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).

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 fl 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.

2.

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.

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.

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.

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

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.

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.

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

4.

5.

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9.

NUREG1021 Revision 9 1

BVPS-2 Form ES-401-2 I ES-401 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 E/APE # / Name Safety Function WA Topids1 025 I L o s s of RHR System I 4 029lATWSll 038 I Steam Gen. Tube Rupture I 3

X 058 I L o s s of DC Power I 6 X

062 I L o s s of Nuclear Svc. Water I 4 X

El 1 I Loss of Emergency Coolant Recirc. I 4

007 I Reactor Trlp - Stabilization - Recovery I 1

008 I Pressurizer Vapor Space Accident I 3

01 5 I 17 I RCP Malfunctions I 4 X

022 I Loss of Rx Coolant Makeup I 2 025 I Loss of RHR System I 4 027 I Pressurizer Pressure Control System Malfunction I 3

029 I ATWS I 1 X

X X

X X

X X

X X

AA2.02 j

EA2.01 r 2.2.25 1

2.2.25 2.4.31 EA2.1 EA1.W AA2.20 AKl.05 AA2.04 AA1.08 w.02 EK3.11 Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:

Leakage of reactor coolant from RHR into closed cooling water system or into reactor building atmosphere Ability to determine or interpret the following as they apply to a ATWS: Reactor nuckar instrumentation Equipment Control Knowledge of bases in technical spec~cations for limiting conditions for operations and safety limits.

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~ Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

Emergency Procedures I Plan Knowledge of annunciators alarms and indications, and use of the response instruct ions.

Abil~ty to determine and interpret the foliowing as they apply to the (Loss of Emergency Coolant Recirculation)

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

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Ability to operate and monitor the following as they apply to a reactor triD: CVCS Ability to determine and interpret the following as they apply to the Prmurizer Vapor Space Accident: The effect of an open PORV on code safety, based on observation of plant parameters Knowledge of the operational implications of the following concepts as they apply to the Reactor Coolant Pump Malfunctions ( L o s s of RC Flow): Natural Circulation in a nuclear power plant Ability to determine and interpret the following as they apply to the L o s s of Reactor Coolant Makeup: How long PZR level can be maintained within limits Ability to operate and I or monitor the following as they apply to the Loss of Residual Heat Removal System:

RHR cooler inlet and outlet temperature indlcatm Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions:

Normal values for RCS pressure Knowledge of the reasons for the following responses as the apply to the ATWS: lnitlating emergency boration I

NUREG-1021 Revision 9 2

K2 I L

K3 X

0 2

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

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EAl.16 Ability to operate and monitor the following as they apply to a SGTR: SIG atmospheric relief valve and secondary PORV controllers and indicators Ability to determine and interpret the following as they apply to the Steam Line Rupture: When ESFAS systems

, may be secured 038 I Steam Gen. Tube Rupture I 3 040 I Steam Line Rupture - Excessive Heat Transfer / 4 AA2.05 AAl.04

' Abillty to operate and I or monitor the following as they l apply to the Loss of Main Feedwater (MFW): HPI, under total feedwater loss conditions Knowledge of the operational implications of the following concepts as they apply to the Station Blackout : Natural circulation cooling X -

054 I Loss of Main Feedwater I 4 055 I Station Blackout I 6 4.4 48 4.1 4

4.3 50 3.7 51 3.4 52 4.0 53 3.4 54 3.7 55 3.6 56 EKl.02 I

I Abiltty to operate and I or monitor the following as they apply to the Loss of Offsite Power: Auxiliarylemergency fcedwater pump (motor driven)

Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: ESF system panel alarm annunciators and channel status indicators Ability to operate and I or monitor the following as they apply to the Loss of DC Power: Cross-tie of the affected dc bus wlth the aiternate supply 056 I Loss of Off-site Power I 6 AAl.10 X -

X X

057 I Loss of Vital AC Inst. Bus 16 r i AA2.04 I

I I

I I

I 058 I Loss of DC Power I 6 AA1.01 Emergency ProcedureslPlan: AMIity to recognize abnormal Indications for system operating parameters which are entry level conditions for abnormal and emergency operating procedures.

Ability to determine and interpret the following as they apply to the (LOCA Outside Containment) Facility condltions and selection of appropriate procedures during abnormal and emergency operations.

Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink) Normal, 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 Coolant Recirculation) Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of Emergency Coolant Recirculation).

EA2.1 I 062 I Loss of Nuclear Svc. Water I 4 2.4.4 X '

E04 I LOCA Outside Containment I 3 I

I E05 I Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 EK3.2 El 1 I Lose of Emergency Coolant Recirc. I 4 I

I x EK1.3 WA Category Point Totals:

1 1 / 3 1 3 6 - 6l3 -

Group Point Total: I 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 Number I WA Topic(s)

X X

X E/APE # / Name Safety Function I G I K1 I K2 037 I Steam Generator Tube Leak 13 I

x Emergency Procedures I Plan Knowledge of which events related to system operationdstatus should be reported to outside agencies.

2.4.30 Abilrty to determine and interpret the following as they apply to the (Reactor Trip or Safety Injection Rediagnosk)

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

EA2,,

E01 & E02 I Rediagnosis and SI Termination I 3 E09 I Natural Circulation Operations / 4 E06 I Degraded Core Cooling I 4 X

001 I Continuous Rod Withdrawal I 1

003 I Dropped Control Rod I 1

037 I Steam Generator Tube Leak I 3 EA2.2 Ability to determine and interpret the following as they apply to the Natural Circulation Operations: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Emergency Procedures I Plan Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

Ability to operate and I or monitor the followlng as they apply to the Continuous Rod Withdrawal: Bank select swltch 3.8 83 4.3 85 3.5 57 2.5 58 3.5 59 3.5 60 3.3 61 3.8 62 3.7 63 2.4.4 AA1.01 Knowledge of the interrelations between the Dropped Control Rod and the followlng: Control rod drive power supplies and logic circuits Knowladge of the operational implications of the following concepts as they apply to Steam Generator Tube Leak:

Leak rate VI.

pressure drop AK2.05 AK1.02 Ability to determlne and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: ARM panel displays Emergency procedures I Plan Knowledge of annunciators alarms and indications, and use of the response Instructlons.

X -

AA2.01 081 I ARM System Alarms I 7 EO1 8 E02 I Redlagnosis and SI Termination I 3 2.4.31 Knowledge of the reasons for the following responses as they apply to the (Saturated Core Cooling) Manipulation of controls required to obtain desired operating results during abnormal and emergency situations.

EK3.3 E07 I Inad. Core Cooling I 4 E08 I RCS OvercooHng - PTS I 4 Knowledge of the reasons for the following responses as they apply to the (Pressurized Thermal Shock)

Manipulation of controls required to obtain desired operating results during abnormal and emergency rltuations.

EK3.3 NUREG-1021 Revision 9 4

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

El3 I Steam Generator Over-pressure I 4 X

068 / Control Room Evacuation I 8 X

WA Category Point Total:

1 2 / 2 1 1 I 2 -

2 EK2.1 2.1.30 Group Point 1 Knowledge of the Interrelatlom between the (Steam Generator Overpressure) and the followlng: Components, and functions of control and safety systems, Including Instrumentatlon, slgnak, Interlocks, failure modes, and automatlc and manual features.

Conduct of Operations: Abilky to locate and operate components, including local controls.

fiai:

3.0 3.9 64 65 at4 -

NUREG-1021 Revision 9 5

F O I ~

ES-401-2 BVPS-2 ES-401 A3 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 Group 1 A4 X

X X

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

Number 4.0 2.1.33 Equipment Control Knowledge of limiting conditions for 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 mitigate the consequences of those malfunctions or 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 those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Air dryer and filter malfunctions 2.2.22 A2.05 A2.01 2.1.1 2 4.1 3.0 2.9 K5.02 A4.12 A4.04 K3.05 A2.12 A4.01 K/A ToDics 1 imD.

Conduct of Operations: Ability to apply Technical followlng concepts as they apply to the RCPS: Effects of RCP coastdown on RCS parameters

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Knowledge of the effect that a loss of malfunction of the I 3.5 RCPS will have on the foUowina: SIG Ability to manually operate and/or monitor in the control room: BoratiotVdilution batch control Ability to manually operate and/or monitor in the control room: Controls and Indication for closed cooling water Pumps 3.8 -

3.1 Knowledge of the effect that a loss or malfunction of the 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 on those predictions, use procedure8 to correct, control, or mitlgate the consequences of those malfunctions or operations: Conditions rsquirlng actuation of ECCS AMlky to manually operate and/or monitor in the control room: PRT spray supply valve 3.7 -

4.0 -

2.7 -

90 -

1 2 -

3 -

4 5 -

0 -

7 NUREG-1021 Revision 9 6

Form ES-401-2 BVPS-2 ES-401 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 Group 1 WA ToDics I IrnD. I Q#

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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, 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 tables Knowledge of the dlect of a loss or malfunction of the fdlowlng will have on the RPS: Sensors and detectors 3.2 -

3.5 -

2.7

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

12 -

13 14 Knowledge of the mect that a loss or manunction of the I 3.6 MRSS will have on the followirxr: RCS 115 Knowledge of the physical connections and/or cause systems: MFW effect relationship between the MRSS and the following 1 2.7 1 16

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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 I 3.1 1 18 Pumps Knowledge of the physical connedions andlor cause effect relationships W e e n the AFW and the following 3.6 I 9 systems: Emergency water source Ability to monitor automatic operation of the A N,

including: RCS cooldown during AFW operations Knowledge of bus power supplies to the following: Major system loads 4.0 20 3,3 2,

NUREG-1 021 Revision 9 7

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

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# 1 063 DC Electrical Distributlon 084 Emergency Diesel Generator 064 Emergency Dlesel Generator K3,0, Knowledge of the effect that a loss of malfunction of the dc electrical system will have on the following: EDlG Knowledge of EDlG system deslgn feature(s) and/or EDIG while operating (normal or emergency)

Ability to monitor automatic operation of the EDlG X

A3.05 system, includlng: Operation of the governor control of 2.8 24 3.7 X

X K4.02 inter-lock@) whlch provlde for the following: Trips for 3.9 23 X

073 Process Radiation Monltorlng I 2.7 I 26 I Knowledge of bus power supplies to the following:

I 076 Servlce Water I

I 1 x 1 I I I

I I I I

I I Service water Ability to (a) predict the Impacts of the fdowlng malfunctions or operations on the PRM sy8ttem; and (b) 1 A 2. U 2 F on those predictlom, use procedures to correct, 2.7 1 25 control, or mitigate the consequences of those malfunctlons or operations: Detector failure Knowledge of the effect that a logs or malfunctlon of the IAS will have on the following: Systems having 1 3.4 1 27 I I I I x I I I I I I 1

1 I pneumatic valves and controls 078 Instrument Air 103 Containment WA Category Point Totals:

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28 A3,01 Ability to monitor automatlc operation of the contalnment 3.9 system, including: Containment isolation X

1 1 3 3 2

5 3

2 1

0 3

1 2

4 4

Group Point Total:

2816 NUREG-IO21 Revision 9

BVPS-2 FORTI ES-401-2 1 ES-401 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 G ~ U D 2

I 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#

001 Control Rod Drive F 035 Steam Generator System 056 Condensate 001 Control Rod Drive 002 Reactor Coolant 01 1 Pressurizer Level Control 041 Steam Dump System 033 Spent Fuel Cooiing 034 Fuel Handling Equipment 045 Main Turblne Generator 068 Liquid Radwaste 071 Waste Gas Dispasal X

X X

X X

i 3.8 I 30 2.8 2.5 NUREG-1021 Revision 9 9

BVPS-2 FOim ES-401-2 I ES-401 I

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I 1

I Knowledae of the Dhvsicai connections and/or cause-I I

075 Circulating Water 2.9 38 K1,02 effect relationshi& between the circulating water system and the following systems: Liquid radwaste discharge X

NUREG-1021 Revision 9 1 WA Category Point Totals:

I 112 I 2 I 1 I 1 I 1 I 0 I 1 10 2

0/1 1

0 Grwp PointTotal:

1013

2'1'14 Facility:

BVPS2 Date of Exam: I I

I Knowledge of system status criteria which require the notification of plant personnel.

Knowledge of conduct of operations KIA# I Topic 1 iadiation Control I

I I

2.1.20 Abilrty to execute procedure steps.

2.4'48

1.

Conduct of Operations veril$ the status and operation of system, and understand how operator actions and directives affect plant and system conditions.

I requirements.

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

I

2.

Equipment Control 2.2.25 2.2.30 Subtotal 1 for limiting conditions for operations and safety

~ limits.

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

, handling area, communication with fuel storage

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

2.3.9 2.3.10 2.3.11 Knowledge of the process for performing a containment purge.

Abiltty to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

Abiltty to control radiation releases.

I Subtotal

4.

Emergency Procedures I Plan 2.4.35 2.4.4 2.4.29 2.4.20 I=

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

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

Knowledge of the emergency plan.

Knowledge of operational implications of EOP warninas. cautions. and notes.

Abilrtv to interpret control room indications to Tier 3 Point Total NUREG-1021 Revision 9 11 2/28/2005 RO I

SRO-Only

I EM01 I

Record of Rejected WAS 1 Form ES-401-4 I 111 112 Reason for Rejection Tier I Randomly Group Selected WA 057AA2 09 OOlAA1.04 The subject WA isnt relevant at the subject facility.

The subject WA isnt relevant at the subject facility.

112 21 1 212 21 1 21 1 3

1 1 1 OO3AK2.03 01 2K6.11 0332.4.6 059K4.14 061 K1.10 G2.2.9 0272.4.49 The subject WA isnt relevant at the subjed facility.

The subject WA isnt relevant at the subject facility.

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 examination. and there isnt a site-specific priority that justitks keeping the WA if its importance rating is below 2.5.

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 examination, and there isnt a site-specifc prionty that justifies keeping the WA if its importance rating is below 2.5.

The subject WA isnt relevant at the subject facillty.

1 I 1 112 1 I 1 0622.1.1 4 067AA2.11 0622.1.23 It isnt possible to prepare a psychometrically sound question related to the subject WA.

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

Random selection of replacement KA was a duplicate topic 3

3 G2.4.29 G2.2.17 Duplicate of KA already selected KA deleted because 3 topics selected for Generic Section 2. Replaced with 2.4.4 111 111 R isnt possible to prepare a psychometrically sound question related to the subject WA. Plant effects are minimal, not operationally valid Double Jeopardy with Question 8.

015 AK1.05 062 AA2.06

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1 12 061 AA2.05 112 El 6 G2.4.4

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DwMe Jeopardy with Question 90. Also, significant number of radiation monitoring questions on exam Procedure contains one step, operationally insignifEant NUREG-1021 Revision 9 112 112 12 037 AK1.O1 E16 G2.1.30 Topic not operationally valid. fhis event does not require use of steam tables.

Procedure contains one step unrelated to topic 21 1 21 1

- R isnt p&Me to prepare a psychometrically sound question relevant to this WA. No procedure exists for OO6 062 A2.15 thi event, and the closest possible topic would duplicate question 54 No procedure guidance for KA statement. and question would test same knowledge as Quesbon 24

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212 212

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045 A2.17 028 Al.01 WA identical to event performed in dynamic simulator System removed (retired in place) at facility 1 1 1 21 1

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062 AA2.06 103 A2.04 No relation to facility procedure requirements It isnt possible to wrlte a psychometrically sound SRO question to this KA

Appendix D Scenario Outline Form ES-D-1 Faci I i ty:

BVPS-2 I Scenario NO.:

I 1 I OpTest No.: I NRC Examiners:

Candidates:

CRS RO Turnover :

Critical Tasks:

t--

Event No.

1 2

3 4

5 6

Malf. No.

I I PO EHCO6 CRFOlA RCS031A MSS047A EHC07 xc210790 PPLOl A PPLOl B 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.

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*

(R) RO (N) PO, US (C) ALL (TS) US (I) RO, US (TS) US (I) PO, us (TS) US (M) ALL (C) RO, US Event Description Power Reduction for Waterbox Cleaning Turbine Control Valve failure (Load Rejection)

Control Rods Fail in Auto Pressurizer Pressure Transmitter Fails High SG Pressure Transmitter Fails Low Turbine Trip - Steam Dump Failure. Reactor Trip required.

Auto and Manual Reactor Trip Failure (C) PO, US One SG Atmosphere Dump Valve Fails Partially Open II I MSS02A I (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

- 1 -

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 I

I PO 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.

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

E-0.1, Start Train B HHSl Facility:

BVPS-2 I Scenario NO.:

I 2 1 OpTest No.:

NRC Examiners:

Candidates:

CRS RO Initial Conditions:

Turnover:

Critical Tasks:

Event No.

1 2

3 4

5 6

7 8

9 Malf. No.

Event Type*

Event Description (R) RO Reduce Power (N) PO, us MSS005A (TS) US SG Level Transmitter Fails High DSGOlB 1 (TS) US I Train " B (2-2) EDG Failure LDS007A (C) RO, Letdown Pressure Control Valve Fails Closed In Auto Ius I

CNvlOB (C) PO, SG " A FRV Controller Fails Closed In Auto Ius I

RCS02A (C) RO, RCS Leak us (TS) US RCS02A (M) ALL SBLOCA PPL07A (C) RO Train " A HHSVCharging Pump Auto Start Failure AFW Start Failure (Auto SI Failure Train "6) ppL07B I (c)po I

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

- 1 -

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 (M) ALL (C) PO Facility:

BVPS-2 I Scenario NO.:

I 3 I OpTest No.:

I NRC Examiners:

Candidates:

CRS RO SG B SGTR (when ANV is initiated).

CIA Fails To Automatically Actuate II 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 :

Critical Tasks:

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

E-O.F, Initiate Feedwater Flow with MDAW E-3.A, Isolate Ruptured SG Event Malf. No.

2 SWSOOG I

I BKRHIVol CFWOO4 DSGOlA PPL07B RCS04B L

PPL08A Event Type*

Event Description (R) RO Raise Power us (N) PO, (C) RO, us manually started.)

(TS) US Train A Service Water Pump Trips. (Backup pump must be (C) ALL (TS) US (M) ALL (C) PO Loss of 4KV Bus 2AE. 2-1 EDG Fails to Auto Start.

MFW Pump A Degradationnrip. Reactor Trip.

2-1 EDG Failure MDAFW Train B Pump Auto Start Failure TDAFW Pump Auto Start Failure (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

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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

DE 5

DER 4s A C nictrihlltinn BV-2 Actions to Establish Station Blackout Cross-Tie to Unit 1 DE 6

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 2/28 -313105 Facility:

BVPS-2 Date of Examination:

Exam Level :

SRO(I)

Operating Test No.:

NRC om Systems (8 for RO; 7 for$RO-1)2 or 3 for SRO-U)

System JPM Title -

00 1 Raise Reactor Power to 10.' Amps Rod Control Safety Function Type Code' NSAL 1

E02 NSA 3

Perform SI Termination IAW ES-1.1 SI Termination E03 Post LOCA C/D and Depressurization 041 Isolate SI Accumulators During a LOCA NSA 4P Initiate Natural Circulation Cooldown DASP 4s Steam Dump 103 DSAP 5

Manually Actuate CIB Containment 064 I s6 DS Synchronize and Load EDG 2-1 6

FDG 7

01 5 NIS DSP Remove Power Range Instrument From Service DS 2

I In-Plant Svstems (3 for RO: 3 for SRO-I: 3 or 2 for SRO-U) 028 HRPS P1 Locally Startup a Containment Hydrogen Analyzer Align Service Water Supply to AFW Pumps Suction I p2 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 s3 s4 s5 S6 57 S8 P1 P2 P3 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.

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.

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.

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.

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

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.

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

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

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).

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-30 1 Control Room/ln-Plant Systems Outline Form ES-301-2

• Type Codes (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power (N)ew or (M)odified from bank including 1(A)

( W A (P)revious 2 exams jS)imulator Criteria for RO J SRO-I J SRO-U 4-6 14-6 12-3

< 9 1 \\ a i. r 4 211) 1 / > I 2 1 1 2 1 l ? l

,21221221 2 3 I < 3 I <: 2 (randomly selected)

> I 1>1 I > 1 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:

Authorization 1.3.40 Knowledge of SROs responsibilities in (4.0) emergency plan implementation Review a Gaseous Waste Discharge 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 A1 b A2 A3 A4 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.

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.

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.

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.

The applicant will be given conditions during performance of Emergency Director duties that require classifying an emergency event and determining the recommended protective action recommendations to 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 (I 3 for ROs; I for 4 for SROs & RO retakes)

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

(P)revious 2 exams (I 1 ; randomly selected)

(S)imulator NUREG-1021, Revision 9