ML030410516

Initial Submittal of the Written Examination for the Monticello Inital Examination - Oct. 2002

ML030410516
Person / Time
Site:	Monticello
Issue date:	10/14/2002
From:	Engen M Nuclear Management Co
To:	Office of Nuclear Reactor Regulation
References
50-263/02-301, ES-401, ES-401-7 50-263/02-301
Download: ML030410516 (181)

Text

INITIAL SUBMITTAL OF THE WRITTEN EXAMINATION FOR THE MONTICELLO INITIAL EXAMINATION - OCT 2002

ES-401 Written Examination Form ES-401-7 Quality Checklist Facility: Monticello Nuclear Generating Plant Date of Exam: 09/09/02 Exam Level: RO Initial Item Description a b* C#

1. Questions and answers technically accurate and applicable to facility

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2. a. NRC K/As referenced for all questions

b. Facility leaming objectives referenced as available

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3. ROISRO overlap is no more than 75 percent, and SRO questions are appropriate per Section D.2.d of ES-401.

4. Question selection and duplication from the last two NRC licensing exams appears consistent with a systematic sampling process

5. Question duplication from the license screening/audit exam was controlled as indicated below (check the item that applies) and appears appropriate:

X the audit exam was systematically and randomly developed; or

_the audit exam was completed before the license exam was started; or

_the examinations were developed independently; or

_the licensee certifies that there is no duplication; or other (explain)

6. Bank use meets limits (no more than 75 Bank Modified New percent from the bank at least 10 percent new, 19 6 75 and the rest modified); enter the actual question distribution at right

7. Between 50 and 60 percent of the questions on Memory C/A the exam (including 10 new questions) are 46 54 written at the comprehension/analysis level; enter the actual question distribution at right

8. References/handouts provided do not give away answers '114e

9. Question content conforms with specific K/A statements in the previously approved examination outline and is appropriate for the Tier to which they are Qy Y'i '*,.

assigned; deviations are justified

10. Question psychometric quality and format meet ES, Appendix B, guidelines )44\

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11. The exam contains 100, one-point, multiple choice items; the total is correct and agrees with value on cover sheet Printed Name / Signature Date

a. Author Marvin R Engen 17.,

b. Facility Reviewer (*) Kurt Markling

c. NRC Chief Examiner (#)

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d. NRC Regional Supervisor 1.. L. r(-t.,_.

Note: *The facility reviewer's initials/signature are not applicable for NRC-developed examinations.

1. Independent NRC reviewer initial items in Column "c;" chief examiner concurrence required.

NUREG-1021, Revision 8, Supplement I

ES-401 Written Examination Form ES-401-7 Quality Checklist Facility: Date of Exam: Exam Level: SRO Initial Item Description a b* c#

1. Questions and answers technically accurate and applicable to facility ___ ___

2. a. NRC K/As referenced for all questions

b. Facility learning objectives referenced as available P _

3. RO/SRO overlap is no more than 75 percent, and SRO questions are appropriate per Section D.2.d of ES-401.

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4. Question selection and duplication from the last two NRC licensing exams appears consistent with a systematic sampling process

5. Question duplication from the license screening/audit exam was controlled as indicated below (check the item that applies) and appears appropriate:

X the audit exam was systematically and randomly developed; or

-- _-the audit exam was completed before the license exam was started; or

_the examinations were developed independently; or

_the licensee certifies that there is no duplication; or other (explain)

6. Bank use meets limits (no more than 75 Bank Modified New percent from the bank at least 10 percent new, 19 6 75 and the rest modified); enter the actual question distribution at right

7. Between 50 and 60 percent of the questions on Memory C/A the exam (including 10 new questions) are 44 56 D written at the comprehension/analysis level; enter the actual question distribution at right

8. References/handouts provided do not give away answers ' /44,t

9. Question content conforms with specific K/A statements in the previously approved examination outline and is appropriate for the Tier to which they are assigned; deviations are justified

10. Question psychometric quality and format meet ES, Appendix B, guidelines k-

11. The exam contains 100, one-point, multiple choice items; the total is correct and A i agrees with value on cover sheet Printed Name / Sign ture Date

a. Author Marvin REngenI

b. Facility Reviewer (*) Kurt Markling "t wAA49, /0,1 /0 2

c. NRC Chief Examiner (#) > 1j, COG . _ _(

d. NRC Regional Supervisor 1,)z Note: *The facility reviewers initials/signature are not applicable for NRC-developed examinations.

1. Independent NRC reviewer initial items in Column "c;" chief examiner concurrence required.

NUREG-1021, Revision 8, Supplement I

BWR SR( xamination Outline Facility: ( .ticello Nuclear Generating Printed: 06/( 02 ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group I l*orm I*_,tNII E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points 295003 Partial or Complete Loss of A.C. Power / 6 X AK2.02 - Emergency generators 4.2* 1 295003 Partial or Complete Loss of A.C. Power / 6 X AA1.01 - A.C. electrical distribution system 3.8 1 295006 SCRAM / I X AK 1.01 - Decay heat generation and removal. 3.9 1 295006 SCRAM / 1 X AK3.03 - Reactor pressure response 3.9* 1 295010 High Drywell Pressure / 5 X 2.4.8 - Knowledge of how the event-based 3.7 1 emergency/abnormal operating procedures are used in conjunction with the symptom-based EOPs.

295013 High Suppression Pool Temperature / 5 X AA2.01 - Suppression pool temperature 4.0 1 295013 High Suppression Pool Temperature / 5 X AK2.01 - Suppression pool cooling 3.7 1 295014 Inadvertent Reactivity Addition / 1 X AK1.01 - Prompt critical 3.8 1 295015 Incomplete SCRAM / 1 X AK3.01 - Bypassing rod insertion blocks 3.7 1 295017 High Off-Site Release Rate / 9 X AK3.01 - System isolations 3.9 1 295023 Refueling Accidents / 8 X AA1.02 - Fuel pool cooling and cleanup system 3.1 1 295023 Refueling Accidents / 8 X 2.2.22 - Knowledge of limiting conditions for operations 4.1 1 and safety limits.

295024 High Drywell Pressure / 5 X EK1.0 - Drywell integrity: Plant-Specific 4.2* 1 295024 High Drywell Pressure / 5 X EA2.02 - Drywell temperature 4.0 1 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 1 of 13

Facility: (MI..celloNuclear Generating BWR SRlI amination Outline Printed: 06/17 2 ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Groun I E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points 295025 High Reactor Pressure / 3 X EKI.04 - Decay heat generation 3.9 1 295025 High Reactor Pressure / 3 X EA2.02 - Reactor power 4.2 1 295026 Suppression Pool High Water Temperature / 5 X EA1.03 - Temperature monitoring 3.9 1 295026 Suppression Pool High Water Temperature / 5 X 2.1.10 - Knowledge of conditions and limitations in the 3.9 1 facility license.

295030 Low Suppression Pool Water Level / 5 X EA2.03 - Reactor pressure 3.9 1 295031 Reactor Low Water Level / 2 X 2.1.32 - Ability to explain and apply system limits and 3.8 1 precautions.

295031 Reactor Low Water Level / 2 X EA 1.06 - Automatic depressurization system 4.4* 1 295037 SCRAM Condition Present and Reactor Power Above X EK2.07 - Neutron monitoring system 4.0 1 APRM Downscale or Unknown / I 295037 SCRAM Condition Present and Reactor Power Above X EK3.04 - Hot shutdown boron weight: Plant-Specific 3.7 1 APRM Downscale or Unknown / 1 295038 High Off-Site Release Rate / 9 X EK2.03 - Plant ventilation systems 3.8 1 295038 High Off-Site Release Rate /9 X EK3.03 - Control room ventilation isolation: 3.9 1 Plant-Specific 500000 High Containment Hydrogen Concentration / 5 X EA2.03 - Combustible limits for drywell 3.8 1 K/A Category Totals: 4 4 5 4 5 4 Group Point Total: 26 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 2 of 13

BWR SRl( amination Outline Printed: 06/17( 2 Facility: i, .icello Nuclear Generating ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Groun 2 Ffll-m l?*l_*lflll E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points 295001 Partial or Complete Loss of Forced Core Flow X AK3.06 - Core flow indication 3.0 1 Circulation / 1 295002 Loss of Main Condenser Vacuum / 3 X AA1.01 - Condensate system 2.6 1 295008 High Reactor Water Level / 2 X AK1.03 - Feed flow/steam flow mismatch 3.2 1 295008 High Reactor Water Level / 2 X 2.1.6 - Ability to supervise and assume a management 4.3 1 role during plant transients and upset conditions.

295012 High Drywell Temperature / 5 X AK2.02 - Drywell cooling 3.7 1 295018 Partial or Complete Loss of Component Cooling X AA2.01 - Component temperatures 3.4 1 Water / 8 295021 Loss of Shutdown Cooling / 4 X AA2.05 - Reactor vessel metal temperature 3.3 1 295022 Loss of CRD Pumps / I X AK2.02 - CRD mechanism 3.1 1 295028 High Drywell Temperature / 5 X EA1.02 - Drywell ventilation system 3.8 1 295029 High Suppression Pool Water Level / 5 X EA2.02 - Reactor pressure 3.6 1 295029 High Suppression Pool Water Level / 5 X EK1.01 - Containment integrity 3.7 1 295033 High Secondary Containment Area Radiation Levels / X EK2.01 - Area radiation monitoring system 4.0 1 9

295034 Secondary Containment Ventilation High Radiation / X 2.4.10 - Knowledge of annunciator response procedures. 3.1 1 9

295034 Secondary Containment Ventilation High Radiation / X EK3.05 - Manual SCRAM and depressurization: 3.9 1 9 Plant-Specific I_ I Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 3 of 13

BWR SR( amination Outline Printed: 06/1* 2 Facility: .icello Nuclear Generating ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 Iinrm F* 3-401-1 1 1 r~Form-TV E/APE # E/APE Name / Safety Function KI K2 K31 A1IA2 G IKATopic I mp. Pnint*z 295035 Secondary Containment High Differential Pressure / 5 X EK1.01 - Secondary containment integrity 4.2* 1 295036 Secondary Containment High Sump/Area Water X EA1.02 - Affected systems so as to isolate damaged 3.6 1 Level / 5 portions 600000 Plant Fire On Site / 8 X 2.3.10 - Ability to perform procedures to reduce 3.3 1 excessive levels of radiation and guard against personnel exposure.

K/A Category Totals: 3 3 2 3 3 3 Group Point Total: 17 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 4 of 13

( BWR SRO 1 nination Outline Printed: 06/( )02 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 / Group I Form F.R-dlTI .I Sys/Ev# System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imp. Points 202002 Recirculation Flow Control System /I X K1.03 - Reactor core flow 3.7 1 203000 RHR/LPCI: Injection Mode (Plant X K6.05 - Condensate storage and transfer system: 2,5 1 Specific) / 2 Plant-Specific 203000 RHR/LPCI: Injection Mode (Plant X 2.1.7 - Ability to evaluate plant performance 4.4 1 Specific) / 2 and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

211000 Standby Liquid Control System / I X A2.01 - Pump trip 3.8* 1 215004 Source Range Monitor (SRM) System X K2.01 - SRM channels/detectors 2.8 1

/7 215004 Source Range Monitor (SRM) System X A4.01 - SRM count rate and period 3.8 1

/7 215005 Average Power Range Monitor/Local X K2.02 - APRM channels 2.8 1 Power Range Monitor System / 7 215005 Average Power Range Monitor/Local X K4.01 - Rod withdrawal blocks 3.7 1 Power Range Monitor System / 7 216000 Nuclear Boiler Instrumentation / 7 X KI.09 - Redundant reactivity control alternate 4.0 1 rod insertion; Plant-Specific 218000 Automatic Depressurization System / X A 1.03 - ADS valve air supply pressure: 3-4* 1 3 Plant-Specific 223001 Primary Containment System and X K6.12 - D.C. electrical distribution 3.0 Auxiliaries / 5 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 5 of 13

( BWR SRO 1 nination Outline Printed: 06/( 102 Monticello Nuclear Generating Facility:

ES - 401 Plant Systems - Tier 2 / Group I Form ES-401-1 Sys/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 223002 Primary Containment Isolation X K4.05 - Single failures will not impair the 3. 1 1 System/Nuclear Steam Supply function ability of the system Shut-Off/ 5 223002 Primary Containment Isolation X A3.01 - System indicating lights and alarms 3.4 1 System/Nuclear Steam Supply Shut-Off/ 5 226001 RHR/LPCI: Containment Spray X A3.05 - Containment pressure 4.0* 1 System Mode / 5 239002 Relief/Safety Valves / 3 X K5.06 - Vacuum breaker operation 3.0 1 241000 Reactor/Turbine Pressure Regulating X K3.02 - Reactor pressure 4.3" 1 System / 3 241000 Reactor/Turbine Pressure Regulating X K5.04 - Turbine inlet pressure vs. reactor 3.3 1 System / 3 pressure 259002 Reactor Water Level Control System / X A 1.05 - FWRV/startup level control position: 2.9 1 2 Plant-Specific 259002 Reactor Water Level Control System X A2.05 - Loss of applicable plant air systems 3.4 1 2

262001 A.C. Electrical Distribution / 6 X 2.2.21 - Knowledge of pre and post 3.5 maintenance operability requirements.

262001 A.C. Electrical Distribution / 6 X A4.01 - All breakers and disconnects (including 3.7 1 available switch yard): Plant-Specific 290001 Secondary Containment / 5 X K3.01 - tOff-site radioactive release rates 4.4* 1 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 6 of 13

( BWR SRO 1( nination Outline Printed: 06/( )02 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 / Group 1 Form ES-401-1 Sys/Ev # System / Evolution Name K1I K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 290001 Secondary Containment / 5 X 2.4.48 - Ability to interpret control room 3.8 1 indications to verify the status and operation of system, and understand how operator action s and directives affect plant and system conditions.

K/A Category Totals: 2 2 2 2 2 2 2 2 2 2 3 Group Point Total: 23 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 7 of 13

c BWR SRO l( nination Outline Printed: 06/(* 102 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 /Groun 2 Sys/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 201001 Control Rod Drive Hydraulic System X K4.1 1- Protection against filling the SDV 3.6 1 1 during non-SCRAM conditions 202001 Recirculation System / 1 X 2.1.33 - Ability to recognize indications for 4.0 1 system operating parameters which are entry-level conditions for technical specifications.

215002 Rod Block Monitor System / 7 X K5.01 - Trip reference selection: Plant-Specific 2.8 1 215002 Rod Block Monitor System / 7 X A1.01 - Trip reference: BWR-3, 4, 5 2.8 1 230000 RHR/LPCI: Torus/Suppression Pool X K3.02 - Suppression pool temperature 3.5 1 Spray Mode / 5 230000 RHR/LPCI: Torus/Suppression Pool X K5.01 - System venting 2.7 1 Spray Mode / 5 245000 Main Turbine Generator and Auxiliary X A2.09 - Turbine vibration 2.8 1 Systems / 4 272000 Radiation Monitoring System / 7 X K2.05 - Reactor building ventilation monitors: 2.9 1 Plant-Specific 272000 Radiation Monitoring System / 7 X A3.09 - Containment isolation indications 3.5 1 286000 Fire Protection System / 8 X K 1.09 - Emergency generator rooms: 3.3 1 Plant-Specific 290003 Control Room HVAC / 9 X 2.3.10 - Ability to perform procedures to reduce 3.3 1 excessive levels of radiation and guard against personnel exposure.

Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 8 of 13

( BWR SRO I nination Outline Printed: 06/( )02 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 / Group 2 I*nrm l*'*_Lln 1 _1 Sys/Ev # System / Evolution Name KI K2 K3 K4 KS K6 Al A2 A3 A4 G KA Topic Imp. Points 400000 Component Cooling Water System X K6.04 - Pumps 3.1 1 (CCWS) / 8 400000 Component Cooling Water System X A4.01 - CCW indications and control 3.0 1 (CCWS) / 8 1 K/A Category Totals: I I 1 1 2 1 1 1 1 1 2 Group Point Total: 13 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 9 of 13

! BWR SRO rnination m Outline Printed: 06/( 902 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 / Group 3 17*rrn l*-dfll-!

Sys/Ev # System / Evolution Name Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 201003 Control Rod and Drive Mechanism / I X 2.2.25 - Knowledge of bases in technical 3.7 1 specifications for limiting conditions for operations and safety limits.

201003 Control Rod and Drive Mechanism / 1 X K1.0I - Control rod drive hydraulic system 3.3 233000 Fuel Pool Cooling and Clean-up / 9 X K4.07 - Supplemental heat removal capability 2.9 233000 Fuel Pool Cooling and Clean-up / 9 X Al.07 - System temperature 2.8 K/A Category Totals: 1 0 0 1 0 0 1 0 0 0 1 Group Point Total: 4 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 10 of 13

Generic Knowledge ard Abilities Outline (Tier 3)

( Printed: 06/17/204(

BWR SRO Examination Outline Form ES-401-5 Facility: Monticello Nuclear Generating Generic Category KA KA Topic Imp. Points Conduct of Operations 2.1.4 Knowledge of shift staffing requirements. 3.4 1 2.1.22 Ability to determine Mode of Operation. 3.3 1 2.1.33 Ability to recognize indications for system operating parameters which are entry-level 4.0 1 conditions for technical specifications.

2.1.31 Ability to locate control room switches, controls and indications and to determine that 3.9 1 they are correctly reflecting the desired plant lineup.

Category Total: 4 Equipment Control 2.2.27 Knowledge of the refueling process. 3.5 1 2.2.32 Knowledge of the effects of alterations on core configuration. 3.3 1 2.2.30 Knowledge of RO duties in the control room during fuel handling such as alarms from 3.3 1 fuel handling area / communication with fuel storage facility / systems operated from the control room in support of fueling operations / and supporting instrumentation.

2.2.11 Knowledge of the process for controlling temporary changes. 3.4* 1 2.2.2 Ability to manipulate the console controls as required to operate the facility between 3.5 shutdown and designated power levels.

Category Total: 5 Radiation Control 2.3.9 Knowledge of the process for performing a containment purge. 3.4 1 2.3.2 Knowledge of facility ALARA program. 2.9 1 2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements. 3.0 1 2.3.7 Knowledge of the process for preparing a radiation work permit. 3.3 1 Category Total: 4 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 11 of 13

Generic Knowledge a-4 Abilities Outline (Tier 3)

( Printed: 06/17/204(

BWR SRO Examination Outline Form ES-401-5 Facility: Monticello Nuclear Generating Generic Category KA KA Topic Imp. Points Emergency Plan 2.4.38 Ability to take actions called for in the facility emergency plan, including (if 4.0 1 required)supporting or acting as emergency coordinator.

2.4.17 Knowledge of EOP terms and definitions. 3.8 1 2.4.8 Knowledge of how the event-based emergency/abnormal operating procedures are used in 3.7 1 conjunction with the symptom-based EOPs.

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm 3.3 1 response manual.

Category Total: 4 Generic Total: 17 Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 12 of 13

ES-401 BWR SRO Examination Outline Printed: 06/17/2002 Facility: Monticello Nuclear Generating Plant Form ES-401-1 Exam Date: 09/09/2002 Exam Level: SRO 7-7 K/A Category Points Tier Group Point I I Total KI K2 K3 K4 K5 K6 Al A2 A3 A4 G 1 4 4 5 4 5 4 26 Emergency

& 2 3 3 2 3 3 17 Abnormal Plant Tier Evolutions Totals 7 7 7 7 8 7 43 I 2 2 2 2 2 2 2 2 2 2 3 23 2.

2 1 1 1 1 2 1 1 1 1 1 2 13 Plant Systems 3 1 0 0 1 0 0 1 0 0 0 1 4 iler Totals 4 3 3 4 4 4 3 6 40 Cat I Cat 2 Cat 3 Cat 4

3. Generic Knowledge And Abilities 4 5 4 4 17 Note:

1. Attempt to distribute topics among all K/A Categories; select at least one topic from every K/A category within each tier.

2. Actual point totals must match those specified in the table.

3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.

4. Systems/evolutions within each group are identified on the associated outline.

5. The shaded areas are not applicable to the category tier.

Monticello Nuclear Generating Plant - BWR SRO Examination Outline - NRC Review Copy Page 13 of 13

Facility: IVL.. 2ello Nuclear Generating BWR RO ( mination Outline Printed: 06/17(

ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Groun I ]l*'r. rm I* qAO 1_")

E/APE # E/APE Name / Safety Function KI K2 K3 Al A2 G KA Topic Imp. Points 295006 SCRAM / I X AK1.01 - Decay heat generation and removal. 3.7 1 295006 SCRAM / 1 X AK3.03 - Reactor pressure response 3.8 1 295014 Inadvertent Reactivity Addition / I X AK1.01 - Prompt critical 3.7 1 295015 Incomplete SCRAM / 1 X AK3.01 - Bypassing rod insertion blocks 3.4 1 295024 High Drywell Pressure / 5 X EKL.01 - Drywell integrity: Plant-Specific 4.1 1 295024 High Drywell Pressure / 5 X EA2.02 - Drywell temperature 3.9 1 295025 High Reactor Pressure / 3 X EK1.04 - Decay heat generation 3.6 1 295031 Reactor Low Water Level / 2 X EA 1.06 - Automatic depressurization system 4.4* 1 295031 Reactor Low Water Level / 2 X EK2. 10 - Redundant reactivity control: Plant-Specific 4.0 1 295037 SCRAM Condition Present and Reactor Power Above X EK2.07 - Neutron monitoring system 4.0* 1 APRM Downscale or Unknown / I 295037 SCRAM Condition Present and Reactor Power Above X EK2.04 - SBLC system 4.4* 1 APRM Downscale or Unknown / I 500000 High Containment Hydrogen Concentration / 5 X EA2.03 - Combustible limits for drywell 3.3 1 500000 High Containment Hydrogen Concentration / 5 X EAL.05 - Wetwell sprays 3.3 1 K/A Category Totals: 4 3 2 2 2 0 Group Point Total: 13 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 1 of 12

(j.0,cello Nuclear Generating BWR RO ( mination Outline Printed: 06/17( ,

Facility:

ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 Form ES-401-2 E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points 295001 Partial or Complete Loss of Forced Core Flow X AK3.06 - Core flow indication 2.9 1 Circulation / 1 295002 Loss of Main Condenser Vacuum / 3 X AA 1.01 - Condensate system 2.6 1 295003 Partial or Complete Loss of A.C. Power / 6 X AK2.02 - Emergency generators 4.1

• 1 295003 Partial or Complete Loss of A.C. Power / 6 X AA1.01 - A.C. electrical distribution system 3.7 1 295008 High Reactor Water Level / 2 X AKI.03 - Feed flow/steam flow mismatch 3.2 1 295012 High Drywell Temperature / 5 X AK2.02 - Drywell cooling 3.6 1 295013 High Suppression Pool Temperature / 5 X AK2.01 - Suppression pool cooling 3.6 1 295017 High Off-Site Release Rate / 9 X AK3.0t - System isolations 3.6 1 295017 High Off-Site Release Rate / 9 X AK1.02 - f Protection of the general public 3.8* 1 295022 Loss of CRD Pumps / 1 X AK2.02 - CRD mechanism 3.1 1 295026 Suppression Pool High Water Temperature / 5 X EA1.03 - Temperature monitoring 3.9* 1 295026 Suppression Pool High Water Temperature / 5 X 2.1.10 - Knowledge of conditions and limitations in the 2.7 1 facility license.

295028 High Drywell Temperature / 5 X EA1.02 - Drywell ventilation system 3.9 1 295029 High Suppression Pool Water Level / 5 X EK1.01 - Containment integrity 3.4 1 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 2 of 12

Facility: 4 ,.-cello Nuclear Generating BWR RO ( mination Outline Printed: 06/17(

ES - 401 Emergency and Abnormal Plant Evolutions - Tier I / Group 2 Form ES-401-2 E/APE E/APE Name / Safety Function KI K2 K3 Al A2 G KA Topic Imp. Points 295033 High Secondary Containment Area Radiation Levels / X EK2.01 - Area radiation monitoring system 3.8 1 9

295033 High Secondary Containment Area Radiation Levels / X EA1.04 - SBGT/FRVS 4.2* 1 9

295034 Secondary Containment Ventilation High Radiation / X EK2.04 - Secondary containment ventilation 3.9 9

295038 High Off-Site Release Rate /9 X EK2.03 - Plant ventilation systems 3.6 1 295038 High Off-Site Release Rate / 9 X EA1.03 - Process liquid radiation monitoring system 3.7 1 K/A Category Totals: 3 7 2 6 0 1 Group Point Total: 19 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 3 of 12

Facility: Molicello Nuclear Generating BWR RO ( mination Outline Printed: 06/17ý ES - 401 Emergency and Abnormal Plant Evolutions - Tier I / Groun 3 Pnrm l*_Afi 1_9 E/APE # E/APE Name / Safety Function K1 K2 K33 Al A2 G KA Topic Imp. Points 295021 Loss of Shutdown Cooling / 4 X AK3.01 - Raising reactor water level 3.3 1 295023 Refueling Accidents / 8 X AA1.02 - Fuel pool cooling and cleanup system 2.9 1 295032 High Secondary Containment Area Temperature / 5 X 2.4.48 - Ability to interpret control room indications to 3.5 1 verify the status and operation of system, and understand how operator action s and directives affect plant and system conditions.

295035 Secondary Containment High Differential Pressure / 5 X EK1.01 - Secondary containment integrity 3.9 1 K/A Category Totals: 1 0 S1 0 1 Group Point Total: 4 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 4 of 12

( BWR RO E( ination Outline Printed: 06/( '02 Facility: Monticello Nuclear Generating ES - 401 Plant Svstems - Tier 2 / Groun I I*r* I*_AI'*I "9 Sys/Ev# System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 201001 Control Rod Drive Hydraulic System X K4.11 - Protection against filling the SDV 3.6 1 1 during non-SCRAM conditions 201002 Reactor Manual Control System / 1 X A3.01 - Control rod block actuation 3.2 1 202002 Recirculation Flow Control System / I X K1.03 - Reactor core flow 3.7 1 202002 Recirculation Flow Control System / 1 X K3.02 - Reactor power 4.0 1 203000 RHR/LPCI: Injection Mode (Plant X K6.05 - Condensate storage and transfer system: 2.5 1 Specific) / 2 Plant-Specific 203000 RHR/LPCI: Injection Mode (Plant X 2.4.27 - Knowledge of fire in the plant 3.0 1 Specific) / 2 procedure.

209001 Low Pressure Core Spray System / 2 X A4.01 - Core spray pump 3.8 1 211000 Standby Liquid Control System / I X A2.01 - Pump trip 3.5 1 211000 Standby Liquid Control System / 1 X A1.02 - Explosive valve indication 3.8 1 215004 Source Range Monitor (SRM) System X K2.01 - SRM channels/detectors 2.6 1

/7 215004 Source Range Monitor (SRM) System X A4.01 - SRM count rate and period 3.9 1

/7 215005 Average Power Range Monitor/Local X K2.02 - APRM channels 2.6 Power Range Monitor System / 7 1 1 1 1 11 1 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 5 of 12

BWR RO E( ination Outline Printed: 06/( 02 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 / Group 1 l*nrm I?.*-All 1 _9 If Form ES~-40l1 -1 Iron. Pnint*

Sys/Ev # System / Evolution Name K1 K2 1K3 jK4 1K5 K6 AlI A2 A3 A4 G KA Topic 215005 Average Power Power Range Range Monitor/Local . - K4.01 -Rod withdrawal blocks 3.7 1 Monitor System / 7 216000 Nuclear Boiler Instrumentation / 7 X K 1.09 - Redundant reactivity control! alternate 3.7 1 rod insertion; Plant-Specific 216000 Nuclear Boiler Instrumentation / 7 X K3.30 - Recirculation system 3.2 1 218000 Automatic Depressurization System / X A1.03 - ADS valve air supply pressure: 3.2 1 3 Plant-Specific 223001 Primary Containment System and X K6.12 - D.C. electrical distribution 2.7 1 Auxiliaries / 5 223002 Primary Containment Isolation X K4.05 - Single failures will not impair the 2.9 1 System/Nuclear Steam Supply function ability of the system Shut-Off/ 5 223002 Primary Containment Isolation X A3.01 - System indicating lights and alarms 3.4 1 System/Nuclear Steam Supply Shut-Off/ 5 239002 Relief/Safety Valves / 3 X K5.06 - Vacuum breaker operation 2.7 1 241000 Reactor/Turbine Pressure Regulating X K3.02 - Reactor pressure 4.2* 1 System / 3 241000 Reactor/Turbine Pressure Regulating X K5.04 - Turbine inlet pressure vs. reactor 3.3 1 System / 3 pressure 259002 Reactor Water Level Control System / X A1.05 - FWRV/startup level control position: 2.9 1 2 Plant-Specific Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 6 of 12

( BWR RO E( iination Outline Printed: 06/( 102 Facility: Monticello Nuclear Generating ES - 401 Plant Svstems - Tier 2 / Gropn 1 l'nRrn, Ii' VAflhi Sys/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imp. Points 259002 Reactor Water Level Control System / X A2.05 - Loss of applicable plant air systems 3.2 1 2

261000 Standby Gas Treatment System / 9 X K1.02 - Drywell 3.2 1 261000 Standby Gas Treatment System /9 X A2.01 - Low system flow 2.9 1 264000 Emergency Generators (Diesel/Jet) / 6 X K5.05 - Paralleling A.C. power sources 3.4 1 264000 Emergency Generators (Diesel/Jet) / 6 X A4.05 - Transfer of emergency generator (with 3.6 1 load) to grid I II K/A Category Totals: 3 2 3 3 3 2 3 3 2 3 1 Group Point Total: 28 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 7 of 12

( BWR RO 1 iination Outline Printed: 06/( 902 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 / Group 2 Form ES-401-2 Sys/Ev # System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 201003 Control Rod and Drive Mechanism / 1 X K1.01 - Control rod drive hydraulic system 3.2 1 204000 Reactor Water Cleanup System / 2 X A4.09 - Reactor water temperature 2.9 1 205000 Shutdown Cooling System (RHR X A 1.08 - Heat exchanger temperatures 3.1 1 Shutdown Cooling Mode) / 4 215002 Rod Block Monitor System / 7 X K5.01 - Trip reference selection: Plant-Specific 2.6 1 215002 Rod Block Monitor System /7 X A 1.01 - Trip reference: BWR-3, 4, 5 2.7 1 219000 RHR/LPCI: Torus/Suppression Pool X K4.10 - Prevention of leakage to the 3.3 1 Cooling Mode / 5 environment through system heat exchanger:

Plant-Specific 226001 RHR/LPCI: Containment Spray X A3.05 - Containment pressure 4.0* 1 System Mode / 5 230000 RHRiLPCI: Torus/Suppression Pool X K3.02 - Suppression pool temperature 3.3 1 Spray Mode / 5 230000 RHR/LPCI: Torus/Suppression Pool X K5.01 - System venting 2.6 1 Spray Mode / 5 245000 Main Turbine Generator and Auxiliary X A2.09 - Turbine vibration 2.5 1 Systems / 4 262001 A.C. Electrical Distribution / 6 X A4.01 - All breakers and disconnects (including 3.4 1 available switch yard): Plant-Specific 271000 Offgas System / 9 X K4.01 - Dilution of hydrogen gas concentration 2.9 1 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 8 of 12

( BWR RO E tiination Outline Printed: 06/( )02 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 / GrouD 2 Fnrm i*+,qAfil _9.

Sys/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA.Topic Imp. Points 272000 Radiation Monitoring System / 7 X K2.05 - Reactor building ventilation monitors: 2.6 1 Plant-Specific 272000 Radiation Monitoring System /7 X A3.09 - Containment isolation indications 3.6 286000 Fire Protection System / 8 X K1.09 - Emergency generator rooms: 3.2 1 Plant-Specific 290001 Secondary Containment / 5 X K3.01 - f Off-site radioactive release rates 4.0 1 290003 Control Room HVAC / 9 X K6.02 - Component cooling water systems 2.7 1 400000 Component Cooling Water System X K6.04 - Pumps 3.0 1 (CCWS) / 8 400000 Component Cooling Water System X A4.01 - CCW indications and control 3.1 1 (CCWS) / 8 K/A Category Totals: 2 1 2 2 2 2 2 1 2 3 0 Group Point Total: 19 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 9 of 12

( BWR RO iiination Outline Printed: 06/( 302 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 / Group 3 Form ES-401-2 Sys/Ev # System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imp. Points 233000 Fuel Pool Cooling and Clean-up / 9 X K4.07 - Supplemental heat removal capability 2.7 1 233000 Fuel Pool Cooling and Clean-up /9 X A1.07 - System temperature 2.7 1 288000 Plant Ventilation Systems / 9 X K1.06 - Plant air systems 2.7 1 290002 Reactor Vessel Internals / 5 X 2.2.23 - Ability to track limiting conditions for 2.6 1 L_ _ _ _ __ Ioperations.

K/A Category Totals: 1 0 0 1 0 0 1 0 0 0 I Group Point Total: 4 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 10 of 12

Generic Knowledge a41- Abilities Outline (Tier 3)

( Printed: 06/17/200(

BWR RO Examination Outline Form ES-401-5 Facility: Monticello Nuclear Generating Generic Category KA KA Topic Imp. Points Conduct of Operations 2.1.33 Ability to recognize indications for system operating parameters which are entry-level 3.4 1 conditions for technical specifications.

2.1.31 Ability to locate control room switches, controls and indications and to determine that 4.2 1 they are correctly reflecting the desired plant lineup.

2.1.29 Knowledge of how to conduct and verify valve lineups. 3.4 1 Category Total: 3 Equipment Control 2.2.30 Knowledge of RO duties in the control room during fuel handling such as alarms from 3.5 1 fuel handling area / communication with fuel storage facility / systems operated from the control room in support of fueling operations / and supporting instrumentation.

2.2.11 Knowledge of the process for controlling temporary changes. 2.5 1 2.2.2 Ability to manipulate the console controls as required to operate the facility between 4.0 1 shutdown and designated power levels.

2.2.22 Knowledge of limiting conditions for operations and safety limits. 3.4 1 Category Total: 4 Radiation Control 2.3.2 Knowledge of facility ALARA program. 2.5 1 2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements. 2.6 1 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against 2.9 1

_____________________________personnel exposure. I___I Category Total: 3 Emergency Plan 2.4.17 Knowledge of EOP terms and definitions. 3.1 1 2.4.8 Knowledge of how the event-based emergency/abnormal operating procedures are used in 3.0 1 conjunction with the symptom-based EOPs.

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm 3.3 1

____________________________response manual.I Category Total: 3 Generic Total: 13 Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 11 of 12

ES-401 BWR RO Examination Outline Printed: 06/17/2002 Facility: Monticello Nuclear Generating Plant Form ES-401-2 Exam Date: 09/09/2002 Exam Level: RO I I K/A Category Points Tier Group l/

1NI I If *1 Ný

[7 N.5 K4 K5 Al A2 A3 A4 Point K6 G Total 1 4 3 2 2 2 0 13 1.

Emergency 2 3 7 2 6 0 19 Abnormal Plant 3 1 0 1 0 1 4 Evolutions Totals 8 10 5 9 2 ,* $ 2 36 Tier I 3 2 3 3 3 2 3 3 2 3 28 2.

2 2 1 2 2 2 2 2 1 2 3 0 19 Plant Systems 3 1 0 0 1 0 0 1 0 0 0 1 4 I 1Tt1 Totals 6 5 6 5 4 6 4 4 6 2 51 Cat I Cat 2 Cat 3 Cat 4

3. Generic Knowledge And Abilities 3 4 3 3 13 Note:

1. Attempt to distribute topics among all K/A Categories; select at least one topic from every K/A category within each tier.

2. Actual point totals must match those specified in the table.

3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.

4. Systems/evolutions within each group are identified on the associated outline.

5. The shaded areas are not applicable to the category tier.

Monticello Nuclear Generating Plant - BWR RO Examination Outline - NRC Review Copy Page 12 of 12

Monticello Nuclear Generating Plant 2002 NRC Operator Licensing Exam Exam Bank Description To assist in your review the following description is provided concerning information provided for each question.

1. Exam field - The exam field identifies whether the question is to be used on the RO or SRO licensing examination. This field contains the following codes.

"* BOTH - the K/A for the question appears on the examination outlines for both the RO and SRO exams and the question is to be used on both the RO and SRO exams.

"* RO - the K/A for the question appears only on the examination outline for the RO exam and the question is be used only on the RO exam.

" SRO - the K/A for the question appears only on the examination outline for the SRO exam and the question is be used only on the SRO exam and was written at the SRO level of knowledge.

" BRO - the K/A for the question appears on the examination outlines for both the RO and SRO exams, the BRO question was written at the RO level of knowledge and is to be used only on the RO exam.

" BSRO - the K/A for the question appears on the examination outlines for both the RO and SRO exams, the BSRO question was written at the SRO level of knowledge and is to be used only on the SRO exam.

" SRO-RO - the K/A for the question appears only on the examination outline for the SRO exam, the SRO-RO question was written at the RO level of knowledge and is to be used only on the SRO exam.

2. System and K/A fields - reference is provided to the NUREG 1123 system/evolution and K/A statements randomly selected for inclusion in the examination outline and addressed by the question.

3. Lesson Plan and Enabling Objective fields - reference is provided to the Monticello lesson plan and learning objective applicable to the question. Where a lesson plan and learning objective was not identified as applicable to the question, then "NONE" is indicated.

4. Reference field - reference description is provided to the supporting technical information for the question. As noted above, copies of the supporting technical information is provided following each question.

5. Question Pedigree field - a cross reference is provided to the Monticello or industry examination bank to indicate the origin of the question. Where a question did not originate from a bank, then "NEW" is indicated.

Page 2 of 14

Monticello Nuclear Generating Plant 2002 NRC Operator Licensing Exam Exam Bank Description

6. Cog Level field - identifies the cognitive level of the question. This field contains the following codes.

0 1 - Level 1 cognitive level (i.e., fundamental knowledge or simple memory).

• 2 - Level 2 cognitive level (i.e., comprehension).

• 3 - Level 3 cognitive level (i.e., analysis, synthesis, or application).

7. Answer and ANSWER/DISTRACTER JUSTIFICATION fields - The "Answer field provides the letter designator for the correct answer. The ANSWER/DISTRACTER JUSTIFICATION field provides justification for the correct answer, as well as justifications as to why the distracters are incorrect.

8. COGNITIVE LEVEL JUSTIFICATION field -justification is provided for the identified cognitive level.

9. SRO LEVEL JUSTIFCATION field - for questions written to the SRO level of knowledge, justification is provided as to why the question is at the SRO knowledge level.

Page 3 of 14

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 1 E)Cam: BOTH System: 201001 K/A: K4.11 Lesson Plan M-8107L-020 Enablina Obiective 7.e Which of the following states the purpose for the Scram Discharge Volume High Level Scram?

A. A Reactor scram is inserted before the Scram Discharge Volume fills and the ability to complete a scram is lost.

B. A Reactor scram is inserted to isolate an open flow path connecting the primary coolant boundary to Secondary Containment.

C. A level increase in the Scram Discharge Volume is indicative of multiple leaking scram outlet valves.

D. A level increase in the Scram Discharge Volume is indicative of primary coolant pressure boundary leakage greater than Tech Spec limits.

Answer: A

Reference:

B.5.6-02.A.1 .f Question Pedigree: INPO Bank QID 8938 Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Per B.5.6-02.A. .f, should the SDV fill with water to the point where not enough space remains for the water displaced during a scram, control rod movement would be hindered in the event of an actual scram. To prevent this, the Rx is scrammed when the water attains a level high enough to verify the volume is filling up yet low enough to ensure that the remaining capacity in the SDV can accommodate a scram.

B. Plausible answer as the SDV does drain to Radwaste Drain Tank located in secondary containment; however this is not stated as the function of the SDV High Level Scram.

Shutting the SDV vent and drain valves can isolate the leakage flow path as occurs on a Reactor Scram and the flow path does not in and of it self necessitate a scram.

C. Plausible answer as per B. 1.3-01. B. 12, a gross measure of seat leakage through the scram outlet valves may be obtained by closing the discharge volume vent and drain valves and observing the length of time required for SDV not drained alarm to annunciate. However this is not stated as the function of the SDV High Level Scram and the drain capability of the SDV could maintain the SDV level below that which impacts the ability to scram.

D. Plausible answer as level increase in the SDV would be indicative of leakage from the RCS and SDV is part of the primary system boundary following a scram; however this leakage path does not meet the TS definition of primary coolant pressure boundary leakage. Shutting the SDV vent can isolate the leakage flow path as occurs on a Reactor Scram and the flow path does not in and of it self necessitate a scram.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recall purpose of system automatic action.

Page 1 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 2 Exam: RO System: 201002 K/A: A3.01 Lesson Plan M-8107L-032 Enablina Obiective Rh Given the following plant conditions.

"* Reactor start-up is in progress with the Mode Switch in STARTUP-TO-HOTSTBY.

"* IRMs are on range 5.

"* While hanging an isolation on Panel Y-20, an Operator incorrectly OPENs the breaker for Circuit 26, SRM/IRM DRIVE CONTROL, causing a loss of SRM and IRM FULL IN/FULL OUT indication.

What is the impact of losing Panel Y-20, Circuit 26 AND why?

A. A half scram signal occurs due to SRMs Inoperable (INOP).

B. A full Reactor scram occurs due to IRMs Inoperable (INOP).

C. A Control Rod withdraw block occurs due to the loss of SRM Full IN indication.

D. A Control Rod withdraw block occurs due to the loss of IRM Full IN indication.

Answer: D

Reference:

B.5.2.2-02.C.l.c & C.4-B.9.13.B Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. An SRM scram is generated by an SRM Hi-Hi signal with SRM shorting links removed, half-scram occurs with SRM shorting links installed on SRM HI-HI.

B. The IRM INOP scram occurs due to IRM mode switch and module interlocks or loss of detector power. IRM scram is bypassed with RX mode switch in Run.

C. With the Mode Switch in Startup to Hot Standby, the SRM not fully inserted rod withdraw block is bypassed if the SRM count level is greater than 100 cps or the IRM range switches are on range 3 or greater, thus the answer is incorrect.

D. A rod withdraw block occurs if the Mode Switch is in Startup to Hot Standby and an IRM channel detector is not fully inserted in the core.

COGNITIVE LEVEL JUSTIFICATION Level 2 based on 1) the examinee must recognize the loss of SRM and IRM detector not full in results in a rod withdraw block and 2) the examinee must recognize the relationship between the Power Range Monitoring System and the RMCS that with the IRMs on range 3 or greater the SRM not full in rod block is bypassed.

Page 2 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 3 Exam: SRO System: 201003 K/A: 2.2.25 Lesson Plan M-8107L-021 Enablina Obiective Given the following plant conditions.

"* The plant is at 100% power.

"* Surveillance Procedure 0074, CONTROL ROD DRIVE EXERCISE, is in progress to perform weekly Control Rod exercise checks.

"* Control Rod 22-19 was at position 30 at the start of the procedure.

"* The operator has attempted to insert and withdraw Control Rod 22-19 with elevated drive water pressures up to 400 psig and no response was obtained with Control Rod 22-19 remaining at position 30.

"* The operator has attempted to single rod scram Control Rod 22-19 and no response was obtained.

"* All conditions have been verified to be normal for the HCU for Control Rod 22-19.

"* All remaining partially withdrawn rods demonstrate proper Control Rod exercising.

What operator action is required AND what is the basis for this action?

A. Electrically disarm Control Rod 22-19, verify shutdown margin, and continue plant operation. Operation may continue with the shutdown reactivity limits satisfied for the ability to shutdown with one rod out.

B. Place the plant in a Hot Shutdown condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. CRD collet housing failure may have occurred which could prevent a scram of the Control Rod.

C. Perform Control Rod exercise checks daily and continue plant operation. Increased surveillance will assure that the Reactor is not operated with a large number of inoperable Control Rods.

D. Place the plant in a Cold Shutdown condition with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The rod is uncoupled and a rod drop accident could occur for this partially withdrawn Control Rod.

Tech Spec to be provide as open reference question without bases.

Answer: B

Reference:

B.1.3-05.H.1 & TS 3/4.3.A.2.b and bases Question Pedigree: New Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Per TS 3.3.A.2(a), shut down is not required if rod is inoperable for reasons other than collet housing failure. Per TS 3.3.2.A.2(c), plant operation may continue with up to 6 inoperable control rods for reasons other than collet housing failure. However, the provided information of inability to move rod from original position with elevated drive pressure or single rod scram indicates CRD collet housing failure per B.1.3-05.H.1, thus action statement per TS 3.3.2.A.2(a) and TS 3.3.A.2(c) are not applicable.

B. The provided information of inability to move rod from original position with elevated drive pressure or single rod scram indicates CRD collet housing failure per B.1.3 05.H.1. TS 3.3.2.A(b) directs hot shutdown in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based on failure can prevent a rod scram and indicates a generic problem.

C. Per TS 4.3.2.A(b) surveillance frequency is to daily for stuck rod because of collect housing failure. However TS 3.3.2.A(b) required hot shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based on failure can prevent a rod scram and indicates a generic problem.

I nbJP. bie.tive 10d Page 3 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank D. Per TS 3.3.B & 3.3.G, Cold shutdown is required in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the actions of specifications 3.3.A through 3.3.D can not be met. The provided information of inability to move rod from original position with elevated drive pressure or single rod scram indicates CRD collet housing failure per B.1.3-05.H.1, not an uncoupled CRD.

COGNITIVE LEVEL JUSTIFICATION Level 2 Recognize Interactions - examinee must recognize 1) elevated drive water pressure is being used to position the rod with out response. 2) if CRD has not moved from its original position that CRD collet housing failure is indicated, and 3) must recognize Tech Spec required action and bases.

SRO LEVEL JUSTIFICATION SRO knowledge level question based on examinee must recall plant Technical Specification Bases information.

"lPag 4Jof156 Page 4 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 4 Exam: BOTH System: 201003 K/A: K1.01 Lesson Plan M-8107L-020 Enablina Objective Reactor power is 100%. NO CRD flow is available, and two CRD accumulator low pressure alarms are received.

Why is a manual Reactor Scram required?

A. The Control Rod mechanism temperatures will begin increasing.

B. The Control Rod accumulator Nitrogen pressure is decreasing.

C. The Control Rod Scram times may be slow.

D. The Control Rod will begin to drift out.

Answer: C

Reference:

C.4-B.1.3.A Question Pedigree: ILT BANK M8107L- Cog. Level: 1 020-006 ANSWER/DISTRACTER JUSTIFICATION A. Control rod temperatures will begin to increase do to loss of cooling water flow, however a scram is inserted to preserve plant safety analysis.

B. Accumulator Nitrogen pressure will decrease due to loss of charging water flow, however a scram is inserted to preserve plant safety analysis.

C. The concern for loss of CRD flow is the inability to maintain accumulator pressure with charging water. If the accumulator discharges and a scram is required, the rods may not insert into the core at the required scram rate, or at all, depending on Reactor pressure. Therefore, the operator is instructed to scram the Reactor after receipt of the second accumulator low pressure alarms.

D. Control rod drift out would be caused be collet failure which requires reactor shutdown to hot shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

COGNITIVE LEVEL JUSTIFICATION Level 1: Procedure bases for C.4-B. 1.3.A, Loss of CRD Pump Flow.

[J T, KI) 1-7 I I.

Hi i S Page 5 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 5 Exam: SRO System: 202001 K/A: 2.1.33 Lesson Plan M8107L-029 Enabling Objective 11.d While operating at 100% power, a lockout occurs on Bus No. 11. Reactor power decreases to 60%. Core and Recirculation Loop flows are as follows.

"* A Recirculation Loop Flow is 0 GPM.

"* A Recirculation Jet Pump Flow is 3.0 X 106 Ibm/hr.

"* B Recirculation Loop Flow is 33,000 GPM.

"* B Recirculation Jet Pump Flow is 26.0 X 106 Ibm/hr.

"* Total Core Flow is 30.5 X 106 Ibm/hr.

1) What operator action is required AND 2) what mechanism is in place to prevent exceeding fuel safety limits?

A. 1) Insert Control Rods to exit the Power-Flow Operating Map Stability Exclusion region.

2) Power distribution controls.

B. 1) Insert Control Rods to exit the Power-Flow Operating Map Stability Exclusion region.

2) The APRM Flow biased scram.

C. 1) Insert Control Rods to exit the Power-Flow Operating Map Stability Buffer region.

2) Power distribution controls.

D. 1) Insert Control Rods to exit the Power-Flow Operating Map Stability Buffer region.

2) The APRM Flow biased scram.

Answer: D

Reference:

C.4-B.5.1.2.A & TS 3.5.F.2 and Bases Question Pedigree: MOD ILT Bank Cog. Level: 2 M8114L-002-040 ANSWER/DISTRACTER JUSTIFICATION A. For the conditions provided, the plant has entered the stability buffer region, not the stability exclusion region, thus the 1 st part of the answer is incorrect. Per TS bases the APRM flow biased scram will suppress oscillation prior to exceeding the fuel safety limit.

B. Distracter is a precaution for single loop operations and is thus plausible. Decrease in pump speed is inconsistent with actions for entry to stability buffer region.

C. Distracter is consistent for actions required by TS 4.6.G.1 for unexplained changes in core flow. However, core flow change is a result of loss of A Recirc Pump and is thus not unexplained.

D. Loss of Buss 11 results in loss of A Recic Pump. Immediate action per trip of one recirc pump is to enter C.4-B.5.1.2.A. Per C.4-B.5.1.2.A, If core flow is < 26 Mlbm/hr, then determine if stability buffer or exclusion regions have been entered. Procedure directs to determine core flow with one Recirc pump running by subtracting idle loop jet pump flow from running loop jet pump (26 Mlbm/hr - 3 MIb/hr = 23 Mlbm/hr). Plant is in stability buffer region, TS direct to insert rods or increase recirc speed to exit with basis being to prevent thermal-hydraulic instability.

I £1 i¶'iI1 T Ji I7 JO

• t.17 'F Page 6 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank COGNITIVE LEVEL JUSTIFICATION Level 2 - Rocognize Interactions: Examinee must 1) recognize loss of A Recirc Pump,

2) recognize core flow indication relationship to loss of A Recirc Pump AND determine proper core flow based on indications, 3) determine entry to stability buffer region and required actions with bases.

SRO Level Justification SRO level based on examinee must recall procedure and plant tech spec bases.

i 2 '7iF<jIi,7PIi EI7:i IPage 7]of 156 Page 7 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 7 Exam: BOTH System: 202002 KIA: K1.03 Lesson Plan M-8107L-047 Enablinq Obiective 5.b & 7.e While the plant is operating at 100% power, a trip of No. 11 Reactor Feed Pump occurs.

Reactor total core flow and Reactor power have decreased. Reactor water level is +20 inches and decreasing. The plant responds as designed and no operator actions have been taken.

What caused the decrease in Reactor total core flow AND what is the basis for the system response?

A. Recirculation MG Set field breakers tripped to decrease total core flow and Reactor power to prevent Reactor Feed Pump run out and a Reactor scram on low Reactor water level.

B. Runback of the Reactor Recirculation Pumps to 50% pump speed decreased total core flow and Reactor power to prevent Reactor Feed Pump run out and a Reactor scram on low Reactor water level.

C. Recirculation MG Set field breakers tripped to decrease total core flow and Reactor power to mitigate the consequences if a failure to scram occurs during the transient.

D. Runback of the Reactor Recirculation Pumps to 50% pump speed decreased total core flow and Reactor power to mitigate the consequences if a failure to scram occurs during the transient.

Answer: B

Reference:

B.5.8-01 .C & C.4-B.6.5.A Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Per B. 1.4-02.F.2, Recirculation MG field breakers trip at -47 inches Reactor level in response for the ATWS scram logic. Distracter is plausible, but fails to be supported by provided information as Rx level is stated as +20 inches and decreasing. In addition, the distracter does not support the basis for Recirc MG field breaker trip.

B. Per B.5.8-01 .C, automatic recirc runback to 50% occurs if feed pump or cond pump trips with steam flow > 60% and both recirc pumps running. Per C.4-B.6.5.A bases, failure to reduce power will result in pump runout and trip of the operating feed water pump and may result in Rx Scram at +9 inches.

C. Per B.1.4-02.F.2, Recirculation MG field breakers trip at -47 inches Reactor level in response for the ATWS scram logic. Per B.5.6-01, basis for ATWS is to mitigate failure of RPS to shutdown the RX (i.e., scram) when required. Distracter is plausible, but fails to be supported by provided information as Rx level is stated as +20 inches and decreasing.

D. Per B.5.8-01.C, automatic recirc runback to 50% occurs if feed pump or cond pump trips with steam flow > 60% and both recirc pumps running. The distracter does not provide the basis for the automatic recirc runback.

COGNITIVE LEVEL JUSTIFICATION Level 2 - examinee must 1) recognize interaction between systems concerning the feedpump trip and recirc runback vs Reactor level and Recirc MG set field breaker trip,

2) recognize interaction between recirc pump speed and total core flow, and 3) recognize basis for recirc pump runback.

Page 8 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 8 Ex:am: RO System: 202002 K/A: K3.02 Lesson Plan M-8107L-029 Enabling Objective 9.j M-81 08L-038 The plant is operating at 100% power. The No. 11 Recirc MG Set Scoop Tube has been locked due to unstable speed control on the No. 11 Recirc MG Set and cannot be reset.

I&C is investigating the speed control problem in accordance with a Work Order for the troubleshooting activity. The No. 11 Recirc MG Set speed is to be lowered 5% to assist troubleshooting per the Work Order.

Who may perform this action AND why?

A. A Reactor Building Operator because the manipulation of equipment in the Reactor Building is required.

B. A Control Room Operator because the manipulation of switches on Control Room panel C-04 is required.

C. An I&C Technician because the Work Order directs and controls this action as part of troubleshooting.

D. A Licensed Operator because the speed change is a reactivity manipulation effecting Reactor power.

Answer: D

Reference:

B.1.4-05.H.1 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. See correct answer.

B. See correct answer.

C. See correct answer.

D. B.1.4-05.H.1, local pump speed changes are to be performed by a licensed operator.

COGNITIVE LEVEL JUSTIFICATION Level 2 - based on examinee must recognize system interaction between core flow (Recirc system) and core power, and implications.

AMPage 1] 9ofif 56 Page 9 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 9 E)cam: SRO-RO System: 203000 K/A: 2.1.7 Lesson Plan M-8107L-023 Enablino Obi~ctiv*.

Enablinn ONective 7r.. 1 7i.1 7k1 Given the following plant conditions.

S A Reactor scram initiated from 100% power.

S Drywell pressure is 2.2 psig.

0 Reactor water level is 0 inches and increasing.

0 Reactor pressure is 550 psig.

S ECCS systems have initiated.

0 MO-2-53A, 11 Recirc Pump Discharge, is OPEN.

0 MO-2-53B, 12 Recirc Pump Discharge, is CLOSED.

What is the current status of LPCI injection, AND which Recirculation loop has been selected for LPCI injection?

A. LPCI is injecting into the 'A' Recirculation loop.

B. LPCI is NOT injecting and the 'A' Recirculation loop has been selected for injection.

C. LPCI is injecting into the 'B' Recirculation loop.

D. LPCI is NOT injecting and the 'B' Recirculation loop has been selected for injection.

Answer: D

Reference:

B.3.4-02 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. LPCI is not injecting because Rx Pressure is greater than low pressure permissive of 460 psig.

B. LPCI is not injecting because Rx Pressure is greater than low pressure permissive of 460 psig. The B Recirculation loop has been selected for LPCI injection based on MO-2-53B CLOSED from the LPCI loop selection logic.

C. LPCI is not injecting because Rx Pressure is greater than low pressure permissive of 460 psig.

D. LPCI is not injecting because Rx Pressure is greater than low pressure permissive of 460 psig. The B Recirculation loop has been selected for LPCI injection based on MO-2-53B CLOSED from the LPCI loop selection logic.

COGNITIVE LEVEL JUSTIFICATION Level 1 - based on: 1) examinee must recognize interlocks and set points for LPCI initiation and 2) examinee must recognize interrelationships between LPCI logic and Recirc system however interrelationship concept is taught in close proximity to other LPCI interlocks..

Page 10 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 10 Exam: RO System: 203000 K/A: 2.4.27 Lesson Plan M-8107L-040 Enabling Objective 4.b M-8107L-023 4.b & 7.b The plant is operating at 100% power. A fire has occurred in the Turbine Building 911 foot Elevation East MCC room in MCC-1 12. The Fire Brigade Leader has notified the control room that the following breakers are to be OPENED to support combating the fire per the fire fighting strategies.

"* Breaker 52-304, MCC-133A FEEDER BREAKER

"* Breaker 52-307, MCC-133B FEEDER BREAKER

"* Breaker 52-903, MCC-1 12 FEEDER BREAKER What impact do these breaker operations have on the status of energized equipment in the vicinity of the fire, AND what means of fire suppression should be used?

A. MCC-133B will re-energize from its alternate source. Portable water extinguishers may be used.

B. MCC-133B will re-energize from its alternate source. Water applied with a fog nozzle may be used.

C. The MCCs are de-energized. Carbon Dioxide (C02) extinguishers may be used.

D. The MCCs are de-energized. Portable dry chemical extinguishers may be used.

Answrer: B

Reference:

A-3-004.1.A.4.1, A.3-13-C, B.9.7-02.B Ques tion Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. See justification for the correct answer.

B. B.9.7-02.B, MCC-1 33B and MCC-1 34B are hard wired together forming the LPCI swing bus and is normally powered from LC-103 via 52-307 and B3300. On loss of voltage sustained for 12 sec, the LPCI swing bus auto transfers to power from LC-104. Thus MCC-1 33B will re-energize on opening 52-307. Per fire fighting strategies, portable water extinguishers are not to used on electrical fires, but water may be applied with fog nozzles.

C. See justification for the correct answer.

D. See justification for the correct answer.

COGNITIVE LEVEL JUSTIFICATION Level 2 - based on examinee must recognize the following relationships 1) on opening breaker 52-307, cross-tied MCC-1 33B and MCC-143B transfer to the alternate source with 12 sec time delay to be powered from LC-1 04 via breakers52-407 and B4300, 2) recall action for fighting fire near energized equipment.

1Page III;of156 Page 11 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 11 Excam: BOTH System: 203000 K/A: K6.05 Lesson Plan M-8107L-023 Enabling Objective 8.d M-8107L-094 4 Given the following plant conditions.

"* The plant is operating at 100% power.

"* Condensate Service Pump P-60A has been isolated due to excessive pump seal leakage.

"* Condensate Service Pump P-60B is in service to support backwash of a Condensate Filter Demineralizer.

"* Annuciator 8-C-34, No. 104 480 LDCTR MCC FEEDER TRIP, alarms.

"* The Control Room team determines that MCC-141 has been lost.

What impact does the loss of MCC-141 have on the LPCI mode of RHR?

A. A Group 2 Primary Containment Isolation will cause LPCI Inboard Isolation Valves MO 2014 and MO-2015 to CLOSE.

B. The Condensate Service jockey pump will auto start to maintain Condensate Service pressurizing station pressure which maintains the RHR system filled with water.

C. LPCI Outboard Isolation Valve MO-2013 will NOT open on a LPCI initiation signal thus the B LPCI injection path is NOT operable.

D. The Condensate Service pressurizing stations for RHR will be unable to maintain the RHR system filled with water and a water hammer could occur on LPCI initiation.

Answer: D

Reference:

B.3.4-05.A.2.a & B.8.9-05.C Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Distracter is plausible, C.4-B.9.7.D, Table MCC-141 indicates a partial group 2 will occur on loss of MCC-141. MO-2014 and MO-2015 receive close signals if SDC is in service. However, SDC is not in service and valves do not close on the partial group 2 signals. (Ref B.3.4-02.F.5)

B. Distracter is plausible, Cond Serv Jockey pump is normally operated to maintain system pressure with back wash operations are not in progress; however, power is lost to the jockey pump on loss of MCC-141.

C. Distracter is plausible, MO-2013 is normally closed however the MOV is powered from MCC-143B not MCC-141.

D. Per B.8.9-05.C, power supplies to Cond Serv Pumps are as follows: P-60A from MCC 111, P-60B from MCC141, Jockey pump from MCC-141. Per B.8.9-01 .B, System design is for pressurizing RHR and core spray systems. Per B.3.4-05.A.2.a, to assure RHR is full of water the RHR system pressure must be maintained > 40 psig by the Service Cond System.

COGNITIVE LEVEL JUSTIFICATION Level 2 - based on 1) examinee must recognize the Cond Serv Water Pump P-60B and Jockey pump are both powered off of MCC-141. 2) With Cond Serv Water Pump P-60A Page 12 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank isolated no pressure source exists for RHR Cond Serv pressurizaing stations, 3) loss of Cond Serv pressurizing stations results in loss of ability to maintain RHR system water filled and pressurized, 4) loss of RHR water fill system pressuization can result in water hammer on LPCI initiation and RHR inop if pressure < 40 psig.

Page 13 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank

"-S.-.- Question: 12 Exam: RO System: 204000 K/A: A4.09 Lesson Plan M-8107L-030 Enablina Obiective 3_c Reactor Shutdown is in progress.

"* Reactor Pressure is 80 psig and steady.

"* Reactor Temperature is 320°F and steady.

What actions can be taken to continue the Reactor cool down to obtain Cold Shutdown conditions?

A. Place Reactor Head Spray in service.

B. Place the RWCU system in the Heat Rejection mode of operation.

C. Place a train of the RHR system in the Shutdown Cooling mode of operation.

D. OPEN the Feedwater pump recirculation valves and CLOSE the Condensate recirculation valves.

Answer: B

Reference:

C.3, Part VI1.B.1 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Placing Rx head spray in service is a plausible answer as this will impact the cooldown rate as indicted by step 39 of Ref B.3.4-05.D.1, however procedure requires temperature to be less than 212OF prior to starting head spray with RHR in service in Shutdown cooling mode.

B. Per C.3, Part VII.B.1, RWCU is to placed in Heat Rejection mode if supplemental shutdown cooling is required.

C. Interlock for the SDC valve MO-2029 and MO-2030 prevents opening valves and establishing SDC if Reactor Pressure is greater than or equal to 75 psig at the steam dome. (Ref B.3.4-02.F.4)

D. Per C.3, Part VII.A.7, Notes 1 &2, Opening the RFP Recirculation to condenser valves will permit closing the condensate recirculation valves to prevent cooling the Main Condenser Hotwell. Thus feedwater temperature would be maintained higher and thus the Rx cooldown would be lowered.

COGNITIVE LEVEL JUSTIFICATION Level 2 - based on 1) examinee must recognize RHR shutdown cooling interlock of 75 psig Reactor pressure is not satisfied and RHR cannot be placed in SDC mode, 2) examinee must recognize that RWCU can be used to provide supplemental shutdown cooling in the Heat Rejection mode.

ii RC Ti;Page 14YoV15 Page 14 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 13 Exam: RO System: 205000 K/A: A1.08 Lesson Plan M-8107L-023 Enablina Obiective. 20 The plant has been shutdown for an outage to inspect the ECCS Pump Torus Suction Strainers due to an identified safety concern. RHR Loop A is in Shutdown Cooling mode.

Annunciator 3-A-12, RHR HX A OR B DISCH WTR HI TEMP, alarms.

The RHR Heat Exchanger high temperature condition can be confirmed by monitoring temperature...

A. recorder on Control Room back panel C-21.

B. indicator in 'A' Residual Heat Removal Room.

C. recorder on Control Room front panel C-04.

D. recorder on Control Room front panel C-03.

Answer: A

Reference:

ARP C.6-003-A-12 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Per ARP 3-A-1 2, confirm RHR HX temp on TR-23-115 located on panel C-21 B. Plausible answer as temperature could be obtained from local indications; however RHR temp indicators do not have local display.

C. Plausible answer as TR-2-167 on C-04 monitors recirc loop temps; however, recirc loop temps would not be a direct indication of RHR HX temps based on Recirc pumps in operation for typical SDC ops providing mixing in RPV, or for SDC with idle loop the TR would be isolated from RHR flow by closing RECIRC pump suction and discharge valves.

D. Plausible answer RHR system control and indications are on C-03.

COGNITIVE LEVEL JUSTIFICATION Level 1 - based on examinee must recall system facts and equipment locations.

'IE5ll

%PageIoTE Page 15 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 14 Exam: RO System: 209001 K/A: A4.01 Lesson Plan M-8107L-005 Enabling Objective 7 M-8114L-01 1 9f .

Given the following conditions:

"* A LOCA has occurred.

"* ECCS systems injected into the Reactor.

"* Reactor water level was restored to normal level and ECCS pumps were secured per C.5-3205, TERMINATE AND PREVENT.

"* Drywell pressure is 3.5 psig.

"* Subsequently, Reactor level dropped to minus 50 (-50) inches.

Which of the following describes the operation of the Core Spray system?

A. The Core Spray system will require manual restarting of the pumps and realignment of the Core Spray injection valves.

B. The Core Spray pumps will automatically restart and inject into the RPV to raise level.

C. The Core Spray pumps will automatically restart and the Core Spray Injection valves must be manually OPENED from the control room.

D. The Core Spray system will inject into the RPV when the MO-1751 CS INJECTION BYPASS and MO-1752 CS INJECTION BYPASS hand switches are placed in AUTO.

Answer: A

Reference:

C.5-3205 Part A Question Pedigree: INPO BANK 16343 Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Per C.5-3205, Part A, CS is terminated by placing hand switches for MO-1751/MO 1752 CS INJECTION BYPASS in bypass, closing MO-1751 and MO-1752, and placing the pumps in P-T-L. Thus to restore CS system, the system must be manually realigned.

B. Distracter is plausible as CS will automatically initiate on Drywell pressure greater than 2.0 psig, but system has been terminated and prevented from injection per C.5-3205.

C. Distracter is plausible as CS pump will automatically start on Drywell pressure greater than 2.0 psig, but system has been terminated and prevented from injection per C.5 3205.

D. Distracter is plausible as returning the MO-1751/MO-1752 CS INJECTION BYPASS would allow the valves to auto open on high Drywell pressure with reactor pressure less than 460 psig. However no information is given on reactor pressure and the pumps have been placed in P-T-L per C.5-3205.

COGNITIVE LEVEL JUSTIFICATION Level 2 - based on 1) examinee must recognize actions taken to terminate and prevent CS system, 2) recognize that with Drywell pressure greater than 2.0 psig Core Spray does receive and auto initiation, 3) determine that CS will not auto start based on actions taken to terminate and prevent CS injection.

Page 16 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 15 Exam: RO System: 211000 K/A: A1.02 Lesson Plan M-8107L-0( Enabling Objective 2.b, 2.d, 7.b, 7.d During the conduct of his normal panel rounds, a Control Room Operator has identified that the continuity meter for the SYSTEM 1 Standby Liquid Control Squib valve is indicating zero (0) amperes. Which of the following explains how system operation would be affected if initiation of Standby Liquid Control SYSTEM 1 was attempted?

A. The 'A' SBLC pump would start, but NOT inject since the squib valve will NOT fire.

B. The 'A' SBLC pump would start and inject since the squib valve has fired.

C. The 'A' SBLC pump would NOT start, but the inject path is open as the squib valve fired.

D. The 'A' SBLC pump would NOT start and the squib valve will NOT fire.

Answer: A

Reference:

B.3.5-01.C, B.3.5-05.G.1, B.3.5-01.D, B.3.5-02.C, ARP 5-B-31 Question Pedigree: INPO BANK 13129 Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Per B.3.5-01.C, 2 SBLC has separate flow loops each having positive displacement pump, an explosive injection valve, piping to support vessel injection. All valves in flow path are manual except squib valves. Per B.3.5-05.G.1, turn SLC System Selector switch to SYS 1 starts the selected pump. No interlocks are present between squib firing circuit and SBLC pumps per B.3.5-01.D. Per B.3.5-02.C, two firing squibs in the primer chamber for explosive valves. The bridge wire continuity is monitored by two continuity relays in Cont. Room Panel C-05. Should loss of continuity occur in either of the explosive valves (indicating loss of firing ability) the continuity relays will energize alarm 5-B-31. Per ARP 5-B-31, meters11-67A and 1167B can be read to determine which circuit of squib valves has lost continuity.

B. Plausible answer as pumps will start when control room selector switch is positioned to desired system. Answer is incorrect based on loss of squib valve continuity does not indicate explosive valve has fired, but rather the valve cannot be fired open.

C. Plausible answer as provided information could be confused to indicate that a discharge path does not exist for the positive displacement pump, thus pumps do not start D. Plausible answer as provided information could be confused with pump interlocks to prevent simultaneous operation of both pumps. Answer is incorrect as interlocks do not exist between the pumps and the squib valve firing circuit, loss of squib valve continuity does not indicate explosive valve has fired, but rather the valve cannot be fired open.

COGNITIVE LEVEL JUSTIFICATION Level 1 - based on knowledge of interlocks, setpoints, and system response.

Page 17 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 16 Exam: BOTH System: 211000 K/A: A2.01 Lesson Plan M-8107L-0C04 Enablina Obiective 8o A plant transient has occurred which has resulted in an automatic Reactor scram. 25 Control Rods failed to fully insert. The Operator-At-The-Controls has placed hand switch 1 1A-S1, SLC SYSTEM SELECTOR, to SYS 1 and verified system operation.

Subsequently, annunciator 8-B-32, NO. 103 480V LDCTR MCC FEEDER TRIP, alarms.

Given the following plant conditions.

"* Standby Liquid Control Tank Level is 1000 gal and steady.

"* No. 11 SBLC Pump status indicating lights are OFF.

"* No. 11 RBCCW Pump status indicating lights are OFF.

"* No. 11 RWCU Pump has tripped.

"* Torus Temperature is 105OF and increasing.

What operator action is required?

A. Terminate boron injection and insert rods per C.5-3101, ALTERNATE ROD INSERATION.

B. Inject boron per C.5-3102, ALTERNATE BORON INJECTION WITH CRD.

C. Inject boron per C.5-3103, ALTERNATE BORON INJECTION WITH RWCU.

D. Place SBLC System 2 in service per B.03.05-05.G.1, MANUAL INITIATION.

Answer: D

Reference:

B.3.5-05.G.1, B.3.5-05.C, C.4-B.9.7.C Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Answer is plausible as C.5-2007, part S directs that shutting down with rods is preferable to using boron; however, Torus temp is approaching boron injection temp, has been initiated, and C.5-2007 directs that once boron injection commences, it must continue until cold shutdown boron weight is injected.

B. Answer is plausible as C.5-2007, Part S, directs to inject with boron per alternate means if SBLC is unavailable. SBLC remains available based on provided tank level and system 2 powered from LC-1 04 and powered MCCs.

C. Answer is plausible as C.5-2007, Part S, directs to inject with boron per alternate means if SBLC is unavailable. SBLC remains available based on provided tank level and system 2 powered from LC-1 04 and powered MCCs.

D. Loss of LC-103 results in No. 11 (System 1) SBLC Pump trip due to loss of power to MCC-133A. Per B.3.5-05.G.1, "If operation of SLC cannot be verified. Then turn the handle of switch 11 A-S1 to SYS 2 or SYS 1 (Postion not used in step 1) AND verify operation, as specified in Step 3."

COGNITIVE LEVEL JUSTIFICATION Level 2 - based on examinee must 1) recognize that loss of MCC-133A based on alarm and SBLC tank not decreasing, 2) 1 1A-S1, SLC SYSTEM SELECTOR, to SYS 1 selects No. 11 SBLC pump for injection, 3) No. 11 SBLC pump trips on loss of MCC PPage 1, of 156 Page 18 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 17 Exam: BOTH System: 215002 K/A: A1.01 Lesson Plan M-8107L-057 Enabling Objective 2.b The plant is operating at 50% power with a non-edge rod selected when the B Recirc Pump controller failed, raising the speed of the B Recirc Pump. Operators respond by locking the B Recirc MG Scoop Tube, but not before Reactor power rises to 75%. No other operator actions are taken.

Which of the following describes the impact of this transient on the amplification applied to the average LPRM inputs in the Rod Block Monitor (RBM) system?

The amount of amplification to BOTH RBM A and B average LPRM input signals...

A. is automatically adjusted to correspond to the reference APRM readings.

B. remains fixed at value at the time of the rod selection.

C. is automatically adjusted to correspond to the new core flow.

D. is bypassed based on the rod selection.

Answer: B

Reference:

B.5.1.2-02.C.2 Question Pedigree: MOD INPO BANK Cog. Level: 2 8752 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer. The RBM Automatic Gain Adjustment automatically amplifies the selected LPRMs to 100%. However, the gain is held until a new rod is selected. The Trip Set Point Circuit selects the RBM rod block trip levels based on average reactor power as input from the APRMs.

B. The RBM Automatic Gain Adjustment automatically amplifies the selected LPRMs to 100%. The gain is held until a new rod is selected.

C. Plausible answer. The RBM Automatic Gain Adjustment automatically amplifies the selected LPRMs to 100%. However, the gain is held until a new rod is selected and the Trip Set Point Circuit selects the RBM rod block trip levels based on average reactor power as input from the APRMs not core flow.

D. Plausible answer. The RBM is automatically bypassed when an edge rod is selected or power is below 27% (referred to as the LPSP: low power setpoint). However, this action bypasses the RBM rod block features and not the automatic gain adjustment, a non-edge rod is identified in the stem, and power is above the LPSP COGNITIVE LEVEL JUSTIFICATION Level 2 - based on examinee must recall system interlocks, setpoints and system response and recognize interactions between Recirc system and Power Range Monitoring system.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank "Question: 18 E (am: BOTH System: 215002 K/A: K5.01 Lesson Plan M-8107L-057 Enablinq Obiective 2.c Plant power has been reduced from 100% power to 75% power to perform procedure 0255 07-IA-1, MAIN STEAM VALVE EXERCISE TESTS, and Control Rod pattern adjustments.

Annunciators 5-A-43, RBM DOWNSCALE, and 5-A-3, ROD WITHDRAW BLOCK, are in alarm.

What operator action is required to reset the alarms and change the Rod Block Monitor (RBM) rod withdraw block set points?

A. Momentarily depress the APRM/RBM TRIP LEVEL pushbuttons on C-05.

B. Place the RBM BYPASS switch on C-05 momentarily to Rod Block Monitor 7 and then momentarily to Rod Block Monitor 8 and return to OFF.

C. Place the Rod Block Monitor 7 and 8 MODE switches momentarily to STANDBY and then return back to OPERATE.

D. Deselect the currently selected Control Rod and then select the desired Control Rod to be positioned.

Answer: D

Reference:

C.6-005-A-43 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer. Depressing the APRM/RBM TRIP LEVEL pushbuttons on C-05 provides indication on the C-05 recorders as to the trip level setpoints for the APRM flow referenced rod block and the RBM setpoint, but does not reset the rod withdraw block or re-initiate the RBM null sequence for reset of the Trip Setpoint.

B. Plausible answer. Bypassing of the RBMs will bypass all of the RBM channel trip output signals but does not re-initiate the RBM null sequence for reset of the Trip Setpoint.

C. Plausible answer. Placing the RBM MODE switch to standby connects the RBM in the same configuration as OPERATE with the exception the INOP trip circuit is placed in the tripped condition. B.5.1.2-02.C.5 (page 18 or 21 rev 3) indicates the RBM INOP trip results in an auto reset trip signal to 5-A-51. This means the trip signal will automatically reset when the INOP condition clears, not that the INOP trip signal automatically resets alarm 5-A-51.

D. Per C.6-005-A-43, The RBM downscale alarm is expected for power changes when inserting rods or reducing Recirc flow. Step 3 of the ARP directs to deselect and reselect a rod if the annunciator is due to a power transient. Per B.5.1.2-02.C.3, the RBM power biased rod block setpoints are selected during the RBM nulling sequence based on APRM average power level input to the RBM.

COGNITIVE LEVEL JUSTIFICATION Level 1 - based on examinee must recall system interlocks, setpoints and system response.

IPag R ofT156 Page 21 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 19 Exam: BOTH System: 215004 K/A: A4.01 Lesson Plan M-8107L-054 Enabling Objective 1 M-8120L-009 3 A Reactor startup is in progress. The following data was obtained during the startup.

TIME SRM 21 SRM 22 SRM 23 SRM24 10:10:00 600 560 360 690 10:11:00 960 900 550 920 10:12:00 1360 1290 780 1320 10:13:00 1920 1860 1100 1900 10:14:00 2720 2580 1560 2640 Control Rod motion stopped at time 10:11:00 and the Reactor was declared critical at 10:12:00.

Which of the following describes the Reactor period at criticality?

A. Between 110 and 130 seconds.

B. Between 130.1 and 160 seconds.

C. Between 160.1 and 190 seconds.

D. Between 190.1 and 220 seconds.

"--.-- Answer: C

Reference:

C.1.111.D.2, C.1.V.A.3.

Question Pedigree: INPO BANK 9547 Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. See correct answer, period could be obtained with SRM 21 and 22 data prior to criticality.

B. See correct answer, period could be obtained with SRM 23 data prior to criticality.

C. Correct answer is obtained by using P(t)=P(O) et"', for data after criticality occurs at time 10:12. Correct answer can also be obtained using the relationship Period = doubling time divided by 0.693 or multiplied by 1.443 and data sets for SRMs 21 and 23 at times 1012 and 1014.

D. See correct answer, period could be obtained with SRM 24 data prior to criticality.

COGNITIVE LEVEL JUSTIFICATION Level 3 - based on examinee must solve a problem using knowledge and its meaning using the following concepts 1) criticality exists when SRM count rate is increasing and Reactor period is constant and positive, 2) equation for P(t)=P(0) et", or Period =

doubling time divided by 0.693 or multiplied by 1.443, 3) stated that Reactor is critical at time 1012 and control rod motion stopped at time 1011.

HIRC 1FIMNH I*AM]IAATII©H JRPC Rave Co1y ow Page 22 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 20 Exam: BOTH System: 215004 K/A: K2.01 Lesson Plan M-8107L-054 Enablina Obiective 4 Which of the following plant electrical systems supplies power to the detector and electronic circuitry of the Source Range Monitoring (SRM) System?

A. 24 VDC station batteries B. 125 VDC station batteries C. 120 V Instrument AC D. 480 VAC station auxiliary Answer: A

Reference:

B.5.1.1-02.C.1, B.5.1.1-06, Fig 4 Question Pedigree: ILT BANK M-8107L- Cog. Level: 1 054-028 ANSWERJDISTRACTER JUSTIFICATION A. Per B.5.1.1-02.C.1, SRM power is supplied from the 24 VDC battery system. B.5.1.1 06, Fig 4, indicates that both the dectector and associated electronic circuit are powered from 24 VDC.

B. Plausible answer, as SRMs are DC powered and do include an internal high DC voltage power supply of 350 VDC. Examinee could confuse this high DC voltage as coming from an external high voltage DC source.

"--J C. Plausible answer, as SRMs detector drive control relays are powered from Y-20.

D. Plausible answers as SRM detector drive motors are power from 480 VAC that is transformed to 208VAC via Lighting Panel L-28.

NOTE: original question had non-significant modification. Original question asked what is power supply to SRM System. The original question had two correct answers as 480 VAC does supply the detector drive motors and 24 VDC supplies the detector and electronic circuitry.

COGNITIVE LEVEL JUSTIFICATION Level 1 - based on examinee must recall system facts.

M¶IC WTUIITNEH EIEiZAMW ll©H JR(C I*vkw (CD*Y Page 23 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank s-' Question: 21 Exam: BOTH System: 215005 K/A: K2.02 Lesson Plan M-8107L-069 Enabling Objective 4 M-8107L-072 3.e The plant is operating at 100% power. A plant electrical fault occurs resulting in the loss of MCC-111.

With the loss of MCC-1 11, power is lost to Average Power Range Monitoring (APRM) system channels...

A. 1, 2 and 3 and can be restored by energizing RPS Bus A from the alternate source.

B. 4, 5 and 6 and can be restored by energizing RPS Bus B from the alternate source.

C. 1, 2 and 3 and cannot be restored until the cause of the loss of MCC-111 is corrected and MCC-1 11 is re-energized.

D. 4, 5 and 6 and cannot be restored until the cause of the loss of MCC-1 11 is corrected and MCC-1 11 is re-energized.

Answer: C

Reference:

B.5.1.2-02.D, B.9.12-02.G, B.9.12-05.C Question Pedigree: New Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer. Per B.5.1.2-02.D, APRM channels 1, 2 and 3 are powered from RPS Bus A. RPS bus A is powered from MCC-111 per B.9.12-02.G and B.9.12-05.C; however Y-60, the alternate source to RPS buses is also powered from MCC-111 and is thus unavailable.

B. Plausible answer. Per B.5.1.2-02.D, APRM channels 1, 2 and 3 are powered from RPS Bus A and channel 4, 5, and 6 are powered from RPS bus B. RPS bus B is powered from MCC-141 per B.9.12-02.G, thus the identified APRM channel have not lost power with a loss of MCC-1 11. In addition, per B.9.12-05.C the alternate source to RPS buses (Y-60) is also powered from MCC-1 11 and is thus unavailable.

C. Per B.5.1.2-02.D, APRM channels 1, 2 and 3 are powered from RPS Bus A. RPS bus A is powered from MCC-111 per B.9.12-02.G and B.9.12-05.C; however Y-60, the alternate source to RPS buses is also powered from MCC-1 11 and is thus unavailable.

The power to the identified APRMs cannot be restored until MCC-1 11 is re-energized.

D. Plausible answer. Per B.5.1.2-02.D, APRM channels 1, 2 and 3 are powered from RPS Bus A and channel 4, 5, and 6 are powered from RPS bus B. RPS bus B is powered from MCC-141 per B.9.12-02.G, thus the identified APRM channel have not lost power with a loss of MCC-1 11. The alternate source to RPS buses is also powered from MCC-1 11 and is thus unavailable.

COGNITIVE LEVEL JUSTIFICATION Level 2 - based on examinee must recognize system interactions between APRMS, RPS, and 480 VAC: 1) APRM channels 1, 2 and 3 are power from RPS bus A and APRM channels 4, 5 and 6 are powered from RPS bus B, 2) A RPS bus is energized from A MG set from MCC-I 11 3) alternate RPS power supply Y-60 is powered from MCC-I 11.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 22 E) cam: BRO System: 215005 K/A: K4.01 Lesson Plan M-8107L-069 EnablinQ Obiective 7.a A plant startup is in progress. Reactor power is currently at 75% power and power is being raised to 100% with Recirc Flow. The following annunciators alarm:

"* 5-A-38, APRM FLOW BIAS OFF NORMAL

"* 5-A-3, ROD WITHDRAW BLOCK The operator observes the following indications.

"* Recirculation flow indications are within expected ranges.

"* The amber Flow Converter 1 COMPARATOR indicating light is ON.

"* The white Flow Converter 1 INOP indicating light is ON.

Based on the above indications, what is the cause of the Control Rod withdraw block?

A. An APRM high upscale trip based on Reactor power and flow relationship exceeding 0.66W+53.6%.

B. An excessive deviation in APRM Recirculation Flow Converter output signals.

C. An APRM high upscale trip based on Reactor power and flow relationship exceeding 0.66W+65.6%.

D. APRM Recirculation Flow Converter 1 mode selector switch is NOT in OPERATE.

Answer: B

Reference:

C.6-005-A-38, B.5.1.2-02. B.4 Question Pedigree: NEW Cog. Level: I ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer based on an APRM high upscale trip does cause a rod block when pwr to flow relationship exceeds Pwr=0.66W+53.6%. Answer is incorrect based on indications for APRM high upscale trip would be alarm 5-A-14, APRM HI FLUX, and HI FlUX lights ON on panel C-05 and the applicable APRM cabinet.

B. Per B.5.1.2-02.B.4, Each Recirc Flow Converter unit has a comparator circuit which gives a COMPARATOR trip ifthe output signals of the two converters deviate by a preset value. Per ARP 5-A-38, for a COMPARATOR trip indications are amber COMPARATOR light and UPSCLIINOP light ON on panel C-37, and Rod Block alarm 5-A-3.

C. Plausible answer based on an APRM high upscale trip does cause a rod block when pwr to flow relationship exceeded. However relationship given is for the High High SCRAM setpoint being exceeded (Pwr=0.66W+65.6%). Answer is incorrect based on indications for APRM high high upscale trip would be alarm 5-A-22, APRM HI HI INOP CH 1, 2, 3 or alarm 5-A-30, APRM HI HI INOP CH 4, 5, 6, and HI HI UPSCALE lights ON on panel C-05.

D. Plausible answer based on an APRM Flow Converter NOT in OPERATE does cause a ROD block. Per B.5.1.2-02.B.4 for Flow converter not in operate, the amber UPSCALE trip light would be on and the White INOP light would be on.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank COGNITIVE LEVEL JUSTIFICATION Level 1 - based on examinee must recall system interlocks, setpoint and system response. System relationship between APRMs and RMCS are taught in close proximity.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 23 Exam: BSRO System: 215005 K/A: K4.01 Lesson Plan M-8107L-069 Enabling Objective 10.c & 10.d Per Plant Technical Specifications, which Reactor mode(s) is the APRM flow biased rod withdrawal block required to be operable, AND what is the basis for the APRM flow biased rod withdrawal block?

A. Reactor Mode Switch in RUN to prevent operation significantly above the licensing basis power level during operation at reduced flow.

B. Reactor Mode Switch in RUN or STARTUP TO HOT STANDBY to prevent operation above the licensing basis power level during operation at reduced flow.

C. Reactor Mode Switch in RUN to provide local core protection against exceeding the MCPR safety limit.

D. Reactor Mode Switch in RUN or STARTUP TO HOT STANDBY to provide local core protection against exceeding MCPR safety limit.

Answer: A

Reference:

Tech Spec Table 3.2.3, TS 3.2 Bases (page 67)

Question Pedigree: NEW Cog. Level: 2 ANSWERIDISTRACTER JUSTIFICATION A. Per TS table 3.2.3, the APRM flow biased rod block is required to be operable in the RUN mode. Per TS bases, "The APRM rod block is referenced to flow and prevents operation significantly above the licensing basis power level especially during operations at reduced flow. The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal sequence."

B. Plausible answer based on the correct bases is provided; however the APRM flow biased rod block is not required in Start to Hot Standby. Further, the basis statement concerning protection a reduced flow provides credibility to the stated mode switch positions and per B.5.6-02 (RMCS) the APRM rod withdraw block does provide a rod block independent of Rx Mode Switch position.

C. Plausible answer based on the correct mode switch position is provided and the APRM rod blocks protects integrity of the MCPR safety limit; however, the APRMs provide gross core protection as stated in the bases whereas the IRM rod block provide local and gross core protection.

D. Plausible answer as the APRM rod block is required in RUN; however operability is not required in Startup to Hot Standby. Further the APRM rod blocks protect integrity of the MCPR safety limit; however, the APRMs provide gross core protection as stated in the bases whereas the IRM rod block provide local and gross core protection. Per B.5.6-02 (RMCS) the APRM rod withdraw block does provide a rod block independent of Rx Mode Switch position.

COGNITIVE LEVEL JUSTIFICATION Level 2 - based on examinee must recall and recognize system facts and relate to consequences stated in Tech Spec bases.

SRO Level Justification SRO level knowledge based on examinee must recall plant Tech Spec Bases information.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 24 E xam: BOTH System: 216000 K/A: K1.09 Lesson Plan M-8107L-071 Enablina Obiective 2.a. 3.a. 3.b. 4.d Given the following plant conditions:

"* Annunciator 5-A-39, ATWS CH A PB ARMED, is in alarm.

"* Pushbutton S-5A, ATWS A MAN, is then depressed.

"* Annunciator 5-A-31, ATWS CHANNEL A TRIP, alarms.

If a second sub-channel on ATWS logic...

A. Channel A trips, then the Recirc MG Set Drive Motor Breakers OPEN and an ARI valve OPENS.

B. Channel A trips, then the Recirc MG Set Generator Field Breakers OPEN and an ARI valve OPENS.

C. Channel B trips, then the Recirc MG Set Drive Motor Breakers OPEN and the ARI valves OPEN.

D. Channel B trips, then the Recirc MG Set Generator Field Breakers OPEN and the ARI valves OPEN.

Answer: B

Reference:

C.6-005.A-31, B.5.6-01I.C.3, B.5.6-02.C.1 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer because the Recirc MG Set Drive Motor breakers do trip OPEN on Low-Low Reactor Water level, however this trip does not initiate from the ATWS logic.

ATWS logic trips the Recirc MG Set Field breakers on Low-Low Rx Water level sustained for 9 seconds. Further the answer is plausible as it describes the correct instrument input logic scheme for ATWS (2 out of 2 once).

B. Per C.6-005.A-31, alarm will annunciate if one of the four sub-channels trip on Low-Low Rx Level or 1135 psig Rx Pressure and both sub-channels must actuate to cause automatic actions. Per B.5.6-01 .C.3, The ATWS logic consists of two trip systems, Channel A and Channel B, with each Channel of trip systems made-up of two sub channels for a total of four sub-channels. A trip of both sub-channels in logic Channel A or a Trip of both sub-channels in logic Channel B will OPEN both Recirc MG Set generator field breakers and OPEN the associated ARI valve. Per B.5.6-02.C.1, each ATWS system logic channel is made up of two sub-channels designated A and C for Channel A, and B and D for Channel B. For each input parameter, there is one independent sensor for each sub-channel. Per Per B.5.6-02.C.l.c, CR annunciators 5 A-31 and 5-A-32 alarm when a logic sub-channel trip occurs. 5-A-31 will alarm when either sub-channel A or C is tripped, 5-A-32 will alarm when either sub-channel B or D is tripped. For this arrangement of the 2-out-of-2-once logic, only one annunciator would alarm for a complete trip in one trip system.

C. Plausible answer because the Recirc MG Set Drive Motor breakers do trip OPEN on Low-Low Reactor Water level, however this trip does not initiate from the ATWS logic.

ATWS logic trips the Recirc MG Set Field breakers on Low-Low Rx Water level URIC WRJI¶NPH I7IX A MATh©O 1N]RC TRvnw Colpy Page 28 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank sustained for 9 seconds. Further, the answer is incorrect as is describes the instrument input logic scheme for ATWS as 1-out-of-2-taken-twice, while ATWS is 2-out-of-2-once logic.

D. Answer is plausible because Per B.5.6-01.C.3, an ATWS logic actuation will OPEN both Recirc MG Set generator field breakers and OPEN the associated ARI valve. The answer is incorrect as is describes the instrument input logic scheme for ATWS as 1 out-of-2-taken-twice, while ATWS is 2-out-of-2-once logic.

Question Note: Question 112 (KA295031EK2.10) is also related to ATWS. This question has been tailored to remove question similarity and uses ARI valves in all 4 answers as Question 112 uses ARI valves in all 4 answers. Thus review of either question will not point to the correct answer in either question. This question is focused on Nuc Inst physical connections via AWTS logic sheme and cause and effect relationships, while question 112 is focused on interrelationships between Reactor Low Water Level and ATWS.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recall system facts 1) ATWS logic scheme and 2) automatic actions initiated by ATWS.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 25 Excam: RO System: 216000 K/A: K3.30 Lesson Plan M-81 07L-029 7n An The plant has been in a maintenance outage. Plant conditions have been established to start No. 11 and No. 12 Recirc Pumps. The Operator places hand switch 2A-S1A, No. 11 MG SET DRIVE MOTOR to START and observes normal indications for start of the MG Set drive motor.

The Operator then places hand switch HS2A-S7A, MO-2-53A PUMP DISCHARGE, to OPEN and observes normal indications for Recirc Pump startup.

Shortly after the following alarms are received.

"* 4-C-1, RECIRC A LOCKOUT

"* 4-C-21, RECIRC A STARTUP SEQUENCE INCOMPLETE

"* 4-C-31, RECIRC DRIVE MOTOR A TRIP What malfunction caused the trip of the No. 11 Recirc MG Set?

A. Loss of Recirc Pump suction valve OPEN indication.

B. Recirc MG Set high lube oil temperature.

C. Recirc Pump low differential pressure.

D. Recirc MG Set low lube oil pressure.

Answer: C

Reference:

B.1.4-02.F.1, C.6-004-C-01, C.6-004-C-08, C.6-004-C 21, C.6-004-C-31 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer based on Recirc pump suction valve < 90% open will cause a Recirc Drive Motor trip and alarm 4-C-31. Failure does not cause 4-C-21 or 4-C-1.

B. Plausible answer based on Recirc MG set high lube oil temperature will cause a Recirc Drive Motor trip and alarm 4-C-31. High lube oil temp does not cause 4-C-21 or 4-C-1.

C. Per B.1.4-02.F.1, if the pump does not develop >7.5 psid within 15 sec of placing the hand switch to start, the MG set will receive a lockout and a drive motor breaker trip.

Per ARP 4-C-21, setpoint for alarm is < 7.5 PSID after drive motor handswitch in START, D. Plausible answer based on Recirc MG set generator low lube oil pressure does cause alarm 4-C-31. Recirc MG set generator low oil pressure not cause alarm 4-C-21 or 4-C 1.

COGNITIVE LEVEL JUSTIFICATION Level 1 - based on examinee must recognize system interlock, set points and response.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 26 Ex:am: BOTH System: 218000 K/A: A1.03 Lesson Plan M-8107L-025 Enabling Objective 5.d, 7.d, 9.b M-8107L-024 9.b The plant is operating at 98% power. Annunciator 3-A-48, N2 LO PRESSURE SRVS, INBD T-RINGS, alarms. The operator dispatched to the Train A Alternate Nitrogen manifold reports that the manifold pressure is 200 psig and decreasing.

With respect to the Safety Relief Valves (SRVs), what action is required?

A. Declare ADS SRV A and Low-Low Set SRV E inoperable.

B. Declare ADS SRV B and Low-Low Set SRV F inoperable.

C. Declare ADS SRV C and Low-Low Set SRV H inoperable.

D. Declare ADS SRV D and Low-Low Set SRV G inoperable.

Answer: A

Reference:

B.8.4.3-05.2.a, B.8.4.3-05.3.b, C.6-003-A-48 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Per B.8.4.3-05.2.a, If one train of the Alt N2 system is inoperable, then the ADS capability of one SRV becomes inoperable and the Lo-Lo Set capability of one SRV becomes inoperable. Per B.8.4.3-05.2.a, Alt N2 Train A supplies ADS SRV A and Lo Lo Set SRV E. Per B.8.4.3-05.3.b, for Alt N2 train to be operable, manifold pressure must be a minimum of 230 psig. Per C.6-003-A-48, alarms setpoint is at 625 psig to indicate Alt N2 low manifold pressure.

B. Plausible answer as loss of alt N2 Train will cause SRVs to be inoperable, However SRV B is not one of the ADS SRVs and SRV F is not one of the low-low set SRVs.

C. Plausible answer as loss of alt N2 Train will cause SRVs to be inoperable, However ADS SRV C and low-low set SRV F are supplied by train B of the Alt N2 System.

D. Plausible answer as loss of alt N2 Train will cause SRVs to be inoperable, However ADS SRV D and low-low set SRV G are not supplied by the Alt N2 System. Their backup supply is from local accumulators.

COGNITIVE LEVEL JUSTIFICATION Level 1 - based on examinee must recall Reactor Pressure Relief operability requirements contained in section 05 of the system operations manual.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank "Question: 27 Exam: RO System: 219000 K/A: K4.10 Lesson Plan M-8107L-016 Enablina Obiective 6d 7di 7a Qi The plant is operating at 100% power. Procedure Enablia

.. 0255-08-IA-1,

. . t.. RCIC SYSTEM 7rPUMP

.1 I I .

FLOW AND VALVE TESTS, was recently completed. The following plant conditions exist.

"* No. 11 RHR pump is aligned for Torus cooling and is operating at 4000 gpm.

"* No. 13 RHRSW pump is operating at 3600 gpm.

"* Torus temperature is 981F and decreasing.

"* No. 11 and No 12 Service Water pumps are operating.

Annunciator 3-B-19, RHR HX B TUBE/SHELL LOW DIF PRESS, alarms.

What operator action is required to prevent RHR to RHRSW intersystem leakage?

A. Increase No. 11 RHR Pump flow to 4400 gpm.

B. Verify CLOSED CV-1729, 12 RHR Heat Exchanger Service Water Outlet, and OPEN RHRSW-32, RHRSW Loop A and B Crosstie.

C. Decrease No. 13 RHRSW Pump flow to 3500 gpm.

D. Shut down No. 11 RHR pump and No. 13 RHRSW pump, and vent the No. 12 RHR Heat Exchanger using the RHRSW vent valves.

Answer: B

Reference:

ARP C.6-003-B-19, B.8.1.3-05.H.5 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer as ARP 3-B-19 directs to increase RHR flow to 4400 gpm if pump is operating; however the alarm has been received for the shutdown loop of RHR. If unable to maintain DP across HX while RHRSW is shutdown, then the ARP directs to take action per B.8.1.3-05.

B. Per B.8.1.3-05.H.5, "Restoration of B RHRSW Following Loss of System Pressure, Part A, actions to restore pressure are to 1) verify CV-1 729 is CLOSED, 2) Verify A RHRSW is available (which it is because it is running supporting Torus cooling, and 3) and then pressurize the B loop by opening RHRSW-32.

C. Plausible answer as ARP 3-B-19 directs to reduce RHRSW flow to 3500 gpm if pump is operating; however the alarm has been received for the shutdown loop of RHR. If unable to maintain DP across HX while RHRSW is shutdown, then the ARP directs to take action per B.8.1.3-05.

D. Plausible answer as procedure B.8.1.3-05.H.5 directs the action to shutdown the A loop of RHRSW if it is not required; however, the loop is required to be in operation to support Tech Spec compliance to restore Torus temperature to less than 90 0 F.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recall procedure steps and cautions and recognize system relationships R1C C P 3Rvow1P56 Page 32 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 28 E) tam: BOTH System: 223001 K/A: K6.12 Lesson Plan M-8107L-041 Enabling Objective 5.c M-8107L-078 4.a Given the following plant conditions:

"* Due to severe winds a loss of off-site power has occurred.

"* No. 11 and No. 12 Emergency Diesel Generators (EDGs) failed to energize the essential busses.

"* The plant has been shutdown by a manual scram.

"* RCIC is being used to control Reactor Water Level between minus 126 (-126) inches and 48 inches.

"* No. 13 Diesel Generator has been determined to be unavailable due to damage sustained during the severe winds.

"* Efforts have been in progress for the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore off-site power and determine the cause of the EDGs loading failure, and have NOT been successful.

Four and one-half (4.5) hours after the initiation of the event annunciator 5-A-15, 24 VDC SYSTEM A UNDERVOLTAGE/OVERVOLTAGE, alarms.

What is the consequence of receiving this alarm concerning Primary Containment integrity?

A. A loss of solenoid power to Division I air operated Primary Containment isolation valves is imminent and the valves should be verified CLOSED.

B. A loss of position indication power to the H20 2 Analyzer Primary Containment isolation valves is imminent and the manual isolation valves should be verified CLOSED.

C. A loss of control power to Division I DC powered Primary Containment isolation valves is imminent and the valves should be verified CLOSED.

D. A loss of power to the solenoid operator for the H20 2 Analyzer Primary Containment isolation valves is imminent and the manual isolation valves should be CLOSED.

Answer: B

Reference:

C.4-B.9.11.A, B.4.3.1-05.C.2 Question Pedigree: NEW Cog. Level: 3 ANSWERIDISTRACTER JUSTIFICATION A. SBO directs to confirm Containment Isolation if CANNOT maintain level above -126 inches, however Solenoid power for air operated valves comes from 125VAC and air operate containment isolation valves fail in safe position.

B. C.4-B.9.1 1.A directs to confirm containment integrity by CLOSEING the manual H202 Analyzer valves. Per B.4.3.1-05.C.2 and C.4-B.9.1 1.A bases, the position indication power for the H202 Analyzer Cntmt Isolation valve is supplied from 24 VDC.

C. SBO directs to confirm Containment Isolation if level CANNOT be maintained above 126 inches, however DC powered valve control power comes from 125VDC center tapped from the respective 250 VDC battery.

D. C.4-B.9.11.A directs to confirm containment integrity by CLOSEING the manual H202 Analyzer valves, however per B.4.3.1-05.C.2, the solenoid operated H202 Analyzer Cntmt Isolation valve solenoid power is supplied from 120 VAC.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank COGNITIVE LEVEL JUSTIFICATION Examinee must recognize 1) the Stations Blackout condition, 2) recognize under the SBO condition that 24 VDC battery chargers are unavailable and the batteries have been discharging to maintain system loads, 3) predict the event outcome that battery chargers can not be returned to service under the given SBO conditions and the 24 VDC batteries have been depleted based on alarm 5-A-15, 6) examinee must identify the correct event outcome for loss of 24 VDC to the H202 Analyzer Containment Isolation valves.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank

"-~~---- Question: 29 Exam: BOTH System: 223002 K/A: A3.01 Lesson Plan M-81 07L-0C 7 r*;

7. --- ,1 V J.I A plant transient has occurred resulting in a loss of No. 11 and No. 12 Feed Water pumps, Reactor scram, and a Group 1 Primary Containment Isolation. RCIC is operating to restore Reactor Water Level when the following alarms and indications are received.

"* Annunciator 4-A-9, RCIC TURBINE TRIPPED, alarms.

"* Annunciator 3-B-56, HIGH AREA TEMP STEAM LEAK, alarms.

"* MO-2078, RCIC Turbine Steam Supply Valve, is OPEN.

"* MO-2080, RCIC Turbine Trip and Throttle Valve, is CLOSED.

"* RCIC turbine speed and pumped flow are decreasing.

"* RCIC pump suction pressure is 20 psig.

"* The Mechanical Overspeed light on C-04 is OFF.

0 The Group 5 Isolation Reset light on C-04 is OFF.

Based on the above indications, what was the cause of the RCIC turbine trip?

A. Mechanical over-speed.

B. High Reactor water level.

C. Group 5 Containment Isolation.

D. Low pump suction pressure.

Answer: C

Reference:

B.2.3-01.C, C.4-B.4.1.E Question Pedigree: NEW Cog. Level: I ANSWER/DISTRACTER JUSTIFICATION A. RCIC Turbine Trip on Mech Over-speed results in indication of the Mech Overspeed light on C-04 to be ON.

B. RCIC Turbine Trip occurs at +48 inches reactor water level and results in closure of MO-2078.

C. Alarm 3-8-56 and Group 5 Isolation Reset light Off are indications of the Group 5 isolation. Group 5 isolation will cause a turbine trip and closure of MO-2080.

D. RCIC Turbine Trip on low pump suction pressure occurs at 15 inches HG vacuum.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recognize interlocks, set points, system response, and indications.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 30 Exam: BOTH System: 223002 K/A: K4.05 Lesson Plan M-8107L-0:30 Enablina Obiective. I1_q The Reactor Water Clean-up (RWCU) high system flow Group 3 Primary Containment Isolation uses a single flow element to measure system flow. Therefore, a failure of a flow element sensing line could prevent detection of a RWCU line break.

What design feature is used to ensure that a single failure of the flow element sensing line will NOT prevent a Group 3 Isolation signal on high system flow?

A. A downscale trip on two of the four RWCU flow transmitters will initiate a Group 3 Isolation signal.

B. A single instrument trip will initiate a high system flow Group 3 Isolation signal to ensure conservative automatic action is initiated.

C. A high Drywell pressure signal at 2 psig provides redundancy to the RWCU high system flow Group 3 Isolation signal.

D. A negative differential pressure signal will initiate a Group 3 Isolation signal at an indicated flow of minus 200 (-200) gpm.

Answer: D

Reference:

B.5.6-02.B.1 .j Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION

"-*-' A. Plausible answer as several instruments such as fuel pool hi rad and RX Bldg Plenum hi rad will initiate isolation on two instruments downscale. RWCU initiates at -200 gpm and is a I-out-of-2-twice logic.

B. Plausible answer as several exceptions to 1-out-of-2-twice logic exists in PCT isolation logics, including RCIC and HPCI high steam flow. In these cases a single logic channel trip initiates the isolation. RWCU logic for high flow is 1 -out-of-2-twice.

C. Plausible answer as a Group 3 isolation does occur due to high Drywell pressure of 2 psig. This signal does not provide redundancy to high RWCU system flow, but rather provides for detection of leaks inside primary containment. The RWCU flow transmitter is located downstream of the primary containment isolation valves and outside primary containment and is not installed to detect leaks inside primary containment.

D. Per B. 5.6-02.B.1 .j, a break in the high side instrument line could prevent the 4 flow transmitters from responding to a high flow condition; therefore, a negative differential pressure condition provides a Group 3 isolation at in indicated flow of -200 gpm.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recall system set point purpose of setpoint. This is new equipment installed on the MNGP to address the single failure criteria of the flow element sensing line.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank "K~-. Question: 31 E)ram: BOTH System: 226001 K/A: A3.05 Lesson Plan M-8107L-023 Enabling Objective 7.m The plant was operating at 100% when a LOCA occurred. The following are the current plant conditions.

"* Torus and Drywell Sprays are in-service via Loop A of RHR.

"* Core Spray and Loop B of RHR are injecting to the vessel.

"* Reactor water level is minus 54 (-54) inches and increasing at 1 inch/minute.

"* Torus temperature is 100OF and increasing at 1°F/minute.

"* Torus pressure is 4 psig and decreasing at 0.5 psig/minute.

"* Drywell pressure is 5 psig and decreasing at 0.5 psig/minute.

If no operator action is taken, what action will occur based on the above trends?

A. The Containment Spray Valves will remain OPEN and Drywell pressure will be adversely impacted.

B. The Containment Spray Valves will automatically CLOSE at approximately 1.0 psig Drywell pressure.

C. The Containment Spray Valve will remain OPEN and ECCS pump NPSH will be adversely impacted.

D. The Containment Spray Valves will automatically CLOSE at approximately 1.0 psig Torus pressure.

Answer: B

Reference:

B.3.4-02.F.7, B.3.4-01 Question Pedigree: NEW Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer as procedure C.5-1200, part L, directs to remove Drywell and Torus spray when pressures drops below 2 psig. Basis for removing Drywell spray a 2 psig is that is override action level provides the operator sufficient time to take action before Drywell pressure drops below atmospheric or NPSH limits are exceeded.

B. Per B.3.4-02.F.7, for decreasing Drywell pressure, the Drywell and Torus spray valves will automatically close at approximately 0.75 psig Drywell pressure if the LPCI isolation/initiation signal is still in. Per B.3.4-01. the LPCI initiation signal occurs at Rx level -47 inches or Drywell press. 2 psig. For trends provided the Drywell pressure will be at 0.75 psig in 8.5 minutes. For the given RX level of -54 inches and level trend of increasing at 1 inch/minute, Rx level will be -45.5 inches in 8.5 minutes thus still below the LPCI isolation/initiation setpoint of -47 inches.

C. Plausible answer as procedure C.5-1200, part L, directs to remove Drywell and Torus spray when pressures drops below 2 psig. Basis for removing Drywell spray a 2 psig is that is override action level provides the operator sufficient time to take action before Drywell pressure drops below atmospheric or NPSH limits are exceeded. Based on given trend in Torus temperature, ECCS pump NPSH would not be impacted prior to auto closure of Containment Spray valves.

D. Per B.3.4-02.F.7, The containment spray valves are interlocked with Drywell pressure WIfCl¶MITEh 1BkAMWIIAUTH©N HRC RvfiwCopy Page 37 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank signals NOT Torus pressure. Answer is plausible based on the given Torus pressure is the automatic containment spray isolation based on Drywell pressure and interrelationships between Torus and Drywell.

COGNITIVE LEVEL JUSTIFICATION Level 3 based on examinee must predict an event outcome based on given trends recalling containment spray valves automatically close on Drywell pressure if a LPCI initiation signal is present.

HIRC WMITIEH 19714\I*I*ATR(0©H

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank

• JY Question: 32 Exam: BOTH System: 230000 K/A: K3.02 Lesson Plan M-8107L-0 23 Enablinri Chip~rtivA 9n 9nl The plant was operating at 75% power when a LOCA and Reactor scram occurred. The A loop of the Residual Heat Removal (RHR) system is operating in the Torus Spray mode.

LPCI and both Core Spray loops have been prevented from injecting to the Reactor per C.5 3205, TERMINATE AND PREVENT. HPCI and RCIC are operating to restore Reactor vessel level.

The following are the current plant conditions.

"* Reactor level is minus 50 (-50) inches and increasing.

"* Reactor pressure is 972 psig and decreasing.

"* Drywell pressure is 3 psig and decreasing.

"* Torus pressure is 2 psig and decreasing at 0.1 psig per minute.

"* Torus temperature is 850 F and increasing at 1°F per minute.

The Control Room Supervisor has directed RHR loop A be transferred to the Torus Cooling mode and RHR loop B be placed in the Torus Cooling mode. While transferring the A RHR loop from Torus Spray to Torus Cooling, MO-2010, TORUS SPRAY - INBOARD, fails to go full close.

What action should be taken to control Torus temperature?

A. Manually override the MO-2010 CLOSE torque switch at the MCC to fully CLOSE the valve and then place A and B loops of RHR in Torus Cooling.

B. Place the B loop of RHR in Torus Cooling AND secure the A loop of RHR Torus Sprays by CLOSING MO-2006, DISCHARGE TO TORUS SPRAY OUTBOARD.

C. Reduce Reactor pressure using the Main Turbine Bypass Valves to stop adding energy to the Torus from the Safety Relief Valves.

D. Place the A loop of RHR in Torus Cooling as adequate RHR pump capacity is available to support Torus Cooling and Spray AND place B loop of RHR in Torus Cooling.

Answer: B

Reference:

C.5-1200 Part L, C.5-3502 Part A, B.3.4-05.D.3 Question Pedigree: NEW Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer as the valve could be overridden closed from the MCC by depressing the close contactor; however the closed torque switch contact must be overridden at the MOV Limit Switch Compartment and is not directed by procedure.

B. Closing MO-2006 would isolate Torus spray on the A loop of RHR and thus prevent Torus pressure from decreasing to atmospheric or impacting pump NPSH as directed by C.5-1200 part L to secure Torus spray when Torus pressure drops below 2 psig.

Procedure C.5-3502 directs to secure Torus spray by closing MO-2010. On failure of MO-2010 to close, must recognize that Torus spray and Torus cooing on A loop must be isolated by closing MO-2006. Placing B RHR in Torus cooling would provide the M*C *fKUMEM TZPae39oHf 5 MRC Revkw C*py Page 39 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank maximum available Torus cooling.

C. Plausible answer as C.5-1 100, part M, directs that IF anticpate a blowdown, then to depressurize via Bypass valves, the bypass valves are not available as a Group 1 isolation has closed the MSIV on low-low RX water level (-47').

D. Plausible answer as the A loop of RHR can support Torus spray and cooling; however, the Torus pressure would continue to decrease and potentially reach atmospheric or impact pump NPSH.

COGNITIVE LEVEL JUSTIFICATION Cognitive level 3 based on the examinee must determine a course of action that is not directly driven by procedure using knowledge of RHR system, available flow paths for Torus cooling, and impact of continued use of Torus spary on Torus pressure.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank

• -- Question: 33 Exam: BOTH System: 230000 K/A: K5.01 Lesson Plan M-8107L-023 Enabling Objective 8.o The plant is operating at 99% power. Procedure 0255-04-IA-1, RHR PUMP AND VALVE TESTS, is in progress. Stroke timing of the RHR valves per Part A of the procedure has been completed.

Part B of the procedure directs to vent the RHR discharge piping per procedure 2145, RHR SYSTEM DISCHARGE VENTING, prior to performing the remaining portions of the procedure that places each of the RHR pumps in operation for surveillance data.

Why is venting required, AND does this have an impact on the RHR modes of operation?

A. The venting is to reduce elevated RHR system pressure due to valve stroking to maintain RHR to RHRSW differential pressure AND does not impact the RHR modes of operation.

B. The venting is to re-establish a water solid system after stroking valves that allow drain paths to the Torus AND does not impact the RHR modes of operation C. The venting is to reduce elevated RHR system pressure due to valve stroking to maintain RHR to RHRSW differential pressure AND the LPCI mode, Containment Spray mode, and Containment Cooling modes are not operable until system venting is complete.

D. The venting is to re-establish a water solid system after stroking valves that allow drain paths to the Torus AND the LPCI mode, Containment Spray mode, and Containment Cooling modes are not operable until system venting is complete.

Answer: D

Reference:

B.3.4-05.2.a, B.3.4-05.4.c, 2145, 0255-04-lA-1 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION CONFIRM CURRENT REV OF PROCEDURE 2145 REQUIRES LCO ENTRY PRIOR TO SUBMITTING QUESTION. VALIDATORS INDICATED THAT 2145 WAS TO BE REIVSED TO ELIMINATE REQMT FOR LCO ENTRY.

A. Plausible answer because after operation of the RHR pumps high pressure is trapped in the system and the high pressure is vented off using valve PC-36, no LCO is required for this action. However, the pumps have not been operated for the conditions given and the need to vent is due to loss of system fill, which does impact operability.

B. Plausible answer because opening valves that communicate with the Torus will results in a loss of system fill. However, this does impact the operability of the modes of RHR as stated in procedure 2145 specifically and can be determined based on general precautions and operating requirements contained in the Ops B-Man sections referenced.

C. Plausible answer because after operation of the RHR pumps high pressure is trapped in the system and the high pressure is vented off using valve PC-36, no LCO is required for this action. However, the pumps have not been operated for the conditions given and the need to vent is due to loss of system fill, which does impact operability.

D. Opening valves that communicate with the Torus will results in a loss of system fill as stated in B.3.4-05.4.c. This does impact the operability of the modes of RHR as stated in procedure 2145 specifically and can be determined based on general precautions MI*C Revhw Copy Page 41 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank and operating requirements contained in the Ops B-Man sections referenced.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recongnize procedure cautions that RHR modes are not operable when system is not full.

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• R(C Revte~w (Cly Page 42 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 34 Exam: BOTH System: 233000 K/A: A1.07 Lesson Plan M-8107L-022 Enablinqj Obiective 7.c The plant is operating at 100% power. The following plant conditions are in effect.

"* Upstream river water temperature is 85 0F.

"* No. 11 Fuel Pool Cooling Water pump is in service.

"* No. 12 Fuel Pool Cooling Water pump is isolated for maintenance.

"* No. 11 Fuel Pool Heat Exchanger is in service with the Reactor Building Closed Cooling Water Valve full OPEN.

"* No. 12 Fuel Pool Filter/Demineralizer is in service.

Annunciator 3-B-47, FUEL POOL HEAT EXCHANGER INLET HI TEMP, alarms and the local indicator has been verified to be at the alarm set point.

What action should be taken in response to the high temperature condition?

A. Remove the No. 12 Fuel Pool Filter/Demineralizer from service.

B. Restore the No. 12 Fuel Pool Cooling Water pump to service and establish cooling with No.11 and No. 12 Fuel Pool Cooling Water pumps via No. 11 Fuel Pool Heat Exchanger.

C. Place the No 12 Fuel Pool Heat Exchanger in service.

D. Increase RBCCW flow to other components cooled by RBCCW and monitor temperatures on other components cooled by Reactor Building Closed Cooling Water (RBCCW).

Answer: C

Reference:

ARP 3-B-47 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer as B.2.1-05.4.c states that system design temp of 140F should not be exceeded to protect filter resins from high temp. Alarm 3-B-47 set point is 120F and is upstream of HX, alarm for HX downstream temp has not been received. Thus no indication that filter resins are in danger of high temp.

B. Plausible answer as increased pump flow would decrease pool temp; however, allowed modes of operation for FPCC per B.2.1-01..B (page 4 of 16) indicates 2 HX are required to be in service with 2 pumps in service.

C. ARP 3-B-47 directs to increase RBCCW to affected Heat Exchanger. Since RBCCW valve for No. 11 FPCC HX is full open, an additional FPCC HX must be placed in service. Per B.2.1-01.8 (page 4 of 16) 2 HX may be in service with 1 pump.

D. Plausible answer as RBCCW provides cooling to FPCC and ARP 3-B-47 directs to monitor all RBCCW cooled equipment adjust cooling as necessary. However this action would have an adverse impact on FPCC cooling.

COGNITIVE LEVEL JUSTIFICATION Level 2 based on examinee must recognize alarm set point, procedure step in response to alarm 3-B-47 and system flow path and relationships between RBCCW and FPCC systems.

M(C WMI*ITTEF 19KA1M/f]HAkM©H H*C Revfihw C*y Page 43 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 35 Exam: BOTH System: 233000 K/A: K4.07 Lesson Plan M-8107L-022 Enablina Obiective 5.a. 9.i The plant is operating at 100% power following a refueling outage, which was completed 20 days ago. During the refueling outage, No. 11 and No. 12 Fuel Pool Cooling pumps were replaced to address an equipment obsolescence issue. Since plant start-up, the vibration readings on the Fuel Pool Cooling pumps have steadily increased and have reached the "Required Action" range per the Fuel Pool Cooling pump surveillance procedure.

Plant engineering personnel were in the process of taking vibration readings on the Fuel Pool Cooling pump when the No. 11 pump MCC breaker tripped OPEN, followed about 20 minutes later by a trip of the No. 12 pump MCC breaker.

Engineering has determined that with the current Fuel Pool heat load, the Fuel Pool temperature will reach 140OF in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Plant maintenance personnel indicate that replacement pumps are in transit from the manufacturer, but will not be on-site and installed until 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from now.

What course of action is required to restore Fuel Pool cooling?

A. Place the Tri-Nuc Underwater Filter in operation in the Fuel Pool to provide supplemental emergency Fuel Pool cooling.

B. Place loop B of the Residual Heat Removal system in service in the Fuel Pool Cooling mode to provide emergency Fuel Pool cooling.

C. Establish a feed and bleed of the pool using local Fire hose stations for make-up and pool drains for discharge.

D. Establish Fuel Pool cooling using loop A of Residual Heat Removal aligned for alternate fuel pool cooling while RBCCW is unavailable.

Answer: B

Reference:

B.2.1-05.H.8 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer because B.2.1-05.H.9, directs use of the Tri-Nuc system as an alternate method for fuel pool filtering. However, the system does not have any heat removal capability.

B. B.2.1-05.H.8 directs if the fuel pool gates are installed (plant is operating) and pool cooling is lost, then monitor fuel pool temperature rates and use procedure 3.2.4 05.G.5, special procedures - Emergency Fuel Pool Cooling, as needed. Design temperature limit of 140OF being reached in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with no pumps available would indicate that emergency cooling is needed. Procedure and system interconnections required loop B of RHR to be used for Fuel Pool cooling.

C. Plausible answer as this could provide some cooling, however per B.2.1-05.H.4 states that Fire protection system should not be used for make-up with other preferred sources available (i.e., Cond Serv or Demin water). In addition, without pump operation temperature stratification could occur. Note of procedure B.2.1-05.H.8 specifies that a HRC W iIF¶ T9 H A1FEI©H FR(C IRevnw (Copy Page 44 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank feed and bleed may be used if FPCC pumps are available, which they are not.

D. Plausible answer as B.2.1-05.H.8 directs if shut down cooling is needed and the cavity is flooded (plant is operating), then use special procedure 8147, ALTERNATE FUEL POOL COOLING WHILE RBCCW IS UNAVAILABLE. Differentiation from correct answer is also provided as loop A of RHR can be used per procedure 8147; however loop A of RHR cannot be used for the emergency Fuel Pool cooling alignment.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on must recall system facts and procedure steps. System inter relationships between RHR and FPCC does not categorize as a level 2 because system interactions are taught in close proximity and in same lesson plan.

• ,*J HMC WM*ITUTEN IEZAI*HNAUHI© h]RC Review Copy Page 45 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 36 Exam: BOTH System: 239002 K/A: K5.06 Lesson Plan M-81 07L-0: 25 Enablina, Ohi*rtiv,. 9h R V Given the following conditions.

"* A Reactor scram and Main Steam Isolation Valve (MSIV) closure from 90% power occurred.

"* The Safety Relief Valves (SRVs) are cycling to control Reactor pressure.

Which of the following Primary Containment parameters indicates that one of the SRV tailpipe vacuum breakers has failed OPEN?

A. Drywell pressure will go up each time the SRV cycles.

B. Torus pressure will go up each time the SRV cycles.

C. Drywell Equipment Drain Sump level will increase each time the SRV cycles.

D. Torus temperature will show rapid localized rises each time the SRV cycles.

Answer: A

Reference:

B.3.3-01.C.7 & B.2.4-01I.C.2 Question Pedigree: INPO BANK 8533 Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Per B.2.4-01 .C.2 the SRV vacuum breakers are located in the DW on elevation 954.

Vacuum breaker stuck open would release steam directly to the DW each the SRV cycles open resulting in an increase in DW pressure.

B. Plausible answer because Torus pressure would increase if SRV vacuum breakers were located in the Torus. DW to Torus vent downcomers would discharge to the Torus, steam would be condensed in the Torus water space and thus Torus pressure would not increase.

C. Plausible answer as the Equipment drain sump collects leakage piped from equipment in the DW; however the floor drain sump would collects leakages from non-piped sources.

D. Plausible answer because steam release to the Torus could cause localized temp increases. However, SRV vacuum breaker open will release steam directly to the DW.

COGNITIVE LEVEL JUSTIFICATION Level 3 based on examinee must predict an event outcome based on knowledge of SRV vacuum breaker function, location, and primary containment interconnections.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 37 Exam: BOTH System: 241000 K/A: K3.02 Lesson Plan M-8107L-048 Enablina Obiective The plant is operating at 100% power with Reactor pressure controlled by the EPR.

If the EPR pressure set point fails high, then which of the following describes the expected final plant conditions assuming no operator action is taken?

A. Reactor pressure will increase and Reactor power will increase.

B. Reactor pressure will decrease and Reactor power will decrease.

C. Reactor pressure will increase and Reactor power will decrease.

D. Reactor pressure will decrease and Reactor power will increase.

Answer: A

Reference:

C.4-B.5.9.B Bases Question Pedigree: Mod Bank M8107L- Cog. Level: 1 048-010 ANSWERPDISTRACTER JUSTIFICATION A. On a failure of the controlling pressure regulator, pressure would increase to the backup pressure regulator set point and stabilize. If the backup regulator was not available, then pressure would continue to increase. An increase in reactor pressure results from TCVs and Bypass valves closing as the set point for the pressure regulator increases.

Reactor power to increase from collapsing steam voids.

--- B. See correct answer.

C. See correct answer.

D. See correct answer.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recall system set points and facts.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank

'~-*-- Question: 38 Exam: BOTH System: 241000 K/A: K5.04 Lesson Plan M-8107L-0z48 Enablina Obiective The plant is operating at 100% power. During restoration of the Reactor Water Clean-Up (RWCU) system following maintenance activities, an inadvertent Group 3 isolation occurs.

The Control Room Operator is in the process of restoring RWCU to service and has OPENED MO-2398, RWCU OUTBOARD ISOLATION. Valves MO-2397, RWCU INBOARD ISOLATION, and MO-2399, RWCU RETURN ISOLATION, are CLOSED. The operator is in the process of verifying that the RWCU pressure and Reactor pressures are equalized and receives the following information.

" The Reactor Building Operator reports that RWCU pump suction pressure is 1010 psig.

" Normal Reactor pressure indication is unavailable due to I&C work that was in progress. The Lead Control Room Operator reports that the Main Steam Line Pressure Averaging Manifold pressure is 960 psig.

What is the next action the Control Room Operator should take to restore the RWCU system?

A. RWCU pressure and Reactor pressure are equalized and MO-2397, RWCU INBOARD ISOLATION, should be fully OPENED.

B. RWCU pressure and Reactor pressure are NOT equalized and MO-2397, RWCU INBOARD ISOLATION, should be throttled OPEN to increase RWCU pressure.

C. RWCU pressure and Reactor pressure are equalized and MO-2399, RWCU RETURN ISOLATION, should be fully OPENED.

D. RWCU pressure and Reactor pressure are NOT equalized and MO-2404, RWCU DUMP TO HOTWELL, should be OPENED and CV-2403, DUMP FLOW, throttled OPEN to lower RWCU pressure.

Answer: A

Reference:

B.5.9-01.C.1, C.4-B.4.1.C Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Per B.5.9-01I.C. 1, at 100% power, the reactor pressure will be 50 psig above the pressure sensed by the pressure regulators due to pressure drop in the steam lines.

Thus for EPR controlling at 960 psig, reactor pressure is 1010 psig and reactor pressure is equalized with RWCU pressure. C.4-B.4. 1.C step 8.f, MO-2397 is fully opened if RWCU and Reactor pressure is equalized.

B. Plausible answer because using information provided could fail to recall that reactor pressure is higher than EPR controlling set point and take inappropriate actions.

C. Plausible answer because using information provided could recognize pressure is equalized but take incorrect action. MO-2399 is opened after inlet side to RWCU is un isolated (MO-2397 open).

D. Plausible answer because using information provided could fail to recall that reactor pressure is higher than EPR controlling set point and take inappropriate actions.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Opening MO-2404 and CV-2403 is performed to depressurize RWCU during RWCU system shutdown.

COGNITIVE LEVEL JUSTIFICATION Level 2 based on examinee must recognize/recall 1) at 100% power the reactor pressure is 50 psig greater than the EPR control set point, 2) procedure actions to restore RWCU system to service, 3) based information determine RWCU pressure with respect to reactor steam dome pressure.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 39 Exam: BOTH System: 245000 K/A: A2.09 Lesson Plan M-8107L-013 Enabling Objective 9.h The plant is operating at 75% power and increasing power to 100% following completion of a refueling outage.

Annunciator 7-A-24, TURB BRG DRN HI TEMP, alarms and the Lead Reactor Operator reports that No. 6 Turbine bearing temperature is 156 0 F and increasing.

Shortly after receiving annunciator 7-A-24, annunciator 7-B-33, TURBINE VIBRATION HIGH, alarms. A Control Room Operator observes that the No. 6 bearing vibration is 13 mils and increasing.

What is the expected plant response, AND what operator action would be taken based on the expected plant response?

A. Turbine trip at 225 0 F lube oil temperature; enter C.3, SHUTDOWN PROCEDURE.

B. Turbine trip at 15 mils vibration, enter C.3, SHUTDOWN PROCEDURE.

C. Turbine trip at 225°F lube oil temperature; enter C.4-A, REACTOR SCRAM.

D. Turbine trip at 15 mils vibration, enter C.4-A, REACTOR SCRAM.

Answer: D

Reference:

C.6-007-B-33, B.6.1-02.4.g, C.6-007-B-25 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. High lube oil temperature does not result in a turbine trip. Answer is plausible as a turbine trip does occur on high exhaust hood temp at 225F. Action to enter C.3 is not appropriate for turbine trip.

B. Answer is incorrect. A turbine trip will occur at 15 mils; however, appropriate action to take is to enter C.4-A as a Rx scram will occur with the turbine trip.

C. High lube oil temperature does not result in a turbine trip. Answer is plausible as a turbine trip does occur on high exhaust hood temp at 225F. Action to enter C.4-A is appropriate for turbine trip.

D. Per ARP C.6-007-B-33 a turbine trip occurs at 15 mils turbine vibration. Per B.6.1 02.4.g, a turbine trip results in a turbine lockout relay trip. C.6-007-B-25, actions on turbine trip are to enter C.4-A as a Rx scram should occur on turbine stop valves closing with the turbine trip.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recall automatic set points and required actions on reaching set point.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank SI Question: 40 E (am: BOTH System: 259002 K/A: A1.05 Lesson Plan M-81 07L-046 Fn~ihlinn Ohi~r~tiv 7 A The plant is operating at 100% power when an inadvertent Reactor scram occurs due to personnel bumping sensitive Reactor Protection Instrumentation sensing lines. The plant responds as designed and conditions are as follows.

"* Reactor pressure is 950 psig and lowering.

"* Reactor level is 0 inches and rising.

"* No. 11 Reactor Feedwater pump is in service.

"* The CV-6-13, Low Flow Valve, controller setpoint is at 5 inches and in AUTO.

"* The Main Feedwater Regulating valves are CLOSED with their controllers in MANUAL.

"* MO-1 133, HP FW Line A Block Valve, and MO-1 134, HP FW Line B Block Valve are CLOSED.

Without further operator action, the Feedwater system will try to restore Reactor level to...

A. 5 inches.

B. 15 inches.

C. 40 inches.

D. 48 inches.

Answer: B

Reference:

B.5.7-02.B.5.a, C.4-A, Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer based on low flow FW reg valve is set a 5 inches. However, per B.5.7-02.B.5.a, the level controller has a low set point clamped value of 15 inches.

B. Per C.4-A, bases, the coordinated action of tripping one FW pump, transferring level control to the Manual Loading Station for the Low Flow Valvie, and closing the HP FW line block valves are intended to support Reactor level control between 15 and 20 inches. C.4-A step 3, operator is directed to place Low Flow set point between 15 and 20 inches. With the low flow controller set at 5 inches per B.5.7-02.B.5.a, the level controller has a low set point clamped value of 15 inches.

C. Plausible answer based on the low flow valve controller has an upper clamped setpoint of 40 inches.

D. Plausible answer based on the feedwater pumps will trip at plus 48 inches.

COGNITIVE LEVEL JUSTIFICATION LEVEL 1 based on examinee must recall system set points, interlocks, and response.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank

-*' Question: 41 Exam: BOTH System: 259002 K/A: A2.05 Lesson Plan M-8107L-046 Enabling Objective 8.a, 9.j The plant is operating at 100% power when annunciator 5-B-30, FW CONTROL VALVE LOCKED, alarms. No other Control Room annunciators are received. The operator observes the Control Room panel C-05 indications as shown on the following page.

What malfunction is indicated, AND what action should the operator take in response to the malfunction?

A. A loss of control air pressure has occurred for CV-6-12A. Control Reactor level with CV 6-12B from the Control Room MANUAL/AUTO STATION, and with local MANUAL control of CV-6-12A.

B. A partial loss of AC Distribution Panel Y-30 has occurred. Control Reactor level with CV 6-12B from the Control Room MANUAL/AUTO STATION, and with local MANUAL control of CV-6-12A.

C. A loss of control air pressure has occurred for CV-6-12B. Control Reactor level with CV 6-12A from the Control Room MANUAL/AUTO STATION, and with local MANUAL control of CV-6-12B.

D. A partial loss of AC Distribution Panel Y-30 has occurred. Control Reactor level with CV 6-12A from the Control Room MANUAL/AUTO STATION, and with local MANUAL control of CV-6-12B.

"- "Answer: A

Reference:

ARP 5-B-40, C.4-B.5.7.A, B.5.7-03.B Question Pedigree: NEW Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Per ARP 5-B-40 and B.5.7-03.B, Low Control Air pressure will provide a lockup of FW Reg Vales. Confirmatory indications for locked FW reg Valve is that the "Reset" pushbotton light is ON. Provided picture provides indication of A FW Reg Valve locked and B FW Reg Valve NOT locked. C.4-B.5.7.A directs as an immediate action to control Rx level in manual using any unlocked FW Reg Valve. Subsequent action of C.4.B.5.7.A directs to take local manual control of FW Reg Valve that is locked up.

B. Plausible answer because a loss of Y-30 does result in a lock of FW Reg Valve; However both FW reg Valves would lock-up and the FW Reg Valve RESET pushbuttons would NOT be ON for either valve. Both FW Reg Valves are powered from the same circuit off of Y-30, thus a partial loss involving FW Reg Valve would still impact both FW Reg vavles.

C. Plausible answer because provided indications are for locked FW Reg Valve; However examinee must recognize the RESET pushbutton light ON indicates valve is locked.

Thus CV-6-12B is NOT in a locked condition.

D. Plausible answer because a loss of Y-30 does result in a lock of FW Reg Valve; However both FW reg Valves would lock-up and the FW Reg Valve RESET pushbuttons would NOT be ON for either valve. Both FW Reg Valves are powered from the same circuit off of Y-30, thus a partial loss involving FW Reg Valve would still impact both FW Reg vavles.

COGNITIVE LEVEL JUSTIFICATION NRC RIvhw jCpy Page 52 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Level 3 based on examinee must solve a problem using knowledge, predict event outcome based on indications, and take correct actions.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 42 Exam: RO System: 261000 K/A: A2.01 Lesson Plan M-8107L-008 Enabling Objective 8.b, 8.c SBGT Train B is to be tested for Post Maintenance Testing by performing procedure B.04.02 05.D.4, MANUALLY INITIATE SBGT B TRAIN. Given the following:

"* HS-2988B, SBGT UNIT B (MANUAL/AUTO), has been placed in POSITION 1 (MANUAL).

"* Additional actions to start the SBGT train are about to be performed.

"* A Reactor scram then occurs due to high Drywell pressure.

"* MCC-133A becomes locked out on overcurrent.

"* Annunciator 24-A-2A, LOW FLOW 3000 CFM, has been alarming for 15 seconds.

What is the status of SBGT and what action should be taken?

A. Low SBGT flow exists therefore HS-2983B, V-EF-17B ON/OFF, control switch should be placed in ON.

B. Low SBGT flow exists therefore HS-2983A, V-EF-1 7A ON/OFF, control switch should be placed in ON.

C. No SBGT flow exists therefore HS-2988A, SBGT UNIT A (MANUAL/AUTO), control switch should be placed in AUTO.

D. No SBGT flow exists therefore HS-2988B, SBGT UNIT B (MANUAL/AUTO), control switch should be placed in AUTO.

Answer: D

Reference:

B.4.2-01.C.3, B.04.02-05.D.1 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Starting the 'B' SBGT fan will not restore adequate system flow since the inlet and outlet dampers would still be closed.

B. Placing the control switch for the 'A' SBGT fan to ON will not start the fan since there is no power to the fan.

C. The 'A' SBGT train should already be in auto however, without power the train will not start.

D. Placing the 'B' SBGT train control switch in auto will allow the train to start and establish adequate system flow.

COGNITIVE LEVEL JUSTIFICATION Level 3 since the student is required to solve the question using knowledge of the procedures after analyzing the initial conditions to determine why the 'B' train didn't start.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 43 Exam: RO System: 261000 K/A: K1.02 Lesson Plan M8107L-04 r%

Given the following:

"* The plant is in a startup following a refueling outage.

"* The containment is still de-inerted.

"* Drywell pressure is 1.4 psig and has been slowly rising during the startup.

Which of the following describes the action to take for the above stated condition AND why?

A. Scram the reactor because Drywell pressure should never rise this high during a startup without a reactor coolant leak.

B. Vent the primary containment using SBGT because Drywell pressure is expected to rise by more than 2 psig during containment heatup.

C. De-inert the primary containment using Procedure 2140, DE-INERTING PRIMARY CONTAINMENT, to reestablish normal Drywell pressure conditions.

D. Shutdown the reactor using C.3, SHUTDOWN PROCEDURE, as required by C.4-B.4.1.F, LEAK INSIDE PRIMARY CONTAINMENT, due to abnormally high Drywell pressure.

Answer: B

Reference:

B.04.01-05.G.4 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. See answer B.

B. During heatup of the primary containment it is expected that primary containment pressure can change as much as 2.3 psig when changing temperature from 70°F to 150 0 F.

C. See answer B.

D. See answer B.

COGNITIVE LEVEL JUSTIFICATION Level 3 since the student is required to solve the problem using knowledge of normal plant conditions during a startup and then use that information and apply it to a reference of a procedure that is normally conducted for this condition.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 44 Exam: SRO System: 262001 K/A: 2.2.21 Lesson Plan M-8108L-0: Enabling Objective 2.s M-8108L-0: 2.z. 2.aa The plant is operating at 99% power. On-line maintenance activities have been in progress during the last two days on Loop A of the Core Spray (CS) system.

You have been directed to review Work Order (WO) 0201354 for the Shift Supervisor Completion review.

Review of the WO identifies the following.

"* The Work Requested/Symptoms section of the WO states: "Perform routine meggering of No. 11 CS pump motor."

"* The WO contains one step to perform meggering of No. 11 CS pump motor, and the step has been completed.

"* The Responsible Supervisor and the workers have authorized release of the WO isolation for isolation tag removal.

"* The WO isolated the following CS Loop A components:

• Breaker 152-505, P-208A (11 Core Spray Pump) 4KV Supply, is REMOVED.

• Breaker 152-505 NR Fuses are REMOVED.

• Hand Switch HS-14A-S5A, 11 Core Spray Pump, is in PULL-TO-LOCK.

What action should be taken concerning Post Maintenance Testing for this WO?

A. Return the WO to the WO preparer for processing of a WO temporary procedure change to include post-maintenance testing of the No. 11 CS pump motor breaker.

B. Return the WO to the Responsible Supervisor for preparation of post-maintenance testing to include a load test of the breaker by starting No. 11 CS pump.

C. Return the WO to Quality Control (QC) for insertion of QC inspection points, and operability and post-maintenance testing requirements for the No. 11 CS pump motor.

D. Return the WO to the Responsible Supervisor for preparation of post-maintenance testing to include protective relay testing of the No. 11 CS pump motor breaker.

Answer: B

Reference:

OWI-03.02 sect 4.1.4, 4AWI-04.05.05, sect 4.4, B.09.06-05.G.1 Part D Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer because 4AWI-02.02.05, section 4.6 requires temporary procedure changes to WO when changes are made after the Ss approval to start work. However, 4AWI-02.02.05 specifically excludes a temp change for the addition of testing requirements. Processing of a WO temp change unnecessarily delays return to service of safety related equipment.

B. Per OWI-03.02, section 4.1.4, 4KV or 480 V breakers removed from their normal in service condition should be considered inoperable until the breaker/bucket is load HIR(C 1F?1UIiN ]MIFHAMNlTOU S-RC R vw (Copy Page 57 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 45 Exam: BOTH System: 262001 K/A: A4.01 Lesson Plan M-8107L-039. Rev. 14 FEnahlinn Fdphimtv An operator is getting ready to transfer Bus 13 from the 1 R to the 2R Transformer, when he observes that the red indicating light for breaker 152-302, NO. 1R RES XFMR SEC ACB, is out.

A check of the local indication shows the red indicating light for breaker 152-302 is also out.

Neither of the bulbs are burned out.

What is this indicative of, and what action should be taken?

A. Loss of electrical tripping capability from the control room only, place 152-302 control switch in PULL-TO-LOCK and trip the breaker with the mechanical trip pushbutton.

B. Loss of electrical tripping capability from the control room and locally, place 152-302 control switch in PULL-TO-LOCK and trip the breaker with the mechanical trip pushbutton.

C. Loss of electrical tripping capability from the control room only, remove all loads from the bus before transferring the bus to the 2R Transformer.

D. Loss of electrical tripping capability from the control room and locally, remove all loads from the bus before transferring the bus to the 2R Transformer.

Answer: B

Reference:

Ops Man B.09.06-05 Question Pedigree: New Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Will lose capability to trip the breaker locally also.

B. Loss of the red indicating lights in the Control Room and locally is indicative of a malfunction of the tripping circuit which prevents the breaker from being tripped electrically from the Control Room and locally. B.09.06-05 has the switch placed in PULL-TO-LOCK and then someone is dispatched to mechanically trip the breaker locally.

C. Will lose capability to trip the breaker locally also, and see correct answer for action.

D. See correct answer for action.

COGNITIVE LEVEL JUSTIFICATION Level 2 - requires the candidate to recognize 1) that the loss of indication is indicative of a tripping circuit malfunction and 2) what the required action is.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 46 Exam: RO System: 264000 K/A: A4.05 Lesson Plan M-8107L-042 Enabling Objective 9.a Given the following:

"* No. 11 EDG is paralleled to Bus 15 for the monthly surveillance.

"* No. 11 EDG load is being held constant at 2500 KW.

"* No. 11 EDG Power Factor is being adjusted to approximately 1.0.

Using the attached drawing (Figure 1, 11 EDG Amperes Limitations), which of the following sets of parameters results in a power factor closest to 1.0 for the No. 11 EDG?

A. Grid voltage is 4100 volts and AO amp reading is 370 amps.

B. Bus 15 voltage is 4125 volts and BO amp reading is 365 amps.

C. Grid voltage is 4150 volts and BO amp reading is 355 amps.

D. Bus 15 voltage is 4175 volts and AO amp reading is 340 amps.

Answer: D

Reference:

Procedure 0187-01 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Grid voltage is not used to determine power factor for the 11 EDG.

~--' B. This would result in a power factor that is acceptable but is 15 amps higher than required for a power factor of 1.0.

C. See answer A.

D. This would result in a power factor that is acceptable but is 6 amps lower than required for a power factor of 1.0. However, this would result in a power factor closest to 1.0 of all the sets of parameters above.

COGNITIVE LEVEL JUSTIFICATION Level 3 since the student is required to solve the question using the attached reference figure from the procedure.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 47 E) cam: RO System: 264000 K/A: K5.05 Lesson Plan M-81 07L-042 Enahlinn ONhitiva a k Which of the following describes the operational implication of paralleling 11 EDG to the grid as part of the monthly surveillance AND why?

A. 11 EDG is operable since it is already supplying power to Bus 15 therefore it is meeting its intended safety function.

B. 11 EDG is inoperable while being barred over because it will not be able to respond within 10 seconds of an ECCS signal.

C. 11 EDG is inoperable because the output breaker could trip on overspeed if the grid were to fail due to overloading of the diesel.

D. 11 EDG is operable but will not be able to handle full load on Bus 15 since speed droop has been adjusted to between 50-60%.

Answer: B

Reference:

B.09.08-05.A.3.k Question Pedigree: New Cog. Level: 1 ANSWERJDISTRACTER JUSTIFICATION A. The 11 EDG is inoperable during the surveillance.

B. This is correct.

C. The output breaker does not trip open on overspeed.

D. The 11 EDG is inoperable during the surveillance.

COGNITIVE LEVEL JUSTIFICATION Level 1 since the student only needs to recognize a System/Component Operability Requirement in the Ops Manual.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 48 Exam: RO System: 271000 K/A: K4.01 Lesson Plan M-8107L-01 I P::nnhlinn K"hir-fivi'm ,);

Which of the following components is designed to reduce hydrogen concentration to < 4%,

AND, if hydrogen concentration exceeds 4% as determined by the hydrogen analyzers, what will be the result?

A. Recombiner Preheater CLOSES AO-1085A/B B. 2nd Stage Steam Jet Air Ejector CLOSES AO-1085A/B C. Recombiner Preheater TRIPS Off-Gas Compressors D. 2nd Stage Steam Jet Air Ejector TRIPS Off-Gas Compressors Answer: D

Reference:

B.07.02.01-01 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Preheater raises the temperature at the Recombiner inlet to prevent condensation in the recombiner and ensures the catalyst is not wetted. AO-1085A/B does not close on a hydrogen concentration of 4%.

B. AO-1085A/B does not close on a hydrogen concentration of 4%.

C. Preheater raises the temperature at the Recombiner inlet to prevent condensation in the recombiner and ensures the catalyst is not wetted.

D. Correct answer COGNITIVE LEVEL JUSTIFICATION Level 2 - Requires the candidate to have knowledge of the function of the components in the Off-Gas system and the interlocks associated with high hydrogen concentration.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 49 E)(am: BOTH System: 272000 K/A: A3.09 Lesson Plan M-81 07L-070 Enablina Obiective The plant is currently shutdown with RHR "A" in shutdown cooling.

A plant transient has occurred causing the following annunciators:

"* 3-A-49, SBGT ANNUNCIATOR

"* 5-A-01, REAC BLDG VENT & F P RAD CH A-HI/LO

"* 3-B-55, REACTOR BLD EXH PLENUM HI RAD Which of the following CLOSED valves would be positive indication that these are valid annunciators?

A. MO-2397, RWCU INLET INBOARD B. MO-2014, RHR DIV 1 LPCI INJECTION INBOARD C. CV-331 1, TORUS TO CTMT RAD MON OUTBOARD D. AO-2541A, DRYWELL FLOOR DRAIN ISOLATION Answer: C

Reference:

Ops Man C.4-B.04.01.B Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. MO-2397 would close on a Group 3 isolation. So it would not show positive indication of a "Partial" Group 2 isolation.

B. The annunciators indicate that a "Partial" Group 2 isolation has occurred. MO-2014 is a Group 2 valve, but does not close on a "Partial" Group 2 isolation, only on a full Group 2 isolation.

C. The annunciators indicate that a "Partial" Group 2 isolation has occurred. Cv-3311 is a Group 2 valve and does close on Rx Building plenum Hi Rad. This would be an indication that the "Partial" Group 2 has occurred.

D. The annunciators indicate that a "Partial" Group 2 isolation has occurred. AO-2541A is a Group 2 valve, but does not close on a "Partial" Group 2 isolation, only on a full Group 2 isolation.

COGNITIVE LEVEL JUSTIFICATION Level 2 - must be able to discern that the annunciators indicate a "Partial" Group 2 isolation, and determine which of the valves would close due to the isolation signal.

This requires analysis of more than one piece of data to correctly answer the question.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 50 Exam: BOTH System: 272000 K/A: K2.05 Lesson Plan M-8107L-077 Enablinq Obiective 4 What is the normal source of power to:

RM-17-452B, REACTOR VENTILATION EXHAUST MONITOR B?

A. RPS Panel Y-40 B. RPS Panel Y-50 C. UPS Panel Y-70 D. UPS Panel Y-80 Answer: A

Reference:

B.9.12-05.C Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Correct Answer.

B. This panel supplies the RM-17-452A Rad Monitor.

C. This panel supplies essential loads but not Radiation Monitoring.

D. This panel supplies essential loads but not Radiation Monitoring.

COGNITIVE LEVEL JUSTIFICATION Level 1 - based on fundamental knowledge of system response and facts.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 52 Exam: RO System: 288000 K/A: K1.06 Lesson Plan M-81 07L-027 ai A loss of control air to the Reactor Building Heating & Ventilation units has occurred.

In accordance with Ops Man B.8.7-05.04, HEATING AND VENTILATION (H&V), an immediate shutdown of the H&V Units is required.

What condition could shutting down the H&V Units cause, AND what action must be taken?

A. General area temperatures could exceed 104 0 F; enter C.4-A, REACTOR SCRAM.

B. General area temperatures could exceed 1040 F; enter C.4-F, RAPID POWER REDUCTION.

C. Main Steam Chase temperatures could exceed 165 0 F; enter C.4-A, REACTOR SCRAM.

D. Main Steam Chase temperatures could exceed 165 0 F; enter C.4-F, RAPID POWER REDUCTION.

Answer: D

Reference:

Ops Man C.4-B.08.07.A Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Area temperatures could exceed 104 0 F but action is to maintain Bus 15 and Bus 16 voltages >4100 V.

B. Area temperatures could exceed 104 0 F but action is to maintain Bus 15 and Bus 16 voltages >4100 V.

C. Correct action is to enter C.4-F and perform Rapid Power Reduction.

D. Correct Answer COGNITIVE LEVEL JUSTIFICATION Level 2 - requires the candidate to understand the interaction between the loss of ventilation and the condition that it would cause and then the correct action that must be taken.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 53 Exam: SRO System: 290001 K/A: 2.4.48 Lesson Plan M-81 14L-005 Enabling Objective 4.b & c The plant is operating at 100% power. A plant transient is in progress creating the following conditions.

"* 3-B-56, HIGH AREA TEMP STEAM LEAK, has alarmed.

"* TR-4926 points 13 through 20 are all reading between 160°F and 185 0 F, and rising.

"* Isolation of Reactor Water Cleanup has been attempted.

"* MO-2397, CLEANUP INLET INBOARD VALVE, has failed to CLOSE.

"* MO-2398, CLEANUP INLET OUTBOARD, has failed to CLOSE.

What would be the next action directed by the Control Room Supervisor, AND why would that action be taken?

A. Shutdown the Reactor, to reduce the rate of energy production.

B. Shutdown the Reactor, to ensure equipment necessary for safe shutdown will operate.

C. Scram the Reactor, to reduce the rate of energy production.

D. Scram the Reactor, to ensure equipment necessary for safe shutdown will operate.

Answer: C

Reference:

C.5.1-1300 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Incorrect because there is a primary system (RWCU) discharging into the reactor building which requires the Scram action, and 2 areas of the same parameter have not reached max safe level.

B. Incorrect because there is a primary system (RWCU) discharging into the reactor building which requires the Scram action, and 2 areas of the same parameter have not reached max safe level.

C. Correct answer D. Incorrect because of the reason. The reason is associated with "Maximum Safe" operating level, which has not been reached yet.

COGNITIVE LEVEL JUSTIFICATION Level 2 because the candidate must be able to recognize that the temperature points puts them in Secondary Containment EOP and then use the information to determine where they are in the EOP.

SRO Level Justification SRO ONLY: candidate is required to know the EOP basis for the action taken, which is only required of the SRO candidates.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 54 E) (am: BOTH System: 290001 K/A: K3.01 Lesson Plan M-81 07L-006 E:nnhlinn Ohinrt"fu r, The plant has been operating at 100% thermal power. A transient occurs resulting in an elevated fission product release. The following conditions exist:

"* 3-B-55, REACTOR BLDG EXH PLENUM HI RAD is in ALARM.

"* SBGT A Train is running.

Given the above conditions, why should V-EF-20 through V-EF-22, REACTOR BUILDING MAIN EXHAUST FANS be verified in the OFF position, and what would be the effect if these fans were not secured?

A. To prevent the possibility of the Turbine Building AP becoming more negative than the Reactor Building which could cause some radioactivity to bypass the SBGT filtration.

B. To prevent the possibility of the Turbine Building AP becoming more negative than the Reactor Building, which could cause backflow through the SBGT train and reduce its effectiveness.

C. To prevent the possibility of the Main Exhaust Plenum Room being at a lower pressure than SCTMT which could cause some radioactivity to bypass the filtration by SBGT.

D. To prevent the possibility of the Main Exhaust Plenum Room being at a lower pressure than SCTMT which could cause backflow through the SBGT train and reduce its effectiveness.

Answer: C

Reference:

B.04.02-05.D.2 bases, C.4-B.2.4.A bases Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. The Turbine Building AP could be increased but would not become > Reactor Building.

B. The Turbine Building AP could be increased but would not become > Reactor Building.

C. Correct answer: the Main Exhaust Plenum Room could be at a lower pressure which could caused some leakage through that system instead of SBGT.

D. Would not cause backflow through SBGT.

COGNITIVE LEVEL JUSTIFICATION Level 2 - The candidate must be able to determine 1) why the fans must be secured and 2) what the effect would be if they are not tripped.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 55 Exam: RO System: 290002 K/A: 2.2.23 Lesson Plan M-8108L-038 Enablina Obiective 2.m The plant has just started up from an outage. During the outage, maintenance was performed on the Jet Pumps, and a special test has been developed to check the Jet Pumps.

The Limiting Condition for Operation (LCO) on Jet Pump operability must be entered and exited repeatedly during the performance of this test.

Which of the following requires tracking the LCO each time it is entered and exited?

A. Shift Supervisor's Log B. Control Room status board C. Reactor and Control Room Log D. Work Execution Center status board Answer: C

Reference:

OWI-02.02 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Required to log the initial entry and final exit.

B. Status is maintained but not required to address each entry and exit of the LCO.

C. Correct answer D. Status is maintained but not required to address each entry and exit of the LCO.

COGNITIVE LEVEL JUSTIFICATION Level 1 - requires the candidate to recall a step in a procedure.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 56 Exam: SRO-RO System: 290003 K/A: 2.3.10 Lesson Plan M-81 07L-049 Pnn h:ihln rl imih-, iv'

I ,.

Manual initiation of the EFT High Radiation Mode shall be performed if radiation level is _>

A. Control Room; 1 mrem/hr B. EFT Building; 1 mrem/hr C. Reactor Building; 3 mrem/hr D. 250 VDC Battery Room; 3 mrem/hr Answer: A

Reference:

B.08.13-05.H1 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Answer is verbatim from Ops Man B.08.13-05.04.

B. EFT Building radiation does not require manual initiation of the EFT high rad mode.

C. Reactor Building radiation of 3 mr/hr is normal in many areas and does not require initiation of the EFT high rad mode.

D. The radiation detectors for automatic initiation of the EFT high rad mode are located in the 250 VDC battery room however, intake air does not come from the battery room and therefore manual initiation of the EFT high rad mode would not be required.

COGNITIVE LEVEL JUSTIFICATION Level 1 - The information that the operator needs to know comes from a single procedure step and is a recognition of that step.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 57 Exam: RO System: 290003 K/A: K6.02 Lesson Plan M-8107L-049 Enablina Obiectivw A line break in the Service Water piping has rendered the Service Water system inoperable.

Which of the following Emergency Service Water (ESW) pumps must be started to keep V-EAC-14A, Division I Control Room Ventilation Train, operable?

A. No. 11 ESW Pump B. No. 12 ESW Pump C. No. 13 ESW Pump D. No. 14 ESW Pump Answer: C

Reference:

P & ID NH-36665, B.08.13-05.A.2.e Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. P-111A (No. 11 ESW) does not supply Control Room Ventilation B. P-I 11 B (No. 12 ESW) does not supply Control Room Ventilation C. Correct Answer D. P-I 110D (No. 14 ESW) supplies V-EAC-14B COGNITIVE LEVEL JUSTIFICATION Level 1 - requires recall of knowledge about loads on ESW.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 58 Exam: SRO System: GENERIC K/A: 2.1.4 Lesson Plan M-8108L-0( Enablina Obiective Postulated scenario:

It is 0400 on a quiet midshift during normal full power operation. The Control Room Supervisor's (CRS's) wife calls to tell him that she has gone into labor and that she must get to the hospital.

"* At 0405 the CRS departs as directed by the Shift Manager (SM).

"* At 0410 the SM calls the Operations Manager to inform him of the reduction in crew composition.

"* At 0420 the SM reaches a relief for the CRS and directs him to come to work.

"* At 0615 the CRS relief arrives and joins in on the turnover.

"* At 0645 the CRS shift turnover briefing is completed.

Which of the following is correct concerning the operating crew's compliance with the shift manning requirements?

A. The operating crew has complied fully with shift manning requirements.

B. The CRS may not leave until the Plant Manager's permission is obtained.

C. The CRS may not leave until his relief has arrived on site and has been briefed.

D. The CRS position must be manned by a relief within two hours of the CRS's departure.

Answer: D

Reference:

Tech Spec Section 6.0 & 4 AWI-04.01.01 Question Pedigree: DAEC 2001 NRC Cog. Level: 3 Exam QID #83 ANSWER/DISTRACTER JUSTIFICATION A. The crew did not comply since they were < 2 licensed senior operators for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. Plant Manager's permission is not required.

C. The CRS may leave without a relief as long as the position is filled within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. This is the correct answer.

COGNITIVE LEVEL JUSTIFICATION Level 3 because the student is required to solve the problem using knowledge of the references.

SRO Level Justification Per Ops Dept work instruction OWI-01.06, section 4.1.11 and 4.5.3.A.1.n, the responsibility to assure shift is properly manned is a Control Room Supervisor (SRO) function; in addition, the RO rating for this KA is below the 2.5 threshold.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 59 Exam: SRO-RO System: GENERIC K/A: 2.1.22 Lesson Plan M-8108L-003 Enablina Obiective 2 Given the following:

"* SRMs indicate 300 cps and steady.

"* Reactor coolant temperature is < 212 0 F.

"* The RPV head has been removed.

"* The reactor cavity is flooded and fuel pool gates are removed.

"* Core alterations are in progress.

Which of the following states the Mode of Operation the plant is in AND why?

A. Refuel because core alterations are in progress.

B. Refuel because the RPV head has been removed.

C. Shutdown because reactor coolant temperature is < 212 0F.

D. Shutdown because reactor power is steady in the source range.

Answer: A

Reference:

Tech Spec Section 1.0 Question Pedigree: New Cog. Level: 1 ANSWERIDISTRACTER JUSTIFICATION

,_ A. The refuel mode of operation applies when the reactor coolant temperature is < 2121F and core alterations are in progress.

B. Having the RPV head removed does not mean that core alterations are in progress.

C. See answer A.

D. Reactor power can be steady in the source range and not be in shutdown.

COGNITIVE LEVEL JUSTIFICATION Level 1 because the student must recall definitions.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 60 Exam: RO System: GENERIC K/A: 2.1.29 Lesson Plan M-8108L-039 Enablino ObiectiwA 2r, A manual throttle valve in a safety related system is to be positioned three turns closed from full open for a test procedure.

The Shift Manager assigns you the job of accompanying the operator who will adjust the throttle valve to assure that the valve is throttled as called for in the test procedure and to sign off the procedure step.

Which of the following are you being asked to perform?

A. PEER CHECK of the operator doing the valve manipulation.

B. WITNESS CHECK of the operator doing the valve manipulation.

C. DOUBLE VERIFICATION of the operator doing the valve manipulation.

D. INDEPENDENT VERIFICATION of the operator doing the valve manipulation.

Answer: B

Reference:

4 AWI-04.04.02 Question Pedigree: Clinton 2000 Exam Cog. Level: 1 INPO Bank #19053 ANSWER/DISTRACTER JUSTIFICATION A. A peer check does not sign off completion in procedures.

B. This is correct.

C. There is no verification called "double verification" in the AWl.

D. Independent verification can't be used when repositioning a throttle valve in a safety related system.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a fact from the AWl for verifications.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 61 E (am: BOTH System: GENERIC K/A: 2.1.31 Lesson Plan M-81 07L-003 IPnnhlinn 0hintr-tivm, M-81071--.--003~

Given that the plant is operating at 100% power, which one of the following RCIC valves is NOT in a normal standby lineup?

A. MO-21 10, RCIC TEST FLOW ISOLATION, indicates CLOSED.

B. MO-2106, RCIC PUMP DISCHARGE OUTBOARD, indicates OPEN.

C. CV-2104, RCIC PUMP MINIMUM FLOW VALVE, indicates CLOSED.

D. MO-2076, RCIC STEAMLINE ISOLATION OUTBOARD, indicates OPEN.

Answer: B

Reference:

B.02.03.03 Question Pedigree: Clinton 2001 NRC Cog. Level: 1 Exam QID #22 ANSWER/DISTRACTER JUSTIFICATION A. This is the expected position.

B. MO-2106 is normally closed when in standby.

C. This is the expected position.

D. This is the expected position.

COGNITIVE LEVEL JUSTIFICATION Level 1 since the question is asking facts about the RCIC system in a normal standby lineup.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 62 Exam: BOTH System: GENERIC K/A: 2.1.33 Lesson Plan M-8107L-029 Enablina Obiective 11_e The plant is operating at 100% reactor power with the following conditions:

"* Generated power has dropped from 613 MWe to 585 MWe.

"* Core plate d/p has dropped.

"* Jet Pump Loop 'A' flow indicates 26.5 Mlbm/hr.

"* Jet Pump Loop 'B' flow indicates 28.0 Mlbm/hr.

"* Jet Pump 1 flow indicates 2.8 Mlbm/hr.

"* Jet Pump 6 flow indicates 2.8 Mlbm/hr.

"* Jet Pump 11 flow indicates 2.8 Mlbm/hr.

"* Jet Pump 16 flow indicates 1.3 Mlbm/hr.

Based on the above indications, what action is required per Tech Specs?

A. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> place the reactor in hot shutdown.

B. Place the reactor in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> adjust jet pump flow in the 'A' loop to match the 'B' loop.

D. Isolate the 'A' recirc loop and adjust the single loop operating setpoints within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Answer: B

Reference:

B.01.04-05.H.3 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Placing the plant in hot shutdown would not be consistent with action directed by B.01.04-05.H.3.

B. This is the correct action as directed by procedure B.01.04-05.H.3.

C. This would not be correct since this would not correct a failed jet pump.

D. Isolating the recirc loop associated with that jet pump would not isolate the jet pump so the problem would still exist.

COGNITIVE LEVEL JUSTIFICATION Level 3 since the student is required to analyze the conditions and determine what the course of action is based on information in Tech Specs. This is based on a recent industry event at Quad Cities for failure of a jet pump in 2002.

HRNC LyfwC(DpY Page 77 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 63 Exam: BOTH System: GENERIC K/A: 2.2.2 Lesson Plan M-81 13L-001 I::nahlin al thiz~i, 01, A reactor startup M-81131--001 is in progress. Conditions just prior to the startup and currently are listed below:

Beginning of Startup Currently

"*SRM 21 at 9 cps SRM 21 at 85 cps

"*SRM 22 at 11 cps SRM 22 at 100 cps

"*SRM 23 at 8 cps SRM 23 at 90 cps

"*SRM 24 at 10 cps SRM 24 at 95 cps

"*Moderator temperature was 1481F Moderator temperature is 149 0 F The reactor is NOT critical and you still have one Control Rod left to pull to complete the current sequence step.

In order to withdraw this Control Rod to continue the startup, what must you do per C. 1, STARTUP PROCEDURE, concerning the method of Control Rod withdrawal?

A. Change from continuous rod withdrawal to single notch withdrawal.

B. Change from single notch withdrawal to continuous notch withdrawal.

C. Use continuous notch withdrawal to notch 24 then single notch withdrawal to notch 48.

D. Use single notch withdrawal to notch 24 then continuous notch withdrawal to notch 48.

Answer: A

Reference:

C.1, Startup Procedure Question Pedigree: DAEC 2001 NRC Cog. Level: 3 Exam QID 92 ANSWER/DISTRACTER JUSTIFICATION A. See C.1, Startup Procedure, Section Ill.E, Control Rod Withdrawal Guidelines.

B. Per C.1 continuous rod withdrawal should have been used to this point.

C. This step only applies to rods designated "HI-LIGHTED/NOTCH" and does not override the requirement to single notch withdrawal to criticality.

D. See answer C.

COGNITIVE LEVEL JUSTIFICATION Level 3 since this question requires the student to solve the question using the knowledge that counts on SRM channel 23 have increased by a factor of 10 and the reactor is close to criticality. This information must then be applied to the restrictions of control rod withdrawal of the C.A procedure.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 64 Exam: BOTH System: GENERIC K/A: 2.2.11 Lesson Plan M-8108L-039 Enabling Objective 2.r Which of the following items is considered a temporary modification (jumper/bypass) requiring Form 3034, JUMPER BYPASS FORM, to be filled out?

A. Tygon hose installed on a hose station to fill a cleaning bucket.

B. Blocking device installed on a relief valve to prevent inadvertent opening.

C. Thermal overloads removed from a breaker specified on a work procedure.

D. Installation of an ECCS Test plug into Panel C-32 for performance of an OC approved procedure.

Answer: B

Reference:

4 AWI-04.04.03 Question Pedigree: Mod INPO Bank Cog. Level: 1 QID #19160 ANSWERIDISTRACTER JUSTIFICATION A. See exceptions in 4 AWI-04.04.03. Section 2.2.

B. See list in 4 AWI-04.04.03 Section 2.1.

C. See answer A.

D. See answer A.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this a fact from 4 AWI-04.04.03.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 65 Exam: RO System: GENERIC K/A: 2.2.22 Lesson Plan M-8108L-003 Enabling Obiective 4.c The Main Turbine Stop Valve RPS Trip function is designed, in combination with other LSSSs, to prevent any anticipated combination of transient conditions that would result in reaching which of the following limits?

A. The RPV pressure SAFETY LIMIT.

B. The RPV high pressure scram setpoint.

C. The fuel cladding integrity Minimum Critical Power Ratio (MCPR) SAFETY LIMIT.

D. The Average Planar Linear Heat Generation Rate (APLHGR) Power Distribution Limit.

Answer: C

Reference:

Tech Spec Section 2.0 Question Pedigree: DAEC 1999 RO NRC Cog. Level: 1 Exam QID #96 ANSWER/DISTRACTER JUSTIFICATION A. The Turbine Stop valve scram does lesson the pressure transient on the RPV but its main purpose is to avoid exceeding the MCPR safety limit.

B. See answer A.

C. This is the correct answer per the bases of D. APLHGR is not the thermal limit of concern during this transient.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is information directly from the Safety Limits section of the Tech Specs.

HJRC WMIlT71H ]FIAMAUMN© MRPC Repw (Cy Page 80 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 66 Exam: SRO System: GENERIC K/A: 2.2.27 Lesson Plan M-8107L-0 EnablinQ Obiective 7.a Which of the following is a responsibility of the Fuel Handling Supervisor during core alterations?

A. Obtaining permission from the Plant Manager if deviation from a refueling procedure is necessary.

B. Checking that portable monitoring instruments are available and their operation is understood.

C. Receiving verbal authorization from the Nuclear Engineer for movement of each fuel assembly both into and out of the core.

D. Delegating his responsibilities to a Licensed Reactor Operator when he cannot be present on the refuel floor during core alterations.

Answer: B

Reference:

D.2-05 Question Pedigree: New Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Plant Manager permission is not required to change a refueling procedure.

B. This is a responsibility of the Fuel Handling Supervisor.

C. Nuclear Engineer's permission is not required to move fuel.

D. Per OWI-01.06, section 4.3, an active SRO shall directly supervise (from the refueling floor) all alterations to the reactor core.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a procedural requirement for the responsibilities of the Fuel Handling Supervisor.

SRO Level Justification SRO knowledge level based on examinee must recall knowledge of Fuel Handling Supervisor duties. The Fuel Handling Supervisor is an SRO licensed individual.

NRC RIvwCjpy Page 81 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 67 Exam: BOTH System: GENERIC K/A: 2.2.30 Lesson Plan M-8107L-019 Enabling Obiective 7.a The plant is performing refueling operations. Which of the following activities is NOT a responsibility of the Operators in the Control Room during refueling?

A. Maintaining direct communications with the Refueling Bridge.

B. Maintaining the fuel accountability tag boards and computer data file.

C. Recording each fuel move on the refueling procedure verification checklist.

D. Monitoring of SRM count rates and notifying supervision of abnormal indications.

Answer: B

Reference:

D.1-05.B.2 and D.1-05.B.5 Question Pedigree: New Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Per D.1-05.B.6, the Control Room Operator is to be in direct communication with the Fuel Handling Supervisor on the Refuel Bridge.

B. Per procedure D.1-05.E.1 and D.2-05.E.2 this is a local visual verification performed by the Fuel Handling Supervisor and Accountability Recorder and is the overall responsibility of the Fuel Handling Supervisor.

C. Per D. 1-05.B.6, the Control Room Operator shall enter the time and date of completion of each fuel move on procedure checklist data file or Form 9027.

D. Per Procedure 9007, precautions 1, 2 and 5.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is procedural requirements.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 68 Exam: SRO System: GENERIC K/A: 2.2.32 Lesson Plan M-81107L-01 9 The following conditions exist.

"° The plant is in REFUEL with core shuffle in progress.

"* Two operable SRMs; located in quadrants in and adjacent to fuel being moved.

"* Quadrant where fuel moves are in progress contains 4 fuel assemblies.

"* Adjacent quadrant contains 5 fuel assemblies.

"* SRM in quadrant where fuel is being moved indicates 2 cps.

"* SRM in adjacent quadrant indicates 10 cps.

Which of the following describes the action(s) to be taken for the above stated conditions AND why?

A. Immediately suspend fuel moves and make event notifications. With <3 cps assurance of neutron flux monitoring is no longer provided to prevent inadvertent criticality.

B. Immediately notify the Nuclear Engineer and continue fuel moves with caution. It is expected that SRM indication in a quadrant with 2 or fewer fuel assemblies will be <2 cps.

C. Immediately halt fuel handling operations and make event notifications to required personnel. The refueling interlocks should have prevented fuel moves with a SRM <3 cps to prevent inadvertent criticality.

D. Immediately notify the Superintendent Nuclear Engineer and perform upcoming fuel moves with caution. With <3 cps assurance of neutron flux monitoring may no longer be provided to prevent inadvertent criticality.

Answer: A

Reference:

Tech Spec Bases 3.10. B &

D.2-05 Question Pedigree: New Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. This is the expected action to take lAW Ops Man D.2-05 Plant Operating Requirements and Bases of Tech Spec 3.10.B.

B. Fuel moves are not allowed to continue.

C. Refueling interlocks would not have prevented fuel moves.

D. Fuel moves are not allowed to continue even with Nuclear Engineer review.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this question is asking for the bases of why fuel moves need to be stopped with SRM indication <3 cps.

SRO Level Justification SRO knowledge based on the examinee must recall plant Technical Specification bases information and RO rating for this KA is below the 2.5 threshold.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 69 E)(am: BOTH System: GENERIC K/A: 2.3.1 Lesson Plan M-7703L-002 1"7 Fnnhlinn rNharftiga An area of the RADWASTE Building is set aside to store some highly radioactive material. It is determined that the entrance into this area could result in personnel receiving 7 Rem in one hour 30 centimeters from the radiation source.

Which of the following is the correct posting for this area?

A. High Radiation Area B. Very High Radiation Area C. Locked High Radiation Area D. Double Locked High Radiation Area Answer: C

Reference:

DEF-L Question Pedigree: DAEC '99 RO NRC Cog. Level: 1 Exam QID #95, INPO Bank QID 8853 ANSWER/DISTRACTER JUSTIFICATION A.

B. HRA is > 100 mrem/hr.

VHRA is > 500 rads in one hour.

C. LHRA is > 1000 mrem/hr.

D. This term is no longer used.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is based on definitions of the various areas.

MJRC Revhw (Copy Page 84 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 70 Exam: BOTH System: GENERIC K/A: 2.3.2 Lesson Plan M-8108L-039 Enabling Objective 2.q You have been directed to independently verify the closed position of a system drain valve.

To complete the task you will have to spend approximately 10 minutes in the general area of the valve. The dose rate in the general area of the valve is 1 Rem/hr.

Which of the following is the correct approach to fulfilling this task?

A. Perform the independent verification since total individual dose is expected to be less than the annual limit.

B. Request the Control Room Supervisor to waive the independent verification since a significant radiation hazard exists.

C. Request the Control Room Supervisor to review previous position checklists for the system drain valve to ensure that the valve is normally closed.

D. Perform the independent verification after installing shielding to reduce your dose to less than 5 mrem for verifying valve position.

Answer: B

Reference:

4 AWI-04.04.02 Question Pedigree: INPO Bank QID Cog. Level: 1

1. 19137 ANSWERIDISTRACTER JUSTIFICATION A. Independent verification may be waived if significant radiation hazard exists.

B. This is correct as written in 4 AWI-04.04.02 section 4.2.8.

C. Position verification from previous checklists may not take into consideration recent changes in the valve's position.

D. Installation of shielding will increase total accumulated dose for performing the task and would not be conservative for this evolution.

SRO Level Justification SRO knowledge level based on examinee must recall actions which are assigned per plant administrative procedures to Shift Managers/Control Room Supervisors (SRO licensed individuals) and RO rating for this KA is below the 2.5 threshold.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a administrative procedural requirement.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 71 Exam: SRO System: GENERIC K/A: 2.3.7 Lesson Plan None Found Enablina Ohiect.iv*.

In accordance with 4 AWI-08.04.05, RADIOLOGICAL WORK CONTROL, which of the following jobs will require its own RWP Request to be filled out specifically for the job?

A. Changing oil in the Fire Diesel Pump.

B. Adding water to the RBCCW Surge tank.

C. Cleaning out the Floor Drain Collector Tank.

D. Changing out of the Nitrogen bottles on the Alternate N2 System.

Answer: C

Reference:

4 AWI-08.04.05 Question Pedigree: New Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. This would be exempt since it is considered a job on an uncontaminated system in an uncontaminated area.

B. This would be exempt since it is a minor, well-defined job in a low hazard area.

C. This would require its own RWP.

D. This would be exempt since it is a job that is normally added to an existing extended (general entry) RWP.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a procedural requirement of 4 AWI-08.04.05.

SRO Level Justification SRO knowledge level based on the responsibility to complete RWP requests is a Shift Manager/Control Room Supervisor function based on plant administrative procedures, and based on low RO rating for KA.

HRC W UInTrzH ZEMIIHA1FII© HRC Reh( Copy Page 86 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 72 Exam: SRO-RO System: GENERIC K/A: 2.3.9 Lesson Plan M-8107L-044 Enablina Obiective 9.a The plant is operating at 5% power with the following conditions.

• C.3, SHUTDOWN PROCEDURE, is in progress.

0 The primary containment is being de-inerted.

• Instrument failure results in PCIS Group 2 isolation.

• 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later the failure is corrected and Group 2 isolation restored.

• It is desired to restart the containment purge.

Which of the following identify the requirement(s) to restart the containment purge?

A. Re-verify Containment Integrity is established per LCO 3.7.A.2.

B. Perform ASME Section Xl tests on containment purge isolation valves.

C. Notify I&C to re-calibrate the HAYS Oxygen Analyzer in the high range.

D. Verify ODCM-03.01, GASEOUS EFFLUENTS, requirements are met prior to restart of purge.

Answer: D

Reference:

Procedure 2140, B.04.01-05.A.2.e, ODCM-03.01 Question Pedigree: INPO Bank QID Cog. Level: 1

1. 13974 ANSWER/DISTRACTER JUSTIFICATION A. See answer D.

B. See answer D.

C. See answer D.

D. When the PCIS Group 2 isolation occurred the purge line-up was secured. Two hours later if the purge is to be reestablished then the requirements of the ODCM again need to be verified to ensure that conditions inside the containment haven't changed since the purge was secured.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a procedural requirement prior to every containment purge.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 73 E) cam: RO System: GENERIC K/A: 2.3.10 Lesson Plan M-81 07L-085 E:nghline, Chi*_r~fiv, Step I of procedure 1279-05, HOT AREA INSPECTIONS, requires reducing the hydrogen injection rate per Ops Man Section B.02.06-05, HYDROGEN WATER CHEMISTRY SYSTEM.

Which of the following describes the reason for performing this step?

A. Hydrogen injection rate is reduced as required by procedure before reducing reactor power and thus feed flow.

B. Hydrogen injection is reduced to minimize the explosion hazard during inspection in areas of suspected steam leaks.

C. Reducing hydrogen injection will reduce dose rates in the Steam Chase, Main Condenser Room and Turbine Operating Floor.

D. Reducing hydrogen injection will help ensure suitable breathing environment exists in confined spaces, such as the Steam Chase, prior to entry.

Answer: C

Reference:

B.02.06-05 & Procedure 1279-05 Question Pedigree: New Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. There is no specific requirement in the Procedure 1279-05 for reducing reactor power.

B. All of the areas being inspected contain ventilation systems and would not allow hydrogen to concentrate to the level necessary for an explosion to occur.

C. This topic is discussed in the Plant Operating Requirements section of the Ops Man specifically for hot side inspections/rounds.

D. The Steam Chase is not a confined space.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a procedural requirement of the Ops Man.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 74 Exam: BOTH System: GENERIC K/A: 2.4.8 Lesson Plan M8114L-005 Enablinq Obiective 3.a, 3.b The following conditions exist.

"* The plant was operating at 100% power when a scram occurred due to a loss of 2RS.

"* The following abnormal procedures are being performed:

"o C.4-A, REACTOR SCRAM "o C.4-B.1.4.B, TRIP OF TWO RECIRCULATION PUMPS

"* Two minutes later suppression pool temperature is 911F and rising.

Which of the following actions are required?

A. Concurrently enter C.5-1200, PRIMARY CONTAINMENT CONTROL.

B. Continue C.4-A, REACTOR SCRAM and monitor suppression pool temperature C. Exit both abnormal procedures and enter C.5-1200, PRIMARY CONTAINMENT CONTROL.

D. Exit C.4-A, REACTOR SCRAM, and enter C.5-1200, PRIMARY CONTAINMENT CONTROL.

Answer: A

Reference:

C.5.1-1000, OWI-01.04 Question Pedigree: INPO Bank QID Cog. Level: 1

1. 19293 ANSWER/DISTRACTER JUSTIFICATION A. EOPs and abnormal procedures can/should be performed concurrently.

B. Suppression pool temperature of 91OF is an EOP entry condition and therefore requires entry into EOP C.5-1200.

C. See answer A.

D. There is no need to exit an abnormal procedure when entering the EOPs.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a procedural use requirement written in C.5.1-1000, INTRODUCTION.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 75 Exam: BRO System: GENERIC K/A: 2.4.17 Lesson Plan M-8114L-01 Enabling Objective 3.d Per the Emergency Operating Procedures, which of the following conditions establishes "Adequate Core Cooling"?

A. Reactor level at 2/3 core height with Core Spray injecting into the Reactor vessel.

B. Reactor level at 2/3 core height with LPCI injecting into the Reactor vessel.

C. Reactor level below minus 126 (-126) inches with HPCI injecting into the Reactor vessel.

D. Reactor level below minus 126 (-126) inches with alternate injection sources injecting into the Reactor vessel.

Answer: A

Reference:

C.5.1-1000 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Correct answer per C.5.1-1000, Part B, Definitions B. Incorrect as adequate core cooling is established with Core Spray injection, not LPCI.

C. For core submergence, Rx level must be above top of active fuel (above -126").

D. For core submergence, Rx level must be above top of active fuel (above -126").

COGNITIVE LEVEL JUSTIFICATION Level I since this is evaluating EOP definition of terms.

HY¶C W*I TEh YEzAI*fIHAkUl©H MRC Re rw CIpy Page 90 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 75.1 Exam: BSRO System: GENERIC K/A: 2.4.17 Lesson Plan M-8114L-005 Enablina Obiective Given the following:

"* Reactor scram has occurred from a LOCA coincident with a Station Blackout.

"* Control Rods have fully inserted with the exception that 1 Control Rod is at position 30.

"* All high pressure injection sources have failed to inject and are not lined up for injection.

"* Reactor pressure is 750 psig and decreasing.

"* Reactor level is minus 95 (-95) inches and decreasing.

"* Torus level is 2.0 inches.

Which of the following identifies the RPV water level that will require opening SRVs, AND what is the basis for that action?

A. Minimum Steam Cooling RPV Water Level (minus 164 (-164) inches); to minimize break flow and reduce RPV pressure for low pressure injection sources.

B. Minimum Steam Cooling RPV Water Level (minus 149 (-149) inches); to minimize break flow and reduce RPV pressure for low pressure injection sources.

C. Minimum Zero-Injection RPV Water Level (minus 164 (-164) inches); to draw excess steam up through the fuel assemblies quenching the fuel and reducing cladding temperature.

D. Minimum Zero-Injection RPV Water Level (minus 149 (-149) inches); to draw excess steam up through the fuel assemblies quenching the fuel and reducing cladding temperature.

Answer: C

Reference:

C.5.1-1100, TS 3.3.A.1, C.5.1-2003 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION NOTE: Per C.5-1100 Part B 1) assurance that the Rx will remain shutdown under all condition is obtained by verifying all control rods are inserted to or beyond position 04 2)

If one or more rods are not inserted to position 04 then may confirm RX will remain shutdown by satisfying TS 3.3.A.1 criteria or rod pattern analysis by the nuclear engineers. With one rod at position 30, the criteria of TS 3.3.A.1 would be satisfied (i.e.,

core loading is such that Rx can be made subcritical with the strongest operable control rod in its full-out position and all other control rods fully inserted.

A. Minimum Steam Cooling RPV Water Level applies to actions taken in C.5-2007 and this procedure has not been entered.

B. Minimum Steam Cooling RPV Water Level applies to actions taken in C.5-2007 and this procedure has not been entered.

C. This is correct.

D. The Minimum Zero-Injection RPV Water Level is not -149 inches.

COGNITIVE LEVEL JUSTIFICATION S!C WiMIi H IZZAMHA¶]I©M HIRC Review CDPy Page 91 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Level 3 since the student is required to analyze the conditions to determine what procedure will direct opening SRVs and then needs to have a knowledge of EOP bases to determine why.

SRO ONLY JUSTIFICATION This is SRO only knowledge since it is based on information from the EOP bases.

HR(*C *JITTIMH Ai]

HIRC RIevw Cpy Page 92 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 76 Exam: SRO System: GENERIC K/A: 2.4.38 Lesson Plan M-7406L-002 Enabling Objective 3,5,7,8 Which of the following is not a responsibility of the acting Emergency Director during an emergency?

A. The final decision to issue potassium-iodide to workers.

B. Declaration and notification of emergency classifications.

C. The authorization of employees to receive radiation exposures greater than I OCFR20 limits.

D. Implementing actions recommended by the Severe Accident Management Guidelines Group.

Answer: D

Reference:

A.2-001 Question Pedigree: INPO Bank QID Cog. Level: 1

1. 6408 ANSWER/DISTRACTER JUSTIFICATION A. See Emergency Director duties and responsibilities A.2-001 page 5.

B. See Emergency Director duties and responsibilities A.2-001 page 5.

C. See Emergency Director duties and responsibilities A.2-001 page 5.

D. See SAMG Group duties and responsibilities A.2-001 page 39-40.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is an administrative procedure step.

SRO Level Justification SRO level knowledge bases on examinee must recall responsibilities of the Emergency Director which is a position filled by an SRO and meeting 10 CFR 55.43(b)(5).

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 77 Exam: BOTH System: GENERIC K/A: 2.4.50 Lesson Plan M-8107L-048 Enablina Obiective Given the following conditions:

"* The plant is operating at 100% power.

"* A failure of the Steam Pressure Control System occurs causing Turbine Control Valves to fully open.

"* A large increase in Main Generator MWe is noticed.

"* Annunciator 5-B-16, REACTOR PRESS HI/LO, is alarming.

Which of the following actions should be taken for the conditions stated above?

A. Manually scram the reactor.

B. Commence a normal reactor shutdown.

C. Push the EPR STOP pushbutton on the C-07 panel.

D. Restore reactor pressure using the Pressure Regulator Override (PRO).

Answer: A

Reference:

C.4-B.05.09.A Question Pedigree: New Cog. Level: 3 ANSWERJDISTRACTER JUSTIFICATION

>" A. Per C.4-B.05.09.A if reactor pressure and MPR setpoints differ by >15 psi from the original setpoints than a manual reactor scram is to be inserted. This is done to prevent an automatic scram from occurring when the MSIVs close on the PCIS Group 1 isolation at 840 psig with the Mode switch in RUN.

B. See answer A.

C. See answer A.

D. See answer A.

COGNITIVE LEVEL JUSTIFICATION Level 3 since the student needs to recognize what failure has occurred, how that failure has caused reactor pressure to lower to the alarm setpoint, and what the procedure reference requires for this condition.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 78 Exam: BOTH System: 295001 K/A: AK3.06 Lesson Plan M-8107L-029 Enabling Objective 9.e, 9.f To perform repairs to the No. 11 Recirc MG Set collector ring, the plant has entered single loop operations with No. 12 Recirculation pump in operation.

The operator is directed by procedure B.01.04-05.E.3, SINGLE LOOP OPERATION, to verify idle Recirculation loop flow by increasing the running Recirculation pump's speed.

What indication confirms idle Recirculation loop reverse flow AND why?

A. A DECREASE in Total Core Flow indicates idle loop reverse flow because the Total Core Flow summing logic can differentiate loop flow direction based on the status of the Recirc MG Set field breaker position.

B. A DECREASE in idle loop Jet Pump Flow indicates idle loop reverse flow because the core flow measurement instrumentation cannot differentiate between reverse flow and forward flow.

C. An INCREASE in Total Core Flow indicates idle loop reverse flow because the Total Core Flow summing logic can differentiate loop flow direction based on the status of the Recirc MG Set Field breaker position.

D. An INCREASE in idle loop Jet Pump Flow indicates idle loop reverse flow because the core flow measurement instrumentation cannot differentiate between reverse flow and forward flow.

Answer: D

Reference:

B.01.04-05.E.3, B.01.04-05.A.3.b.2)g)

Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Plausible answer as some plant designs do use two different summing loops for total core flow; however Monticello does not have this design feature. An increase in total core flow does not give positive indication of idle loop reverse flow.

B. Plausible answer, based on jet pumps due share the high pressure connection for 4 of the 20 jet pumps. However, decrease in idle loop jet pump flow with increase in running loop recirc speed would indicate forward flow in the idle loop.

C. Plausible answer as some plant designs do use two different summing loops for total core flow; however Monticello does not have this design feature. An increase in total core flow does not give positive indication of idle loop reverse flow.

D. Per B.01.04-05.E.3 & B.01.04-05.A.3.b.2)g) idle loop reverse flow is indicated by increase in idle loop jet pump flow as running recirc speed increases because core flow measurement system cannot differentiate forward or reverse jet pump flow.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recall system procedure steps and cautions.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 79 E xam: BOTH System: 295002 K/A: AAI.01 Lesson Plan M-8107L-01 1 Enabling Objective 9.h What temperature should the Condensate pump suction be maintained below in the summer, AND what would be the concern if you are unable to maintain it below that temperature?

A. <100°F; degraded performance of the Steam Jet Air Ejectors, which could lead to loss of Main Condenser vacuum.

B. <100°F; reduction in condensate pump suction pressure which effects Feedwater pump suction pressure and possible loss of Reactor water level.

C. <1 25 0 F; degraded performance of the Steam Jet Air Ejectors, which could lead to loss of Main Condenser vacuum.

D. <1250 F; reduction in condensate pump suction pressure which effects Feedwater pump suction pressure and possible loss of Reactor water level.

Answer: C

Reference:

B.06.03-05, C.4-B.06.03.A Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. <1 00°F is for winter time.

B. 100°F would not substantially affect the condensate pump suction pressure.

C. Correct answer "D. 125 0 F would not substantially affect the condensate pump suction pressure.

COGNITIVE LEVEL JUSTIFICATION Level 2 - Requires the candidate to recall information in the SJAE procedure dealing with SJAE problems and recognize the interaction between the SJAE and the main condenser.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 80 E)cam: BOTH System: 295003 K/A: AA1.01 Lesson Plan M-8114L-003 Enablinq Obiective 2.f Upon a loss of Bus 11, Abnormal Procedure C.4-B.09.06.A, LOSS OF BUS 11 OR BUS 12, directs the operator to place the control switch for the tripped Feedwater pump in the STOP position.

Why is this action directed?

A. Resets the low Feedwater flow automatic runback logic for the Reactor Recirculation pumps.

B. Enables the auto start logic for the Feedwater pump Aux Oil Pump to start on a restoration of Bus 11.

C. Allows troubleshooting Bus 11 to determine if the fault is in the Feedwater pump breaker or elsewhere.

D. Prevents an uncontrolled start of the tripped Feedwater pump if power should be restored to Bus 11.

Answer: D

Reference:

C.4-B.09.06.A Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. See correct answer B. See correct answer C. See correct answer D. The 4160 volt breakers for the Reactor Feedwater Pumps do not trip from undervoltage.

If all other starting interlocks are satisfied and the de-energized bus is subsequently energized, the Feedwater Pump will restart if the breaker is still closed.

COGNITIVE LEVEL JUSTIFICATION Level 1 - requires the candidate to recall reason for performing a step in a procedure.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 81 E)tam: BOTH System: 295003 K/A: AK2.02 Lesson Plan M-8107L-093 Enabling Objective 3.a & 11.c The plant is in normal full power operation with no LCOs, when massive grid instabilities result in a loss of offsite power for the foreseeable future. The plant responds as designed with both Emergency Diesel Generators (EDGs) running and loaded.

Diesel fuel oil delivery is uncertain due to infrastructure problems.

Assume the following in your answer:

"* The Diesel Oil Storage Tank, T-44, level is 40, 500 gallons.

"* The Diesel Oil Receiving Tank, T-83, level is at 9, 000 gallons.

"* The Heating Boiler Oil Storage Tank, T-84, level is 19, 500 gallons.

"* Power is available for necessary fuel oil transfer equipment.

Select the answer that CORRECTLY describes how long the EDGs will be able to operate FULLY LOADED, if no more fuel can be delivered.

A. 1 EDG for 7 days.

This is at the design fuel consumption rate for the Diesels and T-44 only; the fuel in T-84 is normally too low quality for transfer to T-44.

B. 2 EDGs for 7 days.

This is at the design fuel consumption rate for the Diesels and T-44 only; the fuel in T-84 is normally too low quality for transfer to T-44.

C. 2 EDGs for 7 days.

This includes an additional 28, 500 gallons of Diesel Oil from T-83 and T-84 with fuel being used at the design fuel consumption rate for the Diesels and T-44.

D. 2 EDGs for 14 days.

This includes an additional 28, 500 gallons of Diesel Oil from T-83 and T-84 with fuel being used at the design fuel consumption rate for the Diesels and T-44.

Answer: A

Reference:

TS 3.9.B.3.c, B.8.11-05.A.3.a, B.8.11-01.B.1, B.8.11 01.C.1 Question Pedigree: MOD INPO QID 649 Cog. Level: 2 ANSWERJDISTRACTER JUSTIFICATION A. Design fuel consumption per TS 3.9.B.3.c is 34,500 for 1 EDG for 7 days at 2500 KW.

Per B.8.11-01.B.1, Tank T-44 for EDGs has a capacity of 60, 000 gal (minimum T-44 tank level for EDG operability is 38, 300 gal), Tank T-83 for Diesel Oil Receiving is 10, 000 and T-84 for Heating Boiler is 20,000. Per B.8.11-01.C.1 EDG fuel oil must meet spec MPS-49. Oil that does not meet spec is sent to the heating boiler storage tank to be burned or offloaded to tanker truck.

B. This is not correct for T-44, and would only be correct if T-84 could be transferred with T-83; however T-84 oil does not meet EDG specifications.

C. This is correct if T-84 could be transferred with T-83; however T-84 oil does not meet EDG specifications.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank D. This is not correct for any combination of tanks.

COGNITIVE LEVEL JUSTIFICATION Level 2 based on examinee must recall TS value for EDG operability for 7 days and solve a problem using the given information using knowledge of its meaning.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 82 E xam: BOTH System: 295006 K/A: AK1.01 Lesson Plan M-8114L-001 The following conditions exist.

"* A Reactor scram has occurred.

"* A Group 3 Isolation has been received due to low-low Reactor water level.

"* The scram has not been reset.

What is the concern with these specific conditions?

A. The Group 3 isolation coupled with the shrink from the cold water addition could make RPV level control difficult.

B. Cold water injected into the Recirc loops could cause thermal clamping of a Recirc pump discharge valve preventing it from being completely closed.

C. A failure to reset the scram will prevent the draining of the Scram Discharge Volume, which could limit the ability to rescram the Reactor if needed.

D. A low temperature over pressurization of the vessel could occur due to loss of decay heat removal capability and injection of cold CRD flow.

Answer: D

Reference:

C.4-A Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Level would normally be controlled by other methods.

B. Thermal clamping could prevent operation of a closed valve.

C. Volume would be sufficient to attempt rescram if needed, but all rods are in.

D. Correct answer.

COGNITIVE LEVEL JUSTIFICATION Level 2 - candidate is required to know that a Group 3 would isolate RWCU and then also have knowledge of the caution in the Scram Off-Normal.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 83 E Kam: BOTH System: 295006 K/A: AK3.03 Lesson Plan M-8114L-001 Assuming the plant responds as designed, which of the following describes the immediate Reactor pressure response to a manual scram from rated conditions AND the reason?

Indicated Reactor pressure will...

A. raise due to delay in Bypass Valves opening.

B. raise due to rapid closure of the Turbine Control Valves.

C. lower due to Recirc pump runback to 30%.

D. lower due to collapsing voids in the core region.

Answer: D

Reference:

C.4-A Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Pressure would drop and Bypass Valves would not have a delay in them.

B. Pressure would drop and Turbine Control Valves would not rapidly close.

C. Recirc pump runback would occur after feedwater flow has dropped to below 20% for 15 seconds.

D. Correct answer COGNITIVE LEVEL JUSTIFICATION Level 2 - requires the candidate to know the direction the pressure will go and the reason.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 84 Exam: SRO System: 295008 K/A: 2.1.6 Lesson Plan M-8114L-005 Enabling Objective 4.b, 4.c A Reactor scram has occurred and an ATWS condition exists. The following are the current plant conditions.

"* Reactor power is 7%.

"* A Group 1 Isolation has occurred.

"* Torus temperature is 105 0 F and slowly increasing.

"* Drywell pressure is 0.8 psig and slowly increasing.

"* Reactor water level is minus 51 (-51) inches.

"* Standby Liquid Control has been started.

"* SRVs are controlling Reactor pressure.

What would be the level band that you would direct the operator to maintain, AND what is the basis for the bands upper setpoint?

A. Minus 126 (-126) to 48 inches, suppresses Reactor power to the lowest practical level while still ensuring that the core is adequately cooled.

B. Minus 126 (-126) to 48 inches, preserves the availability of Feedwater, HPCI, and RCIC and avoids moisture carryover and loss of boron into Main Steam lines.

C. Minus 149 (-149) to minus 51 (-51) inches, suppresses Reactor power to the lowest practical level while still ensuring that the core is adequately cooled.

D. Minus 149 (-149) to minus 51 (-51) inches, preserves the availability of Feedwater, HPCI, and RCIC and avoids moisture carryover and loss of boron into Main Steam lines.

Answer: B

Reference:

C.5.1-2007 Question Pedigree: NEW Cog. Level: 2 ANSWERIDISTRACTER JUSTIFICATION A. The band is correct but the basis is for the IF-AND-AND-AND leg.

B. Correct answer C. Band and basis are for the IF-AND-AND-AND leg.

D. Band is for the IF-AND-AND-AND leg and the basis is for the other leg.

COGNITIVE LEVEL JUSTIFICATION Level 2 - requires the candidate to determine the correct level band from the information given and to have knowledge of the basis for the upper setpoint of that band.

SRO Level Justification SRO knowledge level based on examinee must recall EOP bases information.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 85 E) cam: BOTH System: 295008 K/A: AKI.03 Lesson Plan M-81 07L-046 Q Al The plant is in steady state operation at 100% power. A Feedwater flow transmitter slowly starts failing low.

How would reactor water level respond and what would the Reactor Level Control (RLC) system do?

Water level would...

A. decrease until there is a large enough mismatch in feed flow signals that causes the RLC system to shift from three-element to single-element and stabilize level.

B. decrease until level reaches 30 inches at which time the Master Level Controller clamps the signal at 30 inches and stabilizes level.

C. increase until there is a large enough mismatch in feed flow signals that causes the RLC system to shift from three-element to single-element and stabilize level.

D. increase until level reaches 40 inches at which time the Master Level Controller clamps the signal at 40 inches and stabilizes level.

Answer: C

Reference:

B.05.07-02 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. See correct answer.

B. See correct answer and Master Level Controller is still trying to control level at original setpoint (~35 inches).

C. Sensed feed flow in decreasing, so RLC system will increase flow. Feed flow will then be higher than steam flow and reactor water level will increase until the mismatch between the good transmitter and the failing transmitter is large enough (570, 000 lb/hr) which will cause an automatic shift to single-element control. This should prevent level from increasing much more because feed flow and steam flow are not used to control level in single-element.

D. Master level controller is still trying to control level at original setpoint (-35 inches).

COGNITIVE LEVEL JUSTIFICATION Level 2 - requires the candidate to know 1) how level will respond under the given circumstances and 2) what the response of the RLC system will be.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 86 Exam: SRO System: 295010 K/A: 2.4.8 Lesson Plan M-8114L-005 Enablina Obieotivw The plant has experienced a transient causing the following conditions in the Primary Containment.

"* Drywell pressure is 2.5 psig.

"* Reactor pressure is 830 psig Given the above conditions, C.5-1 100, RPV LEVEL CONTROL, directs that before RPV pressure drops to 320 psig, prevent Core Spray and LPCI injection not needed for core cooling per C.5-3205.

What is the basis for this action?

A. Reduces the thermal stress on the Reactor vessel through cold water injection.

B. If not needed for core cooling, injection is prevented to facilitate RPV pressure control.

C. If not needed for core cooling, injection is prevented to facilitate RPV water level control.

D. Minimizes the reduction in NPSH of the ECCS pumps needed for core cooling due to Torus water level decrease.

Answer: C

Reference:

C.5.1-1100 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. See correct answer B. See correct answer C. Per the bases in C.5.1-1 100: LPCI and Core Spray would inject into the vessel at a Reactor pressure of approximately 320 psig. If these systems are not needed for core cooling they may be prevented to facilitate RPV water level control.

D. See correct answer COGNITIVE LEVEL JUSTIFICATION Level 1 - requires candidate to recall information on the bases of this procedure.

SRO Level Justification SRO knowledge level based on examinee requires knowledge of the EOP Bases.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 87 E xam: BOTH System: 295012 K/A: AK2.02 Lesson Plan M-8107L-044 Enablina Obiectiw The Reactor was operating at 100% power when a transient occurred causing a Main Turbine trip and Reactor scram. In accordance with C.4-A, REACTOR SCRAM, Part B, Step 8 the operator places the POST SCRAM switch on the C-25 Panel to ON.

Which of the following describes the reason for performing this step?

A. This provides maximum cooling to the general areas of the Primary Containment since these areas become very hot post scram.

B. This provides maximum cooling to the Primary Containment CRD under vessel area since this area becomes very hot post scram.

C. This causes the Drywell coolers to receive maximum cooling flow from RBCCW by fully opening flow control valves due to extra heat loads post scram.

D. This causes Reactor Recirc pumps to receive maximum cooling flow from RBCCW by fully opening flow control valves due to extra heat loads post scram.

Answer: B

Reference:

B.8.16-01 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. See answer B.

B. The post scram mode will open vent dampers to provide maximum cooling to the CRD under vessel area and close dampers to general areas of the drywell since after a scram condition the CRD under vessel areas can heat up in excess of 150 0 F.

C. There is no flow control valve for RBCCW to the drywell coolers.

D. There is no flow control valve for RBCCW to the reactor recirc pumps.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a fact from the Ops Manual.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 88 Exam: SRO System: 295013 K/A: AA2.01 Lesson Plan M-8107L-0, 11 h 11r In response to a leaking Safety Relief Valve (SRV) Enablina a plant0biectivA shutdown has been1hinitiated 1r and all available Torus cooling has been placed in service. Due to high river water temperatures, the Torus Temperature has continued to increase and is currently 92 0 F.

The Reactor is currently holding at 75% power while engineering evaluates the improved trend in SRV discharge line temperatures, with SRV discharge line temperature at 210 0 F and steady.

What action should be taken to maintain compliance with plant Technical Specifications, AND what is the basis for this action?

A. Terminate the plant shutdown and monitor Torus temperature every five minutes because excessive steam condensing loads are avoided for Torus temperatures less than 160 0 F.

B. Continue with the plant shutdown because the Torus temperature is greater than the initial conditions assumed in the LOCA analysis for confirming adequate ECCS pump net positive suction head.

C. Terminate the plant shutdown and monitor Torus temperature every five minutes to ensure Torus temperature is less than the initial conditions assumed in the LOCA analysis for confirming adequate ECCS pump net positive suction head.

D. Continue with the plant shutdown because excessive steam condensing loads may occur if a spurious lift of the leaking Safety Relief Valve were to occur.

Answer: B

Reference:

TS 3.7.A.1.a, TS 3.7.A.1.b, and TS bases page 176, Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Per Tech Spec 4.7.A.1 .b, Torus temp must be logged every 5 minutes. SPOTMOS will record temperature every minute. Action to terminate the shutdown is inappropriate and is inconsistent with plant Tech Specs. Tech Spec bases information is plausible as TS bases page 176 states that excessive steam condensing loads can be avoided if the peak suppression pool temp is maintained below 160°F during any period of relief valve operation. This bases support requirements for Rx Scram if Torus temp exceeds 110°F.

B. Per TS 3.7.A. 1 .a, water temp during normal operating shall be less than or equal to 90 0 F, bases is that 90°F Torus temp was an initial condition in the ECCS LOCA analysis to demonstrate adequate NPSH.

C. Tech Spec allow this condition for addition of heat to the Torus due to testing, which does not apply for the leaking SRV, thus the answer is inconsistent with TS. Bases information is correct concerning 90°F limit is based on ECCS pump NPSH.

D. See correct answer. Bases information provided is not consistent with 90°F Torus temperature.

SRO ONLY JUSTIFICATION SRO only based on examinee must recall Tech Spec Bases for Torus Temp.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank COGNITIVE LEVEL JUSTIFICATION Level 2 because examinee must 1) recognize Torus temp is above TS limit 2) relationship between Torus temp and leaking SRV, and 3) recognize consequences and implications of Torus temp > 90°F when heat is being added from non-test conditions.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 89 Ex :am: BOTH System: 295013 K/A: AK2.01 Lesson Plan M-81 07L-044 Ennhlinn r)ahiniz Whenever Reactor Core Isolation Cooling (RCIC) is to be operated for surveillance testing, Torus cooling is required to be in service as a prerequisite to starting RCIC.

Ensuring that Torus cooling is in service before operating RCIC...

A. assures that the bulk water temperature in the Torus does not exceed Tech Spec limits.

B. allows the maximum average suppression pool water temperature limit to be increased to 1050 F.

C. allows adequate thermal mixing of the water in the suppression pool to limit stress on the Torus shell due to differential thermal expansion.

D. assures that heat added to the suppression pool does not increase Torus air space pressure to the point where the Torus to Drywell vacuum breakers cycle.

Answer: A

Reference:

Test 0255-08-IA-1 Question Pedigree: INPO Bank QID 8883 Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Correct Answer B. Normal limit is increased from 90°F to 100OF when testing is in progress.

C. Torus is designed to take stress of RCIC starting without having Torus cooling in service.

D. The heat added would not increase Torus air space pressure.

COGNITIVE LEVEL JUSTIFICATION Level 1 - requires the candidate to recall the reason for starting Torus cooling.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 91 Exam: BOTH System: 295014 K/A: AK1.01 Lesson Plan M-8108L-031 Enabling Objective 3 M-81 08L-036 1 The Chernobyl Unit 4 event resulted in severe core damage with extensive off-site radioactive releases. A cause of this event was that the Reactor was operated in an unanalyzed configuration, which resulted in a prompt criticality excursion.

Which of the following describes mechanisms that are in place to prevent a Chernobyl like prompt criticality event from occurring at the Monticello Nuclear Generating Plant?

A. Detailed safety reviews are required for tests or experiments; and infrequent tests and evolutions require procedure validation and review, dedicated management oversight, pre-job briefing, and specific guidance for when and how to abort the test.

B. The Anticipated Transient Without Scram (ATWS) and Standby Liquid Control systems are designed to mitigate the consequences of a failure of the primary Reactor shutdown mechanisms.

C. Symptom based emergency procedures, installed post-accident monitoring instrumentation, control room panel human factor design, and emergency response facilities and staffing are required by the Monticello operating license.

D. The Reactor Protection System, in conjunction with the Primary Containment Isolation systems, is designed to ensure radioactive releases do not exceed the guidelines of 10 CFR Part 100.

Answer: A

Reference:

Lesson plan M-8108L-031 section Ill.C, 4AWI 02.03.03, sect 4.14, 4AWI-05.06.02, sect 2.0 & 4.6 Question Pedigree: Mod ILT Bank QID Cog. Level: 2 M-8108L-036-006 ANSWER/DISTRACTER JUSTIFICATION A. Lesson Plan M-8108L-031 section III.C identifies the lessons learned from the Chernobyl event and MNGP controls in place to mitigate such events. Mechanisms in place to address these lesson learned are contained in 4AWI-05.06.02 concerning safety reviews, 4 AWI-04.05.07 section 4.14 concerning infrequent tests and evolutions, and 4AWI-02.03.08 concerning special procedures.

B. Plausible answer because the ATWS and SBLC system do provide alternate mechanisms to control reactivity; however this system due to prevent prompt criticality events.

C. Plausible answer as these are items were put in place to address the TMI event; however these items to not prevent operating the plant in an unanalyzed configuration leading to a potential prompt criticality event.

D. Plausible answer because the RPS and PCIS do provide mechanisms to mitigated the consequences of postulated accidents and transients; however the system do not prevent prompt criticality events or mitigated all events if the plant is operated in an unanalyzed configuration. For example, while the RPS and PCIS will mitigate the consequences of a control rod drop accident, prevention of a control rod drop accident HI C RIvwCopy Page 109 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank is provided by performing coupling checks and operating the plant consistent with the plant license.

COGNITIVE LEVEL JUSTIFICATION Level 2 based on examinee must recognize similarities between MNGP and Chernobyl plant operations as well as differences to prevent occurrences at MNGP.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 92 E (am: BOTH System: 295015 K/A: AK3.01 Lesson Plan M-8107L-001 Enabling Objective 4.b, 6.b M-8114L-01 1 9 r, A scram signal has been received, but not all the rods have inserted. Reactor power is 15%.

C.5-2007, FAILURE TO SCRAM, calls for inserting Control Rods per C.5-3101, ALTERNATE ROD INSERTION.

Which of the parts in C.5-3101 require you to bypass the Rod Worth Minimizer (RWM), AND why must it be bypassed?

A. De-energize Scram Logic, rod positions may not match rod sequence in the RWM B. Increase Cooling Water DP and use RMCS, rod positions may not match rod sequence in the RWM.

C. De-energize Scram Logic, position indication from the Rod Position Information System is lost during a scram.

D. Increase Cooling Water DP and use RMCS, position indication from the Rod Position Information System is lost during a scram.

Answer: B

Reference:

C.5-3101 Question Pedigree: Clinton 2001 NRC Cog. Level: 2 Exam Modified ANSWER/DISTRACTER JUSTIFICATION A. Does not require the RWM to be bypassed.

B. Correct answer C. Does not require the RWM to be bypassed and RPIS position indication is not lost during a scram.

D. RPIS position indication is not lost during a scram.

COGNITIVE LEVEL JUSTIFICATION Level 2 - Candidate is required to know 1) which part of the procedure requires bypassing the RWM and 2) why it must be bypassed. This is 2 discrete pieces of information that must be put together to answer the question.

HIRC EIr¶TIE I ATh©H

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 93 Exam: RO System: 295017 K/A: AK1.02 Lesson Plan M-81 14L-0' IPnnhlinn Irhiar-fi%,rg 1)-

M-81 -

A plant transient has occurred which has resulted in fuel element failure. Drywell pressure has increased to the point that Primary Containment venting is required.

Which of the following methods of venting would be expected to have the highest radioactive release to the public?

A. Hard Pipe Vent B. 2 inch line from Torus.

C. 2 inch line from Drywell.

D. 18 inch line from Drywell.

Answer: A

Reference:

C.5-3505 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Venting through the Hard Pipe Vent results in an unfiltered, ground level release.

B. Vents through the SBGT system to the main stack.

C. Vents through the SBGT system to the main stack.

D. Vents through the SBGT system to the main stack.

COGNITIVE LEVEL JUSTIFICATION Level 2 - Requires the candidate to understand the differences between the different methods of venting to determine which would result in the highest release.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 94 Exam: BOTH System: 295017 K/A: AK3.01 Lesson Plan M-81 07L-077 E~nabling Ohiar~hio 7,m 7 k .Q.7 1 A fuel failure is suspected with the plant operating at 100% power. The following conditions exist:

"* Stack Gas Rad Monitors A, and B indicate 105,000 gc/sec

"* Reactor Building Ventilation (RBV) Exh Plenum Rad Monitors A, and B indicate 23 mr/hr What is the expected plant response and the reason?

A. RBV isolation with Standby Gas Treatment System initiation due to Stack Gas Rad Monitor signals.

B. Off-Gas Line and Compressed Gas Storage Line to Stack isolates due to the RBV Exh Plenum Rad C. RBV isolationMonitor signals.

with Standby Gas Treatment System initiation due to the RBV Exh Plenum Rad Monitor signals.

D. Off-Gas Line and Compressed Gas Storage Line to Stack isolates due to Stack Gas Rad Monitor sioqnals.

Answer: D

Reference:

B.05.11-01 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Do not get the isolation of RBV and initiation of SBGT on either of these signals.

B. Off-Gas Line and Compressed Gas Storage Line do isolate but not due to the RBV Exh Plenum Rad Monitor Signals.

C. Do not get the isolation of RBV and initiation of SBGT on either of these signals.

D. Correct answer - setpoint to isolate is 90,000 p.c/sec.

COGNITIVE LEVEL JUSTIFICATION Level 1 - requires the candidate to recall the trip setpoints and which component isolates which would be taught at the same time or would be considered one fact.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 95 Exam: SRO System: 295018 K/A: AA2.01 Lesson Plan M-8107L-016 Enabling Objective 4.a, 11.d The plant was operating at 100% power when a Loss of Off-Site power occurred. The current conditions are:

"* The No. 12 Standby Diesel Generator started and tied onto Bus 16.

"* The No. 11 Standby Diesel Generator started but failed to tie onto Bus 15.

Alarms received associated with the No. 11 Emergency Diesel Generator start included:

"* 93-A-19, RAW WATER

"* 93-A-1 3, HOT ENGINE

"* 8-B-30, DIESEL ENG TROUBLE

"* 8-B-19, ESW PUMP 11 LO DSCH PRESS

1) What is the cause of the above alarms, AND 2) given the current plant conditions, is the objective met for an adequate supply of power for operation of engineered safeguards equipment per the Technical Specification Bases?

A. Loss of power to ESW pump 11 only, No B. Loss of power to ESW Pump 11 and SW Pump 11, No C. Loss of power to ESW pump 11 only, Yes D. Loss of power to ESW Pump 11 and SW Pump 11, Yes Answer: D

Reference:

NF-36298-1, Tech Spec Bases 3.9 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Both the ESW and SW pumps are lost due to Bus 15 being without power, According to Tech Spec Bases only one source of power is required and that is being supplied by No. 12 EDG. So the answer is Yes.

B. According to Tech Spec Bases only one source of power is required and that is being supplied by No. 12 EDG. So the answer is Yes.

C. Both the ESW and SW pumps are lost due to Bus 15 being without power.

D. Correct Answer.

SRO Level Justification SRO level knowledge based on examinee must recall plant Tech Spec bases information concerning power supplies for ECCS equipment thus meeting 10 CFR 50.43(b)(2).

COGNITIVE LEVEL JUSTIFICATION Level 2 - Requires the candidate to recognize what power supplies are lost to discern what cooling water systems have been lost. Also requires knowledge of Tech Spec bases and how it relates to current conditions.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 96 Exam: SRO System: 295021 K/A: AA2.05 Lesson Plan M-8108L-003 Enablino Obiective The Reactor is currently in Cold Shutdown with the Reactor Vessel head still tensioned.

Normal shutdown cooling has been lost. Other means of shutdown cooling have been unsuccessful and it is decided to establish a cooling flow path through an SRV to the Torus.

This procedure requires the monitoring of the reactor vessel metal temperatures.

What is the minimum Technical Specification temperature for the Reactor Vessel metal temperatures, AND what is this based upon?

A. 70°F, Shell to Flange AT (at greatest stress)

B. 70°F, Nil Ductility Temperature + 60°F C. 78 0 F, Shell to Flange AT (at greatest stress)

D. 78 0 F, Nil Ductility Temperature + 60 0 F Answer: B

Reference:

Tech Spec Bases 3.6/4.6 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION

-. A. Temperature is 70 0 F, but it is not based upon the shell to flange AT.

B. Correct answer.

C. Temperature is not 780 F, and it is not based upon the shell to flange AT.

D. Temperature is not 78 0 F.

COGNITIVE LEVEL JUSTIFICATION Level 2 - requires the candidate to know both the minimum temperature and the basis for that temperature.

SRO Only Justification - Requires candidate to have knowledge of the bases of Tech Specs which is required knowledge for SROs only.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 96.1 Exam: RO System: 295021 K/A: AK3.01 Lesson Plan M-81114-002 Enablina ObiectivA 9_ 9f of Following~~nabin lossv a nomlsudwncoigtefprtri Following a loss of normal shutdown cooling the operator is directed to monitor and control RPV water level from 55 inches to 80 inches.

Maintaining the Reactor water level at a high level provides...

A. the required net positive suction head for restarting the RHR pump.

B. an adequate margin to the low water level shutdown cooling isolation setpoint.

C. a large volume of relatively cool water to control the Reactor temperature increase.

D. an adequate margin to establish a natural circulation path through the Reactor core.

Answer: D

Reference:

C.4-B.3.4.A Question Pedigree: LaSalle 1995 NRC Cog. Level: 1 Exam ANSWER/DISTRACTER JUSTIFICATION A. Adequate NPSH would be provided even at normal levels.

B. An adequate margin is maintained even at normal levels. Isolation setpoint is +9 in.

C. No credit is taken for the introduction of this colder water.

D. Correct answer - with level less than -46 inches the steam separators effectively block the flow of water from inside the shroud to the annulus area. Level must be above the separator turnaround area which is at ~46 inches. Maintaining level above +55 inches maintains adequate margin to establish a natural circulation flow.

COGNITIVE LEVEL JUSTIFICATION Level 1 - requires the candidate to recall the reason for performing a step in the procedure.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 97 Exam: BOTH System: 295022 K/A: AK2.02 Lesson Plan M-8114L-0( 02 Enablina Obiective An earthquake has taken place resulting in a Loss of Coolant Accident (LOCA). Current conditions are:

"* Reactor water level is minus 180 (-180) inches and decreasing.

"* Containment High Range Radiation Monitors are beginning to come on-scale.

"* Annunciator 5-B-17, CHARGING WATER LOW PRESSURE, is in alarm.

"* All off-site power has been lost.

Concerning the CRD system, what action needs to be taken?

A. Start a Condensate Pump.

B. Start a Condensate Service Pump.

C. Close the selected CRD Flow Control Valve.

D. Close CRD-10-1 and CRD-10-2, 11 and 12 CRD Post Filter Inlets.

Answer: D

Reference:

C.4-B.01.03.A Question Pedigree: NEW Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. In accordance with LOSS OF CRD PUMP FLOW, IF a LPCI initiation signal AND evidence of significant fuel damage are present, THEN START a Condensate Pump.

This could not be accomplished because with a loss of off-site power there is no power to the pumps.

B. Starting a Condensate Service Pump is not an action lAW LOSS OF CRD PUMP FLOW.

C. Closing the CRD Flow Control Valve would not fully isolate the CRD pump header D. Correct answer - This action is taken in response to NRC Information Notice 90-078 which describes a potential post-LOCA bypass leakage pathway from the reactor coolant system past the CRD seals and valves in the HCU back to the CRD pump header and out to the CSTs.

COGNITIVE LEVEL JUSTIFICATION Level 3 - Candidate would be required to assemble and integrate multiple pieces of information to make the correct determination of closing the isolation valves.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 98 Exam: SRO System: 295023 K/A: 2.2.22 Lesson Plan M-8107L-0 Enabling Objective 9.d In accordance with the refueling Limiting Conditions for Operations:

More than one control rod may be withdrawn from the reactor core during outages provided that, except for momentary switching to the Startup mode for interlock testing, the Reactor mode switch shall be locked in the Refuel position.

Which of the following states how more than one Control Rod may be removed from the core, AND what will prevent the loading of fuel into the core once the control rods are removed?

A. 1) Refueling interlock from the Control Rod is bypassed.

2) Only administrative controls are in place to prevent loading fuel.

B. 1) Rod Worth Minimizer is bypassed.

2) Interlock from other Control Rod prevents moving the bridge over the core with fuel on the hoist.

C. 1) Refueling interlock from the Control Rod is bypassed.

2) Interlock from other Control Rod prevents moving the bridge over the core with fuel on the hoist.

D. 1) Rod Worth Minimizer is bypassed.

2) Only administrative controls are in place to prevent loading fuel.

Answer: A

Reference:

Tech Spec 3.10 and Bases Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION Page 208 of Tech Specs needs to be removed to prevent a direct look-up of the first part of the question.

A. Correct answer.

B. Bypassing the RWM would not matter while in refueling. The interlocks from the other control rods are either bypassed also or the Refuel position allows 1 rod to be withdrawn.

C. The interlocks from the other control rods are either bypassed also or the Refuel position allows 1 rod to be withdrawn.

D. Bypassing the RWM would not matter while in refueling.

SRO Level Justification SRO level knowledge based on examinee must recall plant Tech Spec bases information concerning power supplies for ECCS equipment thus meeting 10 CFR 50.43(b)(2).

COGNITIVE LEVEL JUSTIFICATION Level 2 - Requires the candidate to know what would allow the rod to be withdrawn and to know what controls are in place to prevent fuel loading. This is 2 separate pieces of information.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 99 Exam: BOTH System: 295023 K/A: AAI.02 Lesson Plan M-8107L-022 Enablina Obiective 9qi The plant is shutdown for a refueling outage with the Reactor Cavity flooded.

Failure of the RPV-to-Drywell Refueling Cavity Bellows could first be identified by which of the following indications?

Fuel Pool Cooling Drywell Floor Drain Skimmer Surge Tank Level Sump Level A. Lowering Rising B. Lowering Lowering C. Rising Rising D. Rising Lowering Answer: A

Reference:

B.02.01-05.H.7 Question Pedigree: Clinton 2001 NRC Cog. Level: 2 NEW Exam QID 85 ANSWER/DISTRACTER JUSTIFICATION A. Correct Answer B. Drywell Floor Drain Sump would be rising due to leakage in the Drywell entering the sump.

C. Surge Tank would be lowering due loss of inventory in the reactor cavity which the Fuel Pool Cooling system is lined up to.

D. Surge Tank would be lowering due loss of inventory in the reactor cavity which the Fuel Pool Cooling system is lined up to.

COGNITIVE LEVEL JUSTIFICATION Level 2 since the student needs to recognize the interrelations between systems the their response to determine the correct answer.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 100 Exam: BRO System: 295024 K/A: EA2.02 Lesson Plan M-8114L-005 A h A major plant transient has occurred causing the following plant conditions.

"* Drywell pressure is 13 psig and rising slowly.

"* Drywell temperature is 270°F and rising slowly.

"* Torus water level is 2.7 feet.

Based on the information above, which of the following would be the next action to take?

A. Initiate Torus Cooling.

B. Initiate a Blowdown.

C. Initiate Drywell Sprays.

D. Initiate Primary Containment Vent.

Answer: C

Reference:

C.5-1200 Question Pedigree: NEW Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Torus Cooling is not an action taken for high Drywell pressure.

B. Blowdown would not be initiated yet, you are well within Fig 0, Pressure Suppression Pressure.

C. Correct answer - per C.5-1200 Drywell pressure leg, the current Drywell pressure is above 12 psig., pressure and temperature falls within Fig. N, and Torus level is below 4.3 psig which allows Drywell Sprays to be used.

D. Venting would not be performed yet because you are well within Fig P, DW Pressure Limit.

COGNITIVE LEVEL JUSTIFICATION Level 3 - Requires candidate to analyze the information and determine that with the given Drywell pressure and temperature you are within Fig. N which will allow Drywell Sprays to be used.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 100.1 Exam: BSRO System: 295024 K/A: EA2.02 Lesson Plan M-8114L-005 Enablina Obiective A major plant transient has occurred causing the following plant conditions:

• Drywell pressure is 13 psig and rising slowly.

0 Drywell temperature is 270°F and rising slowly.

• Torus water level is 2.7 feet.

Based on the information above, which of the following would be the next action to take, AND what is the basis for taking this action?

A. Initiate Drywell Sprays, to prevent a phenomenon known as "chugging".

B. Initiate Primary Containment Vent, to prevent a phenomenon known as "chugging".

C. Initiate Drywell Sprays, to decrease the percentage of noncondensibles in the Drywell.

D. Initiate Primary Containment Vent, to decrease the percentage of noncondensibles in the Drywell.

Answer: A

Reference:

C.5-1200 Question Pedigree: NEW Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Correct answer - per C.5-1200 Drywell pressure leg, the current Drywell pressure is above 12 psig., pressure and temperature falls within Fig. N, and Torus level is below 4.3 psig which allows Drywell Sprays to be used.

B. Venting would not be performed yet because you are well within Fig P, DW Pressure Limit.

C. Drywell sprays are initiated to increase (not decrease) the percentage of noncondensibles in the Drywell.

D. Venting would not be performed yet because you are well within Fig P, DW Pressure Limit, and this is not the basis for initiating Drywell Sprays or Pri Cntmt vent.

SRO LEVEL JUSTIFICATION Requires the candidate to know the EOP Bases for initiating Drywell Sprays.

COGNITIVE LEVEL JUSTIFICATION Level 3 - Requires candidate to analyze the information and determine that with the given Drywell pressure and temperature you are within Fig. N which will allow Dryweil Sprays to be used.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 101 E (am: BOTH System: 295024 K/A: EK1.01 Lesson Plan M-8107L-044 Enablina Obiectiwe The plant transient has occurred resulting in the following conditions.

"* Drywell pressure is 57 psig and rising.

"* Loss of all decay heat removal capability (beyond the design basis) of the plant.

"* Torus water level is 11.5 feet.

Which system would be used to prevent Primary Containment over pressurization under the above listed conditions?

A. Hard Pipe Vent B. Combustible Gas Control System C. Standby Gas Treatment from Torus D. Standby Gas Treatment from Drywell Answer: A

Reference:

B.04.01-02 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Correct answer B. Does not provide for venting capability. Provides for reduction in oxygen concentration of containment atmosphere.

C. The containment vent from the Torus becomes covered at 11.3 feet and therefore would not be able to be used.

D. The containment vent valves become inoperable at containment pressures above 56 psig.

COGNITIVE LEVEL JUSTIFICATION Level 2 - Requires candidate to understand interrelationships between systems and how the initial plant conditions affect operability of those systems.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 102 Exam: SRO System: 295025 K/A: EA2.02 Lesson Plan M-8107L-007 Which of the following is NOT a valid description of the bases for the RPV high pressure scram setpoint?

A. Assures that RCS pressure does not exceed the range of the fuel cladding integrity safety limit.

B. Provides a backup to the APRM neutron flux scram for main steam line isolation type transients.

C. Works in conjunction with the TCV fast closure and TSV closure scrams and the SRVs to ensure the RCS pressure safety limit is not exceeded.

D. Calculated such that with the design base pressure transient, RCS pressure does not exceed 102% of maximum normal operating pressure.

Answer: D

Reference:

Tech Spec 2.4 Bases Question Pedigree: New Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. See Tech Spec 2.4 Bases.

B. See Tech Spec 2.4 Bases.

"C. See Tech Spec 2.4 Bases.

D. Maximum pressure transient will not exceed 110%.

SRO Level Justification This is SRO level of knowledge because the answer comes from the bases of Tech Specs.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a question of the bases of Tech Specs.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 103 Exam: BOTH System: 295025 K/A: EK1.04 Lesson Plan M-8107L-025 Enabling Objective 7.c The plant had been operating on a 340 day run when a reactor scram occurred. 15 minutes after the scram plant conditions are:

0 A Group I isolation has occurred and has not been reset.

• 6 Control Rods are at position 04, and 3 Control Rods are at position 02 0 Reactor power is 2%.

• RCIC tripped on overspeed and cannot be restarted.

• HPCI is out of service.

• Reactor water is 34 inches and has remained steady.

Why is power still at 2% AND, assuming no operator action, what is the status of the Safety Relief Valves (SRVs)?

A. Multiple rods still out, SRVs controlling pressure in Low Low Set mode.

B. Decay heat generation, SRVs controlling pressure in Low Low Set mode.

C. Multiple rods still out, SRVs controlling pressure in the self action mode.

D. Decay heat generation, SRVs controlling pressure in the self action mode.

Answer: B

Reference:

C.4-A, M-8107L-025 Question Pedigree: NEW Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. If control rods are inserted to or beyond 04 the Reactor will remain shutdown under all conditions. So the reactor will be shutdown, but decay heat is still significant.

B. Correct answer C. If control rods are inserted to or beyond 04 the Reactor will remain shutdown under all conditions. So the reactor will be shutdown, but decay heat is still significant. Low Low Set would initiate and control pressure below the self action mode.

D. Low Low Set would initiate and control pressure below the self action mode.

COGNITIVE LEVEL JUSTIFICATION Level 2 - requires the candidate to know what is creating the power and then to know how pressure is being controlled.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 104 E)(am: BOTH System: 295026 K/A: 2.1.10 Lesson Plan M-8107L-044 In accordance with plant Technical Specifications:

If the Torus water temperature exceeds ,then the Reactor shall be A. 100°F immediately scrammed B. 100°F in cold shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. 110°F immediately scrammed D. 110OF in cold shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Answer: C

Reference:

Tech Spec 3.7.A.l.c Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION Delete TS item 3.7.A.l.c from examinee reference material and indicate the paragraph was intentionally deleted.

A. Correct action, but incorrect temperature.

B. Incorrect temperature and action.

C. Correct answer per Tech Spec D. Correct temperature, but incorrect action.

COGNITIVE LEVEL JUSTIFICATION Level 1 - requires the candidate to recall a single limiting condition for operation from Tech Specs.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 105 Exram: BOTH System: 295026 K/A: EA1.03 Lesson Plan M-81 07L-044 Enablina C bi*.ctiv*. 9 h A small break LOCA has occurred and the operators are taking actions to control plant parameters.

"* HPCI is injecting for RPV level control.

"* Drywell pressure is 4 psig and lowering.

"* Torus pressure is 3 psig and lowering.

"* Loop A and B of RHR are in Torus Spray.

The operator reports that Torus temperature 85 0 F and increasing, when annunciator 5-B-52, TORUS WATER HI TEMP SPOTMOS TROUBLE, alarms. The operator then reports that that the observed SPOTMOS indication reads "ERROR 49".

For the conditions given above the operators should verify Torus temperature using...

A. temperature recorder TR-23-115 on Control Room back panel C-21.

B. steam tables and converting Torus pressure to bulk Torus temperature.

C. the other division of SPOTMOS indication on Control Room Panel C-03.

D. the SPOTMOS averaging unit LED display on Control Room back panel C-21.

"--j Answer: C

Reference:

B.4.1-02.B, C.5.1-1200, C.6-005-B-52 Question Pedigree: NEW Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. Per C.5.1-1200, Torus temperature refers to bulk Torus temperature which is available from SPOTMOS. Points plotted on TR-23-115 are local temperature from only two thermocouples. Lack of Torus mixing via Torus cooling results in Torus local temperatures, not bulk temperature.

B. Use of steam tables would not provide bulk temperature as torus water is not under saturation conditions.

C. Per B.4.1-02.B Torus average bulk temp indicated by two divisions of the Suppression Pool Temperature Monitoring System (SPOTMOS) with indications on Control Room panel C-03. Temperature should be verified using Div 2 SPOTMOS if Div 1 is unavailable.

D. Temperature can be obtained from the SPOTMOS averaging unit; however, the unit is located in the cable spreading room, not the Control Room.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recall system indications, locations, and EOP actions based on bulk torus temp.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 106 E)am: BOTH System: 295028 K/A: EA1.02 Lesson Plan M-8107L-044 Enablinn Ohi*.r.tive_

The plant is operating at 100% power with the following conditions:

"* The Drywell Atmosphere Cooling System is in a normal lineup with fan V-RF-1 in standby.

"* A small steam leak exists in the Primary Containment.

"* The operating crew then inserts a manual Reactor scram.

"* During the scram the following occurs:

"o Lockout of Bus 14 "o Bus 16 powered from 1AR

" The following conditions now exist:

"o Reactor water level is 2 inches and steady.

"o Drywell pressure is 1.7 psig and steady.

"o Drywell temperature is 155 0 F and slowly rising.

Based on the above transient, which of the following describes the current condition of the Drywell Atmosphere Cooling system?

A. All four Drywell fans are running and dampers V-D-32 and V-D-34 (Drywell Ventilation Dampers) are OPEN.

B. One Drywell fan (V-RF-3) is running and dampers V-D-32 and V-D-34 (Drywell Ventilation Dampers) are CLOSED.

C. None of the Drywell fans are running and dampers V-D-32 and V-D-34 (Drywell Ventilation Dampers) are CLOSED.

D. Two Drywell fans (V-RF-1 and V-RF-3) are running and dampers V-D-32 and V-D-34 (Drywell Ventilation Dampers) are OPEN.

Answer: A

Reference:

B.8.16-02 Question Pedigree: New Cog. Level: 1 ANSWERIDISTRACTER JUSTIFICATION A. The reactor signal will cause the Drywell cooling system to automatically shift into the scram mode (mode 5), which means all fans start, V-D-31 and V-D-33 go closed and V D-32 and V-D-34 open. When the lockout occurs on Bus 14, the undervoltage sensed on Bus 16 will cause a bus transfer to 1AR or the EDG and the associated Drywell fans will trip. When Bus 16 is powered from an alternate source the Drywell fans powered from Bus 16 will automatically restart. This results in all Drywell fans running with the dampers in the scram mode.

B. See answer for C.

C. See answer for C.

D. See answer for C.

COGNITIVE LEVEL JUSTIFICATION HRC RITTaH I127AofI15 NIRC Rv*w Cgy Page 127 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Level 1 since this question is asking the interlocks associated with the Drywell atmosphere cooling systemO.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 107 Exam: SRO System: 295029 K/A: EA2.02 Lesson Plan M-8114L-Oi Enablina Objective 4_o The following plant conditions exist.

"* PCIS Group 1 isolation has occurred and the Reactor failed to scram.

"* Reactor water level is being maintained above the top of active fuel.

"* Condensate and Feedwater are the only available high pressure injection source.

"* Reactor pressure is being maintained by Safety Relief Valves.

"* Torus water level is 3.5 feet and rising.

"* C.5-3402, DRAINING TORUS WATER TO RADWASTE, is in progress.

Based on the stated conditions, which of the following describes the expected actions the crew will take AND how those actions affect Reactor power and pressure?

A. Stop injection to the RPV with Feedwater and Condensate to prevent exceeding 4.3 feet Torus water level. This will result in lower Reactor power and pressure due to increased voiding in the core.

B. A blowdown will need to be conducted due to Torus level reaching 4.3 feet. This will cause RPV pressure to drop rapidly and should help to lower Reactor power by increased voiding in the core.

C. A blowdown will need to be conducted due to Torus level reaching 4.3 feet. This will cause RPV pressure to drop rapidly and will result in a power increase due to the increased Feedwater flow into the RPV.

D. Stop injection to the RPV with Feedwater and Condensate to prevent exceeding 4.3 feet Torus water level. This will result in a lower Reactor power due to increased voiding in the core but will cause Reactor pressure to rise due to the loss of inlet subcooling.

Answer: B

Reference:

C.5.1-1200, C.5.1-2007, & C.5.1-2002 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Injection into the RPV cannot be stopped because adequate core cooling will be lost.

B. A blowdown should be conducted before 4.3 feet is reached because above that point the Torus to Drywell vacuum breakers will be covered, which will prevent them from restoring pressure to the Drywell when Drywell pressure becomes less than Torus pressure. During the blowdown reactor water level and pressure will drop resulting in maximum voiding of the core. This will cause reactor power to drop.

C. Reactor power will not rise during the blowdown because injection into the RPV is prevented during the blowdown until the pressure reaches the Min Alt RPV Flooding Pressure.

D. See answer A.

SRO Level Justification SRO knowledge level based on examinee must recall EOP bases information.

COGNITIVE LEVEL JUSTIFICATION Level 3 because the student is required to solve a problem using EOP flow charts and HI C WMITIMEage129ofH 15 HR(C R(Me Cop Page 129 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank recognize what step of C.5-1200 will need to be completed next and how that step will affect reactor power and pressure (relationship between pressure and power during a blowdown in an ATWS).

HIRC W*ITITFH T EKAMEAUI©H HIRC Review CopIy Page 130 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 108 Exam: BOTH System: 295029 K/A: EK1.01 Lesson Plan M-8114L-005 Enabling Objective 3.e Given the following:

"* An unisolable LOCA has occurred.

"* The crew has been unable to keep the core covered.

"* The Severe Accident Management Guidelines have been implemented.

"* The crew is in the process of flooding the Primary Containment to keep the core covered.

Based on the conditions given, which of the following describes the consequence of adding too much water to the Primary Containment?

A. It may cause the loss of all accurate level indications inside the RPV due to covering up of all available indicators inside the Primary Containment.

B. It could result in a structural failure of some components of the Primary Containment as a result of excessive force due to airspace pressure and hydrostatic head of the water.

C. It may cause Drywell pressure to rise to the point where the RPV vent valves become inoperable due to a loss of the d/p necessary for the pneumatic operator to work.

D. It could cause a rapid rise in Primary Containment pressure as the steam/nitrogen bubble becomes compressed in the upper part of the containment when the containment goes solid.

Answer: B

Reference:

C.5.1-1000 Question Pedigree: New Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. Completely flooding the Drywell will not cause loss of level indication inside the RPV however, it may cause a loss of level indication inside the Drywell.

B. This is correct; see attached description from C.5.1-1000.

C. The RPV vent valves at MNGP are not pneumatically operated and do not suffer the problems associated with high Drywell pressure.

D. The containment will structurally fail before being able to reach a solid condition therefore this situation is not plausible.

COGNITIVE LEVEL JUSTIFICATION Level 2 since this question is requires the student to understand the implications of adding too much water to the Drywell and its effects on containment integrity in relation to the Drywell Water Level Limit.

NRt C WKIl IE ft H II ATmI© HRRC Review (Copy Page 131 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 109 Exam: SRO System: 295030 K/A: EA2.03 Lesson Plan M-8114L-005 Enablina Obiective A plant transient has occurred which has resulted in the following conditions:

"* Torus level is minus 3 (-3) feet and decreasing.

"* Torus temperature is 160°F and steady.

1) Given these conditions what is the highest that reactor pressure may be to keep from Emergency Depressurizing?

2) The basis for maintaining reactor pressure below this limit is to ensure that a Blowdown will not result in exceeding the...

A. 1) 850 psig

2) Torus design temperature.

B. 1) 950 psig

2) Torus design temperature.

C. 1) 850 psig

2) Drywell design temperature.

D. 1) 950 psig

2) Drywell design temperature.

Answer: A

Reference:

C.5.1-1200 Question Pedigree: NEW Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Correct answer.

B. Maximum pressure would be 850 psig because level is at the 3 foot graph on the Heat Capacity Limit Graph (Fig M) so must take action if in the shaded region above the curve or on the curve C. Correct pressure but the basis is either the Torus Design Temperature or Drywell Pressure Limit.

D. Incorrect pressure and incorrect basis.

SRO Level Justification SRO knowledge level based on examinee must recall knowledge of the basis for EOP steps and graphs to answer the question.

COGNITIVE LEVEL JUSTIFICATION Level 3 - candidate would need to determine where point on the graph from the information given and then understand the basis for the graph.

KRTC Review (Copy Page 132 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 110 Exam: SRO System: 295031 K/A: 2.1.32 Lesson Plan M-8107L-072 Enablina Obiective Given the following conditions.

"* The plant is operating at 100% power.

"* A failure occurs in the Reactor Level Control System.

"* Reactor water level is lowering at 5 inches per minute.

"* A rapid power reduction is performed.

"* In accordance with 4 AWI-04.01.01, GENERAL PLANT OPERATING ACTIVITIES, the Control Room Supervisor orders a manual scram at 15 inches Reactor water level prior to receiving the automatic scram.

Which of the following describes the reason for inserting the Reactor scram on low Reactor water level?

A. At the low Reactor water level scram set point excessive carryunder of steam will occur, which will raise the core boiling boundary and may challenge fuel integrity.

B. The steam dryer skirt starts to uncover allowing excessive carryover of steam into the downcomer region, which reduces NPSH to the Reactor Recirc pumps.

C. A Reactor scram is required at the low Reactor water level set point to reduce the fission heat generation in the core to ensure adequate core cooling will be maintained.

D. At the low Reactor water level scram set point steam will bypass the dryer and separator through the downcomer causing excessive carryover of moisture to the main turbine.

Answer: C

Reference:

B.05.06-02 and Tech Spec 2.1 & 2.3 Bases Question Pedigree: New Cog. Level: I ANSWER/DISTRACTER JUSTIFICATION A. If excessive carryunder of steam occurs this will reduce the core inlet subcooling and cause the core boiling boundary to lower (boiling will occur at a lower point in the core).

B. Carryover is defined as the water entrained in the steam passing through the dryer and separator, which could make it to the turbine and damage turbine blading. Carryover does not occur on low water level; this is called carryunder.

C. This is the correct answer as stated in the bases of Tech Specs.

D. Steam cannot bypass the separator because the separator assembly is bolted to the core shroud.

SRO Level Justification SRO knowledge level based on examinee must recall plant Technical Specication bases information.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this question is stated in the bases of the Tech Specs and is also referenced in the Ops Manual.

Ragv3w (Cofy PRC Page 133 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: ill Exam: BOTH System: 295031 K/A: EA1.06 Lesson Plan M-8107L-025 7 n The following plant conditions exist. Enablina Ohierctivie7~

• 0800:00 A LOCA with a loss of offsite power has occurred.

• 0800:10 No. 11 EDG and No. 12 EDG energize Bus 15 and Bus 16.

• 0800:15 Annunciator 3-A-41, AC INTERLOCK, alarms.

• 0800:45 Annunciator 3-A-38, REACTOR LOW LOW LEVEL, alarms.

• 0800:57 NO. 11 EDG trips on overspeed.

• 0802:00 Annunciator 3-A-38, REACTOR LOW LOW LEVEL, clears.

Based on the above timeline, which of the following describes the expected condition of the ADS Safety Relief Valves (SRVs) AND why?

A. At 0802:02 all three SRVs will open to depressurize the RPV due to the 107 second timer has timed out as indicated by alarm 3-A-38.

B. At 0802:32 all three SRVs will open to depressurize the RPV due to the 107 second timer as indicated by alarm 3-A-41.

C. All three SRVs will remain closed due to the RPV low-low level condition clearing.

D. All three SRVs will remain closed due to the loss of all Div. I low pressure ECCS pumps.

w~> Answer: C

Reference:

B.03.03-01 Question Pedigree: New Cog. Level: 2 ANSWER/DISTRACTER JUSTIFICATION A. See description C.

B. At 0800:15 the AC INTERLOCK annunciator only signifies that low pressure ECCS pumps are running. The 107 second timer has not yet started.

C. The RPV low-low level signal input into the ADS initiation logic is not seal-in. Therefore if this signal clears before the 107 second timer times out the K6A(B) relays will never energize and the ADS function will not occur.

D. With only Div. II low pressure ECCS pumps running contacts will still be closed in both

'A' and 'B' ADS initiation logic channels allowing ADS to function.

COGNITIVE LEVEL JUSTIFICATION Level 2 since this question is asking the student to recognize the relationship between the initiation signals for ADS and what reaction the initiation logic has when these signals change.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 112 Exam: RO System: 295031 K/A: EK2.10 Lesson Plan M-8107L-0O Enablina Obiective Given the following:

"* The plant is operating at 100% power.

"* A transient results in a scram setpoint being exceeded.

"* The Reactor Protection System fails to automatically scram the Reactor.

Without operator action, which of the following describes how the Control Rods will be automatically inserted to shutdown the Reactor?

A. RPV pressure >_1035 psig for 9 seconds will energize the ATWS/ARI system logic to insert Control Rods.

B. RPV pressure > 1135 psig will immediately de-energize the ATWS/ARI system logic to cause Control Rods to insert.

C. RPV level < minus 47 (-47) inches for 9 seconds will energize the ATWS/ARI system logic to cause Control Rods to insert.

D. RPV level < minus 126 (-126) inches, will immediately de-energize the ATWS/ARI system logic to cause Control Rods to insert.

Answer: C

Reference:

B.05.06-02

<"-' Question Pedigree: New Cog. Level: I ANSWER/DISTRACTER JUSTIFICATION A. There is no time delay associated with the high RPV pressure trip of the ATWS/ARI system.

B. The ATWS/ARI system logic is energize to actuate.

C. This is correct; -47" for 9 seconds will energize the ATWS/ARI system logic which will trip the remaining recirc pump and vent off the scram air header allowing all scram inlet and outlet valves to open inserting all control rods.

D. There is a 9 second time delay associated with the trip of ATWS/ARI on RPV level. The ATWS/ARI system logic is energize to actuate.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is an interlock/setpoint question of the ATWS/ARI system.

HIRC W RlRllH TZAIM ATh©IP HRC RevewTCfy Page 135 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 113 Exlam: RO System: 295032 K/A: 2.4.48 Lesson Plan M-81 07L-003 IPnnhlin Mn t-hit..lig,. Given the following:

"* The plant is operating at 100% power.

"* A report comes to the Control Room that a possible steam leak exists in the RCIC room.

Which of the following describes the confirmatory indication AND corresponding action taken in the Control Room?

Based on TR-4926 points 22, 23, & 24 (RCIC EQUIP AREA) on reading above max normal, PCIS Group _ isolation should be performed.

A. Control Room Back Panel C-21; 5 B. Cable Spreading Room Panel C-18; 5 C. Control Room Back Panel C-21; 4 D. Cable Spreading Room Panel C-18, 4 Answer: A

Reference:

B.02.03 Question Pedigree: New Cog. Level: I ANSWER/DISTRACTER JUSTIFICATION A. A confirmatory look at the temperature recorder with temperature above max normal would require isolation of RCIC per C.4-B.2.4 for steam leak outside the primary containment. The recorder is located on the C-21 panel and this would require a PCIS group 5 isolation be initiated per C.5-1300.

B. See answer A.

C. See answer A.

D. See answer A.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this question requires the student to know location of the recorder for temperature monitoring and the applicable PCIS group isolation for RCIC.

HRIC RwC(Cpy Page 136 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 114 Exam: SRO System: 295036 K/A: EAI.02 Lesson Plan M-8114L-0O Ennhlinn rhiatfiv, Ae,-

Given the following:

"* A fire exists on the 935 foot elevation of the Reactor Building in the SW corner.

"* The fire brigade has been fighting the fire for 15 minutes.

"* Alarms in the Control Room indicate high water level exists in the RHR 'B' corner room.

"* Local report indicates that water level is 4 inches above the floor.

"* Source of the leak has been identified as the fire protection header.

Which of the following describes the action required for the high corner room water level AND why?

A. Do not isolate the fire protection header since this system is an alternate injection source into the RPV to keep the core cooled.

B. Do not isolate the fire protection header since fire fighting efforts take precedence over Secondary Containment concerns in this condition.

C. Isolate the fire protection header as directed by C.5-1300, SECONDARY CONTAINMENT CONTROL, to prevent having to scram the Reactor.

D. Isolate the fire protection header as directed by C.5-1300, SECONDARY CONTAINMENT CONTROL, to prevent having to declare the 'B' ECCS equipment inoperable.

Answer: B

Reference:

C.5.1-1300 Question Pedigree: INPO Bank QID Cog. Level: 3

1. 1832 ANSWER/DISTRACTER JUSTIFICATION A. Even though the fire header can be used as an alternate injection source it is not needed at this time and therefore is not the reason not to isolate the fire protection header.

B. This is correct per the bases of C.5-1300.

C. The fire protection header should not be isolated with a fire in the area.

D. The fire protection header should not be isolated with a fire in the area.

SRO Level Justification SRO knowledge level based on examinee must recall EOP bases information.

COGNITIVE LEVEL JUSTIFICATION Level 3 since the student is required to solve the question with the referenced EOP and then use the knowledge of the bases to determine which is the correct answr.

HIRC !MIITTEN 71XA ]HA7T(©H iRC Rehiw (Copy Page 137 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 115 Exam: RO System: 295033 K/A: EA1.04 Lesson Plan M-8107L-008 Enabling Objective 7.b, 8.f The plant is operating at full power with the following conditions.

"* A leak has developed from RWCU.

"* PCIS Group 3 isolation has failed to occur.

"* Reactor Building ventilation has isolated and 'A' SBGT is running.

"* The following indications exist from 'A' SBGT:

"o EXHAUST FAN V-EF-17A is ON "o AIR HEATING UNIT E-34A-1 is OFF "o AO-2945, FILTER TRAIN UPSTREAM ISOLATION, is OPEN "o AO-2979, FILTER TRAIN DOWNSTREAM ISOLATION, is OPEN "o FLOW CONTROLLER FIC-2943 indicates 3500 cfm

"* Annunciator 24-A-3A, NO DT, is alarming and will not reset.

Based on the above indications, which of the following is the correct operator response?

A. Manually start 'B' SBGT since the 'A' train has failed to start correctly.

B. Start AIR HEATING UNIT E-34B-1 manually since it failed to start in auto.

C. Adjust FLOW CONTROLLER FIC-2943 to achieve 4000 cfm, which will allow the flow switch to turn on the air heating unit.

D. Verify the air heating unit cycles to maintain 70-1 10OF filter train inlet temperature since this is an expected response of the heater.

Answer: A

Reference:

ARP 24-A-3A Question Pedigree: New Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. If the train's heater does not start then excessive moisture will be introduced into the charcoal filter. This condition is adverse to the operation of the charcoal filter. This situation would then require the standby train to be started and the operating train to be shutdown.

B. Air heating unit E-34B-1 is for the 'B' train of SBGT and cannot be started manually.

This heater is only used when 'B' SBGT is in operation.

C. Adjusting the flow controller to increase flow through the train will not cause the heater to turn on. The only interlock associated with the heater and flow is that the heater will turn off with low flow (2000 cfm).

D. The air heating unit does not cycle to maintain temperature. The heater is normally on when the train is running and, with the design flow, will maintain the air temperature to the filter train as required.

COGNITIVE LEVEL JUSTIFICATION Level 1 based on examinee must recall SBGT system response.

MRI]C Revew (Copy Page 138 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 116 Exam: BOTH System: 295033 K/A: EK2.01 Lesson Plan M-8107L-077 Enablina ObiectiwA A high radiation condition exists on the 935 foot elevation of the Reactor Building NE due to a hot particle being lodged in the Scram Discharge Volume (SDV) during a scram. C.5-1300, SECONDARY CONTAINMENT CONTROL, has been entered.

Which of the following describes a possible indication available to warn the operators of this condition?

A. SPDS screen 110, EOP AREA RADIATION LIMITS.

B. ARM reading from Control Room back panel C-252.

C. LIQUID PROCESS HI RADIATION annunciator alarming on the C-04 panel.

D. REACTOR BUILDING HI RADIATION annunciator alarming on the C-05 panel.

Answer: A

Reference:

B.5.12, SPDS User Manual Question Pedigree: New Cog. Level: I ANSWER/DISTRACTER JUSTIFICATION A. Area radiation monitor readings are provided on SPDS.

B. The ARMs provide remote indication of radiation levels on the C-1 1 panel.

C. The liquid process hi radiation annunciator is located on the C-04 panel however it should not be alarming for the condition stated above.

D. The Reactor Building high radiation annunciator is located on the C-04 panel.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this question is asking for facts and locations associated with the ARMs.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 117 Exam: SRO System: 295034 K/A: 2.4.10 Lesson Plan M-7406L-0{

The plant is in a refueling outage. Fuel moves began 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago. While removing a spent fuel assembly from its core cell location, the Fuel Grapple failed resulting in release of the fuel assembly and the assembly dropping back into the cell it was being removed from. The following conditions exist.

"* Evidence of gas release is seen from the refueling cavity.

"* Refuel floor ARM indicates 200 mrem/hr.

"* The following annunciators are alarming:

"o 4-A-1, REFUELING FL AREA HI RADIATION "o 5-A-1, REAC BLDG VENT & F P RAD CH A HI/LO "o 5-A-2, REAC BLDG VENT & F P RAD CH B HI/LO

"* Reactor Building ventilation has isolated and SBGT has started.

Which Guideline would be entered AND which Emergency Action Level would need to be declared for the above stated conditions?

A. Guideline 2 (In-Plant Radiation Levels); Alert (Increase by a factor of 500 in plant radiation levels as indicated by Area Radiation Monitoring System or direct measurement.)

B. Guideline 2 (In-Plant Radiation Levels); Site Area Emergency (Increase by a factor of 1000 in plant radiation levels as indicated by Area Radiation Monitoring System or direct measurement.)

C. Guideline 30 (Major Damage to Spent Fuel); Alert (Dropping, bumping or otherwise rough handling of a spent bundle or individual fuel rods and FUEL POOL RADIATION MONITOR CH A or CH B exceeds 50 mr/hr.)

D. Guideline 30 (Major Damage to Spent Fuel); Site Area Emergency (Dropping a heavy object onto spent fuel confirmed by direct observation and FUEL POOL RADIATION MONITOR CH A or CH B exceeds 50 mr/hr.)

Answer: C

Reference:

A.2-101 Question Pedigree: New Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION THIS QUESTION MAY REQUIRE THE STUDENTS BE GIVEN A COPY OF A.2-101.

A. An alert for high in-plant radiation levels would be declared if refuel floor radiation monitors reached full scale (1 RPhr).

B. There is no site area emergency for guideline 2.

C. Dropping of a spent fuel bundle or rod with an associated fuel pool radiation monitor reading of 50 mr/hr or greater would require declaring an alert for major damage to spent fuel (Guideline 30).

D. This would require dropping a heavy object onto spent fuel confirmed by direct observation or a decrease in fuel pool level below 36'9".

SRO Level Justification M.C W UITITEF ]EAae MR(C Revhew (Copy 4AT0f 5 Page 140 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank SRO knowledge level based on examinee must perform event classification which is a responsibility assigned to SRO licensed individuals.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this is a procedural step that the student will need to reference based on the stated indications.

HR C RvhwCopy Page 141 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 118 Exam: SRO System: 295034 K/A: EK3.05 Lesson Plan M-8114L-005 Enabling Objective 4.a, b & c The plant is operating at 100% power with the following conditions.

"* Steam leak has developed from the HPCI system.

"* REACTOR BUILDING HI RADIATION annunciator is alarming.

"* Reactor Building ventilation has isolated and SBGT is running.

"* HPCI Turbine Area temperature has exceeded MAX SAFE.

"* HPCI Turbine Area radiation monitor indicates UPSCALE HIGH.

"* Reactor Building Drain Tank Room radiation monitor indicates 1R/hr.

"* All automatic and manual attempts to isolate the leak have failed.

Which of the following is the correct action to take for this condition AND why?

A. Blowdown the RPV to minimize the radioactivity released into the Secondary Containment.

B. Open the Main Turbine bypass valves to maximize the use of the Off-Gas System for radioactivity control.

C. Shutdown the Reactor as required by C.5-1100, RPV CONTROL, to reduce the break flow into the Reactor Building.

D. Execute C.3, SHUTDOWN PROCEDURE, to reduce the rate of energy production and heat input into the Secondary Containment.

Answer: A

Reference:

C.5-1300 & C.5.1-1300 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. This is correct because dumping the energy into the Torus will minimize the radioactivity deposited into the Reactor Building.

B. A blowdown can be anticipated in this situation allowing the turbine bypass valves to be used however this is done to minimize the heat deposited into the Torus and preserve the HCTL and suppression capability of the primary containment.

C. A reactor scram is required but is not directed by C.5-1100. RPV Control procedure is performed concurrently because a scram is required per C.5-1300.

D. A normal shutdown would not be the correct action to take in this scenario. An immediate removal of the heat source is required which is directed through the use of a reactor scram.

SRO Level Justification SRO knowledge level based on examinee must recall EOP bases information.

COGNITIVE LEVEL JUSTIFICATION Level 3 since the student is required to determine that entry is required into C.5-1300, what procedure steps need to be taken for the stated conditions, and the bases for the actions taken.

FRC % TIE.IN CE4Z IMAM HIRC Review Copy Page 142 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 118.1 Exam: RO System: 295034 K/A: EK2.04 Lesson Plan M-8114L-005 Enablina Obiective Given the following:

"* A high radiation condition exists in the Reactor Building plenum.

"* A Reactor scram and PCIS Group 2 isolation has occurred due to lowering Reactor water level.

"* Both SBGT trains have failed to start.

"* Annunciators REAC BLDG VENT & F P RAD CH A-HI/LO and REAC BLDG VENT &

F P RAD CH B-HI/LO are NOT alarming.

If indications of a significant radioactive leak were identified in the Turbine Building, which of the following would be correct?

A. The Stack Dilution Fans should be started to dilute any radioactivity discharged.

B. The SBGT System should be manually isolated to prevent contaminants from being drawn into the Reactor Building.

C. The Reactor Building plenum exhaust fans should be started to assure the release is monitored by the RBV WRGMs.

D. The Radwaste Building ventilation fans should be restarted to prevent drawing contaminants in from the Turbine Building.

Answer: C

Reference:

C.4-B.4.1 .B Question Pedigree: Mod '99 NRC Exam Cog. Level: 3 QID RO #81 ANSWERIDISTRACTER JUSTIFICATION A. The SBGT System is not capable of taking a suction from the Turbine Building.

B. There is no connection from SBGT to the Turbine Building.

C. This step is required by C.4-B.4.1.B Part C step 13.a.3).

D. The Radwaste Building ventilation fans should not be restarted.

COGNITIVE LEVEL JUSTIFICATION Level 3 since the student is required to solve the problem using knowledge of the systems and EOPs.

HRC IRvaew C(Cpy Page 143 of 156

Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 119 E) ram: BOTH System: 295035 K/A: EK1.01 Lesson Plan M-81 07L-006 Q*

The plant is operating at 100% power with the following conditions.

"* The roving security officer reports that an equipment explosion has occurred on the 935 foot elevation at the SE corner of the Reactor Building and that no plant security compromise has occurred.

"* Annunciator 3-A-27, RX BLDG DP AT OR ABOVE 0" WATER, is alarming

"* Panel C-24B manometer DPI-4424 reads 0 inches H20.

Which of the following describes the effect this transient has on operation of the plant?

A. The Reactor should be immediately scrammed.

B. Verify automatic isolation of Secondary Containment and automatic initiation of SBGT.

C. Commence a normal shutdown per C.3 and be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

D. Throttle supply fan flow by raising the control air pressure for the associated SCTMT isolation dampers.

Answer: C

Reference:

B.04.02-05 H.1, Tech Spec 3.7.C.4 Question Pedigree: New Cog. Level: 1 ANSWER/DISTRACTER JUSTIFICATION A. There is no procedural guidance or any need to scram the reactor. Procedure B.04.02 05 H.1 and Tech Specs require commencing a normal reactor shutdown.

B. Secondary Containment does not auto isolate and SBGT does not auto start on high RX Bldg pressure (low Rx Bldg DP).

C. With a loss of secondary containment integrity, a normal reactor shutdown should be commenced and the plant placed in a condition where secondary containment integrity is not required within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

D. Throttling supply fan flow is an action taken for high pressure within the Reactor Building however the control air pressure should be lowered not raised for the associated isolation damper.

COGNITIVE LEVEL JUSTIFICATION Level 1 since the actions being taken are directed by procedure and Tech Specs.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 120 Exam: BOTH System: 295037 K/A: EK2.07 Lesson Plan M-81 14L-005 P:nnhlinn r)3 hg,'tfi tz Which of the following sets of conditions would require implementation of C.5-2007, FAILURE TO SCRAM?

A. APRMs indicate 1% power Turbine Control Valve oil pressure is 100 psig 4 Control Rods are at position 48 RPS air header is depressurized B. APRMs indicate 8% power West SDV contains 40 gallons of water Control Rods have not inserted Backup scram valves are de-energized C. IRMs indicate 75 on range 7

'A' and 'B' MSIVs are 50% open 2 Control Rods are at position 48 RPS air header is depressurized D. IRMs indicate 75 on range 9 Main Condenser vacuum is 20" Hg Control Rods have not inserted Backup scram valves are energized Answer: D

Reference:

C.5-1100 and C.5-2007 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. APRMs <3% power is not an entry condition for C.5-1100 and therefore not an entry into C.5-2007.

B. SDV level of 40 gallons is a control rod block not a scram signal and the backup scram valves are normally de-energized therefore control rods should not have inserted.

C. IRM indication of 75 on range 7 would be <1% power and is not an entry condition for C.5-1100 and therefore not an entry into C.5-2007.

D. IRM indication of 75 on range 9 would be about 8% reactor power and with Main Condenser vacuum of 20" Hg control rods should have inserted. Backup scram valves energized as expected but no controls rods inserted. This would require entry into C.5 1100 for power above 3% when a scram was required and since rods did not insert it would require entry into C.5-2007.

COGNITIVE LEVEL JUSTIFICATION Level 3 since the student is required to have knowledge of several references and analyze them to come up with an expected course of action.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 121 Exam: SRO System: 295037 K/A: EK3.04 Lesson Plan M-8114L-005 Enabline Obiective The following conditions exist.

"* The Reactor has failed to scram.

"* Reactor power is 12% and steady.

"* SRVs are being used to maintain Reactor pressure below 1056 psig.

"* Torus temperature is 113 0 F and rising.

"* SBLC System is injecting into the RPV.

"* Reactor water level is being maintained minus 126 (-126) inches to minus 149 (-149) inches.

Which of the following describes the condition the operating crew will expect to achieve NEXT AND the affect this condition has on the core?

A. Hot shutdown boron weight will allow the crew to state that the Reactor will remain shutdown under all conditions and exit the failure to scram EOP.

B. All SRVs remaining closed will allow the crew to restore the normal RPV water level control band using preferred and alternate ATWS Injection Systems.

C. SBLC tank level < 675 gallons will allow the crew to restore the normal RPV water level control band to promote mixing of boron in the core via natural circulation.

D. All SRVs remaining closed will allow the crew to transition from the override leg of RPV water level control and restore RPV water level to the normal operating band.

Answer: C

Reference:

C.5.1-2007 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. The reactor is not guaranteed to remain shutdown under all conditions with just Hot Shutdown boron injected.

B. All SRVs remaining closed does not allow restoring the normal water level control band.

C. This is correct.

D. All SRVs remaining closed does not allow restoring the normal water level control band.

COGNITIVE LEVEL JUSTIFICATION Level 3 since this question requires the student to solve the problem using C.5-2007 and with the step in mind ensure that the bases for that step is correct.

SRO Level Justification SRO knowledge level based on examinee must recall EOP bases information.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 121.1 Exam: RO System: 295037 K/A: EK2.04 Lesson Plan M-8114L-005 Enablina OhiectivA_

4h An The plant was operating at full power when the following transient occurred.

"* MSIVs CLOSE from spurious PCIS Group 1 isolation.

"* Control Rods have failed to insert.

"* Reactor pressure peaked at 1150 psig.

"* Reactor power is now 23% and slowly lowering.

"* Feedwater and Condensate has become unavailable.

"* Reactor water level is being lowered by preventing injection.

"* Reactor pressure is being maintained via SRVs.

"* Drywell radiation indicates possible fuel element failure.

Which of the following actions should be performed to minimize heat input into the Primary Containment?

A. Inject SBLC.

B. Open the MSIVs.

C. Blowdown the RPV.

D. Trip the Recirc pumps.

<> Answer: A

Reference:

C.5-2007 & C.5.1-2007 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. Ifterminate and prevent is in progress for level control then Torus temperature is already at or above 1 10°F. This requires the initiation of SBLC. This is done to minimize the possibility of having to perform a blowdown from Torus water temperature approaching the HCTL.

B. The MSIVs should not be opened because of the indications of a fuel element failure.

C. A blowdown is not required because CRD, HPCI and RCIC are still available high pressure sources of injection.

D. The recirc pumps have already tripped due to actuation of the ATWS system on high RPV pressure and is not required by C.5-2007 with the MSIVs closed.

COGNITIVE LEVEL JUSTIFICATION Level 3 because the student is required to determine, based on the conditions stated, what actions have been completed in C.5-2007, what action would be necessary and why that action would be needed to help correct the transient.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 122 Exam: BOTH System: 295038 K/A: EK2.03 Lesson Plan M-8114L-005 A f-Enablinaconditions:

The plant is operating at 100% power with the following ObierctivA A

"* A transient has occurred resulting in fuel failure.

"* Release rates are approaching the Alert level.

"* PCIS Group 2 and secondary containment isolation has occurred.

"* Several Turbine Building ARMs are alarming.

"* All automatic functions have occurred as expected.

Based on the above conditions, which of the following systems should be started to minimize the effect this transient has on the health and safety of the public and plant personnel?

A. Standby Gas Treatment System B. High Radiation Sampling System C. Turbine Building Ventilation System D. Reactor Building Ventilation System Answer: C

Reference:

C.4-B.04.01 .B Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION

/

A. SBGT should already be running and does not impact Turb Bldg Release Rates.

B. HRSS provides no filtering function to protect the public or plant personnel.

C. TB Ventilation will provide for a monitored release that is not ground level.

D. RB Ventilation cannot be started and is not allowed by EOPs.

COGNITIVE LEVEL JUSTIFICATION Level 3 since this question requires the student to solve the problem using knowledge of the EOP actions.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 123 Exam: SRO-RO System: 295038 K/A: EK3.03 Lesson Plan M-8107L-049 Enablina Obiective. 11* 1A The plant is in a refueling outage with the following conditions.

"* Spent fuel is being moved in the Spent Fuel Pool.

"* A spent fuel bundle is dropped on top of several other spent fuel bundles.

"* Large gas bubbles are emanating from the spent fuel bundles.

"* All of the Reactor Building 1027 foot elevation radiation monitors are in alarm.

"* Standby Gas Treatment has started.

"* Control Room air intake radiation monitors are reading 2 mrem/hr.

Which of the following describes the response of the CRV-EFT System AND why?

The CRV-EFT System will...

A. operate in the Normal Mode since normal outside air intake is not affected by the above conditions.

B. shift to the Isolate Mode to prevent any contaminated air from entering the Control Room atmosphere.

C. operate in the Normal Mode to provide filtered outside air to the Control Room for pressurization and prevent in-leakage of contaminated air.

D. shift to the High Radiation Emergency Mode to provide filtered outside air to the Control Room for pressurization and prevent in-leakage of contaminated air.

Answer: D

Reference:

B.08.13-01 Question Pedigree: INPO Exam Bank QID Cog. Level: 1

1. 16334 ANSWER/DISTRACTER JUSTIFICATION A. See description of D.

B. There is no isolation mode of the CRV-EFT System.

C. See description of D.

D. Control Room air intake radiation monitors (RM-9021A1B) monitor radioactivity introduced into the CR HVAC System. When radiation levels reach 1 mr/hr annunciator 20-B-4, CR Air Intake HI-HI Rad, will alarm and the CRV-EFT System will automatically shift into the High Radiation Emergency Mode to provide HEPA/charcoal filtered outside air to the Control Room and EFT Buidling 1 st and 2 nd floors to pressurize them. This will prevent any leakage into the Control Room from adjacent spaces.

COGNITIVE LEVEL JUSTIFICATION Level 1 since this question is checking fundamental knowledge of system interlocks, setpoints and system response to high radiation of CR intake air.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 123.1 E) cam: RO System: 295038 K/A: EA1.03 Lesson Plan M-8107L-077 Enabling Objective 7.k M-8107L-026 8.a the following:

Given the following:

"* The plant is operating at 100% power.

"* The RBCCW Liquid Process Radiation Monitor is alarming on the C-10 Panel and has been verified to be reading greater than the alarm setpoint.

"* The RBCCW Expansion Tank high level alarm is in and tank level has been verified to be greater than the alarm setpoint.

Leakage from which of the following systems/components would NOT result in this condition?

A. RWCU B. Fuel Pool Cooling C. Reactor Recirc Pumps D. Primary Containment Drywell Coolers Answer: D

Reference:

B.5.11 and C.4-B.02.05.B Question Pedigree: Mod INPO Bank Cog. Level: 2 QID#1789 ANSWER/DISTRACTER JUSTIFICATION A. This system would cause the indications given.

B. This system would cause the indications given.

C. This component would cause the indications given.

D. Leakage from the Drywell coolers may cause radioactivity to increase into the RBCCW system if a LOCA were to occur and Drywell pressure was greater than RBCCW pressure however, this condition would not cause RBCCW surge tank level to rise.

COGNITIVE LEVEL JUSTIFICATION Level 2 since this question requires the student to recognize the interaction between the process liquid radiation monitoring system and the process streams it monitors. Using that information the student must then determine which supported system could cause the indications given.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 124 Exam: BOTH System: 400000 K/A: A4.01 Lesson Plan M-8107L-026 Enabling Objective 7.a The plant is operating at 100% power with the following initial conditions.

"* No. 12 RBCCW Pump is operating with No. 11 RBCCW Pump in standby.

"* Procedure 1084, RBCCW STANDBY PUMP AUTO INITIATION TEST, is in progress for No. 11 RBCCW Pump.

"* Isolation valve for PS-1398, RBCCW LOW PRESSURE PUMP START, for No. 11 RBCCW Pump has been CLOSED.

After CLOSING isolation valve for PS-1 398, breaker 52-403 trips on overcurrent causing a loss of MCC 142A and MCC 142B.

Which of the following actions should be taken to correct the stated conditions?

A. Observe auto start of the No. 11 RBCCW Pump.

B. Manually scram the Reactor and verify RWCU System isolates.

C. The standby RBCCW Pump should be manually started from the C-06 Panel.

D. Trip both Reactor Recirc Pumps after 60 seconds and CLOSE RBCCW Drywell isolation valves.

Answer: C

Reference:

C.4-B.02.05.A, Procedure 1084 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. The standby RBCCW Pump will not auto start since the isolation valve to the pressure switch sensing low system pressure has been closed.

B. These actions should not be necessary since the standby pump should be capable of being started from the control room.

C. The standby pump should be started manually from the control room as directed by immediate action #1 of C.4-B.02.05.A.

D. See description for B.

COGNITIVE LEVEL JUSTIFICATION Level 3 based on the fact that the RBCCW system has been placed in an abnormal lineup and the student will need to analyze that, with a loss of power to the operating pump, the standby pump will not auto start. This will require action to be taken per the procedure to start the standby pump manually from the control room.

]R]C RevhI (

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 125 E) cam: BOTH System: 400000 K/A: K6.04 Lesson Plan M-81 07L-01 7 Ennhlinn f'lhiarfiv*=, 7,,A The plant is operating at 100% power during a hot summer day with the following conditions.

"* Low river level has placed the Circulating Water System in closed cycle operation.

"* 480V transformer X-10 is out of service due to overheating.

"* LC 101 is being powered from LC 102 via 101/102 LC BUS TIE ACB 52-209.

"* No. 11 Circulating Water Pump trips due to a loss of power to Bus 13.

Which of the following states the effect on the Circulating Water System?

A. No. 12 CIRC WATER PUMP will trip.

B. No. 12 CIRC WATER PUMP will trip and the MO-1 850 will CLOSE.

C. No. 11 COOLING TOWER PUMP will trip.

D. NO. 11 COOLING TOWER PUMP will trip and the MO-1 154 will move to the intermediate low flow position.

Answer: C

Reference:

B.06.04-02 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. 12 CWP will go to runout condition but there is no trip associated with this condition.

B. MO-1 850 will not close because LC-109 has lost power.

C. 11 CTP will trip on interlock with the trip of the 11 CWP in closed cycle operation.

D. MO-1 154 should move to the intermediate high flow position on a trip of a CWP.

COGNITIVE LEVEL JUSTIFICATION Level 3 based on that the student 1) must recognize what the circulating water system lineup will be in closed cycle operation, 2) must determine that LC 109 will not be automatically cross-tie to LC 102 and will lose power, and 3) understand what interlocks for the Circulating Water System will affect the final sytem lineup.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 126 Exam: RO System: 500000 K/A: EAI.05 Lesson Plan M-8107L-023 Enabling Objective 7.m, 7.q M-81 07L-01 6 7-Given the following plant conditions.

"* A large break LOCA has occurred with a loss of all offsite power.

"* Containment hydrogen and oxygen concentrations require initiation of Torus sprays.

"* The following actions were taken to initiate Torus sprays on Division I of RHR:

"o Control Switch 10A-S17A, CONTAINMENT SPRAY/COOLING LPCI INITIATION BYPASS, was placed in BYPASS.

"o MO-2006, DISCHARGE TO TORUS OUTBOARD, was OPENED.

"o MO-2010, TORUS SPRAY - INBOARD, was OPENED.

"o MO-2008, TORUS COOLING INJ/TEST INBOARD, was OPENED as necessary to maintain RHR pump flow requirements.

"o MO-2002, HX BYPASS, keylock switch was placed in the CLOSE position.

"o 11 RHRSW Pump and 13 RHRSW Pump control switches were placed in the START position.

Which of the following describes the status of Torus sprays?

A. Div I RHR pumps are spraying the Torus and RHRSW pumps are providing cooling to the RHR Hx.

B. Div I RHR pumps are NOT spraying the Torus but RHRSW pumps are providing cooling to the RHR Hx.

C. Div I RHR pumps are spraying the Torus and RHRSW pumps are NOT providing cooling to the RHR Hx.

D. Div I RHR pumps are NOT spraying the Torus and RHRSW pumps are NOT providing cooling to the RHR Hx.

Answer: C

Reference:

C.5-3502 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. RHRSW Pumps cannot be started unless the RHRSW PUMPS LPCI AND ECCS LOAD SHED MANUAL OVERRIDE switch is placed in the MANUAL OVERRIDE position.

B. RHRSW Pumps cannot be started unless the RHRSW PUMPS LPCI AND ECCS LOAD SHED MANUAL OVERRIDE switch is placed in the MANUAL OVERRIDE position.

C. RHR Division I has been aligned for torus spray per C.5-3502, however RHRSW has not been properly aligned as the RHRSW Pumps cannot be started unless the RHRSW PUMPS LPCI AND ECCS LOAD SHED MANUAL OVERRIDE switch is placed in the MANUAL OVERRIDE position with the loss of offsite power condition present.

D. See answer C.

COGNITIVE LEVEL JUSTIFICATION Level 3 because the student is required to solve the question using knowledge of the procedure for initiation of Torus sprays with a load shed of RHRSW pumps.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 127 Exam: BRO System: 500000 K/A: EA2.03 Lesson Plan M-8114L-005 Enablina ObiecivA. A n EnablinaO~iectiv A transient occurred 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago resulting in the following conditions:

0 Drywell pressure is 14.6 psig and slowly lowering.

0 Torus level is 5 inches and steady.

0 Torus temperature is 165 0 F.

0 Drywell temperature is 2750 F.

0 Actual Reactor water level is minus 174 (-174) inches and steady.

0 Upon checking the H2/02 analyzer readings it is determined that hydrogen concentration is 7% and oxygen concentration is 5.5%.

Based on the indications, which of the following describes the implications to plant safety?

A. Exceeding the Drywell Spray Limit challenges Primary Containment integrity.

B. Exceeding the Hydrogen Combustion Limit challenges Primary Containment integrity.

C. Exceeding the Vortex Limit challenges core cooling.

D. Exceeding the NPSH Limit challenges core cooling.

Answer: B

Reference:

C.5-1205, C.5-1000 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. For conditions given, the Drywell Spray Limit is not exceeded.

B. For the conditions given, the hydrogen combustion limit is exceeded.

C. For conditions given, the Vortex Limit is not exceeded.

D. For conditions given, the NPSH Limit is not exceeded.

COGNITIVE LEVEL JUSTIFICATION Level 3 since this question requires the student to formulate a course of action using available references based on the indications given. The student also needs to evaluate core conditions to determine that Drywell and Torus sprays should not be used to lower hydrogen concentration at this time in the scenario.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 127.1 Exam: BSRO System: 500000 K/A: EA2.03 Lesson Plan M-8114L-005 Enablina Objec~tive En..........

The plant is in a LOCA with an emergency depressurization in. cprogress.

.iv .r 4 Systems have been aligned to feed the RPV and water level is 10 inches and rising fast. RPV pressure is currently 300 psig and lowering.

Which of the following situations would require the initiation of Drywell sprays AND why?

A. Drywell pressure is 10 psig and Drywell temperature is 1750 F; to reduce Drywell pressure and temperature through convective cooling.

B. Drywell pressure is 30 psig and Drywell temperature is 3250 F; to reduce Drywell pressure and temperature through evaporative cooling.

C. Containment oxygen concentration is unknown and containment hydrogen concentration is 5.4%; to ensure that a deflagration condition does not occur.

D. Containment oxygen concentration is 5.4% and containment hydrogen concentration is unknown; to reduce the flammability of combustible gasses in the Drywell atmosphere.

Answer: D

Reference:

C.5-1205 and C.5.1-1205 Question Pedigree: Mod INPO Bank Cog. Level: 3 QID #905 ANSWER/DISTRACTER JUSTIFICATION A. Drywell pressure must be above 12 psig and Drywell temperature must be approaching 281IF for initiation of Drywell sprays.

B. The conditions stated place the containment outside of the DSIL.

C. Drywell sprays will not prevent a deflagration condition from occurring but will limit the pressure effects if one does occur.

D. This is correct.

COGNITIVE LEVEL JUSTIFICATION Level 3 because the student is required to solve the question using knowledge of the bases of EOPs for initiation of Drywell sprays with a combustible condition in the containment.

SRO ONLY JUSTIFICATION This question is SRO only level of knowledge because it is taken from the bases of the EOPs.

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Monticello Nuclear Generating Plant 2002 NRC Licensing Exam Bank Question: 128 Exam: SRO System: 600000 K/A: 2.3.10 Lesson Plan M-7406L-01 q 7 Given the following conditions:

"* The plant has just started up after a refueling outage.

"* A fire has broken out in the Radwaste Trash Compacter Area.

"* The Fire Brigade has been fighting the fire for 12 minutes.

"* Levels of airborne contamination have been rising in the Radwaste Building.

"* No safety systems have been affected by the fire.

Which of the following is correct with regards to the above conditions?

A. Declare an Unusual Event for a plant fire and make a site evacuation announcement over the site's PA system.

B. Declare an Unusual Event for a plant fire and make a local evacuation announcement over the plant's PA system.

C. Declare a Site Area Emergency for a plant fire and make a site evacuation announcement over the site's PA system.

D. Declare a Site Area Emergency for a plant fire and make a local evacuation announcement over the plant's PA system.

Answer: B

Reference:

A.2-301 Question Pedigree: New Cog. Level: 3 ANSWER/DISTRACTER JUSTIFICATION A. A site evacuation would not be made for this fire.

B. This is correct.

C. A site area emergency would only be applicable for a fire affecting a safety system.

D. A site area emergency would only be applicable for a fire affecting a safety system.

SRO Level Justification SRO level knowledge as the responsibility to determine need to evacuate plant areas is a responsibility of Emergency Director per A.2-301 and is identified as SM/SS knowledge in lesson plan M-7406L-002.

COGNITIVE LEVEL JUSTIFICATION Level 3 since this question requires the student to solve the problem using knowledge of the references.

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