ML20149D285

Forwards NRC Approved Operator Licensing Exam (Facility Outline & Initial Exam Submittal & as-given Operating Exam) for Tests Administered on 970317-21

ML20149D285
Person / Time
Site:	Beaver Valley
Issue date:	07/14/1997
From:	Burley V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9707170073
Download: ML20149D285 (133)

Text

i July 14, 1997 NOTE T0:

NRC Document Control Desk Mail Stop 0-5-D-24 i

FROM:

Ircoil Oon/ev

.-Licensing Assistant OpeCbting Licen ng Branch. R I'

SUBJECT:

OPERATOf,LICENSINGEXAMINATIONADMINIS'EkDONb ei d 1

Mac 17 <171

. AT &Me n DOCKET #50-d3//-

g t((7e tdeeFor.J//7/97 Operator Licensing Examinations were administered at the refdrehced fac11ity. Attached, you will find the following

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information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:

i Item #1 -

a)

Facility submitted outline and initial exam submittal, designated for distribution under RIDS Code A070.

b)

As given operating ~ examination designated for distribution under RIDS Code A070.

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Rec. 2-/17/ 9 7 i To: Paul Bissett - USNRC From: Rich Brooks, Duquesne Light Co. Date: February 13,1997

Subject:

Beaver Valley Power Station, Unit 1 - Initial Exam Materials. The following materials are being submitted to you for review, comment, and approval for the BVPS Unit 1 NRC Initial License Examination scheduled for the week of March 17,1997. This submittal is in accordance with the instructions in Revision 7, Supplement 1, of NUREG-1021, " Operator Licensing Examiner Standards," Revision 5, of NUREG-BR-1022, " Examiners Handbook for Developing Operator Licensing Written Examinations," and associated " Voluntary Pilot Examination Program Guidelines." )

1. Integrated Examination Outline (rev.1)

2. Written Exam (105 Questions)

3. Written Exam Reference Package, with index.

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4. Oper6lng Test Drills (6)

5. Operating Test JPMs (25)

6. Exam Materials History We request that these materials be withheld from public disclosure until after the completion of the examination.

If you have any questions or require further information please contact me at 412-393-5755. / / qD o 1 )

Exam Materials History Written Exam Questions 1. 105 Questions are submitted,5 additional questions are to be used as replacements I if necessary, The Outline delineates which sections of the exam have extra questions. '2. Breakdown of questions: ] e 74 New e 30 Modified 1 from the LOT Exam bank. Question # 1-97-93. (Question # 1 of the Exam) Question was used on 1/96 for P. Cilli Exam only, other candidates e did not see this question. 3. The specific information regarding this history is identified on each question after the 1-97-XX number on the question in the lower right hand comer. (ex. M-0125 j identifies the question as a Modified version of LOT exam bank question # 125, i New, simply means entirely New. ) Drills All six Drills are entirely New. i JPMs All 25 JPMs are New,6 are Time Critical and 4 are Faulted. Refer to the exam outline i for specifics on which JPMs will be used for the Instants vs Upgrades. Exam Outline The Exam Outline (rev.o) that was sent to you on January 16,1997 has been revised as follows: 1. The JPM for local E-Boration has been deleted and replaced with Local lockout of 4KV emergency bus during a fire. This change was due to existing Hi-Rad and contaminated area around the E Boration valve. 2. The JPM for Sign-out of safety related valve keys was deleted and replaced with ' Fill Out An OMCN'(on the spot procedure change). This was done because the key sign-out JPM was judged to have little discriminatory value. 3. Some cf the original K/A's selected for the written examination have either been deleted or revised. 4. These revisions are designated by Rev bars in the right hand margin of the Outline. - Validation All of the submitted exam materials have been validated by a Licensed operating crew and the exam team. None of the Written Exam Materials are duplicated on the Independent Audit Examination which will be administered on Feb 17,1997. 2 j

) i 6 BVPS Unit 1 Senior Reactor Operator March,1997 - NRC Initial Licensed 4 Operator Examination Exam Outline Exam Breakdown Total Questions Submitted = 105 Plant Wide Generics: 17 Questions (0 Extra) Plant Systems - Group 1: 20 Questions (1 Extra) Plant Systems - Group 2: 17 Questions (0 Extra) Plant Systems - Group 3: 05 Questions (1 Extra) Emergency Plant Evolution's - Group 1: 24 Questions (0 Extra) Emergency Plant Evolution's - Group 2: 18 Questions (2 Extra) Emergency Plant Evolution's - Group 3: 04 Questions (1 Extra) BVPS Rev. I

Knowledgsend Abilities Record Form PLANT-WIDE GENERIC RESPONSIBILITIES PWR - Senior Reactor Operator -17% BVPS - Unit 1 .i Check if 194001 included K/A # Statement Rating / K1.01 ' Knowledge of how to conduct and verify valve lineups. 3.7 / K1.02 Knowledge of tagging and clearance procedures. 4.1 / K1.03 Knowledge of 10 CFR 20 and related facility radiation control 3.4 requirements. / K1.04 Knowledge of facility ALARA program. 3.5 / K1.05 Knowledge of facility requirements for controlling access to 3.4* vital / control areas. K1.06 Knowledge of safety procedures related to rotating equipment. 3.4* K1.07 Knowledge of safety procedures related to electrical equipment. 3.7* / K1.08 Knowledge of safety procedures related to high temperature. 3.4 l K1.09 Knowledge of safety procedures related to high pressure. 3.4 / K1.10 Knowledge of safety procedures related to caustic solutions. 3.3 K1.11 Knowledge of safety procedures related to chlorine. 3.5* j K1;12 Knowledge of safety procedures related to noise. 2.9 -/- K1.13 Knowledge of safety procedures related to oxygen-deficient 3.6 environment. K1.14 Knowledge of safety procedures related to confined spaces. 3.6 /- - K1.15 Knowledge of safety procedures related to hydrogen. 3.8* / K1.16 - Knowledge of facility protection requirements, including fire brigade 4.2* and portable fire-fighting equipment usage. / K1.17 Knowledge of the equipment rotation schedules and reasoning 2.5 behind the rotation procedure. s i i Examiners' Handbook 2-53 BVPS. REV.1 - j

_.. _ _ - - ~ _ _ - _. _. - _. - _ _. _ - _.. ~. _ - - Knowledgn t.nd Abilities Record Form PLANI'-WIDE GENERIC RESPONSIBILITIES j PWR - Senior Reactor Operator (Continued) BVPS - Unit 1 Checkif '194001 included K/A # Statement Rating l / . A1.01 Ability to obtain and verify control procedure copy. 3.4 / A1.02 Ability to execute procedural steps. 3.9 i-A1.03 Ability to locate and use procedures and station directives related to 3.4 shift staffing and activities. ) A1.04 Ability to operate the plant phone, paging system, and two-way 3.2 radio. A1.05 Ability to make accurate, clear and concise verbal reports. 3.8 i A1.06 ' Ability to maintain accurate, clear and concise logs, records, status 3.4 bnrds and report.. / A1.07 Ability to obtain and interpret station electrical and mechanical 3.2 drawings. -/ A1.08 Ability to obtain and interpret station reference material such as 3.1 graphs, monographs, and tables which consin system performance data. I. A1.09 Ability to coordinate personnel activities inside the control room. 3.9* I A1.10 ' Ability to coordinate personnel activities outside the control room. 3.9* A1.11 Ability to' direct personnel activities inside the control room. 4.1* A1.12 Ability to direct personnel activities outside the control room. 4.1* F A1.13 Ability to locate control room switches, controls, and indications, and 4.1 to determine that they are correctly reflecting the desired plant lineup. / ' A1.14 Ability to maintain primary and secondary plant chemistry within 2.9 allowable limits. l i j ( Examiners' Handbook 2-54 BVPS. REY. I

1 Knowledge and Abilities Record Form PLANT-WIDE GENERIC RESPONSIBILITIES PWR - Senior Reactor Operator (Continued) BVPS - Unit 1 Check if - 194001 included K/A # Statement Rating A1.15 Ability io use plant computer to obtain and evaluate parametric 3.4 information on system and component status. { / A1.16 Ability to take actions called for in the Facility Emergency Plan, 4.4

• including (if required) supporting or acting as the Emergency Coordinator. -

i i 4 i Examiners' Handbook 2-55 BVPS. REV. i

___m Knowledgacnd Abilities Record Form PLANT SYSTEMS PWR - Senior Reactor Operator - 40% BVPS - Unit 1 ' t Plant Specific Priorities - System # K/A # ' K/A Topic Rating i 3.01 004. 010.A4.03 Ability to manually operate and/or monitor in the control room: 3.7 Boration/dllution. i Group I Plant Systems-19% 001 Control Rod Drive System 025 Ice Condenser System (N/A BVPS) 003 Reactor Coolant Pump System 056 Condensate System 004 Chemicaland Volume ControlSystem 059 Main Feedwater System 013 ESF Actuation System 061 Auxiliary / Emergency Feedwater System 014 Rod Position Indication System 063 DC Electrical Distribution System 015 NuclearInstrumentation System 068 Liquid Radwaste System { 017 In-Core Temperature Monitor System 071 Waste Gas DisposalSystem 022 Containment Cooling System 072 Area Radiation Monitoring System System # K/A # K/A Topic Rating 3.01 001. 000.K5.08 Knowledge of the following theoretical concepts as they apply 4.4 to the CRDS: Reasons for rod insertion limits and their effect on shutdown margin. 3.01 001, 050.A2.01 Ability to (a) predict the impacts of the following malfunctio'ns 3.9. or operations of the CRDS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Urgent failure alarm, including rod-out-of sequence and motion-inhibit alarms. 3.01 004. 020.K1.02 Knowledge of the physical connections and/or cause-effect 2.8 i relationships between the CVCS and the following systems: Flow path from CVCS to the reactor coolant drain tank and holdup tank. 3.01 004. 010.A4.03 Ability to manually operate and/or monitor in the control 3.7 room: Boration/ dilution. 3.01 014. N/A N/A N/A 3.02 013. 000.K4.12 Knowledge of ESFS design feature (s) and/or interlock (s) 3.9 which provide for the following: Safety injection block. ~ 3.02 013. 000.A4.03 Ability to manually operate and/or monitor in the control 4.7 room: ESFSinitiation. - 3.04 003. 000.K3.02 Knowledge of the effect that a loss of the RCP's will have on 3.8 the following: S/G. 3.04 003. 000.A2.01 Ability to (a) predict the impacts of the following malfunctions 3.9 or operations of the RCP's; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Problems with RCP seals, especially rates of sealleakoff. Examiners' Handbook 1 BVPS REV.1 -

' Knowledge and Abilities Record Form PLANTSYSTEMS PWR - Senior Reactor Operator - 40% BVPS - Unit t Group I Plant Systems - (Continued) System # K/A # ~ K/ Atopic Rating 3.05 056. N/A N/A N/A 3.05 05i 000.K1.04 Knowledge of the physical connections and/or cause-effect 3.4 relationships between the MFW system and the following systems: S/G water level control system. 3.05 061. 000.K2.02 Knowledge of bus power supplies to the following: AFW 3.7 electric-driven pumps. 3.05 061. 000.A3.01 Ability to monitor automatic operation of the AFW system,

4.2 including

AFW startup and flows. 3.% 022. 000.K4.03 Knowledge of HVR design feature (s) and/or interlocks (s) 4.0 which provide for the following: Automatic containment isolation. 3.06 022. 000.A1.04 Ability to predict and/or monitor changes in parameters (to 3.3 prevent exceeding design limits) associated with operating the HVR controls including: Cooling water flow. 3.06 026. 000.G06 Knowledge of bases in technical specifications for limiting 3.8 conditions for operations and safety limits. (Containment Spr y System) 3.07 063. 000.K1.03 Knowledge of the physical connections and/or cause-effect 3.5 relationships between the de electrical system and the following systems: Battery charger and battery. 3.09 015. 000.K2.01 Knowledge of bus power supplies to the following: NIS 3.7 channels, components, and interconnections. 3.09 015. 000.G05 Knowledge of the Limiting conditions for operations and 3.8 safety limits. (Nuclear Instrumentation System) 3.09 017. 020.A3.01 Ability to monitor automatic operation of the ITM system,

3.8 including

indications of normal, natural, and interrupted circulation of RCS. -3.09 072. 000.K3.02 Knowledge of the effect that a loss of the ARM will have on the

3.5 following

Fuel handling operations. 3.11 068. 000.K6.10 DELETED 3.11 071. 000.K5.03 Knowledge of the following theoretical concepts as they apply 2.9 to the GWS: Sources of hydrogen that could accumulate in the decay tank. Examiners' Handbook 2 BVPS. REV. I i

. Knowledgscnd Abilities Record Form PLANT SYSTEMS i PWR - Senior Reactor Operator - 40% BVPS - Unit 1 Group II Plant Systems-17% J 002 Reactor Coolant System 035 Steam Generator System 006 Emergency Core Cooling System 039 Main and Reheat Steam System 010 Pressurizer Pressure Control System 055 Condenser Air RemovalSystem 011 Pressurizer levelControlSystem 062 AC ElectricalDistribution System i j 012 Reactor Protection System 064 Emergency DieselGenerator System 016 Non-Nuclear Instrumentation System 0 73 Process Radiation Monitoring System 027 Containment Iodine RemovalSystem 075 Circulating Water System 028-H2 Recombiner and Purge Control System 079 Station Air System 029 Containment Purge System. 086 Fire Protection System 033 Spent Fuel PoolCooling System 103 Containment System 034 Fuel Handling Equipment System System # K/A # K/A Topic Rating 3.02 002. 000.K6.03 Knowledge of the applicable performance and design 3.6 attributes of the following RCS components: Reactor vessel levelindication. 3.02 006. 020.K4.03 Knowledge of ECCS design feature (s) and/or interlocks (s) 3.6 which provide for the following: Recirculation flowpath of reactor building sump. 3.02 011. 000.A2.10 Ability to (a) predict the impacts of the following malfunctions 3.6 or operations of the PZR LCS; and (b) based on those predictio.ts, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of PZR levelinstrument - high. 4 3.03 010, 000.A1.04. Ability to predict and/or monitor changes in parameters (to 3.8 prevent exceeding design limits) assxiated with operating the PZR PCS including: Effects of temperature change during solid operation. 3.04 035. 010.A3.01 Ability to monitor automatic operation of the S/G including: 3.9 S/G water level control. 3.05 039, 000.K1.02 Knowledge of the physical connections and/or cause-effect 3.3 relationships between the MS system and the following systems: Atmospheric relief dump valves. 3.05 055. N/A N/A N/A 3.06 027. 000.G10 DELETED

3. % 028.

000.K5.03 - Knowledge of the following theoretical concepts as they apply 3.6 to the HCS: Sources of hydrogen within containment. 3.06 103. 000.A4.01 Ability to manually operate and/or monitor in the control 3.3 room: Flow control, pressure control, and temperature control valves, including pneumatic valve controllers. Examiners' Handbook 3 BVPS. REV.1

Knowledgiand Abilities Record Form PLANT SYSTEMS PWR - Senior Reactor Operator - 40% BVPS - Unit 1 Group II Plant Systems - (Continued) _ System # K/A # K/A Topic Rating 3.07 062. 000.K3.02 Knowledge of the effect that a loss of the ac distribution system 4.4 willhave on the following EDG. 3.07 064. 000.A4.01 Ability to manually operate and/or monitor in the control 4.3 room: Local and remote operation of the EDG. 3,08 079. 000.K1.01 Knowledge of the physical connections and/or cause-effect 3.1 relationships between'the SAS and the following systems: IAS. 3.09 012. 000.K2.01 Knowledge of bus power supplies to the following: RPS 3.7 channels, input relays and slave relays. 3.09 016. 000.K3.03 Knowledge of the effect that a loss of the NNIS will have on 3.1 the following: SDS. 3.09 073. N/A N/A N/A 3.10 075. 000.A2.01 Ability to (a) predict the impacts of the following malfunctions 3.2 or operations of the cire water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of intake structure. 3.11 029. 000.A1.03 DELETED 3.11 033. 000.K4.05 Knowledge of SFPCS design feature (s) and/or interlocks (s) 3.3 which provide for the following: Adequate SDM (boron concentration) 3.11 034.. 000.G07 Knowledge of the purpose and function of major system - 3.0 components and controls. (Fuel Handling) 3.11 086. 000.K5.04 Knowledge of the following theoretical concepts as they apply 3.5 to the FPS: Hazards to personnel as a result of fire type and methods of protection. Examiners' Handbook 4 BVPS. REV,1

... ~.. Knowledgm cnd Abilities Record Form Pl. ANT SYSTEMS PWR - Senior Reactor Operator - 40% BVPS - Unit t I - Group III Plant Systems - 4% 005-ResidualHeat RemovalSystem 041 Steam Dump System 007 Pzr Relief Tank / Quench Tank System 045 Main Tm5>ine Generator 008 Component Cooling System 076 Service Water System 078 Instrument AirSystem System # K/A # K/A Topic Rating 3.04 005. 000.K1.09 Knowledge of the physical connections and/or cause-effect 3.9 relationships between the RHRS and the following systems: RCS. 3.05 041. 020.K3.04 Knowledge of the effect that a loss of the SDS will have on the

3.4 following

Reactor power. 3.05 045. 050.A210 Ability to (a) predict the impacts of the following malfunctions 2.9 or operations of the Main Turbine Generator system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Malfunction of electrohydraulic control. 3.05 076. 000.K1.21 Knowledge of the physical conne:tions and/or cause-effe t 2.9 relationships between the SWS and the following systenu: Standby SWS. 3.06 007. N/A N/A N/A 3.08 078. N/A N/A N/A 3.10 008. 000.A1.04 Ability to predict and/or monitor changes in parameters (to 3.2 prevent exceeding design limits) associated with operating the CC system controls including: Surge tank levels. l i t Examiners' Handbook 5 BVPS REV.1 I

Knowledgn.nd Abilities Record Form i EMERGENCY PLANT EVOLUTIONS PWR - Senior Reactor Operator - 43% BVPS - Uniti l Plant Specific Priorities System # K/A # K/A Topic Rating NONE Group I Emergency and Abnormal Plant Evolutions - 24% i 000001 Continuous Rod Withdrawal 000051 Loss of Condenser Vacuum 000003 Dropped Control Rod 000055 Loss of Offsite and Onsite Power 000005 Inoperable / Stuck Control Rod 000057 Loss of Vital AC ElectricalInst. Bus 000011 Large Break LOCA 000059 Accidental Liquid Rad-Waste Release 000015 RCP Motor Malfunction 000067 Plant Fire on Site 000024 Emergency Boration 000068 Control Room Evacuation 000026 Loss of Component Cooling Water 000069 Loss of Containment Integrity 000029 Anticipated Transient Without Scram 000074 Inadequate Core Cooling 000040 Steam Line Rupture 000076 High Reactor Coolant Activity EA # K/A # K/A Topic Rating 3.01.000.001 EK1.03 Knowledge of the following theoretical concepts as they apply 4.0 to the Continuous Rod Withdrawal emergency task: Relationship of reactivity and reactor power to rod movement. 3.01.000.001 EA2.04 Ability to determine or interpret: Reactor pr.er and it's trend. 4.3 3.01.000.003 EK2.05 Knowledge of the following components: Control rod drive 2.8 power supplies and logic circuits. 3.01.000.005 EK3.05 Knowledge of the bases or reasons for the following: Power 4.2 limits on rod misalignment. 3.01.000.024 EA2.05 Ability to determine or interpret: Amount of boron to add to 3.9 achieve required SDM. 3.01.000.029 EA1.12 Ability to operate and monitor the following: M/G set power 4.0 supply and reactor trip breakers. 3.03.000.011 EK3.12 Knowledge of the bases or teasons for the following: Actions 4.6 contained in EOP for emergency LOCA (large break). 3.03.000.011 EA2.13 Ability to determine or interpret : Difference bet.veen 3.7 overcooling and LOCA indications. 3.04.000.015 EK2.10 Knowledge of the following components: RCP indicators and 2.8 controls. 3.04.000.074 EK1.03 Knowledge of the following theoretical concepts as they apply 4.9 to the inadequate core cooling emergency task: Processes for removing decay heat from the core. l 3.04.000.074 EA1.27 Ability to operate and monitor the following: ECCS valve 4.2 control switches and indicators. e Examiners' Handbook 1 BVPS REV.1

Knowledga cnd Abilities Record Form EMERGENCY Pl. ANT EVOLUTIONS l~ PWR - Senior Reactor Operator-43% L BVPS - Unit 1 '

.3 l

Group I Emergency and Abnormal Plant Evohitions -(Continued) EA # K/A # K/A Topic Rating l-3.05.000.040 Eli!.06 Knowledge of the following theoretical concepts as they apply. 3.8 to the steam line rupture emergency task: High-energy. steam 4 - line break considerations. 3.05.000.040 EA1.04 Ability to operate and monitor the following: Isolation of all 4.3 -{ I steam lines from header. 3.05.000.051-EA2.02 Ability to determine or interpret: Conditions requiring reactor 4.1 and/or turbine trip. 3.06.000.069 EK2.03 Knowledge of the following components: Personnel access 2.9 hatch and emergency access hatch. 3.06.000.% 9 EA2.02 Ability to determine or interpret: Verification of automatic and 4.4 manual means of restoring integrity. 3.07.000.055 . EA1.06 Ability to operate and monitor the following: Restoration of 4.5 power with one EDG. 3.07.000.055 EA2.03 Ability to determine or interpret: Actions necessary to restore 4.7 power. . 3.07.000.057 EK3.01 Knowledge of the bases or reasons for the following: Actions. 4.4 contained in ACP and EOP for loss of vital AC electrical instrument bus. l 3.08.000.068 EA1.21 DELETED 3.10.000.026 EK3.03 Knowledge of the bases or reasons for the following: Guidance 4.2 actions contained in the EOP for Loss of CCW/ nuclear service water. 3.10.000.026 G05 Knowledge of the annunciator alarms and indications, and use 3.4 of the response instructions. 3.11.000.059 G02-Knowledge of which events related to system operation / status 3.9 6 should be reported to outside agencies. 3.11.000.067 EK1.02 Knowledge of the following theoretical concepts as they apply 3.9 to the plant fire on site emergency task: Fire fighting. 3.11.000.076 G07 Ability to explain and apply all system limits and precautions 3.4 (High reactor coolant activity). i !2 Examiners' Handbook 2 l BVPS. REV.1 l

Knowledge and Abilities Record Form EMERGENCY PLANT EVOLUTIONS PWR - Senior Reactor Operator-43% BVPS - Uniti Group 11 Emergency and Abnormal Plant Evolutions - 16% 000007 Reactor Trip 000037 Steam Generator Tube Leak 000008 Pzr Vapor Space Accident 000038 Steam Generator Tube Rupture 000009 Small Break LOCA 000054 Loss of Main Feedwater 000022 Loss of Reactor Coolant Makeup 000058 Loss of DC Power 000025. Loss of Residual Heat RemovalSystem 000060 Accidental Gaseous-Waste Release 000027 Pzr Pressure ControlSystem 000061 Area RMS Alarms Malfunction : 000032 Loss of Source Range NIS 000065 Loss ofInstrument Air 000033 Loss ofIntermediate Range NIS EA # K/A # K/A Topic Rating ' 3.01.000.007 EK3.01 Knowledge of the bases or reasons for the following: Actions 4.6 contained in EOP for reactor trip. 3.02.000.022 EA2.01 Ability to determine or interpret. Whether charging line leak 3.8 exists. 3.03.000.008 EA1.01 Ability to operate and monitor the following: PZR spray block 4.0 valve and PORV block valve. 3.03.000.009 EK1.01 Knowledge of the following theoretical concepts as they apply 4.7 to the small break LOCA emergency task: Natural circulation and cooling, including reflux boiling. 3.03.000.009 EK3.23 Knowledge of the bases or reasons for the following: RCP 4.3 tripping requirements. 3.03.000.027 EA2.03 Ability to determine or interpret: Effects of RCS pressure 3.4 changes on key components in plant. 3.03.000.027 G05 Knowledge of the annunciator alarms and indications, and use 3.3 of the response instructions. (PZR PCS malfunction) 3.03.000.037 G11 Ability to recognize abnormal indications for system operating 4.1 parameters which are entry-level conditions for emergency and abnormal operating procedures. (S/G tube leak) 3.03.000.038 EK3.06 ' Knowledge of the bases or reasons for the following: Actions 4.5 contained in EOP for RCS water inventory balance, S/G tube rupture, and plant shutdown procedures. 3.03.000.038 EA1.16 Ability to operate and monitor the following: S/G atmospheric 4.3 relief valve controllers and indicators. 3.04.000.025 EK1.01 Knowledge of the following theoretical concepts as they apply 4.3 to the loss of RHRS emergency task: Loss of RHRS during all modes of operation. 3.05.000.054 EA1.01 Ability to operate and monitor the following: AFW controls, 4.4 including the use of alternate AFW sources. 1 t Examiners' Handbook 3 i BVPS REV.1

~ ~. Knowledge and Abilities Record Form EMERGENCY PLANT EVOLUTIONS PWR - Senior Reactor Operator - 43% j l BVPS - Unit! Group 11 Emergency and Abnormal Plant Evolution's -(Continued) 4 EA # K/A # K/ A Topic Rating 3.07.000.058 EK1.01 Knowledge of the following theoretical concepts as they apply 3.1 to the Loss of DC power emergency task: Battery charger equipment and instrumentation.' I 3.08.000.065 EA1.04 Ability to operate and monitor the following: Emergency air 3.4 compressor. 1 3.09.000.032 EK2.01 Knowledge of the following components: Power-applies, 3.1 including proper switch positions. (Source range NIS) I 3.09.000.033 EA2.11 Ability to determine or interpret: Loss of compensating 3.4 i voltage. 3.09.000.061 G08 Ability to recognize abnormal indications for system operating 3.3 parameters which are entry-level conditions for technical specifications. 3.11.000.060 EK2.02 Knowledge of the following components: Auxiliary Building 3.1 Ventilation System. j l l l l l1 j Examiners' Handbook 4 l BVPS. REV.1 l

Knowledgn.nd Abilities Record Form EMERGENCY PIANT EVOLUTIONS 1 PWR - Senior Reactor Operator - 43% BVPS - Unit t ) 1 Group III Emergency and Abnormal Plant Evolution's - 3% 000028 PZR Level Malfunction 000056 Loss of Offsite Power 000036 FuelHandling Incident i EA # K/A # K/A Topic Rating 3.02.000.028 EK1.01 Knowledge of the following theoretical concepts as they apply 3.1 ) to the PZR level control malfunction emergency task: PZR reference leak abnormalities. 3.02.000.028 EK2.03 Knowledge of the following components: Controllers and 2.9 positioners. j 3.07.000.056 EK3.02 Knowledge of the bases or reasons for the following: Actions 4.7 contained in EOP for loss of offsite power. 3.11.000.036 G04 Knowledge of bases in technical specifications for limiting 3.8 conditions for operations and safety limits. i 1 l Examiners' Handbook 5 BVPS-REV.1

=. ES-301 Administrative Topics Outline Form ES-301-1 l-l L Examination Level (Circle One): RO / SRO Facility: Beaver Valley - Unit 1 Week ofExamination: 03/17/97 Examiner's Name (print): Administrative Describe method ofevaluation: Topic / Subject-

1. ONE Administrative JPM, OR Description

2. TWO Administrative Questions A.1 Plant Parameter Verification JPM: Perform Manual QPTR Calculation Procedure.

Change JPM: Fill Out An Operating Manual Change Notice A.2 Tagging and Clearances JPM: Verify Clearance A.3 Radiation Work Permits IPM: Review and Sign In on RWP A.4 Emergency Event Classification - JPM: Event Classification i Examiner: ChiefExaminer: i [ ( Examiner Standards 21 of 26 Rev. 7, January 1993 DVPS. REv. I

.~ ES-301 Individual Walk-through Test Outline Form ES-301-2 l Examination Level (Circle One): SRO(I) i Facility: Beaver Valley - Unit One Week of Exammation: 3/17/97 Examiner's Name (print): System / JPM Safety Function Planned Follow-up Questions: K/A/G // Importance // Description

1. RHR - Respond to RHR

a. 005 000 A2.04 2.9/2.9 Pump Trip IV RHR Common Mode Failures l

SD

b. 005 000 K1.04 2.9/3.1 RHR - CVCS Interface

2. PZR PCS - Depressurize

a. 010 000 Kl.03 3.6/3.7 RCS during SGTR III Purpose of Action Taken l,

Faulted

b. 000 038 EK1.03 3.9/4.2 Indication of Head Voiding

3. LRS - Contaminated Drain

a. 068 000 K5.04 3.2/3.5 Monitor Alarm XI Basis for RMS Setpoint

b. 068 000 A4.03 3.9/3.8 Conditions Requiring Discharge Isolation

4. CVCS - Verify SHUTDOWN

a. 004 000 K5.20 3.6/3.7 MARGN I

Possible Causes for Reduction in SDM Faulted

b. 004 000 Kl.06 3.1/3.1 VCT Response to Emergency Boration

5. AC DIST-Energize Stub

a. 062 000 K4.07 2.7/3.1 Busses VII Loads Energized by Action Taken b 062 000 K4.02 2.5/2.7 Basis for De energizing Stub Busses

6. CAS - St1rt Containment

a. 078 000 K3.03 3.0/3.4

~ Instrument Air Corrpresser VIII Loss of Containment Instrument Air Compressors

b. 000 065 EA2.08 2.9/3.3 Pressure Control on Loss ofinstrument Air

7. FW - Operate Feedwater

a. 059 000 A4.113.1/3.3 Following SG H1 HIlevel y

Reason for Transfer to AFW ]

b. 059 000 A3.06 3.2/3.3 Reason for Reactor Trip

8. ECCS - Emergency Start

a. 006 000 K2.013.6/3.9 HHSI Pump on Bus I AE II Basis for Selecting Recovery Path

Emer,

b. 006 000 K1.08 3.6/3.9 Basis for Isolating Seal Injection Flow

9. AFW Warm Slow Start

a. 061000 K4.08 2.7/2.9 Turbine Driven AFP y

AFW Recirculation Flow Paths

b. 061000 A2.04 3.4/3.8 Loss of Oil to AFP Turbine

10. CCR - Place CCR Heat

a. 008 000 Al.02 2.9/3.1 i

Exchanger in Service X System Response to Action Taken RCA Entry

b. 008 000 Kl.02 3.3/3.4 l

CCR High Temperature 4 Examiner: ChiefExaminer: 4 Examiner Standards 22 of 26 Rev. 7, January 1993 o BVPS REV. I

l ES-301 - Individual Walk-through Test Outline Form ES-301-2 ~Exammation imel(Circle One): SRO(U) Facility: Beaver Valley'- Unit One Week of Exanunation: 3/17/97 l, Examiner's Name (print): System / JPM Safety Function Planned Follow-up Questions: K/A/G // Importance // Description

1. ECCS - Verify SI Flow Is Not

a. 000 009 EKl.02 3.5/4.2 Required.

II Subcooling. ESF/Fauned

b. 006 020 K6.012.8/3.1 Low Head Injection Pressure.

2. RHR - Perform RHR

a. 005 000 K4.07 3.2/3.5 Pump Performance Test IV RHR Pressure Protection SD

b. 005 000 A1.013.5/3.6 Heat up Rate if RHR is Lost

3. WGDS - Respond to Waste

a. 071000 G7 2.5/2.9 Gas Monitor Annunciator XI Waste Gas System Oxygen Control

{ I Faulted

b. 071000 K4.012.6/3.0 Waste Gas Decay Tank Pressure I

4. CCR-Establish RCP

a. 008 000 Kl.04 3.3/3.3 I

CCR Cooling X ThermalBarrier Rupture. EOP/RCA

b. 008 000 A3.013.2/3.1 CCR Response to JPM Actions

5. Lockout of Affected 4160V

a. 064 000 K4.02 3.9/4.2 Emergency Dus I

Diesel Generator Control

b. 062 000 K4.05 2.7/3.2 Breaker Interlocks 6.

a. b. 7. a. b. 8, a. b. 9. a. b. 10. a. b. I Examiner: ChiefExaminer: Examiner Standards 22 of 26 Rev. 7, January 1993 I BVPS. REV.1 l 1 e

.. ~. p ES-301 Scenario Events Form ES-301-3 4 Simulation Facility: Scenario No.: 97-1 t. Beaver Valley Power Station - Unit 1 Examiners: Applicants: -t l Initial Conditions: 12% power, BOL,1437 ppm baron, XE free, plant startup in progress. Turnover: 1. FW P-3A, MDAFW pump is on clearance for motor replacement. Due to be returned next shift. 2. SA-C-1B, station air compressor on clearance for compressor failure. Not due to be retumed for 2 days. i 3i Rod control is in manual, CBD is at 140 steps. 4. FW-P-4, the dedicated AFW pump, is on clearance for motor bearmg replacement. Not due to be returned for 2 days. Fire Watch for AFW room is required, once per hour. t. 5. Main generator is warmed up and rolling at 1800 rpm, ready to be loaded. 6. SGWLC is in AUTO on main feed on bypass valves. 7. OM $2.4.A, increasing power from 5%, is completed through step 69. Crew starts at Step 70 and continues unit startup. 8. FW 150A&B are failed open and CN 101 is in Manual and closed. 9. Boron is 1437 ppm. Event Malf. Event Event No. No. Type

• Description 1

MALF I N36 IR instrument power fuse blows. Remove instrument from service NIS7B IAW AOP 1.2.lB. Take action IAW TS 3.3.1.1 act. 3.- 2 R Raise power from 12% to sync Main generator. 3 N Sync Main Generator. 4 MAIJ I l PT-MS-464, main steam header press transmitter, fails high causing MSS 11 steam dumps to fail open.~ 5 MALF C 1 steam dump valve fails / sticks open with a steamline break outside of MSS 8 containment. 6 MAIS C Failure of CH-P IB to start. SISSA 7 (INII51 1 Failure of Auto Main Steamline Isolation. M Low steamline pressure Sl; Scenario termmation criteria - S1 termmated/ BIT isolated and letdown is in service. E 0... ES-1.1 (N)ormal,. (R)eactivity, (I)nstrument, (C)omponent, (M)ajor ( L Examiner: ChiefExaminer: h I 1 Exammer Standards 23 of 26 Rev. 7, January 1993 BVPS Rev.O i

I ES-301 Scenario Events Form ES-301-3 Simulation Facility: Scenario No.: 97-2 l Beaver Valley Power Station Unit 1 i Examiners: Applicants: Initial Conditions: 75% power. MOL,722 ppm Boron. Power reduction in progress due to high vibes on FW-P-lB. Main feedwater pump. Xe concentration is building in @ 250 pcm/hr. De shutdown from 100% was perfonned at 1%/ min. and ~25 minutes to perfonn. Turnover: 1. FW-P-3 A, MDAFW pump is on clearance for motor replacement. Due to be returned next shift. 2. SA-C-1B, station air compressor on clearance for compressor failure. Not due to be retumed f'or 2 days. 3. Rod control is in automatic, CBD is at 175 steps. 4. FW P-4, the dedicated AFW pump, is ou clearance for motor bearing replacement. Not due to be retumed for 2 days. Fire Watch required in AFW room once an hour. 5. Management directs power to be reduced at 1%/ min and remove FW-P 1B from service as soon as possible, using IOM-52.B. Load Follow. Event Malf. Event Event No. Ng Type

• Description 1

R Perform power reduction of greater than 5% IAW AOP 51.1, Emergency Shutdown. 2 N Remove FW-P-1B, main feed pump, from service at approximately 60% power. 3 MALF 1 PT-RC-444, RCS pressure transmitter, fails high causing PCV-RC-PRS 8D 455C, PZR PORV, to open. Crews responds to failure IAW IFM 6.4.IF. Possible entry into TS 3.2.5 - DNB Parameters due to low RCS pressure. 4 MALF C Upon clomre, PCV-RC-455C, PZR PORV, develops a leak. Block PRS 3A vahr is to be manually closed within I hour IAW TS 3.4.11 act. a. =5% 5 MALF C 2* main feed pump trip /Rx trip on low SGWL.

FWM1A, C

FW-P-2, teny turbine AFP, fails to start due to mechanical failure of the FWMi1C trip valve. 6 MALF C he DF 4KV Emergency bus trips when the turbine generator output EPS4F breakers open. EPSIIB ne #2 emergency DO starts and loads 60 seconds after the DF bus is M lost, then trips and won't restart. His causes a loss of the DF 4KV emergency bus and FW-P-3B, motor driven AFW pump, resulting in a Loss of Heat Sink condition. Scenario temunation criteria - Feed flow established from the condensate system. E4... ES-0.1... IR-H.1. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor l Examiner: ChiefExaminer; i Examiner Standards 23 of 26 Rev. 7, January 1993 DVPS-Rev 0

ESo301 Scenario Events Form ES-301-3 Simulation Facility. Scenario No.: 97-3 Beswr Valley Power Station - Unit 1 Examiners: Applicants: Initial Conditions: i 100% power, BOL,938 pprn boron, equilibrium conditions. Turnover: 1. FW-P-3A, MDAFW pump is on clearance for motor replacement,8 hours ago. Due to be returned next I shift.-TS 3.7.1.2. 2 SA-C 18, station air compressor on clearance for compressor failure. Not due to be retumed for 2 days. 3. Rod control is in automatic, Control Bank D is at 227 steps. 4. FW P-4, the dedicated AFW pump, is on clearance for motor bearing replacement. Not due to be returned for 2 days. Fire Watch for AFW room is required, once per hour. 5. Maintain 100% power. Event Malf. Event Event No. No. Type

• Description 1-MALF I

N44 powet range detector fails high. Auto Rod control drives rods in. NIS3D C :w removes channel from service LAW AOP 1.2.1C. 2 MALF C During rod insertion, rod F-6 (control bank rod) drops due to blown fuse CRF4 on stationary gripper. No Rx trip occurs 3 N/R Power reduction initiated in response to dropped rod IAW AOP 1.1.5. 4 MALF C Afict N44 is removed ' rom service, Loss of 120 vac vital bus III causing f EPS6C a Rx trip on high PR flux. 5 MALF M Rx trip causes a 350 gpm SGTR on IB S/O. Scenario termination criteria RCS3B - SI termmated and second charging pump stopped in E 3. E-0... E-3. i l i (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Examiner: ChiefExaminer: 1 l l-i E I l ' Examiner Standards 23 of 26 Rev. 7, January 1993 BVPS Rev.0

t - ES 301 Scenario Events Form ES-301-3 Simulation Facility: Scenario No.: 97-4 Beaver Valley Power Station Unit 1 Examiners: Applicants l Initial Conditions: i 75% power, EOL,161 ppm boron, Xenon increasing @ 200pcm/hr due to power decrease from 100% over the past 2 hours. Turnover: i l 1. FW-P-3A, MDAFW pump is on clearance for motor replacement. Due to be returned next shift. b 2. SA-C-18, station air compressor on clearance for compressor failure. Not due to be returned for 2 days. 3. Rod control is in automatic, CBD @ 206 steps, boron = 161 ppm, burnup = 12,000 MWD /MRI 4. FW-P-4, the dedicated AFW pump, is on clearance for motor bearing replacement. Not due to be returned for 2 days. Fire Watch required in AFW room once an hour. 5. N42 Power Range channel inoperable and removed from service IAW AOP 1.2.1C. 6. Crew is to maintain current power level pending request of System Operator. - Event Malf. No. Event Event No. Type' Description 1 MALF I I'T-RC-455, RCS pressure protection channel I fails high. With N42

NIS3B, already inoperable TS 3.0.3 should be entered due to 2/3 OT/DT channels PRS 8A..

beinginop rable. 2 N Power reduction due to T3 3.0.3. 3 R' Power reduction of at least 5%. 4 MALF C Excessive #1 seal leak off of 20 gpm on IRC-P-1B, B loop RCP, requiring RCS5B-a Rx trip IAW ARP 7.4.ABE. 5 MALF M A'IV/S that causes a PZR PORV to open. Local Rx trip to occur at t = 2 CRF12 minutes after requested from crew. A&B 6 MALF C PCV-RC-455C, PZR PORV, leaks. PRS 3A = 50 % i l 7 OVRD C MOV-RC-535, PZR PORV block valve, fails to close resulting in a PZR M vapor space leak /SBLOCA. Scenario termination criteria: Transition to ES 1.2. l E-0... FRS-1... E.O... E-1... ES-1.2. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor l-Examiner: ChiefExaminer: Exammer Standards 23 of 26 Rev 7, January 1993 c l BVPs.Rev 0

. ~. - - ES-301 Scenario Events Form ES-301-3 Simulation Facility: . Scenario No.: 97-5 Beaver Valley Power Station - Unit 1 Examiners: Applicants: Initial Conditions: .J Rx S/U in progress, Rx power is at IE-8 amps, Xenon free, EOL conditions. Turnover: 4 l. SA C-1B, station air compressor on clearance for compressor failure. Not due to be returned for 2 days. 2. Rod controlis in manual. 3. FW-P-4, the dedicated AFW pump, is on clearance for motor bearing replacement. Not due to be returned for 2 days. Fire Watch required once per hour in AFW room. i 4. OM 50.4.D, Plant startup, at step C.I. Crew is to continue Unit startup. 5. FCV-FW-150A&B are failed open and CN 101 is in Manual and shut. Event Malf. Event Event No. No. Type' Description l 1 MALF 1 FT-RC-414, I A loop flow transmitter, fails low. Remove RCS19A instrument from sen.ce IAW IFM 6.4.lF and TS 3.3.1.1 act. 7. ] 2 N Conduct plant startrp. 3 R Raise Rx power to the POAH. i 4 MALF C Uncontrolled Rod withdrawal requiring a manual Rx trip. CRF5B 5 MALF M On Rx trip the C S/G becomes faulted inside containment and is not MSSIC isolable from the S/G. 6' INH 49 C Failure of auto actuation of both trains of CIA, requiring manual actuation of CIA. Scenario termination criteria - SI terminated / BIT isolated. E-0... E-2... E-1... ES-1.1. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Examiner: . j ChiefExaminer: Examiner Standards 23 of 26 Rev. 7, January 1993 BVPS Rev0

ES-301 - Scenario Events Form ES-301-3 l Simulation Facility: Scenario No.: 97 6 l l ; Beaver Valley Power Station - Unit I _ l Examiners: Applicants: i Initial' Conditions: i. i 74% power, MOL Xenon increasing @ 200 pcm/hr Tumover: 1. FW-P 3A, MDAFW pump is on clearance for motor replacement. Due to be retumed next shift. 2. SA-C 1B, station air compressor on clearance for compressor failure. Not due to be returned for 2 days. 3. Rod controlis in automatic. 4. FW-P-4, the dedicated AFW pump, is on clearance for motor bearing replacement. Not due to be retumed for 2 days. Fire Watch required once per hour in AFW room. 5. lead reduction to 75% power has just been completed, using load Follow procedure,10M-52.B from 100% at 12% per hour. 6. He boron concentration is 722 ppm. Control Bank D is at 175 steps. He core age is 8000 MWD /MTU. 7. Maintain current power level per System Operators request. j Event Malf. No. Event Event No. Type

• Description 1

MALF C S/G tube leak of 30 GPD on the B S/O, requiring a plant shutdown. RCS3B N/R Power reduction to comply with primary to secondary leakage AOP { 1.6.4. 2 Power reduction of at least 5%. 3 MALF I Irf 446. Turbine first stage pressure transmitter fails high disabling the TUR18A steam dumps in the Tave load reject mode, and auto rod control. Crew -600 responds IAW IFM procedure to remove channel from senice. 4 OVRD C Train B Rx trip breaker fails closed causing steam dumps to remain in PO711a=1 the Tave load reject mode. PO711b=2 i MALF MSS 7 5 MALF M B S/0 tube leak that tums into a 500 gpm SGTR during the power RCS3B set reduction. to 500 gpm 6 OVRD C S/G safety valve on B S/G opens on R,c trip due to the loss of the steam CBSTSP dumps and fails open; this results in a ruptured / faulted S/G. Scenario 183=0 termination criteria - RCS cool down initi:.ted IAW ECA-3.1. j E-0... E-2... E 3... ECA 3.1... l i i (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Examiner: ChiefExaminer: i i i Examiner Standards 23 of 26 Rev. 7, January 1993 BVPS -Rev 0

RTL #A5.620.H DUQUESNE LIGHT COMPANY Volume 3 Nuclear Power Division Procedure 5-5 Training Administration Manual Figure 5-5.1 Revision 10 Page1of1 WRITTEN EXAMINATION COVER SHEET PROGRAM: Initial Licensed Operator Training l l CLASS NUMBER: 1 LOT-2

SUBJECT:

Senior Reactor Operator. March 1997 - NRC Initial Licensed Operator Exam. l I " l By this signature,I state that all of the work done on this examination is my own. I have neither given nor received aid. SIGNATURE DATE 'NAME Answer Key DLC EMP # (Please Print) COMPANY (if other than DLC) POSSIBLE POINTS 100 SCORE Instructor Initials l TRAINING DIRECTOR / SUPERVISOR j PREPARED Bs David C. Gibson APPROVAL f SIGNATURE i Date

k 1 'ES-402 Policies and Guidelines 4 for Taking NRC Written Examinations 4 1. theating on the examination will result in a denial of your application and could result in more severe penalties. 2. After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. 3. To pass the examination, you must achieve a grade of 80 percent or greater. 4 4. The point value for_ each question is indicated in parentheses after the question number. 5. There is a time limit of 4 hours for completing the examination. 6. Use only black ink or dark pencil to ensure legible copies. 7. Print your name-in the blank provided on the examination cover sheet and the answer sheet. 8. Mark your answers on the answer sheet provided and do not leave any question blank. 9. If the intent of a question is unclear, ask questions of the examiner only. 10. Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheating. 11. When you complete the examination, assemble a package including the examination questions, examination aids, and answer sheets and give it 4 to the examiner or proctor. Remember to sign the statement on the examination cover sheet. 12. After you have turned in your examination, leave the examination area as defined by the examiner. i ' Examiner Standards 5 of 6 Rev. 7, January 1993

~-..,. - -. .. -.,..... ~.. ~ _ ~... _,.... ~ -.. -... -. -....., g Question Number 1 - L f' Select the stateinent that. describes the safety equipment REQUIRED to be worn l when handling sodium hydroxide ~(NaOH). l i A. Safety glasses and appropriate gloves ONLY. !E B.': Safety glasses, appropriate gloves and a paper surgical mask. I i C.. Goggles /faceshield ONLY. D. Goggles /faceshield, appropriate gloves and impervious clothing. -) i i POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: .D. -s l i I: i i i l

REFERENCES:

AOP 1/2 53.C.4A.75.7, Att. 1, page 12 - Issue 1A, Rev. 6. 1LP-SOS-53C.1 OBJECTIVE: 4: NUMBER: 1-97-093, R-0117 i JTA #: i K/A et 194001.K1.10 K/A IMPORTANCE: 3.0/3.3 4 i L - Rev.! l l

- - ~. - - ~-. -. ~....- -. ~... (hiestion Numtw:r 2' .An MOV designated as "VDM", (valve drifts manually), has been closed manually ~ for a: clearance. ^ Which of the following actions is requ1 red to prevent this valve from drifting l 'due to system pressure effects? A. Ensure that the MOV manual operating lever remains in the fully " ENGAGED" position following manual closure. B. Manually return the MOV manual operating lever to the " DISENGAGED" position following manual closure. C. Restore power to the MOV following manual closure and take the control switch'to the."0 PEN" position to re-engage the valve motor operator. D. Restore power to the MOV following manual closure and take the control switch to the " CLOSED" position to re-engage the valve motor operator. i l l t 1 l l. POINTS: 1.00 . TIME ALOTTMENT: udnutes l ANSWER: D. i i i l l \\; I l i i I l

REFERENCES:

NPDAP 3.4, Rev. 6, Page 12, item o. ILP-SQS-48.1 OBJECTIVE: 25 NUMBER: 1-97-094, M-0121 JTA 9: K/A ft 194001.Kl.01 K/A IMPORTANCE: 3.6/3.7 Rev.1

..__.._m.._..,_ .. _ ~. _ _. _ l Question Number 3: lJ Refer to Unit 1 Technical Specification 3.4.7, Chemistry, and 3.4.8 Specific-I Activity, to answer.the following question. J-l . Given the followings i RCS pressure is 225 psig. e ~. RCS temperature is at 210'F. e p Chemistry has just called in the most recent RCS sample results. e Which of the following would exceed the Technical Specification limit for the RCS transient chemistry specifications? j l A. Fluoride = 1.0 ppm. l B. ' Chloride = 1. 6 ppm. 1 C. Dose Equivalent Iodine.- 131 = 0.5 uCi/gm. D. Dissolved Oxygen = 1.1 ppm. I i l 1 1 1 1 l POINTS: 1.00 TIME ALOTTMENT: minutes i ANSWER: B. ] i i 1 l I~

REFERENCES:

Unit 1 Technical Specification 3.4.7, Table 3.4-1, Original. 1LP-SQS-CHEM-19 OBJECTIVE: 4 NUMBER: 1-97-095, M-0125 JTA 9: i

j

'K/A 0 194001.A1.14 K/A IMPORTANCE: 2.5/2.9 4 ) Rev.I 7 w

Question Number 4 l An operator is required to be continuously stationed at a valve in a confined i area for 30 minutes. Radioactive material on/in this valve is exposing the -l operator to 200 mR/hr. Three feet behind the operator is another valve that emits 100 mR/hr at 25 cm. Which of the following is applicable to this situation? A. The valve behind the operator should be labeled a " Hot Spot" and the area posted as a Radiation Area. B. The operator will exceed their 10 CFR 20 dose lLaits. C. The area should be posted as an High Radiation Area and the operator should have an integrating dose rate meter. D. An HP technician should be present to monitor the contamination in the j room with a portable ion chamber while the operator is stationed at the valve. l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: .C.

REFERENCES:

10CFR20.003 and Unit 1 Tech Specs 6.12 Amendment No. 188. 1/2LP-RC-02 Rev. 17 OBJECTIVE: 4-9 NUMBER: 1-97-096, M-0165 JTA #: K/A #: 194001.Kl.03 K/A IMPORTANCE: 2.8/3.4 Rev.1 l

j.. P Qtx:stion Numtxtr 5 l' [ Which of the following is NOT correct regarding a confined space entry? A.' A minimum of two qualified individuals shall enter the confined space, I one of which will act.as only a safet/ man. B. A method of' communication shall be established to maintain contact with personnel within the confined space. l C. An SCBA for emergency use shall be located near the entrance of a 1 confined space when certain tasks are being performed within. i D. For conditions where an SCBA used by rescuers may be impracticable, the ventilation flow can be increased as an additional precaution. i i I i i POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. -i i

REFERENCES:

NGAM 3.7 l 1/2LP-GM-6040 OBJECTIVE: 7 NUMBER: 1-97-097, M-0173 JTA 4: { K/A #: 194001.Kl.13 K/A IMPORTANCE: 3.3/3.6 l Rev.1

= i Question Number 6 - Listed below in random order are steps regarding the preparation of the r Valve / Switching Procedure Form. The NCO checking the Form sh'll check the clearance point (s)- (using -1. a L -. control room prints) to ensure the procedure is proper. I~ l

2. The ANSS or NSS approving the Form shall ensure the equipment is being cleared properly and that required equipment.is not made inoperable.

3. The operator performing the switching will present the Form to NCO.

4. The operator completing the Form shall fill in the pertinent information (i.e., Clearance 9, Clearance Point, Tag Type, etc.).

(. l Which of the following groupa is in the proper order? A. 4, 3, 1, 2 B. 2,4,3,1 C. 4, 2, 1, 3, D. 1, 2, 3, 4, I i l l l l l POINTS: 1.00 TIME ALOTTMENT: minutes i ANSWER: A. i. i l l I'

REFERENCES:

SAP 41, pages 18 and 19, Rev. 6. 1LP-SQS-48.1 OBJECTIVE: 30 NUMBER: 1-97-098, M-0181 JTA 9: K/A 9: 194001.Kl.02 K/A IMPORTANCE: 3.7/4.1 4 Rev.I 1> f

... ~..,,... -. -... ... ~., ~.... -.. ~. ~ - ?' i Question Number 7 - What is the minimum positive threshold indication'of personal contamination when ~1 - using a frisker?. 1 i l A. 100 cpm total. j B. 100 cpm > background. 1 C. 300. cpm total. D. 300 cpm > background. t l I POINTS: 1.00' TIME ALOTTMENT: -minutes ANSWER: B. l l r. ~

REFERENCES:

Rad Con' Manual' Unit 162 - Issue 4, Rev. 11, Chap 1 Part III Page 3. l 1/2LP-RC-02 Rev. 17 OBJECTIVE: 6-4' NUMBER: 1-97-099, M-0382 JTA # K/A #: 194001.K1.04 K/A IMPORTANCE: 3.3/3.5 l \\ I-I -- Rev.1 '.

. - ~. . ~. ~ - ~... -,.,,, !i QuestionNwber 8 'l Who must approve an on the spot change (OMCN) to an operating procedure? l - l.. A. The respective department supervisor of the individual requesting the change ONLY. B. TWO members of the plant management staff, ONE of whom must hold an-SRO license for the affected Unit. C. ONE me:aber of the plant management staff who holds an SRO license' for the affected Unit. D. Both the respective department supervisor and an individual with an ( SRO license on either Unit. i l l 1. i l 1 .i POINTS: 1.00 TIME ALOTTHENT: minutes ) i ANSWER: B. l l. I I I i

REFERENCES:

1/20M48.2.B - Issue 3, Rev. 13. i 1LP-SQS-48.1 OBJECTIVE: 8 NUMBER: 1-91-100, M-0385 i JTA 4: l l'. K/A 42. 194001.A1.01 K/A IMPORTANCE: 3.3/3.4 ' Rev, I l:. l l.

._ ~.. ~,. (hiestion Nr aber 9 When performing a station startup in accordance with ON Chapter 50 " STATION ~ ~STARTUP," steps marked by a filled diamond sign indicate that the step. !.Il. s A. umy.be skipped at the discretion of the NSS. l B. may be omitted by the NSS provided the UOM initials the omitted step. -C. cannot be omitted but may be started out of sequence. 1 D. cannot be omitted and must be performed in the specified sequence. i I i I l\\ POI:1TS: 1.00 TIME ALOTTMENT: . minutes ANSWER: C. I i l REFERENCES : 10M48.2.C - Issue 3, Rev. 13. ILP-SQS-48.1 OBJECTIVE: 10 NUMBER: 1-97-101, M-0386 f JTA 4:

4 K/A 4

194001.A1.02 K/A IMPORTANCE:

4.1/3.9 t t . Rev.I 1 i

m-Question Number 10 'l Which of the following.is the reason why a nitrogen blanket is maintained on the I Pressurizer Relief Tank (PRT)? i [ A. - To reduce the concentration of oxygen in the reactor coolant system. 'B. To provide a driving force for the sampling system. _l l C.' To reduce the potential for an explosive mixture of hydrogen and oxygen. l D. - To reduce the amount. of water required. in the PRT to quench t . pressurizer relief valve discharge, j l l i i i 1 l l i

POINTS:

1.00 TIME ALOTTMENT: minutes ANSWER: C. l 1 l l L il' l REn.RENCES : 10M1.6.1.C - Issue 4, Rev. 1. l4 ILP-SQS-6.4 OBJECTIVE: 7 NUMBER: 1-97-102, M-0661 e JTA 4: K/A 4: 194001.K1.15 K/A IMPORTANCE: 3.4/3.8 Rev.1'

QuestionNumber 11 Heat Stress countermeasures are implemented to reduce heat stress exposure and may enable a person to work beyond their ' Total Action Time' without experiencing the onset of heat stress symptoms. Which group of items below are all heat stress countermeasures? 1. Increased Ventilation / Air Motion. 2. Increased Clothing Requirements. 3. Reduced Work Scope. 4. Reduced Work Demand / Exertion Levels. A. 1, 2 and 3. B. 1, 3 and 4. C. 1, 2 and 4. D. 2, 3 and 4. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B. l l l l

REFERENCES:

NPDAP 3.8, Page 3, Rev. 2 1/2LP-RC-02 Rev.17 OBJECTIVE: 6-2 NUMBER: 1-97-103, New JTA 9: K/A #: 194001.Kl.08 K/A IMPORTANCE: 3.5/3.4 Rev.I

Question Numix:r 12 Refer to the EPP/IP handout to answer the following question. Which of the following criterion would require the Technical Support Center (TSC) to be activated? A. A pressurizer PORV fails to close following a valid open signal with NO Safety Injection actuation required and the associated PORV Block Valve operable. B. The Rx fails to trip when an automatic trip signal is generated but trips when activated manually at the benchboard. l C. A simultaneous loss of ALL annunciators, sequence of events recorders, and SPDS for >15 minutes in Mode 5. D. A report by plant personnel of a chlorine gas release within the site perimeter that renders the chlorine building inaccessible, l l 1 1 1 POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B.

REFERENCES:

EPP Tab 2.2. 1/2LP-EPP-57.81 OBJECTIVE: 1 & 11 NUMBER: 1-97-104, M-6280 JTA 6: K/A 8: 194001.A1.16 K/A IMPORTANCE: 3.1/4.4 Rev.I i

.,s ~. .-_,_-._.-.__m .--._m.. l. I-i l-QuestionNumber 13 r Which of the following approved portable fire fighting equipment should be used I j. ' to combat a flammable liquid fire? ) i 1. Dry Chemical. 2. Water. 3. CO2 4. Foam. l l' f I 'A. 1, 2 or 3. l B. 1,.3 or 4. C. 1,'2 or 4. I D. 2, 3 or 4. 1 I POINTS: 1.00 TIME ALOTTMENT: mimtes ' ANSWER: B. i-

REFERENCES:

1/2.56A.4.H - Issue 3, Rev. 1, and LP 9339. 1LP-SQS-9339 OBJECTIVE: 7 NUMBER: 1-97-105, New - JTA 6: !.. I. K/A 9: 294001.K1.16 K/A IMPORTANCE: 3.5/4.2 o-1 Rev.1

QuestionNumber 14 Who should be contacted if a person is sited without a security badge or if

someone is attempting to enter a Vital Area without using the key card reader?

'{' A. Health Physics field office. B. Your immediate Supervisor. C. The Central Alarm Station (CAS) or a Security Officer. D. The Emergency Control Center (ECC). l l POINTS: 1.00 TIME ALOTTMENT: rinutes I ANSWER: C.

REFERENCES:

1/2LP-RC-02-Rev. 17 1/2LP-RC-02 Rev.17 OBJECTIVE: 5-12 NUMBER: 1-97-106, M-PAT Exam 96-2, #25 JTA #1 1 K/A 6: 194001.Kl.05-K/A IMPORTANCE: 3.1/3.4 i Rev.I 'I i I

' Question Number 15 - Technical Specification Surveillance 4.8.1.1.2 requires in part, that each Emergency Diesel Generator (EDG) be started from ambient conditions at least i once per 31 days, on a STAGGERED TEST BASIS. Which of the following would be in . compliance with this STAGGERED. TEST BASIS requirement? Start #1 EDG on the (1) day of each 31 day period and start #2 EDG on the' (2) day of the (3) 31 day period. (1) (2) (3) A. 1** 8 '" same B. l 15'" same C. l'* 1** next th D. 8 3gst .gggg l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B. l-l l L.

REFERENCES:

Technical Specification Definitions Amendment No. 192. ILP-SOS-TS OBJECTIVE: 3 NUMBER: 1-97-107, New JTA'#: ~ K/A #: 194001.K1.17 K/A IMPORTANCE: 2.1/2.5 .g Rev,1 J I' [. l.

____.~._ _ _ _ _.. _. _ _ _ _. _ _. _ _ _ _. _ .QuestionNumber 16 f Refer to the Unit 1 Curve Book,' Cycle 12, Issue 12 Rev. 1, and the WAG Tables to answer the following question. 1 4 4 ~ Given the followings i Reactor power has just been reduced from 100% to 60%. 4 Control Bank D rods (CBD) are at 122 steps. T.v. -. T.. t error-is 0'F. Borou concentration is 375 ppm. Reactor Engineering reports that over the next hour, Xenon will add a ( negative 115 pcm. a 1 'e core burn-up is 11,500 MWD /MTU. j Over the next hour, how much boric. acid or primary grade water must be add'd to e the RCS in order to withdraw CBD to the fully withdrawn. position AND keep Rx-power'and T... constant? A. 299.0 gallons of Boric Acid. B. 361.5 gallons of Boric Acid, ] C. 5,866 gallons of Primary Water. l D. 7,190 gallons of Primary Water. i l i Y POINTS: 1.00-TIME %OTTMENT: minutes k ANSWER: B. CB10[h - Rod motion will add +550 pcm. j. Xenon will add -115 pcm. Boron must add -435pcm. E CB28 - Boron worth is -8.4pcm/ ppm. e Boron conc. must be increased by 52 ppm for a total of 427 ppm. e T.,. = 5 65'F a t 60 %. powe r. R WAG table 565'F, iteration method: 2 e From 375 ppru to 425 ppm - 347.57 gal of BA. From 425 ppm to 427 ppm = 13.96 gal of BA. Total of 361.5 ga)* ons of boric acid must be added. o h h Unit - 1 Curve book, CB-10f, CB-28, WAG Tables 560*F.

REFERENCES:

j ILP-SQS-RT-6 OBJECTIVE: 15 NUMBER: 1-97-108, M-LRT 2.3.4 JTA 6: K/A'4: 194001.A1.08 K/A IMPORTANCE: 2.6/3.1 1i I i Rev.1 i.

._m-QuestionNumber 17 Refer to UFSAR Figure.7.2-1, Instrumentation and Control System Logic Diagram, Sheet 1 and 13 to answer the following question. I Which of the following describes the control logic necessary to OPEN the feedwater regulating BYPASS valves, FCV-1FW-479, 489, & 4997 Following a Safety Injection (SI) signal. A. ONLY the SI signal needs to be Reset. B. ONLY the Feedwater Interlock signal needs to be Reset. C. BOTH the SI signal AND the Feedwater Interlock signals need to be Reset. D. The.feedwater regulating BYPASS, valves CANNOT be OPENED with a standing SI signal even if it is Reset. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B. l (

REFERENCES:

UFSAR. Figure 7.2-1, Sheet 13, Rev. 10. 1LP-SQS-24.1 OBJECTIVE: 4 NUMBER: 1-97-109, New. ) JTA 6: = i-K/A 8: 194001.A1.07 K/A IMPORTANCE: 2.5/3.2 l 1 Rev.I

Qulsti:nNumber 18-1he plant is at 754 power with the rod bank selector switch in manual and all other systems in automatic. What would be the effect on the Rod Insertion- -- f; Limits (RIL) and the Shutdown Margin (SDM) if the main generator electrical -output was raised by 36, and the only other operator action was to restore T.,. to program using the Boration/ Dilution controls? RIL will (1) and SDM will-(2) (1) (2) A.

raise, lower.

B.

lower, raise.

C.

lower, lower.

D. raise; raise. POINTS: '1.00 TIME ALOTTMENT: minutes ANSWER: A. lj l e

REFERENCES:

Tech. Spec. Definitions ) l ILP-SQS-1.4 . OBJECTIVE: 6 ' NUMBER: 1-97-001, M-0395 ] -JTA #: 0010110104 K/A 8: 3.01.001.000.K5.08 (001K5.04) K/A IMPORTANCE: 3.9/4.4 { Rev. l.

1 Questi:nNumber 19 Given the.following: . j Control Rod D12,'a control bank "C" Group 1 rod, has fallen into the e 4 core due to an equipment failure. 1 The equipment failure has been corrected and all retests are completed l satisfactory. The dropped rod recovery is in progress per AOP-1.1.5, " Dropped RCCA." All applicable switches are in their correct position for the rod i e recovery. l AOP step 1.1.5.17.a directs the operator to " Anticipate rod control system urgent failure alarm." i Determine the cause and effect of the anticipated ROD CONTROL SYSTEM URGENT ALARM, A4-105. T ROD CONTROL SYSTEM URGENT ALARM is caused by a A. Logic cabinet failure and ALL rod motion will be inhibited. B. Logic cabinet failure and ONLY those rods aligned to power cabinet 1AC will move. C. Power cabinet 1AC failure and ONLY those rods aligned to power cab' net. .i i 2AC will move. D. Power cabinet 2AC failure and ONLY those rods aligned to power cabinet 1AC will move. POINTS: 1.00 TIME ALOTTHENT: minutes ANSWER: D.

REFERENCES:

10M-1.1.D - Issue 4 Rev. 1, AOP 1.1.5 - Issue 3A Rev. 5 ILP-SQS-1.3 OBJECTIVE: 10 NUMBER: 1-97-002, M-0055 JTA 8: K/A 8: 3.01.001.050.A2.01 (001A2.14) K/A IMPORTANCE: 3.7/3.9 Rev.I

Questi:n Number 20 1 Given the following: 4 [ The plant has been at.100% power for 20 days. e All systems are in their at power, NSA configurations. e RCS.T.v. is stable at 577'F. I There have been NO sump pump runs.for the past 24 hours. e Pressurizer level and pressure are stable at their program values. e e Charging flow is 89 gpm. Letdown flow is 104 gpm. e RCP seal injection flow is 24 gpm. e Total RCP seal return flow 9 gpm. e There have been three auto make-ups to the VCT in the past 30 minutes.- Which of the following is the most probable malfunction? c. A. Letdown flow control valves LCV-1CH-112 & 115A are partially diverting letdown flow to the CRT's. l B. RV-1CH-203, Letdown line relief valve, has lifted and failed to reseat causing a portion of letdown flow to be diverted to the'PRT. C. HCV-lCH-186, RCP seal injection control valve, has failed open causing excessive seal injection flow and VCT depletion. D. LCV-1C -460A, Letdown isolation valve, has developed a severe packing i leak causing a small loss of reactor coolant condition. l 1 POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A l 4 REFERENCES ' 10M-7.1.D - Issue 4 Rev. 2,-10M Fig. No. 7 Issue 9 Rev. 9. 1LP-SOS-7.1 OBJECTIVE: 6 NUMBER: 1-97-003, New JTA f { K/A f: 3.01.004.020.K1.02 K/A IMPORTANCE: 2.7/2.8 Rev, I

l Question thunber 21 l l Given the following: The plant is at 100% power. l e All systems are in their at power, NSA configurations. l e RCS T.v. High Annunciator is lit. You are directed to adjust RCS temperature using the Boration/ Dilution controls. Determine the mode of makeup control required and the expected corresponding l valve lineup. l l l A. Borates l FCV-1CH-113A, Boric acid flow control valve - OPEN l FCV-1CH-114A, Primary water flow control valve - CLOSED l FCV-1CH-113B, Makeup stop valve to the charging pump suction - CLOSED FCV-1CH-114B, Makeup stop valve to the VCT - OPEN B. Borates i FCV-1CH-113A, Boric acid flow control valve - OPEN FCV-1CH-114A, Primary water flow control valve - CLOSED FCV-1CH-113B, Makeup stop valve to the charging pump suction - OPEN FCV-1CH-114B, Makeup stop valve to the VCT - CLOSED C. Dilutes FCV-1CH-113A, Boric acid flow control valve - OPEN FCV-1CH-114A, Primary water flow control valve - OPEN FCV-1CH-113B, Makeup stop valve to the charging pump suction - OPEN. FCV-1CH-114B, Makeup stop valve to the VCT - CLOSED D. Dilutes FCV-1CH-113A, Boric acid flow control valve - CLOSED FCV-1CH-114A, Primary water flow control valve - OPEN FCV-1CH-113B, Makeup stop valve to the charging pump suction - CLOSED l FCV-1CH-114B, Makeup stop valve to the VCT - OPEN POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B l

REFERENCES:

10M-7.1.D - Issue 4, Rev. 2. ILP-SOS-7.1 OBJECTIVEi 5 NUMBER: 1-97-004, New JTA 9: 1 K/A #: 3.01.004.010.A4.03 K/A IMPORTANCE: 3.9/3.7 Rev.I

1 -- 1 Questi:nNumber 22 f Refer to.BVPS Unit 1, Curve Dook, Cycle 12, Issue 12 Rev. 1, to answer the + following question, a i Given the following A plant startup is in progress with Rx power at 17%. e ,e Steam dumps are in the Main Steam Pressure Control Mode maintaining Main Steam Pressure at 100$ psig. The Main Generator Outpat Breakers have just been closed. e e Rod control is in manual. CBD is at 100 steps with normal rod sequencing. e RCS boron concentration is 705 ppm. e The Rx core is Xenon free. A spurious equipment failure caused an intermittent continuous rod withdrawal of CBD that stopped with CBD at 110 steps. Assuming that no Rx trip occurs and Tav. remains on program, determine the level at which Rx power will stabilize. A. 14% l B. 17% C. 18% D.'20% I POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. CB-13: 705 ppm = 825 MWD /MTU. - Use MOL curves. CB-21: Power Defect is 20 pam/% power. CB-11Et 100 steps on CBD = 839 pcm. 110 steps on CBD = 780 pcm. Delta = 59 pcm. 59 pcm + 20 pcm/t power = 3% change. 17% + 3% = 20%.

REFERENCES:

Unit 1 Curve Book, Cycle 12, Issue 12 Rev.1, curves CB-11E,13,21 and 24B. ILP-SOS-LP-RT-6 OBJECTIVE: 9 NUMBER: 1-97-005, New JTA f: [ K/A f: 3.01.000.001.EK1.03-K/A IMPORTANCE: 3.9/4.0 Rev.t l

QuIstion Number 23 1 l Due to an equipment malfunction in the automatic rod control system, a continuous rod withdrawal was initiated from an initial power level of 726. i After 5 seconds of rod withdrawal, the reactor operator selected " MANUAL" on the I Rod Control Selector Switch which terminated the rod movement. All other systems are aligned in their normal, at power, NSA lineups. Assume that no reactor trip occurs, and NO other operator action is taken. Which of the following parameters will return to essentially the same value that it was before the transient? A. RCS Tave. B. Pressurizer level. C.. Delta I. D. Reactor power. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. i i

REFERENCES:

LP-RT-6 Rev. 2' ILP-SQS-RT-6 OBJECTIVE: 15 NUMBER: 1-97-006, M-0497 JTA i: K/A #: 3.01.000.001.EA2.04 (001EA2.04) K/A IMPORTANCE: 4.2/4.3 ' Rev, I

~.. - - _ - = -..... - ~ QuestionNumber 24 Given the following: {; A reactor.startup is in progress with power at IE-8 amps in the IR. e All systems are aligned in their normal lineups for the current power l e level. Main feed pump FW-P-1B is in seevice with SGWL control in manual using the bypass feedwater regulating valves FCV-FW-479,489, and 499. An electrical fault occurs that causes a sustained loss of the 1A 4KV bus. The IAE' bus is re-energized from the No. 1 Emergency Diesel Generator. e Determine the expected configuration of the Reactor 'and the Rod Drive MG Sets following the loss of the 1A 4KV bus. A. The Reactor will trip on low flow due to the loss of the 1A RCP. Neither rod drive MG sets are affected due to the automatic bus transfer of the 1A 480V bus feed to the 1C 4KV bus via the 480V bus tie breaker. B. The Reactor Will not trip. Rod drive MG set Rod-MG-1 will be lost due to the loss of the 1A 4KV bus but Rod-MG-2 has sufficient capacity to maintain power to all control rods. C. The Reactor will trip due to' loss of Rod Drive power. Rod drive MG set Rod-MG-1 will be lost due to the loss of the 1A 4KV j bus and Rod-MG-2 does not have sufficient capacity to maintain power to all. control rods. D. The Reactor will not trip. Neither rod drive MG sets are affected due to the automatic bus transfer of the 1A 480V bus feed to the 1C 4KV bus via the 480V bus tie breaker. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. i l i~'

REFERENCES:

10M-37.1.D - Issue 4 Rev. O, 10M-1. 5.B.4 - Issue 2 Rev.10. j l' 1LP-SOS-37.1-OBJECTIVE: 2 NUMBER: 1-97-007, New JTA 9: ( K/A #: 3.01.000.003.EK2.0$ K/A IMPORTANCE: 2.5/2.8 i~ l Rev.1

QuestionNumber 25 Technical Specification 3.1.3.1 action c.3 allows continued power oper.ation when there is a control rod that is trippable but misaligned from its group step i counter demand position by more than +-12 steps, provided that THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the hour. The Technical Specification bases for this power level restriction is to-A. ensure adequate shutdown margin during continued operation. B. provide assurance of fuel rod integrity during continued operation. C. reduce the required minimum rod insertion limit during continued operation. D. minimize the power peaks on a subsequent rod drop accident. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B.

REFERENCES:

TS 3.1.3.1 and Bases 1LP-SOS-1.3 OBJECTIVE: 15 NUMBER: 1-97-008, New JTA #: i' K/A f: 3.01.000.005.EK3.05 K/A IMPORTANCE: 3.4/4.2 Rev.1

Qu:sti:nNumber 26 The diagnostic steps of E-0, Reactor Trip or Safety Injection, use steam generator pressure and level, secondary system radiation levels, and containment j parameters (radiation, pressure, and sump level) to determine if a loss of secondary coolant, steam generator tube rupture, or a loss of primary coolant has occurred. Why is Reactor Coolant System (RCS) pressure NOT used in these diagnostic steps in E-07 RCS pressure will_ A. NOT be af fected by a loss of secondary coolant accident. B. ONLY be affected by a loss of primary coolant accident. C. be affected by ONLY a loss of primary coolant accident and a loss of i secondary coolant accident. D. be affected by ALL three accidents; loss of primary and secondary coolant accidents, and steam generator tube rupture accidents. 1 1 j l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D.

REFERENCES:

NOMCD Lesson Plan ILP-SQS-53.3 OBJECTIVE: 3 NUMBER: 1-97-009, M-0360 JTA 8: 3010010601 K/A 8: 3.01.000.007.EK3.01 K/A IMPORTANCE: 4.0/4.3 j Rev, I

~ -. l l QuestionNumber 27 .l A Rx trip from 100% power has occurred at 0700 hours today. During the performance of ES-0.1, Reactor Trip Response, it was determined that two (2) i control rods failed to fully insert and an emergency boration was initiated as required by procedure. Present RCS boron concentration has been determined to be equal to the required boron concentration for the current xenon concentration. In order to obtain the required amount of shutdown margin IAW 10ST-49.2, between 2000 hours and 2100 hours today, the required RCS boron concentration will (1) because the change in xenon concentration will be adding (2) reactivity. (1) (2) i A.

increase, negative l

'B.

increase, positive l'

positive i C.

decrease, i

D.

decrease, negative i

j i-l \\ l l l l 1 i i POINTS: 1.00 TIME ALOTTMENT: minutes j ANSWER: B. l I i l f l

REFERENCES:

10ST-49.2 - Issue 4 Rev. 4. ILP-SQS-49.1 OBJECTIVE: 2 NUMBER: 1-97-010, New JTA #: K/A #: 3.01.000.024.EA2.05 K/A IMPORTANCE: 3.3/3.9 i 5 l: i Rev.1 l m r

- Question Number 28 i An equipment malfunction has caused a demand for a Rx trip. The automatic Rx trip signal did not open the Rx trip breakers (RTB's). Which of the following describes the locations from which the RTB's and Rod Drive MG Set Supply breakers can be opened? 1. Locally on'the front of the respective breaker, t

2. Remote Shutdown Panel.

3. Rod drive MG Control Panel. Rod Drive MG Set j RTB's Supply Breakers i A. 1 ONLY. 1 and 3 ONLY. B. 1 and 2 ONLY. 1 and 3 ONLY. C. 1 ONLY. 3 ONLY. D. 1 and 2 ONLY. 3 ONLY. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. -j

REFERENCES:

10M-1.3.C - Issue 4, Rev. 1. i 1LP-SQS-53.3 OBJECTIVE: 6 NUMBER: 1-97-011, New JTA 8: -K/A 4: 3.01.000.029.EA1.12 K/A IMPORTANCE: 4.1/4.0 Rev.1

- Questisn Number 29 With the plant'st 100% power, all systems in their full power, NSA configuration; " MALFUNCTION" is displayed on the RVLIC Train A plasma display Dynamic Head meter. Which of the following could be the cause of this indication? A. A high volume sensor bellows is leaking causing a hydraulic isolator limit switch to actuate. B. PT-1RC-455, Pressurizer pressure protection channel I has failed high. C. This indication is normal, the RCP breaker auxiliary contacts disable

the. Dynamic Head indications when any RCP is running.

D. TRB-RC-412B1, a T-hot narrow range channel I RTD has failed open. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. RE FERENCES t 10M-6.1.D - Issue 4 Rev. 1 1LP-SQS-6.5 OBJECTIVE: Sh NUMBER: 1-97-012, New JTA 9: K/A is 3.02.000.K6.03 K/A IMPORTANCE: 3.1/3.6 j. Rev.I

as -~- s. .s -,u..- r.n,. .+.au_a s x-m. =u.a w xa-x-a..-s.ua n_s +. ...m- -a._x----a.~.Awus .s .+- w~-- ~ ~ - - -,s.. l- . Question Number 30 A large break LOCA and Safety Injection is in progress in the cold leg injection l - phase. What automatic actions will occur within 2.5 minutes of receiving - annunciator Al-25 "2/4 RWST LO LEVEL & SI AUTO XFR SI INJ TO RECIRC"? ) i i A. - CNMT sump to LHSI pump suction valves MOV-1SI-860A&B CLOSE. B. HHSI to RCL Cold Leg isolation valve MOV-1SI-836 OPENS. -C. RWST to HHSI valves MOV-1CH-115B&D OPEN. ) l D. LHSI to HHSI cross connect valves MOV-1SI-863A&B OPEN. 4 l l 1 I I .i l L l I I i l l l PRINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. s l-

REFERENCES:

10M-53.A.1 Att. 1-G, Issue 1B, Rev. 1 1LP-SQS-11.1 OBJECTIVE: $c NUMBER: 1-97-013, M-0421 i JTA'4: i... K/A 4: 3.02.006.020.K4.03 (006A3.08) , K/A IMPORTANCE: 3.2/3.6 I Rev.I

.~. ~. - - -.~.-. -.. - t 2 QuestionNumber 31 Given the following .The plant is operating at 100% power. e i e All systems are aligned in their at power, NSA configurations. The controlling pressurizer level channel, LT459, fails high. 'e Assuming NO operator actions are taken, what will be the First Out i Annunciator for the ensuing Rx trip? l A. Low Pressurizer Pressure. B. Low Pressurizer Level. C. High-Pressurizer Level. D. High Pressurizer Pressure. i e IJ 2 I 1. 1 J 1 f I p. 1 POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER:- C. j l' J 1 s a i

REFERENCES:

10M-6.1.D - Issue 4, Rev. 1 ILP-SQS-6.4 OBJECTIVE: 14 NUMBER: 1-97-014, M-0658 JTA 0: '( K/A 0: 3.02.011.000.A2.10 (011K3.01) K/A IMPORTANCE: 3.4/3.6 i Rev.1

Ghiestion Numtxt 32 ) l Given the Following: I A normal plant cooldown and depressurization is in progress using the e condenser steam dumps. All S/D rod banks are withdrawn. All RCP's are running. RCS T.v. is 501*F and PZR pressure is 1925 psig. The cooldown and depressurization is temporarily terminated to conduct shift turnover. During the turnover T.. has drifted up to 514*F and PZR pressure has risen to 2005 psig. When the cooldown and depressurization is recommenced, which of the following must be performed? 1 A. Immediately depressurize the RCS to less than 1945 psig to prevent exceeding the S/G tube differential pressure limit. B. Reset the condenser steam dump cooldown valve interlock to restore manual operator control of the cooldown. C. Verify the PZR low-pressure reactor trip is bypassed prior to 1945 psig to keep the shutdown banks withdrawn. D. Re-block the PZR low-pressure SI signal when pressure is reduced below 1980 psig. i l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. REFERENCES : 10M-51.4.C, Issue 4, Rev. 9 - Caution. 1LP-SOS-51.1 OBJECTIVE: 3/9 NUMBER: 1-97-015, M-0255 JTA #1 { K/A #: 3.02.013.000.K4.12 (010K1.02) K/A IMPORTANCE: 3.7/3.9 Rev.1

-. -. - ~ .. - - -. -.. - ~ ~ -... -. _ _ -, ~.-...-.~ l QuestionNumber 33 i i r I The plant is in Mode 3 making preparations for a Rx startup, when a small break loss of coolant accident occurs. All systems function as designed and no other l l transient exists. Which of the following actions will automatically occur if i ~ all four containment pressure instruments reach a maximum of 5.0 psig? A. MSLI actuation, Feedwater Isolation actuation and the Motor Driven j Auxiliary Teodwater Pumps auto start. B. Containment Spray Actuation,' Emergency Diesel Generators auto start, and the Motor Driven Auxiliary Feedwater Pumps auto start. C. MSLI actuation,' Emergency Diesel Generators auto start and load, and the Turbine Driven Auxiliary Feedwater Pump auto starts. D. CREBAPS actuation, Feedwater Isolation' actuation, and the Turbine Driven Auxiliary Feedwater Pump auto starts. l 1 i -POINTS: 1.00 TIME ALOTTMENT: minutes-ANSWER: A.

REFERENCES:

-10M-1,11,12, & 13.2.B - Issue 4, Rev. 4,2,1 & 3 Respectively.

1LP-SQS-11.1/13.1 OBJECTIVE: 5/5 NUMBER: 1-97-016, New JTA 9: K/A 4:> 3.02.013.000.A4.03 K/A IMPORTANCE: 4.5/4.7 Rey. I l

I QuestionNumber 34 Given the following: ( e Rx power is.100%. All systems are at power, NSA configurations. e The following alarms are actuated: Pressurizer Control Low Level Deviation, Letdown Temperature Hi/Lo, e i Containment. Sump Pump Run. e PZR level is 42% and dropping slowly. e TCV-1CH-143, Domineralizer Bypass valve is in the VCT position. -e. FCV-1CH-122, Charging flow control valve is full open. FI-1CH-122, Charging flow meter is peggad high. FI-1CH-150, Letdown flow meter indicates 105 gpm. All three RCP seal injection flow meters indicate between 8.0-8.5 gpm. e All three KOP seal return flow recorders indicate between 2.8-3.2 gpm. The location of the leak is on the_ A. in service RCP seal injection filter inlet line. B. charging line between the containment penetration and the Regen l HX. l C. letdown line between the Regen HX and the Letdown Orifices. i D. MOV-1CH-311, PZR AUX' spray control valve inlet line. l I 1 POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B. 1 i l 1 'l

REFERENCES:

OM Figure 7-1 Rev 9 ILP-SQS-7.1 OBJECTIVE: 2 NUMBER: 1-97-017, New JTA 9: K/A 9: '3.02.000.022.EA2.01 K/A IMPORTANCE: 3.2/3.8 ,i Rev.1

...__.-.._m i ' l L -Question Number 35 Given the following: ' The plant is at 47% power. - e .All systems are at power, NSA configurations, e The pressurizer level selector switch is in position 1, 459/460. e The reference leg for pressurizer level transmitter LT-1RC-460 - e-develops a leak. Assuming no operator actions are taken, which of the following will occur? .A. Pressurizer level will stabilize at the full load setpoint of 59%. B. The Rx will eventually trip on low pressurizer pressure. [-- C. Annunciator A4-1, Pressurizer Control Level High will actuate. I D. All letdown orifice isolation valves will immediately trip closed. I: t i t i i 3 i l l i i POINTS: 1.00 TIME ALOTTMENT: minutes l ANSWER: C. t l ia ( l l t i

REFERENCES:

.lOM-6.4IF Issue 4,.Rev. 5 l-i ILP-SQS-6.5 OBJECTIVE: Sh NUMBER: 1-97-018, M-Oll6 i JTA 9: l. K/A 8: 3.02.000.028.EKl.01 (028AKl.01) K/A IMPORTANCE: 2.8/3.1 m. 4 i-l. .m,.

o...

__-~_...._m QuestionNumber 36 4 1 The pressurizer level controller utilizes an integral control function. Which of the following describes the action of this integral control function? The pressurizer level controller integral control action will (1) the demand signal to FCV-1CH-122, Pressurizer Level Control Valve, (2) A. (1) raise (2) for as long as actual Lpar is below program level B. -(1) raise (2) only as long as actual L,i, is dropping. C. (1) lower (2) only as long as program Lys, is rising. D. (1) provide (2) in proportion to the difference between actual and program L,ar. s 4 I i a d i i s i POINTS: 1.00 TIME ALOTTMENT: minutes i ANSWER: A. k i ) i i 1

REFERENCES:

1/2LP-ICS-1.4 j ILP-ICS-1.4 OBJECTIVE: 2 NUMBER: 1-97-019, New i JTA #: 1 K/A 8: 3.02.000.028.EK2.03 K/A IMPORTANCE: 2.6/2.9 ? 1-a Rev.I R y - -m., m-,- y m2

t Questi:n Number 37 The; Unit is in Mode 5 with the RCS water solid. PZR level is >100% and the RCS. heating up. What valve should be adjusted to control RCS pressure, and in which direction should this. valve be moved in order to maintain RCS pressure stable? A. MOV-lCH-142, RH Letdown to Non Regen HX Inlet Flow Control Valve i should be CLOSED. i B. ' PCV-lRC-455A PZR Spray Valve should be OPENED. d C. FCV-lCH-122, Charging Flow to Regen HX Inlet Control Valve should be OPENED. D. PCV-lCH-145, LP Letdown Back Pressure Regulating Valve should be OPENED. i l l I 1 I POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D.

REFERENCES:

lOM-6.4.F - Issue 4, Rev. 6 ILP-SQS-6.5 . OBJECTIVE: 4 ' NUMBER: 1-97-020, New JTA 4: K/A 4: 3.03.010.000.A1.04 K/A IMPORTANCE: 3.6/3.8 Rev.1

i QuestionNumber 38 When responding.to a stuck open PZR PORV, the associated PORV Block Valve is closed.- Which of the following sets of indications would indicate that the {. Block Valve is fully closed and not leaking? A. PORV Acoustic Monitor Won' t Roset PORV Tailpipe Temperature Dropping PRT Level Dropping PRT Pressure Dropping ) B. PORV Acoustic Monitor Will Reset PORV Tailpipe Temperature At T-sat for PRT pressure PRT Level Rising PRT Pressure Rising C. PORV Acoustic Monitor Will Reset PORV Tailpipe Temperature At T-sat for PRT pressure j PRT Level Stable PRT Pressure Stable D. PORV Acoustic Monitor Will Reset PORV Tailpipe Temperature Rising PRT Level Rising PRT Pressure Rising l POINTS: 1.00 TIME ALOTTMENT: minutes i l ANSWER: C. ) i l~

REFERENCES:

.OM Fig. 6-2 Rev. 9 I' ILP-SQS-6.4 OBJECTIVE: 4 NUMBER: 1-97-021, New JTA it K/A 4: 3.03.000.008.EA1.01 K/A IMPORTANCE: 4.2/4.0 k i Rey,1

QuestionNumber 39 When conducting a natural circulation cooldown following a small break LOCA, which of the following would cause Natural Circulation flow to increase? i A. Raising all SGWL's using Aux Feed. B. Lowering RCS pressure using Aux Spray. C. ' Raising the setpoint on the S/G Atmospheric relief valves. D. Lowering the RCS cooldown rate using the. Condenser Steam Dumps. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A.

REFERENCES:

10M-53B.5.GI Issue 1B, Rev. 1 1LP-SQS-53.2 OBJECTIVE: 11 NUMBER: 1-97-022, New JTA 5: K/A 4: '3.03.000.009.EK1.01 K/A IMPORTANCE: 4.2/4.7 Rev.I

4, 1 Questien Number 40 f. 1 l i What is'the bases for the RCP Trip Criteria Setpoint of RCS/ Highest SG D/P = 150 1 psid (450 psid Adverse]? ' k A. ' Provides for timely RCP trips for a small break LOCA events but reduces the probability of RCP_ trips for SGTR's and non-LOCA events. 'B. Allows the RCPs to remain in service for core cooling until cavitation -damage potential' reaches FSAR limits. 'C. Ensures RCS. fluid level never drops below the elevation of the break during a small break LOCA thus ensuring re-pressurization of the RCS. i l D. Ensures the RCPs are removed from service during a Loss of Secondary Coolant event whi n limits the RCS cooldown rate. 1 f. a l POINTS: 1.00 TIME ALCTTMENT: minutes. 1 l ANSWER: A. i 4 i

REFERENCES:

10M-53.B.5.GI-6,' Issue 1B, Rev. 1 lLP-SQS-53.2 OBJECTIVE: 1 NUMBER: 1-97-023, New JTA #: i I { K/A #: 3.03.000.009.EK3.23 K/A IMPORTANCE: 4.2/4.3 i Rey, I

.____.._._.m.- m.- .._ _ m m .- y... _ _ _ - Question Number 41 1-A large break'LOCA has occurred. The-crew has made the following procedure transitions; E-1, Loss of Reactor or Secondary Coolant, to ES-1.3, Transfer to Cold Leg Recirculation, completed the cold leg recirculation line up and then I i ' returned to the procedure and step in effect, E-1 Loss of Reactor or Secondary-Coolant, step 24. Step 24 of E-1 directs actions to isolate the SI Accumulators I if-at least two RCS hot leg temperatures are less than 390*F. l '- The-bases for this RCS temperature ensures 1-J a A. saturation pressure of the RCS is less than Accumulator nitrogen pressure when the accumulator water volume is fully discharged. B. sufficient Accumulator water and nitrogen volumes are injected into the RCS prior to isolating the accumulators. ? C. adequate core cooling is established prior to isolating the Accumulators as a water injection source. i .D. that the injected Accumulator nitrogen has expanded sufficiently to maintain RCS saturation temperature less than the UFSAR design bases. l 2-f l' ) i 4 l-J' 1 I ti. l-i -I 4 J S { POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. is i 7 0 e d l i i

REFERENCES:

' lOM-53B.4.E Issue 1B, Rev. 3 ILP-SQS-53.3 OBJECTIVE: 3 NUMBER: 1-97-024, New JTA #:

K/A 9: 3.03.000.011.EK3.12 K/A IMPORTANCE: 4.4/4.6 i o, - r i Rev.1 i

) Question Number 42 A major plant transient is in progress with the Unit at full power. The current plant parameters are: j (Assume parameters are stable unless otherwise stated.) Highest Core Exit Thermocouple = 623*F e e RCS T-hot's: A = 608*F, B = 608'F, C = 608'F e RCS T-cold's: A = 54 6*F, E = 54 5*F, C = 54 4'E RCS Press = 2112 psig snd dropping. e s A S/G: Press = 800 psig, Level = 44%, Feed Flow = 3.6 E5 lbm/hr. B S/G: Press = 805 psig, Level = 43%, Feed Flow = 3.7 E5 lbm/hr. C S/G: Press = 790 psig, Level - 44%, Feed Flow = 3.5 E5 lbm/hr. l e Containment: Press = 12 psia, Temp = 110'F, e Radiation: 1 Particulate monitor 1RM-RM-215A = 1.3 E5 cpm with its HI e alarm lit. Gaseous monitor 1RM-RM-215B = 1.9 E5 cpm with its HI-HI alarm lit. In-Core Transfer Device monitor 1RM-RM-204 = 1000 mR/hr with its HI-HI alarm lit. j Which of the following events is occurring

• A.

Only a Loss of Secondary Coolant. l B. Only a Steam Generator Tube Rupture. C. Only a Loss of Reactor Coolant. D. A single S/G is Ruptured and Faulted iniide containment. POINTS: 1.00 TIME ALOTTMENT: minutes i ANSWER: C. REFERENCES : 10M-53A.1.E Issue 1B, Rev. 4 ILP-SQS-53.3 OBJECTIVE: 6 NUMBER: 1-97-025, New JTA #: i K/A #: 3.03.000.011.EA2.13 K/A IMPORTANCE: 3.7/3.7 Rev.I

m l QuestionNumber 43 Given the followings ( The Unit is in Mode 4 with a. plant heat-up in progress. RCS Temperature is 220 'T, being maintained by RHR. There is a bubble in the pressurizer with PZR level at 22%. e RCS Pressure is 340 psig. e 1A RCP is in service. Pressurizer overpressure protection system (OPPS) is in service. i I&C is performing maintenance on the pressurizer pressure master and e slave controllers, and requests the controllers to be placed in the following line-up; PCV-lRC-455A PZR Spray Valve controller in AUTO. PCV-lRC-455B PZR Spray Va'1ve controller in manual and shut. e PZR Group A Heater control switch in AUTO. PZR Group B,C,D, and E Heater control switches in OFF. e Master Pressure Controller in AUTO. j i What effect will this line-up have on PZR pressure with no further operator action? A. No effect. The master pressure controller output is bypassed with the OPPS keyswitches in AUTOMATIC. B. PZR pressure will be' maintained at 340 psig with PCV-lRC-455A and Group A Heaters operating in AUTO. C. PZR pressure will rise to 410 psig and cause ONLY OPPS Relief valve PCV-1RC-455D to open. PCV-lRC-455C will not open. D. PZR pressure will rise to 410 psig and cause BOTH OPPS Relief valves PCV-1RC-455C and PCV-1RC-455D to open. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D.

REFERENCES:

10M-6.1.D Issue 4, Rev. 1 lLP-SOS-6.4 OBJECTIVE: 5 NUMBER: 1-97-026, New JTA 9: K/A 9: 3.03.000.027.EA2.03 K/A IMPORTANCE: 3.3/3.4 i Rev.I

Question Number 44 Given the following: { The Unit is in Mode 1 with all systems in their full power, NSA configuration. The 2B and 2E Pressurizer heaters are in manual and on to equalize the PZR and RCS boron concentrations. The Auto / Man station for POV-RC-455A, PZR spray valve failed, causing the valve to go full open. In responding to the transient, the operator placed both PCV-1RC-455A and B, PZR spray valves, in MANUAL and was able to close both valves. PZR pressure at that point was 2156 psig. If no further operator actions are performed, PZR pressure will A. drop resulting in an OT/AT Reactor Trip. B. drop resulting in a Low Pressurizer Pressure Reactor Trip, C. rise and cause PZR PORV PCV-1RC-455C to open at 2335 psig. D. riso and cause PZR PORV PCV-lRC-455C to open at < 2335 psig. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D.

REFERENCES:

10M-6.4.ABU Step 5 Caution, Issue 3, Rev. O ILP-SQS-6.4 OBJECTIVE: 11 NUMBER: 1-97-027, New JTA #: { K/A 4: 3.03.000.027.G05 K/A IMPORTANCE: 3.3/3.3 Rev. I

Questiin Number 45 With the Unit in MODE 3, T.v. at 547'F, RCS Pressure at 2235 psig, charging flow 120 gpm and letdown-flow is 45 gpre, which of the following conditions would require entry into AOP 1.6.4 - Steam Generator Tube Leakage? RCS Pressure PZR Level S/G Blowdown SGWL Rad Monitor j A. Dropping Dropping Rising Rising B. Stable Rising Dropping Stable C. Dropping Dropping Stable Stable D. Stable Stable Stable Dropping l i l ) i POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A.

REFERENCES:

AOP 1.6.4 Issue 3A, Rev. 7 ILP-S0$-53C.1 OBJECTIVE: 2 NUMBER: 1-97-028, New JTA #: K/A I: 3.03.000.037.Gli K/A IMPORTANCE: 3.9/4.1 Rev.I

Question Number 46 i, 1 A continuous action step in E-3,." Steam Generator Tube Rupture", directs the operator to maintain feed flow to the ruptured S/G until narrow range level is greater than 5%. What is the basis for establishing this minimum level in the ruptured S/G7' j .? l A. ' To compress the ruptured S/G steam bubble and raise its pressure, thus minimizing break flow. B. To minimize the ruptured S/G depressurization during the. subsequent RCS cooldown. j C. To dilute RCS' water with SG water in anticipation of an uncontrolled radiological release to the environment. l D. To prevent thermal stratification in the ruptured S/G which would j extend the time required to stop break flow. l l j POINTS:- 1.00 TIME ALOTTMENT: minutes ANSWER: B. REFERENCES :- 10M-53B.4.E Issue 1B, Rev. 3 ILP-SQS-53.3 OBJECTIVE: 3 UUMBER: 1-97-029, M-0536 JTA #: { K/A #: 3.03.000.038.EK3.06 (038EK3.06) K/A IMPORTANCE: 4.2/4.5 I Rev.!

l Questian Numtwr 47 l The Unit was in' Mode 1 with all systems in their full power, NSA configurations, when a SGTR on the 1A S/G, accompanied by a loss of the 1B 4KV bus occurred. l The operators have transitioned from E-0, Reactor Trip or Safety Injection, to E-3, Steam Generator Tube Rupture, and-have completed isolating 1A S/G. S/G parameters ares e 1A S/Gs Level = 10%, Pressure = 1010 psig e 1B S/Gs Level = 23%, Pressure = 995 psig e 1C S/Gs Level = 15%, Pressure = 1000 psig At this point, a circuit malfunction causes all of the steam dumps to trip closed. Which of the following will occur FIRST in response to the RCS heatup following the loss of the condenser. steam dumps? A. PCV-1MS-101A, the 1A S/G Atmospheric Relief Valve will modulate open l at 1035 psig. B. PCV-1MS-101B, the 1B S/G Atmospheric Relief Valve will trip open at 1060 psig. C. PCV-1MS-101C, the 1C S/G Atmospheric Relief Valve will trip open at 1060 psig. D. PCV-1MS-101C, the 1C S/G Atmospheric Relief Valve will modulate open at 1035 psig. t POINTS: 1.00 -TIME ALOTTMENT: minutes ANSWER: C. c i l j i l i

REFERENCES:

10M-21.3.A - Issue 4, R'ev. 4. ILP-SOS-21.1 OBJECTIVE: 3 NUMBER: 1-97-030, New JTA 9: { K/A 9: 3.03.000.038.EA1.16 K/A IMPORTANCE: 4.4/4.3 Rev.I

' Qusstion Number 48 Which of the following describes the response of 1A S/G NR water level and 1A S/G steam flow when lA RCP is stopped with Rx power initially at 20%7 A. SGWL will initially drop due to shrink and steam flow will drop due to reverse flow.in the loop. B. SGWL will initially rise due to swell and steam flow will rise due to .. reverse flow in the loop. 'C. SGWL will-initially drop due to shrink and steam flow will drop due to the reduced load on the main generator. D. - SGWL will initially rise due to shrink and steam flow will drop due to reverse flow in the loop. l E l l l l POINTS: 1.00 TIME ALOTTMENT: minutes l [ ANSWER: A.

REFERENCES:

lLP-SQS-6.3 ILP-SQS-6.3 OBJECTIVE: 11 NUMBER: 1-97-031, New JTA # K/A #: 3.04.003.000.K3.02 K/A IMPORTANCE: 3.5/3.8 ,j Rev.I

QuestionNumber 49 With the Unit at full power, the number one seal on IB RCP has failed. After . tripping the Rx and stopping the 1B RCP, which of the following is performed to limit the temperature rise of the 1B RCP lower radial bearing? 1 A. Open HOV-1CH-307, RCP Seal bypass valve. B. Open 1CCR-307, 1B RCP Thermal barrier flow throttle valve. C. Close MOV-1CH-378 or 381, Conunon seal return isolation valve. D. Close MOV-1CH-303B, 1B RCP seal leakoff isolation valve. e f f i 1 POINTS: 1.00 TIME ALOTTMENT: minutes ) ANSWER: D. 1

REFERENCES:

lOM-7.4.ABE - Issue 3, Rev. 3 ILP-SOS-6.3 OBJECTIVE: 12 NUMBER: 1-97-032, New JTA # - K/A 8 - 3.04.003.000.A2.01 K/A IMPORTANCE: 3.5/3.9 Rev.I C

__.m.-.._ 1 ) Questi:n Number'50 .When conducting a core off load in a recent refueling outage, the manipulator- . crane. operator experienced the following: ,f e: Minor vibrations of a suspen'ded fuel assembly in one quadrant of the Rx vessel near the outer edge of the vessel. Visual observation.of flow turbulence in the same general location in e .the Rx vessel. With all systems in their normal line up for MODE 6, and an RCS sample in progress, what is the most probable cause of these indications? A. High charging flow from the HHSI/ Charging pumps via the 'B' Loop Hot Leg. B. High flow through the CVCS letdown orifices via the 'A' Loop Cold Leg. C. High letdown flow to the Residual Heat Removal System via the 'A' Loop Hot Leg. D. High RCS sample purge flow to the Primary Sample System via the 'B' Loop Hot Leg. i ) iPOINTS: 1.00 TIME ALOTTMENT: ndnutes ANSWER: C. I, l*

REFERENCES:

lOM Figure 6-1.- Rev. 7 i [ -lLP-SQS-10.1-OBJECTIVE: 1 NUMBER: 1-97-033, New j. JTA 8: 1 K/A 8: 3.04.005.000.Kl.09 K/A IMPORTANCE: 3.6/3.9 i Rev.1-

QuestionNumber 51 The Unit'is at 37% Rx power conducting a power ascension to full power. All systems are aligned in their normal lineups for the current power level with the . i' following exceptions Turbine EHC control is in MANUAL - IMP OUT control due to i a problem with the EHC first stage pressure transmitter, which is de-energized. The operator depresses the Gvt pushbutton for 2 seconds to continue the load ascension. What is the response of the main feedwater regulating valves to this action? The Main Feedwater Regulating Valves will initially throttle-A. CLOSED due to the shrink of the SGWL, and then throttle OPEN when level drops below 44%. l l B. CLOSED due to the~ steam flow - feed flow mismatch, and then throttle OPEN when level drops below 33%. j C. OPEN due to the swell of the SGWL, and then regulate to control level at 44%. 1 D. OPEN due to the steam flow - feed flow mismatch, and then regulate to control level at 44%. l l l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D.

REFERENCES:

lOM-24.1.D - Issue 4, Rev. 1 ILP-SOS-24.1 -OBJECTIVE: 4 NUMBER: 1-97-034, New JTA i: K/A i: 3.04.035.010.A3.01 K/A IMPORTANCE: 4.0/3.9 ~ Rev 1

s . Question Number 52 Which of the following would electrically prevent-the IC RCP breaker from closing?' .l-A. -1C RCP #1 seal dif ferential pressure < 200 psid. -B. 1C RCP #1 seal leakoff flow <0.2 gpm. 'C. 1C RCP Oil Lift pop is de-energized. D. Annunciator A3-75, RCP Lower Bearing-Lube Oil Cooling Water Flow Low, is lit. .i-POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C. L l

REFERENCES:

OM Figure 6-21 Rev. 10 and 6-22 Rev. 8 ILP-SOS-6.3 OBJECTIVE: 6 NUMBER: 1-97-035, New JTA # K/A 6: 3.04.000.015.EK2.10-K/A IMPORTANCE: 2.8/2.8 4 Rev.I

Question Number 53 l 1 While operating in Reduced Inventory /Midloop conditions, a loss of RCS inventory has occurred, resulting in a loss of RHR. The cause of the loss of RCS -inventory has not been identified. Which of the following will be performed in accordance with AOP 10.2, Loss of RER While Operating at Reduced Inventory, to: minimize the possibility of gas _ binding the RER pumps when they are restarted? 1. Raise RCS Level using a Charging /HHSI pump. 2. Shut MOV-1RH-758 and 605, the RHR temperature-and flow control valves. 3. Establish containment integrity.

4. Establish level in at least 2 S/G's >15% narrow range.

A. 1 & 3 ONLY. B. 1 & 2 ONLY. C. 2, 3& 4 ONLY. l .D. 1, 2 & 4 ONLY. 1 POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B. I 'l 1 l

REFERENCES:

10M-53C.4.1.10.1 - Issue 3A, Rev. 5 l ILP-SOS-10.1 OBJECTIVE: 9 NUMBER: 1-97-036, New JTA'#: ) K/A #: 3.04.000.025.EK1.01 K/A IMPORTANCE: 3.9/4.3 Rev.1.

4 - Question Numtwr 54 ' q I 1 Functional Restoration Procedure, FR-C.1, " Response to Inadequate Core Cooling" is designed to reduce core exit thermocouple temperatures and recover Rx vessel level. Which of the following describes the processes used to accomplish this objective in the order in which'they are performed? 1. Establish RCS Bleed and Feed for heat' removal. 2. Perform a rapid secondary depressurization to depressurize the RCS and inject the SI Accumulators.

3.. Restart the RCP's to provide two phase flow through the core.

4. Restore High Pressure Safety Injection flow. A. 2,3,4. B. 2,1,3. C. 4,2,1. D. 4,2,3. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D.

REFERENCES:

loM-53B.4.FR-C.1 - Issue 1B, Rev. 3 ILP-SQS-53.3 OBJECTIVE: 3 NUMBER: 1-97-037, New JTA 1: K/A 4: 3.04.000.074.EK1.03 K/A IMPORTANCE: 4.5/4.9 - Rev.1

m QuestionNumber 55 i When responding to a Degraded Core Cooling condition in FR-C.2, the operator is directed to " Verify SI Valve Alignment" with the SI system in the Cold Leg ! { Injection Mode. Which of the following lists the proper position of the l selected components? MOV-1CH-289 MOV-1SI-867D MOV-1CH-115B TV-1MS-105B Regen HX/Chg BIT Outlet RWST Discharge AIM Turbine Header Inlet Isolation to Charging Steam Supply CNMT Isolation Valve Pumps Suction B Train Valve Valve Trip Valve A. CLOSED OPEN CLOSED OPEN B. CLOSED OPEN OPEN OPEN C. CLOSED CLOSED OPEN CLOSED D. OPEN OPEN OPEN-OPEN o POINTS: ".00 TIME ALOTTMENT: minutes ANSWER: B. 1

REFERENCES:

10M-53A.1.1-A - Issue 1B, Rev. 1 l ILP-SQS-11.1 OBJECTIVE: 3 NUMBER: 1-97-038, New l l JTA f: K/A #: 3.04.000.074.EA1.27 K/A IMPORTANCE: 4.2/4.2 l Rev.I

..... ~.. .-. - -.. - ~ .. ~ ~. - ~ -.~.- ~.-. I l l QuenicmNumber 56 L 1 TV-1MS-101C, 'C' Loop Main. Steam Isolation Valve, inadvertently trips closed ' (. - with the Unit at 40% power /NSA line-up. The Rx tripped on low SGWL in the 'C' ,S/G and NO Safety Injection occurred. Determine which of the following will '3 occur to control S/G pressures. I [: A. The steam dumps will actuate on high RCS temperature and maintain ALL l S/G pressures <1005 psig by maintaining T o at,44 9'F' (2'F dead band j from no-load T.n) using steam from only 'A' 'B' S/G's. 1 B. The steam dumps will actuate on high RCS temperature and maintain 'A' 6.'B',S/G pressures <1005 psig and PCV-1MS-101C, the 'C' S/G atmospheric steam dump valve, will trip open if 'C' S/G pressure reaches 1060.psig. C. The steam dumps will actuate on high steam header pressure and l ' maintain *,' & 'B' S/G pressures <1005 psig and PCV-1MS-101C, the 'C' i A S/G atmospheric steam dump valve, will modulate open if 'C' S/G pressure. reaches 1035 psig. D. The steam dumps will' actuate on high steam header pressure and maintain 'A' & 'B' S/G pressures <1005 psig and the 'C' S/G safety valves will open in succession if 'C' S/G pressure reaches 1075 psig. t i i i l t - POINTS: 1.00 TIME ALOTTHENT: minutes ANSWER: B. 'i i i l

REFERENCES:

10M-21.2.B - Issue 4, Rev. 1, 10M-21.3.A - Issue 4, Rev. 4. l ILP-SQS-21.1-OBJECTIVE: 3 NUMBER: 1-97-039, New JTA #: K/A # 3.05.039.000.Kl.02 K/A IMPORTANCE: 3.3/3.3 i f i L' i i Rev.I

. ~ _.. _ _ -. _... _ - _.. Question Number' 57 -Given the following: _( The Unit-is in MODE 1 with Rx power at 8%. a Boron concentration is 665 ppm.~. e. Tave is 547'F. PZR Pressure is 2235 psig. PZR Level is 22%. -Steam dumps are in Auto in the Main Steam Pressure Control Mode, e S/G atmospheric relief valve controllers are in Manual. e e SGWL control is in manual on main feed. All systems are aligned in their normal lineup for the existing power e-level. l_ e 'An inadvertent MSLI occurs. I i Assuming no Rx Trip occurs and no operator action, determine the response of the. 'l l. following parameters immediately following the MSLI. Rx Power 'A' S/G 'A' SGWL 'PZR Level i Pressure i 1 l-A. DECREASE INCREASE INCREASE INCREASE ] l B. -INCREASE INCREASE INCREASE DECREASE l C. DECREASE INCREASE DECREASE ~ INCREASE D. INCREASE DECREASE DECREASE INCREASE r 4 I l l' POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C.

REFERENCES:

LP-RT-6, Objective 15 ILP-SOS-21.1 OBJECTIVE: 2 NUMBER: 1-97-040, New i ^JTA # K/A #: 3.05.041.020.K3.04 K/A IMPORTANCE: :3.5/3.4 l 3 l i i i Rev.I ~ _.

~- - ~ Question Number $8 4 I - The EHC auxiliary governor speed sensor circuits have failed to 104% of rated speed, causing the 20-1-OPC and 20-2-OPC scienoids to energize. Which of the { _ following describes the rerponse of the EHC system to this malfunction? A. ' The Control valve Emergency Trip Header will be continuously dumped, causing the Governor, Interceptor and Extraction Steam Non-Return Valves to close rapidly and remain closed, j s l B. The Control Valve Emergency Trip Header will be dumped for 1.5 seconds, causing the Governor, Interceptor and Extraction Steam Non-i Return Valves to close rapidly and then re-open. C. The Trip Valve Emergency Trip Header will be continuously dumped, causing the Reheat and Turbine Stop valves to close rapidly und remain j closed. j D. The Trip Valve Emergency Trip Header will be dumped for 1.5 seconds j out of every 31.5 seconds, causing the Reheat and Turbine Stop valves i to reduce load at 200%/ min. i 9 t i j POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. 4 e j

REFERENCES:

10M-26.1.B - Issue 4, Rev. 4 ILP-SQS-26.3 OBJECTIVE: 9d NUMBER: 1-97-041, New JTA 9: K/A 9: 3.05.045.050.A2.10 K/A IMPORTANCE: 2.7/2.9 g Rev.I

~. l, i L 'QuestionNumber 59 I, Given the following: i e The Unit'is stable at 854. l .All systems are in their at power, NSA configuration. e .The operator notices that FCV-1 W-498, 1C S/G feed reg valve, demand' e signal.is 15% lower than FCV-1W-478 and 488, the 1A and 1B S/G feed reg valves. The S/G 1evel chart recorders show that all three S/G levels have been s stable at 44% and-that all three S/G steam flows and feed flows have been stable at 85% for an extended period of time. Which of the following statements explains the observed plant conditions? A. 1C S/G controlling feed flow transmitter has failed low. l-B. FCV-1 W-499, 1C S/G feed reg valve bypass valve, has failed l open. C. There is a feed water leak upstream of FCV-1 W-498, 1C S/G feed reg valve. D. FCV-1FW-498, 1C S/G feed reg valve, valve stem has become I uncoupled from its actuator. l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B. j

REFERENCES:

10M-24.1.D - Issue 4, Rev. 1 l l '1LP-SQS-24.1 OBJECTIVE: 4 NUMBER: 1-97-042, M-0153 (- l JTA #: fj K/A #: 3.05.059.000.K1.04 (059A2.11) K/A IMPORTANCE: 3.4/3.4 e + ? Rey,1

6 )._ Questi:n Number '60 -

-Which of the~ following is the power. supply to ifM-P-3A,. Motor Driven Aux Feed Pump?e 3

'A. AE 4KV Bus. B. DF 4KV Bus. .C. 8N 480V Bus, j.. D. 9P 480V Bus. A i ' t b-1 5' ? N. 4 ii. POINTS: 1.00 TIME ALOTTMENT: minutes { . ANSWER: A. i i f - i i.

REFERENCES:

10M-24.3.C - Issue 4, Rev. 6 1LP-SQS-24.1 OBJECTIVE: 2 NUMBER: 1-97-043, New JTA 9: K/A 0 3.05.061.000.K2.02 K/A IMPORTANCE: 3.7/3.7 Rev I

Question Number 61 - l Which of the following would indicate a degradation in total AFW flow capability following an automatic Low PZR Pressure Rx Trip? + l A. TV-1MS-105A, AW Turbine Steam Supply Trip Valve OPEN. B. Annunciator A7-7, "STM UNAVAILABLE TURB DRIVEN FEED PP", NOT in alar.t.. C. MOV-1W-151C, 1B S/G AW Flow Control Valve, OPEN. D. FCV-1FW-103A, ifW-P-3A Motor Driven AW Pump Recirc Valve, OPEN. s POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. REFERENCES : 10M-24.2.B - Issue 4, Rev. 0 1LP-SQS-24.1 OBJECTIVE: 4 UUMBER: 1-97-044, New JTA #: K/A #: 3.05.061.000.A3.01 .K/A IMPORTANCE: 4.2/4.2 1 Rev.I

-~... -. .~ 1 QuestionNumber 62 Which of the following is the preferred method of restoring Rx Plant River Water 4-Header pressure in the event of a complete loss of the normal River Water Intake f structure? I' A. Aligning Unit 2 'B' Service Water Header to supply Unit 1 'A' Rx Plant River Water Header. 'i B. Aligning 1FP-P-2 Diesel Driven Fire Pump to supply Unit 1 'A' Rx Plant River Water Header. C. Aligning Unit 1 Aux River Water Pumps to supply that pumps respective Unit 1 Rx Plant River Water Header. ] D. Aligning Unit 1 Turbine Plant River' Water Pumps to supply that pumps respective Unit 1 Rx Plant River Water Header. i .i i i i s 1 POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C. 4 1 T -

REFERENCES:

10M-30.4.AAC 1LP-SQS-30.2 OBJECTIVE: 3 NUMBER: 1-97-045, New JTA i: K/A 4: 3.05.076.000.K1.21 K/A IMPORTANCE: 2.7/2.9 i Rev.1-

. = -. ... ~. ~ -.. . ~ 1 l Question Number 63 Which of the following oesign features protect plant personnel and Systems, Structures and Components (SSC's) outside of containment from the effects of a i l '. l 1.- High Energy Line Break? i ( A. Automatic S/G blowdown isolation on High Cable Vault Pipe Tunnel Area temperature. 1 B. Programming of SGWL to reduce the total mass in the S/G's at high power levels. C. The Main Steam Line Isolation on a High-2 containment pressure signal. D. 'Autonatic opening of the Main Steam Valve Room Louvers on high room temperature. ) l ) l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. 1

REFERENCES:

BVPS UFSAR Appendix. Di Pg. D.1 Rev. 1, and Pg. D.1 Rev. 4. I ILP-SQS-25.1-OBJECTIVE: 5 NUMBER: 1-97-046, New JTA # K/A # - 3.05.000.040.EKl.06 K/A IMPORTANCE: 3.7/3.8 l Rev.I r.

i . Question Number 64 l f 'The Unit is at 100% power with all systems in their at power, NSA 1 configurations. Which of the following control switches, at a MINIMUM, need to 2 { be placed in the CLOSE position to preyent 1A S/G from feeding the Main Steam Manifold? A. TV-1MS101A, 1A SG MAIN STEAM A TRN TRIP VLV. .B. TV-1MS101A, lA SG MAIN STEAM A TRN TRIP VLV, MOV-1MS-101A, 1A SG MAIN $ TEAM BYPASS TRIP VLV. C. TV-1MS101A, 1A SG MAIN STEAM J TRN

• RIP VLV,-

] TV-1MS101A, 1A SG MAIN STEAM B TRr4 TRIP VLV, l MOV-1MS-101A, IA SG MAIN STEAM BYPASS TRIP VLV. i D. TV-1MS101A, lA SG MAIN STEAM B TRN TRIP VLV, f MOV-1MS-101A, IA SG MAIN STEAM BYPASS TRIP VLV, TV-1MS-111A, IA MAIN. STEAM LINE PRE-NRTRN DRAIN ISOL VLV. l I l< POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. REFERENCES : 10M Figures 21-1 Rev. 10 and 21-06 Rev. 4 ILP-SQS-21.1 OBJECTIVE: 3 NUMBER: 1-97-047, New JTA #: K/A 4: 3.05.000.040.EA1.04 K/A IMPORTANCE: 4.3/4.3 ' hv,1

QuestionNumber 65 Refer to 10M-26.4. AAS " Condenser vacuum Low" alarm response procedure to answer the following question, f Assuming all turbine vibration reading's are normal, which of the following sustained conditions would require a manual Turbine and/or Rx Trip IAW 10M-26.4. AAS " Condenser Vacuum Low" ARP7 I Main Generator Condenser l Output (MW) Backpressure ] (In. Hg Abs.) l 1. 210 4.0 2. 800 4.5 3. 800 6.5 I 4. 210 2.5 l A. 1, 2 and 4 ONLY. B. 1 and 3 ONLY. C. 3 ONLY. D. 2 and 4 ONLY. i POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C. REFEPENCES: 10M-26.4.AAS - Issue 3, Rev. 3 j. ILP-SOS-26.2 OBJECTIVE: 9d NUMBER: 1-97-048, New JTA #: K/A 8: 3.05.000.051.EA2.02 K/A IMPORTANCE: 3.9/4.1 Rev.1,

~ _4.__ ._ ~ ___.m.. ~ -. -..... _ - - - m_ QuestionNumber 66 Control for MOV-1W-151B, 1C S/G 'A' Train Auxiliary Feedwater Regulating Valve, has been' transferred to the Emergency Shutdown Panel (SDP). What actions are ( necessary to transfer control of MOV-1 W-151B back to the Main Control Room Benchboard (BB-C)? i A. Operating the control switch on BB-C for MOV-1 W-151B out of the NORMAL position. B. Depressing the Transfer Pushbutton for MOV-1W-151B on the SDP. 4 C. Resetting the Master Reset Transfer Relay on the SDP'. D. Resetting the SDP Transfer Relay for MOV-1 W-151B at the respective Aux Relay Panel. I l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D.' 1

REFERENCES:

-10M-24.1.D - Issue 4 Rev. 1 1LP-SQS-24.1 OBJECTIVE: 4 NUMBER: 1-97-049, New

'JTA 8: K/A 8: 3.05.000.054.EA1.01 K/A IMPORTANCE: 4.5/4.4 L h e

~ ~ ~... . - - -.. ~.. - ~. i ' Question Number 67 With the Unit... riode 5,. which of the following automatic actions will occur when RM-IVS-104A, the Containment Purge Exhaust Monitor, goes into a High-High .I alarm conditionir A. A CIA signal will ba generated isolating all Phase-A flowpaths from the containment. B. Containment Purge Supply and Exhaust Fans will trip, and the Purge Supply and Exhaust Dampers will close. C. The Supplemental Leak Collection and Release System (SLCRS) will be aligned to bypass the Main Filter BLnk and provide an elevated release .flowpath, 3-1 D. The Containment Purge Exhaust will be aligned to the SLCRS and then filtered through the Main Filter Banks. 4 4 1 J 1 i i 4 ( I 4 .l.00 TIME ALOTTMENT: ~ minutes POINTS: j j ANSWER: B. 4 i ? I

REFERENCES:

lOM-44C.1.B - Issue 4, Rev. 0 ILP-SQS-44C (2382) OBJECTIVE: 3 NUMBER: 1-97-050,.New JTA # K/A'#: 3.06.022.000.K4.03 K/A IMPORTANCE: 3.6/4.0 I I Rev.1-

Question Number 68 Refer to Unit 1 Technical Specification ficare 3.6-1 to answer the following I., question. Actual containment air partial pressure is 9.8 psia. If river water temperature were to rise from 72*F to 77'r, the required Maximum Allowable Operating Air Partial Pressure would be '(1) psia and the actual containment air partial pressure (2) meet the Technical Specification requirement. (1) (2) A. 9.7 would NOT B. 9.7 would C. 9.9 would NOT D. 9.9 would i i POINTS: 1.00-TIME ALOTTMENT: minutes ANSWER: 'A. I

REFERENCES:

BVPS TS, Figure 3.6-1, Amendment No. 174 1LP-SQS-12.1' OBJECTIVE: 10 NUMBER: 1-97-051, New JTA 4: I .[l K/A'#: 3.06.022.000.A1.04 K/A IMPORTANCE: 3.2/3.3 Rev.I

.. ~ - ..._.-...-....x._.. ..__~.. -... - -.... _ - . -.... ~, -. - ~. 4-l - Question Number 69 I,', The bases for the OPERABILITY of the Containment Quench and Recirculation Spray j Systems is to ensure containment depressurization and subsequent return to which f of the following? s A. subatmospheric pressure in the event of a Main Steam Line 3reak inside containment. I B. atmospheric pressure in the event of a Main Steam Line Break inside containment. C. subatmospheric pressure in_the event cf a Loss of Rx Coolant Accident. j D. atmospheric pressure in the event of a Loss of Rx Coolant Accident. 2 4 4' = I-b l I i I POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C. l ~

REFERENCES:

BVPS Unit 1 TS, 3/4.6.2 Bases, Amendment No. 200 ILP-SOS-13.1 OBJECTIVE: 11 NUMBER: 1-97-052, New JTA f: K/A #: 3.06.026.000.G06 K/A IMPORTAhCE: 2.5/3.8 Rev, I

m.... f QuestisnNumber 70 .The zirc-water reaction of the fuel clad is one of the major sources of Hydrogen 4 . H2) generation in the containment following the design bases Loss of Rx Coolant ( ^ Accident (DBA LOCA). Which of the following are the.other major sources of Hydrogen generation.in the containment following the DBA LOCA? ~ A. Radiolysis of the containment eump and Rx coolant water, H2 gas that has accumulated in the Pzr gas space, Corrosion of aluminum and zine in the containment. - I B. Radiolysis of the containment sump and Rx coolant water, H2" gas that- ] has accumulated in the Pzr gas space, H2 released from the assumed 10% i failed fuel pins, i 1 C. Radiolysis of the containment sump and Rx coolant water, H2 gas that e has come out of solution from the RWST injection water, Corrosion of aluminum and zinc in containment. D. H2 gas that has accumulated in the Pzr gas space, H2 released from the j assumed 10% failed fuel pins, Corrosion of aluminum and zine in j containment. 4 1 i s i t i i h 1 l. i POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. 1 -) i t e j 4-1 i-1 i f i I REFERENCES. UFSAR Section 14.3, page 14.3-39, Rev. 12. ILP-SQS-ATA4.2 OBJECTIVE: 1 NUMBER: 1-97-054, M-0006 4 JTA f: K/A 4: 3.06.028.000.K5.03 (028K5.03) K/A IMPORTANCE: 2.9/3.6 4 4 4 Rev.1.

......... -.. ~ _ _.. ~ -. .~ l l QuestisnNmmber 71 1 During routine' maintenance, TV-1CH-200B, 60 GPM Ltdn Orifice Cnmt Isol Viv, is I - to be stroked' closed and timed by de-energizing SOV-1CH-200. One 60 GPM, and i the 45 gpm Letdown Orifices are in service with PCV-1CH-145, LP Ltdn Back Press j Reg Viv, in MANUAL. In order to maintain letdown system stability when TV-1CH-200B is. going closed, PCV-1CH-145 should be throttled (1) , to (2)- letdown pressure as indicated on PI-1CH-145, Letdown Pressure, (BB-A). (1) (2) 1 A. CLOSED ~ RAISE B. CLOSED LOWER C. OPEN . RAISE D. OPEN LOWER ' e POINTS: 1.00 TIME ALOTTMENT: minutes i ANSWER: A. i l a i i

REFERENCES:

10ST-47.3A - Issue 4, Rev. 15. l ILP-SQS-7.1 OBJECTIVE: 10 NUMBER: 1-97-055, New l l i-' JTA i: K/A is 3.06.103.000.A4.01 K/A IMPORTANCE: 3.2/3.3 f Rev.! --n-,.. -.m.- e ~.,, -, ~ ~., -,

-.. - -.. ~. - -. -.. _ -. -... ~... -.. ~ ~ -. ~.. - -.. -... Question Number 72 I l - ) i 4 The inner and outer doors of which of the following Concainment Airlocks, are l

. interlocked to prevent both doors from being ipened at the same time?

l l

1. Normal Personnel Air Lock-~84-Inch Full Size Doors.

i 1 f 2.~ Normal Personnel Air Lock 18-Inch Escape Manway Doors. 3. Equipment Hatch Emergency Air Lock Doors. i A. 'I and 2 ONLY. i B. 1 and 3 ONLY. 1 C. 2 and 3 ONLY. D. 1, 2 and 3. i l i POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B. i I i 1

REFERENCES:

10M-47.4.B - Issue 4,-Rev. 3, 10M-47.4.C - Issue 4, Rev. 1 and 10M-47.4.D - Issue 4, Rev. 1. 1LP-SQS-47.1 OBJECTIVE: 4 NUMBER: 1-97-056, New JTA 8: K/A 8: 3.06.000.069.EK2.03 K/A IMPORTANCE: 2.8/2.9 1 i Rev.I i r m. i

.~ Chiestion Nmmber 73 Refer to VOND Figures; 7-1 Rev. 9, and 15-3 Rev. 2, and Section 3.9 of the Unit i 1 Technical Specifications to answer the following question. l' Under which of the following conditions would Technical Specification 3.9.4, . CONTAINMENT BUILDING PENETRATIONS, be satisfied and allow CORE ALTERATIONS to commence? NOTE: Assume there are NO blank flanges or pipe caps installed on equipment that is disassembled or removed. A. e 1A S/G secondary side manway removed, TV-1MS-101A and MOV-1MS-101A, the 1A S/G MS Trip and Bypass e Valves are CLOSED,.and All 1A S/G pressure transmitter sensing lines are removed e from the steam lines to install new pressure taps. i B. e RV-1CC-113A, CRDM Shroud Cooling Coil relief valve is

removed, 1CCR-188, CRDM Shroud Cooling outlet isolation valve is CLOSED, and e

MOV-1CC-111A, CRDM Shroud Cooling coil Inlet isolation l valve is CLOSED. C. e The CVCS charging line removed from the Regen Heat Exchanger, 1CH-390, charging line vent valve locked OPEN, and MOV-1CH-289, Charging Line Containment Isolation Valve, is e CLOSED but inoperable due to its motor operator leads being lifted. D. Containment Equipment Hatch installed with three closure bolts, and both doors on the Emergency Personnel Access Hatch CLOSED, and All containment purge dampers are CLOSED. e POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B.

REFERENCES:

10M Figures; 7-1 Rev. 9, 15-3 Rev. 2, and 21-1 Rev. 10, and BVPS Unit 1 TS 3.9.4 Amendment No. 185. 1LP-SQS-6.13 OBJECTIVE: 12 NUMBER: 1-97-057, New JTA 9: K/A 9: 3.06.000.069.EA2.02 K/A IMPORTANCE: 3.9/4.4 Rey,1 I o

I QuestionNumber 74 4 All of ths following Emergency Diesel Generator (EDG) conditions will prevent { the EDG output. breaker from closing during a loss of all offsite power EXCEPT a(n) A. Elects:ical Engine Overspeed Trip Signal. B. Generator Overcurrent Signal. 4 C. Generator Output Voltage at 2.1Xv. D. Engine Oil Low Pressure Trip Signal. 4 f. i d i. t POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. i 1 i si I

REFERENCES:

lOM-36.1.E - Issue 4, Rev. 1. ILP-SOS-36.2 OBJECTIVE ~ 6 NUM*ER: 1-97-058, New JTA #: il K/A 6: 3.07.062.000.K3.02 K/A IMPORTANCE: 4~1/4.4 4 Rev.1

._...._,_.=...____.___..._._m_.- i t Questi:nNumber 75 .Which of the following describes the relationship between the stations 125VDC I batteries and their respective battery chargers? In the'NSA configuration, the Battery Charger Output Breaker is normally A. OPEN, Land CLOSES automatically to charge the battery when battery voltage drops below a preset value. B. CLOSED,' allowing the. charger to supply the normal DC loads, and remains CLOSED on a' loss of AC input power to the charger. C. CLOSED, allowing the charger to maintain a continuous equalizing -charge on the battery, and remains CLOSED on a loss of AC input power to the charger. I D. CLOSED, allowing the charger to maintain a float charge on the battery, and OPENS on a loss of AC input power to the charger. l l J l 1 l POINTS: 1.00 . TIME ALOTTMENT: minutes ANSWER: B. i l i

REFERENCES:

10M-39.1 - Issue 4, Rev. 0 ILP-SOS-39.1 OBJECTIVE: 1 NUMBER: 1-97-059, New JTA #3 .y - K/A 9: 3.01.063.000.Kl.03 K/A IMPORTANCE: 2.9/3.5 Rev,I' l- ~-

- ~. M QuestionNumber 76 b With the Emergency Diesel Generator (EDG) LOCAL / REMOTE keylock switch in the LOCAL position, the EDG will - -{ ~ A. START on an SI or UNDERVOLTAGE signal, but the output breaker-WILL NOT automatically close. B. START on an SI or UNDERVOLTAGE signal, and the otitput breaker WILL automatically close. C. NOT START on an SI signal, but WILL START on an UNDERVO2.TAGE' signal, and the output breaker WILL automatically close. D. NOT START on an SI or UNDERVOLTAGE signal, therefore the output breaker WILL NOT automatically close. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. I

REFERENCES:

lOM-36.1.E - Issue 4, Rev. 1 ILP-SQS-36.2 OBJECTIVE! 7 NUMBER: 1-97-060, M-0664 'JTA 9: K/A 9: 3.07.064,000.A4.01 (064A4.01) K/A IMPORTANCE: 4.0/4.3 Rev.I

Questi8nNumber 77 Given the following f A loss of all AC power-in conjunction with a steam line break inside e containment, has occurred 23 minutes ago. There are indications of a steam void in the Rx Vessel head. All SG narrow range water levels are still <5%. Containment pressure is 8.5 psig. The 1A Emergency Diesel Generator has just been started locally. In accordance with ECA-0.0 " Loss of All AC Power", which of the following states which pump should be started / verified running FIRST, and the bases for this action?_ A. Charging /HHSI pump, to collapse the Rx Vessel head void. B. Motor Driven Aux. Feedwater pump, to establish a Heat Sink. C. River Water pump, to provide cooling to the EDG's. D. Quench Spray pump, to reduce containment pressure. i. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C.

REFERENCES:

10M-53B.4.ECA-0.0 - Issue 1B, Rev. 3a ILP-SQS-53.3 OBJECTIVE: 4 NUMBER: 1-97-061, M-0561 JTA #: K/A is 3.07.000.055.EA1.06 (064K1.02) K/A IMPORTANCE: 4.1/4.5 BWy.1 i l

Qu!stion Number 78 - Given the following -{- There has been an explosion in the high voltage switchyard resulting in the total loss of offsite' power to both BVPS Units. There are NO Safety Injections in progress on either Unit. ' Emergency Diesel Generator (EDG) status'is; Unit 1 fl Running with its output breaker on fire and open. Unit 1 82 On Clearance and disassembled. Unit 2 41 Running and loaded at 2300KW. Unit 2 92 Running and loaded at 2650KW. ERF Black Diesel Running and loaded at 1700KW. Which of the following describes the actions necessary to restore.lbnited power . to the IDF 4KV bus? Establish the SBC Crcss-tie by aligning the IDF bus _ A. directly to the 2DF bus. B. to the ID bus, the 1D bus to.the 2D bus, and the 2D bus directly to the 2DF bus. .C. to the ID bus, the 1D bus to the 2D bus, and the 2D bus directly to the 2AE bus. D. to the 1D bus, the 1D bus to the 2D bus, and the 2D bus ) directly to the ERF Black Diesel. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B.

REFERENCES:

10M-53A.1.2-M-DF - Issue 1B, Rev. 1 1LP-SOS-53.3 -. OBJECTIVE : 3 NUMBER: 1-97-062, New JTA #: .K/A 0: 3.07.000.055.EA2.03 K/A IMPORTANCE: 3.9/4.7 Rev.!

l. __._.___m.__._..___.__._. l l QuestisnNumber 79 l ECA-0.0 " Loss of All Emergency 4KV AC Power" directs the operator to locally close MOV-lCH-381 Seal Water Return Containment Isolation valve. Which of the .{ following is the hares for performing this step? 'A. To prevent-ovar-pressurizing and possibly rupturing the VCT. B. To minimize tie potential for a radioactive release within the Aux. i - . Building. C. To minimize the chance of RCP seal damage when seal injection and CCR flow is isolated. D. To prevent steam formation on'the CCR side of the Seal Water Heat ~ Exchanger. i I l l. l POINTS: L1.00 TIME ALOTTMENT: minutes . ANSWER: B. i

REFERENCES:

lOM-53B.4.ECA-0.0 - Issue IB, Rev. 3 lLP-SQS-53.3 OBJECTIVE: 3 NUMBER: 1-97-063, New JTA 9: K/A 9: 3.07.000.056.EK3.02 K/A IMPORTANCE: 4.4/4.7 I I- . Rev.1 l -~

QuestionNumber 80 A loss of the 120vac Vital Bus II has occurred. The operators are responding to the loss of power in accordance with ARP 10M-38.4.AAC, Vital Bus II Trouble. i l They are directed to place the 120vac Vital Bus II inverter ' Man Bypass Switch' to the 'STBY-ISOL' position. This switch position transfers the Vital Bus power source from the (1) to the (2) and when the normal power cource is restored, allows a(n) (3) transfer back to the normal power source. (1) (2) (3) A. AC Static Line Inverter Output Automatic i Regulator l B. AC Static Line Inverter Output Manual Regulator C. Inverter Output AC Static Line Automatic Regulator D. Inverter Output AC Static Line Manual Regulator POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. l l

REFERENCES:

10M-38.4.AAC - Issue 4, Rev. 1, and Figure 30 Issue 3, Rev. 6. lLP-SOS-38.1 OBJECTIVE: 2 NUMBER: 1-97-064, New JTA 8: K/A 4: 3.07.000.057.EK3.01 K/A IMPORTANCE: 4.1/4.4 Rev.1 I

Questi:n Phunber 81 l 4 All of the following will increase the 125VDC battery capacity EXCEPT. A. disconnecting from the battery, individual loads that have indications of being grounded. B. periodically performing a deep cycle and battery equalizing charge. C. ensuring the individual cell electrolyte levels fall below the minimum level before refilling. D. maintaining the batteries on a continuous float charge. FOINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C.

REFERENCES:

lOM-39.1.D - Issue 4, Rev. 1 ILP-SQS-39.1 OBJECTIVE: 2 NUMBER: 1-97-065, New JTA 4: K/A #: 3.07.000.058.EKl.01 K/A IMPORTANCE: 2.8/3.1 Rev.I

. Question Number 82 i Which of the following describes the flowpath and proper order of components in the Station and Instrument compressed air systems? -A A, The Station Air Compressors, Station Air Receivers, Compressed Air Dryers, Instrument Air System. 'B. The Station Air Compressors, Compressed Air Dryers, Station Air Receivers, Instrument Air System. C. TheInstrumentAir' Compressors,StationAirReceivers,ComphessedAir Dryers, Station Air System. D. - The Station Air Compressors, Compressed Air Dryers, Instrument Air System, Station Air Receivers, 1 POINTS: 1.00 TIME ALOTTMENT: minutes l ANSWER: A. i '

REFERENCES:

10M Figure 34-1 Rev. 6, and 34-2 Rev. 7 - 1LP-SQS-34.1 OBJECTIVE: 2 NUMBER: 1-97-066, New JTA #: ( K/A 6: 3.08.079.000.K1.01 K/A IMPORTANCE: 3.0/3.1 -1 i Rev.I i: r-,.., --n-~- - - -

..- -. - -.. -. - -. - ~ ~ - ~ ~ ---- - - - 1 g Chiestirn Number 83 j 1 Refer to 10M-34.4.L

• Placing Diesel Air Compressor Into Service" to answer the f

following question. 1 4 - - - ' The Diesel Air Compressor has a low engine oil pressure trip for engine protection. What allows the Diesel Air Compressor to start with a standing low oil pressure signal before engine oil pressure has a chance to build up? 'The Engine Low Oil Pressure Trip is A. automatically bypassed for the first 15 seconds of operation to allow i l oil pressure to build up.- B. cleared when the operator starts the auxiliary oil pump prior to diesel engine start up. 1 C. manually bypassed by the operator when the ' Diesel Air Compressor Start Pushbutton' is depressed. -D. manually bypassed by the operator by maintaining the ' Diesel Air Compressor Selector Switch' in the ' HOLD' position. I l j I e I 1 1 i i I 1 . POINTS:. ' TIME ALOTTMENT: minutes 1.00 i . D. j.. ANSWER: }. I l 1 a 4 4 REFERENCES : 10M-34.4.L - Issue 4, Rev. 1 -- 1LP-SQS-3 4.1 OBJECTIVE: $ NUMBER: 1-97-067, New .JTA 6: K/A 9: 3.08.000.065.EA1.04 K/A IMPORTANCE: 3.5/3.4 d Rev.I i

_...__._.__..m= _. _ _ _ _. = _ _ _. _ _ _ _ _ _. _ _ _. _ _ _ _ - = _ _. _ - - _,. l l' Question Number 84 Assume initial conditions of 100% power, with all systems in their NSA configuration. A. ALL of the Train B, Output Bay Slave Relsys will not function on a Safety Injection signal. B. ONLY the 81 Emergency Diesel Generator Load Sequencer will load the required components on a Safety Injection signal. i C. ALL of the Train B, Input Bay Relays will de-energize resulting in a Rx Trip and Safety Injection. D. Train B will function as designed due to the auctioneered power supplies to the Logic and Output Bays. l [ i i: l .I i I POINTS: 1.00. TIME ALOTTMENT: minutes ANSWER: A. j

REFERENCES:

10M Figure 1 Rev. 8,, ILP-SQS-1.2 OBJECTIVE: 2 NUMBER: 1-97-069, New f JTA #: ~ K/A #: 3.09.012.000.K2.01 K/A IMPORTANCE: 3.3/3.7 l.! i Rev.1

l j - Question Num'oer 85 Refer to Unit 1 Technical Specification section 3/4.9 - Refueling Operations, to answer the following question. k~ ~ 'Given the following The Unit is in Mode 6, conducting a core off-load. l N-43 power range instrument is on clearance for maintenance.- e l. ~ N31 and N32 HV Manual On/off switches are in Normal. 1 e-l 'All other systems are in their normal line-ups for the current Mode of ) e operation. 1 The 120vac Vital Bus IV is inadvertently de-energized. j

e-Following the loss of the 120vac Vital Bus IV, Core Alterations may_

A. continue, the power range nuclear instruments.are not required by the refueling Technical Specifications. B. continue, provided boron concentration of the RCS is determined -at least once per 12 hours. C. NOT continue, due to the loss of both source range nuclear j instruments. j i D. ' NOT continue, due to the inability to actuate a complete containment phase A isolation (CIA). i POINTS: 1.00 TIME ALOTTMENT: minutes ANSWFR: C.

REFERENCES:

10M-1.5.B.1, Table 1 Issue 2, Rev. 8 s TS 3.9.2. Amend No. 175. ILP-SOS-2.1 OBJECTIVE: 2 NUMBER: 1-97-070, New JTA 6: l (- K/A #2 3'09.015.000.K2.01 K/A IMPORTANCE: 3.3/3.7 -Rev.1

-.. _ - = - - - -. ~... - Questi:n Numter 86 Refer to Unit 1 Technical Specification 3.3.1.1, Table 3.3-1, to answer the following question. y Given the following: The Unit is in Mode 3 with the shutdown banks fulli withdrawn, preparing to enter Mode 2. Source Range counts are: N31, 120 cps N32, 130 cps e A review of the most recent N36 nuclear intermediate range (IR) e channel functional test (OST 1.2.2) indicates the "as-left" setting for the IR high neutron flux trip is equivalent to 33% power. (3.5 x 10E-4 amps). Which of the following describes the Technical Specifications required actions for this condition? A. Adjust the N36 high flux trip setpoint to the current equivalent of 25% rated thermal power prior to raising power above P-6. B. Adjust the N36 high flux trip setpoint to the current equivalent of 25% l rated thermal power prior to raising power above 5% rated thermal power. C. Place N36 in the tripped condition when greater than P-10 and the low power trip setpoints have been blocked. D. Place the N36 ' Level Trip Bypass' switch in the.' BYPASS' position prior to exceeding P-6, 4 4 F i POINTS: 1.00 TIME ALOTTHENT: minutes ANSWER: A. l 2

REFERENCES:

Unit 1 TS 3.3.1.1, Table 3.3 Amendment No. 195 1LP-SQS-TS OBJECTIVE: 1 NUMBER: 1-97-071, M-0270 JTA # { K/A #: 3.09.015.000.G05 (2.2.22) K/A IMPORTANCE: 3.3/3.8 Rev.I

. _..-~~~ - ~.- -. -..- _ ~.-.-.....~ ., - - -.~ ~.. -. Question Number 87 3

r..

Given the following: ' {' The Unit is'at 100% power. s All systems are in their NSA configuration. A spurious reactor trip with no Safety Injection occurs. Following the Rx tript 1 The turbine trip was delayed for five seconds, causing T... to drop to e 540*F. Two cooling tower pumps have tripped. '1A RCS loop hot leg temperature instrument has failed high. Assuming no operator action, which one of the following describes how RCS T. will be maintained? i A. At 547'T by the steam dumps with ONLY the cooldown valves (PCV-1MS- '106A,B,CB1) armed. B. At 547'T by the steam dumps with the first 2 banks of the steam dump valves armed. C. By ONLY the atmospheric steam dump valves (PCV-1HS-101 A,B,&C) cycling at their trip open setpoint. D. At 543*F by the condenser steam dumps cycling open due to the failed high Taos instrument and closed by the Lo-Lo T.,. interlock. l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: B. -l

REFERENCES:

10M Figure 21.1.D - Issue 4, Rev. 1. 1LP-SQS-21.1 OBJECTIVE: 3, 4 NUMBER: 1-97-072, M-0188 JTA 0: .'h K/A 0: 3.09.016.000.K3.03 (041K4.09) K/A IMPORTANCE: 3.0/3.1 e .Rev.I-

~ - _ - -. =.. QuestionNumber 88 l l l From the parameter trends listed below, which. set of conditions would be positive verification of a natural circulation condition existing within the f RCS? t l S/G Core Exit RCS RCS Pressures TC' s T-Hot's T-Cold' s At T-Sat A. DROPPING RISING STABLE for S/G Pressure l At T-Sat i B. RISING DROPPING for S/G STABLE l Pressure At T-Sat C. DROPPING STABLE STABLE for S/G Pressure At T-Sat D. STABLE for S/G DROPPING DROPPING Pressure i l k POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C.

REFERENCES:

10M-53.A.1 Attachment 2-G - Issue 1B, Rev. 1. l 1LP-SOS-53.2 OBJECTIVE: 11 NUMBER: 1-91-073, M-0171 l JTA ft 'K/A #: 3.09.017.020.A3.01 (E09EA1.20) K/A IMPORTANCE: 3.6/3.8 l Rev.I

. __.--. ~. _ _ ...m.. - Question Number 89 Which of the following actions should occur if RM-1RM-207, Fuel Pool Bridge Crane Radiation Monitor, were to fail high while raising a spent fuel assembly 3 out of the fuel transfer cart? All (1) fuel handling crane movement is (2) and a High-High activity alarm (3) actuated. f \\ (1) (2) (3) 1 A. UPWARD Automatically IS i Stopped B. UPWARD and Automatically IS DOWNWARD Stopped i l C. UPWARD and Unaffected IS NOT DOWNWARD D. UPWARD and Unaffected IS DOWNWARD l l 1 POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D.

REFERENCES:

10M-43.1.E - Issue 4, Rev. 4. l-1LP-SQS-43.1 OBJECTIVE: 6 NUMBER: 1-97-074, New JTA f: (; K/A f: 3.09.072.000.K3.02 K/A IMPORTANCE: 3.1/3.5 3 i-Rev.I

.. ~. -. =..-.. _. - _....-.. -- . Question Number' 90 s Which of the following lists of switch positions is the 100% power NSA switch . lineup for the N31 Source Range Nuclear Instrument Drawer? k HV MANUAL LEVEL HIGH FLUX i ON/OFF TRIP AT SHUTDOWN s 1 ) A. NORMAL NORMAL BLOCK 1 B. HV OFF NORMAL BLOCK C. HV OFF BYPASS BLOCK 4 D. NORMAL NORMAL NORMAL 's i .( i 1 i I t- ? ' POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. i l j

REFERENCES:

10M-2.3.C - Issue 4, Rev. 2. 1 ILP-SOS-2.1 OBJECTIVE: 3 NUMBER: 1-97-075, New i-. JTA #1 K/A # 3.09.000.032.EK2.01 K/A IMPORTANCE: 2.7/3.1 Rev.I m e

.~_. l QuestionNumber 91 l-i ( i N35, Intermediate Range Nuclear Instrument Channel I, compensating voltage is set excessively HIGH. Following a Rx trip, N35 will indicate A. HIGH, preventing the source range from automatically energizing. B. HIGH, and the source range will be energized when N36 is <P-6. C. LOW, and the source range will be energized when N35 is <P-6. Il. D.' LOW, and the source range will be energized when N36 is <P-6.. l l l ll' l POINTS: 1.00 TIME ALOTTHENT: minutes ANSWER: D. ]

REFERENCES:

lOM-2.1.C - Issue 4, Rev. 1. j ILP-SQS-2.1 OBJECTIVE: 4&5 NUMBER: 1-97-076, M-0030 JTA 6: 1 K/A 6: '3.09.000.033.EA2.ll (015K4.07) K/A IMPORTANCE: 3.1/3.4 Rey, I '

l-QuestitnNumber 92 i i Which of the following instrument failures would require entry into a Technical (. Specification Action Statement? l. k. i l A. The Auxiliary Feedwater Pump Turbine Exhaust Radiation Monitor fails (' low during a plant cooldown with RCS temperature at 290*F. B. The Waste Gas Decay Tank Hydrogen Monitor fails low during an RCS degas operation. C. The Pressurizer Surgc Line Temperature instrument fails low during steady-state operations at 100% power. I D. WT-1TK-26, Domineralized Water Storage Tank level transmitter fails low during a steam plant heatup. i POINTS: 1.00 TIME ALOTTMENT: minutes j ANSWERt A. 1

REFERENCES:

BVPS UNIT 3.TS 3.3.3.1, Table 3.3-6-2.c.v - Amendment No. 59 1LP-SQS-43.1 OBJECTIVE: 7 NUMBER: 1-97-077, New JTA #: K/A #: 3.09.000.061. GOB K/A IMPORTANCE: 2.6/3.3 i ! l' Rev.'l

~.... ~ - -.- -..~-.-.. -. -, _ - - -... t i Question Number 93 With the Unit at 100% power and all systems in.their at power, NSA i configuration, a tube leak has developed in the Non-Regenerative Heat Exchanger. 'l With no operator action,'this will result in a (1) in CCR Surge Tank i Level, resulting in the CCR Surge Tank (2) '(1) (2) i A. RISE, overflowing to the Auxiliary Building Sump. l B. RISE, relief valve lifting and relieving to the Gaseous Waste Surge Tank. ) C. DROP, being automatically made up to from the Primary Water Supply Pumps. D. DROP, going out off scale low and etusing cavitation damage to the CCR . Pumps. 1 l i, POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C.

REFERENCES:

10M-15.1.C - Issue 4, Rev l' 1LP-SQS-15.1 ' OBJECTIVE: 369' NUMBER: 1-97-078, New JTA #. K/A et.3.10.008.000.A1.04 K/A IMPORTANCE: 3.1/3.2 Rev.l _. ~, _, -

IQuestirnNumber 94 Complete the following'for the complete loss of the Unit 1 River Water Intake Structure. y -f Cooling Tower Basin water level will-A. remain constant, makeup will be automatically provided by the Turbine ' Plant River Water System. 4 'B. ~ remain constant, makeup will be provided by the automatic start of the Auxiliary River Water System. C. drop-until makeup can be manually aligned from the Turbine Plant River Water System. D. drop until makeup can be provided by the manual start of the Auxiliary River Water System. i /, l POINTS: 1.00 TIME ALOTTHENT: minutes ANSWER: D.

REFERENCES:

10M-30.1.B - Issue 4, Rev. 2. ILP-SQS-30.2 OBJECTIVE: 368 NUMBER: 1-97-079, New JTA'6: e K/A 5: 3.10.075.000.A2.01 K/A IMPORTANCE: 3.0/3.2 Rev.1

~. - Question Number 95 - 1 Under w'hich of the fallowing conditions do the Emergency Operating Procedures allow an operating RCP, to REMAIN running, when ALL CCR flow to that RCP is i lost? j 'A. At ALL times, provided adequate HHSI flow AND seal injection flow-can be maintained. i B. During a SGTR,' AITER the RCS depressurization has commenced, if the s.CS/ Highest SG D/P drops to < 150 paid. ) r C. During a response to Inadequate Core Cooling, if high pressure t injection flow AND an adequate heat sink CANNOT be established. I d D. When responding to a Loss of Emergency Coolant Recirculation, and ALL safety injection flow is lost due to t ~ t depAetion of the RWST. i 4 . i i 7 1 + POINTS: 1.00 TIME ATCTTMENT: minutes ANSWER: C. t

REFERENCES:

10M-53A.1.2 M - Issue IB, Rev. 2 (Note) ILP-SOS-53.3 OBJECTIVE: 4 NUMBER: 1-97-080, New JTA 9:

3..

q K/A # - 3.10.000.026.EK3.03 K/A IMPORTANCE: 4.0/4.2 g , Rev, : 5

..-. ~ QuestionNumtx:r 96 Refer to Window Al-50, 10M-9.4.AAB "Incore Instrument Room Sump Level High" and - Window Al-73,10M-15.4. AAD " Neutron Shield Expansion Tank Level Low" ARP's to j. answer the following question. With the Unit at 100% power and all systems in their at power, NSA configuration, Annunciators Al-73, NEUTRON SHIELD EXPANSION TANK LEVEL LOW, and 71-50, INCORE INSTRUMENT RM SUMP LEVEL HIGH, are received. Additionally, gross leakage is indicated by a rapidly dropping indication on [LIS-INS-101), Neutron i Shield Expansion Tank Level Indicator. What are the required actions in l accordance with the associated ARP's? l A. Perform a plant shutdown at a rate determined by ti NSS/ANSS in ar;ecrdance with 10M-51.4, " Station Shutdown". I B. Perform a plant shutdown at a rate detennined by the NSS/ANSS in accordance with 10M-53C.4, AOP 1.51.1, " Emergency Shutdown". 1 C. Restore level to the normal range by operating [TV-1NS-101), Neutron Shield Tank Makeup Valve. D. Manually trip the reactor and proceed to 10M-53A.1, E-0 " Reactor Trip 4, or Safety Injection." i e i i POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: P. 4 1 '4 1-

REFERENCES:

10M-15.4.AAD - Issue 4, Rev. 1. 1LP-SQS-15.1 OBJECTIVE: 10 NUMBER: 1-97-081, New

JTA #:

K/A #: '3.10.000.026.G05 K/A IMPORTANCE: 3.3/3.4 ' Rev. I

.~ ~. - -. . ~. 4 .QuestionNumber 97 Which of-the following design features ensures K-eff remains less than or equal 'to 0.95 in the spent fuel pool with irradiated fuel in the pool? ~ $ 1. A minimum of 2000 ppm boron concentration in the pool. i. 2. The 'Boral' fuel rack installed neutron poison. 3. A minimum center-to-center distance between fuel assemblies. A. I and 2 ONLY. B. 1 and 3 ONLY. ~ C. 2 and 3 ONLY. ~ D. 1, 2 and 3. POINTS: 1.00 TIME ALOTTMENT: minutes 1 ANSWER: C. i

REFERENCES:

10M-20.1.B - Issue 4, Rev. 3. 1LP-SQS-20.1 OBJECTIVE: 167 NUMBER: 1-97-083, New .JTA 8: K/A 8: 3.11.033.000.K4.05 'K/A IMPORTANCE: 3.1/3.3 g Rev l-

... ~._ _. QuestionNumber 98 Which of the following is a function of the Refueling Manipulator Crane? ..k' Transfer' fuel assemblies froin the Reactor to the Fuel Transfer System A. + Upender. B. Transfer fuel assemblies from the Reactor Containment to the Fuel Handling Building. C. Provide the motive force to raise and lower the Burnable Poison Rod Assembly Handling Tool. D. Provide the motive force to raise and lower the Reactor Upper Internals. i i l h \\ 1 1 POINTS: 1.00 TIME ALOTTMENT: ' minutes ANSWER:- A, ) 1 i l i E7?ERENCES: 1RP-10R-3.3 - Issue 0, Rev. O. 11P-SOS-6.13 OBJECTIVE: 9 NUMBER: 1-97-084, New JTA #: K/A #: 3.11'.034.000.G07 K/A IMPORTANCE: 2.5/3.0 i Rev.1

a ~.,... - -.. I j - Question Number 99 ~ ] 1 l Which of the following are sources of hydrogen that could accumulate in tha l Waste Gas Disposal System? .( - 1. Cover gas on the Volume Control Tank.

2. ; Hydrogen gas in the Main Generator.

\\

3. ' Gaseous Waste Disposal Blower Effluent.

I 4. Degasifier Gaseous Waste Charcoal Bed Effluent. . A. 1 and 3 ONLY. B. 1 and 4 ONLY. C. 1, 3 and 4 ONLY. I D. 2, 3 and 4 ONLY. 1 POINTS: 1.00 TIME ALOTTMENT: minutes - ANSWER: B.

REFERENCES:

10M-19.1.C - Issue 4, Rev. O. 1LP-SQS-19.1 OBJECTIVE: 2 NUMBER: 1-97-086, New JTA 0: 'K/A 6: 3.11.071.000.K5.03 K/A IMPORTANCE: 2.3/2.9 Rev.1

m.___ ._.__._.m.__.. l QuestionN' umber 100 i When combating an electrical fire using foam, which of the following precautions should be exercised by the Fire Brigade members? f

1. Anticipate and avoid the run off from the electrical equipment being sprayed.

2. Maintain a minimum distance of 15 feet from the electrical equipment being sprayed.

3. Always wear rubber boots for electrical insulation. 4. Always use a MSA 401'SCBA when using foam. A. 1 and 2 ONLY. B. 2 and 3 ONLY. C. 1, 3 and 4 ONLY. D. 1, 2, 3, and 4. L F 4 ? POINTS: 1.00 TIME ALOTTMENT3 minutes = ANSWER: A. F i l

REFERENCES:

10M-56B56.B.2 - Is:ue 4, Rev. O. ILP-SQS-33.1 OBJECTIVE: 3 NUMBER:'1-97-007, New .JTA 6: j-K/A 6: 3.11.086.000.K5.04. K/A IMPORTANCE: 2.9/3.5 Rev.I

QuestionNumber 101 'Techniccl Specification 3.9.10, Water Level - Reactor Vessel, requires that at least 23 feet of water be maintained above the reactor pressure vessel flange 'l during refueling operations. What does the bases for this TS requirement ensure? A. In the event of a fuel element rupture, the limits of 10 CFR 100 are maintained. B. In the event one train of RHR is lost, an adequate heat sink is available. C. The refueling operators can perform a full core off-load without exceeding their 10 CFR 20 exposure limits. D. In the event of a fuel element rupture, 99% of the assumed 20% ioditie i gap activity released is removed. 1 l POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: D. .j i i l REFERENCE.ci Unit 1 Technical Specification 3.9.10 Bases, Amendment No. 175. ILP-SQS-6.13 OBJECTIVE: 12 NUMBER: 1-97-088, New JTA #: K/A #: 3.11.000.036.G04. K/A IMPORTANCE: 2.6/3.8 Rev.I

_.._.m ._._.,_._..m... - Question Number 102 4 Refer to NPDAP 5.1, REPORT REQUIREMENTS, to answer the following question. (- Which of the following events would require that a Special Report be submitted to the NRC DCD within 30 days of occurrence? A. A Unit shutdown is commenced to comply with Technical Specification 3.0.3. B. A contract mechanic is found to be intoxicated inside the protected 5 area. C. A radiological liquid release was performed and later found to be-in excess of 10CFR20 limits. D. A fire has occurred in the primary chemistry lab that took 23 minutes to extinguish. POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: C.

REFERENCES:

.NPDAP 5.1 Rev. 5. 1LP-SQS-48.1 . OBJECTIVE: 19 NUMBER: 1-97-089, New JTA 9: -l -K/A 9: 3.11'000.059.G02 K/A IMPORTANCE: 2.6/3.9 Rev.I

.= .~ j . Question Number 103 + Which of the following radiation monitor automatic actions are designed to protect the health and safety of the general public if a highly radioactive Waste Gas Decay Tank were to rupture? - A. RM-IVS-106, Waste Gas Decay Tank Radiation Monitor, will OPEN the Main Filter Bank inlet damper ar.d CLOSE the Main Filter Bank Bypass damper. .B. RM-IVS-106, Waste Gas Decay Tank Radiation Monitor, will TRIP the Leak Collection Area Exhaust Fans. C. RM-IVS-102A, Aux Bldg Vent Gaseous Radiation Monitor, will TRIP the Leak Collection Area AND Aux Bldg Exhaust fans. 1 1 D. RM-IVS-102A, Aux Bldg Vent Gaseous Radiation Monitor, will CLOSE the l Main Filter Bank inlet AND Bypass dampers, f a d I I' i J l i POINTS: 1.00 TIME ALOTTMENT: minutes ANSWER: A. 1 I i i l e

REFERENCES:

10M-43.5.B.2 - Issue 4, Rev. 1. - 1LP-SQS-43.1 OBJECTIVE: 6 NUMBER: 1-97-090, New 1 JTA 8: I K/A is 3.11.000.060.EK2.02 K/A IMPORTANCE: 2.7/3.1 Rn.1

_.~ - _. - _.. ~ ~ _ uestion Number 104 Q 4 - Which of the following'is an advantage of using a straight hose stream over a -fog spray when fighting a large building fire? i i_ } The straight hose strean A. gets the water to the base of the fire before it can vaporize. B. has a better heat absorption / cooling effect. C. _ provides a better. thermal shield to the firefighters. D. uses less water thus minimizes the potential for a re-flash. 4 k 4 J 4 i i 1 l'- i I x POINTS: 1.00 TIME ALOTTMENT: minutes i 1 ANSWER: A. J - i s s 4 4 i i. I-

REFERENCES:

10M-56B56.B.2 - Issue 4, Rev. O and LP 9339. 1LP-SQS-9339 OBJECTIVE: 20 NUMBER: 1-97-091, M-0154 JTA 9: [- K/A #: 3.11.000.067.EK1.02 (2.4.25) K/A IMPORTANCE: 2.9/3.9 Rev.l .- e i r-w y ?-mes

--~-.-.. - - - - -. -. - -- -.~ ~- 1 4 IQuestiam Numtwr 105 i i t' 1. Refer to. Technical Specification 3.4.0, Specific Activity, to answer the following question. Given the following: 1- .e A rapid (5%/ min) power reduction from 1004 to 75% was performed due to e Grid instabilities, o Power has been stable at 75% for seven hours. -The.results from the RCS chemistry samples taken four hours after e l power was stabilized at 75%, reveal the following; t Dose Equivalent I-131 (DEI) is 97 uCi/gm. e-Gross coolant activity is 18 uCi/gm. e e The 100/E-Bar limit is 950 uCi/gm. (E-Bar itself is 0.4 uC1/gm) [ What is the status of the RCS activity levels and what actions are required to comply with Technical Specification 3.4.8? A.. RCS activity levels are within the LCO lLnits and NO actions are necessary to comply with the Technical Specifications, i-B. The DEI limit has been exceeded, power operation may continue provided an isotopic analysis for iodine is performed every four hours. { C. The DEI limit has been exceeded, be in Hot Standby with T..- <500'F within 6 hours. i .D. The 100/E-Bar limit has been exceeded, be in Hot Standby with i T... <500*F within 6 hours. 1 i 6 d POINTS: 1.00-TIME ALOTTMENT: minutes ANSWER: C. 4 e in ? 1 l l

REFERENCES:

Unit 1 Technical Specification 3.4.8, Amendment No.102. ILP-SQS-6.5 OBJECTIVE: 8 NUMBER: 1-97-092, New 'JTA 9: {f K/A 9: 3.11.000.076.G07 K/A IMPORTANCE: 2.9/3.4 Rev.I I i .}}