ML111870267
| ML111870267 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/27/2011 |
| From: | Todd Fish Operations Branch I |
| To: | Exelon Nuclear |
| Hansell S | |
| Shared Package | |
| ML110030666 | List: |
| References | |
| TAC U01831 | |
| Download: ML111870267 (30) | |
Text
ES-401 Written Examination Outline Form ES-401-1 Oyster Creek ILT 10-1 i Facility: Date of Exam: 07/11/11 NRC Exam Outline RO KIA Category Points SRO-Only Points !
Grou Tier K K K K K K A A A A G Tota p A2 G* Total 1 2 3 4 S 6 1 2 3 4
- I
- 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
& Tier Plant Total 4 5 4 4 5 5 27 5 5 10 Evolutions s
1 2 2 2 2 3 2 3 2 2 3 3 26 3 2 S 2.
2 1 1 2 1 1 1 1 1 1 1 1 12 0 1 2 3 Plant Systems Tier Total 3 3 4 3 4 3 4 3 3 4 4 38 4 4 8 s
- 3. Generic Knowledge & Abilities 1 2 3 4 1 2 3 4 10 7 Categories 2 3 2 3 2 2 2 1 Note 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals' in each KIA category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the 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.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, Site-specific systems that are not included on the outline should be added. Refer to section D.1.b of ES-401, for guidance regarding elimination of inappropriate KIA statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
S. 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) 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
- 8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the KIA Catalog, and enter the KIA numbers, deSCriptions, IRs, and point totals (#) on Form ES-401-3. limit SRO selections to KlAs that are
ES-401 Written Examination Outline Form ES-401-1 linked to 10CFR55.43
ES-401 2 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 I EAPE # I Name Safety Function KIA Topic(s)
AA2.05 - Ability to determine and/or interpret the following as they apply to 295006 SCRAM 11 X SCRAM: Whether a reactor SCRAM 4.6 1 has occurred AA2.01 - Ability to determine and/or interpret the following as they apply to 295004 Partial or Total Loss of DCPwr/6 X PARTIAL OR COMPLETE LOSS OF 3.6 2 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 Complete PARTIAL OR COMPLETE LOSS OF LossofAC/6 X A.C. POWER: Whether a partial or 4.2 3 complete loss of A.C. power has occurred 295026 Suppression Pool High 2.2.38 - Knowledge of conditions and Water Temp. I 5 X limitations in the facility license.
4.5 4 295037 SCRAM Conditions 2.4.8 - Emergency Procedures I Plan:
Present and Rec:lctor Power Knowledge of how abnormal operating Above APRM Downscale or X procedures are used in conjunction with 4.5 5 Unknown I 1 EOP's.
2.4.45 - Ability to prioritize and interpret 295021 Loss of Shutdown Cooling 14 X the significance of each annunciator or 4.3 6 alarm.
EA2.03 - Ability to determine and/or 295030 Low Suppression Pool interpret the following as they apply to Water Levell 5 X LOW SUPPRESSION POOL WATER 3.9 7 LEVEL: Reactor pressure AK1.01 - Knowledge of the operational implications of the following concepts as 700000 Generator Voltage and they apply to GENERATOR VOLTAGE Electric Grid Disturbances X AND ELECTRIC GRID 3.3 39 DISTURBANCES and the following:
Over-excitation AK1.05 - Knowledge of the operational implications of the following concepts as 295004 Partial ()r Total Loss of DCPwr/6 X they apply to PARTIAL OR COMPLETE 3.3 40 LOSS OF D.C. POWER: Loss of breaker protection AK1.01 - Knowledge of the operational implications of the following concepts as 295005 Main Turbine Generator Trip I 3 X they apply to MAIN TURBINE 4.0 41 GENERATOR TRIP: Pressure effects on reactor power AK2.02 - Knowledge of the 295023 Refueling Acc Cooling interrelations between REFUELING X ACCIDENTS and the following: Fuel 2.9 42 Mode/8 pool cooling and cleanup system EK2.03 - Knowledge of the 295038 High Olf-site Release interrelations between HIGH OFF-SITE X 3.6 43 Rate I 9 RELEASE RATE and the following:
Plant ventilation systems EK2.01 - Knowledge of the 295028 High Drywall interrelations between HIGH DRYWELL Temperature 15 X TEMPERATURE and the following:
3.7 44 Drywell spray: Mark-I&II EK3.04 - Knowledge of the reasons for 295031 Reactor Low Water the following responses as they apply to Level 12 X REACTOR LOW WATER LEVEL:
4.0 45 Steam cooling
ES-401 2 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 I EAPE # / Name Safety Function KIA Topic{s)
AK3.02 - Knowtedge of the reasons for 295021 Loss of Shutdown the following responses as they apply to X 3.3 46 Cooling /4 LOSS OF SHUTDOWN COOLING:
FeedinQ and bleedinQ reactor vessel EK3.04 - Knowtedge of the reasons for 295024 High D~{We1l Pressure / the following responses as they apply to X 3.7 47 5 HIGH DRYWELL PRESSURE:
Emergency depressurization AA1.09 - 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 AA1.03 - 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 EA1.06 - Ability to operate and/or 295025 High Reactor Pressure monitor the following as they apply to X 4.5 50
/3 HIGH REACTOR PRESSURE: Isolation condenser: Plant-Specific EA2.03 - Ability to determine and/or 295026 Suppression Pool High interpret the following as they apply to X 3.9 51 WaterTemp./5 SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor pressure AA2.16 - Ability to determine and interpret the following as they apply to 600000 Plant Fire On-site / 8 X 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 room 295019 Partial or Total Loss of switches, controls, and indications, and X 4.6 54 Inst. Air /8 to determine that they correctly reflect the desired plant lineup.
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 flow EK2.14 - Knowtedge of the 295037 SCRAM Conditions interrelations between SCRAM Present and Reactor Power CONDITION PRESENT AND X 3.6 58 Above APRM DClwnscale or REACTOR POWER ABOVE APRM Unknown /1 DOWNSCALE OR UNKNOWN and the following: RPIS: Plant-5pecific KIA Category Totals: 3 4 3 3 414 3/3 Group Point Total:
I 20/7
ES-401 3 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE # / Name Safety Function 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 SU~lpression Pool 4.0 Water Levell 5 X explain and apply all system limits and 9 precautions.
2.2.44 - Equipment Control: Ability to interpret control room indications to 295020InadvertentConl verify the status and operation of a X 4.4 10 Isolation /5 & 7 system. and understand how operator actions and directives effect plant and system conditions.
EK1.03 - Knowledge of the operational implications of the following concepts as 295032 High Secondary they apply to HIGH SECONDARY 3.
Containment Area Temperature X CONTAINMENT AREA 5
59
/5 TEMPERATURE: Secondary containment leakage detection: Plant-Specific AK2.01 - Knowledge of the interrelations between HIGH 295013 High Suppression Pool 3.6 X SUPPRESSION POOL 60 Temperature /5 TEMPERATURE and the following:
Suppression pool cooling AK3.05 - Knowledge of the reasons for 295010 High Drywall Pressure the following responses as they apply to 3.
X HIGH DRYWELL PRESSURE: 5 61
/5 Temperature monitoring AA1.06 - Ability to operate and/or monitor the following as they apply to 295002 Loss of Main 3.
X LOSS OF MAIN CONDENSER 0
62 Condenser Vac l 3 VACUUM: Reactor/turbine pressure regulating system AA2.01 - Ability to determine and/or 295022 Loss of CRD Pumps I interpret the following as they apply to 3.
1 X LOSS OF CRD PUMPS: Accumulator 5 63 pressure 2.4.35 - Emergency Procedures / Plan:
295036 Secondary Knowledge of local auxiliary operator 3.
Containment High SumplArea X tasks during emergency and the 8 64 Water Levell 5 resultant operational effects.
2.4.8 - Emergency Procedures I Plan:
295009 Low Reactor Water Knowledge of how abnormal operating 3.
Level 12 X procedures are used in conjunction with 8 65 EOP*s.
KIA Category Totals: 1 1 1 1/1 212 Group Point Total:
I 7/3
ES-401 4 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline 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 A2.04* Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM and (b) 259002 Reactor Water Level X based on those predictions. use 3.1 11 Control System 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 261000SGTS those predictions. use X 3.4 12 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 218000 ADS shutdown implications in X 4.2 13 acckfent (e.g.* loss of coolant acckfent or loss of residual heat removal) mitigation strategies.
2.4.46
- Emergency Procedures 215005 APRM I LPRM / Plan: Ability to verify that the X 4.2 14 alarms are consistent with the
~ant conditions.
A2.01 - Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on 212000 RPS those predictions. use X 3.9 15 procedures to correct. control. or mitigate the consequences of those abnormal conditions or operations: RPS motor aenerator set failure K1.06 - Knowledge of the physical connections and/or cause- effect relationships 218000 ADS X between AUTOMATIC 3.9 1 DEPRESSURIZATION SYSTEM and the following: Safety/relief valves K1.05 - Knowledge of the physical connections and/or cause- effect relationships 205000 Shutdown Cooling between SHUTDOWN X 3.1 2 COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and the following: Component cooling water systems 400000 Component Cooling Water X
0 K2.02 - Knowledge of electrical power supplies to the following:
CCWvalves 2.9 3
ES-401 4 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline 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 K2.01
- Knowledge of electrical 263000 DC Electrical X power supplies to the following: 3.1 4 Distribution i Maior D.C. loads K3.03 - Knowledge of the effect that a loss or malfunction of the LOW PRESSURE CORE 209001 LPCS X 2.9 5 SPRAY SYSTEM will have on following: Emergency aenerators K3.02
- Knowledge of the effect that a loss or malfunction of the ISOLATION (EMERGENCY) 207000 Isolation CONDENSER will have on X 3.8 6 (Emergency) Condenser 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 223002 PCIS/Nuclear X feature( 5) and/or interlocks 3.4 7 Steam Supply Shutoff 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 212000 RPS X 3.0 8 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 264oo0EDGs X 3.4 9 to EMERGENCY GENERATORS (DIESEUJET) :
Load seauencina K5.01 - Knowledge of the operational implications of the following concepts as they 262001 AC Electrical X apply to A.C. ELECTRICAL 3.1 10 Distribution 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 262002 UPS (AClDC) X 2.7 11 UNINTERRUPTABLE POWER SUPPLY (A.C.lD.C.) : A.C.
electrical power K6.04 - Knowledge of the effect that a loss or malfunction of the 215004 Source IRange X following will have on the 2.9 12 Monitor SOURCE RANGE MONITOR (SRH) SYSTEM: Detectors
ES-401 4 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A Imp System # I Name A2 G Q#
1 2 3 4 5 6 1 3 4 A1.05 - Ability to predict and/or monitor changes in parameters associated with operating the 259002 Reactor Water X REACTOR WATER LEVEL 2.9 13 Level Control 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 261000 SGTS X STANDBY GAS TREATMENT 3.0 14 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; 215005 APRM I LPRM X and (b) based on those 3.5 15 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, use procedures to correct.
239002SRVs X 3.2 16 control, or mitigate the consequences of those abnormal conditions or operations: Low reactor pressure A3.03 - Ability to monitor automatic operations of the 2150031RM X INTERMEDIATE RANGE 3.7 17 MONITOR (IRM) SYSTEM includine: RPS status A4.06 - Ability to manually operate and/or monitor in the 211000 SLC X 3.9 18 control room: RWCU system isolation: Plant-Specific A4.01 - Ability to manually 300000 Instrumlent Air X operate and/or monitor in the 2.6 19 control room: Pressure gauges A4.01 - Ability to manually operate and/or monitor in the 263000 DC ElectriCal Distribution X control room: Major breakers 3.3 20 and control power fuses: Plant-Specific 2.1.30 - Conduct of Operations:
223002 PCIS/Nuclear Ability to locate and operate Steam Supply Shutoff X components, including local 4.4 21 controls.
2.4.31 - Emergency Procedures
/ Plan: Knowledge of 261000 SGTS X 4.2 22 annunciator alarms, indications.
or respOnse procedures.
ES-401 4 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A Imp System # I Name 1 2 3 4 5 6 1 A2.
3 4 G on 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 A1.01 - 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: 2E 2 2 2 2 3 313 3 2 3/2 Group Point Total: 26/5 1
ES-401 5 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline 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 #
A2.15 - Ability to (a) predict the Impacts of the following on the REACTOR CONDENSATE SYSTEM; and (b) based on 256000 Reactor those predictions. use X 3.3 16 Condensate procedures to correct. control.
or mitigate the consequences of those abnormal conditions or operations: Abnormal water auality I 2.2.40 - Ability to apply 201oo2RMCS X Technical Specifications for a 4.7 17 I system.
2.1.32 - Ability to explain and 215001 Traversing In-Core Probe I X I :;cly system limits and 4.0 18 ecautions.
K1.10 - Knowledge of the physical connections and/or cause- effect relationships 215001 Traversing In-core X between TRAVERSING IN 2.6 27 Probe CORE PROBE and the following: Area radiation monitoring system: (Not-BWR1)
K2.05 - Knowledge of electrical power supplies to the following:
201001 CRD Hydraulic X 4.5 28 Altemate rod insertion valve solenoids: Plant-Specific K3.16 - Knowledge of the effect that a loss or malfunction of the 239001 Main and Reheat X MAIN AND REHEAT STEAM 3.6 29 Steam SYSTEM will have on following:
Relief/safety valves K4.02 - Knowledge of CONTROL ROD AND DRIVE 201003 Control Rod and MECHANISM design feature(s)
Drive Mechanism X and/or interlocks which provide 3.8 30 for the following: Detection of an uncoupled rod K5.02 - Knowledge of the operational implications of the 202002 Recirculation Flow follOwing concepts as they apply X 2.6 31 Control to RECIRCULATION FLOW CONTROL SYSTEM:
Feedback signals K6.07 Knowledge of the effect that a loss or malfunction of 241000 ReactorlTurbine the following will have on the Pressure Regulating X REACTORITURBINE 3.4 32 System PRESSURE REGULATING SYSTEM: Turbine inlet pressure A1.07 - Ability to predict and/or monitor changes in parameters associated with operating the 219000 RHRlLPCI:
RHRlLPCI:
Torus/Suppression Pool X 3.2 33 TORUS/SUPPRESSION POOL Cooling Mode COOLING MODE controls including: Emergency generator loading
ES-401 5 Form ES-401-1 Oyster Creek ILT 10-1 NRC Exam Outline 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 #
A2.06 - Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM; and (b) based on those predictions, use procedures to 202001 Recirculation X 3.6 34 correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadvertent recirculation flow decrease A3.02 - Ability to monitor 234000 Fuel Handling automatic operations of the X 3.1 35 Equipment FUEL HANDLING EQUIPMENT including: tlnterlock operation A4.02 - Ability to manually operate andlor monitor in the 259001 Reactor Feedwater X 3.9 36 control room: Manually start/control a RFPITDRFP 2.1.28 - Conduct of Operations:
Knowledge of the purpose and 204000 RWCU X 4.1 37 function of major system components and controls.
K3.02 - Knowledge of the effect that a loss or malfunction of the 216000 Nuclear Boiler NUCLEAR BOILER X 4.0 38 Instrumentation Instrumentation will have on following: PCIS/NSSSS KIA Category Totals: 1 1 1211 1 1 1 1/1 1 1 112 Group Point Total:
I 1213
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Oyster Creek ILT 10-1 NRC Facility: Date: 07/11111 Exam Outline RO SRO-Only Category KIA # Topic IR Q# IR Q#
Knowledge of procedures and limitations 2.1.36 4.1 19 involved in core alterations.
Knowledge of procedures, guidelines, or 2.1.37 limitations associated with reactivity 4.6 23 management.
1.
Conduct Knowledge of conservative decision making of Operations 3.6 66 practices.
Knowledge of the purpose and function of 2.1.28 4.1 67 major system comp<>nents and controls.
Subtotal 2 2 Knowledge of limiting conditions for 2.2.22 4.7 20 operations and safety limits.
Knowledge of the process for contrOlling 2.2.11 3.3 25 temporary design changes.
Ability to perform pre-startup procedures for the facility, including operating those controls 2.2.1 4.5 68
- 2. associated with plant equipment that could Equipment affect reactivity.
Control Ability to analyze the effect of maintenance activities, such as degraded power sources, 2.2.36 3.1 69 on the status of limiting conditions for operations.
Ability to manipulate the console controls as 2.2.2 required to operate the facility between 4.6 74 shutdown and designated power levels.
Subtotal 3 2
- 3. 2.~to control radiation releases. 4.3 21 Radiation 234 Knowledge of radiation exposure limits under Control 3.7 24
. . 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.
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Knowledge of Radialogical Safety Principles pertaining to licensed operator duties, such as I 2.3.12 containment entry requirements, fuel handling 3.2 71 responsibilities, access to locked high-radiation areas, aligning filters, etc.
Subtotal 2 2 2.4.6 Knowledge of EOP mitigation strategies. 4.7 22 i
I
- 4. Knowledge of EOP layout, symbols, and I 2.4.19 3.4 72 Emergency icons.
Procedures / Knowledge of EOP entry conditions and 2.4.1 4.6 73 !
Plan immediate action steps.
Knowledge of the bases for prioritizing safety 2.4.22 functions during abnormal/emergency 3.6 75 operations.
Subtotal 3 1 Tier 3 Point Total 10 7
ES-401 Record of Rejected KIA's Form ES-401-4 Randomly Tier / Group Reason for Rejection Selected KJA 211000 A3.06 - Oyster Creek does not have an 2/1 RO 211000 A4.06 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 2/1 RO 263000 A4.01 would discriminate from question #19. A new KIA was randomly selected.
211000 A2.03 - KIA rejected due to overlap with Audit 2/1 RO 211000 K5.04 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 1/1 SRO 295026 2.2.25 was randomly selected. [KIA 295026 2.2.38 un rejected due to being able to tie Suppression Pool temperature to the Facility License].
2950192.2.38 - KIA rejected due to no ties to 1/1 RO 2950192.1.31 Instrument Air in the Facility License. A new KIA was randomly selected.
201002 2.2.4 - KIA rejected due to Oyster Creek not 2/2 SRO 201002 2.2 40 having a multi-unit license. A new KIA was randomly selected.
233000 2.1.31 - KIA rejected due to not having a 2/2 SRO 233000 2.1.34 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 3RO 2.2.1 multi-unit license. A new KIA was randomly selected.
2.2.17 - KIA supports testing at the SRO-Only level, 3RO 2.2.2 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 3SRO 2.3.11 #70 (also KIA 2.3.15). A new KIA was randomly selected.
2.4.1 - KIA rejected due to supporting testing at the 3SRO 2.4.6 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 2/2 RO 241000 K6.07 no specific affect on Fuel Handling Equipment. A new KIA was randomly selected.
256000 A 1.07 - An operationally relevant question 2/2 RO 219000 A1.07 could not be written at an LOD level greater than 1. A new KIA was randomly selected.
2/2 RO 202001 A2.06 202001 A2.13 - KIA rejected due to being associated
ES-401 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 2/2RO 234000 A3.02 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 2/2 RO 259001 A4.02 with Generic Fundamentals. A new KIA was randomly selected.
226001 K3.02 - KIA rejected due to concept overlap 2/2 RO 216000 K3.02 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 1/1 RO 295006 2.4.4 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 1/1 RO 295025 2.4.20 any warnings, cautions, or notes. A new KIA was randomly selected.
295032 EK1.04 - Unable to develop three credible or 1/2 RO 295032 EK1.03 plausible distracters. A new KIA was randomly selected.
295021 2.4.35 - Unable to develop an operationally 1/1 SRO 295021 2.4.45 relevant question. A new KIA was randomly selected.
295030 EA2.04 - KIA rejected due to overlap with 1/1 SRO 295030 EA2.03 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 2/1 SRO 259002 A2.04 system at Oyster Creek. A new KIA was randomly selected.
261000 A2.14 - Could not develop an operationally 2/1 SRO 261000 A2.15 relevant question. A new KIA was randomly selected.
262001 A2.10 - KIA supports testing at the RO level, 2/1 SRO 212000 A2.11 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 2/2SRO 256000 A2.15 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 2/2SRO 215001 2.1.32 Specifications. KIA rejected and a new KIA was randomly selected.
700000 AK1.01 - KIA related to Generic 1/1 RO 700000 AK1.02 Fundamentals; concept tested on NRC GFE exam. A new KIA was randomly selected.
ES-301 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: ILT 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 Control 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)]
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 Codi~s & 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
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Oyster Creek Date of Examination: 7/11/11 Examination Level: RO 0 SRO [2J 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 Control 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 Control D,R Admin JPM4]
Determine Primary Containment Water Level lAW EMG-Emergency Procedures/Plan M,R SP28 and determine required action; G2.4.21 (4.6) [NRC SRO Admin JPM5]
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (s 3 for ROs; :s 4 for SROs & RO retakes)
(N)ew or (M}odified from bank ~ 1)
(P)revious 2 exams (s 1; randomly selected)
ES 301 , Page 22 of 27
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Oyster Creek Date of Examination: 7111/11 Exam level: RO 0 SRO-I D 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)
Safety System I JPM Title Type Code*
Function
- a. Perform Core Spray Surveillance with faulted Core Spray Pump lAW D,A,S 2 610.4.002, Core Spray Pump Operability Test (Alternate Path); 209001 A4.01 (3.8/3.6) [NRC Sim JPM1]
- b. Shutdown of the Automatic Depressurization System lAW 308, P, D, l, EN, 3 Emergency Core Cooling System Operation; 218000 A4.03 (4.214.2) S
[NRC Sim JPM2]
- c. Cool dovvn the RPV using the Isolation Condenser tube side vents lAW M, A, l, EN, 4 EMG-SP15, Alternate Pressure Control Systems - IC Tube Side Vents S (Alternate Path); 295021 AA1.04 (3.7/3. 7f [NRC Sim JPM3]
- d. Place the H2I02 monitoring system in service lAW EMG-SP39, PlaCing N,l,S 5 The H2I02 Monitoring System In Service; 500000 EA 1.01 (3.4/3.3)
[NRC Sim JPM4]
- e. Transfer 4160 VAC Bus 1A to the Startup Transformers (Alternate Path); D,A,S 6 262001 ~~4.02 ~2.9/~.3) [NRC Sim JPM5]
- g. Inject Fire Water via the Core Spray System lAW SP-20, low Pressure N,S 8 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 p, D, A, l, S 9 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)
- i. Vent the scram air header lAW EMG-SP21, Alternate Insertion of Control D,E,R 1 Rods; 295037 EA1.05 (3.9/4.0) [NRC Plant JPM1]
- j. Add makeup from Fire Water to the Isolation Condensers lAW 307, D, l, R 4 Isolation Condenser S~stem; 207000 K1.06 (3.3/3.7) [NRC Plant ..IPM3]
- k. Operate Service Water Pump 1-2 from Local Shutdown Panel 1B3 (LSP D, EN, S* 7 1B3) 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.
@ 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 / 2-3 (C)ontrol room (D)irect from bank ~ 9/ ~8 / ~4 (E)mergency or abnormal in-plant ~ 1/ ~1 / ~ 1 (EN)gineered safety feature - / - / ~ 1 (control room system (L)ow-Power / Shutdown ~ 1/ ~1 / ~1 (N)ew or (M)odified from bank including 1(A) ~2/ ~2 / ~1 (P)revious 2 exams ~3/ ~3 / ~ 2 (randomly selected)
(R)CA ~ 1/ ~1 / ~ 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: 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)
Safety System 1JPM Title Type Code*
Function a, Perform Core Spray Surveillance with faulted Core Spray Pump lAW D,A,S 2 610.4.002, Core Spray Pump Operability Test (Alternate Path); 209001 A4.01 (3.8/3.6) [NRC Sim JPM1l
- b. Shutdown of the Automatic Depressurization System lAW 308, P, D, L, EN, 3 Emergency Core Cooling System Operation; 218000 A4.03 (4.214.2) S
[NRC Sirn ...IPM2]
- c. Cool down the RPV using the Isolation Condenser tube side vents lAW M, A, L, EN, 4 EMG-SP'15, Alternate Pressure Control Systems - IC Tube Side Vents S (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 N, L, S 5 The H2/02 Monitoring System In Service; 500000 EA1,01 (3.4/3.3)
[NRC Sirn JPM4]
- e. Transfer 4160 VAC Bus 1A to the Startup Transformers (Alternate Path); D,A,S 6 262001 K4,02 (2.9/3.3) [NRC Sirn JPM5]
g.
- h. Startup of the Turbine Building Ventilation System lAW 328, The Turbine P, D, A, L, S 9 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)
- i. Vent the scram air header lAW EMG-SP21, Alternate Insertion of Control D, R 1 Rods; 295037 EA1.05 (3,9/4.0) [NRC Plant JPM1]
j, Add makeup from Fire Water to the Isolation Condensers lAW 307, D, L, R 4 Isolation Condenser System; 207000 K1.06 (3.3/3.7) [NRC Plant JPM3]
- k. Operate Service Water Pump 1-2 from Local Shutdown Panel 183 (LSP D, EN, S* 7 1B3) 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.
@ 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
ILT 10-1 NRC Scenario 2 (NEW)
Scenario Outline Facility: Oyster Creek Scenario No.: " Op Test No.: 10-1 NRC Examiners: 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 I Event No. Malt. N -- Event Description 1 NA N BOP 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 to 3 TS SRO NSS026A Isolation Condenser Initiation Logic Failing High. !
MAL- I ATC 4 NIS0218 Respond to APRM 2 failing INOP.
TS SRO MAL-5 MSSOO5A C BOP Respond to trip of Steam Packing Exhauster 1.
C ATC Respond to the E EMRV lifting leading the crew to a MAL-6 BOP NSS025E TS manual scram.
SRO 7 CAEP ATWS M Crew Respond to an Electric ATWS.
PMP SLCOO1A 8 C Crew Respond to a Standby Liquid Control Pump shaft break.
PMP SLC002A
.. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs ILT 10-1 NRC Scenario 2 Page 1 of 23
ILT 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 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 10-1 NRC Scenario 2 Page 2 of 23
ILT 10-1 NRC Scenario 2 (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 3,4,5,6,7,8
- 2. Malfunctions after EOP entry (1-2) 1 8
- 3. Abnormal events (2-4) 2 4,6
- 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 10-1 NRC Scenario 2 Page 30f23
ILT 10-1 NRC Scenario 3 (NEW)
Scenario Outline Facility: Oyster Creek Scenario No.: ~ Op Test No.: 10-1 NRC Examiners: Operators:
Initial Conditions:
- 75% power
- TBCCW Pump 2 is tagged out of service Turnover:~
- 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 ATC 3 CRDOO2 Respond to a CRD Pump A trip TS SRO MAL 4 TCS010 I BOP Respond to the EPR setpoint failing low SWI C 5 TBS027C BOP Respond to a trip of Control Room Vent Fan B ANN-L4f TS SRO 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 BKR 7 CRDOO1 M Crew Respond to a loss of all CRD Flow MAL 8 NSS016A M Crew Respond to a Safety Valve lifting post scram MAL- Respond to a trip of the operating Containment Spray 9 CNSOO4A- C Crew 0
Pump
- ("'I)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Transient, (TS) Tech Specs ILT 10-1 NRC Scenario 3 Page 1 of 28
ILT 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
ILT 10-1 NRC Scenario 3 (NEW) 9 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)
Critical The Crew must manually scram the reactor following a loss of all CRD Task 1 pumps. With no CRD flow at high power, damage to the CRD drive mechanisms will occur potentially inhibiting the ability to successfully scram.
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 drywell lAW SP-29, Initiation of the Containment Spray System for Drywell Sprays.
ES..301-4 Target Quantitative Actual Event Attributes Attributes Numbetls)
- 1. Total malfunctions (5..8) 7 3,4,5,6, 7, 8,9
- 2. Malfunctions after EOP entry (1-2) 2 8,9
- 3. Abnormal events (2-4) 2 4,6
- 4. Major transients (1-2) 2 7,8
- 5. EOPs entered/requiring substantive 1 8 actions (1-2)
- 6. EOP contingencies requiring substantive 0 N/A actions (0-2)
- 7. Critical tasks (2-3) 2 7,9 ILT 10-1 NRC Scenario 3 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 Turnover:~
- No evolutions are planned during this shift
-- N... 1UI"lf. No. EventType* Event Description MAL-I CRD001A C ATC Respond to CRD Flow Control Valve failed dosed.
ICH- C BOP 2 ICSOO1A Respond to a leak in Isolation Condenser Shell A.
TS SRO IN[)' R ATC Condensate Pump A experiences high amps requiring a 3 18 rapid power reduction and securing of Condensate Pump.
-2B C BOP MAL RCPOO3D C BOP 4 Respond to Recirculation Pump 0 total seal failure.
MAL- TS SRO RCPOO4D
- C Respond to multiple drifting control rods.
ATC MAL- Respond to a Torus Leak requiring entry into Primary 6 PCNOO7 M Crew 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 PCNOO7 M Crew 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 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)
ILT 10-1 NRC Scenario 4 Page 2 of 24
IlT 10-1 NRC Scenario 4 (NEW) 8 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 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 configuration.
Critical The crew will Emergency Depressurize the RPV prior to Torus level Task 2 reaching 110 inches.
ES-301-4 Target Quantitative Actual Event Attributes Attributes Numbe~s}
- 1. Totall malfunctions (5-8) 7 1, 2, 3, 4, 5, 6, 7
- 2. Malfunctions after EOP entry (1-2) 2 7,8
- 3. Abnormal events (2-4) 3 1,4,5
- 4. Major transients (1-2) 2 6,8
- 5. EOPs entered/requiring substantive 2 5,6 actions (1-2)
- 6. EOP contingencies requiring substantive 1 8 actions (0-2)
- 7. Critical tasks (2-3) 2 5,8 IlT 10-1 NRC Scenario 4 Page 30f24