate. A and B are incorrect because the FH crane only locks out as described in B with the I
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I actuation and a correct action.
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iQuestion Top~ ISRO 16 I Given the following conditions:
- Unit 2 is stable in MODE 3 at NOT, NOP, lAW S2.0P-IO.ZZ-0003, Hot Standby to Minimum Load.
- Rod Control is energized.
- 23 charging pump is in service.- 21-24MS10 Atmospheric Relief Valves are in Auto controlling SG pressures at 990 psig.
- Main Steam Dumps are in MS Pressure Control - Manual with 0% demand.
- 23TB40 fails full open.
Which of the following describes the initial plant response to this failure, and how should the CRS proceed?
rall23 Main Steam line flow will rise. Direct the PO raise the setpoints of 21 .. 24MS 1Os lAW S2.0P-IO.ZZ-0003 to restore Tavg to 547T I
ib. i IALL Main Steam line flows will rise. Enter S2.0P-AB.STM-0001, Excessive Steam Flow, and direct the PO to initiate a Main Steam line
- Isolation. I
~ IALL Main Steam line pressures will lower. Enter S2.0P-AB.STM-0001, Excessive Steam Flow, and direct the PO to depress the Train A and Train B Off & Reset Bypass Tavg pushbuttons on 2CC3. I i
- d. 123 Main Steamline pressure will lower until 23MS10 is fully shut, then stabilize. Initiate S2.0P-IO.zZ-0006, Hot Standby to Cold Shutdown
- and direct the PO commence a cooldown at <1QO°F 1 hr using the 21-24MS10s. I
,Answer II c , IExam Level II S IICognitive Level .1 Application ~ iFacility: II Salem 1 & 2 I :ExamDate: i I 12/3/20121 IKA: II 039000A204 I ;A2.04 IIROValue: I[~ !SRO valuel~ :Section: II~ ROGroup:ll 111SRO Group:!1 11~~lt~ ~
System/Evolution Title I I Main and Reheat Steam System 11 039
!KA Statement~ Ability to (a) predict the impacts of the following on the Main and Reheat Steam System and (b) based on those predictions, use
. procedures to correct, control, or mitigate the consequences of those abnormal operation:
Malfunctioning steam dump I iExPlanation o~ 55.43(5) The steam dump valve are arranged by condenser, not steam lines. 23TB40 goes into 23 condenser. The distracters with
,Answers: 23 Main steamline flow and pressure will occur, but the actions would not be performed per the lOP in effect. With the Rx trip breakers shut, entry into the Excessive Steam flow AS is correct, and the action in AB.STM with a malfunctioning Steam Dump Valve is to turn off both Trains of Steam Dumps. The action to initiate a MSU would occur if a Rx trip were required based on
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I Reference Title ii Facility Reference Number I iReferenceSection I i Page No. !Revisionl I Excessive Steam Flow II S2.0P-AB.STM-0001 ~I II 11 9 I I IC -JI II II I I I[ ---..JI II II I
- LO. Number 1 MSTEAME015 1-------1 IMaterial Required for Examination ! I II IQUestion Source: II Facility Exam Sank 1 ,Question Modification Method: ;1 Significantly Modified I IUsed During Training Program' I D
[Question Source Comments I I I iComment I I I I I I I
Given the following conditions:
- Unit 2 is in MODE 2 with a startup in progress.
- Welding in the Unit 2 Turbine Building has caused an actual deluge actuation to occur in the Unit 2 Turbine Building.
- The control room receives the following alarms:
- OHA A-7 FIRE PROT FIRE
- OHA A-15 FIRE PUMP 1/2 RUN Coded Fire alarm 2-2-1 TURBINE GEN AREA -88' ELEV roceed?
ONLY one diesel fire pump has started. Enter S2.0P-AB.FIRE-0001, Control Room Fire Response. Place BOTH Unit 1 and Unit 2 CAV in Fire Outside Control Area.
ONLY one diesel fire pump has started. Enter S2.0P-AB.FP-0001, Fire Protection System Malfunction. The Unit startup will be have to be stopped due to the current capability of the Fire Protection system being degraded.
BOTH diesel fire pumps have started. Enter S2.0P-AB.FIRE-0001, Control Room Fire Response. Place BOTH Unit 1 and Unit 2 CAV in Fire Outside Control Area.
BOTH diesel fire pumps have started. Enter SZ.OP-AB.FP-0001, Fire Protection System Malfunction. The Unit startup will be have to be stopped due to the current capability of the Fire Protection system being degraded.
rAns.. . .e r II a I 'Exam Level .1 S I [~ognitive Level .'1 Application .J iFacmty:ll Salem 1 & Z II~xamDate: ! I lSystem/EvolutionTitiel [..;..F... ro...t.;..ec...t;...;jo.;..;n....;s~y...
ir..;..e_p... s.;..;te...m_ _ _ _ _ _ _ _ _ _ _._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ L....****** _~.
IKA Statement~ to (a) predict the impacts of the following on the Fire Protection System and (b) based on those predictions, use procedures ct, control, or miti ate the consequences of those abnormal 0 eration:
iExpfanation of. 55.43(5) A deluge valve opening as stated in stem will cause FP system header pressure to lower to the point that #1 Fire pump will
[Answers: start at 85 pSig, and will restore header pressure. Each Fire Pump is rated to supply all fire protection needs. The second Fire Pump will NOT start, as its auto start pressure is set at 75 psig, When the deluge occurs, the CRS will not know it is inadvertent.
The CRS will respond to the auto start of the pump lAW ARP for A-7 FIRE PROT FIRE and A-15 based on the deluge valve
. _ * '.c. , ._
outside the CR. AB.FP is NOT entered, because there is no indication of a malfunction, but indication of a valid deluge valve actuation. The shutdown distracter is plausible because it is the action required in AB,FP if bot.., normal and backup fire protection systems are unavailable aHA A-15 annunciates when EITHER (or both)Fire Pumps start I
, Reference Title I~~Faci!ity Referenc;Number~'-: :Reference Sectl;;:;;: ~ !Pat;J~1'I<:>.j :Revision:
'"'l~~' -,'-j 1Control Room Fire Response 11 S2.0P-AB.FIRE-0001 II II 117 I I Fire Protection System Lesson Plan II NOS05FIRPRO-07 II II 117 I
[ Fire Protection System Malfunction II S2.0P-AB.FP-0001 --11 II liz I LO. Numb;i=r=J i FIRPROE004
Given the following conditions:
- Unit 2 has experienced a LOCA.
- 21 RHR pump has been CfT for the last 2 days.
Containment pressure is 14 psig and rising slowly.
- 22 RHR pump trips after performing Safeguards Reset actions in EOP-LOCA-1, Loss of Reactor Coolant.
- The CRS transitions to LOCA-S, Loss of Emergency Recirculation.
- The ST A reports a valid PURPLE path on Containment Environment with containment pressure at 15 psig and rising slowly. and no highElr PURPLE or any RED paths present.
Which of the following describes how the CRS should use Containment Spray pumps?
The CRS should start/stop Containment Spray pumps in .....
EOP-FRCE-1, Response to Excessive Containment Pressure, because Containment Spray pumps will not auto start with the SECs reset.
EOP-FRCE-1 because at least one Containment Spray pump is required to be running whenever containment pressure is >15 psig.
EOP-LOCA-S to establish minimum required CS flow in order to conserve RWST inventory.
fcill EOP-LOCA-5 since FRPs are not in effect when LOCA-5 is in effect.
I~~~====~~====~~==~======~==~========~====~======~
~~ D i'Ex am Level !Cogn~tive LevelJ I Memory I ~~ ISalem 1 & 2 I [~~~!5J 1------.. .
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~ystem/Evolu}i9n Titlej I_co_n,.;ta_in_m_e....n....t....s..;.y_st_e_m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _....1
'KA Statement::
jrK-n-o-w-le-d-g-e-o-f-th-e-b-a-s-e-s-fu-r-p-ri-or-it-iz-in-g-s-a-fe-~-fu-n-c-tio-n-s-d-u-ri-n-g-a-bn-o-nm-a-I/-em-e-rg-e-n-cy-o-p-e-ra-t-io-n-s--------------~
IIExplanatioll of 55.43(5) The transition to FRCE-1 is required upon a valid PURPLE path (i5 psig containment). A is incorrect because FRCE-i
- Ans..... ~--" specifically asks if LOCA-S is in effect, and if so. direct CS pumps to be operated lAW LOCA-S. B is incorrect because of the above reason. and additionally. using the table found in LOCA*5 as the bases, there are conditions with cont press>15 psig that NO CS pumps will be directed to be started. C is correct. Dis incorrect because FRPs are in effect after the transition out of TRIP-i. and i thOrQ;" nr> r!;ra.-t;nn fn '" .""onr! "op" o;th",r;n (')('JI.1 M {j('JI_c:.
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'Material RequfredforExamination' l I II
,IOIt4stionsource: j I Previous 2 NRC Exams IiQu~stion~odificationMethod: 1Editorially Modified I~UsedOutirlg TraJning Piogr~ 0 sourcecomrn~nts! I i [Question Vision Q77740 modified to ask what procedure 08-Q1 NRC Exam I
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Given the following conditions:
!- Unit 2 is in MODE 3 preparing for a startup.
- 21 RDMG set motor AND generator breakers are closed, and BOTH Reactor Trip Breakers A and B are shut for rod control testing.
Which of the following identifies how many Reactor Coolant loops are REQUIRED to be in operation lAW Salem Tech Spec 3.4.1.2.c, Reactor Coolant System, Hot Standby, and correctly reflects its Bases?
IFour, because single failure considerations recuire all loops in operation when rod control is energized.
I lone, because it is sufficient to provide positive pressure control of the ReS with a bubble established in the PZR I
'c.] 1Four, because it ensures DNB criteria are satisfied for any conditions under which control rods are not fully inserted.
I lone, because it provides adequate flow to enSJre mixing, prevent stratification, and produce gradual reactivity changes during boron concentration reductions in the Reactor Coola!::.! System. I IiAnsyver II a I rE~am L~\{.!!J IS I ,C"I:!.11ti.."Level i IMemory J 'Facility: I ISalem 1 & 2 I 'ExamDate; 'I 12/3/20121
)<kI1194001G132 1 2 .1.32 i iRQ Value: 'I 3.81 iSRovalu~L~~ !sectio£ll PWG IIRQ Group:] 1 11SRQ Gro~1 11 ~~!~j ~
lSysfem/EvolutioriTitle .:
I I :GENERI i lKA Statement::
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!Explanationof 1155.43(2) All of the choices have their reasons pulled from the Bases section of Tech Spec for RCP operation. The conditions in the'
!Answers:i Istem indicate that Rod Control is energized. With rod control energized, 4 Reps must be in operation. As per the Bases on page
~.... wI 83/4 4-11 it is for single failure criteria, B is incorrect because 4 loops are required. C is incorrect because DNB is not a bases for I Mode 3 RCP operation, it is a Mode 1 and 2 bases.. 0 is incorrect because it is one RCP, but has the correct bases for when only "no i::)('p ;" ",,,, ;ro";
L Reference TitleT Ii Facility ReferenceNurnber: ~Reference SeciiOlli TageNOl .~~
IFS=al=em=T=e=ch=S::::pe=c=s=========i-II 3 11 .4.1.2 113/44-2 I ~14=4=::;1 s=al=em=Te=c=h 1=1 =sP=e=cs=B=as=e=s======l-IIF========:;II~=3/=4.=4.=1======='1183/44-1 1 11=1=97=:::;1 I_________---'!I______~I ____-JII 11_--,1
'Material Required for ~<llllinat.ion j I II rQ,::.siion sour~E!J IFacility Exam Bank I (9pe'stion MO<!ificationMethod: j Significantly Modified IiUsed During Training Progr~m ID I
!QuestionSource Commentsj Q27905 Question originally asked why all RCPs had to be running with energized rod control. Modified so that
' ' , candidate had to determine rod control is energized, then decide how many Reps had to be running, in addition to I Ic"..... ".mt '.
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Given the following conditions:
- Both units are operating at 100% power.
- Reactor Engineering has determined that a single fuel assembly in the Spent Fuel Pool must be moved to a new storage location.
- A Notification has been submitted.
- The assembly has been in the SFP for 100 months.
- The SM has given permission, and Radiation Protection has been notified of the movement.
Which of the following describes the Operations Department requirements for this evolution lAW S2.0P-IO.ZZ-001 0, SPENT FUEL POOL MANIPULATIONS?
uired to observe the fuel movement if a Qualified Reactor Engineer is present.
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~~O Value: 2.21ISR?~~~~<::!ion:! IPWG I~~O Group:1 QSRO Group:i I 11 responsibilities of SROs.
55.43(7) S2.0P-IO.ZZ-0010, Precautions and Limitiations, 2.2 states a Reactor Engineer OR SRO must be assigned for Spent Fuel Pool manipulations. A is incorrect because even if a SRO was assigned, they are only required to supervise from the area, not specifically on the trolley. B is incorrect because SFP boron concentration is not a pre-requisite to who is required to supervise fuel movment. C is incorrect and D is correct because a RE OR SRO is required.
Material Required for Examination, Ii II lSource: j IFacility Exam Bank I ~estion Modification Method: II Editorially Modified I[Used During Training Program] -,
U iQuestion~g.~rc~9~mrrients: IVision Q80747 editorially modified choices I
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._. ~R¥i~~~Jfiwyd!atl6nl..isi . . . .!i I *. :.*.**.uililrri~ic. tlriil.~~*.*.~1 1 [Question TO~ ISRO 21 i~IO=f=t=h=e=fO=II=O=W::::in=g=,wh~i=C=h=id=e=n=tif=ie=s=W=h=O=Wl='=II=A=P=P=R=O=V=E=th=e=i=ns=t=al=la=ti=o=n=o=fa=T=e=m=p=o=ra=ry=c=on=fi=rg=u=ra=tj=o=n=C=h=a=ng=e=IA=W=C=C=-=AA=.1=1=2=-1=0=O=1=,A=t=ta=c=h=m=e=nt=5='=~11 Temporary Configuration Change Package Installation?
I
~ Plant Operations Review Committee (PORC) Chairman. I II System Manager(s) for affected system(s). I
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~ystemlEvo[ution Title; I_____________________________________ ...J _ _ _.....J KAStatement:i r--------------------------------------------"""""""';"
Knowledge of the process for contr=ol=.lin:::g~te=m~p=o=ra=ry~d=es=ig:!:n=ch=a=n~g=e=s.======================~
Expla;ation of. 55.43(3) CC-AA-112 and associated T&RM delineate who approves Temporary Configuration changes. A is incorrect because the iAnsWers: i
' System Manager (SM) performs a review of the TCCP. The SM is also responsible for reviewing the limited duration requirements for temporary changes installed per Maintenance Rule (a)(4). SM is responsible for assisting the RE with post installation testing.
SM is also responsible for ensuring that TCCP Extended Installation Justification is approved. A is incorrect because the PORC is I" ,1+;";; * * ,";' ,~;h fr . ,f~," . +h~' h Ithe potential' to affect nuclear safety. C is incorrect because the Ops Manager is not required to approve TCCs.
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'...i ReferEmce T i t l e . *
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II'TEMPORARYCONFIGURATIONCHANGES JlCC..AA-112* .. II Section 3 II 1112 ITemporary Configuration Change Implementatio IfCC*AA-112-1001 ~ IAttachment 5 1/ i 1~1==::;
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1 IS -'-' source:II_N_e_w_ _ _ _ _ _.....1!QuestionModific<ition Method:
m l ________ .....J1 'Used DurlngTrainingProgram**
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Given the following conditions:
Unit 2 is operating at 100% power.
- 28 EDG is CIT for scheduled maintenance, with a scheduled return to service in 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />.
- It is discovered that 2A EDG monthly surveillance was not performed within its 31 day required periodicity,
- The required 2A EDG surveillance was last performed 33 days ago.
Which of the following identifies the status of 2A EDG lAW Tech Specs, and why?
INOPERABLE because it has exceeded its 31 day surveillance INOPERABLE since the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay time past the 31 day requirement has been exceeded, OPERABLE because the normal surveillance interval plus 25% extension has not been exceeded.
OPERABLE because the surveillance can be performed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay time which starts upon discovery of the missed surveillance,
,RO Value: !L~SRO v!'B~ [S~ection: II PWG IRO Group:11 1llSROGroup:1 Abili to determine operability andlor availability of safet related equipment.
55.43(2) Tech Spec 4.0.2 states ... "Each Surveillance RequirE,ment shall be performed within the specified surveillance interval with a maximum allowable extension nct to exceed 25 percent of tile specified surveillance interval." Since the 25% of 31 days has not L-~-'-._ _ _ _****...JI been exceeded, the EDG remains OPERABLE, since its surv'9i1lance is not required to be performed until 31 +7.75 A is incorrect because of the 25% time. B is incorrect because thE~ 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay time is not applicable until after the ho'ur delay time is N/A first because the allowable time would be longer, and second because it is 'not applicable yet. 2B EDG Ibeing inoperable does not affect the question since 2A is not Qut of eriodiC! et, and is still considered 0 erable.
jMateri.1I Required fO~£§'$arriinationl I I iUsed During Training Program I Vision 60376 made into a specific operability question.
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.(;l~::tjon Topic
- I_S_R_O_2_3_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _....i Given the following condition:
- Unit 2 was manually tripped at 20:00:00 on January 21st to enter a refueling outage If all other requirements are met, which of the following is the EARLIEST time that movement of fuel in the Rx vessel could occur lAW Tech Specs?
12000 I January 24th. I 10400 on January 25th. I 1 0400 on January 26th. I
"- .. ~ 12000 on January 28th. I l~~ E:J ~ Level i ~ I~ognitivelev~i ; IApplication .JFaciiitrJ [Salem 1 & 2 1 lExamDat~: II 12/3(20121
~1194001G313 12.3.13 tRovallle:l~sRovalueL-~ iSection:ll~RoGroup:jUsRo Group::=3 J2J D
~~!r.n/EvOlution Title I I I ~ __
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Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities. access to locked high-radiation areas, aligning filters, etc.
Itl"c.::i.~"
- I"\II",VVC."'.
of I 55.43(7,2) TSAS 3.9.3.a states that for refueling outages between Oct. 15- May 15th, the reactor shall be subcritical for 80 hours3.333 days <br />0.476 weeks <br />0.11 months <br />.
80 hours3.333 days <br />0.476 weeks <br />0.11 months <br /> from 2000 on January 21 is 0400 on January 25th. Per Tech Specs Bases, the minimum requirement for Rx subcriticalily prior to movement of irradiated fuel assemblies in the Reactor Pressure Vessel ensures sufficient decay time has elapsed to allow the radioactive decay of the short lived fission products. The 80 hour9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> decay time (LAR S08-01) is consistent with the assumptions I'ilsec ie tee fllel baedlicc_ 2i:i:ic:lae! . 2C"h/";<O 'A/hem ""ki,;o tho.,,, ..,.,,,,,th,, 1~~ h",,, ';0,.." * '" ",,,,,iro,;
. "hi,..h ie' I"n ')Qth ')(\(\(\
i distracter. The 0400 on Jan 26th is..J~orrect time of day but one day late.
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1* Facility Reference Number' r "h . II Decay Time 3.9.3 II 113/49-3 11 273 1
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I 'Questi~n Source Comments' I Vision 084937 i
1 1
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An explosion and fire at the RAP tank area has resulted in a possible large spill of radioactive water in the area. An Alert has been declared and all required facilities are activated and staffed. The Fire Department has determined that off-site assistance from the local fire department is needed.
lAW S2.0P-AB.FIRE-0001, Control Room Fire Response, which choice identifies who must AUTHORIZE requesting off-site fire department assistance?
Nuclear Fire Protection Supervisor.
SM' Emergency Duty Officer (EDO).
Radiological Assessment Coordinator (RAC).
- System/Evolution
Title:
_______ ._ ______.________________...........II.G~~
fi(A Statement0. ;
Knowledge of facility protection requirements, includin£! fire brili/ade and eortable fire fi~htin~ equipment usage. !
'Explanation of 55.43(7)55.43(5) CAS ATT. 1, Fire Dept. Support, Caution prior to step 3.0 in AB.FIRE-1 states, "In the event of a radiological IAnswers: emergency, the Nuclear Fire Protection Supervisor should obtain permission from the EDO/SM prior to calling for off-site assistance." A is plausible because security is required to be notified whenever off-site assistance is requested (CAS 2.0) B is plausible because they will be leading the fire brigade and will be the person to request the off-site assistance through the EDO. D
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- RererenceTitle lt Facility Reference Numoer ! .Reference Section *. -.J rPagaNo. i ~~Vision!
IControl Room =ire Response II S2.0P-AB.FIRE-0001 II II 117 I II II II II I III II ----.JI II II I
'L.O. Number 1FIRPROE007 1-----'
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Given the following conditions:
- 22 CVCS Monitor Tank was released to the Delaware River earlier this shift.
- The release was secured 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> ago during the performance of S2.0P-SO.WL-0002, Radioactive Release from 22 CVCS Monitor Tank.
- During a sample review after the release was secured, the Chemistry Department recognized that a Radioactive Liquid Release to thE' Delaware River was performed with an isotopic concentration which exceeded the ECG EAL RU1.3 Unusual Event threshold of 2X the ODCM for >60 minutes.
Which of the following describes how this should be addressed?
- a.. 1Initiate a non-emergency one hour report for this After-the-Fact event.
b_ 'I Notify the NJ DEP within one hour of the report by Chemistry Department that the release rate exceeded the ODCM.
'c. IDeclare an Unusual Event based on exceeding the EAL at the time of the event, then terminate the UE because the EAL threshold is no longer being exceeded.
d.: 1Declare an Unusual Event based on exceeding the EAL at the time of the event, then retract the UE because the EAL threshold was NOT [:
~ exceeded when the declaration was made.
- i
.Answer; ~ [EXam LeveilIr=J ICognitive Level II Memory I ;Facility: , ISalem 1 & 2 I ,ExamDate: II 12/3/201211 KA:11194001G429 I:2.4.29 I:RO value:1[JJ:rS~§Y~I\.I~Li:J iSection; Il~ .RO Group:IUlsRo Group:;U II~ ~
ISystem/Evolution Title! I-_______________________________________ ..J '--_ __
IKAS~tementJ~-----~--------------------------------------------_ _ _ _ _--,
Knowledge of the emergency plan.
iExplanation of 55.43.(5) Salem ECG, Introduction and Usage, Section 8.6, Conditions Discovered After-the-Fact, describes an after-the-fact event
!Answers: as an event that exceeded an EAL threshold and was not recognized at the time of occurrence but is identified greater than one
~----- hour after the conditions has occurred and the condition no longer exists. The stem identifies that the liquid release was terminated 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> ago, which terminates the release exceeding the ODCM. After the Fact events that occur will be assessed and evaluated I tf'l "n<Ollr" -th<lt nf'l 1= A I (" Irr"nth <lnnli,,<o An . k I\I()T r"""liro,; ::>n,; " n""_omorf'l"n",, ()"o_1-l",lr ROI"I"rt ",hf'>lll,;
I be initiated. The NJ DEP notification is required for spills, not discharges normally performed.
Reference Title !~FaCiliiyR~f;rence Number 'Reference Section I[PagaNo. I rRevision~
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!Question Source c?rnmentsll Objective fr0f!! VISION is EPTRAININGE1, Given ~p related issues and topics, analyze the issue, classify events L.----_,~ ___.. ~____ and communicate to the States, lAW approved statIon procedures. Not listed In drop down menu . I liComment ..
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