ML111860014
| ML111860014 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/28/2011 |
| From: | Kelly Clayton Operations Branch IV |
| To: | Nebraska Public Power District (NPPD) |
| References | |
| 50-298/11-006 | |
| Download: ML111860014 (44) | |
Text
ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: 2-28-2011 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4 *
- 1. 1 4 3 3 3 4 3 20 2 5 7 Emergency &
Abnormal Plant 2 1 1 1 N/A 1 1 N/A 2 7 2 1 3 Evolutions Tier Totals 5 4 4 4 5 5 27 4 6 10 1 2 2 2 3 2 2 2 3 3 3 2 26 2 3 5 2.
Plant 2 1 1 2 1 1 1 1 1 1 1 1 12 1 1 1 3 Systems Tier Totals 3 3 4 4 3 3 3 4 4 4 3 38 4 4 8
- 3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 Categories 3 2 2 3 10 2 2 1 2 7 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.
The final RO exam must total 75 points and the 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/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (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 K/A categories.
7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the 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 K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 RO E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #
1 2 3 1 2 AA2.05 Ability to determine and/or interpret Jet 295001 Partial or Complete Loss of Forced pump operability as they apply to PARTIAL OR Core Flow Circulation / 1 & 4 X COMPLETE LOSS OF FORCED CORE FLOW 3.1 53 CIRCULATION:
AK3.06 Knowledge of the reasons for Containment 295003 Partial or Complete Loss of AC / 6 isolation as they apply to PARTIAL OR COMPLETE X 3.7 46 LOSS OF A.C.POWER 2.1.27 Knowledge of system purpose and/or 295004 Partial or Total Loss of DC Pwr / 6 X function. 3.9 56 2.1.26 Knowledge of industrial safety procedures 295004 Partial or Total Loss of DC Pwr / 6 (such as rotating equipment, electrical, high X temperature, high pressure, caustic, chlorine, 3.4 56 oxygen and hydrogen).
2.4.4 Ability to recognize abnormal indications for 295004 Partial or Total Loss of DC Pwr / 6 system operating parameters that are entry-level X conditions for emergency and abnormal operating 4.5 56 procedures.
AK2.01 Knowledge of the interrelations between 295005 Main Turbine Generator Trip / 3 X MAIN TURBINE GENERATOR TRIP and RPS 3.8 43 2.4.45 Ability to prioritize and interpret the 295006 SCRAM / 1 X significance of each annunciator or alarm. 4.1 54 AA1.02 Ability to operate and/or monitor the 295016 Control Room Abandonment / 7 reactor/turbine pressure regulating system as they X 2.9 49 apply to CONTROL ROOM ABANDONMENT:
AK2.01 Knowledge of the interrelations between 295018 Partial or Total Loss of CCW / 8 PARTIAL OR COMPLETE LOSS OF COMPONENT X 3.3 42 COOLING WATER and system loads AK2.01 Knowledge of the interrelations between 295019 Partial or Total Loss of Inst. Air / 8 PARTIAL OR COMPLETE LOSS OF INSTRUMENT X 3.8 44 AIR and CRD hydraulics AK3.03 Knowledge of the reasons for increasing 295021 Loss of Shutdown Cooling / 4 drywell cooling as they apply to LOSS OF X 2.9 45 SHUTDOWN COOLING AK3.01 Knowledge of the reasons for the following 295021 Loss of Shutdown Cooling / 4 responses as they apply to LOSS OF SHUTDOWN X 3.3 45 COOLING: Raising reactor water level AK1.03 Knowledge of the operational implications 295023 Refueling Acc / 8 of inadvertent criticality as they apply to X 3.7 41 REFUELING ACCIDENTS EK1.01 Knowledge of the operational implications of 295024 High Drywell Pressure / 5 drywell integrity as they apply to HIGH DRYWELL X 4.1 40 PRESSURE EA2.03 Ability to determine and/or interpret 295025 High Reactor Pressure / 3 Suppression pool temperature as they apply to X 3.9 57 HIGH REACTOR PRESSURE:
EA2.02 Ability to determine and/or interpret Reactor 295025 High Reactor Pressure / 3 power as they apply to HIGH REACTOR X 4.2 57 PRESSURE:
EA2.02 Ability to determine and/or interpret 295026 Suppression Pool High Water Temp. Suppression pool level as they apply to
/5 X 3.8 51 SUPPRESSION POOL HI WTR TEMPERATURE:
295027 High Containment Temperature / 5 EK1.01 Knowledge of the operational implications of 295028 High Drywell Temperature / 5 the reactor water level measurement as they apply X 3.5 39 to HIGH DRYWELL TEMPERATURE EA1.05 Ability to operate and/or monitor HPCI as 295030 Low Suppression Pool Wtr Lvl / 5 they apply to LOW SUPPRESSION POOL WATER X 3.5 50 LEVEL:
EK1.03 Knowledge of the operational implications of 295031 Reactor Low Water Level / 2 water level effects on reactor power as they apply to X 3.7 58 REACTOR LOW WATER LEVEL:
EK1.01 Knowledge of the operational implications of 295031 Reactor Low Water Level / 2 Adequate Core Cooling as they apply to REACTOR X 4.6 58 LOW WATER LEVEL
EK3.03 Knowledge of the reasons for lowering 295037 SCRAM Condition Present reactor water level as they apply to SCRAM and Reactor Power Above APRM CONDITION PRESENT AND REACTOR Downscale or Unknown / 1 X 4.1 47 POWER ABOVE APRM DOWNSCALE OR UNKNOWN:
2.2.36 Ability to analyze the effect of maintenance 295038 High Off-site Release Rate / 9 activities, such as degraded power sources, on the X 3.1 55 status of limiting conditions for operations.
2.2.44 Ability to interpret control room indications to 295038 High Off-site Release Rate / 9 verify the status and operation of a system, and X understand how operator actions and directives 4.2 55 affect plant and system conditions.
AA2.05 Ability to determine and interpret Ventilation 600000 Plant Fire On Site / 8 alignment necessary to secure affected area as they X 2.9 52 apply to PLANT FIRE ON SITE:
AA1.01 Ability to operate and/or monitor Grid 700000 Generator Voltage and Electric Grid frequency and voltage as they apply to Disturbances / 6 X GENERATOR VOLTAGE AND ELECTRIC GRID 3.6 48 DISTURBANCES:
K/A Category Totals: 4 3 3 3 4 3 Group Point Total: 20
ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 RO E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #
1 2 3 1 2 AK2.08 Knowledge of the interrelations between LOSS 295002 Loss of Main Condenser Vac / 3 OF MAIN CONDENSER VACUUM and the Condenser X 3.1 60 circulating water system 295007 High Reactor Pressure / 3 295008.AK1.03 Knowledge of the reasons for the 295008 High Reactor Water Level / 2 X following responses as they apply to HIGH REACTOR 3.4 59 WATER LEVEL: Main turbine trip 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 AK3.01 Knowledge of the reasons for bypassing rod 295015 Incomplete SCRAM / 1 insertion blocks as they apply to INCOMPLETE X 3.4 61 SCRAM 295017 High Off-site Release Rate / 9 AK1.04 Knowledge of the operational implications of 295020 Inadvertent Cont. Isolation / 5 & 7 Bottom head thermal stratification as they apply to X 2.5 59 INADVERTENT CONTAINMENT ISOLATION:
AA2.01 Ability to determine and/or interpret 295022 Loss of CRD Pumps / 1 accumulator pressure as they apply to LOSS OF CRD X 3.5 63 PUMPS:
2.4.1 Knowledge of EOP entry conditions and 295029 High Suppression Pool Wtr Lvl / 5 X immediate action steps. 4.6 65 EA1.05 Ability to operate and/or monitor affected 295032 High Secondary Containment systems so as to isolate damaged portions as they Area Temperature / 5 X apply to HIGH SECONDARY CONTAINMENT AREA 3.7 62 TEMPERATURE:
2.4.9 Knowledge of low power/shutdown implications 295033 High Secondary Containment in accident (e.g., loss of coolant accident or loss of Area Radiation Levels / 9 X 3.8 64 residual heat removal) mitigation strategies.
295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Level / 5 500000 High CTMT Hydrogen Conc. / 5 K/A Category Point Totals: 1 1 1 1 1 2 Group Point Total: 7
ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 K5.02 Knowledge of the operational 203000 RHR/LPCI: Injection implications of core cooling methods as they Mode X 3.5 9 apply to RHR/LPCI INJECTION MODE.
K6.04 Knowledge of the effect that a loss or malfunction of reactor water level will have on 205000 Shutdown Cooling X the SHUTDOWN COOLING SYSTEM (RHR 3.6 11 SHUTDOWN COOLING MODE):
A1.08 Ability to predict and/or monitor 206000 HPCI changes in parameters associated with X operating the HIGH PRESSURE 4.1 14 COOLANT INJECTION SYSTEM controls including system lineup 207000 Isolation (Emergency)
Condenser 207000 Isolation (Emergency)
Condenser K4.01 Knowledge of LOW PRESSURE 209001 LPCS CORE SPRAY SYSTEM design feature(s)
X and/or interlocks which provide for prevention 3.2 7 of over pressurization of core spray piping 209002 HPCS A4.07 Ability to manually operate and/or 211000 SLC X monitor in the control room: Lights and alarms 3.6 23 A2.08 Ability to (a) predict the impacts of Low reactor level on the REACTOR PROTECTION SYSTEM ; and (b) based on 212000 RPS X those predictions, use procedures to correct, 4.1 15 control, or mitigate the consequences of those abnormal conditions or operations:
K5.01 Knowledge of the operational 215003 IRM implications of detector operation as they X apply to INTERMEDIATE RANGE MONITOR 2.6 10 (IRM) SYSTEM:
K2.01 Knowledge of electrical power supplies 215004 Source Range Monitor X to the SRM channels/detectors 2.6 4 K3.07 Knowledge of the effect that a loss or malfunction of the AVERAGE POWER 215005 APRM / LPRM X RANGE MONITOR/LOCAL POWER RANGE 3.2 5 MONITOR SYSTEM will have on the rod block monitor:
A3.02 Ability to monitor automatic operations of the AVERAGE POWER 215005 APRM / LPRM X RANGE MONITOR/LOCAL POWER 3.5 17 RANGE MONITOR SYSTEM including:
Full core display K3.02 Knowledge of the effect that a loss or 217000 RCIC malfunction of the REACTOR CORE X ISOLATION COOLING SYSTEM (RCIC) will 3.6 6 have on Reactor vessel pressure A3.04 Ability to monitor automatic operations 217000 RCIC of the REACTOR CORE ISOL COOLING X 3.6 18 SYSTEM (RCIC) including, system flow.
2.1.19 Ability to use plant computers to 218000 ADS X evaluate system or component status. 3.9 22 A4.02 Ability to manually operate and/or 223002 PCIS/Nuclear Steam monitor in the control room: Manually initiate Supply Shutoff X 3.9 19 the system 2.4.21 Knowledge of the parameters and logic 239002 SRVs used to assess the status of safety functions, such as reactivity control, core cooling and X heat removal, reactor coolant system 4.0 21 integrity, containment conditions, radioactivity release control, etc.
ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 2.4.45 Ability to prioritize and interpret 239002 SRVs X the significance of each annunciator or 4.1 21 alarm.
A2.03 Ability to (a) predict the impacts of 239002 SRVs Stuck open SRV on the RELIEF/SAFETY VALVES; and (b) based on those predictions, X use procedures to correct, control, or mitigate 4.1 26 the consequences of those abnormal conditions or operations:
A4.07 Ability to manually operate and/or 259002 Reactor Water Level monitor in the control room: All individual Control X component controllers when transferring from 3.8 20 automatic to manual mode K1.12 Knowledge of the physical connections 261000 SGTS and/or cause/effect relationships between X STANDBY GAS TREATMENT SYSTEM and 3.1 2 the primary containment purge system:
A2.04 Ability to (a) predict the impacts of 261000 SGTS High train moisture content on the STANDBY GAS TREATMENT SYSTEM; and (b) based X on those predictions, use procedures to 2.5 16 correct, control, or mitigate the consequences of those abnormal conditions or operations:
262001 AC Electrical K2.01 Knowledge of electrical power supplies X to the following: Off-site sources of power 3.3 3 Distribution K4.03 Knowledge of A.C. ELECTRICAL 262001 AC Electrical DISTRIBUTION design feature(s) and/or Distribution X interlocks which provide for the interlocks 3.1 25 between automatic bus transfer and breakers K4.01 Knowledge of UNINTERRUPTABLE 262002 UPS (AC/DC) POWER SUPPLY (A.C./D.C.) design X feature(s) and/or interlocks which provide for 3.1 8 transfer from preferred power to alternate power supplies.
K1.03 Knowledge of the physical connections 263000 DC Electrical and/or cause/effect relationships between Distribution X D.C. ELECTRICAL DISTRIBUTION and 2.6 1 battery ventilation:
A1.04 Ability to predict and/or monitor 264000 EDGs changes in parameters associated with X operating the EMERGENCY GENERATORS 2.6 13 (DIESEL/JET) controls including crank case temperature and pressure A3.02 Ability to monitor automatic operations 300000 Instrument Air of the INSTRUMENT AIR SYSTEM including X 2.9 17 air temperature K6. 07 Knowledge of the effect that a loss or 400000 Component Cooling malfunction of Breakers, relays, and Water X 2.7 12 disconnects will have on the CCWS A3.01 Ability to monitor automatic operations 400000 Component Cooling of the CCWS including: Setpoints on Water instrument signal levels for normal operations, X 3.0 24 warnings, and trips that are applicable to the CCWS A2.04 400000 Component Cooling X 2.9 16 Water K/A Category Point Totals: 2 2 2 3 2 2 2 3 3 3 2 Group Point Total: 26
ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS K4.08 Knowledge of CONTROL ROD 201003 Control Rod and Drive 2.6 30 Mechanism AND DRIVE MECHANISM design X feature(s) and/or interlocks which provide for the following: Monitoring CRD mechanism temperature 201004 RSCS 201005 RCIS 201006 RWM K2.02 Knowledge of electrical power 202001 Recirculation 3.2 28 X supplies to the following: Plant Specific -
MG Set K2.02 Knowledge of electrical power 202002 Recirculation Flow Control 2.6 28 X supplies to the following: Hydraulic power unit:
204000 RWCU K6.01 Knowledge of the effect that a loss 214000 RPIS 2.5 32 or malfunction of the A.C. electrical X power will have on the ROD POSITION INFORMATION SYSTEM:
A2.07 Ability to (a) predict the impacts of 215001 Traversing In-core Probe 3.4 34 failure to retract during accident conditions on the TRAVERSING IN-X CORE PROBE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations 215002 RBM K3.29 Knowledge of the effect that a loss 216000 Nuclear Boiler Inst. 3.1 38 or malfunction of the NUCLEAR BOILER X Instrumentation will have on Jet pump flow monitoring:
K3.01 Knowledge of the effect that a loss or malfunction of the RHR/LPCI: TORUS 219000 RHR/LPCI: Torus/Pool Cooling / SUPPRESSION POOL COOLING X 3.9 29 Mode MODE will have on Suppression pool temperature control 223001 Primary CTMT and Aux.
226001 RHR/LPCI: CTMT Spray Mode K1.01 Knowledge of the physical 230000 RHR/LPCI: Torus/Pool Spray 3.6 27 Mode connections and/or cause/effect relationships between RHR/LPCI:
X TORUS/SUPPRESSION POOL SPRAY MODE and the following: Suppression pool K5.07 Knowledge of the operational 233000 Fuel Pool Cooling/Cleanup 2.5 31 implications of the Maximum (abnormal)
X heat 102d load as they apply to FUEL POOL COOLING AND CLEAN-UP:
234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control A1.06 Ability to predict and/or monitor 241000 Reactor/Turbine Pressure 3.2 33 Regulator changes in parameters associated with operating the REACTOR/TURBINE X PRESSURE REGULATING SYSTEM controls including: Main turbine steam flow 3.2 245000 Main Turbine Gen. / Aux.
ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 A3.05 Ability to monitor automatic 256000 Reactor Condensate 3.0 35 operations of the REACTOR X CONDENSATE SYSTEM including:
Lights and alarms 259001 Reactor Feedwater 268000 Radwaste 271000 Offgas 272000 Radiation Monitoring 2.4.50 Ability to verify alarm setpoints 4.2 37 X and operate controls identified in the alarm response manual.
286000 Fire Protection 288000 Plant Ventilation A4.04 Ability to manually operate and/or 290001 Secondary CTMT 2.6 36 X monitor in the control room: Auxiliary building area temperature:
A4.01 Ability to manually operate and/or 290001 Secondary CTMT 3.3 36 monitor in the control room: Reactor X building differential pressure: Plant-Specific 290003 Control Room HVAC 2.2.22 Knowledge of limiting conditions 4.0 37 X for operations and safety limits. (Note 1)
ES-401-4 form 290002 Reactor Vessel Internals K/A Category Point Totals: 1 1 2 1 1 1 1 1 1 1 1 Group Point Total: 12
S-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 SRO E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #
1 2 3 1 2 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 295003 Partial or Complete Loss of AC / 6 2.2.25 Knowledge of the bases in Technical 295004 Partial or Total Loss of DC Pwr / 6 X Specifications for limiting conditions for operations 4.2 76 and safety limits.
AA2.01 Ability to determine and/or interpret the 295004 Partial or Total Loss of DC Pwr / 6 X following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Cause of partial or 3.6 76 complete loss of D.C. power.
295005 Main Turbine Generator Trip / 3 295006 SCRAM / 1 295016 Control Room Abandonment / 7 AA2.03 Ability to determine and/or interpret the 295018 Partial or Total Loss of CCW / 8 X Cause for partial or complete loss as it apply to PARTIAL OR COMPLETE LOSS OF COMPONENT 3.5 78 COOLING WATER.
295019 Partial or Total Loss of Inst. Air / 8 295021 Loss of Shutdown Cooling / 4 AA2.05 Ability to determine and/or interpret the 295023 Refueling Acc / 8 entry conditions of emergency plan as they apply to X 4.6 77 REFUELING ACCIDENTS.
2.4.6 Knowledge of EOP mitigation strategies.
295024 High Drywell Pressure / 5 X 4.7 80 EA2.03 Ability to determine and/or interpret 295025 High Reactor Pressure / 3 suppression pool temperature as they apply to X 4.1 78 HIGH REACTOR PRESSURE:
EA2.03 Ability to determine and/or interpret the 295026 Suppression Pool High Water Temp. Reactor Pressure as they apply to SUPPRESSION
/5 X 4.0 76 POOL HIGH WATER TEMPERATURE 2.1.27 Knowledge of system purpose and/or 295027 High Containment Temperature / 5 X function 4.0 81 2.1.39 Knowledge of conservative decision making 295027 High Containment Temperature / 5 X practices. 4.3 81 2.4.9 - Knowledge of low power/shutdown 295032 High Secondary Containment Area implications in accident (e.g., loss of coolant Temperature / 5 X accident or loss of residual heat removal) mitigation 4.2 81 strategies.
295028 High Drywell Temperature / 5 295030 Low Suppression Pool Wtr Lvl / 5 2.4.3 Ability to identify post-accident 295031 Reactor Low Water Level / 2 X instrumentation. 3.9 82 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 2.2.42 Ability to recognize system parameters that 295038 High Off-site Release Rate / 9 are entry-level conditions for Technical X 4.6 79 Specifications.
600000 Plant Fire On Site / 8 700000 Generator Voltage and Electric Grid Disturbances / 6
K/A Category Totals: 2 5 Group Point Total: 7 ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 SRO E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #
1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 295015 Incomplete SCRAM / 1 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation / 5 & 7 2.2.22 Knowledge of limiting conditions for operations X 4.7 84 and safety limits.
295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Wtr Lvl / 5 EA2.01 Ability to determine and/or interpret X Suppression pool water level as they apply to HIGH 3.9 85 SUPPRESSION POOL WATER LEVEL. 3.9 295032 High Secondary Containment Area Temperature / 5 EA2.01 Ability to determine and/or interpret Area 295033 High Secondary Containment radiation levels as they apply to HIGH SECONDARY Area Radiation Levels / 9 X 3.9 83 CONTAINMENT AREA RADIATION LEVELS.
295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Level / 5 500000 High CTMT Hydrogen Conc. / 5 K/A Category Point Totals: 2 1 3
ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 SRO System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 203000 RHR/LPCI: Injection Mode A2.09 Ability to (a) predict the impacts of Reactor low water level on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN 205000 Shutdown Cooling X COOLING MODE); and (b) based on those 3.8 86 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
206000 HPCI 207000 Isolation (Emergency)
Condenser A2.09 A2. Ability to (a) predict the impacts of 209001 LPCS low suppression pool level on the LOW PRESSURE CORE SPRAY SYSTEM ; and X (b) based on those predictions, use 3.3 86 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations.
209002 HPCS 211000 SLC A2.19 Ability to (a) predict the impacts of Partial system activation (half-SCRAM) on the REACTOR PROTECTION SYSTEM ; and (b) 212000 RPS X based on those predictions, use procedures 3.9 87 to correct, control, or mitigate the consequences of those abnormal conditions or operations.
215003 IRM 215004 Source Range Monitor 2.4.20 Knowledge of the operational 215005 APRM / LPRM X implications of EOP warnings, cautions, and 4.3 90 notes.
217000 RCIC 218000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff 239002 SRVs 259002 Reactor Water Level Control 261000 SGTS 262001 AC Electrical 2.2.7 Knowledge of the process for X conducting special or infrequent tests. 3.6 89 Distribution 262001 AC Electrical 2.2.40 Ability to apply Technical X Specifications for a system. 4.7 89 Distribution 262002 UPS (AC/DC) 263000 DC Electrical Distribution
2.2.36 Ability to analyze the effect of 264000 EDGs maintenance activities, such as degraded X power sources, on the status of limiting 4.2 88 conditions for operations.
K/A Category Point Totals: 2 3 Group Point Total: 5
ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 SRO System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-core Probe 215002 RBM 216000 Nuclear Boiler Inst.
A2.16 Ability to (a) predict the impacts of high suppression pool level on the RHR/LPCI: TORUS/SUPPRESSION 219000 RHR/LPCI: Torus/Pool Cooling POOL COOLING MODE; and (b) based X on those predictions, use procedures to 3.2 93 Mode correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.07 Ability to (a) predict the impacts of High drywell pressure on the PRIMARY CONTAINMENT SYSTEM AND 223001 Primary CTMT and Aux. X AUXILIARIES; and (b) based on those 4.3 91 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Regulator 245000 Main Turbine Gen. / Aux.
256000 Reactor Condensate 259001 Reactor Feedwater 2.1.23 Ability to perform specific system 268000 Radwaste X and integrated plant procedures during 4.4 92 all modes of plant operation. 4.4 271000 Offgas 272000 Radiation Monitoring 286000 Fire Protection 288000 Plant Ventilation 290001 Secondary CTMT
ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 SRO System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 K/A Category Point Totals: 2 1 Group Point Total: 3
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam: 2-28-2011 Category K/A # Topic RO SRO-Only IR # IR #
2.1.5 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.
2.9 66
- 1. 2.1.17 Ability to make accurate, clear, and concise verbal 2.1.17 reports. 3.9 67 Conduct of Operations 2.1.27 2.1.27 Knowledge of system purpose and/or function. 3.9 75 Knowledge of individual licensed operator responsibilities related 2.1.4 to shift staffing, such as medical requirements, no-solo 3.8 94 operation, maintenance of active license status, 10CFR55, etc.
2.1.34 Knowledge of primary and secondary plant chemistry limits. 3.5 99 Subtotal 3 2 2.2.14 Knowledge of the process for controlling equipment 2.2.14 configuration or status. 3.9 69 Knowledge of the process for controlling equipment configuration 2.2.14 or status.
- 2. 3.9 68 Equipment Control Ability to perform pre-startup procedures for the facility, including 2.2.1 operating those controls associated with plant equipment that 4.4 95 could affect reactivity.
2.2.43 Knowledge of the process used to track inoperable alarms. 3.3 98 Subtotal 2 2 2.3.11 Ability to control radiation releases.
2.3.11 3.8 70 2.3.4 Knowledge of radiation exposure limits under normal or 2.3.4 emergency conditions. 3.2 71
- 3. Knowledge of radiological safety procedures pertaining to Radiation licensed operator duties, such as response to radiation monitor Control 2.3.13 alarms, containment entry requirements, fuel handling 3.8 96 responsibilities, access to locked high-radiation areas, aligning filters, etc.
Subtotal 2 1 2.4.43 Knowledge of emergency communications systems and 2.4.43 techniques. 3.2 72 2.4.22 Knowledge of the bases for prioritizing safety functions
- 4. 2.4.22 during abnormal/emergency operations. 3.6 73 Emergency 2.4.32 Knowledge of operator response to loss of all Procedures / 2.4.32 annunciators. 3.6 74 Plan Knowledge of procedures relating to a security event (non-2.4.28 safeguards information).
4.1 97 Knowledge of the organization of the operating procedures 2.4.5 network for normal, abnormal, and emergency evolutions. 4.3 100 Subtotal 3 2 Tier 3 Point Total 10 7
ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 2/1 207000.A3.02 Cooper does not have an isolation condenser 2/1 207000.K3.02 Cooper does not have an isolation condenser 2/1 SRO 207000 Cooper does not have an isolation condenser 2/1 209002.A2.08 Cooper does not have a high pressure core spray system 2/1 SRO 209002 Cooper does not have a high pressure core spray system 2/2 SRO 201004 Cooper no longer uses a Rod Sequence Control System 1/1 295027.EK1.02 Cooper has a Mark I containment design not a Mark III.
This Generic K/A deals with the effects of maintenance on the 295038 Generic 1/1 status of LCOs - A psychometrically valid question could not 2.2.36 be developed based on 10CFR55.41.
3 2.2.4 Cooper is not a multi-unit facility Replaced Generic K/A with a randomly selected A2 K/A to fill 2/2 2.4.45 all blocks on form 401-1. Note 1 Could not come up with a psychometrically valid question for 2/1 400000.A2.04 this K/A, Also CCW was sampled three times in this group.
Randomly selected another system.
Unable to develop a psychometrically valid question for this 1/1 295004 G 2.1.27 K/A. Randomly selected another generic K/A.
Unable to develop a psychometrically valid question for this 1/1 295021.AK3.03 K/A. Cooper has no actions for ventilation on a loss of SDC.
Randomly selected another K/A.
Unable to develop a psychometrically valid question for this 1/1 295025.EA2.03 K/A. Randomly selected another K/A.
Unable to develop a psychometrically valid question for this 1/1 295031.EK1.03 K/A. This question and question 47 were too similar.
Unable to develop a psychometrically valid question for this 1/2 295020.AK1.04 K/A. Randomly selected another system.
Unable to develop a psychometrically valid question for this 2/1 239002. G 2.4.21 K/A. Randomly selected another K/A.
Unable to develop a psychometrically valid question for this 2/1 300000.A3.02 K/A. Randomly selected another system.
Unable to develop a psychometrically valid question for this K/A. Cooper does not have any buildings named or 2/2 290001.A4.04 designated as Auxiliary Building. Randomly selected another K/A.
This topic is for a BWR 5/6, Replaced it with Equivalent BWR 2/2 202002.K2.02 3/4 topic 202001.K2.02 Unable to develop a psychometrically valid question for this 1/1 SRO 295025.EA2.03 K/A. Too similar to another RO question. Randomly selected another K/A.
Unable to develop a psychometrically valid question for this 1/1 SRO 295026.EA2.03 K/A. Too similar to another RO question. Randomly selected another system.
Unable to develop a psychometrically valid question for this 1/1 SRO 295027. G 2.1.27 K/A. Randomly selected another generic K/A.
Not on the approved list of K/As in NUREG 1021 for Generics for Tier 1 Group 1. Randomly selected and rejected again 1/1 RO 295004 G2.1.27 Generic K/A 2.2.26, as a psychometrically valid question could not be written. Randomly reselected the set again and got 2.4.4 Cooper is a Mark I Design and this K/A is for a Mark III.
1/1 SRO 295027 Randomly selected another topic 295032 and randomly selected Generic 2.4.9 Not on list of approved K/As for Tier, Randomly selected 2/1 SRO 2.2.7 2.2.40.
Double Jeopardy with RO question, Randomly reselected 2/1 SRO 205000.A2.09 209001 as the Topic and maintained the same K/A A2.09.
Unable to write a psychometrically valid question for this K/A.
3/1 SRO G 2.1.34 Randomly selected G 2.1.34 Could not develop a generic RO question concerning 3/2 2.2.36 maintenance affects on LCOs. Randomly selected K/A 2.2.14.
Could not develop a question for the system and the generic K/A selected. Randomly selected one of the open tier 2 2/2 2.2.22 systems and Randomly selected a new generic K/A. The system selected was 272000 Radiation Monitoring, The generic K/A was 2.4.50.
Could not develop an SRO question using this K/A. Talked 1/1 SRO 295004.AA2.01 with NRC Lead Examiner and we selected a generic K/A to replace the old K/A. The Generic K/A selected was 2.2.25
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 06/06/11 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*
Document the Maintenance of your active Reactor Conduct of Operations N,R Operator License Conduct of Operations D,C Perform RO (SRO) Review of Daily Logs Equipment Control N,R Initiate an Operator Aid request.
Radiation Control D,R Determine Dosage On Workers For ALARA Emergency Procedure/Plan N/A NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, then all 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs; < 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (> 1)
(P)revious 2 exams (< 1; randomly selected)
ES-301, Page 22 of 27
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 06/06/2011 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*
Conduct of Operations N,R Determine If A Working Hour Waiver Is Required Conduct of Operations N,R Determine if Mode Change is Allowed Determine Refueling LCO for CRD Equipment Control D,R Drive Removal Radiation Control D,R Determine Dosage On Workers For ALARA Emergency Procedure/Plan D,R Reportable Occurrence to NRC NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, then all 5 are required.
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs; < 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (> 1)
(P)revious 2 exams (< 1; randomly selected)
ES-301, Page 22 of 27
ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 6/06/11 Exam Level: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
Safety System / JPM Title Type Code*
Function
- b. SKL034-20-XXX Withdraw SRMs during a Start-up (Alt A,N,L,S 7 Path)
- c. SKL034-20-XXX Shift REC Heat Exchangers N,S 8
- d. SKL034-21-47 Operate RCIC in Press Control (Alt Path) A,D,S 3
- e. SKL034-21-06 Respond to a HPCI System Automatic A,D,En,S 4 Initiation.(Alt Path)
- f. SKL034-20-XX Monitor SGT System Following A,N,S 9 Automatic Initiation (Alt Path)
- g. SKL034-20-XXX Placing SDG In Service From N, S 6 Control Room
- h. SKL034-21-XXX Vent Primary Containment using 2.4PC N, S 5 In-Plant Systems @ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. SKL034-XX-XX Reboot OPC A Computer Server N 7
- j. SKL034-10-XX Transfer CPP power supply from CDP-1B N 6 (normal) to CDP-1A (emergency)
- k. SKL034-10-85 Place standby CRD Flow Control Valve in D,R 1 Service When In service Valve Fails Closed
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-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: Cooper Nuclear Station Date of Examination: 6/06/11 Exam Level: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
Safety System / JPM Title Type Code*
Function
- b. SKL034-20-XXX Withdraw SRMs during a Start-up (Alt A,N,L,S 7 Path)
- c. SKL034-20-XXX Shift REC Heat Exchangers N,S 8
- d. SKL034-21-47 Operate RCIC in Press Control (Alt Path) A,D,S 3
- e. SKL034-21-06 Respond to a HPCI System Automatic A,D,En,S 4 Initiation (Alt Path)
- f. SKL034-20-XX Monitor SGT System Following A,N,S 9 Automatic Initiation(Alt Path)
- g. SKL034-20-XXX Placing SDG In Service From N, S 6 Control Room h.
In-Plant Systems @ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. SKL034-XX-XX Reboot OPC A Computer Server N 7
- j. SKL034-10-XX Transfer CPP power supply from CDP-1B N 6 (normal) to CDP-1A (emergency)
- k. SKL034-10-85 Place standby CRD Flow Control Valve in D,R 1 Service When In service Valve Fails Closed
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-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
Appendix D Scenario Outline Form ES-D-1 IC-121 Facility: Cooper Nuclear Station Scenario No.: NRC 1 Op-Test No.: 1___
Examiners: ____________________________ Operators: _____________________________
Initial Conditions: The plant is operating at approximately 85% power in single valve control with the Main Turbine Governor Valves. It is a red light day because record grid loads are expected.
Turnover: After turnover, the crew is to place the governor valves in Sequential GV control (Load Dispatcher has been notified) and continue the power ascension to 747 MWE net ~95% at <10 MWe per minute. Maintenance will be calling to start the HPCI Aux Oil Pump for Pump PM later in shift.
Event Malf. Event Event No. No. Type* Description 1 N/A N Place MT Gov Valves into sequential valve control 2 N/A R Raise power with RR flow 3 3 C,TS HPCI Aux Oil system leak 4 4 I,TS CRD FC-301 Auto Output fails low, three CRDs with High Temps 5 5 C,TS SRV Fails Open 6 6 C, Two Rods Drift In, Manual Scram Failure, Manual ARI inserts rods 7 7 M, Earthquake; Suppression Pool Rupture: ED on Low SP/L
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Critical Tasks are located in Events 6 and 7.
1
Appendix D Required Operator Actions Form ES-D-1 Scenario Objective Evaluate the crews ability to perform normal operations and raise power during non-emergency operations.
Evaluate the crews response to a lowering Suppression Pool Level and their ability to emergency depressurize the Reactor to suppress the steam in the torus before level gets too low and its energy is released into the Reactor Building.
Scenario Summary Initial Conditions:
- The plant is operating at approximately 80% power
- No equipment out of service.
- It is a red light day because record grid loads are expected.
Events:
- Place MT Gov Valves into sequential valve control
- Raise power with RR flow
- HPCI Aux Oil System Leak, TS
- SRV Fails Open, TS
- Two Rods Drift In, Manual Scram Failure Manual ARI works
- Earthquake; Suppression Pool Rupture: ED on Low SP/L Scenario Sequence
- Normal evolution - Place MT into Sequential Gov. Valve Control
- Reactivity manipulation - Raise Power with RR Flow
- Component Failure before the EOPs - Two Rods Drift In, Manual Scram Failure Manual ARI works
- Major Event - Earthquake;
- Component Failure after the EOPs - Suppression Pool leak, below the water line:
Emergency Depressurization on low Suppression Pool Level.
2
Appendix D Required Operator Actions Form ES-D-1 Event One: Place MT Gov Valves into sequential valve control Malfunction Required:
No malfunction required; this is a normal manipulation for the BOP.
Objective:
Evaluate the crews ability to select the Digital Electro Hydraulic Controller and perform the required manipulations on the touch screen to shift the main turbine valve governor control to a Sequential alignment.
Success Path:
The Operator, in accordance with Procedure 2.2.77.1, performs the steps necessary to make the Turbine Governor Valves transition smoothly from single valve control to sequential valve control which causes turbine efficiency improvement (MWatts increase).
Event Two: Raise Power with Reactor Recirc to 737 MWE Net ~ 90%
Malfunction Required:
No malfunction required, this is a normal manipulation for the BOP.
Objective:
Evaluate the RO ability to adjust Reactor Recirculation Pump flows to make power increase from ~80% to ~90%, by alternately raising flow on one RR Pump, letting the plant respond then making a similar adjustment on the other pumps controller.
Success Path:
Both Reactor Recirculation Pumps speeds are raised from ~65 to ~80; maintaining them within the required 5% allowed by Tech Specs, and it is done in accordance with Procedure 2.1.10.
Event Three: HPCI Aux Oil System Leak Malfunction Required:
HP12 Active set at 100%, at the start of the scenario. This failure will not start until the HPCI Aux Oil Pump is started.
Objective:
Evaluate the crews response to an oil leak in the HPCI control oil system while performing the section of the Operating Procedure for the HPCI system 2.2.33 to start 3
Appendix D Required Operator Actions Form ES-D-1 the Aux Oil Pump for maintenance. During this event the Operator will be required to secure the pump.
Evaluate the CRSs ability to determine that HPCI is Inoperable.
Success Path:
The HPCI system is declared Inoperable per TS 3.5.1. Condition C 14 day LCO. RCIC is checked to be operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the Aux Oil Pump is secured to prevent operation.
Event Four: CRD FC-301 Auto Signal Failure Malfunction Required:
Override 03A35A1 CRD-FC-301 CRD System Flow Control Setpoint set at 0%.
Malfunction PMIS N200 set at 380, N213 set at 402 and N214 set at 480.
Objective:
Evaluate the crews response to CRD-FC-301 Automatic control signal failure resulting in high charging water pressure and low cooling water flow.
Evaluate the ROs ability to diagnose the controller has failed.
Evaluate the SROs ability to determine if the CRDs which have high temperatures should be declared slow or inoperable.
Success Path:
CRD-FC-301 placed in Manual and CRD system flows and pressures returned to normal. SRO addresses CRDs with high temperatures.
Event Five: SRV Fails Open Malfunction Required:
AD06b set at 50% to start the relief valve leaking then it is modified to 20% to minimize the heat addition to the torus.
Objective:
Evaluate the crews performance of Abnormal Procedure 2.4SRV for a leaking SRV which has the Operator cycle the leaking valve to reseat it.
Evaluate the CRS addressing Technical Specification 3.6.2.1 for Suppression Pool Average Temperature, if the SRV is not closed within a short period of time.
Success Path:
The Operator notices and responds to the leaking SRV in accordance with Annunciator Procedures and 2.4SRV. The SRV will be cycled open then taken back to the closed position to reseat the valve. Once the valve has been cycled the tail pipe temperature starts lowering and the heat addition into the Torus is secured. The CRS determines 4
Appendix D Required Operator Actions Form ES-D-1 that Technical Specification 3.6.2.1 for Suppression Pool Average Temperature may apply depending on whether Suppression Pool Temperature at 95F.
Event Six: Two Rods Drift In, Manual Scram Failure Manual ARI works Malfunction Required:
RD162615 CRD 26-15 Slow Scram Time RD162623 CRD 26-23 Slow Scram Time RD162615 CRD 26-15 Scrammed RD162623 CRD 26-23 Scrammed Override Reactor Manual Scram Push Button position Push Out Override Reactor Manual Scram Push Button light Off Objective:
Evaluate the crews response in accordance with Abnormal Procedure 2.4CRD and their ability to recognize Scram Actions with more than one rod is drifting in. Also to recognize that the One Channels Manual Scram Push Button failed to work and the need to manually initiate ARI.
Success Path:
The Operator determines the ATWS when a manual scram fails to work when more than one rod is drifting and initiates ARI manually.
Event Seven: Earthquake; Suppression Pool leak: ED on low SP/L Malfunction Required:
HV02b Major Earthquake set to 25%
PC08 Suppression Pool Water Leak 25% level lowers at -0.2/min Objective:
Evaluate the crews response to a major earthquake in accordance with Emergency Procedure 5.1Quake.
Evaluate the crews ability to monitor and control the consequences of an Unisolable leak in the torus below the normal water level.
Evaluate the crews ability to anticipate emergency depressurization and transfer as much energy to the condenser prior to emergency depressurizing the Reactor.
Success Path:
The crew emergency depressurizes the reactor when suppression pool level lowers to 9.6.
5
Appendix D Required Operator Actions Form ES-D-1 Scenario Termination:
When the reactor is depressurized (50 psig above Torus pressure) and level is being maintained between +3 to +54 and the lead examiner has seen enough, the scenario may be stopped.
6
Appendix D Scenario Outline Form ES-D-1 IC -122 Facility: Cooper Nuclear Station Scenario No.: NRC 2 Op-Test No.: 1____
Examiners: ____________________________ Operators: _____________________
Initial Conditions: The plant is operating at approximately 100% with Surveillance 6.1DG101 in progress:
Turnover: After turnover, the crew is to secure DG#1 following its monthly test; LCO 3.8.1 Condition B is in effect.
Event Malf. Event Event No. No. Type* Description 1 N/A N DG-1 Monthly Surveillance 2 2 I,TS RR Pump speed lowers 3 3 C Combustion in the Off-Gas system 4 4 I,TS FW and Main Turbine high water level instruments fail 5 5 M Loss of Turbine High Pressure Fluid, requiring Manual Scram 6 6 C Loss of Emergency Transformer, Loss of Critical Bus 1F 7 7 C,5 DG-1 Fails to Auto Start 8 8 M,5 Small Break LOCA, Containment Sprays
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 5 Critical Tasks are located in Events 7 and 8.
Appendix D Scenario Outline Form ES-D-1 IC -122 Scenario Objective Evaluate the crews ability to perform normal surveillances and to respond to instrument and component failures during non-emergency conditions.
Evaluate the crews response to a rising reactor power level and a lowering Turbine High Pressure Fluid Reservoir.
Evaluate the crews response to a small break LOCA which causes Drywell temperatures and pressures to rise and before 280°F in the Drywell the crew initiates Drywell Sprays to maintain temperature below 280°F.
Scenario Summary Initial Conditions:
- The plant is operating at approximately 100% power
- No equipment out of service.
- 6.1DG.101 is in progress.
- DG-1 load was lowered to 1000 kW 5 minutes ago.
Events:
- DG-1 Monthly Surveillance
- RR Pump speed lowers - TS
- Combustion in the Off-Gas system
- FW and Main Turbine high water level instrument fails- TS
- High Pressure Fluid leak requiring Manual Scram
- Loss of Emergency Transformer, Loss of Critical Bus 1F
- DG-1 Fails to Auto Start
- Small Break LOCA, Containment Sprays Scenario Sequence
- Normal activity - DG-1 Monthly Surveillance
- Instrument Failure - RR speed lowers
- Component Failure before EOPs - Combustion in the Off-Gas System
- Instrument Failure - FW and Main Turbine high water level instrument fails
- Major Failure - Loss of Turbine High Pressure Fluid, requiring Manual Scram
- Component Failure after EOPs - Loss of Emrg Transfmr, Loss of Critical Bus 1F
- Component Failure after EOPs - DG-1Fails to Auto Start
- Accident mitigation strategy - Containment Sprays
Appendix D Scenario Outline Form ES-D-1 IC -122 Event One: DG-1 Monthly Surveillance Malfunction Required:
No malfunction required; this is a normal manipulation for the BOP.
Objective:
Evaluate the crew during normal surveillance activities.
Evaluate the BOP Operator unloading and securing DG1 in accordance with 6.1DG.101, 31 day load test.
Success Path:
The #1 DG is unloaded from 1000 KW and the engine is secured in accordance with the surveillance procedure.
Event Two: RR Pump speed lowers Malfunction Required:
RR17b set at 70%.
Override ZDIRRMGSWS16B(1) set to ON.
Objective:
Evaluate the crews response to a failed Jordan RR controller.
Evaluate the CRS addressing Technical Specifications for RR loop mismatch.
Success Path:
RR Pump A speed is lowered until both loop flows are balanced within specification.
The CRS will review Tech Specs LCO 3.4.1 and consider RR Pump B not in service until pump speeds are matched.
Event Three: Combustion in the Off-Gas System Malfunction Required:
OG02 SUSTAINED H2 BURN Objective:
Evaluate the crews response to combustion in the off-gas train as indicated by Annunciator B-3 / E-3 AUG OFFGAS TROUBLE and elevated temperatures on the Off-gas lines.
Evaluate the BOP Operators ability to accurately communicate steps of the Abnormal Procedure to the Turbine Building Station Operator.
Success Path:
The steps on the Abnormal are performed and the combustion is extinguished.
Appendix D Scenario Outline Form ES-D-1 IC -122 Event Four: FW and Main Turbine high water level instruments fail Malfunction Required:
RR27a set at 0%.
RR27c set at 0%.
Objective:
Evaluate the crews response to the failure 2 of the 3 RFP Turbine and Main Turbine high water level trip instrumentation.
Evaluate the CRS addressing Technical Specifications.
Success Path:
With only one instrument available (NBI-LT-52B) the crew will recognize that there is a loss of the two out of three logic for the high level trip for the RFP Turbines and the Main Turbine The CRS declares the instruments inoperable in accordance with Tech Specs LCO 3.3.2.2. Condition A. Required Action A.1 Place channel in trip within 7 days, and Condition B Required Action B.1 Restore feedwater and main turbine high water level trip capability within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Event Five: Loss of Turbine High Pressure Fluid, requiring a Manual Scram Malfunction Required:
TC10 Turbine High Press Fluid leak increased from 75%.
Objective:
Evaluate the pre-staging and conservative decision making prior to the need to scram the Reactor prior to losing Turbine High Pressure Fluid Pumps and control of Turbine GVs, Stop Valves and Bypass Valves.
Success Path:
Reactor is scrammed and pressure control is transferred to HPCI and SRVs.
Event Six: Loss of Emergency Transformer, Loss of Critical Bus 1F Malfunction Required:
ED06 LOSS OF POWER (EMERGENCY 69KV TRANSFORMER)
ED08A 4160 BUS 1A FAILURE Objective:
Appendix D Scenario Outline Form ES-D-1 IC -122 Evaluate the crews response to the loss of the Emergency Transformer and a subsequent loss of 1F Critical Bus, during the Scram recovery.
Evaluate the BOPs ability to enter Procedure 5.3EM-PWR and ensure the Critical Busses are powered by an emergency power source.
Evaluate the crews ability to shift RPV level control to the High Pressure ECCS and RCIC systems due to a loss of all Condensate and Booster pumps.
Success Path:
Startup Transformer is supplying the Critical Busses until a fault causes the loss of 1F.
RPV Level is being controlled within the +3 to 54 inch range with RCIC, CRD and HPCI.
Event Seven: DG-1 Fails to Auto Start Malfunction Required:
DG06A Diesel Generator #1 Fails to Auto Start Objective:
Evaluate the crew recognition that Diesel 1failed to auto start when required and to perform the necessary steps to start Diesel 1 and energize the Critical Bus.
Success Path:
Both Critical Busses are Energized; one from the Diesel Generator, the other from the Startup Transformer.
Event Eight: LOCA Containment Sprays Malfunction Required:
RR20A Coolant Leakage Inside Primary Containment @ 12% with a ramp time of 10 minutes.
Objective:
Evaluate the crew response to a slow increase in Drywell Temperature and pressure and to vent Primary Containment in an attempt to control the pressure rise.
Evaluate the crews ability to spray the Drywell in accordance with the EOPs to control pressure and temperature, as the LOCA gradually worsens.
Success Path:
Torus and Drywell Sprays are initiated prior to DW temperature reaching 280°F.
Scenario Termination:
When Reactor water level is being controlled between +3 and +54 inches and Drywell Sprays are controlling Drywell Pressure between 2 and 10 psig.
Appendix D Scenario Outline Form ES-D-1 Facility: Cooper Nuclear Station Scenario No.: NRC 3 Op-Test No.: 1 Examiners: Operators:
Initial Conditions: The plant is operating at approximately 2.0% power, holding here to perform the 500 psig Reactor Building walkdown. After turnover, the crew is to shift CRD Pumps.
Event Event Event No. Malf. No.
Type* Description 1 N/A N Shift CRD Pumps 2 2 C,TS REC Pump B trip 3 3 I IRM Inop Trip 4 4 I,TS Drywell Radiation Monitor Isolation 5 5 C,R Rod drop and stuck rods non-EOP rod driving 6 6 M RCIC Steam Line Leak 7 7 C, Fuel failure ATWS EOP rod driving 8 8 C, Reactor Recirc Pump speed fails high 9 9 C, ED on Secondary Containment 2 Areas
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Critical Tasks are located in Events 7, 8 and 9.
1
Appendix D Scenario Outline Attachment 2 Scenario Objective Evaluate the Crews ability to operate at low power levels.
Evaluate the Crews response to a rod drop accident and fuel failure.
Evaluate the crews actions when RCIC develops a steam line leak that will not fully isolate and when two areas in Secondary Containment exceed their Max Safe values the RPV is Emergency Depressurized to limit the release of highly radioactive steam through the RCIC steam line break.
Scenario Summary Initial Conditions:
- The plant is operating at approximately 2.0% power
- The 500 psig Reactor Building walkdown is in progress.
- Shift CRD Pumps following turnover.
Events:
- Shift CRD Pumps
- REC Pump B trip
- IRM Inop Trip
- DW Vent Radiation Monitor Isolation
- Rod drop and stuck rods requiring EOP rod driving
- RCIC Steam Line Leak
- Reactor Recirc Pump speed fails high
- ED on Secondary Containment 2 Areas Scenario Sequence
- Normal activity - Shift CRD Pumps B to A
- Component Failure before EOPs - Rod drop and stuck control rods requiring non-EOP driving
- Major Event - RCIC Steam Line Leak
- Component Failure after EOPs - Reactor Recirc Pump speed fails high
- Accident mitigation strategy - Emergency Depressurize the RPV 2
Appendix D Scenario Outline Attachment 2 Event One: Shift CRD Pumps Malfunction Required:
No malfunction required; this is a normal manipulation for the RO.
Objective:
Evaluate the crew during normal equipment shifting.
Evaluate the Reactor Operator shifting from the B CRD Pump running to the A CRD Pump running and securing the B Pump.
Success Path:
The A CRD Pump is running and the B CRD Pump is secured. All CRD parameters indicated on Panel 9-5 restored to within their normal band.
Event Two: REC Pump B trip Malfunction Required:
SW11B - REC Pump Trip 1B.
Objective:
Evaluate the crews response to the tripping of one of the three Reactor Equipment Cooling Pumps and takes appropriate action in accordance with the Annunciator Procedure to restart another pump prior to receiving an REC Isolation.
Evaluate the CRS addressing Technical Specifications.
Success Path:
The BOP Operator either responds quickly enough (within 1 minute) to the tripping of the pump and starts an additional REC Pump in accordance with the Annunciator Card.
Or, the REC system isolation is reset following the restart of the third REC Pump and system flows and pressures are returned to normal.
The SRO will address Tech Specs and determine that LCO 3.7.3 Condition B, a 30 day LCO on one sub system.
Event Three: IRM Inop trip Malfunction Required:
Malfunction NM13C - IRM INOP Channel-C.
NM06G IRM G Stuck and fully withdrawn Objective:
Evaluate the crews response to a failed Intermediate Range Monitor (IRM).
Evaluate the At the Controls (ATC) Operators actions to determine the cause of the 1/2 Scram, and bypasses the failed IRM. This allows resetting the half-scram.
Evaluate the CRS addressing Technical Specifications for the failed IRM.
3
Appendix D Scenario Outline Attachment 2 Success Path:
IRM - C is bypassed, and the CRS initiates an LCO on IRM C in accordance with Technical Specifications 3.3.1.1 (RPS Instrumentation) Table 3.3.1.1-1 Function 1. Also TRM T3.3.1 Function 2 potential LCO with one INOP IRM there remains the minimum number required of 6.
Event Four: Drywell Vent Radiation Monitor isolation Malfunction Required:
Override 02S87 ZDIRMARA10AV[1] = CLOSE Drywell Vent Rad Mon. Isol Vlv.
Override 12A2AR1 RMV-RR-4 DW Vent Rad Monitor - Gas to 4.0E-006 Objective:
Evaluate the CRS addressing Technical Specifications for the failed Drywell Radiation Monitor.
Success Path:
CRS initiates an LCO on DW Radiation Monitor in accordance with Technical Specifications 3.4.5 (RCS Leakage Detection Instrumentation) b. Condition B, Required Action B.1 Grab Sample once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and B.2 Restore within 30 days.
Event Five: Rod drop and stuck rods non-EOP rod driving Malfunction Required:
RD02B ATWS South Bank set at 75%
CR023827 Increased Rod Worth on rod 38-27 set at 40%
RD133827 Rod Uncoupled RD123827 Rod Stuck - Delete CR01 Fuel Failure at 100%
Objective:
Evaluate the crews response to approximately half of the control rod insertion on a reactor scram signal.
Evaluate the ROs ability to perform 2.4CRD and drive control rods using RMCS.
Evaluate the CRS implementing strategy for reactivity controls outside the EOPs.
Evaluate the crews teamwork in installing jumpers and controlling Reactor Pressure and level.
Success Path:
Control Rods are inserted using RMCS.
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Appendix D Scenario Outline Attachment 2 Event Six: RCIC Steam Line Leak Malfunction Required:
RC06 RCIC Steam Line Break in at 100%
RC07 Failure of RCIC Auto-Isolation OR ZDIRCICSWS2(2) MO-16 C/S to OPEN OR ZDIRCICSWS1(2) MO-15 C/S to OPEN RF RC06A RCIC-MO-16 Control Power De-energized Objective:
Evaluate the crews response to a failure of RCIC to fully isolate during a RCIC Steam Line Break.
Evaluate the BOPs ability to monitor and report Secondary Containment Temperatures and Radiation Levels to the CRS.
Evaluate the crews ability to continue Control Rod insertion in accordance with 2.4CRD and Emergency Depressurize the RPV when 2 Areas in Secondary Containment exceed Max Safe values.
Success Path:
RPV Level is being controlled within the +3 to 54 inch range with CRD and HPCI and Condensate. The Reactor is depressurized to <50 psig above Torus Pressure when two areas in Secondary Containment reach and exceed Max Safe values.
Event Seven: Fuel Failure ATWS EOP rod driving Malfunction Required:
None Objective:
Evaluate the crew recognition of two areas being above Max Safe Operating Temperature and entry into EOP 5A.
Evaluate the crews ability work through EOP 5A to the point where it transitions to EOP 1A and eventually to 6A and 7A to give directions to drive control rods.
Success Path:
The Crew will enter EOPs and start driving control rods full in with EP 5.8.3 versus 2.4CRD.
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Appendix D Scenario Outline Attachment 2 Event Eight: Reactor Recirc Pump speed fails high Malfunction Required:
RR17B RR MGSet Jordan Controller failure to 100 ZDIRRSWS2B[1] = ON Reset Scooptube Lockout Objective:
Evaluate the crew recognition of the increasing speed of the B RR Pump.
Evaluate the crews ability to monitor nuclear instrumentation and RR Pump speed.
Success Path:
The Crew will trip the RR Pump when it is determined that either it is at full speed or that the scoop tube lock will not stop the increase.
Event Nine: ED on Secondary Containment 2 Areas Malfunction Required:
None Objective:
Evaluate the crew response to a slow increase in Reactor Building Temperatures and Radiation levels to the point where the RPV must be Emergency Depressurized.
Evaluate the crews ability to manually open 6 SRVs and reduce RPV Pressure to less than 50 psig above Torus pressure, in accordance with the EOPs.
Success Path:
RPV is depressurized to 50 psig above Torus pressure.
Scenario Termination:
When Reactor water level is being controlled between +3 and +54 inches and the RPV has been Emergency Depressurized and all but one Control Rod have been inserted.
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Appendix D Scenario Outline Form ES-D-1 Facility: Cooper Nuclear Station Scenario No.: NRC 4 (Spare) Op-Test No.: 1 Examiners: Operators:
Initial Conditions: The plant is operating at approximately 100% power. There are thunderstorms in the area and Nemaha Co. is in a thunderstorm watch. After turnover, the crew is to continue with the Core Spray Surveillance.
Event Event Event No. Malf. No.
Type* Description 1 N/A N,TS CS Surveillance 2 2 C CS Pump Trip 3 3 C,TS Torus to Drywell Vacuum Breaker Opens 4 4 N Severe Thunderstorm Warning, increased winds, shift Air Comps.
5 5 C,TS Lightning Strike - 125V B Ground loss of HPCI Starter Rack 6 6 M, Lightning Strike - Both Reactor Feed Pump Suction Pressure Trip 7 7 C Loss of 125VDC A Emergency Depressurization to restore level with low pressure 8 N/A systems
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Critical Task are located in Event 5 and 7.
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Appendix D Scenario Outline Form ES-D-1 Scenario Objective Evaluate the crews ability to perform surveillance and normal operations and respond to instrument failure which has an input to the Reactor Vessel Level Control System.
Evaluate the crews response to a loss of RPV level indication and entry into the RPV Flooding procedures.
Evaluate CRS ability to determine Tech Spec LCOs are not met and take appropriate Required Actions.
Scenario Summary Initial Conditions:
- 100% Power
- Surveillance Procedure 6.1CS.101, CORE SPRAY TEST MODE SURVEILLANCE OPERATION (IST) (DIV 1) (92 day) is due to be performed at the beginning of shift.
Events:
- CS Surveillance
- CS Pump Trip
- Torus to Drywell Vacuum Breaker Opens
- Severe Thunderstorm Warning, increased winds, shift Air Comps.
- Lightning Strike - 125V B Ground loss of HPCI Starter Rack
- Lightning Strike - Both Reactor Feed Pump Suction Pressure Trip
- Loss of 125VDC A
- Emergency Depressurization to restore level with low pressure systems Scenario Sequence
- Normal activity - CS Surveillance
- Component Failure before EOPs - Torus to Drywell Vacuum Breaker Opens
- Abnormal event with Normal Activity - Severe TS Warning, shift Station Air Compressors.
- Major Event - Lightning Strike - Both Rx Feed Pump Suction Pressure Trip circuit initiates
- Component Failure after EOPs - 125VDC Bus A failure
- Accident mitigation strategy - Alternate RPV Injection to control level 2
Appendix D Scenario Outline Form ES-D-1 Event One: CS Surveillance Malfunction Required:
None Objective:
Evaluate the crew during normal equipment surveillance.
Evaluate SRO Tech Spec entry for CS surveillance Success Path:
The A Core Spray Pump is started and the minimum flow valve is verified to close as flow is raised.
The SRO will address Tech Specs for inoperable subsystem and pump.
Event Two: CS Pump Trip Malfunction Required:
CS01a Core Spray A pump trip.
Objective:
Evaluate the crews response to the tripping of the CS Pump.
Evaluate the CRS directing the steps of the surveillance that must be performed and those to be N/Aed.
Success Path:
The Crew places the A Core Spray system valves back to the normal alignment.
Event Three: Torus to Drywell Vacuum Breaker NRV-21 opens Malfunction Required:
PC02b Torus To Drywell Vacuum Breaker Failure NRV-21 (open) set at 100%
Objective:
Evaluate the crews response to a vacuum breaker failing open.
Evaluate the CRS addressing Technical Specifications for inoperable vacuum breaker.
Success Path:
The SRO addresses Tech Specs and declares the vacuum breaker inoperable.
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Appendix D Scenario Outline Form ES-D-1 Event Four: NAWAS issues Severe Thunderstorm Warning Malfunction Required:
HV03 Lightning Strike Objective:
Evaluate the crews response to a weather emergency and performance of mitigating steps.
Success Path:
The Crew enters 5.1Weather and prepares the site for severe weather. The crew shifts the Station Air Compressors and aligns REC to the running compressor.
The CRS ensures that a safety announcement is made over the Gaitronics to help protect the personnel on site.
Event Five: Lightning Strike - Lightning Strike - 125VB Ground loss of HPCI Starter Rack TS-3.0.3 Shutdown Malfunction Required:
HV03 Lightning Strike ED12A HPCI Starter Rack Loss of Power OR RA:MUX01C072 125V DC B Bus Ground Annunciator Objective:
Evaluate the crews response to a ground on the 125 V DC System and subsequent loss of HPCI Starter Rack.
Evaluate the CRS addressing Technical Specifications.
Success Path:
The Crew identifies the loss of the HPCI Starter Rack by walking down the control board.
The CRS determines that HPCI is inoperable Event Six: Lightning Strike - Both Rx Feed Pump Suction Pressure Trip circuit initiates Malfunction Required:
HV03 Lightning Strike FW01A Time Delay 10 seconds FW01B Time Delay 15 seconds Override Annunciator A1/A6 RFP A Suction Pressure Low ON Override Annunciator A2/A3 RFP B Suction Pressure Low ON 4
Appendix D Scenario Outline Form ES-D-1 Objective:
Evaluate the crews response to the failure of the suction pressure switches for both Reactor Feed Pumps and timed sequence of tripping. This Results in the subsequent loss of feedwater to the vessel.
Success Path:
The Crew identifies the loss of feedwater and scrams the reactor prior to the automatic trip. The Crew should use their only available injection sources to maintain and recover level.
Event Seven: 125VDC Bus A failure Malfunction Required:
ED12D Loss of 125 VDC Bus Distribution Panel Objective:
Evaluate the crews ability to recognize the loss of RCIC and inability of the remaining high pressure injection sources to restore level.
Success Path:
The Crew identifies the loss of RCIC and the inability to restore level. The Crew transitions to Alternate RPV Injection using SLC and CRD.
Event Eight: Anticipate emergency depressurization, or emergency depressurizes Malfunction Required:
None Objective:
Evaluate the crews ability to recognize the inability of remaining high pressure injection sources to restore RPV water level and either Alternate Emergency Depressurizes or Emergency Depressurizes the RPV to allow low pressure systems to restore level.
Success Path:
The Crew identifies the loss of loss of RCIC to restore level. The Crew transitions to Alternate RPV Injection using SLC then to Emergency Depressurization to refill the vessel.
Scenario Termination:
When the rate of level decrease has slowed because of the use of alternate injection sources or the crew has emergency depressurized the vessel and injected with low pressure systems or at the direction of the lead examiner.
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