ML20241A247
| ML20241A247 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 08/04/2020 |
| From: | Greg Werner Operations Branch IV |
| To: | Nebraska Public Power District (NPPD) |
| References | |
| Download: ML20241A247 (51) | |
Text
ES-401 1
Form ES-401-1 Rev 3 Facility: Cooper Nuclear Station Date of Exam: 04/2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
1 4
4 N/A 4
4 N/A 3
20 7
2 1
1 1
2 1
1 7
3 Tier Totals 2
5 5
6 5
4 27 10
- 2.
Plant Systems 1
2 2
2 3
2 2
2 3
3 2
3 26 5
2 1
1 2
1 1
1 1
1 1
1 1
12 3
Tier Totals 3
3 4
4 3
3 3
4 4
3 4
38 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
2 3
2 3
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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 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 outline. Systems or evolutions that do not apply at the facility should be deleted with justification. 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. 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 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 a category 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.
G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-1 Rev 3 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X
Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:
AK3.03 Idle loop flow 2.8 38 295003 (APE 3) Partial or Complete Loss of AC Power / 6 X
Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER:
AA2.05 Whether a partial or complete loss of A.C. power has occurred 3.9 28 295004 (APE 4) Partial or Total Loss of DC Power / 6 X 2.2.22 Knowledge of limiting conditions for operations and safety limits.
4.0 52 295005 (APE 5) Main Turbine Generator Trip /
3 X
Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP:
AA1.01 Recirculation system: Plant-Specific 3.1 58 295006 (APE 6) Scram / 1 X
Ability to operate and/or monitor the following as they apply to SCRAM:
AA1.03 Reactor/turbine pressure regulating system 3.7 48 295016 (APE 16) Control Room Abandonment
/ 7 X
Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:
AK2.01 Remote shutdown panel: Plant-Specific 4.4 45 295018 (APE 18) Partial or Complete Loss of CCW / 8 X
Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:
AK3.06 Increasing cooling water flow to heat exchangers 3.3 46 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X
Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:
AA2.01 Instrument air system pressure 3.5 60 295021 (APE 21) Loss of Shutdown Cooling /
4 X
Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING:
AA2.06 Reactor pressure 3.2 29 295023 (APE 23) Refueling Accidents / 8 X
Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS:
AA2.01 Area radiation levels 3.6 42 295024 High Drywell Pressure / 5 X
Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE:
EK3.06 Reactor SCRAM 4.0 32
ES-401 3
Form ES-401-1 Rev 3 295025 (EPE 2) High Reactor Pressure / 3 X
Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:
EK2.09 Reactor power 3.9 51 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X
Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:
EK1.02 Steam condensation 3.5 49 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 N/A for CNS 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 X
Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following:
EK2.04 Drywell ventilation 3.6 41 295030 (EPE 7) Low Suppression Pool Water Level / 5 X 2.2.40 Ability to apply Technical Specifications for a system.
3.4 44 295031 (EPE 8) Reactor Low Water Level / 2 X
2.4.3 Ability to identify post-accident instrumentation.
2.4.2 Knowledge of system setpoints, interlocks and automatic actions associated with EOP entry conditions.
4.5 57 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X
Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:
EA1.07 RMCS: Plant-Specific 3.9 36 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X
Knowledge of the reasons for the following responses as they apply to HIGH OFF-SITE RELEASE RATE:
EK3.03 Control room ventilation isolation:
Plant-Specific 3.7 34 600000 (APE 24) Plant Fire On Site / 8 X
Ability to operate and / or monitor the following as they apply to PLANT FIRE ON SITE:
AA1.08 Fire fighting equipment used on each class of fire 2.6 64 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X
Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following:
AK2.02 Breakers, relays 3.1 39 K/A Category Totals:
1 4
4 4
4 3
Group Point Total:
20/7
ES-401 4
Form ES-401-1 Rev 3 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 295002 (APE 2) Loss of Main Condenser Vacuum / 3 295007 (APE 7) High Reactor Pressure / 3 X
Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE:
AA1.05 Reactor/turbine pressure regulating system 3.7 33 295008 (APE 8) High Reactor Water Level / 2 295009 (APE 9) Low Reactor Water Level / 2 295010 (APE 10) High Drywell Pressure / 5 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 295012 (APE 12) High Drywell Temperature /
5 X
Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE:
AA2.03 Drywell humidity: Plant-Specific 2.8 40 295013 (APE 13) High Suppression Pool Temperature. / 5 X 2.4.18 Knowledge of the specific bases for EOPs.
3.3 55 295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 295017 (APE 17) Abnormal Offsite Release Rate / 9 X
X Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE:
AK1.03 Meteorological effects on off-site release Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following:
AK2.10 Process radiation monitoring system 3.3 35 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 X
X Knowledge of the interrelations between LOSS OF CRD PUMPS and the following:
AK2.03 Accumulator pressures Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS:
AK1.02 Reactivity control 3.6 62 295029 (EPE 6) High Suppression Pool Water Level / 5
ES-401 5
Form ES-401-1 Rev 3 295032 (EPE 9) High Secondary Containment Area Temperature / 5 X
Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE:
EK3.03 Isolating affected systems 3.8 53 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 X
Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:
EA1.02 SBGT/FRVS 3.8 56 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 K/A Category Point Totals:
1 1
1 2
1 1
Group Point Total:
7/3
ES-401 6
Form ES-401-1 Rev 3 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 1 (RO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 203000 (SF2, SF4 RHR/LPCI)
RHR/LPCI: Injection Mode X
Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) controls including:
A1.01 Reactor water level 4.2 26 205000 (SF4 SCS) Shutdown Cooling X
Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following:
K3.05 Fuel pool cooling assist: Plant-Specific 2.6 19 206000 (SF2, SF4 HPCIS)
High-Pressure Coolant Injection X
2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.
2.4.31 Knowledge of annunciator alarms, indications, or response procedures.
4.2 10 207000 (SF4 IC) Isolation (Emergency) Condenser N/A for CNS 209001 (SF2, SF4 LPCS)
Low-Pressure Core Spray X
Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM:
K5.04 Heat removal (transfer) mechanisms 2.8 25 209002 (SF2, SF4 HPCS)
High-Pressure Core Spray N/A for CNS 211000 (SF1 SLCS) Standby Liquid Control X X Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.03 A.C. power failures Ability to monitor automatic operations of the STANDBY LIQUID CONTROL SYSTEM including:
A3.06 RWCU system isolation: Plant-Specific 3.2 4.0 3
17 212000 (SF7 RPS) Reactor Protection X
Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.11 Main steamline isolation valve closure 4.0 20
ES-401 7
Form ES-401-1 Rev 3 215003 (SF7 IRM)
Intermediate-Range Monitor X
Knowledge of the operational implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM)
SYSTEM:
K5.03 Changing detector position 3.0 13 215004 (SF7 SRMS) Source-Range Monitor X
Knowledge of the effect that a loss or malfunction of the SOURCE RANGE MONITOR (SRM) SYSTEM will have on following:
K3.04 Reactor power and indication 3.7 11 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X
Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following:
K4.07 Flow biased trip setpoints 3.7 24 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X
Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC):
K6.01 Electrical power 3.4 21 218000 (SF3 ADS) Automatic Depressurization X
Knowledge of the physical connections and/or cause-effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following:
K1.02 Low pressure core spray: Plant-Specific 4.0 2
223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X X Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.02 D.C. electrical distribution failures Ability to monitor automatic operations of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including:
A3.03 SPDS/ERIS/CRIDS/GDS: Plant-Specific A3.02 Valve closures 2.9 2.5 3.5 14 6
ES-401 8
Form ES-401-1 Rev 3 239002 (SF3 SRV) Safety Relief Valves X
X Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including:
A1.06 Reactor power 2.1.32 Ability to explain and apply system limits and precautions.
3.7 3.8 4
16 259002 (SF2 RWLCS) Reactor Water Level Control X
Ability to manually operate and/or monitor in the control room:
A4.11 High level lockout reset controls: Plant-Specific 3.5 15 261000 (SF9 SGTS) Standby Gas Treatment X
Knowledge of the physical connections and/or cause-effect relationships between STANDBY GAS TREATMENT SYSTEM and the following:
K1.02 Drywell 3.2 7
262001 (SF6 AC) AC Electrical Distribution X
Knowledge of electrical power supplies to the following:
K2.01 Off-site sources of power 3.3 8
262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)
X X
Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) design feature(s) and/or interlocks which provide for the following:
K4.01 Transfer from preferred power to alternate power supplies Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.):
K6.03 Static inverter 3.1 2.7 5
12 263000 (SF6 DC) DC Electrical Distribution X
Ability to manually operate and/or monitor in the control room:
A4.01 Major breakers and control power fuses: Plant-Specific 3.3 23 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X
X Ability to monitor automatic operations of the EMERGENCY GENERATORS (DIESEL/JET) including:
A3.06 Cooling water system operation 2.4.46 Ability to verify that the alarms are consistent with the plant conditions.
3.1 4.2 22 18 300000 (SF8 IA) Instrument Air X
Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following:
K4.02 Cross-over to other air systems 3.0 1
ES-401 9
Form ES-401-1 Rev 3 400000 (SF8 CCS) Component Cooling Water X
Knowledge of electrical power supplies to the following:
K2.02 CCW valves 2.9 9
K/A Category Point Totals:
2 2
2 3
2 2
2 3
3 2
3 Group Point Total:
26/5
ES-401 10 Form ES-401-1 Rev 3 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 2 (RO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 201001 (SF1 CRDH) CRD Hydraulic X
Knowledge of CONTROL ROD DRIVE HYDRAULIC SYSTEM design feature(s) and/or interlocks which provide for the following:
K4.12 Controlling CRD system flow 2.9 65 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism X
Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROD AND DRIVE MECHANISM:
K6.02 Reactor pressure 3.0 54 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup X
2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
4.5 37 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe X
Ability to (a) predict the impacts of the following on the TRAVERSING IN-CORE PROBE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.01 Low reactor water level: Mark-I&II (Not-BWR1) 2.7 61 215002 (SF7 RBMS) Rod Block Monitor 216000 (SF7 NBI) Nuclear Boiler Instrumentation X
Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.09 Jet pump flow: Design-Specific 3.1 61 219000 (SF5 RHR SPC) RHR/LPCI:
Torus/Suppression Pool Cooling Mode X
Knowledge of the operational implications of the following concepts as they apply to RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE:
K5.04 Heat exchanger operation 2.9 43 223001 (SF5 PCS) Primary Containment and Auxiliaries
ES-401 11 Form ES-401-1 Rev 3 226001 (SF5 RHR CSS) RHR/LPCI:
Containment Spray Mode 230000 (SF5 RHR SPS) RHR/LPCI:
Torus/Suppression Pool Spray Mode X
Knowledge of the physical connections and/or cause-effect relationships between RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE and the following:
K1.06 Keep fill system 3.0 50 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel-Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam X
Knowledge of the effect that a loss or malfunction of the MAIN AND REHEAT STEAM SYSTEM will have on following:
K3.05 Condenser air removal 2.8 59 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate X
Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CONDENSATE SYSTEM controls including:
A1.10 Condenser vacuum A1.07 System lineup 3.1 31 259001 (SF2 FWS) Feedwater X
Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including:
A1.06 Feedwater heater level 2.7 31 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas X
Ability to manually operate and/or monitor in the control room:
A4.02 System flows 2.9 63 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection X
Knowledge of electrical power supplies to the following:
K2.02 Pumps 2.9 47 288000 (SF9 PVS) Plant Ventilation X
Knowledge of the effect that a loss or malfunction of the PLANT VENTILATION SYSTEMS will have on following:
K3.01 Secondary containment temperature: Plant-Specific 2.8 27 290001 (SF5 SC) Secondary Containment
ES-401 12 Form ES-401-1 Rev 3 290003 (SF9 CRV) Control Room Ventilation X
Ability to monitor automatic operations of the CONTROL ROOM HVAC including:
A3.01 Initiation/reconfiguration 3.3 30 290002 (SF4 RVI) Reactor Vessel Internals K/A Category Point Totals:
1 1
2 1
1 1
1 1
1 1
1 Group Point Total:
12/3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Rev 3 Facility: Cooper Nuclear Station Date of Exam: 04/2020 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps.
4.6 74 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management.
4.3 67 Subtotal 2
- 2. Equipment Control 2.2.7 2.2.14 Knowledge of the process for conducting special or infrequent tests.
2.2.14 Knowledge of the process for controlling equipment configuration or status.
3.9 72 2.2.40 Ability to apply technical specifications for a system.
3.4 75 2.2.43 Knowledge of the process used to track inoperable alarms.
3.0 73 Subtotal 3
- 3. Radiation Control 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
2.9 71 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.
3.2 69 Subtotal 2
- 4. Emergency Procedures/Plan 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
4.2 66 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
4.6 68 2.4.3 Ability to identify post-accident instrumentation.
3.7 70 Subtotal 3
Tier 3 Point Total 10 7
ES-401 14 Form ES-401-1 Rev 4 Facility: Cooper Nuclear Station Date of Exam: 04/2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
N/A N/A 20 4
3 7
2 7
2 1
3 Tier Totals 27 6
4 10
- 2.
Plant Systems 1
26 2
3 5
2 12 1
2 3
Tier Totals 38 3
5 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
2 2
1 2
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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 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 outline. Systems or evolutions that do not apply at the facility should be deleted with justification. 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. 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 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 a category 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.
G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 15 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X
Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:
AA2.04 Individual jet pump flows: Not-BWR-1&2 3.1 82 295003 (APE 3) Partial or Complete Loss of AC Power / 6 295004 (APE 4) Partial or Total Loss of DC Power / 6 295005 (APE 5) Main Turbine Generator Trip /
3 295006 (APE 6) Scram / 1 295016 (APE 16) Control Room Abandonment
/ 7 295018 (APE 18) Partial or Complete Loss of CCW / 8 X 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.
4.0 90 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X
Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:
AA2.02 Status of safety-related instrument air system loads (see AK2.1 - AK2.19) 3.7 91 295021 (APE 21) Loss of Shutdown Cooling /
4 295023 (APE 23) Refueling Accidents / 8 295024 High Drywell Pressure / 5 X
2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
4.7 81 295025 (EPE 2) High Reactor Pressure / 3 X
Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE:
EA2.06 Reactor water level 3.8 77 295026 (EPE 3) Suppression Pool High Water Temperature / 5 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 NA for CNS 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 X 2.2.12 Knowledge of surveillance procedures.
4.1 83 295030 (EPE 7) Low Suppression Pool Water Level / 5 295031 (EPE 8) Reactor Low Water Level / 2
ES-401 16 Form ES-401-1 Rev 4 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 600000 (APE 24) Plant Fire On Site / 8 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X
Ability to determine and/or interpret the following as l they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:
AA2.07 Operational status of engineered safety features 4.0 88 K/A Category Totals:
4 3
Group Point Total:
20/7
ES-401 17 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 295002 (APE 2) Loss of Main Condenser Vacuum / 3 X
2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
4.6 87 295007 (APE 7) High Reactor Pressure / 3 295008 (APE 8) High Reactor Water Level / 2 295009 (APE 9) Low Reactor Water Level / 2 295010 (APE 10) High Drywell Pressure / 5 X
Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:
AA2.02 Drywell pressure 3.9 89 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 295012 (APE 12) High Drywell Temperature /
5 295013 (APE 13) High Suppression Pool Temperature. / 5 295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 X
Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM:
AA2.02 Control rod position 4.2 85 295017 (APE 17) Abnormal Offsite Release Rate / 9 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 295029 (EPE 6) High Suppression Pool Water Level / 5 295032 (EPE 9) High Secondary Containment Area Temperature / 5 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 K/A Category Point Totals:
2 1
Group Point Total:
7/3
ES-401 18 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 1 (SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 203000 (SF2, SF4 RHR/LPCI)
RHR/LPCI: Injection Mode 205000 (SF4 SCS) Shutdown Cooling 206000 (SF2, SF4 HPCIS)
High-Pressure Coolant Injection 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS)
Low-Pressure Core Spray X
Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.01 Pump trips 3.4 84 209002 (SF2, SF4 HPCS)
High-Pressure Core Spray 211000 (SF1 SLCS) Standby Liquid Control 212000 (SF7 RPS) Reactor Protection 215003 (SF7 IRM)
Intermediate-Range Monitor 215004 (SF7 SRMS) Source-Range Monitor 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X 2.2.38 Knowledge of conditions and limitations in the facility license.
4.5 78 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling 218000 (SF3 ADS) Automatic Depressurization 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff 239002 (SF3 SRV) Safety Relief Valves 259002 (SF2 RWLCS) Reactor Water Level Control X
2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.
2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.
2.4.31 Knowledge of annunciator alarms, indications, or response procedures.
4.1 76
ES-401 19 Form ES-401-1 Rev 4 261000 (SF9 SGTS) Standby Gas Treatment 262001 (SF6 AC) AC Electrical Distribution 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG 300000 (SF8 IA) Instrument Air X
2.2.40 Ability to apply Technical Specifications for a system.
2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
4.7 86 400000 (SF8 CCS) Component Cooling Water X
Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
A2.01 Loss of CCW pump 3.4 92 K/A Category Point Totals:
2 3
Group Point Total:
26/5
ES-401 20 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 2 (SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe X 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
4.7 80 215002 (SF7 RBMS) Rod Block Monitor X 2.2.22 Knowledge of limiting conditions for operations and safety limits.
4.7 79 216000 (SF7 NBI) Nuclear Boiler Instrumentation 219000 (SF5 RHR SPC) RHR/LPCI:
Torus/Suppression Pool Cooling Mode 223001 (SF5 PCS) Primary Containment and Auxiliaries 226001 (SF5 RHR CSS) RHR/LPCI:
Containment Spray Mode 230000 (SF5 RHR SPS) RHR/LPCI:
Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel-Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate
ES-401 21 Form ES-401-1 Rev 4 259001 (SF2 FWS) Feedwater X
Ability to (a) predict the impacts of the following on the REACTOR FEEDWATER SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.02 Feedwater heater isolation 3.3 93 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals K/A Category Point Totals:
1 2
Group Point Total:
12/3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Rev 4 Facility: Cooper Nuclear Station Date of Exam: 04/2020 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.
3.8 98 2.1.35 Knowledge of fuel handling responsibilities of SROs.
3.9 100 Subtotal 2
- 2. Equipment Control 2.2.
2.2.38 Knowledge of conditions and limitations in the facility license.
4.5 95 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
4.6 94 Subtotal 2
- 3. Radiation Control 2.3.11 Ability to control radiation releases.
4.3 97 Subtotal 1
- 4. Emergency Procedures/Plan 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
4.5 96 2.4.19 2.4.40 Knowledge of EOP layout, symbols, and icons.
Knowledge of SRO responsibilities in emergency plan implementation.
4.5 99 Subtotal 2
Tier 3 Point Total 10 7
ES-401 Record of Rejected K/As Form ES-401-4 Rev 4 Tier / Group Randomly Selected K/A Reason for Rejection RO T1/G1 295037 EA1.08 295037 EA1.07 (Q#36) CNS does not have RC&IS. Replaced 295037 EA1.08 (RC&IS) with 295037 EA1.07 (RMCS). Page 1 point totals not affected by this change. (Rev 1)
RO T2/G1 212000 A2.17 212000 A2.11 (Q#20) RPS does not receive input from MSL high radiation at CNS. Replaced 212000 A2.17 (MSL high radiation) with 212000 A2.11 (MSIV closure). Page 1 point totals not affected by this change. (Rev 1)
RO T2/G1 239002 A1.02 239002 A1.06 (Q#4) CNS does not have acoustical monitors. Replaced 239002 A1.02 (acoustical monitor noise) with 239002 A1.06 (reactor power). Page 1 point totals not affected by this change. (Rev 1)
RO T2/G1 206000 G2.4.35 206000 2.4.31 (Q#10) There are no local auxiliary operator tasks associated HPCI during emergencies except for reading REC flow locally, which is not operationally significant.
[NRC] Replaced with 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.
RO T1/G2 295017 AK1.03 295017 AK2.10 (Q#35) Could not develop discriminatory question at the RO level regarding meteorological effects on off-site release.
ROs are not involved in dose assessment or PARs at CNS.
[NRC] Replaced with K/A AK2.10: Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: AK2.10 Process radiation monitoring system.
NOTE: No other AK1 K/As were acceptable for RO testing for this E/APE, so AK2 K/A was chosen. Because this dropped the Tier 1 K1 K/As to less than 2, another T1/G2 System with a K2 K/A was changed to K1 for Q#62.
RO T1/G2 295022 AK2.03 295022 AK1.02 (Q#62) [NRC] System 295022 AK2.03 Loss of Control Rod Drive Pumps:
Replaced with K1 K/A AK1.02, Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS: Reactivity control This is to make up for loss of 1 K1 K/A per Q#35 question change, above.
ES-401 Record of Rejected K/As Form ES-401-4 Rev 4 RO T2/G2 295017 AK1.03 215001 A2.01 216000 A2.09 (Q#61) TIP question overlaps SRO Q#80 and has low IR (2.7). Recommend replacing with A2 from K/A 216000 NBI (same safety function)
[NRC] Agree, replace system with 216000 A2.09, Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Jet pump flow: Design-Specific RO T3 G2.2.7 G2.2.14 (Q#72) Could not develop discriminatory question at the RO level regarding Infrequently Performed Tests and Evolutions.
ROs have no detailed responsibilities regarding IPTEs at CNS.
[NRC] Replaced with 2.2.14 Knowledge of the process for controlling equipment configuration or status.
SRO T2/G1 259002 G2.4.9 259002 G2.4.30 (Q#76) Could not develop question at SRO level for K/A 2.4.9 related to RVLCS due to K/A is for low power/shutdown accident mitigation strategies.
[NRC] Replaced with G2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.
SRO T2/G1 215005 G2.2.38 N/A (Q#78) There are no License Conditions [10 CFR 55.43(b)(1) topics] related to APRMs.
[NRC] Technical Specifications are part of the facility license, test on Tech Spec aspects of APRMs to satisfy conditions and limitations in the facility license to satisfy the original K/A.
SRO T3 G2.4.19 G2.4.40 (Q#99) Could not develop question at SRO level for knowledge of EOP layout, symbols, and icons.
[NRC] Replace with 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation.
RO T2/G2 259001 A1.06 256000 A1.10 (Q#31) [NRC Requested] K/A 259001 is sampled on the SRO exam. The same system cannot be sampled on both RO and SRO exams until all systems in a tier/group have been sampled between the two exams. (12-8-19)
SRO T2/G1 300000 G2.2.40 300000 G2.1.7 (Q#86) Per NRC direction from free review because of K/A mismatch, for Q#86 replaced G2.2.40 with G2.1.7 for K/A 300000, since Instrument Air is not a TS system. (1-6-20)
ES-401 Record of Rejected K/As Form ES-401-4 Rev 4 RO T2/G2 256000 A1.10 256000 A1.07 (Q#31) Could not develop a psychometrically sound question at the RO level for this K/A (Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CONDENSATE SYSTEM controls including:
Condenser vacuum). Condensate supplies the SJAE condensers, but there are no control room controls for aligning the specific flow path. Also, reduction in condensate flow would tend to always reduce SJAE efficiency and cause condenser pressure to rise, so developing three plausible distractors may be unattainable. Recommend replacing with A1.06 or A1.07.
[NRC] Replace with 256000 A1.07 System lineup RO T1/G1 295031 G2.4.3 295031 G2.4.3 (Q#57) G2.4.3 Ability to identify post-accident instrumentation is already sampled for Q#70. Replaced G2.4.3 with randomly selected G2.4.2.
RO T2/G1 223002 A3.03 223002 A3.02 (Q#6) Replaced 223002 A3.03 with 223002 A3.02 due to low operational relevance. CNS Operators are required to verify PCIS status from control panels and not use SPDS to verify isolation status.
SRO T2/G1 259002 G2.4.30 259002 G2.4.31 (Q#76) Could not develop question at SRO level for K/A 2.4.30 due to no reporting requirements directly related to RVLCS. Obtained replacement K/A 259002 G2.4.31 from NRC.
ES-401 Record of Rejected K/As Form ES-401-4 Rev 4 Revision statement:
Rev 1 For Q#36, replaced original K/A 295037 EA1.08 related to RC&IS with 295037 EA1.07 related to RMCS, since CNS does not have RC&IS, and documented on ES-401-4. Tier totals were not affected by this change.
For Q#3, corrected K/A 211000 from A3.03 to A2.03, since A2 category was indicated by the first X in A2 column. Tier totals were not affected by this change.
For Q#20, replaced original K/A 212000 A2.17 related to MSL high radiation with 212000 A2.11 related to MSIV closure, since MSL radiation is not an input to RPS at CNS, and documented on ES-401-4. Tier totals were not affected by this change.
For Q#4, replaced original K/A 239002 A1.02 related to acoustical monitoring noise with A1.06 related to reactor power, since CNS does not have acoustical monitors, and documented on ES-401-4. Tier totals were not affected by this change.
For Q#23, corrected K/A 263000 from A3.01 to A4.01, since A4 category was indicated by X in A4 column. Tier totals were not affected by this change.
For Q#63, corrected K/A 271000 from A3.02 to A4.02, since A4 category was indicated by X in A4 column. Tier totals were not affected by this change.
For Q#83, corrected description and IR of listed K/A 2.2.12.
Rev 2 The following K/As were requested to be replaced by CNS, and NRC selected the replacement K/As (This is the status of the written exam outlines at the time of draft operating test submittal):
For Q#10, replaced 206000 G2.4.35 with 206000 G2.4.31.
For Q#35, replaced 295017 AK1.03 with 295017 AK2.10. No other AK1 K/As were acceptable for RO testing for this E/APE, so AK2 K/A was chosen. Because this dropped the Tier 1 K1 K/As to less than 2, another T1/G2 System with a K2 K/A was changed to K1 (ref. Q#62).
For Q#62, replaced 295022 AK2.03 with 295022 AK1.02 to make up for loss of one K1 K/A per Q#35 question change, above.
For Q#61, replaced 215001 (TIPS) A2.01 with 216000 (NBI) A2.09.
For Q#72, replaced G2.2.7 with G2.2.14.
For Q#76, replaced 259002 G2.4.9 with 259002 G2.4.30.
For Q#99, replaced G2.4.19 with G2.4.40.
ES-401 Record of Rejected K/As Form ES-401-4 Rev 4 Rev 3 Per NRC direction from draft Op Test outline submittal, for Q#31 replaced K/A 259001 A1.06 with K/A 256000 A1.10 because 259001 was already sampled on the SRO exam.
Per NRC direction from free review, for Q#86 replaced G2.2.40 with G2.1.7 for K/A 300000, since Instrument Air is not a TS system.
For Q#31, per request by CNS, replaced K/A 256000 A1.10 with 256000 A1.07 provided by NRC.
For Q#57, rejected K/A 2.4.3 since it is sampled in Q#70, and replaced it with randomly selected K/A 2.4.2.
For Q#6, rejected K/A 223002 A3.03 with randomly selected K/A 223002 A3.02 due to low operational validity.
Rev 4 For Q#76, per request by CNS, replaced K/A 259002 G2.4.30 with 259002 G2.4.31 provided by NRC.
ES-301 Administrative Topics Outline Form ES-301-1 Rev 2 Facility:
Cooper Nuclear Station Date of Examination:
4/27/2020 Examination Level: RO SRO Operating Test Number:
CN-2020-4 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations R, M A1, modified SKL034-21 Drywell Average Temperature Calculation (Normal Method) IAW 5.8.10 K/A G2.1.23 (4.3)
Conduct of Operations N/A Equipment Control R, N A2, Determine mechanical and electrical isolation boundaries for impeller replacement on REC Pump B K/A G2.2.13 (4.1)
Radiation Control R, P, D From Previous 2 exams randomly selected 3/2017 ILT A4 A3. Determine Workers Projected Total Dose (using survey map) 2.3.14 (3.4)
Emergency Plan R, D A4, Calculate DW release rate IAW Emergency Plan Implementing Procedure 5.7.16 K/A G2.4.39 (3.9)
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1, randomly selected)
ES-301 Administrative Topics Outline Form ES-301-1 Rev 1 Facility:
Cooper Nuclear Station Date of Examination:
4/27/2020 Examination Level: RO SRO Operating Test Number:
CN-2020-4 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations R, P, D A5, From Previous 2 exams randomly selected 9/2018 NRC SA5 Review IST stroke time surveillance for RRMG ventilation dampers (plant OE)
K/A G2.1.20 (4.6)
Conduct of Operations R, N A6, Determine reportability for Control Room Emergency Filtration System BF-C-1A Emergency Booster Fan non-functional in Mode 1. (Group 6 occurs due to inadvertent short (invalid ESF actuation) while I&C working in panel, CREF fan motor fails) (Procedure 2.0.5)
K/A 2.1.18 (3.8), 2.1.20 (4.6)
Equipment Control R, M A7, SKL0345061 (modified to rod 10-11) -
Determine TS Actions for Removal of a Single CRD during Refueling K/A 2.2.40 (4.7)
Radiation Control R, N A8, Determine emergency dose limit, estimated exposure, and whether KI authorization is required for operator to manually close RCIC-MO-15 K/A G2.3.4 (3.7)
Emergency Plan R, D A9, SKL034-30-64 (Emergency Classification)
K/A G2.4.41 (4.6)
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1, randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 page 1 of 2 Rev 3 Facility:
Cooper Nuclear Station Date of Examination:
4/27/2020 Exam Level: RO SRO-I SRO-U Operating Test Number:
CN-2020-4 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function S1. Respond to two rods drifting during 6.CRD.301 (Alt Path - Two control rods will begin drifting during control rod exercise testing, requiring entry into AOP 2.4CRD and manual scram)
K/A 201003 A2.03 (3.4/3.7)
RO, SRO-I S, N, A 1
S2. Restart RWCU following Group 3 isolation IAW SOP 2.2.66 sect 6 (9/2018 ILT NRC JPM S2 randomly selected)
K/A 204000 A4.01 (3.1/3.0)
RO S, P, D 9/2018 NRC ILT S2 2
S3. SKL034-20-29R0, Open MSIVs with RPV pressurized (Alt Path
- Steam leak will occur while warming MSLs, automatic Group 1 isolation fails, requiring manually isolating steam supply IAW procedure 2.0.3, Conduct of Operations)
K/A 239001 A4.01 (4.2/4.0)
RO, SRO-I S, D, A, EN, L 3
S4. SKL034-20-137, Operate RCIC in pressure control (Alt Path -
When RCIC is manually started, the speed controller fails low, requiring securing RCIC to prevent turbine damage IAW a different section of the SOP)
K/A 217000 A4.01 (3.7/3.7)
RO, SRO-I, SRO-U S, D, A, L 4
S5. Perform 5.2FUEL Att. 2 (Actions to minimize rad release)
KA 295036 EK3.04 (3.1/3.4)
RO, SRO-I S, N, L 5
S6. Energize 4160V critical Bus 1G from Emergency Station Service XFMR (ESST) (SKL034-20-43 Rev 0)
K/A 262001 A4.01 (3.4/3.7)
S7. Defeat RPS Logic Trips During Failure-to-Scram (Restoration -
Removal of RPS jumpers) IAW EP 5.8.3 section 5 (SKL034-21-79) 212000 K4.12 (3.9/4.1)
RO, SRO-I S, D 7
S8. Start SGT B for HPCI operation for level control post-scram.
2.2.73 Section 8. (Alt Path - High Moisture will alarm for the preferred SGT filter train due to heater failure, requiring shifting SGT trains per the alarm card)
KA 261000 A4.07 (3.1/3.2)
RO, SRO-I, SRO-U S, N, A, EN, L 9
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 page 2 of 2 Rev 3 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1. SKL034-10-128 Startup the RPS Motor Generator Set (RPS MG A)
KA 212000 A2.01 (3.7/3.9)
RO, SRO-I, SRO-U D, A 7
P2. SKL034-10-65R9, Conduct alternate rod insertion by manually venting scram air header K/A 295037 EA1.05 (3.9/4.0),
K/A 201003 A2.05 (4.1/4.1)
RO, SRO-I, SRO-U D, R, L, E 1
P3. SKL034-10-34 Place the A Air Dryer in Service K/A 295019 AA1.02 (3.3/3.1)
RO, SRO-I D
8 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for R /SRO-I/SRO-U (Actual)
(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 (5/5/3) 9/ 8/ 4 (8/7/4) 1/ 1/ 1 (1/1/1) 1/ 1/ 1 (control room system) (3/3/2) 1/ 1/ 1 (5/5/3) 2/ 2/ 1 (3/3/1) 3/ 3/ 2 (randomly selected) (1/0/0) 1/ 1/ 1 (1/1/1)
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 1 of 41 Page 1 of 41 Rev 2 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CN-2020-04 Examiners: ____________________________ Operators:
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
- 2. Respond to narrow range level transmitter NBI-LT-52C failure
- 3. Respond to Critical Power Panel CPP-1 loss of power
- 4. Respond to Condensate Pump B trip, Reactor Recirc Pump A runs back too far, trips
- 5. Respond to Reactor Recirc Pump B trip with Thermal Hydraulic Instability
- 6. Respond to a Main Steam Line D leak in the drywell
- 7. Respond to DW Spray A permissive switch failure
- 8. Respond to Torus to Drywell Vacuum Breaker failure with failure of RHR Loop B containment sprays Initial Conditions: Plant operating at 100% power near end of cycle.
Inoperable Equipment: Core Spray Pump A is tagged out of service for motor PMs.
Turnover:
Plant operating at operating at 100% power near end of cycle.
Planned activities for this shift are:
Continue steady state power operation.
Continue maintenance on CS Pump A.
Scenario Notes:
This is a new scenario.
Validation Time: 60 minutes
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 2 of 41 Page 2 of 41 Rev 2 Event No.
Malf. No.
Event Type Event Description 1
Shift CRD Pumps (B to A) 2 rr27c TS (CRS)
Narrow range level transmitter NBI-LT-52C failure 3
A (CREW)
Critical power panel CPP-1 loss of power 4
A (CREW)
TS (CRS)
Condensate Pump B trip, RR Pump A runs back too far, Reactor Recirc Pump A trip 5
A (CREW)
Reactor Recirc Pump B shaft binding with Thermal Hydraulic Instability CT#1 Manually scram the reactor when both recirculation pumps trip, prior to exceeding 25% peak to peak neutron flux oscillations.
6 ms02d M (CREW)
Main Steam Line D leak in the drywell 7
(Override) zdirhrsws17a C (BOP,CRS)
DW Spray A permissive switch failure 8
pco2a rh01b rh08d C
Torus to Drywell Vacuum Breaker failure, RHR Pumps B and D trip CT#2 When torus pressure cannot be maintained below the Pressure Suppression Pressure (PSP)(approximately 25 psig in this scenario),
the crew completes Emergency Depressurization prior to exceeding PCPL-A (62.7 psig). (Anticipating ED and fully opening Bypass valves satisfies this Critical Task)
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D)
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 3 of 41 Page 3 of 41 Rev 2 Quantitative Attributes Table Attribute ES-301-4 Target Actual Description Malfunctions after EOP entry 1-2 2
- 1.
DW Spray A permissive switch failure
- 2.
Torus to Drywell Vacuum Breaker failure, RHR Pumps B and D trip Abnormal Events 2-4 3
- 1. Critical power panel CPP-1 loss of power
- 2. Condensate Pump B trip, RR Pump A runs back too far, Reactor Recirc Pump A trip
- 3. Reactor Recirc Pump B shaft binding with Thermal Hydraulic Instability Major Transients 1-2 1
- 1. Main Steam Line D leak in the drywell EOP entries requiring substantive action 1-2 2
- 1. EOP-1A
- 1. EOP-2A Contingency #2 - Emergency Depressurization Pre-identified Critical Tasks 2
2
- 1. (CT#1) Manually scram the reactor when both recirculation pumps trip, prior to exceeding 25% peak to peak neutron flux oscillations.
- 2. (CT#2) When torus pressure cannot be maintained below the Pressure Suppression Pressure (PSP)(approximately 25 psig in this scenario), the crew completes Emergency Depressurization prior to exceeding PCPL-A (62.7 psig).
(Anticipating ED and fully opening Bypass valves satisfies this Critical Task)
Normal Events N/A 1
- 1. Shift CRD pumps Reactivity Manipulations N/A 0
- 1. none Instrument/
Component Failures N/A 5
- 1. Critical power panel CPP-1 loss of power
- 2. Condensate Pump B trip, RR Pump A runs back too far, Reactor Recirc Pump A trip
- 3. Reactor Recirc Pump B shaft binding with Thermal Hydraulic Instability
- 4. DW Spray A permissive switch failure
- 5. Torus to Drywell Vacuum Breaker failure, RHR Pumps B and D trip Total Malfunctions N/A 5
- 1. Critical power panel CPP-1 loss of power
- 2. Condensate Pump B trip, RR Pump A runs back too far, Reactor Recirc Pump A trip
- 3. Reactor Recirc Pump B shaft binding with Thermal Hydraulic Instability
- 4. DW Spray A permissive switch failure
- 5. Torus to Drywell Vacuum Breaker failure, RHR Pumps B and D trip
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 4 of 41 Page 4 of 41 Rev 2 Top 10 systems and operator actions important to risk that are tested:
Primary Containment, RHR, RHRSW
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 5 of 41 Page 5 of 41 Rev 2 SCENARIO
SUMMARY
The plant is operating at 100% power near the end of the operating cycle.
After the crew takes the watch, the ATC will shift CRD pumps from CRD Pump B in operation to CRD Pump A in operation IAW Procedure 2.2.8.
After CRD pumps have been shifted, narrow range level transmitter NBI-LT-52C fails high, resulting in a half Main Turbine/RFPT trip. The crew will respond IAW the alarm card and Procedure 4.4.1 to diagnose the failure using the RVLCS HMI. The CRS will enter TS 3.3.2.2 Condition A.
After TS for the failed level transmitter have been addressed, 120V AC power panel CPP-1 will lose power. The crew will respond IAW Procedure 5.3AC120.
The BOP will start a turbine building exhaust fan to restore TB HVAC.
After response to loss of CPP-1 is complete, Main Condensate Pump 1B will trip. The crew will respond to CP B trip and enter 2.4MC-RF to address the tripped MC Pump B. At this power level the Reactor Recirc pumps should run back towards 45% speed and stop running back once Condensate pump discharge header pressure restores, which should occur approximately 50%
Reactor Recirc pump speed. However, Reactor Recirc Pump A will continue to runback for several seconds. If the crew identifies RR Pump A continues to run back, they will lockout the RR A scoop tube. This will require entry into 2.4RR.
After a delay RR Pump A will trip. The CRS will enter TS 3.4.1 Condition A &
B due to single loop operation.
After the TS for RR Pump A failure have been addressed, RR Pump B will trip.
The ATC will respond by inserting a manual scram IAW procedure 2.4RR due to no RR Pumps operating and power above 1%. If a manual scram is delayed, thermal hydraulic instability will occur, also requiring a manual scram per 2.4RR (CT#1).
When the reactor is scrammed, a leak develops on MSL D in the drywell.
Drywell pressure and temperature rise. The crew will enter EOP-1A and EOP-3A due to high drywell pressure. Torus Spray will be required. Only RHR Loop B is available for containment spray due to failure of the spray valve control permissive for RHR Loop A. A Torus to Drywell Vacuum Breaker will fail open, requiring use of Drywell Spray.
Drywell Spray will fail a short time after it has been started due to trip of RHR Pumps B and D. Drywell pressure will rise above the Pressure Suppression Pressure (PSP) limit. Emergency Depressurization will be required (CT#2).
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 6 of 41 Page 6 of 41 Rev 2 The exercise ends when the reactor has been depressurized, RPV water level is being controlled +3 to +54 inches.
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 7 of 41 Page 7 of 41 Rev 2 CRITICAL TASK BASIS Critical Task #1 Manually scram the reactor when both recirculation pumps trip, prior to exceeding 25% peak to peak neutron flux oscillations.
Safety Significance Analyses of neutronic/thermal-hydraulic instabilities(THI) during failure-to-scram conditions have been performed. Instabilities are manifested by oscillations in reactor power, which, if the reactor cannot be shut down, may increase in magnitude. If the oscillations remain small or moderately sized, they tend to repeat on approximately a two second period. Under certain circumstances, however, the oscillations may continue to grow and become sufficiently large and irregular to cause localized fuel damage.
Analytical results indicate that the fuel clad may experience boiling transition during THI but that it subsequently rewets and is adequately cooled even for oscillations that resemble reactivity excursion events. For an occasional large pulse, however, rewetting of some of the highest-powered locations within the highest-powered fuel bundles may not occur; the clad could then continue to heat up over several oscillation cycles.
Cues SRM period alarms.
Oscillating power indications on neutron monitoring instrument.
Measurable Performance Indicators Operator depresses both manual scram pushbuttons, or places the Reactor Mode Switch to SHUTDOWN on panel 9-5.
Performance Feedback On panel 9-5:
RPS status lights de-energized.
Reactor power level trend.
Control rods inserted.
Applicability Operation in the Stability Exclusion Region with no Recirc pump in operation.
Justification for the chosen performance limit The threshold of 25% peak-to-peak neutron flux oscillations has been chosen to ensure an attempt to shut down the reactor using RPS for the same conditions that require boron injection during failure to scram events. Analysis has shown that APRM peak-to-peak amplitudes reach 25% of rated thermal power well before any individual LPRM signal reaches an amplitude for which fuel damage might be possible. Requiring manual reactor shutdown before 25%
peak-to-peak neutron flux oscillations is reasonable to allow a short time to recognize and diagnose THI, communicate plant conditions, and effect the manual scram.
BWR Owners Group Appendix App. B, step RC/Q-6
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 8 of 41 Page 8 of 41 Rev 2 Scenario Guide Requirements Initial power level following the recirc pump trips, with RPV level in the normal band, should be in the Stability Exclusion Region. THI malfunction CR04A or CR04B must be inserted at a ramped severity that will eventually cause 25% peak-to-peak neutron flux oscillations, but at long enough delay and/or ramp rate to allow the crew to recognize and diagnose THI, then communicate and insert a manual scram before 25% peak-to-peak neutron flux oscillations are reached.
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 9 of 41 Page 9 of 41 Rev 2 Critical Task #2 When torus pressure cannot be maintained below the Pressure Suppression Pressure (PSP)(approximately 25 psig in this scenario), the crew completes Emergency Depressurization prior to exceeding PCPL-A (62.7 psig). (Anticipating ED and fully opening Bypass valves satisfies this Critical Task)
Safety Significance If suppression pool and/or drywell sprays cannot be initiated or are ineffective in reversing the increasing trend of primary containment pressure, as evidenced by not being able to maintain torus pressure below the Pressure Suppression Pressure, the RPV is depressurized to minimize further release of energy from the RPV to the primary containment. This action serves to terminate or reduce as much as possible any continued primary containment pressure increase.
The Pressure Suppression Pressure (PSP) is the lesser of:
- The highest torus pressure which can occur without steam in the torus air space.
- The highest torus pressure at which initiation of RPV depressurization will not result in exceeding Primary Containment Pressure Limit before RPV pressure drops to the Minimum RPV Flooding Pressure.
- The highest torus pressure which can be maintained without exceeding the suppression pool boundary design load if SRVs are opened.
The Primary Containment Pressure Limit (PCPL) is the lesser of:
- The pressure capability of the primary containment.
- The maximum primary containment pressure at which vent valves sized to reject all decay heat from the containment can be opened and closed.
- The maximum primary containment pressure at which SRVs can be opened and will remain open.
The PCPL is a function of primary containment water level.
Exceeding the limit may challenge primary containment vent valve operability, SRV operability, or the structural integrity of the primary containment.
Cues Rising torus pressure indicated on SPDS and panel 9-3 recorder PC-LRPR-1A.
Cursor approaching unsafe boundary on PSP graph display on SPDS.
Measurable Performance Indicators Manipulation of any six SRV controls on panel 9-3:
SRV-71A SRV-71B SRV-71E SRV-71G SRV-71H
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 1 Page 10 of 41 Page 10 of 41 Rev 2 SRV-71C SRV-71D SRV-71F If ED is anticipated, fully opening Bypass valves on Panel B DEH HMI Bypass valves selected to MANUAL on HMI Bypass valves A, B, C indicate ~100% on HMI mimic Performance Feedback Crew will observe SRV light indication go from green to red, amber pressure switch lights illuminate, reactor pressure lowering on SPDS and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.
If ED is anticipated, fully opening Bypass valves on Panel B DEH HMI, crew will observe Bypass valves A, B, C in MANUAL at
~100% position and reactor pressure lowering on SPDS and panel 9-3 and 9-5 meters and recorders.
Applicability EOP-3A conditions with energy being added to primary containment (LOCA) and torus/drywell pressure rising.
Justification for the chosen performance limit When torus pressure cannot be maintained below PSP is the EOP-3A, step PC/P-4 criteria requiring transition to emergency depressurization. Failure to ED early enough to avert exceeding PCPL-A could result in failure of the containment barrier.
Depressuring to the Decay Heat Removal Pressure (50 psig) before PCPL-A is exceeded ensures energy discharged to the primary containment will be within the capacity of the containment vent.
BWR Owners Group Appendix App. B, step PC/P-2.
Scenario Guide Requirements A leak with failure of torus/drywell sprays and failure of a Torus to Drywell vacuum Breaker is normally required to effect exceeding PSP. It is important to design the scenario such that the crew has information early during the LOCA event to determine the rising DW pressure cannot be mitigated and that PSP will be exceeded.
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 1 of 49 Rev 3 Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: CN-2020-04 Examiners: ____________________________ Operators:
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
- 1. Raise reactor power using Reactor Recirculation flow
- 2. Respond to a control rod drifting out
- 3. Respond to Service Water Pump C trip
- 4. Respond to Circulating Water pump inlet screen clogging and partial loss of CW pumps, resulting in degraded condenser vacuum
- 5. Respond to complete loss of Circulating Water pumps, loss of condenser vacuum
- 6. Respond to hydraulic block ATWS
- 7. Respond to failure of Reactor Recirc Pump B to trip on ATWS/RPT
- 8. Respond to SLC B relief valve failing open
- 9. Respond to trip of RHRSWB pump operating for Suppression Pool Cooling Initial Conditions: Plant operating at 97% power following a sequence exchange near middle of cycle.
Inoperable Equipment: Core Spray Pump A is tagged out of service for motor PMs.
Turnover:
Plant operating at 97% power following a sequence exchange near middle of cycle.
Core Spray Pump A is tagged out of service for motor PMs.
Planned activities for this shift are:
Raise power to 100% IAW Procedure 2.1.10. There are NO preconditioning holds.
Continue maintenance on CS Pump A.
Scenario Notes:
This is a new scenario.
Validation Time: 60 minutes
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 2 of 49 Rev 3 Event No.
Malf. No.
Event Type Event Description 1
Raise reactor power using Reactor Recirculation flow 2
rd101819 C
A (CREW)
TS (CRS)
Control rod 18-19 drifts out CT#1 When an inadvertent reactivity addition occurs (in this case, rod drift out) that would cause power to rise above 100% rated thermal power, crew lowers power to restore and maintain power 100% rated thermal power by inserting the control rod and/or lowering Reactor Recirc flow before PMIS point NSSRP641 (30 minute power average) exceeds 2419 MWt.
TS (CRS)
Service Water Pump C trip 4
mc06a mc05a mc03a,b,c,d C (ATC,CRS)
A (CREW)
Circulating Water pump inlet screen clogging and partial loss of CW pumps, resulting in degraded condenser vacuum 5
mc05b,c,d mc01 rd02a rd02b (override) zdiarsw150mv M (CREW)
Complete loss of Circulating Water pumps, loss of condenser vacuum, hydraulic block ATWS CT#2 When control rods fail to scram and energy is discharging to the primary containment (e.g.
SRVs, LOCA), crew initiates SLC injection or inserts all control rods to at least position 02 before exceeding the Boron Injection Initiation Temperature (BIIT) curve.
Failure of Reactor Recirc Pump B to trip on ATWS/RPT 7
SLC B relief valve fails open 8
sw04a(b, c, or d)
First RHRSWB pump started for Suppression Pool Cooling trips (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D)
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 3 of 49 Rev 3 Quantitative Attributes Table Attribute ES-301-4 Target Actual Description Malfunctions after EOP entry 1-2 3
- 1.
Failure of Reactor Recirc Pump B to trip on ATWS/RPT
- 2.
SLC B relief valve fails open
- 3.
First RHRSWB pump started for Suppression Pool Cooling trips Abnormal Events 2-4 2
- 1. Control rod 18-19 drifts out
- 2. Circulating Water pump inlet screen clogging and partial loss of CW pumps, resulting in degraded condenser vacuum Major Transients 1-2 1
- 1. Complete loss of Circulating Water pumps, loss of condenser vacuum, hydraulic block ATWS EOP entries requiring substantive action 1-2 2
- 1. EOP-3A
- 1. EOP-7A Contingency #5 - Level/Power Control Pre-identified Critical Tasks 2
2
- 1. (CT#1) When an inadvertent reactivity addition occurs (in this case, rod drift out) that would cause power to rise above 100% rated thermal power, crew lowers power to restore and maintain power 100% rated thermal power by inserting the control rod and/or lowering Reactor Recirc flow before PMIS point NSSRP641 (30 minute power average) exceeds 2419 MWt.
- 2. (CT#2) When control rods fail to scram and energy is discharging to the primary containment (e.g. SRVs, LOCA), crew initiates SLC injection or inserts all control rods to at least position 02 before exceeding the Boron Injection Initiation Temperature (BIIT) curve.
Normal Events N/A 1
- 1. none Reactivity Manipulations N/A 1
- 1. Raise reactor power using Reactor Recirculation flow Instrument/
Component Failures N/A 6
- 1. Control rod 18-19 drifts out
- 2. Service Water Pump C trip
- 3. Circulating Water pump inlet screen clogging and partial loss of CW pumps, resulting in degraded condenser vacuum
- 4. Failure of Reactor Recirc Pump B to trip on ATWS/RPT
- 5. SLC B relief valve fails open
- 6. First RHRSWB pump started for Suppression Pool Cooling trips
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 4 of 49 Rev 3 Total Malfunctions N/A 6
- 1. Control rod 18-19 drifts out
- 2. Service Water Pump C trip
- 3. Circulating Water pump inlet screen clogging and partial loss of CW pumps, resulting in degraded condenser vacuum
- 4. Failure of Reactor Recirc Pump B to trip on ATWS/RPT
- 5. SLC B relief valve fails open
- 6. First RHRSWB pump started for Suppression Pool Cooling trips Top 10 systems and operator actions important to risk that are tested:
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 5 of 49 Rev 3 SCENARIO
SUMMARY
The plant is operating at 97% power during power ascension.
After the crew takes the watch, the crew raises power to 100% using Reactor Recirculation flow control.
After the crew has raised power, control rod 18-19 drifts out. The crew responds IAW Procedure 2.4CRD and inserts rod 18-19, but the rod does not latch. The crew scrams rod 18-19 IAW Procedure 2.4CRD and it then latches.
The crew must insert rod 18-19 and/or lower Reactor Recirc flow to prevent the 30-minute power average from exceeding the License Condition limit (CT#1).
The CRS enters TS 3.1.3 Condition C for rod 18-19.
After the TS for rod 18-19 has been addressed, Service Water Pump C will trip.
The BOP will start SW Pump B IAW the alarm card. The CRS will enter TS 3.7.2 Condition A for SW Pump C.
After the TS for SW Pump C has been addressed, Circulating Water travelling screens will begin to block due to river debris. After a delay, CW Pump A will trip, followed later by CW Pump B. Condenser vacuum will lower. The crew will respond IAW Procedure 2.4VAC and lower power to stabilize condenser vacuum.
After power has been lowered and vacuum stabilized, CW Pumps C and D will trip due to travelling screen failure. The crew will insert a manual scram IAW Procedure 2.4VAC when vacuum cannot be maintained above 23 Hg.
When the reactor is scrammed, control rods will fail to insert due to blockages in both scram discharge volumes. Reactor power will be approximately 30%.
EOP-6A and 7A are entered via EOP-1A. Reactor Recirc Pump B to trip on ATWS/RPT requiring the operator to manually tip the Recirc Pump. The crew injects SLC and installs the necessary PTMs to bypass interlocks and insert control rods individually via RMCS. MSIVs will close on low condenser vacuum. With MSIVs closed and SRV operating to control reactor pressure, either SLC must be initiated or all controls rods inserted to at least position 02 before Suppression Pool temperature exceeds the Boron Injection Initiation Temperature (BIIT) curve (CT#2). SLC B relief valve will fail open when SLC Pump B is started.
Stop and Prevent is required because reactor power is above 3%. RPV level is intentionally lowered below -60 inches wide range in order to lower core inlet subcooling and lower reactor power. ADS is inhibited when EOP-7A is entered
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 6 of 49 Rev 3 to avert uncontrolled depressurization due to lowering level. HPCI and RCIC are available for RPV level control.
Suppression Pool Cooling will be required due to SRV operation. The RHRSWB pump associated with the first loop of SPC attempted to be placed into service will trip, requiring the operator to start the other SWB pump in that RHR loop or transition to the other RHR loop for SPC.
Once several control rods have been inserted, the ATC begins alternately resetting RPS, driving rods individually while allowing the SDV to drain, and reinserts a manual scram. All control rods will fully insert upon the scram. The CRS will transition from ATWS to non-ATWS flowcharts, SLC pumps are stopped, and RPV level restoration is directed.
The exercise ends when control rods are inserted or Hot Shutdown Boron weight has been injected, and the CRS has reset the level band to +3 to +54.
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 7 of 49 Rev 3 CRITICAL TASK BASIS Critical Task #1 When an inadvertent reactivity addition occurs (in this case, rod drift out) that would cause power to rise above 100% rated thermal power, crew lowers power to restore and maintain power 100%
rated thermal power by inserting the control rod and/or lowering Reactor Recirc flow before PMIS point NSSRP641 (30 minute power average) exceeds 2419 MWt.
Safety Significance License Condition C.1 states 2419 MWt is the Maximum Power Level authorized. The licensed 100% power limit is a basis for assumptions in the plants safety analysis. Sustained operation above 2419 MWt may place unit operation outside of the plant design basis. NUREG 1021 App. D, section D states a CT must be essential to safety, and lists actions to for which operation or correct performance prevents violation of a facility license condition as one example of a CT.
Cues Annunciators 9-5-1/C-4, Rod Drift, Red drift LED lit for rod 18-19 on full core display on panel 9-5, When selected, control rod 18-19 position rising on panel 9-5, Reactor power rising indicated on IRM/APRM recorders NM-NR-46A-D, SPDS, PMIS.
Measurable Performance Indicators Operator selects rod 18-19 and places Emergency Notch Override switch to EMER ROD IN on panel 9-5, and/or Operator Selects S on RR flow controllers RRFC-SIC-16A(B) on panel 9-4 and lowers RR pump flow (by turning speed demand counter-clockwise on one speed controller at a time) until power stabilizes below 100% on IRM/APRM recorders NM-NR-46A-D.
Performance Feedback Reactor power stabilizes below 100% on IRM/APRM recorders NM-NR-46A-D and on PMIS points NSSRP640, NSSRP641, NSSRP642, NSSRP643, and NSSRP645.
Applicability Any time a rod drift out would cause reactor power to exceed 100% with no operator intervention.
Justification for the chosen performance limit License Condition C.1 lists 2419 MWt as the Maximum Power Level for CNS. Procedure 2.1.10, Station Power Changes, sections 10 and 11 describes the methodology for adherence to this limit. A note at step 11.1 states Minor power fluctuations due to automatic control system response, random processes such as bi-stable flow, and flow meter measurement uncertainties are inherent to BWR operating characteristics. Small, short-term fluctuations in power that are not under the direct control of a Licensed Reactor Operator are not considered intentional. Step 11.1 states It is prohibited to intentionally operate greater than the applicable licensed power limit as determined in Section 10. If core thermal power 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average calculation exceeds applicable licensed power limit, action shall be taken to ensure subsequent hourly average remains less than or equal to applicable limit. Step 11.4 directs monitoring and maintaining PMIS Point NSSRP643 (running 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average) and PMIS Point NSSRP645 (running 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average) below 2419 MWt. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> point is the legal
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 8 of 49 Rev 3 record for plant power level. PMIS point NSSRP641 (30 minute average) was chosen because it is more limiting and provides timely update to maintain adequate margin to the core thermal power limit.
BWR Owners Group Appendix N/A Scenario Guide Requirements Initial power level must be near 100% and the rod drift out must cause reactor power to rise above 100% with no operator intervention.
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 9 of 49 Rev 3 Category EOPS Critical Task #2 When control rods fail to scram and energy is discharging to the primary containment (e.g. SRVs, LOCA), crew initiates SLC injection or inserts all control rods to at least position 02 before exceeding the Boron Injection Initiation Temperature (BIIT) curve.
Safety Significance Failure to effect shutdown of the reactor when a RPS setting has been exceeded would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. Action to shut down the reactor is required when RPS and control rod drive systems fail.
The Boron Injection Initiation Temperature (BIIT) is the greater of:
- The highest suppression pool temperature at which initiation of boron injection will permit injection of the Hot Shutdown Boron Weight of boron before suppression pool temperature exceeds the Heat Capacity Temperature Limit.
- The suppression pool temperature at which a reactor scram is required by plant Technical Specifications.
The BIIT is a function of reactor power. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Since failure-to-scram conditions may present severe plant safety consequences, the requirement to initiate boron injection is independent of any anticipated success of control rod insertion. When attempts to insert control rods satisfactorily achieve reactor shutdown, the requirement for boron injection no longer exists. (Control rod insertion is directed under Step RC/Q-7 concurrently with Step RC/Q-6.)
Cues Manual scram is initiated and numerous control rods indicate beyond position 00 and reactor power not downscale on panel 9-5 indications.
Suppression Pool temperature rising on PMIS and panel indications.
Measurable Performance Indicators Operator manipulates keylocked switch for SLC A pump to START on panel 9-5. (SLC Pump B relief valve fails open in this scenario.)
Operator selects individual control rods by depressing the respective pushbutton on the panel 9-5 matrix and inserts the rod by manipulating the emergency in switch on panel 9-5.
Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-4 Scenario 3 Page 10 of 49 Rev 3 Performance Feedback SLC A pump red light illuminated, SLC discharge pressure rising, SLC tank level lowering by 26% on panel 9-5 (from ~80% to below
~54% for this scenario).
Operator selecting and inserting control rods indicated by rod position decreasing to 00 for selected rod on panel 9-5.
Applicability ATWS with power >3% following trip of both recirc pumps per EOP-7A, energy being discharged to Primary Containment causing Torus water temperature to rise.
Justification for the chosen performance limit If boron injection is initiated or all control rods are inserted to position 02 before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Since failure-to-scram conditions may present severe plant safety consequences, the requirement to initiate boron injection is independent of any anticipated success of control rod insertion.
If the failure to scram EOP were to be exited, other procedures would not provide the guidance for control rod insertion necessary to achieve reactor shutdown. Before exiting EOP-6A ensures guidance to effect reactor shutdown is not removed.
BWR Owners Group Appendix App. B, step RC/Q-6 Scenario Guide Requirements Initial conditions, combined with the ATWS severity, should result in power >3% following trip of both recirc pumps per EOP-7A.
Suppression Pool temperature must be rising due to unstoppable condition such as loss of the main condenser or LOCA. The scenario should be validated to exceed BIIT; therefore, ability to achieve control rod insertion may need to be hampered or delayed.