ML053330592

GG-08-2005 - Initial Final SRO Written Examination

ML053330592
Person / Time
Site:	Grand Gulf
Issue date:	10/03/2005
From:	Operations Branch IV
To:
References
NRC-082005-3
Download: ML053330592 (207)

Text

Exam Data:Access to this exam is restricted to:lCharles BelllMikeRaschlMickey EllislTommy HarrelsonExam History:lCreated by jbell at Fri Jun 17 08:59:22 CDT 2005lReviewed by mrasch at Mon Jun 20 14:44:19 CDT 2005lApproved by mellis at Mon Jun 20 14:46:49 CDT 2005lQuestions added by mrasch at Mon Jul 18 15:28:20 CDT 2005lQuestion 57 removed by mrasch at Mon Jul 18 15:29:00 CDT 2005lQuestion 100 renumbered to 57 by mrasch at Mon Jul 18 15:29:18 CDT 2005lQuestion 100 renumbered to 58 by mrasch at Mon Jul 18 15:31:57 CDT 2005lQuestion 100 renumbered to 59 by mrasch at Mon Jul 18 15:33:34 CDT 2005lQuestion 100 renumbered to 60 by mrasch at Mon Jul 18 15:35:19 CDT 2005lQuestion 61 renumbered to 62 by mrasch at Mon Jul 18 15:36:28 CDT 2005lQuestion 100 renumbered to 61 by mrasch at Mon Jul 18 15:37:00 CDT 2005lReviewed by mellis at Thu Aug 04 07:58:11 CDT 2005lApproved by mrasch at Thu Aug 04 07:59:21 CDT 2005Comments:Exam Name: NRC August 2005 - 3ExamID: NRC-082005-3Exam Owner: Charles BellExam Date: 08/12/2005Created by: Charles BellCreated at: Fri Jun 17 08:59:22 CDT 2005Last Revised by: Charles BellLast Revised at: Fri Jun 17 08:59:22 CDT 2005Reviewed By: Mickey EllisReview Date: Thu Aug 04 07:58:11 CDT 2005Exam Review Status:ReviewedApproved By: MikeRaschApproval Date: Thu Aug 04 07:59:21 CDT 2005Exam Approval Status:ApprovedEdit ExamMail MessageCompare This ExamPrint Exam DataPrint Exam and KeyGenerate Exam Cover SheetEB QUESTION: 1 (1.0 Points)The reactor is at 95% power and I&C is performing a surveillance on the ReactorRecirc Flow Control System. Suddenly, the A Reactor Recirc Flow Control Valve(FCV) unexpectedly ramps back to its minimum position.Based on this: A.Page1 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-ONEP Objective: 2.0KA

References:

1.295001 AK1.02 Power/flow distribution [3.3/3.5]2.GENERIC 2.4.4 Ability to recognize abnormal indications for system operatingparameters [4.0/4.3]3.GENERIC 2.4.11 Knowledge of abnormal condition procedures [3.4/3.6]

References:

1.05-1-02-III-3sect 3.1; 3.4; 3.5; Figure 1the reactor will scram on Flow Control Trip Reference (FCTR) and ONEP 05-1-02-I-1, Reactor Scram, should be entered. B.reactor power will drop and ONEP 05-1-02-III-3, Reduction in RecirculationSystem Flow Rate should be entered. C.the reactor will scram on high reactor water level and ONEP 05-1-02-I-1, ReactorScram, should be entered. D.reactor power will drop and Reactor Recirc FCV B should immediately be closedto its minimum position.Answer:BQuestionComments:The reactor will not scram on a runback to minimum position. There areno immediate actions for this event, however it is an entry condition toONEP Reduction in Recirculation System Flow Rate. Therefore B is thonly correct answer. This is a NEW question. Tier 1 Group 1 CFR41.5/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00833Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage2 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.B33: Reactor Recirculation System2.C51-6: Period Based Detection SystemCategories:1.Off Normal Event Procedures2.SystemsTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 14:58:38 CDT 2005Question History:1.Created by tharrelso at Wed Apr 20 14:49:03 CDT 20052.Modified by tharrelso at Wed Apr 20 14:58:13 CDT 20053.Modified by tharrelso at Thu Apr 28 08:31:36 CDT 20054.Modified by mrasch at Tue May 24 07:52:01 CDT 20055.Modified by mrasch at Tue May 24 08:22:58 CDT 20056.Modified by mrasch at Tue May 31 14:40:03 CDT 20057.Question Reviewed by mellis at Tue May 31 14:57:01 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200510.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200511.Modified by mrasch at Thu Jul 28 06:34:57 CDT 200512.Modified by tharrelso at Fri Jul 29 06:27:20 CDT 200513.Modified by tharrelso at Fri Jul 29 10:26:57 CDT 200514.Modified by mrasch at Fri Jul 29 14:58:38 CDT 200515.Question Reviewed by mellis at Thu Aug 04 07:23:44 CDT 200516.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 2 (1.0 Points)Which one of the following describes the reasons for transferring Divisions 1 and 2Page3 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-R2100 Objective: 22.CourseID: GLP-OPS-R2100 Objective: 3.1buses from offsite power sources to the diesel generators during degraded grid voltagesituations? A.Safety related loads are only allowed to be supplied power from Emerency DieselGenerators during accident conditions. B.To preclude damage to safety related electrical equipment as a result of extendedoperation at reduced voltage levels. C.To prevent core damage by ensuring safety systems are connected to a reliablepower source and minimize response time during an accident. D.To ensure the safety related equipment is supplied with a power source ofsufficient capacity to ensure the plant can be maintained in a safe shutdowncondition during an accident.Answer:BQuestionComments:Answer A is INCORRECT because Offsite power sources are thepreferred power sources for the ESF buses. Answer B is CORRECTbecause voltage degradation in conjunction with operating equipmentcan cause damage to safety related equipment normally operated onthese buses. Answer C is incorrect because the transient timesassociated with transferring power to alternate feeds (Diesel Generators)is consistent with the accident analyses. Answer D is incorrect becausethe offsite power sources have sufficient capacity to handle all loadsrequired for safe shutdown during accident conditions and do not havethe lower limitations of the diesel generators. TIER 1 GROUP 1 This is aNEW question. CFR 41.7/41.8Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00464Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page4 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.CourseID: GLP-OPS-R2100 Objective: 1KA

References:

1.295003 AK3.01 Manual and auto bus transfer [3.3/3.5]

References:

1.Tech Spec Bases 3.3.8.1-12.UFSAR 8.3.1.1.2.2TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.R21: 4.16 KV AC Power SystemCategories:1.SystemsTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 10:45:04 CDT 2005Question History:1.Used on December 2000 NRC Exam2.Used on August 2002 Audit RO Exam3.Converted from MSWord on Tue May 25 14:16:47 CDT 20044.Imported at Tue May 25 14:24:02 CDT 20045.Modified by tharrelso at Wed Apr 20 15:54:30 CDT 20056.Modified by mrasch at Tue May 10 13:42:42 CDT 20057.Modified by mrasch at Tue May 24 07:55:38 CDT 20058.Question Reviewed by mellis at Tue May 31 14:56:59 CDT 20059.Modified by jbell at Thu Jun 16 16:50:13 CDT 200510.Modified by mrasch at Mon Jun 20 07:34:08 CDT 200511.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 200512.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200513.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200514.Modified by mrasch at Fri Jul 01 11:04:00 CDT 200515.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 2005Page5 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 16.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200517.Modified by mrasch at Tue Jul 26 12:14:06 CDT 200518.Modified by mrasch at Wed Jul 27 07:48:01 CDT 200519.Modified by mrasch at Wed Jul 27 12:27:53 CDT 200520.Question Reviewed by mellis at Thu Aug 04 07:23:42 CDT 200521.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200522.Modified by mrasch at Tue Aug 09 10:45:04 CDT 200523.Question Reviewed by mellis at Tue Aug 09 11:59:48 CDT 2005Comments:EB QUESTION: 3 (1.0 Points)With RHR A in standby, how does a complete loss of Division I 125 VDC affect placingRHR A in Shutdown Cooling (SDC) mode of operation? A.All necessary component manipulations can be performed at the RemoteShutdown Panel. B.The RHR A pump can be started from the Control Room. The SDC Return ValveE12-F053A will require local operation. C.All necessary component manipulations can be performed in the Main ControlRoom D.The RHR A pump can be started from the breaker cubicle. The SDC Return ValveE12-F053A can be manipulated from the Control Room.Answer:DQuestionComments:RHR A Pump to be operated anywhere other than the circuit breakerrequires DC Control power. The E12-F053A control power is 120 VACand is available for control room operation. Based on these conditionsAnswer D is the only correct answer. TIER 1 GROUP 1 This is a NEWquestion. CFR 41.7/41.8Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamPage6 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-P4100 Objective: 12.62.CourseID: GLP-OPS-L1100 Objective: 8aKA

References:

1.295004 AA1.02 Systems necessary to assure safe plant shutdown [3.8/4.1]2.295004 AA1.03 AC electrical distribution [3.4/3.6]

References:

1.E-1181TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E12: Residual Heat Removal System2.L11: Plant DC Electical SystemCategories:1.SystemsTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 15:00:26 CDT 2005Question History:1.Created by tharrelso at Thu Apr 21 09:43:45 CDT 20052.Modified by mrasch at Tue May 10 13:54:14 CDT 20053.Modified by mrasch at Tue May 24 10:10:51 CDT 20054.Question Reviewed by mellis at Tue May 31 14:57:01 CDT 20055.Modified by mrasch at Mon Jun 20 07:29:14 CDT 20056.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005QuestionID:GGNS-NRC-00834Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page7 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 9.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200510.Modified by mrasch at Wed Jul 27 16:59:35 CDT 200511.Modified by mrasch at Thu Jul 28 06:54:35 CDT 200512.Modified by tharrelso at Fri Jul 29 06:28:31 CDT 200513.Modified by mrasch at Fri Jul 29 15:00:26 CDT 200514.Question Reviewed by mellis at Thu Aug 04 07:23:44 CDT 200515.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 4 (1.0 Points)The plant is operating at rated conditions when a turbine trip occurs.Immediately following the transient, the reactor water level shrink associated with thereactor pressure change will: A.cause reactor water level to initially drop, then recover to a value lower than theinitial level. B.not be seen because of the sharp drop in steam flow. C.not be seen because of setpoint setdown. D.cause reactor water level to rapidly drop and then recover to it's initial level.Answer:AQuestionComments:Answer A is CORRECT because reactor water level will initially drop asthe rise in reactor pressure collapses voids. When the bypass valvesopen to combat the reactor pressure rise, voids reform and reactor waterlevel recovers. However, because of the setpoint setdown feature,reactor water level stabilizes at a new value lower than is normal for anoperating reactor. Answer B is INCORRECT because the sharp drop insteam flow following a reactor scram reduces the void content in thereactor, resulting in a drop in reactor water level. Answer C isINCORRECT because the setpoint setdown feature, by design, willPage8 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP02 Objective: 7aKA

References:

1.295005 AK1.03 Pressure effects on reactor level [3.5/3.7]

References:

1.GGNS PSTG, Appendix B Pages B-6-27 and 282.02-S-01-27 Step 6.6.8.k3.Simulator and plant operating dataTrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Shift Technical Advisor Training ProgramSystems:1.B21: Nuclear Boiler System2.N19: Condensate System3.N21: Feedwater System4.N32: EHC Control SystemCategories:1.Fundamentals2.Off Normal Event Procedures3.Operating Experience4.Systems5.STA Trainingchange reactor water level following a scram. Answer D is INCORRECTbecause reactor water level will initially drop, but setpoint setdown isdesigned to recover reactor water level to a value lower than the levelnormal for an operating reactor. This is a NEW question. TIER 1 GROUP1 CFR 41.3/41.5/41.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00835Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage9 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Task

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 10:28:23 CDT 2005Question History:1.Created by tharrelso at Thu Apr 21 12:17:26 CDT 20052.Modified by tharrelso at Thu Apr 21 13:52:50 CDT 20053.Modified by mrasch at Tue May 10 13:58:26 CDT 20054.Question Reviewed by mellis at Tue May 31 14:57:02 CDT 20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Modified by tharrelso at Thu Jul 28 12:59:40 CDT 20059.Modified by tharrelso at Fri Jul 29 06:29:46 CDT 200510.Modified by tharrelso at Fri Jul 29 10:28:23 CDT 200511.Question Reviewed by mrasch at Thu Aug 04 07:21:41 CDT 200512.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 5 (1.0 Points)In which one of the following Scram Reports can it be determined that ShutdownMargin is assured without Reactor Engineering assistance? A.The Mode Switch is in Shutdown, reactor power is dropping, all control rodsindicate full in except four control rods indicate position 08. B.The Mode Switch is in Shutdown, reactor power is dropping, all control rodsindicate full in except forty-nine control rods indicate position 02. C. The Mode Switch is in Shutdown, APRMs indicate downscale, NO control rodposition indication is available. D.Page10 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP02 Objective: 11.0KA

References:

1.295006 AK1.02 Shutdown margin [3.4/3.7]

References:

1.GE SIL No. 529 dated February 19, 19912.GE SIL No. 529 Supplement 1dated March 14, 19973.Technical Specification Bases 3.1.1TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.B13: Reactor Pressure Vessel2.C11-2: Rod Control and Information System3.C51-5: Average Power Range Nuclear Instrumentation SystemThe Mode Switch is in Shutdown, reactor power is dropping, all control rodsindicate full in except two peripheral control rods indicates position 04.Answer:BQuestionComments:Answer A is incorrect because all control rods must be inserted to at leastposition 02 to ensure adequate shutdown margin. Answer B isCORRECT because all control rods are inserted to at least position 02and by analysis, this ensures adequate shutdown margin. Answer C isincorrect because even though the APRMs indicate downscale, nocontrol rod position indication is available. Answer D is incorrect becauseall control rods must be inserted to at least position 02 to ensureadequate shutdown margin. No distinction is made for lower worthperipheral control rods. This is a NEW question. TIER 1 GROUP 1 CFR41.1/41.2/41.5/41.6/41.10/43.1/43.2/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00836Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page11 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Categories:1.Emergency Procedure Training2.FSAR3.Mitigation of Core Damage4.Systems5.Technical SpecificationsTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 15:04:04 CDT 2005Question History:1.Created by tharrelso at Thu Apr 21 16:40:40 CDT 20052.Modified by mrasch at Tue May 10 14:00:12 CDT 20053.Modified by mrasch at Tue May 24 10:14:09 CDT 20054.Question Reviewed by mellis at Tue May 31 14:57:02 CDT 20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Modified by tharrelso at Fri Jul 29 06:31:03 CDT 20059.Modified by tharrelso at Fri Jul 29 10:34:48 CDT 200510.Modified by mrasch at Fri Jul 29 15:04:04 CDT 200511.Question Reviewed by mellis at Thu Aug 04 07:23:45 CDT 200512.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 6 (1.0 Points)A small fire with heavy smoke has forced the Operations crew to abandon the ControlRoom before the reactor could be shutdown.Which of the following will scram the reactor from outside the Control Room? A.Open breakers CB3A and CB7B. B.Open breakers CB2A and CB8A.Page12 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C7100 Objective: 5.42.CourseID: GLP-OPS-C7100 Objective: 6.43.CourseID: GLP-OPS-C7100 Objective: 5.3KA

References:

1.295016 AA1.01 RPS [3.8/3.9]

References:

1.05-1-02-II-1 Step 3.32.E-1173-014TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training Program C.Open breakers CB2B and CB8B. D.Open breakers CB2A and CB2B.Answer:DQuestionComments:Answer A is incorrect because this will have no effect on the RPS logic.Answer B is incorrect because opening only one division of breakers willresult in only a half scram. Answer C is incorrect because opening onlyone division of breakers will result in only a half scram. Answer D isCORRECT because at least 1 RPS breaker in division (1 and 2) isrequired for full scram. This is a NEW question. TIER 1 GROUP 1 CFR41.6/41.7/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00837Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage13 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Systems:1.C61: Remote Shutdown System2.C71: Reactor Protection SystemCategories:1.Off Normal Event Procedures2.SystemsTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 10:39:06 CDT 2005Question History:1.Created by tharrelso at Fri Apr 22 09:41:15 CDT 20052.Modified by mrasch at Tue May 10 14:01:49 CDT 20053.Modified by mrasch at Tue May 24 10:30:20 CDT 20054.Question Reviewed by mellis at Tue May 31 14:57:03 CDT 20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Modified by mrasch at Fri Jul 01 12:10:03 CDT 20058.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20059.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200510.Modified by tharrelso at Fri Jul 29 06:38:02 CDT 200511.Modified by tharrelso at Fri Jul 29 10:39:06 CDT 200512.Question Reviewed by mrasch at Thu Aug 04 07:21:06 CDT 200513.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 7 (1.0 Points)The plant is operating at rated conditions when a rupture on the discharge header ofthe Component Cooling Water (CCW) System results in a complete loss of CCW.Which of the following would be affected by the loss of CCW? A.Control Rod Drive Hydraulic pump oil temperatures. B.Page14 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-ONEP Objective: 2.02.CourseID: GLP-OPS-P4200 Objective: 10KA

References:

1.295018 AK1.01 Effects on component/system operations [3.5/3.6]

References:

1.05-1-02-V-1 Section 4.0TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training ProgramSystems:1.B33: Reactor Recirculation System2.C11-1A: CRD Hydraulic SystemReactor Water Clean Up containment penetration temperatures. C.Ambient Drywell temperatures. D.RHR A pump seal temperatures.Answer:AQuestionComments:The only component cooled by CCW listed in the answers is CRD pumplube oil. Drywell cooling is from Drywell Chilled Water. The RWCUpenetration are cooled by Plant Chilled Water. RHR A pump seal iscooled by SSW A.Therefore A is the only correct answer. This is a NEWquestion. TIER 1 GROUP 1 CFR 41.4/41.5/41.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00838Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage15 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.G33: Reactor Water Cleanup4.P42: Component Cooling Water SystemCategories:1.Off Normal Event Procedures2.SystemsTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 10:47:33 CDT 2005Question History:1.Created by tharrelso at Fri Apr 22 11:04:55 CDT 20052.Modified by tharrelso at Fri Apr 22 13:11:21 CDT 20053.Modified by mrasch at Tue May 24 08:02:55 CDT 20054.Question Reviewed by mellis at Tue May 31 14:57:03 CDT 20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Modified by mrasch at Fri Jul 01 15:38:53 CDT 20058.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20059.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200510.Modified by mrasch at Thu Jul 28 07:29:23 CDT 200511.Modified by tharrelso at Fri Jul 29 09:52:03 CDT 200512.Modified by tharrelso at Fri Jul 29 09:53:32 CDT 200513.Modified by tharrelso at Fri Jul 29 09:59:16 CDT 200514.Modified by tharrelso at Fri Jul 29 10:01:13 CDT 200515.Question Reviewed by mrasch at Thu Aug 04 07:21:47 CDT 200516.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200517.Modified by mrasch at Tue Aug 09 10:47:33 CDT 200518.Question Reviewed by mellis at Tue Aug 09 11:59:49 CDT 2005Comments:EB QUESTION: 8 (1.0 Points)Unit I Instrument Air Compressor is tagged out for maintenance and the UnitII Instrtument Air Compressor just failed.With NO operator action the minimum Instrument Air pressure the operator wouldexpect to see is approximately: A.80-85 psigPage16 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-P5300 Objective: 4KA

References:

1.295019 AA2.01 Instrument air system pressure [3.5/3.6]2.295019 AA1.01 Backup air supply [3.5/3.3]

References:

1.M-1068D2.05-1-02-V-9 section 3.2.3TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training Program B.90-95 psig C.100-105 psig D.110-115 psigAnswer:BQuestionComments:As instrument air pressure lowers, it will eventually reach the setpointfor auto opening of the service air to instrument air cross tie valve. Oncethis valve opens service air wil maintain instrument air pressure atapproximately 90 psig per the loss of Instrument Air ONEP. Tier 1Group 1 This is a NEW question. CFR 41.4/41.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00548Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage17 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Systems:1.P52: Service Air System2.P53: Instrument Air SystemCategories:1.SystemsTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 10:41:54 CDT 2005Question History:1.No exam history found for this question during conversion. Converted from MSWordon Tue May 25 14:16:50 CDT 20042.Imported at Tue May 25 14:24:22 CDT 20043.Modified by tharrelso at Thu Apr 28 08:20:24 CDT 20054.Modified by mrasch at Tue May 24 11:11:40 CDT 20055.Question Reviewed by mellis at Tue May 31 14:57:00 CDT 20056.Modified by mrasch at Thu Jun 09 16:38:33 CDT 20057.Modified by jbell at Thu Jun 16 16:51:16 CDT 20058.Modified by mrasch at Mon Jun 20 07:45:31 CDT 20059.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 200510.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200512.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200513.Modified by mrasch at Thu Jul 28 07:51:00 CDT 200514.Modified by mrasch at Thu Jul 28 08:28:06 CDT 200515.Modified by tharrelso at Fri Jul 29 10:41:54 CDT 200516.Question Reviewed by mrasch at Thu Aug 04 07:21:22 CDT 200517.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 9 (1.0 Points)The plant is in Mode 4 and all forced circulation has been lost.Which of the following is the reason for raising reactor water level to +82 inches per 05-1-02-III-1, Inadequate Decay Heat Removal?Page18 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-ONEP Objective: 2.02.CourseID: GLP-OPS-ONEP Objective: 20.03.CourseID: GLP-OPS-B1300 Objective: 5.124.CourseID: GLP-OPS-B3300 Objective: 42.0KA

References:

A.This is the indicated level required to establish flow through open Safety ReliefValves to the Suppression Pool. B.This is the indicated level required to establish alternate cooling using the ReactorWater Cleanup system. C.This is the indicated level used in the FSAR accident analysis for the "Loss ofForced Circulation" Time to Boil Curves. D.This is the indicated level required to allow natural circulation through the core andfeedwater annulus.Answer:DQuestionComments:Answer A is INCORRECT because the indicated level required toestablish flow through open safety relief valves to the suppression pool isbetween +101 to 129 inches. Answer B is INCORRECT because theReactor Water Cleanup System can be used as an alternate coolingmethod at normal water level. Answer C is INCORRECT because eachTime to Boil Curve specifies the initial conditions of time after shutdownand RPV level for thier valid analysis. Answer D is CORRECT becausethis is the level at which sufficient driving head exists to establish naturalcirculation through the core. Tier 1 Group 1 MODIFIED idWRI 515 NRCExam June 2001 Question 33 10CFR 41.5/41.10/41.14/42.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00078aReview Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage19 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.295021 AK3.01 Raising reactor water level [3.3/3.4]

References:

1.05-1-02-III-1, Inadequate Decay Heat Removal Step 3.3.2TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training ProgramSystems:1.B13: Reactor Pressure Vessel2.B21: Nuclear Boiler System3.E12: Residual Heat Removal SystemCategories:1.Off Normal Event Procedures2.SystemsTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 10:43:27 CDT 2005Question History:1.Created by tharrelso at Mon Apr 25 10:18:40 CDT 20052.Created by tharrelso at Mon Apr 25 10:18:40 CDT 2005 from parent QuestionIDGGNS-NRC-000783.Modified by mrasch at Tue May 24 08:31:58 CDT 20054.Question Reviewed by mellis at Tue May 31 14:56:57 CDT 20055.Modified by mrasch at Thu Jun 09 16:16:34 CDT 20056.Modified by mrasch at Thu Jun 09 16:37:24 CDT 20057.Modified by mrasch at Thu Jun 09 16:39:33 CDT 20058.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20059.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200511.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200512.Modified by tharrelso at Fri Jul 29 10:43:27 CDT 200513.Question Reviewed by mrasch at Thu Aug 04 07:20:36 CDT 200514.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:Page20 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm EB QUESTION: 10 (1.0 Points)In Mode 5 during a Shutdown Margin Demonstration (SMD), which one of thefollowing would be the immediate concern in the event of inadvertent criticality? A.Unplanned mode change B.Fuel damage C.Inadequate decay heat removal D.High in-plant doseAnswer:DQuestionComments:Answer D is CORRECT because of the near proximity of workers in thedrywell and on 208' of containment relative to the reactor core during arefueling outage. Localized criticality would raise dose rates in the drywelland possibly within line of sight of the core on elev. 208' of containment.Procedural guidance clearly prioritizes dose rate monitoring for aninadvertent criticality event. Answer A is INCORRECT since in Mode 5, amode change is not made based on the effects of criticality, rising flux orcoolant temperature, but only by Reactor Mode Switch position. Answer Bis INCORRECT because any postulated criticality would be localized, notglobal, since control rod density would be ~99% for the SMD. Also, IRMswould generate a rod block and/or scram to limit the power excursion.Manual control rod insertion or system interlocks would mitigate the localpower rise. Answer C is INCORRECT due to the lengthy amount of time itwould take for the large volume of reactor coolant during Mode 5 to reacha temperature of significant concern. Other methods of decay heatremoval could be placed in service during that time. Moreover, the controlrod(s) withdrawn for the SDM would be quickly inserted to terminate theevent. Tier 1 Group 1 This is a NEW Question. 10CFR41.1/41.12/43.4/43.6Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00848Page21 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 8.30KA

References:

1.295023 AK1.03 Inadvertent criticality [3.7/4.0]

References:

1.01-S-06-2 step 6.7.122.FSAR 15.4.1.1.4TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.J11: Reactor FuelCategories:1.Off Normal Event Procedures2.Refueling Training3.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 10:50:27 CDT 2005Question History:1.Created by mrasch at Thu Jun 09 16:52:06 CDT 20052.Modified by mrasch at Mon Jun 20 07:47:42 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by tharrelso at Fri Jul 29 10:44:56 CDT 20058.Modified by mrasch at Fri Jul 29 15:07:01 CDT 2005Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page22 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 9.Question Reviewed by mellis at Thu Aug 04 07:23:47 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200511.Modified by mrasch at Tue Aug 09 10:50:27 CDT 200512.Question Reviewed by mellis at Tue Aug 09 11:59:51 CDT 2005Comments:EB QUESTION: 11 (1.0 Points)The plant is operating at 100% power when a spurious Division 1 ECCS initiationoccurs.Drywell pressure is rising due to Drywell Purge Compressor operation.What must be reset first to perform a normal shutdown of Drywell Purge CompressorA? A.Division 1 Load Shedding and Sequencing panel B.Division 1 Combustible Gas Control System logic C. Division 1 Containment/Drywell isolation logic D.Division 1 ECCS logicAnswer:DQuestionComments:Answer A is incorrect because the LSS LOCA initiation signal originatesfrom ECCS Logic. Answer B is incorrect because CGCS initiation logicLOCA signal originates from the ECCS logic. Answer C is incorrectbecause the CGCS and ECCS logics are separate from Containmentisolation logic. Answer D is CORRECT because it is the top tier logic.Tier 1 Group 1 This is a NEW question. 10CFR 41.5/41.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00849Review Status:ReviewedPage23 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E6100 Objective: 6.8, 19.0KA

References:

1.295024 Generic 2.1.23: 3.9/4.0

References:

1.17-S-06-5 Att. I pgs 3, 4; Att. II pgs 16,19,21,27,37,382.04-1-01-E12-1 Att. IX3.M-1077BTrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E12: Residual Heat Removal System2.E61: Combustible Gas Control System3.M71: Containment and Drywell Instrumentation SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 15:08:57 CDT 2005Question History:1.Created by mrasch at Thu Jun 09 17:02:20 CDT 20052.Modified by mrasch at Mon Jun 20 13:33:34 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Wed Jul 27 15:19:10 CDT 20058.Modified by tharrelso at Fri Jul 29 10:45:38 CDT 2005Difficulty:2: Comprehension or AnalysisDifficulty Rating:2Page24 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 9.Modified by tharrelso at Fri Jul 29 10:46:10 CDT 200510.Modified by tharrelso at Fri Jul 29 10:52:43 CDT 200511.Modified by mrasch at Fri Jul 29 15:08:57 CDT 200512.Question Reviewed by mellis at Thu Aug 04 07:23:47 CDT 200513.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 12 (1.0 Points)Which one of the following describes the reason for automatic actuation of AlternateRod Insertion (ARI) in response to rising reactor pressure? A.ARI acts independently of RPS to insert control rods during conditions that shouldhave resulted in a reactor scram. B.ARI utilizes the RPS trip units to provide an alternate vent path for the scram airheader as a backup to RPS. C.ARI anticipates a reactor scram and assists RPS by establishing alternate ventpaths for the scram air header. D.ARI acts in conjunction with the Recirculation pumps to prevent exceeding theRPV pressure safety limit.Answer:AQuestionComments:Answer A is CORRECT because ARI is an independent system to RPSwhich utilizes separate B21 trip units to open three independent ventpaths in the event reactor pressure exceeds the 1064.7 psig scramsetpoint. The ARI setpoint is at 1126 psig which is sufficiently above thescram setpoint to minimize inadvertent trips. Answer B is incorrectbecause ARI uses separate B21 trip units from those used by RPS.Answer C is incorrect because ARI actuation setpoint is higher than theRPS actuation. Answer D is incorrect because the Safety Relief Valveswill prevent challenging the RPV Pressure Safety Limit. TIER 1 GROUP1 This is a NEW question. CFR 41.5/41.6Image

Reference:

NoneClosed Reference QuestionPage25 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C111A Objective: 3.13KA

References:

1.295025 EK3.06 Alternate rod insertion: Plant-Specific [4.2/4.4]

References:

1.UFSAR 4.6.1.1.1.1.1(d)2.UFSAR 4.6.1.1.2.4.5TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.C11-1A: CRD Hydraulic SystemCategories:1.SystemsTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 10:55:25 CDT 2005Question History:1.Used on NRC 2004 Exam2.Converted from MSWord on Wed May 26 18:04:19 CDT 20043.Imported at Wed May 26 18:04:47 CDT 20044.Modified by tharrelso at Wed Apr 27 13:41:14 CDT 20055.Modified by mrasch at Tue May 10 13:47:54 CDT 20056.Question Reviewed by mellis at Tue May 31 14:57:00 CDT 20057.Modified by mrasch at Mon Jun 20 06:38:30 CDT 20058.Modified by mrasch at Mon Jun 20 10:58:22 CDT 20059.Modified by mrasch at Mon Jun 20 12:49:23 CDT 200510.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 200511.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamHandout Not Required with ExamQuestionID:GGNS-NRC-00690Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage26 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Date: 08/12/200512.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200513.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200514.Modified by mrasch at Wed Jul 27 11:30:12 CDT 200515.Modified by mrasch at Wed Jul 27 14:15:49 CDT 200516.Modified by tharrelso at Fri Jul 29 10:55:25 CDT 200517.Question Reviewed by mrasch at Thu Aug 04 07:21:28 CDT 200518.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 13 (1.0 Points)Which one of the following is the basis to emergency depressurize the RPV prior toexceeding the Heat Capacity Temperature Limit (HCTL) of the Suppression Pool? A.The emergency depressurization adds negative reactivity to the core and therebyreduces the total energy addition into the containment. B.Prevent introduction of energy into the suppression pool at a rate that exceeds theenergy removal rate of plant systems. C.Emergency depressurizing the RPV prior to exceeding HCTL ensures the ECCSpumps will maintain NPSH if required to operate. D.Prevent exceeding containment pressure and temperature limits in the eventemergency depressurization is required.Answer:DQuestionComments:HCTL requires emergency depressurization of the RPV in the eventconditions are in the UNSAFE region to protect equipment in thepressure suppression chamber (Containment) from the maximumtemperature and pressure capability per the PSTGs. Therefore AnswerD is the only correct answer. Answer B is incorrect because it says whenthe reactor is still pressurized. Tier 1 Group 1 This is a NEW question.10CFR55. 41.9Image

Reference:

NonePage27 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP02A Objective: 22.CourseID: GG-1-LP-RO-EP02A Objective: 5KA

References:

1.295026 EK3.01 Emergency/normal depressurization [3.8/4.1]

References:

1.GGNS PSTG App. B 17.5TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.M41-1: ContainmentCategories:1.Emergency Procedure Training2.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 12:17:42 CDT 2005Question History:1.Created by tharrelso at Mon May 02 07:18:33 CDT 20052.Modified by mrasch at Tue May 10 14:19:27 CDT 20053.Question Reviewed by mellis at Tue May 31 14:57:04 CDT 20054.Modified by jbell at Fri Jun 17 14:31:55 CDT 20055.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00840Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:3Page28 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Date: 08/12/20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Modified by mrasch at Thu Jul 28 09:50:15 CDT 200510.Modified by tharrelso at Fri Jul 29 06:53:01 CDT 200511.Modified by tharrelso at Fri Jul 29 07:46:38 CDT 200512.Modified by mellis at Fri Jul 29 12:13:17 CDT 200513.Modified by mellis at Fri Jul 29 12:16:15 CDT 200514.Modified by mellis at Fri Jul 29 12:17:42 CDT 200515.Question Reviewed by mrasch at Thu Aug 04 07:21:53 CDT 200516.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 14 (1.0 Points)ATWS conditions exist.Average Containment temperature is 187°F and slowly rising.Suppression Pool temperature is 147°F and rising.Suppression Pool level is 20 feet and rising.Which one of the following describes the action that should be taken tocontrol Containment parameters? A.Vent the Containment B.Emergency depressurize the reactor C.Operate all containment coolers utilizing EP Attachment 7 D.Initiate Containment SprayAnswer:BQuestionComments:Answer A is incorrect since the vent path for EP attachment 14 is isolatedand cannot be opened under the listed conditions.Answer B is correctsince EP-3 step 28 requires emergency depressurization if containmentPage29 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP03 Objective: 3KA

References:

1.295027 EA1.03 Emergency depressurization: Mark-III [3.5/3.8]

References:

1.PSTG B 7-192.05-S-01-EP-3 Step 29TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E12: Residual Heat Removal System2.M41-1: ContainmentCategories:1.Emergency Procedure TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 12:18:13 CDT 2005Question History:temperature cannot be restored below 185°F. Containment temperaturecannot be restored below 185°F since P71 is isolated to the containmentunder these conditions, so normal containment cooling is unavailable.Answer C is incorrect since operation of containment coolers without P71as a heat sink would be ineffective. Answer D is incorrect since EP-3Step 29 requires emergency depressurization and any other action wouldbe inappropriate. Tier 1 Group 1 This is a NEW question. 10 CFR41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00844Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page30 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Created by mrasch at Mon May 16 15:27:06 CDT 20052.Question Reviewed by mellis at Tue May 31 14:57:04 CDT 20053.Modified by mrasch at Fri Jun 10 07:59:51 CDT 20054.Modified by mrasch at Mon Jun 20 13:35:23 CDT 20055.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Modified by mrasch at Mon Jul 11 08:11:54 CDT 20059.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 200510.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200511.Modified by tharrelso at Wed Jul 27 13:39:49 CDT 200512.Modified by tharrelso at Wed Jul 27 13:44:08 CDT 200513.Modified by tharrelso at Fri Jul 29 07:08:25 CDT 200514.Modified by mellis at Fri Jul 29 12:18:13 CDT 200515.Question Reviewed by mrasch at Thu Aug 04 07:21:57 CDT 200516.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 15 (1.0 Points)A LOCA occurred, the following conditions exist:Drywell temperature at the 166 foot elevation is 200 degrees F.Reactor Recirc Pumps are off.Reactor pressure is 600 psig.Containment temperature at the 139 foot elevation is 95 degrees F.If the above conditions remain stable, the Wide Range Reactor Water Levelinstrument:CAUTION 1 of EP-2 is provided. A.is limited to use below -130 inches indicated level but will indicate higher thanactual water level. B.can be used as long as it is on scale but will indicate higher than actual waterPage31 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP02 Objective: 32.CourseID: GLP-OPS-PROC Objective: 59.3KA

References:

1.295028 EK1.02 Equipment environmental qualification [2.9/3.1]

References:

1.05-S-01-EP-2 Caution 12.02-S-01-27 section 6.1.8TrainingPrograms:1.Reactor Operator Training Programlevel. C.is limited to use above -130 inches indicated level but will indicate higher thanactual water level. D.can be used as long as it is on scale but will indicate lower than actual water level.Answer:BQuestionComments:Answer A is incorrect because since the RPV Saturation Curve is stillsafe, the instrument is allowed full use as long as the instrument tracksper the Operations Philosphy. Answer B is CORRECT because theelevated temperature of the reference leg will cause the instrument toread higher than actual and Operations Philosophy allows its use.Answer C is incorrect because Operations Philosophy allows full use ofthe instrument as long as it tracks even with the elevated reference legtemperatures. Answer D is incorrect because indicated level will behigher than acutal level. Tier 1 Group 1 This is a NEW question. 10CFR41.7/41.10/43.5Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00845Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:4Page32 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.B21: Nuclear Boiler SystemCategories:1.Emergency Procedure TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 12:18:58 CDT 2005Question History:1.Created by mrasch at Mon May 16 15:36:22 CDT 20052.Modified by mrasch at Mon May 16 15:54:53 CDT 20053.Question Reviewed by mellis at Tue May 31 14:57:05 CDT 20054.Modified by mrasch at Fri Jun 10 08:04:08 CDT 20055.Modified by mrasch at Mon Jun 13 13:12:39 CDT 20056.Modified by mrasch at Thu Jun 16 12:51:16 CDT 20057.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200510.Modified by mrasch at Mon Jul 11 08:46:21 CDT 200511.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 200512.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200513.Modified by mrasch at Wed Jul 27 13:21:06 CDT 200514.Modified by mrasch at Wed Jul 27 13:30:51 CDT 200515.Modified by mrasch at Wed Jul 27 14:16:09 CDT 200516.Modified by mellis at Fri Jul 29 12:18:58 CDT 200517.Question Reviewed by mrasch at Thu Aug 04 07:22:03 CDT 200518.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 16 (1.0 Points)The reactor is at rated conditions.An RO is performing the daily Technical Specification surveillance rounds.Page33 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-M4101 Objective: 11KA

References:

1.2950302.GENERIC 2.2.12 Knowledge of surveillance procedures [3.0/3.4]

References:

1.TS3.3.3.12.TS3.6.2.1A Suppression Pool Limiting Condition for Operation (LCO) would require entry per thetechnical specifications if: A.One narrow range Suppression Pool level recorder pen on H13-P870 is stuck. B.Suppression Pool Average Temperature is 93 degrees F. C.Suppression Pool Water Level is 18 feet 2 inches. D.Suppression Pool Water Level is 18 feet 8 inches.Answer:CQuestionComments:TS 3.6.2.1 Suppression Pool Average Temperature limit is 95 degrees F.TS 3.6.2.2 Suppression Pool Level must be between 18 feet 4 1/2inches and 18 feet 9 3/4 inches. TR3.6.2.2 for Suppression Pool LevelAlarms and TS 3.3.3.1 Post Accident monitoring are InstrumentationSpecifications not affecting the operability of the Suppression Pool itself.Therefore Answer C is the only correct answer. Tier 1 Group 1 This is aNEW question. 10 CFR 41.9/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00846Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:3Page34 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.TS3.6.2.24.TR3.6.2.2TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.M41-1: ContainmentCategories:1.Administrative Requirements2.Systems3.Technical SpecificationsTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 10:51:21 CDT 2005Question History:1.Created by mrasch at Mon May 16 15:53:01 CDT 20052.Question Reviewed by mellis at Tue May 31 14:57:05 CDT 20053.Modified by mrasch at Fri Jun 10 08:09:11 CDT 20054.Modified by mrasch at Mon Jun 20 13:37:44 CDT 20055.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Modified by mrasch at Tue Jul 12 15:41:48 CDT 20059.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 200510.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200511.Modified by mrasch at Thu Jul 28 08:26:37 CDT 200512.Modified by tharrelso at Fri Jul 29 07:14:44 CDT 200513.Modified by tharrelso at Fri Jul 29 07:15:16 CDT 200514.Modified by mellis at Fri Jul 29 12:19:43 CDT 200515.Question Reviewed by mrasch at Thu Aug 04 07:22:07 CDT 200516.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200517.Modified by mrasch at Tue Aug 09 10:51:21 CDT 200518.Question Reviewed by mellis at Tue Aug 09 11:59:49 CDT 2005Comments:Page35 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm EB QUESTION: 17 (1.0 Points)Following a reactor scram on high Drywell pressure, RPV water level initially droppedto - 20 inches on wide range.HPCS automatically initiated.As level was being restored, the HS for the HPCS Pump was stopped and the HPCSInjection Valve E22-F004 was taken to close.RPV level is now +60 inches Narrow Range and Drywell pressure is 1.5 psig.In this situation, if the HPCS INIT RESET pushbutton is depressed, the: A.HPCS pump will auto start immediately but the HPCS Injection Valve E22-F004will have to be opened using the handswitch. B.HPCS pump will auto start immediately and the HPCS Injection Valve E22-F004will automatically open. C.HPCS pump will auto start after reactor water level drops below Level 2. D.HPCS pump will auto start after reactor water level drops below Level 8.Answer:CQuestionComments:Answer A is incorrect because depressing HPCS INIT RESET wouldreset the initiation signal and override high drywell pressure. HPCS pumpwould not auto start until -41.6 in. level was reached. Answer B isincorrect because the HPCS HI LVL RESET push button does notbypass high level but only resets the seal in when level falls below 53.5in.. Answer C is CORRECT because the High Drywell pressure signal isbypassed such that HPCS is reset to a standby condition and will autoinitiate on level 2. Answer D is incorrect because depressing HPCS INITRESET would reset the initiation signal and override high drywellpressure the system will remain in standby until level 2 is reached. Tier 1Group 1 This is a NEW question. 10CFR 41.7/41.8Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamPage36 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E2200 Objective: 9.32.CourseID: GLP-OPS-E2200 Objective: 9.43.CourseID: GLP-OPS-E2200 Objective: 204.CourseID: GLP-OPS-E2200 Objective: 21KA

References:

1.295031 EA1.04 High pressure core spray: Plant-Specific [4.3/4.2]

References:

1.E-1183-032.E-1183-233.E-1188-19TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E22-1: High Pressure Core Spray SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 16:10:27 CDT 2005Question History:1.Created by mrasch at Mon May 16 16:08:02 CDT 20052.Question Reviewed by mellis at Tue May 31 14:57:06 CDT 20053.Modified by mrasch at Thu Jun 09 17:22:58 CDT 20054.Modified by jbell at Thu Jun 16 16:52:57 CDT 20055.Modified by mrasch at Mon Jun 20 13:38:53 CDT 20056.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamQuestionID:GGNS-NRC-00847Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page37 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Date: 08/12/20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200510.Modified by mrasch at Thu Jul 28 08:44:45 CDT 200511.Modified by tharrelso at Fri Jul 29 07:21:21 CDT 200512.Modified by mellis at Fri Jul 29 12:21:05 CDT 200513.Question Reviewed by mrasch at Thu Aug 04 07:22:11 CDT 200514.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200515.Modified by mrasch at Tue Aug 09 10:52:52 CDT 200516.Question Reviewed by mellis at Tue Aug 09 11:59:50 CDT 200517.Modified by mrasch at Tue Aug 09 16:10:27 CDT 200518.Question Reviewed by mellis at Tue Aug 09 16:14:20 CDT 2005Comments:EB QUESTION: 18 (1.0 Points)The plant has experienced a transient and currently: Reactor power stabilized at 8% following the reactor scram. MSIVs are open. RPV water level is - 20 inches and lowering. Offsite power is available.In this situation, the EOPs direct the operators to defeat the low level auto closure ofthe MSIVs.One of the bases for this direction is to: A.facilitate a reactor plant cooldown to 200 degrees F. B.prevent the requirement of injection of Standby Liquid Control. C.keep the main condenser available to reduce the probability of exceeding theHCTL.Page38 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP02A Objective: 5KA

References:

1.295037 EK3.06 Maintaining heat sinks external to the containment [3.8/4.1]

References:

1.EP-2A STEP 402.PSTG pg B-14-13TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:Categories:1.Emergency Procedure Training2.Continuing TrainingTask

References:

D.keep the Turbine Bypass Valves available in the event a rapid depressurization isrequired.Answer:CQuestionComments:The MSIVs are kept open in order to keep the Main Condenseravailable as a heat sink and maintain Condensate and Feedwateravailable. This reduces the heat input into the containment andminimizes the probability of exceeding the HCTL. This makes Answer Ccorrect. Tier 1 Group 1 This is a NEW question. 10CFR 41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00850Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage39 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Question Last Revised By:Tommy Harrelson at Fri Jul 29 07:22:09 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 08:18:07 CDT 20052.Modified by jbell at Thu Jun 16 16:54:03 CDT 20053.Modified by mrasch at Mon Jun 20 07:58:06 CDT 20054.Modified by mrasch at Mon Jun 20 13:40:07 CDT 20055.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Modified by mrasch at Thu Jul 28 09:08:20 CDT 200510.Modified by tharrelso at Fri Jul 29 07:22:09 CDT 200511.Question Reviewed by mrasch at Thu Aug 04 07:22:21 CDT 200512.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 19 (1.0 Points)A radioactive liquid leak has formed a puddle inside the protected area fence.Eventually, the puddle flowed under the protected area fence, into a storm drain, andthen into a waterway.This would be considered a liquid effluent release when the radioactive water: A.initially started to leak. B.crossed the protected area fence. C.entered the storm drain. D.entered the waterway.Answer:BPage40 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-TS001 Objective: 34KA

References:

1.295038 EA2.01 Off-site [3.3/4.3]

References:

1.TS6.11.1TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:Categories:1.Technical Specifications2.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 10:15:30 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 08:24:53 CDT 20052.Modified by mrasch at Mon Jun 20 13:42:54 CDT 2005QuestionComments:Answer B is correct because because the protected area fence isconsidered to be the boundary between "onsite" and "offsite" with respectto releases. Answer A is incorrect because it is not transgressed outsidethe protected area fence. is Answers C, and D are incorrect because theareas listed in those answers are well beyond the protected area fence,which would be crossed by the liquid first, before reaching the otherlisted locations. Tier 1 Group 1 This is a NEW question. 10CFR41.10/41.13/43.4/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00851Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage41 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 09:14:56 CDT 20058.Modified by tharrelso at Fri Jul 29 10:15:30 CDT 20059.Question Reviewed by mrasch at Thu Aug 04 07:22:25 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 20 (1.0 Points)Which one of the following operator actions is required in relation to a fire inside theprotected area? A.If a fire is reported in the Division 3 Diesel Generator room, start the Division 3Diesel Generator Room Outside Air Fan from 1H13-P870. B.If a fire occurs in the Upper Cable Spreading Room, always place Transfer Switchfor Lockout Transfer Relay C61-HSS-M150 at 1H22-P152 to ON. C.Before manning the Remote Shutdown Panel due to a fire in the Main ControlRoom, defeat the Division 3 Switchgear Room CO2 system. D.If a fire occurs in the Main Control Room, secure the Control Building Fan CoilUnit, Z17-B002.Answer:CQuestionComments:Answer A is incorrect because the fan in the room with the fire isrequired to be stopped by 04-1-01-P81-1 step 3.14. 04-1-01-P81-1 step3.14 Answer B is incorrect because it is not required by 05-1-02-II-1 step3.5.1 for a fire outside the main control room. Answer C is correctbecause it is required by 05-1-02-II-1 step 3.4.1. Answer D is incorrectPage42 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 67.3KA

References:

1.600000 AK3.04 Actions contained in the abnormal procedure for plantfire on site[2.8/3.4]

References:

1.04-1-01-P81-1 step 3.142.05-1-02-II-1 Steps 3.4.1; 3.5.13.10-S-03-2 Step 6.2.2.fTrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.P64: Fire Water Protection SystemCategories:1.Administrative Requirements2.Emergency Plan Training3.Off Normal Event Procedures4.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 10:09:05 CDT 2005Question History:because the control room is not an area listed in 10-S-03-2 step 6.2.2frequiring Z17-B002 to be secured, since that does not directly serve thecontrol room. Tier 1 Group 1 This is a NEW question. 10 CFR 41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00852Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage43 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Created by mrasch at Fri Jun 10 08:33:06 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Modified by tharrelso at Thu Jul 28 08:02:53 CDT 20057.Modified by tharrelso at Fri Jul 29 10:09:05 CDT 20058.Question Reviewed by mrasch at Thu Aug 04 07:22:28 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 21 (1.0 Points)The plant is operating at 20% power when a loss of condenser vacuum occurred dueto air inleakage.What will be the first cause of a direct reactor scram as condenser vacuum slowlylowers? A.Main Turbine Trip B.Low Reactor water level C.High Reactor Water Level D.Group 1 isolationAnswer:BQuestionComments:Answer A is incorrect because the Main Turbine Trip Scram signals arebypassed. Answer B is CORRECT because the reactor feed pumpturbine trip will cause reactor level to drop causing a Level 3 Scramsignal. Answer C is incorrect because the reactor will drop when theRFPTs trip. Answer D is incorrect because the reactor will scram on lowwater level well before the MSIV closure setpoint of 9"Hg vacuum isreached. Tier 1 Group 2 This is a NEW question. 10 CFRPage44 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-ONEP Objective: 39.02.CourseID: GLP-OPS-C7100 Objective: 9; 10KA

References:

1.295002 AK2.01 RPS [3.5/3.5]

References:

1.05-1-02-V-8 Section 5.02.Tech Spec Bases B3.3.1.1 Functions 9; 10TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.C71: Reactor Protection System2.N62: Condenser Air Removal SystemCategories:1.Off Normal Event Procedures2.Systems3.Technical Specifications4.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Wed Jul 27 15:25:51 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 08:55:35 CDT 20052.Modified by mrasch at Mon Jun 20 13:43:36 CDT 200541.4/41.5/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00854Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage45 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Wed Jul 27 15:25:51 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:23:48 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 22 (1.0 Points)The plant was at the end of the current operating cycle following a 400 day run whenthe plant was scrammed due to bus 15AA lockout.SRVs are being used for reactor pressure control.The time estimated to restore bus 15AA is 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.Which one of the following should be performed to control reactor pressure? A.Install nitrogen bottles on the Instrument Air header supply to the ADS airreceivers in the Auxiliary Building. B.Bypass the MSIVs and bleed reactor pressure to the Main Condenser through theMain Steam drains from the control room. C.Allow Low-Low Set function of the SRVs to maintain pressure by cycling only twoSRVs. D.Install air jumpers in the Auxiliary Building to open Instrument Air Isolation valvesrestoring Instrument Air to the SRV receivers .Answer:AQuestionComments:This task is performed by Non-Licensed Operators and Mechanics in theAuxiliary Building to supply air (nitrogen) to the ADS valves for reactorPage46 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-ONEP Objective: 2.02.CourseID: GG-1-LP-RO-EP02A Objective: 5; 6KA

References:

1.GENERIC 2.4.35 Knowledge of local auxiliary operator tasks during emergencyoperations [3.3/3.5]2.295007

References:

1.05-S-01-EP-2 Att 7 Step 2.42.EP-2A Steps 40; 543.05-1-02-V-9 Step 3.124.05-1-02-I-4 Step 3.2.4TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.E22-2: Automatic Depressurization System2.P53: Instrument Air Systempressure control. Answer A is CORRECT since the estimated time torestore bus 15AA and restore instrument air to supply SRVs exceeds the6 hour limit in the referenced plant procedures. This installs Nitrogenbottles to ADS receivers allowing them to be used for pressure control.Answer B is incorrect because the loss of bus 15AA prevents opening theSecondary Containment Isolation Air Operated Valves. Answer C isincorrect because the air supply for the Non-ADS SRVs would beexhausted after a short period of time. Answer D is incorrect becausewould require work instructions to control installation of air jumpers anddefeat secondary containment. GGNS Scram April 2003 Tier 1 Group 2This is a NEW question. 10 CFR 41.4/41.7/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00855Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage47 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Categories:1.Emergency Procedure Training2.Off Normal Event Procedures3.Systems4.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Wed Jul 27 16:32:16 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 09:04:56 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Mon Jul 11 09:39:57 CDT 20056.Modified by mrasch at Wed Jul 13 09:05:11 CDT 20057.Modified by mrasch at Wed Jul 13 10:19:52 CDT 20058.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20059.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200510.Modified by mrasch at Tue Jul 26 13:33:20 CDT 200511.Modified by mrasch at Tue Jul 26 15:41:35 CDT 200512.Modified by mrasch at Wed Jul 27 16:32:16 CDT 200513.Question Reviewed by mellis at Thu Aug 04 07:23:49 CDT 200514.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 23 (1.0 Points)A three (3) rod out ATWS condition exists with MSIVs closed.Instrument Air is in normal lineup to the Containment and Drywell.Which one of the following describes the use of SRVs to control reactor pressure? A.Allow Low-Low Set to cycle under these conditions. B.If there is only one leaking/weeping SRV, as designated by a colored key, onlyPage48 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 59.32.CourseID: GLP-OPS-B1300 Objective: 14.1; 14.2KA

References:

1.295013 AA2.02 Localized heating/stratification [3.2/3.5]

References:

1.04-1-01-B21-1 Step 4.2.2c2.02-S-01-27 Step 6.2.4TrainingPrograms:that SRV should be used. C.Only non-ADS SRVs should be used to preserve ADS Air Receiver pressure, andthey should be rotated. D.Only ADS SRVs should be used, and they should be rotated.Answer:CQuestionComments:Answer A is incorrect because low-low set is not allowed during anATWS, as defined by being in EP-2A. Answer B is incorrect because useof only 1 SRV is limited to situations when suppression pool cooling is inservice for pool circulation. Answer C is correct because non-ADS valvesare specifically preferred when instrument air is available and rotation isrequired since Suppression Pool Cooling is not in service. Answer D isincorrect because ADS valves are specifically preferred only wheninstrument air is unavailable. Tier 1 Group 2 This is a NEW question. 10CFR 41.3/41.7/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00856Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page49 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E22-2: Automatic Depressurization SystemCategories:1.Emergency Procedure Training2.Off Normal Event Procedures3.Systems4.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 10:55:30 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 09:16:04 CDT 20052.Modified by jbell at Thu Jun 16 16:57:45 CDT 20053.Modified by mrasch at Mon Jun 20 13:45:28 CDT 20054.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Modified by mrasch at Thu Jul 28 13:58:05 CDT 20059.Modified by mellis at Fri Jul 29 12:36:41 CDT 200510.Question Reviewed by mrasch at Thu Aug 04 07:22:42 CDT 200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200512.Modified by mrasch at Tue Aug 09 10:55:30 CDT 200513.Question Reviewed by mellis at Tue Aug 09 11:59:51 CDT 2005Comments:EB QUESTION: 24 (1.0 Points)The plant is in startup with reactor pressure at 80 psig.RCIC has developed a steam leak within the RCIC room that cannot be isolated.The CRS has ordered a reactor scram due to elevated RCIC room temperature.Page50 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP04 Objective: 7Which one of the following is the reason for the reactor scram for these conditions?These actions prevent: A.An excessive release of radioactive material. B.Exceeding any applicable TRM operability limits. C.Failure of safe shutdown equipment and allows personnel access to the RCICroom. D.Contaminating the Auxiliary Building and manually starting the Standby GasTreatment system.Answer:CQuestionComments:Answer A is incorrect because there is no evidence of fuel failure andtherefor no source for gross amounts of radioactive material. Answer B isincorrect because EP-4 does not require entering EP-2 to effect a scramuntil the maximum safe temperature, 212°F, is reached. 185°F is theoperating limit, only. Answer C is correct because 212°F is the maximumsafe temperature for the RCIC room, and with an unisolable leak, asystem that cannot be isolated from the RPV is discharging outsideprimary containment. Per EP-4 step 14, EP-2 should be entered, and itwill require manual scram (EP-2 step 3). This will limit the heat added tothe area preventing a challenge to equipment in the area and allowingpersonnel access. Answer D is incorrect since only one area is affected,and temperature would have to exceed maximum safe levels in at least 2areas before contaminating the Auxiliary Building would becom an issue.Tier 1 Group 2 This is a NEW question. 10 CFR 41.5/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00857Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page51 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm KA

References:

1.295032 EK3.02 Reactor SCRAM [3.6/3.8]

References:

1.EP-4 Steps 9, 10, 13, 142.03-1-01-1 Step 6.2.6f3.PSTG Appendix B 16.14TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E51: Reactor Core Isolation Cooling SystemCategories:1.Emergency Procedure Training2.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 12:38:04 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 09:25:18 CDT 20052.Modified by mrasch at Mon Jun 20 13:46:09 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by mrasch at Wed Jul 13 10:16:02 CDT 20057.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Modified by tharrelso at Wed Jul 27 14:07:56 CDT 200510.Modified by tharrelso at Fri Jul 29 07:25:06 CDT 200511.Modified by mellis at Fri Jul 29 12:38:04 CDT 200512.Question Reviewed by mrasch at Thu Aug 04 07:22:45 CDT 200513.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Page52 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Comments:EB QUESTION: 25 (1.0 Points) How will the Standby Gas Treatment System respond to an event resulting in thefollowing radiation levels detected in ventilation systems: A B CDContainment Vent Exhaust Radiation Monitors 3.8 mr/hr 4.0 mr/hr 3.0 mr/hr4.0 mr/hrFuel Handling Area Exhaust Radiation Monitors 2.0 mr/hr 4.0 mr/hr 3.0 mr/hr4.0 mr/hrFuel Pool Sweep Exhaust Radiation Monitors 32 mr/hr 20 mr/hr INOP 22mr/hr A.A and B will remain in standby. B.A will automatically start and B will remain in standby. C.B will automatically start and A will remain in standby. D.A and B will automatically start.Answer:AQuestionComments:Initiation setpoint for SGTS from Fuel Handling Area Vent Exh RadMonitors is 3.6 mr/hr. Initiation setpoint for SGTS from Fuel HandlingArea Pool Sweep Exh Rad Monitors is 30 mr/hr. SGTS does not receivean auto start from Containment/Drywell Vent Exhaust Radiation Monitors.The initiation logic requires Channels 'A' and 'D' to initiate SGTS 'A', or 'B'and 'C' to initiate SGTS 'B'. With D17K618C removed, a channel 'B' tripof FPS Rad Monitor (K618B) would start SGTS 'B'. Answer A is correct(and answers B,C, and D are incorrect) because D17K618B, D17K618D,D17K618A, and D17K617C do not reach their trip setpoints, so neitherdivision completes a full logic initiation. Tier 1 Group 2 This is a NEWquestion. 10CFR 41.4/41.7Image

Reference:

NonePage53 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-T4801 Objective: 8.6KA

References:

1.295034 EK2.03 SBGT/FRVS: Plant-Specific [4.3/4.5]

References:

1.17-S-06-5 Att II pages 35 and 36TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.D17: Process Radiation Monitoring System2.T48: Standby Gas Treatment SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 07:42:21 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 09:32:37 CDT 20052.Modified by mrasch at Mon Jun 20 08:07:53 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Closed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00858Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage54 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 6.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Wed Jul 27 13:33:00 CDT 20058.Modified by mrasch at Wed Jul 27 13:47:18 CDT 20059.Modified by mrasch at Wed Jul 27 13:52:29 CDT 200510.Modified by tharrelso at Fri Jul 29 07:26:52 CDT 200511.Modified by tharrelso at Fri Jul 29 07:40:57 CDT 200512.Modified by tharrelso at Fri Jul 29 07:42:21 CDT 200513.Question Reviewed by mrasch at Thu Aug 04 07:22:48 CDT 200514.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 26 (1.0 Points)The plant is in a refueling outage when a fuel handling accident occurs.Standby Gas Treatment System (SGTS) A is manually initiated.on 1H13-P870, amber alarm ENCL BLDG NEG PRESS LO (2A-E3) is subsequentlyreceived and does not clear.What is the operational implication of this alarm? A.Greater personnel industrial safety hazard when entering or exiting the auxiliarybuilding. B.Possible damage to the enclosure building due to high pressure. C.Possible excessive filtered leakage from secondary containment. D.Possible excessive unmonitored leakage from secondary containment.Page55 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-T4801 Objective: 2.0; 11.02.CourseID: GG-1-LP-RO-EP04 Objective: 6KA

References:

1.295035 EK1.02 Radiation release [3.7/4.2]

References:

1.ARI 04-1-02-1H13-P870 2A-E32.EP-4 Bases pg B-8-2, 6TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.T48: Standby Gas Treatment SystemAnswer:DQuestionComments:Answer A is incorrect because the alarm is indicative of a higher pressurein secondary containment (i.e. lower dp relative to outside secondarycontainment). A safety concern only exists when there is a higher dp,resulting in high forces on doors which could cause them to open quicklywhen unlatched. Answer B is incorrect because the alarm is indicative ofa low dp with respect to outside, not a high dp which could causeexcessive forces on enclosure building coverings. Answer C is incorrectbecause no amount of filtered leakage would be excessive. SGTS flowrates would be governed predominantly by the flow control circuit, soapproximately the same flow rate through the SGTS filter train would existafter the 120 sec timer had expired. Answer D is correct becauseprevention of exfiltration could not be assured if pressure was higher than-0.25"wc. The given alarm occurs at -0.2"wc or higher pressure. Thatleakage would bypass the SGTS filter train and exhaust radiationmonitoring system. Tier 1 Group 2 This is a NEW question. 10CFR41.8/41.10/41.13/43.4/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00859Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page56 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Categories:1.Emergency Procedure Training2.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 12:42:47 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 09:40:21 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Modified by mellis at Fri Jul 29 12:39:22 CDT 20057.Modified by mellis at Fri Jul 29 12:42:47 CDT 20058.Question Reviewed by mrasch at Thu Aug 04 07:22:51 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 27 (1.0 Points)RHR A room sump pumps are each in AUTO with their selector switch in ALTERNATE.RHR A is operating in Shutdown Cooling when a 10 gpm leak develops inside RHR Aroom.RHR Room A Floor Drain Sump Pump A, P45-C013A, started and pumped downalmost to the low sump level but its breaker tripped due to a fault.How will this affect subsequent operation of the other sump pump in RHR A room,P45-C013B? A.P45-C013B will alternate starting on high-high and high sump levels, each timepumping until low sump level is reached. B.P45-C013B will come on at high-high sump level and off at high sump level andcontinue to cycle that way.Page57 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-P4500 Objective: 7.1; 11.1; 11.2KA

References:

1.295036 EA1.01 Secondary containment equipment and floor drainsystems [3.2/3.3]

References:

1.04-1-01-P45-2 Steps 3.5; Note 4.2.2a2.M-1094A3.M-1098BTrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.P45: Floor and Equipment Drain SystemCategories: C.P45-C013B will come on at high-high sump level and off at low sump level andcontinue to cycle that way. D.P45-C013B will come on at high sump level and off at low sump level one time. Itwill not automatically start after that.Answer:DQuestionComments:If the High alarm has cleared on the initial start from pump A then the Bpump will start on a subsequent high level alarm but only once since thelogic thinks pump A should start next. Therefore Answer D is correct.Tier 1 Group 2 This is a NEW question. 10 CFR 41.4/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00860Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage58 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 07:46:20 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 09:46:57 CDT 20052.Modified by mrasch at Fri Jun 10 13:23:02 CDT 20053.Modified by mrasch at Mon Jun 20 08:13:30 CDT 20054.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Modified by mrasch at Fri Jul 29 07:46:20 CDT 20059.Question Reviewed by mellis at Thu Aug 04 07:23:50 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 28 (1.0 Points)A plant shutdown is in progress and conditions are as follows:Reactor pressure 30 psigReactor water level 36 inches Narrow RangeBoth reactor recirculation pumps are shutdown.Shutdown cooling was lost due to a failure of the Shutdown Cooling common suctionisolation valve E12-F008.Which of the following would provide an alternate method to ensure core cooling ismaintained? A.Allowing natural circulation to initiate.Page59 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E1200 Objective: 9.7KA

References:

1.203000 K5.02 Core cooling methods [3.5/3.7]

References:

1.E-1181-67TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training Program B.Initiating RCIC. C.Placing one loop of RHR in the LPCI mode and opening SRVs. D.Starting one of the Reactor Recirculation pumps.Answer:CQuestionComments:Answer A is incorrect because level is not high enough to allow naturalcirculation. Answer B is incorrect because Reactor pressure is belowRCIC isolation setpoint making RCIC unavailable. Answer C isCORRECT since these actions are consistent with guidance in ONEPInadequate Decay Heat Removal 05-1-02-III-1. Answer D is incorrectbecause the restoration of a Recirc pump does not result in heatremoval. Tier 2 Group 1 This is a NEW question. 10CFR41.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00861Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page60 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Systems:1.E12: Residual Heat Removal SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:12:23 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 09:53:23 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Modified by tharrelso at Fri Jul 29 06:26:05 CDT 20057.Modified by tharrelso at Fri Jul 29 08:34:42 CDT 20058.Modified by mellis at Fri Jul 29 12:44:03 CDT 20059.Question Reviewed by mrasch at Thu Aug 04 07:23:02 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200511.Modified by mrasch at Tue Aug 09 11:12:23 CDT 200512.Question Reviewed by mellis at Tue Aug 09 11:59:52 CDT 2005Comments:EB QUESTION: 29 (1.0 Points)The plant is in Mode 5.RHR B has been placed in Shutdown Cooling with suction from Recirc Loop B,returning to the reactor via RHR B SHUTDN CLG RTN TO FW valve E12-F053B.RHR B ADHRS MODE TRIP ENABLE hand switch on 1H13-P618 is in NORMAL.Which one of the following will cause RHR B pump to trip?Page61 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E1200 Objective: 8.1 A.Opening the power supply to RHR SHUTDN CLG INBD SUCT VLV E12-F009. B.RHR SHUTDN CLG OTBD SUCT VLV E12-F008 were to close to 50%. C.RHR B FPC ASSIST SUCTION VLV E12-F066B were to open to 50%. D.RHR B ADHRS MODE TRIP ENABLE hand switch on 1H13-P618 were to beplaced in ADHRS position.Answer:BQuestionComments:The purpose of the RHR pump suction path interlocks is to trip the RHRpump when no fully open suction path exists. Answer A is incorrectbecause the trip circuit for RHR pump 'B' looks directly at the limit switchcontact for E12F009. It does not rely on control power for E12F009, butuses RHR B/C DC logic power. Answer B is correct because the tripcircuit for RHR pump 'B' looks directly at the limit switch for E12F008.RHR pump B trip coil is energized when the valve closes to "not full open"position, ~95% open. Answer C is incorrect because either a suction pathfrom the reactor, one from the suppression pool, or one from the fuel poolhas to exist for RHR pump to remain running. In this case, a suction pathexists from the reactor, so isolation of the suction path from the fuel pooldoes not energize the pump trip coil. Answer D is incorrect becauseplacing RHR B ADHRS MODE TRIP ENABLE hand switch in ADHRSposition only removes E12F066B as a permissive to run RHR pump B.As long as the suction path from the reactor is aligned fully open, thepump will continue to run. Tier 2 Group 1 This is a NEW question. 10CFR41.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00862Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage62 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm KA

References:

1.205000 K4.04 Adequate pump NPSH [2.6/2.6]

References:

1.E-1160-09; 102.E-1181-05; 44; 68TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.E12: Residual Heat Removal SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 10:18:44 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 09:58:56 CDT 20052.Modified by mrasch at Mon Jun 20 08:15:20 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by tharrelso at Fri Jul 29 08:36:03 CDT 20058.Modified by tharrelso at Fri Jul 29 10:14:03 CDT 20059.Modified by tharrelso at Fri Jul 29 10:17:45 CDT 200510.Modified by tharrelso at Fri Jul 29 10:18:44 CDT 200511.Question Reviewed by mrasch at Thu Aug 04 07:23:06 CDT 200512.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:Page63 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm EB QUESTION: 30 (1.0 Points)Objectives:1.CourseID: GLP-OPS-E2100 Objective: 9.2; 10.2; 16.0KA

References:

1.209001 A4.01 Core spray pump [3.8/3.6]

References:

1.ARI 04-1-02-1H13-P601 20A-H6TrainingPrograms:Due to a blown fuse, logic power for RHR A has been lost.In this situation, if a LOCA were to occur, the LPCS pump: A.would auto start but RHR Pump A would have to be manually started. B.would have to be manually started, and RHR Pump A is unavailable. C.and RHR Pump A are both available but each would have to be manually started. D.and RHR Pump A are both unavailable.Answer:CQuestionComments:The loss of logic power for RHR A will preclude an auto start of eitherpump. However,, control power is still available so both pumps could bemanually started if desired. This makes Answer C the only correctanswer. Tier 2 Group 1 This is a NEW question. 10CFR 41.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00863Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page64 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E21: Low Pressure Core Spray SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:12:54 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 10:04:09 CDT 20052.Modified by mrasch at Fri Jun 10 12:40:43 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 09:36:27 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:23:50 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200510.Modified by mrasch at Tue Aug 09 11:12:54 CDT 200511.Question Reviewed by mellis at Tue Aug 09 11:59:53 CDT 2005Comments:EB QUESTION: 31 (1.0 Points)The plant is in Mode 1.RHR C Jockey Pump has been shutdown for 5 days.How will this affect High Pressure Core Spray (HPCS) pump suction automatic re-alignment to Suppression Pool if actual Suppression Pool Level rises? A.HPCS suction will automatically re-align when Suppression Pool level reaches thePage65 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E2201 Objective: 22.0KA

References:

1.209002 A4.09 Suppression pool level: BWR-5,6 [3.4/3.5]

References:

1.04-1-01-E22-1 Sections 6.3; 6.4TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training Programsetpoint. B.HPCS suction will automatically re-align at a Suppression Pool level that isactually higher than the setpoint. C.HPCS suction will automatically re-align at a Suppression Pool level thatis actually lower than the setpoint. D.HPCS suction will automatically re-align on Low Condensate Storage Tank levelonly.Answer:AQuestionComments:Answer A is correct. RHR C jockey Pump does not affect HPCSSuppression Pool Level Instrumentation. Tier 2 Group 1 This is aNEW question. 10 CFR 41.7/41.8Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00864Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page66 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Systems:1.E22-1: High Pressure Core Spray SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:15:05 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 10:11:05 CDT 20052.Modified by mrasch at Fri Jun 10 12:40:11 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Wed Jul 27 16:01:51 CDT 20058.Modified by mellis at Fri Jul 29 12:47:02 CDT 20059.Question Reviewed by mrasch at Thu Aug 04 07:28:59 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200511.Modified by mrasch at Tue Aug 09 11:15:05 CDT 200512.Question Reviewed by mellis at Tue Aug 09 11:59:53 CDT 2005Comments:EB QUESTION: 32 (1.0 Points)Why does E22-F004, HPCS Injection Valve, automatically close on high reactor waterlevel? A.Prevent over pressurizing the Reactor Pressure Vessel (RPV). B.Prevent excessive cool down rates for RPV internals. C.Page67 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E2201 Objective: 5.5; 9.4; 17.0KA

References:

1.GENERIC 2.1.28 Knowledge of the purpose and function of major systemcomponents and controls [3.2/3.3]2.209002

References:

1.Tech Sec Bases B3.3.5.1 function 3c2.FSAR 7.3.1.1.1.3.6TrainingPrograms:Prevent overflow into the main steam lines. D.Prevent Main Turbine and Reactor Feed Pump trips.Answer:CQuestionComments:The FSAR design function and Tech Spec bases for the High ReactorWater Level isolation of E22-F004 HPCS Injection Valve is to preventwater introduction into the Main Steam Lines. Answer A is incorrectbecause this is not the reason listed in Tech Spec bases and FSAR. Inthis case, SRVs would prevent over-pressurization. Answer B is incorrectbecause this is not the reason listed in Tech Spec bases and FSAR. Thedesign function of HPCS is to provide adequate core cooling, not tocontrol cool down rates. Answer C is CORRECT because this is thereason listed in Tech Spec bases and is assumed in the accidentanalysis. Answer D is incorrect because this is not the reason listed inTech Spec bases and FSAR, and would not prevent reaching level 9 inall cases due the time required for E22 F004 to stroke closed. Tier 2Group 1 This is a NEW question. 10CFR 41.8/43.2Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00865Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page68 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.E22-1: High Pressure Core Spray SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:15:45 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 10:16:45 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Mon Jul 11 09:58:23 CDT 20056.Modified by mrasch at Wed Jul 13 10:59:56 CDT 20057.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Modified by mrasch at Tue Jul 26 13:37:59 CDT 200510.Modified by mrasch at Tue Jul 26 13:39:15 CDT 200511.Modified by tharrelso at Fri Jul 29 08:37:33 CDT 200512.Modified by mellis at Fri Jul 29 12:48:12 CDT 200513.Question Reviewed by mrasch at Thu Aug 04 07:29:00 CDT 200514.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200515.Modified by mrasch at Tue Aug 09 11:15:45 CDT 200516.Question Reviewed by mellis at Tue Aug 09 11:59:54 CDT 2005Comments:EB QUESTION: 33 (1.0 Points)Both squib valves have failed to fire on a Standby Liquid Control (SLC) initiation.What affect will this have on the ability of SLC to inject and what acceptable alternatePage69 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C4100 Objective: 6; 10.1; 10.4; 12KA

References:

1.211000 A2.02 Failure of explosive valve to fire [3.6/3.9]

References:

means of SLC injection should be utilized? A.SLC will inject at a much lower injection rate. Inject Sodium Pentaborate directlyinto piping downstream of the Condensate demineralizers such as the ReactorFeedwater pump suction. B.SLC will NOT inject. Dump required amounts of boric acid and borax into theRefueling Water Storage Tank (RWST) and inject with the Condensate TransferPumps. C.SLC will inject at a much lower injection rate. Dump required amounts of boric acidand borax into the Condensate Storage Tank (CST) and inject with theCondensate Transfer pumps. D.SLC will NOT inject. Dump required amounts of boric acid and borax into theCondensate Storage Tank (CST) and mix the contents then inject with RCIC.Answer:DQuestionComments:If neither squib valve fires then SLC will not inject requiring the use ofEP-2A Attachment 28 Alternate SLC. This involves adding 5000 lbs.each of Boric Acid and Borax to the CST and mixing and injecting witheither the RCIC pump or HPCS. Therefore Answer D is Correct. Tier 2Group 1 This is a NEW question. 10 CFR 41.7/41.8/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00866Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page70 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.04-1-01-C41-1 ATT. VI2.05-S-01-EP-2 Attachment 28TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.C41: Standby Liquid Control SystemCategories:1.Emergency Procedure Training2.Systems3.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 12:54:23 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 10:22:40 CDT 20052.Modified by mrasch at Mon Jun 20 08:18:48 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by mrasch at Mon Jul 11 11:17:18 CDT 20057.Modified by mrasch at Thu Jul 14 08:50:36 CDT 20058.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20059.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200510.Modified by mrasch at Thu Jul 28 14:25:00 CDT 200511.Modified by mellis at Fri Jul 29 12:54:23 CDT 200512.Question Reviewed by mrasch at Thu Aug 04 07:29:01 CDT 200513.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 34 (1.0 Points)Page71 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm A plant startup is in progress at 36% of rated thermal power.The A Main Bypass Control Valve (BCV) has failed open and startup has beensuspended.The Baxter Wilson 500 KV line into GGNS trips causing breakers J5228 and J5232 toopen. The Turbine Initial Pressure Control (IPC) system responds as designed, except the AMain Bypass Control Valve remains open. How would the plant automatically respond to this event and what operator actionsshould be taken? A.The Reactor Recirculation Pumps would transfer to slow speed and the reactorwould scram on high water level. Enter the Reactor Scram ONEP. B.The Reactor Recirculation Pumps would transfer to slow speed and the reactorwould scram due to the Turbine Control Valve closure. Enter the Reactor ScramONEP. C.The Reactor Recirculation Pumps would remain in fast speed and the reactorwould scram on high neutron flux. Enter the Reactor Scram ONEP. D.The Reactor Recirculation Pumps would remain in fast speed and the reactorwould continue to operate. Suspend plant startup until the BCV A failure iscorrected.Answer:DQuestionComments:Though at 36% power, turbine 1st stage inlet pressure was equivalent to~26% power with one bypass valve full open (~13%) worth). TSV/TCVscrams bypassed until 40% power based on Turbine 1st stage inletpressure. Therefore, EOC/RPT and scram from TCV/TSV closure isautomatically bypassed. B and C bypass valves can accommodate steamflow for the remaining 26% power, so reactor pressure is relativelyunaffected. Answer A and B are incorrect because EOC/RPT is bypassedunder given conditions. Answer C is incorrect because bypass valves arefast acting and can accommodate all steam flow for the specified powerlevel, so the pressure/flux transient is minimal, well below what wouldcause a high flux scram. Answer D is CORRECT for the reasonsPage72 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-N3202 Objective: 162.CourseID: GLP-OPS-B3300 Objective: 27.53.CourseID: GLP-OPS-C7100 Objective: 10KA

References:

1.212000 A2.12 Main turbine stop control valve closure [4.0/4.1]

References:

1.FSAR 7.2.1.1.4.4.2TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.B33: Reactor Recirculation System2.C71: Reactor Protection System3.N32: EHC Control SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Wed Jul 27 14:40:48 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 10:30:09 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 2005previously stated. Tier 2 Group 1 This is a NEW question. 10CFR41.5/41.6/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00867Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage73 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Thu Jul 14 09:04:20 CDT 20056.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Modified by tharrelso at Wed Jul 27 14:40:48 CDT 20059.Question Reviewed by mellis at Thu Aug 04 07:24:12 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 35 (1.0 Points)Which of the following combinations of plant activities would be allowed by plantprocedures? A.Driving in IRM 'A' while I&C performs a surveillance for Scram Discharge VolumeWater Level High channel B B.Driving in IRM 'A' with the under-vessel service platform out of its standby positionand the under-detector grid sections installed C.Driving out IRM 'A' while driving out IRM 'C' in Mode 2 D.Driving out IRM 'A'and IRM 'B' simultaneously in Mode 1Answer:CQuestionComments:Answer A is incorrect because driving IRM might cause a Div 2 halfscram during the Div 1 half scram surveillance, resulting in a full scram.Answer B is incorrect because this might cause damage to the IRM drivecables. Answer C is correct because both IRMs are Div 1, so the worstconsequence for driving one IRM, a Div 1 half scram, is no worse thanthat for driving the two IRMs simulataneously. Answer D is incorrectbecause driving IRM 'A' might cause a Div 1 half scram, driving IRM 'B'might cause a Div 2 half scram, resulting in a full scram. Tier 2 Group 1This is a NEW question. 10CFR 41.7/41.10/43.5Page74 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C5102 Objective: 12.2KA

References:

1.215003 K5.03 Changing detector position [3.0/3.1]

References:

1.04-1-01-C51-1 sections 3.5, 3.7, cautions section 4.2.2TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.C51-2: Intermediate Range Nuclear Instrumentation SystemCategories:1.Administrative Requirements2.Systems3.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Mon Jun 20 08:22:45 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 10:40:26 CDT 20052.Modified by mrasch at Mon Jun 20 08:22:45 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00868Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage75 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 6.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 36 (1.0 Points)With the plant in Mode 2, which of the following will result in a control rod block? A.SRM A is upscale and bypassed.SRM E is NOT full in and reads 70 cps.All IRMs are on Range 2. B.SRMs A and E are INOP.IRMs A and E are on Range 9. C.SRM A is upscale and bypassed.SRM B is NOT full in and reads 275 cps.All IRMs are on range 2. D.SRMs A and B are INOP.IRMs A and B are on Range 9.Answer:AQuestionComments:With SRM E not full in and less than 70 CPS a rod block will occurbecause SRM s are single coincidence logic and SRM E will generate arod block. None of the other conditions would generate a control rodblock, therefore Answer A is the only correct answer. Tier 2 Group 1This is a NEW question. 10CFR 41.2/41.6Image

Reference:

NoneClosed Reference QuestionPage76 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C5101 Objective: 8.2; 11.1KA

References:

1.GENERIC 2.2.33 Knowledge of control rod programming [2.5/2.9]2.215004

References:

1.04-1-01-C51-1 Step 3.82.E-1171-203.FSAR 7.6.2.5.1.1TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.C11-2: Rod Control and Information System2.C51-1: Source Range Nuclear Instrumentation SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 12:58:07 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 10:53:52 CDT 20052.Modified by mrasch at Mon Jun 20 08:24:17 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Handout Not Required with ExamQuestionID:GGNS-NRC-00869Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page77 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 6.Modified by mrasch at Mon Jul 11 13:03:15 CDT 20057.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Modified by mrasch at Thu Jul 28 10:08:00 CDT 200510.Modified by mellis at Fri Jul 29 12:58:07 CDT 200511.Question Reviewed by mrasch at Thu Aug 04 07:29:02 CDT 200512.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 37 (1.0 Points)Objectives:1.CourseID: GLP-OPS-C5104 Objective: 11.3The power supply for APRM D is: A.1Y87 B.1Y88 C.1Y95 D.1Y96Answer:CQuestionComments:per the referenced L62 SOI and C51 SOI 1Y95 Inverter is the powersupply to APRM cabinet D. Tier 2 Group 1 This is a NEW question.10 CFR 41.6Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00870Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage78 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm KA

References:

1.215005 K2.02: 2.6/2.8 Knowledge of power supplies to APRM Channels.

References:

1.04-1-01-L62-1 Att VI Table 3TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.C51-5: Average Power Range Nuclear Instrumentation System2.L62: Uninterruptible Power Supply SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Thu Jul 28 10:14:55 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 11:02:59 CDT 20052.Modified by mrasch at Fri Jun 10 13:40:49 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 10:14:55 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:24:20 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 38 (1.0 Points)Page79 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E5100 Objective: 3.1; 8.17; 19.0KA

References:

1.217000 A1.02 RCIC pressure [3.3/3.3]

References:

1.M-1083A2.M-1085AThe reactor is at 1000 psig and slowly lowering.RCIC is in operation in automatic control.In response to the RPV pressure lowering, RCIC discharge pressure will: A.slowly rise due to the governor valve slowly opening. B.slowly lower due to RCIC turbine speed slowly lowering. C.remain constant due to RCIC speed slowly rising. D.remain constant due to the RCIC throttle valve slowly opening.Answer:BQuestionComments:As RPV pressure lowers, the flow controller will close down on thegovernor valve thereby reducing RCIC speed and discharge pressuremaking B the correct answer. This makes the other three distractorsincorrect. Tier 2 Group 1 This is a NEW question. 10 CFR41.4/41.7/41.14Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00871Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage80 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.M-1077DTrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E12: Residual Heat Removal System2.E51: Reactor Core Isolation Cooling System3.N21: Feedwater SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Thu Jul 28 10:24:22 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 11:17:17 CDT 20052.Modified by mrasch at Mon Jun 20 08:26:05 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 10:24:22 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:24:23 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 39 (1.0 Points)The power supply for Division 1 ADS Logic is Division 1: A.RPS.Page81 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E2202 Objective: 9.1; 9.2; 10.2; 15.0; 19.3; 25.0; 27.0KA

References:

1.218000 K2.01 ADS logic [3.1/3.3]2.218000 K6.06 D [3.4/3.6]

References:

1.E-1161-04TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E22-2: Automatic Depressurization System B.UPS. C.120 VAC. D.125 VDC.Answer:DQuestionComments:The power supply to ADS Logic is from the Divisional DC power supplyper the facility electrical drawings. Therefore Answer D is the onlycorrect answer. Tier 2 Group 1 This is a NEW question. 10CFR 41.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00872Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage82 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 2.L11: Plant DC Electical SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Thu Jul 28 10:40:13 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 11:34:17 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Mon Jul 11 13:08:46 CDT 20056.Modified by mrasch at Thu Jul 14 09:18:07 CDT 20057.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Modified by mrasch at Thu Jul 28 10:40:13 CDT 200510.Question Reviewed by mellis at Thu Aug 04 07:24:26 CDT 200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 40 (1.0 Points)A condenstate system transient occurred while at 100% power which caused reactorwater level to drop to -50 inches wide range.Level 2 Trip units for Groups 6, 7, 8, 10 and Auxiliary Building (B21-N682A and B21-N682B) failed to trip.A large leak on Drywell Chilled Water piping inside containment has been reported.Regarding building isolations only, what is the appropriate response for theseconditions? A.All Division 1 valves have failed to isolate. All Division 2 valves have isolated.Manually close all Division 1 valves that failed to isolate.Page83 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 59.3KA

References:

1.223002 A2.03 System logic failures [3.0/3.3] B.Only Division 1 and 2 primary containment valves have failed to isolate. Allsecondary containment valves have closed. Isolate all primary containmentvalves. C.No primary or secondary containment valves have isolated. Close at least onevalve in each penetration, except for Instrument Air and Plant Service Water. D.No primary or secondary containment valves have isolated. Close at least onevalve in each penetration, except for Drywell Chilled Water, Instrument Air andPlant Service Water.Answer:CQuestionComments:Many isolations should have occurred due to low reactor water level,level 2. Operations Philosophy, 02-S-01-27, states in a failure to isolatesituation, only one valve in each penetration that should be isolatedneeds to be shut, and to not isolate P53, P44, or P72 if those systemsare intact and are going to be unisolated per the EPs. Answer A isincorrect because only one valve in each penetration is required to beclosed. Answer B is incorrect because some isolation valves that failedare MOVs, and loss of air has no effect on them. Answer C is correctbecause it mirrors Operations Philosophy as stated above. P72 shouldbe isolated because of a system breach. Answer D is correct becauseP72 should be isolated because of a system breach. Tier 2 Group 1 Thisis a NEW question. 10CFR 41.9/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00873Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:3Page84 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm

References:

1.02-S-01-27 Step 6.1.32.17-S-06-5 Att I page 3; Att II 28, 30, 31, 32, 34, 35, 36TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.P44: Plant Service Water System2.P53: Instrument Air System3.P72: Drywell Chill Water SystemCategories:1.Administrative Requirements2.Systems3.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 12:28:02 CDT 2005Question History:1.Created by mrasch at Fri Jun 10 12:20:51 CDT 20052.Modified by mrasch at Mon Jun 20 08:27:42 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 17:46:39 CDT 20058.Modified by tharrelso at Fri Jul 29 08:52:04 CDT 20059.Modified by mrasch at Fri Jul 29 12:23:26 CDT 200510.Modified by mrasch at Fri Jul 29 12:28:02 CDT 200511.Question Reviewed by mellis at Thu Aug 04 07:24:29 CDT 200512.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:Page85 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm EB QUESTION: 41 (1.0 Points)Objectives:1.CourseID: GLP-OPS-E2202 Objective: 18KA

References:

1.239002 K4.09 Manual opening of the SRV [3.7/3.6]

References:

1.E-1161-13; 16TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramIn the event of a loss of Instrument Air to Containment, what is the minimum number ofmanual actuations the Automatic Depressurization System air system is designed toprovide Safety Relief Valve B21-F051D over a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period? A.100 B.50 C.25 D.10Answer:AQuestion Comments:Tier 2 Group 1 This is a NEW question. 10CFR 41.3/41.5/41.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00874Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage86 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Systems:1.E22-2: Automatic Depressurization System2.P53: Instrument Air SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 10:23:47 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 07:37:50 CDT 20052.Modified by mrasch at Mon Jun 13 07:39:03 CDT 20053.Modified by mrasch at Mon Jun 20 08:29:28 CDT 20054.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Modified by mrasch at Mon Jul 11 13:19:16 CDT 20058.Modified by mrasch at Thu Jul 14 10:04:14 CDT 20059.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 200510.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200511.Modified by mrasch at Wed Jul 27 16:06:08 CDT 200512.Modified by tharrelso at Fri Jul 29 10:23:47 CDT 200513.Question Reviewed by mellis at Thu Aug 04 07:24:32 CDT 200514.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 42 (1.0 Points)The plant is at 90% power.Annunciator RX LVL HI/LO is received on 1H13-P680.You notice the output of the Master Level Controller is slowly trending down, the outputof RFP A speed controller is slowly trending up, and the output of RFP B speedcontroller is slowly trending down.Which of the following describes the cause of these indications and the operatorPage87 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm response? A.RFP A runout due to the Master Level Controller failing in automatic operation.Immediately insert a manual scram. B.RFP B failing to the low speed stop due to the Master Level Controller failing inautomatic operation. Take manual control of the Master Level Controller andcontrol level 32 inches to 42 inches. C.RFP A runout due to the RFP A speed control failing upscale. Attempt to takemanual control of RFP A, and control level 32 inches to 42 inches. If it CANNOTbe controlled manually, trip RFP A and verify a Recirc Flow Control Runbackoccurs. D.RFP B failing to the low speed stop due to the RFP B speed control failingdownscale. Attempt to take manual control of RFP B, and control level 32 inchesto 42 inches. If it CANNOT be controlled manually, trip RFP B and verify a RecircFlow Control Runback occurs.Answer:CQuestionComments:The condition is indicative of failure upward of the RFP A speedcontroller. The master level controller and RFP B speed controller areresponding to the increased inventory added by RFP A. Answer A isincorrect because The Master Controller output is responding to RPVlevel. The transient is stated to be slow, and scramming is notconservative if it can be avoided. Procedural guidance exists to avoid ascram. Answer B is incorrect because the master level controllercontroller is responding properly to the failure of RFP A speed controller.Answer C is CORRECT because ONEP 05-1-02-V-6 specifically directsthis action for the given failure. Answer D is incorrect because RFP Bcontroller is responding properly to the failure of A. Tier 2 Group 1 This isa NEW question. 10 CFR 41.4/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00875Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page88 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-N2100 Objective: 17; 31.2; 372.CourseID: GLP-OPS-ONEP Objective: 37KA

References:

1.259002 A2.04 RFP runout condition: Plant-Specific [3.0/3.1]

References:

1.05-1-02-V-6 Step 2.2TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.C34: Feedwater Level Control System2.N21: Feedwater SystemCategories:1.Off Normal Event Procedures2.Systems3.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 12:30:52 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 07:45:26 CDT 20052.Modified by mrasch at Mon Jun 20 08:31:07 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:19 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by mrasch at Thu Jul 14 09:35:26 CDT 20057.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Modified by mrasch at Tue Jul 26 14:16:14 CDT 200510.Modified by mrasch at Wed Jul 27 09:47:44 CDT 2005Page89 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 11.Modified by mrasch at Wed Jul 27 12:40:35 CDT 200512.Modified by mrasch at Wed Jul 27 12:46:39 CDT 200513.Modified by mellis at Fri Jul 29 12:30:52 CDT 200514.Question Reviewed by mrasch at Thu Aug 04 07:29:03 CDT 200515.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 43 (1.0 Points)The plant is at 100% power.Which one of the following will cause the Feedwater THREE ELEMENT pushbuttonDISABLED back light to illuminate on 1H13-P680-2C? A.B Feedwater flow signal drifts up to 11 mlbm/hr. B.The difference between highest and lowest Main Steam Line Flow signalsdiverges to 1 mlbm/hr. C.The difference between Feedwater Flow A and B diverge to 0.6 mlbm/hr greaterthan normal. D.Plant power reduction to 45% rated thermal power.Answer:AQuestionComments:Answer A is correct because when a feedwater flow signal goes above10.6 mlbm/hr it is a hard failure which disables three element control.Answer B is incorrect because Steam flow divergence is 1.6 mlbm/hr todisable three element control. Answer C is incorrect because feedwaterflow divergence is 0.8 mlbm/hr to disable three element control. AnswerD is incorrect because on a power decrease three element control isdisabled at 30% power (4.95 mlbm/hr). Tier 2 Group 110CFR41.7/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00233aPage90 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C3400 Objective: 6.32.CourseID: GLP-OPS-C3400 Objective: 5.3KA

References:

1.295009 AA4.06: 3.1/3.22.259002 A4.06 DP/Single/three element control selector switch:Plant-Specific[3.1/3.2]

References:

1.04-1-02-1H13-P680 2A-C9TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.C34: Feedwater Level Control System2.B21: Nuclear Boiler SystemCategories:1.SystemsTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 08:53:29 CDT 2005Question History:1.Created by tharrelso at Wed May 04 16:49:27 CDT 20052.Created by tharrelso at Wed May 04 16:49:27 CDT 2005 from parent QuestionIDGGNS-NRC-002333.Modified by mrasch at Tue May 24 08:46:46 CDT 20054.Modified by mrasch at Tue May 24 10:05:45 CDT 20055.Question Reviewed by mellis at Tue May 31 14:56:58 CDT 20056.Modified by tharrelso at Tue Jun 07 15:36:27 CDT 20057.Modified by mrasch at Mon Jun 13 07:56:49 CDT 20058.Modified by mrasch at Mon Jun 13 08:03:40 CDT 20059.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 2005Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage91 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 10.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200512.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200513.Modified by mrasch at Thu Jul 28 15:15:56 CDT 200514.Modified by tharrelso at Fri Jul 29 08:53:29 CDT 200515.Question Reviewed by mrasch at Thu Aug 04 07:20:40 CDT 200516.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 44 (1.0 Points)A planned move of radioactive material is planned on the refueling floor.As a precaution, Standby Gas Treatment System (SGTS) A has been started andapplicable parameters have stabilized.Enclosure Building differential pressure is -0.6 inches wc and stable.An operator is about to manually start Standby Gas Treatment system B.How should the Steam Tunnel Outside Containment damper, T48-F005 respond? A.T48-F005 will immediately throttle to its intermediate position. B.T48-F005 will travel full open and throttle to its intermediate position after 90seconds. C.T48-F005 will travel full open and throttle to its intermediate position after 120seconds. D.T48-F005 will travel full open and throttle to its intermediate position after 120seconds or Enclosure Building pressure reaches -.75 inches wc.Page92 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-T4801 Objective: 8.4; 8.5; 8.7KA

References:

1.261000 A3.03 Valve operation [3.0/2.9]

References:

1.04-1-01-T48-1 Step 5.2.1b NOTETrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.T48: Standby Gas Treatment SystemCategories:1.Systems2.Continuing TrainingTask

References:

Answer:AQuestionComments:Answer A is correct because enclosure building pressure is less than -0.2 inches wc when SGTS B is initiated, so T48F005 goes tointermediate position right away. Answer B is incorrect because theT48F005 damper will not travel to it full open position before throttling.Answer C is incorrect because T48F005 would go to intermediateposition right away instead of 120 sec later. Answer D is incorrectbecause T48F005 would go to intermediate position right away instead of120 sec or -.75 inches wc in the enclosure building. These times areactually for flow control vane T48F500B. Tier 2 Group 1 This is a NEWquestion. 10CFR 41.4/41.13Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00876Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage93 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Question Last Revised By:Mickey Ellis at Fri Jul 29 13:02:31 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 07:54:52 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Modified by tharrelso at Wed Jul 27 15:02:15 CDT 20057.Modified by mellis at Fri Jul 29 13:01:50 CDT 20058.Modified by mellis at Fri Jul 29 13:02:31 CDT 20059.Question Reviewed by mrasch at Thu Aug 04 07:29:04 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 45 (1.0 Points)The plant is operating in mode 1 with HPCS running for a surveillance with suctionfrom the suppression pool.A fire in the Division 3 Diesel Generator room de-energized bus 11DC.How does this affect the operation of the HPCS Pump breaker 152-1702 and theHPCS Injection Valve E22-F004? A.Pump breaker can be tripped using its control room hand switch but the injectionvalve must be operated locally. B.Pump breaker must be tripped locally at the breaker and the injection valve mustbe operated locally. C.Pump breaker can be tripped from the control room and the injection valve can beoperated from the control room. D.Pump breaker must be tripped locally at the breaker and the injection valve can bePage94 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E2201 Objective: 13.2; 13.3KA

References:

1.262001 K6.01 D [3.1/3.4]

References:

1.E-1183-032.E-1188-19TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.E22-1: High Pressure Core Spray System2.L11: Plant DC Electical System3.R21: 4.16 KV AC Power SystemCategories:operated from the control room.Answer:DQuestionComments:Answer A is incorrect because 11DC supplies control power for 152-1702, and E22F004 could be operated from the control room since it isAC. Answer B is incorrect because 11DC supplies control power for 152-1702, and E22F004 could be operated from the control room since it isAC. Answer C is incorrect because 11DC supplies control power for 152-1702. Answer D is correct because there is no control power to energizethe trip coil for 152-1702, and E22F004 control power is AC power. Tier2 Group 1 This is a NEW question. 10CFR 41.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00877Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage95 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Wed Jul 27 16:37:13 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 09:08:22 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Modified by mrasch at Wed Jul 27 14:10:33 CDT 20057.Modified by mrasch at Wed Jul 27 16:37:13 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:24:42 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 46 (1.0 Points)A LOCA occurred shortly after placing static inverter 1Y95 on its alternate powersupply.Fifteen minutes later Bus 16AB locked out.Based on this information, the reactor operator should conclude: A.1Y80 is receiving power from its alternate supply. B.1Y95 is receiving power from its normal supply. C.1Y88 is receiving power from its normal supply. D.Page96 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-L6200 Objective: 4.1; 4.2; 8; 9.1; 9.2; 10.1; 15KA

References:

1.262002 K4.01 Transfer from preferred power to alternate powersupplies [3.1/3.4]

References:

1.04-1-01-L62-1 Steps 3.4; 3.5 Attachment IIITrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.L62: Uninterruptible Power Supply SystemCategories:1.Systems2.Continuing Training1Y87 is receiving power from its alternate supply.Answer:CQuestionComments:1Y95 was placed on its alternate power supply and does not have anauto re-transfer feature therefore when 16AB bus is locked out the loadsfor 1Y95 are lost. 1Y88 is also a division 2 powered inverter however itwill remain on its DC Normal power supply when its alternate supply islost. 1Y87 is division 1 powered and would have had a momentary lossof its alternate power source during the LOCA signal but remain on itsnormal power supply. Tier 2 Group 1 This is a NEW question. 10 CFR41.4Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00878Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage97 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Task

References:

Question Last Revised By:MikeRasch at Thu Jul 28 10:52:29 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 09:17:16 CDT 20052.Modified by mrasch at Mon Jun 20 08:34:35 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 10:52:29 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:24:47 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 47 (1.0 Points)The plant DC system is in its normal lineup when a loss of Load Control Center 15BA6occursWhat effect will this have on Division 1 battery voltage and specific gravity? A.Both will remain stable. B.Both will lower. C.Voltage will remain constant and specific gravity will go down. D.Voltage will lower and specific gravity will remain constant.Answer:AQuestionComments:Answer A is correct because load is shared between chargers 1A4 and1A5, and either charger is rated to maintain battery parameters by itself.Answer B is incorrect because with load sharing, 1A5 will pick up loadPage98 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-L1100 Objective: 2; 7; 16KA

References:

1.263000 K4.01 Manual/ automatic transfers of control: Plant-Specific [3.1/3.4]

References:

1.04-1-01-L11-1 Attachment IIIA2.Tech Spec Bases B3.8.4TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.L11: Plant DC Electical SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Thu Jul 28 11:00:25 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 09:22:05 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/2005automatically. Answers C and D are incorrect because charger 1A5 is100% duty rated and will maintain battery parameters. Tier 2 Group 1This is a NEW question. 10CFR 41.4Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00879Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage99 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 4.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Modified by mrasch at Thu Jul 28 11:00:25 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:24:54 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 48 (1.0 Points)The plant is at 50% power due to grid disturbances.16AB bus is separated from the grid with DG12 carrying the bus.With respect to DG 12, if a valid LOCA were to occur in this electrical lineup, the DG 12output breaker would: A.open, load shedding would occur, and the output breaker would re-closeautomatically. No operator action is required. B.remain closed and load shedding would occur, and loads are sequenced on. Nooperator action is required. C.remain closed and load shedding would NOT occur. The operator would manuallysecure the unnecessary loads. D.open, load shedding would occur, and the operator would have to re-close theoutput breaker.Answer:BQuestionComments:With the Diesel carrying the bus by itself the diesel output breaker willremain closed but the Load Shedding and Sequencing system will shedthe loads on the bus resequencing the loads per the LOCA sequencingprocess. Therefore Answer B is the only correct answer. Tier 2 Group 1This is a NEW question. 10CFR 41.7Image

Reference:

NonePage100 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-R2100 Objective: 11; 14; 342.CourseID: GLP-OPS-P7500 Objective: 26KA

References:

1.264000 A2.10 LOCA [3.9/4.2]

References:

1.04-1-01-P75-1 ATT V page 52.M-1077B3.E-1109-244.E-1120-04TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.B21: Nuclear Boiler System2.P75: Div 1 and 2 Diesel Generator System3.R21: 4.16 KV AC Power SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 13:05:01 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 09:28:08 CDT 20052.Modified by mrasch at Mon Jun 20 09:04:22 CDT 2005Closed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00880Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page101 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 11:10:46 CDT 20058.Modified by tharrelso at Fri Jul 29 08:56:44 CDT 20059.Modified by mellis at Fri Jul 29 13:05:01 CDT 200510.Question Reviewed by mrasch at Thu Aug 04 07:29:05 CDT 200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 49 (1.0 Points)The plant is operating normally at 100% power when the normal supply breaker to bus16AB trips open. As expected, DG 12 starts and comes up to 450 RPM. However, theReady to Load light is not lit and the generator frequency and voltage meters are bothreading downscale. The output breaker did not close and the LSS failure alarm hasannunciated.Based on these indications the operator would suspect: A.generator frequency is low. B.generator voltage is low. C.the Parallel Reset Switch needs to be reset. D.the lockout relay for bus 16AB has tripped.Answer:BQuestionThe ready to load light is powered off the AC bus and will not light untilPage102 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-P7500 Objective: 27KA

References:

1.2640002.GENERIC 2.4.48 5 EDG/Diagnose EDG status from control room indications.[3.5/3.8]

References:

1.ARI 04-1-02-1H13-P864 2A-H1TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.P75: Div 1 and 2 Diesel Generator SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 13:09:20 CDT 2005Question History:Comments:the DG output breaker closes. Because the engine is at 450 RPM thegenerator is turning at the electrical equivalent to 60 Hz. However, thefrequency meter will not indicate a value if the generator has no voltageoutput. Therefore the immediate suspicion would be the generator fieldflash failed and the generator output voltage is zero. Tier 2 Group 1 Thisis a NEW question. 10CFR55.41 (7)Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00881Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:4Page103 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Created by mrasch at Mon Jun 13 09:37:34 CDT 20052.Modified by mrasch at Mon Jun 20 09:05:22 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by tharrelso at Fri Jul 29 08:04:47 CDT 20058.Modified by mellis at Fri Jul 29 13:09:20 CDT 20059.Question Reviewed by mrasch at Thu Aug 04 07:29:09 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 50 (1.0 Points)The plant is operating at 100% power when instrument air is lost to the AuxiliaryBuilding.Pressure is now approximately 60 psig, slowly lowering and the restoration efforts thusfar have been unsuccessful.If pressure continues to lower, the Auxiliary Building air bleed off valve (PV-F531) is setto automatically open at approximately: A.20 psig. B.30 psig. C.40 psig. D.50 psig.Answer:BQuestionComments:The referenced lesson plan states the bleed off valve will open atapproximately 30 psig, making Answer B the correct answer. Tier 2Group 1 This is a NEW question. 10 CFR 41.7Page104 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-P5300 Objective: 5.6; 14KA

References:

1.300000 K3.01 Containment air system [2.7/2.9]

References:

TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.P53: Instrument Air SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:17:02 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 09:42:58 CDT 20052.Modified by mrasch at Mon Jun 20 09:17:44 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/2005Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00882Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page105 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 7.Modified by mrasch at Thu Jul 28 17:24:47 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:25:05 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200510.Modified by mrasch at Tue Aug 09 11:17:02 CDT 200511.Question Reviewed by mellis at Tue Aug 09 11:59:54 CDT 2005Comments:EB QUESTION: 51 (1.0 Points)A break in the instrument air header in the Turbine Building has occurred.Unit 2 Instrument Air Compressor and Plant Air Dryer B are in service.Service Air Compressor A is in service.The resulting high air flow through the in service drying tower causes the desiccant toclog the tower.Dryer outlet pressure reaches 45 psig.What will occur as a result of this condition? A.Plant Air Dryer A will automatically align itself to Unit 2 Instrument Air Compressor. B.Unit 1 Instrument Air Compressor will automatically start and align through PlantAir Dryer A. C.Service Air Compressor A will automatically align through Plant Air Dryer A. D.Plant Air Dryer B will undergo an Executed Stop and its other tower will go intodrying mode.Answer:DQuestionComments:There is no automatic shifting or starts for the Instrument AirCompressors or Air Dryers. The in service air dryer will go through anexecuted stop. Therefore Answer D is correct. Tier 2 Group 1 This is aNEW question. 10CFR 41.4/41.10/43.5Page106 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-P5300 Objective: 3; 4; 23; 27; 31; 332.CourseID: GLP-OPS-ONEP Objective: 40KA

References:

1.300000 K6.13 Filters [2.8/2.3]2.300000 A2.01 Air dryer and filter malfunctions [2.9/2.8]

References:

1.M-1067G2.M-11263.M-1068D4.M-1067A5.ONEP 05-1-02-V-9 section 3.6; 3.8.1; 3.206.ARI 04-1-02-1H13-P870-7A-E3TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.P51: Plant Air System2.P52: Service Air System3.P53: Instrument Air SystemCategories:1.Off Normal Event Procedures2.Systems3.Continuing TrainingTask

References:

Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00883Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:4Page107 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Question Last Revised By:MikeRasch at Tue Aug 09 11:17:49 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 09:51:52 CDT 20052.Modified by mrasch at Mon Jun 20 05:29:03 CDT 20053.Modified by mrasch at Mon Jun 20 09:37:57 CDT 20054.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 14 10:46:22 CDT 20058.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20059.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200510.Modified by mrasch at Thu Jul 28 15:32:03 CDT 200511.Modified by mellis at Fri Jul 29 13:10:28 CDT 200512.Question Reviewed by mrasch at Thu Aug 04 07:29:11 CDT 200513.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200514.Modified by mrasch at Tue Aug 09 11:17:49 CDT 200515.Question Reviewed by mellis at Tue Aug 09 11:59:55 CDT 2005Comments:EB QUESTION: 52 (1.0 Points)The plant is in Mode 3 with RHR A in Shutdown Cooling mode.Reactor coolant temperature is 338°F.Reactor Water level and temperature were stable before the event.Which one of the following would be indicative of an RHR A Heat Exchanger tuberupture if a tube leak occurred? A.RHR A pump discharge pressure trending up B.Reactor water level trending up C.SSW A radiation monitor readings trending upPage108 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E1200 Objective: 4.2; 212.CourseID: GLP-OPS-P4100 Objective: 21KA

References:

1.400000 K1.04 Reactor coolant system, in order to determine source (s)of RCSleakage into CCWS [2.9/3.1]

References:

1.ARI 04-1-02-1H13-P601 18A-F62.SFD-1085-0013.SFD-1085-0024.SFD-1061CTrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training Program D.Reactor coolant temperature trending downAnswer:CQuestionComments:Coolant temperature 338°F equates to ~ 100 psig. Tube leakage wouldbe from RHR to SSW. Answer A is incorrect because a leak into SSWwould be less flow restriction, RHR flow would go up and dischargepressure would go down. Answer B is incorrect because RHR would beat a higher pressure than SSW, so RPV level would go down. Answer Cis CORRECT because RHR would leak into SSW. Reactor water is ofhigher activity than SSW, so rad monitor readings would go up. AnswerD is incorrect because less RHR flow to the reactor, which would be at ahigher temperature, would result. So, higher temperature IS indicative ofa tube leak. Tier 2 Group 1 This is a NEW question. 10CFR41.7/41.13/43.4Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00884Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page109 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Systems:1.D17: Process Radiation Monitoring System2.E12: Residual Heat Removal System3.P41: Standby Service Water SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:18:51 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 09:58:09 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Thu Jul 14 11:00:04 CDT 20056.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Modified by mrasch at Tue Jul 26 15:39:19 CDT 20059.Modified by mrasch at Wed Jul 27 07:50:34 CDT 200510.Question Reviewed by mellis at Thu Aug 04 07:25:11 CDT 200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200512.Modified by mrasch at Tue Aug 09 11:18:51 CDT 200513.Question Reviewed by mellis at Tue Aug 09 11:59:56 CDT 2005Comments:EB QUESTION: 53 (1.0 Points)The plant is at 100% power.The Turbine Building Cooling Water (TBCW) temperature control valve, P44-F513, failsclosed.Which one of the following will necessitate plant shutdown first, assuming Loss ofTBCW ONEP actions are performed, but P44-F513 remains closed?Page110 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-ONEP Objective: 1; 2; 44KA

References:

1.400000 A1.02 CCW temperature [2.8/2.8]

1.0 5-1-02-V-2TrainingPrograms

A.Main Turbine Lube Oil temperature B.Loss of Instrument Air Compressors C.Reactor Feed Pump Oil temperature D.Generator Seal Oil temperatureAnswer:DQuestionComments:This question is based on actual plant events and the changes made tothe ONEP for Loss of TBCW. At 100% power, seal oil temperature isnormally ~ 115°F. This is only 10°F margin to the limit of 125°F specifiedin plant procedures where plant shutdown is required. The ONEP liststemperatures in their expected order of priority. That sequence isexpected based on plant data for a universal degradation of TBCW heatremoval capacity. Seal oil is expected to reach its limit first based onplant and simulator data, given its normal operating temperature. That iswhy answer D is correct and answers A, B, and C are incorrect. Tier 2Group 1 This is a NEW question. 10CFR 41.4/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00885Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page111 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.N42: Seal Oil Syste2.P43: Turbine Building Cooling Water System3.P44: Plant Service Water SystemCategories:1.Off Normal Event Procedures2.Systems3.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 13:23:59 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 10:10:30 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Fri Jul 15 11:19:02 CDT 20056.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Modified by mellis at Fri Jul 29 13:23:59 CDT 20059.Question Reviewed by mrasch at Thu Aug 04 07:29:12 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 54 (1.0 Points)With reactor power above the high power setpoint, what is the purpose for the controlrod withdrawal limiter? A.Mitigates the effects from a control rod drop eventPage112 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C1102 Objective: 2; 6KA

References:

1.201005 K5.10 Rod withdrawal limiter: BWR-6 [3.2/3.3]

References:

1.Tech Spec Bases F3.3.2.1 Function 1aTrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems: B.Prevents inadvertently exceeding fuel preconditioning limits C.Prevents violating the Minimum Critical Power Ratio (MCPR) Safety Limit D.Ensures core exposure burn rates are evenly distributedAnswer:CQuestionComments:The basis for RWL as specifically stated in Tech Spec bases is toprevent exceeding the MCPR safety limit. That is why answer C iscorrect. Answers A, B, and D are not related to that limit, so they areincorrect. Tier 2 Group 2 This is a NEW question. 10CFR41.2/41.6/43.2Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00886Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage113 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.C11-2: Rod Control and Information SystemCategories:1.Systems2.Technical Specifications3.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 09:04:26 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 10:15:34 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Modified by tharrelso at Wed Jul 27 15:20:46 CDT 20057.Modified by tharrelso at Fri Jul 29 09:04:26 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:25:25 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 55 (1.0 Points)The plant has been performing a special core physics test that involves varying reactorrecirc pump flow. A human performance error by the lead test engineer leads to areactor transient.Currently: Reactor power is 28% and stable. Core Flow is 8% and stable. RPV pressure is 990 psig and stable. RPV water level is +36 inches and stable. The Turbine is in operation and the bypass valves are closed.Page114 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-TS001 Objective: 29KA

References:

1.GENERIC 2.4.11 Knowledge of abnormal condition procedures [3.4/3.6]2.202001

References:

1.Tech Spec 2.1.1TrainingPrograms:1.Reactor Operator Training Program One Reactor Recirc Pump is operating in slow speed.Based on this information, the crew should: A.Raise core flow by starting the second Reactor Recirc pump. B.Reduce reactor power to 25%. C.Raise core flow by opening the operating pump flow control valve. D.Insert all control rods.Answer:DQuestionComments:Power greater than 25% with core flow less than 10% is a safety limitviolation. This requires all control rods be inserted within two hours.Tier 2 Group 2 This is a NEW question. 10CFR55.41 (5)Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00887Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page115 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.B33: Reactor Recirculation SystemCategories:1.Systems2.Technical Specifications3.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:21:32 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 10:28:15 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Thu Jul 14 13:00:42 CDT 20056.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Modified by mrasch at Thu Jul 28 16:48:58 CDT 20059.Modified by tharrelso at Fri Jul 29 08:26:14 CDT 200510.Modified by tharrelso at Fri Jul 29 09:05:15 CDT 200511.Modified by mellis at Fri Jul 29 13:23:09 CDT 200512.Question Reviewed by mrasch at Thu Aug 04 07:29:13 CDT 200513.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200514.Modified by mrasch at Tue Aug 09 11:21:32 CDT 200515.Question Reviewed by mellis at Tue Aug 09 11:59:56 CDT 2005Comments:EB QUESTION: 56 (1.0 Points)The plant is operating at 100% rated thermal power with the Reactor Recirc FlowControl Valves at 68% valve position when a LOCA in the Drywell occurs.As a result, drywell pressure reached 4.5 psig.Page116 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-B3300 Objective: 27.5; 28.2; 28.3; 47KA

References:

1.202002 K1.01 Recirculation system [3.5/3.6]

References:

RPV water level reached - 80 inches.Five minutes later the Reactor Operator should observe: A.Recirc Pump breakers CB-3A/B are closed and Recirc Flow ControlValves are 20% open. B.Recirc Pump breakers CB-3A/B are closed and Recirc Flow ControlValves are 68% open. C.All the Recirc Pump breakers are open and Recirc Flow Control Valves are 20%open. D.All the Recirc Pump breakers are open and Recirc Flow Control Valves are 68%openAnswer:BQuestionComments:Drywell pressure will cause the Recirc FCV HPUs to trip preventingvalve motion on the FCVs (remain at 68%). The Drywell pressure willcause a reactor scram preventing the EOC actuation. RPV leveldropping will cause ATWS RPT to occur opening CB-1, 2, 4, 5 for bothRecirc pumps. Therefore Answer B is the correct answer. Tier 2 Group 2This is a NEW question. 10CFR 41.3/41.5/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00888Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page117 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.ARI 04-1-02-1H13-P680 3A-E32.Tech Spec Bases B3.3.4.1; B3.3.4.23.17-S-06-5 Att II pages 12; 13TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.B33: Reactor Recirculation SystemCategories:1.Off Normal Event Procedures2.Systems3.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:22:45 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 10:39:55 CDT 20052.Modified by mrasch at Wed Jun 15 12:21:58 CDT 20053.Modified by mrasch at Wed Jun 15 12:26:43 CDT 20054.Modified by mrasch at Wed Jun 15 12:32:08 CDT 20055.Modified by mrasch at Mon Jun 20 09:41:46 CDT 20056.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200510.Modified by mrasch at Thu Jul 28 14:39:03 CDT 200511.Modified by tharrelso at Fri Jul 29 09:07:58 CDT 200512.Modified by mellis at Fri Jul 29 13:24:51 CDT 200513.Question Reviewed by mrasch at Thu Aug 04 07:29:14 CDT 200514.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200515.Modified by mrasch at Tue Aug 09 11:22:45 CDT 200516.Question Reviewed by mellis at Tue Aug 09 11:59:57 CDT 2005Comments:Page118 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm EB QUESTION: 57 (1.0 Points)Objectives:Which one of the following conditions associated with the RWCU system will causea reduction in plant efficiency when operating at 100% power?Assume none of the listed actions cause a RWCU filter/deminerilizer isolation. A.A rise of 5°F in Component Cooling Water temperature at the inlet of the RWCUNon-Regenerative Heat Exchangers. B. The RWCU Filter/Demineralizer Bypass valve is leaking by a value of 100 gpm. C.An RWCU flow controller malfunction resulting in a reduction in total RWCU flowof 100 gpm. D.The RWCU Regenerative Heat Exchanger Bypass valve is throttle open 25%.Answer:DQuestionComments:Answer A is incorrect because a rise in CCW temperature results in arise in RWCU temperature that causes a less heat having to be made upby the core thus a rise in efficiency. Answer B is incorrect becausebypassing the Filter Demineralizers does not affect the heat loss ofRWCU such that there is no effect on plant efficiency. Answer C isincorrect because a reduction in RWCU flow means less cool water to bereheated by the core thus a rise in plant efficiency. Answer D isCORRECT because less cooling flow through the Regenerative heatexchanger results in less preheating of the return RWCU water to thereactor causing a loss in plant efficiency. Tier 2 Group 2 This is a NEWquestion. 10CFR 41.5/41.14Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00889aReview Status:ReviewedDifficulty:2: Comprehension or AnalysisPage119 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.CourseID: GLP-OPS-G3336 Objective: 4.2; 9.9; 18; 19KA

References:

1.204000 A4.06 System flow [3.0/2.9]

References:

1.SFD-1079TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.G33: Reactor Water CleanupCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Fri Jul 15 16:04:42 CDT 2005Question History:1.Created by mrasch at Fri Jul 15 11:57:02 CDT 20052.Created by mrasch at Fri Jul 15 11:57:02 CDT 2005 from parent QuestionID GGNS-NRC-008893.Modified by mrasch at Fri Jul 15 11:58:32 CDT 20054.Modified by mrasch at Fri Jul 15 16:04:42 CDT 20055.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 58 (1.0 Points)Prior to emergency depressurizing using the SRVs, EP-2, RPV Control, requires theoperating crew to verify Suppression Pool level greater than 10.5 feet.Page120 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-EP02A Objective: 72.CourseID: GLP-OPS-EP03 Objective: 3KA

References:

1.223001 K1.08 Relief/safety valves [3.6/3.8]

References:

1.PSTGThe basis for this verification is to ensure: A.there is sufficient heat capacity in the Suppression Pool to absorb the energy froma rapid depressurization. B.there is sufficient water to preclude the rise in Suppression Pool temperatureaffecting the operation of the ECCS injection pumps. C.there is sufficient backpressure to limit steam flowrates and potential damage tothe SRVs. D.the SRV discharge quenchers in the Suppression Pool are submerged.Answer:DQuestionComments:10.5 feet in the suppression pool is based on coverage of the SRVquenchers being covered to prevent the introduction of steam intoContainment which would present a challenge to the ContainmentSturcture. Therefore Answer D is the correct answer. Tier 2 Group 2This is a NEW question. 10CFR 41.3/41.4/41.7/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00890Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage121 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E22-2: Automatic Depressurization System2.M41-1: ContainmentCategories:1.Emergency Procedure Training2.Systems3.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Thu Jul 28 12:16:42 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 10:58:48 CDT 20052.Modified by mrasch at Mon Jun 20 09:44:41 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 12:16:42 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:25:43 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 59 (1.0 Points)The plant is operating at 100% power when the steam supply to the Offgas Preheatersis lost.What effect will this have on the Offgas System? A.Less hydrogen will be removed from the process stream by the catalyticPage122 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-N6465 Objective: 5.1; 14.3KA

References:

1.239001 K3.04 Offgas system [2.8/2.8]

References:

1.04-1-02-1H13-P845 1A-A2TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Programrecombiners. B.Fewer radioactive particulates will be removed from the process stream by thecatalytic recombiners. C.Offgas Pre-treatment radiation levels will rise. D.Offgas Post-treatment radiation levels will rise.Answer:AQuestionComments:The loss of steam to the Preheaters will result in a less efficiency in therecombiner component causing offgas hydrogen content to rise. TheOffgas recombiners do not affect radioactive materials in offgas.Therefore Answer A is correct. Tier 2 Group 2 This is a NEW question.10 CFR 41.4/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00891Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage123 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 4.Licensed Operator Requalification Training ProgramSystems:1.N11: Main Steam System2.N64: Offgas SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Thu Jul 28 12:36:49 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 11:10:19 CDT 20052.Modified by mrasch at Mon Jun 20 09:49:29 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by mrasch at Thu Jul 14 13:39:50 CDT 20057.Question Reviewed by mrasch at Fri Jul 15 16:31:39 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Modified by mrasch at Thu Jul 28 12:36:49 CDT 200510.Question Reviewed by mellis at Thu Aug 04 07:25:47 CDT 200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 60 (1.0 Points)The plant was at 100% power when a LOCA occurred.Severe Accident Procedures have been entered.The Control Room Supervisor has directed initiation of the Outboard Main SteamIsolation Valve Leakage Control System (MSIV LCS).Which one of the following describes the effect of Standby Gas Treatment System(SGTS) on the effluent of the MSIV LCS? A.Page124 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E3200 Objective: 12.3; 13.1; 13.2KA

References:

1.239003 K1.02 Standby gas treatment system: BWR-4,5,6(P-Spec) [2.9/3.0]

References:

Either Standby Gas Treatment System (SGTS) A or B will process most of theeffluent from either MSIV LCS subsystem. B.Effluent of the Outboard MSIV LCS is piped directly to Standby Gas TreatmentSystem (SGTS) A ducting, therefore SGTS A is the preferred system to operate. C.Simultaneous operation of both Inboard MSIV LCS and Outboard MSIV LCS isprohibited unless both Standby Gas Treatment Systems (SGTS) are in operation. D.Operation of the Outboard MSIV LCS in conjunction with Standby Gas TreatmentSystem (SGTS) B is preferred.Answer:AQuestionComments:Answer A is correct because MSIV LCS exhausts to auxiliary buildingcorridors on 119' elev. SGTS takes suction on these areas and maintainsnegative pressure in the auxiliary building, therefore essentially all MSIVLCS exhaust will be eventually processed by SGTS. Answer B isincorrect because MSIV LCS is not piped directly to SGTS, but only inthe vicinity of an intake to SGTS ductwork. Answer C is incorrectbecause simultaneous operation of both inboard and outboard MSIV LCSis always prohibited. Answer D is incorrect because SGTS A is preferredwith the outboard MSIV LCS due to the shorter associated transport time.Tier 2 Group 2 This is a NEW question. 10CFR 41.4/41.14/43.4Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00892Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page125 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.04-1-01-E32-1 steps 3.1; 3.2; 3.7; 3.8; 5.2.1cTrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E32: MSIV Leakage Control System2.T48: Standby Gas Treatment SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 16:11:37 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 11:23:14 CDT 20052.Modified by mrasch at Mon Jun 20 09:52:39 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by tharrelso at Fri Jul 29 09:12:42 CDT 20058.Modified by tharrelso at Fri Jul 29 09:19:47 CDT 20059.Question Reviewed by mellis at Thu Aug 04 07:25:53 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200511.Modified by mrasch at Tue Aug 09 11:23:52 CDT 200512.Question Reviewed by mellis at Tue Aug 09 11:59:58 CDT 200513.Modified by mrasch at Tue Aug 09 16:11:37 CDT 200514.Question Reviewed by mellis at Tue Aug 09 16:14:22 CDT 2005Comments:EB QUESTION: 61 (1.0 Points)Following a scram from rated conditions, pressure is being controlled with the MainPage126 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Bypass Valves.There is a steam leak in the RHR A room that CANNOT be isolated.EP-2 and EP-4 are being performed.The CRS has determined Emergency Depressurization is anticipated due to EP-4concerns.What action should be taken for this condition? A.Open 6 Safety Relief Valves to lower the pressure band to 450 psig to 650 psig. B.Throttle open Main Bypass Valves to reduce the driving head of the leak, stayingwithin Technical Specification cooldown rate limitations. C.Fully open the Main Bypass Valves using the Manual Bypass Jack to fullydepressurize the reactor. D.Lower Pressure Reference to 450 psig to reduce the driving head of the leak.Answer:CQuestionComments:The PSTGs state it is preferred to discharge steam to the maincondenser to limit the challenge to containment. No conditions exist thatrequire closing MSIVs. EP-2A gives guidance for using bypass valvesand maintaining the MSIVs open. Operations Philosophy disallows use oflow-low set operation of SRVs during ATWS conditions. EHC pumps canbe restarted from the control room since bus 14AE undervoltage lockoutsare reset. Answer A, B, and D are incorrect because EP-2A basesprefers using bypass valves and maintaining the MSIVs open. Answer Cis correct because it includes guidance for using bypass valves andmaintaining the MSIVs open. Tier 2 Group 2 This is a NEW question.10CFR 41.5/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00893Review Status:ReviewedPage127 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-EP02A Objective: 2; 52.CourseID: GLP-OPS-PROC Objective: 59.3KA

References:

1.GENERIC 2.4.6 Knowledge symptom based EOP mitigation strategies [3.1/4.0]2.241000

References:

1.PSTG B-6-25, 38, 41, 42, 432.PSTG B-14-9, 113.PSTG B-16-54.02-S-01-27 steps 6.1.6; 6.2.4; 6.6.8dTrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.B21: Nuclear Boiler SystemCategories:1.Administrative Requirements2.Emergency Procedure Training3.Systems4.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:24:55 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 11:35:20 CDT 20052.Modified by mrasch at Mon Jun 20 09:59:28 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by mrasch at Thu Jul 14 14:03:56 CDT 2005Difficulty:2: Comprehension or AnalysisDifficulty Rating:2Page128 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 7.Modified by mrasch at Thu Jul 14 14:21:37 CDT 20058.Modified by mrasch at Thu Jul 14 14:22:52 CDT 20059.Question Reviewed by mrasch at Fri Jul 15 16:31:40 CDT 200510.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200511.Modified by mrasch at Thu Jul 28 15:38:39 CDT 200512.Modified by tharrelso at Fri Jul 29 09:21:49 CDT 200513.Question Reviewed by mellis at Thu Aug 04 07:25:57 CDT 200514.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200515.Modified by mrasch at Tue Aug 09 11:24:55 CDT 200516.Question Reviewed by mellis at Tue Aug 09 11:59:59 CDT 2005Comments:EB QUESTION: 62 (1.0 Points)Which one of the following would automatically occur during Main Turbine roll up from400 rpm to 1800 rpm? A.Primary Water Circulating Pump will start. B.Turbine Turning Gear valves will close. C.Shaft Lift Oil Pump will stop. D.Generator Field Breaker will close.Answer:CQuestionComments:Answer A is incorrect because the Primary Water Ciruclating Pumpstops on turbine roll up. Answer B is incorrect because the Turning gearvalves close around 240 RPM and must be closed before 400 RPM.Answer D is incorrect because the Generator Field Breaker is onlyenabled at 1710 RPM but must be manually closed from the controlroom. Answer C is CORRECT because the Shaft Lift Oil Pump willautomatically stop at about 540 RPM. Tier 2 Group 2 This is a NEWquestion. 10CFR 41.4/41.10Image

Reference:

NoneClosed Reference QuestionPage129 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-N4151 Objective: 8.22.CourseID: GLP-OPS-N4300 Objective: 103.CourseID: GLP-OPS-N3402 Objective: 6.3: 6.44.CourseID: GLP-OPS-IOI01 Objective: 33KA

References:

1.245000 A3.02 Turbine roll to rated speed [2.8/2.8]

References:

1.03-1-01-2 sections 7.1.4; 7.1.5; 7.22.04-1-01-N40-1 section 3.12TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.N30: Main Turbine2.N34: Main Turbine Lube Oil System3.N40: Main Generator4.N43: Primary Water SystemCategories:1.Integrated Plant Operations2.Systems3.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:25:33 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 11:43:21 CDT 20052.Modified by mrasch at Mon Jun 20 10:02:26 CDT 2005Handout Not Required with ExamQuestionID:GGNS-NRC-00894Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page130 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Tue Jul 26 17:41:15 CDT 20058.Modified by mrasch at Wed Jul 27 09:39:56 CDT 20059.Modified by tharrelso at Fri Jul 29 09:23:50 CDT 200510.Question Reviewed by mellis at Thu Aug 04 07:26:01 CDT 200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200512.Modified by mrasch at Tue Aug 09 11:25:33 CDT 200513.Question Reviewed by mellis at Tue Aug 09 12:00:00 CDT 2005Comments:EB QUESTION: 63 (1.0 Points)The plant is operating at 100% power when Condensate Pump A spuriously trips.What is the preferred action for this condition? A.Scram the reactor due to inevitable Reactor Feed Pump trip on low suction flow. B.Initiate Reactor Core Isolation Cooling due to the inevitable Reactor Feed Pumptrip on low suction pressure. C.Re-start Condensate Pump A in accordance with the SOI to prevent inevitableCondensate Booster Pump trips on low suction pressure. D.Lower reactor power to raise margin to Reactor Feed Pump trip setpoints, stayingwithin power to flow limitations.Answer:DQuestionComments:A loss of a single Condensate pump will cause a lower suction pressurefor the Condensate Booster Pumps and Reactor Feed pumps but willnot cause a reactor scram. GGNS is designed to operate full power witha single condensate and condesate booster pump out of service. Tier 2Page131 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-N1900 Objective: 2; 25KA

References:

1.259001 A2.03 Loss of condensate pump(s) [3.6/3.6]

References:

1.05-1-02-V-72.04-1-02-1H13-P680-1A-A1TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.N19: Condensate System2.N21: Feedwater SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Thu Jul 28 15:55:48 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 11:48:54 CDT 20052.Modified by mrasch at Mon Jun 20 10:07:44 CDT 20053.Modified by mrasch at Mon Jun 20 10:12:56 CDT 20054.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 2005Group 2 This is a NEW question. 10CFR 41.4/41.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00895Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage132 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 5.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20058.Modified by mrasch at Thu Jul 28 15:55:48 CDT 20059.Question Reviewed by mellis at Thu Aug 04 07:26:07 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 64 (1.0 Points)A LOCA has occurred causing Drywell pressure to rise to 4.2 psig.Which one of the following describes a possible flowpath to transfer the contents of theContainment Floor Drain sump under these conditions? A.The sump cannot be transferred. B.The sump will automatically transfer directly to the Floor Drain Collector tank. C.The sump will automatically transfer to the Auxiliary Building Floor Drain Transfertank then to the Floor Drain Collector tank. D.The sump can be transferred to the Suppression Pool via AUX BLDG EQ/FL DRPMPBK TO SUPP POOL valves, P45-F273 and P45-F274.Answer:AQuestionComments:With Drywell pressure at 4.2 psig, an NSSSS isolation signal exists. Thiscloses the Primary and Secondary Containment isolation valves for P45and cannot be overridden except P45-F273 and F274 which may beoverridden 30 seconds after the isolation signal. Answer A is CORRECTbecause the flowpath from the containment cannot be opened with theisolation signal present. Answer B is INCORRECT because theContainment Floor Drain Sump must pass through the Auxiliary BuildingTransfer tank before exiting the Auxiliary Building to Radwaste. Answer Cis INCORRECT because the flowpath is correct, however the primary andPage133 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-P4500 Objective: 9, 10.9KA

References:

1.268000 K1.04 Reactor building floor drains: Plant-Specific [2.7/2.9]

References:

1.M1094TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.P45: Floor and Equipment Drain SystemCategories:1.Administrative Requirements2.Systems3.Technical Specifications4.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 09:24:20 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 11:56:48 CDT 20052.Modified by mrasch at Mon Jun 20 14:01:47 CDT 2005secondary containment isolation valves are closed and cannot bereopened. Answer D is INCORRECT because the Primary Containmentisolation valves are closed and cannot be reopened. Tier 2 Group 2 Thisis a NEW question. 10CFR 41.4/41.10/43.2/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00896Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage134 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by mrasch at Mon Jul 11 14:19:19 CDT 20057.Modified by mrasch at Thu Jul 14 14:56:24 CDT 20058.Modified by mrasch at Thu Jul 14 16:37:19 CDT 20059.Question Reviewed by mrasch at Fri Jul 15 16:31:40 CDT 200510.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200511.Modified by tharrelso at Wed Jul 27 15:36:45 CDT 200512.Modified by tharrelso at Fri Jul 29 09:24:20 CDT 200513.Question Reviewed by mellis at Thu Aug 04 07:26:14 CDT 200514.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 65 (1.0 Points)Which one of the following identifies where the Auxiliary Building Steam Tunnel FloorDrains are directly routed? A.The RHR A room Floor Drain sump B.The Auxiliary Building Floor Drain Transfer tank C.The Containment Steam Tunnel Floor Drain sump D.The RCIC room Floor Drain sumpAnswer:DQuestionComments:Answer A, B, and C are incorrect because the Auxiliary Building SteamTunnel Floor Drains are hard piped directly to the RCIC room FloorDrain sump, which makes answer D correct. Tier 2 Group 2 This is aPage135 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-P4500 Objective: 3.13KA

References:

1.290001 K4.03 Fluid leakage collection [2.8/2.9]

References:

1.M1098ATrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.P45: Floor and Equipment Drain SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 09:25:47 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 12:01:57 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamNEW question. 10 CFR 41.9Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00897Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage136 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Date: 08/12/20056.Modified by tharrelso at Wed Jul 27 16:51:23 CDT 20057.Modified by tharrelso at Wed Jul 27 16:53:58 CDT 20058.Modified by tharrelso at Thu Jul 28 06:16:34 CDT 20059.Modified by tharrelso at Fri Jul 29 09:25:47 CDT 200510.Question Reviewed by mellis at Thu Aug 04 07:26:21 CDT 200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 66 (1.0 Points)The plant was at 95% power when Reactor Feed Pump B tripped.Operation is in the Restricted Region of the Power-Flow Map.Which of the following indications on 1H13-P680 would allow continued operation ofthe unit? A.APRM oscillations of 14% peak-to-peak. B.Annunciators PBDS A INOP (5A-A6) and PBDS B INOP (7A-A6) in due to failedcards. C.PBDS A has generated a HI-HI DECAY RATIO (5A-A10) only and PBDS B hasgenerated a HI DECAY RATIO (7A-C4) only. D.Alarm PBDS B HI-HI DECAY RATIO (7A-C6) and PDS computer point PBDSChannel B Highest Counts reads 12 counts.Answer:CQuestionComments:PBDS is normally a computer indication of APRM oscillations. Answer Ais INCORRECT because APRM oscillations of > 10% is an indication ofthe onset of instability requiring plant shutdown. Answer B isINCORRECT because with operation in the Restricted Region of thePower to Flow Map and both PBDS channels INOP action with FCTR inNORMAL, calls for immediately placing the reactor mode switch toshutdown. Answer C is CORRECT because Channel A has a HI-HIwithout a HI which indicates a bad channel and Channel B only has a HIalarm. These conditions allow continued operation but require actions tobe taken to immediately exit the region. Answer D is INCORRECTbecause both the HI-HI alarm on Channel B and indication of 12 countson the computer point are indications of the Onset of Instability requiringPage137 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C5106 Objective: 1; 14.1; 14.2; 23.2KA

References:

1.GENERIC 2.1.19 Ability to use plant computer to obtain and evaluate parametricinformation on [3.0/3.0]2.GENERIC 2.4.7 Knowledge of event based EOP mitigation strategies [3.1/3.8]

References:

1.Power to Flow Map (Restricted Region)2.05-1-02-III-3 Steps 2.1; 4.93.ARI 04-1-02-1H13-P680 5A-A6, A10; 7A-A6, C4, C6TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.B33: Reactor Recirculation System2.C51-3: Local Power Range Nuclear Instrumentation System3.C51-5: Average Power Range Nuclear Instrumentation System4.C51-6: Period Based Detection SystemCategories:1.Off Normal Event Procedures2.Systems3.Continuing TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 13:38:58 CDT 2005a reactor scram. Tier 3 This is a NEW question. 10CFR41.1/41.5/41.6/41.10/43.5/43.6Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00898Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page138 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Question History:1.Created by mrasch at Mon Jun 13 12:14:50 CDT 20052.Modified by mrasch at Mon Jun 20 10:18:27 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by tharrelso at Fri Jul 29 09:27:45 CDT 20058.Modified by mellis at Fri Jul 29 13:38:58 CDT 20059.Question Reviewed by mrasch at Thu Aug 04 07:29:15 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 67 (1.0 Points)The plant is operating at 100% power when the System Load Dispatcher calls andrequests Grand Gulf change their generator power factor from1.0 to 0.975 capacitive tohelp with grid stability.Generator hydrogen pressure is 60 psig.Given NO generator operating limits will be exceeded, the operating crew should:Generator V Curves are provided. A.depress the voltage regulator LOWER pushbutton until approximately -300MVARs are being carried by the generator. B.depress the voltage regulator RAISE pushbutton until approximately 300 MVARsare being carried by the generator. C.depress the voltage regulator RAISE pushbutton until approximately 250 MVARsare being carried by the generator. D.Page139 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-N4151 Objective: 7; 13KA

References:

1.GENERIC 2.1.25 Ability to obtain and interpret station reference materials such asgraphs / [2.8/3.1]

References:

1.04-1-01-N40-1 step 3.82.03-1-01-2 Figure 2TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.N40: Main Generator2.N41: Generator AuxiliariesCategories:1.Integrated Plant Operations2.Systemsdepress the voltage regulator LOWER pushbutton until approximately -250MVARs are being carried by the generator.Answer:DQuestionComments:From a unity power factor to 0.975 capacitive requires the generator tobe under excited so the voltage regulator must be lowered. The limit onthe GGNS Generator reactive load is 250 MVARs. Therefore Answer Dis the Correct answer. Tier 3 This is a NEW question. 10CFR41.4/41.10/43.5Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00899Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage140 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Task

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 09:28:32 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 12:21:49 CDT 20052.Modified by mrasch at Mon Jun 20 10:23:16 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 13:03:59 CDT 20058.Modified by tharrelso at Fri Jul 29 09:28:32 CDT 20059.Question Reviewed by mellis at Thu Aug 04 07:26:33 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 68 (1.0 Points)An operator is restoring a red tag clearance on a manual valve.The required position for the valve is THROTTLED, 1 1/2 TURNS OPEN.There are NO red tie wraps available at GGNS.Which one of the following is an acceptable method to lock the valve in the requiredposition? A.Cable with a padlock B.Yellow valve seal C.Black tie wrapPage141 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 49.14KA

References:

1.Generic 2.1.29: 3.4/3.3

References:

1.02-S-01-2 Att IIITrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:Categories: D.Blue tie wrapAnswer:AQuestionComments:Yellow plastic seals are used on Fire Protection Valves. Blue Tie-Wrapsare used as a locking devices for valves other than throttled valves.Lockwire is used by I&C for sealing instrument valves in position. Chainsand/or cables with padlocks are acceptable alternatives to Red and BlueTie Wraps in the event the appropriate tie-wraps are unavailable. This isper Attachment III of 02-S-01-2 Component Position Verification. Basedon this the ONLY CORRECT answer is answer A. This question isMODIFIED. Stem changed to Locked Throttled Valve from LockedClosed Valve. Answer Lockwire was replaced with Black Tie Wrap whichis normally used when attaching red tags to components. Originalquestion used RO Audit Examination December 2000 Question # 82 IDWRIA082. Similar question used RO NRC Examination April 2000Question # 89 ID WRI289. Tier 3 10CFR 41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00237aReview Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage142 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Administrative RequirementsTask

References:

Question Last Revised By:MikeRasch at Thu Jun 09 09:25:39 CDT 2005Question History:1.Created by tharrelso at Tue May 10 13:33:43 CDT 20052.Created by tharrelso at Tue May 10 13:33:43 CDT 2005 from parent QuestionIDGGNS-NRC-002373.Modified by mrasch at Thu May 12 15:52:49 CDT 20054.Question Reviewed by mellis at Tue May 31 14:56:58 CDT 20055.Modified by mrasch at Thu Jun 09 08:53:48 CDT 20056.Modified by mrasch at Thu Jun 09 09:25:39 CDT 20057.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200510.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 69 (1.0 Points)Which one of the following situations presents an operability concern regardingIntermediate Range Monitors? A.In Mode 2 preparing to transfer to RUN, all IRMs are fully inserted and indicate 15to 20 on range 10 while all APRMs indicate 6% to 8% power. B.During startup in Mode 2 with IRMs fully inserted, all IRMs indicate approximately10 on range 1 when all SRMs, fully inserted, indicate approximately 2 x 104 cps. C.During a "soft" shutdown with IRMs fully inserted, preparing to go from Mode 1 toMode 2, APRMS indicate 4% power while the highest reading IRM indicates 110on range 10 and the lowest IRM indicates 30 on range 10.Page143 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C5102 Objective: 9; 10; 152.CourseID: GLP-OPS-IOI01 Objective: 20KA

References:

1.GENERIC 2.2.1 Ability to perform pre-startup procedures for the facility / includingoperating those [3.7/3.6]

References:

1.06-OP-1000-D-00012.03-1-01-1 sections 5.28, 6.2.163.03-1-01-3 sections 5.4.7, 5.4.8, 5.4.9 caution step 5.74.TS Bases 3.3.1.1 SR3.3.1.1.5 and 3.3.1.1.6TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program D.In Mode 2 with IRMs fully inserted, the highest reading IRM indicates 70 on range8 and the lowest IRM indicates 20 on range 7.Answer:CQuestionComments:Answer A is INCORRECT because overlap of IRMs to APRMs is ofconcern because system design will initiate rod blocks if adequateoverlap is not maintained during power increases. Answer B isINCORRECT because proper overlap is being observed between SRMsand IRMs on plant startup. Answer C is CORRECT because APRMS arereading 4% with at least one IRM >108/125, this does not meet theoverlap requirements of 03-1-01-3 section 5.4.7 of Attachment I. The IRMchannel check is within a factor of 4 (limit is 10). Answer D isINCORRECT because the IRM channel check is within a factor of 4 (limitis 10). 20 on range 7 is 20 on range 8. Even ranges are just anexpansion of the previous odd range. Tier 3 This is a NEW question.10CFR 41.2/41.6/43.2Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00900Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page144 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.Licensed Operator Requalification Training ProgramSystems:1.C51-2: Intermediate Range Nuclear Instrumentation SystemCategories:1.Integrated Plant Operations2.Systems3.Technical Specifications4.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Mon Aug 01 16:03:50 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 12:29:04 CDT 20052.Modified by mrasch at Mon Jun 20 10:27:49 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Mon Aug 01 16:03:50 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:26:38 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 70 (1.0 Points)The plant is in Mode 5.Core Alterations are in progress.High Pressure Core Spray (HPCS) is tagged out of service to perform an internalinspection of HPCS TESTABLE CHK VLV E22-F005.Fuel Pool Cooling and Clean Up (FPCC) pumps A and B are in service.NO FPCC Filter Demin is in service.Page145 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm RHR B is in Shutdown Cooling, returning to the RPV via RHR B SHUTDN CLG RTNTO FW valve E12-F053B.Reactor Water Clean Up (RWCU) pump A is tagged out of service for breakerpreventive maintenance, only.Which one of the following activities would require notification by control roompersonnel to Refuel Floor supervision? A.Clearing red tags and restoring HPCS to standby. B.Returning RWCU pump A to standby. C.Placing Standby Service Water B in service to FPCC heat exchangers. D.Making necessary adjustments to RHR HX B OUTLT VLV E12-F003B to maintainconstant reactor coolant temperature.Answer:AQuestionComments:Answer A is CORRECT because there is a potential of an air bubblerising into the reactor could cause problems on the Refuel floor. AnswerB is INCORRECT because this realignment of RWCU would not changeflows to the Reactor. Answer C is INCORRECT because this will onlyalter the cooling medium for FPCCU not the actual flow of the system.Answer D is INCORRECT because this is only changing the amount offlow from Shutdown Cooling not swapping the system providingshutdown cooling. Tier 3 This is a NEW question. 10CFR41.10/43.5/43.6/43.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00901Page146 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-IOI05 Objective: 2.32.CourseID: GLP-OPS-PROC Objective: 8.30KA

References:

1.GENERIC 2.2.30 Knowledge of RO duties in the control room during fuel handlingsuch as alarms [3.5/3.3]

References:

1.01-S-06-2 step 6.7.292.03-1-01-5 steps 2.24; 2.34; 2.35TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.E12: Residual Heat Removal System2.E22-1: High Pressure Core Spray System3.G33: Reactor Water Cleanup4.G41: Fuel Pool Cooling and Cleanup System5.P41: Standby Service Water SystemCategories:1.Administrative Requirements2.Integrated Plant Operations3.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Fri Jul 29 09:40:46 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 12:36:19 CDT 20052.Modified by mrasch at Mon Jun 20 10:29:40 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/2005Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage147 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 5.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by tharrelso at Fri Jul 29 09:32:42 CDT 20058.Modified by tharrelso at Fri Jul 29 09:40:46 CDT 20059.Question Reviewed by mellis at Thu Aug 04 07:26:46 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 71 (1.0 Points)The Auxiliary Building Operator must enter a High Radiation Area to hang red tags.The general area dose rate is 50% of the maximum dose rate that could beexperienced for a High Radiation Area.The operator is male, 30 years old, and has accumulated 200 mrem year-to-date TotalEffective Dose Equivalent (TEDE).His lifetime TEDE is 1200 mrem.NO extension of the operator's dose limit will be granted.Which one of the following times is the longest the operator could stay in the generalarea dose rate for the High Radiation Area without exceeding the administrative doselimit for TEDE? A.1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B.3.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> C.4.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D.9.6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sAnswer:BQuestion50% of the maximum radiation dose rate of a High Radiation Area of 999Page148 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: ELP-GET-RWT Objective: RWT30; 32; 43; 44KA

References:

1.GENERIC 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation controlrequirements [2.6/3.0]

References:

1.01-S-08-2 section 6.5.1d2.NMM ENS-RP-201 section 5.2.3.1TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:Categories:1.Administrative Requirements2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Mon Jun 20 10:30:45 CDT 2005Question History:Comments:mrem/hr is 499 mrem/hr. Worker's dose margin to the annualadministrative limit is 1800 mrem based on 200mrem already receivedand a 2000 mrem per year administrative TEDE limit. Entry into a499mrem/hr field with an 1800 mrem maximum dose gives a stay time of3.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This is Answer B. Answer A is based on 1000 mrem/hr.Answers C and D are INCORRECT because they are based on the NRCTEDE Limit of 5 Rem/Yr. Tier 3 This is a NEW question. 10CFR41.10/41.12/43.4/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00902Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage149 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Created by mrasch at Mon Jun 13 12:41:38 CDT 20052.Modified by mrasch at Mon Jun 20 10:30:45 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 72 (1.0 Points)A Site Area Emergency has been declared.You must leave the control room to perform operations at an Alternate ShutdownPanel.For this situation, your administrative annual dose limit is automatically extended to: A.2.5 Rem B.4.5 Rem C.5 Rem D.10 RemAnswer:CQuestionComments:Emergency Support personnel are administratively extended to thefederal limits of 10 CFR20 at the declaration of an Alert or above. Thefederal limit is 5 Rem TEDE per year Based on the above discussionAnswer C is the only CORRECT answer. Tier 3 This is a NEW question.10CFR 41.10/41.12/43.4/43.5Image

Reference:

NonePage150 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: ELP-GET-RWT Objective: RWT36KA

References:

1.GENERIC 2.3.4 Knowledge of radiation exposure limits and contamination control /including [2.5/3.1]

References:

1.10-S-01-17 section 6.8.1TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:Categories:1.Administrative Requirements2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Thu Jul 28 17:09:15 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 12:50:27 CDT 20052.Modified by mrasch at Mon Jun 20 10:31:56 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by mrasch at Thu Jul 14 15:05:11 CDT 20057.Question Reviewed by mrasch at Fri Jul 15 16:31:40 CDT 20058.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00903Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage151 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Date: 08/12/20059.Modified by mrasch at Thu Jul 28 17:09:15 CDT 200510.Question Reviewed by mellis at Thu Aug 04 07:26:52 CDT 200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 73 (1.0 Points)Objectives:An ATWS is in progress.Emergency Depressurization has just commenced.The Control Room Supervisor has directed you to inject with 2 mlbm/hr usingCondensate through the Start-Up Level Control Valve when reactor pressure lowers tothe Minimum Steam Cooling Pressure (MSCP).What is the basis for initializing injection at the rate of 2 mlbm/hr? A.To prevent damage to reactor internals due to thermal shock. B.To prevent exceeding DP restrictions for the Start-Up Level Control Valve. C.To prevent substantial fuel damage due to large power excursions. D.To prevent exceeding the upper end of the allowed level band.Answer:CQuestionComments:EP-2A Caution 5. Answer C is the only correct answer. Tier 3 This isa NEW question. 10 CFR 41.7/41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00904Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page152 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.CourseID: GG-1-LP-RO-EP02A Objective: 2; 9KA

References:

1.GENERIC 2.4.20 Knowledge of operational implications of EOP warnings / cautions /and notes [3.3/4.0]

References:

1.05-S-01-EP-2 Caution 5TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:1.B21: Nuclear Boiler SystemCategories:1.Administrative Requirements2.Emergency Procedure Training3.Systems4.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 16:12:35 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 12:55:42 CDT 20052.Modified by mrasch at Mon Jun 20 10:35:55 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by mrasch at Thu Jul 28 16:11:24 CDT 20058.Modified by mrasch at Thu Jul 28 17:33:52 CDT 20059.Question Reviewed by mellis at Thu Aug 04 07:26:57 CDT 200510.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200511.Modified by mrasch at Tue Aug 09 11:27:46 CDT 2005Page153 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 12.Question Reviewed by mellis at Tue Aug 09 12:00:01 CDT 200513.Modified by mrasch at Tue Aug 09 16:12:35 CDT 200514.Question Reviewed by mellis at Tue Aug 09 16:14:23 CDT 2005Comments:EB QUESTION: 74 (1.0 Points)Both Standby Gas Treatment systems have automatically started because of a smallbreak LOCA in the drywell.The Roving operator was verifying that the systems started properly when the followingannunciators alarmed. SGTS FLTR TR A CHAR TEMP HI SGTS FLTR TR A CHAR TEMP HI-HIWhat immediate action should the operator take in response to these alarms afterplacing handswitch SGTS FLTR TR A HTR to the OFF position? A.Dispatch a building operator to open the exhaust filter train fan A breaker. B.Secure SGTS train A per the T48 SOI. C.Dispatch a building operator to start the motor driven fire pump. D.Secure both SGTS trains per the T48 SOI.Answer:AQuestionComments:The SGTS automatically started due to exceeding a NSSSS LOCAsignal setpoint and cannot be secured until the NSSSS signal is clearedand reset. Therefore answers B and D are incorrect. Answer A is correctbecause, per the ARI, the exhaust filter train fan for the affected systemis opened to minimize the amount of air available to feed the fire. AnswerC is incorrect because the motor driven fire should automatically startonce the filter train deluge system is activated. Tier 3 This is a NEWquestion. 10 CFR 41.7Page154 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-P6400 Objective: 3.8KA

References:

1.GENERIC 2.4.25 Knowledge of fire protection procedures [2.9/3.4]

References:

1.10-S-03-22.04-1-02-1H13-P870-2A-C23.04-1-02-1H13-P870-2A-B2TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:1.P64: Fire Water Protection System2.T48: Standby Gas Treatment SystemCategories:1.Systems2.Continuing TrainingTask

References:

Question Last Revised By:Tommy Harrelson at Thu Jul 28 06:43:33 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 13:00:50 CDT 20052.Modified by mrasch at Mon Jun 20 10:38:38 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamImage

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00905Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage155 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Date: 08/12/20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20057.Modified by tharrelso at Thu Jul 28 06:43:33 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:27:05 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 75 (1.0 Points)Objectives:Which of the following is the preferred back-up method for notifications to state andlocal agencies when the Operational Hot Line (OHL) is inoperative duringimplementation of the Emergency Plan? A.UHF radio B.Satellite telephone C.Commercial telephone D.Entergy fiber optic linesAnswer:CQuestionComments:Per 10-S-01-6 all of the above are backup communications but per6.3.1 lists in order of use the backup communications listingCommercial Telephone as the first method. Therefore answer C isCORRECT. Tier 3 This is a NEW question. 10CFR 41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00906Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryPage156 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.CourseID: GLP-EP-EPT6 Objective: 3KA

References:

1.GENERIC 2.4.43 Knowledge of emergency communications systems and techniques[2.8/3.5]

References:

1.10-S-01-6 section 6.3TrainingPrograms:1.Nonlicensed Operator Training Program2.Reactor Operator Training Program3.Senior Reactor Operator Training Program4.Licensed Operator Requalification Training ProgramSystems:Categories:1.Emergency Plan Training2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Mon Jun 13 13:04:50 CDT 2005Question History:1.Created by mrasch at Mon Jun 13 13:04:50 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Question used on ExamName: NRC August 2005 - 1 ExamID: NRC-082005-1 ExamDate: 08/12/20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/2005Comments:EB QUESTION: 76 (1.0 Points)The plant was operating at 100% power in a normal configuration with NO inoperableequipment.Page157 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Then, at 0800 this morning, the Division 1 125 VDC battery was declared inoperableand removed from service to allow maintenance to work on an emergency batteryissue.To perform the battery work, both Division 1 battery chargers were also tagged out.Later in the day, at 1500, a problem was discovered on the HPCS pump breaker andHPCS was declared inoperable.In this situation, the plant must enter Mode 3, NO later than:Tech Specs are provided. A.2000 B.2100 C.2200 D.0400 (the following day)Answer:CQuestionComments:When Division 1 125 VDC was declared inoperable and forced an entryinto TS 3.8.4C giving 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore the DC system. When the 2hours expire, this forces an entry into 3.8.4.E and this gives the crew 12hours to get into Mode 3. Therefore based on this they have 14 hour1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />sfrom 0800 or 2200. When HPCS is declared inoperable this forces anentry into 3.0.3 per TS 3.5.1. This requires the plant to be in mode 3within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> of discovery (0400). distractor A is based on entering3.8.4.E immediately and not taking into consideration the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> allowedto meet the LCO. While this is permitted, it does not change the actualTS due date. Therefore C is the only correct answer because it is themost limiting LCO. Tier 1 Group 1 This is a NEW question. 10CFR 43.2Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00908Review Status:ReviewedPage158 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-TS001 Objective: 34KA

References:

1.295004 AA2.02: 3.5/3.9; AK2.03: 3.3/3.3

References:

1.Tech Specs 3.8.4; 3.5.3; 3.5.1TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:Categories:1.Technical SpecificationsTask

References:

Question Last Revised By:MikeRasch at Thu Aug 04 08:11:56 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:05:36 CDT 20052.Modified by jbell at Thu Jun 16 14:06:58 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Fri Jul 29 06:34:25 CDT 20056.Question Reviewed by mellis at Thu Aug 04 07:27:17 CDT 20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Modified by mrasch at Thu Aug 04 08:11:56 CDT 20059.Question Reviewed by mellis at Thu Aug 04 08:17:51 CDT 200510.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 77 (1.0 Points)Difficulty:2: Comprehension or AnalysisDifficulty Rating:2Page159 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: Rad Worker Training, ELP-GET-RWT Objective: RWT36KA

References:

The plant was at 100% power when a steam line break occurred in the TurbineBuilding.A site evacuation was conducted, however, one mechanic who was logged into theTurbine Building at the time of the event could NOT be accounted for.NO one on the search and rescue team would voluntarily accept the assignment.What is the highest administrative dose limit extension that may be approved for NON-Volunteers by the Emergency Director if he believes it to be for a life saving activity? A.5 Rem B.10 Rem C.25 Rem D.50 remAnswer:CQuestionComments:During an emergency the Emergency Director /Offsite EmergencyCoordinator may authorize extensions of dosse limits based on thesituation. Further dose limmit extensions are applicable only tovolunteers. Up to 25 Rem may be authorized for NON-volunteers ofsearch and rescue teams and repair teams. The highest dose limitextension for a Non-Volunteer is 25 Rem for protection of populations orsaving a life. Protection of property is only authorized up to 10 Rem.Greater than 25 Rem is strictly voluntary for protection of populations andsaving a life. Based on the above discussion Answer C is the onlyCORRECT answer. Tier 1 Group 1 This is a NEW question.10CFR41.10/41.12/43.4/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00909Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page160 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.GENERIC 2.3.4 Knowledge of radiation exposure limits and contamination control /including [2.5/3.1]2.295005

References:

1.10-S-01-17 Section 6.1TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:Categories:1.Administrative Requirements2.Emergency Plan TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:30:49 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:09:50 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Mon Jul 18 16:30:24 CDT 20055.Modified by mrasch at Wed Jul 27 15:03:23 CDT 20056.Question Reviewed by mellis at Thu Aug 04 07:27:22 CDT 20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Modified by mrasch at Thu Aug 04 08:12:40 CDT 20059.Question Reviewed by mellis at Thu Aug 04 08:17:53 CDT 200510.Modified by mrasch at Tue Aug 09 11:30:49 CDT 200511.Question Reviewed by mellis at Tue Aug 09 12:00:02 CDT 200512.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 78 (1.0 Points)The plant is in an ATWS.Page161 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP03 Objective: 2, 3, 6KA

References:

1.295026 EA2.03: 3.9/4.02.Generic 2.4.6: 3.1/4.0Main Steam Isolation Valves are closed.Suppression Pool level is 19.2 feet.Average Suppression Pool temperature is 160 degrees F.Reactor pressure is 1000 psig.What is the highest priority action for this condition?Figure 1 HCTL is provided. A.Enter EP-3 and place both loops of pool cooling in service B.Enter EP-2A emergency depressurization leg C.Enter EP-3 and inhibit Suppression Pool Make-Up D.Enter EP-2A and lower the pressure control band to 450-600 psigAnswer:BQuestionComments:With Suppression Pool temperature at 160 degrees F and SuppressionPool Level at 19.2 feet and the Reactor at 1000 psig conditions areUNSAFE on the HCTL curve requiring emergency depressurization perEP-2A. Tier 1 Group 1 This is a NEW question. 10 CFR 41.10/43.5Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00910Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page162 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm

References:

1.HCTL curveTrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:Categories:1.Emergency Procedure TrainingTask

References:

Question Last Revised By:MikeRasch at Thu Aug 04 08:13:17 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:12:42 CDT 20052.Modified by jbell at Thu Jun 16 15:38:32 CDT 20053.Modified by mrasch at Mon Jun 20 10:44:37 CDT 20054.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by mrasch at Fri Jul 29 06:45:02 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:27:27 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Modified by mrasch at Thu Aug 04 08:13:17 CDT 200510.Question Reviewed by mellis at Thu Aug 04 08:17:54 CDT 200511.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 79 (1.0 Points)The basis for the Technical Specification limit for Average PrimaryContainment temperature is: A.To maintain containment air temperature below 185 degrees F during a LOCA.Page163 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-M4101 Objective: 4, 11, 122.CourseID: GLP-OPS-M4100 Objective: 17KA

References:

1.295027 Generic 2.2.22: 3.4/4.12.EK1.03: 3.8/3.8

References:

1.Tech Spec Bases B 3.6.1.5TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:Categories: B.To prevent degradation of equipment important to safety during plant operation. C.To maintain containment pressure less than 22.4 psig during a LOCA. D.To maintain errors in reactor water level instrumentation within the initialassumptions of the accident analyses.Answer:AQuestionComments:This the bases of Tech Spec 3.6.1.5 and answer A is the only correctanswer. Tier 1 Group 1 This is a NEW question. 10CFR 43.2Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00911Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page164 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Technical SpecificationsTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 16:13:24 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:15:44 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Thu Jul 14 15:38:29 CDT 20055.Modified by mrasch at Fri Jul 29 06:55:37 CDT 20056.Question Reviewed by mellis at Thu Aug 04 07:27:49 CDT 20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Modified by mrasch at Tue Aug 09 16:13:24 CDT 20059.Question Reviewed by mellis at Tue Aug 09 16:14:24 CDT 200510.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 80 (1.0 Points)A strong earthquake has occurred, causing large LOCA in the drywell and aSuppression Pool leak into the Low Pressure Core Spray (LPCS) pump room.The LPCS room watertight door has failed.Also, electrical bus 15AA has tripped due to a fault.The following indications of containment parameters exist:Division 1 Drywell pressure is 8.5 psigDivision 2 Drywell pressure is 8.6 psigAverage Drywell temperature is 170°F.Average Containment temperature is 135°F.Division 1 Suppression Pool level is 14.6 feet.Division 2 Suppression Pool level is 14.0 feet.Average Suppression Pool temperature is 135°F.Page165 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP03 Objective: 3, 6KA

References:

1.295030 EA2.02: 3.9/3.92.Generic 2.4.3: 3.5/3.8

References:

1.EP caution 2What action should be directed based on evaluation of these conditions? A.Enter EP-4 and direct operation of all available sump pumps. B.Execute emergency depressurization per EP-2. C.Execute EP-2 Attachment 29, Primary Containment Water Level Determination. D.Rapidly depressurize using the Bypass Jack per EP-2.Answer:BQuestionComments:Given there is a 7 inch difference between Division 1 and 2 SuppressionPool level instruments and a loss of power to the Division 1 SuppressionPool level instrument reference leg Division 1 Suppression Pool Levelinstrument is incorrect. Suppression Pool level on the correct readinglevel instrument is below 14.25 ft per Caution 2 of the EmergencyProcedures Suppression Pool Temperature Instruments are suspect andsince they are reading the same temperature as Containment Airtemperature the Suppression Pool Temperature readings are inaccurate.Since Answer B is the only answer listing these two items it is the onlycorrect answer. 10CFR 41.7; 41.10; 43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00912Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page166 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 2.PSTG Appendix B Caution 23.04-1-01-E21-1 step 3.84.ARI 04-1-02-1H13-P601-21A-C7 step 4.1.3TrainingPrograms:Systems:Categories:Task

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Question Last Revised By:MikeRasch at Tue Aug 09 11:33:10 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:18:53 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Fri Jul 29 12:29:54 CDT 20055.Modified by mellis at Fri Jul 29 14:30:45 CDT 20056.Modified by mrasch at Fri Jul 29 14:50:32 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:27:53 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Modified by mrasch at Tue Aug 09 11:33:10 CDT 200510.Question Reviewed by mellis at Tue Aug 09 12:00:03 CDT 200511.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 81 (1.0 Points)The plant is in day seven of a refueling outage.Secondary Containment is NOT required operable.During Core Alterations, a malfunction of the Refueling Bridge has caused a fueldamaging event in the containment fuel racks.Which one of the following actions is required to limit the offsite release rate? A.Close the 208' containment airlock B.Page167 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 8.292.CourseID: GLP-OPS-ONEP Objective: 23.CourseID: GLP-RF-F1105 Objective: 3.7KA

References:

1.295038 Generic 2.2.28: 2.6/3.5

References:

1.05-1-02-II-8 steps 3.2, 3.32.01-S-06-2 steps6.7.26c, 6.7.233.Tech Specs 3.6.1.1; 3.6.1.2; 3.6.4.1 and basesClose the containment equipment hatch C.Start Fuel Pool Sweep Ventillation system D.Close any open Secondary Containment doorsAnswer:DQuestionComments:Conduct of Operations states during a fuel handling accident to startStandby Gas Treatment, establish secondary containment and performthe secondary containment verification surveillance to ensure allpenetrations are closed since during an outage work can be in progresswith these penetrations open. Primary Containment is NOT required to beOperable per Technical Specifications in modes 4 or 5. SecondaryContainment is NOT required to be Operable per TechnicalSpecifications due to being greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown.Answers A, B, and C are not required per Conduct of Operations, 01-S-06-2, and the High Radiation during Fuel Handling ONEP, 05-1-02-II-8.Answer D is CORRECT because Conduct of Operations, 01-S-06-2, andHigh Radiation During Fuel Handling, 05-1-02-II-8, require closingopenings in Secondary Containment. 10CFR 41.10/43.2/43.4/43.5/43.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00913Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page168 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 4.FSAR 15.7.4TrainingPrograms:Systems:Categories:Task

References:

Question Last Revised By:MikeRasch at Fri Jul 29 12:31:13 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:22:52 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mellis at Tue Jul 26 11:03:03 CDT 20055.Modified by mrasch at Fri Jul 29 12:31:13 CDT 20056.Question Reviewed by mellis at Thu Aug 04 07:28:00 CDT 20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 82 (1.0 Points)There is a fire in the Control Room.Which one of the following describes the basis for separating the Main Control Roomfrom the Remote Shutdown Panels? A.Essential equipment for both divisions in the Main Control Room is isolated toprevent a fire from causing equipment malfunctions preventing placing the plant incold shutdown. B.All Division I equipment associated with Remote Shutdown and systems requiredto place the plant in cold shutdown can be isolated from the Main Control Roomcontrols.Page169 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-C6100 Objective: 3; 9; 10; 11KA

References:

1.600000 AA2.16: 3.0/3.5

References:

1.UFSAR 7.4.1.5.1.12.05-1-02-II-1 Attachment IVTrainingPrograms:Systems:Categories: C.All Division II equipment associated with Remote Shutdown and systems requiredto place the plant in cold shutdown can be isolated from the Main Control Roomcontrols. D.All Division I equipment and selected Division II equipment associated withRemote Shutdown and systems required to place the plant in cold shutdown canbe isolated from the Main Control Room controls.Answer:BQuestionComments:GGNS design utilized Division I Remote Shutdown Panel as the SafeShutdown Systems for placing the unit in Cold Shutdown following a Firein the Main Control Room that could potentially compromise control ofsystems required to maintain the reactor in a safe cold shutdowncondition. 10CFR50 Appendix R sections III.G and III.L require only onetrain (division). Answers A, C, and D include Division II equipmentmaking them INCORRECT. Answer B is CORRECT because it identifiesthe correct division and reason for separating Remote ShutdownSystems from the Main Control Room. 10CFR41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00914Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page170 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Task

References:

Question Last Revised By:MikeRasch at Fri Jul 29 12:43:10 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:26:04 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Fri Jul 29 12:43:10 CDT 20055.Question Reviewed by mellis at Thu Aug 04 07:28:05 CDT 20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 83 (1.0 Points)The plant is in Mode 2 during startup, preparing to transfer to Mode 1.PCW SPLY TO SMPL WTR CLRS/CTMT CLRS valve P71-F150 has failed closed dueto a blown control power fuse.Average Containment Temperature is 97 degrees F.Transferring to Mode 1 is:Tech Specs are provided. A.Not allowed only due to containment temperature. B.Not allowed only due to inoperability of P71-F150. C.Allowed only within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of containment temperature exceeding the limit. D.Allowed provided containment temperature is returned within limits within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sof entering Mode 1.Answer:APage171 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-TS001 Objective: 34KA

References:

1.295011 AA2.01: 3.6/3.92.Generic 2.1.12: 2.9/4.0

References:

1.Tech Spec 3.6.1.52.Tech Spec 3.6.1.33.Tech Spec 3.0.4TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:Categories:1.Technical SpecificationsTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 07:16:12 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:29:08 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Fri Jul 29 07:16:12 CDT 2005QuestionComments:Tech Spec 3.6.1.5 is out of spec and is not 3.0.4 exempt for plant modechange. Tech Spec 3.6.1.3 is met for P71-F150 because the valve isclosed. Therefore Answer A is the only correct answer. Tier 1 Group 2This is a NEW question. 10CFR 43.2Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00915Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage172 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 5.Question Reviewed by mellis at Thu Aug 04 07:28:09 CDT 20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 84 (1.0 Points)The plant was operating at 100% power when a control rod drift annunciator wasreceived.The ACRO determined that a single control rod was drifting out of the core fromposition 12 to position 30.He selected the control rod and attempted to insert it.An RC&IS malfunction prevented the control rod from fully inserting so core flow wasreduced to 67 Mlbm/hr.Cyclops shows NO challenges to Thermal Limits, but reactor power did rise by 0.6%before lowering Recirc Flow.What level of Reactivity Management Event/Precursor is this classified as and who at aminimum is expected to be notified of the occurrence?NMM OP-103 is provided. A.The event is classified as a level 3 reactivity event. Operations Management andReactor Engineering are notified. B.The event is classified as a level 4 precursor. Reactor Engineering and NRCRegion IV Headquarters are notified. C.The event is classified as a level 3 reactivity event. Operations Managementand NRC Region IV Headquarters are notified. D.The event is classified as a level 4 precursor. Operations Management andReactor Engineering are notified.Answer:APage173 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 59.3; 74.3; 76.22.CourseID: GLP-OPS-ONEP Objective: 31KA

References:

1.295014 Generic 2.1.14: 2.5/3.32.Generic 2.1.6: 2.1/4.3

References:

1.02-S-01-27 section 6.5.32.ONEP 05-1-02-IV-1 section 3.2.33.NMM Policy ENS-PL-163 Attachment 9.1 section 2.2 item 194.NMM OP-103 section 5.3.3 and Attachment 9.1 Level 3 bullet 3.TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Shift Manager Training Program4.Licensed Operator Requalification Training ProgramQuestionComments:Subsequent actions of the ONEP for Control Rod/Drive Malfunctions for acontrol rod drift requires notification of Reactor Engineering for analysis ofthe core conditions with an out of position control rod. OperationsPhilosophy Operations Section Procedure requires analysis by ReactorEngineering. Entergy Policy PL-163 section 2.2 standards requires thenotification of Operations Management and Reactor Engineering ifabnormal reactivity changes occur in the core. Entergy NuclearManagement Manual Procedure OP-103 Reactivity Management setsthis as a Level 3 Reactivity Management Event since it is an unintendedpower rise by greater than 0.5 percent. It is NOT a Level 4 ReactivityManagement Precursor because this is the lesser severity of reactivitymanagement event. This event does not reach the level to makenotification to the NRC Regional Headquarters. It would include acourtesy call to the NRC Resident Inspector. Answer A is CORRECTbecause it is the only answer identifying the correct notifications and levelof reactivity management event. 10CFR41.10/43.5/43.6Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00916Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page174 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Systems:Categories:1.Administrative Requirements2.Off Normal Event Procedures3.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:36:07 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:33:17 CDT 20052.Modified by mrasch at Mon Jun 20 12:38:21 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by tharrelso at Thu Jul 28 07:02:17 CDT 20056.Modified by mrasch at Fri Jul 29 12:36:33 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:28:14 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Modified by mrasch at Tue Aug 09 11:36:07 CDT 200510.Question Reviewed by mellis at Tue Aug 09 12:00:05 CDT 200511.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 85 (1.0 Points)The plant is at 100% power when secondary containment instrument air isolation valveP53-F026B fails closed due to loss of control power and CANNOT be re-opened.The first action that should be directed by the Control Room Supervisor for thiscondition is: A.start Reactor Core Isolation Cooling (RCIC) in anticipation of entry into the Loss ofFeedwater Flow ONEP due to Main Steam Isolation Valve closure. B.attempt to restore power to P53-F026B in accordance with the Loss of AC PowerONEP.Page175 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-ONEP Objective: 1KA

References:

1.295020 AA2.03: 3.7/3.72.Generic 2.4.48: 3.5/3.8

References:

1.05-1-02-V-92.05-1-02-IV-1TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems: C.scram the reactor in anticipation of entry into the Control Rod Drive MalfunctionsONEP due to multiple control rod drifts/scrams. D.enter EP-2 and lower reactor pressure to 450-600 psig in anticipation of losing themain condenser as a heat sink.Answer:CQuestionComments:When air is isolated to the containment it is also lost to the Scram Airheader this allows slow depressurization of the header and control rodsto begin to drift/scram due to the scram inlet and outlet valves opening.Therefore Answer C is the correct answer for the given conditions. Tier1 Group 2 This is a NEW question. 10CFR 41.10/43.2/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00917Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page176 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Categories:1.Off Normal Event ProceduresTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 12:38:16 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:38:07 CDT 20052.Modified by mrasch at Mon Jun 20 14:21:57 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Fri Jul 29 07:28:33 CDT 20056.Modified by mrasch at Fri Jul 29 12:38:16 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:28:18 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 86 (1.0 Points)The plant was at 100% power when a LOCA occurred.EP2 and EP3 are being executed.The only available ECCS systems are RHR A and B.Current parameters are as follows:Reactor level is - 194 inches compensated fuel zone and falling.Pressure Suppression Pressure (PSP) is in the unsafe region.Both loops of Suppression Pool Cooling are in service, but not maximized.Reactor Core Isolation Cooling (RCIC), Standby Liquid Control (SLC) and Control RodDrive (CRD) are injecting at their maximum rates.How should the CRS direct use of RHR Systems to limit the offsite release rate underthese conditions? A.Page177 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GG-1-LP-RO-EP02 Objective: 4, 112.CourseID: GG-1-LP-RO-EP03 Objective: 2, 3, 6KA

References:

Execute EP-3 and direct initiation of both loops of Containment Spray. B.Execute EP-3 and direct initiation of one loop of Containment Spray, and leaveone loop of Suppression Pool Cooling in service. C.Execute EP-3 and direct maximizing both loops of Suppression Pool Cooling inservice. D.Execute EP-2 and direct alignment of RHR A and B for injection into the Reactor.Answer:DQuestionComments:Background for ECCS This question involves the difference between useof RHR for control of Containment parameters to prevent a challenge tothe Containment structure and the control of RPV parameters to preventthe degradation of the first fission product barrier. EP-3 step 8 says toinitiate those loops of RHR not required for adequate core cooling. WithRPV level at -194 inches and falling all ECCS will be aligned to the RPVuntil no longer needed. Suppression Pool temperature, while above thepoint to need Suppression Pool cooling, is not critical. RHR is stillrequired for the adequate core cooling issue. Once adequate core coolingis assured, other priorities can be established. Answer D is the onlyanswer to align ECCS Systems (RHR 'A' and 'B' for LPCI Injectionmode). Answers A, B, and C align RHR 'A' and 'B' for Containment Sprayand Suppression Pool Cooling which are lower priorities than recoveringadequate core cooling. This is a NEW question.10CFR41.8/41.10/43.2/43.4/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00918Review Status:ReviewedDifficulty:2: Comprehension or AnalysisPage178 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.203000 Generic 2.3.11: 2.7/3.22.Generic 2.4.22: 3.0/4.0

References:

1.TS Bases 3.5.1 Background2.EP-23.EP-34.PSP curve,5.CSIPL curve, PSTG pgs B-7-7TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Shift Manager Training Program4.Licensed Operator Requalification Training Program5.Shift Technical Advisor Training ProgramSystems:1.E12: Residual Heat Removal SystemCategories:1.Emergency Procedure Training2.FSAR3.Integrated Plant Operations4.Mitigation of Core Damage5.Technical Specifications6.Continuing Training7.STA TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 13:58:09 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:42:50 CDT 20052.Modified by mrasch at Mon Jun 20 12:52:50 CDT 20053.Modified by mrasch at Mon Jun 20 13:03:33 CDT 20054.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by tharrelso at Thu Jul 28 11:00:12 CDT 20057.Modified by mrasch at Fri Jul 29 12:41:38 CDT 20058.Modified by mrasch at Fri Jul 29 12:44:41 CDT 20059.Modified by mellis at Fri Jul 29 13:58:09 CDT 2005Page179 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 10.Question Reviewed by mrasch at Thu Aug 04 07:29:16 CDT 200511.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200512.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 87 (1.0 Points)The plant is in Mode 2 at the point of adding heat.I&C is removing a chart recorder in 1H13-P629, when incidental electrical contactinside the panel causes Low Pressure Core Spray to initiate and inject into the reactor.The resulting power excursion causes an IRM scram.Notification of this event to the NRC Operations Center is required within:(01-S-06-5 Attachment III is provided) A.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. B.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. C.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. D.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.Answer:BQuestionComments:Answer B is correct since valid RPS actuations with the reactor criticalare 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reports per 10 CFR 50.72(b)(2)(iv)(B). Answer C is incorrectbecause a 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report is required only for RPS initiations when theReactor is NOT critical. Answers A and D are incorrect since this doesnot fall into one of the events listed for 1 hr or 24 hr reports. SRO only10CFR 41.10/43.5Page180 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 11.13KA

References:

1.209001 Generic 2.1.15: 2.3/3.0

References:

1.01-S-06-5 Att. III2.10 CFR 50.72(b)(2)(iv)(B)TrainingPrograms:Systems:Categories:Task

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:41:20 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:45:59 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Fri Jul 29 12:45:16 CDT 20055.Modified by mellis at Fri Jul 29 15:11:44 CDT 20056.Modified by mellis at Fri Jul 29 15:16:11 CDT 20057.Question Reviewed by mrasch at Thu Aug 04 07:29:16 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Modified by mrasch at Tue Aug 09 11:41:20 CDT 200510.Question Reviewed by mellis at Tue Aug 09 12:00:06 CDT 200511.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00919Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page181 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm EB QUESTION: 88 (1.0 Points)The plant had been operating at 100% power when a scram occurred due to failure ofan optical isolator causing a Recirculation Pump double downshift.The Control Room Supervisor (CRS) has received the scram report and entered EP-2due to reactor water level.Reactor water level is + 18 inches narrow range and stable.The CRS has made NO assessments other than what was provided in the scramreport.NO Reactor Scram ONEP subsequent actions have been performed.Which one of the following is proper with respect to procedure execution and hierarchyin this situation? A.Since reactor water level is back above +11.4 inches, EP-2 should be exitedimmediately. Then, the CRS should enter the Reactor Scram ONEP and directsubsequent actions. B.The CRS should complete the primary control loop in EP-2. The CRS shoulddirect complimentary subsequent actions from the Reactor ScramONEP concurrently during EP-2 execution. C.The CRS should complete the primary control loop in EP-2. Then, if NOemergency exists, he should direct performance of subsequent actions from theReactor Scram ONEP only in the order listed in the ONEP. D.The CRS should complete the primary control loop in EP-2. The ACRO mayinsert IRMs without CRS direction or permission as soon as he completes allReactor Scram ONEP immediate actions.Answer:BQuestionComments:Answer A is INCORRECT because to exit the Emergency Procedures adetermination must be made per 05-S-01-EP-2 section 2.3 that anemergency NO longer exists and that is not stated in the answer. AnswerB is CORRECT because it ensures control is established per EP-2 thenallows other actions that are not in contradiction to the EPs to beperformed. In addition, 02-S-01-2 allows ONEP subsequent actions to beperformed out of sequence at the discretion of an SRO . Answer C isPage182 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 49.52.CourseID: GG-1-LP-RO-EP02 Objective: 4, 93.CourseID: GLP-OPS-ONEP Objective: 3.0KA

References:

1.215003 2.4.16: 3.0/4.0

References:

1.02-S-01-2 step 6.9.6d2.05-S-01-EP-2 steps 2.3, 2.4TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Shift Manager Training Program4.Shift Technical Advisor Training ProgramSystems:Categories:1.Administrative Requirements2.Emergency Procedure Training3.Integrated Plant Operations4.Off Normal Event Procedures5.STA TrainingTask

References:

Question Last Revised By:Mickey Ellis at Fri Jul 29 14:00:10 CDT 2005INCORRECT because subsequent actions of an ONEP are not requiredto be performed in order if an ON-Shift SRO authorizes out of orderperformance. Answer D is INCORRECT because the CRS gives thedirections for action during the execution of the EPs and all otherprocedures are subordinate to the EPs. 10CFR 41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00920Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page183 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Question History:1.Created by jbell at Thu Jun 16 14:49:34 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by tharrelso at Thu Jul 28 09:53:14 CDT 20055.Modified by mrasch at Fri Jul 29 12:46:28 CDT 20056.Modified by mellis at Fri Jul 29 13:46:57 CDT 20057.Modified by mellis at Fri Jul 29 14:00:10 CDT 20058.Question Reviewed by mrasch at Thu Aug 04 07:29:17 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200510.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 89 (1.0 Points)Source Range Monitor (SRM) A failed to respond greater than 300 cps during plantstartup.I&C has reworked the connector where the detector signal cable meets the SRM Adrawer.Only the flux signal to the SRM was affected (e.g. detector position function wasunaffected).The plant startup is still in progress, and Intermediate Range Monitors are on range 7.Which one of the following would satisfy the minimum retest requirements for SRM A toreturn it to operable status for ALL plant modes?(Do NOT consider whether plant conditions would support a particular re-test.)Technical Specifications 3.3.1.1; 3.3.1.2 and TR 3.3.2.1 are provided. A.SR 3.3.1.1.5 and SR 3.3.1.2.1 and SR TR 3.3.2.1.3 and SR TR 3.3.2.1.9 B.SR 3.3.1.1.5 and SR 3.3.1.2.4 and SR TR 3.3.2.1.3 and SR TR 3.3.2.1.9 C.SR 3.3.1.1.5 and SR 3.3.1.2.1 and SR 3.3.1.2.4 and SR TR 3.3.2.1.3Page184 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-TS001 Objective: 3.3; 4.2; 4.3; 4.4; 4.11; 13;KA

References:

1.215004 Generic 2.2.21: 2.3/3.5

References:

1.Tech Spec SR 3.3.1.1.5, SR 3.3.1.2.1 through 3.3.1.2.6, TRM table TR3.3.2.1-2,2.Tech Spec 1.0 Definitions - Channel Check; Channel Calibration; ChannelFunctional Test, Operable-Operability3.LCO3.0.54.SR3.0.1TrainingPrograms:Systems: D.SR 3.3.1.2.1 and SR 3.3.1.2.4 and SR TR 3.3.2.1.3 and SR TR 3.3.2.1.9Answer:DQuestionComments:To declare an SRM operable it must pass the surveillance requirementsfor Channel Checks, Channel Calibration, Channel Functional Test and aminimum count rate. SRM/IRM overlap surveillance per 3.3.1.1 is notrequired for SRM operability it is for IRM operability. Answer A isINCORRECT because it does not include a Minimum Count Rate and aChannel Calibration. SRM /IRM overlap is not required for SRMOperability. Answer B is INCORRECT because it does not include aChannel Check and Minimum Count Rate. SRM /IRM overlap is notrequired for SRM Operability. Answer C is INCORRECT because it doesnot include a Minimum Count Rate and a Channel Calibration. SRM /IRMoverlap is not required for SRM Operability. Answer D is CORRECTbecause it includes the necessary Channel Checks, Calibration,Functional Test and Minimum Count Rate checks. SRM /IRM overlap isnot required for SRM Operability. 10CFR 41.10/43.2/43.5Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00921Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:3Page185 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Categories:Task

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:55:32 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:53:41 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Fri Jul 29 12:47:49 CDT 20055.Question Reviewed by mellis at Thu Aug 04 07:30:28 CDT 20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Modified by mrasch at Tue Aug 09 11:51:23 CDT 20058.Modified by mrasch at Tue Aug 09 11:55:32 CDT 20059.Question Reviewed by mellis at Tue Aug 09 12:00:07 CDT 200510.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 90 (1.0 Points)The plant is in Mode 2 with reactor pressure 125 psig.An I&C technician just reported he discovered a bad relay in the RCIC actuation logic.Subsequently, RCIC was declared inoperable and HPCS was verified to be operable.In this situation, the Technical Specifications: A.require the reactor to be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. B.allow the startup to continue, but exceeding 150 psig is not permitted. C.allow the startup to continue to normal operating pressure, but entering Mode 1 isnot permiteed. D.Page186 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-TS01 Objective: 34KA

References:

1.2170002.GENERIC 2.1.10 Knowledge of conditions and limitations in the facility license[2.7/3.9]

References:

1.Tech Spec 3.5.32.Tech Spec 3.0.4TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:1.E51: Reactor Core Isolation Cooling SystemCategories:1.Integrated Plant Operations2.Technical Specificationsallow the startup to continue into Mode 1, however RCIC must be restored within14 days or LCO condition 3.5.5.B must be entered.Answer:BQuestionComments:Answer B is the only correct answer because RCIC may be out ofservice in this condition as long as RPV pressure is less than 150 psig.Since RCIC is required to be operable when RPV pressure is above150 psig the change is not allowed. Tier 2 Group 1 This is a NEWquestion. 10CFR 41.10/43.5Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00922Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page187 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Task

References:

Question Last Revised By:MikeRasch at Fri Jul 29 14:07:38 CDT 2005Question History:1.Created by jbell at Thu Jun 16 14:56:39 CDT 20052.Modified by mrasch at Mon Jun 20 13:12:37 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Fri Jul 29 12:49:38 CDT 20056.Modified by mrasch at Fri Jul 29 14:07:38 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:30:28 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 91 (1.0 Points)A large break LOCA with failure of Divisions 1 and 2 ECCS has occurred.Emergency Depressurization has been conducted.High Pressure Core Spray is injecting at rated flow, but it is determined that reactorwater level is unable to be restored above -192 inches.Reactor water level is currently -210 inches and slowly lowering.Severe Accident Procedure (SAP) 3 is entered.Which one of the following is the reason why SAP-3 is entered? A.Core damage is occurring requiring the Containment to be flooded and theseactions are described in the SAPs. B.Emergency Procedures have insufficient systems allocated for restoration of RPVlevel. C.Conditions in the reactor require venting of the vessel to facilitate RPV flooding.Page188 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-EP-EPT19 Objective: 7KA

References:

1.290002 Generic 2.4.14: 3.0/3.9

References:

1.05-S-01-SAP-1 step 2.1TrainingPrograms:Systems:Categories:Task

References:

Question Last Revised By:MikeRasch at Thu Aug 04 08:15:37 CDT 2005Question History: D.Reactor conditions fail to assure adequate core cooling.Answer:DQuestionComments:Emergency Procedures are based upon maintaining adequate corecooling. Severe Accident Procedures are based upon conditions nolonger assure adequate core cooling and the focus is to shift operationsto ensure Primary Containment is not challenged. Answer A isINCORRECT because core damage is not indicated per definitions in thegiven information at this time. Answer B. is INCORRECT because thesame systems are listed for use in RPV level control in both the EPs andSAPs. Answer C is INCORRECT because RPV Flooding is contained onEP-2. Answer D is CORRECT because the entry conditions of 05-S-01-SAP-1 state that adequate core cooling cannot be assured and with thegiven conditions adequate core cooling is not assured. 10CFR41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00923Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page189 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Created by jbell at Thu Jun 16 14:58:47 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Fri Jul 29 12:51:27 CDT 20055.Question Reviewed by mellis at Thu Aug 04 07:30:29 CDT 20056.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20057.Modified by mrasch at Thu Aug 04 08:15:37 CDT 20058.Question Reviewed by mellis at Thu Aug 04 08:17:57 CDT 20059.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 92 (1.0 Points)The plant is in Mode 1.Containment Spray A time delay relay E12-K93A has been declared inoperable due toit failing to actuate during an electrical surveillance and the relay has been removed fortroubleshooting.An alarm and status light are received on 1H13-P601 indicating CTMT SPRBSPARGER INL VLV E12-F028B control power fuses have blown.What is the most limiting impact of these conditions regarding RHR ContainmentSpray?Tech Specs and 02-S-01-17 Att II are provided. A.Containment Spray A will not automatically initiate; the inoperable channel mustbe placed into trip within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and E12-F028B restored within 7 days. B.Containment Spray A can be considered OPERABLE via manual initiationcapability; E12-F028B must be restored within 7 days. C.Containment Spray A will not automatically initiate; Mode 3 must be entered within20 hours. D.Containment Spray A will not automatically initiate; Mode 3 must be entered within21 hours.Page190 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-E1200 Objective: 8.11, 10.1, 10.2, 13.2KA

References:

1.226001 A2.13 Valve logic failure [2.8/2.9]

References:

1.E-1181-40; 42; 44; 68; 692.Tech Specs 3.3.6.33.Tech Specs 3.6.1.7TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:Categories:Answer:DQuestionComments:This involves a safety function determination for instrumentation. 02-S-01-17 says if a safety function determination involves instrumentation,take action per the instrumentation Technical Specification. Here,automatic initiation capability for containment spray has been lost due toa non-functional relay on Div 1 and a non-functional spray valve on Div 2.The appropriate action would be TS 3.3.6.3 action B.1, which requires,after 1 hr, declaring both sprays inop. TS 3.6.1.7 requires, after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,entering a 12 hr shutdown statement for 2 inop sprays. Thus 1+8+12=21which is action d. It would be inappropriate to declare both sprays inopimmediately, as represented by answer c. All other answers are wrongsince they state times other that 21 hrs. Tier 2 Group 2 This is a NEWquestion 10CFR 43.2Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00924Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page191 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.Technical SpecificationsTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:57:41 CDT 2005Question History:1.Created by jbell at Thu Jun 16 15:01:13 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Fri Jul 29 08:18:52 CDT 20055.Modified by mrasch at Fri Jul 29 12:52:42 CDT 20056.Question Reviewed by mellis at Thu Aug 04 07:30:30 CDT 20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Modified by mrasch at Tue Aug 09 11:57:41 CDT 20059.Question Reviewed by mellis at Tue Aug 09 12:00:08 CDT 200510.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 93 (1.0 Points)The Refueling Floor SRO must grant permission to bypass a Refueling Platforminterlock manifested by which one of the following Refueling Platform Interlock Displayindicating lights? A.Safety Travel interlock B.Backup Hoist interlock C.Rod Block Interlock No.1 D.Bridge Reverse Stop No.1Answer:AQuestionComments:Fuel handling precaution and limitation specifies the travel interlock asrequiring Refuel SRO permission to bypass. Answer A is the correctanswer. Tier 3 10CFR 43.2/43.7Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00925Page192 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-RF-F1101 Objective: 15.3KA

References:

1.234000 A2.01 Interlock failure [3.3/3.7]

References:

1.04-1-01-F11-1 step 3.19TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:1.F11: Reactor Refueling EquipmentCategories:1.Administrative Requirements2.SystemsTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 12:54:23 CDT 2005Question History:1.Created by jbell at Thu Jun 16 15:04:03 CDT 20052.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20053.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20054.Modified by mrasch at Thu Jul 28 18:13:07 CDT 20055.Modified by mrasch at Fri Jul 29 12:54:23 CDT 20056.Question Reviewed by mellis at Thu Aug 04 07:30:30 CDT 20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page193 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm EB QUESTION: 94 (1.0 Points)Objectives:1.CourseID: GLP-OPS-PROC Objective: 55.4KA

References:

1.K/A Generic 2.1.2: 3.0/4.0The plant is in Mode 5 with core alterations in progress.A major evolution involving use of the Polar crane is scheduled.Who must the Refueling Floor Supervisor notify before the evolution may begin? A.The Shift Supervisor and the Outage Director B.The Shift Manager and the Outage Director C.The Shift Manager and the Refueling Floor SRO D.The Shift Supervisor and the Refueling Floor SROAnswer:CQuestionComments:01-S-06-2, Section 6.7.5 requires the Refueling Floor Supervisor to notifythe Shift Manager of any refueling floor major activity. When corealterations are in progress, he must also notify the Refueling Floor SROtherefore, Answer C is CORRECT. Answer A is INCORRECT becauseneither of these positions is required to be notified. Answer B isINCORRECT because only one of thses positions must be notified.Answer D is INCORRECT because only one of thses positions must benotified. 10CFR 41.10/43.5Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00926Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:3Page194 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm

References:

1.01-S-06-2 sections 6.7.5TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Shift Manager Training Program4.Licensed Operator Requalification Training ProgramSystems:Categories:1.Administrative Requirements2.Industrial Safety3.Operating Experience4.Refueling Training5.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 12:55:20 CDT 2005Question History:1.Created by jbell at Thu Jun 16 15:06:04 CDT 20052.Modified by mrasch at Mon Jun 20 13:14:29 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by tharrelso at Thu Jul 28 07:24:58 CDT 20056.Modified by mrasch at Fri Jul 29 12:55:20 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:30:31 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 95 (1.0 Points)The plant is in Mode 1.Standby Service Water Pump A is tagged out of service for planned maintenance whenRHR B Jockey Pump trips.Page195 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-TS001 Objective: 34; 35KA

References:

1.GENERIC 2.1.12 Ability to apply technical specifications for a system [2.9/4.0]

References:

1.Tech Specs 3.5.1; 3.6.1.7; 3.6.1.8; 3.6.2.3; 3.7.12.02-S-01-173.Tech Specs 3.0.6; 3.0.3TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:Categories:1.Administrative RequirementsWhat is the latest time until the plant must be in Mode 3 for this condition?Tech Specs are provided. A.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B.13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> C.20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> D.84 hour9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />sAnswer:BQuestion Comments:This is a NEW question. Tier 3 10 CFR41.10/43.3/43.5Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00927Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:3Page196 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 2.Continuing TrainingTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 12:57:14 CDT 2005Question History:1.Created by jbell at Thu Jun 16 15:08:46 CDT 20052.Modified by mrasch at Mon Jun 20 05:25:01 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Thu Jul 28 17:55:30 CDT 20056.Modified by mrasch at Fri Jul 29 12:57:14 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:30:32 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 96 (1.0 Points)A surveillance test on a Control Room Air Conditioning Unit indicates the system hassignificantly degraded but is still operable.The work to repair the unit will require the unit to be out of service for 32 days.Which one of the following describes the proper work prioritization?EN-WM-100 and Tech Spec 3.7.4 are provided. A.Priority 1 Work on-line. B.Priority 2 work on-line. C.Priority 2 work in outage. D.Priority 3 work in outageAnswer:CQuestionThe work per the prioritization matrix of EN-WM-100 for safety relatedPage197 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 24.4KA

References:

1.2.2.19: 2.1/3.1

References:

1.EN-WM-100 Attachments 9.1 and 9.22.Tech Spec 3.7.4TrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:Categories:1.Administrative RequirementsTask

References:

Question Last Revised By:MikeRasch at Fri Jul 29 13:01:47 CDT 2005Question History:1.Created by jbell at Thu Jun 16 15:12:44 CDT 20052.Modified by mrasch at Mon Jun 20 13:16:38 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Thu Jul 28 13:14:00 CDT 2005Comments:equipment covered by Tech Specs that is operable and available isYellow Priority 2 per attachment 9.2 of EN-WM-100 work that would takea component out of service for a period of time that would exceed theTS shutdown requirements must be deferred to an outage. This is aNEW question. 10CFR 41.10/43.5Image

Reference:

NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00928Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page198 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 6.Modified by mrasch at Thu Jul 28 13:16:26 CDT 20057.Modified by mrasch at Fri Jul 29 13:01:47 CDT 20058.Question Reviewed by mellis at Thu Aug 04 07:30:32 CDT 20059.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200510.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 97 (1.0 Points)A Temporary Alteration will be used to install specialized test equipment in place of apermanent plant recorder for testing activities.The testing will last one week.One of the individuals responsible for approving the installation of this TemporaryAlteration is the: A.Control Room Supervisor. B.Manager, Planning and Scheduling. C.Manager, System Engineering. D.Radiation Protection Manager.Answer:AQuestionComments:The following personnel are required to approve a temporary alterationprior to its installation per 01-S-06-3: Manager, Operations (or designeeAssistant Operations Manager - Shift), Shift Manager, Control RoomSupervisor. The Manager, System Engineering is a process owner andrequests Temporary Alterations but doe not have to approve them.Therefore Answer B is CORRECT. This is a NEW question. 10CFR41.10/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamPage199 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-PROC Objective: 9.9KA

References:

1.Generic 2.2.16: 1.9/2.6

References:

1.01-S-06-3 sections 2.4.1; 2.5; 6.1.6; 6.1.8; 6.1.13; 6.1.142.FSAR 13.1.2.3.7; 18.1.13 response a.TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:Categories:1.Administrative Requirements2.Industrial Safety3.Plant ModificationsTask

References:

Question Last Revised By:MikeRasch at Tue Aug 09 11:59:08 CDT 2005Question History:1.Created by jbell at Thu Jun 16 15:15:42 CDT 20052.Modified by mrasch at Mon Jun 20 05:19:07 CDT 20053.Modified by mrasch at Mon Jun 20 13:18:19 CDT 20054.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20055.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20056.Modified by tharrelso at Thu Jul 28 08:24:29 CDT 20057.Modified by mrasch at Fri Jul 29 13:03:19 CDT 20058.Modified by mrasch at Fri Jul 29 14:37:39 CDT 20059.Modified by mrasch at Fri Jul 29 14:39:52 CDT 200510.Modified by mrasch at Fri Jul 29 14:47:59 CDT 200511.Question Reviewed by mellis at Thu Aug 04 07:30:33 CDT 2005QuestionID:GGNS-NRC-00929Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page200 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 12.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/200513.Modified by mrasch at Tue Aug 09 11:59:08 CDT 200514.Question Reviewed by mellis at Tue Aug 09 12:00:08 CDT 200515.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 98 (1.0 Points)Objectives:1.CourseID: GLP-OPS-PROC Objective: 36 and 37KA

References:

1.Generic 2.3.6: 2.1/3.1

References:

The individual responsible for authorizing the commencement of a release on a BatchLiquid Radwaste Discharge Permit is the: A.Radwaste Specialist. B.Chemistry Supervisor. C.Control Room Supervisor. D.Shift Manager.Answer:DQuestionComments:The Shift Manager is the approval signature for Radwaste Dischargesper 01-S-08-11. Tier 3 This is a NEW question. 10CFR41.10/41.13/43.4/43.5Image

Reference:

NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00930Review Status:ReviewedDifficulty:1: Fundamental Knowledge or MemoryDifficulty Rating:2Page201 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 1.01-S-08-11TrainingPrograms:1.Senior Reactor Operator Training ProgramSystems:Categories:1.Administrative RequirementsTask

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Question Last Revised By:MikeRasch at Fri Jul 29 14:13:03 CDT 2005Question History:1.Created by jbell at Thu Jun 16 15:18:23 CDT 20052.Modified by mrasch at Mon Jun 20 13:23:28 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Fri Jul 29 13:04:44 CDT 20056.Modified by mrasch at Fri Jul 29 14:13:03 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:30:33 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 99 (1.0 Points)A plant startup is in progress and all reactor recirc system indications are normal.When Reactor Recirc Pumps are shifted to fast speed, the reactor operator reports: The REACTOR PMP A SEAL STG FLO HI/LO alarm has annunciated. Seal staging flow indicates higher than normal. Reactor Recirc pump A seal cavity # 1 is 1050 psig. Reactor Recirc pump B seal cavity # 2 is 1000 psig. Indicated Reactor Recirc seal temperatures are lower than normal.Page202 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-OPS-B3300 Objective: 29.1; 29.4KA

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1.Generic 2.4.47: 3.4/3.7Based on this information, the SRO should enter procedure: A.05-1-02-III-3, Reduction in Recirculation System Flow Rate because of improperseal cavity differential pressures. B.04-1-01-B33-1, Reactor Recirculation System, because of improper seal cavitydifferential pressures. C.04-1-01-C11-1, Control Rod Hydraulic System, because of improper seal stagingflow. D.04-1-01-B33-1, Reactor Recirculation System, because of lower than normal sealtemperatures.Answer:BQuestionComments:This is an indication of a failure of Seal #1 due to the following RECIRCPMP A SEAL STG FLO HI/LO annunciator being in, seal staging flowindicating high, Seal # 2 pressure approaching Seal # 1 pressure andseal temperatures dropping. If there are signs of seal de-staging theRecirc pump is to be shifted back to slow speed to attempt to restage theseals then return its operation to fast speed. Even if the seals do notrestage the pump may continue operation in fast speed allowing the plantto raise power with monitoring of the seals. Answer B is the correctanswer. The procedure to be entered to downshift the recirc pumps toattempt to fix the seal staging. This is not a change in recirc flowrequiring entry into the ONEP. This is a NEW question. 10CFR41.10/43.5Image

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NoneClosed Reference QuestionHandout Not Required with ExamQuestionID:GGNS-NRC-00931Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:3Page203 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm

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1.04-1-01-B33-1 sect 3.13.1; 4.2.2a(12), (13), (14), (15)2.ARI 04-1-02-1H13-P680-3A-B53.M-1081B4.M-1078ATrainingPrograms:1.Senior Reactor Operator Training Program2.Licensed Operator Requalification Training ProgramSystems:1.B33: Reactor Recirculation SystemCategories:1.SystemsTask

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Question Last Revised By:MikeRasch at Thu Aug 04 08:17:26 CDT 2005Question History:1.Created by jbell at Thu Jun 16 15:21:23 CDT 20052.Modified by mrasch at Mon Jun 20 13:25:42 CDT 20053.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Fri Jul 29 13:07:05 CDT 20056.Modified by mrasch at Fri Jul 29 14:27:49 CDT 20057.Question Reviewed by mellis at Thu Aug 04 07:30:34 CDT 20058.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20059.Modified by mrasch at Thu Aug 04 08:17:26 CDT 200510.Question Reviewed by mellis at Thu Aug 04 08:18:01 CDT 200511.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:EB QUESTION: 100 (1.0 Points)A Reactor Recirc LOCA and Feed Water line break in the drywell has occurred.Reactor Core Isolation Cooling (RCIC) has tripped due to high RCIC room ambientPage204 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm temperature that occurred five minutes ago and failed to isolate.RCIC room blowout panel alarms are in on the Fire Computer in the control room.NO systems listed on Table 1 of the Alternate Level Control leg of EP-2 are availablefor injection. RHR B Injection Valve E12-F042B has lost power in the closed position.Control Rod Drive flow and Standby Liquid Control flow are maximized.The following conditions exist: Reactor Power 0% Drywell Pressure 7.6 psig slowly falling Reactor Pressure 750psig slowly falling Drywell Temperature 192°F slowly falling Reactor level -170 inches slowly falling Drywell Radiation 6R/hr slowly rising Containment pressure 2.0 psig slowly falling Containment Temperature 108°F slowly falling Suppression Pool Temperature 125°F stable Suppression Pool Level 20.8 ft. stableDose commitment at the site boundary is 20 mr/hr TEDE and 130 mr/hr CEDE Thyroid.What are the Emergency Classification and Protective Action Recommendation (PAR),if any, for this event?10-S-01-1 is provided. A.Alert, NO PAR B.Site Area Emergency, NO PAR C.General Emergency, Evacuate 2 miles all sectors and 5 miles downwind andshelter the remainder of the 10 mile Emergency Planning Zone D.General Emergency, Evacuate 2 miles all sectors and 10 miles downwind andshelter the remainder of the 10 mile Emergency Planning ZoneAnswer:CQuestionComments:With RPV level below TAF and falling (Potential Loss of Fuel Cladding),high Drywell Pressure/ RCIC Steam line Break outside ContainmentPage205 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm Objectives:1.CourseID: GLP-EP-EPTS6 Objective: 1; 2KA

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1.Generic 2.4.44: 2.1/4.0

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1.10-S-01-1 EAL 3.4, 10-S-01-1 step 6.1.4k(1)TrainingPrograms:1.Reactor Operator Training Program2.Senior Reactor Operator Training Program3.Licensed Operator Requalification Training ProgramSystems:Categories:1.Emergency Plan TrainingTask

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Question Last Revised By:MikeRasch at Fri Jul 29 13:08:04 CDT 2005Question History:1.Created by jbell at Thu Jun 16 15:23:55 CDT 20052.Modified by mrasch at Mon Jun 20 13:29:02 CDT 2005(Loss of Reactor Pressure boundary), RCIC has failed to isolate (Loss ofPrimary Containment), Loss of 2 of 3 fission product barriers with apotential of the 3rd is a General Emergency per EAL 3.4. Due to theoffsite radiation levels the Standard PAR is issued for a GeneralEmergency. Answers A and B are INCORRECT because the wrongemergency classification. Answer D is INCORRECT because the wrongPAR is issued for the given radiation conditions. Answer C is CORRECTbecause the correct emergency classification and PAR are identified.This is a NEW question. 10CFR41.10/43.5Image

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NoneOpen Reference QuestionHandout Required with ExamQuestionID:GGNS-NRC-00932Review Status:ReviewedDifficulty:2: Comprehension or AnalysisDifficulty Rating:2Page206 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm 3.Question Reviewed by mrasch at Mon Jun 20 14:44:20 CDT 20054.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20055.Modified by mrasch at Fri Jul 29 13:08:04 CDT 20056.Question Reviewed by mellis at Thu Aug 04 07:30:35 CDT 20057.Question used on ExamName: NRC August 2005 - 3 ExamID: NRC-082005-3 ExamDate: 08/12/20058.Question used on ExamName: NRC August 2005 - 2 ExamID: NRC-082005-2 ExamDate: 08/12/2005Comments:END OF EXAMTotal Number of Questions: 100 Total Point Value: 100.0Total Exam Question Difficulty: 163.0Average Exam Question Difficulty: 1.63Questions with Difficulty Level 1: 37Questions with Difficulty Level 2: 63Go to Action Menu.Go to Home Menu.Exit.Page207 of207View Exams8/9/2005file://F:\SROExamRev3Final.htm