ML112620102

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Final Outlines (Folder 3)
ML112620102
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 08/25/2011
From:
Exelon Nuclear
To: Caruso J
Operations Branch I
JACKSON D RGN-I/DRS/OB/610-337-5306
Shared Package
ML110030671 List:
References
TAC U01829
Download: ML112620102 (26)
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Text

ES-402 Written Examination Outline Form ES-401-2 Facility: TM I Date of Exam: 08/29/11 RO KiA Category Points SRO-Only Points Tier K K A A A A G Total A2 G* Total 6 2 3 4

  • 1.

1 3 3 3 18 3 3 6 Emergency 1 2 9 2 2 4 Plant Evaluations 4 5 27 5 5 10 2 3 28 2 3 5 2.

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

3. Generic Knowledge & Abilities 1 2 3 4 10 1 2 3 4 7 2 2 3 3 2 2 1 2 Note 1. Ensure that at least two topics from every 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 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 section D.1.b of ES-401, 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 KIA categories.

7: The generic (G) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIA's

8. On the following pages, enter the KIA 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. If fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the KIA Catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KlAs that are linked to 10CFR55.43

ES-402 Form ES-401-2 TMI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function K1 I K21 K31 A1 A2 G KIA Topic(s) Imp. IQ# I I

AA2.01 Ability to determ ine and 077 / Generator Voltage and E:lectric I interpret the following as they apply to X Generator Voltage and Electric Grid 3.6 79 Grid Disturbances Disturbances: Operating point on the I qenerator capability curve EA2.1 - Ability to determine and i

interpret the following as they apply to

  • E02 / Reactor Trip - Stabilization - the (Vital System Status Verification):

Recovery i 1 X Facility conditions and selection of 4.0 84 appropriate procedures during abnormal and emergency operations i EA2.05 Ability to determine or interpret the fotJolNing as they apply to a' 055 i Station Blackout / 6 X 3.7 80 I Station Blackout: When battery is approaching fully discharged i 2.1.20 Conduct of Operations: Ability 0381 Steam Generator Tube Rupture X to interpret and execute procedure 4.6 93 I 13 steps.

2.1.25 - Conduct of Operations: Ability 015 117 I Reactor Coolant l:lump X to interpret reference materials. such as 4.2 91 Malfunctions I 4 graphs, curves, tables, etc.

2.4.45 - Emergency Procedures I Plan:

Ability to prioritize and interpret the 0561 Loss of Off-site Power i 6 X 4.3 81 significance of each annunciator or alarm.

EK1.3 - Knowledge of the operational implications of the following concepts as they apply to the (Post-Trip E 10 I Reaclor Trip - Stabilization Recovery 11 X Stabilization) Annunciators and 4.0 8 conditions indicating signals, and remedial actions associated with the (Post -Trip Stabilization).

AK1.01 - Knowledge of the operational implications of the following concepts 008 I Pressurizer Vapor Space as they apply to a Pressurizer Vapor X 3.2 65 Accident I 3 Space Accident: Thermodynamics and flow characteristics of open or leaking valves AK1.01 - Knowledge of the operational implications of the following concepts as they apply to Loss of Main 0541 Loss of Main Feedwater I 4 X 4.1 11 I Feedwater (MFW): MFW line break depressurizes the SIG (similar to a i

steam line break)

AK2.03 - Knowledge of the interrelations between the Loss of 025 I Loss of Residual Heat Removal X Residual Heat Removal System and 2.7 52

,System 14 the follolNing: Service water or closed i cooling water pumps EK2.2 - Knowledge of the interrelations between the (Excessive Heat Transfer) and the following: Facility's heat E05 I Steam Line Rupture Excessive removal systems, including primary X 4.2 4 Heat Transfer I 4 coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

AK2.05 - Knowledge of the 077 I Generator Voltage and Electric interrelations between Generator

  • 31 X 40 Grid Disturbances Voltage and Electric Grid Disturbances i .

and the following: Pumps

ES-402 Form ES-401-2 TMI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function G KIA Topic(s) Imp. I I Q#

AK3.02 Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water 062 I Loss of Nuclear Service. Water I X The automatic actions (alignments) 3.6 53 4

within the nuclear service water resulting from the actuation of the ESFAS EK3.03 Knowledge of the reasons for 0291 Anticipated Transient Without the following responses as they apply Scram (ATWS) 11 X to the ATWS: Opening BIT inlet and 3.7 46 outlet valves AK3.02 - Knowledge of the reasons for the following responses as they apply 058 1 Loss of DC Power 1 6 X 4.0 21 to the Loss of DC Power: Actions contained in EOP for loss of dc power EA1.05 - Ability to operate and monitor 009 1 Small Break LOCA I 3 X the following as they apply to a small 3.4 25 break LOCA: CCWS EA1.02 - Ability to operate and monitor 055 I Station Blackout 1 6 X the following as they apply to a Station 4.3 17 Blackout: Manual ED/G start AA1.05 - Ability to operate and 1 or 057 I Loss of Vital AC Electrical monitor the following as they apply to X 3.2 60 Instrument Bus 16 the Loss of Vital AC Instrument Bus:

Backup instrument indications AA2.13 - Ability to determine and 027 I Pressurizer Pressure Control interpret the following as they apply to System Malfunction I 3 X 2.8 42 the Pressurizer Pressure Control Malfunctions: Seal return flow EA2.1 Ability to determine and interpret the following as they apply to E04 I Inadequate Heat Transfer - the (Inadequate Heat Transfer) Facility X 3.2 74 Loss of Secondary Heat Sink I 4 conditions and selection of appropriate procedures during abnormal and emergency operations.

EA2.01 Ability to determine or interpret the following as they apply to a 011 I Large Break LOCA I 3 X Large Break LOCA: Actions to be 4.2 67 taken, based on RCS temperature and pressure - saturated and superheated 2.4.31 - Emergency Procedures 1 Plan:

065 I Loss of Instrument Air 18 X Knowledge of annunciator alarms, 4.2 37 indications, or response procedures.

2.1.28 - Conduct of Operations:

038 I Steam Generator Tube Rupture Knowledge of the purpose and function 13 X 4.1 9 of major system components and controls.

2.2.44 - Equipment Control: Ability to interpret control room indications to verify the status and operation of a 056 I Loss of Off-site Power I 6 X 3.7 47 system, and understand how operator actions and directives affect plant and system conditions.

KIA CategoryTotals 3 3 3 3 3/3 3/3 Group Point Total: 18/6 I

ES-402 Form ES-401-2 TMI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 L -_ _ E_A_P_E_#~/N_a_m_e__S_a_~_ty~Fu_n_c_ti_o_n__~__~~~~~~~____~__~______~~~_A__T_OP_i_C(_S_)_________ 1 Imp. I Q# I I AA2.01 - Ability to determine and I interpret the follovving as they apply to 005 Ilnoperable/Stuck Control Rod / the Inoperable f Stuck Control Rod:

X 1 4 .1 90 1 Stuck or inoperable rod from in-core and ex-core NIS, in-core or loop

~.

temperature measurements AA2.1 - Ability to determine and interpret the follovving as they apply to 1

.AOS f Fuel Handling AccidEnt 18 X the (Refueling Canal Level Decrease)

Facility conditions and seection of 4.0 i 78 appropriate procedures during i

I abnormal and emergency operations.

2.1.32 - Conduct of Operations: Ability 037 I Steam Generator Tube Leak! 3 X to explain and apply all system limits 4.0 92 and precautions.

2.4.18 - Emergency Procedures I Plan:

! A06! Control Room Evac. I 8 X Knowledge of the specific bases for 4.0 98 EOPs.

EK1.2 - Knowledge of the operational I implications of the following concepts I as they apply to the (Natural E09 I Natural Circ. 14 X Circulation Cool down) Normal, 3.7 63 abnormal and emergency operating procedures associated with (Natural Circulation Cooldown).

AK2.1 Knovviedge of the interrelations between the (Shutdown Outside I Control Room) and the following:

Components, and functions of control A061 Control Room Evac. 18 X 3.8 44 and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and I manual features.

" AK3.02 - Knowledge of the reasons for the following responses as they apply  !

067 I Plant Fire On-site I 8 X to the Plant Fire on Site Steps called 2.5 35 out in the site fire protection plan, FPS manual, and fire zone manual 1 1 AA1.22 - Ability to operate and I or I monitor the following as they apply to 0241 Emergency Boration 11 X the Emergency Boration: Safety 3.2 injection valves, switches, flow meters, 45 I I and indicators I EA2.1 - Ability to determine and  !

i I E14! EOP Rules and Enclosures 15 X interpret the follovving as they apply to the (EOP Enclosures): Facility 3.4 69 conditions and selection of appropriated procedures during abnormal and emergency operations.

2.4.8 - Emergency Procedures I Plan:

Knowledge of how abnormal operating A04 I Turbine Trip I 4 X 3.8 12 procedures are used in conjunction with EOP*s.

2.2.38 - Equipment Control: Knowledge

  • 059 / Accidental Liquid RadWaste X of conditions and limitations in the 3.6 30 Release 19 facility license.

AA1.01 - Ability to operate and I or monitor the following as they apply to 032 I Loss of Source Range Nuclear X the Loss of Source Range Nuclear 3.1 61 Instrumentation / 7 Instrumentation: Manual restoration of power

ES-402 Form ES-401-2 TMI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function KIA Topic(s)

AK1.3 Knowledge of the operational implications of the following concepts as they apply to the (Loss of NNI-Y)

A031 Loss NNI-x/Y 17 X Annunciators and conditions indicating 3.0 75 signals, and remedial actions associated with the (NNI-Y).

KIA Category Totals 2 1 1 2 1/2 2/2 Group Point Total: 9/4 I

ES-402 Form ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s) Imp. I Q# I A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and {b} based on those 026 Containment Spray predictions. use procedures to X 3.0 99 correct, control, or mitigate the consequences of those malfunctions or operations: Reflux boiling pressure spike when first going on recirculation A2.08 - Ability to (a) predict the impacts of the following ma.lfunctions or operations on the AFW; and (b) based on those predictions. use procedures to 061 Auxiliary/Emergency X correct, control, or mitigate the 2.9 100 Feedwater consequences of those malfunctions or operations: Flow rates expected from various combinations of AFW pump discharoe valves 2.2.37 - Equipment Control:. Ability 006 Emergency Core to determine operability and lor Cooling X 4.6 89 availability of safety related equipment i 2.1.32 - Conduct of Operations:

064 Emergency Diesei Generator X Ab.lity to explain and apply all 2.8 82 limits and precautions.

2.2.36 - Equipment Control: Ability to analyze the effect of 003 Reactor Coolant Pump maintenance activities. such as X 4.2 83 degraded power sources. on the status of limiting conditions for operations.

K 1.13 - Knowledge of the physical

  • connections and/or cause-effect 003 Reactor Coolant Pump X relationships between the RCPS 2.5 41 and the following systems: RCP bearing lift oil Dump K1 .04 - Knowledge of the physical 061 Auxiliary/Emergency connections and/or cause-effect X 3.9 2 Feedwater relationships between the AFW and the following systems: RCS K2.01 - Knowledge of bus power 005 Residual Heat Removal X supplies to the following: RHR 3.0 51 Dumps

! 004 Chemical and Volume K2.01 - Knowledge of bus power X supplies to the following: Boric 2.9 23 Control acid makeup pumps K3.01 Knowledge of the effect 007 Pressurizer that a loss or malfunction of the X 3.3 64 Relief/Quench Tank PRTS will have on the following:

Containment K3.01 - Knowledge of the effect 026 Containment Spray that a loss or malfunction of the X 3.9 24 CSS will have on the following:

CCS K4.06 Knowledge of RPS design feature(s) and/or interlock(s) 012 Reactor Protection X which provide for the following: 3.2 56 Automatic or manual enable/disable of RPS trips

ES-402 Form ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 1 System #lName KJA Topic(s) Imp. I Q# I K4.04 - Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the 039 Main and Reheat following: Utilization of steam X 2.9 14 Steam pressure program control when steam dumping through atmospheric relief/dump valves, including T-ave. limits K5.02 - Knowledge of the operational implications of the 013 Engineered Safety X following concepts as they apply 2.9 26 Features Actuation to the ESFAS: Safety system logic and reliability K4.20 - Knowledge of ECCS design feature(s) and/or 006 Emergency Core interlock(s) which provide for the X 3.2 49 Cooling following: Automatic closure of common drain line and fill valves to accumulator.

K6.07 - Knowledge of the effect of 064 Emergency Diesel a loss or malfunction of the 2.7 20 Generator following will have on the ED/G system: Air receivers K6.03 - Knowledge of the effect of 010 Pressurizer Pressure a loss or malfunction of the X 3.2 43 Control following will have on the PZR pcs: PZR sprays and heaters A 1.02 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) 076 Service Water X associated with operating the 2.6 13 SWS controls including: Reactor and turbine building closed cooling water temperatures.

A 1.07 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) 059 Main Feedwater X associated with operating the 2.5 6 MFW controls including: Feed Pump speed, including normal control speed for ICS AZ.15 - Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) 062 AC Electrical based on those predictions, use X 2.8 18 Distribution procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Consequence of paralleling out-of phase/mismatch in volts A2.08 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to 008 Component Cooling X correct, control, or mitigate the 2.5 59 Water consequences of those malfunctions or operations:

Effects of shutting (automatically or otherwise) the isolation valves of the letdown cooler

ES-402 Form ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s)

A3.01 Ability to monitor automatic operation of the dc 063 DC Electrical Distribution X electrical system, including: 2.7 22 Meters, annunciators, dials, recorders, and indicating lights A3.01 Ability to monitor 078 Instrument Air X automatic operation of the lAS, 3.1 38 including: Air pressure A4.02 Ability to manually operate 073 Process Radiation and/or monitor in the control room X 3.7 28 Monitoring Radiation monitoring system control panel A4.04 Ability to manually operate 103 Containment X andlor monitor in the control room: 3.5 54 Phase A and phase B resets 2.4.2 Emergency Procedures /

Plan: Knowledge of system set 022 Containment Cooling X points, interlocks and automatic 4.5 72 actions associated with EOP entry conditions.

2.4.47 - Emergency Procedures /

Plan: Ability to diagnose and L

062 AC Electrical recognize trends in an accurate Distribution X and timely manner utilizing the 4.2 68 appropriate control room reference material.

K5.26 - Knowledge of the operational implications of the 004 Chemical and Volume following concepts as they apply X 3.1 48 Control to the CVCS: Relationship between VCT pressure and NPSH K3.03 Knowledge of the effect that a loss or malfunction of the 059 Main Feedwater X 3.5 5 MFW will have on the following:

S/GS A3.03 - Ability to monitor 003 Reactor Coolant Pump X automatic operation of the RCPS, 3.2 39 including: Seal DIP A 1.03 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) 005 Residual Heat Removal X associated with operating the 2.5 50 RHRS controls including: Closed cooling water flow rate and temperature 2.4.46 - Emergency Procedures /

039 Main and Reheat Plan: Ability to verify that the X 4.2 10 Steam alarms are consistent with the plant conditions.

K4.01 - Knowledge of de electrical system design feature(s) and/or 063 DC Electrical

'; ~

X interlock(s) which provide for the Distribution following: Manual/automatic transfers of control KIA Category Totals 2 2 3 4 2 2 3 2/2 2, 2 3/3 Group Point Total:

ES-401 Form ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name KIA Topic{s) Imp. I Q# I I  ! A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System' and (b) based on those predictions.

071 Waste Gas Disposal X use procedures to correct. control. 3.6 86 or mitigate the consequences of tilose malfunctions or operations:

Use of waste gas release monitors. radiation. gas flow rate.

and totalizer 2.2.12 Equipment Control:

086 Fire Protection X Knowledge of surveillance 3.4 94 procedures.

A2.03 - Ability to (a) predict the impacts of the following mal functions or operations on the Containment Purge System: and (b) based on those predictions.

029 Containment Purge X 3.1 77 use procedures to correct, control.

or mitigate the consequences of those malfunctions or operations:

Startup operations and the associated required valve lineups K5.01 - Knowledge of the operational implications of the following concepts as they apply 014 Rod Position Indication X 2.7 58 to the RPIS: Reasons for differences between RPIS and step counter A 1.05 - Ability to predict and/or monitor changes in parameters (to!

prevent exceeding design limits) associated with operating the 045 Main Turbine Generator X 3.8 15 MT/G system controls including:

Expected response of primary plant parameters (temperature and pressure) followinQ T/G trip A3.01 - Ability to monitor 072 Area Radiation automatic operation of the ARM

! X 2.9 29 Monitoring system. including: Changes in ventilation alignment K3.01 - Knowledge of the effect that a loss or malfunction of the

! 017 In-core Temperature X ITM system will have on the 3.5 66 I Monitor following: Natural circulation indications I I K1.01 - Knowledge of the physical 027 Containment Iodine connections and/or cause-effect 3.4 70 Removal XI relationships between the CIRS I and the following systems: CSS 1 I 2.4.8 - Emergency Procedures I 011 Pressurizer Level Plan: Knowledge of how abnormal X 3.8 71 Control operating procedures are used in I I conjunction with EOP*s.

A4.05 - Ability to manually operate 035 Steam Generator Ix and/or monitor in the control room:

Level Control to enhance natural 3.8 3 J i circulation K4.03 - Knowledge of design feature(s) and/or interlock(s) 029 Containment Purge X I which provide for the following:

Automatic purge isolation 3.2 27 I I

ES-401 Form ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name KIA Topic(s) Imp. I Q# I K2.02 - Knowledge of bus power supplies to the following: One-line 001 Control Rod Drive X 3.6 57 diagram of power supply to trip breakers A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system: and (b) based on those predictions, use 075 Circulating Water X procedures to correct, control, or 2.5 16 mitigate the consequences of those malfunctions or operations:

Safety features and relationship I

between condenser vacuum, turbine trip, and steam dump KIA Category Totals J1 1 1 1 1 0 1 1/2 1 1 1/1 Group Point Total:

I 10/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: TMI Date:

Category KA# Topic RO SRO-Only IR Q# IR Q#

Ability to locate control room switches, controls, and 2.1.31 indications, and to determine that they correctly reflect 4.6 19 the desired plant lineup.

Ability to use procedures to determine the effects on 2.1.43 reactivity of plant changes, such as RCS temperature, 4.1 7 secondary plant, fuel depletion, etc.

1. Cond uct of Operations 2,1.35 Knowledge of the fuel-handling responsibilities of 39 76 SRO's.

Ability to use procedures related to shift staffing. such 2,1.5 as minimum crew complement, overtime limitations, 39 96 etc.

Subtotal 2 2 Ability to manipulate the console controls as required 2.2.2 to operate the facility between shutdown and 4.6 62 designated power levels.

Knowledge of the process for controlling equipment 2.2.14 3.9 31 configuration or status.

2. Equipment Knowledge of the process for making design or Control 2.2.5 operating changes to the facility.

3.2 95 Ability to track Technical Specification limiting 2.2.23 4,6 85 conditions for operations.

Subtotal 2 2 Ability to comply with radiation work permit 2.3.7 3.5 32 requirements during normal or abnormal conditions.

Knowiedge of Radiological Safety Principles pertaining to licensed operator duties. such as containment entry 2.3.12 3.2 34 requirements. fuel handling responsibilities. access to locked high-radiation areas, aligning filters, etc.

Knowledge of radiation exposure limits under normal 2.3.4 3.2 33 or emergency conditions.

3. Radiation Control 2.3.6 Ability to approve release permits. 3.8 87 Subtotal 3 1

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Knowledge of RO tasks performed outside the main I 2.4.34 control room during an emergency and the resultant 4.2 73 operational effects.

Knowledge of emergency communications systems 2.4.43 3.2 36 and techniques.

2.4.17 Knowledge of EOP terms and definitions. 3.9 1

4. Emergency Knowledge of the organization of tile operating Procedures I 2.4.5 procedures network for normal. abnormal, and 4.3 97 emerqencyevolutions.

Plan Knowledge of low power! shutdown implications in 2.4.0 accident (e.g., loss of coolant accident or loss of 4.2 88 residual heat removal) mitigation strategies.

I Subtotal 2 iTier 3 Point Total: Wo* 7

E8-401 Record of Rejected KIA's Form E8-401-4 Randomly Selected I Tier 1 Group Reason for Rejection KA 039 / K4.02 replaced The Facility does NOT have Tave control(MS and reheat is not 2/1 by 039 1 K4.04 an input to Tave control) 028 / K2.01 replaced The Hydrogen Recombiners have been re-classified as having 2/2 by 001 I K2.01 no safetx: related im~act at TMI.

028 I A2.03 replaced The Hydrogen Recombiners have been re-classified as having 2/2 by 029/ A2.03 no safety related im~act at TMI. I 001 1 K2.01 replaced 2/2 The subject KIA isn't relevant at the subject facility.

by 001 / K2.02 064 1 2.1 .19 replaced It isn't possible to prepare a psychometrically sound question !

2/1 by 064 I 2.1.32 related to the subject KIA.

071 1 A2.03 replaced It isn't possible to prepare a psychometrically sound question 2/2 by 071 I A2.02 related to the subiect KIA.

010 1 K6.04 replaced RCDT operations were oversampled on the RO Exam. Question 2/1 by 010 1 K6.03 5 and 40 also focus on RCDT operations.

0071 K5.02 replaced 2/1 KAs involving RCDT Operations have been over-sampled.

by 0041 K5.26 063 I K2.01 replaced NRC Examiner Feedback indicating that this KA would not yield 2/1 by 0631 K4.01 a discriminating question, and the appropriate level of difficulty.

034 I A3. 0 1 replaced NRC Examiner requires re-select based on Refueling activities 2/2 by 072 I A3.01 being performed by contracted personnel rather than ROs.

NRC Examiner Feedback indicating that this KA may not allow G1 I 2.1.45 replaced 3/1 the construction of a written exam question at the level required by G1 12.1.35 by 10CFR55.43(b)

NRC Examiner Feedback indicating that this KA may not allow G1 12.1.25 replaced 3/1 the construction of a written exam question at the level required by G1/2.1.5 by 10CFR55.43(b) 0621 AK3.04 1 11 replaced by 062 1 Effect of a Loss of CCW on NR Flow is non existent at TMI.

AK3.02 0561 2.4.46 replaced NRC Examiner Feedback indicated a potential to over sample 1 11 by 056 1 2.2.44 between writteln and operational evaluated; and replaced KIA.

G41 2.4.40 replaced NRC Examiner feedback indicated this may not yield a greater 3/4 by G4 1 2.4.43 than LOD1 guestion if asked at the RO level.

079/2.2.12 replaced NRC Examiner Feedback discussion agreed that Station Air at 2/2 by 086/2.2.12 TMI is very low safety significance re-picked system. I 006 I K5.01 replaced Randomly selected new K in same system after original attempt

  • 2/1 by 006 K4.20 to write to K5.01 rejected by NRC Examiner i

E141 EA2.2 replaced NRC determine tie to FSAR did not meet facility license and 1/2 by E14/ EA2.1 amendment portion of KIA, keep question changed KIA I

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: August 2011 Examination Level: RO ~ SRO D Operating Test Number: 289-2011-301 Administrative Topic Type Describe activity to be performed (See Note) Code*

Verify watch standing requirements - Work-hour Rules Conduct of Operations N/R 2.1.5 (2.9)

Calculate an Estimated Critical Rod Position.

Conduct of Operations N/R 2.1.37 (4.a)

Isolate a component for maintenance.

Equipment Control N/R 2.2.41 (3.~;)

Radiation Control N/A Category not selected for RO applicants Perform State and Local Event Notification Emergency Procedures/Plan D/S 2.4.43 (3.~~)

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, (S)imulator, or Class(R)oom (D)irect from bank (:s 3 for ROs; S 4 for SROs & RO retakes)

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

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

ES 301, Page 22 of 27

ES-301 Administrative Outline Form ES-301-1 THREE MILE ISLAND 2011 NRC RO EXAMINATION CONDUCT OF OPERATIONS (A1-1): Verify watch standing requirements - Work-hour rules.

Given a shift schedule for the cycle and reference LS-AA-11 g, Overtime Controls, identify which requested days the candidate can work while not violating Work-hour rules. This JPM is a new JPM.

License is evaluated against properly identifying the allowable and forbidden overtime shifts lAW LS-AA 119.

Safety significance, failure to select the correct shifts will lead to Work-hour rule violations and possible fatigue issues. Failure to properly staff a Control Room crew is also a violation of 10CFR.

CONDUCT OF OPERATIONS (A1-2): Calculate an Estimated Critical Position.

Given plant conditions and references OP-TM-300-000, "Reactivity and power distribution calculations" and OP-TM-300-403, "Estimated Critical Rod Positions" calculate an estimated critical rod condition.

This is a New J PM updated for the current core load and also using a new series of procedures. The old JPM used one procedure with all attached graphs, the new operations procedures uses a calculation procedure and a reference procedure additionally the Calculated point IS out side the tolerance band for the old calculation.

License is evaluated against a job done by reactor operators with measurements of using correct graphs and doing math properTy. Several opportunities for error exist in similar named graphs and applying curve data in the wrong direction.

Safety significance, failure to calculate an estimated critical rod position properly could lead to reactor being taken critical at an unexpected rod height.

EQUIPMENT CONTROL (A2): Isolate a component for maintenance.

This is a new JPM. The candidate must determine the isolation pOints for removal of a relief valve to repair a flange leak.

Safety significance This failure to adequately isolate the leak prior to removing the relief or opening vent and / or drain prior to isolating the relief would result in RCS water leaking into containment and a reduction in shielding water above the core.

RADIATION CONTROL (A3): Category not selected for RO Candidates.

EMERGENCY PROCEDURES/PLAN (A4): ERO Notification with a faulted sheet. This JPM is from the LORT bank. RO Candidate will be given a State and Local Notification Form with Errors that need to be identified. The Candidate will then Simulate initiating a State and Local notification on the NARS line.

This is a time critical JPM.

Failure to make timely and accurate notification to State and Local emergency management agencies could lead to a degradation of safety of public if agencies can not plan and execute tneir responsibilities in a timely fashion.

Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: August 2011 Examination Level: RO D SRO [gI Operating Test Number: 289-2011-301 Administrative Topic Type Describe activity to be performed (See Note) Code*

Maintain Minimum Shift Staffing, Control Overtime Conduct of Operations N/R 2.1.5 (3.9)

Review an Estimated Critical Rod Position Calculation for Conduct of Operations N/R approval, identify any errors.

2.1.37 (4.6)

Approve Isolation Points for a Component for Equipment Control N/R Maintenance 2.2.41 (3.9)

Authorize emergency exposure in excess of 5 REM.

Radiation Control N/R 2.3.4 (3.7)

Identify and Declare an Emergency Classification and Emergency Procedures/Plan N/R Associated PAR.

2.4.44 (4.4)

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, (S)imulator, or Class(R)oom (D)irect from bank (~ 3 for ROs; ~ 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (;:: 1)

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

ES 301, Page 22 of 27

ES-301 Administrative Topics Outline Form ES-301-1 THREE MILE ISLAND 2011 NRC SRO EXAMINATION CONDUCT OF OPERATIONS (A1-1): Verify watch standing requirements - Work-hour Rules.

Given. plant conditions and referef')ces O~-TM-101-111-1 001, Shift t'!1anning Requirements, LS-AA-119, Overtime Controls, a prepared Shift Staffing Report, LMS Qual Matrix Report, and a prepared Over time List, identify required actions to restore mimmum staffin'g, and select personnel in accordance with requirements to control overtime. This JPM is significantly modified in that it incorporates Work-Hour rules lAW LS-AA-119, changing the requirements of whether personnel can report to work or not.

License is evaluated against properly manning a shift lAW 10CFR and LS-AA-119. Several opportunities for error exist in the fact that they must review 5 different documents to determine if personnel can stand watch.

Safety significance, failure to select the correct personnel will lead to Work-hour rule violations and possible fatigue issues. Failure to properly staff a Control Room crew is also a violation of 10CFR.

CONDUCT OF OPERATIONS (A1-2): Review an Estimated Critical Rod Position Calculation for approval, identify any errors.

Given an Estimated Critical Rod Position previously calculated and references OP-TM-300-000, "Reactivity and power distribution calculations" and OP-TM-300-403, "Estimated Critical Rod Positions" review for SRO approval and identify errors if any.

This is a New JPM updated for the current core load and also using a new series of procedures. The old Bank JPM used one procedure with all attached graphs, the new operations procedures uses a calculation procedure and a reference procedure and additionally the numbers have changed to cause the new answer to be outside the range of the old answer.

License is evaluated against a job done by senior reactor operators with measurements of using correct graphs and doing matli properly. Several o~portunities for error exist in similar named graphs and applying curve data in the wrong direction. SRO should recognize that errors were made, identify errors and return for recalculation.

Safety significance, failure to calculate an estimated critical rod position properly could lead to reactor being taken critical at an unexpected rod height.

EQUIPMENT CONTROL (A2): Approve Isolation Points for a Component for Maintenance.

Given a Worker Tag Out Clearance Form, review for approval. A new ~IPM created for ILT 10-1. The submitted Tag Out Form omits required pOints and leavEls a group of breakers in the Off position on restoration.

Operations Authorization would come from an SRO. Safety significance is two valves do not have their remote controls indicated with info tags nor their breakers open this could result in operation that would drain the fuel handling canal if operated. The restoration points are also wrong and could result in not being able to appropriately respond to an excessive seal leak if the breakers were not closed.

RADIATION CONTROL (A3): Authorize emergency exposure in excess of 5 REM.

Conditions are given for a non-contaminated injured individual in a high radiation area, rescue personnel must go through a highly contaminated area. Six volunteers to carry stretcher, candidate must recognize one individual can not volunteer as they have already received a once in a life time exposure, a second volunteer is Inot respirator qualified.

Safety significance failure to properly remove these volunteers could result in excessive exposure.

EMERGENCY PROCEDURES/PLAN (A4): Identify and Declare an Emergency Classification Given a set of plant conditions declare an Emergency Classification. This new JPM deals with declaring an emergency classification (Time Critical) baseo on given plant conditions, and declaring the associated PAR also time critical.

License level is SRO as the Emergency Director is the position to declare an event.

Safety significance events must be recognized and declared in a timely fashion to get the required notifications and support. declaration of the PAR in a timely manor allows for the Slate to make appropriate recommendations for public protection.

ES-301 Control Room/In-Plant Outline Form ES-301-2 II Facility: Three Mile Island Auaust 2011 Exam Level: RO 0 SRO-I IZl SRO-U 0 Operating Test Number: 289-2011-301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / ~IPM Title Type Code*

Function

a. Respond to an Inoperable/Stuck Rod (005 AA1.01) D/S/L 1
b. Respond to a loss of Pressurizer Level Controller with Failures (011 N/S/A 2 A2.03)
c. Transfer to Reactor Building Sump Recirculation (011 EA1.05) D/S/LlAlEN 3
e. RCP #1 Seal Failure (003 A2.01) N/S/A 4P
f. Restore 1 D BUS from SBO Operations (064 A2.09) N/S 6
g. Respond lAW OP-TM-MAP-C01 01 Alarm Response with Failure (072 M/S/A 7 A3.01)
h. Cross Connect the Secondary River Water System to the Nuclear River MIS/AIL 8 Water System (026 AA2.02)

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

i. Manually Operate MU-V-20 and IC-V-4 (015/017 AA1.07) D/E/R 4P II' j . Local Start of EG-Y-1 B and Loading of 1E 4160V bus (068 AA1.10) N/E 8
k. Respond to a failure of EF-P-2A, and EF-V-30D (061 A2.04) M/E 4S I

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlaQ those tested in the control room.

  • Type Codes I Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C}ontrol room I

(D)irect from bank :5.9/ :5.8 I :5. 4

()E mer gency or abnormal in- plant >1/ >1 / > 1 (EN)gineered safety feature - / / 2: 1 (control room system (L)ow-Power / Shutdown 2: 1 / 2:1 / 2: 1 (N)ew or (M)odified from bank including 1 (A) 2: 2 / 2:2 / 2: 1 (P)revious 2 exams :5. 3/ :5.3 / :5. 2 (randomly selected)

(R)CA 2: 1 / 2: 1 I 2: 1 ES-301, Page 23 of 27

.--------------~~------------------------.--------

ES*301 Control Room/ln*Plant ....".~~4 ..... Outline C! Form ES*301*2 THREE MILE ISLAND 2011 NRC SROI EXAMINATION JPM A - Respond to an Inoperable/Stuck Rod. Bank JPM.

Safety significance failure to correct misalignment would result in operation out side of Technical Specification requirements designed to ensure core power limits.

JPM B - Respond to a Loss of Pressurizer Level Controller with Failures. New Alternate path.

Safety significance failure to complete JPM will result in a loss of RCS inventory.

JPM C - Transfer to Reactor Building Sump Recirculation, Bank Alternate path involving Engineered Safety Features equipment.

Safety significance failure to recognize DH-V-6A failure to open will result in damage to one of two available LPI trains during post accident conditions, seriously degrading safety margin. PRA SAHSR2 HSROA switch over to sump following LOCA.

JPM D - Not used for Instant SRO.

JPM E - RCP #1 Seal Failure. New Alternate path JPM.

Safety significance failure to properly address excessive sealleakoff could reSUlt in Seal LOCA.

JPM F - Restore 1D BUS from SBO Operations. New Involves restoring the "A" train ES Bus from Station Blaclkout Diesel power to normal Off-Site power.

Safety significance is improper operation could lead to loss of 1 train of ES components through loss of ES bus, causing significant degradation in coping capability.

JPM G - Respond lAW OP-TM-MAP-C01 01 Alarm Response with Failure. Modified Alternate path Similar to 2007 NRC JPM however this involves alternate path.

Safety significance Failure to place control tower on "recirculation" following high airborne contamination in the Control Room may result in unnecessary dose for the personnel that must remain to operate the plant.

JPM H Cross Connect the Secondary River Water System to the Nuclear River Water System. Modified JPM, significantly updated to current procedures, change from 2007 NRC is the alternate path.

Safety significance failure to complete the task would result in loss of adequate cooling for Nuclear Services components.

JPM I Manually Operate MU-V-20 and IC-V-4. Bank JPM In Plant Engineered Safety Features JPM.

Safety significance failure to complete the task could result in a loss of Reactor Coolant Pump seal cooling, and subsequent LOCA.

JPM J - Local Start of EG-Y-1 B and Loading of 1E 4160V bus. New JPM supplies power to the protected busses on Gontrol Room Evacuation. Safety significance Remote shutdown capability only exists if "B" tram power IS available.

JPM K - Respond to a failure of EF-P-2A, and EF-V-30D. Modified Bank JPM, old JPM involved steam bound pump, when restarted had to be done from switchgear. Switchgear start is no longer in procedure, added failed open discharge control valve on start.

st Safety significance failwe n

to complete 1 part of task could lead to inadequate secondary feed for loss of steam dnven pump, 2 part failure could lead to excessive heat transfer.

ES-301 Control Room/In-Plant ""I<C!TOrnQ Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: August 2011 Exam Level: RO D SRO-I D SRO-U [gI Operating Test Number: 289-2011-301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / ~IPM Title Type Code*

Function

a. Respond to an Inoperable/Stuck Rod (005 AA1.01) D/S/L 1 Sump Recirculation (011 EA1.05) I D/S/L/NEN 3
  • lAW OP-TM-MAP-C0101 Alarm ReSpOnSE! with Failure (072 M/S/A 7 A3.01) t Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. Manually Operate MU-V-20 and IC-V-4 (015/017 AP.1.07) D/E/R 4P
j. Local Start of EG-Y-1 B and Loading of 1 E 4160V bus (068 AA1.10) N/E 8

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO I SRO-I / SRO-U (A)lternate path 4-6/ 4-6 / 2-3 (C)ontrol room (D)irect from bank  :::.9/ :::.8 / :::. 4 (E)mergency or abnormal in-plant ~ 1 / ~1 / ~1 (EN)gineered safety feature - / - / ~ 1 (control room system (L)ow-Power / Shutdown ~1 / ~1 / ~1 (N)ew or (M)odified from bank including 1 (A) ~2/ ~2 / ~1 (P)revious 2 exams  :::. 3/ :::.3 / :::. 2 (randomly selected)

(R)CA ~ 1 I ~1 / ~ 1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room/In-Plant SV!~'AmA Outline Form ES-301-2 THREE MILE ISLAND 2011 NRC SROU EXAMINATION JPM A- Respond to an Inoperable/Stuck Rod. Bank JPM.

Safety significance failure to correct misalignment would result in operation out side of Technical Speclficalion requirements designed to ensure core power limits.

JPM B - Not used for SRO Upgrade.

JPM C - Transfer to Reactor Building Sump Recirculation, Bank Alternate path involving Engineered Safety Features equipment.

Safety significance failure to recognize DH-V-6A failure to open will result in damage to one of two available LPI trains during post accident conditions, seriously degrading safety margin. PRA SAHSR2 HSROA switch over to sump following LOCA JPM D - Not used for SRO Upgrade.

JPM E Not used for SRO Upgrade.

JPM F - Not used for SRO Upgrade.

JPM G Respond lAW OP-TM-MAP-C01 01 Alarm Response with Failure. Modified Alternate path Similar to 2007 NRC JPM however this involves alternate path.

Safety significance Failure to place control tower on "recirculation" following high airborne contamination in the Control Room may result in unnecessary dose for the personnel that must remain to operate the plant.

JPM H - Not used for SRO Upgrade.

JPM I Manually Operate MU-V-20 and IC-V-4. Bank JPM In Plant Engineered Safety Features JPM.

Safety significance failure to complete the task could result in a loss of Reactor Coolant Pump seal cooling, and subsequent LOCA.

JPM J - Local Start of EG-Y-1 B and Loading of 1E 4160V bus. New JPM supplies power to the protected bu~ses on Gontro! Room Evacuation. Safety significance Remote shutdown capability only exists if "8" train power IS available.

JPM K Not used for SRO Upgrade.

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

.------.----~--~~~~------.~-------------------

Facility: Three Mile Island Date of Examination: August 2011 Exam Level: RO r8] SRO-ID SRO-U 0 Operating Test Number: 289-2011-301 i

Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title I Type Code* Function i

n.

nC;:)jJullU Inoperable/Stuck Rod (005 AA1.01) D/S/L i 1

b. Respond to a loss of Pressurizer Level Controller with Failures (011 N/S/A 2 A2.03)
c. Transfer to Reactor Building Sump Recirculation (011 EA 1.05) D/S/LINEN 3 I

I

--itt- Rt:::>lult: TBVs and ADVs to ICS Auto Control (041 A4.06) D/S/L 4S

e. RCP #1 Seal Failure (003 A2.01) N/S/A i 4P
f. Restore 10 BUS from SBO Operations (064 A2.09) N/S I 6

I n

OP-TM-MAP-C0101 Alarm Response with Failure (072 M/S/A 7 t.~:cr1 )"U the S'-::vUIIU, River Water System to the Nuclear River M/S/NL 8 Water SY;::::I (026 AA2.02) I In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i

i. Manually Operate MU-V-20 and IC-V-4 (015/017 AA1.07) D/E/R 4P
j. Local Start of EG-Y -1 B and Loading of 1 E 4160V bus (068 AA 1.1 0) N/E 8
k. Respond to a failure of EF-P-2A, and EF-V-30D (061 A2.04) M/E 4S I

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must 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-II SRO-U (A)lternate path 4-61 4-6 / 2-3 (C)ontrol room (D)irect from bank ~91 ~8 I ~4 (E)mergency or abnormal in-plant ~ 1I ~1 / ~1 (EN)gineered safety feature - I - / ~ 1 (control room system (L)ow-Power I Shutdown ~1 / ~1 / ~1 (N)ew or (M)odified from bank including 1 (A) ~2/ ~2 / ~1 i

, (P)revious 2 exams ~3/ ~3 / ~ 2 (randomly selected)

II (R)CA ~ 1/ ~1 / ~1 (S)imulator ES-301, Page 23 of 27

---~~~~~--------~~------ .. -- .. -----~-------.--

ES-301 Control Room/In-Plant ~\I~;ttAlmCl Outline Form ES-301-2 THREE MILE ISLAND 2011 NRC RO EXAMINATION JPM A - Respond to an Inoperable/Stuck Rod. Bank JPM.

Safety significance failure to correct misalignment would result in operation out side of Technical Specification requirements designed to ensure core power limits.

JPM B - Respond to a Loss of Pressurizer Level Controller with Failures. !\lew Alternate path.

Safety significance failure to complete JPM will result in a loss of RCS inventory.

JPM C - Transfer to Reactor Building Sump Recirculation, Bank Alternate path involving Engineered Safety Features equipment.

Safety significance failure to recognize DH-V-6A failure to open will result in damage to one of two available-LPI trains during post accident conditions, seriously degrading safety margin. PRA SAHSR2 HSROA switch over to sump following LOCA.

JPM D - Restore TBVs and ADVs to ICS Auto Control. Bank JPM.

Safety significance failure to complete the task would result in unnecessary steaming to atmosphere resultmg In a loss of secondary inventory used for heat sink.

JPM E - RCP #1 Seal Failure. New Alternate path JPM.

Safety significance failure to properly address excessive seal leakoff could result in Seal LOCA.

JPM F - Restore 1D BUS from SBO Operations. New Involves restoring the "A" train ES Bus from Station Blackout Diesel power to normal Off-Site power.

Safety significance is improper operation could lead to loss of 1 train of ES components through loss of ES bus, causing significant degradation in coping capability.

JPM G - Respond lAW OP-TM-MAP-C0101 Alarm Response with Failure. Modified Alternate path Similar to 2007 NRC JPM however this involves alternate path.

Safety significance Failure to place control tower on "recirculation" following high airborne contamination in the Control Room may result in unnecessary dose for the personnel that must remain to operate the plant.

JPM H - Cross Connect the Secondary River Water System to the Nuclear River Water System. Modified JPM, Significantly updated to current procedures, change from 2007 NRC is the alternate path.

Safety significance failure to complete the task would result in loss of adequate cooling for Nuclear Services components.

JPM I - Manually Operate MU-V-20 and IC-V-4. Bank ,JPM In Plant Engineered Safety Features JPM.

Safety significance failure to complete the task could result in a loss of Reactor Coolant Pump seal cooling, and subsequent LOCA.

JPM J Local Start of EG-Y-1 B and Loading of 1E 4160V bus. New JPM supplies power to the protected busses on Control Room Evacuation. Safety significance Remote shutdown capability only exists if "B" train power is available.

JPM K - Respond to a failure of EF-P-2A, and EF-V-30D. Modified Bank JPM, old JPM involved steam bound pump, when restarted had to be done from switchgear. Switchgear start is no longer in procedure, added failed open discharge control valve on start.

Safety significance failwe n

to complete 1st part of task could lead to inadequate secondary feed for loss of steam driven pump, 2 part failure could read to excessive heat transfer.

D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 2 Op Test No.: 289-2011-301 Examiners: Operators:

Initial Conditions: * (Temporary IC-52)

  • 100% Power, MOL Turnover: Maintain 100% Power Operations I

I Critical Tasks:

  • Establish FW Flow and Feed SG(s) (CT-10)
  • Electrical Power Alignment (CT-8)
  • Protect against RCP Seal LOCA (CT-*)

Event No. Malt. No. Event Type* Event Description 1 MU01B CCRS MU-P-1 B Trips (TS), entry into OP-TM-AOP-041 CURO (URO: ensures MU-V-32 is in HAND and closed, ARO: Starts MU P-1A)

CARO TSCRS 2 ED18B CCRS Loss of the 8 Bus (TS) with EG-Y-1 B failing to start, entry into OP TM-AOP-014 CARO (ARO: Starts the SBO)

TSCRS I 3 NI27A ICRS Narrow Range Pressure Instrument Fails high with SASS failure to IC48 actuate, entry into OP-TM-MAP-G0308 IURO (URO: Closes Spray Valve, selects Alternate Pressure Instrument, IARO ARO: Manual control of Pressurizer Heaters) 4 I FW15B NCRS Loss of FW-P-1 B, entry into OP-TM-MAP-M01 07, Runback Fails to Occur, Power Reduction Performed, entry into OP-TM-MAP RURO H0101 and 1102-4 NARO 5 RD0230 TSCRS Stuck Rod (TS),. entry into OP-TM-AOP-062 6 ED18A MCRS Loss of OffsitaF'ower (tripping SBO output breaker), entry into OP TM-AOP-020 op- TM-EOP-001 MURO MARO '\

7 ICR02 CCRS EFW Valves for "A\ and "B" OTSG's fail to 0% in Auto, manual ICR04 control required. (AtlO)

CARO 8 EG01A CCRS "A" EDG fails to start. manual start required. (URO)

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

-1

Appendix D Scenario Outline Form ES-D-1

-.~. ----~-----------------------

Facility: Three Mile Island Scenario No.: 3 Op Test No.: 289-2011-301 Examiners: Operators: I Initial Conditions: * (Temporary IC-54)

  • 100% Power, MOL
  • EF-P-1 OOS for bearing replacement, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> into a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> clock T.S. 3.4.1.1 (2)

Maintain 100% Power Operations

  • Reduce Steaming/Isolate Affected SGs (includes

.... __use.. of SG drains) (CT-22)

  • Minimize SCM (CT-7)

Eve~ Malf. No. Event Type* Event Description 1 03A4S01 I CRS Inadvertent ES Actuation, "B" Train (TS), entry into OP-TM-AOP 046 IURO ZDIPB1 R (URO: Defeats signal, ARO: Opens MU-V-2NB)

CBON IARO TSCRS 2 RC08B ICRS Tc Instrument Fails High, SASS Fails to Actuate, entry into OP IC51A TM-AOP-070 IURO (URO: Manual control of Control Rods, ARO: Manual control 01 IARO Feedwater) 3 TH17B CCRS -30 gpm "B" OTSG Tube Leak (TS), entry into OP-TM-EOP-005 CURO (URO: Guide 9)

TSCRS 4 N/A NCRS Power Reduction lAW 1102-4 R URO I NARO 5 CC04A CCRS Loss of ICCW, entry into OP-TM-AOP-032, OP-TM-EOP-001

  • CC04B CURO (URO: Trips Reactor) 6 TH16B M CRS -800 gpm "B" OTSG Tube Rupture, entry into OP-TM-EOP-005 M URO MARO 7 MS09A CCRS "B" TBV's fail closed, entry into OP-TM-421-451 MS09B CARO (ARO: Places ADV on Backup Loader)

MS09C

  • (N)ormal, (R)eactivity, (I)nstrument, (C)ompon l21nt, (M)ajor

Appendix D Scenario Outline Form ES-D-1

~

Facility: Three Mile Island Scenario No.: 4 Op Test No.: NRC Examiners: Operators:

i Initial Conditions: * (Temporary IC-55)

  • 5% Power, MOL
  • EF-P-1 OOS for bearing replacement Turnover: Continue with Power escalation Critical Tasks:
  • Control HPI (CT-5)
  • Isolate Overcooling SGs (CT-1 '7)

I Event No. Malf. No. Event Type* Event Description i

1 N/A NCRS Raise Reactor Power from 5% to 10%

RURO 2 IC37A CCRS Invalid "A" OTSG Low Level (TS), "A" EFW actuation, entry into OP-TM-424-90 1 IC41A . CARO

' (ARO: defeats invalid signal, secures EF-P-2A) i TSCRS 3 lAO? CCRS Loss of Instrument Air, entry into OP-TM-AOP-028 IA01C CARO (ARO: Starts IA-P-1A or B) 4 MS02B CCRS Steam Leak into the Reactor Building, entry into OP-TM-AOP-051 CARO RR-P-1 B Fails to start (TS)

TSCRS (ARO: Initiate RI3 Emergency Cooling) 5 THOS CCRS PORV fails open, entry into OP-TM-MAP-G01 06 CURO (LlRO: Closes PORV Block Valve) 6 R003C CCRS Uncontrolled outward rod motion, group 7, entry into OP-TM-AOP 064 and OP-TM-EOP-001, Reactor Trip CURO

, (URO: Selects Sequence Override, performs IMA's of EOP-001) 7 MS02B MCRS Steam Rupture in Reactor Building, entry into OP-TM-EOP-003,

. M URO Excessive Heat Transfer, and OP-TM-EOP-01 0, Rule 3 to isolate "B" OTSG.

MARO 8 MUR6? CCRS MU-V-36, MU-P-1A/1 B/1C RECIRC ISOL VALVE, breaker opens, Valve fails closed. Alternative minimum flow path for Makeup MUR94 CURO i Pump established.

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

-1

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