Text

Draft - Outlines (Folder 2)
	ML15335A328
Person / Time
Site:	Three Mile Island
Issue date:	06/30/2015
From:	; NRC Region 1
To:	; Exelon Generation Co
Shared Package
ML15135A253	List: IR 05000289/2015301; ML15267A771; ML15292A369; ML15292A370; ML15335A327; ML15335A328; ML15335A329; ML15335A330; ML15335A331; ML15335A332; ML15352A064; ML15352A065;
References
	TAC U01913, TMl-15-075
	Download: ML15335A328 (30)
	v • d • e

Exelon Gene rat ion Three Mile Island Unit 1 Telephone 717-948-8000 Route 441 South, P.O. Box 480 Middletown, PA 17057 June 30, 2015 TMl-15-075 USNRC, Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMl-1)

RENEWED OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289

SUBJECT:

SUBMIITAL OF INITIAL OPERATOR LICENSING EXAMINATION OUTLINES Enclosed are the examination outlines, supporting the Initial License Examination scheduled for the week of September 28, 2015, at Three Mile Island Unit 1.

This submittal includes all appropriate Examination Standard forms and outlines in accordance with NU REG 1021, "Operator Licensing Examination Standards for Power Reactors," Revision 10.

In accordance with NUREG 1021, Revision 10, Section ES-201, "Initial Operator Licensing Examination Process," please ensure that these materials are withheld from public disclosure until after the examinations are complete.

Should you have any questions concerning this letter, please contact Mike Fitzwater of Regulatory Assurance at {717) 948-8228. For questions concerning examination materials, please contact Rich Megill, Exam Author, at (717) 948-2023.

Respectfully, Thomas Haaf Site Vice President (Acting), Three Mile Island Unit 1 Exelon Generation Co., LLC TPH/mdf

Enclosures:

(Mailed to Peter Presby, Chief Examiner, NRC Region I)

Examination Security Agreement (Form ES-201-3)

Administrative Topics Outline (Form ES-301-1)

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

PWR Examination Outline (Form ES-401-2)

SUBMITIAL OF INITIAL OPERATOR LICENSING EXAMINATION OUTLINES TMl-15-075 Page 2 Generic Knowledge and Abilities Outline (Tier 3) (Form ES-401-3)

Statement detailing method of Written Outline generation Scenario Outline (Form ES-D-1)

Record of Rejected K/As (Form ES-401-4)

Completed Checklists:

Examination Outline Quality Checklist (Form ES-201-2)

Transient and Event Checklist (Form ES-301-5) cc: (without attachments)

Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector-TM! Unit 1

Statement Detailing Method of Written Outline Generation:

All original and replacement K/A's for Three Mile Island ILT Class 14-01 NRC Written Examination have been randomly selected utilizing Westinghouse NRC KIA Exam Generator (NKEG) software.

ES-401 PWR Examination Outline Form ES-401-2 Facility: Three Mile Island Date of Exam: 09/28/2015 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4 *

1. 1 3 3 3 3 3 3 18 3 3 6 Emergency

& Abnormal 2 1 2 2 N/A 2 1 N/A 1 9 2 2 4 Plant Evolutions Tier Totals 4 5 5 5 4 4 27 5 5 10 1 3 2 3 3 3 2 2 3 2 3 2 28 3 2 5 2.

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

3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 10 7 Categories 2 3 2 3 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO outlines (i.e. except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category.)

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 NRG 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/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A 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 K/As 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 K/A categories.

7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A 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 K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (I Rs) 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 K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals(#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As ES-401, Page 21 of 33

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO I SRO)

E/APE #I Name I Safety Function K K K A A G KIA Topic(s) IR #

1 2 3 1 2 EK1 .2 - Knowledge of the operational implications of the following concepts as 000007 (BW/E02&E10; CE/E02) Reactor they apply to the (Post-Trip Stabilization):

Trip - Stabilization - Recovery I 1 x Normal, abnormal and emergency 3.5 3 operating procedures associated with (Post-Trip Stabilization)

AK2.03 - Knowledge of the interrelations 000008 Pressurizer Vapor Space between the Pressurizer Vapor Space Accident/ 3 x Accident and the following: Controllers 2.5 4 and positioners EA2.04 - Ability to determine or interpret 000009 Small Break LOCA I 3 x the following as they apply to a small 3.8 13 break LOCA: PZR level EK2.02 - Knowledge of the interrelations 000011 Large Break LOCA I 3 x between the and the following Large 2.6 5 Break LOCA: Pumps AA 1.02 - Ability to operate and I or monitor the following as they apply to the 000015/17 RCP Malfunctions I 4 x Reactor Coolant Pump Malfunctions (Loss 2.8 10 of RC Flow): RCP oil reservoir level and alarm indicators 2.2.39 - Knowledge of less than or equal 000022 Loss of Rx Coolant Makeup I 2 x to one hour Technical Specification action 3.9 16 statements for systems.

AK3.02 - Knowledge of the reasons for the following responses as they apply to the 000025 Loss of RHR System I 4 x Loss of Residual Heat Removal System: 3.3 7 Isolation of RHR low-pressure piping prior to pressure increase above specified level AK3.02 - Knowledge of the reasons for the following responses as they apply to the 000026 Loss of Component Cooling Loss of Component Cooling Water: The Water I 8 x automatic actions (alignments) within the 3.6 8 CCWS resulting from the actuation of the ES FAS 000027 Pressurizer Pressure Control 2.4.18 - Knowledge of the specific bases System Malfunction I 3 x for EOPs.

3.3 17 EK1 .01 - Knowledge of the operational implications of the following concepts as 000038 Steam Gen. Tube Rupture I 3 x they apply to the SGTR: Use of steam 3.1 1 tables AK1 .06 - Knowledge of the operational 000040 (BW/E05; CE/E05; W/E12) implications of the following concepts as Steam Line Rupture - Excessive Heat x they apply to Steam Line Rupture: High 3.7 2 Transfer I 4 enerav steam line break considerations AA 1.01 - Ability to operate and I or monitor the following as they apply to the 000054 (CE/E06) Loss of Main Feedwater I 4 x Loss of Main Feedwater (MFW): AFW 4.5 11 controls, including the use of alternate AFW sources 2.2.42 - Ability to recognize system 000055 Station Blackout I 6 x parameters that are entry-level conditions 3.9 18 for Technical Specifications.

AA2.07 - Ability to determine and interpret the following as they apply to the Loss of 000057 Loss of Vital AC Inst. Bus I 6 x Vital AC Instrument Bus: Valve indicator 3.3 14 of charging pump suction valve from RWST AK3.01 - Knowledge of the reasons for the following responses as they apply to the 000058 Loss of DC Power I 6 x Loss of DC Power: Use of de control 3.4 9 power by D/Gs ES-401, Page 22 of 33

ES-401 2 Form ES-401-2 AA2.02 - Ability to determine and interpret the following as they apply to the Loss of 000062 Loss of Nuclear Svc Water I 4 x Nuclear Service Water: The cause of 2.9 15 ossible SWS loss EK2.1 - Knowledge of the interrelations between the (Inadequate Heat Transfer) and the following: Components, and BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink I 4 x functions of control and safety systems, 3.8 6 including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

AA 1.04 - Ability to operate and/or monitor 000077 Generator Voltage and Electric the following as they apply to Generator Grid Disturbances I 6 x Voltage and Electric Grid Disturbances:

4.1 12 Reactor controls 2.4.41 - Knowledge of the emergency 000009 Small Break LOCA I 3 x action level thresholds and classifications.

4.6 79 EA2.01 - Ability to determine or interpret the following as they apply to a Large 000011 Large Break LOCA I 3 x Break LOCA: Actions to be taken, based 4.7 76 on RCS temperature and pressure -

saturated and su erheated AA2.09 - Ability to determine and interpret the following as they apply to the Reactor 000015/17 RCP Malfunctions I 4 x Coolant Pump Malfunctions (Loss of RC 3.5 77 Flow): When to secure RCPs on high stator tern eratures 2.4.30 - Knowledge of events related to system operation/status that must be reported to internal organizations or 000057 Loss of Vital AC Inst. Bus I 6 x external agencies, such as the State, the 4.1 80 NRG, or the transmission system o erator.

2.4.11 - Knowledge of abnormal condition 000058 Loss of DC Power I 6 x procedures.

4.2 81 EA2.1 - Ability to determine and interpret the following as they apply to the (Vital 000007 (BW/E02&E1 O; CE/E02) Reactor System Status Verification): Facility Trip - Stabilization - Recovery I 1 x conditions and selection of appropriate 4.0 78 procedures during abnormal and emer enc o erations.

ES-401, Page 22 of 33

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO I SRO)

E/APE #I Name I Safety Function K K K A A G KIA Topic(s) IR #

1 2 3 2 AK2.01 - Knowledge of the interrelations between the Inoperable I Stuck Control 000005 Inoperable/Stuck Control Rod I 1 x Rod and the following: Controllers and 2.5 20 ositioners AA 1.19 - Ability to operate and I or monitor the following as they apply to 000024 Emergency Boration I 1 x Emergency Boration: Makeup control 3.2 23 system selector switch for eves isolation valve AK3.3 - Knowledge of the reasons for the following responses as they apply to the (Loss of NNl-X): Manipulation of controls BW/A02&A03 Loss of NNl-X/Y I 7 x required to obtain desired operating 3.7 25 results during abnormal, and emergency situations.

AK1 .1 - Knowledge of the operational implications of the following concepts as 000036 (BW/A08) Fuel Handling Accident I 8

x they apply to the (Refueling Canal Level 3.7 19 Decrease): Components, capacity, and function of emer enc s stems.

AK3.01 - Knowledge of the reasons for the following responses as they apply to the 000051 Loss of Condenser Vacuum I 4 x Loss of Condenser Vacuum: Loss of 2.8 22 steam dump capability upon loss of condenser vacuum AK2.03 - Knowledge of the interrelations between the Loss of Containment Integrity 000069 (W/E14) Loss of CTMT Integrity I 5 x and the following: Personnel access 2.8 21 hatch and emer enc access hatch 2.1.25 - Ability to interpret reference BW/A01 Plant Runback I 1 x materials, such as graphs, curves, tables, 3.9 27 etc.

AA2.2 - Ability to determine and interpret the following as they apply to the (Emergency Diesel Actuation): Adherence BW/A05 Emergency Diesel Actuation I 6 x to appropriate procedures and operation 3.5 26 within the limitations in the facility*s license and amendments.

EA 1.2 - Ability to operate and I or monitor the following as they apply to the (EOP BW/E13&E14 EOP Rules and Enclosures x Enclosures): Operating behavior 2.8 24 characteristics of the facili 2.4.6 - Knowledge of EOP mitigation 000024 Emergency Boration I 1 x strata ies.

4.7 84 AA2.02 - Ability to determine and interpret 000036 (BW/A08) Fuel Handling Accident I the following as they apply to the Fuel 8

x Handling Incidents: Occurrence of a fuel 4.1 82 handlin incident AA2.1 - Ability to determine and interpret the following as they apply to the (NNl-X):

BW/A02&A03 Loss of NNl-X/Y I 7 x Facility conditions and selection of 4.0 83 appropriate procedures during abnormal and emer enc o erations BW/E08; W/E03 LOCA Cooldown - 2.4.18 - Knowledge of the specific bases De ress. I 4 for EOPs.

ES-401, Page 23 of 33

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO I SRO)

System #I Name K K K K K K A A A A G KIA Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 K5.02 - Knowledge of the operational x implications of the following concepts 003 Reactor Coolant Pump 2.8 39 as they apply to the RCPS: Effects of RCP coastdown on RCS parameters K6.14 - Knowledge of the effect of a x loss or malfunction on the following will 003 Reactor Coolant Pump 2.6 42 have on the RCPS: Starting reauirements K4.10 - Knowledge of CVCS design feature(s) and/or interlock(s) which 004 Chemical and Volume x provide for the following: Minimum 3.2 36 Control temperature requirements on borated systems (prevent crystallization)

A2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based 005 Residual Heat Removal x on those predictions, use procedures to 2.9 46 correct, control, or mitigate the consequences of those malfunctions or operations: RHR valve malfunction A4.04 - Ability to manually operate and/or monitor in the control room:

005 Residual Heat Removal x 3.1 51 Controls and indication for closed coolino water pumps K1 .13 - Knowledge of the physical x connections and/or cause-effect 006 Emergency Core Cooling 3.3 28 relationships between the EGGS and the followino systems: CSS K5.02 - Knowledge of the operational 007 Pressurizer Relief/Quench implications of the following concepts x 3.1 40 Tank as they apply to PRTS: Method of formino a steam bubble in the PZR K1 .04 - Knowledge of the physical connections and/or cause-effect x relationships between the CCWS and 008 Component Cooling Water 3.3 29 the following systems: RCS, in order to determine source(s) of RCS leakage into the CCWS A4.1 O - Ability to manually operate x and/or monitor in the control room:

008 Component Cooling Water 3.1 52 Conditions that require the operation of two CCW coolers K3.02 - Knowledge of the effect that a 010 Pressurizer Pressure x loss or malfunction of the PZR PCS will 4.0 33 Control have on the followino: RPS K6.11 - Knowledge of the effect of a x loss or malfunction of the 012 Reactor Protection 2.9 43 following will have on the RPS: Trip setpoint calculators K2.01 - Knowledge of bus power 013 Engineered Safety Features x supplies to the following: 3.6 31 Actuation ESFAS/safeauards eauipment control A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based 022 Containment Cooling x on those predictions, use procedures to 2.5 47 correct, control, or mitigate the consequences of those malfunctions or operations: Fan motor over-current A4.05 - Ability to manually operate x and/or monitor in the control room:

022 Containment Cooling 3.8 53 Containment readings of temperature, pressure, and humidity system ES-401, Page 24 of 33

ES-401 4 Form ES-401-2 K4.06 - Knowledge of CSS design feature(s) and/or interlock(s) which 026 Containment Spray x 2.8 37 provide for the following: Iodine scaven in via the CSS K4.07 - Knowledge of MRSS design feature(s) and/or interlock(s) which 039 Main and Reheat Steam x 3.4 38 provide for the following: Reactor build in isolation A3.02 - Ability to monitor automatic 039 Main and Reheat Steam x operation of the MRSS, including: 3.1 49 Isolation of the MRSS 2.2.44 - Ability to interpret control room indications to verify the status and 059 Main Feedwater x operation of a system, and understand 4.2 54 how operator actions and directives affect lant and s stem conditions.

K5.01 - Knowledge of the operational implications of the following concepts 061 Auxiliary/Emergency x as the apply to the AFW: Relationship 3.6 41 Feedwater between AFW flow and RCS heat transfer A2.08 - Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to 061 Auxiliary/Emergency x correct, control, or mitigate the 2.7 48 Feedwater consequences of those malfunctions or operations: Flow rates expected from various combinations of AFW pump dischar e valves K1 .02 - Knowledge of the physical connections and/or cause effect 062 AC Electrical Distribution x relationships between the ac 4.1 30 distribution system and the following s stems: ED/G K3.01 - Knowledge of the effect that a loss or malfunction of the ac distribution 062 AC Electrical Distribution x 3.5 34 system will have on the following:

Ma*or s stem loads A 1.01 - Ability to predict and/or monitor changes in parameters associated with 063 DC Electrical Distribution x operating the DC electrical system 2.5 44 controls including: Battery capacity as it is affected b dischar e rate A 1.01 - Ability to predict and/or monitor changes in parameters (to prevent 064 Emergency Diesel exceeding design limits) associated x 3.0 45 Generator with operating the ED/G system controls including: ED/G lube oil tern erature and ressure K3.01 - Knowledge of the effect that a 073 Process Radiation loss or malfunction of the PRM system x 3.6 35 Monitoring will have on the following: Radioactive effluent releases K2.08 - Knowledge of bus power 076 Service Water x supplies to the following: ESF-actuated 3.1 32 MOVs A3.01 - Ability to monitor automatic 078 Instrument Air x operation of the IAS, including: Air 3.1 50 ressure 2.2.37 - Ability to determine operability 103 Containment and/or availability of safety related 3.6 55 e ui ment.

ES-401, Page 24 of 33

ES-401 4 Form ES-401-2 A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) 01 O Pressurizer Pressure based on those predictions, use x 4.2 86 Control procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PORV failures A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) 013 Engineered Safety Features x based Ability on those predictions, use 4.8 87 Actuation procedures to correct, control, or mitigate the consequences of those malfunctions or o erations: LOCA 2.4.20 - Knowledge of the operational 063 DC Electrical Distribution x implications of EOP warnings, cautions, 4.3 89 and notes.

A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and 073 Process Radiation x (b) based on those predictions, use 2.9 88 Monitoring procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Calibration drift 2.4.1 - Knowledge of EOP entry 076 Service Water x 4.8 90 conditions and immediate action steps.

ES-401, Page 24 of 33

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO I SRO)

System # I Name K K K K K K A A A A G KIA Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 K3.01 - Knowledge of the effect that a 001 Control Rod Drive x loss or malfunction of the GROS will 2.9 58 have on the followin : eves x K2.02 - Knowledge of bus power 011 Pressurizer Level Control 3.1 57 supplies to the following: PZR heaters A4.04 - Ability to manually operate 029 Containment Purge x and/or monitor in the control room: 3.5 65 Containment evacuation si nal K6.02 - Knowledge of the effect of a loss or malfunction on the following 034 Fuel Handling Equipment x will have on the Fuel Handling 2.6 61 System: Radiation monitoring s stems K5.01 - Knowledge of operational implications of the following concepts 035 Steam Generator x as the apply to the S/GS: Effect of 3.4 60 secondary parameters, pressure, and tern erature on reactivi A 1.02 - Ability to predict and/or monitor changes in parameters (to 041 Steam Dump/Turbine Bypass x prevent exceeding design limits) 3.1 62 Control associated with operating the SOS controls includin : Steam ressure K4.01 - Knowledge of design feature(s) and/or interlock(s) which provide for the following: Safety and 068 Liquid Radwaste x 3.4 59 environmental precautions for handling hot, acidic, and radioactive Ii uids A3.03 - Ability to monitor automatic operation of the Waste Gas Disposal 071 Waste Gas Disposal x System including: Radiation 3.6 64 monitoring system alarm and actuating si nals K1 .01 - Knowledge of the physical connections and/or cause effect 075 Circulating Water x relationships between the circulating 2.5 56 water system and the following s stems: SWS A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those 086 Fire Protection x 3.0 63 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Low FPS header ressure A2.15 - Ability to (a) predict the impacts of the following malfunction or operations on the CRDS- and (b) based on those predictions, use 001 Control Rod Drive x 4.2 91 procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Quadrant ower tilt ES-401, Page 25 of 33

ES-401 5 Form ES-401-2 A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use 016 Non-nuclear Instrumentation x 3.1 92 procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure 2.4.45 - Ability to prioritize and 035 Steam Generator x interpret the significance of each 4.3 93 annunciator or alarm.

0 KIA Category Point Totals: 1 1 1 1 1 1 1 ¥.! 1 1 I Group Point Total: 10/3 1

ES-401 , Page 25 of 33

ES-401 General Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Three Mile Island Date of Exam: 09/28/2015 Category KIA# Topic RO SRO-Only IR # IR #

Ability to use plant computers to evaluate system or 2.1.19 3.~ 66 component status.

Ability to interpret reference materials, such as 2.1.25 qraphs, curves, tables, etc.

3. 't 67 Ability to use procedures related to shift staffing,

1. 2.1.5 such as minimum crew complement, overtime 3/~ 94 Conduct limitations, etc.

of Operations Knowledge of the fuel-handling responsibilities of 2.1.35 3,t) 95 SROs.

2.1.

Subtotal 2 2 2.2.13 Knowledge of tagging and clearance procedures. 4.1 68 Knowledge of conditions and limitations in the 2.2.38 3.b 69 facility license.

Ability to recognize system parameters that are 2.2.42 3 'C( 70 entry-level conditions for Technical Specifications.

2.

Equipment 2.2.12 Knowledge of surveillance procedures. 4.1 96 Control Knowledge of the bases in Technical Specifications 2.2.25 for limiting conditions for operations and safety 4.i 97 limits.

2.2. 3 2 Subtotal Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment 2.3.13 3,~ 71 entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Knowledge of radiation monitoring systems, such as

3. 2.3.15 fixed radiation monitors and alarms, portable survey 2.1 72 Radiation instruments, personnel monitoring equipment, etc.

Control 2.3.6 Ability to approve release permits. 3.¥ 98 2.3.

2.3.

Subtotal 2 1

4. Ability to recognize abnormal indications for system Emergency 2.4.4 operating parameters that are entry-level conditions 4,$" 73 Procedures I for emerqency and abnormal operating procedures.

ES-401, Page 26 of 33

ES-401 General Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Plan Knowledge of the organization of the operating 2.4.5 procedures network for normal, abnormal, and 3,7 74 emerqencv evolutions.

Ability to diagnose and recognize trends in an 2.4.47 accurate and timely manner utilizing the appropriate 4'1, 75 control room reference material.

2.4.18 Knowledge of the specific bases for EOPs. 4.0 99 2.4.25 Knowledge of fire protection procedures. 3.7 10 2.4.

Subtotal 3 2 Tier 3 Point Total 10 7 ES-401, Page 26 of 33

ES-401 Record of Rejected K/As Form ES-401-4 Tier I Original KIA Reason for Rejection Group The subject KIA is not relevant at the subject facility.

1I 1 054 I AA1 .03 Replaced with 054IAA1.01 The subject KIA is not relevant at the subject facility.

2I 1 008 I A4.11 Replaced with 008 I A4.1 O Topic overlaps with the Audit Written Exam.

2I 1 059 I K3.03 Replaced with 062 I K3.01 Topic overlaps with the Audit Written Exam.

2I 1 103 I A1 .01 Replaced with 003 I K5.02 Generic KIA oversampled. 078 I 2.2.44 overlaps with 059 I 2.2.44.

2I 1 078 I 2.2.44 Replaced with 022 I A4.05 Original KIA is LOD 1.

2I 1 013 I 2.1.27 Replaced with 103 I 2.2.37 Original KIA is LOD 1.

212 014 I 2.1.27 Replaced with 086 I A2.02 The subject KIA is not relevant at the subject facility.

212 015 I K1.04 Replaced with 075 I K1 .01 The subject KIA is not generic enough for Tier level at the subject facility.

3I 1 2.1.30 Replaced with 2.1.25 Topic overlaps with the Audit Written Exam.

3/3 2.3.11 Replaced with 2.3.13 Generic KIA oversampled. 009 I 2.4.45 overlaps with 035 / 2.4.45.

1 I 1 SRO 009 I 2.4.45 Replaced with 009 I 2.4.41 The subject KIA is not relevant to the topic at the subject facility.

1 I 1 SRO E05 I 2.2.36 Replaced with 057 I 2.4.30 Original KIA is RO LOK.

1I2 SRO 051 I 2.4.1 Replaced with E08 I 2.4.18 Topic oversampled. 060 I 2.4.47 overlaps with RO Generic KIA 2.4.47.

1I2 SRO 060 I 2.4.47 Replaced with 024 I 2.4.6 Original KIA is RO LOK.

2I1 SRO 006 I A2.04 Replaced with 010 I A2.03 Topic overlaps with the Audit Written Exam I NRG Operational.

2 I 2 SRO 075 I A2.03 Replaced with 001 I A2.15 Original KIA is RO LOK.

3 /2 SRO 2.2.1 Replaced with 2.2.21 Topic oversampled. 2.3.15 overlaps with RO Generic KIA 2.3.15.

3/ 3 SRO 2.3.15 Replaced with 2.3.6 ES-401, Page 44 of 50

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: 09/28/2015 Examination Level: RO~ SRO D Operating Test Number: 2015-301 Administrative Topic Type Describe activity to be performed (See Note) Code*

Verify Watchstanding Requirements - Work-Hour Rules Conduct of Operations M/R 2.1.5 (2.9): Ability to use procedures related to shift staffing, such as minimum crew compliment, overtime limitations, etc.

Given a Dropped Rod at Power, Calculate SOM Conduct of Operations M/R 2.1.25 (3.9): Ability to interpret station reference materials such as graphs, curves, tables, etc.

Given Intermediate Closed Cooling Water System Electrical and Mechanical Print Drawings, Identify the Equipment Control N/R Status of Associated Containment Isolation Valves 2.2.41 (3.5): Ability to obtain and interpret station electrical and mechanical drawings.

Radiation Control Category Not Selected for RO Applicants.

Perform State and Local Event Notification Emergency Plan M/S 2.4.43 (3.2): Knowledge of Emergency Communications Systems and Techniques.

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics (which would require all five items).

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 b 1; randomly selected)

ES 301, Page 22 of 27

JPM A1 Conduct of Operations: Verify Watchstanding Requirements - Work-Hour Rules Given plant conditions and references OP-TM-1010-111-1001, Shift Manning Requirements, and LS-AA-119, Overtime Controls, identify which requested days of overtime the candidate may work while staying within the requirements of Work-Hour rules.

Safety Significance: Exelon procedures associated with work-hour rules implement requirements for managing fatigue and controlling work hours in accordance with 10 CFR 26, Subpart I, "Managing Fatigue." The requirements are intended to provide reasonable assurance that worker fatigue will be avoided and that all individuals will be able to safely perform their duties and maintain the health and safety of the public.

This JPM has been modified to ensure that the allowable days are completely different than the previous JPM.

JPM A1 Conduct of Operations: Given a Dropped Rod at Power, Calculate SDM Given a dropped rod at power, calculate Shutdown Margin IAW OP-TM-300-205, Shutdown Margin for Hot Shutdown Conditions.

Safety Significance: Tech Specs require that a Shutdown Margin of> 1% L1k/k must be maintained at all times.

This JPM has been modified to ensure that the data given and resultant calculations are completely different than the previous JPM.

JPM A2 - Equipment Control: Given Intermediate Closed Cooling Water System Electrical and Mechanical Print Drawings, Identify the Status of Associated Containment Isolation Valves Given a set of conditions, a timeline of events, and Intermediate Closed Cooling Water System electrical and mechanical prints, the candidates will determine the status of multiple Containment Isolation Valves.

Safety Significance: Tech Specs require that containment integrity shall be maintained whenever all three of the following conditions exist:

a. Reactor coolant pressure is 300 psig or greater.

b. Reactor coolant temperature is 200 degrees F or greater.

c. Nuclear fuel is in the core.

Containment Integrity exists when the following conditions are satisfied:

c. All active CIVs, including power-operated valves, check valves, and relief valves, are OPERABLE or locked closed. Normally closed active CIVs (other than the purge valves) may be unisolated intermittently or manual control of power-operated valves may be substituted for automatic control under administrative control.

This is a new J PM created for the ILT 14-01 N RC examination.

JPM A4 - Emergency Plan: Perform State and Local Event Notification Given a faulted State and Local Notification form, the candidate will identify the faulted errors on the form and then will simulate performance of making State and Local Event notifications.

Safety Significance: Provides prompt and accurate notification of nuclear station emergencies to local, state and federal agencies.

This JPM has been modified to ensure that the combination of faults given is completely different than the previous JPM.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: 09/28/2015 Examination Level: RO D SRO [gl Operating Test Number: 2015-301 Administrative Topic Type Describe activity to be performed (See Note) Code*

Maintain Minimum Shift Staffing - Control Overtime Conduct of Operations M/R 2.1.5 (3.9): Ability to use procedures related to shift staffing, such as minimum crew compliment, overtime limitations, etc.

Given a Dropped Rod at Power, Review Submitted SOM for Approval Conduct of Operations M/R 2.1.25 (4.2): Ability to interpret station reference materials such as graphs, curves, tables, etc.

Given Intermediate Closed Cooling Water System Electrical and Mechanical Print Drawings, Identify the Status of Associated Containment Isolation Valves with Equipment Control N/R Tech Spec LCO 2.2.41 (3.9): Ability to obtain and interpret station electrical and mechanical drawinqs.

Review RB Entry Survey Log 2.3.13 (3.8): Knowledge of radiological safety procedures Radiation Control M/R pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Determine the Emergency Action Level (EAL) and Make a Protective Action Recommendation (PAR) IAW the Facility Emergency Plan M/S Emergency Plan 2.4.44 (4.4): Knowledge of Emergency Plan Protective Action Recommendations.

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics (which would require all five items).

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

JPM A1 Conduct of Operations: Maintain Minimum Shift Staffing- Control Overtime Given plant conditions and references OP-TM-1010-111-1001, Shift Manning Requirements, and LS-AA-119, Overtime Controls, a prepared Shift Staffing Report, LMS Qual Matrix Report, and a prepared overtime List, identify the required actions to restore minimum staffing and select personnel IAW the requirements to control overtime.

Safety Significance: Exelon procedures associated with work-hour rules implement requirements for managing fatigue and controlling work hours in accordance with 10 CFR 26, Subpart I, "Managing Fatigue." The requirements are intended to provide reasonable assurance that worker fatigue will be avoided and that all individuals will be able to safely perform their duties and maintain the health and safety of the public.

This JPM has been modified to ensure that the allowable personnel and circumstances are different than the previous JPM.

JPM A1 Conduct of Operations: Given a Dropped Rod at Power, Review Submitted SOM for Approval Given a dropped rod at power, review the submitted Shutdown Margin calculation for approval (by calculating Shutdown Margin IAW OP-TM-300-205, Shutdown Margin for Hot Shutdown Conditions, to verify accuracy) and identify the faults. Additionally determines Tech Spec action, and does not approve the submitted SOM.

Safety Significance: Tech Specs require that a Shutdown Margin of > 1% 11k/k must be maintained at all times.

This JPM has been modified to ensure that the data given and resultant calculations are completely different than the previous JPM.

JPM A2 - Equipment Control: Given Intermediate Closed Cooling Water System Electrical and Mechanical Print Drawings, Identify the Status of Associated Containment Isolation Valves with Tech Spec LCO Given a set of conditions, a timeline of events, and Intermediate Closed Cooling Water System electrical and mechanical prints, the candidates will determine the status of multiple Containment Isolation Valves.

Additionally determines Tech Spec action.

Safety Significance: Tech Specs require that containment integrity shall be maintained whenever all three of the following conditions exist:

a. Reactor coolant pressure is 300 psig or greater.

b. Reactor coolant temperature is 200 degrees F or greater.

c. Nuclear fuel is in the core.

Containment Integrity exists when the following conditions are satisfied:

c. All active CIVs, including power-operated valves, check valves, and relief valves, are OPERABLE or locked closed. Normally closed active CIVs (other than the purge valves) may be unisolated intermittently or manual control of power-operated valves may be substituted for automatic control under administrative control.

This is a new JPM created for the ILT 14-01 NRG examination.

JPM A3 - Radiation Control: Review RB Entry Survey Log Given a faulted Reactor Building Entry Survey Log and while referencing RP-TM-460-1007, Access to TMl-1 Reactor Building, identify the faults. Additionally, does not approve the RB entry.

Safety Significance: The Material describes sampling, equipment and conditional requirements needed prior to entry into the TMl-1 Reactor Building. Possible hazards which may exist in Reactor Building include gamma and neutron radiation (reactor critical), airborne radioactive contamination, and explosive or oxygen-deficient atmosphere.

This JPM has been modified to ensure that the combination of faults given is completely different than the previous JPM.

JPM A4 - Emergency Plan: Determine the Emergency Action Level (EAL) and Make a Protective Action Recommendation (PAR) IAW the Facility Emergency Plan Given a set of conditions, declare the appropriate Emergency Classification (a Time Critical component).

Additionally, declare the associated Protective Action Recommendation (also a Time Critical component).

Safety Significance: As required in the conditions set forth by the Nuclear Regulatory Commission (NRG) for the operating licenses for the Exelon Nuclear Stations, the management of Exelon recognizes its responsibility and authority to operate and maintain the nuclear power stations in such a manner as to provide for the safety of the general public. The Exelon Emergency Preparedness Program consists of the Exelon Nuclear Standardized Radiological Emergency Plan ("Standard Plan"), Station Annexes, emergency plan implementing procedures, and associated program administrative procedures. The Standard Plan outlines the basis for response actions that would be implemented in an emergency. Planning efforts common to all Exelon Nuclear stations are encompassed within the Standard Plan.

This JPM has been modified to ensure that the conditions given and the method for deciding the PAR are completely different than the previous JPM.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: 09/28/2015 Exam Level: RO ~ SRO-I D SRO-U D Operating Test No.: 2015-301 Control Room Systems: *8 for RO; 7 for SRO-I; 2 or 3 for SRO-U Safety System I JPM Title Type Code*

Function

a. Respond to an Inoperable/Stuck Control Rod (005) AA 1.01 M/S 1

b. Respond to a Loss of Pressurizer Level Control with Failures (011) D/A/S 2 A2.03

c. Restore Seal Injection with a Loss of ICCW (003) K6.02 P/S/A 4P

d. Respond to an OTSG Overfeed (035) A2.04 N/A/S 4S

e. Initiate RB Spray (026) A2.03 D/US/A/EN 5

f. Lower CFT Level and Pressure from the Control Room (006) A4.02 N/S 3

g. Startup Reactor Protection System Channel (012) A4.02 D/S 7

h. Respond to Loss of SCCW (026) AA 1.05 D/A/S 8 In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Initiate Emergency Boration IAW EOP-020 (004) G2.1.30 D/E/L/R 1

j. Respond to a Loss of Instrument Air (078) A3.01 D/E 8

k. EFW from Fire Service using FS-P-15 (061) A2.04 D/E/L 4S

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

Type Codes Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank ::::. 9 I :::_8 I ::::. 4 (E)mergency or abnormal in-plant .::: 1 I !:'.. 1 I .::: 1 (EN)gineered safety feature .::: 1 I .::: 1 I .::: 1 (control room system)

(L)ow-Power I Shutdown .::: 1 I > 1 I .::: 1 (N)ew or (M)odified from bank including 1 (A) .::: 2 I .:::2 I .::: 1 (P)revious 2 exams ::::. 3 I :::_3 I ::::. 2 (randomly selected)

(R)CA .::: 1 I !:'.. 1 I .::: 1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: 09/28/2015 Exam Level: RO D SRO-I [8] SRO-U D Operating Test No.: 2015-301 Control Room Systems: *8 for RO; 7 for SRO-I; 2 or 3 for SRO-U Safety System I JPM Title Type Code*

Function

a. Respond to an Inoperable/Stuck Control Rod (005) AA 1.01 M/S 1

b. Respond to a Loss of Pressurizer Level Control with Failures (011) D/A/S 2 A2.03

c. Restore Seal Injection with a Loss of ICCW (003) K6.02 P/S/A 4P

d. Respond to an OTSG Overfeed (035) A2.04 N/A/S 4S

e. Initiate RB Spray (026) A2.03 D/US/A/EN 5

f. Lower CFT Level and Pressure from the Control Room (006) A4.02 N/S 3 g.

h. Respond to Loss of SCCW (026) AA 1.05 D/A/S 8 In-Plant Systems * (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Initiate Emergency Boration IAW EOP-020 (004) G2.1.30 D/E/UR 1

j. Respond to a Loss of Instrument Air (078) A3.01 D/E 8

k. EFW from Fire Service using FS-P-15 (061) A2.04 D/E/L 4S

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five 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 I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank 5':. 9 I 5':. 8 I 5':. 4 (E)mergency or abnormal in-plant 2: 1 I 2: 1 I 2: 1 (EN)gineered safety feature 2: 1 I 2: 1 I 2: 1 (control room system)

(L)ow-Power I Shutdown 2: 1 I 2: 1 I 2: 1 (N)ew or (M)odified from bank including 1 (A) 2: 2 I ?.2 I 2: 1 (P)revious 2 exams 5':. 3 I 5':,3 I 5':. 2 (randomly selected)

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

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: 09/28/2015 Exam Level: RO 0 SRO-I D SRO-U [8J Operating Test No.: 2015-301 Control Room Systems: *8 for RO; 7 for SRO-I; 2 or 3 for SRO-U Safety System I JPM Title Type Code*

Function a.

b.

c.

d. Respond to an OTSG Overfeed (035) A2.04 N/A/S 4S

e. Initiate RB Spray (026) A2.03 D/US/A/EN 5

f. Lower CFT Level and Pressure from the Control Room (006) A4.02 N/S 3 g.

h.

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

i. Initiate Emergency Boration IAW EOP-020 (004) G2.1.30 D/E/UR 1

j. Respond to a Loss of Instrument Air (078) A3.01 D/E 8 k.

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five 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 I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank s91 sB I s 4 (E)mergency or abnormal in-plant ~1 I >1 I ~1 (EN)gineered safety feature ~ 1I ~ 1 I ~ 1 (control room system)

(L)ow-Power I Shutdown ~ 1 I ~1 I ~1 (N)ew or (M)odified from bank including 1(A) ~21 ~2 I ~ 1 (P)revious 2 exams s 3I s3 I s 2 (randomly selected)

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

JPM A: Respond to an Inoperable/Stuck Control Rod (Modified JPM): The candidate will take control of an individual Control Rod which is greater than 7 inches off from the rest of the rod group and return it to within the acceptable band IAW OP-TM-622-414, Exercising One or More Control Rods.

Safety Significance: Nine inches is a Tech Spec limit that requires the Control Rod to be declared inoperable. The alarm comes in at 7 inches in order to take action prior to reaching the Tech Spec limit. The Tech Spec axial power imbalance, quadrant power tilt, and control rod position limits are based on LOCA analyses which have defined the maximum linear heat rate. These limits are developed in a manner that ensures the initial condition LOCA maximum linear heat rate will not cause the maximum clad temperature to exceed 10 CFR 50 Appendix K.

This JPM is modified from the previous JPM in that this is designed for the Digital Control Rod Drive System, whereas the original JPM was designed for the older Analog Digital Control Rod System.

JPM B: Respond to a Loss of Pressurizer Level Control with Failures (Bank JPM): The candidate will take manual control of the Pressurizer makeup valve to avoid improper Pressurizer level while at power.

Once the alternate instrument is selected and the Pressurizer makeup valve is placed back in automatic control, the upstream Pressurizer makeup valve will fail closed, forcing the candidate to control Pressurizer level with an HPI valve IAW OP-TM-EOP-010, Guide 9, RCS Inventory Control.

Safety Significance: Tech Specs require that the reactor shall be maintained subcritical by at least one percent delta k/k until a steam bubble is formed and an indicated water level between 80 and 385 inches is established in the pressurizer. If level is too low, there will not be enough inventory to keep the core covered on an event, even a reactor trip, which would cause a transfer of the bubble from the Pressurizer to the Reactor Vessel. If level is too high, there would not be sufficient room in the Pressurizer to prevent severe overpressurization in the event of any single failure.

This JPM is alternate path because the candidate must identify that the upstream valve has gone closed, leave the alarm response, and enter Guide 9.

JPM C: Restore Seal Injection with a Loss of ICCW (Previous two JPM's): The candidate will restore seal injection IAW OP-TM-AOP-041, Loss of Seal Injection. The first Makeup Pump (MU-P-1A) will not start and the candidate will continue in the procedure to start MU-P-1 C. As soon as MU-P-1 C starts, a loss of Intermediate Closed Cooling Water occurs. The candidate will identify no seal cooling to the Reactor Cooling Pumps and will trip the reactor and secure the reactor coolant pumps.

Safety Significance: To avoid seal damage, seal injection water flow is required to all RCPs when reactor coolant temperature is above 190°F and pressure is above 100 psig, except while operating in the loss of injection mode. Operating the RCPs in the loss of seal injection mode without intermediate cooling water operating may result in damage to the pump bearing and/or seals from particles in the reactor coolant.

This JPM is alternate path because the candidate must identify that the Reactor Coolant Pumps have no seal cooling and return from Section 6 to Section 3 of OP-TM-AOP-041. The candidate will then enter OP-TM-EOP-001, Reactor Trip.

JPM was randomly chosen via selecting from playing cards representing JPM's from the last two NRC examinations.

JPM D: Respond to an OTSG Overfeed (New Alternate JPM):

The candidate will take manual control of Feedwater regulating valves. While manipulating the "B" set of valves, Feedwater to the "B" OTSG will become excessive, causing the candidate to trip the reactor based on OTSG isolation. Additionally, a Main Feedwater Valve does not automatically isolate on high level and the operator must manually close the valve.

Safety Significance: If FW is not being controlled and level exceeds 97.5% operating range, actions are taken to promptly stop the overfeed and minimize possible water carryover or main steam line flooding. At 97.5%, HSPS should have stopped FW flow by closing the main FW valves.

This JPM is alternate path because the candidate must identify that the "B" OTSG hi level has occurred and that the reactor may not remain critical. Therefore, the candidate will exit the alarm response and trip the reactor IAW OP-TM-EOP-001, Reactor Trip.

JPM E: Initiate RB Spray (Bank Alternate JPM): The candidate will initiate Reactor Building Spray IAW OP-TM-214-901, RB Spray Operation.

Safety Significance: The Reactor Building Spray System is a Safety Related system that provides for protection of the integrity of the Reactor Building and limits the release of radioactivity to less than 10CFR100 limits following a Loss of Coolant Accident. The Building Spray system accomplishes this by:

a. Maintaining Reactor Building pressure less than 55 psig

b. Absorbing Iodine

c. Providing a means for measurement of Reactor Building pressure

d. Providing a means to establish post LOCA liquid inventory in an acceptable long term pH range This JPM is alternate path because the candidate must identify that the "B" Building Spray Train has not properly actuated and then route from section 4.1 to 4.2 of OP-TM-214-901 to take the compensating actions.

JPM F: Lower CFT Level and Pressure from the Control Room (New JPM): The candidate will restore Core Flood Tank "A" level and pressure IAW OP-TM-213 series procedures.

Safety Significance: Unlike any of the other ECCS components, which require that only train be operational, Tech Specs state that both CFTs are required because a single CFT has insufficient inventory to reflood the core for hot and cold line breaks.

This JPM is a new JPM, created for ILT 14-01 NRC examination.

JPM G: Startup Reactor Protection System Channel (Bank JPM): The candidate will startup Reactor Protection System Channel "D" IAW OP-TM-641-404, De-energizing RPS Channel D.

Safety Significance: IAW Tech Specs, There are four reactor protection channels. Normal trip logic is two out of four. Minimum required trip logic is one out of two. Every reasonable effort will be made to maintain all safety instrumentation in operation.

This JPM is a bank JPM.

JPM H: Respond to Loss of SCCW (Bank Alternate JPM): The candidate will identify that a Secondary Closed Cooling Water Pump has tripped with no automatic start of the standby pump. The candidate will manually start the standby pump and then recognize that SCCW surge tank level has dropped and will secure SCCW cooled components IAW OP-TM-AOP-033, Loss of Secondary Component Cooling.

Safety Significance: OP-TM-AOP-033, Loss of Secondary Component Cooling, is designed to mitigate the effects of loss of cooling to the components cooled by the secondary closed cooling system. This procedure provides the mitigation strategy for events that challenge the system function. If secondary closed flow is lost (or pumps must be shutdown), then each component cooled by secondary closed is shutdown or otherwise protected from loss of cooling. CSF 4, Core Heat Removal, is affected by the following means:

Main Feedwater capability is lost (Condensate, Condensate Booster & Main FW Pumps are not available). Condenser Vacuum may be lost. EFW and ADVs are used for RCS heat removal via OTSGs.

This JPM is alternate path because the candidate must identify that the the SCCW Surge Tank has lowered to the point where it is no longer providing net positive suction head to the SCCW Pumps and then must exit the alarm response and route to OP-TM-AOP-033, Loss of Secondary Component Cooling.

JPM I: Initiate Emergency Boration IAW EOP-020 (Bank JPM): The candidate will perform the in-plant steps required to initiate Emergency Boration. This task includes signing onto an RWP.

Safety Significance: Emergency boration is desired to insert negative reactivity and ensure the reactor remains shutdown during a cooldown.

This JPM is a bank JPM.

JPM J: Respond to a Loss of Instrument Air (Bank JPM): The candidate will perform the in-plant steps required to start and maintain backup Instrument Air Compressors.

Above 60 psig, the actions are focused on restoring IA system pressure and identifying the problem, Below 60 psig, the actions are focused on safety and equipment protection. The most challenging post trip threat is the potential to lose all means of RCP seal cooling. Actions must be quickly performed to maintain seal injection or thermal barrier cooling. Primary inventory control (letdown and bleed capabilities), RCS heat removal (OTSG feeding and steaming capabilities) are also affected. CSF-5, Containment Integrity, is affected by the following means:

All containment isolation valves fail closed on loss of IA.

This JPM is a bank JPM.

JPM K: EFW from Fire Service using FS-P-15 (Bank JPM): The candidate will perform the in-plant steps required to align a portion of the Emergency Feedwater System to be supplied with Fire Service Water.

Safety Significance: The purpose is to provide a means to remove decay heat following a complete loss of the control tower, AC/DC power and licensed operators. When OTSG pressures are approximately 250 psig, FS-P-15 will be capable of providing sufficient head/flow for adequate decay heat removal. Once FS-P-15 is aligned and capable of providing adequate cooling flow (i.e., OTSG pressure< 250 psig and flow requirements less than 200 gpm), OTSG feed will be swapped to the portable pump with flow from the pump matching decay heat.

This JPM is a bank JPM.

Appendix D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 1 Op Test No.: 14-01 NRG Examiners: Operators:

Initial Conditions: * (Temporary IC-175)

85% Power, MOL

BS-P-1A is OOS for maintenance, expected to return to service in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Crane work is occurring on the West side of the Plant to stage new piping Turnover: Maintain 85% Power Operations Critical Tasks:

PORV Control for Heat Transfer (CT-13) (If conditions are met)

Shutdown Reactor - ATWS (CT-24)

Restore Feed to a Dry OTSG (CT-26)

Event No. Malt. No. Event Type* Event Description 1 RM0323 TSCRS Reactor Building Hi Range Radiation Monitor, RM-G-23, Failure 2 ZAIRC1LIC CCRS MU-V-17 Fails Closed in Auto, entry into OP-TM-211-472 CURO (URO: Controls Pressurizer Level with MU-V-17 in Manual) 3 ED09D TSCRS Loss of D Inverter, Loss of VBD, entry into OP-TM-AOP-018 CARO (ARO: Place Rad Monitors Interlock switches to Defeat, Restore Control Building Ventilation) 4 02A5S81 CCRS Low Makeup Tank Pressure, entry into OP-TM-MAP-D0303 CURO (URO: Raise Makeup Tank pressure) 5 IC23 I CRS SG/RX Demand Station fails to 0 Volts, Entry into OP-TM-AOP-070 IURO IARO (URO/ARO: ICS station to Manual, Stabilize Power) 6 MU29 CCRS RCS leak through the Letdown Line, entry into OP-TM-AOP-050 RURO CARO (URO: Lowers power in Manual ARO: Isolate the Letdown Line) 7 FW15B MCRS "B" Main Feed Pump trips, "A" Main Feed Pump Runs to 0 rpm, RD28 ATWS, Lack of Primary to Secondary Heat Transfer.

MURO RD32 MARO 8 FW19 CCRS EFW Control Valves fail to operate, EF-V-52A-D Closed CARO (ARO: Establish PSHT via Condensate Booster Pump flow) 9 (if MU35B CCRS MU-P-1A/C will not start, MU-P-18 trips.

required)

CURO (URO: Establish PORV control for Heat Transfer)

(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 2 Op Test No.: 14-01 NRG Examiners: Operators:

Initial Conditions: * (Temporary IC-176)

100% Power, MOL

BS-P-1 A is OOS for maintenance, expected to return to service in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Crane work is occurring on the West side of the Plant to stage new piping Turnover: Maintain 100% Power Operations Critical Tasks:

Electrical Power Alignment (CT-8)

Turbine Trip (CT-18)

Protect against RCP Seal LOCA (CT-*)

Event No. Malt. No. Event Type* Event Description 1 Nl27A I CRS Pressurizer Pressure Instrument Fails High, entry into OP-TM-IURO MAP-G0106, OP-TM-MAP-G0107 IARO (URO: Blocks PORV, closes Spray Valve, Pressurizer Heater Control in Manual, ARO: "A" RPS to Manual Bypass) 2 IC12 CCRS Total RCS Flow IN Fails to Zero Volts, entry into OP-TM-AOP-070 CURO and 1102-4.

CARO (URO/ARO: ICS station to Manual, Stabilize Power) 3 03A3S09 TSCRS Loss of 1 E 4KV Bus, Entry into OP-TM-AOP-014

- CURO (URO: Manual control of Makeup valves. ARO: Restore Seal ZDl1SAE Injection)

CARO 2(1) 4 TU01D CCRS High Vibrations on Main Turbine, entry into OP-TM-MAP-K0201 RURO and 1102-4 NARO (URO/ARO: Power reduction with ICS in Manual) 5 EG04A I CRS Loss of Stator Coolant Pumps, Main Turbine fails to automatically EG04B I URO runback and trip (URO: Trip Reactor) 6 HVB-1-1 TSCRS Fire in EG-Y-1 B Room, entry into OP-TM-AOP-001 HVB-2-1 CARO A-1-4 (ARO: Secure EG-Y-1 B) 7 ED01 MCRS Loss of Offsite Power, entry into OP-TM-AOP-020.

MURO MARO 8 EG01A CCRS "A" EOG fails to start, SBO start required. (URO)

CURO

(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 3 Op Test No.: 14-01 NRC Examiners: Operators:

Initial Conditions: * (Temporary IC-177)

85% Power, MOL

BS-P-1 A is OOS for maintenance, expected to return to service in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Crane work is occurring on the West side of the Plant to stage new piping Turnover: Maintain 85% Reactor Power Critical Tasks:

Control HPI (CT-5)

Establish FW Flow and Feed SG(s) (CT-10)

Natural Circulation RCS Flow (CT-12)

Event No. Malt. No. Event Type* Event Description 1 DHR32 TSCRS BWST level lowers, entry into OP-TM-MAP-E0204 2 03A4S01 TSCRS Inadvertent ES Actuation, "B" Train (TS), entry into OP-TM-AOP-

- IURO 046 ZDIPB1R (URO: Defeats signal, ARO: Opens MU-V-2A/B)

CBON IARO 3 RC08B ICRS Tc Instrument Fails High, SASS Fails to Actuate, entry into OP-TM-IC51 AOP-070 IURO (URO: Manual control of Control Rods, ARO: Manual control of IARO Feedwater) 4 MU19 CCRS RC-P-1A #1 Seal Leak, leak at 6.5 gpm, Entry into OP-TM-AOP-040 RURO (URO/ARO: Power reduction in manual)

NARO 5 MU19 CCRS RC-P-1A #1 Seal Failure, leak at 10 gpm, Entry into OP-TM-AOP-040 CURO (URO: Secure RC-P-1A) 6 MS19A C CRS lsolable Steam Leak in Turbine Bldg, entry into OP-TM-AOP-051.

CARO (ARO: Isolate Steam Leak) 7 TH06 MCRS RCS LOCA, entry into OP-TM-EOP-001.

MURO MARO 8 CC06A CCRS NSCCW Rupture in RC-P-1A Motor Air Cooler, Loss of NSCCW, Reactor trip, entry into OP-TM-AOP-031, and OP-TM-EOP-001 CURO (URO: Reactor Trip IMA's) 9 ICR02 CCRS HSPS fails to feed OTSG's to 50%

ICR04 CARO (ARO: Feed OTSG's to >50% in manual)

(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Exelon Gene rat ion Three Mile Island Unit 1 Telephone 717-948-8000 Route 441 South, P.O. Box 480 Middletown, PA 17057 June 30, 2015 TMl-15-075 USNRC, Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMl-1)