ML111870267

Draft - Outlines (Folder 2)

ML111870267
Person / Time
Site:	Oyster Creek
Issue date:	05/27/2011
From:	Fish T H Operations Branch I
To:	Exelon Nuclear
Hansell S
Shared Package
ML110030666	List: IR 05000219/2011301 ML111810110 ML111870186 ML111870261 ML111870267 ML111870274 ML111870278 ML112130525 ML112130528 ML112130534 ML112130541
References
TAC U01831
Download: ML111870267 (30)

Text

i Written Examination Outline Form ES-401-1 Oyster Creek IL T 10-1 Date of Exam: 07/11/11 NRC Exam Outline Tier Grou p K 1 K 2 RO KIA Category Points K K K K A A A 3 4 S 6 1 2 3 A 4 G

• Tota I SRO-Only Points A2 G* Total ! 1. 1 3 4 3 3 4 3 20 4 3 7 Emergenc y 2 1 1 1 1 1 2 7 1 2 3 & Plant Evolutions Tier Total 4 5 4 4 5 5 27 5 5 10 s 1 2 2 2 2 3 2 3 2 2 3 3 26 3 2 S 2. Plant 2 1 1 2 1 1 1 1 1 1 1 1 12 0 1 2 3 Systems Tier Total 3 3 4 3 4 3 4 3 3 4 4 38 4 4 8 s 3. Generic Knowledge

&Abilities Categories 1 2 2 3 3 2 4 3 10 1 2 2 2 3 2 4 1 7 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 ltable. 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.1.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) KlAs 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, IRs, and point totals (#) on Form ES-401-3.

limit SRO selections to KlAs that are ES-401 Written Examination Outline Form ES-401-1 linked to 10CFR55.43 2 Form ES-401-1 Oyster Creek IL T 10-1 NRC Exam Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group I EAPE # I Name Safety KIA Topic(s) AA2.05 -Ability to determine and/or interpret the following as they apply to 295006 SCRAM 11 4.6 1 SCRAM: Whether a reactor SCRAM has occurred AA2.01 -Ability to determine and/or interpret the following as they apply to 295004 Partial or Total Loss of X PARTIAL OR COMPLETE LOSS OF 3.6 2 DCPwr/6 D.C. POWER: Cause of partial or complete loss of D.C. power AA2.05 -Ability to determine and/or interpret the following as they apply to 295003 Partial or PARTIAL OR COMPLETE LOSS OF4.2 3 A.C. POWER: Whether a partial complete loss of A.C. power 295026 Suppression Pool 2.2.38 -Knowledge of conditions and 4.5 4 Water Temp. I 5 limitations in the facility 295037 SCRAM Conditions 2.4.8 -Emergency Procedures I Present and Rec:lctor Power Knowledge of how abnormal 4.5 5 Above APRM Downscale or procedures are used in conjunction Unknown I 1 2.4.45 -Ability to prioritize and interpret 295021 Loss of Shutdown X the significance of each annunciator or 4.3 6 Cooling 14 EA2.03 -Ability to determine 295030 Low Suppression Pool interpret the following as they apply 3.9 7 Water Levell 5 LOW SUPPRESSION POOL LEVEL: Reactor AK1.01 -Knowledge of the implications of the following concepts 700000 Generator Voltage and they apply to GENERA TOR 3.3 39 Electric Grid AND ELECTRIC DISTURBANCES and the AK1.05 -Knowledge of the operational implications of the following concepts as 295004 Partial ()r Total Loss of they apply to PARTIAL OR COMPLETE 3.3 40 DCPwr/6 LOSS OF D.C. POWER: Loss of breaker protection AK1.01 -Knowledge of the operational implications of the following concepts as 295005 Main Turbine they apply to MAIN TURBINE 4.0 41 Generator Trip I 3 GENERATOR TRIP: Pressure on reactor AK2.02 -Knowledge of 295023 Refueling Acc Cooling interrelations between 2.9 42ACCIDENTS and the following: pool cooling and cleanup EK2.03 -Knowledge of 295038 High Olf-site interrelations between HIGH OFF-SITE 3.6 43 Rate I RELEASE RATE and the Plant ventilation EK2.01 -Knowledge of 295028 High interrelations between HIGH DRYWELL 3.7 44 Temperature 15 TEMPERATURE and the Drywell spray:

EK3.04 -Knowledge of the reasons 295031 Reactor Low Water the following responses as they apply 4.0 45 Level REACTOR LOW WATER Steam 2 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group I EAPE # / Name Safety KIA Topic{s) AK3.02 -Knowtedge of the reasons 295021 Loss of Shutdown the following responses as they apply 3.3 46 Cooling LOSS OF SHUTDOWN FeedinQ and bleedinQ reactor EK3.04 -Knowtedge of the reasons 295024 High Pressure the following responses as they apply to 3.7 47HIGH DRYWELL PRESSURE:

Emergency depressurization AA 1.09 -Ability to operate and/or monitor the following as they apply to 295016 Control Room X CONTROL ROOM ABANDONMENT:

4.0 48 Abandonment

/ 7' Isolation/emergency condenser(s):

Plant-Specific AA 1.03 -Ability to operate and/or monitor the following as they apply to 295003 Partial or Complete X PARTIAL OR COMPLETE LOSS OF 4.4 49 Loss of AC /6 A.C. POWER: Systems necessary to assure safe plant shutdown EA 1.06 -Ability to operate and/or 295025 High Reactor monitor the following as they apply to 4.5 50HIGH REACTOR PRESSURE:

condenser:

EA2.03 -Ability to determine 295026 Suppression Pool interpret the following as they apply to 3.9 51 SUPPRESSION POOL HIGH TEMPERATURE:

Reactor AA2.16 -Ability to determine interpret the following as they apply 600000 Plant Fire On-site / 8 PLANT FIRE ON SITE: Vital equipment 3.0 52 and control systems to be maintained and operated durinQ a fire AA2.01 -Ability to determine and/or interpret the follOwing as they apply to 295018 Partial or Total Loss of X PARTIAL OR COMPLETE LOSS OF 3.3 53 CCW/8 COMPONENT COOLING WATER : Component temperatures 2.1.31 -Ability to locate control 295019 Partial or Total Loss of switches, controls, and indications, 4.6 54 Inst. Air to determine that they correctly the desired plant 2.4.4 -Ability to recognize abnormal indications for system operating 295006 SCRAM /1 X parameters that are entry-level 4.5 55 conditions for emergency and abnormal operating procedures.

2.4.20 -Emergency Procedures

/ Plan: 295025 High Reactor Pressure X Knowtedge of operational implications 3.8 56/3 of EOP warninQs, cautions, and notes. AA2.03 -Ability to determine and/or 295001 Partial or Complete interpret the following as they apply to Loss of Forced Core Flow X PARTIAL OR COMPLETE LOSS OF 3.3 57 Circulation / 1 & 4 FORCED CORE FLOW CIRCULATION

Actual core EK2.14 -Knowtedge of 295037 SCRAM Conditions interrelations between Present and Reactor Power CONDITION PRESENT 3.6 58 Above APRM DClwnscale or REACTOR POWER ABOVE Unknown /1 DOWNSCALE OR UNKNOWN and following:

RPIS:

KIA Category Totals: 3 4 3 3 414 3/3 Group Point Total: I 20/7 3 Form ES-401-1 Oyster Creek IL T 10-1 NRC Exam Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE # / Name Safety KIA Topic(s) EA2.01 -Ability to determine and/or 295033 High Secondary interpret the following as they apply to Containment Area Radiation X HIGH SECONDARY CONTAINMENT 3.9 8 Levels/9 AREA RADIATION LEVELS: Area radiation levels 2.1.32 -Conduct of Operations:

Ability to 295029 High Pool X explain and apply all system limits and 4.0 9 Water Levell 5 precautions.

2.2.44 -Equipment Control: Ability to interpret control room indications to 295020InadvertentConl verify the status and operation of a 4.4 10 Isolation

/5 & system. and understand how operator actions and directives effect plant and system conditions.

EK1.03 -Knowledge of the implications of the following concepts 295032 High Secondary they apply to HIGH 3.Containment Area Temperature CONTAINMENT AREA 59 5TEMPERATURE:

Secondary containment leakage detection:

Plant-Specific AK2.01 -Knowledge of the interrelations between HIGH 295013 High Suppression 3.6SUPPRESSION POOL 60 Temperature

/5 TEMPERATURE and the Suppression pool AK3.05 -Knowledge of the reasons 295010 High Drywall Pressure the following responses as they apply to 61/5 HIGH DRYWELL PRESSURE:

5 Temperature monitoring AA 1.06 -Ability to operate and/or monitor the following as they apply to 295002 Loss of 3.X LOSS OF MAIN 62 Condenser Vac l 0 VACUUM: Reactor/turbine pressure regulating system AA2.01 -Ability to determine and/or 295022 Loss of CRD Pumps interpret the following as they apply to 3.63LOSS OF CRD PUMPS: Accumulator 5 pressure 2.4.35 -Emergency Procedures

/ Plan: 295036 Secondary Knowledge of local auxiliary operator 3.Containment High SumplArea 64 tasks during emergency and the 8 Water Levell 5 resultant operational effects. 2.4.8 -Emergency Procedures I 295009 Low Reactor Water Knowledge of how abnormal operating 65 Level procedures are used in conjunction with 8 EOP*s. KIA Category Totals: 1 1 1 1/1 Group Point Total: I 7/3 ES-401 4 Form ES-401-1 System # / Name 259002 Reactor Water Level Control System 261000SGTS 218000 ADS 215005 APRM I LPRM 212000 RPS 218000 ADS 205000 Shutdown Cooling Oyster Creek ILT 10-1 NRC Exam Written Examination Plant Systems -Tier 2 Group K K K K K K A A A A2. G 1 2 3 4 5 6 1 3 4 A2.04* Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM and (b) X based on those predictions.

use procedures to correct. control. or mitigate the consequences of those abnormal operation:

RFP runout condition:

Plant-Specific A2.15 -AbiHty to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on X those predictions.

use procedures to correct. control, or mitigate the consequences of those abnormal conditions or operations:

High area radiation bv refuel bridge: Plant.specffic 2.4.9* Emergency Procedures

/ Plan: Knowledge of low power I X shutdown implications in acckfent (e.g .* loss of coolant acckfent or loss of residual heat removal) mitigation strategies.

2.4.46

• Emergency Procedures X / Plan: Ability to verify that the alarms are consistent with the conditions.

A2.01 -Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on X those predictions.

use procedures to correct. control. or mitigate the consequences of those abnormal conditions or operations:

RPS aenerator set failure K1.06 -Knowledge of the physical connections and/or cause-effect relationships X between AUTOMATIC DEPRESSURIZATION SYSTEM and the following:

Safety/relief valves K1.05 -Knowledge of the physical connections and/or cause-effect relationships X between SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and the following:

Component Imp 3.1 3.4 4.2 4.2 3.9 3.9 3.1 Q# 11 12 13 14 15 1 2 0 cooling water systems K2.02 -Knowledge of electrical 400000 Component Cooling X power supplies to the following:

2.9 3 Water CCWvalves 4 Form ES-401-1 System # / Name 263000 DC Electrical Distribution 209001 LPCS 207000 Isolation (Emergency)

Condenser 223002 PCIS/Nuclear Steam Supply Shutoff 212000 RPS 264oo0EDGs 262001 AC Electrical Distribution 262002 UPS (AClDC) 215004 Source IRange Monitor Oyster Creek ILT 10-1 NRC Exam Written Examination Plant Systems -Tier 2 Group K K K K K K A A A A2 G 1 2 3 4 5 6 1 3 4 K2.01

• Knowledge of electrical X power supplies to the following: Maior D.C. loads K3.03 -Knowledge of the effect that a loss or malfunction of the LOW PRESSURE CORE X SPRAY SYSTEM will have on following:

Emergency aenerators K3.02

• Knowledge of the effect that a loss or malfunction of the ISOLATION (EMERGENCY)

CONDENSER will have on X following:

Reactor water level (EPG's address the isolation condenser as a water source): BWR-2.3 K4.06 -Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(5) and/or interlocks which provide for the following:

Once initiated.

system reset requires deliberate operator action K4.03 -Knowledge of REACTOR PROTECTION SYSTEM design feature(s) and/or interlocks which provide X for the following:

The prevention of supplying power to a given RPS bus from multiple sources simultaneously K5.06

• Knowledge of the operational implications of the following concepts as they apply X to EMERGENCY GENERATORS (DIESEUJET)

Load seauencina K5.01 -Knowledge of the operational implications of the following concepts as they apply to A.C. ELECTRICAL DISTRIBUTION:

Principle involved with paralleling two AC. sources K6.01 -Knowledge of the effect that a loss or malfunction of the following will have on the X UNINTERRUPTABLE POWER SUPPLY (A.C.lD.C.)

A.C. electrical power K6.04 -Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRH) SYSTEM: Detectors Imp 3.1 2.9 3.8 3.4 3.0 3.4 3.1 2.7 2.9 Q# 4 5 6 7 8 9 10 11 12 ES-401 4 Form ES-401-1 Oyster Creek IL T 10-1 NRC Exam Outline Written Examination Outline Plant Systems -Tier 2 Group 1 System # I Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A2 A 3 A 4 G Imp Q# 259002 Reactor Water Level Control 261000 SGTS 215005 APRM I LPRM 239002SRVs 2150031RM 211000 SLC 300000 Instrumlent Air 263000 DC ElectriCal Distribution 223002 PCIS/Nuclear Steam Supply Shutoff 261000 SGTS A 1.05 -Ability to predict and/or monitor changes in parameters associated with operating the X REACTOR WATER LEVEL CONTROL SYSTEM controls including:

FWRV/startup level control position:

Plant-Specific.

A 1.04 -Ability to predict and/or monitor changes in parameters associated with operating the X STANDBY GAS TREATMENT SYSTEM controls including:

Secondary containment differential pressure A2.05

• Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM; X and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions Loss of recirculation flow signal A2.05 -Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, X use procedures to correct. control, or mitigate the consequences of those abnormal conditions or operations:

Low reactor pressure A3.03 -Ability to monitor automatic operations of the X INTERMEDIATE RANGE MONITOR (IRM) SYSTEM includine:

RPS status A4.06 -Ability to manually X operate and/or monitor in the control room: RWCU system isolation:

Plant-Specific A4.01 -Ability to manually X operate and/or monitor in the control room: Pressure gauges A4.01 -Ability to manually operate and/or monitor in the X control room: Major breakers and control power fuses: Plant-Specific 2.1.30 -Conduct of Operations:

X Ability to locate and operate components, including local controls.

2.4.31 -Emergency Procedures X / Plan: Knowledge of annunciator alarms, indications.

or respOnse procedures.

2.9 3.0 3.5 3.2 3.7 3.9 2.6 3.3 4.4 4.2 13 14 15 16 17 18 19 20 21 22 ES-401 4 Form ES-401-1 Oyster Creek IL T 10-1 NRC Exam Written Examination Plant Systems -Tier 2 Group K K K K K K A A A Imp System # I Name A2. G on1 2 3 4 5 6 1 3 4 K5.04 -Knowledge of the operational implications of the following concepts as they apply 211000 SLC X 3.1 23 to STANDBY LIQUID CONTROL SYSTEM: Explosive valve operation A3.01 -Ability to monitor automatic operations of the UNINTERRUPTABLE POWER 262002 UPS (AG/DC) X 2.8 24 SUPPLY (A.C.lD.C.)

including:

Transfer from preferred to alternate source A 1.01 -Ability to predict and/or monitor changes in parameters associated with operating the 264000 EDGs X 3.0 25 EMERGENCY GENERATORS (DIESEL/JET) controls including:

Lube oil temoerature 2.4.47 -Emergency Procedures I Plan: Ability to diagnose and recognize trends in an accurate 218000 ADS X 4.2 26 and timely manner utilizing the appropriate control room reference material.

KIA Category Totals: 2 2 2 3 313 3 2 3/2 Group Point Total: 26/5 2E 2 1 ES-401 5 Form ES-401-1 System # I Name 256000 Reactor Condensate 201oo2RMCS 215001 Traversing In-Core Probe 215001 Traversing In-core Probe 201001 CRD Hydraulic 239001 Main and Reheat Steam 201003 Control Rod and Drive Mechanism 202002 Recirculation Flow Control 241000 ReactorlTurbine Pressure Regulating System 219000 RHRlLPCI:

Torus/Suppression Pool Cooling Mode Oyster Creek ILT 10-1 NRC Exam Written Examination Plant Systems -Tier 2 Group K K K K K K A A A A2 G 1 2 3 4 5 6 1 3 4 A2.15 -Ability to (a) predict the Impacts of the following on the REACTOR CONDENSATE SYSTEM; and (b) based on X those predictions.

use procedures to correct. control. or mitigate the consequences of those abnormal conditions or operations:

Abnormal water auality I 2.2.40 -Ability to apply X Technical Specifications for a I system. I X 2.1.32 -Ability to explain and I:;cly system limits and ecautions.

K 1.10 -Knowledge of the physical connections and/or cause-effect relationships X between TRAVERSING CORE PROBE and the following:

Area radiation monitoring system: (Not-BWR1)

K2.05 -Knowledge of electrical X power supplies to the following:

Altemate rod insertion valve solenoids:

Plant-Specific K3.16 -Knowledge of the effect that a loss or malfunction of the X MAIN AND REHEAT STEAM SYSTEM will have on following:

Relief/safety valves K4.02 -Knowledge of CONTROL ROD AND DRIVE X MECHANISM design feature(s) and/or interlocks which provide for the following:

Detection of an uncoupled rod K5.02 -Knowledge of the operational implications of the X follOwing concepts as they apply to RECIRCULATION FLOW CONTROL SYSTEM: Feedback signals K6.07 Knowledge of the effect that a loss or malfunction of the following will have on the X REACTORITURBINE PRESSURE REGULATING SYSTEM: Turbine inlet pressure A 1.07 -Ability to predict and/or monitor changes in parameters associated with operating the X RHRlLPCI:

TORUS/SUPPRESSION POOL COOLING MODE controls including:

Emergency generator loading Imp. 3.3 4.7 4.0 2.6 4.5 3.6 3.8 2.6 3.4 3.2 Q # 16 17 18 27 28 29 30 31 32 33 ES-401 5 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline Written Examination Outline Plant Systems -Tier 2 Group 2 System # I Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A2 A 3 A 4 G A2.06 -Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM; and (b) based on those 202001 Recirculation X predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Inadvertent recirculation flow decrease A3.02 -Ability to monitor 234000 Fuel Handling Equipment X automatic operations of the FUEL HANDLING EQUIPMENT including:

tlnterlock operation A4.02 -Ability to manually 259001 Reactor Feedwater X operate andlor monitor in the control room: Manually start/control a RFPITDRFP 2.1.28 -Conduct of Operations:

204000 RWCU X Knowledge of the purpose and function of major system components and controls.

K3.02 -Knowledge of the effect 216000 Nuclear Boiler Instrumentation X that a loss or malfunction of the NUCLEAR BOILER Instrumentation will have on following:

PCIS/NSSSS KIA Category Totals: 1 1 1211 1 1 1 1/1 1 1 112 Group Point Total: Q Imp. # 3.6 34 3.1 35 3.9 36 4.1 37 4.0 38 I 1213 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Oyster Creek IL T 10-1 NRC Facility:

Date: 07/11111 Exam Outline Category KIA # 2.1.36 2.1.37 Topic Knowledge of procedures and limitations involved in core alterations.

Knowledge of procedures, guidelines, or limitations associated with reactivity management.

RO IR Q# SRO-Only IR Q# 4.1 19 4.6 23 1. Conduct of Operations 2.1.28 Knowledge of conservative decision making practices.

Knowledge of the purpose and function of major system comp<>nents and controls.

3.6 4.1 66 67 Subtotal 2.2.22 2.2.11 Knowledge of limiting conditions for operations and safety limits. Knowledge of the process for contrOlling temporary design changes. 2 4.7 3.3 2 20 25 2. Equipment Control 2.2.1 2.2.36 2.2.2 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. 4.5 3.1 4.6 68 69 74 3. Radiation Control Subtotal control radiation releases. 2 3 4 Knowledge of radiation exposure limits under . . normal or emergency conditions.

3 4.3 3.7 2 21 24 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. 2.9 70 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Knowledge of Radialogical Safety Principles pertaining to licensed operator duties, such as I containment entry requirements, fuel handling 3.2 71 responsibilities, access to locked high-radiation areas, aligning filters, etc.

2 2 Knowledge of EOP mitigation strategies.

4.7 22 i I Knowledge of EOP layout, symbols, and I3.4 72 icons.Emergency Procedures

/ Knowledge of EOP entry conditions and 4.6 73 ! immediate action steps. Knowledge of the bases for prioritizing safety functions during abnormal/emergency 3.6 75 operations.

3 1 Tier 3 Point 10 7 ES-401 Randomly Tier / Group Selected KJ A 2/1 RO 211000 A4.06 2/1 RO 263000 A4.01 2/1 RO 211000 K5.04 1/1 SRO 295026 2.2.25 1/1 RO 2950192.1.31 2/2 SRO 201002 2.2 40 2/2 SRO 233000 2.1.34 3RO 2.2.1 3RO 2.2.2 3SRO 2.3.11 3SRO 2.4.6 2/2 RO 241000 K6.07 2/2 RO 219000 A1.07 2/2 RO 202001 A2.06 Record of Rejected KIA's Form ES-401-4 Reason for Rejection 211000 A3.06 -Oyster Creek does not have an automatic SLC initiation.

A new KIA was randomly selected.

300000 A4.01 -KIA rejected due to overlap with RO question #19. Could not generate a question which would discriminate from question #19. A new KIA was randomly selected.

211000 A2.03 -KIA rejected due to overlap with Audit Exam question #17 and NRC Simulator Scenario #2. A new KIA was randomly selected.

2950262.2.38

-KIA rejected due to no ties to Suppression Pool in the Facility License. A new KIA was randomly selected.

[KIA 295026 2.2.38 rejected due to being able to tie Suppression Pool temperature to the Facility License].

2950192.2.38

-KIA rejected due to no ties to Instrument Air in the Facility License. A new KIA was randomly selected.

201002 2.2.4 -KIA rejected due to Oyster Creek not having a multi-unit license. A new KIA was randomly selected.

233000 2.1.31 -KIA rejected due to not having a 10CFR55.43(b) link and KIA 2.1.31 being RO level of knowledge.

A new KIA was randomly selected.

2.2.3 -KIA rejected due to Oyster Creek not having a multi-unit license. A new KIA was randomly selected.

2.2.17 -KIA supports testing at the SRO-Only level, but NOT at the RO level due to job responsibilities.

A new KIA was randomly selected.

2.3.15 -KIA rejected due to overlap with RO question #70 (also KIA 2.3.15). A new KIA was randomly selected.

2.4.1 -KIA rejected due to supporting testing at the RO level, but not SRO-Only level due to job responsibilities.

A new KIA was randomly selected.

234000 K6.07 -An operationally relevant question could not be written due to a loss of RBHVAC having no specific affect on Fuel Handling Equipment.

A new KIA was randomly selected.

256000 A 1 .07 -An operationally relevant question could not be written at an LOD level greater than 1. A new KIA was randomly selected.

202001 A2.13 -KIA rejected due to being associated ES-401 2/2RO 234000 A3.02 2/2 RO 259001 A4.02 2/2 RO 216000 K3.02 1/1 RO 295006 2.4.4 1/1 RO 295025 2.4.20 1/2 RO 295032 EK1.03 1/1 SRO 295021 2.4.45 1/1 SRO 295030 EA2.03 2/1 SRO 259002 A2.04 2/1 SRO 261000 A2.15 2/1 SRO 212000 A2.11 2/2SRO 256000 A2.15 2/2SRO 215001 2.1.32 1/1 RO 700000 AK1.02 Record of Rejected KIA's Form ES-401-4 with Generic Fundamentals.

A new KIA was randomly selected.

290003 A3.02 -KIA rejected due to the Control Room HVAC not having any automatic initiations/failures during a fire. A new KIA was randomly selected.

259001 A4.06 -KIA rejected due to being associated with Generic Fundamentals.

A new KIA was randomly selected.

226001 K3.02 -KIA rejected due to concept overlap with NRC Simulator scenario #3 and Audit simulator scenario #1. A new KIA was randomly selected.

295006 2.4.34 -There are no RO tasks outside the Control Room for a Scram, only non-Licensed Operator Tasks. A new KIA was randomly selected.

2950302.4.20

-Low Torus Level EOP does not have any warnings, cautions, or notes. A new KIA was randomly selected.

295032 EK1.04 -Unable to develop three credible or plausible distracters.

A new KIA was randomly selected.

295021 2.4.35 -Unable to develop an operationally relevant question.

A new KIA was randomly selected.

295030 EA2.04 -KIA rejected due to overlap with Audit SRO question #2 (also KIA 295030 EA2.04). A new KIA was randomly selected.

400000 A2.04 -Could not develop an operationally relevant question connecting monitors to a CCW system at Oyster Creek. A new KIA was randomly selected.

261000 A2.14 -Could not develop an operationally relevant question.

A new KIA was randomly selected.

262001 A2.10 -KIA supports testing at the RO level, but no the SRO-Only level due to job responsibilities.

A new KIA was randomly selected.

214000 A2.01 -There are no operationally relevant abnormal, emergency, or Tech Spec actions for a failed RPIS reed switch. A new KIA was randomly selected.

233000 2.1.34 -Oyster Creek does not have any chemistry specifications in the Technical Specifications.

KIA rejected and a new KIA was randomly selected.

700000 AK1.01 -KIA related to Generic Fundamentals; concept tested on NRC GFE exam. A new KIA was randomly selected.

Administrative Topics Outline Form ES-301-1 Facility:

Oyster Creek Date of Examination:

7/11/11 Examination Level: RO I3J SRO 0 Operating Test Number: IL T 10-1 Administrative Topic Type Describe activity to be performed (See Note) Code* Perform Week 4 of 680.4.007, Safety Related Equipment Conduct of Operations P,S Verification; G2.1.29 (4.1) [NRC Admin JPM1 (RO)l Perform Core Thermal Limits Verification; G2.1.7 (4.4) Conduct of Operations D,R [NRC Admin JPM2 (RO)] Equipment Control Application of Radiation Exposure Limits lAW Procedure Radiation D,R RP-AA-203; G2.3.4 (3.2) [NRC Admin JPM3 (RO)l Determine Primary Containment Water Level lAW EMG-Emergency ProcedureS/Plan M,R SP28; G2.4.21 (4.0) [NRC Admin JPM4 (RO)] 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

& Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank 3 for ROs; 54 for SROs & RO retakes) (N)ew or (M)odified from bank 1) {P)revious 2 exams (5 1; randomly selected)

ES 301, Page 22 of 27 Administrative Topics Outline Form ES-301-1 Facility:

Oyster Date of Examination:

7/11/11 Examination Level: RO 0 SRO Operating Test Number: ILT 10-1 Administrative Topic Type Describe activity to be performed (See Note) Code* Review the Technical Specification Log Sheet; G2.1.3 Conduct of Operations N,R (3.9) [NRC SRO Admin JPM1] Review a Completed Pre-Critical Checkoff lAW Procedure Conduct of Operations P, R 201; G2.1 .23 (4.4) [NRC SRO Admin JPM2] Review the acceptance criteria for surveillance procedure 609.3.022, "Au Isolation Condenser Isolation Test and Equipment D,R Calibration A1 Sensors First; G2.2.12 (4.1) [NRC SRO Admin JPM3] Authorize TIP Room Entry; G2.3.13 (3.8) [NRC SRO Radiation D,R Admin JPM4] Determine Primary Containment Water Level lAW Emergency Procedures/Plan M,R SP28 and determine required action; G2.4.21 (4.6) SRO Admin All items (5 total) are required for SROs. RO applicants require only 4 items unless they retaking only the administrative topics, when 5 are

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or (D)irect from bank (s 3 for ROs; :s 4 for SROs & RO (N)ew or (M}odified from bank (P)revious 2 exams (s 1; randomly ES 301 , Page 22 of 27 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Oyster Creek Date of Examination: Exam level: RO 0 SRO-I D SRO-U D Operating Test Number: IlT Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System I JPM Title Perform Core Spray Surveillance with faulted Core Spray Pump lAW 610.4.002, Core Spray Pump Operability Test (Alternate Path); 209001 A4.01 (3.8/3.6)

[NRC Sim JPM1] Shutdown of the Automatic Depressurization System lAW 308, Emergency Core Cooling System Operation; 218000 A4.03 (4.214.2)

[NRC Sim JPM2] Cool dovvn the RPV using the Isolation Condenser tube side vents lAW EMG-SP15, Alternate Pressure Control Systems -IC Tube Side Vents (Alternate Path); 295021 AA 1.04 (3.7/3. 7f [NRC Sim JPM3] d. Place the H2I02 monitoring system in service lAW EMG-SP39, PlaCing The H2I02 Monitoring System In Service; 500000 EA 1.01 (3.4/3.3)

[NRC Sim JPM4] Transfer 4160 VAC Bus 1A to the Startup Transformers (Alternate Path); 262001

[NRC Sim JPM5] Perform an APRM Gain Adjustment; 215005 A4.03 (3.213.3)

[NRC Sim JPM 6] g. Inject Fire Water via the Core Spray System lAW SP-20, low Pressure Injection During An ATWS; 286000 A4.06 (3.4/3.4)

[NRC Sim JPM7] h. Startup of the Turbine Building Ventilation System lAW 328, The Turbine Building iHeating And Ventilation System (Alternate Path); 288000 A4.01 (3.1/2.9)

[NRC Sim JPM8] In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) Vent the scram air header lAW EMG-SP21, Alternate Insertion of Control Rods; 295037 EA1.05 (3.9/4.0)

[NRC Plant JPM1] Add makeup from Fire Water to the Isolation Condensers lAW 307, Isolation Condenser 207000 K1.06 (3.3/3.7)

[NRC Plant ..IPM3] Operate Service Water Pump 1-2 from Local Shutdown Panel 1 B3 1 B3) lAW 346, Operation of the Remote and Local Shutdown Panels; 295016 AA1.07 (4.2/4.3)

[NRC Plant JPM2)

• This plant JPM will be performed on the Simulator replica of LSP-1 B3. Safety Type Code* Function D,A,S 2 P, D, l, EN, 3 S M, A, l, EN, 4 S N,l,S 5 D,A,S M,S N,S 8 p, D, A, l, S 9 D,E,R D, l, R D, EN, S* ! 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 oVElriap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6/ 4-6 / (C)ontrol (EN)gineered safety feature -/ -/ 1 (control room (P)revious 2 exams / 2 (randomly (D)irect from bank 9 / / (E)mergency or abnormal in-plant 1 / / (L)ow-Power

/ Shutdown 1 / / (N)ew or (M)odified from bank including 1 (A) / (R)CA 1 / / ES-301, Page 23 of Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Oyster Creek Date of Examination:

7/11/11 Exam Level: RO D SRO-I [8l SRO-U D Operating Test Number: ILT 10-1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title a, Perform Core Spray Surveillance with faulted Core Spray Pump lAW 610.4.002, Core Spray Pump Operability Test (Alternate Path); 209001 A4.01 (3.8/3.6)

[NRC Sim JPM1l Shutdown of the Automatic Depressurization System lAW 308, Emergency Core Cooling System Operation; 218000 A4.03 (4.214.2)

[NRC Sirn ...IPM2] Cool down the RPV using the Isolation Condenser tube side vents lAW EMG-SP'15, Alternate Pressure Control Systems -IC Tube Side Vents (Alternate Path); 295021 AA1.04 (3.7/3.7)

[NRC Sim JPM3] d. Place the H2/02 monitoring system in service lAW EMG-SP39, Placing The H2/02 Monitoring System In Service; 500000 EA 1,01 (3.4/3.3)

[NRC Sirn JPM4] Transfer 4160 VAC Bus 1A to the Startup Transformers (Alternate Path); 262001 K4,02 (2.9/3.3)

[NRC Sirn JPM5] Perform an APRM Gain Adjustment; 215005 A4.03 (3,213.3)

[NRC Sirn JPM 6] g. h. Startup of the Turbine Building Ventilation System lAW 328, The Turbine Building Heating And Ventilation System (Alternate Path); 288000 A4.01 (3.1/2.9)

[NRC Sirn JPM8] In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) Vent the scram air header lAW EMG-SP21, Alternate Insertion of Control Rods; 295037 EA1.05 (3,9/4.0)

[NRC Plant JPM1] Add makeup from Fire Water to the Isolation Condensers lAW 307, Isolation Condenser System; 207000 K1.06 (3.3/3.7)

[NRC Plant JPM3] Operate Service Water Pump 1-2 from Local Shutdown Panel 183 1 B3) lAW 346, Operation of the Remote and Local Shutdown Panels; 295016 AA1.07 (4.214.3)

[NRC Plant JPM2)

• This plant JPM will be performed on the Simulator replica of LSP-1 83. Type Code* Safety Function D,A,S 2 P, D, L, EN, 3 S M, A, L, EN, 4 S N, L, S 5 D,A,S 6 M,S 7 P, D, A, L, S 9 D, R 1 D, L, R 4 D, EN, S* 7 All R:O and SRO-I control room (and in-plant) systems must be different and serve different safety func'tions; 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-I / SRO-U (A)lternate path 4-6/ 4-6 / 2-3 (C}ontrol room (D)irect from bank '5:. 91 '5:.8 1 '5:. 4 (E}mergency or abnormal in-plant 2. 1 / 2. 1 I 2. 1 (EN}gineered safety feature -I -1 2. 1 (control room system (L)ow-Power 1 Shutdown 2. 1 / 2. 1 I 2. 1 (N}ew or (M}odified from bank including 1 (A) 2. 21 2. 2 / 2. 1 (P)revious 2 exams '5:. 31 '5:.3 / '5:. 2 (randomly selected) (R)CA 2. 1 1 2.1 1 2. 1 (S)imulator ES-301, Page 23 of 27 IL T 10-1 NRC Scenario 2 (NEW) Scenario Outline Facility:

Oyster Scenario No.: " Op Test No.: 10-1 NRC Operators:

Initial Conditions: , . 97% power

• Main Generator voltage control is in Manual Turnover:

• Place the amplidyne in automatic service lAW 336.1, section 8

• Raise reactor power to 100% with reCirculation flow Event No. Malt. N --Event Description 1 NA N Return the Amplidyne to service lAW 336.1. Raise reactor power to 100% with recirculation flow 2 NA R ATC (REMA). ICH-Respond to RPV High Pressure Instrument RE15 toTS SRO NSS026A Isolation Condenser Initiation Logic Failing High. ! MAL-I ATCRespond to APRM 2 failing INOP.NIS0218 TS 5 MAL-C BOP Respond to trip of Steam Packing Exhauster C Respond to the E EMRV lifting leading the crew to a 6 MAL-BOP NSS025E TS manual scram. 7 CAEP M Crew Respond to an Electric A ATWS C Crew Respond to a Standby Liquid Control Pump shaft break. .. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs IL T 10-1 NRC Scenario Page 1 of 23 IL T 10-1 NRC Scenario 2 (NEW) Simulai4)r Summary Event Event Summary 1 The BOP will return the amplidyne to automatic service lAW 336.1. The BOP will place the control switch in TEST, zero the amplidyne voltmeter, then place the control switch to ON. (BOP: Normal Evolution) 2 The ATC will raise reactor power to rated power (100%) with recirculation flow using the Master Recirc Speed Controller. (ATC: Reactivity Manipulation) 3 RPV high pressure instrument RE15A to Isolation Condenser Initiation Logic fails high. No Isolation Condenser initiation will occur from this failure. The SRO will review and apply Tech Spec Table 3.1.1 part C.1. (SRO: Tech Specs) 4 The ATC will respond to an APRM HI-HI/INOP alarm and report that APRM 2 has failed INOP with a Y2 scram on RPS 1. The ATC will bypass APRM 2 lAW procedure 403 and reset the Y2 scram. The SRO will review and apply Tech Spec Table 3.1.1. (ATC: Instrument Malfunction; SRO: Tech Specs) 5 The BOP will respond to the failure of the in-service steam packing exhauster.

The BOP will start the standby Exhauster and throttle open its discharge valve to maintain the correct vacuum. (BOP: Component Malfunction) 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 3.5.A for Torus Temperature limits. (ATC: Component Malfunction; BOP: Component Malfunction; SRO: Tech Specs) 7 The Crew will diagnose an electric A TWS 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) IL T 10-1 NRC Scenario 2 Page 2 of 23 8 IL T 10-1 NRC Scenario 2 (NEW) 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)

With reactor power> 2% during an A TWS, terminate and prevent Task injection into the RPV to intentionally lower RPV water level which will lower reactor power.

Crew directs the Reactor Building EO to vent the scram air header. Task (The Lead Examiner will direct the Booth to vent the scram air header at their discretion).

ES-301-4 Target Actual Event Attributes Number{s} Total malfunctions (5-8) 6 3,4,5,6,7,8 Malfunctions after EOP entry (1-2) 1 8 Abnormal events (2-4) 2 4,6 Major transients (1-2) 1 7 EOPs entered/requiring substantive 2 7 actions (1-2) EOP contingencies requiring substantive 1 7 actions (0-2) Critical tasks (2-3) 2 7 ILT 10-1 NRC Scenario Page 30f23 IL T 10-1 NRC Scenario 3 (NEW) Scenario Outline Facility:

Oyster Scenario No.: Op Test No.: 10-1 NRC Operators:

Initial Conditions: 75% power TBCCW Pump 2 is tagged out of service Lower power to 70% using recirculation flow lAW 1001.22-3, Core Maneuvering Daily Instruction Sheet Backwash the Main Condenser Half B South I Event No. Malf. No. EventType*

Event Description 1 NA R ATC Lower reactor power to 70% using recirculation flow 2 NA N BOP Continue backwashing Main Condenser Half B South BKR-C ATCRespond to a CRD Pump A trip CRDOO2 TS I BOP Respond to the EPR setpoint failing low BOP 5 TBS027C C Respond to a trip of Control Room Vent Fan B TS SRO ANN-L4f MAL-I ATC Respond to a variable leg leak in the A and C GEMAC 6 NSS012E TS SRO RPV level indicators ID13A and ID13C M Crew Respond to a loss of all CRD Flow M Crew Respond to a Safety Valve lifting post scram Respond to a trip of the operating Containment Spray 9 CNSOO4A-C Crew Pump 0 * ("'I)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs IL T 10-1 NRC Scenario Page 1 of 28 IL T 10-1 NRC Scenario 3 (NEW) Simulator Summary Event Event Summary 1 The ATC will lower reactor power to approximately 70% 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 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) 4 The BOP will respond to the EPR setpoint failing low (indicating 890 psig) lAW ABN-9, Electric Pressure Regulator Malfunction.

The BOP will transfer RPV pressure control to the MPR and secure power to the EPR. The BOP will then raise RPV pressure as directed by the SRO. (BOP: Instrument Malfunction) 5 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) 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 water 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 CRO Pump B will trip on overload and the Crew will respond to a loss of all CRO flow lAW RAP H-2-c. Upon receipt of two HCU accumulator alarms, the Crew will manually scram the reactor perform post scram actions. (Major Evolution) (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)

ILT 10-1 NRC Scenario 3 Page 2 of 28 9 ILT 10-1 NRC Scenario 3 (NEW) When initiating Containment Spray lAW the Primary Containment 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 altemate Containment Spray Pump. (Component Failure After EOP)

The Crew must manually scram the reactor following a loss of all CRD Task pumps. With no CRD flow at high power, damage to the CRD drive mechanisms will occur potentially inhibiting the ability to successfully scram.

When Drywell or Torus pressure exceeds 12 psig, OR before Drywell Task bulk temperature is determined it cannot be maintained below 281°F, spray the drywell lAW SP-29, Initiation of the Containment Spray System for Drywell Sprays. ES..301-4 Target Actual Event Attributes Numbetls) Total malfunctions (5 ..8) 7 3,4,5,6, 7, 8,9 2. Malfunctions after EOP entry (1-2) 8,9 3. Abnormal events (2-4) 4,6 4. Major transients (1-2) 7,8 EOPs entered/requiring substantive 1 8 actions (1-2) EOP contingencies requiring substantive 0 N/A actions (0-2) 7. Critical tasks (2-3) 7,9 IL T 10-1 NRC Scenario Page 3 of 28 ILT 10-1 NRC Scenario 4 (NEW) Scenario Outline Facility:

Oyster Creek Scenario No.: Op Test No.: 10-1 NRC Examiners:

Operators:

Initial Conditions:

• 100% power

• Dilution pump 2 is tagged out of service

• No evolutions are planned during this shift --N... 1UI"lf. No. EventType*

I MAL-C ATC Respond to CRD Flow Control Valve failed dosed. ICH-C 2 Respond to a leak in Isolation Condenser Shell A.ICSOO1A TS 18 R ATC Condensate Pump A experiences high amps requiring a 3 -2B C BOP rapid power reduction and securing of Condensate Pump. C RCPOO3D BOP 4 Respond to Recirculation Pump 0 total seal failure. MAL-TS -C ATC Respond to multiple drifting control rods. MAL-Respond to a Torus Leak requiring entry into Primary 6 M Crew PCNOO7 Containment Control. VLV-Respond to Core Spray system suction valves being 7 CSS001, C Crew 009 mechanically seized when lining up the CST to the Torus. MAL-Respond to a Torus leak increase requiring tV 8 M Crew PCNOO7 Emergency Depressurize.

• (N)ormal, (R)eactlvity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs ILT 10-1 NRC Scenario 4 Page 1 of 24 ILT 10-1 NRC Scenario 4 (NEW) Simulator Summary Event Event Summary 1 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) 2 The BOP will respond to lowering level in Isolation Condenser A shell lAW RAP-C6a. The BOP will fill the IC shell from Pane15F/6F.

The IC shell level will not recover above the required point of 7.3 ft and the SRO will declare IC A inoperable and apply Tech Spec 3.B.C. (BOP: Component Malfunction; SRO: Tech Specs) 3 Condensate Pump A will experience rising motor amps.The crew will receive annunciator K-2-f, CONDENSATE PUMP B OL, and the BOP will diagnose the Condensate Pump A current indication in the Control Room is high. The field operator (on request) will report loud noise coming from Condensate Pump A. The SRO will direct the ATC to perform a rapid power reduction and the BOP to remove one Feedwater Pump and the affected Condensate Pump from service. (ATC: Reactivity Manipulation; BOP: Component Malfunction) 4 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 0 Recirculation Loop. The SRO will review and apply Tech Specs 3.3.0 and 3.3.F for unidentified leak rate and recirculation loop operability. (BOP: Component Malfunction; SRO: Tech Specs) 5 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) 6 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 System. (Major Evolution) 7 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 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) IL T 10-1 NRC Scenario 4 Page 2 of 24 8 IlT 10-1 NRC Scenario 4 (NEW) 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)

The ATC will manually scram the reactor following control rods drifting Task into the core. There is no manual scram associated with this casualty and the core is not analyzed for the resultant abnormal rod configuration.

Critical The crew will Emergency Depressurize the RPV prior to Torus level Task 2 reaching 110 inches. ES-301-4 Target Actual Event Attributes Totall malfunctions (5-8) 7 1, 2, 3, 4, 5, 6, 7 2. Malfunctions after EOP entry (1-2) 7,8 Abnormal events (2-4) 3 1,4,5 4. Major transients (1-2) 6,8 EOPs entered/requiring substantive 2 5,6 actions (1-2) EOP contingencies requiring substantive 1 8 actions (0-2)

7. Critical tasks (2-3) 5,8 Il T 10-1 NRC Scenario Page 30f24