ML051890160

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Draft - Outlines (Folder 2)
ML051890160
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 02/18/2005
From: Mccall K
AmerGen Energy Co
To: D'Antonio J
Operations Branch I
Conte R
References
ES-401
Download: ML051890160 (33)
v • d • e


Text

ES-401 PWR Examination Outline Form ES-401-2

& 2 1 2 2 2 1 1 9 2 2 4 Abnormal Plant Tier 2 3 3 9 8 2 27 6 4 10 Evolutions Totals 1 2 1 3 5 1 3 3 5 2 2 1 28 2 3 5 2.

Plant 2 1 1 1 1 2 1 0 0 1 1 1 10 2 1 3 Systems Tier 3 2 4 6 3 4 3 5 3 3 2 38 4 4 8 Totals

3. Generic Knowledge and 1 2 3 4 1 2 3 4 10 7 Abilities Categories 3 2 2 3 2 2 1 2 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 (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each WA category shall not be less than two).
2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by k 1 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. Systems/evolutions 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 inamrorxiate WA statements.

I I I 4.

5.

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.

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

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

8.

9.

NUREG-1021 1

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 1 m f e t (

y Function K/A Topic(s) 025 I Loss of RHR System I 4 X I I 2114 I Conduct of Operations Knowledge of system status criteria which require the notification of plant personnel 33 76 027 IPressurizer Pressure Control System Malfunction /

Ix AA205 I Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions PZR heater setpoints 1 33 77 040 ISteam Line Rupture - Excessive Heat Transfer I 4 I 1-2114 I Conduct of Operations Knowledge of system status criteria which require the notification of plant personnel Ability to determine and interpret the following as they 33 78 057 / Loss of Vital AC lnst Bus I 6 AA2 12 apply to the Loss of Vital AC Instrument Bus PZR level 37 79 controller instrumentation.and heater indications Ability to operate and / or monitor the following as they apply to the (Vital System Status Verification) Facility E02 / Reactor Trip - Stabilization - Recovery 1 EA2 1 4.0 80 conditions and selection of appropriate proceduresduring Conduct of Operations Ability to explain and apply all E05 I Steam Line Rupture - Excessive Heat Transfer / 4 2.1.32 system limits and precautions 38 81 Ability to determine or interpret the following as they apply to a reactor trip: If reactor should have tripped but has not 007 IReactor Trip - Stabilization - Recovery I1 EA2.04 4.4 39 done so, manually trip the reactor and carry out actions in ATWS EOP

+

Ability to determine and interpret the following as they 008 IPressurizer Vapor Space Accident I 3 AA2.04 apply to the Pressurizer Vapor Space Accident: High- 3.2 40 temperature computer alarm and alarm type 009 ISmall Break LOCA I 3 I EA2.10 Ability to determine or interpret the following as they apply to a small break LOCA: Airborne activty 3.1 41 Ability to operate and monitor the following as they apply 01 1 ILarge Break LOCA I 3 4.1 42 to a Large Break LOCA: A M I and SWS pumps Ability to operate and Ior monitor the following as they 015 I17 IRCP Malfunctions I 4 AAl.11 apply to the Reactor Coolant Pump Malfunctions(Loss of 2.5 43 RC Flow): RCP onloff and run indicators Ability to operate and Ior monitor the following as they 022 I Loss of Rx Coolant Makeup I 2 AA1.01 apply to the Loss of Reactor Coolant Pump Makeup: 3.4 44 CVCS letdown and charging Knowledge of the interrelations between the Loss of 025 I Loss of RHR System I4 45 water or closed cooling water pumps Ability to determine and interpret the following as they 026 ILoss of Component Cooling Water I 8 46 of a leak in the CCWS NUREG-1021 2

PWR Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 1 029lAlWSll i i EK1.01 Knowledge of the operational implications of the following concepts as they apply to the ATWS: Reactor nucleonics and thermo-hydraulics behavior 2.8 47 Ability to operate and monitor the following as they apply 038 I Steam Gen. Tube Rupture 13 EA1.09 to a SGTR: PZR tank levellpressure indicators, gauges, and recorder.

~ ~~ _____~~______

Conduct of Operations: Knowledge of operator 040 I Steam Line Rupture - Excessive Heat Transfer I 4 2.1.2 responsibilities during all modes of plant operation.

054 ILoss of Main Feedwater I 4 055 I Station Blackout 1 6 057 I Loss of Vial AC Inst. Bus I 6 tt -t X AA1.01 EA2.02 AA2.18 Ability to determine or interpret the following as they apply to a Station Blackout: RCS core cooling through natural circulation cooling to SIG cooling Ability to determine and interpret the following as they apply to the Loss of Vial AC Instrument Bus: The indicator, valve, breaker, or damper position which will occur on a loss of power 4.4 3.1 51 52 Ability to operate and I or monitor the following as they apply to the Loss of DC Power: Static inverter dc input 058 I Loss of DC Power I 6 AA1.02 breaker, frequency meter, ac output breaker, and ground fault detector Knowledge of the reasons for the following responses as 065 ILoss of Instrument Air I8 AK3.08 they apply to the Loss of Instrument Air: Actions contained in EOP for loss of instrument air Ability to operate and Ior monitor the following as they I apply to the (vnal System Status Verification): Adherence E02 IReactor Trip - Stabilization - Recovery I1 X EA2.2 to appropriate procedures and operation within the 3,2 1 55 limitations in the facilitys license and amendments. I Ability to operate and Ior monitor the following as they EO4 Ihadequate Heat Transfer - Loss of Secondary EA1.2 apply to the (Inadequate Heat Transfer) Operating 3.4 56 X

Heat Sink I 4 behavior characteristics of the facility.

I KIA Cateaow Point Totals: I 7 714 Group Point Total: I 1816 NUREG-I021 3

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 I I ElAPE # / Name Safety Function I G I K1 I K2 I K3 I A1 1 A2 1 Number I - KIA T Imp. I Q# I Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions 028 / Pressurizer Level Malfunction I 2 X AA2.,o 34 82 Whether the automatic mode for PZR level control IS functioning improperly, necessity of shifl to manual modes Conduct of Operations Ability to recognize indications for 051 / Loss of Condenser Vacuum I4 2.1.33 system operating parameters which are entry-level 40 83 conditions for technical specifications Ability to determine and interpret the following as they 059 I Accidental Liquid RadWaste Re1 I 9 X AA2.04 apply to the Accidental Liquid Radwaste Release. The 3.5 84 valve lineup for a release of radioactive liquid Equipment Control Knowledge of bases in technical E08 / LOCA Cooldown - Depress I 4 2.2.25 specifications for limiting conditions for operations and 3.7 85 safety limits Knowledge of the interrelations between the Dropped 003 I Dropped Control Rod I1 X AK2.05 Control Rod and the following: Control rod drive power 2.5 57 supplies and logic circuits Ability to operate and I or monitor the following as they 028 I Pressurizer Level Malfunction I 2 X AA1.06 apply to the Pressurizer Level Control Malfunctions: 3.3 58 Checking of RCS leaks Knowledge of the reasons for the following responses as 060 IAccidental Gaseous RadWaste Rel. I 9 X AK3.02 they apply to the Accidental Gaseous Radwaste: Isolation 3.3 59 of the auxiliary building ventilation Ability to operate and I or monitor the following as they 076 I High Reactor Coolant Activity I 9 X AA1.04 apply to the High Reactor Coolant Activity: Failed fuel- 3.2 60 monitoring equipment Knowledge of the reasons for the following responses as they apply to the (Loss of NNI-X) Manipulationof controls A02 I LOSSOf NNI-XPI I 7 , AK3.3 , required to obtain desired operating results during 3.7 61 I I I I X I I I I I I I I abnormal. and emergency situations. I I'  !

Knowledge of the interrelations between the Fuel Handling 036 I Fuel Handling Incidents X AK2,0, 2.9 63 Incidents and the following: Fuel Handling Equipment Conduct of Operations: Ability to perform specific system A06 I Shutdown Outside Control Room I 8 X 2.1.23 and integrated plant procedures during all modes of plant 3.9 62 operation Ability to determine and interpret the following as they apply to the (Inadequate Subcooling Margin) Adherence E03 I Inadequate Subcooling Margin 1 4 X EA2,2 3.5 64 to appropriate procedures and operation within the limitations in the facility's license and amendments.

NUREG-1021 4

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 I E/APE # / Name Safety Function I G 1 K1 1 K2 I K3 I A I I A2 1 Number I K/A Topic(s) I Imp. I Q# I Knowledge of the operational implications of the following concepts as they apply to the (EOP Rules) Normal, X EK' '2 3.0 65 abnormal and emergency operating procedures associated with (EOP Rules).

K/A Category Point Total: 1/2 1 2 2 2 1/2 Group Point Total: 914 NUREG-1021 5

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 Group 1 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS. and (b) based A2.02 on those predictions. use procedures to correct control 42 86 or mitigate the consequences of those malfunctions or t

operations Loss of PZR level (failure mode)

Emergency ProceduresiPlan Knowledge of which events related to system operation/statusshould be 36 a7 reported to outside agencies Emergency ProcedureslPlan Ability to perform without 2 4.49 reference to procedures those actions that require 40 90 immediate operation of system components and controls 1

Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system, and (b) based on those predictions, use proceduresto correct, A2.04 3.0 88 control, or mitigate the consequences of those malfunctions or operations Unloading prior to securing an EDlG Equipment Control Knowledge of limiting conditions for operations and safety limits

, 89 Knowledge of RCPS design feature@)and/or interlock(s) which provide for the following Adequate 2a 2 t

cooling of RCP motor and seals Knowledge of the effect that a loss or malfunctionof the 3,9 1 RCPS will have on the following: RPS Knowledge of the effect of a loss or malfunction on the K6.14 following CVCS components: Recirculationpath for 2.7 3 charging pumps I K2.05 Knowledge of bus power supplies to the following:

I I 2.7 4 t

MOVs Knowledge of the efhxt of a loss or malfunction on the K6.03 2,5 5 following will have on the RHRS: RHR heat exchanger Knowledge of ECCS design feature@) andlor K4.20 interlock(s) which provide for the following: Automatic 3.2 6 h

closure of common line and fill valves to accumulator Ability to manually operate and/or monitor in the control room: Recognition of leaking PORV/code safety 3.6 a Knowledge of the effect that a loss or malfunctionof the K3.01 PRTS will have on the following: Containment 3.3 7 6

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 Group 1 A

Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based 008 Component Cooling Water X A2.02 on those predictions, use procedures to correct. control, 3.2 9 or mitigate the consequences of those malfunctions or operations: Highllow surge tank level 01 0 Pressurizer Pressure Control X I I I I K5.02 I Knowledge of the operational implications of the following concepts as the apply to the PZR PCS:

Constant enthalDv exDansion throuah a valve 1 1 2.6 10

-t X Knowledge of RPS design feature@) andlor interlock(s) 01 2 Reactor Protection K4.09 11 Separation of control and protection circuits Knowledge of ESFAS design feature(s) and/or 013 Engineered Safety Features X K4.13 interlock(s) which provide for the following MFW Actuation 12 isolationlreset Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based 022 Containment Cooling X A2.06 on those predictions, use proceduresto correct, control, 2.8 13 or mitigate the consequences of those malfunctions or operations: Loss of CCS Pump 022 Containment Cooling Al.02 Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Containment pressure 3.67 14 026 Containment Spray A3.01 Ability to monitor automatic operation of the CSS, including: Pump starts and correct MOV positioning 1 4.3 I 15 Knowledge of the physical connections andlor cause-026 Containment Spray X K1.01 effect relationships between the CSS and the following 4.2 16 systems: ECCS

~~~~

Ability to (a) predict the impacts of the following mal-functions or operations on the MRSS; and (b) based on 039 Main and Reheat Steam A2.01 predictions. use procedures to correct, control, or 3.1 17 mitigate the consequences of those malfunctions or operations: Flow paths of steam during a LOCA Conduct of Operations: Ability to perform specific 059 Main Feedwater X 2.1.23 system and integrated plant procedures during all 3.9 18 modes of plant operation.

Ability to (a) predict the impacts of the following matfunctions or operations on the AFW; and (b) based IIH 061 AuxillarylEmergency Feedwater A2.01 on those predictions, use procedures to correct, control, 2.5 19 or mitigate the consequences of those malfunctions or operations: Startup of MFW pump during AFW operation NUREG-1021 7

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 Group 1 I I I I Ability to predict andlor monitor changes in parameters 061 AuxillarylEmergency Feedwater I I I I -

(to prevent exceeding design limlts) associated with operating the A W controls including: S/G level Knowledge of ac distribution system design feature(s) 3.9 062 AC Electrical Distribution X and/or interlock(s)which provide for the following: 2.8 Interlocks between automatic bus transfer and breakers Ability to monitor automatic operation of the dc electrical 063 DC Electrical Distribution system, including: Meters, annunciators, dials, 2.7 recorders, and indicating lights 063 DC Electrical Distribution 064 Emergency Diesel Generator I l l /

Ability to predict andlor monitor changes in parameters associated with operating the dc electrical system controls including: Battery capacity as it is affected by Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers sl

- ~~

Knowledge of the effect that a loss or malfunction of the 073 Process Radiation Monitoring PRM system will have on the following: Radioactive 3.6 effluent releases.

Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based 3

076 Service Water 078 Instrument Air I IxI I I I I I on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: LOSSof SWS Knowledge of the physical connections andlor cause-effect relationshipsbetween the IAS and the following systems: Cooling water to compressor 3.5 2.6 1

I I I I 1 Ability to manually operate and/or monitor in the control 103 Containment I I I I I I I I room: Phase A and Dhase B resets 3,5 WA Category Point Totals: 113 2 1 2 5 'oint Total:

NUREG-1021 8

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 2 Group 2 001 Control Rod Drive control or mitigate the consequences of those malfunctions or operations Quadrant power tilt Ability to (a) predict the impacts of the following mal-functions or operations on the GS and (b) based on 035 Steam Generator X A2 02 those predictions use procedures to correct control or 44 91 mitigate the consequences of those malfunctions or operations Reactor trip/turbine trip Emergency Procedures/Plan Knowledge of symptom 056 Condensate X 93 based EOP mitigation strategies Knowledge of bus power supplies to the following 001 Control Rod Drive X K2 o2 36 29 One-line diagram of power supply to trip breakers Knowledge of the operational implications of the 011 Pressurmer Level Control X K5 09 following concepts as they apply to the PZR LCS 26 32 Reason for manually controlling PZR level Knowledge of the effect of a loss or malfunction on the 015 Nuclear Instrumentation X K6 01 following will have on the NIS Sensors, detectors, and 29 30 indicators Emergency Procedures 1Plan Knowledge of 016 Non-nuclear Instrumentation x 2 4 31 annunciators alarms and indications, and use of the 33 31 response instructions A4 o1 Ability to manually operate and/or monltor in the control 33 034 Fuel Handling Equipment room Radiation levels Knowledge of the physical connections andlor cause-035 Steam Generator X K1 14 effect relationships between the S1GS and the following 39 34 systems ESF 041 Steam Dumvurbine Bypass Knowledge of SDS design feature(s) andlor interlock(s)

X K4 O1 which provide for the following RRGKS system 29 35 Control Knowledge of the operational implications of the following concepts as the apply to the MTB System 36 045 Main Turbine Generator X K5 o1 Possible presence of explosive mixture in generator if hydrogen purity deteriorates Knowledge of the effect that a loss or malfunction of 055 Condenser Air Removal X K3 o1 37 the CARS will have on the following Main condenser Ability to monitor automatic operation of the Waste Gas 071 Waste Gas Disposal X A3 03 Disposal System including Radiation monitoring 36 38 system alarm and actuating signals K/A Category Point Totals 1/1 1 1 1 1 2 1 0 012 1 1 Group Point Total 1013 NUREG-1021 9

Facility: Three Mile Island Date of Exam: 5/9/2005 RO SRO-Only Category KIA # Topic IR Q# J-rG-

  • Ability to supervise and assume a management role during plant transients and upset conditions Ability to recognize indications for system operating ntry-level conditions for 1.

Conduct of Operations Equipment Control 30 1 99 Radiation Control I

4.

Emergency Procedures I Plan Subtotal 3 Tier 3 Point Total 10 NUREG-1021 10

2/ 1 003 / K4.11 The subject K/A isn't relevant at the subject facility.

2/ 1 004 / K2.07 The subject K/A isn't relevant at the subject facility.

2/ 1 I 006 / K4.02 I Double jeopardy with audit exam item 2/ 1 007 / A4.01 The subject K/A isn't relevant at the subject facility.

2/1 026 / A3.02 The subject K/A isn't relevant at the subject facility.

2/ 1 059 / 2.1.30 Topic better suited for In-Plant JPM 2/ 1 061 / A l . 0 3 The subject K/A isn't relevant at the subject facility.

2/ 1 073 / A2.01 Double Jeopardy with audit exam question 2/ 1 I 078 / K1.03 I The subject K/A isn't relevant at the subject facility.

2/ 1 I 001 K 2 . 0 2 1 Double jeopardy with question 29.

211 086 12.4.6 The subject WA isn't relevant at the subject facility.

2/2 028 / K1.01 The subject K/A isn't relevant at the subject facility.

313 G3 12.3.6 Double jeopardy with audit exam JPM It isn't possible to prepare a psychometrically sound question related to the 3I 4 G4 12.4.21 sub.ect WA, It isn't possible to prepare a psychometrically sound question related to the 111 02' 2.4.4 subject WA.

1/ 1 027 I AA2.13 The subject K/A isn't relevant at the subject facility.

1/ 1 025 / AK2.01 Double jeopardy with question 5 It isn't possible to prepare a psychometrically sound question related to the 1 12 028 I AA2.06 sub.ect WA, It isn't possible to prepare a psychometrically sound question related to the 1 12 059 I AA2.03 sub.ect WA, It isn't possible to prepare a psychometrically sound question related to the 1/ 2 A07 / 2.1.23 sub.ect wA.

It isn't possible to prepare a psychometrically sound question related to the 1/ 2 103 12.4.49 subiect wA, 2 I2 071 I A2.03 System over sampled It isn't possible to prepare a psychometrically sound question related to the 211 073 K5'03 subject WA.

2/ 1 022 I A2.01 Question matching WA could not be written for the RO level.

311 G I 2.1.14 NRC request (over-sampling of WA)

NUREG-1021 11

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: 05/09/2 005 Examination Level SRO Operating Test Number: NRC Administrative Topic Type Describe activity to be performed (see Note) Code*

JPM: Verify and approve an Estimated Critical Boron Conduct of Operations M Calculation IAW 1103-158, estimated Critical Conditions.

2.1.25 Ability to obtain and interpret station reference materials such as graphs, nomographs, and tables which contain perfonnance data (3.1)

JPM: Verify the Equipment Status Tag Log supports Conduct of Operations N startup IAW OP-AA-108-108, Unit Restart Review, Section 5,Step 8.

2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation (4.0)

JPM: Evaluate a completed surveillance procedure and Equipment Control perform appropriate actions IAW AP 1041, IST Program Requirements, Section 4.4.3.

2.2.12 Knowledge of surveillance procedures (3.4)

JPM: Authorize emergency personnel radiation exposure in Radiation Control excess of 5 REM TED IAW EP-AA-113, Personnel Protective Actions, Section 4.3.3.

2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized (3.1)

JPM: Given a set of conditions, upgrade an EAL and make a Protective Action RecommendationIAW the facility Emergency Plan.

2.4.44 Knowledge of Emergency Plan Protective Action Recommendations (4.0)

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 8 Criteria: (C)ontrol room (D)irect from bank (s 3 for ROs; 5 for SROs & RO retakes)

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

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

(S)imulator NUREG-1021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: 05/09/2005 Exam Level (circle one): SRO(I) Operating Test No.: NRC System / JPM Title Type Code' 1 Safety Function Transfer the Group 6 control rods from the Auxiliary Power 1 Supply to the Normal Power Supply in accordance with (IAW)

OP-TM-622451, Transferring Rods to Auxiliary Power Supply.

001 G2.1.23 (3.9 / 4.0)

Initiate HPVLPI IAW OP-TM-EOP-010, Emergency Rules Guides 2 and Graphs, Guide 2 - HPllLPl initiation.

006 A4.07 (4.4 / 4.4)

Place the Decay Heat Removal (DHR) System in operation and 4P establish a cooldown in accordance with OP-TM-212-111, Shifting DHR Train A from DHR Standby to DHR Operating Mode.

005 A4.01 (3.6 / 3.4)

Take corrective action for a failure of the Main Generator to trip IAW OP-TM-EOP-001, Reactor Trip.

045 A3.04 (3.4 / 3.6)

Initiate RB Emergency Cooling with RR-P-1A and the "A"Cooler IAW 1104-38, Reactor Building Emergency Cooling Water System.

022 A4.04 (3.1 / 3.2)

~

Energize 1C 4160V Bus using the SBO Diesel IAW 6 OP-TM 864-901, Loss of Station Power.

APE 056 AA2.37 (3.7 / 3.3)

~ ~~ ~ ~

--I----

Reset RPS Channel "A" IAW OP-TM-641-421, Tripping and Resetting RPS Channels.

012 A4.04 (3.3 I3.3)

N/A (SRO-I Candidates)

NUREG-I 021, Revision 9

ES-301 Control Roomh-Plant Systems Outline Form ES-301-2 In-Plant Systems* (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

i. I Respond to a loss of "A" DC Distribution IAW 1202-9A, Loss of "A"DC.

I P*E I 6 System 063 KIA 2.1.30 (3.9 13.4)

APE 058 058 AA2.03 (3.5 13.9)

. Terminate an approved waste gas release IAW 1102-27, Waste Disposal - Gaseous, due to an unexplained pressure drop in a N, R 9 tank not being released.

System 071 KIA 2.1.30 (3.9 13.4)

APE 060 060 AA2.06 (3.6 13.8) k' Establish 8'hstage heating following a reactor trip IAW 4s OP-TM-421-102, Placing 8'" Stage Feedwater String A In Service.

System 039 KIA 2.1.30 (3.9 13.4) 039 Al.05 (3.2 13.3)

I @ 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. I

  • Type Codes Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank 91 ai 4 (E)mergency or abnormal in-plant 11 I/ 1 (L)ow-Power 11 11 1 (N)ew or (M)odified from bank including 1(A) 21 21 1 (P)revious2 exams L 3i t 3 I :2 (randomly selected)

(R)CA ,I/ 11 1 (S)imulator NUREG-1021, Revision 9

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Three Mile Island Date of Exam: 5/9/2005 Operating Test No.: NRC Scenarios 3 4 T M 0 1 CREW POSITION I CREW POSITION A B S A 6 I R T 0 R T 0 0 C P 0 C P

' RX 5 NOR I 4 I i 2 2 1 3 2 I/ RX 4 NOR 5 I/ SROI-4 , I/C 1.3, NUREG 1021 Revision 9

ES-301 Transient and Event Checklist Form ES-301-5 Facility: BVPS-2 Date of Exam: 2/28/2005 Operating Test No.: NRC A Scenarios P

P E -

L N 1 2 3 4 T M I T 0 I C I T N A T CREW POSITION CREW POSITION CREW POSITION CREW POSITION . A I N Y L M T P U M

-1' 1'

4' SROI-5 2

SROI-6 2

1' I _

1' 4'

SROI-7 2

Instructions 1 Circle the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type, TS are not applicable for RO applicants ROs must service in both the "at-the-controls (ATC)" and "balance-of-plant (BOP)"

positions, Instant SROs must do one scenario, including at least two instrument or component (VC) malfunctions and one major transient, in the ATC position 2 Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D 5 d) but must be significant per Section C 2 a of Appendix D

  • Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis 3 Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicant's competence count toward the minimum requirement.

Author.

NRC Reviewer:

NUREG 1021 Revision 9

I Appendix D Scenario Outline Form ES-D-1 I Facility: Three Mile Island Scenario No.: 1 OpTestNo.: NRC Examiners: Operators:

Initial Conditions: 80% power, MOC.

0 AH-E-1A is tagged out of service.

R Turnover:

Continue 80% power operation.

Critical Tasks: Cross Tie LPI Loops and balance LPI flows Trip RCPs within 1 minute following loss of subcooling Malf Event Event No. Type* Description RCOGA I CRS Pressurizer temperature instrument failure (high).

I URO CHOl B TS CRS AH-E-1 B RB Air Handling Unit Trip. (TS)

MU19D C CRS RCP Seal #1 failure.

C URO RDlOC N CRS Manual reactor power reduction due to ICS-CRDS interface problem.

R, URO RF DH32 TS CRS BWST Low Level Alarm Condition. (TS)

TH03B C CRS 50 gpm RCS leak (TS) requires plant shutdown.

C URO TH04A M CRS Large Break LOCA.

M URO DHOlA C CRS LPI Pump failure.

C URO

)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

[ Appendix D Scenario Outline Form ES-D-1 I THREE MILE ISLAND MAY 2005 NRC EXAM SIMULATOR SCENARIO 1 GENERAL DESCRIPTION The crew will take the watch with reactor power at 8O%, with ICs in Full Automatic.

On cue from the Lead Evaluator, initiate the Pressurizer temperature instrument failure. The crew should respond to alarm G-2-4, PZR LEVEL HVLO, and 1202-29, Pressurizer System Failure, transferring Pressurizer level control to manual, selecting valid instrumentation for automatic control, and then returning level control to automatic. An EQUIPMENT DEFICIENCY TAG should be applied to the pressurizer temperature selector pushbutton, but logging and documentation of configuration control procedures is NOT required to continue with the scenario.

After the failed Pressurizer temperature instrument is declared inoperable, and an Equipment Deficiency Tag is applied, the Lead Evaluator can cue the RB Cooling Fan failure. The crew should respond to alarms B-2-5,480V ES MOTOR OVERLOAD, and B-1-5, 480V ES Motor TRIP. Since there is no standby air handling unit to start, the SRO should evaluate for Technical Specification compliance (7-day TS time clock). An EQUIPMENT DEFICIENCY TAG should be applied to the AH-E-1 B control switch, but logging and documentation of configuration control procedures is NOT required to continue with the scenario.

After the SRO has declared the 7-day TS time clock,the Lead Examiner can cue initiation of the RCP Seal failure.

Seal leakoff rate will be set to exceed 1203-16 entry conditions (> 6 GPM), but negate the (> 8 GPM) requirement to trip the reactor if power is reduced to less than the required limit within 5 minutes. Based on these conditions, 1203-16 will require the crew to reduce power to less than 75%, shutdown the affected pump using OP-TM-226-154, SHUTDOWN RC-P-1D. and isolate seal #l leak-off. During the power reduction the CRD system will fail to respond to automatic insert signals. Operators should take MANUAL control of the Diamond Rod Control Panel and additional ICs controls to effect the power reduction in accordance with requirements of OP-TM-621-471, ICs MANUAL CONTROL.

On cue from the Lead Evaluator following the power reduction and RCP seal isolation, the Borated Water Storage Tank (BWST) level will be reduced to activate MAP E-34, BWST TEMP/LVL HI/LO. The crew should confirm the validity of the low level alarm using redundant control room and plant computer indications, and perform the actions described in the alarm response procedure for MAP E-3-4. The SRO should determine that conditions do not satisfy T.S. 3.3.1 .l.a. If BWST level is not restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, T.S. 3.0.1 requires initiation of plant shutdown and the plant must be placed in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If BWST is still not restored, the plant must be placed in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Makeup to the BWST is not required to proceed with the scenario.

Shortly after BWST level reduction, the Lead Evaluator can cue initiation of the 50 gpm RCS leak. The SRO is required to recognize the (TS 3.1.6.1) 1 gpm unidentified and the (TS 3.1.6.2) 10 gpm total RCS leakage limits are exceeded, and announce requirements to place the reactor in hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The SRO should announce the more restrictive time limits to initiate plant shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and reach HOT STANDBY within the time limits of T.S. 3.0.1.

After the leakage tech spec evaluation, the Lead Examiner can cue initiation of the large break LOCA with delayed trip of one LPI Pump. The crew should enter OP-TM-EOP-001, REACTOR TRIP, perform the immediate manual actions and recognize loss of RCS subcooled margin. Actions of RULE 1 should be completed, followed by transitions to OP-TM-EOP-002, LOSS OF SUBCOOLED MARGIN, and then to OP-TM-EOP-006, LOCA COOLDOWN. Among other actions the team will perform the following high level activities: trip all operating RCPs, verify HPI actuation and EFW actuations, raise OTSG levels to 7585%, cross-tie Train A and Train B LPI systems and balance LPI flows between the two systems in response to trip of DH-P-1A after the ES actuation.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

Appendix D NUREG 1021 Revision 9

Appendix D Scenario Outline Form ES-D-1 I Facility: Three Mile Island Scenario No.: 2 OpTest No.: NRC Examiners: Operators:

Initial Conditions: 0 100% power, EOC.

0 AH-E-1A tagged 00s.

0 DC-P-1A is operating in accordance with OP-TM-543-401to support chemistry sampling.

Turnover:

Maintain 100% power operations I

Critical Tasks: Manually actuate ES Train B Throttlenerminate HPI flow to the RCS OTSG - 1A Isolation

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Malf. No. Event Type* Event Description RCR42 I CRS Pressurizer Spray Valve Failure.

RCR43 I URO MU01B C CRS Makeup Pump Trip, Loss of RCS Makeup and Seal Injection.

C URO CC02A C CRS Decay Heat Closed Cooling Pump Failure. (TS)

C URO RC08B I CRS RCS Cold Leg RTD Failure.

I URO 110 N CRS Manual Power Reduction Due to Heat Balance Calculation Error. (TS)

Override R URO TCOl M CRS Main Turbine Trip.

M URO MSMA C CRS Stuck Open Main Steam Safety Valves with Excessive FW Flow MSMB C URO IC26 I/o Override ESOlB I CRS ES Train B Actuation Failure.

I URO 1 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

I Appendix D Scenario Outline Form ES-D-1 I THREE MILE ISLAND MAY 2005 NRC EXAM SIMULATOR SCENARIO #2 GENERAL DESCRIPTION The crew will take the watch with reactor power at 100% and ICs in Full Automatic.

On cue from the Lead Evaluator, the Pressurizer spray valve failure will be activated. The crew should respond to alarm G-3-8, RCPRESS NARROW RNG HI/LO or indications in accordance with 1202-29, PRESSURIZER SYSTEM FAILURE. Spray valve control should be shifted Manual, and the valve closed. From this point forward, OP-TM-220-503, MANUAL CONTROL OF PRESSURIZER PRESSURE, will be applicable. An EQUIPMENT DEFICIENCY TAG should be applied to the pressurizer spray valve Autohlanual control selector, but logging and documentation of configuration control procedures is NOT required to continue with the scenario..

When ready, the Lead Evaluator can cue initiation of the Makeup Pump failure. The crew should respond to alarms 8-2-2,4KV ES MOTOROVERLOAD, and 8-1-2,4KV ES MOTOR TRIP, and implement 1203-15, LOSS OF R.C.

MAKEUPISEAL INJECTION. Following completion of 1203-15 immediate manual actions, MU-P-1A will be started with cooling water established either from DCCS or the NSCC system. An EQUIPMENT DEFICIENCY TAG should be applied to the failed Makeup Pump control switches, but logging and documentation of configuration control procedures is NOT required to continue with the scenario.

When ready, the Lead Examiner can cue initiation of DC-P-1A trip. The crew should respond to alarms B-2-5,480V ES MOTOR OVERLOAD, and B-1-5,480V ES MOTOR TRIP. If not performed with startup of MU-P-1A previously, OP-TM-543-439, SWAPPING MU-P-1A COOLING TO NS, will provide guidance to re-establish cooling to MU-P-1A.

The SRO should review TS 3.3 to determine that a 72-hour TS time clock has been started, with possible plant shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. An EQUIPMENT DEFICIENCY TAG should be applied to the DC-P-1A control switch, but logging and documentation of configuration control procedures is NOT required to continue with the scenario.

After the SRO has declared the TS for the ECCS, the Lead Evaluator can cue the RCS cold leg RTD failure. The crew should respond to alarm H-3-2, SASS MISMATCH, in accordance with OP-TM-MAP- H0302, to select an alternate instrument. An EQUIPMENT DEFICIENCY TAG should be applied to the affected Tcold selector pushbutton, but logging and documentation of configuration control procedures is NOT required to continue with the scenario.

When cued by the Lead Examiner the SOS will call the control room to direct the CRS to reduce reactor power by 5%

due to discovery of an error in the Plant Computer heat balance calculation. The SRO will apply T.S. 1.1 and direct a power reduction in accordance with 1102-4, POWER OPERATION. During the power reduction the operators will be required to take MANUAL control of the ICs to effect the power reduction, using OP-TM-621-471, ICs MANUAL CONTROL, and OP-TM-622-471, MANUAL CONTROL ROD OPERATIONS.

Shortly after the reduction to 95% power, the Lead Evaluator can cue the Main Turbine trip, which will result in an automatic reactor trip. OTSG 1A safety valves will stick open and NV-P-1A will not respond in automatic to cause RCS overcooling conditions. The crew should respond in accordance with OP-TM-EOP-001, REACTOR TRIP, followed by OP-TM-EOP-003, EXCESSIVE PRIMARY-TO-SECONDARY HEAT TRANSFER. Once the crew enters OP-TM-EOP-003, among other actions they will perform the following high level activities: isolation of OTSG 1A feed and steam pathways, manual actuation of HPI Train B (simulator malfunction blocks automatic actuation), HPI throttling, and actions required to prevent RCS reheat and repressurizationfollowing isolation of the OTSG.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

Appendix D NUREG 1021 Revision 9

I Appendix D Scenario Outline Form ES-D-1 I 1 Facility: Three Mile Island Scenario No.: 3 OpTestNo.: NRC Examiners: Operators:

Initial Conditions: 0 68% power, EOC.

0 3 RCPs operating.

0 AH-E-1A tagged 00s.

0 ICs in Manual control due to control svstem.

Turnover:

Maintain 68% power until FW-P-16 is repaired.

Critical Tasks: Initiate Manual Reactor Trip Reduce RCS Subcooled Margin L

Event No.

Malf.

No.

Event (Type' I ICRS I Event Description I Pressurizer Level Controller Failure.

1 I/O Override 1 , URO I 2 RDOl C CRS Dropped Control Rod. (TS)

C URO 3 IC23 N CRS Manual reactor power reduction.

R URO 4 DHOGA TS CRS Core Flood Tank Nitrogen Leak. (TS) 5 W15A MCRS Loss of Main Feedwater.

M URO IlCCRS I RD28 C CRS RPS Auto Trip Failure (ATWS).

C URO TH15A MCRS OTSG Tube Leak.

M URO Emergency Feedwater Pump Failure.

Override I'O (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

I Appendix D Scenario Outline Form ES-D-1 I THREE MILE ISLAND MAY 2005 NRC EXAM SIMULATOR SCENARIO #3 GENERAL DESCRIPTION The crew will take the watch with reactor power at 68% power while FW-P-16 repairs are completed. ICs is in manual due to control system instability with three RCPs operating.

When ready, the Lead Evaluator can cue initiation of the Pressurizer level controller malfunction. The crew will respond to MAP G-2-5, PZR LEVEL HI/LO, or indication of zero RCS makeup flow with lowering Pressurizer level.

The team can implement any of the three following procedures to respond to the malfunction by establishing manual control of makeup valve MU-V-17: 1202-29, PRESSURIZER SYSTEM FAILURE, OP-TM-211-472, MANUAL PRESSURIZER LEVEL CONTROL, or OP-TM-EOP-010, Guide 9, RCS INVENTORY CONTROL. Diagnosis of the malfunction is not required to proceed with the scenario if manual control of MU-V-17 is established to control Pressurizer level. If diagnosed, An EQUIPMENT DEFICIENCY TAG should be applied to the MU-V-17 controller, but logging and documentation of configuration control procedures is NOT required to continue with the scenario.

When ready to proceed, the Lead Evaluator can cue initiation of the dropped control rod. The crew should respond to alarm G-2-1, CRD PAlTERN ASYMMETRIC, in accordance with OP-TM-MAP-GO201 and/or indications of dropped rods as described in 1202-8, CRD EQUIPMENT FAILURE, symptoms. The team should determine that 1202-8 immediate manual actions require reactor power to be reduced to less than 45% (60% of the thermal power allowable for 3-RCP operating conditions) within 2-hours. The SRO should determine that TS 3.5.2.2.e is applicable, requiring power to be reduced to less than 45% within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and the RPS high flux trip setpoints to be reduced less than or equal to 52.3%. The SRO may decide to attempt to recover the dropped rod prior to reducing reactor power, or may opt to reduce power first - either approach is permitted by procedures and Tech Specs; either option satisfies reactivity manipulation requirements of ES-301. If the SRO elects to recover the rod before reducing power, the Lead Examiner should cue a telephone call from the Shifl Operations Superintendentwho will direct the team to reduce power prior to attempting to recover the dropped rod. If the Diamond rod control station is in manual, the URO is required to depress the LATCH pushbutton in order to affect control rod in-motion due to the dropped rod actuating the Group 7 Group In-Limit rod in-motion block interlock.

On cue from the Lead Evaluator, the Core Flood Tank nitrogen leak will begin to reduce CF-T-1A pressure. The crew should respond to alarm D-2-7, CF TANK 1A LEVEUPRESS HIILO. The SRO should review TS 3.3.1.2.a to determine tank pressure is lower than the Tech Spec LCO limit, and TS 3.0.1 is applicable. Tank pressure is not required to be restored to normal to proceed with the scenario.

After the SRO has reviewed the TS for Core Flood Tank pressure, the Lead Evaluator can w e the Main Feedwater Pump failure coupled with initiation of an OTSG tube leak. The crew should recognize the reactor should have tripped (ATWS) and initiate a manual reactor trip in accordance with OP-TM-EOP-001, REACTOR TRIP. Following completion of the immediate actions, the team should diagnose the existence of the OTSG Tube leak, and the SRO should make the transition to OP-TM-EOP-005. OTSG TUBE LEAKAGE.

Coincident with the Feedwater Pump trip Emergency Feedwater Pump EF-P-26 will fail to automatically start. The crew should enter OP-TM-EOP-005,OTSG TUBE LEAKAGE, and, among other actions, perform the following high level activities: Manually start EF-P-26, reduce RCS pressure to minimize RCS subcooled margin, and begin RCS Cooldown to Cold Shutdown.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

Appendix D NUREG 1021 Revision 9

ES-401 PWR Examination Outline Form ES-401-2

-acility: Three Mile Island Date of Exam: 5/9/2005 I I RO WA Category Points I SRO-Only Points Abnormal Plant Evolutions Totals

3. Generic Knowledge and 2 3 4 1 2 3 4 10 7 Abilities Categories 2 2 3 2 2 1 2 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 (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each WA category shall not be less than two).
2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by k1 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. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding elimination of inappropriate WA statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only Dortions. resoectivelv.
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, the topics importance ratings (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the WA Catalog, and enter the WA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to WAS that are linked to 10CFR55.43 NUREG-1021 1

PWR Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 1 Conduct of Operations Knowledge of system status 025 ILoss of RHR System I 4 X 2.1.14 3.3 76 criteria which require the notification of plant personnel Ability to determine and interpret the following as they 027 IPressurizer Pressure Control System Malfunction I X AA2 05 apply to the Pressurizer Pressure Control Malfunctions 33 77 3 PZR heater setpoints Conduct of Operations Knowledge of system status 040 ISteam Line Rupture - Excessive Heat Transfer I 4 X 2.1.14 criteria which require the notification of plant personnel 3.3 78 Ability to determine and interpret the following as they 057 ILoss of Vital AC Inst. Bus I 6 X AA2.12 apply to the Loss of Vdal AC Instrument Bus PZR level 3.7 79 controller instrumentation,and heater indications Ability to operate and Ior monitor the following as they apply to the (Vital System Status Verification) Facildy E02 IReactor Trip - Stabilization - Recovery I1 X EA2.1 conditions and selection of appropriate procedures during 4.0 80 abnormal and emergency operations Conduct of Operations Ability to explain and apply all E05 ISteam Line Rupture - Excessive Heat Transfer 1 4 2.132 system limits and precautions 3.8 81 Abiltty to determine or interpret the following as they apply to a reactor trip: If reactor should have tripped but has not 007 IReactor Trip - Stabilization - Recovery I1 X EA2.04 done so, manually trip the reactor and carry out actions in 4.4 39 ATWS EOP Abihty to determine and interpret the following as they 008 IPressurizer Vapor Space Accident I3 X AA2.04 apply to the Pressurizer Vapor Space Accident: High- 3.2 40 temoerature comDuter alam and alarm tme

~~~~

Abiltty to determine or interpret the following as they apply 009 ISmall Break LOCA I3 X EA2.10 3.1 41 to a small break LOCA: Airborne activQ Abiltty to operate and monitor the following as they apply 01 1 ILarge Break LOCA I 3 X EAl.10 4.1 42 to a Large Break LOCA: AFW and SWS pumps Ability to operate and Ior monitor the following as they 015 I17 IRCP Malfunctions I 4 X AAl.11 apply to the Reactor Coolant Pump Malfunctions (Loss of 2.5 43 RC Flow): RCP onloff and run indicators Ability to operate and Ior monitor the following as they 022 ILoss of Rx Coolant Makeup I 2 X AA1.01 apply to the Loss of Reador Coolant Makeup: CVCS 3.4 44 letdown and charging Knowledge of the interrelations between the Loss of 025 ILoss of RHR System I 4 X AK2.03 Residual Heat Removal System and the following: Service 2.7 45 water or closed cooling water pumps Ability to determine and interpret the following as they 026 ILoss of Component Cooling Water I 8 X AA2.01 apply to the Loss of Component Cooling Water: Location 2.9 46 of a leak in the CCWS NUREG-I021 2

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 I E / A P E # / Name Safety Function I G I K1 I K2 I K3 I A I I A2 I Number I WA Topic(s) t029lATWS11 038 ISteam Gen. Tube Rupture I 3 040 ISteam Line Rupture - Excessive Heat Transfer I 4 X

X EK1.01 EAl.09 2.1.2 Knowledge of the operational implications of the following concepts as they apply to the ATWS: Reactor nucleonics and thermo-hydraulics behavior Ability to operate and monitor the following as they apply to a SGTR: PZR tank levellpressureindicators, gauges, and recorder.

Conduct of Operations: Knowledge of operator responsibilitiesduring all modes of plant operation.

Ability to operate and Ior monitor the following as they 2.8 3.2 3.0 47 48 49

~

054 ILoss of Main Feedwater I 4 X AA1.01 apply to the Loss of Main Feedwater (MFW): AFW 50 controls, including the use of alternate AFW sources Ability to determine or interpret the following as they apply 055 IStation Blackout I 6 X EA2.02 to a Station Blackout: RCS core cooling through natural 51 I

circulation cooling to SIG cooling Ability to determine and interpret the following as they apply to the Loss of Vial AC Instrument Bus: The 057 ILoss of Vial AC Inst. Bus I 6 I X AA2.18 indicator, valve, breaker, or damper position which will 3.1 52 occur on a loss of power

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Abillty to operate and Ior monitor the following as they 058 ILoss of DC Power 16 X apply to the Loss of DC Power: Static inverter dc input AA1.02 breaker, frequency meter, ac output breaker, and ground fault detector 065 I Loss of InstrumentAir I 8 7 X AK3.08 Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Actions contained in EOP for loss of instrument air Ability to operate and Ior monitor the following as they apply to the (Vial System Status Verification): Adherence E02 IReactor Trip - Stabilization - Recovery I1 X EA2.2 3.2 55 to appropriate procedures and operation within the limitations in the facility's license and amendments.

E04 IInadequate Heat Transfer - Loss of Secondary Ability to operate and Ior monitor the following as they X EAl.2 apply to the (Inadequate Heat Transfer) Operating 3.4 56 Heat Sink I 4 behavior characteristicsof the facility.

KIA Category Point Totals: 112 1 1 7 714 Group Point Total: 1816 NUREG-1021 3

I ES-401 Three Mile Island Form ES-401-2 1 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions 028 IPressurizer Level Malfunction I 2 X AA2 ,o Whether the automatic mode for PZR level control is functionina imDroDerlv. necessitv of shift to manual modes 051 ILoss of Condenser Vacuum I 4 X

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

1 Ability to determine and interpret the following as they 059 I Accidental Liquid RadWaste Rel. I 9 apply to the Accidental Liquid Radwaste Release The 3.5 a4 valve lineup for a release of radioactive liquid Equipment Control Knowledge of bases in technical E08 ILOCA Cooldown - Depress I 4 X specificationsfor limiting conditions for operations and safety limits t: t Knowledge of the interrelations between the Dropped q

003 I Dropped Control Rod I 1 AK2.05 Control Rod and the following: Control rod drive power supplies and logic circuits Ability to operate and I or monitor the following as they 028 I Pressurizer Level Malfunction I 2 AA1.06 apply to the Pressurizer Level Control Malfunctions:

Checking of RCS leaks ttt Knowledge of the reasons for the following responses as 060 IAccidental Gaseous RadWaste Rel. I 9 AK3.02 they apply to the Accidental Gaseous Radwaste: Isolation of the auxiliary building ventilation Ability to operate and I or monitor the following as they 076 I High Reactor Coolant Activlty I 9 AA1.04 apply to the High Reactor Coolant Activlty: Failed fuel-monitoring equipment Knowledge of the reasons for the following responses as they apply to the (Loss of NNI-X) Manipulation of controls A02 I LOSSO f NNI-WY I 7 X AK3'3 3.7 required to obtain desired operating results during abnormal and emergency situations.

036 I Fuel Handling Incidents I I I I I AK2.01 I Knowledge of the interrelationsbetween the Fuel Handling Incidents and the following: Fuel Handling Equipment ttt Conduct of Operations: Abillty to perform specific system A06 I Shutdown Outside Control Room I 8 X 2.1.23 and integrated plant procedures during all modes of plant operation

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Ability to determine and interpret the following as they apply to the (Inadequate Subcooling Margin) Adherence to E03 I hadequate Subcooling Margin I 4 EA2.2 3.5 appropriate procedures and operation within the limitations in the facilitv's license and amendments.

I NUREG-1021 4

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Emergency and Abnormal Plant Evolutions Tier 1 Group 2 Knowledge of the operational implications of the following E13 1 Rules and Enclosures concepts as they apply to the (EOP Rules) Normal, X EK1,2 3.0 65 abnormal and emergency operating procedures associated with lEOP Rules).

WA Category Point Total: 112 1 2 2 2 112 Group PointTotal: 914 NUREG-1021 5

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Plant Systems - Tier 2 Group 1 I System #/Name I G I K1 I K2 I K3 I K4 I K5 I K 6 I A1 1 A2 I A3 I A4 I Number I KIA Topics Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS. and (b) based 004 Chemical and Volume A2.02 on those predictions. use procedures to correct, control 4.2 86 Control or mitigate the consequences of those malfunctions or operations Loss of PZR level (failure mode) 006 ECCS Ix 2 4 30 Emergency ProcedureslPlan Knowledge of which events related to system operation/status should be reported to outside agencies I 3.6 87 013 ESFAS Ix 2449 I Emergency ProcedureslPlan Ability to reference to procedures those actions that require immediate ODeration of svstem comDonents and controls 90 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system, and (b) based on those predictions, use procedures to correct, 064 Emergency Diesel Generator A2.04 3.0 88 control. or mitigate the consequences of those malfunctions or operations Unloading prior to securing an ED/G Equipment Control Knowledge of limiting conddions for 076 Service Water X 2.2.22 4.1 89 operations and safety limits Knowledge of RCPS design feature(s) andlor 003 Reactor Coolant Pump K4.04 interlock@)which provide for the following: Adequate coolina of RCP motor and seals Knowledge of the effectthat a loss or malfunction of the 003 Reactor Coolant Pump K3.04 3.9 RCPS will have on the following: RPS Knowledge of the effect of a loss or malfunctionon the 004 Chemical and Volume K6.14 following CVCS components: Recirculation path for 2.7 Control charging pumps 004 Chemical and Volume X K2.05 I Knowledge of bus power supplies to the following:

MOVs I I2.7 4 Knowledge of the effect of a loss or malfunctionon the 005 Residual Heat Removal K6.03 2,5 following will have on the RHRS: RHR heat exchanger

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Knowledge of ECCS design feature@)andlor interlock(s) which provide for the following: Automatic 006 Emergency Core Cooling K4.20 3.2 6 closure of common drain line and fill valves to accumulator 1007 Pressurizer RelieflQuench K3.01 Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: Containment 3.3 7 NUREG- 1021 6

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Plant Systems - Tier 2 Group 1 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based 008 Component Cooling Water A2.02 on those predictions, use procedures to correct, control, 3.2 9 or mitigate the consequences of those malfunctions or operations: Highllow surge tank level rl Knowledge of the operational implications of the 010 Pressurizer Pressure Control 2.6 10 012 Reactor Protection 3.6 8 012 Reactor Protection I K4.09 I Knowledge of RPS design feature(@ andlor interlock(s)

Separation of control and protection circuits I 2.8 I 11 013 Engineered Safety Features Actuation K4.13 Knowledge of ESFAS design feature(s) andlor interlo&(s) which provide for the following MFW isolationlreset 3.7 1 12 Abiltty to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based 022 Containment Cooling A2.01 on those predictions, use procedures to correct, control, 2.5 13 or mitigate the consequences of those malfunctions or operations: Fan motor over-current

~~

Abiltty to predict andlor monitor changes in parameters (to prevent exceeding design limits) associated with 022 Containment Cooling A1.02 3.6 14 operating the CCS controls including: Containment pressure Abiltty to monitor automatic operation of the CSS, 026 Containment Spray A3.01 4.3 15 including: Pump starts and correct MOV positioning Knowledge of the physical connections andlor cause-026 Containment Spray K1.01 effect relationshipsbetween the CSS and the following systems: ECCS Ability to (a) predict the impacts of the following mal-functions or operations on the MRSS; and (b) based on 039 Main and Reheat Steam A2.01 predictions, use proceduresto correct. control, or mitigate the consequences of those malfunctions or owrations: Flow paths of steam durina a LOCA 059 Main Feedwater 2.1.23 Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

I 1 3.9 18 NUREG-102 1 7

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Plant Systems - Tier 2 Group 1 I System #/Name WA Topics I Imp. I Q# I Ability to (a) predict the impacts of the following malfunctions or operations on the A M I ; and (b) based 061 Auxillary/Emergency on those predictions, use procedures to correct, control, 2.5 19 Feedwater or mitigate the consequences of those malfunctions or operations: Startup of MFW pump during AFW operation Ability to predict and/or monitor changes in parameters 061 Auxillary/Emergency (to prevent exceeding design limits) associated with 3.9 20 Feedwater operating the AFW controls induding: SIG level Knowledge of ac distribution system design feature(s) 062 AC Electrical Distribution I X and/or interlock(s)which provide for the following:

Interlocks between automatic bus transfer and breakers 2.8 21 Ability to monitor automatic operation of the dc electrical 063 DC Electrical Distribution system, including: Meters, annunciators, dials, 2.7 22 recorders, and indicating lights

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Ability to predict and/or monitor changes in parameters associated with operating the dc electrical system 063 DC Electrical Distribution 2.5 23 controls including: Battery capaclty as it is affected by discharge rate Knowledge of the effect of a loss or malfunction of the 064 Emergency Diesel Generator 2.7 24 following will have on the EDlG system: Air receivers

~~ ~

Knowledge of the operational implications as they apply to concepts as they apply to the PRM system:

073 Process Radiation Monitoring X 2.9 25 Relationship between radiation intensity and exposure limits Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based 076 Service Water on those predictions, use procedures to correct, control, 3.5 26 or mitigate the consequences of those malfunctions or I

078 Instrument Air X operations: Loss of SWS Knowledge of the physical connections andlor cause-effect relationshipsbetween the IAS and the following systems: Cooling water to compressor Ability to manually operate and/or monitor in the control 2.6 27 103 Containment 3,5 28 room: Phase A and phase B resets WA Category Point Totals: 2 1 2 5 2 I 2815 NUREG-1021

ES-401 Three Mile Island Form ES-401-2 PWR Written Examination Outline Plant Systems Tier 2 Group 2 K1 WA TODiCS Ability to (a) predict the impacts of the following v

E- - Q#

malfunction or operations on the CRDS- and (b) based 001 Control Rod Drive on those predictions, use procedures to correct. 4.2 92 control, or mitigate the consequences of those malfunctions or operations Quadrant power tilt Ability to (a) predict the impacts of the following mal-functions or operations on the GS; and (b) based on 035 Steam Generator those predictions, use procedures to correct, control. or 4.4 91 mitigate the consequences of those malfunctions or operations: Reactor tripiturbine trip Emergency ProceduresIPlan: Knowledge of symptom 4.0 056 Condensate 93 based EOP mitigation strategies Knowledge of bus power supplies to the following:

001 Control Rod Drive X 3.6 29 One-line diagram of power supply to trip breakers Knowledge of the operational implications of the 01 1 Pressurizer Level Control following concepts as they apply to the PZR LCS 2.6 32 Reason for manuallv controllina PZR level Knowledge of the effect of a loss or matfunction on the 015 Nuclear Instrumentation following will have on the NIS: Sensors, detectors, and 2.9 30 indicators Emergency Procedures / Plan Knowledge of 016 Non-nuclear Instrumentation annunciators alarms and indications, and use of the 3.3 31 response instructions.

Abiltty to manually operate and/or monitor in the control 034 Fuel Handling Equipment 3.3 33 room: Radiation levels Knowledge of the physical connections and/or cause-035 Steam Generator X effect relationshipsbetween the S/GS and the following 3.9 34 svstems: ESF 041 Steam Dumpnurbine Bypass Knowledge of SDS design feature@)and/or interlock(s)

, which provide for the following: RRGllCS system 2.9 35 Control -

Knowledge of the operational implications of the following concepts as the apply to the MTlB System:

045 Main Turbine Generator 2.8 36 Possible presence of explosive mixture in generator if hydrogen purity deteriorates Knowledge of the effect that a loss or malfunction of 055 Condenser Air Removal 2.5 37 the CARS will have on the following: Main condenser Abiltty to monitor automatic operation of the Waste Gas 071 Waste Gas Disposal Disposal System including: Radiation monitoring 3.6 38 system alarm and actuating signals K/A Category Point Totals: 1 1 ItTotal: 1 1013 NUREG-I 021 9

I 1

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Facility: Three Mile Island D a m  ! 5/9/2005 Category K/A #

2'1'6 Topic Ability to supervise and assume a management role

+I I I I SRO-Only 4.3 I

94 during plant transients and upset conditions.

Knowledge of system status criteria which require the 2,1,14 33 95 notification of plant personnel.

2.1.29 Knowledge of how to conduct and venfy valve lineups. 3.4 66 1.

Conduct of Operations Ability to coordinate personnel activities outside the 2.1.8 3.8 67 control room.

Ability to locate control room switches, controls and 2.1.31 indications and to determine that they are correctly 4.2 68 reflecting the desired plant lineup.

Subtotal 3 2 2.2,21 Knowledge of pre- and post-maintenanceoperability 3.5 96 requirements.

2.

2'2'5 I Knowledge of the process for making changes in the facility as described in the safety analysis report.

2.7 97 Equipment Control 2.2.12 Knowledge of surveillance procedures. 3.0 69 Knowledge of bases in technical specifications for 2,2,25 2.5 70 limiting conditions for operations and safety limits.

Subtotal 2 2 Knowledge of 10 CFR: 20 and related facility radiation 3.0 99 2.3.1 control requirements

3. 2.3.11 Ability to control radiation releases. 2.7 71 Radiation Control Knowledge of the process f ility to recognize abnormal indications for system 4.

Emergency Procedures I Plan conditions.

Knowledge of EOP entry conditions and immediate 4.3 75 2.4.1 action steps.

Subtotal 3 2 Tier 3 Point Total 10 7 NUREG-102 1 10

Tier / Randomly Reason for Rejection Group Selected WA It isn't possible to prepare a psychometrically sound question related to the 2/ 1 059 / 2.4.30 subiect K/A.

2/ 1 003 / K4.11 The subject WA isn't relevant at the subject facility.

2/ 1 004 / K2.07 The subject WA isn't relevant at the subject facility.

2/1 006 / K4.02 Double jeopardy with audit exam item 1/ 1 027 / AA2.13 The subject KIA isn't relevant at the subject facility.

1/ 1 025 / AK2.01 Double jeopardy with question 5 I "

I I IC, L

I U L O I, M L . V-, O I It isn't possible to prepare a psychometrically sound question related to the

,mn A . .

subject KIA.

It isn't possible to prepare a psychometrically sound question related to the 112 059 I AA2.03 sub.ect wA, t It isn't possible to prepare a psychometrically sound question related to the subject WA.

It isn't possible to prepare a psychometrically sound question related to the subject KIA.

2 12 071 I A2.03 System over sampled 2/1 007 A4.10 Concept over sampled (2 other items with same concept # l o , 40)

NUREG-1021 11

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