ML113340040
| ML113340040 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/11/2011 |
| From: | Public Service Enterprise Group |
| To: | Operations Branch I |
| JACKSON D RGN-I/DRS/OB/610-337-5306 | |
| Shared Package | |
| ML110030673 | List: |
| References | |
| TAC U01830 | |
| Download: ML113340040 (104) | |
Text
{{#Wiki_filter:u.s. Nuclear Regulatory Commission 5 ite-S pecifi c Written Examination
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Applicant Information
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Name: Region: I
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Date: 9/26/2011 Facility: Salem 1 & 2
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License Level: RO Reactor Type: W I Start Time: i Finish Time: Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected SIX hours after the examination starts. 1-------::: .. . ~ ... -- -- Applicant Certification
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All work done on this examination is my own. I have neither given nor received aid. Applicant's Signature Results Examination Value Points
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Applicant's Score Points f---------------.----~-------- . .- - -..---------- . .------.-.. .- -.. .-.-.---..... ------- .---------_._._----- Applicant's Grade Percent
SALEM 2011 EXAM - REACTOR OPERATOR WRITTEN EXAM KEY
- 1. C 33. A 65. B
- 2. B 34. A 66. D
- 3. A 35. C 67. C
- 4. A 36. B 68. B
- 5. D 37. A 69. C
- 6. A 38. C 70. C
- 7. A 39. B 71. B
- 8. B 40. D 72. B
- 9. C 41. D 73. D
- 10. D 42. A 74. D
- 11. D 43. C 75. A
- 12. C 44. B
- 13. B 45. B
- 14. C 46. D
- 15. A 47. B
- 16. B 48. D
- 17. C ~A.
- 18. A 50. D
- 19. A 51. A
- 20. D 52. A
- 21. B 53. B
- 22. A 54. C
- 23. A 55. D
- 24. B 56. D
- 25. D 57. A
- 26. C 58. A
- 27. B 59. C
- 28. B 60. B
- 29. C 61. C
- 30. D 62. A
- 31. A 63. C
- 32. D 64. B
<0 , ~U;Stion T?pic, I RO 1 !
During at-power operation, a control bank rod drops fully into the core V\~thout causing a reactor trip, Which of the follo~nli! is the basis for recording the f.jroup step counter readinli! prior to attemptin);! to recover the droeeed rod? I I Allows bank overlal : to be its <vall after tile. ""V v "'. y I lb. I I Document that the rod insertion limit has not been violated during the recovery< Ensure the operator knows where the dropped rod should be positioned at the end of the recovery < I I I I I Verify operability of the dropped rod after it has been moved mor'3 than 10 steps during the recovery. I I iAnswerq c I 'Exam Le'l~i IR I iCognitiveLevef *IMemory I :FadlitYlI Salem 1 & 2 I [ExamDate; <I 9/26/20111
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iKAStatement: ! Knowledge of the reasons for the follo~ng responses as the~ aPelz: to Droeeed Control Rod: ! Recording of group bank position for dropped rod (reference point used to ~thdraw dropped rod to equal height with other rods in the bank) Expianaiionof ! 55.41(6,10) Verifying the operability of the rod is incorrect, but plausible, since the 10 step movement is the SAT criteria for II Answers: < performing the rod movement surveillance, which ~II be perfomled after the rod has been recovered (Step 3.50) The Bank Overlap
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is incorrect because the overlap computer is removed from the circuit in individual bank select, and will remain at its current setting. Rod insertion limits is incorrect but plausible because it is fed from the PIA converter, and the RIL 10 and 10-10 OHAs will activate if a < I (:::"" ,.... "no rA" ic ....,ff",-+o" h, ,t Ie n, th ",,,c,,n f". tho (:!"" ' " <::'0" (", ,to. ""< ,,,i,,,, I_~m_ . . . . :----ReferenceTItle --~---~l [..** *. Facillty ReferencElNumber.*. ,/1 [Reference'Section .*.***.. PagaNo. I fRevisiofl ,J
- I Dropped Rod !IS2.0P-AB.ROD-0002 II II 11 10 --.J II II II II_~
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nn"",,((-:~ :':':i0'- ',,'D l(;lu~stion Topif! IRO 2 I Given the following conditions:
- Unit 2 is operating at 100% power.
- 21 SGFP trips, and a manual Rx trip is initiated based on lowering SG levels.
- The Rx operator confirms the Rx trip during performance of the immediate actions of EOP-TRIP-1, Rx Trip or Safety Injection.
- While performing the RCS Temperature Control section of EOP-TRlp**2, Rx Trip Response, the RO reports Tavg is -551 0 and stable.
Which of the following describes why Tavg is -551 ° F? 1Reactor Trip Breaker A failed to open. I IReactor Trip Breaker B failed to open. I I II I ITurbine Steam line Inlet Pressure PT-505 failed at 808 psig prior to the Rx trip. I Turbine Steam line Inlet Pressure PT-506 failed at 808 psig prior to the Rx trip. I
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- I Application 1 ~iiitY:J 1Salem 1 & 2 I 'E"am~te:'1 9/26/2011 1
~:] I000007A203 I_~~*o~. =- ~2\1alue;H 4.21 [SR§'value: L~~ ~~n:]1 EPE .JRo~~~~jl 1ISRO~roupJI 11 Fff!' 0 ISystem/Evolution Title
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IReactor Trip I 007
~ Statef!1ent: I Ability to determine and interpret the followinQ as they ao~ll': to Reactor Tri~: I Reactor trip breaker position [
Explanation of i 55.41(5,6,7) A manual Rx trip is initiated by the Rx Trip-landles. This sends a signal to BOTH Solid State Protection System Trains
- Answers: : to open their respective reactor trip breakers. The P-4 signal is developed by the opening of each RTB. One of the functions of the I
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.. P-4 signal is to arm the steam dumps (Train A) and place them in the Plant Trip mode of operation (Train B), to control RCS Tavg at 547°. Since the reactor trip was confirmed, at least one RTB opened. B is correct because the B P-4 signal not being present I " , m ,Irl ":"",,ft in tno ~A<oin do"""" rl,,"""nc . ' ; ' " tn<> I "<orl 6",;",,-+
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f,," T". '" _ ". "n"* "" j programmed Tavg of 547 at no load. Additionally, the MS10s would start to open as steam pressure rose to 1015 pSig, so Tavg would be -551 vs 552 if strictly steam dumps were controlling, A is incorrect because the load reject would arm the steam dumps, and Train B P-4 would place them in Plant Trip Mode and control at 547°, C and 0 are incorrect because steam dumps would have to be in MS Pressure Control mode for the setpoint on tl1e control console to be the driving system signal. Both C and 0 are elausible if candidate does not know normal mode of steam dump operation at power, or function of Tav!=) Control.
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1Reactor Trip Response 112-EOP-TRIP-2 1114 f27 I IReactor Protection System Steam Dump Contro 11221059 II II 1,13 I I II II II I I
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iRQ:~k~Strab~'zl ~~$RQ~kV~~j~QefLI Questi~-nTopiCl
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. RO 3 Given the following conditions:
- Unit 1 is operating at 100% power.
.. Spent Fuel Pool fuel moves are being performed lAW S1.0P-IO.ZZ-0010, Spent Fuel Pool Manipulations.
PZR spray demand lowers.
- PZR level is rising very slowly.
The CRS enters S1.0P-AB.PZR-0001, Pressurizer Pressure Malfunction.
- The crew identifies 1PR2, PZR Power Operated Relief Valve, is leaking and shuts 1PR7, PORV Block Valve.
- ARCS leakrate performed identifies the Unidentified Leak Rate as 0.07 gpm, and the Identified Leak Rate as 0.1 gpm.
Which of the foilowing is required? Assume 1PR2 cannot be restored to an OPERABLE condition. Enter TSAS 3.4.5 for 1 PR2, and maintain power to the 1PR7. Immediately suspend movement of irradiated fuel lAW S1.0P-IO.zZ-0010. Within one hour initiate action to place the unit in Hot Shutdown within the next 6 hours. Enter TSAS 3.4.7.2 for Reactor Coolant System identified leakage and be in Hot Standby within the next 6 hours.
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[FaCility; I Salem 1 & 2 t iE~arnDate:) I 9/26/2011J 1000008G239 I !2.2.39ROVa~ ~;sROVai~E[B[Se~~J'RO*G~~[J]rSROGroup~OJ I
~ystem/E:v~lllti(:mTitl~J Pressurizer Vapor Space Accident .0.0..8 ~Sti~~~~r-______________________________________________________________________________~
IKnowledge of less than or equal to one hour Technical Specification action stEitements for systems. IiExplanationof ........ .i 55.41 (10) 8 is incorrect because there is no requirement to stop moving fuel for this condition, nor is there a general caveat that
~swet.s-.:......-----.J says stop moving fuel for any Abnormal Procedure entry. C is incorrect but plausible, since TS 3.0.3 is used when there is no specific TSAS which applies, and fits witih the other one hour requirements. A is correct because power is maintained to the block valve lAW TSAS 3.4.3 action a for excessive seat leakage. D is incorrect because the source of the leakage has been isolated
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ITSAS 3.4.7.2 action b. Refere~~ce-T-itle~*~, '~~"".---: '. . .. .
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_________________________________,__________________________________________________~ Given the following conditions:
- Unit 2 is operating at 100% power. - 23 AFW pump is CIT, - A 1,000 gpm LOCA occurs, - The Rx does not trip from either Rx Trip handle, nor do the Reactor Trip Breakers open, - Operators successfully open the PZR Heater Bus Infeed breakers 2E6D and 2G6D and initiate a Safety Injection, - NO AFW pumps automatically start, Assuming that both MDAFW pumps could be manually started if demanded, which of the following describes the MINIMUM required AFW pump status, and the reason wh that minimum is re uired lAW 2-EOP-LOCA-l, Loss of Reactor Coolant?
ONE MDAFW pump must be started to ensure sufficient heat ITWO MDAFW pumps must be started to ensure sufficient heat removal. I ONE MDAFW pump must be started to prevent primary to secondary leakage, TWO MDAFW pumps must be started to prevent primary to <:"('nnrl"'r~1
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system/Evoiutionlitle I !Small Break LOCA I VV'=1 I
,KA Statement:! Knowledge of the reasons for the following responses as they apply to Small Break LOCA:
Manual ESFAS initiation requirements iExplariationof iI55.41(4, 10) A is correct because one MDAFW pump will provide at least 22E4 Ibm/hr flow which is the minimum required to satisfy
,Answers:' ..... i the Heat Sink CFST until at least one SG NR level is restored, Additionally, for a small to intermediate sized LOCA, this flow is , J required to ensure a secondary heat sink is available, The distracters all contain either 2 pumps or the incorrect reason, or both. C and D reason is the basis for a LBLOCA, not a SBLOCA. The distinction must be made that a SBLOCA is occun-ing, and apply that tA 'r,,,o <:>1 tho """0"+ Iv,,,;,, fA, tho """o,,! fin' . rO"Hi,o,; T\II/I"\ 1\.1=\1\, "'H""'' '" ""I., ,,,ihlo hor.,,,,,o ;,. '" 1\.1=\, ,~""" ..,,,,
required ~during FRSM when >44E4 Ibm/hr AFW flow is required, and the Rx did ~ot inhially'trip from the trip handles or open'ing the I RTBs', I~ ;FaCilityReferenceNum~ ~nCeSecfu)n'C] fP8g'~ .[j;.wte~i~~~ll r-~~~~---c"---c"~.--:c'~~~.----,--~ I ,< ReferenceTitle ,'.... I~L=OS=S=O=f=Re=a=ct=o=rC=o=o=la=nt=B=a=Si=S=D=OC=u=m=e=nt====~111=2-=E=O=P=-L=O=C=A=-1====:=====~H============~L~1=2==~1128 I I=~I 1=1==========::;11=1=======:::;IC= I________ ---'Il_ _ _ _ _----rlc= IIII 11== 11 __---'1
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1 LOCA01 E009 1------' ! IMaterial Required for Examination' I I 11 I~tionSou~ IFacility Exam Bank I [questio~Moalfication .nethod:!1 Concept Used Iillsed Duii~gTrainingpro9rcll'nl 0
~~==~~==~~:~~======~======~==~I lQuestion-So~tCe90rnme~tsi IVision Q80803 Used the concept of "why" AFW flow is required, and added "how much" and made it an operational I ~.--,.,.,~,-~-'~'_ _.,.,.,._J type question instead of a simple "Why is .. ,.. ,required?"
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I PRO;St&S~r:~i: .".!SR,~;st&rs~efl [~istion '1:o piC! IRO 5 Given the following conditions: Unit 2 has experienced a LBLOCA.
- The crew has performed actions in response to the lBLOCA, and now is preparing for the transfer to Hot Leg Recirculation.
- It has been 3 hours since the SI actuation occurred.
- The control room crew is reviewing the EOPs performed to ensure all actions required have been taken up to this point.
Of the Tr'IIIITr,n. which is the ONLY action which the control room has NOT directed an NEO to the EOP's? Minimized CCW heat load by isolating non safety related Cleared and tagged 21SJ44 and 22SJ44, CONT SUMP SUCT VA.LVES breakers. Removed control power from pumps taking suction from the Containment Sump.
~temlEvOlutionTjtt;;"i I_L_ar....g:....e_B_r_ea_k_L_O_C_A _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-1 _0,1._1. __,.........c KAStatement:1 ....-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-...,.
I Ability to coordinate personnel activities outside the control room. iExplanation of 1 55.41 (10,4) LOCA*1 step 17 sends operator to close CC37 to SFP and CC48 to evaporator. (mimimize CCW heat 10ads)Removing
'Ar1swers: control power from pumps taking suction form containment is wrong, but plausible because LOCA*5 has operators remove control
- - - - - - - power from any pumps taking suction from RWST after it is depleted. It is plausible if operator thinks keeping pumps in recifc mode is so important that automatic pump trips should be defeated. 2MS52 will be resel at step 19 of LOCA-1 after CLR has been
",.,f.""II"hor! <::r- "I;::) 10- ,01" "III "0 "''0 00/_ T"o c:: IA ;"raakers ara opened and CIT at step 18 of I OCA-3
. . Reference Title _.. _---.J ~i.lit¥ Reference N;mJberi 'Reference Section 1[f.~ 'Jllii~
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Qlle~!iOn.!opicJ !R06 I I Which of the following validated conditions will. by itself, require tripping the affected RCP(s) lAW S1.0P-AB.RCP-0001, Reactor Coolant Pump Abnormali!}:, without delay? I 1#1 Seal Leakoff flow >6 gpm. I ISeal Water Outlet temperature 175°F. I IALL charging pumps out of service for 7 minutes. I rctll OHA D-22, 13 RCP BRG CLG WTR FLO LO, received 2 minutes ago. I IAnsWer] a I I ~am LevirJ IR I Icogrlitlve. Level II Memory I iFac-mt>':1 j Salem 1 & 2 I !ExamDatezJ I 9/26/20111 iKA:l1000015K207 I AK2.07 "Rovalue:l~~Rovah.If::iI2.91:seet!on:II~,ROGroup:1] 1ISROG~oUP:]1 11 IE 0 System/Evolution Title IReactor Coolant Pump Malfunctions
'KAl)t§tement~ ~edge of the interrelations between Reactor Coolart Puml2 Malfunctions and the followin~: I seals ,I ~""" .... ,
- Explanation of I 55.43(3)C is incorrect because the loss of charging pumps. and hence seal injection flow. is not a RCP trip criteria, since thermal Answers:
,........._ _ _,.____....JI barrier flow is still established. AB.CVe will not direct stopping the RCPs either. A is correct because #1 Seal Leakoff flow of greater than or equal to 6 gpm is RCP trip criteria based on Att. 1 of ABRCP. When sealleakoff flow rises above 6 gpm. the console digital indication reads "OVER". but the bar graph indication shows it > 6gpm. D is incorrect because the ARP for D-22 I rl,..,,,,,, ",..,t rlimf't entr;' int,.., AR R(,P IInl",,,<:: .' ":' ri"in::; t",m;> Th", ".:.tpo;nt for D_22 is 140 gpm SO the -,PP rp,;"ir",,,, ri",in::;
temp in addition to the alarm, since even with the alarm in there could still be a substantial amount of cooling flow. The normal flow to the RCP bearings is set at -150 gpm, and the OHA actuates at 140 gpm. 8 is incorrect because the setpoint for high seal outlet temp is 190°F.
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I Reactor Coolant Pump Abnormality lis 1.0P-AB.RCP-0001 II II 11 15 I I Loss of Charging II S1.0P-AB.CVC-0001 II II 117 I I II II II II-~ I Material Required for Examination II Question SO!Jrce~j I:F=a:ci=lit:y:;-E-;x:a:m:B:a:n:k==I~~Q::':::ue.=s=t=io=n::M=*. *:::()d=i::fi=c.=at:::lo.=n=*.*:~:=e::th=****O:::d:::::::::J1:E:d:it:o:ri:al:IY:M:O:d:ifi:e:d==:I.:LU::s::e::d::. . D::*u::J:::.rn=g=_T::~~='a=in=i=l1.:g=p=ro::g=r=a::m::*.~=-!~I . ;QuestionSource Comments Added "without delay" to stem to ensure choice c is incorrect.
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~~'~~m'-'l Question Topic RO 7 I I During a total loss of charging event. what is the reason for closing the RCP seal injection isolation valves if RCP seal inlet temperature rises to 225 d",\,J! ",,,,,;;0 !
IThe RCP seals and shafts may be damaged due to thermal shock when a charging pump is started. I ITo prevent steam formation in the Thermal Barrier from impacting CCW system when a charging pump is started. I I To prevent steam binding of the charging pumps due to RCP sealleakoff flashing to steam when a charging pump is started. I (d.' IThe RCP Thermal Barrier heat exchangers may rupture due to thermal shock and water hammer when a charging pump is started. I
,AIlswer II a I [Exam Level' !R I iCognitive L~vel . i IMemory I Facifity: 11 Salem 1 & 2 I !ExamDate: '11 9/26/20111 ~I KA:I _________
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*SyStemlEvolution TItle .! ILoss of Reactor Coolant Makeup 11 022 :
!KA Statement:, Knowledge of the operational im plications of the followiniil conce~ts as the:t apel:i to Loss of Reactor Coolant Makeup: I Consequences of thermal shock to RCP seals I IExpla~ation of 1 55.41(10.7) AS.CVC provides guidance in Att 3 and CAS 5.0 to isolate RCP seal injection if any seal inlet temp is greater than or rAnswers: '.J equal to 225°F. This guidance is consistent with the guioance contained in LOP A-i. basis document, page 32. which states. I "Isolating the RCP seal injection lines prepares the plant for recovery while protecting the RCPs from seal and shaft damage that may occur when a charging pump is started as part of the recovery. With the RCP seal lines isolated. a centrifugal charging pump
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"",n "0 "brtori in t"", n('\rm<>i """'rninn mf'lMO IAlH"f'I! ,I rrm""rn fAr ",,1M ,,0,,1 ini",,!inn flf'I\AI but plausible since starting the movement of hot water in the seal area past th'3 thermal barrier might be dtermined to cause steam formation in the Thermal Barrier. C is incorrect but plausible since the high temperature in the seal area on an extended loss of seal cooling may cause flashing in the seal area, and the gas could be transported back to CVCS system. but is not why the seals are isolated. D is incorrect but plausible because the CCW thermal barrier return is isolated for this reason during a LOPA.
Reference Title i rFacillty RefE!renc~ Number.
- iReference I" ' . . ' Section
, j i. Pageml\l~ :~'- "';;,," ,
ILoss of Charging II S2.0P-AS.CVC-0001 II II 11 9 I ILoss of Power Accident 112-EOP-LOPA-1 II II 11 26 I II II II II II I
,LO. Number IABCVC1 E003 ILOPAOOE003 1------'
QuestioriToPicll RO 8 ! Given the following conditions:
- Unit 2 is in MODE 4 exiting a forced outage.
- The reactor has been shutdown for 96 hours.
- 23 Rep is in service.
- 21 RHR loop is in service supplying shutdown cooling.
- 22 RHR loop is aligned for ECCS,
- RHR HX inlet temperature is 290°F and stable.
- ReS pressure is 325 psig and stable.
1- ALL MS-10's are set for 200 psig in AUTO. The NCO trips 21 RHR pump due to indications of cavitation. With NO operator action, determine how long it will take for MODE 3 will be reached? Assume only heat added to RCS is decay heat. r
-a~ 1< 14 minutes.
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\15-19 linutes.
I Ie] 120-25 minutes. I
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~11~ I'D' ~:II IR I!Cognitive Ley~LJ 1Application I I Salem 1 & 2 I teX~rnDate.:11 9/26/20111 1~I000025K101 I;AK1.01m:=J*ROVaJue:I~SRovalue::@iSeC:tiorl:II~IROGrouP:lUlsRaGroup:JU
~Yrl7EVOluti;;n:nii~' ILoss of Residual Heat Removal System iKA Sf1:'ltement:i Knowiedge of the operational implications of the foliowin;J concepts as they apply to Loss of Residual Heat Removal System:
Loss of RHRS during all modes of operation
'.ExPla,n,atian ,?f I 55,41(10)The HUR will be determined using Attachment 5 of AB.RHR-1, page 2, with PZR level (RCPs are running). The 96 hour
',Answers: ' mark is 4 days, and the HUR will be -3.8- 3.9°/minute. 60 degrees of temp change are needed to get from 290 to 350 (Mode 3) 60/3.8=15.8 minutes. 4°/minute would yield 15 minutes, but 4°1 minute is clearly above where the lines intersect on the graph. If the I after core reload line is used this was a forced outage, not a refueling outage per stem}, then the HUR would be 2.7/2.8 deg/min, which 11l1Q' rid take 21 4_22 2 min **Ies If page 3 HI IR is I.sed then the HIIR we I rid h" 'i n "hif'h "'"\. ,Id (Ii"" 2n min' ,toe If th" I candidate uses the first page of attachment without checking to see if it is for the right conditions as stated in the stem, then they I would use -0.310 which would give over 3 hours. Modifie,d distracters to make all plausible. Provided Attachment 4 in addition to correct Attachment 5 to make question harder. If Reference T i t l e j rF;;tiUty Re~fe~,renceNl1mber >1 [ReferenceSection*'ll*PageNo*.1 ~~ ! ILoss of RHR II S2.0P-AB.RHR-0001 II II II~-.J 11 II II II IC-I
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'Qllestionlr opic] IR09 I Given the following conditions:
- Unit 2 is in MODE 3, NOP, NOT, returning from a refueling outage.
- 21 charging pump is in service.
I- ALL CCW pumps trip due to a faulty electrical relay which was replacE,d on all CCW pumps during the outage. Which of the following identifies an action that must be performed lAW S2.0P-AB.CC-0001, Component Cooling Abnormality, and the correct reason it must be performed? IStop all RCPs to prevent seal package damage. I fb7llnitiate Surge Tank makeup as level drops due to system pressure decay to maintain tank level. 1 IIsolate letdown and swap charging pump suction to the RWST to ensure RCP seal injection is maintained, I IPlace 23 charging pump in service since it is a significantly smaller heat load on the CCW system than 21 charging pump. I IiAnsweri I c I :Examlevel, IR I ,vugm"y", Level II Memory I iFacmtill Salem 1 & 2 I iExarnDate:j I 9/26/20111 II KA: 000026G132 ji2.1.~~!R~value:13.81,sRov~l~e:iL~!Secti()n;iJ~ROGr~up:il 11!S~OGroupJ! 11 _I 0
~Syst~mlEvoluth:m Titlej ILoss of Component Cooling Water KAStatement-
~ ... ..i Ability to explain and apply all system limits and precautions. I
'Expianation of: 55.41{4,7,10) A is incorrect because while RCPs will be stopped, it is beCaUSE! there is no bearing cooling flow. The seal package is iAnswers: " still receiving seal injection flow from charging system. C is correct because in a total loss of CCW scenario where the loss of cooling to the letdown and seal water return heat exchangers (in addition to the loss of cooling to the Thermal Barrier Heat Exchanger) leads to a rising VCT temperature, which ultmately results in the loss of RCP seal injection, and failure of the Reactor r".... IOInt*PIlmn <::"",1" n j,,' h",,.,,,,,,,,,,,, rh"'rninn nllmn h'",,,,t I",,,, .... j" .,,.,I,,,,,lh. ",o.,tor lAth ')-:1 ,.,h.,,,.,i,,,., ""mn Tho ,..o"trif"",,1 1pumps have SW cooling, and the PDP has CCW coaTing.. Bis incorrect because surge tank MIU a is in~iated if leak is present, not for a loss of all the eumps. The stem states the eumes tri~eed from an electrical fault, not because a ststem leak made them trip.
kII Component Cooling Abnormality Reference Title ,. jl Facility Referel'lCeNtnnbef>ii ;Refert?nceSeclion *. *1[Page No:1 ~;:A*~ II S2.0P-AB.CC-0001 II II 1114 I I II II II !I 1 I II IC= 11 II I
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- Q~estion Topic! RO 10 I Given the following conditions:
. Unit 1 is operating at 100% power.
- 13 Main Turbine Governor Valve fails shut.
- The PZR Master Pressure Controller output fails as is at 2235 psig before any response
! to the Governor Valve closure occurs. Which of the fo/lowiniil will occur naturall}': in the Pressurizer to help limit Itle ma9nitude of the resultin£! pressure transient on the ~rima!1: s},:stem? IAn outsurge cools the PZR. This allows some steam to condense to "vater and limits the resulting pressure increase in the RCS. I IAn insurge of water heats the PZR. More liquid then flashes to steam helping to limit the resulting pressure drop in the ReS. IAn outsurge causes the steam space to expand in the PZR. This allows some liquid to flash to steam and limits the resulting pressure drop in 1 the RCS . I
*1An insurge of water compresses the steam space in the PZR. Steam is condensed to water helping to limit the overall pressure increase in the RCS.
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- KA Statemel'lt:11 Ability to operate and 1 or monitor the following as they a:Jpl},: to Pressurizer Pressure Control Malfunction: !
IPressure control when on a steam bubble I 1ExPI~m:ttionOfi 55.41 ( ) A is incorrect because the outsurge cause 10wE~r pressure in the PZR, which causes more flashing to restore pressure. B
~~~ers:~ is incorrect because an outs urge occurs, and if an insurge occurs it would be of colder water, not hotter. C is incorrect because an insurge would occur, but the action is correct. D is correct because the governor valve closing would cause an insurge based on the rising RCS pressure from the load reject. The insurge compresses the steam bubble, which cause condensing of the steam in the I 070 . -h . '. 'h ri....
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Se:cti()n.J . Page NO*1 IRevision IPZR Pressure and Level Control Lesson Plan II NOS05PZRP&L-09 II II I I~- . 'j
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Question Topic: RO 11
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I Given the following conditions:
- Unit 1 is operating at 100% power.
- 12 charging pump is in service,
- A manual Rx trip is initiated after 11 BF19 fails shut.
- The Rx does not trip, and can NOT be tripped from the control room.
Which of the following would be an expected control console indication for the Charging Pumps after the Rapid Boration Initiation steps have been completed in 1-EOP-FRSM-1, Response to Nuclear Power Generation? Assume Safety Injection is not actuated. 11 Charsing Pump 12 Charging Pump 13 Charging Pump I I STOP, START, STOP. I [b~c STO~, START, START, I If:J jSTART, START, START, I ISTART, START, STOP. I cAn~~ [:J ~ Leve!.c
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I Explanation ofl 55.41(10) 0 is correct because FRSM-1 has operator start 11 and 12 chargin£1 pumps. 13 charging pump is not operating from
'Answers: ~ stem and would only be started if neither centrifugal charging pump could is operating. A is plausible if the candidate thinks the
'-"'--'" procedure starts ONLY one centrifugal charging pump, and stops the PDP. B is plausible if the candidate sees 13 PDP is in service I and thinks the procedure just needs at least one centrifugal charging pump in operation. C is plausible if the candidate thinks the I ncr"""ri,,,,, riirl>rtc: ",h,rtin""f ",ith.." 11 Af\ln 1? t"'h"'minn J~llmn'" ",";ri .",nui"",,,, d,...,,,,,inn 1"< ,..h"cninn nllmn" IPDP charging pump is plausible because it is stopp-ediri other EOPs a'nd ABs.
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.* ~.'~.- .-.-."7"~1 ;L.O. Number * "!
I FRSMOOE004
- ~aterial Requ[red for Exall'!illation . iI 11 I
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Given the following conditions:
- Salem Unit 2 is performing a Rx startup by control rods. - The reactor is critical at 3000 cps in the source range. - Control rods are withdrawn to raise power.
No additional operator action is taken and a reactor trip eventually occurs. If a Source Range Nuclear Instrument Discriminator voltage was lost immediately after the rod pull was complete, would a reactor trip occur sooner or later than if the loss of voltage did not occur, and why? later. The recovery period of the instrument following the rod pull """'f'\n",,, sooner. The recovery period of the instrument following the rod pull becomes shorter. er. The gamma level is added to the neutron level. later. The gamma level is deducted from the neutron level. ~ ~ ~.~~ ~ ~itive Level II Applicatiol~ ~I~~ r. Salem 1 & 2 I fEx.amOateTI 9f26/2011 I ~I 000032K1 01 I~K1.01 *)roValue:1 @SRO V'a'u;~TID~~§J~~u~~~D l~ysteM[~volutiO~ ILoss of Source Range Nuclear Instrumentation iKA Stateme nt:: r--::--.....;.."'.....;..--'..;;,..,;;---'-'...:..;...;.;....;....;.;;;=.;:...;...;;--'--'---'--'---'.;...;...........><....:...:;..;.....::...:..L:'O:;'..:...:.....:......'-;;.u;;.;.L..=-::.;;.:;,.:....;:..;.....::..:...:;;;..;;,,;;...;...;;:;;;.;.O';;;;..;.;;.=.;;,;....;.;.;..;;..;;..;;;;;...:...:....:...:..=~ [Explanation of 55.41{1,6) SR discriminator "screens out" gamma radiation at low power levels because it is not proportional to Rx power. If this 'AnsWers: voltage was lost, the SR NI would see all those extra gammas, and would indicate a higher power level. This would bring the SR NI to the trip level sooner than if the voltage had not been lost. D is incorrect because the gamma radiation is added, not deducted. A is incorrect because the trip would come sooner. The Pulse Amplifier has 3 functions: to amplify the signal output from the
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- The BF3 Proportional Counter used in the SR produces pulses proportional to the energy of the ionization reaction in the chamber.
A neutron will react with the B10 in the BF3 gas, to prodJce a large pulse. A gamma will react with the gas to produce a smaller pulse. The Discriminator portion of the Pulse Amplifier acts as a gate, to pass the large magnitude neutron pulses, and to reject the smaller amplitude gamma pulses. Because of its design, it will eliminate any pulses smaller than a neutron pulse. and so removes LO:Nunlb~~ 1EXCOREE005 1--------'
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~estion Topi~J IRO 13 I Given the following conditions: - Unit 1 is performing a unit startup. - Reactor power is 11%. - The low power trips have NOT been blocked. - Intermediate Range (IR) Channel I N35 previously failed and was removed from service lAW S1.0P-SO.RPS-0001, Nuclear Instrumentation Channel Trip I Restoration.
Predict what will occur if the Hifilh VOltafile Power SUEpJi: for IR Channell! N36 fails. IThe high IR flux. bJ IThe reactor will NOT trip, 1N36 indication will drop to zero, I The reactor will NOT trip, but OHA F-25 SR FLUX HI will annunciate. i I Both Source Range channels will automatically energize. and the reactor will trip on high SR flux. I I[AnsWerll b 1 :Eiam'~evei] 1R I [CogniiiveLevel '11 Comprehension 1 1F(icllity:i I~salem 1&2 I !EXamDatill r 9/26/20111 IOOO033A201 I AA2.01 RO Vah.le:11 3.01 :SRf:)\lalue~1 3.51 :Section: 1[!~_~.§_JR6GrouP~1I 21 rS-~OGrouP:11 21 II. D
!System/EvolutioriTItiel ILoss of IntermedIate Range Nuclear Instrumentation ~,~,~,-~,-,--,-,
iKA ~tatement:: Ability to determine and interpret the following as they apply to Loss of Intermediate Range Nuclear Instrumentation: Equivalency between source-range, intermediate-range, and power-ran!i!e channel readinEs I LE:xpllanation of 55.41(7) This is a valid KJA match because the candidatEl is required to understand the relationship between where power is in the
~swe(s: _ _, power range, the level at which automatic re-energization of the SR channels would occur from the IRNI's failing low and/or having their bistables tripped because of the channel failure, and what the significance of the 11 % power in the power range has on NI automatic operation. This question requires knowledge of what levels in the 3 nuclear instruments would be present at 11 % power I
I I r"'n~" nC\\AI<:>r, ' ", thOlt th", SR will h'" ",hml" th,,', will
. '01 if !blil~' llllilCIil to clil.eolilcgize IAibicb is ooe of tbe dislcOIclf>r" 'AJh",t th" IR if they fail low and how that affects the SR instrumentation, and the P-10 permissive, which automatically blocks SR energization with Rx power >10% in the power range. Loss of Instrument power (high voltage DC 300-1500) to N36 will cause its indication to go low, The rx will NOT trip, because the au':omatic energization of P-10 permissive at 214 PRNI at 10% will already I have occurred. (Stem 11 % power), This will prevent the SRNls from automatically reenergizing when the second IRNI channel lowers less than 7x10-11Amps. F-25 will not annunciate, and if it did it would trip the Rx.
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.l..O.NUmber IEXCOREE007 EXCOREE009 IEXCOREE010
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Which of the following identifies why the ReS is depressurized during the response to a SGTR lAW 2-EOP-SGTR-1, Steam Generator Tube Rupture? Maximize boron injection rate and minimize subcooling. Terminate primary-to-secondary leakage and refill the PZR. Terminate primary-to-secondary leakage and reduce secondary plant contamination. ~~clj ~ ~~~ ~ ~!tive Level II Memory I [FaCiiity:'11 Salem 1 & 2 I [E~am~ I 9/26/20111 iRO !ajueiJ! 3.211~RO Value~1 3.51 ~ctionDI EPE I!RO~~OUP:II 1: iSRO Gr~~~1 11 _Sys~~m/EvolutforlTitie-, ISteam Generator Tube Rupture KA StatemeriliJ Knowtedqe of the operational implications of the foHowinq concepts as they apply to Steam Generator Tube Rupture: I Leak rate VS. pressure drop I Explanation of l 55.41(5,10) ReS depressurization is done to eliminate the pressure differencEl between the ReS and the ruptured SG, which stops Answers: ! the primary to secondary leakage. ReS pressure is actually reduced to LESS than ruptured SG pressure to allow backfill of the , _..... ~ m"'-_.,~_
... ..., Res (PZR)from that SG. While secondary plant contanlination will be reduced by stopping the leakage into the SG and its transport to the secondary plant, it is not WHY the depressurization is performed. The boron injection distracters are plausible I",,,,,.,,,,,,, .,,, .,rc:: n,,,,,,,,,,.,,, j"",,,,,,, r;::rrc:: ;,.,i",,..ti,,,., fl"" ,., "., ",n;t" r;"o" "nn ,..,'" n,r",;n", "'0 "'",,,,., ;n;o"';,,n ,,, >n,,, oro C::I'"\~A Iremains adequate. " . "
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matic action and its correct initiating coincidence which will occur if a Main Steam piping rupture occurs at rm<""t,r\r action? Main Steam line Isolation. 1/2 High Steam Flow on 2/4 SGs, with 2/4 RCS loops <543 Q F. Main Steam line Isolation. 2/4 High Steam Flow on 2/4 SGs with 2/4 Steam line pressure <600 psig. 2/3 detectors on 1/4 SG 100 psig lower than corresponding detectors on 2/3 remaining SGs. 2/3 detectors on 1/4 SG 100 psig lower than corresponding detectors on 1/3 remaining SGs. rKA:1 I000040K201 Ir~J~~§~9~[~~~~~~~U~~~][JJ SysterilJEvolutionl1t1e'
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-KA Sta~ment:J KnowledQe of the interrelations between Steam Line Rupture and the followin9: I Valves I I ExplanationOf 55.41(7) Main Steam line Isolation occurs as described in A. S is incorrect because the High steam Flow coincid,snce is 1/2, not
'1 iAnswers: I~'" _...:.:.._~.c 2/4. C is incorrect because there will be no Safety Injection on SG DIP because the rupture is downstream of the MSIVs and ALL SGs will have lowering pressure. C coincidence is correct. D is incorrect for the same reason as C, and additionally has the wrong coincidence for the remaining SGs.
,. Reference Title ". ,j F<tcniiy R.eferericE!Null1berlReferenceSe~ctioniJ iflage No.1 ~eVi~
j RPS Steam Generator Trip Signals 11221056 II II 11 8 J ILicensed Operator Fluency List II NOS05FLUNCY II II 117 I I II II II IC I L.O. ,Number IFLUNCYE002
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fiR~~~Scrapel'c~t.-u. :***.RO SV$temfEvoj.~'~IIml SRO svs,JeInlEvolution List 1ioUt'.~ballglll . . . ......................... It:lUestlon ~pic:~ I.;,..R;.,:.O_1;.,:.6_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _...I1 Given the following conditions:
- Unit 2 is operating at 100% power.
I- Main Condenser backpressure begins to rise at 2 "Hg every 5 minutes. 1- Operators start available vacuum pumps and initate a Turbine load reduction at 3% per minute in an attempt to stabilze vacuum lAW S2.0P-AB.COND-0001, Loss of Condenser Vacuum.
- Condenser backpressure continues to rise.
Of the following, which will occur FIRST if NO OTHER operator action is taken, and why? Assume the Main Turbine does NOT trip at any time. I i",:~ BOTH SGFPs will trip to prevent damage to the SGFP blading. fbJ IInterlock C-g will block Steam Dump operation to prevent damaging the condenser on overpressure. I SGFP Master Speed Controller output will lower to 50 psid to ensure Permissive P-14 does not actuate. I fci111nterlock C-? will arm the Steam Dumps to ensure auto rod insertion below the Rod Insertion Limit does not occur during the down power. I' I I l~slrVe;.:::*i-;:l=b=::;:I-;:::======::-;::=::::::::;~~~=m"'_::-~=m~=~='_=.=_.== ..~=~rl=A=pp=l=ica=t=io=n==:;:-I-:~=a=ci=~=y=:j~l=s=al=em=1=&=2==:;:i-:.E=x=*a=m=.~=3t=eJ=:::-;:::I-=-=-=-=-=-9~/=2-6~/2~0~1:::;1-'1 rKA:II_00_0_0_51_K_30_1_----'L..._~_...JL---JII.2-.8-*.1I . S_R~O~**~Va~'u.e.~i.I, __ ..*-~-1*~I[.s~e~c~ti~onm_E-~ill-E-PE-..J.~R~o-~.-r.o_u-=..p_..J,,-2...1L.~.~R~O~G~ro~u.:.pI.,I----121 'm 0 ,.-.--"~~",- iSystemlEvolutioflTitie i ILoss of Condenser Vacuum I 1051 lKA ~t~em~rlt: I Knowiedge of the reasons for the following responses as they apply to Loss of Condenser Vacuum: Loss of steam dump capability upon loss of condenser vacuum ~~:~j 55.41(4) Control Grade Interlock C-g energization allows Steam Dump operation by unblocking steam dumps at 20" vacuum rising. IAn It also blocks the steam dumps at 20" vacuum lowering. which equates to 10" Hg backpressure. The condensers are not designed . .. to operate at any pressure. Control Grade interlock C-? arms the steam dumps on a 5% I minute ramp. The action is correct but the setpoint would not be reached with a 3% per minute ramp. The assumption stated in the stem is required since the Main I T, rl-.~ A >ri, ,,-I .1-.. 0 dlllJJ!=lS 11101 lid block. aod lbe caodida!e cOlild lbiok Ccctrnl r.!. ("'_7 would actuate on the large load rejection caused by the Turbine trip if load were less than 49% and the reactor did not trip to place steam dumps in plant trip controller. A is incorrect because the SGFP trip is at 0" vacuum. C is incorrect because SGFP speed will lower, but not to 50 psid, which is the bottom end of its control band, and would not be expected to be at that level until at very loweower. [:---
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Reference Title - i Facil!t)'~!ference Nu~~r[ rRefen~nc~~~!!§n ..-] [*Pag~,.,o.[ }le~SiOr!] ILoss of Condenser Vacuum II S2.0P-AB.COND-0001
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m'" L~~esiion Topic I jRO 17 I Given the following conditions: Unit 2 is operating at 35% power with both SGFPs in service.
- BOTH SGFPs trip simultaneously, Which of the following describes the actions reguired in S2.0P-ARCN-0001, Main Feedwater/Condensate Si:stem Abnormalitl;':, and whi:? I I Trip the Rx and manually start 21 and 22 AFW pumps to maximizE; SG .... "',.. y J
I Trip the Main Turbine and manually start 21 and 22 AFW pumps to maximize SG inventory. I I Trip the Rx to ensure only decay heat and RCP heat are being adcled to the RCS. AFW pumps will automatically start. I
~II Trip the Main Turbine to ensure only decay heat and RCP heat are being added to the ReS. AFW pumps will automatically start.
Answer! I c I [EXam Level! I R I Eognltive Level II Memory I IIFacHity:l [salem 1 & 2 I ,Exami)ate:ll 9/26/2011J
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)<A State~_erit: i KnowiedQe of the reasons for the following responses as they apply to Loss of Main Feedwater: I Reactor and/or turbine trip, manual and automatic l
,Explanation ofl 55.41(10) In many cases, when a condition arises which calls for removal of steam flow, the Main Turbine would be tripped if Rx
'!'ns~e~~~ power is <49% (P-9). The Loss of Main Feed is NOT one of those cases, since SG inventory would quickly be depleted without the removal of steam demand from the Main Turbine. For a loss of feed, the rx is tripped to ensure the only heat beirg added to the RCS is decay and RCP pump heat, which minimizes the amount of heat removal required from the SG's, and allows AFW system
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!not possible nor required, since the 'procedu're knows thei: will already be r~nning. . .
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iQuestlonTopicl RO 18 I Given the following conditions:
- Unit 1 is operating at 100% power.
- All 3 Chillers have power, and are operating as expected for temperature conditions.
- 12 chilled water pump is in service.
- A loss of all off-site power occurs.
Which of the following identifies how the loss of off-site power will affect chiller and chilled water pump operation? Chillers will have their 460V breaker closed b~ their reseective SEC, and Chilled Water Pump(s) will be operating. jALL' 11 and 12. I 1ALL. ONLY 12. INO, 11 and 12. I
@JINO, ONLY 12.
II a ! iExamLeveO IR I :Cognitive Le.vel i IApplication~ IFac mtyS I ISalem 1 & 2 Ii [Answer I [examoirte: II 9/26/2011 iKA: II 000056A 108 IIAA1.08 I:RO VaJue: n2.5*1 rsR(}Val~e:q ~ ,Section:!! EPE J fRO Grou~ I 1'~OGrotip: I 11 &Ill [] I
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r~-~~ System/Evolution Title I Loss of Off-Site Power
~
KASt<i!!lmentj Ability to operate and / or monitor the following as they appl;t to Loss of Off-Site Power: I HVAC chill water pump and unit I !iEXPlanat~on Ofl 55.41(7) The SECs will shed then reclose the chiller 460 volt breakers on a MODE II (Blackout) SEC initiation following the EDG iAnsW.!~._"] start. The SEC will close the Chiller 460V breakers and start both Chilled Water Pumps sequentially to satisfy the Chiller start permissive. This is unlike the ECAC interlock start. where the standby pump (11) starts only if the dedicated pump (12) fails to start. I
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R.eference Title " l; f'acilltyRefere~ce _. Numi:lei". ="Reterencesectiol1~ !page.~ :~eviSion' I Chilled Water System Lesson Plan II NOS05CHLWAT i 123 ,34 11 10 I I No 1 & 2 Units Control Area AC Chilled Water P 11228031 ISafeguards Emergency Loading Sequence 11203668 "IIII II II 11 13 11_6_~ 1 r:-:-~._.m;._-.--~. -, ILO. Numb~r I CHLWATE008 IMateriaLRequiredforEXarrifnation'! I II IQuestio,n Source:} I:N=e=w:::::~====~I~iQ=u=e:::s=ti::O=ri=*.*~:::.o:::* *iJ::i!i::IC:.:'a=ti=o=n::.M=e=th=o:::P=:::::=~:========I.:!u:.s:::e:d:D:~u: :(:in: 9: :T:ra:: i: :n: :in: 9: p:r: :o: 9r:: a: :m. '::..]-=.
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- Questi~ Topic_;
The operating crew has entered S2.0P-AB.RAD-0001, Abnormal Radiation, due to RMS alarms on the plant vent. Operators have shifted both the Auxiliary Building and Fuel Handling Building Ventilation controls to HEPA PLUS CHARCOAL lAW their respective procedures. Which statement describes the change, if any, in the release rate resulting from shifting the ventilation systems, if the problem is a stuck open relief valve on a Waste Gas Decay Tank (WGDT)? None - the WGDT relief valves discharge to the plant vent. The release rate is reduced by passing through the Auxiliary Building Ventilation System. The release rate is reduced by passing through the Fuel Handling Suilding Ventilation System. None - the WGDT relief valves discharge to the Auxiliary Building Ventilation exhaust fan suction plenum.
~-"m Levell :C09nitiv~Level I iKA::looo060K202 !SYstem/Evolution Title~ I Accidental Gaseous Radwaste Release ~ ,060 i iKA StatelTlent:' Knowledge of the interrelations between Accidental Gasl,ous Radwaste Release and the following: !
Auxiliary building ventilation system I
!exptanationofi 55.41(13) The WGDT relief valves (205340 sheet 2, E-TI combine into a single header and go to (205322 sheet 1, G-3), which is Answers: ! the FHB Ventilation exhaust downstream of the exhaust fans. It then fiows to the plant vent (205337 sheet 1, G-11 into the plant vent release point.
A Release is detected by Plant Vent Monitors but does not pass through ventilation (B), (C), (D) Flowpath is not through filters nor
"""I'; th" filh"" h", ,f'I',' th" n"hl" "''''''
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.* FacUity Reference Number. .1 :Reference Sec~iori
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I Waste Disposal - Gas i1205340-2 II II I~I IAux Bldg, EDG Room, and Fuel Handling Bldg v 11205322-1 II II il~~ IAuxiliary building Ventilation 11205337-1 II II I~-I
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*****-*~*****~~I Question Topic l .RO 20 I Given the following conditions:
- Unit 2 is operating at 65% power.
- 21 Charging pump is in service.
1- #4 SW Bay has been isolated due to a bay leak lAW S2.0P-AB.SW-0003, Service Water Bay Leak.
- All #2 SW Bay pumps trip.
Which of the followin~ describes an action that will be eerformed in S2.0P-AB.SW-0005, Loss of All Service Water, and wh:t:? Iisolate CVCS letdown. This prevents resin damage in the CVCS demineralizers when letdown temperature exceeds 136°. I IIsolate CVCS letdown. This prevents heating of the fluid in the VCT to a temperature which will cause a loss of NPSH to the charging I ISwap charging to 23 charging pump if it is available. This allows lowering RCP seal injection flow below the minimum stop for the 2CV55 and extends the time available to perform a controlled shutdown of the plant. I Itemporary Swap charging to 23 charging pump if it is available. This maintains seal injection flow to the RCPs, which allows sufficient time to install cooling hoses to centrifugal charging pumps prior to RCP seal temps rising >225 0. I [Answer' I d 1~:'<:C3m Level II R I [Cognitive Level i IMemory --.-J ifaclHty;! [salem 1 2 & 1 tExamDate: i I 9/26/20111 [¥.:;[0OO062A206 I:AA2.06 ROVa[ue !riliiSROvalue:!13.1*j lSectlon:HEPE J R~.Group:H 1ISROGrOup:j[J] III 0
~tem/Evolution Title ILoss of Nuclear Service Water KA Statement:, Ability to determine and interpret the following as they apply to Loss of Nuclear Service Water: I The length of time after the loss of SWS flow to a component before that component may be damaged I Explanat'ionofi 55.41(10,6,7) A and B are incorrect because while letdown is isolated, it is to minimize the heat input to the CCW system to prolong iAnswers: I cooling to the Thermal Barriers and RCP seals. The eves demineralizers are automatically bypassed on high temperature at 136°F. C is incorrect because a swap to the PDP is made since it is cooled by cew, and an immediate action in AB.SW-005 is to trip the Rx, there will be no controlled shutdown. D is correct because the basis document states that the transfer to the Positive I n, on' ",.,..,n;" ,,,r!. h"" '"'" ", Ih", one ;" ", ,r! h\l cell\! not S\I\I This means seal iniection flow ShOI ,Ie h ,h h I maintained by'the PDP until temporary cooling hoses to the CCP's can be installed from the DM system .
. Reference Title. ] Reference Section i Page.N()~~
ILoss of All Service Water 05 II 11 ;::==~
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[Question Topic I IRO 21 I Given the following conditions:
- Unit 2 is in MODE 3, NOT and NOP, - A total loss of all Control Air occurs.
Which of the following describes the basis for transferring the chaqiling eump suction to the RWST lAW S2.0P-AB.CA-0001, Loss of Control Air? I The quantity of water required to maintain PZR level during the RCS cooldown will be greater than the capacity of the CVCS makeup system. I I The RCS must be borated to CSD conditions, and RWST water is at a higher concentration than the VCT, so boron will be added to RCS at a faster rate, I Ilf a centrifugal charging pump is running, its recirc line back to the VCT will short circuit any boron addition and raise the time required to achieve CSD boron concentration. I I CVCS letdown will be isolated, and VCT makeup is unavailable. This would cause a rapid depletion of VCT level and subsequent air induction to the CVCS charging header. Answer II b I iExam Level! m==J E~nitive level jMemory I Facility: IISalem 1 & 2 IIExamDate: I 9/26/2011J
<KA:J! 000065K308 IAK3.08 ~IROvail.Je:113.7Ls,~ovah.le:!l3.9irsectlo'nill~iROGrOUp~I 1ISROGrouP~1 11 ~~111 [J I ILoss of Instrument Air
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SystemiEvolution Title I~ -' L~~tlitement: i Knowiedge of the reasons for the following responses as thex appl:! to Loss of Instrument Air: I Actions contained in EOP for loss of instrument air ! jEX'planation of , 55.41 (10) AB.CA basis document states, "Prior to commencing a cooldown, the RCS must be borated to cold shutdown conditions.
- ~s~e~-=-_.--J Without Control Air, there is no way to reduce RCS inventory. Thus, the amount of boron that can be added before the cooldown commences will depend on the available space in the PZR. Due to the slow rise in level, this may become limiting. Therefore, the charging pump suction Is transferred to the RWST early in the event. This ensures that any addition to the RCS is at RWST
,..nn,..ontr"tinn r; i<:: . . ho,.."" """ tho n, ,mn """It"n \Jill ",,,t,, ",,,,,,n nM)V,,\ I" th", k>1II1<::T " In In 1<>,,01 in tho \/(,T C is incorrect because while the recirc line does go back. to the VCT, it is not the basis for the transfer. A is incorrect but plausible. PZR level will actually be rising due to the loss of letdown, and loss of PZR inventory during the cooldown is not a concem.
'Referellc~ Title . . . ~
. FacilityReferE~nce Number .* JIReference Section I!pageNo. IRevision!1:....:._.<.<<**,,"
II S2.0P-AB.CA-0001 ILoss of Control Air Ii- Ii 11~~ I II II il II I I II 1'-= II Ir=~ iL~O. Number ~ IABCA01E002 [-----' Material ReqUired for EXarninatiQrl : I II I
- n.. "",tinnSollrce:! Previous 2 NRC Exams! :que~~ton Modific~tioIlM~thOd:;1 Direct From Source ItU!~d Duling Training Program' 0
- Que'stion SoiJrce Comments I NRC Exam August 2008
~ ........ ****1
*****1200F.
Prior to opening the PZR PORVs in an attempt to lower CET temperatures, how should RCPs be utilized? RCPs will NOT be started at this point since they are required later in the procedure. Start ONLY one RCP In any RCS loop. Continue operation of ONLY one RCP until core exit thermocouples are less than 1200 deg. Start one RCP in an available RCS loop. If core exit thermocouples remain> 1200F, start another RCP in an available loop. Continue until all available RCPs are running or CETs are <1200 deg. Answer II a I (Exam Level II R I 'Cognitive Lev~iJ IApplication ISalem 1 & 2 I iExamOatel and the followin : 55.41(10) With SG WR levels at -5%, there is no secondary heat sink. At step 13/14 of FRCC-1, no SG level and no AFW flow directs operators to step 23. Step 25 caution states that RCP's are only to be started in loops with SG NR levels >9%, of which
~~.......~_ _"""':"...J there are none. No RCPs can be started. It is NOT because that they will be needed later in procedure.
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iQljesti6n Topic; IRO 23 i Given the following conditions:
. Unit 2 is performing 2-EOP-LOCA-2, POST LOCA COOLDOWN AND DEPRESSURIZATION. - 22 Charging Pump has been stopped. - Conditions are met for stopping one Safety Injection (SI) Pump.
Which of the following identifies the SI pump to be stopped and the reason for the selection?
~1121 SI Pump to ensure ECCS injection flow if any single vital bus failure occurs.
122 SI Pump to ensure ECCS injection flow if any Single vital bus failure occurs. 121 51 Pump to ensure one full train of injecting ECCS equipment is maintained in service. [J 1.22 SI Pump to ensure one full train of injecting ECCS equipment is maintained in service. ,AnsVlier ~ !J:xam Level! IR I !....vl:I... Leve.I.* I IMemory Uy , ; I 'Fa<:;ilityrj ,"Salem 1 & 2 I .ExamD§ I 9/26/20111 l!<k!100WE03K101 IEK1.1 I i :Rcivaluedl 3.4I,SROVallj!D[ 4. OJ iSection:H~ [ROGro~1 21 !s~OGroup:11 21 II~ c -Syste.I1'l!Evolution Title ILOCA Cooldown and Depressurization --------, '!<AStatell1_~~ Knowiedge of the operational implications of the followin9 concepts as they apply to LOCA Cooldown and Depressurization: I Components, capacity, and function of emergency syste'l1s. Explanatiollof ! 55.41(8,5) A is correct per LOCA-2 NOTE at Step 21 (E:CCS PUMPS SHOULD BE STOPPED ON ALTERNATE VITAL BUSES).
- Answers: . , 22 charging pump is powered from *C" vital bus. So is 22 SI pp. This means 21 SI pump should be stopped. The reason it is stopped is because .. ,,"Balancing the load between ECCS trains increases the probability that there will still be some RCS injection flow if a loss of one train occurs." Salem specific ERG deviation changes" altemate ECCS trains" to "alternate vital busses" I
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;Question.s()ur~eCommeni$] Salem 8/2008 NRC Exam RO Q21, Modified stem to "given the following conditions" type format from of format.
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-~--I QuestionTopicj< RO 24 l~ .. ~ __ ~~~vv" ******* __ ~~< I Given the following conditions:
- Unit 2 is attem pting to identify and isolate a LOCA outside containmen:.
1- 2-EOP-LOCA-6, LOCA Outside Containment, has just been entered<
- The source of the water is backleakage from the 23 cold leg injection line.
Assuming that any valves required to be operated during LOCA-6 operate correctly, which of the following leak locations would NOT be isolated while usin£! 2-EOP-LOCA-6? [~ Ion the valve inlet flange on 22SJ49, RHR DISCH TO COLD LEGS. I Ion the valve outlet flange on 21SJ49, RHR DISCH TO COLD LEGS, I IBetween the 2RH20, RHR HX BYP VALVE and the 2RH26, HOT c.EG ISOL VALVE. I
~~J IBetween the 2RH2, RHR COMMON SUCT VALVE, and 22RH4, RHR PMP SUCT VALVE.
I [Ans-'ier11 b I (Exam Levell IR I tCogniiiveLeve'll Comprehension I 'Facility:! ISalem 1 & 2 I iEx.amDate:, I 9/26/20111
...~~JlOOWE04A102 IEA1.2 '-Rovalue:il3.61.SROVli!lle:[3,81,section:.ll~~OGrollP:II 11~ROGrouP:lI 11 1111 [J ~SystemlEvolution Title] ILOCA Outside Containment E04 ~--
KA Sta.~rT1~nt: J Ability to operate and I or monitor the following as they apelL to LOCA Outside Containment: I Operating behavior characteristics of the facility. I Ire;pl~nation of 55.41(3,8) 2-EOP-LOCA-6 closes/checks closed the following valves: 2RH1 OR 2RH2, 21 and 22RH19s, 2RH26, 21 and I Answers:.~.:......J 22SJ49s. Using drawing 205332-SIMP, it shows that any leak between the RH1/2 and the SJ49s will be isolated with the above valves closed, The only location which wouldn't be affected by those valve be:ng closed in the downstream/outlet side of the SJ49 valves. The stem statement of proper valve operation was inserted to preclude a candidate from assuming a leaking valve may not "I""" f.oIh,, ,"'on , I I',Reference.Section . I! P<l~eNo*1 *. ~~
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Reference Title ..;~FacilitYRefererlce NUmber "v'W ILOCA Outside Containment 112-EOP-LOCA-6 II II II I IRHR Simplified Drawing 11205332-SIMP II II 112 I 1 II II II II I
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1 :Que~ti()fT~PiC" IRO 25 I Given the following conditions:
- Unit 1 has experienced a LSLOCA.
- 11 RHR pump is CIT electrically by it's 4KV breaker only.
- RWST level is 14.8', and operators are performing 1-EOP-LOCA-3, Transfer to Cold Leg Recirculation.
Which of the followin9 conditions will erevent transfer to CL recirc, and cause the crew to go to 1-EOP-LOCA-5, Loss of Emergency Recirculation? 112SJ44, Cont Sump Suct Valve, does NOT open when its Sump Auto Arm PB is depressed. I 11 RP4 Lockout Switch for 12RH4, RHR Pump Suction, is stuck in "Locked Out" position I 112SJ44 open PS is depressed before the 12RH4 close PB is depressed. ! I d.1 112RH4 can NOT be closed from control room OR locally. Answe~ II d I :Exarnl..eveli I R I ;CognitiveLeve'ill Application 1 :Facility: , Salem 1 & 2 r I 1 :Exam Date: 1 I 9/26/20111
!KA:IOOWE11K201 I,EK2.1 , [RO Value:J! 3.6IsRovalue::I 3.9!Section: 11~ iROciioup:j 0 !SRO.<;I'oI.lP:] I 111~j4!l 0
'KA Statern~..!1!:.l Knowledge of the interrelations between Loss of EmergEincl Coolant Recirculation and the followinfij: I Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
-.~- ..,
Explanation of : 55.43(7) A is incorrect because Unit 1 SJ44 valves do not have auto swap function like unit 2 has. C is incorrect because the close iArswer~_. __..~ signal will remain to the SJ44 even though it is interlocked to not open until the RH4 is shut, and there is no lockout to prevent it opening. S is incorrect because there is no lockout switch for RH4s on RP5. D is correct because if the 12RH4 cannot be closed, the 12SJ44 cannot be opened. and recirc cannot be established, and procedure directs transition to LOCA-5, This tests I~ "f tr.o "nit ;U, ",inf' I Init " ""'" t"o C;:om Ul. , ,t" C;:",,,n,,,,,::.r "f RI-lhl e, ,,,tif'\n t" ~ ' , o n t C;:, ,m" '>nll I Ina 1 Il ...." " Ino!. - ILdC~- ReferenceTith~ J ~* .*. Fa;;iiitY Referent?~Null1berl iRefel'e~~Section / i 'PageNO,? ~S!§l I Transfer to Cold Leg Recirculation 111-EOP-LOCA-3 Ij II 11~2=8===:::. IINo.1 Unit 12SJ44, 12RH4, and 1RH1 11211507-1 Ij II 1135 )I _______----lil 11 _ _ _ _ il II~::=::;
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, iMaterial Required forExamination! I II Q~~on~6u~Ejl:F~a:ci:lity~E~x:a:m:B:a:n:k~~I~~=u~~=i=1~@~p~M=.~O~~~!=~=*=_~=n~M~~~.~=.6=~~j~I~E:d:~:O:ri:al~~:'M:O:d:l:fie:d~~~I~~~~=**:d:D~*~=!:~~~~T:~r:a:in=l=n~~~p=ro~g~r=~:m=~~I~ 0 Questions~~rcefomrrientsll Vision Q122456, removed step number from distracter C since it is a non-existent step I 1 iComment ...*..**.. '. . . . . . . . . . . >1 I I I
IQuestion Topi~: I_R...;,O_2_6~_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _._ _ _ _ _ _--..l Given the following conditions: Unit 2 has experienced an event which has resulted in 24 SG pressure rising to 1115 psig.
- A MSLI has been performed.
[KJ.::'1 OOWE13A102 rROV:~~~v~@secti911:II~J [ROGrooP:fJJ~~~~
'SysfemlEvol;';tion Title I ISteam Generator Overpressure
- KAStatementJ Ability to operate and I or monitor the following as they apply to Steam Gt::Ilt::ldlUI O\tt::I~'t::""ure: I Operating behavior characteristics of the facility. I IExplanationof ! 55.41 (4. 7} Each SG has 5 safeties, with lift setpoints of 1070, 1100. 1110, 1120, and 1125 psig.
IAnswers: ...** *. *
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- Reference Title ii ,,>: Facility Refel'ence Number i IReferenc~ section .01 'lageNo.!,1 IMain. Reheat. and Turbine Bypass Steam 11205303-2 II II 11 61 I I l! II II II I II II II II II [
I FRHSOOE009 I STMGENE008 iMaterial~equ[redfor~amination{11 'I
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Salem Unit 2 experienced a LOCA with failed fuel.
- Containment pressure peaked at 18 psig. and is currently 3 psig and slowly lowering.
- Containment radiation level is 1.1 E5 Rlhr and rising very slowly.
- The TSC has not given the control room any direction on use of Adverse numbers while in the EOPs.
Which of the following describes how containment conditions will affect use of "Adverse" numbers while in the EOP's? Adverse numbers in effect because IARE; containment pressure rose above 4 psig. IARE; radiation levels have reached their threshold. IARE NOT; the TSC has not provided any direction yet. IARE NOT; they reset on lowering containment pressure and the radiation threshold has not been reached. I AnSwer i~ iExam Levefj ~ ~co~nitive Level i IMemory ~ lFacilitY~J rSalem 1 & 2 I :~~a~Date:' I 9/26/20111
'KA;l\ 00WE16A202 l'EA2.2RO VaIUl:l:Jr:TIi'SRovCllue:i[3.3 i
~ectio~ll EPE JRO GrouP~[]iSRo.Group:,[JJ'I. 0 S~tem/E~olution Title. 1High Containment Radiation I I' KA Statement: ! Ability to determine and interpret the following as they apply to High Containment Radiation:
,~"".------.----,
1 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments I i Explanation of i 55.41 (9,12.10) Adverse containment numbers are used when containment pressure reaches 4 psig or containment radiation level
~~swer~J reaches 1E5 Rlhr. The Adverse condition resets when containment pressure lowers less than 4 psig, but does NOT reset on high radiation. The conditions in stem are such that Adverse1umbers have are In effect based on the containment radiation threshold being exceeded.
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'QJeStiOrts~Comments~ Similar to NRC Exam questions from Sept 2001 Cook exam, Nov. 2002 Salem exam, Dec 2002 Beaver Valley exam .
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Given the following conditions: Unit 2 is operating at 100% power.
- A failure in the automatic Rod Control circuit causes Sank D Control rods to step in at 72 steps per minute. - Control rods insert 20 steps before the RO places rod control in manual and rod motion stops.
Which of the following describes the initial effect of the rod insertion with NO other operator action? RCS pressure will lower and PZR spray valves will shut. Main Generator MW load will lower and governor valves will not throttle open. RCS temperature will lower to the RC Loops Tavg-Tref Deviation alarm setpoint of 1SF. ~~~eiJ ~ iEXamL-eVel' ~ -~~9~ffi;e-LeveiJ IApplication -FacmiY: ISalem 1 & 2 I rexamDai&:' I 9/26/20111 ~~:]I 001000K302 I~~~~_~J [it<?.,~rIfI~~a~@~~l~ ~~~,~OJ ~SROGrOiiPJ OJ ~iStem/Evoiution Title. IControl Rod Drive System 1---- , ' IKA Statemellt~ Knowiedge of the effect that a loss or malfunction of the Control Rod Drive System will have on the following: RCS 55.41 (5) A is incorrect because the RIL for 100% power is -170 steps on Sank D, so 20 steps of insertion from 230 steps (ARO) will not cause this alarm. C is incorrect because Main GenE,rator load will lower due to lower steam pressure, but it will not remain lower since the Governor valves will automatically open to restore Turbine Steamline pressure to the value corresponding to 100% power. S is correct because the rod insertion will cause a lowering of RCS pressure sufficient to full close the Spray valves, which are n"r"" <:I II., - .,0/. "oon ,;, '0 I" rn<:linbininf"l "no tV", ,0 "f D7R AI! 1 bealers in man, 'al ON 0 is incorroct beca' 'S" IAthil" ,,,mn,,,,,h ,r;" ill Ilower, the setpoi'nt of 1.5°F is wrong. 1S 10;;' Tavg=Tref is when auto rod outward motion would occur. . Reference l1tieJ, __ =~:~-]' Facility Refe:el'!.~e_ Number' 1[Reference S~~~~ I Page No.1 !~!~i~!,_~~_ r=====================~r====~==========~F==~== I Continuous Rod Motion ilS2.0P-AS.ROD-0003 Ir== Flo=v=e=rh=ea=d=A=n=n=u=nc=ia=ro=r=W~i=nd=o=w=E~======~IFls~2=.O~P=-=A~R~.ZZ~-0=OO=5========~II====:========~II,11 IIF===::::!;II~~ II~~ IControl Console CC2 II S2.0P-AR.ZZ-0012 II 11,36 II~~
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;Ra1s~~~p~r *ll;CSR6s~gF~p~~'1 ~l.~J~o*rVs!~~~&l~l~nllst)~J I;~§Rq;s~iWE~6111~~~:~~p~tW~~J~h~~~~~M ~ueStionTOpiCJ IRO 29 Given the following conditions: - Unit 2 was operating at 100% power when a 500 KV grid disturbance caused a Rx trip. - 2H 4KV Group bus deenergized upon the Rx trip.
21 AFW tripped immediately upon starting. SGBD flow is zero from all SGs.
- All radiation monitors indicate as expected following a Rx trip.
During performance of EOP-TRIP-2, Reactor Trip Response, the RO reports that 21 SG NR level is reading 8% higher than the other SGs. Which of the following identifies what has occurred? 21 SG is steaming less than the other generators since its RCP is no longer running. The loss of 21 AFW pump has caused Pressure Override Protection circuit activation on 22 AFW pump.
~~ ~ ~~~ ~ ~qgnitive Level Ii Comprehel~ l~ac:mtY:11 Salem 1 & 2 I!ExamDate: : ! 9/26/20111
- KA~l 003000K302 11K3.02::-~ 'RO value:Jr:IN !SRO value:I[~ [Section: l~ §:OG~ U ISRO Group:ID
~ystem/EVolutio:i!'~1 IReactor Coolant Pump System lKAStaterrient:: Knowied e of the effect that a loss or malfunction of the Reactor Coolant Pump S stem will have on the foliowin S/G II Explamltion of 55.41 (5) The loss of 2H bus, which supplies power to 21 RCP, will cause the steaming rate in that SG to go down. With AFW flow Answers: : the same to all SGs, the level in 21 SG will rise markedly, as it will not be steaming. A tube rupture would be indicated if there were no other reason for the level rise, as the diagnostic step asks if SG NR level is rising uncontrollably. In this instance, it is not uncontrolled, since it is the natural reaction ofthe reduced steaming rate. SGBD flow was included in the stem to lend support to I . . . . .
a different Group bus, and SGFP flow has been isolated by FVliI at 554* Tavg following the trip anyway. The 'preS5ure override 1circuit only affects 2 SGs, in this case 21 and 22, and would be seen equally on each SG, and in the opposite direction since it acts
'1 to reduce flow to raise discharge pressure .. The 2R15 Condenser Air Ejector would NOT be indicating normal, it would be elevated if there were a SGTR, since following a Rx trip from 100% power there will be demand on the steam Dumps and steam flow would be occuring to the condenser, which will still be sampled from the R15.
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-queStion Topic I i RO 30 I Given the following conditions:
.. Unit 2 is operating at 100% power.
.. 2CC190, RCP THERM BAR CC OUTLET V, fails shut.
Which one of the following describes the effect on RCP temperatures, if any, as a result of this failure? ALL RCP ... Ilower motor bearing temperatures will rise. I
~ Imotor winding temperatures will rise.
I 1#1 sealleakoff temperatures will rise. I Ibearing temperatures will remain the same . I
.Arlswer I ~ LExamLevel ~ :CoSlnitive level .. II Application IFaciJity~11 Salem 1 & 2 I;';-'c;:, .~
- 1'>:.11 9/26/20111 iKA:JIOO3000K604 liK6 .04 j [Ro Value: II 2.81 iSRovalue:l[ 3.1j[Section:11 SYS IR6GrotJP~1 11 !SROGr9lJP~1 11 '~llj [J SystemlEvolutionTitle!J IReactor Coolant Pump System iKA Statemel1t: ! Knowiedge of the of the effect of a loss or malfunction on the following will have on the Reactor Coolant Pump Sz:stem: I Containment isolation valves affecting RCP operation i Explanation of I 55.41(3) The CCW line supplying the RCPs is a single line supplying both bearing cooling and Thermal Barrier cooling. Once the Answ~s: , line inside containment splits, the CCW from the Thermal Barriers has its own, separate return line, which is isolated by the 2CC190 (inside containment) and 2CC131 (outside containment.) The Thermal Barrier CCW flow acts to cool reactor coolant fiowing upwards through the thermal barrier upon a loss of seal injection flow. With normal seal injection, the loss of CCW to the thermal I I h",~ior ,,1M "". ",ff,,~J 'O,"", 0("'0 ,,,"',.,. ,,..,,,,,,'e Facility R~ference* NUrTl!Jer/' *1 ,RefE!renceSectiori I!'Page No,! iRevi~iOO1
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Reference Title 1 ~:.;.:.:..
!Unit 2 Component Cooling Ir-
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11205331-3 II 11 35 -..J I II IC= II II_-..J I II II II II I LO. Number I RCPUMPE004 I RCPUMPE015 i.______ RCPUMPE016 ---l IMaterial ReqOirE:d for EXamination *. : [ II iCnrrjment .................. '.' .*.................... .... ......**.*.** **.'<1 I. I II I II 1
Given the following conditions: Unit 2 is operating at 100% power. VCT level transmitter 2LT-112 fails low. Which of the following identifies how this failure affects the CVCS System with NO operator action? VCT auto makeu ... starts. 2CV35 will modulate to keep actual level VCT level between 77 and 87%. starts. 2CV35 remains shut until VCT level reaches 87%. will not initially start. 2CV35 retains its modulation capability. will not initially start. 2CV35 loses its modulation capability. Answer [a lExam Lfi!1ffi!I] R I I [Cognitive Levell 1004000K605 IIK6.05 ! iRQ Value: 1~~valll~:J[2.5T~';~'I~ :RQ,Group:':0 rSRO Grot.iP:] 0 IChemical and Volume Control System 55.41 (3,5) The LT*112 inputs to the auto makeup control, and failing low will initiate an auto makeup. The LT-112 also provides "trip open" of CV35 at 87% level. The second VCT level transmitter LT-114 provides modulation of CV35 between 77-87% and remains available. B is incorrect because the LT114 W, LL modulate starting at 77% level. C and 0 are incorrect because an auto makeup WILL occur. fRe~ F===================~F=============~F=======~~==~1~35===
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Question Topic; Given the following conditions:
- Unit 2 is operating at 100% power.
- 21 RHR pump is CIT to repair an oil leak.
- 22 SG Main Steamline ruptures in containment.
- The crew trips the Rx, initiates a MSLI, and initiates a Safety Injection.
30m after the Rx was which of the followi locked in alarms would be consistent with lant conditions?
~ OHA E-5, SR DET VOLT TRBL.
b.: OHA C-S, 21 CFCU WTRFLO TRBL. c.: Console alarm RWST CH III (IV) LEVEL LO.
@J Console alarm CCW OUTLET FLOW LO for 21 RHR HX.
Answer I Exam Level!
- KA:l1 005000G446 I 4.21 [SRO va'u~~r4;2 I Section: l! SYS J :ROGroup: I 11 iSRO Group:] I 11
~!~,,?I~~i()n 'Title j KA Statement': "
Ability to verify that the alarms are consistent with the plant conditions. I iExplanation of ! 55.41 (7, 10) D is correct because the CCW 10 flow alarm would be present since the RHR HX CCW flow isolation valve CC16 will be [Answers: . shut, so the CCW 10 flow alarm will be locked in, as it normally is. A is incorrect because the SR will energize 20 minutes following a trip. E-5 would be in alarm until the SR energized, then clear. B is incorrect because it indicates the CFCU is running in low speed, which it would be following a SEC signal. C is incorrect because RWST level will not be at 1S.2' when the initiating casualty i<: " "innl" <<r;: h[",AJinn " ....\A1n in . RI-lR i" ">::",,tom vhi('h nrrwirlo" h"th R(,<:: ('nnlinn ::lnrl 1=('('<:: fllnr-tinn<: /"If Inormal 'Operating conditions (alarm is locked in with the !)ump O/S) and that the CC16s (cooling-to the RHR HX) are normally closed makes this guestion match the KJA. I~'" Reference Title' '~ Facility Reference Num~~.~.:{II~eferenc~~~ction-.J [!lage: No: j IRevisiO; i 1Overhead Annunciator Window E II S2.0P-AR.ZZ-0005 I! II 11 19 1 I Control Console 2CC1 II S2.0P-AR.ZZ-0011 II II I~I II I! II II-~ I I LO,Number I RHROOOE008
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;Facility: I I Salem 1 & 2
*. ExPlanatio.*.n 1 Answers:
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...;nJ leg 55.41(7) At 15.2' RWST level on Unit 2, the 21 and 22CC16 valves will automatically open in anticipation of placing the RCS on cold recirc. The large (-8,000) gpm additional CCW flow required will auto start the CCW pump in auto. By the time this level is
* ~~~ reached, the SI signal would have been reset, restoring automatic operation of the pumps auto start circuit at 70 psig system pressure. B is incorre~t but plausible if the reduced SW,flow were thought to I~:;~e~ \~;,;,~~~~!~t t~nm~?n~i~y~:n parameters. The I through CCHX. C is incorrect because the splitting of CCW system headers in LOCA-3 occurs after the transfer to'CLR has been performed, and does not add to system flow nor reduce system pressure. D is incorrect because operation of the RH20 increases the flow bypassing the RHR HX, so there would be less:oollng required, and would not cause system pressure to lower.
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QU\:Jstion Topic ! IRO 34 I Given the following conditions:
- During a SBlOCA in MODE 1, a Safety Injection signal is actuated successfully.
I* ReS pressure is 2,000 pslg and lowering slowly. Which of the following descrIbes how the Safet:r: Injection eumes are erevented from overheatIng? l~] jA portion of discharge flow is recirculated back to the RWST. I I Injection flow from the RWST into the RCS passes through the pump. I I Injection flow from the RWST into the RHR system passes through the pump. I
'd,' IA small portion of discharge flow is recirculated to the suction of th'" pump through a Recirculation Flow HX.
I j'AnS\'/er t Ia I fE;am Level r R ICognitivel-evell1 Application IFaciU~y;11 Salem 1 & 2 I 'E?Ca~Date~11 9f26/2011 I 11006000K406 ti K4.06 i((ovahle:112.7r:SRov~'ue:j[~sec~~§I~,Rc?GrouP~1 11~SROGrOllP:11 11 ~I1c. [] System/EvolutionTItle IEmergency Core Cooling System 006
'KASt~ementd~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~__________~
RecIrculatIon of mInImum flow throu humps Explanation of, 55.41(8) Each SI pump has an orificed 1 1/2" line off its pump discharge piplnfJ, whIch connect Into a common 2" line. The iAnswers: "! common line passes through 2 normally open and deactivated MOVs (SJ67 and SJ68) and return to the RWST. The flow through
-,-----~ the pump cools It until the shutoff head for the SI pumps -1560 psig is reaches as RCS pressure lowers during the lDCA. B Is incorrect but plausible if the 2,000 pslg condition in the stem is ignored, as it is the normal f10wpath when RCS pressure lowers to where ECCS accumulators also join RHR system into RCS cold legs. 'D is incorrect but'plausible, since there is no'recirculation cooier, but other pumps (specificall:r: AFW) have recirculation coolers .
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~.~*-*~l ~uestion TOPifi RO 35 I Which of the following relief valves lifting would cause a rise in PreSSUriZE!r Relief Tank temperature or level?
IRHR HX outlet. ISI pump discharge. ISeal Water Return line. [J I Containment Spray Pump discharge.
~~....e,r II c I ~9arnl..e~ I R I :Cognitiv~Le.Jel . ! IMemory I !Facility: II Salem 1 & 2 I iExamOate:i I ~~I 007000A301 1r§=6"C~= [B~~t~i[TIIlsection:!I~ ~~~~Ui~R6Group~[JJ sySternlEvolution*~ r Pressurizer Relief Tank/Quench Tank System
- KA§tatement:: Abili to monitor automatic operations of the Pressurizer Relief Tank/Quench Tank System including:
Components which dischar e to the PRT
- EXplanation of i 55.41(3) The 3 distracters are components which used 10 be directed to the PRT but were re..routed to the containment trench AnSIJV~rs:~ during a OCP.
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.Que!i;t~~.~ource.C6m~:tS' IVision Q61390, inputs to PRT, changed correct answer and all distracters, iCO fJ)IH"'"< .*, . . . ( ....... *. ..... ....J' ******>i*< . . ....). t .:'!.>,'> I I
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Given the following conditions:
- Unit 2 is operating at 65% power. - An intersystem leak develops between the CCW system and the Spen'~ Fuel Pool Heat Exchanger.
CCW Console Alarm SURGE TANK LEVEL HI-La alarms on 2CC1, Which of the following describes the action required lAW S2,OP-ARZZ-0011 Control Console 2CC1? Open 2DR107, MAKE UP TO CC SURGE TANK from the Control Room to restore CC surge tank level. Close 2CC149 CC SURGE TK VENT to prevent CCW from being Introduced Into the ventilation system, Drain the CCW surge tank from local drain valve to a 55 gallon drum to prevent overflowing tank to the In-service WHUT. ~~~ ~ ~.~ ~ ,pognifive Level.! IAppllcatlor iIIty:11 Salem 1 & 2 I lE?CamDate: i I KA~I008000K408 I[K4.Oa ~,SR.OValue:I@:sectiOO!ll~ROGn::lllP;!D[SRQ~r~D m !SYstem/EvolutioriiitiE;l I 1Component Cooling Water System !KA Statement: I Knowledge of Component Cooling Water System design feature~s) and or interlock(sl which erovide for the followln9: I Operation of the surge tank, Including the associated valves and controls I 1 iExplanation of i 55.41(7,8) The first thing that must be deduced is which way the leak will go. In this case, the SFP HX will be at a lower pressure Answers~ I than the CCW system, so the leak will be OUT of the CCW system. This causes surge tank level to go DOWN. The second part of the question requires the operator to know that CCW surge tank makeup is performed from the control console by opening the M/U valve. The local manual valves are normally aligned with one of them open so that remote operation can fill the CCW surge bink A.-h ,,,I I<fA '" 1<"- f1? "nt A. f1A IMnrrllr'rI Ie: "nm>"t fl'\r "- f1? "t1 th""' ,",,,,t ""mnl", nl<l" "':H/" i :R~ference Titre* ..,.i L~a(;ilitYReferef)ceNumber '. i IReferenceSe~tioll'] [Page No.! [Revlsi0l"\' IComponent Cooling Water 11205331-1 II iI 11 52 i IControl Console 2CC1 II S2.0P-AR.ZZ-0011 II II 1157 ~ 1 II II II II I
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I CCWOOOE008 I CCWOOOE012 II Direct From Source
"Ros~'Sc:~~f,"C'1 ~SRO~kVsC~~Pei;1 ;§~!~i~nTOPiC! IRO 37 Given the following conditions:
Unit 2 was operating at 75% power. Plant conditions develop which required a shutdown and depressurization.
- At 1,100 psig during the depressurization, intermittent loss of shut indication for the 2PR1 and 2PR2, PZR PORVs, was observed.
The pressurizer PORV block valves 2PR6 and 2PR7 were shut, with power maintained to the valves.
- 2PR1 and. 2PR2 were placed in manual.
RCS pressure is being maintained at 800 psig due to problems with isolating the SI accumulators.
- The RCS cooldown continues to below 312°F.
Which of the following describes the effect if the operator were to arm the Pressurizer Overpressure Protection System (POPS) under these 2PR6 and 2PR7 would OPEN; 2PR1 and 2PR2 would remain CLOSED 2PR6 and 2PR7 would remain CLOSED; 2PR1 and 2PR2 would I'::,KO emu " , K I WUUIU f!::nm:ilfl vLV0CU; " , K ) e:IIIU ",K':: WUUIU ItHTlGllf1 vLV0CU
- AnS;;..!£] E=J ~ :Cognitive Levelull Applicatior~FacilitY: ! ,.Salem 1 & 2
.RO Varue: I' 4.01sRO Value: :[~section:! I~
iS~~~/Evolution Title i IPressurizer Pressure Control System I 010 I iKA ~+, i Ability to manually operate and/or monitor in the control room: I PORV and block valves I Explanation of i 55.41(7) With pressure above 375 and temperature < 312, arming POPS would cause the PR1,2,5,610 open, regardless of the 'Answers: I MANUAL selected for PR1 and 2. ';" ..,"""',.... Reference Title =:J' I FacilityReference Number, .' 'Reference Section J t Page No.,1 ~~", *."'..... '1 ~lp=R1=,2=P=Z=R=PO=R=VS======~====~111=23=1=35=7======~=====~ilr==~====~11 1114 I [ 1\ II II II I I==:=:=:'=:'=:'=:'=:'=:'=:"=:"~~=:'-=_=_=--,~II!~======:::;II'==:'=:'=:'-~~=-=::II II I
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- Question Topic
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"""';~~.......J Given the following conditions: - Unit 2 is operating at 100% power. - I&C Technicians are in the middle of performing PZR Pressure Channel II (two) calibration lAW S2.IC-CC.RCP-0018, 2PT-456 PRESSURIZER PRESSURE PROTECTION CHANNEL II.
What would be the effect on RCS pressure if the controlling PZR pressure channel instrument were to fail LOW, with no operator action? RCS will ... ICognitivel.evel ,I iKA:i 1010000K301
- System/Evol.':lti~n Title i !Pressurizer Pressure Control System KA Statement: ' Knowledge of the effect that a loss or malfunction of the Pressurizer Pressure Control System will have on the RCS I Explanation of! 55.41 (7) PZR pressure controller will see a low pressure condition from the failed instrument. This will cause spray valves to close Answers.:......_J fully, and all PZR heaters to energize, C is correct becasue actual RCS pressure will rise in response to the heaters being on, and will continue to rise until reaching the Rx trip setpoint. With Channel II 0/5, the automatic function of 2PR2 has been rendered Iinoperable (Note prior to Step 2.5 of procedure) will be shut, so PR2 cannot control RCS pressure, PORV actuation is on 2/2 channels (1 &3 for PR1. 2&4 for PR2) Since one of the ,'0 Ir'" I ch"noQI" (I N' III) '" hil"d I",,, i/<: P('\RI "'"""'0,,,1 ";'00 t:.. .
because neither PORV will act to lower pressure. Sprays will not open due to the minimum demand from the MPC. B is incorrect because pressure would rise, but plausible because OT/DT would actuate first to trip the Rx on lowering pressure. D is incorrect because pressure would rise, but plausible because it is a low pressure trip, and operator would also have to know that OT/DT would act first on lowerin~ pressure. II ., .... . . . ' ReferenceTit.le 'I
, , . Facility Referencl)f>lI.I~ber ....... iReferenceSection it Page No. 1
- Revisi0!1.J I
! Pressurizer Pressure Malfunction II S2.0P-AB.PZR-0001 II Attachment 2 !122 IUC-J I 11 II 11 II-~ 1 II II II Ir=~ IL.O,Number* IABPZR1E001 J_ _ _- - l
For each of the listed PZR heaters, identify their 460V bus power supply. 11 control group __ 11 BtU group NORMAL _ _ 11 B/U group EMERGENCY _ _ 12 BtU group NORMAL _ _ 1E, E, A, G, C. IG, E, A, G, C. iAnswer' ~ Exam Level] ~ ~t!YILeVeill Memory I iFacility: II Salem 1 & 2 I [ExamDate: , I 9/26/2011J IKA:[I011000K202 1K2.02 iiRQValUe::l3.1IisROValt.ie~I~.:~rsectionll~ROG;:;;i;p?1
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2!IsRCrGroup:11 21 i... ."..... ,~ ~'Title; i"!.l'....'v~.u_y~.u .. ~** IPressurizer Level Control System 1 011 IKA<l'~ ,,;. ;';:.,t~ Knowledge of bus power supplies to the following: PZR heaters
'1:,.,,,,. lof 55.41 (3, 7)The distracters all contain the possible busses, but in incorrect orders.
i !:ri~:~';~?,,~ Reference Title :> "C", I Facility Reference Nun-:~~r2J f.~eference Section. 1 "1! Pl:l~e No*1 :Revis~ INo.1 Unit 1GP 480V bus PZR heaters controlle i1203347-1 , I! II 11 13 1 INo 1 Unit Backup Group 11 11203348-1 II II 11 9 I INo 1 & No 2 Units Group 12/22 backup groups 11247992-1 II II 11 3 I [Material Requil'erJ for Examination .,! [
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R6'skyScralre~' I "SRO Skyscraper [Question Topic IRO 40 I Which of the following identifies why a Safety Injection signal on Unit 2 is reset early in the EOPs after transition out of EOP-TRIP-1, Rx Trip or Safety Injection, following a LOCA? SI is reset because ... ra.: Ithe Phase A signal must be reset to allow the ECCS pumps suction valves to be realigned from the RWST to the VCT after SI termination L--' criteria have been met, and Phase A cannot be reset until the SI signal is reset. I I
~ it allows operators to regain control over plant equipment and prevents any equipment from automatically repositioning when RWST reaches I
15.2'. I c.: the Phase A signal must be reset to allow sampling of the SGs and RCS, and Phase A cannot be reset until the SI signal is reset. I
@J Iit allows operators to regain control over plant equipment and restore a sustained compressed air supply to containment.
I Answer II d I IExam level' IR I iCognitive level II Memory ~ :Facility: II Salem 1 & 2 I [ExamOateJ I 9/26/20111
.~ iRO Value: i[TI' ~RO Value: ;82J ,Section: !~ J RO Group:! U ,SRO Group:; U
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~KA:II 012000G406 1;2.4.6 I ~========i !System/Evolution Title I_R_e_a_c_to_r_p_ro_t_e_ct_io_n_s"-ys_t_e_m_ _ _ _ _ _ _,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ --'1 ro~
- KA Statement* . ,I Knowledge of EOP mitigation strategies.
Explanation of I 55.41 (7)The Ph.,e A "9'" CAN be c",et with"tee,ett"9 5' " it h" aceteot"e memo,", o'cooit. See Note 10 0" 221057. Sa,e, Answers: 1 document states that the SI reset function is so that equipment can be aligned, and to restore a sustained, compressed air supply to I j allow control of air operated equipment in containment (e.g., charging and letdown valves, PZR PORV's, etc.) A is wrong both because the reason and the Phase A reset logic is incorrect. B is incorrect because it does not remove the standing "s" signal "hi 'h . ., ,~II;' th ~ 111 'l ,~~, "onrl tho ('("'1 A ('('~lY ~ . ,tI.,t,~ ,~~"t~N,~tI,'~I"
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Reference Title i!FaCility Reference Number ! [Reference Section I' Page No*IIRevision. 1Reactor Protection Signal Safeguards actuation 11221057 II II I~-I I Loss of Reactor Coolant Bases Document 112-EOP-LOCA-1 II 11 18 I~~ I II II II Ir=~ LO. Number J I RXPROTE024 I I I 1-----'1 [Material Required for Examinatio~ I II IQuestion Source: IINew I IQuestion Modification Method: ~ " I [Used During Training Program, I D
- Question Source Comments I I I
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_~~stionTopic: . RO 41
'~-~~-~-I I During a power ascension from an initial power level of 4%, Permissive P*10 does not actuate when expected. *Which of the following describes the effect of this malfunction on the Reactor Protection System if the power ascension were to continue to 14% Rx \
power?
'a. 1 IA valid OT/DT signal would NOT trip the Rx.
IHigh Steam Flow Safety Injection will remain blocked. I IThe low power Rx trips could NOT be blocked until P-13 energized during Main Turbine startup. I
~J 1A Loss of Off-Site power would NOT initially cause the Reactor Protection System to trip the Rx.
I 'Answer I d
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I I IExa~Leveli IR
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I :Cogniti1l'e.Leyel i IComprehension I iFac.ility:jI ISalem 1 & 2 I ExamDate:; ! 9/26/2011J I KA:'I012000K406 I MQ6-- @.~.~~£]0]~1~ ~~.[J]~6Gr<fuP~U [!~~. oiM-<;;-;"" " 0 !YsternJEvolution Title I !.;..R,;.;;e,;;.a,;;.ct;;;.o,;..r,;..P,;.;;ro.;;;te:..:c:..:;ti,;;.o.;;;n...;;S.t..y,;;.st;;;.e.;..m~_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _...J _01__2____ ....1 IKAStatement:1 Knowledge of Reactor Protection System design feature{s} and or interlock(s) which provide for the foliowin;iF I Automatic or manual enable/disable of RPS trips I rExPlanationof! 55.41 (7) C is incorrect because the low power trips can I\JOT be blocked until P-10 is actuated, D is correct because the "at power" Answ~~L..2i Rx trips (which are different from the "low power" trips) are not unblocked until P-7 is actuated, which gets its input from either P 10(which is not actuated) or P-13, (which is not actuated because the turbine is not online at 15% power), At 10% power the AFW pumps will have been secured, and while the loss of off site power will cause a loss of the SGFPs, the SGs will not shrink to the 10 c" , ',;'; ,II, .. .h ,~, .~ f'"'ITIf"\T ' '~h"..,,,., ~I 'I..,h C* ;~. .h *.." ,~, +h h; ~I I::I~ ~I ' L ,<;;u >543 (P-1-2.) Reference Title' III *.**.. **Facility Ref~rence Number* .;. I ;'ReferenceSection! !Page No.1 Revision
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l I 1RPS logic 11221053 II II lis I IRPS logic 11221054 II II 11 10 I I , II II II II I , - - -....... ~-....,..,-~~~-
- L.O. NUmber I RXPROTE02S IRXPROTE027 Material Required for~~aJ1liriCltio!l".! I II
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l~uestion.!3~yrcecomment~i IOS01 C41 j i I I I I I I
- ~~~~~ IRO 42 I Given the following conditions:
!- Unit 1 Rx power is 8.1 %.
- Power is being raised slowiy in preparation for rolling the Main Turbine.
- 11 SGFP is in service supplying FW to SGs,
- 12 SGFP is latched and at idle speed,
- ALL AFW pumps are aligned for normal standby operation.
11 SGFP trips, Which of the following describes the effect, if any, this will have on the AFW pumps with NO operator action?
~_I The MOAFW pumps and the TDAFW pump will start when SG levels drop to the 10 10 level setpoin!.
I IThe MOAFW pumps will start when 11 SGFP trips. The TOAFW pump will start when SG levels lower following the Rx trip. I ALL AFW pumps will remain in standby. Sufficient steam will be supplied through the 11-14MS18s, MS STOP BYP VALVES to supply 12 I SGFP. \
! ALL AFW pumps will remain in standby. 12 SGFP will remain in service since at this power level it is being supplied with steam from the Heating Steam System. I i
,A.."..". ~EXam Levell~cognitiv~Tevel "I IApplicalion I iFacility:11 Salem 1 & 2 I :E~al1loate:ll 9/26/2011J
!~:iI013000A104 ISystemfEvolution ~" 'KA Statement: ~--
- i'Explanation
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i Title I A1.04 I
,'RovaTr:reDr.TI":SROV31ue::~lSection:]I~iROGrOup:lU~Group:jU ~11; 0 I Engineered Safety Features Actuation System Ability to predict andlor monitor changes in parameters associated with operating the Engineered Safety Features Actuation System controls including:
S/G level of- J: 55.41(4,7,8)0 is incorrect because the operating SGFP(s) will be placed on Main steam supply prior to exceeding 5% power (IOP-3, I iAnswers: """~ step 5.4.10). and 12 SGFP will not provide sufficient discharge pressure at 1100 rpm (idle speed), and the stem states with no operator action so speed will not be raised. A is correct because the MOAFW pumps and TDAFW will start on 10 10 level in SGs as the source of FW is lost, and the SGs continue to steam. C is incorrect because the MS18s could provide sufficient steam flow to
, om IMOAFW pumps
'c:::r.:>:p ",t
- is a trip of BOTH SGFPs.
AO/, nn\Al." h"t Ih" ,,,,,,,,,,1 ill nf'>t ric:A nn 1? c:r.:>:p R j<: ' hAf"'" "' tho ""I" ct"rt "i"",,1 f", th" 1-..-.~""-~~-RefertmceTitle . . J I Facility~efen~nceNumbe7Zl ~~eSection-.J ~~ ~~ irRPSAFWPum;~StartU~" 11221064 II q 11 8 I II II 111=::::::;11 I 1_ _ _ _ _ _ _ _-""11_ _ _ _ _----111_ _ _----111 II_--.J 1 AFWOOOE015 I 1 I 1-----'1
~erial RequiredforExaminatio,: iI II
~~esti.on Sour~~;i IFacility Exam Bank IiQuestion M6~.ifld~tionMethO;:l: *:1 Editorially Modified IiUse~ During Trajning~rogram" =.J
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~estion sourcecomlTlentsl\Vision 085462. Modified stem from MSLI to trip of operating SGFP with other SGFP latched. Correct answer
'--~- ... . remains the same. Enhanced distracters .
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Given the following conditions:
- Unit 2 is operating at 4% power during a startup.
- A RCS leak causes PZR pressure and level to lower rapidly.
- The operating crew initiates a Rx trip and Safety Injection.
- When the SI is initiated, a loss of off-site power occurs.
One minute after the loss of off-site power, which of the following describes a condition which indicates a failure of equipment to actuate as is taken? 24 SW pump running and 23 SW pump stopped. Charging Systems SI Flowmeter reads 100 gpm. 21 ABV Supply Fan running, 22 ABV Supply Fan NOT running. IR I ~()Slllitive Level J did not stroke full open, or less than 2 charging pumps started. TRIP-1 page 1, and TRIP-2 page 2 flowcharts, page 1, each Table IA shows Accident Loads and Blackout loads respectively
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'. Facility Referl'!llc:;eNuml>er . IIRef~re~ce SeCtio~: LpageNoJ lr
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i ...... Reference Title iI IRx Trip or Safety Injection 112-EOP-TRIP-1 II II F1 1127 1 ,IRx Trip Response 112-EOP-TRIP-2 II II F1 1127 I 1 II II II II I
'Material RequirEldfo(EJcamiriatioG !
****.*~o'~lIs~~?I}t&RO$k,,~~iQ~~ll "';f~bsv~ij~(tit19ri:t1~t}0f.1 i~sRO~~t~irllE~~llIfi~~~i~ttbl ;,JQ~tltQ~'t~~nA~'s'.'1 ~-. "--'--1 ~estion To!:!~.i RO 44 Given the following conditions:
o Salem Unit 1 is operating at 100% power.
- A small LOCA (300 gpm) occurs, and a Rx trip and Safety Injection are initiated.
- A loss of off-site power occurs when the Main Turbine trips.
o NONE of the CFCUs start in Low Speed. How will the failure of the CFCUs affect containment instrumentation reaclings? Containment will be readin than it would be expected to read with all CFCUs in service IRadiation; higher. ,Kkl1022000K302 IK3.02 fROVaJUe:] [I§jfSRO :va~@ [~nl~Ro.~roue:J[JJ iSRC:li3rOUp? 0 1~~~V()lution Tit~ IContainment Cooling System I.KA Sta~l!lenill Know1ed e of the effect that a loss or malfunction of the Containment Cooling System will have on the following: Containment instrumentation readin s
- E~pf.am:~tion of I 55.41 (5,7)The CFCU HEPA filter, which is in service following accident initiation, is designed to remove minute radioactive
'Answers: . . particulate matter from the atmosphere so that offsite doses do not exceed the limits set by 10CFR100. A is incorrect because the lack of cooling function provided by the CFCU will cause the pressure to rise. C is incorrect because the Dew Point will rise due to higher
,. temperature. in containment. D is incorrect because the leak detection system will rise as condensation occurs on its cooling
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rQuestionToPlcl L...- -, IRO 45 I Given the following conditions:
- Unit 1 has experienced a LOCA.
- Control room operators have progressed to the point where the SEC has been reset.
- Containment pressure was 4.5 psig when the SECs were reset.
Which of the following describes the containment spray system response should a hi-hi containment pressure signal be generated at this point in the
. accident?
CS system valves ... I
~ would realign for spray, the CS pumps would be started by the SEC.
I Iwould realign for spray, the CS pumps would have to be manually started. I I ic.: must be manually realigned for spray, the CS pumps would be started by the SEC. I i Imust be manually realigned for spray, the CS pumps would have to be manually started. I l'A.ow '" : I b I tEiam l~...e51 R I iCogrlltiin:levefCII Applicatior~ tFaciliiy: II Salem 1 & 2 I :ExamDate:! i
~I026nnn-\301 l'A3.01!Royalue:i@[!R.()value:i@,Secti"~I~;ROGrof.ip:iUlSROGroup:jU ~. '
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~-~ .... -~.
Title' IContainment Spray System I '026 I mm lKA Statement:! Ability to monitor automatic operations of the Containment Seral S:tstem including: I Pump starts and correct MOV positioning I jlExPlanation ofi 55.41(7) The SEC controller operates the CS pumps at:2 different point in the sequence UNTIL the SEC is reset The SEC ONLY [Answers: ..... i controls the CS pumps, not the CS valves. The valves realign on the hi-hi cant pressure signal whenever it is received, but once the ,. ~ SEC is reset it will not start the CS pumps since the sequencer is no longer active .
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Reference Titre ]\" Facility Refenince Number J ;Reference section ,. 1rfi~ge No*1 Revisionl IRPS Safeguards Actuation Signal 11221057 I! II 1122 I ISafeguards Emergency Loading Sequence 11203668 Ir== il IE-I ISEC Lesson Plan (CS pump start seq explanati !INOS05-SECOOO 1\ i117 11_6~
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iL.O. Number .' i ICSPRAYEOOa I CSPRAYE009 I !Material Required for Examination ",1 I I fI"",.,tiMSource:) Facility Exam Bank I:9u~sti()n~Odifi!=ation'Method: :J Direct From Source Question sourC~?Orrimelltsll_V_IS_I_O_N_Q_8_0_5_6_7_ _ _ _ _ _ _ _._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ---1
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lfluestibn Topic! IR046 I IWhich of the followin9 identifies the coincidence re9uired for manual Containment Sera~ actuation? I 11/2 'I,, -, ' i BOTH safeguards trains, I 1212 keyswitches on BOTH safeguards trains, I Ie. 1 11/2 keyswitches on EITHER safeguards train. I 12/2 keyswitches on EITHER safeguards train. I I ~s;:"er ! d I I ;Exam,:~vel] IR If',-.,' .:.. Levell j Memory I Facllityll Salem 1 & 2 Ii I 9/26/20111
!KA::1026000A401 I:A4.01 IRO~14*WSRova,U;nf 4.3T:Seqtioll~H~R~Gr()up'!1 1aSROGrOuP~1 11 at' 'System/Evolution Title! IContainment Spray System iKAState;ne;rt:l r-:-;:;:.;.;.:;.<:...:..;;....:..;.;.;;;.;.;.;;;;;;;.;;.J.-=;.;:;;..;;=-;;;;.;.;..;::..;:;.;....:..;.;.;:;.;.;.;.;.;;.;....;;..;...;;;..:.::....;:..::..;.;.::...;;.;....;.;;.;;;;..;.:..;------------------------+
- explanation 1- _"c'_--'_
of 55.41 (7) Salem has two safeguards trains, either of which performs its safety related function. Containment Spray requires 2/2 i :Answers: .' .. i keyswitches turned simultaneously to the operate position to activate containment spray, due to the severe consequences which would occur if it were inadvertently actuated without it being required. Either train of safeguards will perform actuation. 2/2 I keyswitches on EITHER train. ICSPRAYE008 1-------1 Material Required for EX~lllir'lation iI Questions<:urce;~ : Facility Exam BankI IIQuestion Modificaticl0 Method:, IConcept Used IlQlIestion Source cO~TE~TslI VISION 61822 concept used .
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!<=IuestiqnTo~I<;j ,RO 47 Of the following, which one describes how the Spent Fuel Pool level will be lowered if required, lAW S1.0P-SO.SF-0001, Fill and Transfer of the Pumped with Spent Fuel Pool Cooling pump to RWST.
Pumped with Refueling Water Purification pump to in service CVCS HUT, Drained via the Spent Fuel Pool Skimmer loop to the Drain Header to the in service cves HUT. ICognitive Level JIMemory ISalem 1 & 2 1 [Examoate: 'I 9126120111 ! '. .E......xPI.a.na.tion Oi155,41(8,13.4) The RWST normal level is -41', The puts the water level at 141'. The Spent Fuel Pool is -128', There can be no
!Answers: '. . gravity drain TO the RWST, although it will work coming FROM the RWST to the SFP. Transferring water from the SFP to the ,~.~- RWST is done by manipulating valves within the SFP CelOling system, and pumping the water with the Spent Fuel Pool pumps to the RWST. The other method is to drain it to the CVCS HUT. The Refueling Water Purification pump cannot take a suction on the Iperformed by procedure and it would d'rain to the Fuel-Handling Bldg sump not the eves HUT ! . Reference Title '" Fac.:ility Ref~rence Number.; if~~!~~~nce SectfoG ~ Page No.! :Revis!Onl IFill and Transfer of the Spent Fuel Pool II S1.0P-SO.SF-0001 II II II~~
IUnit 1 Spent Fuel Cooling 11205223 II II 1\26 ~ ITank Capacity Data II S1.0P-TM.ZZ-0002 II II
'I 11_8_~
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**RO skVscraper:'~1 . sIr116c;a~~ I ,~U~_~~0!l_!~~ I_R;..;.O_4;..;.8~_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ---Jr Unit 2 is performing a plant startup to full power.
Predict the Steam Generator Narrow Range Levels for the following REACTOR power levels. Assume a normal power ascension. 10% REACTOR Power = NR level. 60% REACTOR Power = % NRlevel. 138 .5%. 1 fb.138.5%. 44%. Explanationofl 55.41 (4) SG NR level is programmed so that as Turbine power rises from 0-20% (Turbine steamline inlet pressure) it rises from 33 Answers: J 44%. On Unit 1 only. 20-100 it rises from 44-48. On Unit 2. A is incorrect but plausible if candidate thinks programmed level is
~-..~ based on RX power. not Turbine power. and uses Unit 1 ramp from 20-100 (calculated at 60% power. lurb power =Rx power after HDPs put in service at 50% power.
-~--..- Reference*'-t~~e . ~-~-~_.~ ..........J~i: FacilityRElference Num-ber-'~:Reference Section] -:P-a-g-e-NOl .~~~
IMain Feedwater/Condensate System Abnormalitll S1.0P-AB.CN-0001 II 1121 11 18 I IMain Feedwater/Condensate System Abnormalitll S2.0P-AB.CN-0001 II ~== 1121 11 26 I I II 11------'11 11f===:::;1 '.'.---.~-.----,.,~-.,,~ .. ,L.O. Number
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I STMGENE007 I IADFWCSE015 I I I Material Required fo.!.~amlnation I II -Question Source: ] INew IIQuestion Modification Method: L 11 I:Used During-Training Progr(lm'] 0
-Question 'Source Comments I
_._ _ _ _..__._.._ _....__ ..--J I I~u ......",... '. .e I I I I I I I
iijQ:~vs~r~p~ril ;~SROS~S~raDe, I '~<~~OSvStemjE~olu.i~n'ustil~osv~~trtlE¥~F.tftltr~'list 'I'" butlioePhariQ~$ IR049 Given the following conditions:
- Unit 2 is operating at 50% power. - 24MS167, Main Steam Isolation Valve Fast Close PB is depressed on the control console.
Which of the following describes SG pressures 5 minutes after the PB is depressed with NO other op ator action? Main Steamline Pressure in 21-23 loops is ... lower because of the increased steam flow from 21*23 SGs ONLY. higher because Hi Steamline Flow Safety Injection signal has occurred. higher because a Steam line Delta Pressure Safety Injection Signal has 0 lower because of the increased steam flow from 21-23 SGs AND ttl steam flow from 24MS10.
!F",ciiiiY:11 Salem 1 & 2 I !8{amDate;' !
c~1 039000A 106 I.~_~ .ROValue:l ~~. ~ITTI~~.~~l~J fROGrOlJP~D[SR(fGiOuP:]O
~~§'/~§~ Main and Reheat Steam System ~
arameters associated with operating the Main and Reheat Steam System controls
'EXplanation of! 55A1{5) The closure of the 24MS1 will cause its pressure to rise. The 3 other loops will have to make up for the loss of flow
- Answers: . i from 24 loop, and the increased s am flow from each loop will lower steam header pressure as the MT governer valves open to
, maintain first stage pressure co tant. The steam line DIP SI signal is on one SG LOWER tha all the others not higher. The High Steam line flow SI would not oc ur because even if the high steam line FLOW were present, it is coincident with LOW steam I 11- 11l=~II-~
I=======I==:::::!,'F'======::::;1[ II II I 1-------+----11 IC= II II I L.o,Nutriber [ MSTEAME015
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"RO Skv~c~Ire{lsRbskvscraper "RO sV$t~~v~'ruij~nJf~~:: ,I ~o SvsteIn/Eyoluti\i~:uSt c. 9'tIt1irie:. Ch~Q~li~1 cQuestionTopic: [:...R;..:.O:.....;,.50:.....;.._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _--'
Given the following condition:
- Unit 2 is preparing to open the 21-24MS167s, Main Steam Isolation Valves during a plant heatup.
Which of the following identifies why the 21-24MS7s, SG DRAIN VALVES are opened before opening the MS167s lAW S2.0P-SO.MS-0001, Main, Ensures a vacuum pathway to the Main Condenser is available free of potential loop seals. Removes any collected corrosion products or impurities to ensure Main Turbine blading is not impinged. Preheats susceptible components such as steam traps prior to exposing them to full system temperature. Prevents pressurized steam from forcing residual water in the piping to cause water hammer on downstream components. iCognitive Level [Facility: II~alem 1 & 2
- ~~Value: !12.91.SROVal;:;e:l[:UJ!Sec~~~,d SYS l~oUP::1 11
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~,Statement: i Knowiedqe of the operational implications of the followinq concepts as they apply to the Main and Reheat Steam System: I Definition and causes of steamlwater hammer I of i 55.41 (4,10) The steam lines are designed to pass 99.25% quality steam at full power. Water which has accumulated in the piping Explanation Answers: ,'.' "I during cooldown would be transported downstream, where it would impact inner piping walls and turbine blades if not removed. A is I
- - - - - - - - incorrect because the vacuum path would be from the condenser back through turbine control and stop valves which would be
,shut. B is incorrect because while it remove anything in the condensed steam in the piping, it is not the reason why. C is incorrect I bec::;allse it is oot ll/b)! steaoo lioes ;ace dr;aioed ;aad lliOllld hQ ,..,,111 ,,,,t,,,r n,,1 u"rm/h,,1 1, 1
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FacilitYReference Number 'Reference Section : Page No. lR.. "';;"'~"I II Main, Reheat, Turbine bypass Steam Warmup. II S2,OP-SO.MS-0001 II 1114 1 !22 J I II II II 1 1
------------'11----------'11------'11----'
IMSTEAME013 IWATHAME006
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[Question~TOPiCII RO 51
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Given the following conditions:
- Unit 2 is operating at 100% power.
- Main Steam Dumps are in MS Pressure Control - AUTO, set at 1005 psig.
AsstuYle- II~ If
- 2PT-505, Main Turbine Steamline Inlet Pressure Channel, fails LOW. (),aer~ a..cnC"'h
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Which of the following identifies how the Main Steam Dumps will respond to this failure prior to any Reactor Protection System response? Main Steam Dump valves wilL .. I remain shut throughout the event. I IALL trip open and then modulate in the closed direction in response to lowering Tavg. I I I ALL trip open and remain open until they automatically shut when Tavg lowers to 543 of initially remain shut, then modulate open as steam pressure rises from the load reduction. I II i ~swer II a , [Exam Level. j R ICognitiv~ Level .' i IApplication I :Facility: i ISalem 1 & 2 I 'Exam Date:' 'I 9/26/2011J I KA:i I041 00OK603 P<6.03 IIRO Value: I 2.71 ;SRO Value: l[z.9!section: I!~R()GrouP:iI 21 iSRo~roup:!! 21 ~!li 0 M Statement: I Knowiedge of the of the effect of a loss or malfunction on the following will have on the Steam Dump System and Turbine Bypass Control:
*Controller and positioners, includin ICS, S/G, CRDS r:<----.. -'~.~.~-~"""1 .Explanation of: 55.41(5,6,7) When the Main Steam Dumps are placed in MS Pressure Control Mode, that "arms" them, and they are set up to iAnswers: respond to a deviation from its setpoint, which in the sten is stated as 1005 psig. That is, until Steam Pressure as sensed by a --.~-<.~<<-~ different steam pressure detector (PT-507, Steam Header Pressure, not PT-505 Steamline Inlet Pressure) rises above 1005 psig, the steam dumps will remain closed. 100% power steam pressure is - 800 psig, and it would rise as a load reduction occurred.
I .. Icorrespondingly, secondary side steam pressure.
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i :~.!stion Topic I I RO 52 I, I i! Given the following condition: I - Unit 2 is performing actions to prepare the Main Turbine for operation, Main Steam header is currently depressurized, Which of the following actions, if performed when directed lAW the aporopriate procedure, would cause all the Main Turbine Stop Valves to open? The NCO selects LATCH at the TURBINE PROTECTION Bezel. The NCO selects TRIP RESET at the TURBINE E-H CONTROL & STATUS panel. The NEO at the Front Standard places the RESET-NORMAL-TRIP lever in the RESET position. The NEO at the Front Standard places the Vacuum Trip Latch Handle in the LATCHED (up) position. procedure, S2.0P-SO.TRB-0003, Main Turbine the actions are contained in the procedure, and nArtnrrnAI1 l1.,r",",,,,r,,., the latched PB with the steam header not open the valves. rRefetenceSection LC. Number 'j 1 MNTURBE008 I I
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QuestionTopic! I..;..R.;..;0..;..5;",;3~_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _---Irl Given the following conditions:
- Unit 2 is operating at 80% power. steady state. - Power was reduced 2 days ago when 21 Condensate Pump tripped.
21 Condensate Pump remains O/S. The Condensate Polisher is in service with full flow. Which of the following identifies the initial concern if 22 Condensate Pump were to trip. and the action which would be performed lAW S2.0P-AB.CN 0001. Main Feedwater/Condensate System Abnormality, In response to that concern? Lowering SGFP suction pressure. Open 21-23CN108 Polisher Bypass Valves. Lowering SG NR level. Initiate rapid load reduction at up to 5% I min to <49% power. Lowering SGFP suction pressure. Initiate rapid load reduction at up to 15% / min to <30% power.
~~ ((:J ~E! ....~JC()gnitive Leve" i IMemory ---I ISalem 1 & 2 I IExamD~te:i I 9/26/20111 rKk1056000A204 I A2.04 [ROVal~~§!SRdVal~l§~~I~ ~~3U*SROGr<>uP=.lU
[SystemlEvolutionTitlei I Condensate System
~ Statement] Ability to (a) predict the impacts of the following on the Condensate System and (b) based on those predictions, use procedl to correct. control, or mitigate the consequences of those abnormal operation:
Loss of condensate pumps I
.Explanation.of; 55.41 (4, 10)The limit per AB.CN for 2 cond pump and 3 HOP's in service is 85'i'opower. When the second pump trips as in the stem iAriswers: .... "1 above, the power limitation is 30%. SGFP suction pressure will rapidly lower. SG NR levels will lower, but the SGFP will speed up ~""~.------~---".,...:
and the BF19s will open, which will tend to restore level but degrade SGFP suction pressure mote. The polisher is bypassed, then the HP heater strings are bypassed in an effort to restore suction pressure. A load reduction will be performed, but it will be to I
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L Reference Title I' II Facility Reference Number * :j [Referencfj Section **. **"lpageNo-=- !Revisiohl [ Main Feedwater/Condensate system Abnormalit II S2.0P-AB.CN-0001 II 11 10 IE I [ II II II II-I I II II r1 II_--.J L:O,*NlI.mber L: 1ABCN01 E004 1-----' iMaterial Required for Examination II Ii0""".ttnn.Sourpe:l INew I fQu~stio~.~odification Method: mml
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Q~~~i.~_n~T-o~p_k_!!_R...;;O;...5;...4_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _....JI Given the following conditions:
- Unit 1 is operating at 40% power during a power ascension.
- Both SGFPs are in service.
- 11BF19 fails shut with valve demand remaining normal.
- 11 BF19 cannot be opened from the control room.
Which of the following describes how 11 SG NR level will be affected, ant what action should the control room crew take? 111 SG NR level will The crew will I lower. Reduce power lAW OHA G-15 ADFCS TROUBLE to establish feed demand within the capacity of 11BF40. I Ib. 11remain within 5% of program. Reduce power lAW OHA G-15 ADFCS TROUBLE to ensure margin is available on operation of 11 BF40. I
-C.: !Iower. Trip the Rx lAW S1.0P-AB.CN-0001 Main Feedwater/Condensate System Abnormality based on inability to maintain 11 SG NR level ,.
c- .>16%. . 1 i 1remain within 5% of program. Lower 11 BF19 valve demand lAW S1.0P-AB.CN-0001, Main Feedwater/Condensate System Abnormality to 1 10% to ensure malfunction cannot reopen valve. I iAn~wer: [G [Exam l!,,~ ~ iCognitixe Lever I! Application I tF~ ISalem 1 & 2 I lExamDate: I! 9/26/20111
,~I059000A212 JrA.2.12 -Rovalue:I@J3;sR(jval~e:l~rs;cti()n:il~'ROGrouP:'U[sR6Group~iU _ _ [J
- System/Evolution
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Title IMaIO Feedwater System
'KA
- ... Statement:'
--------.-.~---~ ....
Ability to (a) predict the impacts of the following on the Main Feedwater System and (b) based on those predictions, use procedures to correct. control, or mitigate the consequences of those abnormal operation: Failure of feedwater regulating valves I iExplanation Of] 55.41(5,7,10) The valve trim for the BF19 and BF40 show that from 0-7% power, the BF40 opens from 0-70% while the BF19 I An~~ers; ___! remains shut. From 7-8% power, the BF19 opens from 0-1.8% with the BF40 opening fully. With the stem indicating 40% power, it is well beyond the capacity of the BF40 to maintain prog;ammed SG NR level in that loop. Insufficient time would be available to lower power to within the capacity of the BF40, and there is no direction to do so in either the ARP for program deviation of the I\.n.,rc: t,,,, ,hi" "I",,,,, I\. I" :-,"" ,dhl", ho,.." , "", H h"" th", debt effect bllt tbaca I" in", ,m"" ,nt tI""" ,,, :-",rf",,,,., ":-' >r ,,,rI, -'i, ,,.., ,,, 7_ 8% to prevent a Rx trip setpoint from being reached. B is incorrect but plausible if candidate thinks the BF40 has sufficient capabilty to maintain SG NR level at 40% steam demand. D is incorrect but plausible since removal of the demand signal to an already faulted valve is prudent, but is not directed to be perfored, as well as incorrect SG NR level trend .
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FacjlityReference NlIrnber. *.'i L~efeJ~nceSection
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iRevisi_()~ ., Main Feedwater/Condensate system Abnormalit S1.0P-AB.CN-0001 II 1112 II~--.J IControl Console 1CC2 I! S1.0P-ARZZ-0012 II rI 11 34 I i I , II II II II_--.J i
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~ue~~~!6pici l_R_O_5_5__________________________________________________________________________~ll Given the following conditions: i
- Unit 2 tripped from 100% power.
- Neither MDAFW pump started or could be started.
- Total AFW flow is 23E4lbm/hr.
- SG NR level on all SGs is 14% and rising slowly.
Which of the following describes the effect on 23 AFW pump when the PO lowers AFW flow to the SGs by throttling shut the 21-24AF11. AUX FEED - SfG LEVEL CONTROL VALVES? 23 AFW pump .... i.1 speed demand will lower to maintain stable discharge pressure. I I speed demand will raise to maintain stable discharge pressure. I Idischarge pressure will lower and remain lower. I Idischarge pressure will rise and remain higher. II i Aris';~r! ~ExamLeveli ~Co(]nitiVaLeveli IComprehension 1;F~salem 1 & 2 I !ExamDate:11 9/26/20111 iKA:!l061000K503 1~5.Q3 iRovaiue:]§~Rova'ue:I~[s~l~fRoGr()upI=]:s*ROGr~up~D Iii D SYStem/Evolution T~ Auxiliary I Emergency Feedwater System I i 1
*KA Stat~ment: I Knowiedge of the operational implications of the following concepts as they apply to the Auxiliary I Emergency Feedwater System:
_ " _ " _ _ _ _ _ _ _ _ _ _.. M~ Pump head effects when control valve is shut I
'EXPlilnat::J 55.41(5,4) 23 AFW pump Terry Turbine has its governor set to maintain a certain speed, not discharge pressure. As the AF11 ~!!"~: valves are throttled shut, the discharge pressure of the pump will rise, and remain at the new higher pressure as less wcrk is required of the turbine. A and B are incorrect because speed demand will remain constant, and discharge pressure will rise. Cis incorrect because discharge pressure will rise.
Reference Title Il.,FacjlityR~fElrenceNlirriberl ~ReMencE!Section ..*.* I'pageNo.! iRevision. IAuxiliary Feedwater System operation II S2.0P-SO.AF-0001 II 11 19 11 35 I lin Service Testing - 23 AFW Pump !IS2.0P-ST.AF-0003 II II 11 48 I IAFW System Lesson Plan II NOS04AFWOOOO II 11 38-39 11 9 1
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~L.O. Number IAFWOOOE004 I AFWOOOE008 1-----' 'I Material Required'for Examination II iQuesli,on ~oufce: '.. I INew 1§~~sti()'l,M§dification M~thO~d: .11 I QuestionSo.Jrce .ComrTlEmtS1
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IQuestion Topic: IRO 56 I Given the following conditions:
- Unit 2 is operating at 100% power. - 2C Vital 4KV Bus is aligned to 24SPT (breaker 24CSD closed). - Power is lost to 2C Vital 125 VDC Bus. - Prior to restoring power to the 2C DC Bus, 24 SPT is deenergized.
Which of the following describes the status of 2C 4KV Vital Bus for these conditions? I la. : Energized from the 2C EDG. I
,b.; IDeenergized with all in-feed breakers tripped.
I I
'c. i Energized from 23SPT (breaker 23CSD closed).
I
@] IDeenergized with in-feed breaker 24CSD closed.
I Answer 11 d I Exam Levell I R I iCognitive Level II Comprehension I iFacility: II Salem 1 & 2 1 ;ExamDate: II 9/26/2011 1 ~I 062000K103 iRQ Value: I@SRQValue: 1[~Sechon: i~ IRQ Group:,U,SROGroup:IU ~~r:l D
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- SystemJEvolution Title i IA C. Electrical Distribution ~ '062 ,
IKA Statement: r Knowledge of the physical connections and/or cause-effect relationships between A.C. Electrical Distribution and the following: I DC distribution I iExplanation of 55.41 (7) DC power is required to operate relays and contacts for the 4KV vital bus breakers. When DC power is lost, breakers will [Answers: ---I remain "as is". The EDG breaker can not close onto the bus even though it is deenergized because one of the interlocks to shut the EDG output breaker is both infeed breakers open. The cther (23) SPT cannot close its infeed breaker to the bus because it has an interlock which requires the other SPT's in feed breaker to be open.
~
l Reference Title Facility Reference Number I.Reference Section l! Page No ! iRevision I; '; t ..i 11C 4160 VAC Emergency Diesel Generator 11203036 II II I~I I II II II II-~ I II II II II I LO. Number 1 DCELECE013 IMaterial Required for Examination ]I II I IQuestion source:] Previous 2 NRC Exams 1 ~uestion iVlodification MethC?d: IEditorially Modified I ~d Du~ing Training Program 1 D iQuestionSource commentsll"J" ILOT RO NRC Exam - August 2008. Added "breaker" and "closed" to original distracter b, and swapped places with distracter c due to its being longer now. Added words to lessen confusion and make it the same style as found I
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[RO:S~$c:raPe~ I SROSkV~lri:p~f sl R()SV~te.h/E";olution Ust ,fl tisRds~~tnIEvoruUon L~t it <l',e.::lf.l~: iY~~~k1t~~ I ,. Question Topi';;; l.........:.. .............. II RO 57 I Which of the following choices identifies an adverse effect of a ground on a 125VDC bus/battery, and the method in which operators perform ground isolation lAW S2.0P-SO.125-0004, 125VDC Ground Detection? A ground. Ion the bus causes a higher level of current to flow in the system. l'1dividually deenergize each load on the bus, then re-energize if that load is not the source of the gr?~ Ion the battery associated with the bus causes a higher level of current to flow in the system. Transfer to the backup battery charger to determine if the liS charger is the cause of the ground. I Ion the bus causes voltage reading on the bus to become unreliable due to the excessivE' current flow. Transfer to the backup batt to determine if the I/S charger is the cause of the ground. Ion the battery associated with the bus causes voltage reading on the bus to become unreliable due to the excessive current flow. Individually i "'''c' \1- each load on the bus, then re-energize if that load is not the source of the ground. iAnswer I ~Exam Lever! !R ICogl"I!tive Le~ef II Memory I :Facility:11 Salem 1 & 2 I !ExamDat~ I 9/26/20111
,KA:II.A")('\f\f\ 'AI 11A2*01 RO Value:!1 2.51 [SROVatus: :!3.2*llSec::tic:m: :] SYS IRO Group:;] 11 'SRO Group:] I 11 iii~. []
<>. .IE:. IItinnTitle I D.C. Electrical Distribution 1 063 KA Statement: I Ability to (a) predict the impacts of the following on the D.C. Electrical Distribution and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Grounds IExplanationof i 55.41(7) The ground detection procedure has operators isolate individual loads. The ARP for low battery voltage has operators [A':'~wers:/ ....... transfer to the standby battery charger if bus voltage is low, and battery current is present, so these are plausible distracters. The bus voltage is higher than the battery voltage, so a ground on the battery would not cause bus current to rise.
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IRefer~nce Sectfo,n
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Reference Title Ii Facility Reference Number '., *. l:'a96 No. !Revision! I125VDC Ground Detection II S2.0P-SO.125-0004 II II 11 13 I I B Window Alarm Response II S2.0P-AR.ZZ-0002 II II II~~ I II il 11 IC~ Material Required for Examination 1 I
~~~~~~~~ IPrevious 2 NRC Exams IQu~~~~nModification Met~9d;1 Editorially Modified I!Used DU'rlllgTraining ProgiarriJ New for "J" ILOT RO Exam - August 2008 Reordered distracters based on length of answer.
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- Question Topic I Given the following conditions:
- Salem Unit 1 is in MODE 3, NOT, NOP. - A 500 KV grid disturbance causes a SEC MODE II actuation. - During the electrical bus realignment. the 2A 4KV vital bus to 460VAC bus breaker trips.
Which of the following describes how this will affect 2A EDG operation, if NO corrective action is taken? 2A EDGwill ...
- a. ;
b.' c.' trip because its lube oil supply pressure was lost when the EDG Lube Oil pumps lost power. d.. run until its Fuel Oil Day Tank empties due to the loss of power to the Diesel Fuel Oil Transfer Pumps. Answeri Ia IIExam Level!
..~j I064000K202
- SystemiEvolution Title i IEmergency Diesel Generators IKA Statement: ' Knowledge of bus power supplies to the following: I Fuel oil pumps I Explanation of 55.41(7,8) A is correct because there are 2 Diesel Fuel Oil Transfer pumps, powered from A and 8 Vital 230V, each of which Answers: j supplies fuel oil to all three EDGs. The loss of power to :2A DFO xfer pump, even if selected as the Lead Pump, will not affect EDG operation as the second pump has power from 2B bus, and will start on 10 level. The DFOSTs have cross connect capability, but the cross connect is always shut. B is incorrect because the EDG Fuel Oil Pumps are shaft mounted. mechanically driven pumps.
(' ;" ;n"""",,"'. h",,,,,* ",'" ,,hn,,, thoro ;" 'On ",lo",lr;", Pro. "k, > OllelC l41bicb is io ocpratioo lAlbeo tbe EDG is sbttfdOl41O T" ;"",III<>to ;::nr-I starting, the lube Oil pumps for operation are shaft driven' mechanical pumps. D is incorrect because there is still power to one DFO xfer pump, which will fill all the EDG Da~ tanks based on level signals .
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. Reference Title i ** ' "FaciliiYReference Number" 11 & 2 Units Diesel Engine Auxiliaries 11205241-1 II II 1142 I 111 & 2 Units Diesel Generator Fuel Oil 11211306 II II 1111 I
!IFueiOil q211307 II 11 1122 1 L.O. Number I EDGOOOE013 I EDGOOOE005 Material Required for Examination II II Question Source: i INew , IQuestion Modificatio",Method: I IiUsed{)uring Training Pr.?~I":':~ 0 I IQuestionSource comments! I r., ,;; .
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[~ues!~~f~el£J IRO 59 I Given the following conditions:
- Unit 2 is operating at 100% power when OHA C-1 GAS ANLY TRBL is received in the control room. - The NEO sent to investigate reports local alarm B-3 OXYGEN HIGH/LOW on Waste Disposal Gas Analyzer PNL 110 is in alarm. - Local indication for in service Waste Gas Decay Tank (WGDT) 02 concentration is 4.1 %.
lAW Tech Specs, which choice describes a required action, and wh)::?
~ iPlace the Standby GOT in service and commence preparations to release the affected GOT.
I
~ IPlace the second Waste Gas Compressor in service to raise the total volume of gas in the WGDT in order to dilute the 02 concentration to below 2%. I
~ IReduce the oxygen concentration of the in service WGDT without delay to prevent potential releases of radioactive materials due to explosion I of the GOT.
[d:.ymmediateI Y suspend all additions to the in service WGDT since the 02 concentration is above the 2% required to sustain combustion when I _. mixed with hydrogen. iA"swer II c I Exam Level; I R I ICognitive Level II Memory I [FaCilit¥: !ISalem 1 & 2 I ExamDate:11 9/26/20111 l~.11 071 000K504 I.K5.0~ __ ~ :iRa Value.*11 ~Ra Value. [2:.!J ~~~ IRa Group. U U ~6,43 D c
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~==========: ~~~~~!~~C:>~':I!~O~_!!!~~J IWaste Gas Disposal System lKAStatement: I Knowledge of the operational implications of the followin~ conceets as the):: apel):: to the Waste Gas Disposal S)::stem: I Relationship of hydrogen/oxygen concentrations to flammabilit):: I I Explanation of i 55.41 (13) Tech Spec 3.11.2.5 requires the 02 concentration in the waste gas holdup system to be <2%. The action for 2-4% 02 is ,Answers: to reduce the 02 concentration to <2% in 48 hours. Distracter 0 is incorrect because while the immediate suspension of additions to the waste gas system is REQUIRED when >4% 02, the flammability percentage is wrong. nictr",,,,ter 11 ie in"",rre",t he",,,,, ,ee the fl",rnrn",hle "",n",en"r",tinn "n In he ren, or*en hefnre ",,,,tinne ",re bven tn nre.,,,,re the bnv fnr release. C is the correct answer because the Tech Spec REQUIRES the reduction of 02 from >4% to less than 2% without delay. Also, the bases section for this tech specs describes that a potential explosion and release of radioactive materials from this explosion would not be lAW GDC 60, 10CFR50 Appx.A. Distracter 0 is incorrect because the TS states to immediatel):: stoe additions to the WG HU system. 1_ .. -_. Reference Title ._.m . _-- I Facility Reference Number ] ~ference Secti()n _. i l~-"g.e .~()J ~e.v.is.~<>'1'!1 ISalem Tech Specs
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I.WhiCh of the following Area Radiation Monitors (ARM) will cause a ventilation system alignment change when it reaches its High Radiation Alarm
~ ~ ~
, ~ ~ ~ ~
I setpoint? I 12R52, Liquid PASS Room. t
.j2R9, N, Fuel Storage.
I
~J 2R32 Fuel Handling Crane. I 1 A, I
12R44A, Containment High Range. I
'Answer ~ I b I :Exam Levell [ R I ,Cognitive Level II Memory I facl!ity;11 Salem 1 & 2 I !ExamDatell 9/26/2011j I
IKA:' 072000K403 I,K4.03 .Rovalue:113.2*1.s~() Value: I~i.6*1 ~cti()l'l! I~ (ROGroup~ I 211SROGroup: U Btl D ISystemJEvo!utionJitle: IArea Radiation Monitoring System
- KA-
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Statement:
...... -~-
Knowiedge of ARM system design feature(s) and or interlock(s) which provide for the following: Plant ventilation systems I I, :Explanationofl
'Answers:
55.41(11) B is correct because it realigns FHB ventilation through the charcoal filters and starts both FHB Exhaust fans. Cis
~~~~--'- ....~
! incorrect but plausible because its auto function is to prevent Fuel Crane motion except in downward direction. A is incorrect since it only has alarm light outside the PASS room which activates, but plausible because of the high radiation levels which would be expected in that area of the aux building following an accident. D is incorrect since it has no automatic function, but is plausible ci""o. "'no, n;,..,n . . ",1<",,.,,, ",,,,,,,,..;,,,,,,,-1 "Hn' n; .<>rI 1.,,,,,1,, i" """binmo"! "ortMm '" "",.,.,<>+,,, .,,,,,,,,,,, It'> 1",,1.,10 "ani;"";",,
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1~~~StiOnToPitl IRO 61 Given the following conditions: I
- Both Salem Units are operating at 100% power.
- 2R1B-1, Unit 2 Intake Duct Rad Monitor Channel 1 loses power.
- BOTH Salem units Control Area Ventilation (CAV) systems remain in NORMAL Mode.
Which of the following identifies the status of the CAV systems, and how will the Control Rooms respond to the loss of power? i The CAV systems ... ia.J
~
I are designed to remain in Normal Mode upon a loss of power to a s.ingle duct radiation monitor. No actions other than normal Corrective Action Program actions to troubleshoot and repair the power supply are required, with no specific time limitation. I b~i tare designed to remain in Normal Mode upon a loss of power to a single duct radiation monitor. The 2R1B-1 must be restored to operable
- - status within its allowable Action Time, or CAV must be placed in Accident Pressurized (AP) Mode on BOTH Units.
I should have swapped to AP Mode. Initiate AP Mode of operation by depressing initiating pushbutton on 2RP2 lAW S2.0P-SO.CAV-0001, Control Area Ventilation Operation. I should have swapped to AP Mode. Initiate AP Mode of operation by aligning individual system components to their correct positions on 2RP2 I I lAW S2.0P-SO.CAV-0001.
~swer ~ lEx~mLevel ~ ~ I [Cognitive L~velj IApplication 1 IFacility:11 Salem 1 & 2 1 iExamDate:) ! 9/26/20 11 1
,~ll 073000A201 11A2.01 m l iRO vallJe;:l0'.SROvalti~~ Isecti()n:.ll~ IRD. Gro~p:DlsROGroITp~JU _ _ [J
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-----~---- ..----.. -~-
Ability to (a) predict the impacts of the following on the Process Radiation Monitoring System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Erratic or failed power supply
~ !Explanation of ! 55.41 (7) Any R1 B channel losing power will automatically initiate Accident Pressurized Mode on BOTH Units. B 's incorrect but !Answers: '~-- ...
I plausible because the Tech Spec for R1 B says if one channel is inoperable, you have 14 days to restore before placing CAV in AP Mode. C is correct because manual initiation of AP Mode is accomplished by depressing the Accident pushbutton on RP2. D is incorrect because individual components are not aligned, still plausible if the candidate thinks the failure has affected the whole
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Revisio.".l: I Control Area Ventilation Operation II S2.0P-SO.CAV-0001 II 11 6,23 11 38 1 ,I Salem Tech Specs II 11 3.3 .3 . 1 113/43-38 11282_1 II II II II II I I
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li:l......tit)nToPic:!R()62 I Given the following conditions:
. Unit 1 is operating at 25% power.
. 18 EDG is running in parallel with station power on 1B 4KV Vital Bus .
13 and 16 SW pumps are in service. 11 SW pump is in AUTO. 1A 4KV Vital bus becomes deenergized due to a Bus Differential signa:. 1 minute after the 1A 4KV Vital bus deenersizes, with NO oeerator action, which of the following contains ALL the SW pum ps which will be funning? 111 13. 111 15. II 113 ,16. I 115,16. t IAnswer ! a I I ~Exam tevelj IR I Cognitive level j IApplication I IFaciifiY:l1 Salem 1 & 2 I Exa~Date;~ I 9/26/2011J
~: t I076000K201
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- 55.41 (7) A bus powers 15, and 16 SW pumps. On a single bus UV as described in the stem, only that bus would load in blackout Answers: loading. A bus is locked out on Bus Differential (deenergized), and the loss of 16 SW pump would cause header pressure to lower to where the auto pump (11) would start. Only one SW pump is aligned for AUTO which is the normal at power configuration for the I;;:1 ft~~~!; ~~t~!ni:~~; ~~~b~~:!~}~~ ~:~~~;-o,,;o~~' pump would never start unless 11 pump did not on. a S~c: i~iti~tion, that is I any of the choices. Th~re can be confusion abo'ut the running EDG and the loss of A vital bus' causing a MODE II (Blackout), which I would strip busses and load the primary SW eume on each bus.
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': :QuestionSourcec0rt.:'~ IJ*ROC61 Changed which pump is in auto (11 instead of 15) and that makes different choice correct. I
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List "',;~" " TI :Question!C>,pic II RO 63 j Given the following conditions:
- All 3 Station Air Compressors have become unavailable.
- The NORMAL cooling water supply to the Unit 1 Emergency Control Air Compressor (ECAC) has been lost.
Describe the status of the Unit 1 ECAC. 0eeration of the Unit 1 ECAC ... I can continue since cooling water will automatically swap to Service Water through
- check valve.
1can continue since cooling water will automatically swap to Demine,ralized Water througr a check valve. I I must be disccntinued until cooling water can be manually aligned through a spool piece from Service Water. I must be discontinued until cooling water can be manually aligned through a spool piece from Demineralized Water. I IAnsWer~ Ic 1:Exam Levell IR j 'cognitiveLevelll Applicatior~ iFacility:11 Salem 1 & 2 1[Ex3;Oate:11 9/26/20111
!KA: 1078000K104 I~ I 'RO Value.: II 2.6jSRO vah,ie:i [~ !S~~t!on: iI~ lROGrOUP: I mmm 11 iSRO Group: D" 1~!~J 0 L~~.~.§.~~~on Title I IInstrument Air System
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I Coolin water to ccmpressor
!Explanation of, 55.41 (4)The normal source of cooling water to the ECAC is the Chilled Water system. Upon a loss or unavailability of the Chilled Answers: *...* J Water system, SERVICE WATER can be supplied by Mi',NUALL Y installing a supply and a return spool piece. Demin water is
~-- plausible because it is as a backup cooling system for other systems (SI and charging pumps when normal cooling is lost.)
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1 n.".~~;~~ Topic i IR064 Given the following conditions:
- Salem Unit 1 is operating at 100% power.
1- OHA A-i5 FIRE PUMP 1/2 RUN, and OHA A-23 FIRE PUMP 1/2 TRBL alarm in the control room.
- NO other fire system alarms are received.
- An NEO dispatched to the Fire Pump House reports fire main header pressure is 132 psig, and both fire pumps are operating.
- Fire Protection reports there are NO fire system actuations.
Which of the followin~ choices identifies the cause of the condition described above? 1Trip ofthe Fire 'V''V' Jockey Pump. I ILoss of control power to BOTH Fire Pumps. I I IA fire protection hose reel is partially opened in the Turbine Building. I I IMomentary (1 second) drop in Fire Protection header pressure to 70 psig. I I;Ans""{~r! ~ Exam Level' ~~(),gnitiveLevel I i Application 1 I
,FaCility:: Salem 1 & 2 1 \ExamDate: II 9/26/20111 lKA:! I086000A302 I:A3.0Z- *iR2Y~~~ :SRO Vallie:! @ ISectio~: II~ ~ Groitij [JjISRO GrQup: I2T ~~ 0 SystetnJEvolution Title IFire Protection System "
iKA .1 Ability to monitor automatic operations of the Fire n "'L1VI System including I Actuation of the FPS I Explanation ofl 55.41 (7)Fire pumps will BOTH auto start upon a loss of power to their battery chargers. Losing the power from the ATS, supplied Answers:
- from #1 and #2 Misc yard panels will cause loss of both battery chargers. Distracter D is incorrect because a momentary drop in pressure will start the #1 pump, and the #2 pump has a time delay. Distracter A is incorrect in that the #1 pump will start as header pressure lowers, and the #2 pump will not start Distracter C is wrong because a hosereel is a small system load, it is only partial;ly I .
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! Fire Pump House Diesel Engine Control 11203776 II II I i 9 i
1 i Alarm Response Window A II S 1.0P-AR.ZZ-0001 II II I I II Fire Protection 11205222-4 II 11 11 59 1 iLO. Number}] I FIRPROE007 I I FIRPROE008 I IFIRPROE009 I
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I IRO 65 O",...tirmTopic I Given the following conditions:
- Unit 1 is operating at 100% power. - Charging flow rose 6 gpm from its steady state value of 88 gpm in 10 minutes, and is now steady at 94 gpm. - Computer trends show the increased RCS leakage started 10 minutes ago. - The CRS enters S1.0P-AB.RC-0001, Reactor Coolant System Leak. - Based on elevated 1R11 A indications coincident with the rise in charging flow, the crew determines a small RCS leak in containment is occurring.
Which of the following describes an action that will be performed lAW S1.0P-AB.RC-0001 based on the determination of the leak location and size, and wh:e I Place a centrifugal charging pump in service to maintain PZR level stable. I I Place 2 CFCUs in slow speed and 2 CFCUs in high speed to prevent containment pressure from rising to the automatic Safety Injection setpoint.
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~
Place 2 CFCUs in slow speed and 2 CFCUs in high speed to minimize the rise in containment humidity and prevent equipment damage from moisture levels. I "d~ Place a centrifugal charging pump in service to allow additional flow through the Mixed Bed Demineralizers to minimize the potential off-site release from containment.
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Containment pressure, temperature, and humidity I Explanatiorlof i 55.41(9) AB.RC-1 Attachment 1, Continuous Action Summary, 4.0, states that at any time if leak is suspected to be in containment,
*Answers:
., then place 2 CFCUs in slow and 2 in fast. The Technica: Bases Document, page 3, states that this is "to prevent an automatic Safety injection and minimize the potential for off-site releases when the leak is in Containment." C is incorrect because while it has the right action, the reason is wrong. Distracter A is incorrect but plausible. because a centrifugal charging pump is placed in I """*",hl,, f,,', "'"
I""",1,..", if P70 L",,,01 ".,nn,,+ ho ,.,.,.,in"';"""' "".,hl", r.r n"in" ....'" "'.. ., "ro..",,.,..,,+'" .,,,1,,,,... incorrect because of the wrong reason for CFCU operati"O"n. D is incorrect as per B above: while additional flow through the demins would reduce RCS activity levels, it is performed in AB.RC-2, High RCS Activity. f", lo.,~ r"l, . '0<0 ( ' ,<0
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Reference Title ;i: Facmty Reference Number . *1 :R~fe.renceSection ] [~Clge No] Revision: IReactor Coolant System leak
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II S1.0P-AB.RC-0001 II 1113,BASE119 I I !I~I=================I~FI============;I~I======~IIF=~IC=
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[Questi(ln Top}c:j IR066 I .* IS2.0P-IO,ZZ-0007, Of the following, which would be considered the FIRST Core Alteration lAW Cold Shutdown to Refueling? I
!When the first stud (during first pass of Reactor Head U<::L"I I",IV' '" I\'I process) is detensioned.
I I
!>~ When the RPV Head is lifted (1-2') to check for CRDM drive dib-v' '~"'~"" 1':"
I 1.lnsertion of a camera to the level of the RPV flange prior to fuel movement. I
~,~~J1Lifting of the firstfuelbundle in the RPV. ' .. ~. . .... . ." ......... " . '. "
I I Answer! Id I Iexam tevel'[ R I [Cognitive Lever II Memory ~ iFacilitY:ll Salem 1 & 2 1 rEXamDate:! I 9/26/20111 KAll194001G136 ! ~2:1.36 LRovalu~e.:JI 3.01 ,sRbvalu~j4.1 1Section: I PWG 1ROGroup:!1 11 ~OGrollP:J 11 8111 c System/EvolutlonTitle I I______________________________________...J ,~'c"cr"1 iKASt~~~rIK-n-o-~-e-d-g-e-O-f-p-ro-c-e-du-r-e-s-a-nd-lim-ita-t-io-n-s-in-v-o-Iv-e-d-in-w-r-El-a-Ite-r-a-tio-n-s----------------------~ IExPlanationon 55.41(10,13) D is correct because "CORE ALTERATIOI\I shall be the movement of any fuel, sources, or reactivity wntroi I
~s~~:~ components, within the reactor vessel with the vessel head removed and fuel in the vesseL" A is inwrrect because it is when MODE 6, Refueling, is officially entered. B is incorrect because Core Alts cannot occur until the vessel head has been removed.
I C is incorrect because "Refueling activities that would constitute a CORE ALTERATION do NOT include the mOV'3ment or rn'On;n"bti"" ,.,f no,{i"o<> ",,,,,,h '0'" -I;"ht<> "0",1 n" sac£ic:icc aCllicwac! Iisec tQ axawiCJa eaaclQc ecessll[a Icomponents, except when these d-;'vices would result in the movement or manipulation of fuel, sources. or reactivity control
\/,,<>,,01 com~onents within the Reactor Pressure Vessel. , ~ ..
Reference Title I r-"FacilityRefeienceN~mberj ~Reference SecUo.n . *... '1 'Page No*1 ~eviSio~ IS2.0P-SO.SF-0009
,I II Refueling Operations II II II~.-J II S2.0P-IO.ZZ-0007 II COLD SHUTDOWN TO REFUELI II !I 1117 J II SC,MD-FR.FH-0004 11 PREPARATION FOR R.EACTOR II !I II~.-J LO. Number IREFUELE012 i IOP007E004 Material Required for examinati0l} iI II
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Question Topic! IR067 I Given the following conditions:
- Unit 2 is operating at 66% power.
- Power was reduced when 22 SGFP tripped 2 days ago, and has remained at this level for 2 days.
- Operators are preparing to start a Main Turbine power ascension using dilution and automatic rod control to maintain Tavg and AFD on program.
Tavg-Tref deviation is OaF. lAW OP-AA-300, Reactivity Management, how should the power ascension be started? Assume this is a normal !20wer ascension with all ~uiement available.
§J jThe crew concurrently initiates the Main Turbine load ascension and a ReS dilution. 1 I
~ iThe crew initiates a ReS dilution. As soon as the dilution is in progress the Main Turbine load ascension is initiated.
I ie. "IThe crew initiates a ReS dilution and waits until a ReS temperature rise is detected. Then the Main Turbine load ascension is initiated. I il.=:::ew initiates the Main Turbine load ascension and waits until a ReS temperature lowering is detected. Then the Res dilution is initiated . I
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! Explanation of.: 55.41 (5,10) Listed under the responsibilities of the Reactor Operator, page 8, 4.6.5, "Typically. during planned load changes where Answers:'" dilution or boration is required. start with the dilution or boration. The initial effects (ReS temperature change) of the reactivity change should be seen prior to initiating the load change." All the distracters have the correct actions in the wrong order.
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Ques;tion Topic i [RO 68 t Given the following conditions:
- On your shift today, a component's normal monthly surveillance item is determined to have been scheduled incorrectly, and it has been 34 days since it was last performed, Which of the following statements describes the status of the affected component?
IThe component must be declared INOPERABLE at the time of discovery because the 24 hour time limit allowed past the normal surveillance interval has been exceeded. i IThe component remains OPERABLE because Technical Specifications allow a 25% time extension of the normal surveillance interval for surveillance completion, which has not been exceeded. IThe component remains OPERABLE ONLY for the next 24 hours after discovery, during which a SAT surveillance must be performed to ensure OPERABILITY, otherwise the component must be declaree INOPERABLE. I IThe component must be declared INOPERABLE at the time of discovery ONLY if any redundant structure, system, or component (SSC) is also INOPERABLE for the system in which the affected component is required to be OPERABLE,
~11~\Ner i ! b liExam Levell! R ! ;vogrmm, Level IjApplication I iFa<:ili~rll Salem 1 & 2 I iExamDate~ II 9/26/2011 KA:!194001G212 11 2.2 .12 Rovalue:113.7jSRovalue:}'.:.:!Jsection:ilpwG I ROGrouP~1 11 [SFioGr()up:I 11.' [J System/Evolution Title ,
IKASbrt~~r- ______~____~______~________________________________________________________________~ Kno'NIed e of surveillance procedures,
- Explanation of 55.41 (10) Tech Spec 4,0.2 states that... " Each Surveillance Requirement shall be performed within the specified surveillance Answers
- ,~ interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance intervaL" Tech Spec 4.0.3 states that. .. " If it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hours or S*urveillance. If the S~rveillance is not performed within the delay period, the L.imiti~g Condition for Operation must immediately be declared not met and the applicable Actions must be entered." A is incorrect because the 25% "Grace Period" (which is 7.75 days for a 31 day monthly surveillance) has not been exceeded, C is incorrect (and different from Distracter A) because the 24 hour time limit is incorrect AND it says it would be applied from time of discovery, not added to the normal surveillance interval. D is incorrect because there is no requirement to check other SSC of that system with regards to how it affects the com ponent which has exceeded its monthly surveillance interval.
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~lle~ti0l'!~!~ic:J RO 69 I Given the following conditions on Unit 2:
- Reactor power is 75%.
- A RCS leak rate surveillance indicates the following:
- Total Corrected Volume leak rate is 4.0 gpm.
- Leakage to PRT is 2.0 gpm.
- Leakage to the Reactor Coolant Drain Tank is 1.3 gpm.
- Total primary to secondary leakage is 0.1 gpm.
Which one, if any, of the following Technical Specification leakage limits has been exceeded? I la.~! None. !
~J jldentified.
Ii I Unidentified. IPrimary-to-Secondary. iAnswe'll c I tExam level :\ R I ~C09nitiVEllevef I [ Application ISalem 1 &2 I !ExamDate:. I [SyshmllEvolution
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Knowledge of limiting conditions for operations and safety limits. [ 1~'~1 iExPlanation of 55.41(5,10) Salem TSAS 3.4.7.2 states the limits on ReS leakage to be 1 gpm for unidentified leakage, 10 gpm for identified iAnswers:*
' - '- '- '- ...... --""~ .....
.,i leakage, and 150 gallons per day through anyone steam generator. The 0.1 gpm pri-to-sec leakage is 144 gpd. PRT leakage is
-~---~--,!
identified, so while it is >1 gpm, its limit is 10 gpm. RCDT leakage is unidentified, so since it is >1 gpm, its limit is exceeded. I .-,~._.... - - 7 .-,- - .~~~'-l
.~
Reference Title '~-;C--.* -*. [Faciiiti:.~efere;,ce Number' :RefElrElnce Section J' PageNo.!Revisionj 1Salem Tech Specs Ii 11 3 .4.7.2 113/44-17 11 282 I I II Ij II Ii-I II II II II II-~ I TECHSPE015 I _ _ _ _ _ _......J M,'lterial RequiredforExamina'tion' : I II iQu~sti()ns()~ IFacility Exam Bank I~~stion~od'ifidatiolllVlrthOd: 1Direct From Source IlUse~ DuringTraini~g~rogram ; [] IQuestion~~~rceC()mril~Il~~ ISalem 2002 RO NRC Exam (4 NRC Exams ago) I I I [ II
- Ouestion Topici IRO 70 I Which of the following Unit 2 conditions will require entry into an ACTIVE Tech Spec LCO Action?
With Unit 2 in ... I IMODE 1, 26 SW pump is CIT, ALL other SW pumps remain OPERABLE. I IMODE 5, performing a shutdown, 2N32 Source Range NI control power supply fuses blow. I IMODE 3, 21AF21 AUX FEED - S/G LEVEL CONTROL VALVE is discovered in the jacked shut position.
@JIMODE4, 2PR3 PZR Code Safety Valve is declared INOPERABLE with 2PR4 and 2PR5, PZR Code Safety Valves OPERABLE.
I i jAnswerl1 c I ~E~am L~,:,eli IR I tc;0!:J['\itiveLevel ! IApplication~ [Facility: II Salem 1 & 2 I'.~A~... '-"nc. iI 9/26/20111 iKA:II194001 G242 1 2.2 .42 IRO Yal~ell 3.91 iSRO V~I~14.61 isecfion~JI PWG IIRO Group: I 1llsRO~roUP:lI 11 ~!:Il 0 rSystemlEvolu.ti.<>~ Title.! I ! !GENERI \
"KA Statement;.. ~ ,
IAbility to recognize indications for system operating parameters which are entry-level conditions for technical specifications. I jlExPlanation OflI55.41(10) In MODE 4, only ONE PZR Code Safety is required to be OPERABLE. 3.4.2 In MODE 1, 2 independent SW loops [Answers: I are required. An OPERABLE SW loop consists of one pump from each vital bus plus 2 pumps per bay. 3.7.4. SO.SW-0005, Att 2 I I page 1 of 36. In MODE 3,4,5 only ONE SR is required. 3.3.1.1 Table 3.3-1.6.b. ALL 3 AFW pumps and flow paths are required ! to be OPERABLE in MODES 1-3. 3.7.1.2
~ ...... -.. ~~.
Reference Title .'
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, . Facility Reference Nu~ber i [Reference Section I LpageN~Revi!i!O~
ISalem Tech Specs II II II II I 1Service Water System Operation II S2.0P-SO.SW-0005 II II 11 40 I I II II II II I
; : -------- ... ::.-.~
L~:O. Number ...! [ TECHSPE015 1_ _ _---1 I[Material Regoired for Exal1'lination .**1 I
'~~~~:JI~~E~m:m=B:a:~==I~IQ~.u=e=s=~=o=~~~~J=J=W=k~y=fi~o=n=M=e=t~.~~~=:F:r:o:m:s:o:u:~:e=~~=~~~==~~~=~~I IQuestion Souffe COlllments:
RO$IIYScra~( I . SRO SkVse£aoet)J !"T~ {v~MlEv~!@qrtli~i\I;~~RO Sv~U;@E;qlution~ist *6utli~~"ChahQ~ I 1 OIt,,~ti~~T;pi~ II R071 . ... . . ..... . .......................... ... m Given the following conditions:
- Units 1 and 2 are at 100% power. - Unit 2 has experienced several fuel pin failures. - A leak must be repaired on a pipe in the Aux. Bldg. pipe tunnel. - The general area dose rate in the location of the repair is 600 mrem/hr. - In order to reach the location of the repair the worker must transit through a 6 Rem/hr high radiation area for 2 minutes and return via the same path. - The worker currently has an accumulated annual dose of 400 mrem.
Calculate the MAXIMUM allowable time that the worker can participate in the repairs AT THE JOB SITE and NOT exceed the Current Annual TEDE Administrative Dose Control Level at PSEG Nuclear LLC.
~ 170 minutes.
1120 minutes.
/140 minutes.
I \160 minutes. ! Cl'-"lswcr* fEJr:,!,,!?~~.?v;;;i; r:s=J :ee@~!t!'!e I e'!c! m; 1.'\;?;?ii'22tl0A : ifaCilliy:}1 Salem 1 & 2 HExa§:~te; H____9_/2_6_/2_0_1--'11
~1194001G304 ~~~~fJill!5eCtiQn: 11~~~u~~JU
______________________________________________________________~I~ENERI Knowledge of radiation exposure limits under normal or emergency conditions. I Explanation bf: 55.41(12) B is correct because: The Current Annual TEDE Administrative Dose Control Level is 2000 mrem for PSEG Nuclear Answers: . i LLC. Transient exposure is 400 mrem (6000mrem/hr x 4/60hr). (transit to and from the job). (Current) 400 mrem (transit) 400 mrem = 800 mrem. ADL of 2000 mrem - 800 mrem 1200 mrem allowable before reaching the Annual TEDE Administrative Dose Control Level. 1200 mrem 1600 mrem/hr '" 2 hours. A is incorrect, based on using limit of 1500 versus correct ADL (2000): Cis In""""", _ Q.,,,. ,,4 r." ,..., I.,+;n,., ,In,.,., "n,,_ ' 1,<0 ,,,11.-1, . n ;e . Q',e, ,,,;,.,, ,,,;n,... ,h" foonn ,., TCnC A . n",*" IControl Level (2000) and NO transit dose. I Reference Title .', Facility ReferenceNumber 'i [Refere~~I~~ tp;g;NO] 'i'"'Revlsciion i Exposure Control and Authorization II RP-AA-203 II rI I5 I , II Ii 111==:::::;'1 F===::::;,
------------'11---------'11--------.111------' 1_----'1 !Mat~!iat RequiredforExamirlati0rt~J I r Facility Exam Bank I [Question MOdifi9ation:Meth~d: iDirect From Source ~~~~
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Question Topic] !RO 72 I Given the following conditions:
- Unit 1 is in Mode 2, with all Shutdown Banks withdrawn.
- Unit 2 is at 100% power i - All Unit 1 & 2 Circulating Pumps are in service
- 21,23 & 26 Service Water Pumps are running ,
- 21 CVCS Monitor Tank is being released via 22 CCW Hx to the Circulating Water System Choose the condition that would require termination of the liguid release. [
121 CCW Pum trips. 1,11A & 11 B Circulators trip, I t:: I,unit 1 Rx trips inadvertently during an instrument calibration,
~ ~L3 service WaterPump headerpressure drops from 115 to 105 psig.
I ! :~ Ib 1 Ex,am ,Level I :~Cognitive Level! Application~ I I
;Facility:i Salem 1 & 2 I ,ExamDate:! I 9/26/20111 iKA:ll194001G311 I~2.3.11 IRO Vlillue: d3.8!SROValue:il 'WiSection:l/ PWG IROGroup:!1 1J 'SRO ~rouP:i/ 11 ~!rll 0 I' . ~tem/Evolution Title I I 1-___________________________________- 1 l.~ ~_,_ ___
i~-St~~~~tJr_--------------------------_-------------- Ability to control radiation releases. ____~ Explanation of 1 55.41(13) S2.0P-SO.WL-0001 states at step 5,5,9 that the released is to be terminated upon loss of dilution water flow, 22 SW Answers: ! header discharges into the outlet of 11A and 118 circulators. A Unit 1 trip does not constitute an "emergency", so the requirements
~---.-~.~ of OP-SA-106-101-2001 Operating with an Emergency on the Opposite Unit are not applicable, Normal SW header pressure is 105 125, the SW flow from the remaining 2 pumps in service will maintain flow to the CCHX. Trip of an operating CCW pump would i ro", ,It in :on ""tn """rt nf th", "brvlh . nl,mn <Inri th", (,(,11\1 """tAm rl>m",in<: 1M" .
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~9u;~~nSo~rcElC(:;mmenl~ I.VISION 0119138 I
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r Jf [9uestion !OPIC r IR073 1 Given the following conditions:
- Unit 2 has experienced a steam line break inside containment
- Operators have entered 2-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock.
Why will the operators be instructed to terminate SI and start a RCP if po:>sible? , ? IThe soak required by FRT8-1 requires 81 to be secured and a RCP to be running to provide ability to use spray to depressurize primary, IThe soak required by FRTS-1 requires SI to be secured. A RCP should be started to equalize boron concentration throughout the primary to ensure proper shutdown margin as the RCS cools. ISafety Injection flow is a significant contributor to any cold leg temperature decrease or overpressure condition and must be terminated. A RCP is started to minimize temperature gradient across 8/G tube sheets. I ) IRCP Safety Injection flow is a significant contributor to any cold leg temperature decrease or overpressure condition and must be terminated. A is started to provide mixing of cold 81 and warm reactor coolant water. iA. ..... ~. i I d I ii:2'am Levell) R I Cognitive Level, IMemory I !Facility: I I Salem 1 & 2 I ExamDate: I r 9/26/20111 KA::i194001 G418 12:~,18 IR~value:iI3.3ISROValue::I~.:2J,sectiOri:lpWG JROGroup:1 1IiSROGrOuP~1 11 ~~r.~ [J r~,~~~.~""-'-.~'-'~-.
.. I ISYS~/~n Title j j GENERI ~St!~~nt:;r- ____________________ ~~ __________________________________________________________-,
Knowledqe of the specific bases for EOPs. I Exp,ana.tionofl 55.41(10) A is incorrect because the purpose for RCP operation is not the priority in FRTS-1, and the soak is not the basis for
,Answers: ..... . terminating SI. B is incorrect because the soak is not \t1e basis for terminating 81. C is incorrect because the SI basis correct but the basis for RCP start is not correct. D is correct because Basis Document states for step 9 starting a RCP ... "in order to mix the cold incoming ECCS water and the warm reactor coolant water, and therefore decrease the likelihood of a PTS condition, an RCP CT..:srt i~ II firlrliti,...,n~Ih.' c:.;1 !e torrnin!:.ttir.n Ie ";ocired jf inventaDt and Sl !brQoliao are adee! late heca! '5e !1~("'f""<:: fit -'.0:11
\ have contrib~ted to the RCS cooldown or mat prevent a subsequent reduction in RCS -pressure.';
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..:, Topic' IRO 74
~, I I
1 Given the following conditions: Unit 2 is operating at 100% OHA A-7, FIRE PROT annunciates. Panel 2RP5 is checked and indicates the following:
- Zone 59 Air and Water Deluge, Containment EI. 100 Panel 335 is lit.
- Zone 74 - Smoke and Fire Detector, Containment EI. 100 Panel 335 is lit.
Which of the following describes how the Control Room crew should respond? Iassociated
'a.
Verify OHA A-i5, FIRE PUMP 1/2 RUN is in alarm signifying a Diesel Fire Pump has auto started to supply Fire Protection water to deluge valves in containment. I
'I lb. Dispatch a NEO to place both PZR PORV breaker keyswltches In I::MER CLOSE, I
IDispatch a NEO to open the associated deluge valves. I id.llopen the 2FP147 from the control room. I [Ans~er. ~E.'<arillevei ! ~ [<:()~l'litiveLeveil !Memory 1 lF~ciljty: II Salem 1 & 2 I [exam Date: i I 9/26/20111
§I194001G427 12.4.27 .Rovalue:!~I~~ovallle;l[0secrioh:il~ROGrOup:l[JJ~~~oUP:JD l~;t~: 0 ,~ ,I";volutionl'ltle I I IGENERI 'KA Statement" .....~j Knowiedge of "fire in the plant" procedures. !Explanation ,Answers:
3 55.41(10) Overhead alarm A-7 states that If both zones :59 and 74 are received, then open 2FP147, This is controlled from the control room on 2RP5. C is incorrect because the deluge valves are automatic and in containment. B is incorrect because the I PORV breakers are operated when the fire is in the relay room, but is plausible becasue they are in containment where the fire is. A is incorrect because the Fire Pumps will not have auto started since the FP147 has to be opened, and is not an action contained i , ;n tho ,.,00
-~~--~---- .... ........"
. Reference Title Facility Reference Number i ~ce Section ,I,"'c~geNo.l Rio"."......
INo.1 &2 Units Fire Protection 11205222-2 II II 11 60 IOverhead Annunciators - Window A il S2.0P-AR.ZZ-0001 II 1122 11 51 I IControl room fire Response II S2.0P-AB.F1RE-0001 Ir== II 117 I Question~~~u;:ce-q IFacility Exam Bank IQuesti()~ M()(Ji,fic~t!ori M~e~t~h~od~.. :"1 Editorially Modified Question sourceC9m~m~en~t~s~(~V~is~iO~n~Q~6~9~8~0~5~,m~o~di~fi;ed~st~e~m~t~o~m~a~k~e~iftwh~a~t~w;o~U~ldfco~n~tr~o~1r~o~o~m~c;r;ew~d~o~in~s~te~a~d~o~f~wh~a~~~~~~~~~l
..--..... ,-~.-~~.~' ..~--~.~, system. Modified choices and added valve identifiers. Replaced dispa to 147
I
\'ROSkVSc/j~t/ ~cSR()~~~r.;h~rl ".Ros1}stJililEyallrtloilList ~~est~~IoPiC\ I_R~O __75__________________________________________________________________________________~
Given the following conditions:
- An Alert has been declared at Salem, and all required notifications have been made by the Primary Communicator.
Conditions degrade to the pOint where a Site Area Emergency is declared. Which of the following identifies the PRIMARY method which the Primary Communicator will use to make notifications to the States of Delaware and New Jersey, and how long from the SAE declaration do they have to make those notifications lAW Attachment 6, Primary Communicator Log of the Salem ECG? IESSX phones within 15 minutes. IESSX phones within 60 minutes. [AnSwer] ~ :Exarril~~ ~ 'C69hitiW'i.:ev~ 1Memory I 'Fadlity:ll Salem 1& 2 1 tExamD<lt~: *1 9/26/2011J
~1194001G443 I~== [TIIi$Ro\fa'ue:j01~~B§~D~'~[J} !systemIEvoluti§:ii§] 1______________________________________________________________________-1
- KAStatement: I
,-------------------------------------------------------------------------------~
Knowied e of emer enc communications s stems and techniques.
- Explanation of i 55.41(10) Salem ECG, lists the communications systems in order of preference. The NETS (Nuclear Emergency Telecommunications System) is the primary closed circuit communication system for off-site notifications. The ESSX is also a IIAnsw."" closed circuit system, which is used as a backup for NETS. The notifications to the States must be made within 15 minutes of the declaration of an Emergency, even if a lower classification emergency is already in progress. The 60minutes is plausible if the
[I Reference Title ISalem ECG - Attachment 6
~=
-I FacilitY-Reference Number J:~eferenceSecti(m 11__
II lf~~,~ ~ISiOi1' 1~15=5=::; IEmergency Preparedness Training Communicat II NEPCOMMDTYSC 11____ I[ 1[ 04 I II 1I-__----l11 1~I~ l.O.Number IGENISSE013
u.s. Nuclear Regulatory Commission Site-Specific Written Examination ~--- Applicant Information Name: Region: I Date: 9/26/2011 License Level: SRO Start Time: Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the HO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours to complete the combined examination, and 3 hours if you are only taking the SRO portion. All work done on this examination is my own. I have neither given nor received aid. Applicant's Signature
................--- -------..... *:*-::-:--==-::-:==========--=======--=:=-=-===-==*:-1 Results ~-------- --- _.- _._----- _.......................* ~- .. --.
RO/SRO-OnlylTotal Examination Values 1 I Points Applicant's Score 1 1 Points Applicant's Grade 1 1 Percent
SALEM 2011 EXAM - SENIOR REACTOR OPERATOR WRITTEN EXAM KEY
- 76. (SRO #1) D
- 77. (SRO#2) A
- 78. (SRO #3) C
- 79. (SRO #4) C
- 80. (SRO #5) D
- 81. (SRO #6) A
- 82. (SRO #7) B
- 83. (SRO #8) D
- 84. (SRO #9) A
- 85. (SRO #10) B
- 86. (SRO #11) B
- 87. (SRO #12) C
- 88. (SRO #13) D G.tiSCf (stet> tL/0 dt!-~U. ~ u/~/t(
- 89. (BRO #14) G
- 90. (SRO #15) C
- 91. (SRO #16) B
- 92. (SRO #17) D
- 93. (SRO #18) D
- 94. (SRO #19) B
- 95. (SRO #20) C
- 96. (SRO #21) D
- 97. (SRO #22) A
- 98. (SRO#23) B
- 99. (SRO #24) A 100. (SRO#25) B
! ~~-'~----'~ I ~ .-. -.~ ---. .-.. -- ; ,-~~~~
'Question Topic -~--~----,-
I SRO 1 I Given the following conditions:
- Unit 2 is performing aRCS heatup and pressurization lAW S2.0P-IO.ZZ-0002, Cold Shutdown to Hot Standby.
- RCS Tavg is 210"F.
- RCS pressure is 310 psig.
- 23 RCP is in service.
- 21 Charging pump is in service.
- 23 CCW pump is crr.
- 2A and 2C 4KV Vital buses are powered from 24 SPT.
- 2S 4KV vital bus is powered from 23 SPT,
- 24 SPT loses power, and NEITHER 2A nor 2C 4KV vital buses transfer to alternate power due to faulty power available relay for 23 SPT.
- 2A vital bus locks out on bus differential.
- 2S EDG fails to start.
Which of the following describes how the control room will respond? TheCRSwill ... F;;ljenter S2.0P-AB.CVC-0001 Loss of Charging, and take actions to start 22 charging pump prior to 23 RCP seal package temperature rising to ! the point where seal injection flow is prohibited and 23 RCP trp is required. !
;h ,I "ntAr :::;? (,)P-AR Rt:p_nnn1 R"",dnr t:nnl"'nt Pllmn ",nri !rin ?::l Rt:P
~
15 minutes after the receipt of OHA's 020-023 21(22,23,24) RCP BRG CLG WTR FLO LO. enter S2.0P-AS.CVC-0001, and check 22 charging pump has started to supply adequate seal injection to allow continued RCP operation. enter S2.0P-AS.RCP-0001, and trip 23 RCP based on the 10S:5 of seal injection and thermal barrier flow to RCPs.
~~~~j II:] ~Levell Is! IC09~itiveLeven IApplication I iFacm~J ISalem 1 & 2 I :ExamDa:~ I 9/26120111 ~:jl 000015A210 !;AA2.10-- iROyallie:J @ [SROValUer:J1j[se~l~ lRci-Gro~ 0 [SRo(;rouP~D lSystemlE;olution
Title:
!Reactor Coolant Pump Malfunctions iKA Statement!] Ability to determine and interpret the following as thf;3Y apply to Reactor Coolant Pump Malfunctions: When to secure RCPs on loss of cooling or seal injection
'Explanation of: 55.43(5)The initial electrical lineup, combined with the conditions occurring in the stem, will result in only 2C 4KV vital bus having
~Jwe!~:' ____i power. With the RCS temperature << 312", all charging pumps except one are crr. 21 charging pump is powered from 2B 4KV vital bus. NO CCW pumps will be running since the only Dowered 4KV vital bus CCW pump is crr. With the only operating (or available) charging pump tripped, there is no seal injection flow to RCPs and no CCW flow to RCP thermal barriers. The CAS II ,~;;"n in AI:! 0;"0 '. t, trin P('P~ J:l;" ~", ....",,,t II '. . , h",,.,,, ",,,, .,na, AI'l P("'D <Of",,,,,, 'h", if ("'('\1\, fin" . n. ,,,I ,,-I "ith' 5 minutes of the alarms annunciating, go to Attachment 2, Tripping RCPs, the loss of both seal injection AND CCW directs RCP trip without any time delay. C is incorrect because while AS,CVC will be entered on the trip of 21 charging pump, 22 charging pump will not have auto started due to its power supply being crr. D is incorrect because while the action will be taken to restore a charging to available status, the CAS action of AS.RGP will be the priori&, and seal package temperature will not stop that action.
!--c:~-.~,,-c-~ "'ReferenceTitle---;---l FTFadlityReferenceNumberJ 'Referel'lceSection--:; !Page No. iRevisioni
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i :~ _ _ ~ _ _---' I Reactor Coolant Pump Abnormality 11 S2.0P-AB.RCP-0001 1\ II 1121 I I Loss of Charging II S2.0P-AB.CVC-0001 II II ! 19 I I 1\ \I 'I II I I I fL.().. Number c_"_" ~~_~--,-~ _ _ _'
-'71 IASRCP1E003 I___ ---l
Given the following conditions:
- Unit 2 is operating with 21 Residual Heat Removal (RHR) pump and HX in service for coofing. - RHR HX inlet temp is 170°F. - RCS pressure is 280 psig. - The RO reports that Pressurizer (PZR) level is slowly lowering unexpectedly. - NO Overhead Annunciator alarms have been received. - Refueling Water Storage Tank (RWST) level is stable. - 21 Waste Hold Up Tank level is rising slowly.
Which of the following identifies the procedure which will be used and the action(s) taken in that procedure which will Isolate this leak?
§J !S2.0P-AB.RHR-0001, Loss of RHR. Close 2CV8, RHR Letdown I
IS2.0P-AB.LOCA-1, Shutdown LOCA. Close 2CV8, RHR Letdown. , IS2.0P-AB.LOCA-1. Place 22 RHR pump and HX in service and isolate 21 RHR loop. I
'doll S2.0P-AB.RHR-0001.
L-J Place 22 RHR pump and HX in service and isolate 21 RHR loop. I ~sWerlI a I !xarll Level i IS I ~ogniti":l9l..ev~!J IApplication I tFa(;iUtY~11 Salem 1 & 2 I [ExamD~i:e: i I
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* - - - ' Knowl elge d 0 fEOP m,T:lga rIon strat egles. .
i ieXPlanation of: 55.43(5) The 2 AB.LOCA distracters are inccrrect because AS.LOCA is applicable in MODE 3 with Accumulators isolated or MODE ~sw~rs: c~.J 4. With MODE 5 indicated in stem, it is the wrong procedure, although it has the correct action in distracter B. A is correct because the 2CV8 will be isolated and it is the correct action for a leak which is causing the WHUT level to rise. Distracter C is inccrrect because it has the wrong action, with the right procedure. The action in the AB.RHR is to remove BOTH loops from service and
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I IShutdown LOCA II S2.0P-AB.LOCA-GOO1 II II lis I Ie II I[ I I i.Lo,Numberd IABRHR1 E005 I jABRHR1E004 I I I
'I ~on S~_urge:: IFacility Exam Bank IrQues~ion MOdi~cati?nMethod:~i Significantly Modified I'Used Duringtr~jriing PIogrlitri] [J ~tj()l'l~B~~~~ ISalem 2003 SRO NRC Exam. (5 NRC Exams ago) Modified to include which procedure to use. I 'Coml",,'" ..... . . . ........ ...........' . *.*. ,..... ... i*'* * *i.'*** .**.**.... '... . ....***.*..'**." ... '..*.~. .,..; .....1 I
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Operators observe the following:
- OHA E-28 PZR HTR ON PRESS LO alarms.
I\ - PZR level is 28% and lowering slowly.
- Control Rods begin stepping out in automatic at 8 spm.
- OHA C-38 CFCU LK DET HI alarms.
- Rx power is rising at 0.35% per minute.
Which of the following identifies the correct procedure and actions which will be performed in that procedure?
~\Enter S2.0P-AB-RC-0001, Reactor Coolant System Leak. If VCT level cannot be maintained 11 %, swap charging pump suction to the RWST I and trip the Rx, confirm the Rx trip, initiate Safety Injection. I I Enter S2.0P-AB.STM-0001, Excessive Steam Flow. Trip the Turbine, initiate AFW, lower power to <5%. I I Enter S2.0P-AB.STM-0001. Trip the Reactor, confirm the Rx trip. initiate Main Steamline Isolation, initiate Safety Injection.
I I Enter S2.0P-AB-RC-0001. If PZR level cannot be maintained stable or rising, trip the Rx and initiate Safety Injection. j
~e\i!!!r! ~ ~~~!ci;..!!f rs==t :C!:!'!l~!'.':.L~*"!2! ~ !A::::';~ti'?n I tacility: HSalem 1 & 2 hEx~mOate:H ____9_/2_6_/_20_1_1J
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~Stat;,;~nt:j Ability to determine and interpret the following as they apply to Steam Line Ruolure: I Conditions requiring a reactor trip 1
.Explanation of 1 55.43(5) C is correct because the indications in the stem are of a steam I,eak in containment. Rx power is rising at -21 % per hour,
,AnSwers~ . .! while the turbine load increase is at 10% per hour. AB.STM states that if Rx power is rising uncontrollably. take the actions as
- -.-~~ __ - 1 described in C above. A is incorrect because only some of the indications present in stem would be present for a RCS leak, and the action is incorrect also. B is incorrect because the actions described would be taken if the steam leak were outside containment
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j Excessive Steam Flow II S2.0P-AB.STM-0001 II 1\ II~~ IReactor Coolant System Leak II S2.0P-AB.RC-0001 1\ II 11 10 I I II II II 11_---.1
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rQu~stion Topic l SR04 I I Given the following conditions:
- 21 CVCS Monitor Tank is in recirc.
- 21 CVCS Monitor tank will be released via the Waste Discharge Cross Connection to Unit 1 Liquid Radwaste system lAW S2,OP-SO.WL-0001, Release of Radioactive Liquid Waste from 21 cves Monitor Tank.
Which of the following describes an action which would result in an unmonitored Radioactive Liquid Release if it was performed AFTER the Liquid Release was started, and which would require operator action to stop the release? la.* 1Unit 2 eRS authorizes a tagging request which results in the 2R18. Liquid Waste Disposal Rad monitor losing its power source,
~.
ib.: I izes a tagging request which results in the 1R.i8, Liquid Waste Disposal Rad monitor losing its power source. 1Unit 2 CRS authorizes a tagging request which results in the 2FR1 064, Rad Waste Liquid Monitor Pumps Discharge flowmeter losing its power source.
'd'lunit 1 eRS authorizes a tagging request which results in the 1FR1 064, Rad Waste Liquid Monitor Pumps Discharge flowmeter losing its power source.
[~Il:;\'IIe~j 1c I [Exam Lev~l_ ~ ClIl:j.lit.... Leveli 1Memory IIFapility: i ISalem 1 & 2 I i~an'lOate: i I 9/26/2011 IKklI 000059G236 11 2 .2 .36 :IRQVatue:l[ITI SROValuet 4.2j...."'..... v .. *I~ROGr()upJI 2IiSROGroup:l/ 21 l~t~: ~
~System/Evolution
Title:
IACCIdental LIqUId Radwasle Release ...0_5_..9........,
- !KA Statement*.'!
Ability 10 analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations, I
.Explanation of i 55.43(4) The Salem ODCM is a supporting document to the Unit 1 and Unit 2 Technicai Specifications. The previous LeOs that Answers: I were contained in Radiological Effuent Tech Specs are now included in the ODCM. Using ODeM 3.3.3.8, lAW Salem Tech Specs - 6.8.4.1.g.1, and Table 3.3-12 on page 19, the R18 a"d the FR1064 required to be operable. The R18 is interlocked Vlith the WL51 so that if the R18 loses power, the WL51 will shut, preventing an unmonitored release, and no operator action would be required to I
Ibetween
. the flow recorder and the WL51, so the release
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would be considered unmonitored based on the fact thai a required
! component was not available, and the release is ongoing, Loss of the flow recorder during the release requires operators to close Icross the 2WL51 lAW step 5.5.9. The Unit 1 distracters require knowledge of the flow path for a release which is directed through the connect line, Reference Title l Facility Re,ference .Number . IiRefe~nceSection I,;Page No.* iR'm;iiim, ISalem ODeM II II II I !26 I IRELEASE OF RADIOACTIVE LIQUID WASTE II S2.0P-SO,WL-0001 11 II 1124 1 I
IWaste Disposal WL51 11203679 II il I 18 I I
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Que!tion Topic i ISR05 I Given the following conditions: I- Unit 1 is operating at 100% power.
- 21,24 and 26 SW pumps are in service.
. 21 and 22 SW header pressures are 108 psig .
- The following OHAs annunciate within 10 seconds of each other in this order:
- 8*13,21 SW HDR PRESS LO
. 8-14,22 SW HDR PRESS LO
- B-15, TURB AREA SW HDR PRESS LO.
- B-48, SW VLV RM FLOODED.
The standby SW pump starts automatically, and OHAs B-13, B-14, and B-15 clear. Which of the following describes the status of the SW system, and which procedure will provide direct actions which will mitigate the event? I
~ A SW leak upstream of the 2ST901, TURB LO CLR ST RET \J has occurred. S2.0P-AB.SW-0002. Loss of Service Water-Turbine Header.
IA SW leak upstream of the 2ST901, TURB LO CLR ST RET \J has occurred. S2.0P-SO.SW-0001, Loss of Service Water Header Pressure. I I A SW leak in the CFCU piping in the 78' Mechanical Penetration Area has occurred. S2.0P-AB.SW-0002, Loss of Service Water-Turbine Header. J I A SW leak in the CFCU piping in the 78' Mechanical Penetration Area has occurred. S2.0P-SO.SW-0001, Loss of Service Water Header Pressure.
;AnSWflrll d I iExamLevel! IS ,cojJnitt-'e*L!v~lj IApplication I iFacility:] ISalem 1 & 2 I ~amDat:'J I 9/26/20111
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[SystemlEvolutiOn}itle J ILoss of Nuclear Service Water 1 062 rKA
-- Statement:! .. _-- Ability to determine and interpret the following as they apply to Loss of Nuclear Service Water: 1 The cause of possible SWS loss Explanatiollofl 55.43(5) The leak location could be in the TGA with the conditions in the stem except for the SW valve room flooding. Knowledge lAnswe~:_.~..'-' of where the SW valve room and what piping is there is needed to answer question. The 2ST901 would respond on a TGA leak, and depending on leak size could cause a restoration of header pressures. If the leak is determined to be in the TGA, the AB.SW-2 procedure would be used. Although AB.SW-1 is entered for all the conditions in the stem, it does NOT provide direct actions (as
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....! F .facility~efer~l1ce .~~mber-l :~eferenceSectiOri **.. ] : P~~~.No,J :Re~iS~
ILoss of SW Header Pressure !IS2.0P-AB.SW-0001
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II II 112~ ILoss of SW- Turbine Header II S2.0P-AB.SW-0002 II II 1111 I IOverhead Annunciator Window B II S2.0P-AR.ZZ-0002 II II 11 35 I i.O.Number:__ ~- ...*-:
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I §~~~tiO~ Source Comments] I
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- ~ue~~on T~~J Given the following conditions:
- Unit 2 is in MODE 6 entering a Refueling Outage.
- No fuel assemblies have been removed from the Rx.
- Rx cavity level is 26' above the PRV flange.
- 21 RHR loop is in service in Shutdown Cooling.
- 22 RHR loop is O/S and available.
Which one of the f"I1"",inl"! would initiation of Rx Oi) Unit 2? Loss of Control Air to containment. Loss of Plant Page capability in containment. Racking down the 22 RHR pump 4KV breaker. Only one of the two SRNl's can provide audible indication in the Control Room.
,---,""'-c""'_...- .-.'. --I Eognitive Le\lE!~--, -KAJI000065G136 ~T~.* -.'~-*~*T**-~ .. ;System/EvolutionTItle i I_L_oS_S_o_f_l_ns_t_ru_m_e_n_t_A_ir_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-' C-*..__. _ ._ _' ! KA Statement' !.........--~~~--, .. ~
Knowledge of procedures and limitations involved in core alterations. I I. IAnswers: [Elcplanation
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_ _ _ .:..~J i 55.43(7,6) The requirement for SRNl's is BOTH operable and providing VISUAL indication in the Control room, with ONE providing AUDIBLE indication in the control room. The manipulator crane is air powered for gripping, so the loss of air to containment would preclude being able to perform core alts. Onfy ONE RHR loop is required to be in operation in MODE 6. 2 RHR loops are required I to be OPERABLE when <23' level above the flange. Cc-~ __- ____ ,Reference Title ,***. -:;_____ --~i ~-;FaciliiyR~'rence Number: fRet;renceSection] [page No.. , *Revi:;!on] ___ ~"." ,_""_"" _ _ _ ~__ _
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IRefueling Operations
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'I S2.0P-IO.ZZ-01 07 II 11 13 10 7 I r Reac Pene Area & Cant Control Air i
11205347-1,3 II II 11 42 ,36 I i II II II 1I I FL.O.. Number IIOP009E002 I REFUELE007 1-_---' Material Required forEXami l1 ation .*. j I 'I
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Question Topicj . SRO 7 ! Given the follovving conditions:
- Unit 1 has experienced a large Main Steamline Break (MSLB) inside containment from 100% power.
- Safety Injection was initiated.
- MSLi has failed to shut ANY MS167 and they remain open.
- 11 AFP is crr.
- 12 and 13 AFP's tripped after starting.
- RCS pressure is 700 psig.
- ALL RCPs have been tripped.
- Containment pressure is 18 psig and rising.
- ALL Wide Range SG levels are 35% and dropping.
- ALL SG pressures are 120 psig and dropping.
- RCS Tc's have dropped from 540°F to 230°F in 20 minutes.
IEvaluate the data and select the procedure to be entered and action to be taken upon transition out of 1-EOP-TRIP-1, Reactor Trip Response.
~ !1-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions; shut all MS10's and steam dump valves.
!1-EOP-FRHS-1, Response to Loss of Secondary Heat Sink; initiate feed and bleed immediately.
J 1-EOP-FRHS-1, initiate feed and bleed ONLY when 3/4 SG WR levels have dropped <32% . id,111-EOP-FRTS-1, reset Safeguards Actuation and restore normal charging and letdown.
~swerll b I ~rJjLevel' !S I [Cogniti"e~eveIJ IApplication I :~ac!,lity:11 Salem 1 & 2 I :ExamDaf~ I 9/26/20111 I .~ 00WE05G406 1 2.4.6 1 i :ROV;alue:!1 3.71 iSRCrWlu~ 4.71 [S;ction~l~ !ROGrouP~1 11 .SRO GiCili~1 1\ ~\f1 ,System/Evolution Ti~
f"~' "~'-'.-.~, ~~~'~'-'. -',' . ~ ILoss of Secondary Heat Sink KA Statement: I
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! "".,~ jExpta l1ation of i 55.43(5) B is correct because the conditions given in stem would transition to FRHS*1 due to a RED path of no AFW flow and <9%
~~wers:,d NR level. The Bleed and Feed initiation criteria are when S/G WR levels are <36%(adverse), NOT 32% as in distracter c.
Distracters A&D are incorrect because it is a lower priority RED path, though it's action is correct for the procedure. I
,','ReferenceTitle'i' ~l i,FacilityReferenciNUll1ber 'I [ReferenceS~ctionj ~ge No.' [Revision'---~~
IResponse to Loss of Secondary Heat Sink 1I1-EOP-FRHS-1
~ c II il i 124 1 IResponse to Imminent Pressurized Thermal Sh 1I1-EOP-FRTS-1 II II 11 25 I II Critical Safety Function Status Trees 111-EOP-CFST-1 II II I 125 I IFRHSOOE005
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Quei#i~p TOPiC'll SRO 8 I Given the following conditions:
- A Rx trip and SI were initiated based on a LOCA 1 hour ago. - Operators have transitioned out of EOP-TRIP-1, Reactor Trip or Safety Injection. - RCS pressure is 350 psig and lowering very slowly.
1- All RCPs are stopped.
- RCS highest CET is 290*F and lowering very slowly. - Rx power is 300 cps and stable. - RVLlS Full Range is 95% and stable. - 21-23 SG NR levels are 33% and stable, and 24 SG NR leve! is 0%. - Containment radiation levels are 500 mr/hr and stable. - Containment pressure is 5 psig and lowering very slowly. - Containment Sump level is 46% and stable.
Which of the following identifies the highest priority CFST for these conditions, and actions taken in that associated EOP? a.1 !yeliOW Path Core Cooling. Isolate ECCS accumulators, depressurize SGs to atmospheric pressure. I 1Purple Path Containment Environment. Isolate Containment Pressure Relief flowpath (2VC5, 2VC6) and return to procedure in effect. IPurple Path Containment Environment. Isolate fluid sources from outside containment which have corroborating indications of lower than levels. '~ \.-1lt:!(;K C\.-\.-;:' IIUW lUI (;Ullt:!ln M.\.-;:' WIIUllIurr:s a::; t:!X[Jt:!(;lt:!U UI :Slall fJUlllfJ::; aflU allYflllUW [Jalll::;, ariU t:!(I:surt:! r-4.M. RPV Head Vents are shut. Answer 'I d I ~ExamLe"elll S I ICognitive Level** II Application I )::aCilit~J ISalem 1 & 2 I '~x'~~Date:! I 9/26/2011J ~:JI 00WE07A201 IEA2.1 ROvaiUel@lsRovalu~fSection:ll~ ,RO Gro~r=:~j:SR()Group:JU System/E'IolJtldrtTItre I ISaturated Core Cooling i KA Statement:j Ability to determine and interpret the following as thez applz to Saturated Core Cooling: ! Facility conditions and selection of appropriate procedures during abnormal and emergency operations. I 1~I"'*Clnation.of . ! 55.43(5) Core Cooling is the highest priority based on: All RCPs off and no subcooling and CETs <700* with RVLlS level >39%.
.....'!'..".. The actions in FRCC-3 check first if you need to be in a different procedure (SGTR-4 or AB.RHR-1), then has you reset safeguards, check ECCS flow vs RCS pressure and align valves if less than expected. Then it checks proper PZR PORV and RPV head vents shut The CFST Tables for subcooling HAVE to be provided since they HAVE to be used to determine subcooling. Purple Path for I! ('1= ,.."" r,,,, ",it;'"".,,1 t; n<>in ' " ('nnt <::lImn ..,7t;0/, ("rI"""",,,,\ Th", V",II", . ""th fnr (';:: rm "nnt r"rI ie ., ')I::ilr,r correct FRCE-1 actions-are partially cOrrect. in' that you would isolate 'VC5 and 6 and return to procedure in effect, but the conditions for entering FRCE-1 were never met so you wouldn't be in procedure < 15 pSig cont pressure,
;::r.>(';::~" ",..HAn" '
h '.0,." Reference Title ° .1 Facility Reference Number
- Reference Section : i Page -,No.!
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[Revisionl
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ICritical Safety Function Status Trees Ie
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rI2-EOP-CFST-1 II 11 25 I IResponse to Saturated Core Cooling conditions 112-EOP-FRCC-3 II II 11 20 I I II II II II I
@r&kvSeraper%l, SR9 skV~ct.rIJ'e(4 '(, R(tsvs{~m/ij~r!1tiont,~f~ "1~~~5s~teluil;vqlu~~ft list~~OUtlin~ {;~it~~i "
ISR09 Given the following conditions:
- Unit 2 was operating at 15% power prior to synchronizing the Main Generator. - A Main Steam line rupture occurred that resulted in multiple Steam Generators depressurizing in containment before 2 steam generators could be isolated from the 2 faulted SGs. - The 2 faulted SGs Tcolds are reading 270°F and lowering, - The intact 2 SG Tcolds are 330°F and stable. - RCS pressure is 500 psig and slow1y lowering. - Containment pressure is 16 psig and slowly lowering. - All SG NR levels are <9%. - Total AFW flow is 24E4 Ibm/hr. - Source Range Nls are NOT energized. - Intermediate Range SUR is 0.0 DPM.
identifies the procedure ent re uired, and actions which will be erformed in that procedure? 2-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions, Maintain AFW flow >22E4 Ibm/hr until at least ONE intact SG NR level is >15%, stop all ECCS pumps except 21 or 22 charging pump. . 2-EOP-FRSM-2. Response to Loss of Core Shutdown. Energize Source Range channels and verify SR SUR is 0 or negative. 2-EOP-FRTS-1. Isolate any faulted SGs, depressurize RCS w1th ONE PORV to within 100 0 /hr Cooldown Curve. 2-EOP-FRSM-2. Establish AFW flow >44E4 Ibm/hr. borate RCS until IR SUR is negative.
~~~~~~"'~~ ~cation ISalem 1 & 2 RO vai;;ij I 3.41 SRO valueJ~ !Secu?n: [~PE , *ROGroup:1 Facili conditions and selection of appropriate rocedures during abnormal and emergency operations.
- Explanation of i 55.43(5) A is correct because the stem conditions result in a PURPLE path on FRTS. Actions for maintaining AFW flow (Step 3.5)
Answers: and ECCS pump reduction (Step 12) are correct. C is incorrect because depressurizing the RCS to restore conditions w1thin the 100°F/hr curve is performed in FRTS-2 in response to a Yellow Priority Condition. FRTS-1 is entered from either RED or PURPLE conditions. and with SG NR levels <9% (which means you're less than 15% adverse) you are directed to maintain AFW flow> 22E4 2. The FRTS is a higher priority. and is a PURPLE path. B is incorrect because it is the 'Nl"ong procedure with the correct actions of that procedure. D is incorrect because it is the 'Nl"on procedure. and the actions are performed in FRSM-1. I Reference, Title. I
',", __ ~cilitYReference Num~iReference Secti()nj ~!!~~~J r_R._-e_vi_S_iO~
~IC:ri:ti~::I:sa:f~e~~~F:un:c:tlo:n:s:t:at:us::Tr:e:es::::::~IFI2=-E=o=P=-C=F=S=T=-1============;ill==============;1I IFI2=5==~ IResponse to Imminent Pressurized Thermal Sh !l2-EOP-FRTS-1 II II 11 25 I~R~es~p~o~ns~e~ID~L~o~SS~O~f~c~o~re~S_h~ut~do~w_n______~IFI2=-E=O=P=-F=R=S=M=-2==========~IFI============~q IFI2=0==~
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*Question TC:PiC] ISRO 10 Given the following:
- Unit 2 is operating at 100% power.
- 21 SPT is crr.
Time T=O A Rx trip occurs when the 2H 4KV Group bus loses power. Time T*1 0 A breaker failure relay actuates for 13KV North ringbus breaker 3-4. Time T-25 The follOwing indications are present:
- All RCS WR Thot's are 559°F and rising slowly.
- All RCS WR Tcold's are 549°F and stable.
- All SG pressures are 1015 psig and stable.
- All SG NR levels are 39% and stable.
- PZR level is 23% and rising slowiy.
Which of the following identifies the action that must be Qerformed? ra:11 Lower 21-24MS10 setpoints to establish CET's stable or lowering lAW S2.0P-IO.ZZ-0008, Maintaining Hot Standby. ILower 21-24MS10 setpoints to establish CET's stable or lowering lAW S2.0P-AB.RC-0004, Natural Circulation. i I Lower Main Steam Dump pressure setpoint to stabilize or reduce RCS Thots lAW S2.0P-AB.RC-0004. i '-Lower Main Steam Dump pressure setpoint to stabilize or reduce RCS Thots lAW S2.0P-IO.ZZ-OOOS,
!Answer i ~ ~~~~J Is I ~ogrliti"e Levell! Application I :F~cility:! ISalem 1 & 2 I tEXamDate:j 1_ _ _9_/2_6_/2_0_1....)1/
- KA~IOOWE09G411 li~~~=§~~~~~~~;H~~~~JD[~~~~DJ lY'J
~fEVOlution]t(e1 ! Natural Circulation Operations ' -_ _ . _.~_ *.J , - -_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- . . . , . .KAStatement*;
Knowiedge of abnormal condition procedures. I I,ExPlanati<:mofl 55.43(5) The breaker failure will cause a loss of 13KV SPTs #2 and #4, and #12 and #22, and 23 SPT, The 23CW bus will lose Answers: < .... I power and deenergize all "A Circulators, but all "B" circulators remain in service and the Main Steam Dumps remain available
~~~.-~-. initially. However, all condenser vacuum pumps lose power, (powered from H,F,G busses) and condenser vacuum will rise to the paint where steam dumps are blocked (20" vaCUUIT! lowering) This is why SG pressure has risen to 1015 psig vs. the 1005
., "', i n ' ? 1 C:;PT "",in,., r.!T'(, ,II,* h"'" nf'\ I""r! 'Anth "nit "t 1 11/,\0/, nf'\'Al",r\ ,rill ",ee, !It in nn nf'\\AIQr tf'l 21' and 22 RCPs. The b~eaker failure will deenergize 22 SPT causing a loss of the other 2 RCPs. A'and are incorrect because B 1MS 10 operation is not the preferred mode of temperature control, and the steam dumps remain available. 0 is incorrect with right action but wrong procedure. AB.RC-4 is entered on a loss of forced circulation in MODES 3 and 4. The stem shows nat elr is NOT 1present. so the correct answer has to be increase steam flow, but knowing how is required, with the correct procedure.
Reference Title _~ .~Z-~ ~~illty~;rr~NUmbetl [R~riCese;ctf~ ~~~ ~~~ I Natural Circulation !IS2.0P-AB.RC-0004" II 117 I I q II II iI I I " II II II I
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i'~ "Y' ., ~ ." -.--'~>~ ;;~1, i~uestionTo~ 'SRO 11
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I Given the following conditions:
- Salem Unit 1 has performed a Rx trip and SI based on rapidly lowering PZR pressure and level. - The crew has transitioned to 1-EOP-lOCA-1, loss of Reactor Coolant.
Which of the following describes why the first several steps of lOCA*1 check for reasons OTHER than a lOCA for being in lOCA-1? lOCA-1 actions assume ... rai Ia loss of RCS inventory. If the ECCS injection flow is due to a loss of Secondary Coolant, an unnecessary transition to FRHS-1, Loss of Secondary Heat Sink could be required. Ia loss of RCS inventory. If the ECCS injection flow is due to a Loss of Secondary Coolant, the event could be terminated by isolating the faulted SG. Ia SGTR is NOT causing the LOCA condition. If the ECCS injection flow is due to a SGTR. an unnecessary transition to LOCA-3, Transfer to Cold Leg Recirculation would be made. Ia SGTR is NOT causing the lOCA condition. If the ECCS injection flow is due to a SGTR, the tube rupture must be addressed before returning to LOCA-1 to terminate ECCS flow,
,Answer IIb I [i::)(am Lev~ IS I icognitiveLe"el 11 Memory I 'FaImtii t Salem 1 & 2
( ~ ---- ,- ,J 1 Exam[)ate: 11 9/26/20111
~J I006000G422 1,2.4.22 jROVallIe:J13.6I'sRQvalueJ 4.4I:secti~IS~IROGr()uP:1 11'sROGroup::1 11 !I~n ~
SyStem/~vol~tiOI1 Title i 1Emergency Core Cooling System 11 006 I
;KA Statement?
r-----~--------------~~----~----~~----------------------------__, . Explanation of i 55.43(5) The Major Action Categories for LOCA-1 are: 1. Check for Subsequent Failure. 2. Monitor Plant Equipment for Optimal
,Answers: J Mode of Operation, and 3. Determine optimal Method of Long-Term Plant Recovery. The first item checks that a faulted or ruptured SG is not the reason for ECCS injection, and either fixes it on the spot (faulted) or transitions to another more appropriate procedure With a faulted SG being the cause of the ECCS flow, LOCA-1 has a "do loop" which will wait until the SG has blown performe-d to do actions which will terminate'the primary to secondary leakage, which is not performed in LOCA-1. Additionally, I
, there is no transition from SGTR-1 to LOCA-1. The unnecessary transition to FRHS-1 is plausible because it happens on other procedures (FRCE,LOSC-2) when minimizing AFW now to SGs when all are faulted. The KJA matches because the logical extension of safety function prioritization is the way the EOPs are ordered when different casualties can have the same indication or I system response, but need to be addressed in a certain order.
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.~. .... .:.' Reference Title .*. .... . *..* '.' .*.f . . . Facility Referenc,e Numbef..1j iReference ~ectionJ :Page No.! !Revisicm; ILoss of Reactor Coolant 111-EOP-LOCA-1 II "11 25 1
'1 Steam Generator Tube Rupture 111-EOP-SGTR-1 IC II 1124 I ILoss of Secondary Coolant i!1-EOP-LOSC-1 IC II 11 20 I l!c~_!;!~__.. ~ I LOCA01E005 I ISGTR01 E003 I ILOSC01 E002 I
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~---"~~I Question!Opic j SRO 12 I Given the following conditions:
- Unit 2 is operating at 100% power.
- 2PR1 fails open and remains open.
Which of the following Identifies how this affects the PZR Master Pressure Controller (MPC) response, and what consequences, if any, are associated with the actions performed by the crew lAW S2.0P-AB.PZR-0001, Pressurizer Pressure Malfunction? [] IMPC output will LOWER. The unit may continue to operate indefinitely after the initial mitigative actions are completed. IMPC output will RISE. The unit may continue to operate indefinitely after the initial mitigative actions are completed. IMPC output will LOWER. A unit shutdown will be required if 2PR1 cannot be restored to operable status. IMPC output will RISE. A unit shutdown will be required if 2PR1 cannot be restored to operable status.
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.~."~' I LExam LevelJ l S i ;C091lIt,ve I iF,ac,htY:ll Salem 1 & 2 I 'ExamDate:11 9/26/201111 ~I010000A203 1iA2*03 t¥o,valllE~:G~RO'lalll~[S-e~I~R6~U~ROGr~[JJ ISystemlEvolutionTitle!, IPressurizer Pressure Control System iKA Statemel'lt1 Ability to (a) predict the impacts of the following on the Pressurizer Pressure Control System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: PORV failures I
- Expfanation of ,I 55.43(2) This question is SRO level because of the Tech Spec knowiedge required, and what actions TS directs for different PORV I
~swe"" ~ m,II"";,n" Add;!;,n,lIy, ""If, th' ",," d,,,n'l 'p",,,,lIy os, what p""d"co I, "" (100 ,osy (0, AB,PZR), It " ' ' ' coq"'re knowledge of the actions IN that procedure. The MPC raises output when actual pressure rises, and lowers as actual pressure II lowers. As pressure lowers due to the open PORV, the output will lower to turn on heaters and close spray valves. When the PORV 1'l1"l'v "",I"",;c "h"t t" ;""I"to tho P('IQ\/;" t.1'l P7R *r..",h ~"",,, ':! A " ",..H"n h ;f th", P('IQ\ *;c "n' ,oct,." .." Hh;" 7') h", ore eh, ,I"""," is required. A PORV isolation that DOESN'T require shutdown if not fixed is a leaking PORV, which is isolated by its Block Valve I with power maintained to the Block valve. Reference J"\tle77--::--~ :~y~hceNurnber J~ete~S~"l Wa9lr~~,,;;r~'~"i IS2.0P-AB.PZR-0001 il Pressurizer Pressure Malfunction II II 1118 I ISalem Tech Specs II 113.4*5 II II I I II Ie II II I ~~7~ I ABPZR1E002 I PZRP&LE010 IMaterial Reql.lired for Examination l!
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- Unit 2 is operating at 100% power. - 23 charging pump is in service. - A control circuit malfunction has resulted in the 2CV3, 45 gpm Letdown Orifice Isolation opening, and cannot be shut from the control room.
Which of the following describes the impact this failure will have on operation of the PZR Level Control System, and how will operators respond? 23 charging pump speed will rise. Place an additional charging pump in service in response to rising VCT level lAW S2.0P-SO.CVC-0002, ! 1Charging Pump Operation. I 2CV55 will modulate in the open direction. Isolate letdown and place Excess Letdown in service lAW S2.0P-SO.CVC-0003, Excess Letdown 1Flow.
!2CV55 will modulate in the open direction. Place an additional charging pump in service in response to rising VCT level lAW S2.0P-SO.CVC 0002.
, 123 charging pump speed will rise. Isolate letdown and place Excess Letdown in service lAW S2.0P-SO.CVC-0003.
[AnswerJ II::]" :EXamLevel i rr=J rcogriiii"~I..~ IApplication I :Facillty~ ISalem 1 & 2 I :Exam[)ate:.! I 9/26/20111 t!<A~1011000A201 !iA2*01 ~iROV~[~3:nSROVafueLEj:sectiOn:lI~~RO.GrOUp:~U~~0' li4jl ~ rSyster:nIE\{olutio~Titlel :::.::.J IPressurizer Level Control System I
;KAst~tement:1 Ability to (a) predict the impacts of the following on the Pressurizer Level Control System and (b) based on those predictions, use ~-"--~--.-~--
procedures to correct, control, or mitigate the consequences of those abnormal oQ..eration: Excessive letdown Ir ....... ~
!Explanation of: 55.43(5) Letdown orifii are Containment Isolation valves, and will require letdown isolation to isolate the CIV. For long term 'Answers*<*l operation Excess Letdown must be placed in service. The 23 charging pump (PDP) is normally in service at power. The CV55 is I '~-~----=-..~-~
placed in manual and open to prevent it from responding to the Master Flow Control/er when a centrifugal charging pump is not in ;1 service. It has a maximum flow of -96 gpm. Normally, letdown flow is 75 gpm with one 75 gpm orifice in service. With the 45 gpm "F. * * ;n ~. ,,.;, fl, _.
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the CV55 when a centrifugal pump is in service. With a CCP O/S, the CV55 is in Manual ope;;, and wi(1 not be affected by PZR I level. Transferring to a centrifugal charging pump is a correct choice for an action, but would not be in response to rising VCT level. While starting an additional charging pump would certainly raise charging flow, Salem does not run with a PDP and a centrifugal charsins E:ump in service concurrentl:t except for the short period of time when transferring between pumps.
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- Unit 1 is operating at 100% power. - 1PR5, PZR Safety Valve, fails open. /
I Which of the following identifies what procedure will be used upon the transition out of 1-EOP-LOCA-1, Loss of Reactor C~: and the actions in M ,,,,oede,. whioh "" mlti,,\e ** elf", of tho ",eo!. /
,!1-EOP-LOCA-2, Post LOCA Cooldown and Depressurization. Establish a <100"Flhr RCS cooldown y{ restore subcooling. Depressurize the I.RCS and stop the depressurization when PZR level reaches 77%. /
I fb.' 11-EOP-TRIP-3, Safety Injection Termination. Determine minimum required subcooling, then r~ce ECCS pumps to minimum require per
- Table C to equalize ECCS and break flow. / J 1.1-EOP-LOCA-2. Establish a continuous RCS cooldown to restore subcooling. Depr~7ize the RCS ONLY until PZR level is >25%. I 11-EOP-TRIP-3. Sequentially stop ECCS pumps while checking RCS pressure 7 e or rising to equalize ECCS and break flow. I !Answer j E::J ~L.eTeiJ ILJ lfclgnltlveLev!!.J I Application L1acllity: i ISalem 1 & 2 I iExamliate?i ! 9/26/20111 LKA:j1013000A203 it2~_~_J~~JB~y!t~~~~~~I~~E~~D1l~~~~i~uhlU !Ii! ~ ~ystemlEvolutioriTitle j I Engineered Safety Features Actuation SystemL I :013 iKAS~tem.!l.btJ Ability to (a) predict the impacts of the following on th!l'Engineered Safety Features Actuation System and (b) based on those predictions, use procedures to correct, control. or l1')./figate the consequences of those abnormal operation:
Rapid depressurization / fexPlariatio.h.O. f 1 55.43(5) The PZR Safety failing open~'11 ca e a SBLOCA. Pressure will rapidly lower, and a Rx trip will occur, followed shortly by
- Answers: . I a Low Pressure SI. The Safety Injection wi actuate, and break flow will equal ECCS flow at - 800 psig. With no subcooling, the c~_~~ _ _ _ _ _ ~ transition to TRIP-3 cannot be made, an OCA-2 will initiate a cooldown/depressurization. The depressurization is ONLY performed until positive control of RCS' ventory is established at 25% PZR level, as any continued depressurization may cause a ii,,,,,, "f "', ,h....""n"" If "', . "" I", ..h win" .he. don.-o",,,,, ",.,,,II,,n It* ill h" .-""t".-o'-{.,,' tho ('(')l\ITIl\III(')1 1<: ,..."nti", ",,,,
The LOCA-2 distracter con~ains-th c orrect cooldowl' rate limitation (which is intentionally absent from the correct choice) but the 77% PZR level is wrong, It is th evel at which a depressurization in LOCA-5 (step 24.1) would be stopped. Choice B is wrong because wrong procedure and rang action. Table C is found in LOCA-2. Choice 0 is wrong because wrong procedure, with correct action for that "foce reo ! Ii. ....' <Ref".. m.:~Title ........// . i:~fa~lityRefei'enceNti~t;r-*"7] ~~~ren~ Se:Cti6ri~"-2J t~~J:i~l ~~~ Post LOCA Cooldown and Dep res/sur,J6tion " 1-EOP-LOCA*2 II II 11 23 I I Safety Injection Termination II 1-EOP-TRIP-3 Ie II 11 25 I 1 1 - - - - - - - - - : ' 1 - / - -.. . . . 1. 1
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!Question Topic' ISRO 15 I Given the follo'Ning conditions:
- Unit 2 is in MODE 6.
- Core off-load is in progress and is 2/3 complete.
- A fuel assembly is in the transfer cart at the Spent Fuel Pool.
- The mast tube of the manipulator crane is empty.
- Refueling cavity level begins to drop rapidly, approximately 6 inches per minute.
- Radiation Protection reports flooding in the lower elevations of containment.
Which of the follo'Ning describes how the Fuel Transfer Tube Gate Valve will be operated lAW S2.0P-AB.FUEL-0002, Loss of Refueling Cavity or Spent Fuel Pool Level?
'\IMMEDIATELY shut the gate valve. Maintain the fuel assembly in a horizontal position.
I
'IIMMEDIATELY shut the gate valve. Transfer the fuel assembly to the Fuel Handling crane and place in designated position.
I ISTART the transfer cart moving to containment, then shut the gate valve when the cart is clear. The fuel transfer cart 'Nill stop when the gate valve open limit is lost. J
, 1START the transfer cart moving to containment, then shut the gate valve when the cart is clear. The fuel transfer cart will continue until it reaches its normal travel stop on the containment side. I Answer II c I ~am Le'lelJ Is I rCognitiveLeveTl IMemory I iFa~ility: II Salem 1 & 2 I fExamDate:.] j 9/26/20111 l't<A:, I034000G435 1 12 .4.35 :Rb..Value:1\ 3.8IsROValuei~isectiOn!I/SYS liROGroup:il 21~ROGrouP:il 21 ~I IYi ..~"~-- ----.- .. -.-.'-," ' - . - . - . .,
i.SystemJ~volution T!!I~ t Fuel Handling EqUipment System 034 IKA Statement*: .... ~- Kno'Niedge of local auxiliary operator tasks during an emergency and the resultant operational effects. I IIEXPlanatio ll Of] 55.43(7) With no fuel assembly in the mast tube and a fuel assembly in the transfer cart, the operator is directed to close the fuel 1---"'-- [Answers:** ...... pool gate valve at step 3.13 AFTER starting the cart to the containment side. The discussion section of S2.0P-AB.FUEL-0002 (2.4.C) states that the gate valve is not to be closed until the fuel handling cart is clear of the gate valve path. It also states that while the transfer cart may stop prior to reaching containment, it is acceptable because the water around the fuel 'Nith shield and i "01 it , I I ~'FacilityRefE!rence Number ! ;Reference Section .**.~ [Page-t4 0 , I Revision l
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Reference Title .....
- i ILoss of Refueling Cavity or Spent Fuel Pool Lev II S2.0P-AB.FUEL-0002 II II 11 10 I I II II II II I II I
II II II II I iL.O;Number 1 I ABFUE2E002 1------'
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'Rl) $k~S~ta~'~li*(sri6'sk~~~per<1 .~-,~~.~ "~ . I 'Question Topic I SRO 16 Given the following conditions: - Unit 2 is performing a Rx startup. - Rx power is currently stable at 6%.
- 22 SGFP is supplying Main Feed to all S/G's.
- Steam dumps are controlling Tave in MS Pressure control- Auto at 980 psig.
- All MS10's are closed in AUTO at 1015 psig.
- 23TB40 fails 50% open.
- RCS Tavg lowers to 540.9°F and continues to lower.
Which of the choices completes the following statement?
,If RCS Tavg cannot be restored >541 within 15 minutes. the Rx trip breakers must be 0 ened within the next 15 minutes because ...
minimum shutdown margin cannot be assured. protective instrumentation may not operate when expected. calculations determining adequate DNBR are no longer valid. control rod worth is calculated only over the range of programmed Tavg. AO~~~EJ ~ ~~~~~=U IS I ~,~!tiVeLeveIJ IMemory 1iF~J ISalem 1 & 2 I ~E~ 1--------' KA:jI041000A202 11A2.02~-:~ ~~!J]"~ov-alU~~nl~ ~RO~~O'iSROGr()up~[]
~~tell11§volutio~TitleJ ISteam Dump System and Turbine Bypass Control AAStatemEmt:l ~.~~.~-
Ability to (a) predict the impacts of the following on the Steam Dump System and Turbine Bypass Control and (b) based on those i predictions. use procedures to correct, control, or mitigate the consequences of those abnormal operation: Steam valve stuck open I IExplan~tionofl 55.43(2) Tech Spec 3.1.1.4 states that the lowest RCS Tavg must be maintained greater than or equal to 547°F in Modes 1 and 2. i~swe~:~~J The stem states 6% power, so the unit is in Mode 1 and the TSAS is applicable. The action is to restore Tavg to 541 within 15 I minutes or be in Hot Standby in the next 15 minutes. The bases for TSAS 3.1.1.4 list 5 reasons why Tavg is required to be at this I temperature. One of them is ..."ensures the protective instrumentation is within its normal analyzed band". The control rod worth is i "I",,"ihlo if Ih", ""nriiri",l" h"Ii"" '0" Ih"t Ih., offo/'t ,.,f lom".,r"t. > nn ('('ml,,,r "vi M".-!n '" . 'c. shutdown. Minimum shutdown margin applies to control rod position and is not included in the bases for Minim~m temp for on..." ,,,h t,., ro,." ,ira" ',nit criticality. DN8R is the bases for maintaining parameters in 3.2.5 (PZR press, RCS Tavg, and ReS total flow rate) C'0/.**R;f~enceTiUe* . . . '*.i"~i ~~ F~cmtyRetE;;~nc;Nornb;8~~1 :Refef6rice'secti~~1J ~geN~l [Re~~ron; ISalem Tech Specs II IITSAS 3.1.1.4 113/41-6 811 1 IExcessive Steam Flow !IS2.0P-A8.STM-0001 Ie 11 16 11 9 I I II II II I ISTDUMPE016 I_ _ _ .........J
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Qu~stion Topic! SRO 17 I Given the following conditions: Salem Unit 2 is responding to a loss of all off-site power.
. 2B EDG is supplying 2B 4KV vital bus,
. 2A and 2C EDGs have tripped .
I 2A and 2C SECs have been deenergized.
- A C02 actuation occurs in 28 EDG room.
Which of the following identifies the effect the C02 actuation will have on 28 EDG operation, and what action(s) will be perfonmed? 2B EDG ... I Iwill automatically trip. Deenergize 2B SEC lAW 2-EOP-LOPA-1, Loss of All AC Power. I I
~:, will NOT automatically trip, Deenergize 2B SEC lAW S2.0P-SO.DGV-0001, Diesel Generator Area Ventilation Operation.
I Iwill automatically trip. Place the 2B Diesel Generator Supply Fans Emerg Bypass of C02 Shutdown Switch at 2RP5 in Emergency lAW 2 I EOP-LOPA-1. I lwill NOT automatically trip. Place the 2B Diesel Generator Supply Fans Emerg Bypass of C02 Shutdown Switch at 2RP5 in Emergency lAW S2.0P*SO.DGV-0001. I I Answerj Id I :ExamLevelJ IS I ~nitive\evelll Application IIt:'acility:] ISalem 1 & 2 I !EXamDate:ll 9/26/ 2011 1
'KA: iI064000A222 IrAz.22 - . ; IRQ Value: II 2.41 'SR()va1ue] 2.8*1 fSectlon;ll~~ ;RQGroup:JI 11 !SRQGrOu~~! 11 It'll ~
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System/Evolution Title I 1'064
--------~'-~-~ ~ .J ~.Statel11e~ Ability to (a) predict the impacts of the following on the Emergency Diesel Generators and (b) based on those predictions. use procedures to correct, control, or mitigate the consequences of those abnonmal operation: J Potential automatic safety sequences (water/C02)and electrical damage (loose wires)
! Expl.:matioIlOfl There is no automatic EDG trip on C02 actuation, either in the EDG control room or the EDG area. The SEC does not control the
- Answers: ' . 1 EDG Area/control room supply fan. The SEC deenergization of the unaffected SECs/EDGs in stem is to make the EDG trip
---.----~ plausible if the candidate thinks that the SEC would prevent an EDG trip, since there would be a standing Mode II SEC Signal present with 2 vital buses deenergized. Bypassing the C02 shutdown of the supply fans allows them to restart since their control
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Ithere is an exhaust fan for the EDG FO Storage area.
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IDiesel Generator Area Ventilation Operation II S2.0P-SO.DGV-0001 1\ II 115 II II Ir II~=::;II:====; 1:=1-=-=-='::=_='::_=_=_=_=_=~::::::::~~ili=:::=:'=:=:'=:'=:'::'::'~~:;I~[::'=~~~==-=_=_=:;ll_----lII=:~
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I Given the following conditions:
- Unit 2 is operating at 100% power.
- 25 SW pump is INOPERABLE for scheduled maintenance.
- The appropriate TSAS Tracking entry for TSAS 3.7.4 has been made in OP-SA-1 08-115-1 001, Operability Assessment and Equipment Control Program.
- 23 SW pump strainer motor fails.
Which choice identifies the action that is required to be performed and the Bases for that action? 1Exit the TSAS Tracking statement for 25 SW pump, and enter TSAS 3.7.4 as an ACTIVE entry due to 1 SW loop being INOPERABLE. I Exit the TSAS Tracking statement for 25 SW pump, and enter TSAS 3.7.4 as an ACTIVE entry due to 2 SW pumps being INOPERABLE. I Make a second, separate TSAS Tracking entry for 23 SW pump. Each 4KV vital bus requires only one OPERABLE SW pump to ensure full OPERABILITY of the SW system. I Make a second, separate TSAS Tracking entry for 23 SW pump. IF ANY other SW pump were to be declared INOPERABLE, THEN entry into active TSAS 3.7.4 is required since only one OPERABLE SW loop would remain. [Answe::~ Id 1 .Exam level, j S I ~gnitive~e"el;: Memory I I !FacilitY: II Salem 1 & 2 I 'ExamDaiel ! 9/26/20111 IKA:1076000G237 112 .2:37 :~O\laluE!J13.6Ill:)ROValue] 4.6Irsectiolill~:R()Grou~1 11:SRo~roup:'l 11 l~ ~
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.~.~.~- -~-~--:-:l KA Statement- I '-~_~. __ ~_*.. J Ability to determine operability andlor availability of safety related equipment.
lExplanation of i 55.43(2) Each SW pump must have its strainer operable for the SW pump to be operable. (SO.SW-005, page 97). With 2 SW iAhswer~-=-..:...."j pumps inoperable on different vital buses AND in different SW bays as described in stem, then BOTH SW loops remain operable. The requirement for 2 operable SW LOOPS are: One operable pump on A bus, One operable pump on B bus, One operable pump on C bus, and 2 operable pumps per bay. The requirement for ONE operable SW loop is 2 operable pumps powered from different
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. ., ~. . ,;romon" fror 'I operable SW loops. C is incorrect because the 'bases is Wrong. If there were only one pump per vital bus, then there would only be one operable SW loop. A and B are wrong because entry into an active Tech Spec is not required for 2 SW pumps inop on different vital buses in different SW bays.
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, Salem Tech Specs II 113.7.4 II II I IService Water System Operation Ii S2.0P-SO.SW-0005 II II 11 40 I
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Given the following conditions:
- Unit 1 is performing a Rx startup lAW S 1.0P-IO.ZZ-0003, Minimum Load to Hot Standby. - Core burn up is 8560 EFPH, - The Estimated Critical Condition calculation performed by BEACON predicts a critical rod height of Control Bank D at 61 steps. - When the ICRR value reaches the 0.125, the predicted rod height is 150 steps on Control Bank D.
Which of the following identifies the effect, if any, this will have on continuing the startup? 1Continue the startup with no additional action required. I I Continue the Rx startup and evaluate the post startup data for trend. I IInitiate Rapid Soration, insert control rod banks, and recalculate the ECC. i I Obtain permission from the Reactor Engineer and the Shift Operations Manager prior to continuing with the startup. J fAOswer i II::] iExam~eveillI:J ~itiVe L.evel-l ~tion I faciHty:ll Salem 1 & 2 I 'ExamDai~:ll 9/26/2011J
,1<A:II194001 G123 I~2.1.23-= iR9Value;i[~ ;SRO valu~:[I~}S~cticinj I£'~ fROGroup~ 0 lSRO Gtoup:1 0" iii; ~
rsysiemlEwl;;-@~~ntiE;'l I II~~E~] iKA Statement*:
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Ability to perform specific system and integrated plant procedures durin:a all modes of plant operation. I IExplanation of,! 55.43(6) IOP-3 contains the actions required when evaluating the deviation between the estimated critical rod position from the iAnswers:
'_~_~_'_.~J
'." ! ECC and the predicted critical rod position from the ICRR. The candidate will have to calculate the reactivity (in pcm) deviation based on estimated and predicted position, while choosing the correct table based on core burnup. They then have to apply that deviation to actions required for that deviation. For a deviation of 300-400 pcm, the startup may continue and post startup data I I """I".,H"" i" ro,",,,i.,,,r! 'Tho '< r!i, <oro "II .,,,H,.,,,,, i" IflD f ..... rliff",ront .. Tho IPIAI .,.t A1 do"" i<> 01 A a """" ".,inn Table 1-8 page 4 of 5 for burnup of 8533 EFPH. The predicted rod height of 150 is just below the 151 var"ue of 527 pem, so wOUld be Slightly CLOSER to 916.9, and within 300-400 pem.
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.h.0.Nu~~ .J IIOP003E007 ~Material REilquiredf()r~aminatiorid ISRO 19 S1.RE-RA.ZZ-0016 Curve Book ~:.:!i~n~~:!:iJ IPrevious 2 NRC Exams I~~~B~~~~~LJ Editorially Modified I ~p~r~rlI!J!~~!t~~gl~~ [J lQue~~_ns~0"c~C()m~~ 108- 01 NRC exam. Replaced one distracter. Correct answer remains the same. J I
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I ~~~~J~IS=R=O~20:ft+/-:t~==~~~~:f~~~:f~~==~~==~~~==~~=t:f~~~~~~~~~i Given the following Initial Conditions al 0800:
- Unit 2 is in MODE 5. - Containment Purge is in service.
At 0830, Containment Purge is stopped to support a maintenance visual inspection of purge ductwork inside containment. At 1015, a Containment Ventilation Isolation (CVI) signal inadvertently actuates due to a rad monitor switch misposilioning. At 1030: The duct inspection is complete. Radiation Protection reports containment was not breached in the past 2 hours except to allow entry and exit of workers through the airlock. Chemistry reports there is operational assurance that no radiological changes to containment environment have occurred and requests containment purge be restored. I lAW S2.0P-SO'wG-0006, Containment Pur e to Plant Vent, is a new containment pur e release form re uired and wh ? Yes, because of the CVI signal. lYes, because the purge was secured. No, because the termination was <4 hours and there is operational assurance of unchanged radiological conditions. No, because the termination was <12 hours (one shift) and containment has not been breached except as allowed for worker ingress/egress through containment airlocks.
- Answ~J E:J ,~I§l Level} IS I r<;ognitiveLeve,11 Memory I ~aci~~ll Salem 1 & 2 I ~a~~~ 1-------'
iKAl! 194001G132 I~~~_~ __ ~~ ~:Yi!~~Q]~~~~~~~I~ ~~O~~CiJ
-Syste;}J~~illIon.:TItle~
--~--
KA Statement"
~~-"-.~.--~. ~'-~;
IAbility to explain and appjy" all sy.stem limits and precautions. I ! ExPlanationOfj 55,43(4 )Procedure S2.0P-SO'wG-0006 allows for reinstatement of purge if of short duration (-4 hours) and containment rad I
'Answers: " i conditions have not changed ( P&L 3.3) Short duration is NOT 12 hours, and distracter does not address rad conditions. New
---~---~~ effluent permit not always required. Can block CVI signal lAW Att 2, Temporary Termination and Reinstatement of Containment Purge.
/ Reference Title ~7~1 rFacilityRef~eflceNumber .' ,,~ iRa!erence Section .' J GP;ge N6J fReviSiO~
,! Containment Purge to Plant Vent II S2.0P-SO'wG-0006 II 1\ i 130 I I II II II I II I
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IWASGASE011 1_ _ _---,
Given the following conditions:
- Unit 1 is operating at 100% power, Operators have just satisfactorily completed SC.OP-PT.OG-0001 DIESEL GENERATOR MANUAL BARRING on 1B EDG, The only Tech Spec entered was 3.8.1,1 ,b for the 1B EDG being declared INOPERABLE when the LOCKOUT SW was CIT in the LOCKOUT position.
when the 1B EDG will be declared OPERABLE? When the Acceptance Criteria in SC,OP-PT.OG-0001 is reviewed and signed off as SAT by the CRS. When the LOCKOUT SW is released to the IN SERVICE position after successful completion of SC.OP-PTDG-0001. When the NCO is directed by procedure to log the EDGs availability for operability testing In the Control Room Narrative Log. has successfully met the acceleration, voltage and frequency Acceptance Criteria in the ST procedure used for the retest, and to the Vital Bus.
~1194001 G223 IL2E:C= ~~~W'SRO~!(TI]~iJl~ ;RO'Gr:ooP:lu~~~D ~ystem/EV;;lutio". Title 1 I I ~ERI~
I KA Statement~ i
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- ExpJanatio~ ofl 55.43(2) The EDG becomes inoperable when the Lockout Switch is placed in the lockout position. The EOG remains inoperable
;Ans",,~r:.~_.,~l during the performance of the barring. It cannot be declared operable until it has proved it meets the requirements for an operable EDG, since it has been affected by the barring operation. The requirements are successfully reaching rated speed, voltage and frequency requirements. AND being synchronized to the vital bus.
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Reference Number: ,Refererlce.Section.j [Page No. i Revision:
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iQuestionTopic i ISRO 22 i A Hope Creek Station employee has received 1900 mrem routine TEDE for the current calendar year, ALL at Hope Creek Station. The employee is expected to receive an additional dose of 450 mrem on his current job assignment at SALEM. His lifetime exposure is 5500 mrem. lAW RP-AA-203, Exposure Control and Authorization, and prior to performing the job, written approval for increasing his dose limit to 3000 mrem TEDE for the calendar year must be received from the work group supervisor and ... Ithe RP Manager ONLY. I Ithe Site Vice president ONLY. I Ithe RP Manager and Station Manager. I Ithe Station Manager and Site Vice President r AnSW~ Ia I ::ExamL~veIJ IS I !£o~l'litiveLevel j IMemory j :Facility: ] ISalem 1 & 2 I !ExamDate: j , 9/26/2011 i
~J1194001G304 1'2.3A_~__~~ l~(l~'!l~1 3.21 §!~~Val~~ 3.71 iSecti~1 PWG I!~dGr9upll 1! 'SRO GrotPil[-1l Iftl 52J
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- ~st~~~eritJr_-----------------------------------------------------~
Knowledge of radiation exposure limits under nonmal or emergency conditions. rExplartation of j 55.43(4) A is correct because the approval requirements are: Up to 3,000 mrem- RP Manager; up to 4,000 mrem- RP Manager and
- Ar!~e~~~J Station Manager, >4,000 Site Vice President Distracters are all some form of combo of each of the positions.
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II~-=-=-=-=-=-=-=-=-=-=-=-=-=-=-=-=-=-=- . . . ."II______--'IIII ____--'II_-..l1 1\ IIi~==:; (::-~~-~~Tj Ll::9..2.c!-l~~~~ I RADCONE002 I I i
! I II e~:~~~nSo~!~~iJ IPrevious 2 NRC Exams I~~r:!i~~~iO~~~i:.11 Direct From Source iQU6;rtiooso~~co~_~etrtS11 08-01 Salem ILOT SRO Exam May 2010 I
I I
Given the following conditions: Unit 2 is operating at 100% power.
- The operating crew entered S2.0P-AB.RC-0002, High Activity in Reactor Coolant System, when RMS channel2R31, Letdown Line Monitor, went into WARNING.
Which of the following is required to be performed lAW S2.0P-AB.RC-0002, and why? The will.. direct Radiation Protection Technician to survey the letdown line in the vicinity of 2R31 to confirm the suspected rise in RCS activity. direct Radiation Protection Technician to take surveys to determine if radiation levels may have changed access requirements, Technician to initiate confirmatory sample analysis because the 2R31 reads in CPM and therefore has no correlation to dose direct Chemistry Technician to sample hourly for isotopic analysis to determine predominant radiation hazard (gamma, neutron, beta, alpha).
~! 194001G314 !~i14~--' iRova[uelQ]~~~~~l2.:.~ ~31~ ~Groiij)~ufSROGr~D ~ys~;m[E~~~tio~~tleJ !____________________________________________________________________________ ~ ~-.-.-'~
c~State~en~r_------------------------------------------------.--------------------------__________~ Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.
;EXplanation ofl 55.43(4) B is correct as described in S2.0P-AB.RC-002 basis, so that prompt identification and subsequent notification of plant iAnswers~ i personnel is ensured.
_ . _ - - '--~ A is incorrect because chemistry sampling confirms 2R31 readings, not survey results. C is incorrect because rising counts does indicate dose level changes. 0 is incorrect because the hourly isotopic analysis performed is for gamma to determine DEI for I '.onrHn~ (~"'n 'l HI\ I,~.'--c'~-:-c*~~-~~~-~:-'~:_:~~~-~~~,C7'~~~7:_:_;:"7"-.. -.,. ~~'~-.. .-'-~, ' - - .- . ~'~'." ., ~-'--~~~~ -,~~,.--~-~ -~--~.~- I F': H=ig='h: A: 'c: ti: ':i=~: i: n= : : :~e: ;=: :~c: Ce: O;: I~: t~: e;: s=,y=s: te: 'm: ' t=s=~: .: : =~ : (i: : :.: :;: ~: :o: r~=~: c=,e=t:'=u=,m=b=e=!,=. IFlR=e=~~=r=en=.c=e=s=e=ct=io=n=~.~.=~.'::;! 1\=~p=a=g: e: N: O:;i.
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IABRC02E003
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~ue~~~TopBJ!SR024 I Given the following conditions:
. Unit 2 is in Mode 4 performing a plant cooldown .
. 21 RHR pump and loop is supplying Shutdown Cooling .
22 RHR pump is aligned for ECCS.
- A fire is reported in Unit 2 Relay Room.
- The Operations Department Fire Department Liason who was dispatched to the scene verifies an actual fire is present.
Which of the following~ describes the required action which must be j)erformed? i IPlace both PZR PORVs in manual and shut, and shut their Block Valves when directed in S2.0P-AB.FIRE-0001, Control Room Fire Response. I
~:.J Place both PZR PORVs in manual and shut, and shut their Block Valves when directed in S2.0P-AB.FIRE-0002, Fire Damage Mitigation.
I IStop 21 RHR pump when directed in S2.0P-AB.FIRE-0001. Control Room Fire Response. I
!Stop 21 RHR pump when directed In S2.0P-AB.FIRE-0002, Fire Damage Mitigation.,
I
- Answer i a I I ~am' Levelj IS I ;Cognitive ~ve!ll Memory ! :Facility;! ISalem 1 & 2 I iExamDat~: II 9/26/20111 KAJI194001G427 F ~
, ,2.4.27
~
-=,Rovaf~13.4I~civalue:m:J'sectiOil:llpWG I'ROGro~1 111SROGrol.lp:ll 11 I'll 2J
~ '-----1 KA Statement:!
~ I~K-n-o-wi-e-d-g-e-o-f-"fi-Ir-e-in--th-e-p-Ia-n-t-"-pr-o-ce-d-u-r-e-s------------------------------------------------------------~
I,~.~~ I 'Explanation of' I . --.
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. .. -.J 55.43(5) AB-FIRE-1provideds the actions necessary to minimize the consequences of a fire from impacting plant operations and I
- A.nswers~ support equipment. AB.FIRE-2 gets the Rx and MT tripped and all RCPs stopped, initiates actions to achieve hot standby, and then cold shutdown. The stopping of the inservice RHR pump is incorrect but plausible since it would be performed if the fire was in the ,
RHR pump area in AS.FIRE-1. This eliminates the potential for a scenario whereby the spurious operation of the SJ44 valves (due it" '" fi.",_inrh 'I"",rl <:ht'!rt\ \/')! ,Irl nr,wirl", " n",thw<I" fr.r " I,.,,,,, "f I::U,<:: . ~ '" .mn ,..., tn th", .
. . <IrQ rlir",,-.torl t,.,
close the PORVs in manual and shut the block valves because spurious operation of either PORV could occur priorto any actual control room evacuation decision or prior to the operations crew reaching the Control Room Evacuation step to operate the transfer switches for the block valves. I:~ .~ .-. _~. ~ ~ ~'~.Referenc~Title ' :.-;i' "I", **.. Facility Reference Nlllntier '.1 [Reference seCtion
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'i Control Room Fire Response II S2.0P-AB.FIRE-0001 II II 117 I IFire Damage Mitigation II S2.0P-AB.FIRE-0002 II 'I \14 I IiI II II II II I
~O.Number IABFP1E003
~Materiaf Required@r ExaminatIon .~~ I 1/
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Given the following conditions:
- Salem Unit 2 is experiencing an event which has resulted in an ALE,=n EAL being exceeded and reccgnized by the operating crew, After 5 minutes, the actions taken by the control room crew result in this ALERT EAl no longer being met.
- Conditions for an UNUSUAL EVENT are present.
- No emergency declaration has been made yet.
Which of the following identifies the actions required by the Emergency Coordinator?
.\ Declare an ALERT, terminate the ALERT, declare an UNUSUAL EVENT.
I I [b~ Declare an ALERT, then reduce the Emergency Classification to an UNUSUAL EVENT. r I Declare an UNUSUAL EVENT, and make a non-emergency After-The-Fact 1 hour report to document the ALERT EAL. ,
. .' J g:~~a~~:;t.UNUSUAL EVENT, andensure the ALERT EAL which was present for the 5 minutes is communicated to the NRC via the NRC I II:Ans~e~ EJ ~amT;VeI:] E:J LCogniti~e Le~ ~ry I iFaCilitYll Salem 1 & 2 I ,ExamDate:ll 9/26/20111 ~Kk]1194001G438 12.4.38 ~[Rovatue:l[EFsRova'u~~SectirinJ~fROGr~U~SROGrOu~DJ l~~ ~
KA Statement: 1
~--~-'-~-.::
Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency ccordinator if required. I iExplanation of i 55.43(5,1) Per the Salem ECG, Section I, page 8 of 10, Section IV, Event Classification Guide (ECG) Use, Subsection 0.2,
- ~iSwer~=--~~1 ... "Short duration events that occur will be assessed and emergency classifications made, if appropriate, within about 15 minutes of control room indications or the receipt of the information, indicating that an EAL, has or had been exceeded. This classification is to I
be made even if no EAL's are currently being exceeded, (Le. actions have been taken to stabilize the Plant such that no EAL's I ! ...,
". "ronth, """,Ii<>" \
<::,.., "",..,,,,ril,,,,,,, ,..,f th", hr't Ih'" tho AI I=QT 1= AI "'.., Inn,.,,,,, """,ii"" th"" AI I=RT "" ,,,t ",HI! h"" riorl",&>ri EP.ZZ-0405, EMERGENCY TERMINATION - REDUCTION - RECOVERY, describes how to reduce the Emergency Classification if "II" I=P.
the emergency is not being terminated. Since the UE is still being met, the classification will be reduced. not terminated.
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