ML051030201

Draft - Outlines (Folder 2)

ML051030201
Person / Time
Site:	Beaver Valley
Issue date:	12/27/2004
From:	Hynes C FirstEnergy Nuclear Operating Co
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
Download: ML051030201 (28)

Text

ES-401 PWF3 Examination Outline F

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ES-401-2 Facility:

BVPS-2 Date of Exam:

2/28/2005 Vote:

1.

Ensure that at least two topics from evw applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the 'Tier Totals' in each KIA 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.

Systemdevolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the faalty should be deleted and justified; operationally important, sitespecific systems that are not included on the outline should be added. Refer to ES401, Attachment 2, for guidance regarding elimination of inappropriate KIA statements.

2.

3.

4.

5.

6.

7..

a.

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, me topics' importance ratings (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totak for each category in the table above. Use duplicate pages for RO and SRmnly exams.

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

9.

NUREG-1021 Revision 9

BVPS-2

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E S 4 Wrttten Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 UAPE # / Name Safety Function I Number I WA Topic(s)

A l A2 L

X X

X X

X X

X X

Q#

76 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 bullding atmosphere 3.8 025 I Loss of RHR System I 4 AA2.02 029 I ATWS I 1 EA2.01 77 7

78 Ability to determine or interpret the following as they apply to a A W S : Reactor nuclear instrumentation Equipment Control Knowledge of bases in technical specifcations for limiting condtions for operations and safety limits.

038 I Steam Gen. Tube Rupture I 3 2.2.25 Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety Ilmits.

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

Abillty to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation)

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

Ablllty to operate and monitor the followlng as they apply to a reactor trlp: CVCS 058 I Loss of DC Power 16 2.2.25 2.4.31 062 / Loss of Nuclear Svc. Water I 4 E l 1 I Loss of Emergency Coolant Recirc. I 4 EA2 1 EAI.09 007 I Reactor Trlp - StaMllzatlon - Recovery I 1

Ablllty to determine and Interpret the following ae they apply to the Pressurizer Vapor Space Accldent: The effect of an open PORV on code safety, bared on observation of plant parameten Knowledge of the operational implkatlm of the followlng conceptm am they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Natural Circulation In a nuclear power plant Ablllty to determine and Interpret the followlng as they apply to the Lobs of Reactor Coolant Makeup: How long PZR level can be malntalned wtthln llmlts 008 I Prerwrlzer Vapor Space Accldent I 3

01 5 I 17 I RCP Malfunctlonm I 4 022 I Loss of Rx Coolant Makeup I 2 AA2.20 AKI.E AA2.04 Ablllty to operate and I or monitor the followlng ae they apply to the Lo-of Resldual Heat Removal Symtem:

RHR cooler Inlet and outlet temperature lndlcaton AMllty to detwmlne and Interpret the followlng am they apply to the Pro8surlzw Preaaure Control Malfunctions:

Normal MIUW for RCS warwre AA1.08 AA2.02 025 I L m of RHR System / 4 027 I Prereurlzer Pressure Control System Malfunctlon I 3

0 2 Q I A l W S I l Knowledge of the reasons for the followlng responses as the apply to the A W S : lnitlating emergency boratlon 4.2 EK3.11 NUREG-1021 Revision 9 2

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BVPS-2 Fo, LS-401-2 Written Examination Outllne Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 1

WAPE # / Name Safety Function I G Ablllty to operate and monitor the following as they apply to a SGTR: SI0 atmospherlc relief valve and secondary PORV controllers and Indicators Ablllty to determine and Interpret the following as they apply to the Steam Line Rupture: When ESFAS rystems mav be Mcured EA1.16 038 I Steam Gem Tube Rupture I 3 040 I Steam Llne Rupture - Excessive Heat Transfer I 4 054 I Loaa of Maln Feedwater I4 X

AA2.05 Ablllty to operate and I or monitor the fdlowlng as they apply to the Loss of Maln Feedwater (MFW): HPI, under total feedwater loas conditions Knowledge of the operational Implicationsofthe followlng

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concepts as they apply to the Station Blackout : Natural circulation cooling Ablllty to operate and I or monitor the followlng as they feedwater pump (motor drlven)

EKI.02 4.1 AAI.10 apply to the Lose of Offslte Power: Auxillarylemergency 4 3 055 I Station Blackout I 6 056 I Loss of Off-site Power I 6 057 I Loss of Vital AC Inst. Bus I6 3.7 1 51 Ablllty to determlne and Interpret the followlng as they apply to the Loss of Vltal AC Instrument Bus: ESF system panel alarm annunciators and channel status Indicators Ablllty to operate and I or monltor the followlng a8 they apply to the Lose of DC Power: Crosbtle of the affected dc bur with the alternate rupply Ablllty to determlne and Interpret the following as they apply to the Lose of Nuclear Service Water: Length of t h e before equipment damage Ablllty to determlne and Interpret the followlng as they apply to the (LOCA Outdde Contalnrnent) Faclltty condltlorm and selectlon of appropriate procedures during abnormal and emergency operatlons.

058 I Loss of DC Power I 6

062 I Loss of Nuclear Svc. Water I 4 E04 I LOCA Outslde Containment I 3 AAI.Ol AA2.08 X

EA2. I Knowledge of the reasons for the followlng responses as they apply to the (Loss of Secondary Heat Slnk) Normal, abnormal and emergency operatlng procedures associated with (Loss of Secondary Heat Slnk).

E05 I Inadequate Heat Transfer - Loss of Secondary Heat Slnk I 4 EK3.2 3.7 55 Knowledge of the operatlonal Implications of the foliowlng concepts as they apply to the (Loss of Emergency Coolant Reclrculatlon) Annunciators and condfflons lndtcatlng dgnab, and remedlal actlons a88oclated with the (Loaa of Emergency Coolant Reclrculatlon).

EK1.3 3.6 56 El 1 I L m of Emergency Coolant Reclrc. I 4 I WA Category Polnt Totalr:

I o n 7t3 I Grou~PolntTotal: I I lW6 NUREG-1021 Revision 9 3

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is-401-2

[ESJG BVPS-2 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 b

X X

X I

VAPE#/NameSafety Function 1 G I K1 Number WA Topic(s)

I Imp.

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

Ability to determine and interpret the following as they apply to the (Reactor Trip or Safety Injection Rediagnosis)

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

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 facilitv's license and amendments.

3.6 4,0 3.8 037 I Steam Generator Tube Leak 1 3 X

E01 8 E02 / Rediagnosis and SI Termination I 3 E09 / Natural Circulation Operations I 4 2.4 30 EA2.1 EA2.2 Emergency Procedures / Plan Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

4 3 2.4.4 E06 I Degraded Core Cooling 14 X

l l

Ability to operate and I or monitor the followlng a8 they swltch apply to the Contlnuous Rod WRhdrawal: Bank select 1

3.5 AA1.Ol 001 I Contlnuous Rod Withdrawal I 1 003 I Dropped Control Rod I 1 I

I Knowledge of the lntenelatlons between the Dropped Control Rod and the following: Control rod drive power suppitem and loglc clrcub Knowledge of the operational Impllcations of the followlng concept8 w they apply to Steam Generator Tube Leak:

Leak rate VI.

pressure drop X -

AK2.05 AK1.02 1 037 I Steam Generator Tube Leak 13 I

I x Abillty to deterrnlne and Interpret the following w they apply to the Area Radlatlon Monitorlng (ARM) System Alarmr: ARM panel dlrplaya Emergency Procedures I Plan Knowledge of annunciatore alarms and lndlcatlone, and use of the response Instructions.

X -

081 I ARM System Alarm I 7 EO1 8 EO2 I Redlagnods and SI Termlnatlon I3 X

E07 I Inad. Core Cooling I 4 AA2.01 2.4.31 Knowledge of the reasons for the following responses as they apply to the (Saturated Core Cooling) Manipulation of controls required to obtaln deslred operatlng reaub during abnormal and emergency eltuatlone.

3.8 EK3.3 Knowledge of the reasons for the followlng rerponses as they apply to the (Pressurized Thermal Shock) operatlng results durlng abnormal and emergency situatlons.

Manlpulatlon of controls required to obtaln dealred 3.7 l

l EK3.3 E08 I RCS Overcoollng - PTS I 4 L

1 I

NUREG-1021 Revision 9 4

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Fc 3-401-2 ESdC BvPS-2 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 1

UAPEWNameSafety Function 1 G I K1 I K2 El3 I Steam Generator Over-pressure I 4 X

068 I Control Room EvecuaUon 18 WA Category Point Total:

2 / 2 1 2

X K3 I A I / A 2 1 Number I WA Tooic(s)

Knowledge of the lnterrelatlonr between the (Steam Generator Overpressure) and the followlng: Components, Inatrumentatlon, dgnals, Interlocks, hllure mod-,

and automatlc and manual features.

EK2.1 and functions of control and safety rystema, lncludlng 3.0 64 I 3.9 I I

Conduct d Opratlonr: AMllty to locate and operate components, lncludlng local controls.

I *I4 I NUREG-IO21 Revision 9 5

Equipment Control Knowledge of limiting conditions for operations and safety limks.

Ability 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 Ability 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 4,1 3.6 2 9 i

ES-40.

BVPS-2 Fu.

S-401-2 Wdtten Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 Gmup 1 A4 I Number I WA Topics I Imp. I Q#

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

4.0 1 86 2.1.33 2.2.22 87 88 A2.05 A2.01 89 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-correct, control, or mtigate the consequences of those malfunctions or operations Containment evacuation (includlna recoontion of the alarm)

(b) based on those predictions. use procedures to 3.6 90 Knowledge of the operational Impllcltiom of the following concepts as they apply to the RCPS: Effects I

I of RCP coastdown on RCS parametera 2.8 1

Knowledge of the effect that a 1080 or malfunction of the RCPS will have on the fdlowlng: SI0 3.8 -

3.1 3 -

4 AMilty to manually operate and/or monltor In the control room: Boratlon/dilutlon batch control Ablllty to manually operate and/or monitor In the control room: Controls and Indication for closed coollng water puma Knowledge of the effect that a 1088 or malfunction of the RHRS will have on the followina: ECCS I A2.12 Ability to (a) predict the impact8 of the fdlowlng malfunctlorn or operatlorn on the ECCS; and (b) bared on those predlctlon8, uw procedure8 to correct, control, or mttlgata the consequence8 of thore malfunctlons or I

I omrationa: Condltlom, reaulrlna actuation of ECCS X I A4.01 AMllty to manually operate and/or monltor In the control room: PRT spray wpply valve

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ES4C Emergency K2 K3 K4 K5 X

X X

X x

i and Abnormal Plant Evolutions - Tier 2 Group 1 K6 AI A2 A3 A4 Number K/A Topics Imp.

Q#

Ablllty to (a) predlct the Impacts of the followlng malfuunctlons or operations on the CCWS, and (b) based or mttlgate the consequences of those malfunctlons or operatiom: HlgMow surge tank level Knowledge of the operatlonal implications of the fdlowlng concepts as the apply to the PZR PCS:

Determination of condltlon of fluid In PZR, using steam tables Knowledge of the effect d a loss or malhrnctlon of the 2.7 K606 followlng will have on the RPS: Sensors and detectors Ability to manually operate and/or monttor In the control A4.01 room: ESFAS-lnttlated equlpment which falls to actuate 4S Knowledge of CCS design feature(r) and/or interiock(8)

K4.01 whlch provide for the following: Cooling of containment 2.5 12 penetratlonr Knowledge of the physlcal connections and/or cause-systems: SEClremote monltoring systems Conduct of Operatlons: AMilty to perform speclfk modes of plant operatlon.

Knowledge of the effect that a loss or malfunction of the MRSS will have on the followlng: RCS Knowledge of the physlcai connectlom andlor cause-K1.08 effect relatlonrhlpa between the MRSS and the following 2.7 16 sptoms: MFW Ablllty to monltor automatic operatlon of the MFW, 2.9 17 A3D2 Includlng: Programmed levels of the SIG Knowledge of MFW deslgn feature@) andor interlock($)

Pump Knowledge of the phplcal connectlons and/or cause-systems: Emergency water source AMilty to monltor automatk operatlon of the AFW, Including: RCS cooldown during AFW operatiom Knowledge of bw power supplies to the followlng: Major X

A2.02 on thorn predictions, use procedures to correct, control, 3.2 8

3.5 9

X K1.02 effect relationrhlps between the CCS and the followlng 3.7 13 2.1.23 eystem and Integrated plant procedures during all 3.9 14 3,6 15 K305 X

K4.16 which provlde for the followlng: Automatlc M p for MFW 3.1 18 Kl.07 effect relatlonshlpa between the AFW and the followlng 3.6 19 4.0 20 3,3 21 A3.02

~2.

system loads X

I 008 Component Coollng Water 01 0 Pressurizer Pressure Control 01 2 Reactor Protectlon 01 3 Engineered Safdy Features Actuation 022 Containment Cooling 022 Containment Coollng 026 Contalnment Spray I x I 039 Maln and Reheat Steam 039 Maln and Reheat Steam 059 Maln Feedwater 059 Maln Feedwater 081 AuxlliarylEmergency Feedwater 081 AuxlllarylEmergency Feedwater D62 AC Electrical Distrlbutlon 1

I NUREG-1021 Revision 9 7

ESJC BvPS-2 Fc, S-401-2 Wrttten Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 Group 1 System #/Name G

K1 K2 K3 K4 K5 K6 A I A2 A3 A4 Number KJA Topics Imp Qb I 063 DC Electrlcai DMrlbution X

Knowledge of the effect that a I088 or malfunction of the dc electrlcal system wlll have on the followlng EDlG Knowledge of EDlG system design feature(@) andor EDlG whlle operatlng (normal or emergency)

Ablllty to monitor automatlc operation of the ED/G frequency and vottage control In parallel opratlon Ablllty to (a) predlct the Impacts of the following malfunctions or opratlons on the PRM system; and (b) control, or mitigate the consequence8 of thore maifunctlons or operattons. Detector failure Knowledge of bus power auppllea to the following Servlce water Knowledge of the effect that a I088 or malfunctlon of the pneumatic valves and controls AMllty to monltor automatlc operatlon of the containment 22 K3 O1 064 Emergency Diesel Generator X

K4 02 Inter-lock(s) whlch provlde for the followlng Trlpa for 3 9 23 064 Emergency Dleael Generator X

A3 05 system, Including Operatlon of the governor control of 2 8 24 073 Process Radiatlon Monitorlng X

A2 02 based on those predlctlons, use procedures to correct, 2 7 25 2 7 26 K2 o, 076 Service Water X

078 Instrument Alr X

K3 02 IAS wlll have on the followlng Systems having 3 4 27 A3 o, 28 103 Contalnrnent X

system, lncludlng Containment Isolation WA Category Polnt Totals In 3 Zd, 5

3 2

1 0 3pbj,' 4 4

Group Polnt Total 2816 r

F" NUREG-1021 Revision 9

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i Fo. 401-21 BvPS-2 F

0 Wtitten Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 Group 2 Conduct of Operations: Ability to explain and apply all system limits and precautions Ability to (a) predict the impacts of the following malfunctions or operation on the SG/S system; and (b) based on those predictions, use procedures to correct.

control, or mitigate the consequences of those malfunctions or operations: Faulted or ruptured SG Emergency Procedures / Plan Ability to verify system alarm reswnse manual.

3,8 4.5 alarm setpoints and operate controls identified in the 3 3 1

I I Svstem#/Name I G I K1 I K2 I K3 I K4 I K5 I K6 I A1 I A2 I A3 I A4 I Number I K/A Totics I Imn I CM 1 91 92 93 Knowledge of bus power supplies to the following: One-001 Control Rod Drive X

3,5 K2'01 line diagram of power supply to MIG sets.

041 Steam Dump System 002 Reactor Coolant 033 Spent Fuel Cooling 034 Fuel Handling Equipment Knowledge of RCS design feature+) and/or Interiock(s) leakaae X

K4.05 which provide for the following: Detection of RCS 045 Maln Turblne Generator I K3.01 Knowledge of the effect that a loss or malfunction of the 01 1 Prsasurlzer Level Control X

X 3.2 31 1

1 I

1 I

I 2'1 '27 K6.02 K1.18 2.9 I

I Ablltty to predlct and/or monitor changes in parameter I I

(to prevent exceeding design limb) associated wlth operating the SDS controls Including: Tavg; verlfkation above 10-10 setpoint Conduct of Operations: Knowledge of system purpose and or function.

Knowledge of the effect of a 1088 or malfunction on the following will have on the Fuel Handling System :

Radiation monitorlng systems Knowledge of the physlcal connections and/or CausB-effect relatlonshlps between the MTlG system and the following systems: RPS 2.8 2.6 068 Liquid Radwaste 3.6 X

3.6 36 Ablllty to monitor automatic operation of the Liquid A3'02 Radwaste Sydem Including: Automatic isolatlon Ability to predlct and/or monttor changes In 32 33 -

34 35 I

2.5 I 37 parametere(to prevent exceeding d i g n Ilmlh) associated with operating the Waste Gas D h p a l I

I I

I I I

I l x I I I

/ A '. s l System controls Including: Ventilation system 071 Waste Gas Dlsposlll NUREG-1021 Revision 9 9

WA Category Point Totalr:

I 23 I 38 I effect relationehip between the circulating water I 075 Circulating Water 1

1 x

1 I

I I

I I

I I

I I

K 1

4 system and the following systems: Liquid radwaste I dlschargr 112 2

1 1

1 0

1 2

0/1 1

0 Group Point Total: I 1013 NUREG-IO21 Revision 9 10

1 E M 0 1 I

Generic Knowledge and Abilities Outline (Tier 3)

I Form ES-401-3 I BVPS-2 Date of Exam:

Facilrty: I 2/26/2005 Q#

66 67 2

68 69 70 3

71 72 2

IR 4.2 3.3 3.5 4.1 3.4 3.5 Subtotal 2.4.35 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.

2.4'4 2.4.29 2.6 Knowledge of operational implications of EOP Abildy to interpret control room indications to venfy the status and operatiin of system, and understand how operator actions and directives affect plant and system conditions.

2.4'20 warnings, cautions, and notes.

2.4.48 Su btdal 3.3 3.5 Category KIA# I Topic Q#

I I

2.1.20 I Abiltty to execute procedure steps.

1 94 95 -

Knowledge of system status criteria which require the notification of plant personnel Knowledge of condud of operations r w uirements "14 3.7

2. '

I.

ondud of 3perations 2

96 97

2.

Equipment Control 2

98

3.

Radiation Control 1

99 4.3 100

4.

Emergency Procedures I Plan 2

L 7

Tier 3 Point Total NUREG1021 Revision 9 11

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

BVPSQ Date of Examination :

2/28/2005 Examination Level RO Operating Test Number:

NRC Administrative Topic (see Note)

Conduct of Operations Conduct of Operations E q u i pm en t Con t ro I Radiation Con t ro I Emergency Plan TY Pe Code*

N M

M N

Describe activity to be performed 2.1.25 Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance data (2.8)

JPM: Perform RCS Cooldown Verification 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation (3.9)

JPM: Perform an ECP Calculation 2.2.13 Knowledge of Tagging and Clearance Procedures (3.6)

JPM: Review a Tagging Request 2.3.2 Knowledge of facility ALARA program (2.5)

JPM: Determine Maximum Allowable Stay Time 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.

• Type Codes & Criteria:

(C)ontrol room (D)irect from bank (I 3 for ROs; 5 for 4 for SROs & RO retakes)

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

(P)revious 2 exams (5 1 ; randomly selected)

(S)imu lator NUREG-1021, Revision 9

Administrative Topics Outline Task Summary A1 a Given a set of plant conditions and a required RCS cooldown, the applicant will be required to determine the cooldown rate and acceptability within specified limits. This is a new JPM.

A1 b Given plant conditions prior to a reactor startup, the applicant will be required to calculate the estimated critical boron concentration. 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 Given a task to perform in the RCA, the applicant will be required to select the appropriate RWP, evaluate the RWP and a survey map, and determine maximum stay time in the work area. This is a new JPM.

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 Faci I i ty :

BVPS-2 Date of Examination:

2/2 812 005 Examination Level Administrative Topic Conduct of Operations E q u i p m en t Con t ro I Radiation Control SRO Operating Test Number:

NRC Type Code*

D, P M

M N

N Describe activity to be performed 2.1.12 Ability to apply Technical Specifications for a system (4.0)

JPM: Determine Action Required For Failed AC Sources Surveillance 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation (4.0)

JPM: Review an ECP Calculation 2.2.13 Knowledge of Tagging and Clearance Procedures (3.8)

JPM: Approve a Tagging Request 2.3.8 Knowledge of the process for performing a planned Gaseous Radioactive release (3.2)

JPM: Review a Gaseous Waste Discharge Authorization 2.4.40 Knowledge of SROs responsibilities in emergency plan implementation (4.0)

JPM: Terminate an Emergency Classification 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

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

• Type Codes & Criteria:

(C)ontrol room (D)irect from bank (5 3 for ROs; I for 4 for SROs & RO retakes)

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

(P)revious 2 exams (5 1 ; randomly selected)

(S)imulato r NUREG-1021, Revision 9

Administrative Topics Outline Task Summary A1 a Alb 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 allow the termination of an emergency classification. The conditions of this JPM are based on a Unit 2 Unusual Event as documented in LER 2-000-03. This is a new JPM.

NUREG-1021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Facility BVPS-2 Date of Examination 32812 005 Exam Level RO / SRO(I)

Operating Test No NRC Control Room Systems (8 for RO; 7 for SRO-I: 2 or 3 for SRO-U)

JPM K System Type Code Safety Function JPM Title Raise Reactor Power to IO Amps 00 1 Rod Control SI NSAL 1

s2 E02 SI Termination Perform SI Termination IAW ES-1.1 NSAE 3

s3 E03 Isolate SI Accumulators During a LOCA NSAE 4P Post LOCA C/D and Depressurization S 3 04 1 Initiate Natural Circulation Cooldown DASEP 4s Steam Dump 103 s5 Manually Actuate CIB DSAEP 5

Containment S6 064 EDG Synchronize and Load EDG 2-1 DS DS P s7 015 NIS Remove Power Range Instrument From Service Perform Manual Makeup to the VCT S8 004 cvcs DS 2

In-Plant Systems (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

PI 028 HRPS 06 1 A FW 062 AC Distribution Locally Startup a Containment Hydrogen Analyzer DER 5

Align Service Water Supply to AFW Pumps Suction DE 4s BV-2 Actions to Establish Station Blackout Cross-Tie to Unit 1 DE 6

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 P2 P3 NUREG-1021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Type Codes (A)lternate path (C)ontrol room (D)irect from hank (E)mergency or abnormal in-plant (L)ow-Power (N)ew or (M)odified from bank including I(A)

(P)revtous 2 exams I

( W A Criteria for RO I SRO-I I SRO-U 4-6 I 4-6 12-3 91 a1 4 1 1 l i 1 1 1 i i 1 2 1 2; 1

3 i 3 I 2 (randomly selected) l i l i 1 NUREG-1021, Revision 9

C o n t ro I Room/ I n - Plan t S ys tems 0 ut I i ne Task Summarv s1 s2 s3 s4 s5 S6 s7 S8 P1 P2 P3 The applicant will raise reactor power using control rods to 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.

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 the BIT and re-establish HHSI flow. 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 (GIB) actuation.

The alternate path of this task is to 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 makeup to the VCT. This is a bank JPM The applicant will locally start a containment hydrogen analyzer. This is a bank JPM that will require entry into the Radiation Control Area (RCA).

The applicant will be required to align plant service water supply to the auxiliary feedwater pumps. This is a bank JPM.

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 3

Appendix D Scenario Outline Form ES-D-1 EHCOG

'acility:

BVPSQ Scenario No.:

1 OpTest No.:

NRC Examiners:

Candidates:

CRS RO PO (R) RO (N) PO, US (C)ALL (TS) US Power Reduction for Waterbox Cleaning Turbine Control Valve failure (Load Rejection)

I nit ial Conditions:

RCS031 A MSS047A Turnover:

Critical Tasks:

(I) RO, US (TS) us (I) PO, US (TS) US Pressurizer Pressure Transmitter Fails High SG Pressure Transmitter Fails Low BOL, 100% Power 2CHS*P21 C, HHSl Pump 00s.

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

flood warnings from heavy rains.

Maintenance investigating 2SWS*P21 A, Service Water Pump abnormal vibratiodnoise.

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 MSS02A Event No.

(C) PO, US One SG Atmosphere Dump Valve Fails Partially Open 1

2 3

4 5

6 7

Malf. No.

Event Event Description Turbine Trip - Steam Dump Failure. Reactor Trip required xc2 EHC07 I0790 I (M)ALL I PPLOlA

~ I (C) RO, US I Auto and Manual R e a c h Trip Failure PPLOlB I 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-1.1.

EOP Flow Path: E-0, FR-S.1, E-0, E-2

Appendix D Scenario Outline Form ES-D-1 (R) RO (N) PO, us Scenario No.:

2 OpTestNo:

NRC Examiners:

Candidates:

C RS RO PO Reduce Power Initial Conditions:

MOL, 48% Power.

2CHS+P21C, HHSl Pump 00s.

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

Flood warnings due to heavy rains.

Maintenance investigating 2SWSP21A, 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 Turnover:

Critical Tasks:

CFWlOB RCS02A RCS02A PPL07A IL (C) PO, us (C) RO, RCS Leak us (TS) US (M) ALL SBLOCA SG A FRV Controller Fails Closed In Auto (C) RO Train A HHSVCharging Pump Auto Start Failure 1

2 3

4 5

6 7

8 9

E-1.C, Stop RCPS Event Description MSS005A I (TS) US I 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 1

AFW Start Failure (Auto SI Failure Train B) ppL07B 1 (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 HHSI/Charging Pump will fail to automatically start and must be started manually. RCPs must be tripped when criteria is met due to the LOCA. Train 6 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-I, ES-1.2

Appendix D Scenario Outline Form ES-D-1 SWSOOG Facility.

BVPSP Scenario No.:

3 OpTest No.:

NRC Examiners:

Candidates:

CRS RO PO (C) RO, us (TS) US Initial Conditions:

MOL, 25% Power.

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

Flood watch remains in effect.

Raise power to 100°/o after a trip due to loss of all circulating water.

E-O.F, Initiate Feedwater Flow with MDAFW Turnover:

Critical Tasks:

CFWOO4 DSGOlA Event 1

2 3

4 5

6 7

(M) ALL E-3.A, Isolate Ruptured SG E-0.0, Initiate CIA AFW03A PPLO7B Malf. No.

Event (C) PO BKR HlVOl (C) ALL I (TS) us RCS04B 1 (M) ALL x PPLO8B Event Description Raise Power Train A Service Water Pump Trips. (Backup pump must be manually started.)

Loss of 4KV Bus 2AE. 2-1 EDG Fails to Auto Start MFW Pump A DegradatiodTrip. Reactor Trip 2-1 EDG Failure MDAFW Train 6 Pump Auto Start Failure TDAFW Pump Auto Start Failure

~

~~

~~

~

SG B SGTR (when AFW is initiated).

CIA Fails To Automatically Actuate (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

- 1 -

Scenario Event Description NRC Scenario 3 The crew will assume the shift at approximately 25% power with instructions to raise power to 10Oio.

L 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

ADpendix D Scenario Outline Form ES-D-1 Facility:

BVPS-2 Scenario No.:

~

~

-~~

4 OpTestNo.

NRC Examiners:

Candidates:

CRS RO PO Initial Conditions:

MOL, 75% power.

2RCS*PCV455D leakage. 2RCS*MOV537, Block Valve closed with power maintained.

River water level has receded. Flood watch cancelled on last shift.

2SWS*P21 A, Service Water Pump 00s.

Reduce power at 12%/hr. in preparation for circulating water pump removal.

Turnover:

Critical Tasks:

E-Z.A, Close MSIVs Terminate ECCS prior to water relief through PORVs Event No.

1 2

3 4

5 6

7 Malf. No.

Event (R) RO (N) PO, US CRF03-K6 (C) RO, US

-I-(TS)

US (C) RO, US pcsloB I (TS) US CFW051 XMT A I (I) us CFWlOA (M) ALL MSS03 (C) PO PPLl OA PPLl OB Event Description Main Feedwater Pump (2FWS-P21 B) Trip Rapid Load Reduction Rod K-6 Drops (Reactor does not trip )

Pressurizer Master Pressure Controller Output Fails High SG A Feedwater Flow Transmitter Fails High SG A Feedwater Reg Valve failure (Unrecoverable). Reactor Trip Required.

Main Steam Break Downstream of MSIVs MSIVs Fail To Close Automatically (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

- 1 -

Scenario Event Description NRC Scenario 4 The crew will assume the shift with instructions to raise power to 100%

A main feedwater pump will trip requiring the crew to initiate a rapid load reduction. After the load reduction, one control rod will drop requiring action to realign and the Unit Supervisor to refer to Technical Specifications.

~ - - -

After the plant is stabilized, the Pressurizer master pressure controller output will fail high requiring the RO to take action to manually control Pressurizer pressure with backup heaters and spray valves. The Unit Supervisor will refer to Technical Specifications.

When Pressurizer pressure is restored to program, a SG feed flow transmitter failure will require the PO to take manual control of the affected SG main feedwater control valve.

When the affected SG level is under control, an unrecoverable main feedwater control valve failure will require a reactor trip.

Upon reactor trip, a steam break will develop downstream of the MSIVs. SI will actuate; however, main steamline isolation will not occur automatically.

The steam line break will be terminated after manual actuation of main steamline isolation by the PO.

The scenario may be terminated when the crew stops HHSl pumps in ES-1.l.

EOP Flow Path: E-0, ES-0.1, E-0, ES-1.1

I ES-401 I

Record of Rejected WAS I Form ES-401-4 1

[ Tier I Randomly 00lAAl.04 112 I 003AK2.03 01 2K6.11 0332.4.6 211 I 059K4.14 211 I 061Kl.10 I

G2*2-9 1 I 1 I 0272.4.49 0622.1.14 067AA2.11 0622.1.23 G2.4.29 3

I G2.2.17 062 AA2.06 061 AA2.05 E16 G2.4.4 112 I 037AK1.01 112 I E16G2.1.30 211 I 006A2.11 211 I 062A2.15 045 A2.17 028 A1.Ol Reason for Rejection The subject WA isnt relevant at the subject facility.

The subject WA isnt relevant at the subject facility.

The subject WA isnt relevant at the subject facility.

The subject WA isnt relevant at the subject facility.

~~

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 sitsspecific priority that justifies 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 sitespecific priority that justifies keeping the KIA if its importance rating is below 2.5.

The subject WA isnt relevant at the subject facility.

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 Random selection of replacement KA was a duplicate topic Duplicate of KA already selected KA deleted because 3 topics selected for Generic Section 2. Replaced with 2.4.4 It isnt possible to prepare a psychometrically sound question related to the subject KIA. Plant effects are minimal, not operationally valid Double Jeopardy with Question 8.

Double Jeopardy with Question 90. Also, significant number of radiation monitoring questions on exam Procedure contains one step; operationally insignificant Topic not operationally valid. This event does not require use of steam tables.

Procedure contains one step unrelated to topic It isnt possible to prepare a psychometrically sound question relevant to this WA. No procedure exists for this event, and the closest possible topic would duplicate question 54 No procedure guidance for KA statement, and question would test same knowledge as Question 24 WA identical to event performed in dynamic simulator System removed (retired in place) at facility NUREG-1021 Revision 9 12