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
- -,/£'
Exel n Three Mile Island Unit 1 Telephone 717-948-8000 Route 441 South, P.O. Box 480 Middletown, PA 17057 TMI-12-012 U.S. NRC Region I Administrator 475 Allendale Road King of Prussia, PA 19406 Three Mile Island Unit Renewed Facility Operating License No. NRC Docket No.
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. R. W. Libra Site Vice President, Three Mile Island Unit I RWUmdf -*1 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 Simulator Scenario Quality Checklist (Form Transient and Event Checklist (Form Competencies Checklist (Form Written Exam Quality Checklist (Form Exam ination Security Agreements (Form Record of Rejected KlAs (Form (without attachments)
Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector
-TMI-1 Operations Training Manager T-7s-Exelon Nuclear Three Mile Island Unit 1 Telephone 717-948-8000 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. Site Vice President, Three Mile Island Unit I GEC/mdf
Enclosures:
(Mailed to John Caruso, Chief Examiner, NRC Region I) Examination Security Agreements (Form Administrative Topics Outlines (Form Control Room/In-Plant Systems Outline (Form PWR Examination Outline (Form Generic Knowledge and Abilities Outline (Tier 3) (Form Statement detailing method of Written Outline Scenario Outlines (Form Record of Rejected KlAs (Form Completed Examination Outline Quality Checklist (Form ES-201-2)
Transient and Event Checklist (Form ES-301-5)
(without attachments)
Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector
-TMI Unit 1 Three Mile 2012 NRC Initial License Written Written Examination Outline 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.
Facility:
TMI Date of Exam: 4/2012 RO KIA Category Points Tier Group KKK KKK A A A 2 3 456 1 2 3 1. , Emergency
& Plant Evaluations
- 2. Plant Systems A G 4
- Total 18 9 27 28 10 38 SRO-Only Points A2
- 3 3 2 2 5 5 3 2 0 2 4 4 Total 6 4 10 5 3 8 PWR Examination Outline Form ES-401-2 2 3 4 10 3. Generic Knowledge
& Abilities f..-_1_-+-_2_-+-_3
__ 2 2 2 332 Note 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). 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. 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. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution. Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected.
Use the RO and SRO ratings for the RO and SRO-only portions, respectively. Select SRO topics for Tiers 1 and 2 from the shaded systems and KiA categories. 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 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. For Tier 3, select topics from Section 2 of the KiA Catalog, and enter the KiA numbers, descriptions, I Rs, and point totals (#) on Form ES-401-3.
Limit SRO selections to KiAs that are linked to 10CFR55.43 7
ES-402 Form ES-401-2 PWR Examination Emergency and Abnormal Plant Evolutions
-Tier 1 Group EAPE#/Name Safety Function 062 I Loss of Nuclear Service. Water! 4 058 I Loss of DC Power I 6 065 / Loss of Instrument Air I 8 077 I Generator Voltage and
- Electrical Grid Disturbances!
6 054 I Loss of Main Feedwater
/ 4 *038/ Steam Generator Tube
- Rupture i 3 E05 / Steam Line Rupture Excessive Heat Transfer / 4 007/ Reactor Trip /1 008/ Pressurizer Vapor Space Accident/3 x x x x KIA Topic(s) AA2.02 -Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The cause of possible SWS loss AA2.02 Ability to determine and interpret the following as they apply 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 to the Loss of Instrument Air: Failure Modes of air-operated 3.6
- 76 3.6 77 3.3 78 4.7 79 4.6 80 4.1 81 3.8 39 3.3 40 3.1 41 2.9 42 029 I Anticipated Transient Without Scram (A TWS) /1 ES-402 Form ES-401-2 PWR Examination Emergency and Abnormal Plant Evolutions
-Tier 1 Group AK2. 1 0
- Knowledge of the 015/17/ Reactor Coolant Pump interrelations between the Reactor 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*
X within the control room team as 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 009/ Small Break LOCA / 3 X reasons for the following 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 AA 1.04 -Ability to operate and / or monitor the following as they apply 056/ Loss of Off-site Power / 6 to the Loss of Offsite Power: 3.2 49 Adjustment of speed of ED/G to maintain frequency and voltage levels AA 1.01 -Ability to operate and / or 027 / Pressurizer Pressure monitor the following as they apply to the Pressurizer Pressure 4.0 : Control System Malfunction / 3 Control Malfunctions:
PZR heaters, AA2.02 -Ability to determine and 026 I Loss of Component Cooling interpret the following as they apply 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 PWR Examination Emergency and Abnormal Plant Evolutions
-Tier 1 Group 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 PWR Examination Emergency and Abnormal Plant Evolutions
-Tier 1 Group EAPE#/Name Safety Function G KIA Topic(s) I K1 I K2 I K3 I A1 I A2 I 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 037 I Steam Generator Tube to the Steam Generator Tube Leak / 3 Leak: System status, using 3.4 83 independent readings from redundant Condensate air ejector exhaust monitor 2.1.23 -Loss of NNI-Y -Ability to A03 ! Loss of NNI-Y /7 perform specific system and 4.4 84 integrated plant procedures during all modes of n. 005 I Inoperable/Stuck Control 2.2.38 -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 X following concepts as they apply 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, A07 / Flooding / 8 X including primary coolant, 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 X monitor the following as they apply 3.3 60 / 1 to the Continuous Rod Withdrawal:
RPI AA2.02 -Ability to determine and 051 / Loss of Condenser Vacuum interpret the following as they apply to the Loss of Condenser Vacuum: 3.9 61 /4 Conditions requiring reactor and/or turbine tri ES-402 Form ES-401-2 PWR Examination Emergency and Abnormal Plant Evolutions
-Tier 1 Group 061 / Area Radiation Monitoring (ARM) System Alarms / 7 060 / Accidental Gaseous RadWaste Release / 9 E09 / Natural eirc. / 4 x A04/ Turbine Trip / 4 KIA Category Totals 2 Group Point Total: 63 3.8 64 ES-402 Form ES-401-2 PWR Examination Plant Systems -Tier 2 Group 006 Emergency Core Cooling 010 Pressurizer Pressure Control 012 Reactor Protection 026 Containment Spray 013 Engineered Safety Features Actuation 012 Reactor Protection X 061 Auxiliary/Emergency X Feedwater A2.10 -Ability to (a) predict the impacts of the following malfunctions or operations on the and (b) based on those predictions, use procedures to correct, control, 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 on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or : Heater failures 2.2.12 -Equipment Control: Knowledge of surveillance 2.4.21 Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling 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 Ability on those predictions.
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 between the RPS and the following systems: 120V 3.9 86 3.6 87 4.1 88 4.6 4.2 3.4 3.4 2 ES-402 Form ES-401-2 PWR Examination Plant Systems -Tier 2 Group System #/Name KIA Topic(s) 006 Emergency Core Cooling 076 Service Water 064 Emergency Diesel Generator 013 Engineered Safety Features Actuation 063 DC Electrical Distribution OOB Component Cooling Water 007 Pressurizer Relief/Quench Tank 010 Pressurizer Pressure Control 004 Chemical and Volume Control 005 Residual Heat Removal x x x x K2.02 -Knowledge of bus power supplies to the following:
Valve operators for accumulators K2.04 -Knowledge of bus power supplies to the following:
Reactor building closed cool water K3.01 -Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following:
Systems controlled automatic loader K3.03 -Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following:
Containment K4.02 -Knowledge of dc electrical system design feature(s) and/or interlock(s) x which provide for the following:
Breaker interlocks, permissives, bypasses and cross-ties.
K4.01 -Knowledge of CCWS design feature(s) and/or x interlock(s) which provide for the following:
Automatic start of K5.02 -Knowledge of the operational implications of the following concepts as the apply to PRTS: Method of x forming a steam bubble in the PZR K5.01 -Knowledge of the operational implications of the following concepts as the x apply to the PZR PCS: Determination of condition of fluid in PZR, using steam tables K6. 17 -Knowledge of the operational implications of the x following concepts as they apply to the CVCS: Flow paths for boration K6.03 -Knowledge of the effect of a loss or malfunction x on the following will have on the RHRS: RHR heat 2.5 3 2.5 4 3.B 5 4.3 6 2.9 7 3.1 B 3.1 9 3.5 10 4.4 11 2.5 12 ES-402 Form ES-401-2 PWR Examination Plant Systems -Tier 2 Group System #/Name 003 Reactor Coolant Pump 103 Containment 026 Containment Spray 078 Instrument Air 062 AC Electrical Distribution 039 Main and Reheat Steam 073 Process Radiation Monitoring 059 Main Feedwater KIA Topic(s) A 1.02 -Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including:
RCP pump and motor bearing tem A 1.01 -Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) 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 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 procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Failure of spray 2.9 13 3.6 14 2.9 15 3.9 16 17 3.0 18 2.8 19 3.9 20 4.1 21 022 Containment Cooling ES-402 Form ES-401-2 System #/Name 078 Instrument Air 062 AC Electrical Distribution 008 Component Cooling Water 039 Main and Reheat Steam 006 Emergency Core Cooling 064 Emergency Diesel Generator 005 Residual Heat Removal KIA Category Totals TMI PWR Examination Outline Plant Systems -Tier 2 Group 1 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 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) associated with operating the 2.5 23 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 those predictions, use 3 procedures to correct, control, 24
- or mitigate the consequences of those malfunctions or operations:
High/low CCW A3.02 Ability to monitor automatic operation of the MRSS, including:
Isolation of K3.01 -Knowledge of the effect that a loss or malfunction of the ECCS will have on the ReS K4.02 -Knowledge of ED/G system design feature(s) and/or inter-Iock(s) which provide for the following:
Trips for ED/G while operating 2.1.7 -Conduct of Operations:
Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument int'>rnrot"ti",n Group Point Total: 3.1 25 4.1 26 3.9 27 28 ES-401 Form ES-401-2 PWR Examination Plant Systems Tier 2 Group KIA Topic(s) 016 Non-Nuclear Instrumentation System 068 Radwaste 071 Waste Gas Disposal System 002 Reactor Coolant 014 Rod Position Indication 041 Steam Dump!Turbine Bypass Control 075 Circulating Water 071 Waste Gas Disposal 029 Containment Purge x x x A2.01 -Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or " ...,,,,,,;f''''*
Detector failure 2.2.40 -Equipment Control: Ability to apply technical 2.1.20 -Waste Gas Disposal System (WGDS): Ability to interpret and execute ste . A 1.03 -Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCS controls including:
K1.02 Knowledge ofthe physical connections and/or cause-effect relationships between the RPIS and the f"lI,mAt_.inn
.,,,,,,t,,,,..,,,,*
NIS A2.09 -Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System; and (bl based on 3.1 91 4.7 92 4.6 93 3.7 29 3.0 30 4.6 31 2.6 32 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) which provide for the 3.2 following:
Automatic purge isolation ES-401 Form ES-401-2 PWR Examination Plant Systems -Tier 2 Group System #/Name 072 Area Radiation Monitoring 086 Fire Protection 015 Nuclear Instrumentation 034 Fuel Handling Equipment KJA Category Totals KJA Topic(s) K3.02 -Knowledge of the effect that a loss or malfunction of the ARM system will have on the following:
Fuel handling K6.04 -Knowledge of the effect of a loss or malfunction on the Fire Protection System following will have on the: Fire, smoke, and heat A3.04 -Ability to monitor automatic operation of the NIS, including:
Maximum disagreement allowed between channels A4.01 -Ability to manually operate and/or monitor in the control room: Radiation levels Group Point Total: 3.1 35 2.6 36 3.3 37 3.3 38 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# 2.1.45 Ability to identify and interpret diverse indications 4.3 66 to validate the response of another indicator.
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 I procedures.
3.4 94 Ability to evaluate piant performance and make 2.1.7 operational judgments based on operating 4.7 98 characteristics!
reactor behavior, and instrument interpretations.
Subtotal 3 ......* 2 2.2.13 Knowledge of tagging and clearance 4.1 68 I procedures.
Knowledge of the process for managing 2.2.18 maintenance activities during shutdown 2.6 69 operations, such as risk assessments, work prioritization, etc. Knowledge of less than or equal to one hour I 2.2.39 technical specification action statements for 3.9 75 2. Equipment systems. Control 2.2.21 Knowle?ge of pre-and post-maintenance operability requirements. Subtotal y+; 3 ..........
1...*..
ES-401 Generic Knowleclge and Abilities Outline (Tier 3) Form ES-401-3 3. Radiation Control 2.3.13 Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aliQninQ filters, etc. 3.4 70 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emerQency conditions or activities.
3.4 71 2.3.11 Ability to control radiation releases.
4.3 96 2.3.13 Knowledge of Radiological Safety Procedures 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. 3.8 99 Subtotal 2 2 4. Emergency Procedures
/ Plan 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
4.2 72 2.4.1 Knowledge of EOP entry conditions and immediate action steps. 4.6 73 2.4.27 Knowledge of "fire in the plant" procedures.
3.9 97 2.4.41 Knowledge of emergency action thresholds and classifications.
4.6 100 Subtotal 2 2 Tier 3 Point Total: 10 7 ES-401 Record of Rejected KIA's Form ES-401-4 Tier / Group Randomly Selected Reason for Rejection KA 1/1 077 / 2.4.3 replaced The subject KIA isn't relevant at the subject facility.
by 077 / 2.2.37 2/2 071 / A2.07 replaced The subject KIA isn't relevant at the subject facility.
by 071 / A2.09 2/1 059/2.4.3 replaced The subject KIA isn't relevant at the subject facility.
by 059 / 2.4.45 056/ M1.12 1 / 1 replaced by 056 / 056/M 1.12 overlaps with a JPM on the Operating Exam. M1.04 2/1 008/ A2.02 replaced There is an overlap issue between the KA selected and a JPM by 008 / A2.03 on the Operating Test 2/1 064/ K2.02 replaced It isn't possible to prepare a psychometrically sound question 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 1/1 SRO replaced by 015/017 M2.08 sampled, randomly chosen from same system. 1/1 SRO 025/2.1.27 replaced Could not write SRO level question to 2.1.27 for evolution 025, by 025/2.4.1 randomly chose from generics against same evolution.
1/1 SRO E05 / 2.1 .32 replaced No system limits and precautions associated with this evolution, by 054/2.2.37 randomly chose new KIA from NKEG exam outline generator.
1/2 SRO 059/2.4.8 replaced No EOP for accidental liquid release, randomly chose new KIA by A03 /2.1 .23 from NKEG exam outline generator.
A04/ 2.4.9 replaced Could not write test question to low power turbine trip affect on 1/2 SRO by 005/2.2.38 mitigation strategy, randomly chose new KIA from NKEG exam outline generator.
2/1 SRO 004/ A2.04 replaced No accidental gas release associated with CCVC, randomly by 006 / A2.1 0 chose new KIA from exam outline Qenerator.
2/1 SRO 059/2.4.47 replaced Could not write E-plan related question to MFW trending, 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.
045/2.2.37 replaced Could not write operability of safety system question against 2/2 SRO turbine, randomly chose new KIA from NKEG exam outline by 017 I A2.02 Qenerator.
0 017/ A2.02 again Could not write a discriminating question, randomly chose new 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 / M2.08 picked a new KIA.
- 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 3SRO 2.1.14 replaced by Could not write a discriminating question at the SRO level. 2.1.7 Randomly chose new KIA from tier 1. 3SRO 2.4.35 replaced by Could not write a discriminating question at the SRO level. 2.4.41 Randomly chose a new KIA from tier 4
- 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 (See Note) Type Code* Describe activity to be performed Conduct of Operations N/R Perform a Manual Power Range Calculation 2.1.37 (4.3) Conduct of Operations M/R Perform a Transient Leak Rate Calculation 2.1.23 (4.3) Isolate a Component for Maintenance Equipment Control N/R 2.2.41 (3.5) Calculate Dose Limit Stay Times M/R Radiation Control 2.3.4 (3.2) Category not selected for RO applicants n/a Emergency Procedures/Plan 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 manually check Power Range Calibration using hand This JPM is a new ...1License is evaluated against properly calculating Power Range Calibration and identifying an Error Power greater than Safety significance, failure to identify an Error Linear Power >2.0% would result in continued with Offset Error outside of the acceptance criteria lAW T.S. CONDUCT OF OPERATIONS (A1-2): Perform a Transient Leak Rate Given plant conditions and reference OS-24 Conduct of Operations During Abnormal and Events, the candidate will be directed to perform a transient RCS leak rate calculation that will accurately determine the current RCS leak JPM is modified from a Bank License is evaluated against properly calculating an RCS leak rate when given multiple data 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 Safety significance, failure to calculate an accurate RCS leak rate could cause a lower than realistic and redirect a Control Room crew away from appropriate Procedures lAW EQUIPMENT CONTROL (A2): Isolate a Component for Maintenance.
Given a plant component needing to be isolated identify mechanical and electrical isolation points. This a new JPM developed for this Safety significance is failure to properly identify the correct points could lead to a loss of nuclear river water and/or personnel RADIATION CONTROL (A3): Calculate Dose Limit Stay Given plant conditions, a dose history, and references RP-M-460 Controls For High and Very Radiation Areas, EP-M-112-100-F-01, Shift Emergency Director Checklist, and EP-M-113.
Protective Actions, the candidate is directed to determine maximum stay time for performing a operation without exceeding the limit approved by the TSC Radiation Protection Manager ...IPM is modified from a Bank License is evaluated against properly identifying the maximum increased dose exposure limit calculating stay time ta1<ing into account current exposure.
Failure to correctly identify stay time result in a dose limit being 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 (See Note) Conduct of Operations Conduct of Operations Equipment Control Type Code* N/R M/R M/R I Describe activity to be performed Perform and Approve a Manual Power Range Calculation 2.1.37 (4.6) Perform a Transient Leak Rate Calculation with a T.S. Call 2.1.23 (4.4) Evaluate a Completed Surveillance Procedure and Perform Appropriate Actions 2.2.12 (4.1) Ii I! Radiation Control Emergency Procedures/Plan N/R M/R Review and Approve a Gaseous Release Permit for a Waste Gas Tank 2.3.6 (3.8) Identify and Declare an Emergency Classification with a 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 manually check and approve Power Range Calibration using hand This JPM is a new License is evaluated against properly calculating Power Range Calibration, identifying several errors the given data sheet, and not approving the hand Safety significance, failure to identify the errors would result in continued operation with Power instrumentation outside of the acceptance criteria lAW T.S. CONDUCT OF OPERATIONS (A1-2): Perform a Transient Leak Rate Calculation with a T.S. Given plant conditions and reference OS-24, Conduct of Operations During Abnormal and Events, the candidate will be directed to perform a transient RCS leak rate calculation that will accurately determine the current RCS leak rate and identify any JPM is modified from a Bank License is evaluated against properly calculating an RCS leak rate when given multiple data 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 Safety significance, failure to calculate an accurate RCS leak rate could cause a lower than realistic and redirect a Control Room crew away from appropriate Procedures and EAL's lAW EQUIPMENT CONTROL (A2): Evaluate a Completed Surveillance Procedure and Appropriate Given plant conditions, a data sheet, and reference ER-TM-321-1041, TMI-1 1ST Program the candidate will be directed to evaluate a completed surveillance procedure and perform JPM is modified from a Bank License is evaluated against properly reviewing the data sheet against ER-TM-321-1 041, and out-of-spec data Safety significance, failure to identify out-of-spec data points would lead to unknown violation of Specs and could lead to possible equipment damage and/or personnel RADIATION CONTROL (A3): Review and Approve a Gaseous Release Permit for a Waste Gas Given plant conditions, data sheets, and reference 661 0-ADM-4250.11 , Releasing Radioactive Effluents
-Waste Gas Tanks AlB/C, the candidate is directed to review and approve a filled-out release This JPM is a new ..License is evaluated against properly reviewing the given data against 661 0-ADM-4250.11 , identifying incorrect data Safety significance, failure to identify incorrect data points would lead to a gaseous release with above Tech Spec allowed levels ana possible adverse effects to the EMERGENCY PROCEDURES/PLAN (A4): Identify and Declare an Emergency Given a set of conditions, and references EP-AA-112-1 OO-F-01Shift Emergency Director Checklist, AA-1009 Exelon Nuclear Radiological Emergency Plan Annex r-or Three Mne Isrand (TMI) Station, MA-114-1 OO-F-01, State/Local Event Notification Form, EP-AA-111, Emergency Classification Protective Action Recommendations, and EP-AA-111-F-09, TMI Plant Based PAR Flowchart, candidate is directed to determine the Emergency Action Level (EAL) and make a Protective Recommendation (PAR) lAW the facility Emergency Plan ..IPM is modified from a Bank License is evaluated against properly identifying the Emergency Classification and Protective Recommendations (PAR). Failure to correcfly laentify the Emergency Classification and Protective Recommendations could result in unnecessary harm to the general II I Control Systems Outline Form ES*301*2 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 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 1 E Inverter (Sys 062) A4.01 -Place 8 th Stage Heating On-Line (Sys 039) G2.1.30 N 0 6 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. Criteria for RO I SRO-II SRO-U* Type Codes (A) Item ate path 4-61 4-6 I 2-3 * {C}ontrol room * (D)irect from bank 59/ 58 / 54 2=, 1 I 2=,1 / 2=, (EN)gineered safety * (E}mergency or abnormal in-plant -/ -/ 2=, 1 (control room system 2=, 1 I 2=,1 / > 1 (N)ew or (M)odified from bank including 1 (A) (L)ow-Power I Shutdown 2=,2/ 2=,2 531 53 / 5 2 (randomly
- (P}revious 2 exams (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 JPM B -Respond To High Pressure Injection Initiation (Alt Path -MU-V-14A Fails To Open). Alternate Safety significance failure to complete this JPM will result in a loss a" three HPI JPM C -Respond to an RCS leak into ICCW. New Alternate path, alternate path leak gets worse, actions required. Safety significance failure to properly respond will lead to a continued RCS leak outside JPM D -RCP #1 Seal Failure. Previous NRC 2011 Alternate path "IPM. Alternate path failure pump must be shut Safety significance failure to properly address excessive seal leakoff could result in Seal LOCA. randomly by drawing playing cards representing JPMs from last two JPM E -Perform the Required Actions for EF-P-1 trip. Bank Safety significance failure to p'roperly complete the task would lower the safety margin for EFW by only motor driven pumps JPM F -Place an RPS cabinet in Manual Bypass. New path. Alternate path failure in another cabinet requires placing cabinet in tripped state to place 1 5 cabinet in Manual Safety significance with a failed instrument in the other cabinet, incorrect operation in this cabinet lead to reactor trip, an initiating event for JPM G -OP-TM-EOP-020 IMA's Safety significance failure to complete the IMA's of the EOP could result in failure to adequately and transfer control of the reactor to the remote shutdown JPM H -Establish Alternate RB Emergency Cooling. New Alternate Safety significance failure to properly complete the task would result in containment temperatures than assumed in the structural analysis, see Technical Specification 3.17. Alternate path involves complete failure of the emergency cooling systems and restoration of a portion of normal cooling alternate power JPM I -Supply VBC from the 1 E inverter.
New In Plant Safety significance failure to complete the task properly could result in inadvertent safety actuations, or other transients (accident initiators) due to reduction of Vital Power JPM J -Place 8 th Stage Heating On-Line. Bank Safety significance failure to complete the task properly could result in water quality reduction to OTSGs, fong term result could be damage to JPM K -Prepare for Transfer for RB Sump Recirculation.
Bank Safety significance failure to complete the task may result in inability to complete these actions at a time due 10 inaccessibility due to radiation levels post accident on recirculation.
This is a time action, approximately 25 minutes from start of event reference UFSAR 14.2.2.5.d and 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) System 1 J PM Title Type Code* Safety Function rom CRD Sequence Fault (Sys 001) A2.18 N/S 1 b. Respond To 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) AA 1.02 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.
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 I N 6 j. Place 8 th Stage Heating On-Line (Sys 039) G2.1.30 D t+/-j k. Prepare for Transfer to RB Sump Recirculation (Sys 006) K4.08 D/E/LIR @ 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 1 I I 1 (EN)gineered safety feature -/ I 1 (control room system (L)ow-Power
/ Shutdown 1 / I 1 (N)ew or (M)odified from bank including 1 (A) / 1 (P)revious 2 exams 53/ 53 I 52 (randomly selected) (R)CA 1 I I 1 (S)imulator ES-301, Page 23 of 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 JPM B -Respond To High Pressure Injection Initiation (Alt Path -MU-V-14A Fails To Open). Alternate Safety significance failure to complete this JPM will result in a loss all three HPI JPM C -Respond to an RCS leak into ICCW. New Alternate path, alternate path leak gets worse, actions Safety significance failure to properly respond will lead to a continued RCS leak outside JPM D -RCP #1 Seal Failure. Previous NRC 2011 Alternate path JPM. Alternate path failure pump must be shut Safety significance failure to properly address excessive seal leak off could result in Seal LOCA. randomly by drawing playing cards representing JPMs from last two JPM E -Perform the Required Actions for EF-P-1 trip. Bank Safety significance failure to properly complete the task would lower the safety margin for EFW by only motor driven pumps JPM F -Place an RPS cabinet in Manual Bypass. New path. Alternate path failure in an other cabinet requires placing cabinet in tripped state to place 1 S cabinet in Manual Safety significance with a failed instrument in the other cabinet, incorrect operation in this cabinet lead to reactor trip, an initiating event for JPM G -OP-TM-EOP-020 IMA's Safety significance failure to complete the IMA's of the EOP could result in failure to adequately and transfer control of the reactor to the remote shutdown JPM H -Not selected for JPM I -Supply VBC from the 1 E inverter.
New In Plant Safety significance failure to complete the task properly could result in inadvertent safety actuations, or other transients (accident initiators) due to reduction of Vital Power JPM J -Place 8 th Stage Heating On-Line. Bank Safety significance failure to complete the task properly could result in water quality reduction to OTSGs, rong term result could be damage to JPM K -Prepare for Transfer for RB Sump Recirculation.
Bank Safety significance failure to complete the task may result in inability to complete these actions at a time due to inaccessibility due to radiation levels post accident on sup recirculation.
This is a time action, approximately 25 minutes from start of event reference UFSAR 14.2.2.5.d and Scenario Outline Form ES-D-Facility:
Three Mile Island Scenario No.: Op Test No.: 10-02 NRC Examiners:
Operators:
nn<:,nn.v D 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-----------------------.
Maintain 100% Reactor Power ! Critical Tasks: Control SG Pressure (adjust TBVs/ADVs) to: Maintain RC Temperature Constant or*
- Control RCS Inventory (CT-30) : Event No. Event Description Malf. No. i Event Type* 1 NR-P-1C Trips, NR-P-1 B Fails to Auto-Start, entry into MAP-B01 05, and OP-TM-MAP-B020S TSCRS RW02C CARO (ARO: Starts NR-P-1 B from CR) 2 ICS Auto Power ICCW Subfeed Failure, entry into i
- ED22G IARO (ARO: Restores Letdown following a Loss of ICS AUTO Power) 3 Loss of EG-Y-1A Starting Air, entry into OP-TM-MAP-A0102, and OP-TM-MAP-A0201 TSCRS I EGR30
- RURO (URO: Reduce Reactor Power) i I 4 IC09 IC53 ICRS IURO ! IARO MW Generated Input Fails to Zero Volts, entry into 070 (URO/ARO:
Control ICS in Manual lAW AOP-070) 5 6 ED12 I ED13
- FW09A CCRS CURO CARO MCRS M URO Loss of ICS Hand and Auto Power, entry into OP-TM-AOP-02S, and OP-TM-EOP-001 (URO: Reactor Trip IMA's, ARO: Control OTSG Pressures)
FW Line Break Inside RB, Excessive Heat Transfer, entry into TM-EOP-003.
7 ZDISSM UV37(1) ! MARO CCRS CURO MU-V-37 Fails Closed (URO: Throttle an MU-V-16 for minimum MU flow) * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, {M)ajor -1 Appendix 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
- Turbine Trip (CT-18) * "e" RPS Cabinet is Maintain 100% Power Operations
- 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 Restore Feed to a Dry OTSG i* I Event Type* Event DescriptionEvent No. Malt. No. I 1 ZAIRC1L1C CCRS CURO MU-V-17 Fails Closed in Auto, entry into OP-TM-211-472 (URO: Controls Pressurizer Level with MU-V-17 in Manual) 2 i ED09D TSCRS CARO Loss of D Inverter, Loss of VBD, entry into OP-TM-AOP-018 (ARO: Place Rad Monitors Interlock switches to Defeat, Restore Control Building, Auxiliary Building, Fuel Handling Building Ventilation) 5 TU01D NCRS RURO , NARO 6 i FW15A CCRS
- FW15B i CURO TC02 7 FW17 i MCRS FW18A M lIRO FW18B MARO 8 MS09A-F CCRS CARO 3 ' NI15B Nuclear Instrument, NI-6, Failure (TS) TSCRS I SG/RX Demand Station fails to 0 Volts, Entry into OP-TM-AOP-070
! i ,llIRO i (URO: ICS station to Manual, ARO: Controls temperature with SG 4 IC23 I CRS , A &B FW DEMAND stations in Manual)IARO High Vibrations on Main Turbine, entry into OP-TM-MAP-K0201 and 1102-4, Reactor shutdown (URO/ARO:
Power reduction with ICS in Manual) Loss of both Main Feedwater Pumps, Turbine fails to OP-TM-EOP-001 (URO: IMA's of OP-TM-EOP-001)
Loss of Emergency Feedwater Pumps, entry into OP-TM-EOP-004, Lack of Heat Transfer.
Turbine Bypass Valves fail Closed, OTSG Pressure control via Atmospheric Dump Valves * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor -1 Appendix D Scenario Form ES-D-1 Facility:
Three Mile Island Scenario No.: 4 Examiners:
Initial Conditions:
Operators:
- (Temporary IC-234)
- 100% Power, MOL
- MO-P-1 C and MO-P-1 F are OFF for work is occurring on the West Op Test No.: 10-02 NRC
- 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 T M aln. t'am 1DooA Po wer° i (URO: IMA's of OP-TM-EOP-001) 5 i Sheared shaft on RC-P-1A, entry into OP-TM-MAP-F0301 and OP-TM-226-151.(URO: Trips 1A 6900V 6 MCRS OTSG Tube Rupture on "A" OTSG with a Loss of Subcooling . Margin, entry into o P-TM-EOP-OOS, OP-TM-EOP-002. : M lIRO MARO : . Critical Tasks: Event No. Malf. No. 1 TH15A L i i 2 I ES08A 3 ED08S 4 TC01 *
- Trip all RCPs (CT-1)
- Minimize SCM (CT-7) Maintain SG availability (CT-29) Event Type* TSCRS RURO NARO ICRS ! Inadvertent 500# ESAS Signal, entry into OP-TM-AOP-046.
IURO i (URO: OP-TM-AOP-046IMA's.
ARO: Restores Letdown following IARO TSCRS : Loss of "8" DC, entry into OP-TM-AOP-024.
CARO CCRS CURO Event Description 20 gpm tube leak on "An OTSG, entry into OP-TM-EOP-005. (lIRO: commences a reactor shutdown with ICS in Manual, ARO: i Place FW-P-1N18 in HAND) an inadvertant ESAS signal) (ARO: Energizes 1 M DC from "N DC) Main Turbine Trip, Reactor does not automatically trip, Entry into OP-TM-EOP-001.
i I I CCRS EF-P-2A Trips, entry into OP-TM-EOP-010, Rule 4.7 I FW18A , CARO (ARO: Feeds OTSG's with Main Feedwater)
- (N}ormal, (R)eactivity, (I)nstru ment, (C}omponent, (M)ajor -1 i Appendix Scenario Outline Form ES-D-1 Facility:
Three Mile Island Scenario No.: 5 Op Test No.: 10-02 NRC Examiners:
Initial Conditions:
Turnover:
Critical Tasks: Event No. Malf. No. 1 DHR32 2 RC37A 3 MS12C 4 FW04B 5 RD0153 IC16 6 RC37A 7 TH06 8 02ABS28 02ABS22 Operators:. (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 Maintain 100% Reactor Power FW Flow Control (CT-16) Maintain RS Radiation Boundary (includes SG tubes) (CT-19)* Event Type* TSCRS CCRS CURO CCRS CARO ICRS IURO IARO TSCRS RURO NARO CCRS CURO CARO M CRS M LlRO MARO CCRS C URO Event Description BWST level lowers, entry into OP-TM-MAP-E0204 NSCCW Leak in RC-P-1A Motor Air Cooler, entry into MAP-F0201 (URO; Starts DW-P-1) Hi Level in Moist. Sep. Tank, entry into OP-TM-MAP-N0201 (ARO: Start MO-P-i C) FW Temperature transmitter failure, entry into OP-TM-AOP-070 (URO/ARO:
Controls reactivity and feedwater in manual. ARO: Controls Feedwater Flow in manual) Dropped Safety Rod, runback fails to occur, entry into H01 01, and OP-TM-AOP-OB2 (URO; Reactivity manipulation, ARO: Feedwater manipulation)
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 (URO: Reactor Trip IMA's) RCS LOCA, Loss of Subcooling Margin, entry into 002. NSCCW Containment Isolation valves fail to close on ES signal with low level. (URO: Manually closes NSCCW Containment Valves) * (N)ormal, (R)eactivity, (I)nstrument, {C)omponent, (M)ajor -1 i Three Mile Island Unit 1 Telephone Route 441 South, P.O. Box Middletown, PA December 27, 2011 TMI-11-156 U.S. NRC Region I Administrator 475 Allendale Road King of Prussia, PA 19406 Three Mile Island Unit Facility Operating License DPR NRC Docket No.
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 (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 no rod bank step counters 2.9/3.2 atTMI AK2.06 T-ave./ref.
deviation 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 3.4/3.9 lout indication Closing CCW surge lank ----*..-------.--------------.------
...--...---....... 3.1*' /3.6'" 2.3/2.6* 3.0*/3.3*
[ OIl Large Break LOCA Malfunctions 024 022 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 Emergency Boration
___ j Loss of AA2.03 The valve lineups necessary no procedural 2.6/2.9 restart the eews while actions for this bypassing the portion of the evolution system causing the abnormal
____________________________
L__________________ r'_
Pressure Control (PZR PCS) Malfunction AA2.17 Allowable ReS temperature not applicable to 3.1/3.3 difference vs. reactor power TMI 028 Pressurizer (PZR) Level Control Malfunction AAl.Ol PZR level reactor protection bistables no PZR level input to RPS at TMI 3.8*/3.9 AAl.04 I AAl.05 L Regenerative heat exchanger and temperature limits Initiation of excess letdown per the eves I no such component atTMI not applicable at TMI 2.7/2.8 2.8/2.9 i not performed at I TMJ BIT inlet valve of control rods into lhe not performed at TMI 4.0/3.6 4.2/3.9 EA2.10 3.1 */3.4* Range Nuclear Instrumentation Loss of Intermediate Range Nuclear Instrumentation 3.4*/3.7*
AK3.04 Tube Leak Collection of Condensate ejector monitor due to its Reset and check of air ejector exhaust
...---.*..
r;;-;;;:;--------------------------------------------
2.3/2.6 3.2/3.5 Steam Generator Tube Rupture (SGTR) RCS loop isolation values Steam flow indicators S/G into the RCS, using the "feed and bleed" method Causes and consequences of Vacuum no steam flow no RCS loop 3.4*/3.1-1 isolation valves at TMI shrink and swell in steam supply Loss of Main Feedwater (MFW) of feedwater and 3.4*/3.7*
...______+_ ..
-..... ....
...... --. trip first-out panel 3.4*/3.9 no such component at TMI 2.9/3.3*
055 Loss of Offsite and Onsite Power (Station Blackout)
AK3.t4 AK3.16 the feed water system closing the AFW pump valve 3.2*/3.4*
2.8*/3.3*
076 High Reactor Coolant Activity AlG.02 CCWtlow K2.04 K2.0S KS.l1 K..'l.71 KS.76 KS.97 One*line diagram of power to MIG sets Relationship between reactivity worth of shaping control rod group and other control rod groups (power-shaping, or sets for maintaining cross-tie breaker between rod drive MIG sets; reliability of control rod I drive trip breakers during set APSRs No longer have No longer have APSRs no T-ref at TMI no T-ref at TMI 1*/2.7 1/3.6* 3.4*/4.1
- 2.9*/2.9*
K6.10 ____
.............
no MIG sets at TMI 3.1 */3.3 no T-ref at TMI
...... "Prepower dependent insertion limit" and power dependent 3.l/3.4 no metroscope at 4.0?/4.2?
insertion limit, determined with metrosco e 3.2*/3.8*
l.9?/1.9 A2.04 A2.05 A2. +-' Positioning of axial shaping rods and their effect on SDM Fractured split 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
014 Protection System K4.07 First out indication
3.6*/3.7*
3.9*/3.9* 1.9*/2.2 2.5*/2.7*
2.1'/2.6 2.9*/3.1 2.6*/3.0*
2.6*/2.7*
026 _,Containment Spray System (eSS) no automatic 4.1 */4.3*Automatic swapover to containment sump suction K4.08 swapover at TMI recirculation phase after LOCA (RWST low-low level alarm) Prevention of path for escape of radioactivity from containment to the outside (interlock on RWST isolation after K4.09 Prevention of path for escape of radioactivity from containment to the outside (interlock on RWST isolation after swapover)
A4.02
- I no such interlock at 3.7*/4.1
- TMI no such components 2.3*/2.6*
atTMI 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 '---..
Purge Control System (HRPS) A3.01 A4.05 Moisture separator steam supply 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 T-ave. mode not applicable to TMI 2.9*/3.1 2.4*12.5 Main Turbine Generator (MT/G) System K4.10 045
__ 045 1.6/1.7 .........-................ Main Turbine Generator (MT/G) System K4.1S Steam blanketing (atmospheric not performed at pressure) moisture separator TMI reheater to drive out air and condensables prior to starting 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 overspeed trip test lever Generator amplidyne balance no such equipment 1.6*/US* system at TMI ,-_. no interface with atTMI ----........no air ejectors at T\1I no air ejectors at
- TMI Loss of air ejector cooling water Operation of hotwell pump and air ejector recirculation line valve to maintain no atTMI no such alTMI no such equipment atTMI no such equipment atTMI no such equipment 1.7*/1.6*
1.8"/1.7*
_
...._ ....at 1.9/1.9 1.7/1.7 L8*!2.0* i K2.03 L pump
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. 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 TMI f-----_. lee buildup on intake structure no intake structure 2.0*/2.0*A2.08 for Circ Water at TMI 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 I system 1.8*/1.8*1 A4.04 Air eductor system no such component C atTMI Water box vacuum priming no vacuum priming isolation valves, control for Circ Water at switches, and indicators TMI
075 Circulating Water Svstem A4.07 Vacuum priming tank/priming compressor controller no vacuum priming for Circ Water at TMI 1.7* /1.6
- A4.08 , Gland seal water supply system A4.14 Lube oil pumps for circulating water pump A4.lS Operation of the vacuum priming system no such component 1.6/1.6 atTMI no such component 1.5*il.7*
atTMI no vacuum priming 1.4/l.5 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 K4.04
___---,---:------:---i---------------t-River intake water level recorders Fire Protection K1.01 K1.02 Raw service vent system system to TMI not applicable to 2.7*/2.7*
TMI not applicable to 2.4*/2.2*
TMI not applicable to 2.4*/2.7*
2.1*/2.2*
-103 Containment System A4.07 A4.0S A4.09 Use of the air lock rate test panel Operation of refueling drain valves (for draining refueling canal to lower containment sump) Containment vacuum system no operated or monitored from the control room at TMI no operated or monitored from the control room at TMI not applicable to TMI 2.4 */2.5* 1.9/2.2 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. Generic K/As associated with emergency and abnormal plant (E/APE) and plant systems for both RO and SRO examinations should randomly selected from the following:
2.1.7,2.1.19,2.1.20,2.1.23,2.1.252.1.27, 2.1.28, 2.1.30, 2.1.31, 2.1.32, 2.2.3, 2.2.4. 2.2.12. 2.2.22, 2.2.36, 2.2.37, 2.2.38, 2.2.39, 2.2.40, 2.2.42, 2.2.44, 2.4.1, 2.4.2, 2.4.4, 2.4_6, 2.4.8, 2.4.9, 2.4.11. 2.4.18, 2.4.20, 2.4.21, 2.4.30, 2.4.34, 2.4.35.
2.4.41, 2.4.45, 2.4.46. 2.4.47, 2.4.49, and 2.4.50. All generic K/As for systems and evolutions may be suppressed.
The generic KlAs that can be suppressed for the generic section of the (Tier 3) are KlAs 2.2.3 and 2.2.4, but only at single-unit NOTE: TMI is a Single Unit Facility (2.2.3 and 2.2.4) should be suppressed)