ML12159A258

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Draft - Outlines (Folder 2)
ML12159A258
Person / Time
Site: Oyster Creek
Issue date: 06/01/2012
From: D'Antonio J
Operations Branch I
To:
Exelon Generation Co
Jackson D
Shared Package
ML120230007 List:
References
50-219/12-301, ES-401, TAC U01848 50-219/12-301
Download: ML12159A258 (29)
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{{#Wiki_filter:ES-401 Written Examination Outline Form ES-401-1 Facility: Oyster Creek Date of Exam: 05/14112 RO KJA Category Points SRO~On1y 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

  • i L 1 3 4 3 3 4 3 20 3 4 7 I Emergency 2 1 1 1 2 1 1 7 2 1 3
    &                                                                                                                            I Plant                                                                                                                         I Tier Evolutions                4      5    4                      5     5                 4      27         5          5         10 Totals                                                                                                            i 1       2    3     3     2     3    2      2     2   2      3     2      26         2         3          5    I 2.

Plant 2 2 1 1 1 1 1 1 1 1 1 1 12 0 2 1 3  ! Systems Tier 4 4 4 3 4 3 3 3 3 4 3 38 4 4 8 Totals

3. Generic Knowledge & Abilities 1 2 3 4 1 2 3 4 10 7 Categories 3 2 3 2 2 2 2 1 Note: L Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA category shall not be less than two).
2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/- I from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section D.l.b of ES-401, for guidance regarding elimination ofinappropriate KIA statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

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

8. On the following pages, enter the KIA numbers, a briefdescription 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. Iffuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note
               #1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 ofthe KIA Catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KlAs that are linked to IOCFR55.43

ES-401 2 Form ES-401-1 ILT 11-1 NRC Written Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # I Name Safety Function KiA Topic(s) AA2.02 - Ability to detennine and/or 295023 Refueling Acc Cooling interpret the following as they apply to X 4.6 J Model 8 REFUEUNG ACCIDENTS: Entry conditions of eme\lIency plan EA2.04 - Ability to detennine and/or 295031 Reactor Low Water Levell interpret the foUowing as they apply to X 4.8 2 2 REACfOR LOW WATER LEVEL: Adequate core cooling AA2.05 - Ability to detennine and/or 295021 Loss of Shutdown Cooling interpret the following as they apply to X 3.5 3 14 LOSS OF SHUTDOWN COOUNG : Reactor vessel metal temperature 295026 Suppression Pool High 7 - Equipment Control: Ability to X .ne operability and 1 or availability of 4.6 4 Water Temp. 1 5 related equipment 295018 Partial or Total Loss of 2.2.22 - Equipment Control: Knowledge of CCW/S X limiting conditions for operations and safety 4.7 5 limits. 2.2.3S - Equipment Control: Knowledge of 295004 Partial or Total Loss of DC X conditions and limitations in the facility 45 6 Pwr/6 license. 600000 Plant Fire On-site I 8 X 2.4.29 - Knowledge of the emergency plan. 4.4 7 AKI.03 - Knowledge of the operational 295021 Loss of Shutdown Cooling implications ofthe following concepts as X 3.9 39 14 they apply to LOSS OF SHUTDOWN COOLING: Adequate core cooling I EKI.02 - Knowledge ofthe operational 295030 Low Suppression Pool implications ofthe following concepts as Water Levell 5 X they apply to LOW SUPPRESSION POOL 3.5 40 I WATER LEVEL: Pump NPSH AKI.O I - Knowledge ofthe operational 295023 Refueling Acc Cooling implications ofthe following concepts as ModelS X they apply to REFUELING ACCIDENTS : 3.6 41 Radiation exposure hazards AK2.03 - Knowledge ofthe interrelations 600000 Plant Fire On-site I 8 X between PLANT FIRE ON SITE and the 2.5 42 following: Motors EK2.04 - Knowledge of the interrelations between HIGH REACTOR PRESSURE 295025 High Reactor Pressure I 3 X and the following: ARIIRPT/ATWS: 3.9 43 Plant-Specific AK2.02 - Knowledge of the interrelations 295006 SCRAM 11 X between SCRAM and the following: 3.8 44 Reactor water level control system AK3.01 - Knowledge of the reasons for 700000 Generator Voltage and the follOwing responses as they apply to Electric Grid Disturbances X GENERATOR VOLTAGE AND 3.9 45 ELECTRIC GRID DISTURBANCES: Reactor and turbine trip criteria EK3.01 - Knowledge of the reasons for the following responses as they apply to 295037 SCRAM Conditions SCRAM CONDITION PRESENT AND Present and Reactor Power Above X REACTOR POWER ABOVE APRM 4.1 46 APRM Downscale or Unknown I 1 DOWNSCALE OR UNKNOWN: Recirculation pump trip/runback: Plant-Specific 295005 Main Tutbine Generator AK3.04 - Knowledge of the reasons for TripI 3 X the following responses as they apply to 3.2 47 MAIN TURBINE GENERATOR TRIP:

ES-401 2 Form ES-401-1 ILT 11-1 NRC Written Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # / Name Safety Function KIA Topic(s) Main generator trip AA1.03 - Ability to operate and/or 295016 Control Room monitor the following as they apply to X 3.0 48 Abandonment 17 CONTROL ROOM ABANDONMENT: RPIS EA1.17 - Ability to operate and/or monitor the following as they apply to 295024 High Drywell Pressure I 5 X 3.9 49 HIGH DRYWELL PRESSURE: Containment spray: Plant-Specific EA1.03 - Ability to operate and/or 295028 Higb Drywell Temperature monitor the following as they apply to X 3.9 50 15 HIGH DRYWELL TEMPERATURE: Drywall coolina system AA2.01 - Ability to determine and/or 295001 Partial or Complete Loss interpret the following as they apply to of Forced Core Flow Circulation j 1 X PARTIAL OR COMPLETE LOSS OF 3.5 51 

&4                                                          FORCED CORE FLOW CIRCULATION
Power/flow map AA2.03 - Ability to determine and/or 295004 Partial or Total Loss of interpret the following as they apply to X 2.& 52 DC Pwrl 6 PARTIAL OR COMPLETE LOSS OF D.C. POWER: Battery voltage EA2.01 - Ability to determine and/or 295031 Reactor Low Water Level interpret the following as they apply to X 4.6 53 12 REACTOR LOW WATER LEVEL:

Reactor water level 2.4.18 - Emergency Procedures / Plan: 295026 Suppression Pool High X Knowledge of the specific bases for 54 Water Temp. I 5 EOPs. 2.4.50 - Emergency Procedures / Plan: - 295019 Partial or Total Loss of Ability to verify system alarm setpoints X Inst. Air/& and operate controls identified in the alarm response manual. 2.1.23 - Conduct of Operations: Ability 295018 Partial or Total Loss of to perform specific system and X 56 CCW/8 integrated plant procedures during all modes of plant operation. AA2.02 - Ability to determine and/or interpret the following as they apply to 295003 Partial or Complete Loss X PARTIAL OR COMPLETE LOSS OF 4.2 57 ofACJ6 A.C. POWER: Reactor power, pressure, and level EK2.06 - Knowledge of the interrelations 295038 High Off-site Release Rate between HIGH OFF-SITE RELEASE X 3.4 58 19 RATE and the following: Process liquid 

                                                        ~ radiation monitoring system KIA CategolY Totals:                3 4    3   3   4/3                      Group Point Total:

I 2017

ES-401 3 Form ES-401-1 ILT 11-1 NRC Written Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE # I Name Safety Function KIA Topic(s) AA2.03

  • Ability to detennine andIor 295009 Low Reactor Water level I interpret the following as they apply to X 2.9 8 2 LOW REACTOR WATER LEVEL:

Reactor water cleanup blowdown rate 2.4.41 - Emergency Procedures I PIan: 295036 Secondary Containment Ability to diagnose and recognize trends in High Sump/Area Water Levell 5 X an accurate and timely manner utilizing the 4.2 9 

                                                                  . te control room reference material.

AA2.04 - Ability to determine andIor interpret the following as they apply to 295014 Inadvertent Reactivity Addition I I X INADVERTENT REACTIVITY 4.4 10 ADDITION: Violation of fuel thermal limits AKI.03 - Knowledge of the operational implications of the following concepts as 3. 295015 Incomplete SCRAM 11 X they apply to INCOMPLETE SCRAM: 59 8 Reactivity effects AK2.08 - Knowledge of the interrelations between INADVERTENT 295020 Inadvertent Cont. 2.5 X CONTAINMENT ISOLATION and the 60 Isolation / 5 & 7 following: Traversing in-core probes: Plant-Specific EK3.01 - Knowledge of the reasons for the following responses as they apply to HIGH 295032 High Secondary 3. Containment Area Temperature / 5 X SECONDARY CONTAINMENT AREA 5 61 TEMPERATURE: Emergency/normal depressurization AA1.02

  • Ability to operate andlor monitor 295012 High Drywcll the fonowing as they apply to HIGH 3.

Temperature / 5 X 62 DRYWELL TEMPERATURE: Drywell 8 

                                                            -" system AA2.04 - Ability to determine andlor 295008 High Reactor Water Level                          interpret the fonowing as they apply to         3.

/2 X HIGH REACTOR WATER LEVEL: 63 1 Heatup rate: Plant-Specific 2.2.12 - Equipment Control: Knowledge of 3. 295001 High Reactor Pressure 13 X surveillance procedures. 1 64 AAI.lO - Ability to operate andIor monitor 295011 High Off-site Release 3. Rate I 9 X the following as they apply to HIGH OFF- 65 6 SITE RELEASE RATE: RPS KiA Category Totals: 1 1 1 2 1/2 1/1 Group Point Total: I 7/3

ES-401 4 Form ES-401-1 ILT 11-1 NRC Written Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A Imp System # / Name A2 G Q# 1 2 3 4 5 6 1 3 4 I A2.08 - Ability to (a) predict the impacts of1he fonowing on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on 262001 AC Electrical X those predictions, use procOOures to 3.6 II Distribution correct, control, or mitigate the consequences of those abnonnal conditions or operations: Opening a disconnect under load A2.09 - Ability to (a) predict the impacts of the fonowing on the REACTOR PROTECTION SYSTEM; and (b) based on those 212000RPS X predictions, use procedures to 4.3 12 correct, control, or mitigate the consequences ofthose abnonnal conditions or operations: High containmentldrvwell pressure 2.2.40 - Equipment Control: Ability 207000 Isolation (Emergency) X to apply Technical Speciftcations 4.7 13 Condenser for a SYStem. 400000 Component Cooling 2.4.11 Knowledge ofabnonnal X 4.2 Water condition orocedures. 2.4.41 - Emergency Procedures I Plan: Knowledge of the emergency 2150031RM X action lev4~1 thresholds and 4.6 15 classifications. K1.I5 - Knowledge of the physical connections andlor cause- effect 259002 Reactor Water Level relationships bctween REACTOR X WATER LEVEL CONTROL 3.2 I Control SYSTEM and the following: Recirculation flow control system K1.05 - Knowledge of the physical connections and/or cause- effect relationships between SHUTDOWN 205000 Shutdown Cooling X 3.1 2 COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and the following: Component coolina water sYStems K2.01 - Knowledge of electrical 262001 AC Electrical X power supplies to the following: 3.3 3 Distnbution Off-site sources of Dower K2.01

  • Knowledge of electrical IRM X power supplies to the following: 2.5 4 IRM channels/detectors K3.03 - Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION 212000 RPS X 3.3 5 SYSTEM will have on following:

Local power range monitoring system: Plant-5pecific K3.01 - Knowledge of the effect that a loss or malfunction of the 300000 Instrument Air X (INSTRUMENT AIR SYSTEM) 2.7 6 will have on the following: Containment air sYstem

ES-401 4 Form ES-401-1 ILT 11-1 NRC Written Exam Written Examination Outline Plant Systems -- Tier 2 Group 1 K K K K K K A A A Imp System # / Name A2 G 0# 1 2 3 4 5 6 1 3 4 K4.01 - Knowledge of CCWS design feature(s) and or 400000 Component Cooling X interlocks wI'lich provide for the 3.4 7 Water following: Automatic start of standby pump K4.03 - Knowledge of EMERGENCY GENERATORS (DIESEUJET) design feature(s) 264000EOOs X 2.5 8 and/or interlocks which provide for the following: Speed droop control K5.01 - Knowledge of the operational implications of the following concepts as they apply 215004 Source Range Monitor X 2.6 9 to SOURCE RANGE MONITOR (SRM) SYSTEM: Detector operation K5.01 - Knowledge of the operational implications of the following concepts as they apply 218000 ADS X 3.8 10 to AUTOMATIC DEPRESSURIZATION SYSTEM

ADS logic operation K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the 262002 UPS (ACIDC) X UNINTERRUPTABLE POWER 2.7 11 SUPPLY (A.C.ID.C.): A.C.

electrical pOwer K6.07 - Knowledge of the effect that a loss or malfunction of the 207000 Isolation (Emergency) following will have on the X 3.0 12 Condenser ISOLATION (EMERGENCY) CONDENSER: A.C. power: BWR-2,3 A1.01 - Ability to predict and/or monitor changes in parameters associated with operating the 263000 DC Electrical X D.C. ELECTRICAL 2.5 13 Distribution DISTRIBUTION controls including: Battery charoinQ/discharaina rate A 1.05 - Ability to predict and/or monitor changes in parameters associated with operating the AVERAGE POWER RANGE 215005 APRM / lPRM X 3.3 14 MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including: Lights and alarms A2.03 - Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, 239002 SRVs X use procedures to correct, 4.1 15 control, or mitigate the consequences of those abnormal conditions or operations: Stuck open SRV

ES-401 4 Form ES-401-1 ILT 11-1 NRC Written Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K!K K A A A Imp System # / Name A2 G Q# 1 2 3 415 6 1 3 4 A2.06 - Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM 223002 PCISlNuclear Steam SUPPLY SHUT-OFF; and (b) Supply Shutoff X based on those predictions. use 3.0 16 procedures to correct control. or mitigate the consequences of those abnormal conditions or operations: Containment instrumentation failures A3.04 - Ability to monitor automatic operations of the 261000 SGTS X STANDBY GAS TREATMENT 3.0 17 SYSTEM including: System tem perature A3.02 - Ability to monitor automatic operations of the 209001 LPCS X LOW PRESSURE CORE 3.8 18 SPRAY SYSTEM including: Pump start A4.01 - Ability to manually 211000 SLC X operate and/or monitor in the 3.9 19 control room: Tank level A4.02 - Ability to manually 209001 LPCS X operate and/or monitor in the 3.5 20 control room: Suction valves 2.2.22 - Equipment Control: 205000 Shutdown Cooling X Knowledge of limiting conditions 4.0 21 for operations and safety limits. 2.1.28 - Conduct of Operations: Knowledge of the purpose and 239002 SRVs X function of major system 4.1 22 components and controls. K5.01 - Knowledge of the operational implications of the following concepts as they apply 263000 DC Electrical Distribution X to D.C. ELECTRICAL 2.6 23 DISTRIBUTION: Hydrogen generation during battery charging. A4.05 - Ability to manually 218000 ADS X operate andlor monitor in the 4.2 24 control room: ADS timer reset K2.02 - Knowledge of electrical 211000 SLC X power supplies to the following: 3.1 25 Explosive valves K3.05 - Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE 215005 APRM 1LPRM X MONITOR/LOCAL POWER 3.8 26 RANGE MONITOR SYSTEM will have on following: Reactor tion KJ A Category Totals: 2 3 3 2 3 2 2 212 2 3 213 Group Point Total: I 2615

ES-401 5 Form ES-401-1 ILT 11-1 NRC Written Exam Written Examination Outline Plant Systems - Tier 2 Group 2 System # / Name K K K K K K A A2 A A G Imp. a 1 2 3 4 5 6 1 3 4 # A2.10 - Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES; 223001 Primary CTMT and and (b) based on those predictions, X 3.8 16 Aux. use procedures to correct, control, or mitigate the consequences of those aboonnal conditions or operations: High drywell temperature 2.1.36 - Conduct ofOperations: Knowledge of procedures and 214000 RPIS X 4.1 17 limitations involved in core alterations. A2.13

  • Ability to (a) predict the impacts ofthe following on the RHRlLPCI:

TORUS/SUPPRESSION POOL 219000 RHRILPCI: COOLING MODE; and (b) based X 3.7 18 ToruslPool Cooling Mode on those predictions, use procedures to correct, control, or mitigate the consequences ofthose aboonnal conditions or operations: High suppression pool temperature KI.15 - Knowledge of the physical connections andlor cause- effect relationships between NUCLEAR 216000 Nuclear Boiler lust. X BOILER INSTRUMENTAnON 3.9 27 and the following: Isolation condenser: Plant-Specific K2.0 I

  • Knowledge of electrical 256000 Reactor Condcosate X power supplies to the following: 2.7 28 System pumPS K3.l6 - Knowledge ofthe effect that a loss or malfunction ofthe 239001 Main and Reheat X MAIN AND REHEAT STEAM 3.6 29 Steam SYSTEM will have on following:

Reliefi'safety valves K4.01 - Knowledge of RADIATION MONITORING 272000 Radiation Monitoring X System design feature(s) and/or 2.7 30 interlocks which provide for the following: Redundancy K5.12 - Knowledge of ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) 201006RWM X design feature(s) and/or 3.5 31 interlocks which provide for the following: Withdraw block: P Spec(Not-BWR6) K6.04* Knowledge of the effect that a loss or malfunction of the follOwing will have on the FIRE 286000 Fire Pmtection X 2.8 32 PROTECTION SYSTEM Diesel fuel transfer system: Plant-Specific A1.02

  • Ability to predict and/or monitor changes in parameters 268000 Radwaste X associated with operating the 2.6 33 RADWASTE controls including:

Off-site release

ES-401 5 Form ES-401-1 ILT 11-1 NRC Written Exam Written Examination Outline Plant Systems - Tier 2 Group 2 K K K K K K A A A Q System # I Name A2 G Imp. 1 2 3 4 5 6 1 3 4 # A2.03 - Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT 

                                                                ; and (b) based on those 234000 Fuel Handling                                           predictions. use procedures to X                                                 2.8     34 I Equipment                                                      correct. control. or mitigate the consequences of those abnormal conditions or operations: Loss of electrical power A3.02 - Ability to monitor automatic operations of the 201002RMCS                                           X        REACTOR MANUAL                     2.8     35 CONTROL SYSTEM induding:

Rod movement sequence lights A4.01 - Ability to manually operate andlor monitor in the 271000 Off-gas X 2.8 36 control room: Reset system isolations 2.1.23 - Conduct of Operations: Ability to perform specific 215001 Traversing In-core X system and integrated plant 4.3 37 Probe procedures during all modes of K 1.08 plant operation. Knowledge of the physical connections and/or cause- effect relationships 245000 Main Tuibine Gen. / between MAIN TURBINE X 3.4 38 Aux. GENERATOR AND AUXILIARY SYSTEMS and the following: Reactor/turbine pressure control system: Plant-Specific KJA Category Totals: 2 1 1 1 1 1 1 112 18 1/ 1 Group Point Total: I 12/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: ILT 11-1 NRC Written Exam Date: 05/14/12 RO SRO...()nly Category KIA # Topic IR Q# IR Q# Knowledge of new and spent fuel movement 2.1.42 3.4 19 procedures. Knowledge of criteria or conditions that 2.1.14 require plant-wide announcements, such as 3.1 24 pump starts, reactor trips, mode changes, etc. 1. Knowledge of conduct of operations Conduct 2.1.1 3.8 66 reQuirements. of Operations Knowledge of procedures and limitations 2.1.36 3.0 67 involved in core alterations. Knowledge of the station's requirements for 2.1.38 verbal communications when implemeting 3.7 74 procedures. Subtotal 3 2 Ability to analyze the effect of maintenance activities, such as degraded power sources, 2.2.36 4.2 20 on the status of limiting conditions for operations. Knowledge of maintenance work order 2.2.19 3.4 23 reQuirements. 2. Equipment Control Knowledge of tagging and clearance I 2.2.13 4.1 68 procedures. Ability to determine operability and I or 2.2.37 3.6 69 availability of safety related eQuipment. Subtotal 2 2

3. 2.3.11 Ability to control radiation releases. 4.3 21 Radiation Knowledge of radiation exposure limits under Control 2.3.4 3.2 25 normal or emergency conditions.

Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, 2.3.15 2.9 70 portable survey instruments, personnel monitoring equipment, etc. Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, 2.3.5 2.9 71 portable survey instruments, personell monitoring eQuipment, etc.

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Knowledge of radiation or containment hazards that may arise during normal, 2.3.14 3.4 75 abnormal, or emergency conditions or activities. Subtotal 3 2 Knowledge of the bases for prioritizing safety 2.4.22 functions during abnormal/emergency 4.4 22 operations. 4. Emergency Knowledge of low power I shutdown Procedures I implications in accident (e.g., loss of coolant 2.4.9 3.8 72 Plan accident or loss of residual heat removal) mitigation strategies. 2.4.42 Knowledge of emergency response facilities. 2.6 73 Subtotal 2 1 Tier 3 Point Total 10 7

ES-401 Record of Rejected KIA's Form ES-401-4 Randomly Tier I Group Reason for Rejection Selected K/A 295023 AA2.02 Unable to develop 3 credible distractors. 111 SRO 295023 AA2.05 Rejected KIA and randomly selected a new KIA. 600000 2.4.18 - KIA supports testing at the RO level, but not the SRO-Only level due to job responsibilities. EOP 111 SRO 600000 2.4.29 bases are RO required knowledge. A new KIA was randomly selected. 295018 2.2.38 - KIA rejected due to CCW not being 111 RO 295018 2.1.23 referenced in the Facility License. A new KIA was randomly selected. 207000 2.2.3 - KIA rejected due to Oyster Creek not being 2/1SRO 207000 2.2.40 a multi-unit site. A new KIA was randomly selected. 400000 2.4.41 - KIA supports testing at the RO level, but not the SRO-Only level due to job responsibilities. EOP 2/1SRO 400000 2.4.11 bases are RO required knowledge. A new KIA was randomly selected. 215003 2.4.41 - Unable to develop 3 credible distractors. 2/1SRO 215003 2.1.20 Rejected KIA and randomly selected a new KIA. 214000 2.1.31 - KIA supports testing at the RO level, but 2/2 SRO 214000 2.1.36 not the SRO-Only level due to job responsibilities. A new KIA was randomly selected. 215001 2.4.6 There are no EOP actions associated with 2/2RO 215001 2.1.23 the TIP system therefore a question could not be written. A new KIA was randomly selected. 2.2.3 - KIA rejected due to Oyster Creek not being a 3/RO 2.2.13 

                           multi-unit site. A new KIA was randomly selected.

2.3.5 - KIA rejected due to overlap with NRC question 71. 3/SRO 2.3.11 A new KIA was randomly selected. 2.3.15 - KIA rejected due to overiap with NRC question 70. 3/SRO 2.3.4 A new KIA was randomly selected.

ES-301 Administrative I ODIC:S Outline Form ES-301-1 Facility: O~ster Creek Date of Examination: 5/14/2012 Examination Level: RO ~ SRO D Operating Test Number: 11-1 NRC Administrative Topic Type Describe activity to be performed (See Note) Code* Calculate Identified Leak Rate lAW 351.2; 2.1.20 (4.6) Conduct of Operations M,R [NRC RO Admin JPM 1] Perform Core Thermal Limit Verification; 2.1.7 (4.4) [NRC Conduct of Operations P, R RO Admin JPM 2] Determine Vortex and NPSH Impacts on the Core Spray Equipment Control D,R System; 2.2.44 (4.2) [NRC RO Admin JPM 3] Radiation Control Review a Completed State/Local Notification Form; 2.4.39 Emergency Procedures/Plan M,R (3.9) [NRC RO Admin JPM 4] NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (: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 Facility: Oyster Creek Date of Examination: 5/14/2012 Examination Level: RO D SRO ~ Operating Test Number: 11-1 NRC Administrative Topic Type Describe activity to be performed (See Note) Code* Review / Approve a Completed Reactor Heat Balance; Conduct of Operations D,R 2.1.7 (4.7) [NRC SRO Admin JPM 1] Review Request to Allow LPRM (input into APRM) Bypass Conduct of Operations D,R lAW 403; 2.1.9 (4.5) [NRC SRO Admin ~IPM 2] Review Completed Surveillance Procedure 610.3.105 Equipment Control D,R (Core Spray Sys 1 Inst Cal and Operability); 2.2.12 (4.1) [NRC SRO Admin JPM 3] Authorize Emergency Exposures lAW EP-AA-113; 2.3.4 Radiation Control M,R (3.7) [NRC SRO Admin ~IPM 4] Determine Primary Containment Water Level lAW EMG-Emergency Procedures/Plan M,R SP28 and Determine Required Action; 2.4.21 (4.6) [NRC SRO Admin JPM 5] NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ~ 3 for ROs; ~ 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (? 1) (P)revious 2 exams (~ 1; randomly selected) ES 301 , Page 22 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: O~ster Creek Date of Examination: 05/14/2012 Exam Level: RO I:8l SRO-I D SRO-U D Operating Test Number: 11-1 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) Safety System / JPM Title Type Code* Function

a. Perform Recirculation Pump Trip Circuitry Test lAW 603.4.001 with P,A,S 1 Multiple Recirculation Pumps Trip (Alternate Path); 202001 A2.04 (3.7/3.8) [NRC Sim JPM 1]
b. Place a second RWCU Pump in service with a high temperature alarm M,A,S 2 and isolation failure lAW 303 (Alternate Path); 204000 A4.01 (3.1/3.0)

[NRC Sim ,JPM 2]

c. Shutdown of the Automatic Depressurization System lAW 308; 218000 D,EN,S 3 A4.03 (4.2/4.2) [NRC Sim JPM 3]
d. Perform Core Spray Surveillance with faulted Core Spray Pump lAW P,A,S 4 610.4.002 (Alternate Path); 209001 A4.01 (3.8/3.6) [NRC Sim ,JPM 4]
e. Purging the Primary Containment with Elevated Stack Radiation P,A, EN,S 5 (Alternate Path); 223001 A4.07 (4.214.1) [NRC Sim JPM 5]
f. Restore 4160VAC Bus 1C to normal with EDG-1 supplying power D, A, L, EN, 6 (Alternate Path); 264000 A4.04 (3.7/3.7) [NRC Sim JPM 6] S
g. Swap Instrument Air Compressors; 300000 K4.04 (2.8/2.9) [NRC Sim N,S 8 JPM 7]
h. Re-establishing Off-Gas System Flow after an Off-Gas System D, L,S 9 Explosion; 271000 A2.06 3.5/3.9 [NRC Sim JPM 8]

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

i. Control CRD in the plant Post-Scram lAW SP-3; 201001 A 1.03 (2.9/2.8) D, L, R, E 1

[NRC Plant ..IPM 1]

j. Line up to vent the Torus through the Hardened Vent lAW SP-35; 295024 D,E 5 EA1.14 (3.4/3.5) [NRC Plant JPM 2]
k. Bypass the Air Dryers and the Pre/Post Filters; 300000 A2.01 (2.9/2.8) D,R 8

[NRC Plant JPM 3] 

@        All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO I SRO-II SRO-U (A)lternate path 4-61 4-6 I 2-3

{C)ontrol room (D)irect from bank :5:.9/ :5:.8 I :5:.4 {E)mergency or abnormal in-plant ~11 ~1 1 !::. 1 (EN)gineered safety feature - 1 "' I !::. 1 (control room system (L)ow-Power I Shutdown  !::. 1 1 ~1 1  !::. 1 {N)ew or (M)odified from bank including 1 (A)  !::.21 !::.2 1  !::. 1 (P)revious 2 exams :5:. 31 :5:.3 1 :5:. 2 (randomly selected) (R)CA ~ 1 I !::.1 I  !::. 1 (S)imulator ES-301 , Page 23 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Oyster Creek Date of Examination: 05/14/2012 Exam Level: RO D SRO-I ~ SRO-U D Operating Test Number: 11-1 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) Safety System / JPM Title Type Code* Function

a. Perform Recirculation Pump Trip Circuitry Test lAW 603.4.001 with P,A,S 1 Multiple Recirculation Pumps Trip (Alternate Path); 202001 A2.04 (3.7/3.8) [NRC Sim JPM 1]
b. Place a second RWCU Pump in service with a high temperature alarm M,A,S 2 and isolation failure lAW 303 (Alternate Path); 204000 A4.01 (3.1/3.0)

[NRC Sim "IPM 2]

c. Shutdown of the Automatic Depressurization System lAW 308; 218000 D,EN,S 3 A4.03 (4.2/4.2) [!\IRC Sim JPM 3]
d. Perform Core Spray Surveillance with faulted Core Spray Pump lAW P,A,S 4 610.4.002 (Alternate Path); 209001 A4.01 (3.8/3.6) [NRC Sim JPM 4]
e. Purging the Primary Containment with Elevated Stack Radiation P,A,EN,S 5 (Alternate Path); 223001 A4.07 (4.214.1) [NRC Sim JPM 5]
f. Restore 4160VAC Bus 1C to normal with EDG-1 supplying power D, A, L, EN, 6 (Alternate Path); 264000 A4.04 (3.7/3.7) [NRC Sim JPM 6] S
g. Swap Instrument Air Compressors; 300000 K4.04 (2.8/2.9) [NRC Sim N,S 8 JPM 7]

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

i. Control CRD in the plant Post-Scram lAW SP-3; 201001 A 1.03 (2.9/2.8) D, L, R, E 1

[NRC Plant JPM 1]

j. Line up to vent the Torus through the Hardened Vent lAW SP-35; 295024 D, E 5 EA 1.14 (3.4/3.5) [NRC Plant JPM 2]
k. Bypass the Air Dryers and the Pre/Post Filters; 300000 A2.01 (2.9/2.8) D,R 8

[NRC Plant "IPM 3] 

@        All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO 1 SRO-II SRO-U (A)lternate path 4-61 4-6 1 2-3 (C)ontrol room (D)irect from bank  :;. 91 $8 1$4 (E)mergency or abnormal in-plant 2:, 1 I 2:,1 I 2:, 1 (EN)gineered safety feature - I - I 2:, 1 (control room system (L)ow-Power 1 Shutdown 2:,11 2:,1 / 2:, 1 (N)ew or (M)odified from bank including 1(A) 2:,2/ 2:,2 / 2:, 1 (P)revious 2 exams  :;. 3/ :;.3 1 $ 2 (randomly selected)

(R)CA 2:,11 2:,1 / 2:, 1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Ollster Creek Date of Examination: 05/14/2012 Exam Level: RO 0 SRO-I 0 SRO-U ~ Operating Test Number: 11-1 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) Safety System 1 JPM Title Type Code* Function a.

b. Place a second RWCU Pump in service with a high temperature alarm I M,A,S 2 i

and isolation failure lAW 303 (Alternate Path); 204000 A4.01 (3.1/3.0) [NRC Sim JPM 2] i II ~ c. d. e.

f. Restore 4160VAC Bus 1C to normal with EDG-1 supplying power D, A, L, EN, 6 (Alternate Path); 264000 A4.04 (3.7/3.7) [NRC Sim JPM 6] S g.

I h.

  • In-Plant Systems@ (3 for RO); (3 for SRO*I); (3 or 2 for SRO-U)
i. Control CRD in the plant Post-Scram lAW SP-3; 201001 A 1.03 (2.9/2.8) D,L, R, E 1

[NRC Plant JPM 1]

j. Line up to vent the Torus through the Hardened Vent lAW SP-35; 295024 D, E 5 EA1.14 (3.4/3.5) [NRC Plant JPM 2]
k. Bypass the Air Dryers and the PrelPost Filters; 300000 A2.01 (2.9/2.8) D,R 8

[NRC Plant JPM 31 

@        All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO I SRO-II SRO-U (A)lternate path 4-6/ 4-6 / 2-3 (C)ontrol room (D)irect from bank '5:. 9 / '5:.8 / '5:. 4 (E)mergency or abnormal in-plant  !::. 1 / !::.1 1 !::. 1 (EN)gineered safety feature 1 !::. 1 (control room system (L)ow-Power / Shutdown  !::. 1 / !::.1 / !::. 1 (N)ew or (M)odified from bank including 1(A)  !::.21 !::.2 I !::. 1 (P)revious 2 exams '5:. 31 :s3 1 '5:. 2 (randomly selected)

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

ILT 11-1 NRC Scenario 1 (NEW) Scenario Outline Facility: Oyster Creek Scenario No.: 1 Op Test No.: 11*1 NRC Examiners: Operators: Initial Conditions:

  • 15% power with mode switch in RUN (IC 153)
  • RWM is inoperable and bypassed i
  • Control Room HVAC System A is inoperable Turnover:
  • Continue with rod withdrawal. Complete step 24 Group 5-1. When rod pulls are complete wait for further direction from Reactor Engineering.

Event No. Malf. No. EventType* Event Description 1 NlA N BOP Swap Sentice Water Pumps. 2 N/A R ATC Withdraw control rods to raise reactor power. MAL 3 CRDOO8_ C ATC Respond to an uncoupled control rod >10% power. 3451 MAL- C BOP 4 EOSOO4B Respond to the loss of VMCC 1B2. TS SRO MAL RCPOO3D C BOP Respond to Recirculation Pump D inner seal failure, then 5 MAL- TS SRO outer seal failure. RCP004D MAL- Respond to the E EMRV lifting leading the crew to a 6 C ATC NSS025E manual scram. CAEP 7 ATWS.CAE M Crew Respond to an Electric ATWS. PMP SLCOO1A 8 C Crew Respond to Standby Liquid Control Pump shaft break. PMP SLCOO2A 

 *         (N)onnal,   (R)eactivity,  (I}nstrument,  (C)omponent,    (M)ajor TranSient,   (TS) Tech Specs ILT 11-1 NRC Scenario 1                                                                   Page 1 of 28

ILT 11-1 NRC Scenario 1 (NEW) Simulator Summary Event Event Summary 1 The BOP will swap Service Water Pumps to equalize run times. The BOP will start the standby pump, stop the running pump, and then verify expected conditions locally with the EO. (BOP: Normal Evolution) 2 The ATC will withdraw control rods to raise reactor power lAW the pull sheet and 302.2. (ATC: Reactivity Manipulation) 3 The ATC will respond to an uncoupled control rod (rod 34-51) at position 48 and will re-couple the control rod lAW ABN-6, Control Rod Drive System. (ATC: Component Malfunction) 4 The crew will respond to the trip of VMCC 1B2 and enter ABN-51, Loss of VMCC 1B2. The BOP will restore power to PSP-2 and the ATC will reset the % scram. The SRO will review TS 3.7 and enter a 30 hr cold shutdown LCO. (BOP: Component Malfunction; SRO: Tech Specs) 5 The BOP will respond to a leak in Recirculation Pump D outer seal, followed by a leak in the inner seal. The SRO will direct entry into ABN-2, Recirculation System Failures, to trip Recirculation Pump D and Isolate the D Recirculation Loop. The SRO will review and apply Tech Specs 3.3.D and 3.3.F for unidentified leak rate and recirculation loop operability. (BOP: Component Malfunction; SRO: Tech Specs) 6 The ATC and BOP will respond to the E EMRV lifting lAW ABN-40, Stuck Open EMRV. The ATC will take manual control of the master feedwater controller. The BOP will cycle the E EMRV then disable it. The ATC will return the master feedwater controller to automatic operation and insert a manual reactor scram. The SRO will review Tech Specs 3.4 for ADS operability and TS 3.S.A for Torus Temperature limits. (ATC: Component Malfunction) 7 The Crew will diagnose an electric ATWS and the SRO will direct entry into RPV Control - with ATWS EOP. The ATC will perform actions to insert control rods and the BOP will perform actions to control Torus water temperature and RPV water level. (Major Evolution) (PRA) ILT 11-1 NRC Scenario 1 Page 2 of 28

ILT 11-1 NRC Scenario 1 (NEW) 8 Due to the Torus water temperature heating up from the E EMRV stuck open, Standby Liquid Control (SLC) injection will be directed. The first SLC pump started will have a broken shaft and the Applicant will start the second SLC pump. (Component Failure After EOP) Critical With reactor power> 2% during an ATWS, terminate and prevent Task 1 injection into the RPV to intentionally lower RPV water level which will lower reactor power. Critical Crew directs the Reactor Building EO to vent the scram air header. Task 2 (The Lead Examiner will direct the Booth to vent the scram air header at their discretion). ES..301-4 Target Quantitative Actual Event Attributes Attributes Number{s}

1. Total malfunctions (5-8) 6 3-8
2. Malfunctions after EOP entry (1-2) 1 8
3. Abnormal events (2-4) 4 3-6
4. Major transients (1-2) 1 7
5. EOPs entered/requiring substantive 2 7 actions (1-2)
6. EOP contingencies requiring substantive 1 7 actions (0-2)
7. Critical tasks (2-3) 2 7 ILT 11-1 NRC Scenario 1 Page 30f28

ILT 11-1 NRC Scenario 3 (Modified) Scenario Outline Facility: Oyster Creek Scenario No.: ~ Op Test No.: 11-1 NRC Examiners: Operators: Initial Conditions:

  • 85% power
  • 'B' RWCU Pump is OOS Turnover:
  • Lower power to 80% using recirculation flow lAW 1001.22-3, Core Maneuvering Daily Instruction Sheet
  • Backwash Main Condenser Half B South Event No. Malf. No. EventType* Event Description 1 NA R ATC Lower reactor power to 80% using recirculation flow.

2 NA N BOP Continue backwashing Main Condenser Half B South. MAL-3 TCS010 I BOP Respond to the EPR setpoint failing high. BKR- C ATC 4 CRDOO2 Respond to CRD Pump A trip. TS SRO psw TBCOO1A Respond to the trip of TBCCW Pump 1-3 and auto start 5 C BOP BKR- failure of TBCCW Pump 1-2. TBCOO3 MAL- I ATC Respond to a reference leg leak in the A & C GEMAC 6 NSS012E TS SRO RPV level indicators ID13A and ID13C MAL-7 CRDOO6 M Crew Respond to a multiple rod drift. MAL- M 8 NSS016A Crew Respond to a Safety Valve lifting post scram C MAL- Respond to a trip of the operating Containment Spray 9 CNSOO4A- C Crew 0 Pump

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs ILT 11-1 NRC Scenario 3 Page 1 of 30

ILT 11-1 NRC Scenario 3 (Modified) Simulator Summary Event Event Summary 1 The ATC will lower reactor power to approximately 80% with recirculation flow using the Master Recirc Speed Controller. (ATC: Reactivity Manipulation) 2 The BOP will backwash condenser B South lAW procedure 323.6, Backwashing Condensers. This will require several switch manipulations by the BOP. (BOP: Normal Evolution) 3 The BOP will respond to the EPR setpoint failing high. The crew will enter ABN-9, Electric Pressure Regulator Malfunction. The BOP will secure the EPR and restore reactor pressure to the normal band on the MPR. (BOP: Instrument Malfunction) 4 The ATC will respond to a trip of CRO Pump A lAW RAP H-1-c. The ATC will start CRO Pump B. The SRO will review and apply Tech Spec 3.4.0.2. (ATC: Component Malfunction; SRO: Tech Specs) 5 The BOP will respond to a trip of TBCCW PUMP 3 and a failure of TBCCW PUMP 2 to auto start on low TBCCW header pressure. The BOP will start the standby pump lAW ABN-20, TBCCW Failure Response. (BOP: Component Malfunction) 6 The ATC will diagnose a rising RPV water level. Indications of actual RPV water level will rise on Panel 4F and Panel 5F/6F Yarway indications. The ATC will perform actions to stabilize RPV water level lAW ABN-17, Feedwater System Abnormal Conditions. The.ATC will take manual control of RPV watE~r level and swap Feedwater Level Control to the alternate water level instrument 1013B. The increased Primary Containment leakage will result in a rise in unidentified leak rate and the SRO will review and apply Tech Spec 3.3.0.2. (ATC: Instrument Malfunction; SRO: Tech Specs) 7 The ATC will identify/report multiple drifting control rods into the core and lAW ABN-6, Control Rod Malfunctions, manually scram the reactor lAW ABN-1, Reactor Scram. (ATC: Component Malfunction) (PRA) 8 Post scram, the crew will respond to a Safety Valve lifting. This will result in rising drywell pressure and temperature requiring Orywell Sprays lAW the Primary Containment Control EOP. (Major Evolution; Component Failure After EOP) 9 When initiating Containment Spray lAW the Primary Containment ILT 11-1 NRC Scenario 3 Page 2 of 30

ILT 11-1 NRC Scenario 3 (Modified) Control EOP, the Containment Spray pump in the system the Crew starts will trip after 30 seconds. The Crew must initiate containment spray using an alternate Containment Spray Pump. (Component Failure After EOP) Critical The ATC will manually scram the reactor following control rods drifting Task 1 into the core. There is no manual scram associated with this casualty and the core is not analyzed for the resultant abnormal rod configu ration. Critical When Drywell or Torus pressure exceeds 12 psig, OR before Drywell Task 2 bulk temperature is determined it cannot be maintained below 281°F, spray the drywelllAW SP-29, Initiation of the Containment Spray System for Drywell Sprays. ES-301-4 Target Quantitative Actual Event Attributes Attributes Number{s}

1. Total malfunctions (5-8) 7 3-9
2. Malfunctions after EOP entry (1-2) 2 8-9
3. Abnormal events (2-4) 4 3,5-7
4. Major transients (1-2) 1 8
5. EOPs entered/requiring substantive 2 8 actions (1-2)
6. EOP contingencies requiring substantive 0 N/A actions (0-2)
7. Critical tasks (2-3) 2 7,9 ILT 11-1 NRC Scenario 3 Page 3 of 30

ILT 11-1 NRC Scenario 4 (NEW) 

                                       -{Backup 86e na rio) ~

Scenario OuUlne Facility: Oyster Creek Scenario No.: ! Op Test No.: 11-1 NRC Examiners: Operators: Initial Conditions:

  • The plant is at 95% power
  • Dilution Pump 2 is tagged out of service
  • Air Compressor #3 is tagged out of service in PTL
  • The RWM is inoperable and bypassed Turnover:
  • Perform Turbine Valve testing lAW 625.4.002 Event No. Malf. No. EventType* Event Description 1 N/A N BOP Tests MPR lAW 625.4.002 MAL-2 CRD001A C ATC Respond to a CRD Flow Control Valve failed closed.

LOA RPS001 MAL CRD011_1 C ATC 3 415 Respond to trip of RPS MG Set 1 and a single rod scram. TS SRO MAL CRD014_1 415 SWI C 4 TBS027C BOP Respond to a trip of Control Room Vent Fan B ANN-L4f TS SRO PSW- R ATC Respond to a major oil leak on 'B' Feed Pump requiring a 5 CFW015A C BOP rapid power reduction. MAL CFWOO6C M Respond to a trip of the 'C' Feed Pump requiring a reactor 6 Crew MAL- C scram and a failure of all control rods to insert. CRD022 MAL- Respond to a Torus Leak requiring entry into Primary 7 PCN007 M Crew Containment Control. VLV-Respond to Core Spray system suction valves being 8 CSS001. C Crew 009 mechanically seized when lining up the CST to the Torus. MAL- Respond to a Torus leak increase requiring the crew to 9 PCNOO7 M Crew Emergency Depressurize.

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs ILT 11-1 NRC Scenario 4 Page 1 of 31

ILT 11-1 NRC Scenario 4 (NEW) (Backup Scenario) Simulator Summary Event Event Summary 1 The shift turnover will direct the performance of the Mechanical Pressure Regulator (MPR) test lAW 625.4.002, Main Turbine Surveillances, section 12. The BOP will lower the MPR setpoint, verify the MPR on control, then adjust the setpoint backup to place the EPR back in service. (BOP: Normal Evolution) 2 The ATC will respond to in-service CRD Flow Control Valve failing closed. The ATC will swap Flow Control Valves lAW procedure 302.1, Control Rod Drive System. (ATC: Component Malfunction) 3 The crew will respond to the trip of RPS MG Set 1 and a single control rod (14-15) scram to pOSition 04 (RAP-G2c, ABN-6, CRD Failures). The BOP will re-power the RPS Bus and will reset % scram and % isolations. The Crew will attempt to manually insert the scrammed control rod from 04-00. When attempted, the rod will not move. The SRO will review TS 3.2.B.4 and 3.2.A. and will declare the control rod inoperable, and valve-out the control rod. (ATC: Component Malfunction; SRO: Tech Specs) 4 The Control Room HVAC Fan B will trip. The SRO will direct the BOP to place Control Room HVAC System A in service lAW 331.1, Control Room and Old Cable Spreading Room Heating, Ventilation, and Air Conditioning System. The SRO will apply Tech Spec 3.17.B. (BOP: Component Malfunction; SRO: Tech Specs) 5 The BOP will respond to a low oil pressure condition on the 'B' Feed Pump. The crew will have the Turbine Building (TB) Operator investigate. The TB Operator will report a large oil leak from the 'B' Feed Pump and that there is no oil in the sight glass, oil is being contained and no oil has gotten into the floor drains. The ATC will perform a rapid power reduction and the BOP will trip the 'B' Feed Pump. (ATC: Reactivity Manipulation; BOP: Component Malfunction) 6 The crew will respond to a trip of the 'C' Reactor Feed Pump requiring a reactor scram. Some control rods will fail to insert (power < 2%) but can be manually inserted using CRD. (Major Evolution; Component Failure After EOP) 7 A leak in the Torus will develop requiring the crew to enter the Primary Containment Control EOP. The crew will commence makeup to the Torus lAW SP-37, Makeup To The Torus Via Core Spray ILT 11-1 NRC Scenario 4 Page 2 of 31

ILT 11-1 NRC Scenario 4 (NEW) (Backup Scenario) System. (Major Evolution) 8 When the crew attempts to line up Core Spray System to make up to the Torus, Core Spray suction valve for System 1(2) V-20-3(4)and V 20-32(33), Core Spray System 1(2) suction valves will not close. The crew will place the alternate Core Spray Pump System in service. (Component Failure After EOP) 9 After the Crew places Core Spray Pump/System 2 in service to makeup to the Torus, the Torus leak will increase leading the crew to Anticipate Emergency Depressurization and/or Emergency Depressurize the RPV. (Major Evolution) Critical Insert control rods lAW Support Procedure 21 to achieve a shutdown Task 1 reactor. Critical Emergency Depressurize the RPV when it is determined Torus level Task 2 cannot be maintained> 110 inches. ES-301-4 Target Quantitative Actual Event Attributes Attributes Number{s}

1. Total malfunctions (5-8) 7 2-8
2. Malfunctions after EOP entry (1-2) 2 7-8
3. Abnormal events (2-4) 4 2,3,5,6
4. Major transients (1-2) 2 6-7
5. EOPs enteredlrequiring substantive 2 6-7 actions (1-2)
6. EOP contingencies requiring substantive 1 9 actions (0-2)
7. Critical tasks (2-3) 2 6,9 ILT 11-1 NRC Scenario 4 Page 30f31}}
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