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 o ES-401-2

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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).

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. Systemdevolutionswithin 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.

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.

a. 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.

9. 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 selectionsto WAs that are linked to 10CFR55.43 NUREG-1021 Revision 9

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ES4 BVPS-2 Wrttten Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 1 UAPE # / Name Safety Function L

Al A2 I Number I WA Topic(s) Q#

Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:

025 I Loss of RHR System I 4 X AA2.02 3.8 76 Leakage of reactor coolant from RHR into closed cooling water system or into reactor bullding atmosphere Ability to determine or interpret the following as they apply 029 I ATWS I 1 X EA2.01 77 to a A W S : Reactor nuclear instrumentation 7

Equipment Control Knowledge of bases in technical 038 I Steam Gen. Tube Rupture I 3 2.2.25 specifcations for limiting condtions for operations and 78 safety limits.

Equipment Control Knowledge of bases in technical 058 I Loss of DC Power 16 2.2.25 specifications for limiting conditions for operations and safety Ilmits.

Emergency Procedures I Plan Knowledge of annunciators 062 / Loss of Nuclear Svc. Water I 4 2.4.31 alarms and indications,and use of the response instructions.

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

E l 1 I Loss of Emergency Coolant Recirc. I 4 X EA2 1 Facility conditions and selection of appropnate procedures during abnormal and emergency operations.

007 IReactor Trlp StaMllzatlon Recovery I1 X EAI.09 Ablllty to operate and monitor the followlng as they apply to a reactor trlp: CVCS Ablllty to determine and Interpretthe following ae they apply to the Pressurizer Vapor Space Accldent: The 008 IPrerwrlzer Vapor Space Accldent I3 X AA2.20 effect of an open PORV on code safety, bared on observation of plant parameten Knowledge of the operational implkatlm of the followlng 015 I17 IRCP Malfunctlonm I 4 AKI .E 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 Interpretthe followlng as they 022 I Loss of Rx Coolant Makeup I 2 X AA2.04 apply to the Lobs of Reactor Coolant Makeup: How long PZR level can be malntalnedwtthln llmlts Ablllty to operate and Ior monitor the followlng ae they 025 IL m of RHR System / 4 X AA1.08 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 027 IPrereurlzer Pressure Control System Malfunctlon I X AA2.02 apply to the Pro8surlzw Preaaure Control Malfunctions:

3 Normal MIUW for RCS warwre Knowledge of the reasons for the followlng responses as 02QIAlWSIl EK3.11 4.2 the apply to the A W S : lnitlating emergency boratlon NUREG-1021 Revision 9 2

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V O. BVPS-2 Fo, LS-401-2 Written Examination Outllne Emergency and Abnormal Plant Evolutions Tier 1 Group 1 1 WAPE # / Name Safety Function IG Ablllty to operate and monitor the following as they apply 038 ISteam Gem Tube Rupture I 3 EA1.16 to a SGTR: SI0 atmospherlc relief valve and secondary

- PORV controllers and Indicators Ablllty to determine and Interpret the following as they 040 ISteam Llne Rupture Excessive Heat Transfer I 4 X AA2.05 apply to the Steam Line Rupture: When ESFAS rystems mav be Mcured Ablllty to operate and I or monitor the fdlowlng as they 054 I Loaa of Maln Feedwater I 4 apply to the Loss of Maln Feedwater (MFW): HPI, under total feedwater loas conditions Knowledge of the operational Implicationsofthe followlng ~

055 IStation Blackout I 6 EKI .02 concepts as they apply to the Station Blackout : Natural 4.1 circulationcooling Ablllty to operate and I or monitor the followlng as they 056 ILoss of Off-site Power I 6 AAI .10 apply to the Lose of Offslte Power: Auxillarylemergency 43 feedwater pump (motor drlven) 057 ILoss of Vital AC Inst. Bus I 6 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 3.7 1 51 Ablllty to operate and I or monltor the followlng a8 they 058 ILoss of DC Power I6 AAI .Ol 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 062 ILoss of Nuclear Svc. Water I 4 AA2.08 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 E04 ILOCA Outslde Containment I 3 X EA2.I condltlorm and selectlon of appropriate procedures during abnormal and emergency operatlons.

Knowledge of the reasons for the followlng responses as E05 IInadequate Heat Transfer Loss of Secondary EK3.2 they apply to the (Loss of Secondary Heat Slnk) Normal, 3.7 55 Heat Slnk I 4 abnormal and emergency operatlng procedures associated with (Loss of Secondary Heat Slnk).

Knowledge of the operatlonal Implicationsof the foliowlng concepts as they apply to the (Loss of Emergency E l 1 IL m of Emergency Coolant Reclrc. I 4 EK1.3 Coolant Reclrculatlon)Annunciators and condfflons 3.6 56 lndtcatlng dgnab, and remedlal actlons a88oclated with the (Loaa of Emergency Coolant Reclrculatlon).

I WA Category Polnt Totalr: Ion 7t3 I Grou~PolntTotal: I I lW6 NUREG-1021 Revision 9 3

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[ESJG BVPS-2 is-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 2 I VAPE#/NameSafety Function 1 G I K1 b Number WA Topic(s) I Imp.

Emergency Procedures I Plan Knowledge of which events 037 I Steam Generator Tube Leak 1 3 X 2.4 30 related to system operationdstatus should be reported to 3.6 outside agencies.

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

E01 8 E02 / Rediagnosis and SI Termination I 3 EA2.1 4,0 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 E09 / Natural Circulation Operations I 4 EA2.2 3.8 appropriate procedures and operation within the limitations in the facilitv's license and amendments.

Emergency Procedures / Plan Ability to recognize abnormal indications for system operating parameters E06 I Degraded Core Cooling 1 4 X 2.4.4 43 which are entry-level conditions for emergency and abnormal operating procedures.

001 I Contlnuous Rod Withdrawal I 1 l l X AA1 . O l Ability to operate and I or monitor the followlng a8 they apply to the Contlnuous Rod WRhdrawal: Bank select 1 3.5 swltch 003 I Dropped Control Rod I 1 I I -

X AK2.05 Knowledge of the lntenelatlons between the Dropped Control Rod and the following: Control rod drive power suppitem and loglc clrcub 1 037 I Steam Generator Tube Leak 13 I Ix AK1.02 Knowledge of the operational Impllcations of the followlng concept8 w they apply to Steam Generator Tube Leak:

Leak rate. VI pressure drop Abillty to deterrnlne and Interpret the following w they 081 I ARM System A l a r m I 7 X AA2.01 apply to the Area Radlatlon Monitorlng (ARM) System

- Alarmr: ARM panel dlrplaya Emergency Procedures I Plan Knowledge of annunciatore EO1 8 EO2 I Redlagnods and SI Termlnatlon I 3 X 2.4.31 alarms and lndlcatlone, and use of the response Instructions.

Knowledge of the reasons for the following responses as they apply to the (Saturated Core Cooling) Manipulation of E07 I Inad. Core Cooling I 4 X EK3.3 3.8 controls required to obtaln deslred operatlng reaub during abnormal and emergency eltuatlone.

Knowledge of the reasons for the followlng rerponses as they apply to the (Pressurized Thermal Shock)

L E08 I RCS Overcoollng PTS I 4 l l 1 I X EK3.3 Manlpulatlon of controls required to obtaln dealred operatlng results durlng abnormal and emergency situatlons.

3.7 NUREG-1021 Revision 9 4

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ESdC BvPS-2 Fc 3-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 2 1 UAPEWNameSafety Function 1 G I K1 I K2 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, E l 3 I Steam Generator Over-pressure I 4 X EK2.1 and functions of control and safety rystema, lncludlng 3.0 64 Inatrumentatlon, dgnals, Interlocks, hllure mod-, and automatlc and manual features.

068 I Control Room EvecuaUon 18 X Conduct d Opratlonr: AMllty to locate and operate components, lncludlng local controls. I 3.9 I I WA Category Point Total: 2 / 2 1 2 I *I4 I NUREG-IO21 Revision 9 5

- 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#

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

4.0 1 86 Equipment Control Knowledge of limiting conditions for 2.2.22 4,1 87 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 A2.05 3.6 88 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 A2.01 those predictions, use procedures to correct, control, or 29 89 mitigate the consequences of those malfunctions or operations: Air dryer and filter malfunctions Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-(b) based on those predictions. use procedures to 3.6 90 correct, control, or mtigate the consequences of those malfunctions or operations Containment evacuation (includlna recoontion of the alarm)

I I Knowledge of the operational Impllcltiom of the following concepts as they apply to the RCPS: Effects of RCP coastdown on RCS parametera Knowledge of the effect that a 1080 or malfunction of the 2.8 1 RCPS will have on the fdlowlng: S I 0 AMilty to manually operate and/or monltor In the control 3.8 3 room: Boratlon/dilutlon batch control Ablllty to manually operate and/or monitor In the control room: Controls and Indication for closed coollng water 3.1 4 puma Knowledge of the effect that a 1088 or malfunction of the RHRS will have on the followina: ECCS II A2.12 I

Ability to (a) predict the impact8 of the fdlowlng malfunctlorn or operatlorn on the ECCS; and (b) bared on those predlctlon8, uw procedure8to correct, control, or mttlgatathe consequence8 of thore malfunctlons or 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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i ES4C Emergency and Abnormal Plant Evolutions Tier 2 Group 1 K2 K3 K4 K5 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 008 Component Coollng Water X A2.02 on thorn predictions, use proceduresto correct, control, 3.2 8 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:

010 Pressurizer Pressure Control X 3.5 9 Determination of condltlon of fluid In PZR, using steam tables Knowledge of the effect d a loss or malhrnctlon of the 012 Reactor Protectlon X K606 2.7 followlng will have on the RPS: Sensors and detectors 013 Engineered Safdy Features Ability to manually operate and/or monttor In the control Actuation A4.01 room: ESFAS-lnttlated equlpment which falls to actuate 4S Knowledge of CCS design feature(r) and/or interiock(8) 022 Containment Cooling X K4.01 whlch provide for the following: Cooling of containment 2.5 12 penetratlonr Knowledge of the physlcalconnections and/or cause-022 Containment Coollng K1.02 effect relationrhlps between the CCS and the followlng 3.7 13 systems: SEClremote monltoringsystems 026 Contalnment Spray 039 Maln and Reheat Steam IxI X 2.1.23 K305 Conduct of Operatlons: AMilty to perform speclfk eystem and Integratedplant procedures during all modes of plant operatlon.

Knowledge of the effect that a loss or malfunctionof the 3.9 3,6 14 15 MRSS will have on the followlng: RCS Knowledge of the physlcai connectlom andlor cause-039 Maln and Reheat Steam K1.08 effect relatlonrhlpa between the MRSS and the following 2.7 16 sptoms: MFW 059 Maln Feedwater X A3D2 Ablllty to monltor automatic operatlon of the MFW, 2.9 17 Includlng: Programmed levels of the SIG Knowledge of MFW deslgn feature@) andor interlock($)

059 Maln Feedwater X K4.16 which provlde for the followlng: Automatlc M p for MFW 3.1 18 Pump Knowledge of the phplcal connectlons and/or cause-081 AuxlliarylEmergency Kl.07 effect relatlonshlpa between the AFW and the followlng 3.6 19 Feedwater systems: Emergency water source 081 AuxlllarylEmergency AMilty to monltor automatk operatlon of the AFW, X A3.02 4.0 20 Feedwater I Including: RCS cooldown during AFW operatiom D62 AC Electrical Distrlbutlon 1 I x ~2.

Knowledge of bw power supplies to the followlng: Major system loads 3,3 21 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 AI A2 A3 A4 Number KJA Topics Imp Qb I Knowledge of the effect that a I088 or malfunctionof the 063 DC Electrlcai DMrlbution X K3 O1 dc electrlcal system wlll have on the followlng EDlG 22 Knowledge of EDlG system design feature(@)andor 064 Emergency Diesel Generator X K4 02 Inter-lock(s) whlch provlde for the followlng Trlpa for 39 23 EDlG whlle operatlng (normal or emergency)

Ablllty to monitor automatlc operation of the ED/G 064 Emergency Dleael Generator X A3 05 system, Including Operatlon of the governor control of 28 24 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) 073 Process Radiatlon Monitorlng X A2 02 based on those predlctlons, use proceduresto correct, 27 25 control, or mitigatethe consequence8 of thore maifunctlons or operattons. Detector failure Knowledge of bus power auppllea to the following 076 Service Water X K2 o, Servlce water 27 26 Knowledge of the effect that a I088 or malfunctlon of the 078 Instrument Alr X K3 02 IAS wlll have on the followlng Systems having 34 27 pneumatic valves and controls AMllty to monltor automatlc operatlon of the containment 103 Contalnrnent X A3 o, system, lncludlng Containment Isolation 28 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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F 0 BvPS-2 Fo. 401-21 Wtitten Examination Outline 1

Emergency and Abnormal Plant Evolutions Tier 2 Group 2 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 Conduct of Operations: Ability to explain and apply all 3,8 91 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 proceduresto correct. 4.5 92 control, or mitigate the consequences of those malfunctionsor operations: Faulted or ruptured SG Emergency Procedures / Plan Ability to verify system alarm setpoints and operate controls identified in the 33 93 alarm reswnse manual.

001 Control Rod Drive X Knowledge of bus power supplies to the following: One-K2'01 3,5 line diagram of power supply to MIG sets.

Knowledge of RCS design feature+) and/or Interiock(s) 002 Reactor Coolant X K4.05 which provide for the following: Detection of RCS leakaae-Knowledge of the effect that a loss or malfunction of the 01 1 Prsasurlzer Level Control X K3.01 3.2 31 1 1 I I I 1 I I Ablltty to predlct and/or monitor changes in parameter I I 041 Steam Dump System (to prevent exceeding design limb) associatedwlth operatingthe SDS controls Including: Tavg; verlfkation 2.9 ' 32 above 10-10 setpoint Conduct of Operations: Knowledge of system purpose 033 Spent Fuel Cooling 2'1'27 2.8 33

- and or function.

Knowledge of the effect of a 1088 or malfunction on the 034 Fuel Handling Equipment X K6.02 following will have on the Fuel Handling System : 2.6 34 Radiation monitorlngsystems I

Knowledge of the physlcal connections and/or CausB-045 Maln Turblne Generator K1.18 effect relatlonshlpsbetween the MTlG system and the 3.6 35 following systems: RPS Ablllty to monitor automatic operation of the Liquid 068 Liquid Radwaste X A3'02 3.6 36 Radwaste Sydem Including:Automatic isolatlon 071 Waste Gas Dlsposlll I I I I I I I l x I I I /A'.sl Ability to predlct and/or monttor changes In parametere(to prevent exceeding d i g n Ilmlh) associated with operating the Waste Gas D h p a l System controls Including: Ventilation system I I2.5 37 NUREG-1021 Revision 9 9

I 075 Circulating Water 1 1 x 1 I I I I I I I I I K 1 effect relationehip between the circulating water 4 system and the following systems: Liquid radwaste I dlschargr I 23 I I 38 WA Category Point Totalr: 112 2 1 1 1 0 1 2 0/1 1 0 Group Point Total: I 1013 NUREG-IO21 Revision 9 10

1 EM01 I Generic Knowledge and Abilities Outline (Tier 3) I Form ES-401-3 I Facilrty: I BVPS-2 Date of Exam: 2/26/2005 Category KIA# I I

Topic I

Q# IR Q#

2.1.20 I Abiltty to execute procedure steps. 1 4.2 94 Knowledge of system status criteria which 3.3 95 "14 require the notification of plant personnel -

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2. ' Knowledge of condud of operations r w uirements 3.7 66 3perations 67 2 2 3.5 96 4.1 97 68 69 2.

Equipment Control 70 3 2 3.4 98

3. 71 Radiation Control 72 Subtotal 2 1 Knowledge of local auxiliary operator tasks 2.4.35 during emergency operations including system 3.5 99 geography and system implications.

Ability to recognize abnormal indications for system operating parameters which are entry- 4.3 100 2.4'4 level conditions for emergency and abnormal operating procedures.

4.

Emergency 2.4.29 Knowledgeof the emergency plan. 2.6 Procedures I Plan Knowledgeof operational implications of EOP 3.3 2.4'20 warnings, cautions, and notes.

Abildy to interpret control room indications to venfy the status and operatiin of system, and 2.4.48 3.5 understand how operator actions and directives affect plant and system conditions.

Subtdal 2 Tier 3 Point Total L 7 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 TY Pe Describe activity to be performed (see Note) Code*

2.1.25 Ability to obtain and interpret station reference Conduct of Operations materials such as graphs, monographs, and tables which contain performance data (2.8)

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

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

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

Radiation Cont roI N

JPM: Determine Maximum Allowable Stay Time Emergency Plan 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 (I3 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)imulator 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 FaciIity : BVPS-2 Date of Examination: 2/2812 005 Examination Level SRO Operating Test Number: NRC Administrative Topic Type Describe activity to be performed Code*

2.1.12 Ability to apply Technical Specifications for a Conduct of Operations system (4.0)

D, P 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 M plant operation (4.0)

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

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

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

N 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 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.

Alb 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 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 Safety JPM Title Code Function SI 00 1 Raise Reactor Power to IO Amps NSAL 1 Rod Control s2 E02 Perform SI Termination IAW ES-1.1 NSAE 3 SI Termination s3 E03 Isolate SI Accumulators During a LOCA NSAE 4P Post LOCA C/D and Depressurization S3 04 1 Initiate Natural Circulation Cooldown DASEP 4s Steam Dump s5 103 Manually Actuate CIB DSAEP 5 Containment 064 S6 Synchronize and Load EDG 2-1 DS EDG s7 015 Remove Power Range Instrument From Service DSP NIS S8 004 Perform Manual Makeup to the VCT DS 2 cvcs In-Plant Systems (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

PI 028 Locally Startup a Containment Hydrogen Analyzer DER 5 HRPS 06 1 P2 Align Service Water Supply to AFW Pumps Suction DE 4s A FW 062 P3 BV-2 Actions to Establish Station Blackout Cross-Tie to Unit 1 DE 6 AC Distribution 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

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 I (A)lternatepath Type Codes Criteria for RO I SRO-I I SRO-U 4-6 I 4-6 12-3 (C)ontrol room (D)irect from hank 91 a 1 4 (E)mergency or abnormal in-plant 11 l i 1 (L)ow-Power 11 i i 1 (N)ewor (M)odified from bank including I ( A ) 2 1 2; 1 (P)revtous2 exams 3i 3I 2 (randomly selected)

(WA l i l i 1 NUREG-1021, Revision 9

Co nt roI Room/In- Plant Sys tems 0utIine Task Summarv s1 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.

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 the BIT and re-establish HHSI flow. 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 (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.

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

s7 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 makeup to the VCT. This is a bank JPM P1 The applicant will locally start a containment hydrogen analyzer. This is a bank JPM that will require entry into the Radiation Control Area (RCA).

P2 The applicant will be required to align plant service water supply to the auxiliary feedwater pumps. This is a bank JPM.

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 3

Appendix D Scenario Outline Form ES-D-1

'acility: BVPSQ Scenario No.: 1 OpTest No.: NRC Examiners: Candidates: CRS RO PO I nitial Conditions: BOL, 100% Power 2CHS*P21C, HHSl Pump 00s.

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

flood warnings from heavy rains.

Maintenance investigating 2SWS*P21A, Service Water Pump abnormal vibratiodnoise.

Turnover: Initiate power reduction to 75% for waterbox cleaning.

Critical Tasks: FR-S.l.C, Initiate RCS Boration and/or insert RCCAs E-2.A, Isolate Faulted SG Event Malf. No. Event No. Event Description 1 (R) RO Power Reduction for Waterbox Cleaning (N) PO, US 2 EHCOG (C)ALL Turbine Control Valve failure (Load Rejection)

(TS) US 3 RCS031A (I) RO, US Pressurizer Pressure Transmitter Fails High (TS) us 4 MSS047A (I) PO, US SG Pressure Transmitter Fails Low 5 xc2 EHC07 I0790 I (TS) US (M)ALL I Turbine Trip - Steam Dump Failure. Reactor Trip required 6 PPLOlA ~ I (C) RO, US I Auto and Manual R e a c h Trip Failure PPLOlB I I 7 MSS02A (C) PO, US One SG Atmosphere Dump Valve Fails Partially Open (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 Scenario No.: 2 OpTestNo: NRC Examiners: Candidates: C RS RO PO 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.

Turnover: Reduce power to take the unit off-line due to circulating water intake clogging Critical Tasks: E-0.1, Start Train B HHSl I

L E-1.C, Stop RCPS Event Description 1 (R) RO Reduce Power (N) PO, us 2 MSS005A I (TS) US I SG Level Transmitter Fails High 3 DSGOlB 1 (TS) US I Train B (2-2) EDG Failure 4

5 LDS007A CFWlOB (C) RO, Ius (C) PO, 1 Letdown Pressure Control Valve Fails Closed In Auto SG A FRV Controller Fails Closed In Auto us 6 RCS02A (C) RO, RCS Leak us (TS) US 7 RCS02A (M) ALL SBLOCA 8 PPL07A (C) RO Train A HHSVCharging Pump Auto Start Failure 9

ppL07B 1 (c)po I AFW Start Failure (Auto SI Failure Train B)

(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 Amain 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 Facility. BVPSP Scenario No.: 3 OpTest No.: NRC Examiners: Candidates: CRS RO 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°/o after a trip due to loss of all circulating water.

Critical Tasks: E-O.F, Initiate Feedwater Flow with MDAFW E-3.A, Isolate Ruptured SG E-0.0, Initiate CIA Event Malf. No. Event Event Description 1 Raise Power 2 SWSOOG (C) RO, Train A Service Water Pump Trips. (Backup pump must be us manually started.)

(TS) US 3

4 BKR HlVOl CFWOO4 I (C) ALL (TS)u s (M) ALL Loss of 4KV Bus 2AE. 2-1 EDG Fails to Auto Start MFW Pump A DegradatiodTrip. Reactor Trip DSGOlA 2-1 EDG Failure 5 AFW03A (C) PO MDAFW Train 6Pump Auto Start Failure PPLO7B TDAFW Pump Auto Start Failure 6 RCS04B 1 (M) ALL ~~ ~ ~~

SG B SGTR (when AFW is initiated).

~

7 x PPLO8B CIA Fails To Automatically Actuate (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Event Description NRC Scenario 3 L

The crew will assume the shift at approximately 25% power with instructions to raise power to 10Oio.

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 0 0 s .

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

Critical Tasks: E-Z.A, Close MSIVs Terminate ECCS prior to water relief through PORVs Event Malf. No. Event No. Event Description 1 Main Feedwater Pump (2FWS-P21 B) Trip

-I 2 (R) RO Rapid Load Reduction (N) PO, US 3 CRF03-K6 (C) RO, US Rod K-6 Drops (Reactor does not trip )

(TS)US 4

pcsloB I (C) RO, US (TS) US Pressurizer Master Pressure Controller Output Fails High 5

6 XMT A CFW051 CFWlOA I (I)

(M) ALL us SG A Feedwater Flow Transmitter Fails High SG A Feedwater Reg Valve failure (Unrecoverable). Reactor Trip Required.

7 MSS03 (C) PO Main Steam Break Downstream of MSIVs PPLl OA MSIVs Fail To Close Automatically PPLl OB (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 Reason for Rejection The subject WA isnt relevant at the subject facility.

00lAAl.04 The subject WA isnt relevant at the subject facility.

112 I 003AK2.03 The subject WA isnt relevant at the subject facility.

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

_____ ~~

0332.4.6 The subject WA isnt relevant at the subject facility.

211 I 059K4.14 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.

211 I 061Kl.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 I G2*2-9 level of the proposed examination, and there isnt a sitespecific priority that justifies keeping the KIA if its importance rating is below 2.5.

1I 1 I 0272.4.49 The subject WA isnt relevant at the subject facility.

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

It isnt possible to prepare a psychometrically sound question related to the subject 067AA2.11 0622.1 .23 Random selection of replacement KA was a duplicate topic G2.4.29 Duplicate of KA already selected 3 I G2.2.17 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 062 AA2.06 Double Jeopardy with Question 8.

Double Jeopardy with Question 90. Also, significant number of radiation monitoring 061 AA2.05 questions on exam E16 G2.4.4 Procedure contains one step; operationally insignificant 112 I 037AK1.01 Topic not operationally valid. This event does not require use of steam tables.

112 I E16G2.1.30 Procedure contains one step unrelated to topic 211 I 006A2.11 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 54 211 I 062A2.15 No procedure guidance for KA statement, and question would test same knowledge as Question 24 045 A2.17 WA identical to event performed in dynamic simulator 028 A1 .Ol System removed (retired in place) at facility NUREG-1021 Revision 9 12