ML070120478
| ML070120478 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 12/11/2006 |
| From: | Kafantaris M Public Service Enterprise Group |
| To: | D'Antonio J Operations Branch I |
| Sykes, Marvin D. | |
| References | |
| Download: ML070120478 (109) | |
Text
Urns. Nuclear Regulatory Commission Site-S pecific Written Examination Applicant Information Name: Region: I Date: 12/11/2006 Facility: Salem 1 & 2 License Level: SRO Reactor Type: W Start Time: Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected SIX hours after the examination starts.
Applicant Certification
~~
All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results Examination Va I ue Points Applicant's Score Points A ppIica nt's Grade Percent
[Which ONE of the following - is correct concerning S2.OP-AB.CR-0001, CONTROL ROOM
~EVACUATIoN?
/Does NOT support shutting down the Reactor during any type of accident.
/Supports shutting down the Reactor during ANY type of accident.
".!Supportsshutting down the Reactor during ANY type of accident, EXCEPT loss of coolant (accidents.
jsupports shutting down the Reactor during ANY type of accident, EXCEPT accidents requiring
!entry into SAMGs.
I 12/11/2006j 12.1 ]/Conductof Operations 12.1.61[Ability to supervise and assume a management role during plant transients and upset conditions. pq :4.31 55.43(5) Section 2.0 Immediate Action NOTE:The EOPs are not applicable during Control Room Evacuation. EOPs should be used for information only or as directed by the TSC while performing this procedure.
[Control Room Evacuation I /S2.OP-AB.CR-0001 1E002 (BrownsFeny Unit 2, 9/17/2001 nrc Exam, modified to Salem terminology.
I I I 1 -I I1 I 1- 1 Thursday, October 26,2006 11:31:16 A Page 1 of 30 Prepared by WD Associates, Inc.
'Given the following conditions:
I I
- Unit 2 is operating normally at 100% power when 21 SGFP trips.
- The Main Turbine runs back to 60% as expected.
- All systems respond as expected to the runback.
- 2 minutes after the runback, the PO announces that condenser backpressure is 2.6"Hg and rising at 1.O"every Iminute.
- The CRS directs entry into AB-LOAD, and commences a 1%/minute load reduction, then directs the unloading rate raised to 3%/min when vacuum continues to degrade.
- With the reactor at 52% power, the Secondary NE0 reports that there is a 2" diameter hole in the SGFP exhaust line to 21 condenser, he can hear a loud whistling noise around the hole.
IWhich actions and procedures which should be performed?
- TRIP the reactor, GO TO TRIP-I.
12.1.71Ability to evaluate plant performance and make operational judgments based on operating pj characteristics, reactor behavior, and instrument interpretation.
procedure.
- Rapid Load Reduction I IS2.OP-AB.LOAD-0001 1 7 1I /141IABLOA/
DE003 LOSS of Condenser Vacuum 1 IS2.OP-AB.COND-0001 1 7 17 1 T I1 -
-I = - - , I Thursday, October 26, 2006 11:31:16 A Page 2 of 30 Prepared by WD Associates, Inc.
Which of the following condition(s) would REQUIRE Field Engineering to review a Troubleshooting Plan developed in accordance with SH.OP-AP.ZZ-0008, OPERATIONS TROUBLESHOOTINGAND either directly or as a result of causing a major plant transient. (Very High Risk)
II. Equipment is NOT removed from service or tagged. Could result in an unexpected load reduction, a plant transient, or a reportable event. Should NOT result in a reactor, turbine, or generator trip. (High Risk) 11 and II only.
/I,II, and Ill only.
!I, II, 111 and IV.
[Memory I 12/11/20061 I
/Generic Knowledge and Abilities I I
[GENERlCI1
(' r2/k! m E qi iunim pme net n tControl i2.2.20 IKnowledge of the process for managingtroubleshootingactivities. 112.2) j3.31
/55.43(5) A is incorrect because both Hiqh Risk and VERY High Risk must be evaluated. C is incorrect
\do NOT need to be evaluated.
OPERATIONS TROUBLESHOOTING AND 1SH.OP-AP.ZZ-0008 EVOLUTIONS PLAN DEVELOPMENT EDEOO I
Thursday, October 26,2006 11:31:16 A Page 3 of 30 Prepared by WD Associates, Inc.
Which ONE of the following describes the MAXIMUM time allowed in accordance with Technical
- 15 minutes.
/30 minutes.
[Equipment Control j2.2.221 [Knowledge of limiting conditions for operations and safety limits. :flrn ,4.1, 55.43(2) A is correct per TS when in MODE 3-5. B is incorrect and is action time for several 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS's.
C is incorrect and only provides symmetry of choices. D is the action time if in MODES 1-2.
ISalem Tech Specs I I - l I I Thursday, October 26,2006 11:31:16 A Page 4 of 30 Prepared by WD Associates, Inc.
1 Do not approve the liquid waste discharge, secure the lineup, the liquid waste discharge is not permitted until 2R18 is repaired.
Approve the liquid waste discharge and ensure that continuous effluent sampling is conducted throughout the liquid waste discharge.
Approve the liquid waste discharge as long as a second sample was drawn, analyzed, and calculations were second verified prior to the release.
12/11/2006 F I l R a d i a t i o n Control
/Releaseof Radioactive Liquid Waste 1 /s2.oP-so.wL-ooo1 - 1 j 2 1 /wAsLII QEOl2 Thursday, October 26, 2006 11:31:16 A Page 5 of 30 Prepared by WD Associates, Inc.
Given the fo Ilowing conditions:
- Unit 2 is operating at 100% power.
- Operators receive OHA G-7, ADFCS SWITCH TO MANUAL.
- The board operator notes both SGFPs Speed Controllers have switched to MANUAL.
- All BFl9 and BF40 valves remain in AUTO.
- 21 SGFP speed is lowering slowly, and remains latched.
- All SG NR levels are 40% and dropping slowly Which of the following describes the procedure which would be most effective in responding to these indications?
S2.OP-AR.ZZ-0007, OVERHEAD ANNUCIATORS WINDOW G to address the SGFP switch to MANUAL.
'2-EOP-TRIP-I, REACTOR TRIP OR SAFETY INJECTION, to respond to the Rx trip caused by ISG 10-10level.
S2.OP-AB.CN-0001, MAIN FEEDWATER / CONDENSATE SYSTEM ABNORMALITY, to address the imminent loss of 21 SGFP.
S2.OP-AR.ZZ-0012, CONTROL CONSOLE 2CC2, to respond to the PROGRAM DEVIATION SETPOINT ACTUAL alarms on all SGs.
n I 12/11/2006' IGeneric Knowledge and Abilities 1- 1 12.4 IIEmergency Procedures / Plan 12.4.10 (Knowledgeof annunciator response procedures. pq13.lj actions besides the swapping to manual of controllers. B is incorrect because the CAS step 4.0 to take manual control of affected controller in AB.CN should preclude having a Rx trip. C is correct because the immediate action contained in AB.CN would trip the malfunctioning SGFP and cause an automatic turbine runback which would allow SG levels to recover, and give operators time to assess the ADFCS failure. D is,I incorrect because the control console alarm response would only take time away from entering the CN AB. ,
IOVERHEAD ANNUCIATORS WINDOW G I jS2.OP-AR.ZZ-0007 -1 115-1711 3 81
'EDEOO
'MAIN FEEDWATER / CONDENSATE
!SYSTEM ABNORMALITY
$2.OP-AB.CN-0001 1 7 /2 :-41 1E004
!USE OF PROCEDURES ISH.OP-AP.ZZ-0102 1 7 ' : 11 1 5 1T I Thursday, October 26, 2006 11:31:16 A Page 6 of 30 Prepared by WD Associates, Inc.
Given the following conditions:
- Salem 1 and 2 are operating at 100% power.
- Hope Creek is operating at 100% power.
- Fire Brigade manning consists of 6 qualified personnel, which includes one Fire Brigade Leader.
- A Fire Brigade member falls ill, and is transported off-site by Medical Department personnel.
Which of the following describes the status of the Fire Brigade, and action(s), if any, which are
[required to be performed IAW NC.FP-AP.ZZ-0001, Fire Protection Organization, Duties, and Staffing?
The Fire Brigade remains adequately staffed. Only five members are required IAW Salem FSAR. No compensatory measures are required.
.,The Fire Brigade remains adequately staffed. The assumption is made that concurrent fires at Salem and Hope Creek are not plausible events. No compensatory measures are required.
'The Fire Brigade staffing is inadequate. Initiate call-out of qualified personnel to ensure manning '
!is restore to six members within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. otherwise submit a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report to the NRC. I
'The Fire Brigade staffing is inadequate. Initiate call-out of qualified personnel to ensure manning
'is restore to six members within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise initiate an Action Request and review for liicensing commitment violation.
Knowledge of facility protection requirements including fire brigade and portable fire fighting equipment "2.9/ 3 12.4.261 usage. 1 Fire Protection Organization, Duties, and (NC.FP-AP.ZZ-0001 -1 COND taffing?
1, OPE00 Thursday, October 26, 2006 11:31:16 A Page 7 of 30 Prepared by WD Associates, Inc.
Given the following conditions:
,-- Unit 2 is in Mode 5 with 21 Residual Heat Removal (RHR) pump in service for cooling.
The RO reports that Pressurizer (PZR) level is slowly lowering unexpectedly.
- NO Overhead Annunciator OR Auxiliary Typewriter alarms have been received.
- Refueling Water Storage Tank (RWST) level is stable.
- 21 Waste Hold Up Tank level is rising slowly.
Which of the following describes:
- 1. The effect this would have if NO operator action were to be taken.
- 2. What procedure and action would terminate this problem.
E Eventual caviation of 21 RHR pump and gas binding of BOTH RHR pumps. Close 2CV132, IExcess Letdown IAW S2.OP-S0.CVC-0003, EXCESS LETDOWN FLOW.
Loss of pressure control when the PZR heaters deenergize. Remove 21 RHR Loop from service I and put 22 RHR loop in service IAW S2.OP-SO.RHR-0001, INITIATING RHR.
ICONCENTRATION CONTROL.
action to close the CV8. B is incorrect because the CV132 is not physically located at an elevation which I could provide flow into the WHUT. C is incorrect because pressure control is provided by the RHR pump 1 discharge, and putting the redundant loop in service would just cause it to become gas bound too. D is incorrect because dilution would not occur with blended makeups.
I--lr------
I r------
I Thursday, October 26, 2006 11:31:16 A Page 8 of 30 Prepared by W D Associates, Inc.
I
Given the following conditions:
- Unit 2 is operating at 100% power.
- ALL station Air Compressors trip.
- BOTH Units Emergency Control Air Compressors start.
- 2CC71 LTDWN HX CC CONT VALVE, sensed a low header pressure on its primary air supply, and transferred to its backup supply. When it transferred, the valve diaphragm failed, and the valve moved to its failed position.
- NO other air operated valves have been adversely affected by the air system perturbation.
Which of the following describes the effect this will have on the CVCS system, and what actions are required?
!Letdown must be manually isolated due to the inability to control letdown temperature, and IExcess Letdown must be placed in service IAW S2.OP-S0.CVC-0003, EXCESS LETDOWN
/FLOW.
/VCTtemperature will lower, causing less effective aeration of letdown flow into the VCT through ithe spray nozzle. Additional RCS lithium control adjustments will be required IAW SC.CH-AP.RC-10106, IMPLEMENTATION OF SALEM LITHIUM CONTROL PROGRAM.
ILetdown temperature will rise until OHA E-41, LTDWN HX OUT TEMP HI alarm is received.
I CVCS Mixed Bed Demineralizers must be removed from service by manually repositioning the
'2CV2.1, LTDWN DM BYP V from Control Console 2CC2 IAW S2.OP-AR.ZZ-0005, OVERHEAD
/ANNUNCIATORS WINDOW E.
'Failure of the 2CC71 will cause the 2CV7, LTDWN HX INLET VALVE, to automatically close due to the OPEN interlock between the two valves. Letdown must be further isolated by closing 2CV2 and 2CV277, LETDOWN LINE ISOL VALVES, IAW S2.OP-S0.CVC-0001, CHARGING LETDOWN, AND SEAL INJECTION.
/Application I
1 12/11/2006]
\Plant Systems 1 jA2. Ability to (a) predict the impacts of the following on the Component Cooling Water System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
D is incorrect because the interlock works the opposite way, that is, when the CV7 is closed, it closes the CC71. A is correct because the ARP directs the manual letdown isolation and directs excess letdown be placed in service.
[OverheadAnnunciators Window E ~ IS2.OP-ARZZ-0005 1 7 1 157-5811 1 7 1 /ccwo]
OOEOO ILoss of Control Air I $2.OP-AB.CA-0001 1 1 11 3 11-u Thursday, October 26,2006 11:31:16 A Page 10 of 30 Prepared by WD Associates, Inc.
I I I I
Thursday, October 26,2006 11:31:16 A Page 11 of 30 Prepared by WD Associates, Inc.
PRIOR to taking any action IAW S2.0P-AB.NIS-0001, NUCLEAR INSTRUMENTATION SYSTEM MALFUNCTION, which of the following identifies the required action, if any, to be taken IAW Tech
!Suspend CORE ALTERATIONS ONLY.
T I [Reactor Protection System F l I E q u i p m e n t Control NIS has operators select the other channel, but the stem states that no action have been performed in the AB yet. The TSAS ACTION must be taken since there is no audible indication in the control room. A is incorrect because the channel selector has not been transferred to the other channel yet, and there is no way of knowing if the Audio Count Rate Monitor itself is broken for control room audible indication. B is incorrect because the audible indication is required in both the containment AND the CR. C is incorrect because it encompasses only half the action required by TS. D contains the correct action for TSAS 3.9.2.
- Technical Specifications I /UnitTwo Tech Specs - 1 -'
Nuclear Instrumentation System Malfunction (S2.OP-AB.NfS-0001 I I (
I 7
I1
' ~ r I1
- l r - l I 1 --I e-&nect answer into a distracter.
Ir--T --1 X i I I -
-~ -~
I Thursday, October 26,2006 11:31:16 A Page 12 of 30 Prepared by WD Associates, Inc.
- Unit 2 is in MODE 5.
- The NE0 dispatched to investigate reports SFP level just below the alarm setpoint, and appears to be stable.
- No leak identification action has been initiated.
Which of the following describes the actions required for this condition?
FUEL LEVEL , to isolate the SFP cooling pumps individually to isolate the most likely source of leakage.
Since occasional SFP low level alarms are to be expected due to the leak on the SFP liner, refill the SFP using CVCS HUT water if available to maintain boron concentration as high as possible IAW S2.OP-S0.SF-0001.
/Monitor 2R5 and 2R32 SFP Area Radiation Monitors, which will initiate 22 HEPA PLUS CHAR lmode of FHV IAW S2.OP-AB.FUEL-0002.
I 12/11/2006 E] Ability to (a) predict the impacts of the following on the Spent Fuel Pool Cooling System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
r ~-
Fill and Transfer of the Spent Fuel Pool
!Overhead Annunciator Window C ~ $2.OP-AR.ZZ-0003 II 7 11 1 3 1T I I
I I 1-Thursday, October 26,2006 11:31:16 A Page 13 of 30 Prepared by WD Associates, Inc.
normal operating bands when 21MSIO, fails open.
Which of the following describes the INITIAL system parameter response assuming NO operator action were taken initially and all effected control systems are in MANUAL, and which procedure would be most effective in responding to this event?
Tave lowers, PZR level lowers, 21 SG NR level rises; S2.OP-AB.STM-0001 EXCESSIVE STEAM FLOW.
iTave rises, PZR level rises, 21 SG NR level rises; S2.OP-AB.CN-0001, MAIN iFEEDWATEWCONDENSATE SYSTEM ABNORMALITY.
Tave rises, PZR level lowers, 21 SG NR level lowers; S2.OP-AB.CN-0001, MAIN IFEEDWATEWCONDENSATE SYSTEM ABNORMALITY.
k r -
I :Comprehension I 12/11/2006'
/Plant Systems 2/
55.43(5) With all control systems in manual, the effect of a Main steam Atmospheric Relief valve failing open would be to raise steam flow from that SG, lower SG pressure, lower Tc of that loop. Lowering Tc would cause Tave to lower. Tave (auct hi) is the input to control PZR level, and while 21 loop may not be the auctioneered hi loop, ALL loops will be affected by any loop Tave dropping, and cause a corresponding
!specific steps to address Galfunctioning MSIO.
~ EXCESSIVE STEAM FLOW I IS2.OP-AB.STM-0001 11 I 14 Id I Thursday, October 26, 2006 11:31:16 A Page 14 of 30 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit Iis at 90% power steady state
- 14BF19 fails full closed over a period of 1 minute
- All other controls respond as expected With NO operator action, which of the following responses will be apparent FIRST to the operators, land what procedure will be used to respond to this event?
BFI9 DEMAND rises on unaffected SGs. S2.OP-AB.CN-0001, MAIN FEEDWATER I
!CONDENSATESYSTEM ABNORMALITY iPZR B/U heaters turn ON in AUTOMATIC. S I .OP-AB.PZR-0001, PRESSURIZER PRESSURE IMALFU NCTI ON.
ISGFP Master Speed Controller demand signal rises. S I .OP-AR.ZZ-O012, CONTROL CONSOLE
,ICC2 Alarm Response.
,,/Reactor trip at 14% NR level on 2/3 channel on 14 SG, I-EOP-TRIP-1 REACTOR TRIP OR ISAFETY INJECTION 12111l2006 1- \Main Feedwatersystem m / / A b i l i t yto (a) predict the impacts of the following on the Main Feedwater System and (b) based on those predictions, luse procedures to correct, control, or mitigate the consequences of those abnormal operation:
[Failure of feedwater regulating valves ; p F p]
cause PZR pressure to rise. Heaters will turn off, not on. C is incorrect because SGFP speed control is based on average steam flow, and is lag compensated. Steam flow should not change much, if any. The Facl-gE I .\X.iW.
Rebren (I
v*- W --e-
?geNumTe?$i F%
I i . ,
r--
,Reactor Trip or Safety Injection 1- FOP-TRIP-1 1 7 11 - :CN&F DWEO 08 I MAIN FEEDWATER / CONDENSATE
,SYSTEM ABNORMALITY IS2.OP-AB.CN-0001 I
Dm IEditorially Modified Added procedure flow path to ensure 55.43(5)was applicable U l 1- 1 Thursday, October 26, 2006 11:31:16 A Page 15 of 30 Prepared by WD Associates, Inc.
1 1 Given the following conditions:
Salem Unit 2 has experienced a LBLOCA coincident with numerous equipment failures and losses of power.
A majority of CETs have exceeded 1200 deg. F.
2-EOP-LOCA-I was in progress when a transition to 2-FRCC-1, RESPONSE TO INADEQUATE CORE COOLING was made 2-FRCC-1 has been ineffective at lowering CET temperatures.
The TSC is activated.
Which of the following describes how this condition will be addressed?
Return to Step 1 of FRCC-1 and continue in a "do" loop until any action has reduced CET temperatures less than 1200 deg. F.
Return to LOCA-I proedure in effect until transfer to HL recirc is required while continuing any available mitiaation actions.
'Transition to SAMG-CRG-1 CONTROL ROOM INTIAL RESPONSE FOR SEVERE ACCIDENT, lsince the normal EOP network has been ineffective at Drotectina the core.
[&e Break LOCA
/EA2. IlAbility to determine and interpret the following as they apply to Large Break LOCA: I jEA2.081 :Conditions necessary for recovery when accident reaches stable phase ;mE 55.43(5)This question is designed to test the candidates ability to determine when the accident can NOT be recovered from by using the normal EOP network. In this sense it meets the WA since knowing when the accident is essentially "non-recoverable" is logically tied to conditions which would let a normal recovery happen. With CET temps >1200 degrees in FRCC, the transition is made to SAMG-CRG-1 at step 27.
The onlv wav into SAMG-CRG-2 is to enter SAMG-CRG-1 first.
/SevereAccident Mitigation Guidelines ~ 12-SAMG-CRG-1
'-=lo_llFRccoi OTOOl l ' ~
l ---r I ( I , ( l Thursday, October 26, 2006 11:31:16 A Page 17 of 30 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 is operating at 100% power.
- 21 charging pump is in service.
- At 0200, an automatic VCT makeup occurs.
- At 0400, another auto makeup occurs.
- PZR level and charging flow are stable and have remained constant.
/Which of the following - describes what is occurring, - and what Drocedure shcAd be imtiemented?
> 1 gpm RCS leak, S2.OP-AB.RC-0001, REACTOR COOLANT SYSTEM LEAK.
g/2LT-112has failed to 14%. S2.0P-AR.ZZ-OOl2.2CC2 CONTROL CONSOLE Alarm Response. I A 2 gpm leak on the 21 charging pump discharge check valve flange, S2.OP-AB.CVC-0001 LOSS OF CHARGING.
'PZR Master Flow Controller setpoint has drifted high, S2.OP-S0.CVC-0001, CHARGING,
/LETDOWN. AND SEAL INJECTION.
55.43(5) A is incorrect because both PZR level and charging flow would have changed to compensate for the loss of fluid from the system. B is incorrect because the auto makeup would never stop since it would be seen as VCT level. Using 20 gallons per % in the VCT, and the makeup band of 14-24%,the system is losing 10% every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or 200 gallons evey 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or 100 gal/hour, or 1.67 gpm. D is incorrect because charging flow would have risen and PZR level would have gone up.
[Loss of Charging / IS2.OP-AB.CVC-0001 1 7 1 r l ICVCSO]
OE012
/Chargingletdown and seal in 1205328 Sheet 2 l ~
r - ir-- 1 Thursday, October 26, 2006 11:31:16A Page 18 of 30 Prepared by WD Associates, Inc.
'Given the following conditions:
Unit Iis operating at 100% power when 12 SGFP trip.
The Main Turbine runs back as expected.
The RO is unable to initiate a normal boration.
Operators receive OHA E-I 6, ROD INSERT LMT LO-LO Control Bank D position is 75 steps.
Reactor power is stable at 64%.
RCS boron concentration is 650 ppm.
Using the attached REM figures, determine the LEAST amount of time a rapid boration through 1CV175 is required IAW S I .OP-S0.CVC-0008, RAPID BORATION, in order to clear OHA E-I6?
L5 minutes.
18 minutes.
&3minutes.
1024 I IEmergency Boration 55.43(6) Using REM Figure 14, the RIL at 64% power is 85 steps. The ARP for RIL lo-lo states that rods must be withdrawn at least 2 steps past the limit to clear the alarm. With rods at 75 steps from stem, rods must be withdrawn 12 steps to the 87 step position. Using REM figure 4, the reactivity from this 12 step movement is -1 00 pcm. Using REM Figure 13, the differential rod worth for 650 ppm is -6.9 pcmlppm.
sNur$e
-~
'Reactor Engineering Manual -- -~T !SI.RE-RA.ZZ-0012
/RapidBoration I / S I.OP-S0.CVC-0008 I / II II II I I I I I - I I I Thursday, October 26, 2006 11:31:16A Page 19 of 30 Prepared by WD Associates, Inc.
>Giventhe following conditions:
- Unit 2 has experienced a LBLOCA.
- 2C EDG reenergizes 2C 4KV Vital bus.
- 2A and 2B EDGs start but BOTH their Vital busses are locked out on Bus Differential.
Which of the following identifies the correct procedure flowpath, starting from when off-site power was lost?
IAB.LOOP-OOOI, LOSS OF OFFSITE POWER Remain in LOCA-1 until transition is required. If LOCA-3 is entered go IMMEDIATELY to LOCA- I
'5, otherwise go directly to LOCA-5, perform in its entirety, transition to IOP-6, HOT STANDBY i
,TO COLD SHUTDOWN.
of RHR pumps is required. GO TO LOCA-5, complete
\Emergencyand Abnormal Plant Evolutions 1025 I ILoss of Residual Heat Removal System 12.4 IlEmergency Procedures / Plan I
[Knowledge of EOP implementation hierarchy and coordination with other support procedures. 153 I
SEC's were reset, so the SECs never loaded in MODE II (Blackout). C is correct because once in LOCA-3, lthe CAS action to go to LOCA-5 should be performed immediately to conserve RWST level and initiate makeup. D is incorrect because there is no transition prior to RWST level of 15.2' or Step 16 of LOCA-1 to I
lao anwhere else with a LBLOCA and no other even t in woaress. Once in LOCA-3 I
/Loss of Emergency Recirculation I 12-EOP-LOCA-5 -'=;LoCAoI OT005
[Loss of Reactor Coolant
[Transfer to Cold Leg Recirculation 1 /2-EOP-LOCA-1 112-EOP-LOCA-3
'--=m OTOOI Thursday, October 26, 2006 11:31:16 A Page 20 of 30 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 is operating at 100% power with a leaking fuel pin.
The specific activity of the RCS has been 0.3 uCi/gram DOSE EQUIVALENT IODINE for 1 week.
- A radiation protection technician reports the latest RCS sample indicates that specific activity has jumped to 70 uCi/gram.
- Prior to any action being taken, a 300 gpm tube rupture occurs on 22 SG
- A MANUAL Rx trip and a MANUAL Safety Injection were initiated successfully.
- IMMEDIATELYfollowing the reactor trip, 22MS10, SG Atmospheric Relief Valve failed open.
- Operators cannot enter the affected penetration area to manually isolate the malfunctioning valve until TWO hours have passed.
Which of the following describes how radiological conditions will be affected by this failure?
A person located at any point on the outer boundary of the low population zone during the entire time of the release may be exposed to more than an acceptable portion of the 25 Rem whole body dose limit.
10381/SteamGenerator Tube Rupture
\Tech Specs 1'3.4.9 Specific activity
/and L
bases I I
[Radiation Protection Program I lNC.NA-AP.ZZ-0024 7 17 1@ l l p I
- Code of Federal Regulations 1 :IOCFRI 00 p i m - - - - I ~ i l I I 1 Thursday, October 26, 2006 11:31:I 6A Page 22 of 30 Prepared by WD Associates, Inc.
Thursday, October 26, 2006 11 :31:I 6A Page 23 of 30 Prepared by WD Associates, Inc.
- Unit 2 is operating at 100% power.
- A steam leak is identified and the CRS orders the reactor tripped.
Which of the following conditions would require a Safety Injection to be MANUALLY initiated after the Rx trip is attempted, and what procedure would require the Safety injection?
Flow.
The reactor does NOT trip from the Control Room; 2-EOP-FRSM, Response to Nuclear Power Generation.
1040 I !Steam Line Rupture w I / A b i l i t y to determine and interpret the following as they apply to Steam Line Rupture:
[Conditions requiring ESFAS initiation Thursday, October 26, 2006 11:31:16 A Page 24 of 30 Prepared by W D Associates, Inc.
i Given the following conditions:
- Unit 2 is operating a 75% power.
- PZR pressure channel II is removed from service for calibration.
- An electrical fault causes the 500 KV switchyard to become deenergized.
- The PZR Master Pressure Controller (MPC) fails low, causing sprays to close and all heaters to energize.
If pressure rises above their lift setpoint, which of the following describes how this will affect PZR PORV operation?
ONLY 2PR1 will open. Since the PORVs are not designed to prevent exceeding RCS design pressure, one OPERABLE PORV is an acceptable plant configuration.
IONLY 2PR2 will open. Since the Rx has already tripped due to the Loss of Off-Site power, the IPORVS are not necessarv for Dlant control.
IBOTH PORVs will open since the MPC does not control PORV operation. The plant will NOT exceed design parameters since two PORVs have enough relief capacity to prevent exceeding parameters as long as the PZR Safety Valves function properly.
/Emergencyand Abnormal Plant Evolutions 1 1 I
j 0 5 6 1/LOSS of Off-site Power m / I A b i l i t yto determine and interpret the following as they apply to Loss of Off-Site Power:
IAA2.011 IPORV controller indicator and setpoint ]PIj3.41 55.43(2) A is correct because the PR2 will not open since it is a 2/2 coincidence and it has one of its channels out for calibration. The PORV's are designed to prevent PZR pressure from reaching the High Pressure Rx trip, not for exceeding design pressure. B is incorrect because of (a) above, even though it has the right reason. C is incorrect because of (a) above, and also that 2 PORV's are NOT designed to keeD RCS Dressure less than 2485 Dsia. D is incorrect because the MPC does not directlv control PORV loperation. 'Each PORV is 2/2 to operat'e on alternate PZR pressure channels. 1 and 3 , -and 2 and 4.
~ -... ?
Referen,?
_-.. .Title _-
Reactor Protection System PZR pressure and 1221060 l--rlEIl level control
-czIzzl-mm 1
r- ~~
II 1 Thursday, October 26,2006 11:31:16 A Page 25 of 30 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 was operating at 100% power when a steam leak upstream of 22MS167 occurred.
- The Rx was tripped and a MSLl performed successfully.
- Operators have transitioned out of EOP-TRIP-1.
- The PO is attempting to open 21-24SS94s, SG BID Sample Valves, but they will not open.
- SGBD sample isolation bypass has been RESET.
Which of the following conditions identifies the reason the valves won't open?
- 22 SG NR level is ~ 9 % .
- SIwas not reset properly.
IPhase A isolation failed to reset. 1 jCA330s have not been reopened.
/mlAdherence amendments.
to appropriate procedures and operation within the limitations in the facility's license and 14.3/
Thursday, October 26, 2006 11:31: I 6 A Page 26 of 30 Prepared by WD Associates, Inc.
Establish charging and letdown to stabile PZR level, ensure both PORVs are closed and PORV stop valves open.
12/11/2006/
/Emergency and Abnormal Plant Evolutions
/E07 1 /Saturated Core Cooling
/EA2. //Abilityto determine and interpret the following as they apply to Saturated Core Cooling:
amendments.
to appropriate procedures and operation within the limitations in the facility's license and 'w I-- @
F:
Response to Saturated Core Cooling ~ -
VISION Q48662 I II I I 1 1 Thursday, October 26,2006 11:31:16 A Page 27 of 30 Prepared by W D Associates, Inc.
/Which of the following choices identifies the condition which would result in the highest priority CFST? I
,f?x tripped 15 minutes ago, 21-23 SG levels are 4% NR, 24SG level is 80% NR.
,Rx is tripped from 100% power, IR SUR is +0.1 dpm, Rx power is 5x10-9 Amps, PZR level is 122%.
..kx trip and SI from 100% power due to LOCA, ALL RCPs stopped, RVLIS Full Range 35%,
!highest CET in each quadrant reading 600 deg.
]F?xwas tripped from 80% power due to steam rupture, ALL RCPs are stopped, RCS Tc's are 230 I jdeg. RCS pressure is 1200 psig.
12/11/2006 1-1 IPressurizedThermal Shock JEA2./IAbility to determine and interpret the following as they apply to Pressurized Thermal Shock:
wl IFacility conditions and selection of appropriate procedures during abnormal and emergency operations. 113,m, path, C is incorrect because it is a PURPLE Path CFST. D is correct because it is the only RED path CFST. The Figure 4A Thermal Shock Limit Curve shows that ant cold leg temperature of 230 deg with pressure > 0 psig is to the left of Limit A. The combination of > 100 deg/hr cooldown rate, and pressure temp to the left of Limit A results in RED path.
/Critical Safety Function Status Trees : I2-EOP-CFST-1 Thursday, October 26,2006 11:31:16A Page 28 of 30 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 is at 100% power.
- With SSPS testing and troubleshooting in progress a Phase B Containment Isolation signal was generated and all related valves closed.
- Before operators could re-open any of the Phase B valves, the operating Charging Pump breaker
/tripped on an electrical fault.
!Which of the following describes the required operator actions?
!!~~lmmediatelystart the other Charging Pump and monitor RCP bearing and seal inlet temps.
'Initiate a MANUAL reactor trip and stop all RCP's if Phase B can NOT be reset. RCP's can be re-
!started anytime after seal injection has been restored.
Start the other Charging Pump OR restore CCW to the thermal barrier within five minutes or initiate a MANUAL reactor trip.
Initiate a MANUAL reactor trip and stop all RCP's. Cooldown to desired temperature IAW EOP-ITRIP-4, NATURAL CIRCULATION COOLDOWN.
because the actions are correct per AB.RCP, and the cooldown will have to be performed naturally circulated.
\Reactor Coolant Pump Abnormality 1 IS2.OP-AB.RCP-0001 4T001
/Natural Circulation Cooldown 112-EOP-TRIP-4 P I EO0 l l - - I Z I I I l D D Thursday, October 26, 2006 11:31: I 6 A Page 29 of 30 Prepared by WD Associates, Inc.
1 Given the following conditions:
- Unit 2 has experienced a Large Break Loss of Coolant Accident.
- The Reactor trip and Safety Injection occurred successfully.
EOP-LOCA-1 LOSS OF REACTOR COOLANT is in effect.
- PZR pressure is 35 psig.
- 1 CET is reading 900°F, ALL other CET's are reading -550°F.
- RVLIS Full Range is reading 74%.
- Containment pressure is 13 psig.
- Containment sump level is 62%.
- R44A radiation monitor is indicating 50 Whr.
I IFRCC-2, RESPONSE TO DEGRADED CORE COOLING.
JFRCE-1,RESPONSE TO EXCESSIVE CONTAINMENT PRESSURE.
IFRCE-~,RESPONSE TO HIGH CONTAINMENT RADIATION.
R level will be offscale low with 700 and RVLIS level is NOT IResponse to High Containment Radiation 112-FRCE-3
/Critical Safety Function Status Trees 1 i2-CFST-1 liii1251D I 1 7 - I - - - - - l - m Thursday, October 26, 2006 11:31:16 A Page 30 of 30 Prepared by WD Associates, Inc.
r Material Requised for Examination Administration Exam Level KA MaterialRequiredforExamination Exam section S 000024A205 REM Figures 4,13,18.S1.OP-S0.CVC-0008 1 Monday, October 30, 2006 Page I of I
U S . Nuclear Regulatory Commission Site-S pecific Written Examination Name: Region: I Date: 12/11/2006 Facility: Salem 1 & 2 License Level: RO Reactor Type: W Start Time: Finish Time:
All work done on this examination is my own. I have neither given nor received aid.
II Applicant's Signature Results Examination Va Iue Points (I Applicant's Score Points Applicant's Grade Percent
Given the following conditions:
- You are a Licensed Reactor Operator, assigned to an off-shift administrative position.
- During the first calendar quarter, you have stood the following duties, all 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> plus turnover:
- 1/6/06 U1 RO
- 1/30/06 U2 PO
- 2/2/06 U2 RO
- 3/10/06 U2 RO Today is 4/1/06 With regards to watch standing hours, which of the following describes the status of your license in
[accordancewith OP-AA-105-102, NRC ACTIVE LICENSE MAINTENANCE?
IActive. You may stand watch with no restrictions.
\Active. You must regain qualification as RO by standing one additional 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift in the RO or PO position.
either the RO or PO.
Inactive. You must reactivate your license by standing at least five 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts under instruction in the ROIPOMICC RO position.
/c /R /Application 12/11/2006
[Generic Knowledge and Abilities m
' I
\GENERIC 11 F I l C o n d u c t of Operations j2.1.11 [Knowledge of conduct of operations requirements. (p[ 13.81 A is incorrect because the WCC NCO is not a Licensed Position. OP-AA-105-102, NRC Active License Maintenance, specifically states that the quarterly watch requirements are to be stood by "performing the duties of the Unit RO and/or the Unit Assist RO" ( Plant Operator at Salem). As such, the individual has only stood 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of watch, plus -1 hour of turnover, and has not met the requirement of 5 12-hour shifts per calendar quarter. Distracter b is incorrect because the 1st calendar quarter is completed, and NO watchstanding can be performed until the license is re-activated. C is correct because 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> must be stood at the RO or PO position. (OP-AA-105-102, Rev. 7, Attachment 2, Reactivation of License Log).
Distracter d is incorrect because the WCC RO position does not count towards Licensed Duties hours.
INRC Active License Maintenance I IOP-AA-105-102 OPE00 I
Monday, October 30, 2006 11:07:30 A Page 1 of 79 Prepared by WD Associates, Inc.
The Unit is in MODE 4. Operators have just removed control power from 22 SI pump. 21 SI pumps is C/T.
(GenericKnowledge and Abilities MODE 4 consisting of ONE charging pump and flow path from RWST, and capable of taking suction from RHR discharge piping, and discharging into each RCS cold leg, AND ONE RHR pump and flowpath from
- I ~ C I K I Monday, October 30, 2006 11:07:30A Page 3 of 79 Prepared by WD Associates, Inc.
bwap to Cold Leg Recirculation 1 12-EOP-LOCA-3 I / - - - - l ~ j 2 6 p G -
1E004 Monday, October 30,2006 11:07:30 A Page 4 of 79 Prepared by WD Associates, Inc.
/2001 on November 15th. I 12.2 ](EquipmentControl (2.2.26)!Knowledge of refueling administrative requirements.
(TechS p e c s I ITSAS 3.9.3.a and b I
Monday, October 30,2006 11:07:30 A Page 5 of 79 Prepared by WD Associates, Inc.
(Knowledgeof radiation exposure limits and contamination control, including permissible levels in excess of those authorized.
12.5/ F l REM/ yr are illegal. HOWEVER, 10.CFR.20.1206, Planned special exposures, directs that this dose shall be maintained separate from the yearly occupational dose, as long as the special exposure dose plus the occupational dose does not EXCEED the occupational dose numbers found in 1201(a). This means that the 5 REM/yr TEDE dose cannot be exceeded by more than 5 REM TEDE. D is correct because even though the Planned special exposure dose will NOT be added to his occupational dose, and as a result, his loccupational dose will not rise above 5 REM for the current year. Distracters a, b, and c are both wrong because it would raise the workers dose limit above 5 REM for the year, which is illegal.
IRadiation Protection Program 1 lNC.NA-AP.ZZ-0024 1 7 17 11 1 3 1 Fl ONE00
/Codeof Federal Regulations 1 jl OCFR20 ] i c z I I I I L _ _ 1 I l - - - - - l ~ , ~ ' l I II EIIIIl Monday, October 30, 2006 11:07:30A Page 6 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 is operating at 75% power.
- Unit Iis operating at 100% power.
- A MANUAL Rx trip and SI are initiated on UNIT 2 due to a LOCA.
- At step 15 of 2-EOP-TRIP-I , operators discover the Control Room Ventilation system is in the NORMAL Mode.
Which of the following identifies the actions required IAW EOP-TRIP-1, if any, and why?
\No action is required, NORMAL is the correct post Rx trip alignment.
- 'No action is required, since the R1B channels will automatically isolate the Control Room
!Envelopeif outside air radiation levels rise.
I
,Depress EITHER units Accident Pressurized PB. This will isolate ALL outside air supplied to the bontrol Room. and habitabilitv reauirements will be met.
j2.3.101'Ability to perform procedures to reduce excessive levels of radiation and guard against personnel pqj3.31 exposure.
/Tech Specs Bases 1 3/4.7.6Control Room Emergency Air Conditioning Units IS2.OP-SO.CAV-0001 ~ Control Area Ventilation Operation
~
Monday, October 30,2006 11:07:30A Page 8 of 79 Prepared by W D Associates, Inc.
- Unit 2 is operating at 100% power.
- A release of 21 WGDT is in progress, 2WG41 is OPEN.
- Containment pressure is 0.21 psid.
IWhich choice states whether or not a Containment Pressure Relief may be performed, and why?
/a E 12/11/2006
]GenericKnowledge and Abilities EENERIC ~ 11 F I l R a d i a t i o n Control (2.3.111]Abilityto control radiation releases.
&+z-% I Ta&$R
-Num 1-EM i o n . ~+i-~TiibG$2j+
,Cx . E O .
~
Dischargeof 21 Gas Decay Tank to Plant VentS2.OP-SO.WG-0008
__- - IF- I r 1 [ 2 6 11WASG 1 r--. -- ----___
Containment Pressure-Vacuum Relief System ,S2.OP-SO.CBV-0002 __I r---.--,,---.-,117-,1 E
I Operation r---- --
_- --_J Monday, October 30,2006 11:07:30 A Page 9 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 is in MODE 5.
- 21 RHR pump is in service for shutdown cooling.
- A large fire is reported in 21 RHR pump room.
Which of the following describes the required action which must be performed IAW S2.0P-AB.FIRE-
)001, CONTROL ROOM FIRE RESPONSE, and why?
ilsolate the PZR PORV's for RCS inventory and pressure control.
RHR cooling must be terminated in order to transfer shutdown cooling to 22 RHR pump. Initiate S2.0P-AB.RHR-001, LOSSof RHR.
/R !Memory I
I 12/11/2006' W l l E m e r g e n c y Procedures / Plan B.FIRE-1 directs the isolation of the RCS-RHR from the the SJ44s and the RH4s runs in the room, and spurious hot I II I
1- J Monday, October 30, 2006 11:07:30A Page 10 of 79 Prepared by WD Associates, Inc.
- Unit 2 is operating at 100% power.
- An interior Control Bank rod drops fully into the core.
- A Rx trip signal is not generated, NOR is it required.
- Which of the following alarms is NOT consistent with these conditions?
IP-250 Computer Alarm.
IOHA E-48, ROD BOTTOM.
___ ~~
1 IOHA E-24, ROD DEV OR SEQ.
ZIRod Control NON-URGENT FAILURE 2CC2 Bezel Alarm.
12.4 !/EmergencyProcedures / Plan 12.4.461bbility to verify that the alarms are consistent with the plant conditions. 113.5:j3.61 C is incorrect because the alarm is driven by ANY rod more than 12 steps deviation from its group counter. B is incorrect because it is driven by ANY rod being e20 step with Control Bank D > 35 steps. A lis incorrect because the P-250computer will alarm for both the deviation and rod bottom alarms. d is correct because the non urgent failure is driven from the loss of DC power supplies, none of which would cause anv rods to dron.
mbe@ EFv>%&
nnunciators Window E , 1 1 7 1 OOEOI I Control Console 2CC2 : IS2.OP-AR.ZZ-0012 1 7 1I 1301T /
u I l -
Monday, October 30, 2006 11:07:30 A Page 11 of 79 Prepared by W D Associates, Inc.
- Unit 2 is operating at 100% power.
- A fire has broken out in the 104 panel, and has spread to the overhead.
- Due to the location of the fire, Fire Protection cannot control the fire.
Which choice identifies an indication that will be present in Unit 2 control room after placing the CVCS cross-connect in service from Unit 1 IAW S2.OP-AB.FIRE-0002 FIRE DAMAGE MITIGATION, and why?
/Slowly rising VCT level due to seal injection supplied from Unit 1 with letdown isolated.
PZR PORV and block valves closed to prevent potential loss of RCS inventory and RCS pressure control.
- ALL CCW pumps stopped to ensure the CCW system is available to support achieving and lmaintaining HSB conditions within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the fire event.
12/11/2006/
!Ability to interpret control room indications to verify the status and operation of system, and understand j3.5:
lhow operator actions and directives affect plant and system conditions.
- 1 r---
- Fire Damage. Mitigation
. .- -. -- , S2.0P-AB.FIRE-0 I I 1 - I I I I Monday, October 30, 2006 11:07:30 A Page 12 of 79 Prepared by WD Associates, Inc.
I Given the following conditions:
- Unit One is in Mode 5, and has just started a heatup to NOT/NOP.
- 11 CVCS HUT level is 70%.
- 13 CVCS HUT level is 12% and in service.
Using the attached tank curves and assuming 50,000 gallons will be letdown from the RCS to the CVCS HUT'S, what will be the final level of the CVCS HUTSwhen the RCS heatup to NOT/NOP is complete?
111 CVCS HUT 13 CVCS HUT 170% 84%
170% 100%
180% 90%
IF% 74%
/Plant Systems 1 c.2 2j 1- 1 [Reactor Coolant System w I / K n o w l e d g e of Reactor Coolant System design feature(s) and or interlock(s) which provide for the following:
I IS2.OP-TM.ZZ-0002
- Tank Curves ICoId Shutdown to Hot Standby I js2.oP-Io.zz-ooo2 I O
P m
ir=- ' ' ~ 7 E002
]
OEOl3 I l ' ~
I I
Monday, October 30, 2006 11:07:30A Page 13 of 79 Prepared by WD Associates, Inc.
I Given the following conditions:
- A steam generator tube rupture occurred on Unit 2 from 80% power.
- The control room operators have tripped the plant and initiated an SI.
- Operators are performing a controlled cooldown of the RCS per EOP-SGTR-1 and are NOT at the target temperature.
The following indicated parameters are present:
- Ruptured SG pressure is 985 psig and stable.
- Ruptured SG level is 42% and rising.
- PZR level is 5% and lowering.
- All RCP's are operating.
- RCS pressure is 1300 psig and lowering slowly.
- RCS subcooling is 25°F and rising.
- High alarms are standing on R15 and the affected R19.
(Selectthe proper crew action for the given conditions.
!Verify ECCS flow established and trip the RCP's.
F I I E m e r g e n c y Procedures / Plan 1 I
/2.4.20 [Knowledge of operational implications of EOP warnings, cautions, and notes. pq Distracter a is incorrect because the stem states that the cooldown depressurization is in progress, so the SGTR-1 CAS item to trip RCPs is NOT in effect. Distracter d is incorrect because the cooldown is required to be continued to target temperature. The standing rad monitor signals are due to the ruptured generator, 1 not from intact SGs. Distracter c is incorrect because subcooling is adequate per the stem, so transition to SGTR-3 won't be necessary.
Steam Generator Tube Rupture Basis ISGTR Basis Document 1 7 ' 1 1261r l Document Monday, October 30,2006 11:07:30 A Page 14 of 79 Prepared by WD Associates, Inc.
I I
Given the following conditions:
- Unit 1 is in MODE 3, at NOPINOT.
- I 1 and 12 CCW pumps are in service.
- 13 CCW pump is O/S and in AUTO.
- A Bus differential fault occurs on 1C 4KV Vital bus.
Which of the following identifies the CCW pumps which will be running 1 minute after the vital bus fault?
jll and 12.
[ I 1and 13. I
'12 and 13.
jl1, 12, and 13.
I 12/11/2006'
\ReactorCoolant Pump System E' J~Knowledge of bus power supplies to the following:
jK2.02,~CCWpumps 1 i q3 1 I'1vital 1-13 CCW pumps are powered from A,B, and C 4kv vital busses respectively. A bus differential will cause 1 the bus to remain deenergized. A single 4KV vital bus being deenergized will not affect the other 2 powered,I busses. 11 and q2 pumps will remain running. Distracters are incorrect because 13 CCW pump has
/no power.
/No.1 Unit 4160V Vital Busses One Line I /203002 OOEOO I I ! I 1-Monday, October 30,2006 11:07:30 A Page 15 of 79 Prepared by WD Associates, Inc.
- Unit 1 is operating at 100% power.
- Operators are transferring CVCS Letdown from 1CV4, LETDOWN ORIFICE ISOLATION VALVE,
- 1CV18, LETDOWN PRESSURE CONTROL VALVE, is in MANUAL.
temperature.
temperature.
throttled CLOSED as pressure lowers, and the 1CC71 modulates closed in response to lower
,temperature.
/A4.[/Ability to manually operate and/or monitor in the control room:
1- /Letdown pressure and temperature control valves 1pj7 1 A is correct because as more system (RCS) pressure is felt as the 2nd orifice is opened, the Pressure Control Valve must be opened to reduce pressure to maintain at NOP of 300 psig. As the letdown flow rises, the temperature control valve must modulate open to maintain setpoint temperature of 100 degrees.
The distracters are all incorrect combination of directions for valve movement and system pressure and temperature changes.
~CVCSSystem 11205228-2 I ~ \ r - - - - q ~ , ; c v OE004 cso/
/ComponentCooling System 11205231-2 OE008 I II I Monday, October 30, 2006 11:07:30A Page 16 of 79 Prepared by WD Associates, Inc.
- Unit 1 has experienced a LBLOCA.
- 11 RHR pp is.C l/
- While attempting to transfer to Cold Leg Recirc IAW I-EOP-LOCA-3, neither the IISJI 13 NOR the 12SJ113 can be opened.
!What effect will this have on the operation of the CVCS pumps?
l Iand 12 CVCS pumps will continue to operate with their discharge aligned to the four RCS hot j
'I 1 and 12 CVCS pumps will continue to operate with their discharge aligned to the four RCS cold jlegs.
lone of the CVCS pumps will have to be stopped to prevent runout of the only operating RHR IPumP.
BOTH CVCS pump will have to be stopped since their suction cannot be aligned to the discharge of 12 RHR pump.
IK1.1Knowledge of the physical connections and/or cause-effect relationships between Residual Heat Removal System and the following:
/Transfer to Cold Leg Recirculation 1 1 -EOP-LOCA-I Monday, October 30, 2006 11:07:30 A Page 17 of 79 Prepared by WD Associates, Inc.
r.
Given the following conditions:
- Salem Unit 2 has experienced a LBLOCA.
- All equipment functioned properly EXCEPT 21 RHR pump, which seized 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after SI was initiated. It will take 3 days to repair.
- After consultation with the TSC, 21SJ45 was closed, and no other operational action related to 21 RHR pump trip has been taken.
Which of the following identifies the lineup which will be present AFTER the transfer to Hot Leg Recirc is complete?
'Containment spray header flow through 21CS36, Hot Leg Recirc from 21 SI pump through
/21SJ40,Cold Leg Recirc through 22 RHR pump and 22SJ49.
NO flow through EITHER containment spray headers, Hot Leg Recirc from 22 SI pump through 22SJ40, Cold Leg Recirc through 22 RHR pump and 22SJ49.
ITransfer to Hot Leg Recirculation I (2-EOP-LOCA-4 14E003 ITransfer to Cold Leg Recirculation 112-EOP-LOCA-3 2E004 Monday, October 30, 2006 11:07:30 A Page 18 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 has experienced a LOCA.
- The reactor was tripped and MANUAL SI initiated sucessfully 10 minutes ago.
- Containment pressure peaked and is stable at 3.0 psig.
- Operators have just entered EOP-LOCA-1, Loss of Reactor Coolant.
Which of the following choices identifies an effect if ECCS injection flow were lost by inadvertent
,closing of the 2SJ69, RHR SUCT FROM RWST, and containment pressure rose to 4.6 psig before it (couldbe reopened and ECCS injection flow restored?
IAdverse containment conditions exist. The criteria for SI flow reduction are less restrictive.
/Control air to the Containment would be isolated when the 21122CA330 valves closed on the
/PhaseA isolation signal.
/AnAUTOMATIC SI would be initiated from the High Containment pressure signal since the initial I IS1was MANUALLY intiated.
[Containmentequipment is subjected to a harsher environment. A higher level of instrument error lcauses indicated subcooling to lower.
Plant Systems E E 1006 I \Emergency Core Cooling System 1 W l l K n o w l e d g e of the effect that a loss or malfunction of the Emergency Core Cooling System will have on the following!
second SI, manual or automatic es are AUTOMATICALLY inserted een reset following SI reset, and
\Loss of Coolant Accident 112-EOP-LOCA-1 MTEOO Monday, October 30, 2006 11:07:30 A Page 19 of 79 Prepared by WD Associates, Inc.
1007: /Pressurizer Relief TanWQuench Tank System r n ' K n o w I e d g e of the physical connections and/or cause-effect relationships between Pressurizer Relief TanWQuench ITank System and the following:
jK1.011 jcontainment system 13.1;
/Reactor Coolant I 1205301-1 TE009
'Davis Besse 5/10/2004 NRC Exam.
Monday, October 30, 2006 11:07:30A Page 20 of 79 Prepared by WD Associates, Inc.
-During normal power operations, which of the following describes how Pressurizer Relief Tank (
PRT) temperature is reduced if required IAW S2.0P-SO.PZR-0003, Pressurizer Relief Tank operation?
i '
[Ensure open 2NT25, PRT N2 SUPPLY, then open 2PR14 to establish a drain path. Start a IPrimary Water Pump and verify 2WR80 and 2WR82 cycle to maintain 54-87% during Idrainingkooling.
I Establish gravity feed and bleed from the PWST by opening 2WR80, CONT PRI WATER STOP VLV, and 2WR82, PRT WATER SUPPLY, and opening 2PR14, PRT Drain.
Open the 2PR14 to start the RCDT pumps, and open the 2PR15, PRT VENT TO RCDT VENT HDR, which will start the gravity feed through the normally open 2WR80 and 2WR82.
w
&Open the 2WR80 and 2WR82, start a Primary Water Pump, and fill the PRT to the Hi level alarm (89%). Secure the Primary water pump and open the 2PR14 until the lo level alarm is reached jK4.1 Knowledge of Pressurizer Relief TanWQuench Tank System design feature(s) and or interlock(s) which provide for the following:
Reactor Coolant 205301-1 EIIZlEIIIlFlCI Chemical and Volume Control Primary Water 205330 C I I I - F l E I I
' I II 1 II 1I 1 - I Monday, October 30,2006 11:07:30 A Page 21 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 1 is operating at 100% power.
- Pressurizer level is dropping slowly.
- CCW Surge tank level is rising slowly.
- Radiation Monitor R I 7A, CCW Process Radiation Monitor is rising.
Which of the following identifies the component which is the source of in-leakage to the CCW system,
[and what action(s) will prevent the release of radiation to the atmosphere?
RHR Heat Exchanger; 1R41D will swap Aux Bldg Exh ventilation to HEPA plus Charcoal in service.
RCP Thermal barrier heat exchanger; 2CC149 Surge Tank Vent Valve will auto close on rising radiation.
'RCP seal water return heat exchanger; 2CC149 Surge Tank Vent Valve will auto close on rising
~ radiation.
r 1 \ComponentCooling Water System
\ m l A b i l i t y to (a)
~ . , .predict the impacts of the following on the Component Cooling Water System and (b) based on those "predictions, use procedures to correct, control,-or mitigate the consequences of those abnormal operation:
p/ ~PRMSalarm IR 13.5j condition. B is correct because the Thermal Barrier is exposed to full seal injection pressure, any leak in the thermal barrier would be into the CC system. C is incorrect because CCW pressure is higher than seal Ireturn pressure, and any leakage would be out of the CCW system. D is incorrect for the same reason.
I IS1.OP-AB.RAD-0001
~~ ~
IAbnormal Radiation E -
Monday, October 30, 2006 11:07:30A Page 22 of 79 Prepared by WD Associates, Inc.
'Which of the following choices describes an evolution which will require the GREATEST magnitude
'(in percent from normal) of correction signal be applied to the PZR Master Pressure Controller AND Ireturn PZR pressure to normal?
/A 1,000 gallon continuous dilution at 90% power @ EOL.
lPZR goes solid after inadvertent SI with ALL RCPs tripped. 1
- A single RCP trips while in MODE 3 with rod control deenergized.
- A single Main Turbine Governor Valve fails shut in one second at 100% power.
jolo] lpressurizer Pressure Control System E a lK6.llKnowledge of the of the effect of a loss or malfunction on the following will have on the Pressurizer Pressure A is incorrect because at EOL, with very little boron in the core, diluting has a much smaller effect on RCS temp/power/pressure. A 1,000 gallon dilution at the normal flowrate of 62 gpm will take 16 minutes to available to operate the spray valves, AND no motive force for the sprays, (RCPs tripped), the PZR pressure control system can NOT return pressure to normal, even though it will have a 100% pressure reduction signal applied to it. C is incorrect because the pressure pertubation will be assuaged by the other 3 RCP's. D is correct because the instantaneous (1 sec) downpower will cause a 25% load rejection, heatup, insurge, and pressure rise. The PZR pressure control system will have to respond with a large IReactor Enginneering Manual I IS1.RE-RA.ZZ-0012 1 7 1 1 'PZRP&;
(LE008 I I t Monday, October 30, 2006 11:07:30A Page 23 of 79 Prepared by WD Associates, Inc.
'Given the following conditions:
I
- Unit 2 is in MODE 3, NOT/NOP.
- The North 13KV bus section 6 becomes deenergized, and remains deenergized.
available for PZR pressure control?
~~ ~
121 BackuD heaters and Control Group heaters ONLY.
j22 Backur, heaters and Control Group heaters ONLY.
IR 12/11/2006\
1011 i jpressurizer Level Control System I lK2. /\Knowledgeof bus power supplies to the following:
IK2.02; /PZR heaters 1m/3.2/
Using drawing 203000-SIMPJwhen 13KV north ring bus is deenergized, the power to 22 SPT is lost. The Unit Main Generator is not online, so there is no alternate source of power to the F and G 4KV group busses. G bus supplies power to the control group and 27 B/U group of PZR beaters (dwg 601398). This leaves only the 22 B/U heaters powered from E 4KV group bus (601397) available for pressure control. 21 IB/U heaters does have a manually transferable power supply to a vital bus, but the question stem lspecifically says with no operator action. The distracters are wrong because they contain the incorrect (heaterarouw.
il"f-R%iiezNi tnp ea F&pqgii
- ]EL&,
L I ^
Simplified Oneline f i - m c z a l Distribution
'No.2 Unit Aux Building Penetration Area=
1 z3000-SIMP
-~
601397 1 I
L
-- -I E 1 1
[------lp--] ' 7 1 2
7-
- 4KV PZR htr. Bus Oneline I __---
Ic z I z I I ' F l D
=Unit Aux Building Pen-
'4KV PZR htr. Bus Oneline
'601398 L
1 L I
r z I I -
I Monday, October 30, 2006 11:07:30 A Page 24 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 1 is operating at 100% power.
- SSPS testing is in progress IAW S I .IC-ST.SSP-009, Solid State Protection System Train B Functional Test.
- Rx Trip BYPASS breaker B is racked in and SHUT.
- Rx Trip breaker B is racked in and SHUT.
- OHA A-42, SSPS TRN B TRBL is in alarm as expected.
The 48VDC power supply from B Vital bus to SSPS Train B Logic Cabinet becomes deenergized.
Which of the following describes the impact of this power supply becoming deenergized while in this Iconfiguration?
The Rx will trip when the UV coils for BOTH Rx Trip Breaker B AND Rx Trip BYPASS breaker
!becomedeenergized.
'The Rx will trip when the shunt trip coils become deenergized for BOTH Rx Trip Breaker B AND IRx Trip BYPASS breaker.
The Rx will NOT trip because the 48VDC supplied to the UV coils is powered from the SSPS Output Bay, not the Logic Bay.
The Rx will NOT trip because the loss of the auctioneered Logic Bay 48VDC power supply to trip the Rx has not been occurred.
1211112006' jA2.I Ability to (a) predict the impacts of the following on the Reactor Protection System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
IPowerDistribution Monday, October 30, 2006 11:07:30A Page 25 of 79 Prepared by WD Associates, Inc.
I Given the following conditions:
- Unit Iis performing a reactor startup.
- Initial power PRIOR to the startup was 100 cps.
- Rx power is now 4200 cps in the Source Range.
After performing a rod pull, rod motion continues when the RAISE pushbutton is released.
IIAW Salem
_ _ UFSAR. how is the dant Drotected from this event with NO operator . action, and what will I Ibe the effect on the'margin to Departure from Nucleate Boiling (DNB)?
- The Source Range High Neutron Flux Trip will ensure a large margin to DNB is maintained.
- The Power Range High Neutron Flux Trip (low setting) will ensure a large margin to DNB is maintained.
lminimal margin to DNB is reached.
l"There is a large margin to DNB during the transient since the rod surface heat flux remains below the design value, and there is a high degree of subcooling at all times in the core."
1 15 Accident
/Salem Updated Final Safety Analysis Report ;Salem UFSAR
/Analysis 11 I (22 1 ITAAOO '
IDEO15 I I l I ~ l ~
I l - ~ ' ~ r - l I II - - - - I Monday, October 30, 2006 11:07:30A Page 26 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 1 is operating at 100% power.
- PT-948C, Containment Pressure detector Channel II, fails LOW.
!Safety Injection- 2/3; Containment Spray- 213
- Safety Injection- 2/2; Containment Spray- 213
[Safety Injection- 1/3; Containment Spray- 113 jK5.1 Knowledge of the operational implications of the following concepts as they apply to the Engineered Safety Features Actuation System:
Reactor Prot. & Process Cont. Systems 1 /220026 Safety Injection interconnections I I I I
Monday, October 30, 2006 11:07:30 A Page 27 of 79 Prepared by WD Associates, Inc.
- Unit 2 is operating at 100% power.
- A Control Bank control rod drops partially into the core.
- The reactor does NOT trip.
Which of the following describes a condition that will result if the reactor is left in this configuration?
and a Xenon oscillation is occurring.
The radial flux tilt will cause certain areas of the core to burn out faster than others, leading to a jK5.I Knowledge of the operational implications of the following concepts as they apply to the Nuclear Instrumentation System :
C",
I Monday, October 30, 2006 11:07:30A Page 29 of 79 Prepared by WD Associates, Inc.
- Unit 2 has tripped from 100% power due to a Main Turbine trip.
! A 3 7 IIAbility to monitor automatic operations of the Non-Nuclear Instrumentation System including:
IA3.021IRelationship between meter readings and actual parameter value I T q p, In the time it takes to acknowledge a Reactor Trip, perform the IAs, then repeat the IAs, AFW pumps will have started on SG lo-lo level, and flow will have settled. Indicated flow to each SG will be -15E4 Ibm/hr land 60E4 Ibm actual flow to all SG. 25E4 is the range of the indicator.
Monday, October 30, 2006 11:07:30 A Page 30 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 1 has experienced a Large Break LOCA from 100% power operation.
- Before operators can respond, containment pressure rises to 10 psig.
- Off-site power remains available.
Assuming ALL automatic actions occur as expected, which of the following describes CFCU operation BEFORE operators take any MANUAL actions?
CFCUs running in HIGH speed.. .
-'receive a simultaneous HIGH speed STOP signal and a LOW speed START signal, all airflow will (bedirected throuah the ROUGHING filters ONLY.
receive a HIGH speed STOP signal, followed 20 seconds later by a LOW speed START signal, all airflow will be directed throuah the HEPA filters ONLY.
-\remain in HIGH speed with all airflow directed through the Roughing Filters, all other available ICFCUs IMMEDIATELY start in LOW speed with all airflow through the HEPA Filters ONLY.
/remain in HIGH speed until their respective Vital Bus EDG is up to speed, then receive a HIGH Ispeed STOP signal, followed 5 seconds later by a LOW speed START signal, all airflow will be
/directedthrough theHEPA AND ROUGHING filters.
IPlant Systems 1022 1 /ContainmentCooling System w I ( K n o w l e d g e of Containment Cooling System design feature(s) and or interlock(s) which provide for the following:
ICorrelation of fan speed and flowpath changes with containment pressure pJp/
The stem states that no operator manual action has been taken, and all automatic actions occur as
/expected. There will be an AUTOMATIC Safety Injection when containment pressure reaches 4 psig, and
'the stem states cont pressure is 10 psig. This will start the SEC MODE 1 sequencer. In MODE 1, most
/equipmentis immediately loaded onto its vital busses. However, the CFCUs normally operate in HIGH
'speed. In order to protect the motors when shifting to LOW speed, a 20 second time delay is incorporated lint0 the automatic LOW speed start signal. (See dwg 203673 C-1). The normal air flow path in HIGH speed is through the rouging filter, when the LOW speed breaker is shut, the HEPA filter is placed in service and the rouahina filter is removed from service bv automatic damDer oDeration.
No. 1& 2 Units Safeguards Emergency (203673 /(Sheet] )6)CONT Loading Sequence No. 1& 2 Units Safeguards Emergency 1203670 ] p Z r - - l l l p i T l r I Loading Sequence
--zxIzIlmm
, 1 I 1 Monday, October 30, 2006 11:07:30 A Page 31 of 79 Prepared by WD Associates, Inc.
I
In addition to pressing the STOP PB on CC1, which ONE of the following identifies ALL required actions for stopping the Containment Spray Pumps following automatic initiation of Containment
[Spray (cs)?
iReset Safety Injection and reset associated SEC.
_ _ _ _ _ ~ ~~
[Reset Containment Isolation Phase B Isolation signal ONLY.
[Reset Safety Injection Signal, then reset Containment Isolation Phase B.
Reset Containment Isolation Phase A, ensure containment pressure less than 14 psig (CS and lPhase B initiating signal clear), reset associated SEC I
\Memory I 12/11/2006' Reactor Protection System Reactor Trip Signals 1221051 l D l i / 1 1 3 l l Logic Containment Spray System Containment spray Pumps (239950 1~~~~
'OtherFacility-----
t I o r 'Significantly Modified I
Beaver Valley-2 2002 NRC RO Exam Question 24, Modified correct answer to reflect Salem logic for Cont Spray pump stopand editorial mod to reflect Salem terminology.
ni=- b I
Monday, October 30, 2006 11:07:30 A Page 32 of 79 Prepared by WD Associates, Inc.
1 Given the following conditions:
- A LBLOCA has occurred.
- In response to a RED path on the CORE COOLING Critical Safety Function Status Tree, FRCC-1, "Response to Inadequate Core Cooling", is currently in progress.
- Containment hydrogen concentration is 4.5%.
Which of the following states the action that is to be taken in regards to operation of the hydrogen I recombiners?
iPlace ONE hydrogen recombiner in service to reduce the hydrogen concentration.
IPlace BOTH hydrogen recombiners in service to reduce the hydrogen concentration. I ID0 NOT operate the hydrogen recombiners since they could result in ignition of the hydrogen.
/DO NOT operate the hydrogen recombiners since the hydrogen recombiner system will not be
/effectiveat this concentration.
~~
j Plant Systems 1 D E 10281\HydrogenRecombiner and Purge Control System (A2.IlAbilitv to (a) predict the impacts of the following on the Hydrogen Recombiner and Purge Control System and (b)
'The hydrogen air concentration in excess of limit flame propagation or detonation with resulting
[equipment damage in containment
[q14.01 Ireaching 4% H2. It also states that 6niy ONE is to be run at a time.
Loss of Coolant Accident 2-EOP-LOCA-I ~ E I Z I l ~ l E I 7 11 1 7 1 Hydrogen Recombiner Operation S2.OP-SO.CAN-0008 Monday, October 30, 2006 11:07:30 A Page 33 of 79 Prepared by WD Associates, Inc.
'Given the following conditions:
- Unit 1 is operating at 100% power.
- Control room operators are preparing to perform a Containment Pressure Relief IAW S I .OP-SO.CBV-0002, CONTAlNMENT PRESSURE-VACUUM RELIEF SYSTEM OPERATION.
- Containment radiation levels are NORMAL for 100% power operation with no failed fuel.
After opening the 1VC5 and 1VC6 to initiate the pressure relief, which choice describes how the respective radiation monitors indication will respond?
IR12A - Containment Gas Effluent IR41B - Plant Vent Noble Gas Intermediate Range IR41D - Plant Vent Noble Gas Release Rate Containment Pressure-Vacuum Relief System $2.OP-SO.CBV-0002 operation 1
Monday, October 30, 2006 11:07:30A Page 34 of 79 Prepared by WD Associates, Inc.
- Unit 2 is in MODE 6.
- Fuel movement is in progress in both the containment and Fuel Handling buildings.
- Operators in containment report lowering Rx cavity level.
- Due to a mis-communication, the Fuel Pool Gate valve is closed with the transfer cart in its way.
- The Fuel Pool Gate valve cannot be closed further than 40 turns closed.
Which of the following choices identifies the condition which will happen FIRST if the leak is in the RHR system, with NO other operator action?
IRHR pumps will cavitate and become air bound.
IFuel in the Spent Fuel Pool racks will become uncovered.
/The umer and lower Reactor Cavitv will comtietelv drain.
1 12/11/2006j w) Knowledge of the physical connections andlor cause-effect relationships between Spent Fuel Pool Cooling System and the following:
leak is from RHR since the lower cavity is below the level of the vessel flange (104'). A and d are incorrect because once the level drops below the Rx vessel flange, it will not affect SFP level, since the cavity level connected to the SFP through the open Gate Valve will never go below 104'. The bottom of the SFP is on the 89'6 level. The height of all the fuel racks in the SFP is 185 114". (VENDOR DWG 316748) These 2 combined is 104'. The spent fuel assembly fits down inside the rack, so it can never become uncovered if iP0d Layout for Spent Fuel Storage Racks 1 jVTD 316748 2E001
[Drainingthe Reactor Coolant System I
J 1
Monday, October 30, 2006 11:07:30 A Page 35 of 79 Prepared by WD Associates, Inc.
1
[Giventhe following conditions:
- Unit 2 is at 100% power.
- Surveillance testing is in progress at the turbine front standard.
- A combination of equipment failure and human error causes an automatic SI signal to be generated only on RPS Train A.
- Reactor Trip Breaker A (RTB A) fails to open, and is stuck in the shut position.
With NO operator action, which of the following indications will be present in the control room?
lOHA F-40 RX TRIP clear.
iPZR level below program. I
'ALL Steam DumP Valves are shut. I
- ALL Main Turbine Stop Valves shut.
I I I I 12/11/2006~
1039 j /Main and Reheat Steam System jA4. /;Abilityto manually operate andlor monitor in the control room:
I iA4.01 /Mainsteam supply. valves 12.8j The AUTO SI on Train A causes the RPS system to send a trip signal to RTB A, and Bypass breaker B ONLY. A MANUAL SI,on the other hand, sends trip signals to all 4 trip and bypass breakers, which would
'trip the reactor even with RTB A stuck shut. In these circumstances, the Rx will still trip, because a signal is (sentto trip the Main Turbine. The Main Turbine trip >P-9 will cause a Rx trip, which will send a trip signal to BOTH RTBs. The AUTO SI signal on Train A sends a signal to the AUTO STOP OIL section of the Turbine Itrips. (221065, D-4) This causes Auto Stop Oil to be dumped from the Main Turbine Stop Valves, and they will shut. Distracter c is incorrect because with the Main Turbine tripped, the steam dump valves will
/openfully. Distracter a is incorrect because the Rx Trip OHA is annunciated when EITHER Rx Trip Breakei
'and its associated Bypass breaker are open, and the turbine trip will open RTB B. Distracter b is incorrect
'because when the turbine trips, programmed PZR level will go from 52% to 22.3%. PZR level will remain
!above program until it reaches program.
Reactor Protection System Reactor Trip I221051 OTE02
'Reactor Protection System Turbine Trips, 1221065 1-1 ~ ~ 1
!Runbacks & Gen Protection
- ~ E I z I l ' l I I z l ~
I 1 Monday, October 30, 2006 11:07:30A Page 36 of 79 Prepared by WD Associates, Inc.
I 1 Given the following conditions:
- Unit 2 is operating at 100% power.
- 21 SGFP must be removed from service to repair a small steam leak.
IWhich of the following is the MAXIMUM Rx power which will allow 21 SGFP to be removed from service IAW S2.OP-SO.CN-0002, STEAM GENERATOR FEED PUMP OPERATION?
166%. J 155%. 1 135%. I 1Al.IAbility to predict andlor monitor changes in parameters associated with operating the Main Feedwater System controls including:
Ion I
a single sGFP trip is automatically disarmed. Distracter d is incorrect because it is the power level at lwhich the feed pump is removed from service IAW IOP-4.
Isteam Generator Feed Pump Operation I lS2.OP-SO.CN-0002 ) 1 I lpower Operation I ~s2.oP-Io.zz-ooo4 DWEO Monday, October 30, 2006 11:07:30 A Page 37 of 79 Prepared by WD Associates, Inc.
[Given the following conditions:
I
- Unit 1 is operating at 85% power.
- 11 charging pump is in service.
- During a manual bus swap prior to clearing and tagging a Station Power Transformer, the 1A 4KV vital bus is inadvertently deenergized, and the SEC loads 1A bus in Mode 2*.
I- All other electrical bus transfers expected to occur from the loss of 1A 4KV vital bus are successful.
With NO oDerator action, which of the following lloss of 1A 4KV vital bus?
- identifies the plant condition 5 minutes after the initial
!Reactor power is >85%.
IPZR level is rising at 1 % per minute.
IThe Main Turbine will have run back to 60%.
IPZR Backup heaters have cvcled on due to low Dressure and remain ON. I I
/R I
/Application I i 12/11/2006~
IA2.IAbility to (a) predict the impacts of the following on the Auxiliary I Emergency Feedwater System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
I I
Monday, October 30, 2006 11:07:30 A Page 38 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 1 was tripped from 100% power due to a steam leak.
- A MSLl was successful in isolating the leak.
- The PO idles 23 AFW pp, and throttles AFW flow in EOP-TRIP-2 to 6E4lbm/hr to each SG.
Which of the following describes how AFW flow will be affected if 21 AFW pump trips with NO operator action?
~AFWflow will.. .
!remain the same to 21 and 22 SGs, and lower to 0 Ibm/hr to 23 and 24 SGs as all AFW flow will
!be lost to these 2 SGs.
IdroD to zero on 23 and 24 SGs, and rise on 21 and 22 SGs due to the lower pressure in those ISGs when 23 and 24 SGs stop steaming.
b o p to zero on 21 and 22 SGs, and lower on 23 and 24 SGs due to the higher pressure in those I
,SGswhen 23 and 24 SGs steam more as they heat up with no AFW flow.
remain the same to 21 and 22 SGs, and lower to some value > 0 Ibm/hr to 23 and 24 SGs lregardless of the relationship between 23 AFP outlet pressure at minimum speed and SG inlet Ipressure.
4 /Comprehension I 1 1
12/11/2006 IK6.1Knowledge of the of the effect of a loss or malfunction on the following will have on the Auxiliary I Emergency Feedwater System:
own oil coolers. Since the distracter says regarless of the relationship, it is not always true.Distracter b is incorrect because AFW flow will not rise. Distracter c is incorrect because AFW flow will not lower.
Distracter d is incorrect because of the discharge piping being backwards.
\Reactor Trip Response I 12-EOP-TRIP-2 OEOl6
[Auxiliary Feed System Operation r I1 l- - r- - l m Monday, October 30, 2006 11:07:30 A Page 39 of 79 Prepared by WD Associates, Inc.
- 2C EDG is operating in parallel with the 500KV grid for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> endurance run IAW S2.0P-ST.DG-0014, 2C DIESEL GENERATOR ENDURANCE RUN, following a complete overhaul.
- Cumulative run times for all individual EDG load limits are less than 10% of rated.
- While operating at 2525 KW three hours into the test, the operator mistakenly adjusts 2C EDG speed control resulting in MW loading increasing to 2800 KW.
'Which choice describes the consequences, if any, of continued EDG operation at this KW load?
[will result in exceeding the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> load limitation for 2C EDG.
lwill result in exceeding the 30 minute load limitation for 2C EDG I
/will result in exceedina the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> load limitation for 2C EDG.
12/11/2006~
1062 I1A.C. Electrical Distribution w I [ A b i l i t y to predict and/or monitor changes in parameters associated with operating the A.C. Electrical Distribution
/controls including: I IA1.01; \Significanceof DIG load limits 1p4; j3.8j (2CDiesel Generator Endurance Run 1 rS2.OP-ST.DG-0014 OEOl2 I II I I)
Monday, October 30, 2006 11:07:30A Page 40 of 79 Prepared by WD Associates, Inc.
- Units 1 and 2 are operating at 100% power.
- 4KV Vital buses 1A and I C are powered from 13 SPT. 1B is powered from 14 SPT.
- 4KV Vital buses 2A and 2B are powered from 24 SPT. 2C is powered from 2C EDG running in parallel with the grid.
- All other electric system lineups are normal for full power operation.
- A fault occurs which sends a trip signal to the North 13KV ring bus breaker 1-6, but it does NOT open.
Which of the following describes the effect this will have on the plant with NO Operator action?
lMain Generator Mwe output will lower.
vital bus will never see a UV condition. B is correct because the loss of 3 circulators per unit will cause MWE output to lower. Distracter d is incorrect because TSAS 3.8.2.3 for DC busses in MODES 1-4 reauires one OPERABLE batterv charaer. and if it is not, connect the backur, batterv charaer.
, I
- *e ^5 P.7 E!? Page hurnhid&if 3evisf$iT C.'O, d-Facilrty-_R
- AC Electrical Distribution Simplified One-L'Lng: '203000-SIMP
- , ;y2%i'13KVA '
1 CEO16
- I
/Technical Specifications j 13.8.2.3 158 Monday, October 30, 2006 11:07:30A Page 41 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- U2 is performing actions to isolate a 125VDC ground on 2A 125VDC bus,
- An Equipment Operator depresses the local pushbutton for 2A 125VDC bus to read bus resistance-
/2A 125VDC bus ground indication will indicate infinity.
[2A 125VDC bus ground indication will indicate zero ohms.
/OHA B-2.2A 125VDC CNTRL BUS VOLT LO will annunciate. I voltage indication for the bus and won't actuate the low voltage alarm. D is incorrect because there is no such AAT alarm. However, if the student does not know the correct answer, it is a plausible distracter.
Monday, October 30, 2006 11:07:30 A Page 42 of 79 Prepared by WD Associates, Inc.
The EDG is INOPERABLE until 125VDC control power has been transferred to its alternate source.
I 12.2 /(EquipmentControl F I /Knowledgeof limiting conditions for operations and safety limits.
blSlON Q82890 I
1 Monday, October 30,2006 11:07:30A Page 43 of 79 Prepared by WD Associates, Inc.
'Given the following conditions:
- BOTH Air Receivers were left isolated.
Which of the following describes the effect this will have on 1A EDG OPERABILITY IAW Technical Specifications?
I lThe EDG.. .
Iremains OPERABLE, since either air start receiver is designed to provide 3 cold starts.
starting air if required.
became INOPERABLE when the Air Compressor supply valve to the second Air Receiver was closed since NO starting air is available to the EDG.
' 1 IEmergency Diesel Generators IK6.\/Knowledge of the of the effect of a loss or malfunction on the following will have on the Emergency Diesel
/Generators:
v/ /Air receivers 112.71 j2.9; A is correct. As shown on dwg 211315, the isolation of air from the compressors to the tanks will not affect the air supply path to the starting air motors. Each Starting air tank IS designed for three cold starts when at 160 psig. Distracter b is incorrect because the EDG remains OPERABLE, but the turbo boost air receivers can NOT be cross connected with the starting air receivers. Distracter c is incorrect because the air "in" to the tank is a separate line from the air "out" of the tank to the EDG. Distracter d is incorrect because the lo pressure alarm is at 182 psig, not 160 psig.
IComponent Design Basis / IDE-CB.DG-0024, DTL 1 7 1 (5-21 No. 1 Unit-1A Diesel Generator Start and Turbo Boost air System 7 7 11 / 1 7 1 I I Monday, October 30, 2006 11:07:30 A Page 44 of 79 Prepared by WD Associates, Inc.
RELEASE OF RADIOACTIVE LIQUID WASTE FROM #I WASTE MONITOR HOLDUP TANK?
/A2.\Ability to (a) predict the impacts of the following on the Liquid Radwaste System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
'RELEASE OF RADIOACTIVE LIQUID
'WASTE FROM #I WASTE MONITOR HOLDUP TANK?
Is1.OP-SO.WL-0003
=
' -"lY QEOl2 I :-c-DE=l
- m m m l - l 1- I Monday, October 30, 2006 11:07:30A Page 45 of 79 Prepared by WD Associates, Inc.
Fuel Storage Area Radiation Monitor, the counts seen by the monitor rise above the Hi Radiation
[Which of the following describes the consequences of this action?
(The Fuel Handling Building (FHB) supply and exhaust fans will receive an auto start signal.
The FHB Hi Radiation Evacuation Horn will sound, but no ventilation system realignment will The FHB Hi Radiation Evacuation Horn will sound, but no ventilation system realignment will iMernory I j I 12/11/2006' IA4. !(Abilityto manually operate and/or monitor in the control room:
(A4.01\Alarmand interlock setpoint checks and adjustments I p - 3 p-q A is incorrect because the fans do NOT receive an auto start signal. B is correct, and c and d are incorrect because the check source test does not automatically block an actuation nor does the procedure have the RPI block switch placed in BLOCK for the test.
[IWDIATION MONITORS - CHECK SOURCES[\Sl.OP-ST.RM-0001 Monday, October 30,2006 11:07:30 A Page 46 of 79 Prepared by WD Associates, Inc.
1 Given the following conditions:
- Unit 2 is operating at 100% power.
- Unit 1 is in MODE 5.
- All Unit 1 circulators are secured.
- Unit 2 Waste Liquid release is in progress from 21 CVCS MT to UNIT 2 CW system via the cross
/2WL51AND 1WL51 will shut.
'1WL115 and 2WL115 will shut. I j2WL51 will shut, but release will continue through IWL51.
I 12/11/2006]
IO73 I 1 /Process Radiation Monitoring System jK3. ~
Knowledge of the effect that a loss or malfunction of the Process Radiation Monitoring System will have on the following:
Distracter b is incorrect because 2R18 will not auto close the opposite unit WL51. Distracter c is incorrect because the WL115s are manually operated valves. Distracter d is incorrect because the release path will be isolated when the 2WL51 shuts.
'Release of Radioactive Liquid Waste from 22 CVCS Monitor Tank S2.OP-S0.WL-0001 I -- -I r y! II 1 VISION (250403 Monday, October 30,200611:07:31A Page 47 of 79 Prepared by WD Associates, Inc.
predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
[CirculatingWater System Malfunction 1 /SI .OP-AB.CW-0001 I I I
I I Monday, October 30, 2006 11:07:31A Page 48 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 is operating at 100% power.
- A large earthquake 5 miles from the site causes a loss of off-site power.
- The reactor trips, and a MANUAL Safety Injection is initiated.
- 2B EDG output breaker does NOT close.
With NO other operator action, which choice contains the system lineup for the Service Water System I 15 minutes after the SI?
12SW26 SHUT, 21SW122 SHUT, 24SW223 SHUT.
i2SW26 SHUT, 22SW122 OPEN, 25SW223 OPEN.
12SW26 OPEN, 22SW122 SHUT, 23SW223 OPEN.
12SW26 OPEN, 21SW122 OPEN, 22SW223 SHUT.
(A3.IIAbility to monitor automatic operations of the Service Water System including:
the "A"and "B" SECs respectively. The SEC, while not having vital loads to sequence, still performs its ancillary control functions of closing the SW122.
No. 2 Unit - Auxiliary Building #22 Component 1218912 Cooling Heat Exchanger
' 'UCEOO INo. 2 Unit - Auxiliary Building # 21 and 22 ~ 1220942 I 7 1j 18 i!SWON 1
~CCHXInlet Control lUCEo0 17 i Modified a distracter to even out all the openlshuts.
Monday, October 30,2006 11:07:31 A Page 49 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 1 is operating at 100% power.
- 12 charging pump is in service.
- Normal letdown must be secured to troubleshoot a control problem with 1CV18, LETDOWN PRESSURE CONTROL VALVE.
- Excess letdown has been placed in service.
Prior to securing Normal Letdown, which of the following actions MUST be performed IAW S I .OP-S0.CVC-0001, CHARGING, LETDOWN, AND SEAL INJECTION, and why?
Fully close the 1CV55, CENT CHG PMP FLOW CONT VALVE, to ensure RCP seal injection remains above 6 gpm per pump.
'Place 1CA2015, CONTROL AIR SUPPLY TO CV55 BYPASS VALVE, in BYPASS, to allow the
!ICV55 to control flow less than the normal minimum flow position.
'Adjust the position of the speed control linkage for 13 charging pump to a lower pressure position, lto prevent exceeding the Tech Spec limit of 40 gpm total flow to the RCP seals.
12/11/2006
/K4.//Knowledge of Instrument Air System design feature(s) and or interlock(s) which provide for the following:
/Manual/automatictransfers of control /j2.71'pq Distracter a is incorrect because flashing is only a concern on the LETDOWN line when charging flow is reduced below 60 gpm with normal letdown flow established due to the cooling of letdown flow in the Regenerative Heat Exchanger. The Excess letdown HX is only cooled by CCW, there is no Regen function. Distracter b is incorrect because the CV55 will lower flow to the RCP seals, and the minimum flow stop will be in excess of the required CVCS flow to minimize PZR level rise AND is not required by the procedure. C is correct because at step 5.3.2 of the procedure, between the steps for placing Excess Letdown in service and securing Normal Letdown, is the step(s) to change the min flow stop for either the centrifugal charging pump FCV CV55 or the speed linkage for the PDP, whichever is in service. Distracter d is incorrect because the speed linkage is adjust to allow lower charging flow to minimize the gain in PZR level while maintaining >6 gpm per pump to the RCP seals. Also the pump is not in service, and the linkage adjustment is only required when it is.
\Charging, Letdown, and Seal Injection 1 IS1.OP-S0.CVC-0001 1 1 1 !I4
[Excess Letdown Flow 1 Is1.oP-so.cvc-ooo3 I j Im 1 5 1 OEOl3 l ' ~ ~ ~ E I l u
Monday, October 30, 2006 11:07:31 A Page 50 of 79 Prepared by WD Associates, Inc.
Which of the following events would require the transfer of spent fuel elements to the Spent Fuel Pool '
to be suspended during MODE 6 refueling operations IAW S2.OP-S0.SF-0009, REFUELING
/OPERATIONS?
- Fuel Handling Area Rad monitor 2R5 fails low.
I
,Only one FHB Supply Fan and 2 FHB Exhaust Fans are running.
'An SRO over-seeing Spent Fuel Pool manipulations leaves the area under supervision of a
/qualified Reactor Engineer.
'21 Spent Fuel Pool Cooling pump is discovered to have no oil in its pump oil bubbler with 22
!Spent Fuel Pool Cooling Pump in service.
m I / E q u i p m e n tControl 1 OPERABLE IAW TSAS 3.3.1 .I, Table 3.3-6. Distracter b is the complement of fans required to be running to have an OPERABLE FHB ventilation system. Distracter c is incorrect because the requirement for supervision of loads in the Spent Fuel Pool is a SRO OR a Qualified RE. D is correct because in S2.0P-S0.SF-0009, REFUELING OPERATIONS, P&L 3.12 specifically requires suspension of irradiated fuel into (RefuelingOperations I IS2.OP-S0.SF-0009 LE012 ,
/Technical Specifications 113.3.1 .I Monday, October 30,2006 11:07:31 A Page 51 of 79 Prepared by WD Associates, Inc.
/Whichof the following choices identifies the relationship between a Rx trip and a Turbine trip?
!A Turbine triD will ONLY cause a Rx trip if Dower is P-9.
iA Rx trip will ONLY cause a Turbine trip if Rx power is >P-9.
,A Turbine trip will ALWAYS cause a Rx trip to prevent lifting the PZR safeties.
- A Rx trip will ALWAYS cause a Turbine trip to prevent an uncontrolled cooldown of the RCS. ~~
12/11/2006/
1- IReactor Trip
'-]\Knowledge of the operational implications of the following concepts as they apply to Reactor Trip:
lEKl.031 /Reasonsfor closing the main turbine governor valve and the main turbine stop valve after a reactor trip in A is incorrect because the power level needs to be ABOVE P-9 for a turbine trip to cause a Rx trip. B is incorrect because a Rx trip ALWAYS causes a turbine trip. C is incorrect because a turbine trip <P-9 will not cause a Rx trip. D is correct because a Rx trip always causes a turbine trip and the reason is IAW the Bases Document w $
- ,w
-su
. li.P f " v'.
1 F i m e - k c e n p Nu
&Trip- or Safe'& Injection ~-
Bases
__ - D o G g n q I2-EOP-TRIP-1 I
I I
Monday, October 30,2006 11:07:31 A Page 52 of 79 Prepared by WD Associates, Inc.
The crew has diagnosed a Pressurizer (PZR) Vapor Space Accident.
The following is the procedural flowpath followed:
-EOP-TRIP-1, REACTOR TRIP OR SAFETY INJECTION
-EOP-LOCA-1, LOSS OF REACTOR COOLANT
-EOP-LOCA-2, POST LOCA COOLDOWN AND DEPRESSURIZATION 90 minutes after the reactor trip, performing the COOLDOWN and DEPRESSURIZATION of the plant per LOCA-2 will result in ...
/a stable flowrate out the vapor space leak, and lowering PZR level.
/a reduction in flowrate out the vapor space leak, and increasing PZR level.
/a reduction in flowrate out the vapor space leak, and PZR level offscale high.
- Post Loca Cooldown and Depressurization I 12-EOP-LOCA-2 1,:T---l125,JLOCAO1
,2E002 ,
Monday, October 30, 2006 11:07:31 A Page 53 of 79 Prepared by WD Associates, Inc.
I Given the following conditions:
Unit 2 is operating at 100% power.
RCS Tavg is 573 degrees.
The unit experiences a SBLOCA.
RCS pressure has dropped from NOP to 1825 psig.
Using trended data, the highest CET has dropped from 614 degrees to 560 degrees.
/Subcoolinghas gone from to .
j39;64 164; 39 181;93 193;81 I
1 ja IApplication [Salem I& 2 I 12/11/20061
'-][Knowledge of the operational implications of the following concepts as they apply to Small Break LOCA:
jEKl.02: \Use of steam tables 13.51/4.21 Saturation temperature at NOP (2250 psia) is 653 deg. Highest CET in stem is 614. 653-614=39.
Saturation temp at 1840 psia is 624 deg. Highest CET in stem is 560. 624-560=64. 81 degrees is subcooling if use TAVG of 573 instead of highest CET. 93 is if use 2235 psig for current pressure and current temp of 560.
I II Isteam Tables 1 I ' I 1-Monday, October 30, 2006 11:07:31 A Page 54 of 79 Prepared by W D Associates, Inc.
- Salem Unit 2 is operating at 100% power.
- 22 RHR pp is C/T.
- A catastrophic failure of RCS loop 21 cold leg piping occurs.
- RCS pressure is 35 psig.
- Initial RWST level was 41 .Ifeet.
EiBetween 22-23 minutes.
/Between 18-19 minutes.
=//Knowledge of the interrelations between Large Break LOCA and the following:
7-- ____
,TANK CAPACITY DATA ECCSOl
/OE0*8 I
/2-EOP-LOCA-3 1E004 Monday, October 30, 2006 11:07:31 A Page 55 of 79 Prepared by WD Associates, Inc.
~ 1 Given the following conditions:
- Unit 2 is operating at 30% power, steady state.
- OHA D-29,22 RCP BKR OPEN/FLO LO is received.
- All 22 loop RC flows are 85% and dropping.
- The red START bezel for 22 RCP is illuminated.
- The reactor has NOT tripped.
lWhich of the following identifies what has occurred?
/AnATWT. The Rx should have trimed on 1/4 RC Loom Lo Flow ~ 9 0 % . I
/22 RCP shaft has sheared.
I22 RCP shaft has seized.
1 L R [Comprehension I 12/11/20061 W j l A b i l i t y to operate and / or monitor the following as they apply to Reactor Coolant Pump Malfunctions:
jRCP onloff and run indicators pq12.4j With a RCP shaft shear, there is no event that would cause the RCP breaker to open. For this reason, that is why the START bezel will still be illuminated, even though loop flows are all dropping. Distracter b is incorrect because between lO%(P-IO) and 36%(P-8), 1/4 RCS loop lo flow will NOT cause a Rx trip, the coincidence is 2/4. Distracter a is incorrect because there are 3 low pressure flow taps, and 1 common high pressure flow tap. Distracter d is incorrect because a seized RCP shaft would cause its supply breaker to ltrip on ovewrcurrent. The indication in the stem is that the breaker is closed. C is correct because a
/sheared shaft would cause that loop flow to drop, even while the bezel indication showed the breaker is still Iclosed.
Nii Section
- e x
- MPEOO
~
Reactor Coolant System ~ ---J;205301-21[-----I 1 - 1 31 ~RCPVI U
Overhead Annunciators Window D IS2.OP-AR.ZZ-0004 1I 71 ! ;231T I
/New I I I1 I 1-I I1 I 1-Monday, October 30, 2006 11:07:31 A Page 56 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 is operating at 40% steady state power.
- 23 CVCS Pp I/S
- 21,23 CC PPSIIS
- 21,24 SW PPSI/S
- 22 SW Pp in AUTO I
iPZR level dropping -1 %/minute.
IControl Console alarm, 21 (22) CC HDR PRESSURE LO.
[OHA B13 21 SW HDR PRESS LO, andlor OHA B-14,22 SW HDR PRESS LO.
I 12/11/20061 IEmergency and Abnormal Plant Evolutions 1; rIlil Reactivity control 2. Core cooling and heat removal 3. Reactor coolant system integrity 4. Containment conditions 5. Radioactivity release control.
'requires ALL 3 charging pump breakers to be open, it does not operate on no flow. B is correct because
!with no letdown isolation and normal letdown flow of 75 gpm, with no charging flow PZR level will drop. A lthumbrule is 75 gallons per percent in the PZR at NOT. C is incorrect because 21 CC pump is powered
/from 2A 4KV bus. and is not affected. D is incorrect because no SW Dumps will be lost.
1
- - ~
'NO.1&2 Units-CVCS No. 1CV4&2CV4 Letdown Oriface Isolation Valves I I IL - 3 -
I II - - ' - I
/
\Direct From Source Monday, October 30,2006 11:07:31 A Page 57 of 79 Prepared by WD Associates, Inc.
concentration. The RWST distracter is incorrect because the boron concentration is much less than the BAST'S,which is the source of the other 3 methods. Using the REM figures, the differential boron worth is -
6.325, -6.725, and -6.9 pcm/ppm respectiveley for a,b, and c. C is correct because it has the highest reactivity worth for the same boron flow rate.
\Figures 1 b2.RE-RAZZ-001 OEOl5 Monday, October 30,2006 11:07:31 A Page 58 of 79 Prepared by W D Associates, Inc.
- Unit 1 is operating at 100% power.
- Rod Control is in MANUAL.
- A PZR Code Safety valve fails full open.
'Which of the following describes the consequences of this event?
iWith rod control in MANUAL, the operator will not be able to insert negative reactivity fast enough 1
[Emergencyand Abnormal Plant Evolutions 1/
Distracter a is incorrect because the UFSAR says the minimum DNBR of 1.24 will be maintained, since the reactor will trip on OT/DT. B is correct because the minimum DNBR of 1.24 will be maintained, since the reactor will trip on OT/DT. Distracter c is incorrect because while the 95% part is correct (UFSAR Section 15.4.4.1 .I, page 4-4-2a) it has nothing to do with the low pressure SI. Distracter d is incorrect because the
[UFSAR- DNB Design_-___-__-
,Frequency
~
__----_---__I Basis UFSAR- Condition II Faults of Moderate 2
' 'Section 15.2- -- -71 - 1 _- '18 1 TAAOO OEOl5 7 7 Monday, October 30, 2006 11:07:31 A Page 59 of 79 Prepared by WD Associates, Inc.
- Unit 2 is operating at 100%.
- Reactor Trip Breaker "A" and Reactor Trip Bypass Breaker "B" are racked in and shut.
- Reactor Trip Breaker "B" is open.
- A feedwater problem has developed, and the CRS directs the RO to trip the reactor.
- The RO depresses the OPEN pushbuttons for the Rx Trip Breakers, but the Rx does NOT trip.
Assuming no automatic trip demand has been generated, and the RO has NOT attempted a trip by
,any other means, which of the following conditions prevented the Rx from tripping? -
[Reactor Trip Breaker " A UV coil did not de-energize.
/ReactorTrip Bypass Breaker "B"shunt coil did not energize.
/Reactor Trip Bypass Breaker "B" UV coil did not de-energize. I (Comprehension ] 12111/20061 GGgGcyandAbnormal - Plant Evolutions__ ' 1- '-7 _--_
Anticipated Transient Without_Scram
- - 7
_Anticipated
_ _ _ _ Transient __- Scram and the following: -A h???Knowledge<f the interrelations between -Without
-~ _ _ _ __________ --__
I - -_-
hK2.06' 'Breakers,relays,anddisconnects --_ -_ _ -_ -- - - _- -
12-FE
= I - - - -
The control console PB are only control-function for the Reactor Trip Breakers. The Reactor Trip Bypass
'breakers are indicate only on 2CC2. The Reactor Trip breaker CC2 PB ONLY energizes the shunt coil of its specific breaker. The correct answer is "a" because the shunt coil for Reactor Trip Breaker "A" did not energize to open the breaker. Opening EITHER of the 2 breakers in series will remove power to the lrods and cause a rx Trip. Distracter "b" is incorrect because the UV coil IS not expected to de-energize when the breaker bezel PB is depressed. Distracters 'IC" and "d" are incorrect because the Bypass IBreakers do not have a control function from the 2CC2, only breaker ~ -- _-_
_ - position indication.
Protection System Reactor Trip Sianals I II Monday, October 30, 2006 11:07:31 A Page 60 of 79 Prepared by WD Associates, Inc.
,Giventhe following conditions:
- Unit 2 was tripped from 100% power 20 minutes ago.
- 2N35 indicates 2.OE-9 amps.
- 2N36 indicates 5.OE-11 amps.
- Intermediate range Nl's are indicating.. ..
,-Ithat N35 is under compensated. Manually reset Source range channels.
/that N36 is over compensated. Manuallv reset Source ranae channels. I lcorrectlv. Ensure P-6 is blocked when the second NI channel goes below 7E-1Iamps.
correctly. Ensure Source Range channels reset when the second NI channel goes below 7E-11 amps.
ja /R I I 12/11/2006~
]Emergency and Abnormal Plant Evolutions j 7 1 /Lossof Source Range Nuclear Instrumentation
[MI./;Abilityto operate and / or monitor the following as they apply to Loss of Source Range Nuclear Instrumentation: I iAA1.01/ IManual restoration of power dpm = 6.6 decades. 100% power is 5E-5Amps, so in 20 minutes power should have dropped at least to 5E
- 11. With the N35 channel reading 2E-9 it is over 2 decades above where it should be. This points to undercompensationas the problem, since more of the gamma pulses are being seen by the detector which are not "screened out". Distracter b is incorrect because if N36 was overcompensated it would read low off
]Reactor Trip Response 1 IEOP-TRIP-2
'Modified VlSlON Q61948 to different detector having different compensation problem, making a distracter right and the previously correct answer wrong.
Monday, October 30, 2006 11:07:31 A Page 61 of 79 Prepared by WD Associates, Inc.
- A loss of BOTH Intermediate Range NI channel indications an 2CC2 has occurred. Indication is still available at the NI racks, and NO IR NI bistables have tripped.
- Prior to losing the IR NI indication, BOTH channels were reading 1E-5 Amps.
$ource Range Nls indicating 9E5 cps.
[Power Range Nls indicating 50% power.
/0331\Lossof Intermediate Range Nuclear Instrumentation pz2: \\Abilityto determine and interpret the following as they apply to Loss of Intermediate Range Nuclear Instrumentation:;
IEquivalency between source-range, intermediate-range, and power-range channel readings 11 3 .013.1 51 The Intermediate Range NI indication is expected to overlap with the Power Range between 4 and 6 E-6 Amps (IOP-3 PAGE 27). 2E-5 is a little more than 1/2 a decade higher, which is approximately 7% power in the Power Range. Distracter a is incorrect because it is the average 100% power core D/T. Distracter b is incorrect because it is approx 1E-IO in the IR. C is incorrect and D is incorrect because programmed SGFP D/P is 50 psid minimum, ramping to 150 psid up to 100% feed flow. Feed flow follows steam flow, and steam flow in the PR is Rx Dower.
[Hot Standby to Minimum Load 1 ~s2.oP-Io.zz-ooo3 lr1]5----/25pm-I 1REEoo 9 I (Reactor Engineering Tables 1 IS2,RE-RA.ZZ-0011 ]7 11 3 1 /2041 I I1 ----I I I 1 I I
Monday, October 30, 2006 11:07:31 A Page 62 of 79 Prepared by WD Associates, Inc.
To prevent high alarm on 2R40, RAD MON CONDENSATE PRCS FILTER, from isolating the Condensate Polisher.
To prevent backfeeding contamination from 21 S/G to any other S/G through the unaffected SIG's blowdown lines.
To prevent the spread of contamination from a Steam Generator Tube Rupture (SGTR) on 21
$/G to secondary systems.
12/11/2006 IEK3. ljKnowledge of the reasons for the following responses as they apply to Steam Generator Tube Rupture:
Isteam Generator Tube Rupture j /EOP-SGTR-I r, 1E001
/Steam Generator Blowdown Operation I
1IS2.OP-SO.GBD-0002 I [ I/ 5 1 1rT I
- -~
-- ~
'Facility Exam Bank IDirect From Source 1- I Monday, October 30,2006 11:07:31 A Page 63 of 79 Prepared by W D Associates, Inc.
Given the following conditions:
- Unit 2 is operating at 85% power.
- Rx power is rising slowly.
- RCS Tave is dropping slowly.
- Containment pressure is 0.1 psig and steady.
Which of the following is causing these indications, and what actions are required? I IA Main Turbine Governer Valve is slowly failing open, trip the RX IAW S2.OP-AB.STM-0001.
iA RCS leak > 10 gpm in the letdown piping OUTSIDE containment, isolate letdown IAW S2.0P-IAB.RC-OOOI, REACTOR COOLANT SYSTEM LEAK.
A normal dilution of 100 gallons was set as 1,000 gallons in the Primary Water Flow Register and Performed. Initiate a ratid Boration IAW S2.OP-S0.CVC-0006, RAPID BORATION.
An inadvertent boration is occurring, place the CVCS Make-up Control in MANUAL and close malfunctioning valves IAW S2.OP-S0.CVC-000, BORON CONCENTRATION CONTROL.
12/11/2006 (040 ][SteamLine Rupture lAA2. /[Abilityto determine and interpret the following as they apply to Steam Line Rupture: J p i IConditions requiring a reactor trip 1p-q14.71 IA is correct because it would cause all the indications in the stem. The AB.STm states at steps 3.3-3.6 that if an EH problem is causing turbine load to be improperly controlled with the turbine >49% power, the trip the reactor. Distracter b is incorrect because an RCS leak would not cause a power change or Tave change, even though it would cause charging flow to rise. Distracter c is incorrect because the dilution would not cause Tave to lower, it would rise along with Rx power and charging flow as PZR level rose.
Distracter d is incorrect because a boration would cause Rx power to lower and charging flow to lower as PZR programmed level dropped while lowering RCS temperature.
\ExcessiveSteam Flow ~ IS2.OP-AB.STM-0001 1 7 11 2 1 1 9 1 y i M I EO0 II I 7
I 7 1 II I 1-Monday, October 30, 2006 11:07:31 A Page 64 of 79 Prepared by WD Associates, Inc.
'Given the following conditions:
I I
- Unit 2 is operating at 100% power when a simultaneous loss of BOTH SGFPs occurs.
- The Rx is manually tripped.
Which of the following choices describes how AFW flow will be controlled after the Immediate Actions of EOP-TRIP-I are performed, and why?
Manually reduce total AFW flow to no less than 22E4 Ibm/hr to prevent an excessive RCS cooldown.
Ensure total AFW flow is no less than 44E4 lbmlhr to prevent an un-needed transfer to FRHS-1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.
'The Pressure Overide Defeat PBs will be required to be depressed since runout flow cannot be
!preventedto the SGs when they shrink and depressurize following the loss of feed.
AFW flow from 2 MDAFW pumps is sufficient for decay heat removal following ANY Rx trip, so only 23 AFW pump flow should be reduced to zero by idling the 23 AFW pump to prevent over ifeedina the SGs.
p54 !!Lossof Main Feedwater E l k n o w l e d g e of the reasons for the following responses as they apply to Loss of Main Feedwater:
m] IManual control of AFW flow control valves ]I3.8jF ]
maintained. 23 AFW pump speed is reduced to idle. The Basis Document for TRIP-2 references C0542, which provides direction to throttle AFW flow to minimize cooldown from excessive feedwater flow. A is correct because it contains both parts of the requirement in TRIP-2. Distracter b is incorrect because 44E4 Ibm/hr is the AFW flow required in FRSM-1, and will over cool the RCS. Transfer to FRHS-1 would leaving 2 MDAFW pumps
]Reactor Trip Response I [EOP-TRIP-2
,2E002 I
I 1 Monday. October 30, 2006 11:07:31 A Page 65 of 79 Prepared by W D Associates, Inc.
- Unit 2 has lost all off site power.
- 2A EDG failed to start.
- 2B EDG tripped on overcrank.
- 2C EDG started but its output breaker tripped on 2C Vital bus differential current.
- After isolating SW to the Turbine Building at Step I 9 of EOP-LOPA-I , Loss of All AC Power, 2A EDG is successfully started and its output breaker is shut.
[Which of the following describes the next action(s) to be performed, and why?
/Start 21 or 22 SW pump to provide cooling to 2A EDG.
$tart 25 or 26 SW pump to provide cooling to 2A EDG. I
!Close 22 and 24 SW20 valves, NUC HDR IS0 VLVS to prevent water hammer to the 2A EDG, start ONE SW pump, and throttle open the 22SW20 to repressurize the nuc header prior to putting full flow to the header.
Close 22 and 24 SW20 valves, NUC HDR IS0 VLVS to prevent water hammer to the 2A EDG, start ALL available SW pumps, and throttle open the 22SW20 to repressurize the nuc header prior to putting full flow to the header.
E El /Application I I
lSalem 1 & 2 1 I
I 12/11/2006 I I 1-r /Station Blackout M j l A b i l i t y to operate and / or monitor the following as they apply to Station Blackout:
haDDened later in the Drocedure.
Monday, October 30, 2006 11:07:31 A Page 66 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 is operating normally at 100% power.
- 23 charging pump is in service.
- 21 and 24 SW pumps are in service.
- A loss of the 500KV switchyard occurs.
Which of the following contains ONLY equipment that will be running as determined by the 2RP4 status lights?
- 2 ECAC, 21 CFCU, 22 AFP.
124 SW pump, 21 CC pump, 22 Chiller.
p5 CFCU, 23 Rx Nozzle Support Fan, 2 ECAC.
1-1 [Loss of Off-site Power I
[a2/IAbilityto determine and interpret the following as they apply to Loss of Off-Site Power: J rAA2.021 IESF load sequencer status lights IF, CFCU. For Distracter c it is 25 CFCU. For Distracter d it is the 23 SW pump. The 24 SW pump is selected as the "Lead" pump, and will start on B vital bus. The 23 pump will only start if the 23 pump does not start. (Dwg 203668 contains the tables of loads sequenced on and the SW pump start logic.)
Safeguards Emergency Loading Sequence 11203668 Logic Diagram Monday, October 30,2006 11:07:31 A Page 67 of 79 Prepared by WD Associates, Inc.
- 1A EDG is set up for normal standby operation.
I
/The EDG will mechanically.. .
/start, but will NOT be capable of flashing its field, due to not having a PMG on the shaft.
[NOT start from ANY start signal, since the DUTR must be energized to allow the EDG to start.
start ONLY if the Fire Bypass Switches are placed in BYPASS, which allows local starting of the EDG.
start ONLY from a SEC signal, since the SEC start circuitry is independent of local or control room start circuitry.
D n I 1211112006
/Emergency and Abnormal Plant Evolutions 1058 1 \Lossof DC Power
[=\\Knowledge of the reasons for the following responses as they apply to Loss of DC Power:
lm /Use of dc control power by D/Gs j l A & 2A EDG Unit Trip and Breaker Failure
- Protection 11223680 l--nFl ICE014 1
[No. 1 Unit 1AADC Distribution Cabinet j221408 - C I I I I l F l D I1A & 2A EDG Alarms 1223693 / ( 1 ( ( 1 9 l 1 I f I I
Monday, October 30, 2006 11:07:31 A Page 68 of 79 Prepared by WD Associates, Inc.
Given the following conditions:
- Unit 2 is in MODE 6, with core reload in progress.
- Containment Purge is in service.
- Water level over the Rx vessel flange is > 23'.
- The Spent Fuel Pool Gate Valve is open.
Which of the following identifies a condition which would require IMMEDIATE suspension of irradiated fuel movement in containment IAW Technical Specifications?
[BOTH the inner and outer 100' Airlock doors are opened.
jA valid 2R5 alarm is received in the Spent Fuel Handling Building.
'22 SG secondary side manway is opened, and the entire Main Steam line C/T, vented, and
/drained.
a F--i [Memory I
[Salem It? 2 12111l2006 I
/Emergencyand Abnormal Plant Evolutions 1069 j /Loss of Containment Integrity Distracter a is incorrect because TSAS 3.9.4.b states that a minimum of one door in each airlock must be CAPABLE of being closed. Distracter b is incorrect because a valid area radiation alarm in the FHB requires suspension of fuel movement in the FHB only, not containment. (S2.OP-I0.Z-0010 PAGE 4) C is correct because the secondary side of the SG, when open in containment, will provide a direct path to the loutside if the steam line is vented, drained and C/T, because a drain line in the outer penetration area MUST be open to C/T the steam line. Distracter d is incorrect because IAW TSAS 3.9.4.a the equipment hatch door only needs to be CAPABLE of being closed and secured with 4 bolts, which it can be if it is already closed and secured with 3 bolts with a 4th bolt in containment.
/Technical Specifications 1 )Building TS 3.9.4 Containment Penetrations I tr ---- I 1
Monday, October 30,2006 11:07:31 A Page 69 of 79 Prepared by WD Associates, Inc.
12R31- Letdown Line-Failed Fuel.
12R41D- Plant Vent Release Rate. I 12R17B- ComDonent Coolina Header 22. I i 12111/20061
/R
[Emergency and Abnormal Plant Evolutions I076 I \High Reactor Coolant Activity lAK2. //Knowledge of the interrelations between High Reactor Coolant Activity and the following:
/AK2.011 IProcess radiation monitors 112.6:p j All of the distracters have an automatic function that acts to close the pathway from the process to the atmosphere EXCEPT the R31, which does not. With no automatic isolation of the letdown line, RCS water will end UD in the VCT. and be Dumned from the charaina I)umDs back throuah the reaen HX to the RCS.
- Abnormal Radiation I /S2.OP-AB.RAD-0001 Monday, October 30, 2006 11:07:31A Page 70 of 79 Prepared by W D Associates, Inc.
Cooldown and Depressurization.
Dump steam using 21-24MS10 at a rate to ensure RCS subcooling remains greater than 20 degrees. This will prevent an unwanted transition to FRCC-2, Response to Degraded Core Cooling, which would lead to a RCP start which would raise RCS inventory loss rate.
Operate the Main Steam Dumps in MS PRESSURE CONTROL - MANUAL mode, dump steam at rate not to exceed 100 degree I hr cooldown rate. This will prevent entry into FRTS-1, Response To Imminent Thermal Shock, which would require an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> soak and raise the amount of RCS inventory loss.
7 I
2'
/E03jILOCA Cooldown and Depressurization IEKl. I Knowledge of the operational implications of the following concepts as they apply to LOCA Cooldown and Depressurization:
Monday, October 30, 2006 11:07:31 A Page 71 of 79 Prepared by WD Associates, Inc.
PZR level is dropping 0.1% every 45 seconds.
- 22 RHR sump pump run alarm is locked in.
Which of the following choices describes the proper course of action for these conditions, and why?
IEA2.2 'Adherence to appropriate procedures and operation within the limitations in the facility's license and 14.2/
jamendrnents.
RCS leakage in excess of ONE centrifugal charging pump with minimum letdown, will require a Rx trip and SI. The RHR DumD sumD runs indicate the leak is in RHR.
ILOCA Outside Containment 1 12-EOP-LOCA-6 IReactor Coolant System Leak ; /S2.OP-AB.RC-0001 1 7 1I Monday, October 30,200611:07:31A Page 72 of 79 Prepared by W D Associates, Inc.
When establishing flow to available SGs, which of the following describes the AFW feed strategy to restore SG levels?
Initiate AFW flow.. .
'at maximum rate until WR level is greater than II%, then feed at desired rate to recover levels lint0 the NR.
!at 1.0 - 5.0 E4 Ibm/hr until WR level is greater than 15%, then feed at desired rate to recover
/levels into the NR.
at 1.O - 5.0 E4 Ibm/hr until WR level is greater than II%, then feed at desired rate to recover levels into the NR.
I
[Loss of Heat Sink Functional Recovery i 12-FRHS-1 l~~~~~
'uLum OE009 Monday, October 30, 2006 11:07:31 A Page 73 of 79 Prepared by W D Associates, Inc.
Monday, October 30,2006 11:07:31 A Page 74 of 79 Prepared by WD Associates, Inc.
- Unit 2 was operating at 100% power.
- A small break LOCA occurred.
- The reactor has tripped and SI has been initiated.
- Numerous ECCS components did not startheposition as required.
- FRCC-2, "Response to Degraded Core Cooling", is entered.
You have been directed to place SI Valves in Safeguards position using Table A, Safeguards Valve 12CV68 AND 2CV69, CHARGING LINE I b S J 4 0 AND 22SJ40, HOT LEG i
Monday, October 30, 2006 11:07:31A Page 75 of 79 Prepared by WD Associates, Inc.
Which of the following is the reason why the PZR PORVs are closed regardless of PZR pressure during performance of FRCC-3, Response To Saturated Core Cooling?
- To terminate the unwarranted flow of RCS inventory.
IOPen PORVs are the onlv wav to reach saturation with a constant RCS temperature. 7 IExit from FRCC-3 can only be obtained with the PORVs closed and PZR pressure rising.
Response to Saturated Core Cooling Conditions IEOP-FRCC-3
-rIzIIIl-mm
- -=m OE006 7 ' n - m I I 1- I I I I l i J Monday, October 30,2006 11:07:31 A Page 76 of 79 Prepared by WD Associates, Inc.
During the performance of LOSC-2, Multiple Steam Generator Depressurization, the following plant condition exists:
- Cooldown rate of the RCS is greater than 1OOF/hour.
/How is the control room crew directed to control feedwater flow?
IFeedwater flow is.. .
/maximizedto all S/Gs until narrow range level in any SG is >9%.
/maintainedat least 22E4 Ibm/hr total until any SG narrow range is >9%. J jterminated to all but a single intact S/G, which is fed at no less than IE4 Ibm/hr.
/reducedto no less than 1E4 Ibm/hr to each S/G with narrow ranae level less than 9%.
IApplication I 12/11/2006 Ability to operate and / or monitor the following as they apply to Uncontrolled Depressurization of all Steam Generators:
lfeed flow indication corresponding to 25 gpm)
\MultipleSteam Generator Depressurization I IEOP-LOSC-2 Indian Point NRC Exam 3/10/2003, modified to Salem procedure title and AFW flow units.
Monday, October 30, 2006 11:07:31 A Page 78 of 79 Prepared by WD Associates, Inc.
- Unit 2 has been tripped due to a secondary system malfunction.
- Operators are performing actions in EOP-TRIP-2, Reactor Trip Response.
- The CRS elects to enter FRHS-2, Steam Generator Overpressure, for a YELLOW PATH on the Heat Sink Status Tree.
Which of the following contains ONLY conditions which would allow steam release from the affected SG after entering FRHS-2?
[Affected SG pressure is 1130 psig; NR level is 93%.
[Affected SG pressure is 1140 psig; NR level is 77%.
\Affected SG pressure is 11I O psig; NR level is 100%.
\AffectedSG Dressure is 1090 Psis: NR level is 68%. I m r 12/11/2006'
- El 3 I \SteamGenerator Overpressure I-/ /Adherenceto appropriate procedures and operation within the limitations in the facility's license and 13.0/
,amendments.
I I Monday, October 30, 2006 11:07:31 A Page 79 of 79 Prepared by WD Associates, Inc.
Question Source-RO Ouestion Source Modification Method RO Number Facility Exam Bank Direct From Source 11 Facility Exam Bank Editorially Modified 4 Facility Exam Bank Significantly Modified 4 New 47 Other Facility Concept Used 2 Other Facility Editorially Modified 4 Other Facility Significantly Modified 2 Previous 2 NRC Exams Direct From Source 1
~~~~~ - - -=
Material Requiredfor Examination Administration Exam Level Kil MaterialRequiredforExamination Exam section R 000009KI02 Steam Tables 1 000011K202 S2.OP-TM.ZZ-0002, PAGE 28, RWST TANK CURVE 1 000024K102 SZ.RE-RA.ZZ-O012(Q) FIGURES 1 002000K407 S2.OP-TM.ZZ-0002, Rev. 7, Page 7 of 33 CVCS HUT curve 2 Monday, October 30,2006 Page 1 of 1