ML14141A104
| ML14141A104 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 04/01/2014 |
| From: | Peter Presby Operations Branch I |
| To: | Exelon Generation Co |
| Jackson D | |
| Shared Package | |
| ML13333A179 | List: |
| References | |
| TAC U01895 | |
| Download: ML14141A104 (20) | |
Text
ES-401 PWR Examination Outline FORM ES-401-2 Facility Name: 17J,eel1t-fe &'k.Dife of Exam: ~/77 2o;L.j RO KlA Category Points SAO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4 *
- 1. Emergency 1 3 3 3 3 3 3 18 3 3 6 Abnormal 2 2 2 1 N/A 2 1 N/A 1 9 2 2 4 Plant Evolutions Tier Totals 5 5 4 5 4 4 27 5 5 10 1 3 2 3 3 3 2 3 3 2 2 2 28 3 2 5 2.
2 1 1 1 1 1 1 1 1 1 1 0 10 0 2 1 3 Plant Systems Tier Totals 4 3 4 4 4 3 4 4 3 3 2 38 5 3 8 1 2 3 4 1 2 3 4
- 3. Generic Knowledge and Abilities 10 7 Categories 3 3 2 2 2 2 2 1 Note: 1. Ensure that at least two topics from every applicable KJA category are sampled within each tier of the RO and SAO-only outlines (i.e., except for one category in Tier 3 of the SAO-only outline, the "Tier Totals" in each KJA 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 SAO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KJA statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating {lA) of 2.5 or higher shall be selected.
Use the RO and SAO ratings for the RO and SAO-only portions, respectively.
- 6. Select SAO topics for Tiers 1 and 2 from the shaded systems and KJA categories.
7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the KJA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the KJA numbers, a brief description of eactl topic, the topics' importance ratings {IRs) 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 SAO-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 SAO-only exams.
- 9. For Tier 3, select topics from Section 2 of the KJA catalog, and enter tl1e ~JA numbers, descriptions, IRs, and point totals {#) on Form ES-401-3. Limit SAO selections to K/As that are linked to 10 CFR 55.43.
ES-401, 21 of 33
ES-401 2 Form ES-401-2 PWR Examination Outline Form ES-401 Emergency and Abnormal Plant Evolutions- Tier 11Group 1 (RO) lr-----------------------~~~--r- -----------~~~-----------------r---r--~1 EIAPE #I Name I Safety Function K K K KIA Topic(s) IR #
2 3 Reactor Trip - Stabilization - Recovery I 1 0 inttl're,lations between a reactor trip and the lollowing: Breakers, 2.6 2
Vital System Status Verification I 1 the reasons for the following responses as they apply to the Pressurizer Vapor Space Accident I 3 3.7 Accident: Wh'l PZR level may come back on scale if RCS is Small Break LOCA I 3 3.4 0
i concepts as 5 RCP Malfunctions I 4 0 Pump Mall unctions (Loss of RC Flow): Basic steady state 2.9 7 RCP Malfunctions (loss of RC Flow) I 4 4 relationship between RCS loops and S/Gs resulting from unbalanced determine and interpret the following as they apply to the Loss of Reactor Loss of Rx Coolant Makeup I 2 3.2 Whether charging line leak exists 0 interrelations between the Loss of Residual Heat Removal System Loss of RHR System I 4 3.2 2 LPI or Deoay Heat RemovaVRHR pumps and I or monitor the following as they apply to the Loss of
3.6 Water
SWS as a backup to the CCWS Pressure Control System 0 2.8 2
4 Existence of Steam Gen. Tube Rupture I 3 4.2 Steam Line Rupture - Excessive Heat 0 the interrelations between the Steam Line Rupture and the following:
2.6 2
Excessive Heat Transfer I 4 Loss of Main Feedwater I 4 0 and monitor the lollowing as they apply to a Station Blackout:
4.3 operate and I or monitor the following as they apply to the Loss of Offsite Loss of Off-site Power I 6 3.2 of ED/G load by selectively energizing PZR backup heaters for the following responses as they apply to the Loss of Vital AC lnst. Bus I 6 Bus: Actions contained in EOP for loss of vital ac electrical 4.1 0 of the operational implications ol the following concepts as they apply to Loss of DC Power I 6 2.8 Power: Battery charger equipment and instrumentation Loss of Nuclear Svc Water I 4 0 Loss of Instrument Air I 8 4.1 4.2 Voltage and Electric 3.3 16 18 ES-401, 22 of 33
ES-401 3 Form ES-401-2 EIAPE #I Name I Safety Function IR #
2 3 Continuous Rod Withdrawal I 1 0 Dropped Control Rod I 1 0 of the interrelations between the Inoperable I Stuck Control Rod and the Inoperable/Stuck Control Rod I 1 01 2.5 I
Emergency Boration I 1 0 Pressurizer Level Malfunction I 2 3.7 of the following concepts as they apply to Loss of Source Range Nl I 7 01 Range NuciE!ar Instrumentation: Effects of voltage changes on 2.5 Loss of Intermediate Range Nl I 7 0 Fuel Handling Accident I 8 0
Refueling Canal Level Decrease I 8 Steam Generator Tube Leak I 3 0 Loss of Condenser Vacuum I 4 0 Accidental Liquid RadWaste Rei. I 9 0 Accidental Gaseous Radwaste Rei. 1 9 0 ARM System Alarms I 7 0 Plant Fire On-site I 9 8 3.6 for the following responses as they apply to the Control Control Room Evac. I 8 01 Sys1em re~sponse to reactor trip 3.9 Shutdown Outside Control Room I 8 Loss of CTMT Integrity I 5 0 lnad. Core Cooling I 4 0 High Reactor Coolant Activity I 9 0 Plant Runback I 1 0 Loss of NNI-X I 7 the operational implications of lhe following concepts as lhey apply to Loss of NNI-Y I 7 02 NNI-Y): Normal, abnormal and emergency operating procedures 3 i (Loss o' NNI-Y).
Turbine Trip I 4 0 between the (Emergency Diesel Actuation) and the and functions of control and safety systems, including Emergency Diesel Actuation I 6 01 signals, interlocks, failure modes, and automatic and manual 4
Flooding I 8 0 Inadequate Subcooling Margin I 4 0 LOCA Cooldown I 4 Natural Circulation Cooldown I 4 3.5 and IN monitor the following as they apply lo the (EOP Rules):
3 EOP Rules characteristics of the facility.
2.8 EOP Enclosures 2 2 9 ES-401, 23 of 33
ES-401 4 Form ES-401-2 Form ES-401 KIA Topic(s) IR #
AePS design feature(s) and/or interlock(s) which provide for i : A<Jequ&te cooling of AeP motor and seals; Knowledge of the 2.8; Reactor Coolant Pump 2 loss or malfunction on the following will have on the ACPS: AeP 2.7 seal water :;upply eves jesign feature(s) and/or interlock(s) which provide for
- ln:errelationships and design basis, including fluid flow splits in Chemical and Volume Control (e.g., charging and seal injection flow); Knowledge of 3.3; 4 2 Residual Heat Removal 3.7 the ope~rational implications of the following concepts as they Emergency Core Cooling ECGS: Tllermodynamics of water and steam, including 3.3 PRTS design feature(s) and/or interlock(s) which provide for Pressurizer Relief/Quench Tank : Quench tank cooling 2.6 Component Cooling Water 2.5 effect that a loss or malfunction of the PZA PeS will have AeS 3.8 3.3 3.6; 2
3.4 2.5; 2
4.1 relationships 4.1; and the following systems: Cooling water, Ability to 2
- and/or monitor in the control room: Containment reset 3.5 the operational implications of the following concepts as they 3.6; Main and Reheat Steam MASS: Efrect of steam removal on reactivity; Ability to monitor 2 of the MASS, including: Isolation of the MASS 3.1 the phy~ical connections and/or cause-effect relationships MF\V anc the following systems S/GS water level control system 3.4 of thE! effect that a loss or malfunction of the AFW will have on Auxiliary/Emergency Feedwater 4.4 following: AeS to predict and/ot monitor changes in parameters (to prevent exceeding limits) associatE!d with operating the ac distribution system controls
. Significance of D/G load limits; Ability to (a) predict the '1rnpacts of 3.4; AC Electrical Distribution malfunctions or operations on the ac distribution system; and (b) 2 predictions, use procedures to correct, control, or mitigate the 2.9 ICOiose.mJ<ence' of those malfunctions or operations: Consequences of Vlrhen transferring to or from an inverter DC Electrical Distribution 2.5 Emergency Diesel Generator 3.2 opetate and/or monitor in the control room: Radiation Process Radiation Monitoring control panel 3.7 predict and/or monitor changes in parameters (to prevent E!xceeding Service Water associated with operating the SWS controls including: Reactor 2.6 building closed cooling water temperatures the physical connections and/or cause-effect relationships lAS and the following systems: Cooling water to compressor 2.6 predict and/or monitor changes in parameters (to prevent exceeding 1 i associate:j with operating the containment system controls 3.7
- Containment pressure, temperature, and humidity Category Totals: 28 ES-401, 24 of 33
ES-401 5 Form ES-401-2 PWR Examination Outline Form ES-401 KIA Topic(s) IR #
of the following operational implications as they apply to the CADS:
inclucling effects of primary/secondary power miE.match on rod 3.2 0
manually opmate and/or monitor in the control room: Charging pump 1 Pressurizer Level Control controh; 3.5 0
3.3 Non-nuclear Instrumentation 0 7 In-core Temperature Monitor 0 predict th13 impacts of the following malfunctions or operations on and (b) based on those predictions. use Procedures to correct.
Containment Iodine Removal mitigate the consequences of those malfunctions or operations: High 3
tennn" "" in the filte*r system Recombiner and Purge 0
0 0
Fuel Handling Equipment 2.9 0
of the effect of a loss or malfunction on the following will have on the Steam Dump!Turbine Bypass Control II a*.nd positioners, including ICS. S/G, CADS 2.7 Main Turbine Generator 0 effect that a loss or malfunction of the CARS will have on the Condenser Air Removal 2.5 0
automatic operation of the Liquid Radwaste System including:
Liquid Radwaste 3.6 fe>ature(s) and/or interlock(s) which provide for the Waste Gas Disposal 2.9 1 of waste gas release tanks the physical connections and/or cause-effect relationships Area Radiation Monitoring ARM systE>m and the following systems: Fuel building isolation 3.6 Circulating Water 0 0
0 10 ES-401, 25 of 33
ES-401 2 Form ES-401-2 PWR Examination Outline Form ES-401 EIAPE #I Name I Safety Function KIA Topic(s) IR #
2 Reactor Trip - Stabilization - Recovery I 1 determine and interpret the following as they apply to the (Vrtal System Vital System Status Verification I 1 Ver*ific<>tionl: Facility conditions and selection of appropriate procedures during 4 operations.
0 Post Trip Stabilization I 1 Pressurizer Vapor Space Accident I 3 0 0
of the operational implications of EOP warnings, cautions, and notes. 4.3 5 RCP Malfunctions I 4 0
7 RCP Malfunctions (Loss of RC Flow) 1 4 Loss of Rx Coolant Makeup I 2 0 Loss of RHR System I 4 0 0
Pressure Control System 0
0 Steam Gen. Tube Rupture I 3 0 Steam Line Rupture - Excessive Heat interpret control room indications to verify the status and operation of a Excessive Heat Transfer I 4 understand how operator actions and directives affect plant and system 4.4 and interpret the following as they apply to the Loss of Main Loss of Main Feedwater I 4 3.7 Status of MFW pumps, regulating and stop valves 0
Loss of Off-site Power I 6 of the specific bases for EOPs. 4 Loss of Vital AC lnst. Bus I 6 0 Loss of DC Power I 6 0 and interoret the following as they apply to the Loss Loss of Nuclear Svc Water I 4 normal values for SWS-header flow rate and the flow rates to the 2.5 the S>IVS Loss of Instrument Air I 8 0 14 0 Voltage and Electric 0
16 0 0 roup Point Total: 6 ES-401, 22 of 33
ES-401 3 Form ES-401-2 EIAPE #I Name I Safety Function IR #
2 Continuous Rod Withdrawal/! 0 Dropped Control Rod /1 0 Inoperable/Stuck Control Rod /1 0 Emergency Boration /1 0 Pressurizer Level Malfunction I 2 0 Loss of Source Range NI I 7 0 Loss of Intermediate Range Nl/ 7 0 Fuel Handling Accident/ 8 Refueling Canal Level Decrease I 8 4 Steam Generator Tube Leak I 3 0 Loss of Condenser Vacuum I 4 0 Accidental Liquid RadWaste Rei. I 9 0 and recognize trends in an accurate and timely manner utilizing Accidental Gaseous Radwaste Rei. I 9 control room 1*eference material.
4.2 ARM System Alarms I 7 0 Plant Fire On-site I 9 8 0 perform specific sys1em and integrated plant procedures during all modes Control Room Evac. I 8 4.4 Shutdown Outside Control Room I 8 Loss of CTMT Integrity I 5 0 lnad. Core Cooling /4 0 High Reactor Coolant Activity I 9 0 Plant Runback /1 0 Loss of NNI-X /7 0
Loss of NNI-Y /7 Turbine Trip I 4 0 Emergency Diesel Actuation I 6 0 Flooding /8 0 Inadequate Subcooling Margin I 4 0 LOCA Cooldown I 4 0
Natural Circulation Cooldown I 4 determine and interpret the following as they apply to the (EOP Rules):
3 EOP Rules and selection of appropriate procedures during abnormal and 4 4 EOP Enclosures 0 0 0 4 ES-401, 23 of 33
ES-401 4 Form ES-401-2 PWR Examination Outline Form ES-401 (f-------------.,-""T""-r"""T--.-r-P:.,:I:;::a:..::..nt Systems- Tier 2_1G_ro_u.:...p_1_;(~S__R_O.:..)_ _ _ _ _ _ _ _ _ _ _ _ _.......,,......._"T""""_--II KIA Topic(s) IR #
0 Chemical and Volume Control 0 Residual Heat Removal 0 Emergency Core Cooling 0 analyze the effect of maintenance activities, such as degraded power Pressurizer Relief/Quench Tank 4.2 on the statu3 of limiting conditions for operations.
Component Cooling Water 0 0 Pressurizer Pressure Control 0 predict tt e impacts of the following malfunctions or operations on
- and (b) bas~d on those predictions, use procedures to correct, 3.7 0
0 Ice Condenser 0 Containment Spray 0 predict the impacts of the following malfunctions or operations on and lb) based on predictions, use procedures to correct, control, or Main and Reheat Steam conseque:nces of those malfunctions or operations: Increasing 3.6 1ts rel3.tionship to increases in reactor power 0
Auxiliary/Emergency Feedwater 0 AC Electrical Distribution 0 of thB bas>3s in Technical Specifications for limiting conditions for DC Electrical Distribution 4.2 safety limits.
Emergency Diesel Generator 0 Process Radiation Monitoring 0 predict the impacts of the following malfunctions or operations on
- and (b) based on those predictions, use procedures to correct, Service Water 3.1 or mitigate the consequences of those malfunctions or operations
water header :)ressure 0
0 Category Totals: 5 ES-401, 24 of 33
ES-401 5 Form ES-401-2 PWR Examination Outline Form ES-401 KIA Topic(s) IR #
0 Reactor Coolant 0 1 Pressurizer Level Control 0 4 Rod Position Indication 0 to (a) predict the impacts of the following malfunctions or operations on and (b basecl on those predictions, use procedures to correct, control, Nuclear Instrumentation the consequences of those malfunctions or operations: Faulty or 3.5 of dHtectors or compensating components predict the impacts of the following malfunctions or operations on and (b) ba~.ed on those predictions, use procedures to correct, Non-nuclear Instrumentation mitigate thn consequences of those malfunctions or operations:
3.1 failure In-core Temperature Monitor 0 0
Recombiner and Purge 0
0 0
Fuel Handling Equipment 0 0
Steam Dump!rurbine Bypass Control 0 Main Turbine Generator 0 Condenser Air Removal 0 0
Liquid Radwaste 0 Waste Gas Disposal 0 Area Radiation Monitoring 0 Circulating Water 0 0
diagnose ancl recognize trends in an accurate and timely manner 4.2 appropriat,3 control room reference material.
Category Totals: 3 ES-401, 25 of 33
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility Name: Date of Exam:
HU :SHU-Only Category KIA# Topic IR # IR #
12.1. 20 lAbility to interpret and execute procedure steps. 4.6 1 4.6 12.1. 36 Knowledge of procedures and limitations involved in core alterations. 3 1 4.1
.1. 12.1. 42 Knowledge of new and spent fuel movement procedures . 2.5 1 3.4 Conduct of 12.1. 34 Knowledge of primary and secondary plant chemistry limits. 2.7 3.5 1
,OpeldliUII' 12.1. 41 Knowledge of the refueling process. 2.8 3.7 1 12.1.
ISubtotal 2 lAbility to perform pre-startup procedures for the facility, including opera1ing those controls 12.2. 01 *~~nri<>t<><j with plant equipment that could affect reactivity. 4.5 1 4.4 12.2. 13 Knowledge of tagging and clearance procedures. 4.1 1 4.3 12.2. 22 Knowledge of limiting conditions for operations and safety limits. 4 1 4.7 2.
lAbility to analyze the effect of maintenance activities, such as degraded power sources, on Equipment 12.2. 36 3.1 4.2 1 Ithe status of limiting w"uuiv~r Opell"_ations._
Comrol
[Ability to interpret control room indications to verify the status and operation of a system, 12.2. 44 land 4.2 4.4 1 how operator actions and directives affect plant and system conditions.
12.2.
lfuJbtotal 2 12.3. 11 lAbility to control radiation releases. 3.8 1 4.3
!Knowledge of radiation , *v* ,;,v, , <! systems, such :fixed radiation monitors and alarms, 12.3. 15 2.9 1 3.1 Iportable~~i~stn*m<>nk-"" "" "'" """ eqtjipmen_l._etc.
- 3. 12.3. 04 Knowledge of radiation exposure limits under normal or emergency conditions. 3.2 3.7 1 C* ..J:. Knowledge of radiation or contamination hazards that may arise during normal, abnormal,
"*"'ALIUII 12.3. 14 or emergency conditions or activities. 3.4 3.8 1 Control 12 . 3.
12.3.
Subtotal 2 2.4. 03 lAbility to identify post-accident instrumentation. 3.7 1 3.9 2.4. 17 Knowledge of EOP terms and definitions. 3.9 1 4.3
- 4. 2.4. 40 Knowledge of SRO responsibilities in emergency plan implementation. 2.7 4.5 1
,Emergency
'n., 2.4.
'Plan 2.4.
12.4.
ISubtotal 1 Tier 3 Point Total 7 ES-401, Page 26 of 33
ES-401 Record of Rejected KlA's Form ES-401-4 Tier I Group Randomly Selected KA Reason for Rejection 001 I AK1.19 replaced by 1I 2 The subject KIA isn't relevant at the subject facility.
A03 I AK1.2 006 I K6.19 replaced by Topic oversampled. 006IK6.19 overlaps with KIA 005 I K3.05 2I 1 064 I K6.08 on the Written Exam.
073 I A4.01 replaced by Topic oversampled. 073IA4.01 overlaps with KIA 057/ AK3.03 2I 1 004 I K4.05 and 068 I A3.02 on the Written Exam.
078 I A3.01 replaced Topic oversampled. 078 I A3.01 overlaps with KIA 065 I 2.4.45 211 by 039 I A3.02 and 078 I K1.04 on the Written Exam.
003 I 2.2.37 replaced by 2I 1 The subject KIA isn't relevant at the subject facility.
061 I 2.4.30 061 I 2.4.30 replaced by 2I 1 Question could not be written at the RO level.
062 I A1.01 017 I K3.01 replaced by Topic oversampled. 017 I K3.01 overlaps with 038 I EA2.09 on 212 055 I K3.01 the Written Exam.
079 I 2.2.39 replaced by 212 The subject KIA isn't relevant at the subject facility.
015 I K2.01 008 I 2.4.9 replaced by Topic oversampled. 008 I 2.4.9 overlaps with 026 I AA1.03, 008 2 I 1 SRO 063 I 2.2.25 I A2.08, 013 I K2.01 and 026 I K1.02 on the Written Exam.
010 I A2.01 replaced by Topic oversampled. 010 I A2'.01 overlaps with 056 I AA 1.03, 2 I 1 SRO 012 I A2.03 016 I A2.01, and 028 I ~!.4.6 on the Written Exam.
034 I K5.03 replaced by Topic oversampled. 034 I K5.03 overlaps with 025 I AK2.02, 212 SRO 016 I A2.01 and G 2.1.36 on the Written Exam.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: April 2014 Examination Level: RO [gl SRO 0 Operating Test Number: 289-2014-301 Administrative Topic Type Describe activity to be performed (See Note) Code*
Calculate an Estimated Critical Boron Concentration in accordance with 1103-158, ESTIMATED CRITICAL CONDITIONS.
Conduct of Operations M/R 2.1.25 (3.9): Ability to interpret station reference materials such as graphs, curves, tables, etc.
Perform OP-TM-300-202, QUADRANT POWER TILT AND CORE IMBALANCE USING THE OUT-OF-CORE DETECTOR SYSTEM. and compare to COLR limits.
Conduct of Operations M/R 2.1.37 (4.3): Knowledge of procedures, guidelines, or limitations associated with reactivity management.
Reactor Coolant Pump eloctrical print reading to determine pump operation.
Equipment Control N/R 2.2.41 (3.5): Ability to obtain and interpret station electrical and mechanical drawin~
Radiation Control Category not selected *for RO applicants.
ERO Notification Emergency Procedures/Plan M/S 2.4.39 (3.9): Knowledge of RO responsibilities in emergency plan implementation.
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (s 3 for ROs; s 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (?. 1)
(P)revious 2 exams (s 1; randomly selected)
ES 301, Page 22 of 27
2014 TMI RO NRC EXAMINATION CONDUCT OF OPERATIONS: Given a set of plant conditions, calculate the Estimated Critical Boron Concentration for reactor startup. Modified Bank JPM.
CONDUCT OF OPERATIONS: Given a failed computer, perform OP-TM-300-202, QUADRANT POWER TILT AND CORE IMBALANCE USING THE OUT-OF-CORE DETECTOR SYSTEM, and compare to COLR limits. Modified Bank JPM. Calculation is RO/SRO common.
EQUIPMENT CONTROL: Given a set of conditions and Reactor Coolant Pump electrical prints, determine if and where the Reactor Coolant Pump can be operated. New JPM.
EMERGENCY PROCEDURES/PLAN: Given a General Emergency declaration and a failure of the ERO Notification using the World Wide Web, initiate activation of the ERO using the Live Everbridge Agent.
Modified Bank JPM.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: April 2014 Examination Level: RO 0 SRO [8J Operating Test Number: 289-2014-301 Administrative Topic Type Describe activity to be performed (See Note) Code*
Review and Approve an Estimated Critical Boron Concentration.
Conduct of Operations M/R 2.1.25 (4.2): Ability to interpret station reference materials such as graphs, curves, tables, etc.
Review OP-TM-300-202, OUADRANT POWER TILT AND CORE IMBALANCE USING THE OUT-OF-CORE Conduct of Operations M/R DETECTOR SYSTEM, and compare to COLR limits.
2.1.37 (4.6): Knowledge of procedures, guidelines, or limitations associated with reactivity management.
Reactor Coolant Pump electrical print reading to determine pump operation and Tech Spec implications.
Equipment Control N/R 2.2.41 (3.9): Ability to obtain and interpret station electrical and mechanical drawin~
Implement the Requirements of ODCM for RMS Operability.
Radiation Control M/R 2.3.15 (3.1) Knowedge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
Emergency Action Level Identification, Event Declaration, and Protective Action Recommendation.
Emergency Procedures/Plan N/R 2.4.44 (4.4): Knowledge of emergency plan protective action recommendations.
NOTE:
All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (.:s 3 for ROs; :5. 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (~ 1)
(P)revious 2 exams (.:s 1; randomly selected)
ES 301, Page 22 of 27
2014 TMI SRO NRC EXAMINATION CONDUCT OF OPERATIONS: Review and approve an Estimated Critical Boron Concentration calculation with multiple errors. Modified Bank JPM.
CONDUCT OF OPERATIONS: Given a failed computer, perform OP-TM-300-202, QUADRANT POWER TILT AND CORE IMBALANCE USING THE OUT-OF-CORE DETECTOR SYSTEM, and compare to COLR limits. Modified Bank JPM. Calculation is RO/SRO common.
EQUIPMENT CONTROL: Given a set of conditions and Reactor Coolant Pump electrical prints, determine if and where the Reactor Coolant Pump can be operated and any Tech Spec implications.
NewJPM.
RADIATION CONTROL: Given a set of conditions with a gas release in progress and various RMS components out of service, determine the actions to be taken lAW facility procedures. Modified Bank JPM EMERGENCY PROCEDURES/PLAN: Given a set of conditions, determine and declare the appropriate Emergency Action Level and make a Protective Action Recommendation lAW with the TMI Emergency Plan. New JPM.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
~
Facility: Three Mile Island Date of Examination: APR 2014 Exam Level: RO ~ SR0-1 D SRO-U D Opelrating Test Number: 289-2014-301
~
Control Room Systems@ (8 for RO); (7 for SR0-1); (2 or 3 for SRO-U, including 1 ESF)
Safety System I JPM Title Type Code*
Function
- a. Feed from the "C" RCBT and the BAMT (001 A4.02) N/S 1
- b. Manually Initiate ESAS (006 A2.12) DIAlS 2
- c. Restore Seal Injection with a Loss of ICCW (003 K6.02) N/A/S 4P
- e. Perform Emergency Operations of Reactor Building Emergency Cooling M/A/S 5 Water (Sys 022) A4.04
- f. Return 1C 480V Bus to the Normal Power Supply (Sys 062) A4.01 N/S 6
- g. Respond lAW OP-TM-MAP-C0101 with Failure (Sys 072) A3.01 A/P/S 7
- h. Initiate and Isolate a Reactor Building Purge (Sys 029 K1.01) D/UAIS 8 In-Plant Systems@ (3 for RO); (3 for SR0-1); (3 or 2 for SRO-U)
- j. Locally/Manually Operate a Turbine Bypass Valve (Sys 041) A4.08 D/E 4S
- k. Pressurize the Core Flood Tanks (Sys 006) A 1.13 D/R 2
@ All RO and SR0-1 control room (and in-plant) systems must be diffemnt 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 SR0-1 I SRO-U (A)Iternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank ::::_9/ .::~8 I ::_ 4 (E)mergency or abnormal in-plant ~ 1I 2:: 1 I ~ 1 (EN)gineered safety feature - I - I .:::. 1 (control room system (L)ow-Power I Shutdown ~ 1 I .:::1 I ~ 1 (N)ew or (M)odified from bank including 1(A) .:::.2/ .:::2 I ~ 1 (P)revious 2 exams ::::_ 3 I ~3 I : : _ 2 (randomly selected)
(R)CA .:::. 1 I .::. 1 I .:::. 1 (S)imulator ES-301, Page 23 of 27
ES-301 Control Room/In-Plant Systems Out_li_n_e_ _ _ _F_o_r_m_E._S_-3_0_1_-_2 Facility: Three Mile Island Date of Examination: APR 2014 Exam Level: RO D SR0-1 1Z! SRO-U D Operating Test Number: 289-2014-301 Control Room Systems@ (8 for RO); (7 for SR0-1); (2 or 3 for SRO-U, including 1 ESF)
Safety System I JPM Title Type Code*
Function a.
- b. Manually Initiate ESAS (006 A2.12) DIAlS 2
- c. Restore Seal Injection with a Loss of ICCW (003 K6.02) N/A/S 4P
- e. Perform Emergency Operations of Reactor Building Emergency Cooling M/A/S 5 Water (Sys 022) A4.04
- f. Return 1C 480V Bus to the Normal Power Supply (Sys 062) A4.01 N/S 6
- g. Respond lAW OP-TM-MAP-C0101 with Failure (Sys 072) A3.01 A/P/S 7
- h. Initiate and Isolate a Reactor Building Purge (Sys 029 K1.01) D/UAIS 8 In-Plant Systems@ (3 for RO); (3 for SR0-1); (3 or 2 for SRO-U)
- j. Locally/Manually Operate a Turbine Bypass Valve (Sys 041) A4.08 D/E 4S
- k. Pressurize the Core Flood Tanks (Sys 006) A 1.13 D/R 2
@ All RO and SR0-1 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 SR0-1 I SRO-U (A)Iternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank ~9/ ~~8 I ~4 (E)mergency or abnormal in-plant ~1 I .:::1 I ~ 1 (EN)gineered safety feature - I - I ~ 1 (control room system (L)ow-Power I Shutdown ~ 1 I ~1 I ~1 (N)ew or (M)odified from bank including 1(A) ~2/ .::.2 I ~ 1 (P)revious 2 exams ~3/ s3 I ~ 2 (randomly selected)
(R)CA ~ 1 I >1 I ~1 (S)imulator ES-301 , Page 23 of 27
Appendix D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 1 Op Test No.: 289-2014-301 Examiners: Operators:
Initial Conditions: * (Temporary IC-241)
- 100% Power, MOL
- SBO OOS For Maintenance, expected to return to service in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
- Crane work is occurring on the West side of the Plant to stage new piping .
- NRC Authorization Code for today is AB12 .
Turnover: Maintain 100% Reactor Power Critical Tasks:
- Minimize SCM (CT-7)
- Limit Uncontrolled Radiation Release (CT-21)
- Reduce Steaming/Isolate Affected SGs (CT*22)
Event No. Malf. No. Event Type* Event Description 1 MSR01 CCRS MSIV Inadvertent Closure, entry into OP-TM-PPC-L2204.
CURO CARO (URO: Lowers power in ICS Auto, ARO: Opens MS-V-1A) 2 RW02C TSCRS NR-P-1 C Trips, NR-P-1 B Fails to Auto-Start, entry into OP-TM-MAP-B0105, and OP-TM-MAP-B0205 CARO (ARO: Starts NR-P-1 B from CR) 3 MU19D CCRS Reactor Coolant Pump Seal Leakage, entry into OP-TM-AOP-040.
CURO (URO: Lower Reactor Power, ARO: Manually Control Feedwater Pumps)
CARO 4 ED09D TSCRS Loss of Vital Bus D, entry into OP-TM-AOP-018 CARO (ARO: Restore Control Building Ventilation) 5 IC20 ICRS Total FW Demand Fails to Zero Volts, ICS Transient, entry into OP-TM-AOP-070.
IURO (URO/ARO: Coordinate to stabilize plant in ICS HAND control)
IARO 6 TH17B TSCRS -30 gpm "B" OTSG Tube Leak (TS), entry into OP-TM-EOP-005 R URO (URO: Guide 9)
NARO 7 RC39A MCRS RC-P-1 B Trip, Reactor Trip, -500 gpm "B" OTSG Tube Rupture TH16B with an elevated offsite dose, entry into OP-TM-EOP-005 MURO MARO 8 MUR67 CCRS MU-V-36 Fails to Open MUR94 CURO (URO: Maintains HPI flow greater than 115 GPM)
- (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Scenario Event Description NRC Scenario 2 Facility: Three Mile Island Scenario No.: 2 Op Test No.: 12-01 NRC Examiners: Operators:
Initial Conditions: * (Temporary IC-242)
- 100% Power, MOL
- SBO OOS For Maintenance, expected to return to service in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
- Crane work is occurring on the West side of the Plant to stage new piping .
- NRC Authorization Code for today is AB12 .
Turnover: Maintain 100% Power Operations Critical Tasks:
- Isolate Possible RCS Leak Paths (CT-3)
- Natural Circulation RCS Flow (CT-12)
- Turbine Trip (CT-18)
Event No. Malt. No. Event Type* Event Description 1 DHR32 TSCRS BWST level lowers, entry into OP-TM-MAP-E0204 I
2 MS19A C CRS lsolable Steam Leak in Turbine Blclg, entry into OP-TM-AOP-051.
C ARO (ARO: Isolate Steam Leak) 3 IC23 ICRS SG/RX Demand Station fails to 0 Volts, Entry into OP-TM-AOP-070 IURO (URO/ARO: Coordinate to stabilize plant in ICS HAND control)
IARO 4 TU01D NCRS High Vibrations on Main Turbine, entry into OP-TM-MAP-K0201 and 1102-4.
R URO (URO/ARO: Power reduction with ICS in Manual)
NARO 5 ZDIPB1 TSCRS Inadvertent 1600# ESAS Signal, entry into OP-TM-AOP-046.
RCA IURO IARO (URO:Immediate Manual Actions, ARO: Restores Letdown) 6 ED05D TSCRS Loss of 1D 4Kv Bus, entry into OP-TM-AOP-013.
C ARO (ARO: Places LO-P-6 in PTL) 7 EG04A I CRS Loss of Stator Coolant Pumps, Main Turbine fails to automatically runback and trip EG04B I URO (URO: Trip Reactor, ARO: Adjust Main Feedwater) 8 ED01 MCRS Loss of Offsite Power, entry into OP-TM-AOP-020.
MURO MARO
~
9 EG07A CCRS EG-Y-1 B Trips, Station Blackout C URO (URO: Isolates Cooling paths to RCP's)
- (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Facility: Three Mile Island Scenario No.: 4 Op Test No.: 289-2014-301 Examiners: Operators:
Initial Conditions: * (Temporary IC-244)
- 85% Power, MOL
- I&C Maintenance is occurring on HSPS, Train B, currently testing EF-V-30B .
- SBO OOS for Maintenance, expected to return to service in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> .
- Crane work is occurring on the West side of the Plant to stage new piping .
- NRC Authorization Code for today is AB12 .
Turnover: Maintain 85% Reactor Power Critical Tasks:
- PORV Control for Heat Transfer (CT-13) (If conditions are met)
- Shutdown Reactor- ATWS (CT-24)
- Restore Feed to a Dry OTSG (CT-26)
Event No. Mall. No. Event Type* Event Description 1 IA08 TSCRS Instrument Air Leak Requiring Isolation of "A" Side 2-Hour Air and "A" EFW Valves, entry into OP-TM-AOP-028 (ARO:Start IA-P-1 AlB)
CARO 2 IC38B CCRS Invalid "B" OTSG Low Level, "B" EFW inadvertent actuation.
CARO (ARO: defeats invalid signal, secures EF-P-2B) 3 RD10B I CRS Uncontrolled Inward Rod Motion.
IURO IARO (URO: Assumes Manual Control of Control Rods) 4 TSCRS MU-V-18 Fails Closed, entry into Guide 9.
C URO (URO: Controls Pressurizer Level with HPI valve) 5 FW15A CCRS "A" Main Feed Pump Trips, Manual runback required.
R URO NARO (URO: Runback in Manual, ARO: Runback in Manual) 6 TH18B CCRS Sheared Shaft on RC-P-1 B CURO CARO (URO: Secures RC-P-1 B, ARO: re-ratios Main Feedwater) 7 RD28 MCRS "B" Main Feed Pump Runs to 0 rpm, ATWS, Lack of Primary to Secondary RD32 Heat Transfer.
MURO MARO 8 CCRS EFW Control Valves fail to operate, EF-V-52A-D Closed CARO (ARO: Establish PSHT via Condensate Booster Pump flow) 9 CCRS MU-P-1 AIC will not start, MU-P-"18 trips.
C URO (URO: Establish PORV control for Heat Transfer)
- (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor