ML12116A182
| ML12116A182 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 02/28/2012 |
| From: | Libra R Exelon Nuclear |
| To: | Operations Branch I |
| Jackson D | |
| Shared Package | |
| ML113350031 | List: |
| References | |
| Download: ML12116A182 (48) | |
Text
- -,/£' s'dl~
Exel n Three Mile Island Unit 1 Telephone 717-948-8000 Nuclear Route 441 South, P.O. Box 480 Middletown, PA 17057 TMI-12-012 U.S. NRC Region I Administrator *1 475 Allendale Road King of Prussia, PA 19406 Three Mile Island Unit I Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289
Subject:
Submittal of Integrated Initial License Training Examination Materials
-
Examination materials were submitted on February 23,2012, for TMI Unit 1, to support the Initial License Examination scheduled for the week of April 16, 2012, at TMI Unit 1.
The submittal included the Reactor Operator Written Examinations, Job Performance Measures, and Integrated Plant Operation Scenario Guides. The submittal also included the Senior Reactor Operator Written Examinations Job Performance Measures, and Integrated Plant Operation Scenario Guides.
The examination materials were developed in accordance with NUREG-1021 , Revision 9, Supplement 1 "Operator Licensing Examination Standards". Please note that reference materials are attached to each individual examination question or item.
Some minor modifications were made to the Integrated Examination Outline with regards to the operational scenarios in order to improve balance and content. Those changes improved examination quality and were in compliance with NUREG-1 021, Revision 9, Supplement 1, "Operator Licensing Examination Standards."
Some modifications or adjustments to the examination material might be required due to procedural changes.
In accordance with NUREG 1021, Revision 9, Supplement 1, Section ES-201, 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 at 717-948-8228. For questions concerning examination materials, please contact Greg Hoek at 717-948-2027.
Respectf~lIy (4~
R. W. Libra Site Vice President, Three Mile Island Unit I RWUmdf
Control Room Systems and Facility Walk-Through Job Performance Measures with references attached Administrative Topic Job Performance Measures with references attached Integrated Plant Operation Scenario Guides Completed Checklists:
Operating Test Quality Checklist (Form ES-301-3)
Simulator Scenario Quality Checklist (Form ES-301-4)
Transient and Event Checklist (Form ES-301-5)
Competencies Checklist (Form ES-301-6)
Written Exam Quality Checklist (Form ES-401-6)
Exam ination Security Agreements (Form ES-201-3)
Record of Rejected KlAs (Form ES-401-4) cc: (without attachments)
Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector - TMI-1 Operations Training Manager
T- 7s- f~dr~ffiI Exelon Three Mile Island Unit 1 Telephone 717-948-8000 Nuclear Route 441 South, P.O. Box 480 Middletown, PA 17057 01/17/2012 TIVII-12-002 U.S. NRC Region I Administrator 475 Allendale Road King of Prussia, PA 19406 Three Mile Island Unit 1 Facility Operating License DPR -50 NRC Docket No. 50-289 Supject: Submittal of Initial Operator Licensing Examination Outlines Enclosed are the examination outlines, supporting the Initial License Examination scheduled for April 16, 2012, at Three Mile Island Unit 1.
This submittal includes all appropriate Examination Standard forms and outlines in accordance with NUREG 1021, Revision 9, Supplement 1, "Operator Licensing Examination Standards".
In accordance with NUREG 1021, Revision 9, Supplement 1, 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 Greg Hoek, Exam Author, at (717) 948-2027.
at~-
GlenE~
Site Vice President, Three Mile Island Unit I GEC/mdf
Enclosures:
(Mailed to John Caruso, Chief Examiner, NRC Region I)
Examination Security Agreements (Form ES-201-3)
Administrative Topics Outlines (Form ES-301-1)
Control Room/In-Plant Systems Outline (Form ES-301-2)
PWR Examination Outline (Form ES-401-2)
Generic Knowledge and Abilities Outline (Tier 3) (Form ES-401-3)
Statement detailing method of Written Outline generation Scenario Outlines (Form ES-D-l)
Record of Rejected KlAs (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 - TMI Unit 1
Three Mile Island 2012 NRC Initial License Written Examination Written Examination Outline Methodology
- Western Technical services provided the outline to Three Mile Island Station.
The Exam Author at TMI is responsible for the SRO portion of the outline.
Reselections for the SRO portion were completed by manually loading provided outline into NKEG software with the above suppression, then rejecting and having NKEG reselect for SRO rejected topics.
ES-402 PWR Examination Outline Form ES-401-2 Facility: TMI Date of Exam: 4/2012 RO KIA Category Points SRO-Only Points Tier Group KKK KKK A A A A G Total A2
- Total 2 3 456 1 2 3 4 *
~--~----r-----~-4--~--
1.
, Emergency 18 3 3 6
& r----+--~-r--
9 2 2 4 Plant Evaluations 27 5 5 10 28 3 2 5 2.
10 0 2 3 Plant Systems 38 4 4 8
- 3. Generic Knowledge & Abilities f..-_1_-+-_2_-+-_3__ 10 2 3 4 7 332 2 2 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 (Le., 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.l.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) KiAs 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 KiAs that are linked to 10CFR55.43
ES-402 Form ES-401-2 TMI PWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KIA Topic(s)
AA2.02 - Ability to determine and interpret the following as they apply 062 I Loss of Nuclear Service.
to the Loss of Nuclear Service 3.6
- 76 Water! 4 Water: The cause of possible SWS loss AA2.02 Ability to determine and interpret the following as they apply 058 I Loss of DC Power I 6 3.6 77 to the Loss of DC Power: 125V de bus low/critical low, alarm AA2.08 - Ability to determine and interpret the following as they apply 065 / Loss of Instrument Air I 8 to the Loss of Instrument Air: 3.3 78 Failure Modes of air-operated 077 I Generator Voltage and 4.7 79
- Electrical Grid Disturbances! 6 054 I Loss of Main Feedwater / 4 4.6 80
- 038/ Steam Generator Tube 4.1 81
- Rupture i 3 E05 / Steam Line Rupture Excessive Heat Transfer / 4 x 3.8 39 007/ Reactor Trip /1 x 3.3 40 008/ Pressurizer Vapor Space Accident/3 x 3.1 41 029 I Anticipated Transient Without Scram (ATWS) /1 x 2.9 42
ES-402 Form ES-401-2 TMI PWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 AK2. 10
- Knowledge of the interrelations between the Reactor 015/17/ Reactor Coolant Pump Coolant Pump Malfunctions (Loss 2.8 43 Malfunctions / 4 of RC Flow) and the following:
RCP indicators and controls EK2.02 . Knowledge of the 011 / Large Break LOCA /3 X interrelations between the following Break LOCA: Pu EK3.01 Knowledge of the reasons for the following 055 / Station Blackout / 6 X respanses as the apply to the Station Blackout: Length of time for which EK3.4 - Knowledge of the reasons for the following responses as they apply to the (Inadequate Heat Transfer) RO or SRO function E04 I Inadequate Heat Transfer* within the control room team as X 3.5 46 Loss of Secondary Heat Sink / 4 appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.
EK3.12 Knowledge of the reasons for the following 009/ Small Break LOCA / 3 X 3.4 47 responses as the apply to the small break LOCA: Letdown isolation AA1.02 Ability to operate and / or monitor the following as they apply 054 / Loss of Main Feedwater / 4 to the Loss of Main Feedwater 4.4 48 (MFW): Manual startup of electric AFW AA1.04 - Ability to operate and / or monitor the following as they apply to the Loss of Offsite Power:
056/ Loss of Off-site Power / 6 3.2 49 Adjustment of speed of ED/G to maintain frequency and voltage levels AA1.01 - Ability to operate and / or monitor the following as they apply 027 / Pressurizer Pressure to the Pressurizer Pressure 4.0
- Control System Malfunction / 3 Control Malfunctions: PZR heaters, AA2.02 - Ability to determine and interpret the following as they apply 026 I Loss of Component Cooling to the Loss of Component Cooling 2.9 Water 18 Water: The cause of possible CCW loss EA2.11
- Ability to determine or 038/ Steam Generator Tube interpret the following as they apply 3.7 Rupture /3 to a SGTR: Local radiation reading :
on main steam
ES-402 Form ES-401-2 TMI PWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 065/ Loss of Instrument Air /8 077 / Generator Voltage and Electric Grid Disturbances 025 / Loss of Residual Heat Removal System / 4 062 / Loss of Nuclear Service.
Water / 4 KJA Category Totals
ES-402 Form ES-401-2 TMI PWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function I K1 I K2 I K3 I A1 I A2 I G KIA Topic(s) limp. I Q# I AA2.1 - Ability to determine and interpret the following as they apply to the (Shutdown Outside Control A06 I Control Room Evac. / 8 room) Facility conditions and 4.2 82 selection of appropriate procedures during abnormal and emergency AA2.09 - Ability to determine and interpret the following as they apply to the Steam Generator Tube 037 I Steam Generator Tube Leak: System status, using 3.4 83 Leak / 3 independent readings from redundant Condensate air ejector exhaust monitor 2.1.23 - Loss of NNI-Y - Ability to perform specific system and A03 ! Loss of NNI-Y /7 4.4 84 integrated plant procedures during all modes of n.
2.2.38 - Inoperable/Stuck Control 005 I Inoperable/Stuck Control Rod - Knowledge of conditions and 4.5 85 Rod /1 limitations in the license.
AK1.01 - Knowledge of the operational implications of the 028 / Pressurizer Level Control following concepts as they apply X 2.8 57 Malfunction / 2 to Pressurizer Level Control Malfunctions: PZR reference leak abnormalities AK2.2 - Knowledge of the interrelations between the (Flooding) and the following:
Facility's heat removal systems, including primary coolant, A07 / Flooding / 8 X 3.3 58 emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the AK3.3 - Knowledge of the reasons for the following responses as they apply to the (Shutdown Outside A06 / Control Room Evac. / 8 X Control Room) : Manipulation of 4.2 59 controls required to obtain desired operating results during abnormal, situations.
AA 1.07 - Ability to operate and / or 001 / Continuous Rod Withdrawal monitor the following as they apply X 3.3 60
/1 to the Continuous Rod Withdrawal: RPI AA2.02 - Ability to determine and interpret the following as they apply 051 / Loss of Condenser Vacuum 61 to the Loss of Condenser Vacuum: 3.9
/4 Conditions requiring reactor and/or turbine tri
ES-402 Form ES-401-2 TMI PWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 061 / Area Radiation Monitoring (ARM) System Alarms / 7 060 / Accidental Gaseous 63 RadWaste Release / 9 E09 / Natural eirc. / 4 x 3.8 64 A04/ Turbine Trip / 4 KIA Category Totals 2 Group Point Total:
ES-402 Form ES-401-2 TMI PWR Examination Outline Plant Systems - Tier 2 Group 1 A2.10 - Ability to (a) predict the impacts of the following malfunctions or operations on the and (b) based on 006 Emergency Core those predictions, use Cooling procedures to correct, control, 3.9 86 or mitigate the consequences of those malfunctions or operations: Low boron concentration in SIS A2.01 -Abilityto the impacts of the malfunctions or operations on the PZR PCS; and (b) based 010 Pressurizer on those predictions, use 3.6 87 Pressure Control procedures to correct, control, or mitigate the consequences of those malfunctions or
- Heater failures 2.2.12 - Equipment Control:
012 Reactor Protection Knowledge of surveillance 4.1 88 2.4.21 Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as 026 Containment Spray reactivity control, core cooling 4.6 and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based 013 Engineered Safety Ability on those predictions.
4.2 Features Actuation use procedures to correct, control. or mitigate the consequences of those malfunctions or operations:
Loss of instrument bus.
K1.01 - Knowledge of the physical connections and/or cause effect relationships 012 Reactor Protection X 3.4 between the RPS and the following systems: 120V 061 Auxiliary/Emergency X 3.4 2 Feedwater
ES-402 Form ES-401-2 TMI PWR Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s)
K2.02 - Knowledge of bus 006 Emergency Core power supplies to the Cooling x following: Valve operators for 2.5 3 accumulators K2.04 - Knowledge of bus power supplies to the 076 Service Water x following: Reactor building 2.5 4 closed cool water K3.01 - Knowledge of the effect that a loss or 064 Emergency Diesel malfunction of the ED/G Generator x system will have on the 3.B 5 following: Systems controlled automatic loader K3.03 - Knowledge of the effect that a loss or 013 Engineered Safety Features Actuation x malfunction of the ESFAS will 4.3 6 have on the following:
Containment K4.02 - Knowledge of dc electrical system design feature(s) and/or interlock(s) 063 DC Electrical Distribution x which provide for the 2.9 7 following: Breaker interlocks, permissives, bypasses and cross-ties.
K4.01 - Knowledge of CCWS design feature(s) and/or OOB Component Cooling Water x interlock(s) which provide for 3.1 B the following: Automatic start of K5.02 - Knowledge of the operational implications of the 007 Pressurizer following concepts as the Relief/Quench Tank x apply to PRTS: Method of 3.1 9 forming a steam bubble in the PZR K5.01 - Knowledge of the operational implications of the following concepts as the 010 Pressurizer Pressure Control x apply to the PZR PCS: 3.5 10 Determination of condition of fluid in PZR, using steam tables K6. 17 - Knowledge of the operational implications of the 004 Chemical and Volume Control x following concepts as they 4.4 11 apply to the CVCS: Flow paths for boration K6.03 - Knowledge of the effect of a loss or malfunction 005 Residual Heat Removal x on the following will have on 2.5 12 the RHRS: RHR heat
ES-402 Form ES-401-2 TMI PWR Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s)
A 1.02 - Ability to predict and/or monitor changes in parameters (to prevent 003 Reactor Coolant exceeding design limits) 2.9 13 Pump associated with operating the RCPS controls including: RCP pump and motor bearing tem A1.01 - Ability to predict and/or monitor changes in parameters (to prevent 022 Containment exceeding design limits) 3.6 14 Cooling associated with operating the CCS controls including:
A2.05 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-and (b) based on those 103 Containment 2.9 15 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations containment A2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use 026 Containment Spray 3.9 16 procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of spray 078 Instrument Air 17 062 AC Electrical 3.0 18 Distribution 039 Main and Reheat 2.8 19 Steam 073 Process Radiation 3.9 20 Monitoring 059 Main Feedwater 4.1 21
ES-402 Form ES-401-2 TMI PWR Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s) 2.4.21 - Emergency Procedures / Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as 078 Instrument Air reactivity control, core cooling 4.0 22 and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
A 1.03 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) 062 AC Electrical associated with operating the 2.5 23 Distribution ac distribution system controls including: Effect on instrumentation and controls lies A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on 008 Component Cooling those predictions, use 3 Water procedures to correct, control, 24
- or mitigate the consequences of those malfunctions or operations: High/low CCW A3.02 Ability to monitor 039 Main and Reheat automatic operation of the 3.1 25 Steam MRSS, including: Isolation of K3.01 - Knowledge of the 006 Emergency Core effect that a loss or 4.1 26 Cooling malfunction of the ECCS will have on the ReS K4.02 - Knowledge of ED/G system design feature(s) 064 Emergency Diesel and/or inter-Iock(s) which 3.9 27 Generator provide for the following: Trips for ED/G while operating 2.1.7 - Conduct of Operations:
Ability to evaluate plant performance and make 005 Residual Heat operational judgments based 28 Removal on operating characteristics, reactor behavior, and instrument int'>rnrot"ti",n KIA Category Totals Group Point Total:
ES-401 Form ES-401-2 TMI PWR Examination Outline Plant Systems Tier 2 Group 2 KIA Topic(s)
A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on 016 Non-Nuclear those predictions, use 3.1 91 Instrumentation System procedures to correct, control, or mitigate the consequences of those malfunctions or
" ...,,,,,,;f''''* Detector failure 2.2.40 - Equipment Control:
068 Radwaste Ability to apply technical 4.7 92 2.1.20 - Waste Gas Disposal 071 Waste Gas Disposal System (WGDS): Ability to 4.6 93 System interpret and execute ste .
A 1.03 - Ability to predict and/or monitor changes in parameters (to prevent 002 Reactor Coolant exceeding design limits) 3.7 29 associated with operating the RCS controls including:
K1.02 Knowledge ofthe physical connections and/or 014 Rod Position Indication x cause-effect relationships 3.0 30 between the RPIS and the f"lI,mAt_.inn .,,,,,,t,,,,..,,,,* NIS 041 Steam Dump!Turbine Bypass 4.6 31 Control 075 Circulating Water x 2.6 32 A2.09 - Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System; and (bl based on 071 Waste Gas Disposal those predictions, use 3.0 33 procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Stuck-open relief valve K4.03 Knowledge of design feature(s) and/or interlock(s) 029 Containment Purge x which provide for the 3.2 following: Automatic purge isolation
ES-401 Form ES-401-2 TMI PWR Examination Outline Plant Systems - Tier 2 Group 2 System #/Name KJA Topic(s)
K3.02 - Knowledge of the effect that a loss or 072 Area Radiation malfunction of the ARM 3.1 35 Monitoring system will have on the following: Fuel handling K6.04 - Knowledge of the effect of a loss or malfunction on the Fire Protection System 086 Fire Protection 2.6 36 following will have on the:
Fire, smoke, and heat A3.04 - Ability to monitor automatic operation of the 015 Nuclear NIS, including: Maximum 3.3 37 Instrumentation disagreement allowed between channels A4.01 - Ability to manually 034 Fuel Handling operate and/or monitor in the 3.3 38 Equipment control room: Radiation levels KJA Category Totals Group Point Total: 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 identify and interpret diverse indications 2.1.45 4.3 66 to validate the response of another indicator.
~bility to interpret reference materials, such as 3.9 67 raphs, curves, tables, etc.
2 1 36 Knowledge of procedures and limitations 3.0 74
. . involved in core alterations.
- 1. Conduct of Knowledge of new and spent fuel movement Operations 2.1.42 3.4 94 I procedures.
Ability to evaluate piant performance and make operational judgments based on operating 2.1.7 4.7 98 characteristics! reactor behavior, and instrument interpretations.
......*
Subtotal 3 2 Knowledge of tagging and clearance 2.2.13 4.1 68 I procedures.
Knowledge of the process for managing maintenance activities during shutdown 2.2.18 2.6 69 operations, such as risk assessments, work prioritization, etc.
Knowledge of less than or equal to one hour 2.2.39 technical specification action statements for 3.9 75
- 2. Equipment systems. I Control
~
Knowle?ge of pre- and post-maintenance 2.2.21 operability requirements.
y+; ..........
Subtotal 3 ...*.. 1
ES-401 Generic Knowleclge and Abilities Outline (Tier 3) Form ES-401-3 Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, 2.3.13 3.4 70 containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aliQninQ filters, etc.
Knowledge of radiation or contamination 2.3.14 hazards that may arise during normal, abnormal, 3.4 71 or emerQency conditions or activities.
- 3. Radiation Control 2.3.11 Ability to control radiation releases. 4.3 96 Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, 2.3.13 containment entry requirements, fuel handling 3.8 99 responsibilities, access to locked high radiation areas, aligning filters, etc.
Subtotal 2 2 Ability to diagnose and recognize trends in an 2.4.47 accurate and timely manner utilizing the 4.2 72 appropriate control room reference material.
Knowledge of EOP entry conditions and 2.4.1 4.6 73 immediate action steps.
- 4. Emergency Procedures /
2.4.27 Knowledge of "fire in the plant" procedures. 3.9 97 Plan Knowledge of emergency action thresholds and 2.4.41 4.6 100 classifications.
Subtotal 2 2 Tier 3 Point Total: 10 7
ES-401 Record of Rejected KIA's Form ES-401-4 Randomly Selected Tier / Group Reason for Rejection KA 077 / 2.4.3 replaced 1/1 The subject KIA isn't relevant at the subject facility.
by 077 / 2.2.37 071 / A2.07 replaced 2/2 The subject KIA isn't relevant at the subject facility.
by 071 / A2.09 059/2.4.3 replaced 2/1 The subject KIA isn't relevant at the subject facility.
by 059 / 2.4.45 056/ M1.12 1/ 1 replaced by 056 / 056/M1.12 overlaps with a JPM on the Operating Exam.
M1.04 008/ A2.02 replaced There is an overlap issue between the KA selected and a JPM 2/1 by 008 / A2.03 on the Operating Test 064/ K2.02 replaced It isn't possible to prepare a psychometrically sound question 2/1 by 064 / K4.02 related to the subject KIA.
015/017 / AA2.02 Could not write SRO level question to 2.02, also M2.02 over 1/1 SRO replaced by 015/017 sampled, randomly chosen from same system.
M2.08 025/2.1.27 replaced Could not write SRO level question to 2.1.27 for evolution 025, 1/1 SRO by 025/2.4.1 randomly chose from generics against same evolution.
E05 / 2.1 .32 replaced No system limits and precautions associated with this evolution, 1/1 SRO by 054/2.2.37 randomly chose new KIA from NKEG exam outline generator.
059/2.4.8 replaced No EOP for accidental liquid release, randomly chose new KIA 1/2 SRO by A03 /2.1 .23 from NKEG exam outline generator.
Could not write test question to low power turbine trip affect on A04/ 2.4.9 replaced 1/2 SRO mitigation strategy, randomly chose new KIA from NKEG exam by 005/2.2.38 outline generator.
004/ A2.04 replaced No accidental gas release associated with CCVC, randomly 2/1 SRO by 006 / A2.1 0 chose new KIA from 1~I5J:::G exam outline Qenerator.
059/2.4.47 replaced Could not write E-plan related question to MFW trending, 2/1 SRO by 013 / A2.04 randomly chose new KIA from NKEG exam outline generator.
,., 056 / A2.04 replaced Could not write SRO level question, randomly chose new KIA by 016 A2.01 from NKEG exam outline generator.
Could not write operability of safety system question against 045/2.2.37 replaced 2/2 SRO turbine, randomly chose new KIA from NKEG exam outline by 017 I A2.02 Qenerator.
017/ A2.02 again Could not write a discriminating question, randomly chose new 0 071/2.1.20 KIA again.
015/017 AA2.08 Could not write a discriminating SRO level question. Randomly 1/1 SRO replaced by 065 /
picked a new KIA.
M2.08 025/2.4.4 again Could not write a discriminating SRO level question. Randomly
- 1/1 SRO replaced by 077 /
chose a new evolution in the tier.
2.4.4 2.1.14 replaced by Could not write a discriminating question at the SRO level.
3SRO 2.1.7 Randomly chose new KIA from tier 1.
2.4.35 replaced by Could not write a discriminating question at the SRO level.
3SRO 2.4.41 Randomly chose a new KIA from tier 4
- ES*301 Administrative Outline Form ES*301*1 Facility: Three Mile Island Date of Examination: April 2012 Examination Level: RO ~ SRO 0 Operating Test Number: 289-2012-301 Administrative Topic Type Describe activity to be performed (See Note) Code*
Perform a Manual Power Range Calculation Conduct of Operations N/R 2.1.37 (4.3)
Perform a Transient Leak Rate Calculation Conduct of Operations M/R 2.1.23 (4.3)
Isolate a Component for Maintenance Equipment Control N/R 2.2.41 (3.5)
Calculate Dose Limit Stay Times Radiation Control M/R 2.3.4 (3.2)
Emergency Procedures/Plan n/a Category not selected for RO applicants 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 (2 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): Perform a Manual Power Range Calculation.
Given a data sheet and reference 1302-1.1, Power Range Calibration, the candidate will be directed to manually check Power Range Calibration using hand calculations.
This JPM is a new ...1PM.
License is evaluated against properly calculating Power Range Calibration and identifying an Error Linear Power greater than 2.IT%.
Safety significance, failure to identify an Error Linear Power >2.0% would result in continued operation with Offset Error outside of the acceptance criteria lAW T.S. 4.1.
CONDUCT OF OPERATIONS (A1-2): Perform a Transient Leak Rate Calculation.
Given plant conditions and reference OS-24 Conduct of Operations During Abnormal and Emergency Events, the candidate will be directed to perform a transient RCS leak rate calculation that will most accurately determine the current RCS leak rate.
JPM is modified from a Bank JPM.
License is evaluated against properly calculating an RCS leak rate when given multiple data points.
Several opportunities for error exist In: multiple aata pOints of leak rate (leak worsens at one point),
multiple times given (5 minute minimum for most accurate), and calculation errors.
Safety significance, failure to calculate an accurate RCS leak rate could cause a lower than realistic rate and redirect a Control Room crew away from appropriate Procedures lAW T.S.'s.
EQUIPMENT CONTROL (A2): Isolate a Component for Maintenance.
Given a plant component needing to be isolated identify mechanical and electrical isolation points. This is a new JPM developed for this class.
Safety significance is failure to properly identify the correct points could lead to a loss of nuclear services river water and/or personnel injury.
RADIATION CONTROL (A3): Calculate Dose Limit Stay Times.
Given plant conditions, a dose history, and references RP-M-460 Controls For High and Very High Radiation Areas, EP-M-112-100-F-01, Shift Emergency Director Checklist, and EP-M-113. Personnel Protective Actions, the candidate is directed to determine maximum stay time for performing a valve operation without exceeding the limit approved by the TSC Radiation Protection Manager .
...IPM is modified from a Bank JPM.
License is evaluated against properly identifying the maximum increased dose exposure limit and calculating stay time ta1<ing into account current exposure. Failure to correctly identify stay time could result in a dose limit being exceeded.
EMERGENCY PROCEDURES/PLAN (A4): Category not selected for RO Candidates.
ES-301 Administrative Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: April 2012 Examination Level: RO D SRO [?SI Operating Test Number: 289-2012-301 Administrative Topic Type I Describe activity to be performed (See Note) Code*
Ii I!
Perform and Approve a Manual Power Range Calculation Conduct of Operations N/R 2.1.37 (4.6)
Perform a Transient Leak Rate Calculation with a T.S. Call Conduct of Operations M/R 2.1.23 (4.4)
Evaluate a Completed Surveillance Procedure and Equipment Control M/R Perform Appropriate Actions 2.2.12 (4.1)
~
Review and Approve a Gaseous Release Permit for a Radiation Control N/R Waste Gas Tank 2.3.6 (3.8)
Identify and Declare an Emergency Classification with a Emergency Procedures/Plan M/R PAR 2.4.41 (4.6)
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; ~ 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 2012 NRC SRO EXAMINATION CONDUCT OF OPERATIONS (A1-1): Perform and Approve a Manual Power Range Calculation.
Given a data sheet and reference 1302-1.1, Power Range Calibration, the candidate will be directed to manually check and approve Power Range Calibration using hand calculations.
License is evaluated against properly calculating Power Range Calibration, identifying several errors in the given data sheet, and not approving the hand calculation.
Safety significance, failure to identify the errors would result in continued operation with Power Range instrumentation outside of the acceptance criteria lAW T.S. 4.1.
CONDUCT OF OPERATIONS (A1-2): Perform a Transient Leak Rate Calculation with a T.S. Call.
Given plant conditions and reference OS-24, Conduct of Operations During Abnormal and Emergency Events, the candidate will be directed to perform a transient RCS leak rate calculation that will most accurately determine the current RCS leak rate and identify any T.S.
JPM is modified from a Bank JPM.
License is evaluated against properly calculating an RCS leak rate when given multiple data points.
Several opportunities for error exist In: multiple data points of leak rate (leak worsens at one point).
multiple times given (5 minutes for most accurate), and calculation errors, and identifying the proper T.S.
Safety significance, failure to calculate an accurate RCS leak rate could cause a lower than realistic rate and redirect a Control Room crew away from appropriate Procedures and EAL's lAW T.S.'s.
EQUIPMENT CONTROL (A2): Evaluate a Completed Surveillance Procedure and Perform Appropriate Actions.
Given plant conditions, a data sheet, and reference ER-TM-321-1041, TMI-1 1ST Program Requirements, the candidate will be directed to evaluate a completed surveillance procedure and perform appropriate actions.
JPM is modified from a Bank JPM.
License is evaluated against properly reviewing the data sheet against ER-TM-321-1 041, and identifying out-of-spec data points.
Safety significance, failure to identify out-of-spec data points would lead to unknown violation of Tech Specs and could lead to possible equipment damage and/or personnel injury.
RADIATION CONTROL (A3): Review and Approve a Gaseous Release Permit for a Waste Gas Tank.
Given plant conditions, data sheets, and reference 661 0-ADM-4250.11, Releasing Radioactive Gaseous Effluents - Waste Gas Tanks AlB/C, the candidate is directed to review and approve a filled-out gaseous release permit.
License is evaluated against properly reviewing the given data against 661 0-ADM-4250.11, and identifying incorrect data points.
Safety significance, failure to identify incorrect data points would lead to a gaseous release with radiation above Tech Spec allowed levels ana possible adverse effects to the environment.
EMERGENCY PROCEDURES/PLAN (A4): Identify and Declare an Emergency Classification.
Given a set of conditions, and references EP-AA-112-1 OO-F-01 ~ Shift Emergency Director Checklist, EP AA-1009 Exelon Nuclear Radiological Emergency Plan Annex r-or Three Mne Isrand (TMI) Station, EP MA-114-1 OO-F-01, State/Local Event Notification Form, EP-AA-111, Emergency Classification And Protective Action Recommendations, and EP-AA-111-F-09, TMI Plant Based PAR Flowchart, the candidate is directed to determine the Emergency Action Level (EAL) and make a Protective Action Recommendation (PAR) lAW the facility Emergency Plan .
..IPM is modified from a Bank JPM.
License is evaluated against properly identifying the Emergency Classification and Protective Action Recommendations (PAR). Failure to correcfly laentify the Emergency Classification and Protective Action Recommendations could result in unnecessary harm to the general public.
ES-301 Control Room/ln~Plant Systems Outline Form ES*301*2 II Facility: Three Mile Island Date of Examination: April 2012
! Exam Level: RO ~ SRO-I D SRO-U D Operating Test Number: 289-2012-301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
Safety System I JPM Title Type Code*
Function
- a. Recover From CRD Sequence Fault (Sys 001) A2.18 N/S 1
- b. Respond To High Pressure Injection Initiation (Alt Path MU-V-14A D/A/LIS 2
~ails To OpenL(Svs 006) A2.02
- c. Respond to an ReS leak into ICCW (EPE 009) EA2.02 N/A/S 3
- d. RCP #1 Seal Failure (Sys 003) A2.01 I P/A/S 4P
- e. Perform the Required Actions for EF-P-1 Trip (APE 054) AA1.02 DillS 4S
- f. Place an RPS Cabinet in Manual Bypass (Sys 012) A4.03 N/A/S 7
- g. OP-TM-EOP-020 IMA's (APE 068) AA1.23 N/S 8
- h. Establish Alternate RB Emergency Cooling (Sys 022) A4.01 N/A/LIS 5 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
-
- i. Supply VBC From the 1E Inverter (Sys 062) A4.01 N 6
-
Place 8th Stage Heating On-Line (Sys 039) G2.1.30 0 4S
! k. Prepare for Transfer to RB Sump Recirculation (Sys 006) K4.08 D/E/L/R 3
@ 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-II SRO-U I (A) Itemate path 4-61 4-6 I 2-3
- {C}ontrol room
- (D)irect from bank 59/ 58 / 54
- (E}mergency or abnormal in-plant 2=, 1 I 2=,1 / 2=, 1 (EN)gineered safety feature - / - / 2=, 1 (control room system (L)ow-Power I Shutdown 2=, 1 I 2=,1 / > 1 (N)ew or (M)odified from bank including 1(A) 2=,2/ 2=,2 I2=,1
- (P}revious 2 exams 531 53 / 5 2 (randomly selected)
(R}CA 2=, 1 I 2=,1 / 2=, 1 (S}imulator ES-30 1, Page 23 of 27
THREE MILE ISLAND 2012 NRC RO EXAMINATION JPM A - Recover From CRD Sequence Fault. New.
Safety significance failure to correct sequence fault allows operation outside analyzed reactivity addition rates.
JPM B - Respond To High Pressure Injection Initiation (Alt Path - MU-V-14A Fails To Open). Bank Alternate path.
Safety significance failure to complete this JPM will result in a loss a" three HPI pumps.
JPM C - Respond to an RCS leak into ICCW. New Alternate path, alternate path leak gets worse, IAAT actions required. .
Safety significance failure to properly respond will lead to a continued RCS leak outside containment.
JPM D - RCP #1 Seal Failure. Previous NRC 2011 Alternate path "IPM. Alternate path failure worsens, pump must be shut down.
Safety significance failure to properly address excessive seal leakoff could result in Seal LOCA. Chosen randomly by drawing playing cards representing JPMs from last two exams.
JPM E - Perform the Required Actions for EF-P-1 trip. Bank JPM.
Safety significance failure to p'roperly complete the task would lower the safety margin for EFW by having only motor driven pumps available.
JPM F - Place an RPS cabinet in Manual Bypass. New A~ternate 5 path. Alternate path failure in another cabinet requires placing cabinet in tripped state to place 1 cabinet in Manual Bypass.
Safety significance with a failed instrument in the other cabinet, incorrect operation in this cabinet would lead to reactor trip, an initiating event for transients.
JPM G - OP-TM-EOP-020 IMA's New.
Safety significance failure to complete the IMA's of the EOP could result in failure to adequately control and transfer control of the reactor to the remote shutdown area.
JPM H - Establish Alternate RB Emergency Cooling. New Alternate path.
Safety significance failure to properly complete the task would result in containment temperatures greater than assumed in the structural analysis, see Technical Specification 3.17. Alternate path involves the complete failure of the emergency cooling systems and restoration of a portion of normal cooling under alternate power conditions.
JPM I - Supply VBC from the 1E inverter. New In Plant JPM.
Safety significance failure to complete the task properly could result in inadvertent safety system actuations, or other transients (accident initiators) due to reduction of Vital Power supplies.
th JPM J - Place 8 Stage Heating On-Line. Bank JPM.
Safety significance failure to complete the task properly could result in water quality reduction to the OTSGs, fong term result could be damage to tubes.
JPM K - Prepare for Transfer for RB Sump Recirculation. Bank "IPM.
Safety significance failure to complete the task may result in inability to complete these actions at a later time due 10 inaccessibility due to radiation levels post accident on su~ recirculation. This is a time critical action, approximately 25 minutes from start of event reference UFSAR 14.2.2.5.d and 6.1.3.2.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: April 2012 Exam Level: RO D SRO-I [gI SRO-U D Operating Test Number: 289-2012-301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
Safety System 1J PM Title Type Code*
Function rom CRD Sequence Fault (Sys 001) A2.18 N/S 1
- b. Respond To H~gh Pressure Injection Initiation (Alt Path - MU-V-14A DINL/S 2 Fails To Open). (Sys 006) A2.02
- c. Respond to an RCS leak into ICCW (EPE 009) EA2.02 N/NS 3
- d. RCP #1 Seal Failure (Sys 003) A2.01 PINS 4P
- e. Perform the Required Actions for EF-P-1 Trip (APE 054) AA1.02 DillS 4S
- f. Place an RPS Cabinet in Manual Bypass (Sys 012) A4.03 N/NS I
- g. OP-TM-EOP-020 IMA's (APE 068) AA1.23 N/S 8 h.
n,~nt Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. Supply VBC From the 1 E Inverter (Sys 062) A4.01 N 6 I
th t+/-j
- j. Place 8 Stage Heating On-Line (Sys 039) G2.1.30 D
@ 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 / SRO-II SRO-U (A)lternate path 4-6/ 4-6 / 2-3 (C)ontrol room (D)irect from bank 5,91 58 I 54 (E)mergency or abnormal in-plant ~1I ~1 I ~1 (EN)gineered safety feature - / I ~ 1 (control room system (L)ow-Power / Shutdown ~ 1/ ~1 I ~ 1 (N)ew or (M)odified from bank including 1 (A) ~21 ~2 / ~ 1 (P)revious 2 exams 53/ 53 I 52 (randomly selected)
(R)CA ~ 1 I ~1 I ~ 1 (S)imulator ES-301, Page 23 of 27
THREE MILE ISLAND 2012 NRC SRO EXAMINATION JPM A - Recover From CRD Sequence Fault. New.
Safety significance failure to correct sequence fault allows operation outside analyzed reactivity addition rates.
JPM B - Respond To High Pressure Injection Initiation (Alt Path - MU-V-14A Fails To Open). Bank Alternate path.
Safety significance failure to complete this JPM will result in a loss all three HPI pumps.
JPM C - Respond to an RCS leak into ICCW. New Alternate path, alternate path leak gets worse, IAAT actions required.
Safety significance failure to properly respond will lead to a continued RCS leak outside containment.
JPM D - RCP #1 Seal Failure. Previous NRC 2011 Alternate path JPM. Alternate path failure worsens, pump must be shut down.
Safety significance failure to properly address excessive seal leak off could result in Seal LOCA. Chosen randomly by drawing playing cards representing JPMs from last two exams.
JPM E - Perform the Required Actions for EF-P-1 trip. Bank JPM.
Safety significance failure to properly complete the task would lower the safety margin for EFW by having only motor driven pumps available.
JPM F - Place an RPS cabinet in Manual Bypass. New A~ternate path. Alternate path failure in an other cabinet requires placing cabinet in tripped state to place 1S cabinet in Manual Bypass.
Safety significance with a failed instrument in the other cabinet, incorrect operation in this cabinet would lead to reactor trip, an initiating event for transients.
JPM G - OP-TM-EOP-020 IMA's New.
Safety significance failure to complete the IMA's of the EOP could result in failure to adequately control and transfer control of the reactor to the remote shutdown area.
JPM H - Not selected for SROs.
JPM I - Supply VBC from the 1E inverter. New In Plant JPM.
Safety significance failure to complete the task properly could result in inadvertent safety system actuations, or other transients (accident initiators) due to reduction of Vital Power supplies.
th JPM J - Place 8 Stage Heating On-Line. Bank JPM.
Safety significance failure to complete the task properly could result in water quality reduction to the OTSGs, rong term result could be damage to tubes.
JPM K - Prepare for Transfer for RB Sump Recirculation. Bank JPM.
Safety significance failure to complete the task may result in inability to complete these actions at a later time due to inaccessibility due to radiation levels post accident on sup recirculation. This is a time critical action, approximately 25 minutes from start of event reference UFSAR 14.2.2.5.d and 6.1.3.2.
nn<:,nn.v D Scenario Outline Form ES-D-Facility: Three Mile Island Scenario No.: Op Test No.: 10-02 NRC Examiners: Operators:
Initial Conditions: * (Temporary IC-231)
- 100% Power, MOL MO-P-1 C and MO-P-1 F are OFF for Chemistry purposes lAW OP-TM-431-403/406
- Crane work is occurring on the West side of the Plant to stage new piping r-----------------------.
Turnover: Maintain 100% Reactor Power
! Critical Tasks:
- Control RCS Inventory (CT-30)
Event No. Malf. No. i Event Type* Event Description 1 RW02C TSCRS NR-P-1C Trips, NR-P-1 B Fails to Auto-Start, entry into OP-TM MAP-B01 05, and OP-TM-MAP-B020S CARO (ARO: Starts NR-P-1 B from CR) 2
- ED22G ICRS ICS Auto Power ICCW Subfeed Failure, entry into OP-TM-MAP-H i 0108 IARO (ARO: Restores Letdown following a Loss of ICS AUTO Power) 3 I EGR30 TSCRS Loss of EG-Y-1A Starting Air, entry into OP-TM-MAP-A0102, and OP-TM-MAP-A0201
- RURO i
(URO: Reduce Reactor Power)
I 4 IC09 ICRS MW Generated Input Fails to Zero Volts, entry into OP-TM-AOP IC53 070 IURO (URO/ARO: Control ICS in Manual lAW AOP-070)
! IARO 5 ED12 CCRS Loss of ICS Hand and Auto Power, entry into OP-TM-AOP-02S, and I ED13 CURO OP-TM-EOP-001 (URO: Reactor Trip IMA's, ARO: Control OTSG Pressures)
CARO 6
M URO
! MARO 7 ZDISSM CCRS MU-V-37 Fails Closed UV37(1) CURO (URO: Throttle an MU-V-16 for minimum MU flow)
- (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.: 2 Op Test No.: 10-02 NRC Examiners: Operators:
Initial Conditions: * (Temporary IC-232)
- 100% Power, MOL
- MO-P-1 C and MO-P-1 F are OFF for Chemistry purposes lAW OP-TM-431-403/406
- Crane work is occurring on the West side of the Plant to stage new piping
- "e" RPS Cabinet is Maintain 100% Power Operations
- Turbine Trip (CT-18)
I
- Restore Feed to a Dry OTSG (CT-26) i i Event No. Malt. No. Event Type* I Event Description I 1 ZAIRC1L1C CCRS MU-V-17 Fails Closed in Auto, entry into OP-TM-211-472 I CURO (URO: Controls Pressurizer Level with MU-V-17 in Manual) 2 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, Auxiliary Building, Fuel Handling Building i Ventilation) 3 ' NI15B TSCRS Nuclear Instrument, NI-6, Failure (TS)
I 4 IC23 I CRS SG/RX Demand Station fails to 0 Volts, Entry into OP-TM-AOP-070 !
,llIRO i (URO: ICS station to Manual, ARO: Controls temperature with SG i
, A & B FW DEMAND stations in Manual)
IARO 5 TU01D NCRS I High Vibrations on Main Turbine, entry into OP-TM-MAP-K0201 and 1102-4, Reactor shutdown RURO
, NARO (URO/ARO: Power reduction with ICS in Manual) 6 i FW15A CCRS Loss of both Main Feedwater Pumps, Turbine fails to OP-TM-EOP-001
- FW15B i CURO (URO: IMA's of OP-TM-EOP-001)
TC02 7 FW17 i MCRS Loss of Emergency Feedwater Pumps, entry into OP-TM-EOP-004, Lack of Heat Transfer.
FW18A M lIRO FW18B MARO 8 MS09A-F CCRS Turbine Bypass Valves fail Closed, OTSG Pressure control via Atmospheric Dump Valves CARO
- (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.: 4 Op Test No.: 10-02 NRC Examiners: Operators:
Initial Conditions: * (Temporary IC-234)
- 100% Power, MOL
- MO-P-1 C and MO-P-1 F are OFF for work is occurring on the West
- MU-P-1 C is OOS lAW OP-TM-211-432, Removing MU-P-1 C From Service, for bearing replacement
- ICS is in Manual due to a faulted Reactor Demand circuit card, expected to be replaced within 24 hrs Turnover: Maln. tam' 1DooA Power
- ° i
. Critical Tasks:
- Trip all RCPs (CT-1)
~.
- Minimize SCM (CT-7)
- Maintain SG availability (CT-29)
Event No. Malf. No. Event Type* Event Description 1 TH15A TSCRS 20 gpm tube leak on "An OTSG, entry into OP-TM-EOP-005.
RURO (lIRO: commences a reactor shutdown with ICS in Manual, ARO:
Place FW-P-1N18 in HAND)
L i NARO i i 2 I ES08A ICRS ! Inadvertent 500# ESAS Signal, entry into OP-TM-AOP-046.
i IURO i (URO: OP-TM-AOP-046IMA's. ARO: Restores Letdown following an inadvertant ESAS signal)
IARO 3 ED08S TSCRS : Loss of "8" DC, entry into OP-TM-AOP-024.
CARO (ARO: Energizes 1M DC from "N DC) 4 TC01 CCRS Main Turbine Trip, Reactor does not automatically trip, Entry into OP-TM-EOP-001.
CURO i (URO: IMA's of OP-TM-EOP-001) 5 i CCRS Sheared shaft on RC-P-1A, entry into OP-TM-MAP-F0301 and CURO OP-TM-226-151.
(URO: Trips 1A 6900V 8us) 6 MCRS OTSG Tube Rupture on "A" OTSG with a Loss of Subcooling
- M lIRO . Margin, entry into o P-TM-EOP-OOS, OP-TM-EOP-002.
MARO I I 7
I FW18A CCRS EF-P-2A Trips, entry into OP-TM-EOP-010, Rule 4.
,
CARO (ARO: Feeds OTSG's with Main Feedwater)
- (N}ormal, (R) eactivity, (I)nstru ment, (C}omponent, (M)ajor
-1
Appendix D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 5 Op Test No.: 10-02 NRC Examiners:
. Operators:
Initial Conditions: * (Temporary IC-235)
- 100% Power, MOL
- MO-P-1C and MO-P-1 F are OFF for Chemistry purposes lAW OP-TM-431-403/406
- Crane work is occurring on the West side of the Plant to stage new piping Turnover: Maintain 100% Reactor Power Critical Tasks:
- FW Flow Control (CT-16)
- Maintain RS Radiation Boundary (includes SG tubes) (CT-19)
Event No. Malf. No. Event Type* Event Description 1 DHR32 TSCRS BWST level lowers, entry into OP-TM-MAP-E0204 2 RC37A CCRS NSCCW Leak in RC-P-1A Motor Air Cooler, entry into OP-TM MAP-F0201 CURO (URO; Starts DW-P-1) 3 MS12C CCRS Hi Level in Moist. Sep. Tank, entry into OP-TM-MAP-N0201 CARO (ARO: Start MO-P-i C) 4 FW04B ICRS FW Temperature transmitter failure, entry into OP-TM-AOP-070 IURO (URO/ARO: Controls reactivity and feedwater in manual. ARO:
Controls Feedwater Flow in manual)
IARO 5 RD0153 TSCRS Dropped Safety Rod, runback fails to occur, entry into OP-TM-MAP IC16 H01 01, and OP-TM-AOP-OB2 RURO (URO; Reactivity manipulation, ARO: Feedwater manipulation)
NARO 6 RC37A 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)
CARO 7 TH06 M CRS RCS LOCA, Loss of Subcooling Margin, entry into OP-TM-EOP 002.
M LlRO MARO i
8 02ABS28 CCRS NSCCW Containment Isolation valves fail to close on ES signal with 02ABS22 low level.
C URO (URO: Manually closes NSCCW Containment Valves)
- (N)ormal, (R)eactivity, (I)nstrument, {C)omponent, (M)ajor
-1
Three Mile Island Unit 1 Telephone 717-948-8000 Nuclear Route 441 South, P.O. Box 480 Middletown, PA 17057 December 27, 2011 TMI-11-156 U.S. NRC Region I Administrator 475 Allendale Road King of Prussia, PA 19406 Three Mile Island Unit 1 Facility Operating License DPR -50 NRC Docket No. 50-289
Subject:
Submittal of Knowledge and Abilities (KIA) statements that will be suppressed from the random exam generation process It is our intent to develop the upcoming initial license exam scheduled for April 16, 2012 in accordance with NUREG-1021, Revision 9, "Operator Licensing Examination Standards for Power Reactors",
In accordance with NUREG 1021, "Operator Licensing Examination Standards", Three Mile Island Unit 1 is submitting for your review the list of KIA statements that will be suppressed from the random exam generation process in support of our April 16, 2012 license exam.
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 Greg Hoek, Exam Author, at (717) 948-2027.
Res"ectfully
~~
Glen Earl Chick Site Vice President, Three Mile Island Unit I GEC/mdf
Enclosures:
Three Mile Island Unit 1 Suppressed KIA statements cc: (without attachments)
Chief, NRC Operator licenSing Branch NRC Senior Resident Inspector - TMI Unit 1
09108/2011 Facility: TMI 1 IMPORTANCE Suppressed KJAs Basis RO/SRO 001 Continuous Rod Withdrawal AKl.Ol Rod bank step counters no rod bank step counters 2.9/3.2 atTMI AK2.06 T -ave./ref. deviation meter no T-ave./ref. deviation 3.0*/3.1 meter at TMJ Control Rod Control Rod Metroscope Metroscope Re:1cu)r Trip - Stabilization no reactor trip first 3.4/3.9 lout indication
- - - - *..- - - - - - - . - - - - - - - - - - - - - - . - - - - - - - - ...- - -..
... -~ ... --~ .. -~
3.1 *' /3.6'"
Closing CCW surge lank 2.3/2.6*
3.0*/3.3*
[ OIl Large Break LOCA Malfunctions
022 Loss of Reactor Coolant Makeup I ---
AIG.03 Performance of lineup to no excess letdown 3.1 */3.3*
establish excess letdown after path at TMI determining need AAl.04 Speed demand controller and no positive 3.3/3.2*
running indicators (positive displacement displacement pump) pumps used for reactor coolant makeup at TMI
-
AAl.07 Excess letdown containment no such equipment 2.8 */2. 7*
isolation valve switches and at TMI indicators 024 ----~
Emergency Boration ~~;~:___ j
~8-----~~-~-:-~-:-:-~-d-c-o-n-tr-o-ll-e-d-~-p-ro-t-~-t~----------------~I-;-~-:-P-p-l-~-a-hl-e-t-o-J~~-
_._---
026 Loss of Com AA2.03 The valve lineups necessary to no procedural 2.6/2.9 restart the eews while actions for this bypassing the portion of the evolution system causing the abnormal
_~~c~o_nd_it_io_n_ _ _ _ _ _ _ _ _~___________________ L __ _ _ _ _ _ _ ~_ _ _ _ _ _ __ _ _
r'
_Pressur~z:~r Pressure Control (PZR PCS) Malfunction AA2.17 028 Allowable ReS temperature difference vs. reactor power
not applicable to TMI 3.1/3.3 Pressurizer (PZR) Level Control Malfunction AAl.Ol PZR level reactor protection no PZR level input 3.8*/3.9 bistables to RPS at TMI AAl.04 Regenerative heat exchanger no such component 2.7/2.8 I AAl.05 and temperature limits Initiation of excess letdown per the eves atTMI not applicable at TMI 2.8/2.9 L I ---
not performed at I TMJ not performed at TMI BIT inlet valve 4.0/3.6 of control rods into lhe 4.2/3.9 EA2.10 3.1 */3.4*
Range Nuclear Instrumentation i Loss of Intermediate Range Nuclear Instrumentation 3.4*/3.7*
r;;-;;;:;-------------------------------------------- ~----~
Tube Leak Collection of Condensate 2.3/2.6 ejector monitor due to its Reset and check of 3.2/3.5 air ejector exhaust AK3.04
~------------- ...- - -.*.. ~-----
Steam Generator Tube Rupture (SGTR)
RCS loop isolation values no RCS loop 3.4*/3.1-1 isolation valves at TMI Steam flow indicators no steam flow S/G into the RCS, using the "feed and bleed" method Causes and consequences of shrink and swell in
'-'U'UU"'U~"1 Vacuum steam supply Loss of Main Feedwater (MFW) of feedwater and 3.4*/3.7*
..._ _ _ _ _ _+_ - - .. --------- ---~-~+_---~--- -..... ~-- ....-------L-------.--t_~-...... ~ ~-- --.
trip first-out panel 3.4*/3.9 no such component 2.9/3.3*
at TMI
055 Loss of Offsite and Onsite Power (Station Blackout)
J068--*-**-~*--*
the feed water system closing the AFW pump valve AK3.t4 3.2*/3.4*
AK3.16 2.8*/3.3*
076 High Reactor Coolant Activity AlG.02 CCWtlow One*line diagram of power to MIG sets K2.04 1*/2.7 K2.0S KS.l1 Relationship between 1/3.6*
reactivity worth of power APSRs shaping control rod group and other control rod groups (power-shaping, or No longer have 3.4*/4.1
- sets K..'l.71 for maintaining cross-tie breaker between rod drive MIG sets; reliability of control rod I drive trip breakers during set KS.76 No longer have APSRs KS.97 no T-ref at TMI no T-ref at TMI
2.9*/2.9*
_ _ _ _~L-~--~............. ~~--~~--~--~ ~~----~~------------~--~=-----~=-~~~-cc K6.10 no MIG sets at TMI 3.1 */3.3 r*-c-:-::--------t---:c=----:----:---:----:----t~--~......
no T-ref at TMI 3.l/3.4 "Prepower dependent insertion no metroscope at 4.0?/4.2?
limit" and power dependent insertion limit, determined with metrosco e +-'
A2.04 ~~-:---,
Positioning of axial shaping
- --~ -------------~~:~_:__--~---r-~~ ~~~-~ .
3.2*/3.8*
rods and their effect on SDM A2.05 Fractured split l.9?/1.9 A2. no MIG sets at TMI no lift coils at Switch no lift coils at
- /2.5*
011 Pressurizer Level K4.05 K4.06 K6.01
11:~!ctor Protection System I K4.07 First out indication nti;~:tti~()lIt 3.0/3.2* \
i K6.07 Core prOleCllon calculator no such component 2.9*/3.2*
atTMI K6.08 COLSS no such component I
-
3.6*/3.7~
at TMI K6J)9 CEAC no such component 3.6*/3.7*
at TMI A4.07 MIG set breakers no MIG set at TMI 3.9*/3.9*
I -
014 no metroscope at 1.9*/2.2 TMI 2.5*/2.7*
2.1'/2.6 2.9*/3.1 no LVDT at TMI 2.6*/3.0*
not applicable to 2.6*/2.7*
TMI 026
_,Containment Spray System (eSS)
K4.08 Automatic swapover to ! no automatic 4.1 */4.3*
containment sump suction for swapover at TMI recirculation phase after LOCA (RWST low-low level alarm) .~--- -----
K4.09 Prevention of path for escape of
- I no such interlock at 3.7*/4.1
A4.02 Prevention of path for escape of no such components 2.3*/2.6*
radioactivity from containment atTMI to the outside (interlock on RWST isolation after swapover)
The remote location and use of spool pieces and other equipment to set up portable recirculation pump for additive tank, including power supply A4.03 The remote location and use of no such components 2.2*/2.5*
the special tank needed for atTMI
' - - -..
draining CSS
Ke~:onlbiJllerand Purge Control System (HRPS)
A3.01 Moisture separator steam supply A4.05 Moisture separator reheater, checking its temperatures and relative to and operating System (8DS) and Turbine Bypass Control Operation of loss-of-Ioad bistable taps upon turbine loss T-ave., verification above low/low setpoint ICS voltage inverter not applicable to 2.9*/3.1 TMI T-ave. mode 2.4*12.5 045 Main Turbine Generator (MT/G) System K4.10
~- .........- ................ -
045
. Main Turbine Generator (MT/G) System K4.1S Steam blanketing (atmospheric not performed at 1.6/1.7 pressure) moisture separator TMI reheater to drive out air and non condensables prior to starting up K4.44 Impulse pressure mode control no such equipment 2.5*/2.8*
of steam dumps atTMI
~ ..--
K4.46 Defeat of reactor trip by no such equipment 2.5/2.8~
overspeed trip test lever atTMI K6.06 Generator amplidyne balance no such equipment 1.6*/US*
system at TMI
,-_.
no interface with 1.9/1.9 atTMI
---~--T~~~~....................... ---- ........ ~. I-----------**************~*********-- ~-~-~~-
no air ejectors at 1.7/1.7 T\1I Loss of air ejector cooling water no air ejectors at L8*!2.0*
- TMI Operation of hotwell pump and no air ejector recirculation line atTMI i~olation valve to maintain no such alTMI no such equipment atTMI no such equipment 1.7*/1.6*
atTMI no such equipment 1.8"/1.7*
at
_ _ .... ....
i K2.03 pump L __
-- ----- -
064 Emergency Diesel Generator (ED/G) System A3.1O Function of ED/G megawatt no such component 2.8/2.8*
load controller / operation at TMI A3.11 Need for setting offsite power no such component 3.1*/2.9*
breaker to automatic / operation at TMI--
A4.04 Remote operation of the air no such component 3.2*/3.2 compressor switch (different / operation at TMI modes)
_.
075 Circulating Water System --
Kl.02 Liquid radwaste discharge no interface at TMI 2.9/3. L Kl.07 Recirculation spray system no such component 2.2 */2. i-;';-
1---
at TMI Kl.09 Vacuum priming no vacuum priming l.5/ l.4 for Circ Water at TMI -
K2.04 Lube oil pumps no such component 1.4 */ l.4 *
-- .----- ... ---~---.---
at TMI ------
K3.05 Recirculation spray system no such component 2.1 */2.3*
_.. atTMI K4.03 Interlocks between circulating no separate cooling 1.7*/2.1
- water system pumps and cooling tower pumps at
--
tower pumps TMI -
K4.04 Automatic pickup of backup no such component 1.7*/1.9 Hube oil (Jum]2s (ac and dc) atTMI K4.06 I Traveling screen operation no such component 1.6/1.8
._ I -
at TMI _. _.
i Relationship of seawater no seawater at TMI 1.4*/1.6*
~;;
I tem]2erature to marine growth IPurpose of the vacuum priming no vacuum priming 1.6/1.6 system for Circ Water at
- ------
TMI A1.08 Circulating water makeup pump no such component 1.6*/1.6*
motor current (within limits) atTMI --- -_._-
A2.0 1 Loss of intake structure no intake structure 3.0* /3.2 for Circ Water at f-----_.
TMI A2.08 lee buildup on intake structure no intake structure 2.0*/2.0*
for Circ Water at
-
~LO Automatic startup mode of water no priming pumps 1.5*/1.6*
box priming pumps relative to at TMI specified minimum vacuum ._ -" ."
A2.11 Time required for fill of piping not filled by 1.5 */1.6*
by induction of water into induction at TMI circulating system using vacuum system I A4.04 Air eductor system no such component 1.8*/1.8*1 atTMI C Water box vacuum priming no vacuum priming 1.8*/I~
isolation valves, control for Circ Water at switches, and indicators TMI
--
075 Circulating Water Svstem A4.07 Vacuum priming tank/priming no vacuum priming 1.7* /1.6
- compressor controller for Circ Water at
, - TMI A4.08 Gland seal water supply system no such component 1.6/1.6
,~-.
atTMI A4.14 Lube oil pumps for circulating no such component 1.5*il.7*
water pump atTMI ----
A4.lS Operation of the vacuum no vacuum priming 1.4/l.5 priming system for Circ Water at TMI A4.16 Traveling screens in manual no such component 1.6/l.6 operation atTMI A4.20 Blowout preventers no such component 1.7*/1.8*
at.TMI 076 Service Water System (SWS)
KI.03 K1.2S Heat sink pond makeup K1.26 Flood alarm system
- - - - , - , - - - - - i - - - - - : - - : - - - ___ - - - , - - - : - - - - - - : - - - i - - - - - - - - - - - - - - - t -
K4.04 River intake water level recorders 086 Fire Protection K1.01 K1.02 Raw service vent system system 2.4*/2.7*
to not applicable to 2.1*/2.2*
TMI not applicable to 2.7*/2.7*
TMI not applicable to 2.4*/2.2*
,--------~- -
103 Containment System A4.07 Use of the air lock rate test panel no operated or 2.4 */2.5*
monitored from the control room at TMI A4.0S Operation of refueling drain no operated or 1.9/2.2 valves (for draining refueling monitored from the canal to lower containment control room at sump) TMI A4.09 Containment vacuum system not applicable to 3.1 * /3. 7*
-" ~-
The above K! As are the pre-suppressed KIA's at TMI in addition to those allowed by D.l.b of ES 401 and all of system 25 Ice Condenser system as we have no Ice Condensers.
Evolution 003 AK3.06 will need to be unsuppressed as digital control rod drive makes this testable.
System 001 K4.06 suppress no first out panel at TMI.
System 001 K4.16 suppress no longer have Aux/Group power supplies under Digital CRD.
System 001 KS.ll suppress no longer have APSRs.
System 001 KS.12 suppress no longer have APSRs.
System 001 KS.76 suppress no longer have APSRs.
System 001 K6.09 suppress no neutron flux recorder.
System 001 A2.04 suppress no longer have APSRs.
System 001 A4.04 suppress no longer have APSRs.
System 001 A4.08 suppress no mode select switch.
NOTE: Generic K/As associated with emergency and abnormal plant evolutions (E/APE) and plant systems for both RO and SRO examinations should be randomly selected from the following: 2.1.7,2.1.19,2.1.20,2.1.23,2.1.25, 2.1.27, 2.1.28, 2.1.30, 2.1.31, 2.1.32, 2.2.3, 2.2.4. 2.2.12. 2.2.22, 2.2.25.
2.2.36, 2.2.37, 2.2.38, 2.2.39, 2.2.40, 2.2.42, 2.2.44, 2.4.1, 2.4.2, 2.4.3, 2.4.4, 2.4_6, 2.4.8, 2.4.9, 2.4.11. 2.4.18, 2.4.20, 2.4.21, 2.4.30, 2.4.31.
2.4.34, 2.4.35. 2.4.41, 2.4.45, 2.4.46. 2.4.47, 2.4.49, and 2.4.50. All other generic K/As for systems and evolutions may be suppressed. The only generic KlAs that can be suppressed for the generic section of the exam (Tier 3) are KlAs 2.2.3 and 2.2.4, but only at single-unit facilities.
NOTE: TMI is a Single Unit Facility (2.2.3 and 2.2.4) should be suppressed)