ML113330798

Salem, Units 1 & 2 - Draft Written Exam (Folder 2)

ML113330798
Person / Time
Site:	Salem
Issue date:	08/03/2011
From:	Public Service Enterprise Group
To:	Operations Branch I
JACKSON D E RGN-I/DRS/OB/610-337-5306
Shared Package
ML110030673	List: IR 05000272/2011301 ML111610244 ML112590037 ML112650617 ML113330710 ML113330734 ML113330777 ML113330798 ML113330805 ML113340036 ML113340040 ML113340044 ML113340065 ML113500280 ML12167A076
References
TAC U0183
Download: ML113330798 (100)

Text

{{#Wiki_filter:d*;, .' .... ........-. ............... . .. --..-..-.--.. ...... ...... ... ,...In"<:lstionTopic J I RO 1 A dropped rod while critical on Unit 2 has resulted in ReS Tavg lowering to 540.9"F, and the Rx remains critical. Which of the following choices contains ONLY reasons why Tavg must be raised to at least 541' for the Rx to remain critical lAW Salem Raising Tavg to at least 541°F ensures ... 1. MTC is within its analyzed band. 2. protective instrumentation is within its normal operating range. 3. the P-12 interlock is above its allowable setpoint.

4. minimum shutdown margin is maintained.

5. that Heat Flux Hot Channel Factor does not exceed its limits in the lower region of the core. the PZR is able to be operable with a bubble

[J 11 I 11 4,6. 1 2 ,3,6. 1 2 ,4,5. I R 1 Memory I I Salem 1 & 2 IIExamDate: I'jj 9/26/2011 000003K307 IIAK3.07--=: _____________________________ ---1 Tech-Spec limits for T-ave 55.41 (5, 10) The Tech Spec Minimum Temperature for Criticality limit of atleast 541°F (TSAS 3.1.1.4) Bases states that: 1. the moderator temperature coefficient is within its analyzed band, 2-the protective instrumentation is within its normal operating range, 3-the P-12 interlock is above it allowable setpoint, 4-the pressurizer is capable of being in an operable status with a steam bubble, and 5-the reactor pressure vessel is above its minimum RTNDT temperature. Of the 2 incorrect choices, #4 is plausible if the '.. . would tend to create higher power lower in' the core, but Fq-(Z) is a condition-which is kept in its allowable range by maintaining I acceptable power distribution limits as described in TSAS 3.2.2. [Am 197 I I I-__________ ---.JI_-,- ______--'I______---'i__--' I I ..... IABROD2E002 I___---l

SROSkvscraperl'

..* < Topic! Given the following -Unit 2 is operating at 100% -21 SGFP trips, and a manual Rx trip is initiated based on lowering SG -The Rx operator confirms the Rx trip during performance of the immediate actions EOP* TRIP-1, Rx Trip or Safety While performing the RCS Temperature Control section of EOP-TRIP-2. Rx Response, the RO reports Tavg is -5510 and Which of tf1l1nwlnr1 describes is -551° F? Reactor trip breaker position 55.41 (5,6,7) A manual Rx trip is initiated by the Rx Trip Handles. This sends a signal to Solid State Protection System Trains 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 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 " *** ***** 0. programmed Tavg of 547 at no load. A is incorrect because the load reject would arm the steam dumps, and B P-4 would place them in Plant Trip Mode and control at 547". C and D are incorrect because steam dumps would have to be in MS Pressure Control mode for the setpoin! on the control console to be the driving system Signal. Both C and D are plausible if candidate does not know normal mode of steam dump operation at ower, or function of Tav Control. L.Q.ri,jumber i' I TRP002E005 . J I I STDUMPE008 I I STDUMPE007 I ROSl<vScraper: 'I .SROSkvscraper J RO SvstemlEvolutian List, 'Question I RO 3 Given the following conditions: -Unit 1 is operating at 100% -Spent Fuel Pool fuel moves are being performed lAW S1.0P-IO.ZZ-0010, Spent Pool -PZR spray demand -PZR level is rising very -The CRS enters S1.0P-AB.PZR-0001, Pressurizer Pressure -The crew identifies 1 PR2, PZR Power Operated Relief Valve, is Of the following choices, which is REQUIRED to be performed FIRST lAW Salem Tech Specs? Assume 1 PR2 cannot be restored to an OPERABLE condition. Within one hour close 1 PR7, PORV Block Valve, and maintain power to the valve. 1 PR7, PORV Block Valve, and remove power from the valve. I R I 9/26/20111 .--------:1[2.2.39 IRQ \/alue: GrouiillOJ OJ I Pressurizer Vapor Space Accident lJS!.. ,...-_______________________________________ ---; Knowled e of less than or e ual to one hour Technical Specification action statements for s stems. 55.41(10) A is incorrect because there is no requirement to stop moving fuel for this condition, nor is there a general caveat that says stop moving fuel for any Abnormal Procedure entry. B is incorrect but plausible, since it is the 3.0.3 action when there is no specific TSAS which applies, and fits with the other one hour requirements. C is correct. and D inconrect, because power is maintained to the block valve lAW TSAS 3.4.3 action a for excessive seat leakage, whereas power is removed from the block valve ____________________ __ Outline Chanties-[questi;n Topicl IR04 r Given the following conditions: -Unit 2 is operating at 100% -23 AFW pump is -A 1,000 gpm LOCA -The Rx is tripped and a Safety Injection -NO AFW pumps automatically Assuming that both MDAFW pumps could be manually started if demanded, which cf the following describes the MINIMUM required AFW status, and the reason wht that minimum is required lAW 2-EOP-LOCA-1, Loss of Reactor lONE MDAFW pump must be started to ensure a secondary heat sink is available. I I TWO MDAFW pumps must be started to ensure a secondary heat sink is available. I I.ONE MDAFW pump must be started to provide a static head of water to prevent primary to secondary leakage. I @.J ITWO MDAFW pumps must be started to provide a static head of water to prevent primary to secondary leakage. I IAnSlwerl1 a I li5xall1il..exelJ I R 1 I 1 Salem 1 & 2 I I§xaIllDate:! 1 9/26/2011 [ 000009K328 I 4.51 I 4.5[ EPE IlB9GrpUp:i 1 1[ 0 ". I Small Break LOCA Manual ESFAS initiation requirements 55.41(4,10) A is correct because one MDAFW pump will provide at least 22E4 Ibm/hr flow which is the minimum required to satisfy the Heat Sink CFST until at least one SG NR level is restored. Additionally, for a small to intermediate sized LOCA, this flow is required to ensure a secondary heat sink is available. The distracters all contain either 2 pumps or the incorrect reason, or both. The distracters which refer to the static head of water is the basis for a LBLOCA, not a SBLOCA. The distinction must be made that , , . . I plausible because two-AFW pumps are required-during FRSM when >44E4Ibm/hr AFW flow is required. 28 =:::;1=1 .....*...* ..I LOCA01 E009 !MaterialRequiredfor ExalTlination I I II rQuesti()n I Facility Exam Bank i Mopift5ation ] Concept Used I Training pr()grarrtJ

J

[Question sourcecormnentsJ' I Vision Q80803 Used the concept of "why" AFW flow is required, and added "how much" and made it an operational I . .----...--type question instead of a simple "Why is ...... required?" lComment . ' .. ................ ' I I I I ! i RO ..* SRO Skvscraperl 'JRO

• • . J;:CSRO

'Cha;lQes I RO 5 validated conditions will, by itself, require tripping the affected RCP(s) lAW ALL charging pumps out of service for >5 minutes. OHA D-22, 13 RCP BRG CLG WTR FLO LO, received 2 minutes ago. I Memory I Salem 1 & 2 9/26/20111 D 55.43(3)C is incorrect because the loss of charging pumps, and hence seal injecticn flow, is not a RCP trip criteria, since thermal barrier flow is still established. AS.CVC 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 Alt. 1 of AB.RCP. D is incorrect because the ARP for D-22 does not direct entry into AS.RCP unless accompanied by rising temp. B is incorrect because the setpoint for high seal outlet temp is o I---------------------- _________________________________ -"I____________________ I ABRCP1 E003 I

• .

.*. O:::d:..:.:::: ..

• .

IAdded "without delay" to stem to ensure choice c is incorrect. ! .. ' ........ ............. .... I 1 I I if RCP seal inlet temperature rises to 225 To prevent runout of a charging pump when a charging pump is The RCP seals and shafts may be damaged due to thermal shock when a charging pump To prevent steam binding of the charging pumps due to RCP sealleakoff flashing to steam when a charging pump is The RCP thermal barrier heat exchangers may rupture due to thermal shock and water hammer when a charging pump is I Salem 1 & 2 concepts as the apply to Loss of Reactor Coolant Makeup: Consequences of thermal shock to RCP seals 55.41(10.7) AS.CVC provides guidance in Att 3 and CAS 5.0 to isolate Rep seal inection if any seal inlet temp is greaterthan ,Answers+/-;*;):i.'J equal to 225°F. This guidance is consistent with the guidance contained in LOPA-1 basis document. page 32. which ----. "Isolating the RCP seal injection lines prepares the plant for recovery while protecting the RCPs from seal and shaft damage may occur when a charging pump Is started as part of the recovery. With the RCP seal lines isolated. a centrifugal charging * ** ., *

• h * ... ..... .." but plausible since starting the first pump in a system has potential for runout after i..igh points have drained back into system. Cis incorrect but plausible since the high temperature in the sea! area on an extended loss of seal cooling may cause flashing in the seal area. and the gas could be transported back to evcs systern. but is not why the seals are isolated.

0 is incorrect but plausible because the CCW thermal barrier return is isolated for this reason during a LOPA. IABCVC1E003 I I LOPAOOE003 I I 1 .. " I RO 7 I Given the following conditions: -Unit 2 is in MODE 4 exiting a forced -The reactor has been shutdown for 96 -23 RCP is in -21 RHR loop is in service supplying shutdown -22 RHR loop is aligned for -RHR HX inlet temperature is 290°F and -RCS pressure is 325 psig and -ALL MS-10's are set for 200 psig in 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 Assume only heat added to RGS is decay References provided. [EJ 1< 15 minutes. I [EJ 1 15-19 minutes. I 11M 120-25 minutes. I @J I> 25 minutes. fO' !Exl'!m Level! I Application 1000025K101 [EX, ti" ' ,Answers: , 55.41(10)The HUR will be determined using Attachment 5 of AB.RHR-1, page 2, with PZR level (Reps are running). The 96 hour 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 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, . _ ., 2 mjolltes If pace 3 HIIR is used then the H11e ,Mould 1 * *

• 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 would use -0.31° which would give over 3 hours. Modified distracters to make all plausible.

Provided Attachment 4 in addition to correct Attachment 5 to make question harder. . ROSlivScraper 11.:...;;:...:.- _________________________________________ ---1 Given the following conditions: Unit 2 is in MODE 3. NOP. NOT. returning from a refueling 21 charging pump is in -ALL CCW pumps trip due to a faulty electrical relay which was replaced on all CCW during the Which of the following identifies an action that must be performed lAW S2.0P-AB.CC-0001, Component Cooling Abnormality, and the reason it must Isolate letdown and swap charging pump suction to the RWST to prevent Place 23 charging pump in service since it is cooled by SW and both centrifugal charging pumps are cooled by CCW. Initiate Surge Tank makeup as level drops due to system pressure decay to ensure air is not introduced to CCW pump suction. I Memory t!Q>.:11000026G132 I j Loss of Component Cooling Water ____________________________________________________________________________to explain and I all system limits and recautions. 55.41 (4.7.10) A is incorrect because while RCPs will be stopped, it is because them is no bearing cooling flow. The seal package is still receiving seal injection flow from charging system. B 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 ultimately results in the loss of RCP seal injection, and failure of the Reactor . . PDP has CCW cooling. so no pump swap would be requireid in 'this case, but would be if the PDP (23) was in service. 0 is incorrect because surge tank M/U is initiated if a leak is present, not for a loss of all the pumps. The stem states the pumps tripped from an electrical fault, not because a system leak made them trip. I I New IIQuestion Mv", .,.vau" Method: .'I Comments 1 I .......... :.Comment .' I I I I ... RO SRd 1.. ., '" """" ., ... __ _______________________________________________________________________________________Given the following conditions: Unit 1 is operating at 100% power. -13 Main Turbine Governor Valve fails The PZR Master Pressure Controller output fails as is at 2235 psig before any to the Governor Valve closure Which of the following will occur naturally in the Pressurizer to help limit the magnitude of the resultin res sure transient on the prima system? An outsurge cools the PZR. This allows some steam to condense to water and limits the resulting pressure increase in the RCS. An insurge of hotter water heats the PZR. More liquid then flashes to steam helping to limit the resulting pressure drop in the RCS. An outs urge causes the steam space to expand in the PZR. This allows some liquid to flash t*:J steam and limits the resulting pressure drop in the RCS. An insurge of cooler water compresses the steam space In the PZR. Steam is condensed to water helping to limit the overall pressure increase in the RCS. I Memory Salem 1 & 2 [!<A: II 000027A 103 I***** I Pressurizer Pressure Control Malfunction a pi to Pressurizer Pressure Contro! Malfunction: .. ,p.*..*....... 0 ..... :.*.* *..*****.55.41( ) A is incorrect because the outsurge cause lower pressure in the PZR, which causes more flashing to restore pressure. B I IAnswers:. ..... il is incorrect because an outsurge 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. 0 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 . . . IABPZR1E001 SROSkvsiraper I RO 10 Given the following conditions: -Unit 1 is operating at 100% power. 13 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. 13 evcs Pum START, START, START. IAnswerl D lKA:l I 000029A204 I Anticipated Transient Without Scram 55.41(10) D is correct because FRSM-1 has operator start 11 and 12 charging pumps. 13 charging pump is normally operating. All three pumps will indicate START. 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 and tllinks the procedure just needs at least one centrifugal charging pump in operation. C is plausible if the candidate thinks the procedure directs starting of either 11 AND 12 . !jres stoo0100 13 cbamino ollmo Stopping the operatiog . . .I stoppedin other EOPs <imd ASs." --_. . iReVisiqil!

1____________

-<1_________---'1______---11 __-' I-_-l iMaterialRequired for II IJ

• • .

[J..* [9uestion sou.rce commentS] I .. . .. I I II I I I ROSkvsc:raoer] I .... RO list ,. Outline ChaoQes L __.*___*__**_' Given the following conditions: -Salem Unit 2 is performing a Rx startup by control -The reactor is critical at 3000 cps in the source -Control rods are withdrawn to raise -No additional operator action is taken and a reactor trip eventually 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 becomes longer. sooner. The recovery period of the instrument following the rod pull becomes shorter. sooner. The gamma level is added to the neutron level since it is not being discriminated out later. The gamma level is deducted from the neutron level since it is not being discriminated out. Iy to loss of Source Range Nuclear Instrumentation: 55.41(1,6) SR discriminator "screens out" gamma radiation at low power levels because it is not proportional to Rx power. If this 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. 0 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 . . ". . . , . ---.. ... The 8F3 Proportional Counter used in the SR produces pulses proportional to the i3nergy of the ionization reaction in the chamber. A neutron will react with the 810 in the BF3 gas, to produce 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 noise pulses. too. 1 9 RO

• • • t
• lrRO 1 Given the following conditions; Unit 1 is performing a unit

-Reactor power is 11 -The low power trips have NOT been Intermediate Range (IR) Channel I N35 previously failed and was removed service lAW S1.0P-SO.RPS-0001, Nuclear Instrumentation Channel Trip / Channel II N36 fails. The reactor will NOT trip, 1 N36 indication will drop to zero. The reactor will NOT trip, but OHA F-25 SR FLUX HI will annunciate. Both Source Range channels will automatically energize, and the reactor will trip on high SR flux. I000033A201 1 iAA2.01--! 1 Loss of Intermediate Range Nuclear Instrumentation Abili to determine and inter ret the followin as the a pi to Loss of Intermediate Range Nuclear Instrumentation: Ecuivalency between source-ran e, intermediate-ran e, and ower-ran e channel readings 55.41(7) This is a valid KIA match because the candidate is required to understanc the relationship between where power is in the Answers:i<).,*;:*1 power range, the level at which automatic re-energization of the SR channels would occur from the IRNl's failing low and/or having 1** _______ 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. , ., .. . . .... .... .... will read 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 automatic energization of P-lO permissive at 2/4 PRNI at 10% will already 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. I____________ F=:::; 1_11_--, I EXCOREE007 1 I EXCOREE009 1 I EXCOREE01 0 1 RO SkyScraper I..SRO I ;';.. I* OutlinifChahaes I :Questhm Topic II RO 13 the following identifies why the ReS is depressurized during the response to a SOTR lAW Allow more rapid boron injection and refill the Allow more rapid boron injection and minimize Terminate primary-to-secondary leakage and refill the Terminate primary-la-secondary leakage and reduced secondary planl ;.=[27=:::; iL(j;;:Number,::: ji c' I SGTR01 E007 I-----J forEXaminatiqn j ..

• • ..

..*=se=d=* ... Sourc: I , ,. "j I I I II RO SkyScraper ISROSkVscraper '1 RO list:.' *.****SRO '.1** Outline Chimqes I I i r--***--._ .... -Cluestion T?Pic:J RO 14 Which of the following identifies an automatic action and its correct initiating coincidence which will o,::;cur if a Main Steam piping rupture occurs at Steamline Isolation. 1/2 High Steam Flow on 2/4 SGs, with 2/4 RCS loops 2/3 detectors on 1/4 SG 100 psig lower than corresponding detectors on remaining SGs. 2/3 detectors on 1/4 SG 100 psig lower than corresponding detectors on 1/3 remaining SGs. Knowied e of the interrelations between Steam Line Ru ture and the followin : 55.41(7) Main Steamline Isolation occurs as described in A. B is incorrect because the High steam Flow coincidence is 1/2, not 2/4. C is incorrect because there will be no Safety Injection on SG DIP because th'3 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. I--------------------------- ______ I_______________________ _____ [Lo.

• • .... 1 I FLUNCYE002 I I I ) I .MCiltel-laIRe9uin.icl . J I II

..

• .*

.*..*'*=et=h:::o=d=,::: .................... '. ... I iComment i ! I i I I Which of the following describes the reason Main Steam dum ps will be blocked from opening if Main Condenser vacuum degrades to 20" Hg? To prevent damaging the condenser on overpressure. To prevent further degradation of condenser vacuum. To prevent exceeding condenser design temperature. To prevent boiling circulating water which can damage condenser tubes. Loss of steam dum p ca abilit upon loss Of condenser vacuum 55.41(4) All of the distracters are things that can occur on a loss of vacuum, but am not the reason for having Control Grade Interlock C-9 blocking sleam dump operation. The condensers are not designed to operate at any pressure, and steam flow from the steam dump system (as well as a Main Turbine trip) occur at 20"Hg decreasing, since at this level it is clear that there is something significantly wrong with the condenser and it must be removed from service. I rL;9:Number "',0 I ABCONDE004 fot @amination, I J I Other Facility I Editorially Modified I Seabrook May 2003 NRC Exam iel""'" ....... '.. ."'.'..",>. ...!::' " ...., I I I I I

• .. ROSkV$Craoerl
• .SRO.

J ....... j:1 *..*SRO List ... J.. ________________________________________________ -_______________________________Given the following conditions: -Unit 2 is operating at 35% power with both SGFPs in -BOTH SGFPs trip Which of the following describes the actions required in S2,OP-AB,CN-0001, Main Feedwater/CondElnsate S stem Abnormalit , and wh ? Trip the Rx and start all AFW pumps to maximize SG inventory. Trip the Main Turbine and start all AFW pumps to maximize SG inventory. Trip the Rx to ensure only decay heat and RCP heat are being added to the RCS. Trip the Main Turbine to ensure only decay heat and RCP heat are being added to the RCS. Reactor and/or turbine trl ,manual and automatic 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 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 fBed, the rx is tripped to ensure the only heat being 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 . .. . . . . I not possible nor reguired, since the 'procedu're knows they will already be running. :==:===:: If=:=::::::!, ILO,Number' .*.*. I ABCN01 E003 I I I I I ;lIJIaterial Requlredfor Examll1ation J I II Question SoUrce:

• .

I.QuestionSo,::rce I I jComment ... .....**.. .. .. . ...*. ..... .1 I I RO L*', Ust<1 List I RO 17 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 res ective SEC, and Chilled Water Pump s will be operatin to operate and I or monitor the HVAC chill water pum and 55.41(7) The SECs will shed then reclose the chiller 460 volt breakers on a MODE II (Blackout) SEC initiation following the EDG start. The SEC will close the Chiller 460V breakers and start both Chilled Water P'Jmps 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. 1_6____ I CHLWATE008 f()fl?'amination '.. ! I !l I ,Source: ..

• • • • .

D, I I .............. ..........* i Com , .. , .. ,,*. .'. I I I I II I , ... ..... - I ;.Cluestion Topic RO I Given the following conditions: -Unit 2 is operating at 100% -The 125 VDC control power normal supply breaker to the 2E 4KV Group Bus -Prior to any action being taken in response to this breaker trip, a 200 gpm LOCA i -Operators respond lAW the EOP network. Which of the followin9 describes how this 125 VDC breaker failure will affect the mitigation of the LOCA? Operators will have to locally trip 22 RCP breaker at its cubicle to stop forced RCS flow when RCS pressure lowers to 1350 psig in EOP-TRIP-l 1 Rx Trip or Safety Injection, , I.operators will not be able to establish/maintain saturated PZR conditions if required in EOP-LOCA-2, Post LOCA Cooldown I loperators will be required to adjust AFW flow to 22 SG lower than that supplied to the remaining 3 SGs in EOP-TRIP-1. I ,__" I.operators be able to restore normal charging and letdown in EOP-LOCA-2. I ." l@]1000058G406 !_L_O_SS_O_fD_C_p_o_w_e_r ________________________________ --' __.,---' [MStaternent'J ______'_M'; Knowledge of EOP mitigation strategies. i55.41(10,4) Control power for operation of 4KV busses is supplied from the 125VDC system. A normal and emergency supply are ,Answers:" provided, and must be manually transferred when the emergency supply is required. The 125VDC control power supplies power to the TRI P coils of the 4KV breakers, and the breaker cannot trip without that power. However, the 4KV group buses are supplied from the APT (Main Generator output) wihen the unit is operating at 100% power. The same lack of control power will prevent the .dK'\I t.:r,." '" 1"., ,e 'nfeu."; hro"l.-or fr"", Inn",,,,.,.. ... \Jlliln th,., h."" 1.-"",. '",n, ,I ,,., nn\MA. e, onnh'.. ,. nrc"l",.r fr,,;'" <:btirm P"",or "nn"t. shut, and the bus \Nill become deenergized. This will cause 22 RCP to have no power, even though its 4KV breaker remains shut. Loss of forced flow in a RC loop cause less heat transfer in that loop, requiring less AFW flow to supply less steam flow. Distracter A is incorrect because forced flow has already been lost in 22 RC loop AND ReS pressure will not lower to 1350 requiring all RCPs to be tripped. B is incorrect because PZR heater busses are powered from E AND G 460V buses, and 1 group of backup heaters remains available if required for PZR heatup to saturation. D is incorrect because letdown is not restored in LOCA-2, although normal charging is. This is Elausible if the correlation between PZR heaters/leveilletdown is confused. tL.iI FacilitYReferel1cet-,ll.lrnbe(; ... ,.* 12 Unit 125 VDC One Line II I 1 33 I II I Rx Trip or Safety Injection II EOP-TRIP-1 -.JI I 127 I I Post LOCA Cooldown and Depressurization II EOP-LOCA-2 II II II I 1 25 I (RO SkyScraper I i.. SROSk\r$craper I i SRO q T;'§J 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 Waste Tank 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 Building Ventilation None -the WGDT relief valves discharge to the Auxiliary Building Ventilation exhaust fan suction plenum. 000060K202 \Accidental Gaseous Radwaste Release lMStateli]eijtl,j 55,41 (13) The WGDT relief valves (205340 sheet 2, E-7) combine into a single header and go to (205322 sheet 1, G-3). which is the FHB Ventilation exhaust downstream of the exhaust fans. It then flows 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 (8), (C). (D) Flowpath is not through filters nor _____--J 1.=.;;..;;.;;..;;.;....;. ______---11______-' 1___...1 1_4°_---1 !LQ.,.Nurnber '! .1 ABRAD1 I I I I 1 [IyIaterial '1 e quired fo r 5 xa ir!ination . I I[ 1 Facility Exam Bank I'QuestioI1Modifi.citionMethod:1 Direct From S.ource I Duril'lg TrainingProgral11 , 0 jQuestionsource I . .....IC"............ .. II I II I II I

.SRQ. Skvscraper.], ROSvstem/Evolut!onListJ

. I... [Question I RO 20 Given the following conditions: -Unit 2 is operating at 65% -21 Charging pump is in

1. 4 SW Bay has been isolated due to a bay leak lAW S2.0P-AB.SW-0003, Service Bay -All #2 SW Bay pumps Which of the following describes an action that will be performed in S2.0P-AB.SW-0005, Loss of All Service Water, and wh ? Isolate CVCS letdown. This prevents resin damage in the CVCS demineralizers when letdown temperature exceeds 136*. Isolate CVCS letdown. This prevents heating of the fluid in the VCT to a temperature which will cause a loss of NPSH to the charging pumps. Swap charging to 23 charging pump if it is available.

This reduces the heat load on the CCW system and extends the time available to a controlled shutdown of the to the RCPs, which allows sufficient time to install >225*. i KA:'II 000062A206 I Loss of Nuclear Service Water 55.41 (10, 6,7) A and B are incorrect because while letdown is isolated, it is to m inirnize the heat input to the CCW system to prolong cooling to the Thermal Barriers and Rep seals. The CVCS demineralizers are automatically bypassed on high temperature at 1360F. C is incorrect because a swap to the PDP is made since it is cooled by CCW, 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 maintained by'the PDP until temporarl cooling hoses to CCP's can be installed from the DM system. Lq':Number ..... \I ABSW04E004 . ...... ;. ......... I I :'" ..... I RO skvscraper'l

SRO

"'c"" """""" " -,"' I Topi£J RO 21 I -Given the following conditions: Unit 2 is in MODE 3, NOT and I-A total loss of all Control Air Which of the following describes the basis for transferring the charging pump suction to the RWST lAW S2.0P-AB.CA-0001, Loss of Control Air? 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 ReS at a I faster rate. [§] 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 achieve CSD boron concentration. I [£J 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. I itive displacement charging pump is running, the minimum flow supplied at the low speed stop would not match the flow lost to the ,,'-'u am the seal retum relief valve lifting, and RCS inventory will be lost I R IIVUl:lI!l<l..,el ...eve! II Memory Salem 1 & 2 11.E:xamDate:*11 9/26/20111 IAK3.08 11 D [J I Loss of Instrument Air 55.41(10) ARCA basis document slates, "Prior to commencing a cooldown, the RCS must be borated to cold shutdown conditions. Withoul Control Air, there is no way to reduce RCS inventory. Thus, the amount 01' 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 suclion is transferred to the RWST early in the event. This ensures that any addition to the RCS is at RWST . . ." is incorrect because while the recirc line does gO' back to the VCT, it is not the for the transfer. D is incorrect because seal I retum is not isolated on loss of air. 1------------------ ______________ I ABCA01 E002 Requiredfor Examination I I II

• .I Previous 2 NRC Exams I iI Direct From Source IIUsed During 1 0 I"J" ILOT RO NRC Exam August 2008 . ......icv.. ", ",." ." .. ....... .. . '. I I I I
• I I RO

.1. '" ,SROSkYSCrape r ] I.SRO SystemlElloluti O l1Listl I ',' Topic I RO 22 Given the following conditions: The Unit 2 operating crew is responding to an Inadequate Core Cooling condition 2-EOP-FRCC-1, Response to Inadequate Core The crew is unable to re-initiate ECCS -All SG WR levels are -NO AFW pumps can be -All core exit thermocouples indicate Prior to opening the PZR PORVs in an attempt to lower CET temperatures, how should RCPs be utilized? RCPs will NOT be started since possible rupture of hot SG tubes could occur. 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. I Memory 1 & 2 I I [Eg1000074K201 !Inadequate Core Cooling and the followinc: I *.o.'1 5?,41 (1 0) With SG WR levels at -5%, the:e is no secondary heat sink. At step 13114 of FRCC-1., no SG level and no AFW f1?W II' 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 wihlch l.______ there are none. No RCPs can be started. It is NOT because that they will be needed later in procedure. I\==:; I FRCCOOE002 I MateriClI ReQuired(c;if EXfurliriation I I II Facility Exam Bank I 0.*ification'Met.hCld;"l! Editorially Modified ... IQuestion Source c;ommentsl IVISION Q 73445. modified to remove extraneous info. Removed RVLlS level in stem. Added AFW ...--.'-'- status and SG WR levels 10 '. II I II I II I

i-A small break LOCA has occurred on Unit 2. The actions of 2-EOP-LOCA-2. Post-LOCA Cool down and Depressurization are being performed, With ONE Charging Pump and ONE 51 Pump running. PZR level is stabilized at RCS pressure of 900 psi!]. With RCS subcooling at gOF. the operator turns on a set of backup heaters. Which of the following describes the result of turning on the backup heaters? Break , ECCS Increased pressure will j LOCA02E005 I [ [QUestion I Facility Exam Bank I iquEi!stioh Editorially Modified gorments] I Vision Q73852 ',' iCon II "..... .*.*...'.! '.:.: ... , :,,' II II II

• I.SRO SkyScraper I ....

!Question TOpi:J I RO 24 Given the following conditions: -Unit 2 is attempting to identify and isolate a LOCA outside 2-EOP-LOCA-6, LOCA Outside Containment, has just been -The source of the water is backleakage from the 23 cold leg injection Assuming that any valves required to be operated during lOCA-6 operate correctly, which of the following leak locations would NOT be while using On the valve outlet flange on 21SJ49, RHR DISCH TO COLD LEGS. Between the 2RH20, RHR HX BYP VALVE and the 2RH26, HOT lEG ISOL VALVE. Between the 2RH2, RHR COMMON SUCT VALVE, and 22RH4, RHR PMP SUCT VALVE. I Comprehension I I Salem 1 & 2 [ 9/26/20111 00WE04A102 I j LOCA Outside Containment 1M StatemeJl!=1 55.41(3,8) 2-EOP-LOCA-6 closes/checks closed the following valves: 2RH1 OR 2RH2, 21 and 22RH19s. 2RH26. 21 and 22SJ49s. Using drawing 205332-SIMP, it shows that any leak between the RH1/2 and the SJ49s will be isolated with the above ..*J valves closed. The only location which wouldn't be affected by those valve being closed in the downstream/outlet side of the SJ49 valves. The stem statement of proper valve operation was inserted to preclude a candidate from assuming a leaking valve may not __________ ... -... ...I L.O,*Num\:>er "iFi* ..1 I LOCA06E005 , ___---J IIMaterial Required for EXamlt,ation j I II I Facility Exam Bank I. J Direct From Source I progra!}L] 0 Ci I question Source Comments I J-ROC21 ...___' .. ____*____ __---2..J. .. ,,",on,, ...,.,,, I I I I RO ,I 'SRO I;* "R6 List',. '1_Outline Chanaes I

Question Topic I RO 25 Given the following

-Unit 1 has experienced a -11 RHR pump is CIT electrically by it's 4KV breaker -RWST level is 14.8', and operators are performing 1-EOP-LOCA-3, Transfer to Cold Recirculation. Which of the following conditions will prevent transfer to CL recirc, and cause the crew to go to 1-EOP-LOCA-5, Loss of EmergencL Recirculation? 1 12SJ44, Cant Sump Suct Valve, does NOT open when its Sump Auto Arm PB is depressed. 11 RP4 Lockout Switch for 12RH4, RHR Pump Suction, is stuck in "Locked Out" position. g] 1 12SJ44 open PB is depressed before the 12RH4 close PB is depressed. ISystemlEvolutioriTiUel I Loss of Emergency Coolant Recirculation IKAStatemElnt: Knowledge of the interrelations between Loss of Emergency Coolant Recirculation and the following: I Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. A is incorrect because Unit 1 SJ44 valves do not have auto swap function like unit 2 has. C is incorrect because the close signal iEXP,la"ri,atio,.,n,.Of iAnswers: 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. B 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 recire cannot be established, and procedure directs transition to LOCA-5. ... '='=1-1 'u=m=' N=o.=1=U=n=it =12=S=J4=4:::, 1=2R=H=4:::, a=nd=1=R=H=1 ===,1 I I________ ___--'II_-__----'rl_---'II----'I \,":"""[L.a. NUmber. I LCA3U1 E004 I RHROOOE006 1___---' IMaterial Required for Examination II !Question Source: II Facility Exam Bank I!QuestlonModification Method:ll Editorially Modified liUsed During Training Programi D jCluestion Source I Vision Q122456, removed step number from distracter C since it is a non-existent step I IComment I I I I I I I

• • • • • • • .. *T* ..

RO RO SvstemlEvolUtion List SRQ.

1 Outli TOPiciJ 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 How many SG Safety Valves will be open on 24 SG if pressure remains at 1115 psig, and all safety valves operate when ex 55.41 (4,7) Each SG has 5 safeties, with lift setpoints of 1070, 11 DO, 1110, 1120, and 1125 psig.

I FRHSOOE009 I STMGENE008 ....w---------I1 .11 I iUsed DuririgJrafningProgram . 0 Comrnents .., .....:d " ',.1 I I I I II I Given the following conditions: -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 Rihr 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? radiation levels have the TSC has not provided any direction yet. they reset on lowering containment pressure and the radiation threshold has not been reached. Abili to determine and inter ret the followin as the appl to Hi h Containment Radiation: Adherence to ap ropriate procedures and operation within the limitations in the facility's license and amendments. 55.41 (9.12.10) Adverse containment numbers are used when containment pressure reaches 4 psig or containment radiation level lEX. ...O** *.n.*.... *.* **.*. *.Mj* reaches 1 E5 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 Adverse numbers have are in effect based on the containment radiation threshold being exceeded. I PROCEDE004 1__----' j Material Requireci for exarnillatioll*** I 11 New I[QUeS'i?n I Program J 0 to NRC Exam questions from Sept 2001 Coo.k exam, Nov. 2002 Salem exam, Dec 2002 Beaver Valley I IComment . ' .......*..... ....... / ................. II I I I I I Given the following conditions: -Unit 2 is operating at 100% -A failure in the automatic Rod Control circuit cause Bank D Control rods to step in 72 steps per -Control rods insert 20 steps before the RO places rod control in manual and rod Which of the following describes the effect of the rod insertion with NO other operator action? RCS temperature will lower to the RC Loops Tavg-Tref Deviation alarm setpolnt of 1.5"F. E=J I Application 1 & 2 I 1 9/26/20111 I Control Rod Drive System I KA Statement] Knowledge of the effect that a loss or malfunction of the Control Rod Drive System will have on the 55.41(5) A is incorrect because the RIL for 100% power is -170 steps on Bank D, so 20 steps of insertion from 230 steps (ARO) will not cause this alarm. B is incorrect because Main Generator load will lower due to lower steam pressure, but it will not remain lower -----since the Governor valves will automatically open to restore Turbine Steam line pressure to the value corresponding to 100% power. C is correct because the rod insertion will cause a lowering of RCS pressure sufficient to full close the Spray valves, wi1ich are n",,,,,,,, I", _ gO/_ "non ';"0 t" rn . . "no ('\,,-,,,n "f P7Q'RII I h.,,,t,,,,,,, in m"nll,,1 nl\l n ie . ho,.."""" "hilo 10m no,,,+' "'" ,.11 I lower, the setpoi'nt of 1 SF is wrong. 1.5' low Tavg=Tref is wihen auto rod outward motion would occur. j............. .. . ....... " .. ReferenceTitle:;" ,,[1 " "IIPage No*1 ! c=o=nt=in;;;;;uo=u=s S2.0P-AB.ROD-0003 II II 1121 I I Overhead Annunciator Window E II S2.0P-AR.ZZ-0005 It 1111 11 19 I c=o=ntr=o=1 I 1136 11 35 I [L.9.Number I ABROD3E003 I RODSOOE019

• • • • • Given the following conditions:

-Unit 2 was operating at 100% power when a 500 KV grid disturbance caused a Rx -2H 4KV Group bus deenergized upon the Rx -21 AFW tripped immediately upon -SGSD flow is zero to all 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 8% higher than the other 3 steam generators. Loss of 23 MAC Panel has caused SGFPs to supply uneven feed flow to the SGs. 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. I Comprehension I Salem 1 & 2 li<A Knowled e of the effect that a loss or malfunction of the Reactor Coolant Pump System will have on the following: S/G 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 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 uncontrollably. In this instance, it is not uncontrolled, since it is the natural reaction of the reduced steaming rate. SGSD flow was included in the stem to lend support to . ., . .. . . . I a different Group bus, and SGFP flow has been isolated by FWI at 554 0 Tavg following the trip anyway. The pressure override circuit 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 to reduce flow to raise dischar e pressure. 'F====;,-------------------- _____________________ .. i I I I I Given the following conditions: -Unit 2 is operating at 100% -2CC190, RCP THERM BAR CC OUTLET V, fails Which one of the following describes the effect on RCP temperatures, if any, as a result of this ALL RCP 1#1 sealleakoff temperatures will rise. @J bearing temperatures will remain the same. I Reactor Coolant Pump System KAS" dJ IK no wi eJQe a f h f h ff a I If unc t'Ion on th e f II oWing WI'11 h h R eac or C ooan 1 tP d teo tee ect a f ass or ma 0 ave on t e t s i: stem: Containment isolation valves affecting RCP operation I ] 55.41(3) The CCW line supplying the RCPs is a single line supplying both bearing cooling and Thermal Barrier cooling. Once line inside containment splits, the CCW from the Thermal Barriers has its own, separate return line, which is isolated by the (inside containment) and 2CC131 (outside containment.) The Thermal Barrier CCW flow acts to cool reactor coolant flawing upwards through the thermal barrier upon a loss of seal injection flow. With normal seal injection, the loss of CCW to the thermal . A ,U, Dt"O , I i,;i",>!,::" t ii"i"; ,I Unit 2 Component Cooling 11205331-3 I I 11 35 I I I I I II I I I I I II I I RCPUMPE004 I RCPUMPE015 I RCPUMPE016 I I II IQuestion SOl!fce:.11 Facility Exam Bank I Direct From Source I[ysedJJllfingTrai.ning 0 J-ROC30 .... **... ..*.. ..*. ....**. *.... .i , .. .... !I I il I II I Starting a RCP in MODE 4 while on RHR Shutdown 1CV277, Letdown Isolation fails shut with normal letdown in service in MODE 1 CV132, Excess Letdown FCV fails shut with Excess Letdown ;n service at maximum flow in MODE IkA:II004000K629 I Chemical and Volume Control System will have on the Chemical and Volume Control S 55.41(3,5) The KJA match for this was difficult as it seems to be asking for two different things which would make a very long question. Incorporated both the "effect of the malfunction" although there isn't a malfunction listed, of an Excess Letdown (CVCS) c ....._ *....___,.__'"' malfunction and its effect on the CCW system. B is incorrect because the RCP pump heat is a large load on the RHR system AND cooling load on the CCW system (bearing and winding coolers). C is incorrect because normal letdown flow is 75 gpm, and its loss . . . 11 81C=he=m=ica=1 a=nd=V=o=lu=m=e =co=n=tro=1 ====::::;11205228-2 II 11 I C:::::om::::p:::o:=ne:::n:=t c::::o:::::o':;:in9=.:w:::::::at=er======41120523H ,2,3 11 i 1F===1-1166.44,451 1 FI __________-__-JI____---'11 il I perU ,'" I CVCSOOE015 EXa§inatiorl"1 I II !Questiorl so§D I_N_e_w______......__.--_-lllusedDuringTraJnirig*progral11 I 0 Comments I I 1 , IComment' .'......... ' . ........,.' c'" .**.,.. ..'........' I II I I I 1__________________________________ 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. 30 minutes after the Rx was trip ed, which of the followin locked in alarms would be consistent with plant conditions? ________________________________________________________________________Ability to verify that the alarms are consistent with the plant conditions. .1. I 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 ',';,;; 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 15.2' when the initiating casualty ;" ., e;,.,,,I,, c:r:: hl".i,.,n,-l", ",., in I RHROOOE008 I I I I I I__ RO SkyScraper I* SROSvstemlElloluiion List

• 1 * ** H"'H"'" **********

CCW system is split during operations in the EOP network. 2RH20, BYPASS VALVE is opened to raise total RHR system flow with 21 RHR loop operating in Shutdown Cooling mode. [§swerll a I;Exam LeveO I R I I Application I II Salem 1 & 2 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 cia] leg recirc. The large (-a,OOO) 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 incorrect but plausible if the reduced SW flow were though to have a large impact on CCW system parameters. The . ,. \MOllfd anen to maintain S\J\l svstem flo):" through CCHX. C is incorrect because the splitting of CCW system headers in LOCA-3 after the transfer to* CLR has been performed, and does not add to system flow nor reduce system pressure. 0 is incorrect because operation of the RH20 increases I the flow bypassing the RHR HX. so there would be less coolinS reguired. and would not cause s stem pressure to lower. __________________ ROSkvScriuier* I I '-...-... I .1 RO Given the following -During a SBLOCA in MODE 1, a Safety Injection signal is actuated -RCS pressure is 2,000 psig and lowering Which of the following describes how the Safety Injection pumps are prevented from Injection flow from the RWST into the RHR system passes through the pump. A small portion of discharf;je flow is recirculated to the suction of the pump through a Recirculation Flow HX. 006000K406 I Emergency Core Cooling System iKA where ECCS accumulators also join RHR sysiem into RCS cold legs. D is incorrect but'plausible, since there is no'recirculation cooler, but other pumps specificall AFW have recirculation coolers. I I----------------------- ________----------- ____________ iL.().NlJmber I ECCSQ()E008 I II 1 !QuestionSOllrce: .. j I New _________----'1 [Use? D IQuestion c()mments. .1 I C""... "',. " '.' ". I I I I ,RO

I "RO II c 1 RO 35 I Which of the following relief valves lifting would cause a rise in Pressurizer Relief Tank temperature or level? 55.41(3} The 3 distracters are components which used to be directed to the PRT but were re-routed to the containment trench during a DCP. IpZRPRTE008 I !PZRPRTE003 I [ I I iOuestinnSource:iJ I Facility Exam Bank !I Significantly Modified! fUsed D Vision Q61390, inputs to PRT, changed correct answer and all distmeters. I I, ROSkvScraoer. f ' SRO I ,.RO Svstem/Evolution uit., List 1.. ...'... I RO 36 Given the following conditions: -Unit 2 is operating at 65% power. -An intersystem leak develops between the CCW system and the Spent Fuel Pool Heat Exchanger. -CCW Console Alarm SURGE TANK LEVEL HI-LO alarms on 2CC1. Which of the following describes the action required lAW S2.0P-ARZZ-0011 Control Console 2CC1? Drain the CCW surge tank from local drain valve to a 55 gallon drum to prevent overflowing tank to the in-service WHUT. I b I R I I Application I I Salem 1 & 2 tion Modification Method .. ... .. Source 90I'Qmentsii VISION Q50370 . . o""lstion Topic I I RO 37 I Given the following conditions: -Unit 2 was operating at 75% power. -Both pressurizer PORV block valves (2PR6 and 2PR7) have been closed for several weeks due to seat leakage past the PORVs (2PR1 and 2PR2). -The PORVs are being maintained in MANUAL, with ALL Tech Spec required actions for the leaks complete. -Plant conditions develop which required a shutdown and depressurization. -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 conditions? 12PR6 and 2PR7 would OPEN; 2PR1 and 2PR2 would OPEN 12PR6 and 2PR7 would OPEN; 2PR1 and 2PR2 would remain CLOSED 12PR6 and 2PR7 would remain CLOSED; 2PR1 and 2PR2 would OPEN fd.l12PR6 and 2PR7 would remain CLOSED; 2PR1 and 2PR2 would remain CLOSED -55.41(7) With pressure above 375 and temperature < 312, arming POPS would cause the PR1 ,2,5,6 to open, regardless of the MANUAL selected for PR1 and 2. -------------II------------'I--------II-----lI--. Material Examination II. II I Facility Exam Bank liCluestion Direct From Source I[Used During Training Program i I Vision Q87628 RO SkyScraper II . ,.Ro ListlFi sRe SystemiEvoiutlortlist. ., ()uiihle Cha'naes *1 RO 38 With Unit 2 operating normally at 100% power, what would be the effect on RCS pressure if the controlling PZR pressure channel instrument were to fail LOW, with no operator action? T. lower until the Rx trips on Low PZR pressure, .4J rise until PZR spray valves open further to restore pressure to 2235 psig. 010000K301 I IKA.Statemerit':l Knowled e of the effect that a loss or malfunction of the Pressurizer Pressure Control S stem will have on the foliowin RCS 55.41(7) PZR pressure controller will see a low pressure condition from the failed instrument. This will cause spray valves to close fully, and all PZR heaters to energize. Actual RCS pressure will rise in response to the heaters being on, and will continue to until reaching the PZR PORV setpoint of 2335 psig. Only ONE PORV will open. PORV actuation coincidence is 2/2 2335. Since one of the channels is failed low, its PORV cannot open. B is incorrect because pressure would rise, but the trip .. . I incorrect because the spray valves will shut on the channel failure, and be unable to respond to the rising pressure. I---------- ________________________.L..O:NUmbeF';; IABPZR1E001 Material Required for Examination IJ I !Question Source: I Facility Exam Bank Modification.Method: . I Editorially Modified I [Used During Training D I 'Olll""tii-m Source Comments I Vision Q80493, replaced two distracters so that all choices did not include rising ReS pressure. Correct answer remains the same . I .... , ..IComment , i I I I

---

RO I . I .RO I SRO List!Out:lineChanQeS I-c'-----:-. I , , :Question RO 39 I For each of the listed PZR heaters, identify their 460V bus power supply. 11 control group 11 B/U group NORMAL __ 11 B/U group EMERGENCY __ 12 B/U group NORMAL __ 12 B/U group EMERGENCY IE, G, C, E, A. I I G, G, C, E, A. I [] I E, E, A, G, C. I [] I G, E, A. G, C. I ,Answer II b 1 R I [Cognitive Level II Memory I 11 Salem 1 & 2 I [EXamoateJ [ 9/26/20111 IKA: II 011000K202 IIK2.02 21 0 . ..,..-.-.".....__._._..._..._ ISystem/Evolution Title I I Pressurizer Level Control System J IKA Statement:] Knowledge of bus power supplies to the following: I PZR heaters I of 55.41 (3,7)The distracters all contain the possible busses, but in incorrect orders. iAnswers: II .r-.,' Reference Title ,/. I[; ,Facility Reference NUOlber:J II Page No-:I [ReVisionl I No.1 Unit 1 GP 480V bus PZR heaters.controlie 11203347-1 11 13 il II 1 I No 1 Unit Backup Group 11 11203348-1 11 9 JI il I I No 1 & No 2 Units Group 12/22 backup groups [1247992-1 11 3 II [I I IL.O.*Number I PZRP&LE005 I___---J IMaterial Required for Examination II II IQuestion Source: II Facility Exam Bank IlQuestion Modification Method: j Direct From Source Ilused Training 0 [QUestion Source Comments 1 [0801 C39 I I IComment '-I I I I I I I RO ..*.. SRO Skyscraper I.. .

• .*

List** I [Question TopiC] 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? the Phase A signal must be reset to allow the ECCS criteria have been and Phase A cannot be reset it allows operators to regain control over plant equipment and prevents any equipment from automatically repositioning when RWST reaches 15.2', 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. it allows operators to regain control over plant equipment and restore a sustained compressed air supply to containment. [*:11 012000G406 ,--________________________ .______________ --, Knowledge of EOP mitigation strategies, 55.41(7)The Phase A signal CAN be reset without resetting SI as it has a retentive memory circuit. See Note 10 on 221057, Bases docum ent states that the SI reset function is so that equipm ent can be aligned, and to restore a sustained, compressed air supply to 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 i '-h.1I ' . '" <::: 111 "l ,,-l .h. ("("'11': ,.u ,II",* ""I. . ,,; 'h. ,rI" "t-nmAl"r i:) ...

i: Ii.

I Reactor Protection Signal Safeguards actuation 11221057 II II II F 2=Z=::::::; I Loss of Reactor Coolant Bases Document 112-EOP-LOCA-1 II 1118 11 28 I II II II f IF====' . II . Topic 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 High Steam Flow Safety Injection will remain blocked. A Loss of Off-Site power would NOT initially cause the RPS to trip the Rx. The low power Rx trips could NOT be blocked until P-13 energized during Main Turbine startup. Knowled e of Reactor Protection S stem desi n feature s and or interlock s Automatic or manual enable/disable of RPS trips 'E.x.):l..'a.***.*.na *.:.. r..'..*.* *..*. *. 55.41(7) D is incorrect because the low power trips can NOT be blocked until P-10 is actuated. C is correct because the "at power".... .*.. ... 't.i.*O.. O.'.f.g.: .. ':'. Rx trips (which are different from the "low power" trips) are not unblocked until P-? is actuated, wlhich gets its input from either ----10(wlhich 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 wlhile the loss of off site power will cause a loss of the SGFPs, the SGs will not shrink to the 10 . ., .. l¢ageNo:l

===:;1\10 1lJ I RXPROTE028 I , RXPROTE027 1 I 1 iO,,""tir.nC!;;:, ...: ',11 Facility iIID:i:re:c:t DI*.

P=I'O::9::

r:::.ap;]:::m:::'* C41 I

• ...:::.. ......> ., ... : '.' ....*:....... ' ...... ':". ... .:,.: ' ..,,, ..... i I 1 1 RO

__ "A141:.-." IRo42 I Given the following conditions: -Unit 1 Rx power is 8.1 %. 1-Power is being raised slowly in preparation for rolling the Main -11 SGFP is in service supplying FW to -12 SGFP is latched and at idle -All AFW pumps are aligned for normal standby 11 SGFP trips. Which of the following describes the effect, if ant, this will have on the AFW pumps with NO operator action? I The MDAFW pumps and the TDAFW pump will start when SG levels drop to the 10 10 level setpoint. j I.The MOAFW pumps will start when 11 SGFP trips. The TOAFW pump will start when SG levels lower following the Rx trip. I IAll AFW pumps will remain in standby. Sufficient steam will be supplied through the 11-14MS18s, MS STOP BYP VALVES to supply 12 SGFP. I II All AFW pumps will remain in standby. 12 SGFP will remain in service since at this power IEivel it is being supplied with steam from the Heating Steam System. 1 :Anslf.lerll a I R I I Application I [Salem 1 & 2 I ! 9/26/20111 KA5. 013000A104 IIA1.04 -I IRQ Value': 3.4 [§"ifQ Value: 23 [§Rc5'Group:1 _1J 0 .-;:::,.:." .......r .... ------,...

===."=[2]=-==.-

I Engineered Safety Features Actuation System Ability to predict and/or monitor changes in parameters associated with operating the Engineered Safety Features Actuation controls includin S/G 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, step 5.4.10), and 12 SGFP will not provide sufficient discharge pressure at 1100 rpm (idle speed), and the stem stales with no .... operator action so speed will not be raised. A is correct because the MDAFW pumps and TDAFW will start on 10 10 leve! in SGs the source of FW is lost, and the SGs continue to steam. C is incorrect because the MS 185 could provide sufficient steam flow

• 01 *
• 1 MDAFW pumps is a trip of BOTH SGFPs.. 1-------------11------------'

---

1 I AFW000E015-

--I I I I lMaterial R=quiredfor ] I II I I Facility Exam Bank Method: Editorially Modified IlU!ied 0 IQuestion Source Comments I I Vision 085462. Modified stem from MSLI to trip of operating SGFP with other SGFP latched. Correct answer * ...... remains the same. Enhanced distrac!ers. I ., , ..... 'IComment ... ..! !I I !I I 'I I RO ... .... RO 43 I Given the following conditions: -Unit 2 is operating at 4% power during a -A RCS leak causes PZR pressure and level to lower -The operating crew initiates a Rx trip and Safety -When the 81 is initiated, a loss of off-site power 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 exeected, if no further operator action is taken? I NO CCW pumps are running. r [§J 124 SW pump running and 23 SW pump stopped. r IChargin g Systems SI Flowmeter reads 100 gpm. I @]J21 ABV Supply Fan running, 22 ABV Supply Fan NOT running. I c I [Exarn Level. I IR I [Cognitive Level. "11 Comprehension I [FaC;III§ij I Salem 1 & 2 I 9/26/2011 I 1 I KA: II 013000A401 11A4*01 ] I 4.51l sRO yalueil! [!§OGroup:11 111sRO Group:'11 11 D ISystem/Evolution Title I I Engineered Safety Features Actuation I KA StatementJ Abilf!y to manually operate and/or monitor in the control I ESFAS-initiated equipment which fails to actuate I 55.41(7) CCW pumps are not started during MODE III (Accident plus Blackout). 24 SW pump is designated as the Lead SW pump of Answers:.* ..on B 4KV bus, and only one pump per bus is started. 23 receives power from B bus also. 22 ABV Supply Fan is administratively locked out, and is what causes the B bus sequence not complete to occur on every SI initiation. The Charging systems SI flowmeter is BIT flow. 2 centrifugal charging pumps should have started, and even if RCS pressure were normal, which its not. the fin" thrn* ,,.,h tho QITrl h . h;,.,hor r::ahor thoro ;c -nmhl"m ,,;th Iho fln,,,n<olh ." c:: 11? <on'; c: 11 !=lIT n"ll"t ",I""c 1 did not stroke full open, or less than 2 pumps started. TRIP-1 page 1, and TRIP-2 page 2 flowcharts, page 1, each Table A shows Accident Loads and Blackout loads I .*..*. Title' I Rx Trip or Safety Injection I Rx Trip Response I I* .. Facility Reference Number=:] [Reference Section fI2-EOP-TRIP-1 II 12-EOP-TRIP-2 II II JI I[iage No.; IIRevisionl II F1 1127 I II F1 1127 I II II I Number I SECOOOE004 IMaterial Required for Examination IQuestion Source: ! . . . jlNew IQuestion Source comrnentsl I ]Comment I I I II IIQuestion Modification

1 -I I I I I II IlUsed During Training Program I D I Given the following conditions:

-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. NONE of the CFCUs start in Low Speed. How will the failure of the CFCUs affect containment instrumentation readings? CFCUs in service I 022000K302 I Containment Cooling System 55.41(5,7)The CFCU HEPA filter, which is in service following accident initiation, is designed to remove minute radioactive 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 .. .

• .ROskvscraperl

  J L. :' :s:R:o:..

i9uestion I_R;;...o:....-4;;...5 ___________________________________________ ---l Given the following conditions: -Unit 1 has experienced a -Control room operators have progressed to the point where the SEC has been -Containment pressure was 4.5 psig when the SECs were 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? would realign for spray, the CS pumps would be started by the SEC. would realign for spray, the CS pumps would have to be manually started. must be manually realigned for spray, the CS pumps would be started by the SEC. must be manually realigned for spray, the CS pumps would have to be manually started. I Application I Salem 1 & 2 I Containment Spray System Abili to monitor automatic operations of the Containment Pump starts and correct MOV I-EJ(.P.**.I.* a...***.n.... .*....*. ()f...... 155.41(7) The SEC controller operates the CS pumps at 2 di.fferent point in. the sequence the SEC reset. The SEC ONLY ,AIisw'ecrs;

• • . ,.... ..' controls the CS pumps, not the CS valves. The valves realign on the hi-hi cont pressure Signal whenever it IS received, but once the . SEC is reset it will not start the CS pumps since the sequencer is no longer active. i L,dJ'N u'!I ber::>tJ!i" '. I CSPRA YE008 1 CSPRAYE009 1----' II I Facility Exam Bank I JI Direct From Source ,Question Source Comments I VISION Q80567 < .' ....*.*.****.***...**.

'C .............. .. ' ............. / '. I I RO I, 112 keyswitches on BOTH safeguards 2/2 keyswitches on BOTH safeguards 1/2 keyswitches on EITHER safeguards 2/2 keyswitches on EITHER safeguards I d I !E:xarnheYElI R i i C 9s'h"lt!"e '-eyer I Memory I [FaB!!JE [Salem 1 & 2 I Containment Spray System 55.41(7) Salem has two safeguards trains, either of which performs its safety related function. Containment Spray requires 2/2 _----""".*.""'.keyswitches tumed 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 keyswitches on EITHER train. II ROSkyScraper

>R6 ,I f

[§ues_tiinTopic:] I RO 47 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 evcs HUT. fKJilllo33000K105 I Spent Fuel Pool Cooling System s between Spent Fuel Pool Cooling System and the following: I 55.41(8,13,4) The RWST normal level is -41'. The puts the water level at 141'. The Spent Fuel Pool is -128'. There can be no gravity drain TO the RWST, although it will work coming FROM the RWST to the SFP. Transferring water from the SFP to the RWST is done by manipulating valves within the SFP Cooling system, and pumping the water with the Spent Fuel Pool pumps to the RWST. The other method is to drain it to the CVCS HUT. The Refueling Water Purification pump cannot take a suction on the . .. .. !performed by procedure, and it would d-rain to' the Fuel-Handling BidS not the eves HUT . ...*'.' .. . . [page No:1IReyisiori! 1 Fill and Transfer of the Spent Fuel Pool [I S1.0P-SO.SF-0001 I[ II r 1 19 1 I Unit 1 Spent Fuel Cooling 11205223 II I : 11 26 I ITankCapacitYData IIS1.oP-TM.zz-0002 II II ila I I SFPOOOE010 !MaterialRequired for Examination J I Ij I jUsed [)UringTraining 'Program 1 0 , ..' I I I I

.. rQuestionTopiC] I RO 48 Unit 1 is performing a plant startup to full power. From 0-20% power NR level will rise from 33% to 44%. From 20-100% power, NR level will rise to From 0-20% power NR level will rise from 33% to 44%. From 20-100% power, NR level From 0-40% power NR level will rise from 33% to 44%. From 40-100% power, NR level will rise to From 0-40% power NR leve! will rise from 33% to 44%. From 40-100% power, NR level will remain at [i5A;iI035000A101 I Steam Generator System I STMGENE007 I ADFWCSE015 iJIJIClteriaJ i! II !Oll"l!itionsource: i I Facility .* .*** 0 .. 'lComment ". ...' , .' .*.**.. *i.*.i.* *.*. *......* "', I I J J

TOPi,§J I R049 1 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. Assuming the reactor does NOT trip, which of the following describes the initial response of RCS Oelta-T and SG pressure in the AFFECTED loop? I RCS Oelta-T rises md SG rises. I IRCS Oelta-T rises and SG steam pressure lowers. I @] I.RCS Delta-T lowers and SG steam pressure rises. I iJ 1 RCS Oelta-T lowers and SG steam pressure lowers. I I R Ilqognitiyel..evel I Comprehension I I Salem 1 & 2 I 9/26/2011 1 -IkAn I 039000A 106 ! [0 (3roupl 11"[ -w"% I Main and Reheat Steam System Ability to predict and/or monitor changes in parameters associated with operating the Main and Reheat Steam System controls indudin : Main steam ressure 55.41(5) The unaffected SG's and therefore RCS loops provide the same amount of energy to the turbine -raising RCS OIT AD!:,,#er;;;:':"j lowering SG steam pressure. In the affected loop, RCS DIT lowers to zero and SG steam pressure rises because heat removal . minimal. At least one of the conditions is incorrect in each distracter is wrong. 1 MSTEAME015 1-__---' Rtl9uirectJor I I 'I I Facility Exam Bank 1 j Direct From SIJurce Prograr!ij 0 ________________________________. .... .......*.... .'.. .... . ..' ..... ...... ." ...., .. . " I , I RO 'SRO Sk';;craper [aoosFi6n TOpiC] I RO 50 I Which of the following describes why Main Steam Jines are drained prior to admitting steam into the headers? a vacuum pathway to the Main Condenser is available free of potential loop seals. I I Removes any collected corrosion products or impurities to ensure Main Turbine blading is not impinged. I IPreheats susceptible components such as steam traps prior to exposing them to full system 1: lperature. I I Prevents pressurized steam from forcing residual water in the piping to cause water hammer on downstream components. I , [[:J I Memory !I Salem 1 & 2 II 9/26/20111 r¥ll039000K501 _ [] I Main and Reheat Steam System I__Knowted e of the operational im Ii cations of the followin to the Main and Reheat Steam System: Definition and causes of steam/water hammer 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 during cooldown would be transported downstream, where it would impact inner piping walls and turbine blades if not removed. A is incorrect because the vacuum path would be from the condenser back through turbine control and stop valves which would be shut 8 is incorrect because while it remove anything in the condensed steam in the piping, it is not the reason why. C is incorrect I--------------..JI __________ -' ______....J __---' 1_,----, I I MSTEAME013 I WATHAME006 I I I I New I ________--" [J.Question Source L_*__..__-_.._-_..:__-__.. ........****** .*..* i** i. I I 1 ....,....""",.'" J IRQ I Given the following conditions: -Unit 2 is operating at 100% -Main Steam Dumps are in MS Pressure Control* AUTO. set at 1005 2PT-505, Main Turbine Steamline Inlet Pressure Channel. fails LOW . Which of the following identifies how the Main Steam Dum ps will respond to this failure prior to any Reactor Protection System response? Main Steam Dump valves will ... ! remain shut throughout the event I ALL trip open and then modulate in the closed direction in response to lowering Tavg. I ALL trip open and remain open until they automatically shut when Tavg lowers to 543 of.I @d initially remain shut, then modulate open as sleam pressure rises from the load reduction. r rJ'u

• Ia I R

Application I I Salem 1 & 2 1 I 9/26/2011 i 041 000K603 IlK6.03 . 2.7I S Rq'vaIG¢=H 2.91 GroiJp:11 21 r.,. d.* .. e I Steam Dump System and Turbine Bypass Control IkASf' n", Knowledge of the of the effect of a loss or malfunction on the following will have on the Steam Dump System and Turbine Controller and positioners, including ICS, S/G, I 55.41(5,6.7) When the Main Steam Dumps are placed in MS Pressure Control ModEl, that "arms" them, and they are sel up 10 respond to a deviation from its setpoint, which in the stem is stated as 1005 psig. Tl1at is. until Steam Pressure as sensed by different steam pressure detector (PT -507. Steam Header Pressure, nol PT*505 Steamline Inlet Pressure) rises above 1005 the steam dumps will remain closed. 100% power sleam pressure is -800 psig, and it would rise as a load reduction wh.m PT_"n" f::.il" I", thi" ,..", "'" ::'l1tf'\!'T'!"tir mn . ,,..: Alhi,...h ill In\Al",r Reg "nn I correspondingly, secondary side steam pressure. __________________

• [.;O.N4mber

(:-;'.:: I STDUMPE007 I STDUMPE008

I New I _____.___-1llUsed[)ur ing Trainingp r ogral11] [] , :c",,,,,,,,,,...',, ,'. ., ! I I ii, I l------------------ ________________ .-.. "" "" "" " Topic I Which of the following identifies how a Main Turbine trip is confirmed in the EOP network lAW OP-AA-101-111-1003, Use of IOHA F-32. OEHC Tril I I Auto Stop Oil Pressure <45 psig. I I ALL Main Turbine Stop Valves shut. I [J I Main Turbine speed <1800 and lowering. I I c I Le,,:,ell I R 1 "." j I Memory I I Salem 1 &2 I 1 9/26/20111 1 045000A406 11A4*o6 I 2.81 [SRO,Value: 1I 2.71 2] Grdue3] 1 21 0 I Main Turbine Generator System 1[045 J IExplanationofl 55.41(4,10) A is incorrect because it is a result of the Turbine Trip, but is not used for trip confinnation. F-32 is plausible iAnswersi >j one of the conditions which causes it to alarm is the correct answer(AIi main Turbine Stop Valves shut) but because other things '" to be exact) also cause this F-32 to alann, it is not a TT confinnation. B is incorrect because while it is available to the operator RP4 right next to the Turbine Stop Valve indication, it is not used to confirm the TT. It only show that's the oil pressure which " ' . incorrect because speed less than 1800 rpm indicates the Main Generator is no longer connected to the grid, and speed lowering indicates no steam suppl for an unloaded turbine, it is not the definition of confirming the turbine Iri .I rReylsi'On" I 1-------------' _______---' --------1 I 11-----' 1 MNTURBE008 1__---' Material Required for Examinati6n l I Question 1 New I ________-'1 :UsedOuring Training pr0!JE<iTJ 0 COIilrTl§tS] IComment ..., I II I II I II I ROSkvScrallerl ** __.m__***_, I [Question Topic: RO 53 Given the following conditions: -Unit 2 is operating al 80% power, steady -Power was reduced 2 days ago when 21 Condensate Pump -21 Condensate Pump remains -The Condensate Polisher is in service with full Which of the following identifies the initial concern if 22 Condensate Pump were to trip, and the action which would be performed in response to that Lowering SGFP suction pressure. Open 21-23CN1 08 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. 9/26/20111 Ability to (a) predict the impacts of the following on the Condensate System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal 0 leration: Loss of condensate umps 55.41 (4.10)The limit per AB.CN for 2 cond pump and 3 HDP's in service is 85%pow."f. When the second pump trips as in the stem above, the power limitation is 30%. SGFP suction pressure will rapidly lower. SG N R 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 more. The polisher is bypassed. the HP heater strings are bypassed in an effort to restore suction pressure. A load reduction will be performed. but it will be o 0" * ** o* (j I ABCN01 E004 I I i I I .. ==tiofl.....S oI New IlCilJe!itionMClclifiCatioriMethod:ll IIUsed.During 1raining Programl 1... .* .... == .. = ....=... ...=.. ... I I IComment> .... ..*. .. .... . ... ...* .*..... . .....* ... ..' .... '.* .***i. .***.* ..1 II. I !I I RO I ". SRO ... "*'1 List :1

• 1 I RO 54 Given the following conditions:
• Unit 1 is performing a load reduction at 1 % per minute due to lowering condenser -With turbine power at 75%, OHA G-7, ADFCS SWITCH TO MANUAL annunciates, the PO reports 11 BF19, SG FEEDWATER CONT VALVE, has swapped to -The load reduction continues to 70% power. where it is Which of the following describes how this failure will affect 11 SG NR water level, and how will it be corrected lAW S1.0P-AB.CN-0001, Feedwater/Condensate System lower than all other SGs. Manually adjust 11 BF1 9 in the OPEN direction to lower 11 SG NR greater than all other SGs. Manually adjust 11 BFi 9 in the CLOSED direction to lower 11 SG NR level. greater than all other SGs. Place SGFP Master Speed Controller in MANUAL and lower SGFP speed. Unaffected SG BF19s will respond and move in the open direction to re-establish equilibrium conditions.

lower than all other SGs. Place SGFP Master Speed Controller in MANUAL and raise SGFP speed. Unaffected SG BFi9s will respond and move in the closed direction to re-establish equilibrium conditions. I Application I Salem 1 &2 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 mitl ate the consequences of those abnormal 0 eration: __________________ """'"" Failure of feedwater re ulatin valves 55.41 (5,7.10) OHA G-7 will swap the valve to manual, and control will still be available to the operator. AB.CN states that if a BF19 has swapped to manual, you are to establish control over the valve and stabilize SG NR level lAW SG Programmed Level (Att. 2). The valve failing as is at a higher power level that the other BF19s are currently controlling(75 vs 70 %j will cause 11 SG NR level to rise, since feed flow is > steam flow. Lowering SGFP speed would work as described. but it would not be done to pertubate the . .. .. . I Main Feedwater/Condensate system Abnormalit lis 1.oP-AB.CN-0001 II 1112 I 1 18 I I II II II II I I Ii II II II I .ell IADFWCSE012 I I I I I II i j I-N...;.e-w--------I1 I D" [Question Source Comments I I .. . ' ..... ..... ....... I i

I tsROSv:ten;;Evdluli({I1'list' Given the following conditions: Unit 2 tripped from 100% -Neither MDAFW pump started or could be -Total AFW flow Is 23E4 -SG NR level on all SGs is 14% and rising 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, FEED -S/G LEVEL CONTROL 061 000K503 !'ijystem/EvolPtiQh:;J;itie! j Auxiliary I Emergency Feedwater 11 061 I Knowledge of the operational implications of the following concepts as they apply to the Auxiliary / Emergency Feedwater Pump head effects when control valve is 55A1(5,4) 23 AFW pump Terry Turbine has its governor set to maintain a certain speed, not discharge pressure. As the AF11 i valves are throttled shut, the discharge pressure of the pump will rise, and remain at the new higher pressure as less work is required of the turbine. A and B are incorrect because speed demand will remain constant, and discharge pressure will rise. C is incorrect because discharge pressure will rise.

';( FI

=S2:::.0:::P=-::::S:::O=.A=F-:::O=OO=1====:=,' [ 1[19 1[35 !lin Service Testing -23 AFW Pump II S2.0P-ST.AF-0003 iI II 11 48 1 I_A_FW_s"-ys_te_m_L_es_s_on_p_la_n _____---l11 NOS04AFWOOOO 1138-39 11 9 1 I AFWOOOE004 I AFWOOOE008 I I_N_ew______--'1 ]I " ,_______________________________________ ......J ..*.*. >'.' .:..... *...** ***..*. I I I J I I

• .. ,

IR056 Given the following conditions: -Unit 2 is operating at 100% -2C Vital 4KV Bus is aligned to 24SPT (breaker 24CSD -Power is lost to 2C Vital 125 VDC -Prior to restoring power to the 2C DC Bus, 24 SPT is Which of the following describes the status of 2C 4KV Vital Bus for these conditions? [J I from the 2C EDG. I I Deenergized with all in-feed breakers tripped. I I Energized from 23SPT (breaker 23CSD closed). I fdll Deenergized with in-feed breaker 24CSD closed. I *ij\hsvy¥rll d I R I I Comprehension I [Salem 1 &2 I 9/26/20111 062000K103 11 0 55.41(7) DC power is required to operate relays and contacts for the 4KV vital bus breakers. When DC power is lost, breakers remain "as is". The EDG breaker can not close onto the bus even though it is deenergized because one of the interlocks to shut EDG output breaker is both infeed breakers open. The other (23) SPT cannot close its infeed breaker to the bus because it has interlock which requires the other SPT's in feed breaker to be I DCELECE013 l___--l II 1()1I.... I Previous 2 NRC Exams Editorially Modified 1 Training Progr,:ltllQ 0 jQuesti()risourcec()lnmentsll"J" ILOT RO NRC Exam -August 2008. Added "breaker" and "closed" to original distracter b, and swapped places ...---.----with distracler c due to its being longer now. Added words to lessen confusion and make it the same style as **.*..***.*.**

.< **** ....IComrne l1 t .... . " I I I I I I RO ROSvstem/Evolutiori't.ist

'I" '_Outline Chanaes "1 ____________________________________________________________-J 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? on the bus causes a higher level of current to flow in the system. Individually deenergize each load on the bus, then re-energize if that load not the source of on the battery associated with the bus causes a higher level of current to flow in the determine if the I/S is the cause of on the bus causes voltage reading on the bus to become unreliable due to the excessive current flow. Transfer to the backup battery to determine if the lIS charger is the cause of the on the battery associated with the bus causes voltage reading on the bus to become unreliable due to the excessive current flow. deenergize each load on the bus, then re-energize if that load is not the source of the jA;;wGr'j Memory 1 & 2 1 I 9126/20111 I D.C. Electrical Distribution Ability to (a) predict the impacts of the following on the D.C. Electrical Distribution and (b) based on those predictions, use rocedures to correct, control, or miti ate the consequences of those abnormal operation: Grounds 55.41(7) The ground detection procedure has operators isolate individual loads. The ARP for low battery voltage has operators transfer to the standby battery charger if bus voltage is low, and battery current is so these are plausible distracters. The . .cJ bus voltage is higher than the battery voltage, so a ground on the battery would not cause bus current to rise. 113 1 DCELECE008 [Material j I IJ In",'",Hon 80",r,e;11 Previous 2 NRC Exams I JI Editorially Modified I Training CJ INew for "J" ILOT RO Exam -August 2008 Reordered distracters based on length of answer. [Comment . ., '.' ." ... .. ..... .... .... . ...., *******1 I I i 1 1 1 I

,( SRQ 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 run until 22 Diesel Fuel Oil Storage Tank is empty. trip because its fuel oil supply pressure was lost when the EDG Fuel Oil Pumps lost power. trip because its lube oil supply pressure was lost when the EDG Lube Oil pumps lost power. run until its Fuel Oil Day Tank empties due to the loss of power to the Diesel Fuel Oil Transfer Pumps. I Emergency Diesel Generators Knowled(le of bus p_ower to the following: I Fuel oil pumps J 55.41 (7,8) A is correct because there are 2 Diesel Fuel Oil Transfer pumps, powerec from A and B Vital 230V, each of which 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. (,';" . h "h' . *0, _ ,h "h' -h ;" ;n . ",h"n +h "'nr,,' ,h f. .. starting. the Lube Oil pumps for operation are shaft driven" mechanical pumps. D is incorrect because there is stili power to one DFO xfer pump. which will fill all the EDG Day tanks based on level signals. 1 ,;.Ref;erence,;fitIEl.:. ( ...... ;; ,;'.,"j . 111 & 2 Units Diesel Engine Auxiliaries 11205241-1 i I II 11=14=2==.::; I1 F=&=2=u=n=ils=D=ie=s=e=1 G=e=n=er=at=or=F=ue=I=O=il I r 1 i 1=11=1I,;...:Fu:;,.::.:el-:,o..;,..iI _______,--'1211307 iI II i 1_22_..... ... ... I EDGOOOE013 I I EDGOOOE005 I I (Material Required for Exanl.inatioll, I i II II LUsed 0 '" " __ '. C" ** _ ' ** -_ -__ -, ***.....J .. . .' ..... i iI...OIl """" :. i 1 1 I m RO .. SRO ______________________________________________________________________________________Given the following conditions: -Unit 2 is operating at 100% power when OHA C-1 GAS ANL Y TRBL is received the control -The NEO sent to investigate reports local alarm 8-3 OXYGEN HIGH/LOW on Disposal Gas Analyzer PNL 110 is in -Local indication for in service Waste Gas Decay Tank (WGOT) 02 concentration is 4.1 lAW Tech Specs, which choice describes what action is REQUIRED to be performed, and why? I Place the Standby GOT in service and commence preparations to release the affected GOT. I I Place 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 l.b elow2%. I !Reduce the oxygen concentration of the in service WGOT without delay to prevent potential releases of radioactive materials due to explosion I of the GOT. @J Immediately suspend all additions to the in service WGOT since the 02 concentration is above the 2% required to sustain combustion when I mixed with hydrogen. r=:J m==i Memory I &2 I 9/26/20111 c 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 to reduce the 02 concentration to <2% in 48 Distracter 0 is incorrect because while the immediate suspension of additions to the waste gas system is REQUIRED when 02, the flammability percentage is release. C is the correct answer because the Tech Spec REQUIRES the reduction of 02 from 2-4% to less that 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 A is in because TS 11=28=2====,I------------------------ __________ I I following Area Radiation Monitors (ARM) will cause a ventilation system alignment change when it reaches its High Radiation Alarm 12R1A' Control room. 12R9, New Fuel Storage 12R32A, Fuel Handling Crane. . 12R52, Liquid PASS Room. EJ" [K:J I Memory II Salem 1 & 2 I I I.. 0 I Area Radiation Monitoring System I ,_ .._,..rovide for the followin Plant ventilation ...*. a.*.*..

o ,.:!;...... , 55.4.1(11) .*. .* *. I is incorrect because it has It is if it is con:used ,with the 2R1 B: w,hich is a IAnswers:

monitor which realigns control room venlilation on high radiation, B IS correct because It realigns FHB venlilatlon through ' charcoal filters and starts both FHB Exhaust fans. C is incorrect but plausible because its auto function is to prevent Fuel except in downward direction. Dis incorr:ct since it only has light outside the PASS room whi?h activate,S, l' 1 RMSOOOE005 IMaterialR,equiredfor ExamirlatiOn ',II II New I.*11 llUsed 0 fQuestion I .. ..' I I I

I" RO 61 Given the following conditions: Both Salem Units are operating at 100% power. -2R1 B-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? are designed to remain in Normal Mode upon a loss of power to a single duct radiation monitor. No actions other than normal Corrective Action Program actions to troubleshoot and repair the power supply are required, with no speci:1c time limitation. are designed to remain in Normal Mode upon a loss of power to a single duct radiation monitor, The 2R 1 B-1 must be restored to operable within its Action or CAV must be in Accident Pressurized Mode on BOTH Units. should have swapped to AP Mode. Initiate AP Mode of operation by depressing initiating pushbutton on 2RP3 lAW S2.0P-SO.CAV-0001, Control Area Ventilation Operation. Initiate AP Mode of operation by aligning individual system components to their correct positions on 2RP3 Application Salem 1 & 2 55.41 (7) Any R18 channel losing power will automatically initiate Accident Pressurized Mode on 80TH Units. B is incorrect but 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 RP3. 0 is incorrect because individual components are not aligned, still plausible if the candidate thinks the failure has affected the whole , ' FI S2.0P-SO.CAV-0001 JI 11 6 . 23 11 38 I 'i=ls=al::::em=Te::::ch=s=p::::ec=s=======:::;I,f;:1 ==========iI;3.3.3.1 II II il 113/43-38 II 1[282 II I I I RMSOOOE003 I I I I I i IMaterial Required for Examinaflonij II .*.*.. .*.=lJe=s=t:::iO=.h=M= .*I so.urc,e SOll1ments l \ I .Ie",,,,,,,,,,,,, ". "'..' ..... . ***********.21.*** .......**..****...

• .***** *********...*..... I I

I ... I' I 8uestion Given the following conditions: Unit 1 is operating at 25% power. 1 B EDG is running in parallel with station power on 1 B 4KV Vital Bus. 13 and 16 SW pumps are in service, 11 SW pump is in 1A 4KV Vital bus becomes deenergized due to a Bus Differential 1 minute after the 1A 4KV Vital bus deene izes NO nne,""t,,," action which of the contains ALL the SW 115, 16. 1.qJ I 076000K201 li K2*01 1 Service Water 11 076 I Knowledge of bus power supplies to the following: , Service water I On a single bus UV as described in the stem, only that bus would load in blackout I I.E.,*.. p.... .**.r.,..,.a.:.t'.o.[I.***.'.*.O.*...f. I 55.41(7) A bus powers 15, and 16 SW pumps. loading. A bus is locked out on Bus Differential (deenergized), and the loss of 16 SW pump would cause header pressure to lower to where the auto pump (11) would start. Only one SW pump is aligned for AUTO which is the normal at power configuration for the SW pumps, one In auto, and the rest in manual. 12 SW pump would never start unless 11 pump did not on a SEC initiation, that is "h, ;t ;" nnt 1,,+0'-; ;n "n' nf tho -hni ,c 1 A <::11\, n, ,no" "'" ,1'-; ""I "bel cinC'o R h, ," n"","'" I"""" "A' ",," >Am;,.." 1<0 >Am" It ;cn" Iido,-i In any of the choices. There can be confusion about 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 pump on each bus. "e:5";"" ,I I Service Water Pump Operation II S1.0P-SO.SW-0001 II I Unit 1 4KV Vital Buses One line 11 )I I I F 3=4==.. 11 ___----111 11_---' I SWBA YSE005 1_-'-------' I Facility I J-ROC61 Changed which pump is in auto (11 instead of 15) and that makes a different choice correct. I . ....ic\,..........j * ....... 'x .... *..*.... , . .I I I

...... I 'SRdSyslemiEvolutjonLiSf TOP5J I RO 63 Given the following conditions: -All 3 Station Air Compressors have become -The NORMAL cooling water supply to the Unit 1 Emergency Control Air (ECAC) has been Describe the status of the Unit 1 ECAC. o eralion of the Unit 1 ECAC ... can continue since cooling water wiJI automatical1y swap to Demineralized Water through a check valve. must be discontinued until cooling water can be manually aligned through a spool piece from Service Water. must be discontinued until cooling water can be manually aligned through a spool piece from Demineralized Water. s between Instrument Air System and the following: , 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 Water system, SERVICE WATER can be supplied by MANUALLY 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.) UO..Number**** 'I I CONAIRE007 I I I I I MC'iterilil! RequiredforExaminationJ I I) iQuestion Source: ,I Facility Exam Bank I M6dlfication Metl1od;,'j Direct From Source I 0 i I_V_is_iO_n_Q __80_7_0_0______________________________________________________________ ---I ........ { I II I II I II I R.QSkYScraperl

• ..

..* *1 SRO.SvstemIEvolutionU'it* .1 Cl1anqes* I 1 RO 64 Given the following conditions: -Salem Unit 1 is operating at 100% power. OHA A-15 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. I !Fire Protection System iE.X.PI.a.h.'.a. .* .*.***..*.o{11 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 .ti.:O,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. Distracler A is incorrect in that the #1 pump will start as header pressure lowers, and the #2 pump will not start. Distracter B is wrong because a major pi ping rupture will cause header pressure to . . . . !FIRPRDE007 I F1RPROE008 I FIRPRDE009 I l&'Iaterial Required for Examination .'J I II I Facility Exam Bank I Direct From Source I Prograrryj 0 iQuestion I Vision Q67854 .'. .... . ....... I I I I I . RO Ii ____________________________________________________________________________________Given the following conditions: -Unit 1 is operating at 100% -Charging flow rose 1 gpm in 5 minutes and is now steady at 88 -Computer trends show the increased RCS leakage started 10 minutes The CRS enters S 1.0P-AB.RC-0001, Reactor Coolant System Based on elevated 1 R11A indications coincident with the rise in charging the crew determines a small RCS leak in containment is 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, Place 2 CFCUs in slow speed and 2 CFCUs in high speed to prevent containment pressure from rising to the automatic Safety Injection setpoint. Place a centrifugal charging pump in service to ensure sufficient charging flow margin is available to maintain PZR level stable should the leak rate rise. Place 2 CFCUs in slow speed and 2 CFCUs in high speed to minimize the rise in containment humidity and prevent equipment damage from elevated moisture levels. 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. 1103000A101 I ____________________________________________________________L-__[KA Abili to redict and/or monitor chan es in arameters associated with 0 eratin the, Containment System controls Containment pressure, tem eralure, and 1.,J: *.;tiO....*.n.' '.*. 55.41 (9) AB.RC-1 Attachment 1, Continuous Action Summary, 4.0, states that at any time if Ie.ak is suspected to be in containment,..*,....x.p.*. la.".,....a.' ......,c,.,.*.4 Answers;', .:,.' then place 2 CFCUs in slow and 2 in fast. The Technical Bases Document, page 3, states that this is "to prevent an -. Safety Injection and minimize the potential for off-site releases when the leak is in Containment." B is incorrect because while it the right action, the reason is wrong. Distracter C is incorrect but plausible, because a centrifugal charging pump is placed . .... . I incorrect because of the wrong reason for CFCU operation. 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, Hi h RCS Activit. ____________ --11_________--' ______.....J I----J I-----l 'ii.! I CONTMTE003 I ABRC01 E003 iMaterial Requi 'f,?!" ExamiO<itign III I Facility Exam Bank I j[ Concept Used

0 .

I Vision Q85538 I * ......*... ..... . ..... " '.' ,', i'; ..' ***.1 II. I II I II

• • • • RO'sNscraller"1
• ... sRoslWscra6er

,._, .n. -.'--. , Questi.on Topic j IR066 I I Of the following. which would be considered a Core Alteration lAW S2.0P-IO.ZZ-0007. Cold Shutdowll to Refueling? I I When the first stud (during first pass of Reactor Head detensioning process) is detensioned. I I When the RPV Head is lifted (1-2') to check for CRDM drive dis-engagement. I I Insertion of a camera to the level of the RPV flange prior to fuel movement. I Lifting of the first fuel bundle in the I I I d I R I

• • .,.!e:tevel"l I Memory IIFacilit'l:.11 Salem 1 & 2 1 9/26120111 i >-"',,' >1 [FA:l1194001G136 2.1.36 J ED PWG I 11 11 0 11 I I I 55.41(10,13)

D is correct because "CORE ALTERATION shall be the movement of any fuel. sources, or reactivity components, within the reactor vessel with the vessel head removed and fuel in the vessel." A is incorrect because it is MODE 6, Refueling, is officially entered. B is incorrect because Core Alts cannot occur until the vessel head has been C is because "Refueling. activities that would constitu,te a ALTERATION do include the movement components, except wihen these d-evices would result in the movement or manipulation of fuel. sources, or reactivity control components within the Reactor Pressure Vessel. I 1:=17== ____________ I REFUELE012 jlOP007E004 ..' .' [\";" "'" ! I 1 I ...." .. RO i Given the following conditions: -Unit 2 is operating at 66% -Power was reduced when 22 SGFP tripped 2 days ago, and has remained at level for 2 -Operators are preparing to start a Main Turbine power ascension using dilution automatic rod control to maintain Tavg and AFD on -Tavg-Tref deviation is lAW OP-AA-300, Reactivity Management, how should the power ascension be started? Assume this is a normal power ascension with all equi!2ment available. IThe crew concurrently initiates the Main Turbine load ascension and a ReS dilution. I I The crew initiates a ReS dilution. As soon as the dilution is in progress the Main Turbine load ascension is initiated. I I,The crew initiates a ReS dilution and waits until a ReS rise is detected. Then the Main Turbine load ascension is initiated. I I The crew initiates the Main Turbine load ascension and waits until a ReS temperature lowering is detected. Then the ReS dilution is initiated. I r:::--1 [1;':;;;;, .i"';""" "I r;:;-----j " ,." I **1. 1"";,. -{FlICiI!Jy;HSalem 1 & 2 9/26/20111-,-.... .. ... ." !________________________________________________________________________ Knowledge of procedures, guidelines, or limitations associated with reactivity management. 55.41(5,10) Listed under the responsibilities of the Reactor Operator, page 8,4.6.5, "Typically, during planned load changes where ,,,".. dilution or boralion is required, start with the dilution or boralion. The initial effects (ReS temperature change) of the change should be seen prior to initiating the load change." All the distracters have the correct actions in the wrong I

• • ,1 t;j' ..
• .**un ",,:,1

'" ''.'], N6;eJ I RA:.ldivif\, m","",),j-""- len', Ilgp-AA-'lon 18 114 I II II I II I! I" /RXOPERE018 I I RXOPERE020 I I .. ..,g:'*:p=ro='g=r=a:rrll= I If()rn mellt , "', I I .,' I: ..... ,' 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, The component must be declared INOPERABLE at the time of discovery because the 24 hour tme limit allowed past the normal surveillance interval has been exceeded, The component remains OPERABLE because Technical Specifications allow a 25% time extension of the normal surveillance interval for surveillance which has not been exceeded. The component remains OPERABLE ONLY for the next 24 hours after discovery, during v.hich a SAT surveillance must be performed to ensure OPERABILITY, otherwise the component must be declared INOPERABLE. The 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 OPERABL.E. ______________________________________________________________________________________Knowled e of surveillance procedures. 55.41(10) Tech Spec 4.0.2 states that... "Each Surveillance Requirement shall be performed within the specified surveillance 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 Surveillance is not p'erfonmed within the delay period, the Limitin\l 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 sse of that system with regards to how it affects the component which has exceeded its monthly surveillance interval. --_________-ll________---JI_____-l __---' I I I I TECHSPE014 1------' RO RO 69. 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 -Leakage to the Reactor Coolant Drain Tank is 1.3 -Total primary to secondary leakage is 0.125 Which one. if any. of the following Technical Specification leakage limits has been exceeded? Knowied e of limitin conditions for operations and safe limits. EXP.*Ia.n.a*. ..* .**.*.*. ...**.........f..*...55.41(5,10) Modified stem conditions to make a distracler (primary-to-secondary)correct, and the former correct .*.*t.*'.o..**.*..*.n ..o .* Answers:*.* *.* answer incorrect. Salem TSAS 3.4.7.2 states the limits on RCS leakage to be 1 gpm for unidentified leakage, 10 gpm for identified '.,-..' leakage. and 150 gallons per day through anyone steam generator. The 0.125 gpm prHo-sec leakage is 180 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. ITECHSPE015 I I I I I [Material Reguired for Examil'lati()n i I II ! I Facility Exam Bank Metho.d:il Significantly Modified IQuestionSciurc:e C.t:)/11:r§ri§ I Salem 2002 RO NRC Exam (4 NRC Exam ) I ......... .........' .**.. '.' **.. .*. .....* ...... *....... .'> ......... .**1 il I II I I I

r SROSkvscrape( I I I [QUIi;rtionT6p-iCll RO 70 Action? 26 SW pump is crr. ALL other SW pumps remain performing a shutdown. 2N32 Source Range NI control power supply fuses 21AF21 AUX FEED -SIG LEVEL CONTROL VALVE is discovered in the jacked shut 2PR3 PZR Code Safety Valve is declared INOPERABLE with 2PR4 and 2PR5. Code Safety Valves I Application I fFaci'ifir;11 Salem 1 & 2 arameters which are ent -level conditions for technical s ecifications. 55.41(10) In MODE 4. only ONE PZR Code Safety is required to be OPERABLE. 3.4.2 In MODE 1,2 independent SW loops are required. An OPERABLE SW loop consists of one pump from each vital bus plus 2 pumps per bay. 3.7.4. SO.SW-OOOB. Att 2 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 ... ,ISalemTechSpecs II 11=1======1,1\==:::;1:=1 ==:;.'I Service Water System Operation II S2.0P-SO.SW-0005 iI i I 11 40 II----------JI 1-------,1 I II 11_---' ,.. _._ u" _I [TECHSPE015 II ilQuestion Sq ur c: e:'m 1 I Facility Exam 8ank I.lVIethod:l Direct From Source ...---'-.-"-' ..-'---------' ".J--._R-:-::-Oc:-=-flf:)-----------------------------------7 I 1 ! !e.......... ".., ."".'. ........:, .:: .', '. .., ,'. .'. '. .'.',C,> .***.1 !I I II I 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 ____________________________________________________________________________Knowled e of radiation exposure limits under normal or emer enc conditions. 55.41(12) B is correct because: The Current Annual TEDE Administrative Dose Control Level is 2000 mrem for PSEG Nuclear 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 maching 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); C is . .., . . I Control Level (2000) and NO transit dose. I RADCONE002 1___---1 I L Material RequIred for EXamination:':! I II lin,. ... Facility Exam Bank From Source : I Vision Q44899 .... . .... ' ................. ,Cor..."'**1L .............. ! I I

.1 I.::.2,Lflirle 'I I RO 72 Given the following conditions: -Unit 1 is in Mode 2, with all Shutdown Banks withdrawn. -Unit 2 is at 100% power -All Unit 1 & 2 Circulating Pumps are in service 21,23 & 26 Service Water Pumps are 21 cves Monitor Tank is being released via 22 CCW Hx to the Circulating Water Choose the condition that would termination of the release. IAbility to control radiation releases. .E.X. p.....'.8.p. ........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 ..f 22 SW I header discharges into the outlet of 1.1 A and 11 B circulators. A Unit trip not an "emergency", so the requirements of OP-SA-106-101-2001 Operating With an Emergency on the OpPosite Umt IS not applicable. Normal SW header pressure IS 125, the SW flow from the remaining 2 pumps in service will maintain flow to the CCI--lX. Trip of an operating CCW pump would I r"""lt in ..,,.., ""t, d-ort ,f tho ,t* n ,,,, n ..,,..,rl th" ,Y'\A, c"eto"" ro,." ",in" in coni,.." I * """'" II I RELEASE OF RADIOACTIVE LIQUID WASTE 1\ S2.0P-SO'wL-0001 II II 1124 I i II II II II I I Ii J I II II I I WASLlQE012 II I Facility Exam Bank I. JI Direct From Source Ijused During Trainlngj=lrograin I 0 i I VISION Q119138 .... ...... ....., , **** I , I I I

. SFl.O SKyscraper, F RO SROSvstem/Evolution

-. """ .. ," LQuestion Topic IRO 73 I 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. WhZ will the operators be instructed to terminate SI and start a RCP if possible? I The soak required by FRTS-1 requires SI to be secured and a RCP to be running to provide ability to use spray to depressurize primary. I.The 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 ReS cools. JSafety Injection flow is a significant contributor to any cold leg temperature decrease or overpressure condition and must be terminated. A Rep is started to minimize temperature gradient across S/G tube sheets. Safety Injection flow is a significant contributor to any cold leg temperature decrease or overpressure condition and must be terminated. A Rep is started to provide mixing of cold SI and warm reactor coolant water. 1 d II R I I Memory I I Salem 1 & 2 9/26/2011J 1194001 G418 11 III! 0 ,..., I I ,...-_____________________ .____________________ ..,..-,. I Knowledge of the specific bases for EO?s. 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 terminating SI. B is incorrect because the soak is not the basis for terminating SI. 'C is incorrect because the 81 basis correct but the basis for Rep start is not correct. 0 is correct because Basis Document states for step 9 starting a Rep... "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 Rep * ".. * *** * ** f1 I have contributed to the ReS cooldown or mat prevent a sUbseguent reduc1ion in ReS1____________ -' _________ ___*__....J 1 __--' I I ,. I Given the following conditions: Unit 2 is operating at 100% OHA A-7, FIRE PROT FIRE, Panel 2RP5 is checked and indicates the

• Zone 59 -Air and Water Deluge, Containment EI. 100 Panel 335 is -Zone 74 -Smoke and Fire Detector, Containment EI. 100 Panel 335 is Which of the follOwing describes how the Control Room crew should respond? Verify OHA A-15, FIRE PUMP 1/2 RUN is in alarm signifying a Diesel Fire Pump has auto started to supply Fire Protection water to associated valves in containment.

Dispatch a NEO to open the associated deluge valves at their outside containment penetration location. Dispatch a NEO to open the 2FP147, FP CONTAINMENT IV. @illiopen the 2FP147 from the control room. I d I EXclrll"Lj[v;U I R 1 I Memory I Salem 1 & 2 ________________________________________________ .____________________________Knowiedoe of "fire in the plant" procedures. i IE...,..'.x.*.**.p t.)#.**.n...*" p.*'.f.,.'.1 .., ... ... *. .., .* *. 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 IIAnswers:**'. .* ""I control room on 2RP5. B is incorrect because the deluge valves are automatic and in containment. C is incorrect because the . . FP147 is opened from the control room. 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 in the ARP. I No.1 & 2 Units Fire Protection 11205222-2 II I f::l 60 1 1_________11 ______1[ ____----111 11_----' 1 FIRPROE004 I I FIRPROE008 I .. !Question C9mmehts]I Vision Q6980? modifi.ed stem to make it w?uld control room crew do instead of what is the status of the FP I --.... ..----.. I.system. Modified chOices and added valve Identifiers. 1 C.....III1""".:, ...... ..>.. , '.' ' " , :". ... ..":J"", t II J II 1 II J . RO SkyScraper .'. ...*.. I j;lutlirle An Alert has been declared at Salem. Which of the following identifies the PRIMARY method which the Primary Communicator will use to make notifications to the States of Delaware and New and how long from the Alert declaration do they have to make those notifications lAW Attachment 6, Primary Communicator of the I NETS phones within 30 minutes. I ESSX phones within 15 minutes. g:; I ESSX phones within 30 minutes. a I I R I ...... communications s stems and techniques. 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 closed circuit system, which is used as a backup for NETS. The notifications to the Sates must be made within 15 minutes of the declaration of an Emergency. [..".

• '1 I Salem ECG -Attachment 6 II II II r

=:::; , Emergency Preparedness Training Communicat I TYSC I II 11 04 I I ... ' T",'. )\,1.1.,O(t-ilir@(U**** . I GENISSE013 I I I I I Inll.... ....:...1 ..**.. Comm.=ritsll I ;1" '" ...*.*....... > .... '.. '.' ......./ ....**....... ..... i"' ........

• I I I II 1

'\:;' .. .. ",. _ ..._.". ..-........ --... [9uestiofl TopiC] I SRO I Given the following conditions: -Unit 2 is in MODE 4, performing aRCS heatup and pressurization lAW S2.0P-IO.zZ-0002, Cold Shutdown to Hot -RCS Tavg is -RCS pressure is 310 -23 RCP is in -21 Charging pump is in -23 CCW pump is -2A and 2C 4KV Vital buses are powered from 24 -2B 4KV vital bus is powered from 23 -24 8PT loses power, and NEITHER 2A nor 2C 4KV vital buses transfer to alternate due to faulty power available relay for 23 1-2A vital bus locks out on bus -28 EDG fails to Which of the following describes how the control room will respond? TheCRS will. .. BJ I enter S2.0P-AB.RCP-0001 and trip 23 RCP 5 minutes after the receipt of OHA's 020-023 21 (22,23,24) RCP BRG CLG WTR FLO LO. I I enter S2.0P-AB.RCP-0001, Reactor Coolant Pump Abnormality, and trip 23 RCP based on the loss of seal injection and thermal barrier flow Ito Rep;;.,enter S2.0P-AB.CVC-0001, Loss of Charging, and check 22 charging pump has started to supply adequate seal injection to allow continued RCP operation. I [Eill.enter S2.0P-A8.CVC-0001, and take actions to start 22charging pump prior to 23 RCP seal package temperature rising to the point where seal injection flow is prohibited and 23 RCP trip is required. [CJ i II Application [Salen1 & 2 I 9/26/2011J ... . -;r , '.,,: ] 55.43(5)The initial electrical lineup, combined with the conditions occurring in the stem, will result in only 2C 4KV vital bus having power. With the RCS temperature < 312', all charging pumps except one are CIT. 21 charging pump is powered from 2S 4KV vital bus. NO CCW pumps will be running since the only powered 4KV vital bus CCW pump is crr. With the only operating (or charging tripped, the::re is no seal flow to RCPs and no CCW flow to RCP barn.ers. The CAS 5 minutes of the alarms annunciating, go to Attachment 2, Tripping RCPs, the loss of both seal injection AND CCW directs Rep 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. 0 is incorrect because while the action will be taken to restore a char in to available status, the CAS action of AS.RCP will be the nori ,and seal acka e temperature will not stop that action. I Revision ! fL..o, Number I ABRCP1 E003 "" I<.>PiC] I SRO 1 Given the fol/owing conditions: -Unit 2 is in Mode 5 with 21 Residual Heat Removal (RHR) pump and HX in service for -The RO reports that Pressurizer (PZR) level is slowly lowering -NO Overhead Annunciator alarms have been -Refueling Water Storage Tank (RWST) level is -21 Waste Hold Up Tank level is rising Which of the foliowinq identifies the procedure which will be used and the action(s) taken in that procedure which will isolate this leak? i I.S2.0P-AB.RHR-0001. Loss of RHR. Close 2CV8, RHR Letdown. I IS2.0P-AB.LOCA-1, Shutdown LOCA. Close 2eV8, RHR Letdown. I IS2.0P-AB.RHR-0001, Loss of RHR. Place 22 RHR pump and HX in service and isolate 21 RHR loop. II JS2.0P-AB.LOCA-1, Shutdown LOCA. Place 22 RHR pump and HX in service and isolate 21 RHR loop. I [Al'l s.v e f'1 I S I I Application I[FaCilitY] I Salem 1 & 2 1 9/26/20111 I 000025G406 1 12.4.6 __ 111sRoGroup;l\ 11 !lIt ,Syste@$vv,L/tiO'l.Tjtlel I Loss of Residual Heat Removal System llo251 _______________________________________ ---:, 55.43(5) The 2 AB.LOCA distracters are incorrect because AB.LOCA is applicable in MODE 3 with Accumulators isolated or MODE 4. With MOOE 5 indicated in stem, it is the wrong procedure, although it has the correct action in distracter B. A is correct because the 2CVa will be isolated and it is the correct action for a leak which is causing the WHUT leve to rise. Distracter C is incorrect because it has the wrong action, with the right procedure. The action in the AB.RHR is to remove BOTH loops from service and ..... I ABRHR1E005 I IABRHR1E004 I I I .*.. J I '/ I Facility Exam Bank 1[§uestionMO;lifiP'1!ti0 rt Method: Modified 1[USed During Trainini;IProgram! 0 Isalem 2003 SRO NRC Exam. (5 NRC Exams ago) Modified to Include which procedure to use. IIComment ..... ...*.... .... ................ , i I I I < -< <<< ! I SRO 3 Given the following conditions: -Unit 2 is operating at 21 % -Turbine load is being raised at 10% per -Operators observe the -OHA E-28 PZR HTR ON PRESS LO -PZR level is 28% and lowering Control Rods begin stepping out in automatic at 8 -OHA C-38 CFCU LK DET HI -Rx power is rising at 0.35% per Which of the following identifies how these conditions will be addressed? Enter S2.0P-AB-RC-0001, Reactor Coolant System Leak. If VCT level cannot be maintained 11 %, swap charging pump suction to the RWST 1 and trip the Rx, confirm the Rx trip, initiate Safety Injection, GO TO EOP-TRIP-1, Reactor Trip Response , Enter S2.0P-AB.STM-0001, Excessive Steam Flow. Trip the Turbine, initiate AFW. lower power to <5% lAW S2.0P-IO.ZZ-0005, Minimum << Load to Hot Standby, close all MSIVs. I Enter S2.0P-AB-RC-0001. If PZR level cannot be maintained stable or rising. trip the Rx and initiate Safety Injection, GO TO EOP-TRIP-1. I Enter S2.0P-AB.STM-0001. Trip the Reactor. confirm the Rx trip, initiate MSLI, initiate Safety Injection, GO TO EOP-TRIP-1. I,... ,,,<,! .. r::;----j I",. ,,; ., & ,.,. 1 & 2 1I;xamDate:.U 9/2612011 I-=---...J I-=---...J r---m -----1

000040A202

[5<l,,{,arue,: I Steam Line Rupture Conditions requirin a reactor trip ... O..f.l 55.43(5) D is correct because the indications in the stem are of a steam leak in containment. Rx power is rising at -21% per hour, *. .. while the turbine load increase is at 10% per hour. AB.STM states that if Rx power is rising uncontrollably. take the actions described in D above. A is incorrect because only some of the indications present in stem would be present for aa 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 I New I Source c:ommentS] I Given the following conditions: -21 CVCS Monitor Tank is in 21 CVCS Monitor tank will be released via the Waste Discharge Cross Connection to Unit Liquid Radwaste system lAW S2.0P-SOWL-0001, Release of Radioactive Liquid Waste from 21 CVCS Monitor 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 and action to the release? Unit 1 CRS authorizes a tagging request which results in the 1 Ri8, Liquid Waste Disposal Unit 2 CRS authorizes a tagging request which results in the 2R18, Liquid Waste Disposal Rae: monitor losing its power source. CRS authorizes a tagging request which results in the 2FR1064, Rad Waste Liquid Monitor Pumps Discharge flowmeter losing its CRS authorizes a tagging request which results in the 2FRi064, Rad Waste Liquid Monitor Discharge flowmeter losing its source. I Salem 1 &2 'Accidental Liquid Radwaste Release ______________________________________________________________________Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. 55.43{4) The Salem ODCM is a supporting document to the Unit 1 and Unit 2 Technical Specifications. The previous LCOs that were contained in Radiological EffiuentTech Specs are now included in the ODCM. Using ODCM 3.3.3.8, lAW Salem Tech Specs 6.8.4.1.g.1, and Table 3.3-12 on page 19, the RiB and the FR1064 required to be operable. The R18 is interlocked with the WL51 so that if the R1810ses power, the WL51 will shut, preventing an unmonitored release, and no operator action would be required to . . . ", , between the flow recoder and the WL51, so the release would be considered unmonitored based on the fact that a required component was not available, and the release is ongoing. Loss of the flow recorder during the release requires operators to close the 2WL51 lAW step 5.5.9. The Unit 1 distraeters require knowledge of the flow path for a release which is directed through the cross connect line. ____________ _________________________ 11 ____ __IL.6.""un1Qer lJ .1 I WASLlQE009 I I I I I

,: ,,:""'Y "',,'" ,':.X'":

-'-,';i, SRO Skyscraper' .. .... -... _... " . .. "",.,, Topic 1 I SRO I Given the following conditions: -Unit 1 is operating at 100% power. -21, 24 and 26 SW pumps are in service. -21 and 22 SW header pressures are 108 psig. -The following OHAs annunciate within 10 seconds of each other in this order: 13,21 SW HDR PRESS LO 14.22 SW HDR PRESS LO B-15. TURB AREA SW HDR PRESS LO. -B-48, SW VL V RM FLOODED. The standby SW pump starts automatically. and OHAs B-13, B-14, and B-15 clear. Which of the following describes the status of the SW s},:stem, and which erocedure will provide direct actions which will mitigate the event? I A SW leak upstream of the 2ST901, TURB LO CLR ST RET V has occurred. S2.0P-AB.SW-0002, Loss of Service Water-Turbine Header. I I.A SW leak upstream of the 2ST901, TURB LO CLR ST RET V has occurred. S2.0P-SO.SW-C001, Loss of Service Water Header Pressure. I I£J 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. I @j 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. I >'iJ Id 1 1S 1 Application I Salem 1 & 2 I 9/26120111 -. '. ., r"'" ' *. "",1 000062A202 I Loss of Nuclear Service Water Ability to determine and interpret the following as they apply to Loss of Nuclear Service The cause of possible SWS .*.... o.J.*.. 155.43(5) The leak location could be in the TGA with conditions in the stem for the SW valve room flooding. Knowledge .*.*'Answers:':" i**,. of where the SW valve room and what piping is there IS needed to answer question. The 2ST901 would respond on a TGA leak, L .._* and depending on leak size eQuid cause a restoration of header pressures. If the leak is determied 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 **rI f, Tr::,J\

• t, J\Q

') fro" tho Tr:01'. .. " x 1=1 L=OS=s=of=S=W=H=e=ad=e=r1 Loss of SW-Turbine Header II S2.0P-AB.SW-0002 II r 1 11/=1=1.1_o_v_e_rh_e_ad_A_n_nu_n_c_ia_to_r_W_'_n_do_w_B _____: I..;.S_2_.0_P_-_A_R_.Z_Z_-O_O_O_2 ____11 11 35 ,', I ABSW01 1 I 1 II I New I [Useet During Training.Progra.!:lJ 0 .....*. ." ! I II I II I """'--<.," ........--.. ... ,.--------"............. ,,-. LQuestionT()pic ISR06 1 Given the following conditions: 1-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. Which one of the following would [lI'event initiation of Rx defuelina on Unit 2? I Loss of Control Air to containment. I Loss of Plant Page capability in containment. I Racking down the 22 RHR pump 4KV breaker. [J I Only one of the two SRNI's can provide audible indication in the Control Room, I I..;;L.;.oS;;,.;S..;;o;;,.;f...;.l...;.ns;;,.;t...:ru..;;m...:e..;;n..;;t..;;A..;;ir _____________________________ --l .. IIAnswerl I a I I S I Memory 1 I Salem 1 & 2 I 9/26120111 0OOO65G136 1 @:1.36 1[ lIit ..'. ___' ____________________________________________________________________________KnowledQe of procedures and limitations involved in core alterations. I 55.43(7. 6) The requirement for SRNl's is BOTH operable and provideing VISUAL indication in the Control room, with ONE :A11sWe(s:i0::.,<.c: 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 ails. Onl! ONE RHR loop is required to be in operation in MODE 6. 2 RHR loops are required to be OPERABLE when <23' level above the flange. F1R=ef=ue=lin::::9:::0::::p=er=at=ion=s=======:::;lls2,OP-10.ZZ-0107 [113 1117/I I 1________--,1 1______-'iFI===::::::::;'11 ll_---J .. IIOP009E002 I REFUELE007 ___---l !CIJ......"1i .... ," .." " , .. ...*. .... .., I I J _. .-.--.--_... _."",," '::om;Siion TOPiC] 1 SRO 7 1 Given the following conditions: -Unit 1 has experienced a large Main Steam line Break (MSLB) inside containment from 100% power. 1-Safety Injection was initiated. -MSLI has failed to shut ANY MS167 and they remain open. I-ii 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. 1-ALL SG pressures are 120 psig and dropping. -RCS Tc's have dropped from 540°F to 230°F in 20 minutes. Evaluatethe data and select the Rrocedure to be entered and action to be taken upon transition out of 1-EOP-TRIP-i. Reactor Trip Response. 11-EOP-FRTS-i. Response to Imminent Pressurized Thermal Shock Conditions; shut all MSi0's md steam dUl ' valves. I [gJ l.i-EOP-FRHS-1. Response to Loss of Seconday Heat Sink; initiate feed and bleed immediately. I 11-EOP-FRHS-1, initiate feed and bleed ONLY when 3/4 SG WR levels have dropped <32%. I @J li-EOP-FRTS-1, reset Safeguards Actuation and restore normal charging and letdown. I b 1 S I Application I Salem 1 & 2 I 9126/2011/ -.. ---..-,-I ... 1 Loss of Secondary Heat Sink __________________________________________________________________________________Knowtedqe of EOP mitigation strateqies. I 55.43(5) B is correct because the conditions given in stem would transition to FRHS*1 due to a a RED path of no AFW flow and <9% NR level. The Bleed and Feed initiation criteria are when SIG WR levels are <36%(adverse), NOT 32% 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. [Re,yTSjtiilj 1 Response to Loss of Seconday Heat Sink 111-EOP-FRHS-1 II r\=1==:;1124 I Response to Imminent Pressurized Thermal Sh 111=1:::-E:::O:::P=-=F:::RT:::S:::-=1 1125 1 Critical Safety Function Status Trees j!_1-_E_O_P_-C_F_ST_-_1

_____--'II_____.___i 1 11 25 '. '"IJj I FRHSOOE005 I I FRHSOOE013 ! I I ','. .. I SRO I Given the following conditions: -A Rx trip and SI were initiated based on a LOCA 1 hour -Operators have transitioned out of EOP-TRIP-1, Reactor Trip or Safety -RCS pressure is 350 psig and lowering very -All RCPs are -RCS highest CET is 290°F and lowering very Rx power is 300 cps and -RVLlS Full Range is 95% and 23 SG NR levels are 33% and stable, and 24 SG NR level is -Containment radiation levels are 500 mr/hr and -Containment pressure is 5 psig and lowering very -Containment Sump level is 46% and Which of the following identifies the highest priority CFST for these conditions, and actions taken in that associated EOP? Reference provided. []I Yellow Path Core Cooling. Isolate ECCS accumulators, depressurize SGs to atmospheric I I Purple Path Containment Environment. Isolate Containment Pressure Relief flowpath (2VC5, :2VC6) and return to procedure in effect. I I Puq;!le Path Containment Environment. Isolate fluid sources from outside containment which have corroborating indications of lower than I I expected levels. Yellow Path Core Cooling. Check ECCS flow for current RCS conditions as expected or start pumps and align flow paths, and ensure PZR PORVs and RPV Head Vents are shut. !S I Application I Salem._1...;&;...2 -'__...l ______ I LEA2.1 I Saturated Core Cooling as the appl to Saturated Core Cooling: conditions and selection of ap ro riate procedures durin abnormal and emer enc 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 ** .* correct. FRCE-1 actions-are partially cOrrect. in' that 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 < 1.5 si cont pressure. ' I FRCCOOE002 I I I ...... ......". .. J I SRO 9 I Given the following conditions: -Unit 2 was operating at 15% power prior to synchronizing the Main Generator. -A Main 8teamline rupture occurred that resulted in multiple Steam depressurizing in containment before 2 steam generators could be isolated from 2 faulted -The 2 faulted 8Gs Tcs are reading 270"F and -The intact 2 SG Tes are 330°F and -RCS pressure is 500 psig and slowly -Containment pressure is 16 psig and slowly -All SG NR levels are -Total AFW flow is 24E4 -Source Range Nis are NOT -Intermediate Range SUR is 0.0 With CFST's in effect, which of the identifies the procedure entry required, and actions whier will be performed in that procedure? l2-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. I [EJ 1.2-EOP-FRSM-2, Response to Loss of Core Shutdown. Energize Source Range channels and verify SR SUR is 0 or negative. I 12-EOP-FRTS-1. Isolate any faulted SGs, depressurize RCS with ONE PORV to within 1 OOo/hr Cooldown Curve. I J2-EOP-FRSM-2. Establish AFW flow >44E4Ibmihr, borate RCS untillR SUR is negative. I !f.ni>vyer] I a 1 S Application 1 I Salem 1 & 2 I :ll 9/26/20111 f" "\';";'> -';); I OOWE08A201 I Pressurized Thermal Shock IIE08 Ability to determine and interpret the following as they apply to Pressurized Thermal Shock: and selection of apJ:1[opriate2focedures during abnormal and eml9rgency operations. 55.43(5) A is correct the stem conditions in a PURPLE path on Actions for maintaining AF.W flow (Step 3.5) :Answers;)1 and ECCS pump reduction (Step 12) are correct. C IS Incorrect because depressuflzlng the RCS to restore oondltlons W1thn the 1000F/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 Ih",,/h, 'h SR NI" nl'1l . 10 CliO Ie ""., . ,.-l ho "'0 " _n ') "';n"" " V::: I ('\\, ",+h

• f", *j::R<::M_

2. The FRTS is a higher priority, and is a PURPLE path. B is incorrect because it is the wrong procedure with' the correct actions of that procedure.

0 is incorrect because it is the wrong procedure, and the actions are performed in FRSM-1.1 I Critical Safety Function Status Trees 112-EOP-CFST-1 II 11 11 25 I I Response to Imminent Pressurized Thermal Sh [\2-EOP-FRTS-1 " il 1125 I I Response to Loss of Core Shutdown 112-EOP-FRSM"2 JI II 11 20 1 I FRTSOOE002 I I I I -.J "IF,:H .....><, ** ..... ". .., , ........ " I SRO 10 I Given the following: -Unit 2 is in Mode 3 following a shutdown after a 200 day -All RCP's are in -Main steam dumps are in MS PRESSURE CONTROL-AUTO @ 1005 24MS10 setpoints are 1015 -A transformer fault results in the total loss of off-site -15 minutes after the transformer fault, with NO operator action, the following indications are -All RCS WR Thot's are 559°F and rising -All RCS WR Tcold's are 549°F and -All SG pressures are 1015 psig and -All SG NR levels are 39% and -PZR level is 23% and rising Which of the followinlij identifies the action that must be eerformed? [!] I.Lower 21-24MS10 setpoints to establish CET's stable or lowering lAW S2.0P-IO.ZZ-0008, Mai:ltaining Hot Standby. I I.Lower 21-24MS10 setpoints to establish CET's stable or lowering lAW S2.0P*A8.RC-0004, Natural Circulation. I "t.] I Lower Main Steam Dume eressure seteoint to stabilize or reduce RCS Thots lAW S2,OP-AB.RC-0004. I Lower Main Steam Dump pressure setpoint to stabilize or reduce RCS Thots lAW S2.0P-IO.ZZ-OOOB. .*I'D' I S I I Application I 1 & 2 i 1----__-' IKA: i loOWE09G411 I I Natural Circulation Operations ________________________________________________________________________ -. Knowled e of abnormal condition procedures. 55.43(5) The .Ios.s of off-site power will cause all and Circulators. to st?p. Loss of the circulators will a loss of Steam AnsWers: l;5,,1 Dumps Permissive and all steam dump valves Will shut. Steam dumping Will transfer to MS1 Os at 1015 pSlg from Steam Dumps at 1005 psig. C and D are incorrect because steam dumps will have no effect. A 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 cir is NOT present, so the correct . . . . . I ABRC04E001 I ABRCP1 E004 1___--' ,Ro. I, I . I SRO 11 Given the following conditions: -Salem Unit 1 is operating at 30% power, performing a shutdown at 30% per The shutdown is being driven by Tech Spec 3.0.3 being entered when BOTH subsystems were declared -The unit is required to be in MODE 3 within the next 2 hours. Which of the following describes the effect on the Rx if a loss of off-site power occurs, and how will operators perform the required cooldown? WILL NOT trip. Operators will perform a rapid cooldown lAW 1-EOP-TRIP-6, Natural WILL trip. Operators will perform a rapid cooldown lAW 1-EOP-TRIP-6, Natural Circulation Rapid Cool down with RVLlS. WILL NOT trip. Operators will perform a normal cooldown lAW 1-EOP-TRIP-4, Natural Circulation Cool down. Operators will perform a normal cooldown lAW 1-EOP-TRIP-4, Natural Circulation Cooldown. 9126/20 11 1 lCJ I Application 1 & 2 I 002000A203 I 'Iltle':j I Reactor Coolant System I i Ability to (a) predict the impacts of the following on the Reactor Coolant System and (b) based on those predictions, use to correct, control, or mitigate the consequences of those abnormal Loss of forced Ii.e.x.***'p!...**...... . 55.43(5) The Rx will trip on the loss of forced flow, based on any of the rCP trips associated with losing at least 2 RCPs when .. a ..

• and < P-8. In order to prevent violating Tech Specs, the unit must be cooled down to <350 OF (MODE 4) within 6 hours of being . in MODE 3. At the time MODE 3 is entered (on the Rx trip) Tavg is 547°F. 547-350=197° degrees of cooldown required in the next 6 hours, 197/6= 32.8°F/hr cooldown required.

TRIP-4 maintains a maximum cooldown rate of 25 °F/hr, so the transition to TRIP-6 Ii<:: f, . '" r",,,irl ,nth R\/I 1<:: " it . f. 111:: ,1I1:l n.. ...*. I Natural Circulation Rapid Cooldown with RVLlS "1-EOP-TRIP-6 II 1[22 1 I Natural Circulation Cooldown 1/1-EOP-TR'P-4 II 1122 I I RPS Primary Coolant System Trip Signals II II 11 10 1 IL.().. i&ill I RCSOOOE007 I [ TRP004E001 I 1

• =O::cl=if=iC=8=ti:::()=":;:M::::e=*t=h::Od::::
• ..
• .

.. I :Comment: ' .. ........ ..... ... .. ....... ::::..' ',: ..... ..... I I I I I

• .. . -. . .... ".-_.. ""' . , ....*. IQuestion Topic! I SRO I Given the following conditions:

-Salem Unit 1 has performed a Rx trip and SI based on rapidly lowering PZR and -The crew has transitloned to 1-EOP-LOCA-1. Loss of Reactor 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 ... It loss of RCS inventory. If the ECCS injection flow is due to a Loss of Secondary Coolant, the event could be term inated by raising AFW flow. I a loss of RCS inventory. If the ECCS injection flow is due to a Loss of Secondary Coolant, the event could be terminated by isolating the I faulted SG. I la SGTR is NOT causing the LOCA condition. If the ECCS injection flow is due to a SGTR, an uneccesary transition to LOCA-3, Transfer to Cold Leg Recirculation would be made. I I a SGTR is NOT causing the LOCA condition. If the ECCS injection flow is due to a SGTR, the tube rupture must be addressed before , retuming to LOCA-1 to terminate ECCS flow. 'b I i;x;;.. I S I I [Facility:ll Salem 1 & 2 9/26/2011J [.!9';/1006000G422 I 3.61IS R OVafufj 4.41 [g.. 11 .tel *** n ;",,: I Emergency Core Cooling System ______________________________________________________ -. Knowledge of the bases for prioritizing safety functions during abnormal/emergency I 55.43(5) The Major Action Categories for LOCA-1 are: 1. Check for Subsequent Failure, 2. Monitor Plant Equipment for :Ar'!swe'rS:i>,,;J**! Mode of Operation, and 3. Determine optimal Method of Long-Term Plant Recovery. The first item checks that a faulted ruptured SG is not the reason for ECCS injection, and either fixes it on the spot (faulted) or transitions to another more procedure With a faulted SG being the cause of the ECCS flow, LOCA-1 has a "do loop" which will wait unril the SG has blown ,..;, In . In Sf T, ' . \/<:.

• in (V'. _1 . th t,,,nciti,,,,, t, (')('[L?'

""0 11 :.,. I" i" performed to do actions which will terminate"the primary to secondary leakage, which is not performed in LOCA-1, Additionally, there is no transition from SGTR-1 to LOCA-1.1 FI L=os=s=of=R:::;:e=aC=to=r C=O::::O=lan=t======:::;1I:=1=-E=O=P-:::;:LO:::;:C:::;:A=-1=====:;iI.:========'.'

='==:::}11l==2=5I Steram Generator Tube Rupture j/1-EOP-SGTR-1 II II 1124 111==2=0 I LOCA01E005 I SGTR01E003 I LOSC01E002 IMaterial Requifed fqr Examinatiori . :j I I_N_e_w______.....1 I fUsed j 0 I ." I I J I

.. ,.;:-'"' -. ."*...... I __Top]£] I SRO / 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 eerformed by the crew lAW S2,OP-AB.PZR-0001, Pressurizer Pressure Malfunction? IMPC output will LOWER. The unit may continue to operate indefinitely after the initial litigative actions Ipleted. I /.MPC output will RISE. The unit may continue to operate indefinitely after the initial mitigative actions are completed. I [J IMPC output will LOWER. A unit shutdown will be required if 2PR1 cannot be restored to operable status, I IIMPC output will RISE, A unit shutdown will be required if 2PR1 cannot be restored to operable status. I IAf1$wer ll c I S I I Application 1 Salem 1 & 2 1 9/26/2011/j k w<<r,?$c--,./- u ""'-r [KAdlo10000A203 1t2,03 -.. I Pressurizer Pressure Control I i 010 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 55.43(2) This question is SRO level because of the Tech Spec knowledge required, and what actions TS directs for different PORV Ati!;we!s: ..*.malfunctions. Additionally, wihile the question doesn't specifically ask what procedure to use (too easy for AB.PZR), it does require knowledge of the actions IN that procedure. The MPC raises output wihen actual pressure rises, and lowers as actual pressure lowers. As pressure lowers due to the open PORV, the output will lower to turn on heaters and close spray valves. When the PORV . 01 ,I, In' 'h 01"'10\' 1\0 070 T, Q, '1 A e:; h ;f 01"10\ ,,-I Hh;n 7') h, -h"tdnlA/f'1 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, I II Pressurizer Pressure Malfunction II II 11 18 I Salem Tech Specs II 3.4.5 r I 1 \=1=::; II II fl i I-J !ABPZR1E002 I !PZRP&LE010 I I I for I I /1 [§:estion New I ________..... 1 !)uring Training CJ I ..... .'"*IConu"""L I I Given the following conditions: -Unit 2 is operating at 100% -23 charging pump is in -A control circuit malfunction has resulted in the 2CV3, 45 gpm Letdown Orifice and cannot be shut from the control room. Which of the this failure will have on rm,,,,,l';nn of the , and how will rm,>r:.'tnr<:. 23 charging pump speed will rise, Isolate letdown and place Excess Letdown in service lAW S2,OP-SO,CVC-0003, Excess Letdown Flow, 2CV55 will modulate in the open direction, Isolate letdown and place Excess Letdown in service lAW S2,OP-SO.CVC-0003, Excess Letdown Flow, 23 charging pump speed will rise, Transfer to a centrifugal charging pump lAW PZR LEVEL LO alarm response in S2,OP-AR.ZZ*0012, Control Console CC2. 2CV55 will modulate in the open direction, Transfer to a centrifugal charging pump lAW PZR LEVEL LO alarm response in S2,OP-AR.ZZ* 0012, Control Console CC2, I Pressurizer Level Control System IKAS!lilt'ernermJ i Ability to (a) predict the impacts of the following on the Pressurizer Level Control System and (b) based on those predictions, use I procedures to correct, control, or mitigate the consequences of those abnormal operation:

• Excessive letdown 55.43(5) Letdown orifii are Containment Isolation valves, and will require letdown isolation to isolate the CIV. For long term

...'.. a.,...* .*.o.'.,.n..*.****...o.' ..*.*....* ..'.,.*...foperation Excess Letdown must be placed in service. The 23 charging pump (PDP) is normally in service at power. 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 orifice in service, letdown flow will be -110 gpm, The Master Flow Controller controls 23 charging pump speed when in service, and the CV55 when , '1 !CQ:Numbe1i":, .' .j 1 PZRP&LE007 1__-, sourc::=J I New I[§ue!>tionModification'..,ethod:Tl----------l10s e d DLJringTraining prograrril [J 1 I ...... "....*.iComment I I I

• RO SRO J

lisrq . 12 I Given the following conditions: Unit 1 is operating at 100% -1 PR5, PZR Safety Valve, falls Which of the following identifies what will occur if NO operator action is taken until AFTER the Rx trips, and which procedure will be used upon the transition out of 1-EOP-LOCA-1 Loss of Reactor Coolant? The Main Turbine will runback on OT/DT, then the Rx will trip on OT/DT. 1-EOP-LOCA-2, Post LOCA Cooldown and Depressurization. The Main Turbine will run back on OT/DT, then the Rx will trip on OT/DT. 1-EOP-TRIP-3, The Rx will trip on Low PZR Pressure. 1-EOP-LOCA-2, Post LOCA Cooldown and Depressurization. The Rx will trip on Low PZR Pressure. 1-EOP-TRIP-3, Safety Injection Termination. 013000A203 I I Engineered Safety Features Actuation System "l LISA-Statement: I Ability to (a) predict the impacts of the following on the Engineered Safety Features Actuation System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Rapid depressurization 55.43(5) The PZR Safety failing open will cause a SBLOCA. Pressure will rapidly lower, and an OT/DT runback will initiate the OT/DT Rx trip. The Safety Injection will actuate, and pressure will stabilize -985 psig. With no subcooling, the transition TRIP-3 cannot be made, and LOCA-2 will initiate a cooldown. The distracters either contain the wrong procedure, the wrong Rx actuation Signal, or both. D':;;:' .. I Post LOCA Cooldown and Depressurization !l1-EOP-LOCA-2 II 11 23 I I II ,,: I II I I II II II f I I I ESFOOOE021 f.or EXamination* .*.'.*. ! I sourceIlI New .. ,..lC......... ...., .'....... I I I I

• *** ... r-c"-.

'I Topi<:J SR016 I Given the following conditions: -Unit 2 is in MODE -Core off-load is in progress and is 2/3 -A fuel assembly is in the transfer cart at the Spent Fuel -The mast tube of the manipulator crane is -Refueling cavity level begins to drop rapidly, approximately 6 inches per -Radiation Protection reports flooding in the lower elevations of Which of the following describes how the Fuel Transfer Tube Gate Valve should be operated lAW S2,OP-AB,FUEL-0002, Loss of Refueling or Spent Fuel Pool Level, and the consequence of that "START the transfer cart moving to containment, then shut the gate valve when the cart is clear. The fuel transfer cart will stop when the gate vaiveopelllimit is lost. I , START the transfer cart moving to containment, then shut the gate valve when the cart is clear. The fuel transfer cart will continue until it Ireaches its normal travel stop on the containment side. IMMEDIATELY shut the gate valve regardless of transfer cart position. If a fuel assembly is in the Spent Fuel Pool upender, transfer it to the Fuel Handling crane and place it into its designated SFP location. I [J IIMMEDIATEL Y shut the gate valve regardless of transfer cart position. If a fuel assembly is in the Spent Fuel Pool upender. lower is to the horizontal position to reduce any radiation exposure due to dropping SFP level. I !Ariswed I a IlI§!rn.Levefll s 1[Cognithie;t;evE!U I Memory 1 [Salem 1 &2 I 9/26/20111 I [4.35_'_J fROyallieTI[ 3.81 [[[groIJP:.11 21 21 Ita 1 Fuel Handling Equipment System __________________________________________________.__________________________Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. 55.43(7) With a no fuel assembly in the mast tube and a fuel assembly in the transfer cart, the operator is directed to close the fuel pool gate valve at step 3.13 AFTER starting the cart to the containment side. The discussion section of S2,OP-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 with shield and "'........1 I Loss of Refueling Cavity or Spent Fuel Pool Lev II S2.0P-AB.FUEL-0002 -11 II 11 10 1 1/---------'11----11----'1 1--, I ABFUE2E002 [ Facility Exam Bank Direct From Source I [Questionsoor.cecol11mentsJ I Vision Q71988 ....................................... ...... .... .....*** I I . I Given the following conditions: -Salem Unit 2 is responding to a loss of all off-site power. -28 EDG is supplying 28 4KV vital bus. 2A and 2C EDGs have tripped. -2A and 2C SECs have been deenergized. -A C02 actuation occurs in 2B EDG room. Which of the following identifies the effect the C02 actuation will have on 2B EDG operation, and what actlon(s) will be performed? will automatically trip. Deenergize 2B SEC lAW 2-EOP-LOPA*i, Loss of All AC Power. will NOT automatically trip. Deenergize 2B SEC lAW S2.0P-SO.DGV-OOOi, Diesel {;p",p,r,>lI>r will automatically trip. Place the 2B Diesel Generator Supply Fans Emerg Bypass of C02 Shutdown Switch at 2RP5 in Emergency lAW EOP-LOPA-i. will NOT automatically trip. Place the 2B Diesel Generator Supply Fans Emerg Bypass of C02 Shutdown Switch at 2RP5 to Emergency lAW S2.0P-SODGV-0001. m I 064000A222 II Emergency Diesel Generators 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 EDG Area/control room supply fan. The SEC deenergization of the unaffected SECsfEDGs in stem is to make the EDG trip plausible if the candidate thinks that the SEC would prevent an EDG trip, since would be a standing Mode 11 SEC signal present with 2 vital buses deenergized. Bypassing the C02 shutdown of the supply fans allows them to restart since their control . . I there is an exhaust fan for the EDG FO Storage area. '[;Q..,Numbef<<l I EDGOOOE008 1__---' 'Material Required for ..*.*. ] I II lQuestioll I New I ________-.l1 Tr<l,!ningProgrC3 rnLJ .. ' .... .......... I I II I I I .. I i: ,., I __________________________________________________ .______________________________Given the following conditions: Unit 2 is operating at 100% -25 SW pump is INOPERABLE for scheduled -The appropriate TSAS Tracking entry for TSAS 3.7.4 has been made OP-SA-1 08-115-1 001, Operability Assessment and Equipment Control Program. 23 SW pump strainer motor fails. for that action? Exit 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. 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. eking entry for 23 SW pump. Each 4KV vital bus requires only one OPERABLE SW pump to ensure full for 23 SW pump. IF ANY other SW pump were to be declared INOPERABLE, THEN entry OPERABLE SW loop would remain. I Service Water System ______________________.____________________________ -. Ability to determine operability and/or availability of safety related eouipment. 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 pumps inoperable on different vital buses AND in different SW bays as described in stem, then BOTH SW !oops 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 . n . h",,,,, , ":,,, "n" ntb",r. CI. . ",.." ,1M ""+ "... +h", 'J '" "0' h", ."' ".a'" f,*?.. . . . . 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. / Salem Tech Specs 113.7.4 II 1/ I I Service Water System Operation II S2.0P-SO.SW-0005 II 11 40 I 'I 11 r I II I ILO;Num'ber,\> .*...1 .. ___ __ ____J I SWONUCE010 1/ :Questionsource:j !Facility Exam Bank I Significantly Modified I[Used During Tr.:dning LJ iaueStiOnSou;.c{Comrnentsj , Q80615. to a. "2 and 2" question. Replaced 3 distracters with newdistracters. Added I section to each chOice ....,.. .... ............. .. .....,..........I Cor...I I I I I------____________________ Given the following conditions: A heatup is in progress on Unit 2 lAW S2.0P-IO.ZZ-0002, Cold Shutdown to Hot Standby. -All RCPs in Seal injection flow is 6.0 gpm to each During the heatup, VCT gas pressure is lowered from 35 psig to 15 be addressed? Flow to the No.2 RCP seal will lower. Fill the RCP standpipe when the low level alarm annunciates lAW S2.0P-AR.ZZ-0011, Control Console 2CC1. A high RCP standpipe level alarm will actuate. Drain the RCP standpipe until the high level clears lAW S2.0P-AR.ZZ-0011. Control Console 2CC1. A high RCP standpipe level alarm will actuate. If required, raise seal injection flow to maintain at least 6 gpm to each RCP lAW SO.CVC-OOOi. Flow to the No.2 RCP seal will lower. If required, raise seal injection flow to maintain at least 6 gpm to each RCP lAW S2.0P-SO.CVC-0001. I I iGE!:lERIJ ______________________________________________________________________________Ability to perform specific system and integrated plant procedures during all modes. of plant operation. .Cltio"""** n"""".f' 55.43(5) Lowering the VCT pressure reduces the backpressure forcing flow to the #2 seal. With less backpressure. flow to the** ',:f,.,,'.,',',I* number 2 seal will lower (more of #1 sealleakoff is going to VCT) With less flow going to the #2 seal. number 2 sealleakoff lower, and standpipe level cannot rise. The RCP standpipe will not lower to the alarm setpoint as long as there is any flow from #2 I Charging, Letdown, and Seal Injection II S2.0P-SO.CVC-0001 II II II I il IIF====;11 I II fI II 11---1 I RCPUMPE008 I I I [ I Required for Exai'ilillatioll I I I Facility Exam Bank Editorially Modified I illsedDuring

• 1 0" fQuestion sourceCorrln'lentsJ I ViSion Q85461 L_,__ ..

_..._ ... ;Commlimt

• .********i

.*.*....*...... ...... I I I I RO I ":R() ____________________________________________________________________________________Given the following conditions: -Salem Unit 1 is operating at 100% power with equil ibrium core -Power is lowered to 75% over a 15 minute period using the control -The NCO then adjusts control rod height as necessary to maintain average coolant tem perature constant. What will be the rod position and directional trend 30 hours after the power change? Reference rovided. Below the initial 75% position and inserting slowly. Above the initial 75% position and inserting slowly. Below the initial 75% position and withdrawing slowly. Above the initial 75% position and withdrawing slowly. :KA=I!194001 G143 =r 1_________________________________________ -1 __ Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, I secondary plant, fuel depletion, etc. I.E..... .... ...*... Of I.......n .*....1 55.43(6) As Xenon builds in following the power reduction, control rods will have to be withdrawn. After the Xenon peak, Xenon will iAnswers;!:. "J be burning out, and control rods will be inserted to offset the less negative reactivity due to Xenon. Xenon will continue to burnout to establish a new equilibrium level at 75% power, and the control rods will continue to be inserted past where they were inserted at I 75%. /=lc=urv=e=B=oo=k========:::;lls1.RE-RA.ZZ-0016 II I fI II II 1l=1==::, l-----------lll------..........II/-------'Il 'L.O.NUmber "J I RXOPERE019 I I I I ! fOfExamini.ltionj I SRO 2Q S1.RE-RA.ZZ-0016 'I [Ques!L0nsC!l.I!'ce: J 1 Facility

lOuestionSource C",".."."nts i IVision 37077 I IComment ,., "j .,I II I II 1 II----------------------------------------,J

.. .'. I 1 SRO 21 Given the following conditions: Unit 1 is operating at 100% Operators have just satisfactorily completed SC.OP-PT.DG-0001 DIESEL MANUAL BARRING on 1 B -The only Tech Spec entered was 3.S.1.1.b for the 1 B EDG being declared when the LOCKOUT SW was crr in the LOCKOUT Which of the following identifies when the 18 EDG will be declared OPERABLE? CRS. When the LOCKOUT SW is released to the IN SERVICE position after successful completion of SC.OP-PT.DG-0001. When the NCO is directed by procedure to log the EDGs availability for operability testing in the Control Room Narrative Log. When the EDG has successfully met the acceleration, voltage and frequency Acceptance Criteria in the ST procedure used for the retest, and is synchronized to the Vital Bus. 1 Memory I I Salem 1 & 2 9/26/20111 Ability to track Technical Specification limiting conditions for operations. i 55.43(2) The EDG beccmes inoperable when the Lockout Switch is placed in the lockout position. The EDG remains inoperable [!.nS'i\ferS:>>i 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 synchonized to the vital bus. I________---lll II II 11=1====; fL':d.Number.. .:C071... I EDGOOOE010 1 I I I I I"I" ILOT SRO CERT Exam 11/2006 ..... . .......... .... IComment .... c. i II I II I :1 1 ... " ",',.,.. iOIlf!stirln Topic i I SRO 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 Exposure Control and Authorization, and prior to performing the job, written approval for increasing his dose limit to 3000 mrem TEDE for the calendar l'ear must be received from the work supervisor and the ... I RP .V.d. 'd\J!;l1 ONLY. I I Site Vb """,,,,Ye,,, ONLY. I @J I RP Manager and Station Manager. I II Station Manager and Site Vice President. 1 I a I S Memory I[FaCility: Iisalem 1 & 2 J

J 1 9/26/20111 i KA: 111 94001 G304 J 3.21 I PWG 1 I

11 Syste I11/Ev()h:ltIPI1 Ti!!:J l__________________________________ --'I[GEN@ Knowled e of radiation exposure limits under normal or emer enc IEXpla'nati6Hof 55.4:3(4) A is correct because the approval requirements are: Up to 3,000 mrem-RP Manager; up to 4,000 mrem-RP Manager Station Manager, >4.000 Site Vice President. Distracters are all some form of combo of each of the ',," :. I RADCONE002 ..

• ...

Iprevious 2 NRC Exams I J Direct From Source I 0 IOuestion 108-01 Salem ILOT SRO Exam May 2010 LComment .* ' ..... .*... ..... ..*. ,.. '. c. I I I I Given the following conditions: -Unit 2 is operating at 100% The operating crew entered S2.0P-AB.RC-0002, High Activity in Reactor Coolant when RMS channEll 2R31, Letdown Line Monitor, went into Which of the tnllflWlnn is ant areas? Confinmation of 2R31 response. Radiation levels may have changed access requirements. 2R31 reads in CPM and therefore has no correlation to dose level changes. Detenmination of radiation type (gamma vs. neutron) will narrow possible sources of the higher radiation levels. 1_____________________________________ -1 L._____. Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. I 55.43(4) B is correct as described in S2.0P-AB.RC-002 basis, so that prompt identification and subsequent notification of plant !"' ... personnel IS A is incorrect because chemistry sampling confirms 2R31 readings, not survey results. C is incorrect because rising counts indicate dose level changes. D is incorrect because the radiation type won't be determined, as neutrons won't be [1-f 11-II ..... L. 0 .52.**.....I ABRC02E003 I I I . . . ..... .-...... [Questio-; ISRo24 I Given the foliolNing conditions: -Unit 2 has received a FIRE alarm for Zone 69, Fuel Handling Building (FHB). -An operator in the area reports a fire in a bin of Protective Clothing in the FHS truck bay. Which of the identifies the correct 8rocedure for this event, and the actions I I.S2.0P-AB.FIRE-0001, Control Room Fire Response, place Control Room Ventilation in FIRE OUTS IDE CONTROL AREA on both Unit 2 and Unit 1, and secure ALL Unit 2 ONLY FHB supply and exhaust fans. I S2.0P-AB.FIRE-0001, Control Room Fire Response, place Control Room Ventilation in FIRE OUTSIDE l,;UN I KUL AREA Unit 2 ONLY, I and secure BOTH units FH B supply fans ONL Y. I IS2.0P-AB.FIRE-0002, Fire Damage Mitigation, place Control Room Ventilation in FIRE OUTSIDE CONTROL AREA on both Unit 2 and Unit 1, and secure ALL Unit 2 ONLY FHB supply and exhaust fans. J @]ls2.oP-AB.FIRE-0002, Fire Damage Mitigation, place Control Room Ventilation in FIRE OUTSIDE CONTROL AREA on Unit 2 ONLY, and secure BOTH units FHB supply fans ONLY. I [Answer I I s I Memory I I Salem 1 & 2 9/26/20111 r.<A<I194001G427 I l 3.41 3.9! PWG 1 11 11 If I .... 1----------------------------------------' L ...._._* ..J [KAStatenient:! r-:------::--:-------------------------------------, Knowledge of "fire in the plant" procedures. ... ..:*..f.*..........

55.43(5) S2.0P-AB.FIRE-02 is the incorrect procedure because it is used for an uncontrolled fire in the Aux Bldg or 78

...

• .n.*.**'.* **.o Answers:.<j pen, not for a fire in the Fuel Handling Building.

Steps 3.12 and 3.13 of S2.0P-AEI.FIRE-01 places BOTH Units in fire outside, ..._---3.94 and 3.95 direct stopping FHB supply and Exhaust Fans. A is correct becausEI it is the correct procedure and action. incorrect distracters are all a combination of either incorrect action, procedure, or L.c:....... I Control Room Fire II S2.0P-AB.FIRE-0001 II II r 17 II --.J I II i 1=1==::::; ________II II II_--J IABFP1E003 1_-----' II I Facility Exam Bank I Direct From Source commentslJ 07-01 SRO CERT Exam I I I + - .. [Question Topic I I SRO 25 1 Given the following conditions: -Salem Unit 2 is experiencing an event which has resulted in an ALERT EAL exceeded and recognized by the operating -After 5 minutes. the actions taken by the control room crew result in this ALERT no longer being 1-Conditions for an UNUSUAL EVENT are -No emergency declaration has been made Which of the identifies the actions required b:t the Coordinator? I Declare an ALERT. terminate the ALERT. declare an UNUSUAL EVENT. I I.Deciare an ALERT. then reduce the Emergency Classification to an UNUSUAL EVENT. I I Declare an UNUSUAL EVENT, and make a non-emergency Afler-The-Fact 1 hour report to cocument the ALERT EAL. I I Declare an UNUSUAL EVENT, and ensure the ALERT EAL which was present for the 5 minutes is communicated to the NRC via the NRC Data Sheet. . I iAOswe-r] 1 b liExam Le"'[Jl/ S

<1 1 Memory 11 Salem 1& 2 9/26/20111 lKA

11194001 G438 1 2.4.38 ..

11 1111 .7:.71 ______________________________ ---1IIGENER! I _____________________________________________________________________________Ability to take actions called for in the facilit 55.43(5.1) Per the Salem ECG, Section I, page 8 of 10, Section IV, Event Classification Guide (ECG) Use, Subsection 0.2,.AAswers: ..."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 be made if no EAL's are currently being exceeded, (Le. actions have taken to stabilize the \lant such that no EAL's EP.ZZ-040S, EMERGENCY TERMINATION REDUCTION -RECOVERY, d"escribes how to reduce the Emergency Classification if I the emer enc is not bein terminated. Since the UE is still bein met, the classification will be reduced, not terminated.

====;

'...'1 lMateriaIRequiredforExarnina!ioti <, I IJ :Comment. ., '.' .......... ' :.... ..' ...... "':'., ..... . ... , I I I 1 1________________________________________ ,1}}