ML050900364

Final - Outlines (Folder 3)

ML050900364
Person / Time
Site:	Beaver Valley
Issue date:	01/15/2005
From:	Hynes C FirstEnergy Nuclear Operating Co
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
Download: ML050900364 (24)

Text

ES-401 PWR Examination Outline ES-401-2 Facility: BVPS-2 Date of Exam: 2/28/2005 Note: 1. Ensure that at least two topics from every applicable WA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the 'Tier Totals" in each WA category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by f l from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systemdevolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding elimination of inappropriate WA statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.

7.' The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

8. On the following pages, enter the WA numbers, a brief description of each topic, the topics' importance ratings (IR) forthe applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the WA Catalog, and enter the WA numbers, descriptions, IRs, and point totals (#) on Form ES401-3. Limit SRO selections to WAS that are linked to 10CFR55.43 NUREG1021 Revision 9 1

ES-401 BVPS-2 Form ES-401-2 I Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 1 I

E/APE # / Name Safety Function WA Topids1 Ability to determine and interpret the following as they 025 I L o s s of RHR System I 4 X apply to the Loss of Residual Heat Removal System:

AA2.02 Leakage of reactor coolant from RHR into closed cooling water system or into reactor building atmosphere 029lATWSll X j EA2.01 Ability to determine or interpret the following as they apply r

to a ATWS: Reactor nuckar instrumentation Equipment Control Knowledge of bases in technical 038 ISteam Gen. Tube Rupture I3 X 2.2.25 spec~cationsfor limiting conditions for operations and safety limits.

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~ Equipment Control Knowledge of bases in technical 058 I L o s s of DC Power I 6 X 1 2.2.25 specifications for limiting conditions for operations and safety limits.

Emergency Procedures I Plan Knowledge of annunciators 062 I L o s s of Nuclear Svc. Water I 4 X 2.4.31 alarms and indications, and use of the response instructions.

Abil~tyto determine and interpret the foliowing as they E l 1 I Loss of Emergency Coolant Recirc. I4 apply to the ( L o s s of Emergency Coolant Recirculation)

X EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

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007 I Reactor Trlp Stabilization Recovery I1 X EA1.W Ability to operate and monitor the following as they apply to a reactor triD: CVCS Ability to determine and interpret the following as they 008 IPressurizer Vapor Space Accident I3 apply to the Prmurizer Vapor Space Accident: The X AA2.20 effect of an open PORV on code safety, based on observation of plant parameters Knowledge of the operational implications of the following 015 I 17 IRCP Malfunctions I 4 X concepts as they apply to the Reactor Coolant Pump AKl.05 Malfunctions ( L o s s of RC Flow): Natural Circulation in a nuclear power plant Ability to determine and interpret the following as they 022 ILoss of Rx Coolant Makeup I 2 X AA2.04 apply to the L o s s of Reactor Coolant Makeup: How long PZR level can be maintained within limits Ability to operate and Ior monitor the following as they 025 I Loss of RHR System I 4 X AA1.08 apply to the Loss of Residual Heat Removal System:

RHR cooler inlet and outlet temperature indlcatm 027 IPressurizer Pressure Control System Malfunction I Ability to determine and interpret the following as they 3 X w.02 apply to the Pressurizer Pressure Control Malfunctions:

Normal values for RCS pressure Knowledge of the reasons for the following responses as 029 IATWS I 1 X EK3.11 the apply to the ATWS: lnitlating emergency boration NUREG-1021 Revision 9 2

ES-401 Form ES-401-2 BVPS-2 Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 1 E/APE # / Name Safety Function I G I K1 K2 L

I K3 ~

Number

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I K/A Topic(@

Ability to operate and monitor the following as they apply 038 I Steam Gen. Tube Rupture I 3 EAl.16 to a SGTR: SIG atmospheric relief valve and secondary PORV controllers and indicators Ability to determine and interpret the following as they 040 I Steam Line Rupture Excessive Heat Transfer / 4 AA2.05 apply to the Steam Line Rupture: When ESFAS systems

, may be secured

' Abillty to operate and Ior monitor the following as they 054 ILoss of Main Feedwater I 4 X AAl.04 l apply to the Loss of Main Feedwater (MFW): HPI, under 4.4 48

- total feedwater loss conditions Knowledge of the operational implications of the following 055 I Station Blackout I 6 EKl.02 concepts as they apply to the Station Blackout : Natural 4.1 4 circulation cooling 056 I Loss of Off-site Power I 6 I I X AAl.10 Abiltty to operate and Ior monitor the following as they apply to the Loss of Offsite Power: Auxiliarylemergency 4.3 50 ri fcedwater pump (motor driven)

Ability to determine and interpret the following as they 057 I Loss of Vital AC Inst. Bus 16 X AA2.04 apply to the Loss of Vital AC Instrument Bus: ESF system 3.7 51 I I panel alarm annunciators and channel status indicators I I Ability to operate and Ior monitor the following as they 058 ILoss of DC Power I 6 I I X AA1.01 apply to the Loss of DC Power: Cross-tie of the affected dc bus wlth the aiternate supply 3.4 52 Emergency ProcedureslPlan: AMIity to recognize 062 I Loss of Nuclear Svc. Water I 4 X 2.4.4 abnormal Indicationsfor system operating parameters 4.0 53 which are entry level conditions for abnormal and emergency operating procedures.

Ability to determine and interpret the following as they I apply to the (LOCA Outside Containment) Facility E04 I LOCA Outside Containment I 3 I I EA2.1 condltions and selection of appropriate proceduresduring abnormal and emergency operations.

Knowledge of the reasons for the following responses as 3.4 54 E05 IInadequate Heat Transfer Loss of Secondary X EK3.2 they apply to the (Loss of Secondary Heat Sink) Normal, 3.7 55 Heat Sink / 4 abnormal and emergency operating procedures associated with (Loss of Secondary Heat Sink).

Knowledge of the operational implications of the following concepts as they apply to the (Loss of Emergency E l 1 I Lose of Emergency Coolant Recirc. I 4 WA Category Point Totals:

I Ix 11/31 3 0 2 6 6l3 EK1.3 Coolant Recirculation) Annunciators and conditions indicating signals, and remedial actions associated with the (Lossof Emergency Coolant Recirculation).

Group Point Total: I 3.6 56 NUREG-1021 Revision 9 3

I ES-401 BVPS-2 Form ES-401-2 I Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 2 E/APE # / Name Safety Function I G I K1 I K2 Number I WA Topic(s)

Emergency Procedures IPlan Knowledge of which events 037 ISteam Generator Tube Leak 13 I x 2.4.30 related to system operationdstatus should be reported to outside agencies.

Abilrty to determine and interpret the following as they E01 & E02 IRediagnosis and SI Termination I 3 apply to the (Reactor Trip or Safety Injection Rediagnosk)

EA2,,

Facility conditions and selection of appropriate pracedures during abnormal and emergency operations.

Ability to determine and interpret the following as they E09 INatural Circulation Operations / 4 apply to the Natural Circulation Operations: Adherence to EA2.2 3.8 83 appropriate proceduresand operation within the limitations in the facility's license and amendments.

Emergency Procedures IPlan Ability to recognize abnormal indications for system operating parameters E06 IDegraded Core Cooling I 4 X 2.4.4 4.3 85 which are entry-level conditions for emergency and abnormal operating procedures.

Ability to operate and Ior monitor the followlng as they 001 IContinuous Rod Withdrawal I1 X AA1.01 apply to the Continuous Rod Withdrawal: Bank select 3.5 57 swltch Knowledge of the interrelations between the Dropped 003 IDropped Control Rod I1 AK2.05 Control Rod and the followlng: Control rod drive power 2.5 58 supplies and logic circuits Knowladge of the operational implications of the following 037ISteam Generator Tube Leak I 3 AK1 .02 concepts as they apply to Steam Generator Tube Leak: 3.5 59 Leak rate. IV pressure drop Ability to determlne and interpret the following as they 081 IARM System Alarms I 7 X AA2.01 apply to the Area Radiation Monitoring (ARM) System 3.5 60

- Alarms: ARM panel displays Emergency procedures IPlan Knowledge of annunciators EO1 8 E02 IRedlagnosis and SI Termination I 3 2.4.31 alarms and indications, and use of the response 3.3 61 Instructlons.

Knowledge of the reasons for the following responses as they apply to the (Saturated Core Cooling) Manipulation of E07 IInad. Core Cooling I 4 X EK3.3 3.8 62 controls required to obtain desired operating results during abnormal and emergency situations.

Knowledge of the reasons for the following responses as they apply to the (Pressurized Thermal Shock)

E08 IRCS OvercooHng PTS I 4 X EK3.3 Manipulation of controls required to obtain desired 3.7 63 operating results during abnormal and emergency rltuations.

NUREG-1021 Revision 9 4

ES-401 BVPS-2 Form ES-401-2 I Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 2 E/APE # / Name Safety Function I ImD. I Q#

Knowledge of the Interrelatlombetween the (Steam Generator Overpressure) and the followlng: Components, E l 3 I Steam Generator Over-pressure I 4 X EK2.1 and functions of control and safety systems, Including 3.0 64 Instrumentatlon, slgnak, Interlocks, failure modes, and automatlc and manual features.

068 / Control Room Evacuation I 8 X Conduct of Operations: Abilky to locate and operate 2.1.30 3.9 65 WA Category Point Total:

- components, including local controls.

12/21 1 I 2 2 Group Point 1 fiai:

- at4 NUREG-1021 Revision 9 5

ES-401 BVPS-2 F O IES-401-2

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Written Examination Outline A3 A4 Emergency and Abnormal Plant Evolutions Tier 2 Group 1 Number K/A ToDics 1 imD.

Conduct of Operations: Ability to recognize indications 2.1.33 for system operating parameten which are entry-level 4.0 conditions for technical soecifications.

Equipment Control Knowledge of limiting conditions for 2.2.22 4.1 operations and safety limits.

AbilQ to (a) predict the impacts of the following mal-functions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or A2.05 mitigate the consequences of those malfunctions or

3.0 operations

Increasing steam demand, its relationship to increases in reactor power AbilQ to (a) predict the impacts of the following malfunctions or operations on the IAS; and (b) based on A2.01 those predictions, use procedures to correct, control, or 2.9 mitigate the consequences of those malfunctions or operations: Air dryer and filter malfunctions 2.1.12 Conduct of Operations: Ability to apply Technical 90 K5.02 followlng concepts as they apply to the RCPS: Effects 1 of RCP coastdown on RCS parameters

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Knowledge of the effect that a loss of malfunction of the I 3.5 2 RCPS will have on the foUowina: SIG Ability to manually operate and/or monitor in the control X A4.12 room: BoratiotVdilution batch control 3.8 3 Ability to manually operate and/or monitor in the control X A4.04 room: Controls and Indication for closed cooling water 3.1 4 Pumps Knowledge of the effect that a loss or malfunction of the K3.05 3.7 5 RHRS will have on the following: ECCS Ability to (a) predict the Impacts of the following malfunctions or operations on the ECCS; and (b) based A2.12 on those predictions, use procedure8 to correct, control, 4.0 0 or mitlgate the consequences of those malfunctions or operations: Conditions rsquirlng actuation of ECCS AMlky to manually operate and/or monitor in the control X A4.01 room: PRT spray supply valve 2.7 7 NUREG-1021 Revision 9 6

ES-401 Form ES-401-2 BVPS-2 Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 2 Group 1 WA ToDics

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I IrnD. I Q#

Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, 3.2 or mitigate the consequences of those malfunctions or operations: HigMow surge tank level Knowledge of the operational implications of the following concepts as the apply to the PZR PCS:

Determination of condition of fluid in PZR, using steam 3.5 tables Knowledge of the dlect of a loss or malfunction of the 2.7 fdlowlng will have on the RPS: Sensors and detectors

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Ability to manually operate and/or monitor in the control room: ESFAS-initiatedequipment which fails to actuate Knowledge of CCS design feature+) and/or interlcck(8) which provide for the following: Cooling of containment 2.5 12 penetrations effect reletionship between the CCS and the following 13 Conduct of Operations: Ability to perform specific system and integrated plant procedum during all 14 modes of plant operation.

Knowledge of the mect that a loss or manunction of the MRSS will have on the followirxr: RCS I 115 3.6 Knowledge of the physical connections and/or cause effect relationship between the MRSS and the following systems: MFW

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1 12.7 16 Ability to monitor automatic operation of the M F W , p l T including: Proarammed levels of the S/G Knowledge of MFW design feature(s) andlor interlock(s) which provide for the following: Automatic trips for MFW Pumps I 13.1 18 Knowledge of the physical connedions andlor cause effect relationships W e e n the AFW and the following 3.6 I9 systems: Emergency water source Ability to monitor automatic operation of the A N ,

4.0 20 including: RCS cooldown during AFW operations Knowledge of bus power supplies to the following: Major system loads 3,3 2, NUREG-1021 Revision 9 7

ES-401 FOITTIES-401-2 BVPS-2 Written Examination Outline I

Emergency and Abnormal Plant Evolutions Tier 2 Group 1 I

I System #/Name I G 1 K1 I K2 I K3 I K4 I K5 I K6 I A1 I A2 I A3 I A4 I Number I WA Topics I Imp. I Q# 1 Knowledge of the effect that a loss of malfunction of the 063 DC Electrical Distributlon X K3,0, 3.7 dc electrical system will have on the following: EDlG Knowledge of EDlG system deslgn feature(s) and/or 084 Emergency Diesel Generator X K4.02 inter-lock@)whlch provlde for the following: Trips for 3.9 23 EDIG while operating (normal or emergency)

Ability to monitor automatic operation of the EDlG 064 Emergency Dlesel Generator X A3.05 system, includlng: Operation of the governor control of 2.8 24 073 Process Radiation Monltorlng X 1 Ability to (a) predict the Impacts of the fdowlng malfunctions or operations on the PRM sy8ttem; and (b)

A 2 . U 2 F on those predictlom, use proceduresto correct, control, or mitigate the consequences of those malfunctlons or operations: Detector failure 2.7 1 25 I 076 Servlce Water I I 1 x 1 I I I I I I I I I Knowledge of bus power supplies to the following:

Service water I 2.7 I I 26 078 Instrument Air I I I IxI I I I I I 1 1 ~ ~ ~

I Knowledge of the effect that a logs or malfunctlon of the IAS will have on the following: Systems having pneumatic valves and controls 1 13.4

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27 103 Containment A3,01 Ability to monitor automatlc operation of the contalnment 3.9 X system, including: Containment isolation 28 WA Category Point Totals: 1 1 3 3 2 5 3 2 1 0 3 1 2 4 4 Group Point Total: 2816 NUREG-IO21 Revision 9

ES-401 BVPS-2 FORTIES-401-2 1 Written Examination Outline I

Emergency and Abnormal Plant Evolutions Tier 2 G ~ U2D System #/Name 1 G I K1 I K2 1 K3 I K4 I K5 1 K6 I A I I A2 I A3 I A4 I Number I WA Topics I Imp. I Q#

F 001 Control Rod Drive 035 Steam Generator System X

056 Condensate X 001 Control Rod Drive X 002 Reactor Coolant 3.8 I 30 01 1 Pressurizer Level Control 041 Steam Dump System 033 Spent Fuel Cooiing X 2.8 034 Fuel Handling Equipment 045 Main Turblne Generator X 068 Liquid Radwaste 071 Waste Gas Dispasal 2.5 i

NUREG-1021 Revision 9 9

ES-401 BVPS-2 FOim ES-401-2 I I I I I I I I I I I 1 I Knowledae of the Dhvsicai connections and/or cause- I I X effect relationshi& between the circulating water 2.9 38 075 Circulating Water K1,02 system and the following systems: Liquid radwaste discharge 1 WA Category Point Totals: I 112 I 2 I 1 I 1 I 1 I 0 I 1 2 0/1 1 0 Grwp PointTotal: 1013 NUREG-1021 Revision 9 10

Facility: BVPS2 Date of Exam:

I 2/28/2005 I

KIA# II Topic RO I SRO-Only I

I I 2.1.20 Abilrty to execute procedure steps.

Knowledge of system status criteria which 2'1'14 require the notification of plant personnel.

1. ,L. I . I Knowledge of conduct of operations Conduct of I requirements.

Operations I Ability to obtain and interpret station reference 2.1.25 materials such as graphs, monographs, and tables which contain performance data.

Subtotal I

2. 2.2.25 1 for limiting conditions for operations and safety limits.

Equipment Control

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Knowledge of RO duties in the control room during fuel handling such as alarms from fuel 2.2.30

, handling area, communicationwith fuel storage

' faciltty, systems operated from the control room in support of fueling operations, and supporting instrumentation.

Subtotal Knowledge of the process for performing a 2.3.9 containment purge.

Abiltty to perform procedures to reduce 2.3.10 excessive levels of radiation and guard against 1 iadiation Control personnel exposure.

2.3.11 Abiltty to control radiation releases.

I Subtotal Knowledge of local auxiliary operator tasks 2.4.35 during emergency operations including system geography and system implications.

Ability to recognize abnormal indications for system operating parameters which are entry-2.4.4 level conditions for emergency and abnormal operating procedures.

4.

Emergency 2.4.29 Knowledge of the emergency plan.

Procedures I Plan I=2.4.20 2.4'48 Knowledge of operational implications of EOP warninas. cautions. and notes.

Abilrtv to interpret control room indications to veril$ the status and operation of system, and understand how operator actions and directives affect plant and system conditions.

Tier 3 Point Total NUREG-1021 Revision 9 11

I EM01 I Record of Rejected WAS 1 Form ES-401-4 I Tier I Randomly Reason for Rejection Group Selected WA 111 057AA2 09 The subject WA isnt relevant at the subject facility.

112 OOlAA1.04 The subject WA isnt relevant at the subject facility.

112 OO3AK2.03 The subject WA isnt relevant at the subjed facility.

21 1 012K6.11 The subject WA isnt relevant at the subject facility.

212 0332.4.6 The subject WA isnt relevant at the subject facilty.

The subject WAs importance rating isnt equal to or greater than 2.5 for the license level of the proposed 21 1 059K4.14 examination. and there isnt a site-specific priority that justitks keepingthe WA if its importance rating is below 2.5.

211 061K1.10 The subject WA isnt relevant at the subject facility.

The subject WAS importance rating isnt equal to or greater than 2.5 for the license level of the proposed 3 G2.2.9 examination, and there isnt a site-specifc prionty that justifies keeping the WA if its importance rating is below 2.5.

1 11 0272.4.49 The subject WA isnt relevant at the subject facillty.

1I1 0622.1.14 It isnt possible to prepare a psychometrically sound question related to the subject WA.

112 067AA2.11 It isnt possible to prepare a psychometricallysound question related to the subject WA.

1I 1 0622.1.23 Random selection of replacement KA was a duplicate topic 3 G2.4.29 Duplicate of KA already selected 3 G2.2.17 KA deleted because 3 topics selected for Generic Section 2. Replaced with 2.4.4 R isnt possible to prepare a psychometricallysound question related to the subject WA. Plant effects are 111 015 AK1.05 minimal, not operationally valid 111 062 AA2.06 Double Jeopardy with Question 8.

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1 12 061 AA2.05 DwMe Jeopardy with Question 90. Also, significant number of radiation monitoring questions on exam 112 El 6 G2.4.4 Procedure contains one step, operationally insignifEant 112 037 AK1.O1 Topic not operationally valid. fhis event does not require use of steam tables.

112 E16 G2.1.30 Procedure contains one step unrelated to topic 21 1 OO6 - thi R isnt p&Me to prepare a psychometricallysound question relevant to event, and the closest possible topic would duplicate question 54 this WA. No procedure exists for 21 1 062 A2.15 No procedure guidance for KA statement. and question would test same knowledge as Quesbon 24

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212 045 A2.17 WA identical to event performed in dynamic simulator 212 028 Al.01 System removed (retired in place) at facility

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111 062 AA2.06 No relation to facility procedure requirements 21 1 103 A2.04 It isnt possible to wrlte a psychometrically sound SRO question to this KA NUREG-1021 Revision 9 12

Appendix D Scenario Outline Form ES-D-1 FaciIity: BVPS-2 I Scenario NO.: I 1 I OpTest No.: I NRC Examiners: Candidates: CRS RO I I PO BOL, 100% Power 2CHS'P21C, HHSl Pump 00s.

2RCSPCV455D leakage. 2RCS*MOV-537, Block Valve closed with power maintained.

Flood warnings from heavy rains.

Maintenance investigating 2SWS'P21 A, Service Water Pump abnormal vibrationhoise.

-t Turnover:

Critical Tasks:

Event No.

Malf. No.

Initiate power reduction to 75% for waterbox cleaning.

FR-S.l .C, Initiate RCS Boration and/or insert RCCAs E-2.A, Isolate Faulted SG I Event Type* Event Description 1 (R) RO Power Reduction for Waterbox Cleaning (N) PO, US 2 EHCO6 (C) ALL Turbine Control Valve failure (Load Rejection)

CRFOlA (TS) US Control Rods Fail in Auto 3 RCS031A (I) RO, US Pressurizer Pressure Transmitter Fails High (TS) US 4 MSS047A (I) PO, us SG Pressure Transmitter Fails Low (TS) US 5 EHC07 (M) ALL Turbine Trip - Steam Dump Failure. Reactor Trip required.

xc210790 6 PPLOlA (C) RO, US Auto and Manual Reactor Trip Failure PPLOl B I I MSS02A I (C) PO, US One SG Atmosphere Dump Valve Fails Partially Open (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Event Description NRC Scenario 1 The crew will assume the shift at 100% power with instructions to reduce load to 75% for waterbox cleaning.

A turbine load rejection will occur due to a turbine valve position limiter failure requiring the crew to stabilize the plant by matching Tave and Tref, and resetting condenser steam dump valves.

After Technical Specifications have been addressed and the plant is stable, Pressurizer Pressure Channel PT-445 will fail high slowly requiring the RO to take manual control of Pressurizer heaters, spray valves, and PORVs. The Unit Supervisor will then address Technical Specifications.

When RCS pressure is stable, SG Pressure Transmitter PT-476 will fail low causing the steam flow signal to its associated SG main feedwater control valve to fail low. The PO will take manual control of the affected valve to prevent RPS actuation on SG low-low level.

When SG level is under control and Technical Specifications have been addressed, a turbine trip will occur with a steam dump failure requiring a reactor trip.

Upon reactor trip, the reactor trip breakers will not open automatically or manually. The RO must insert control rods and initiate emergency boration. The Unit Supervisor will direct crew response in accordance with the ATWS Functional Recovery procedure.

A faulted SG develops due to a stuck open SG atmospheric dump valve requiring transition to E-2 to isolate the faulted SG. The scenario is terminated upon completion of E-2, or upon transition to ES-1.I.

EOP Flow Path: E-0, FR-S.l, E-0, E-2

Appendix D Scenario Outline Form ES-D-1 Facility: BVPS-2 I Scenario NO.: I 2 1 OpTest No.: NRC Examiners: Candidates: CRS RO I I PO Initial Conditions: MOL, 48% power.

2CHS'P21 C, HHSl Pump 00s.

2RCS*PCV455D leakage. 2RCS*MOV-537, Block Valve closed with power maintained.

Flood warnings due to heavy rains.

Maintenance investigating 2SWS'P21 A, Service Water Pump abnormal vibration/ noise.

Turnover: Reduce power to take the unit off-line due to circulating water intake clogging.

Critical Tasks: E-0.1, Start Train B HHSl Event Malf. No. Event No. Type* Event Description 1 (R) RO Reduce Power (N) PO, us 2 MSS005A (TS) US SG Level Transmitter Fails High 3 DSGOlB 1 (TS) US I Train " B (2-2) EDG Failure 4 LDS007A (C) RO, Ius I Letdown Pressure Control Valve Fails Closed In Auto 5

6 CNvlOB RCS02A (C) PO, Ius (C) RO, I SG " A FRV Controller Fails Closed In Auto RCS Leak us (TS) US 7 RCS02A (M) ALL SBLOCA 8 PPL07A (C) RO Train " A HHSVCharging Pump Auto Start Failure 9

ppL07B I (c)po I AFW Start Failure (Auto SI Failure Train " 6 )

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Event Description NRC Scenario 2 The crew will assume the shift at 48% power with directions to reduce power to take the unit off-line due to circulating water intake clogging.

As power is being reduced, a SG B level transmitter will fail high requiring the Unit Supervisor to refer to Technical Specifications.

When the Unit Supervisor has reviewed Technical Specifications, a fuel oil leak on the 2-2 Emergency Diesel Generator will occur making it inoperable. This failure provides the Unit Supervisor with an additional Technical Specification referral and sets up required actions post-trip.

When Technical Specifications have been addressed, the letdown pressure control valve will fail closed requiring the RO to take manual control to restore letdown flow.

When letdown is restored, SG A main feedwater control valve will fail closed in automatic requiring the PO to take manual control to stabilize SG level.

When SG level is stabilized, an RCS leak will develop. When the Unit Supervisor refers to Technical Specifications, the leak will degrade into a SBLOCA requiring a reactor trip and safety injection actuation by the crew.

The Train A HHSVCharging Pump will fail to automatically start and must be started manually. RCPs must be tripped when criteria is met due to the LOCA. Train B AFW Pump must be started manually by the operator.

The scenario may be terminated upon entry to ES-1.2, Post LOCA Cooldown And Depressurization, or when RCS cooldown is initiated.

EOP Flow path: E-0, E-1, ES-1.2

Appendix D Scenario Outline Form ES-D-1 Facility: BVPS-2 I Scenario NO.: I 3 I OpTest No.: I NRC Examiners: Candidates: CRS RO I I t I PO Initial Conditions: MOL, 25% Power.

2RCSPCV455D leakage. 2RCS*MOV537, Block Valve closed with power maintained.

Flood watch remains in effect.

Turnover: Raise power to 100% after a trip due to loss of all circulating water.

Critical Tasks: E-O.F, Initiate Feedwater Flow with M D A W E-3.A, Isolate Ruptured SG Event Malf. No. Event Type* Event Description (R) RO Raise Power (N) PO, us 2 SWSOOG (C) RO, Train A Service Water Pump Trips. (Backup pump must be us manually started.)

(TS) US I I BKRHIVol CFWOO4 (C) ALL (TS) US (M) ALL Loss of 4KV Bus 2AE. 2-1 EDG Fails to Auto Start.

MFW Pump A Degradationnrip. Reactor Trip.

DSGOlA 2-1 EDG Failure (C) PO MDAFW Train B Pump Auto Start Failure PPL07B TDAFW Pump Auto Start Failure L RCS04B PPL08A (M) ALL (C) PO SG B SGTR (when A N V is initiated).

CIA Fails To Automatically Actuate (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Event Description NRC Scenario 3 The crew will assume the shift at approximately 25% power with instructions to raise power to 100%.

After initiation of the power increase, the running service water pump will trip. The backup pump will not start automatically and must be started manually by the RO.

When Technical Specifications have been addressed, 4KV Emergency Bus 2AE will be de-energized and the crew must manually start EDG 2-1 and reinitiate charging flow.

The Unit Supervisor will refer to Technical Specifications.

When the plant is stable, the running feedwater pump will trip requiring a reactor trip.

The 2-1 EDG will fail de-energizing 4KV Bus 2AE. The Train B MDAFW pump and the TDAFW pump will fail to automatically start requiring manual start by the operator.

When transition is made to ES-0.1 and AFW pumps have been started, a SGTR will develop requiring SI initiation. CIA valves will not automatically close requiring manual closure by the PO while performing Attachment A-0.1 1, Verification of Automatic Actions.

The scenario is terminated when the ruptured SG is isolated in E-3 and the crew has commenced an RCS cooldown.

EOP Flow Path: E-0, ES-0.1, E-0, E-3

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Facility: BVPS-2 Date of Examination: 2/28 -313105 Exam Level : SRO(I) Operating Test No.: NRC om Systems (8 for RO; 7 for$RO-1)2 or 3 for SRO-U)

Safety System JPM Title Type Code' Function 001 Raise Reactor Power to 10.' Amps NSAL 1 Rod Control E02 Perform SI Termination IAW ES-1.1 NSA 3 SI Termination E03 Post LOCA C/D and Isolate SI Accumulators During a LOCA NSA 4P Depressurization 041 Initiate Natural Circulation Cooldown DASP 4s Steam Dump 103 Manually Actuate CIB DSAP 5 Containment 064 I s6 FDG Synchronize and Load EDG 2-1 DS 6 015 Remove Power Range Instrument From Service DSP 7 NIS DS 2 I In-Plant Svstems (3for RO: 3 for SRO-I: 3 or 2 for SRO-U) 028 P1 Locally Startup a Containment Hydrogen Analyzer DE 5 HRPS I ::1 p2 Align Service Water Supply to AFW Pumps Suction DER 4s BV-2 Actions to Establish Station Blackout Cross-Tie DE 6 AC nictrihlltinn to Unit 1

@ All control room (and in-plant) systems must be different and serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

NUREG-1021, Revision 9

Control Room/ln-Plant Systems Outline Task Summary s1 The applicant will raise reactor power using control rods lo approach criticality. Source Range High Flux Trips must be blocked and power indication switched to Intermediate Range channels. The alternate path of this task will be based on continuous rod motion in the OUT direction. The applicant will be required to trip the reactor based on AOP guidance.

This is a new JPM.

s2 SI Termination will be performed requiring the applicant to align normal RCS makeup flowpaths and secure ECCS equipment. The alternate path of this task will require the applicant to diagnose the inability to maintain RCS inventory and based on either EOP or Foldout page guidance, realign HHSl equipment. This is a new JPM.

s3 The applicant will be placed in the EOP network during a Post-LOCA Cooldown and Depressurization. The task is to isolate SI accumulators so that RCS depressurization may continue. The alternate path of this task is to vent one SI accumulator to containment once it is determined that it cannot be isolated. This is a new JPM.

s4 The applicant will initiate an RCS cooldown IAW ES-0.2 during natural circulation conditions. The alternate path of this task is to initiate cooldown using the Residual Heat Release Valve after diagnosing the failure of the condenser steam dump valves. This is a bank JPM. This JPM was performed on the 2001 and 2002 NRC examinations.

s5 The applicant will be required to verify Containment Isolation Phase B (CIB) actuation. The alternate path of this task is lo manually realign equipment required by CIB after determining that it did not actuate either automatically, or manually.

This is a bank JPM. This JPM was performed on the 2002 NRC examination.

S6 The applicant will synchronize EDG 2-1 to its emergency bus and raise load on the EDG.

57 The applicant will perform actions to remove a power range NI channel from service. This JPM was performed on the 2001 and 2002 NRC examinations.

S8 The applicant will manually establish blended makeup flow to the VCT. This is a bank JPM.

P1 The applicant will locally start a containment hydrogen analyzer. This is a bank JPM.

P2 The applicant will be required to align plant service water supply to the auxiliary feedwater pumps. This is a bank JPM that will require entry into the Radiation Control Area (RCA).

P3 The applicant will perform actions to restore emergency AC power using the station blackout cross-tie to Unit 1. This is a bank JPM.

NUREG-1021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2

• Type Codes Criteria for RO J SRO-I J SRO-U (A)lternate path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank <91\ai.r4 (E)mergency or abnormal in-plant 211) 1 / > I (L)ow-Power 21121l?l (N)ew or (M)odified from bank including 1(A) ,21221221 (P)revious2 exams 2 3 I < 3 I <: 2 (randomly selected)

(WA > I 1>1 I >1 jS)imulator NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 Facility: BVPSP Date of Examination: 2/28 -313105 Examination Level SRO Operating Test Number: NRC Describe activity to be performed 1.1.12 Ability to apply Technical Specifications for a (4.0) system.

JPM: Determine Action Required For Failed AC Sources Surveillance 1.1.23 Ability to perform specific system and (4.0) integrated plant procedures during all modes of plant operation.

JPM: Review an ECP Calculation 2.2.1 3 Knowledge of Tagging and Clearance (3.8) Procedures.

JPM: Approve a Tagging Request 1.2.8 Knowledge of the process for performing a (3.2) planned Gaseous Radioactive release.

JPM: Review a Gaseous Waste Discharge Authorization 1.3.40 Knowledge of SROs responsibilities in (4.0) emergency plan implementation JPM: Classify an Event and Determine Protective Action Recommendations.

NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

NUREG-1021, Revision 9

Administrative Topics Outline Task Summary A1 a The applicant will be required to identify procedural errors and determine the required Technical Specification actions for a failed surveillance test. This is a bank JPM. This JPM was performed on the 2002 NRC examination.

A1 b Given plant conditions prior to a reactor startup, the applicant will be required to calculate the boron concentration required for reactor startup. This is a modified bank JPM. A variation of this JPM was performed on the 2001 NRC examination.

A2 Given a tagging request, the applicant will be required to perform a review and identify errors contained within the tagging order. This is a modified bank JPM. A variation of this JPM was performed on the 2001 NRC examination.

A3 The applicant will be required to review a gaseous waste discharge release permit containing errors that must be identified and corrected prior to approval. This is a new JPM.

A4 The applicant will be given conditions during performance of Emergency Director duties that require classifying an emergency event and determining the recommended protective action recommendationsto offsite agencies. This is a modified JPM.

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1

• Type Codes & Criteria: (C)ontrol room (D)irect from bank (I3 for ROs; I for 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (> 1)

(P)revious 2 exams (I1 ; randomly selected)

(S)imulator NUREG-1021, Revision 9