ML16280A460
| ML16280A460 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 09/30/2016 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | |
| References | |
| Download: ML16280A460 (38) | |
Text
ES-401 BWR Examination Outline Form ES-401-1 Facility: River Bend Station Date of Exam: 2016 Tier Group RO K/A Category Points SRO-Only Points K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total
- 1. 1 3 3 4 4 3 3 20 4 3 7 Emergency & N/A N/A 2 1 1 2 1 1 1 7 1 2 3 Abnormal Plant Evolutions Tier Totals 4 4 6 5 4 4 27 5 5 10 1 2 3 3 3 3 2 2 2 2 2 2 26 2 3 5 2.
Plant 2 1 1 1 1 1 1 1 1 1 1 2 12 0 1 2 3 Systems Tier Totals 3 4 4 4 4 3 3 3 3 3 4 38 4 4 8
- 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 3 2 2 2 1 2 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 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). (One Tier 3 Radiation Control K/A is allowed if the K/A 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 associated 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 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 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
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 / SRO)
E/APE # / Name / Safety Function K K K A A2 G* K/A Topic(s) IR #
1 2 3 1 Knowledge of the reasons for the following responses as they apply to partial or complete loss 3.7 14 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X of forced core flow circulation: (CFR: 41.5 / 45.6)
AK3.02 Reactor power response Ability to operate and/or monitor the following as they apply to partial or complete loss of A.C. power: 3.6 4 295003 Partial or Complete Loss of AC / 6 X (CFR: 41.7 / 45.6)
AA1.04 D.C. electrical distribution system Ability to determine and/or interpret the following as they apply to partial or complete loss of D.C. power: 3.2 72 295004 Partial or Total Loss of DC Pwr / 6 (CFR: 41.10 / 43.5 / 45.13)
X AA2.01 Cause of partial or complete loss of D.C.
power 2.1.19 Ability to use plant computers to evaluate 3.9 9 295005 Main Turbine Generator Trip / 3 X system or component status. (CFR: 41.10 / 45.12)
Knowledge of the operational implications of the following concepts as they apply to SCRAM: (CFR: 3.7 26 295006 SCRAM / 1 X 41.8 to 41.10)
AK1.01 Decay heat generation and removal Knowledge of the interrelations between control room abandonment and the following: (CFR: 41.7 / 4.0 36 295016 Control Room Abandonment / 7 X 45.8)
AK2.02 Local control stations: Plant-Specific Knowledge of the reasons for the following responses as they apply to partial or complete loss 3.3 16 295018 Partial or Total Loss of CCW / 8 X of component cooling water: (CFR: 41.5 / 45.6)
AK3.04 Starting standby pump Ability to operate and/or monitor the following as they apply to partial or complete loss of instrument 3.5 17 295019 Partial or Total Loss of Inst. Air / 8 X air: (CFR: 41.7 / 45.6)
AA1.01 Backup air supply 295021 Loss of Shutdown Cooling / 4 Ability to determine and/or interpret the following as they apply to refueling accidents: (CFR: 41.10 / 43.5 3.4 47 295023 Refueling Acc / 8 X / 45.13)
AA2.02 Fuel Pool Level 2.2.44 Ability to interpret control room indications to verify status and operation of a system, and 4.2 51 295024 High Drywell Pressure / 5 understand how operator actions and directives X affect plant and system conditions. (CFR: 41.5 / 43.5
/ 45.12)
Knowledge of the operational implications of the following concepts as they apply to high reactor 3.9 65 295025 High Reactor Pressure / 3 X pressure: (CFR: 41.8 to 41.10)
EK1.01 Pressure effects on reactor power Knowledge of the interrelations between suppression pool high water temperature and the 3.0 5 295026 Suppression Pool High Water Temp.
/5 X following: (CFR: 41.7 / 45.8)
EK2.05 Containment pressure: Mark-III Knowledge of the reasons for the following responses as they apply to high containment 3.7 28 295027 High Containment Temperature / 5 temperature (Mark III containment only): (CFR: 41.5 X / 45.6)
EK3.01 Emergency depressurization: Mark-III Ability to operate and/or monitor the following as they apply to high drywell temperature: (CFR: 41.7 / 3.9 40 295028 High Drywell Temperature / 5 X 45.6)
EA1.02 Drywell ventilation system
ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)
E/APE # / Name / Safety Function K K K A A2 G* K/A Topic(s) IR #
1 2 3 1 Ability to determine and/or interpret the following as they apply to low suppression pool water level: 3.7 46 295030 Low Suppression Pool Wtr Lvl / 5 X (CFR: 41.10 / 43.5 / 45.13)
EA2.03 Reactor pressure 2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 3.6 19 295031 Reactor Low Water Level / 2 X 43.5 / 45.12)
Knowledge of the operational implications of the following concepts as they apply to SCRAM 3.4 12 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or condition present and reactor power above APRM X downscale or unknown: (CFR: 41.8 to 41.10)
Unknown / 1 EK1.07 Shutdown margin Knowledge of the interrelations between high off-site release rate and the following: (CFR: 41.7 / 45.8) 3.7 75 295038 High Off-site Release Rate / 9 X EK2.05 Site emergency plan Knowledge of the reasons for the following responses as they apply to plant fire on site: (CFR 2.8 62 600000 Plant Fire On Site / 8 41.5,41.10 / 45.6 / 45.13)
X AK3.04 Actions contained in the abnormal procedure for plant fire on site Ability to operate and/or monitor the following as they apply to generator voltage and electric grid 3.8 6 700000 Generator Voltage and Electric Grid Disturbances / 6 disturbances: (CFR: 41.5 and 41.10 / 45.5, 45.7, X and 45.8 )
AA1.02 Turbine/generator controls Ability to determine and/or interpret the following as they apply to Suppression pool high water temp: 4.2 92 295026 Suppression Pool High Water Temp X
/5 (CFR:43.2) EA2.01 SP water temp 2.2.40 Ability to apply Technical Specifications for a system. (CFR:43.5) 4.7 77 295004 Partial or Total Loss of DC Pwr / 6 X Ability to determine and/or interpret the following as they apply to main turbine generator trip: (CFR:43.5) 3.9 83 295005 Main Turbine Generator Trip / 3 X AA2.05 Reactor Power 2.4.11 Knowledge of abnormal condition procedures (43.5) 4.2 84 295016 Control Room Abandonment / 7 X Ability to determine and/or interpret the following as they apply to loss of shutdown cooling: (CFR: 43.5) 3.4 80 295021 Loss of Shutdown Cooling / 4 X AA2.02 RHR/shutdown cooling system flow 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the 4.2 78 295024 High Drywell Pressure / 5 X status of limiting conditions for operations.
(CFR:43.2)
Ability to determine and/or interpret the following as they apply to high drywell temperature: (CFR:43.5) 4.1 90 295028 High Drywell Temperature / 5 X A2.01 Drywell temp K/A Category Totals: 3 3 4 4 3/4 3/3 Group Point Total: 20/7
ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 Knowledge of the reasons for the following 3.4 24 295002 Loss of Main Condenser Vac / 3 responses as they apply to loss of main condenser X vacuum: (CFR: 41.5 / 45.6)
AK3.05 Main steam isolation valve: Plant-Specific Ability to operate and/or monitor the following as 3.9 73 295007 High Reactor Pressure / 3 they apply to high reactor pressure: (CFR: 41.7 /
X 45.6)
AA1.04 Safety/relief valve operation: Plant-Specific 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 Ability to determine and/or interpret the following 3.9 13 295012 High Drywell Temperature / 5 as they apply to high drywell temperature: (CFR:
X 41.10 / 43.5 / 45.13)
AA2.02 Drywell pressure 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 2.4.3 Ability to identify post-accident 3.7 59 295015 Incomplete SCRAM / 1 X instrumentation.
Knowledge of the operational implications of the 3.8 35 295017 High Off-site Release Rate / 9 following concepts as they apply to high off-site X release rate: (CFR: 41.8) AK1.02 Protection of general public 295020 Inadvertent Cont. Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 Knowledge of the interrelations between high 2.6 15 295029 High Suppression Pool Wtr Lvl / 5 suppression pool water level and the following:
X (CFR: 41.7 / 45.8)
EK2.08 Drywell/suppression chamber ventilation Knowledge of the reasons for the following 3.8 2 295032 High Secondary Containment Area responses as they apply to high secondary Temperature / 5 X containment area temperature: (CFR: 41.5 / 45.6)
EK3.03 Isolating affected systems 295033 High Secondary Containment Area Radiation Levels / 9 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 2.1.20 Ability to interpret and execute procedure 4.6 98 295020 Inadvertent Cont. Isolation / 5 & 7 X steps. (CFR: 41.10 / 43.5 / 45.12)
Ability to determine and/or interpret the following 4.2 99 295033 High Secondary Containment Area X as they apply to high secondary containment area Radiation Levels / 9 radiation levels: (CFR: 41.10 / 43.5 / 45.13)
EA2.03 Cause of high area radiation 2.4.6 Knowledge of EOP mitigation strategies. 4.7 93 500000 High CTMT Hydrogen Conc. / 5 X (CFR: 41.10 / 43.5 / 45.13)
K/A Category Point Totals: 1 1 2 1 1/1 1/2 Group Point Total: 7/3
ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)
System # / Name K K K K K K A A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 Knowledge of electrical power supplies to 3.5 67 203000 RHR/LPCI: Injection X the following: (CFR: 41.7)
Mode K2.01 Pumps Knowledge of the effect that a loss or 4.6 44 malfunction of the RHR/LPCI: injection 203000 RHR/LPCI: Injection X mode (plant specific) will have on following:
Mode (CFR: 41.7 / 45.4)
K3.04 Adequate core cooling Knowledge of shutdown cooling system 2.6 64 (RHR shutdown cooling mode) design 205000 Shutdown Cooling X feature(s) and/or interlocks which provide for the following: (CFR: 41.7)
K4.04 Adequate pump NPSH K5. Knowledge of the operational 2.8 18 implications of the following concepts as 209001 LPCS X they apply to LOW PRESSURE CORE SPRAY SYSTEM: (CFR: 41.5 / 45.3)
K5.04 Heat removal (transfer) mechanisms Knowledge of the effect that a loss or 3.4 32 malfunction of the following will have on the 209002 HPCS X HIGH PRESSURE CORE SPRAY SYSTEM (HPCS): (CFR: 41.7 / 45.7)
K6.02 Condensate Storage Tank Level Ability to predict and/or monitor changes in 4.0 29 parameters associated with operating the 211000 SLC X STANDBY LIQUID CONTROL SYSTEM controls including: (CFR: 41.5 / 45.5)
A1.09 SBLC system lineup Ability to (a) predict the impacts of the 4.0 31 following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, 212000 RPS X control, or mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
A2.16 Changing mode switch position Ability to monitor automatic operations of 3.3 20 the INTERMEDIATE RANGE MONITOR 215003 IRM X (IRM) SYSTEM including: (CFR: 41.7 /
45.7)
A3.02 Annunciator and alarm signals Ability to manually operate and/or monitor 3.4 8 in the control room: (CFR: 41.7 / 45.5 to 215004 Source Range Monitor X 45.8)
A4.07 Verification of proper functioning/
operability 2.4.34 Knowledge of RO tasks performed 4.2 22 outside the main control room during an 215005 APRM / LPRM X emergency and the resultant operational effects. (CFR: 41.10 / 43.5 / 45.13)
Knowledge of the physical connections 2.6 60 and/or cause/effect relationships between REACTOR CORE ISOLATION COOLING 217000 RCIC X SYSTEM (RCIC) and the following: (CFR:
41.2 to 41.9 / 45.7 to 45.8)
K1.04 Main condenser Knowledge of ADS design features and/or 3.8 3 interlocks which provide for the following:
218000 ADS X (CFR: 41.7)
K4.03:ADS logic control
ES-401 6 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)
System # / Name K K K K K K A A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 Knowledge of the effect that a loss or 2.8 30 malfunction of the PRIMARY CONTAINMENT ISOLATION 223002 PCIS/Nuclear Steam X SYSTEM/NUCLEAR STEAM SUPPLY Supply Shutoff SHUT-OFF will have on following: (CFR:
41.7 / 45.4)
K3.06 Turbine building radiation Knowledge of RELIEF/SAFETY VALVES 3.9 33 design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) 239002 SRVs X K4.01 Insures that only one or two safety/relief valves reopen following the initial portion of a reactor isolation event (LLS logic): Plant-Specific Knowledge of the operational implications 2.6 39 of the following concepts as they apply to 239002 SRVs X RELIEF/SAFETY VALVES: (CFR: 41.5 /
45.3)
K5.05 Discharge line quencher operation Knowledge of the effect that a loss or 3.3 37 malfunction of the following will have on the 259002 Reactor Water Level X REACTOR WATER LEVEL CONTROL Control SYSTEM: (CFR: 41.7 / 45.7)
K6.02 A.C. power Ability to predict and/or monitor changes in 3.8 21 parameters associated with operating the 259002 Reactor Water Level REACTOR WATER LEVEL CONTROL X
Control SYSTEM controls including: (CFR: 41.5 /
45.5)
A1.03 Reactor power Ability to (a) predict the impacts of the 2.9 66 following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to 261000 SGTS X correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
A2.01 Low system flow Ability to monitor automatic operations of 3.0 50 the STANDBY GAS TREATMENT 261000 SGTS X SYSTEM including: (CFR: 41.7 / 45.7)
A3.04 System Temperature Ability to manually operate and/or monitor 3.2 68 262001 AC Electrical in the control room: (CFR: 41.7 / 45.5 to X
Distribution 45.8)
A4.03 Local operation of breakers 2.4.31 Knowledge of annunciator alarms, 4.2 10 262002 UPS (AC/DC) X indications, or response procedures.
(CFR: 41.10 )
Knowledge of the physical connections 3.3 34 and/or cause/effect relationships between 263000 DC Electrical D.C. ELECTRICAL DISTRIBUTION and X
Distribution the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.01 A.C. electrical distribution Knowledge of the effect that a loss or 4.2 45 malfunction of the EMERGENCY 264000 EDGs X GENERATORS (DIESEL/JET) will have on following: (CFR: 41.7 / 45.4)
K3.01 Emergency core cooling systems Knowledge of electrical power supplies to 2.8 11 300000 Instrument Air X the following: (CFR: 41.7)
K2.01 Instrument air compressor
ES-401 7 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)
System # / Name K K K K K K A A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 Knowledge of the operational implications 2.5 58 of the following concepts as they apply to 300000 Instrument Air X the INSTRUMENT AIR SYSTEM: (CFR:
41.5 / 45.3)
K5.01 Air compressors Knowledge of electrical power supplies for 2.9 38 400000 Component Cooling X the following: (CFR: 41.7)
Water K2.01 CCW Pumps Ability to (a) predict the impacts of the 3.4 96 following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use 215005 APRM / LPRM X procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
A2.07 Recirculation flow channels flow mismatch 2.2.25 Knowledge of bases for Tech Specs 4.2 95 239002 SRVs X for LCOs and safety limits (CFR: 43.2)
Ability to (a) predict the impacts of the 2.5 94 following on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) ; and (b) based on those predictions, use 262002 UPS (AC/DC) X procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
A2.02 Over voltage 2.4.20 Knowledge of the operational 4.3 87 264000 EDGs X implications of EOP warnings, cautions, and notes. (CFR: 41.10 / 43.5 / 45.13) 2.2.25 Knowledge of the bases in Tech 4.2 88 212000 RPS X Specs for LCOs and Safety limits K/A Category Point Totals: 2 3 3 3 3 2 2 2/2 2 2 2/3 Group Point Total: 26/5
ES-401 8 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / SRO)
System # / Name K K K K K K A A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 2.2.44 Ability to interpret control room 201001 CRD Hydraulic X indications to verify the status and 4.2 48 operation of a system, and understand how operator actions and directives affect plant and system conditions.
(CFR: 41.5 / 43.5 / 45.12) 2.2.42 Ability to recognize system 201002 RMCS X parameters that are entry-level 4.6 97 conditions for Technical Specifications.
(CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3) 201003 Control Rod and Drive Mechanism 201004 RSCS Knowledge of the operational 201005 RCIS X implications of the following concepts 3.5 53 as they apply to ROD CONTROL AND INFORMATION SYSTEM (RCIS):
(CFR: 41.5 / 45.3)
K5.09 High power setpoints BWR-6 201006 RWM 202001 Recirculation Knowledge of the effect that a loss or 202002 Recirculation Flow Control X malfunction of the RECIRCULATION 3.2 71 FLOW CONTROL SYSTEM will have on following: (CFR: 41.7 / 45.4)
K3.05 Recirculation pump speed:
Plant-Specific Knowledge of REACTOR WATER 204000 RWCU X CLEANUP SYSTEM design feature(s) 2.9 23 and/or interlocks which provide for the following: (CFR: 41.7)
K4.03 Over temperature protection for system components 214000 RPIS 215001 Traversing In-Core Probe 215002 RBM Knowledge of the physical connections 216000 Nuclear Boiler Inst. X and/or cause/effect relationships 3.9 27 between NUCLEAR BOILER INSTRUMENTATION and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.06 Low pressure core spray Knowledge of electrical power supplies 219000 RHR/LPCI: Torus/Pool X to the following: (CFR: 41.7) 2.5 43 Cooling Mode K2.01 Valves Knowledge of the effect that a loss or 223001 Primary CTMT and Aux. X malfunction of the following will have 3.5 55 on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES: (CFR:
41.7 / 45.7)
K6.02 Containment cooling: Mark-III 226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray Mode
ES-401 9 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / SRO)
System # / Name K K K K K K A A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control Ability to (a) predict the impacts of the 241000 Reactor/Turbine Pressure X following on the REACTOR/TURBINE 3.7 74 Regulator PRESSURE REGULATING SYSTEMSYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5)
A2.02 High Reactor Pressure 245000 Main Turbine Gen. / Aux.
256000 Reactor Condensate 259001 Reactor Feedwater Ability to predict and/or monitor 268000 Radwaste X changes in parameters associated with 2.7 41 operating the RADWASTE controls including: (CFR: 41.5 / 45.5)
A1.01 Radiation level Ability to monitor automatic operations 271000 Offgas X of the OFFGAS SYSTEM including: 3.4 1 (CFR: 41.7 / 45.7)
A3.07 Process radiation monitoring system indications 272000 Radiation Monitoring 286000 Fire Protection 288000 Plant Ventilation Ability to manually operate and/or 290001 Secondary CTMT X monitor in the control room: (CFR: 3.2 69 41.7 / 45.5 to 45.8)
A4.09 System status lights and alarms:
Plant-Specific 290003 Control Room HVAC 2.1.32 Ability to explain and apply 290002 Reactor Vessel Internals X system limits and precautions. (CFR: 3.8 57 41.10 / 43.2 / 45.12) 204000 RWCU Ability to (a) predict the impacts of the 234000 Fuel Handling Equipment X following on the FUEL HANDLING 3.7 91 EQUIPMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
A2.01 Interlock Failure 2.2.22 Knowledge of limiting conditions 290003 Control Room HVAC X for operations and safety limits 4.7 89 K/A Category Point Totals: 1 1 1 1 1 1 1 1/1 1 1 2/2 Group Point Total: 12/3
ES-401 10 Form ES-401-3 Facility: River Bend Station Date of Exam: 2016 Category K/A # Topic RO SRO-Only IR # IR #
2.1.3 Knowledge of shift or short-term relief turnover practices. 3.7 49 (CFR: 41.10 / 45.13) 2.1.39 Knowledge of conservative decision making practices. 3.6 70 (CFR: 41.10 / 43.5 / 45.12) 2.1.45 Ability to identify and interpret diverse indications to 4.1 7
- 1. validate the response of another indication. (CFR: 41.7 /
Conduct of 43.5 / 45.4)
Operations 2.1.5 Ability to use procedures related to shift staffing, such as 3.9 76 minimum crew complement, overtime limitations, etc.
(CFR: 41.10 / 43.5 / 45.12) 2.1.42 Knowledge of new and spent fuel movement procedures. 3.4 100 (CFR: 41.10 / 43.7 / 45.13)
Subtotal 3 2 2.2.2 Ability to manipulate the controls as required to operate 4.6 42 the facility between shutdown and power levels 2.2.43 Knowledge of the process used to track inoperable 3.0 52
- 2. alarms. (CFR: 41.10 / 43.5)
Equipment 2.2.14 Knowledge of the process for controlling equipment 3.9 25 Control configuration or status. (CFR: 41.10 / 43.3 / 45.13) 2.2.38 Knowledge of conditions and limitations in the facility 4.5 86 license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)
Subtotal 3 1 2.3.4 Knowledge of radiation exposure limits under normal or 3.2 61 emergency conditions. (CFR: 41.12 / 43.4 / 45.10) 2.3.12 Knowledge of radiological safety principles pertaining to 3.2 54 licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
- 3. 2.3.5 Ability to use radiation monitoring systems, such as fixed 2.9 79 Radiation radiation monitors and alarms, portable survey Control instruments, personnel monitoring equipment, etc. (CFR:
41.11 / 41.12 / 43.4 / 45.9) 2.3.15 Knowledge of radiation monitoring systems, such as fixed 3.1 81 radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR:
41.12 / 43.4 / 45.9)
Subtotal 2 2 2.4.12 Knowledge of general operating crew responsibilities 4.0 63 during emergency operations
- 4. 2.4.21 Knowledge of the parameters and logic used to assess 4.0 56 Emergency the status of safety functions, such as reactivity control, Procedures / core cooling and heat removal, reactor coolant system Plan integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 / 45.12) 2.4.32 Knowledge of operator response to loss of all 4.0 82 annunciators. (CFR: 41.10 / 43.5 / 45.13)
ES-401 11 Form ES-401-3 Facility: River Bend Station Date of Exam: 2016 Category K/A # Topic RO SRO-Only IR # IR #
2.4.38 Ability to take actions called for in the facility emergency 4.4 85 plan, including supporting or acting as emergency coordinator if required. (CFR: 41.10 / 43.5 / 45.11)
Subtotal 2 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 Rejected KAs were not tracked since this was NRC written and the majority of the exam required new questions to be written for the randomly selected KAs because they did not have previously written questions in the facility licensees bank.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: River Bend Nuclear Station Date of Examination: 9/12/2016 Examination Level: RO SRO Operating Test Number: LOT-2016 Administrative Topic Type Describe activity to be performed (see Note) Code*
R-N CRD pump clearance Conduct of Operations R-N Calculation for leakage Conduct of Operations Electrical Print Reading (Determine effect of removing R-N Equipment Control fuses in RPS system)
R-N Determine emergency entry requirements for high dose Radiation Control Emergency Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 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 Administrative Topics Outline Form ES-301-1 Facility: River Ben Nuclear Station Date of Examination: 9/12/2016 Examination Level: RO SRO Operating Test Number: LOT-2016 Administrative Topic Type Describe activity to be performed (see Note) Code*
Print Reading (RPS fuse removal)
Conduct of Operations R-N Leakage calculation and TS call associated with Conduct of Operations R-N leakage.
Evaluate a CRD pump clearance Equipment Control R-N Emergency entry for high dose and who authorizes the Radiation Control R-N entry.
Emergency Classification Emergency Plan R-N NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 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 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: RIVER BEND NUCLEAR STATION Date of Examination: 9/12/2016 Exam Level: RO SRO-I SRO-U Operating Test No.: LOT 2016 Control Room Systems* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U Safety System / JPM Title Type Code*
Function S1. Perform Rod Withdrawal Limiter Test with a MSIV 1 A-N-S-L inadvertent closure.
S2. Align SU FRV and controller drifts open. A-N-S-L 2 S3. Open MSIVs per SOP-011. M-S-L 3 S4. Align Division 3 DG to power ENS-SWGR1A (fault occurs EN - N - S 4 requiring alignment to ENS-SWGR1B)
S5. SBGT fails to start with dampers fail to open. A - EN - N - S 9 S6. Swap electrical bus power supplies N-S 6 S7. Drive In IRM/SRM Detectors Following a SCRAM D-S-L 7 S8. Start RHR in Suppression Pool Cooling Mode with high A - D - EN -
5 pump amps. (FIX PAPERWORK) S-L In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
P1. CRD filter swap with new filter clogged (CR-2016-1724) A-N-R 1 P2. ATC actions to Man the RSP. AOP 31 Att 12 Section 1.1.4-N-E-L 2 1.1.6 P3. Instrument Air Diesel Air compressor backup to safety D-E-P-R 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; 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 / 1 / 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 S1 Need to snap an IC in the HLO Load with the appropriate power and a valid RMP Indicate the next rod to be pulled in the JPM Markup a copy of GOP-1 including SRM counts and calculations S2 Align Gabes version and Daves S3 Modified from Audit JPM: RJPM-AUD-D14-S1, Re-Open MSIVs Following Automatic Isolation Ready for Review S4 JPM IC Pairings 5 ICs will be required:
S1 - S6 680 - 601 100%
S5 - S7 863 - 680 post scram S2 - S8 680 - 601 post scram S3 601 same IC as S1 and S8 S4 601 low power above critical,
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 1 Page 1 of 7 Facility: River Bend Nuclear Station Scenario No.: 1 Op-Test No.: NRC LOT 2016 Examiners: ____________________________ Operators: _____________________________
Initial Conditions: Operating at 100% power.
Inoperable Equipment: None Turnover:
Severe Thunderstorm warning in effect for West Feliciana in effect. All required actions per AOP-29 (Severe Weather Operation) are complete.
Scenario Notes:
This scenario is a NEW Scenario.
Validation Time: 60 minutes
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 1 Page 2 of 7 Event Malf. No. Event Type Event No. Description C (BOP)
HPCS Inadvertent Initiation. AOP 6, Condensate/Feedwater 1 HPCS004 TS (CRS)
Failures. TS 3.5.1 AOP C (ATC) Degraded Grid. Adjust MVARs per AOP 64, Degraded Grid.
Isophase Bus Duct Cooling Fan Trips. SOP 67, Isolated 3 MGEN005A C (BOP)
Phase Bus Duct Cooling System. ARP P870-54-B-1 C (ATC)
FWS Pump A motor failure (amps increasing/secure pump).
AOFWS-4 R (ATC) AOP 6, Condensate/Feedwater Failures. AOP 24, Thermal A03-M Hydraulics Stability Controls. ARP P808-86-H-1, p680-3-C-1.
AOP P680-6A-IA_8 SDV level instrument fails high 1 of 4 instruments. TS 3.3.1.1.
5 TS (CRS)
(1/2 scram ARP P680-5-B-10, P680-5-A-10, P680-6-A-8 signal)
NPS-B Fault/Trip. Complete loss of feedwater. RCIC trips 6 ED002B M (CREW) due to malfunction. EOP 1 RWCU line break in the steam tunnel 100 gpm ramp for 3 minutes after mode switch taken to shutdown. ARPs P601-7 WCS006 M (CREW) 21-A-1, P601-21-A-6, P601-21-B-1, P601-21-A-3, P601-21-B-
- 3. EOP 3 WCS004 8 C G33-MOV-1 and G33-MOV-4 fail to auto close WCS005 MSS024D MSIVs (C-B21-AOVF022D and C-B21-AOVF028D) fail to auto 9 C MSS025D close (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Quantitative Attributes Table Normal Events 2 EOP Contingency Procedures Used 1 Total Malfunctions 7 Simulator Run Time 60 Malfunctions After EOP Entry 2 EOP Run Time 30 AOP Events 3 Critical Tasks 4 Major Transients 2 Reactivity Manipulations 1 EOPs Used (Requiring measurable action) 2
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 1 Page 3 of 7 SCENARIO ACTIVITIES:
HPCS Inadvertent Initiation:
A. After turnover and at the direction of the lead evaluator trigger Event 1.
B. The Unit Operator will override HPCS off. This will prevent further automatic initiation.
Degraded Grid:
A. After CRS enters TS 3.5.1 and/or at the direction of the lead evaluator trigger Event 2.
B. CRS will enter AOP 64 and dispatcher will notify control room to adjust MVARs.
C. MVARs will initially be 250 MVARs. ATC adjusts MVARs to 200 MVARs per AOP 64, Degraded Grid and ARP 680-P808-86-H-1.
Isophase Bus Duct Cooling Fan Trips:
A. After ATC adjusts MVARs and/or at the direction of the lead evaluator trigger Event 3.
B. The Unit Operator will restart Isophase Bus Duct Cooling Fan per SOP 67, Isolated Phase Bus Duct Cooling System and ARP P870-54-B-1.
FWS Pump A Motor Failure:
A. After Unit Operator restarts Isophase Bus Duct Cooling Fan and/or at the direction of the lead evaluator trigger Event 4.
B. The ATC will recognize the increasing FW motor amps and secure the FWS Pump A per AOP 6 (Condensate/Feedwater Failures), AOP 24 (Thermal Hydraulics Stability Controls), ARP P808-86-H-1, and ARP P680-3-C-1.
C. An automatic flow control valve runback will occur when the pump is tripped.
SDV level instrument fails high:
A. After ATC trips the FWS Pump A and/or at the direction of the lead evaluator trigger Event 5.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 1 Page 4 of 7 B. 1 of 4 SDV level instruments fails high. CRS enters TS 3.3.1.1.
NPS-B Fault/Trip, Complete loss of feedwater, RCIC malfunction:
A. After CRS enters TS 3.3.1.1 and/or at the direction of the lead evaluator trigger Event 6.
B. Crew will manually scram the reactor due to the complete loss of feedwater. The automatic RPS scram will occur at 9.7 reactor water level if manual action is not taken first.
C. Crew will restore HPCS injection to maintain reactor water level in expanded level band.
RWCU line break in the steam tunnel:
A. 3 minutes after automatic or manual scram, Event 7 will automatically trigger.
B. Indications of high temperatures in the steam tunnel will cause isolation signals.
G33-MOV-1 and G33-MOV-4 will fail to auto close and MSIVs (C-B21-AOVF022D and C-B21-AOVF028D) will fail to auto close.
C. Crew will shut the failed isolation valves.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 1 Page 5 of 7 Critical Task Number Description Basis Close one RWCU isolation valve (G33-1 MOV-1 OR G33-MOV-4) after failure to close on group 15 isolation before completion of step 5.9 of AOP-3 Restore HPCS injection prior to 2 Emergency Depressurization.
- Critical Task (As defined in NUREG 1021 Appendix D)
Simulator Notes:
Prior to scenario, ensure GML Fan 1 is running Prior to scenario, adjust MVAR to 250 MVARs Event 4, get to 360 amps quicker and then ramp to 440 amps slower.
Event 8, need to delete leak once valve is closed.
Provide graphic for the SPI-REC102 to indicate values consistent with grid.
Screen C11NC036 needs to indicate one trip unit in TRIP.
When restoring HPCS during leak, if asked to fill and vent system, expedite the process.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 2 Page 1 of 7 Facility: River Bend Nuclear Station Scenario No.: 2 Op-Test No.: NRC LOT 2016 Examiners: ____________________________ Operators: _____________________________
Initial Conditions: Operating at 100% power.
Inoperable Equipment: None Turnover:
Severe Thunderstorm warning in effect for West Feliciana in effect. All required actions per AOP-29 are complete.
Main Turbine experienced increased (approximately 1 mil higher than normal) turbine vibrations on previous shift.
Scenario Notes:
This scenario is a NEW Scenario.
Validation Time: 60 minutes
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 2 Page 2 of 7 Event Malf. No. Event Type Event No. Description 1 MSS010 C (BOP) Gland Seal Steam Supply Failure. ARP P870-54-E-05 C (BOP)
RHS isolation failure. ARP P870-56A-H-03. AOP-3 2 RHSAOV64 TS (CRS)
(Automatic Isolations). TS 3.3.6.2 AOP C (ATC)
R (ATC) NJS-J malfunction. MFP A trip. FCV runback. B FCV fails to 3 RCS015B runback. Recirculation flow mismatch. AOP 6 (Condensate /
TS (CRS) Feedwater failures). AOP 14 (Loss of 125VDC). TS 3.4.1 AOP Main Turbine Vibrations increase to manual scram level. EOP 4 ED004J M (ATC) 1A (RPV Control, ATWS). OSP-53.
5 CRD014 M (CREW) Hydraulic ATWS 65%. EOP 1A (RPV Control, ATWS).
6 SLC002A C SLC Pump Failure.
7 MSS005E C SRV (47B) stuck open.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Quantitative Attributes Table Normal Events 0 EOP Contingency Procedures Used 1 Total Malfunctions 5 Simulator Run Time 60 Malfunctions After EOP Entry 2 EOP Run Time 30 AOP Events 2 Critical Tasks 2 Major Transients 2 Reactivity Manipulations 1 EOPs Used (Requiring measurable action) 1
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 2 Page 3 of 7 SCENARIO ACTIVITIES:
Gland Seal Steam Supply Failure:
A. After turnover and at the direction of the lead evaluator trigger Event 1.
B. The Unit Operator will throttle open TME-MOVS2 SSE PRESS REG BYP VLV to restore seal steam header pressure per ARP P870-54-E-05.
RHS isolation failure:
A. After seal steam header pressure is restored and/or at the direction of the lead evaluator trigger Event 2.
B. The Unit Operator will close RHS-AOV64, SPC Discharge Valve due to failure to isolate per ARP P870-56A-H-03 and AOP-3 (Automatic Isolations).
C. The CRS will enter TS 3.3.6.2.
NJS-J malfunction:
A. After RHS-AOV64 is closed, TS 3.3.6.2 entered, and/or at the direction of the lead evaluator trigger Event 3.
B. FWP A will trip. An automatic FCV runback will occur. The B FCV fails to runback.
A large Recirculation flow mismatch will result.
C. The Unit Operator will manually runback the B FCV to eliminate the flow mismatch per AOP 6 (Condensate / Feedwater failures) and AOP 14 (Loss of 125VDC).
D. The CRS will enter TS 3.4.1.
Main Turbine Vibrations:
A. After the B FCV has been manually runback, TS 3.4.1 entered, and/or at the direction of the lead evaluator trigger Event 4.
B. Vibrations continue to increase to 12 mils which require a manual scram and turbine trip per OSP-53.
ATWS:
A. After the ATC manually scrams the reactor, a 65% Hydraulic ATWS will occur automatically.
B. The crew will install EOP attachments and drive the control rods in.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 2 Page 4 of 7 SLC Pump Failure:
A. The first SLC pump started will have an automatic malfunction resulting in a failure to inject.
B. The second SLC pump will operate normally and inject boron as required.
SRV (47B) stuck open:
A. SRV (47B) will automatically open upon the reactor scram. The valve will shut when the hand switch is manually taken to close.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 2 Page 5 of 7 Critical Task Number Description Basis Terminate and prevent all injection 1 sources except boron injection, CRD and RCIC prior to exceeding HCTL.
Inject SLC prior to suppression pool 2 temperature reaching 110F.
- Critical Task (As defined in NUREG 1021 Appendix D)
Simulator Notes:
For the increased turbine vibrations:
TMS003 2 on initiate TMS003 raise to 8.5 over 5 minutes Event 2 needs a trigger to remove RHS AOV 64 malfunction.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 3 Page 1 of 7 Facility: River Bend Nuclear Station Scenario No.: 3 Op-Test No.: NRC LOT 2016 Examiners: ____________________________ Operators: _____________________________
Initial Conditions: Operating at 100% power.
Inoperable Equipment: None Turnover:
Severe Thunderstorm warning in effect for West Feliciana in effect. All required actions per AOP-29 are complete.
RPS B power supply is on alternate due to repairs to B RPS MG set.
STP-203-6305 is in progress commence on step 7.6.3.
Scenario Notes:
This scenario is a NEW Scenario.
Validation Time: 60 minutes
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 3 Page 2 of 7 Event Malf. No. Event Type Event No. Description N (BOP) HPCS STP. HPCS pump trip. STP-203-6305. TS 3.5.1.
1 HPCS001 TS (CRS) ARP P601-16-B-3, P601-16-F-5, and P601-16-G-4.
C (BOP/ATC) ENS SWGR B lockout. Loss of alternate RPS power supply.
Power supply switched back to normal and half scram reset.
2 ED003I TS (CRS) AOP 10 (Loss of RPS Bus). ARP P877-32-C-3, P877-32-E-1, P877-32-E-2, P877-32-F-2, P877-32-G-1, P877-32-H-3, AOP and P877-32-H-4.
C (ATC)
RCS005A Recirc pump seal leak, pump trip, (ATC reduces flow 3 TS (CRS)
RCS002A <33kgpm). ARP P680-4-E-5. GOP-4. AOP-24.
AOP Seal leak increases. THI increases and requires manual 4 RCS001A M (CREW) scram. ARP P680-6-C-1, P680-7-A-5, P680-7-A-6, P680 B-5, and P680-7-B-6. AOP-1. EOP-1.
Loss of offsite power. Loss of feedwater. RCIC trips on 5 M (CREW) overspeed. Failure of Division 2 DG to energize ENS SWGR-1B.
6 C RHR A pump fails to start 7 C LPCS injection valve fails to open (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Quantitative Attributes Table Normal Events 1 EOP Contingency Procedures Used 1 Total Malfunctions 5 Simulator Run Time 60 Malfunctions After EOP Entry 2 EOP Run Time 30 AOP Events 2 Critical Tasks 3 Major Transients 2 Reactivity Manipulations 1 EOPs Used (Requiring measurable action) 1
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 3 Page 3 of 7 SCENARIO ACTIVITIES:
HPCS pump trip:
A. After turnover the Unit Operator will perform STP-203-6305. During the performance of the STP the HPCS pump will automatically trip (Event 1). The Unit operator should secure HPCS STP lineup per ARP P601-16-B-3, P601-16-F-5, and P601-16-G-4.
A. After the Unit Operator secures HPCS, CRS enters TS 3.5.1, and/or at the direction of the lead evaluator trigger Event 2.
B. The ENS SWGR B lockout will result in loss of alternate RPS power supply and a division 2 half scram.
C. The Unit Operator will switch the power supply back to normal and the ATC will reset the half scram. AOP 10 (Loss of RPS Bus). ARP P877-32-C-3, P877-32-E-1, P877-32-E-2, P877-32-F-2, P877-32-G-1, P877-32-H-3, and P877-32-H-4.
D. The CRS will enter TS 3.8.1.
Recirculation pump seal leak:
A. After the Unit Operator swaps the RPS power supply and has completed AOP-10 through step 13 of Attachment 2, the ATC resets the Division 2 half scram, the CRS enters TS XXX, and/or at the direction of the lead evaluator trigger Event 3.
B. The A Recirculation pump #1 seal will degrade.
C. After the crew diagnosis of the seal failure and/or at the direction of the lead evaluator, trigger Event 4 to trip the Recirculation Pump.
D. The ATC will be required to reduce flow to < 33 kgpm and monitor for THI.
ARP P680-4-E-5. GOP-4. AOP-24.
E. The CRS will enter TS XXX.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 3 Page 4 of 7 Thermal Hydraulic Instability:
A. After the ATC reduces flow to < 33 kgpm, the CRS enters TS 3.5.1, and/or at the direction of the lead evaluator trigger Event 5.
B. THI power swings get bigger and requires manual scram. ARP P680-6-C-1, P680-7-A-5, P680-7-A-6, P680-7-B-5, and P680-7-B-6. AOP-1. AOP-24. EOP-1.
Loss of offsite power: LEAK????????????
A. After the manual scram, Event 6 will automatically occur.
B. The loss of offsite power will result in a loss of all feedwater. RCIC will trip on overspeed. The Division 2 DG will fail to energize ENS SWGR-1B.
C. Due to a loss of all high pressure injection sources, the crew will be forced to emergency depressurize and inject with low pressure ECCS systems.
RHR A pump fails to start:
A. The RHR A pump will fail to auto start after automatic initiation. Automatic trigger Event 7.
B. The crew will manually start the RHR A pump to assist restoring reactor water level.
RHR alone will not be enough to restore reactor water level. LPCS will be required as well.
LPCS injection valve fails to open:
A. The LPCS injection valve will fail to open after automatic initiation and emergency depressurization. Automatic trigger Event 8.
B. The crew will manually open the LPCS injection valve to assist restoring reactor water level. LPCS alone will not be enough to restore reactor water level. RHR will be required as well.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 3 Page 5 of 7 Critical Task Number Description Basis Mode switch to shutdown prior to 100%
1 oscillations on an APRM channel.
Emergency Depressurize prior to -185 2
inches compensated fuel zone.
Manually start RHR A pump and open 3 LPCS injection valve prior to transitioning to RPV flooding or entering SAPs.
- Critical Task (As defined in NUREG 1021 Appendix D)
Simulator Notes:
RPS B in alternate using the key switch in the back.
Set up the IC rod line to ensure when the recirc pump trips the crew is deep in the restricted region.
Make the recirc loop rupture big enough not to be turned by RHR alone.
Should not get the ventilation alarms.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 4 Page 1 of 3 Facility: River Bend Nuclear Station Scenario No.: 4 Op-Test No.: NRC LOT 2016 Examiners: ____________________________ Operators: _____________________________
Initial Conditions: Operating at 4% power.
Inoperable Equipment: None Turnover:
Continue withdrawing rods to achieve mode 1.
Severe Thunderstorm warning in effect for West Feliciana in effect. All required actions per AOP-29 are complete.
Scenario Notes:
This scenario is a NEW Scenario.
Validation Time: 60 minutes
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 4 Page 2 of 3 Event Malf. No. Event Type Event No. Description C (ATC)
Rod 24-49 Drifts out. ARP P680-7-B-2 and P680-7-C-1.
1 CRDM2449 TS (CRS)
AOP-61 (Control Rod Mispositioned / Malfunction). TS 3.1.3 AOP 2 NMS006E C (ATC) IRM E fails upscale. Bypass IRM. ARP P680-6-C-10.
CRD pump trip. BOP starts other CRD pump. ARP P601-3 C (BOP) 22-A-1, P601-22-B-1, P601-22-F-1, and P680-7-D-1 C (BOP) CCS pump trip. Standby fails to auto start. AOP-12 (Loss of 4
AOP Turbine Plant Component Cooling Water)
Rad monitor fails. HVF-AOD-104, HVF-AOD-122 and HVF-5 RMS005C C (BOP)
AOD-102 fail to auto close.
6 NMS007A TS (CRS) IRM A fails downscale. ARP P680-6-C-9.
ED001 DG001B Station Blackout. Loss of offsite power. Loss of feed.
7 M Reactor scram. Division 2 and 3 DGs fail. EOP 1. Drywell DG002C leak 200 gpm RCS007 Division 1 DG voltage is low and needs to be raised in order 8 DG006A C to energize the bus.
RCIC manual initiation fails. RCIC must be manually aligned 9 C per hardcard.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Quantitative Attributes Table Normal Events 0 EOP Contingency Procedures Used 1 Total Malfunctions 8 Simulator Run Time 60 Malfunctions After EOP Entry 2 EOP Run Time 30 AOP Events 1 Critical Tasks 2 Major Transients 1 Reactivity Manipulations 1 EOPs Used (Requiring measurable action) 1
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 4 Page 3 of 3 SCENARIO ACTIVITIES:
Rod 24-49 Drifts out:
A. After turnover the ATC will continue with the rod withdraw to transition to mode 1.
The rods will be withdrawn one notch at a time. On the second notch withdraw, Event 1 will automatically insert a continuous rod withdraw for rod 24-49.
B. The ATC will recognize the continuous withdraw and fully insert the rod until it is locally isolated to maintain fully inserted per ARP P680-7-B-2, P680-7-C-1, and AOP-61 (Control Rod Mispositioned/ Malfunction)
IRM E fails upscale:
A. After the rod is isolated, CRS enters TS 3.1.3, and/or at the direction of the lead evaluator trigger Event 2.
B. The ATC will bypass IRM E per ARP P680-6-C-10.
CRD pump trip:
A. After the IRM is bypassed and/or at the direction of the lead evaluator trigger Event 3.
B. The running CRD pump will trip and the other CRD pump does not have a standby feature.
C. The Unit Operator will manually start the other CRD pump per ARP P601-22-A-1, P601-22-B-1, P601-22-F-1, P680-7-D-1.
CCS pump trip:
A. After the CRD pump is started and/or at the direction of the lead evaluator trigger Event 4.
B. The running CCS pump will trip and the standby pump fails to start.
C. The Unit Operator will manually start the standby CCS pump per AOP-12 (Loss of Turbine Plant Component Cooling Water).
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 4 Page 4 of 3 Rad monitor fails:
A. After the CCS pump is started and/or at the direction of the lead evaluator trigger Event 5.
B. The rad monitor fails and HVF-AOD-104, HVF-AOD-122 and HVF-AOD-102 fail to auto close. ARP-RMS-DSP230/4GE005 C. The Unit Operator manually closes the three valves.
IRM A fails downscale:
A. After the three HVF dampers are shut and/or at the direction of the lead evaluator, trigger Event 6.
B. The CRS will enter TS 3.3.1.1.
Station Blackout:
A. After the CRS enters TS 3.3.1.1 and/or at the direction of the lead evaluator trigger Event 7.
B. A loss of offsite power will result in a complete loss of feed and an automatic reactor scram on low reactor water level. Division 2 and 3 Diesel Generators will malfunction and not start. A slow drywell leak will occur.
Division 1 Diesel Generator fails to energize ENS-SWGR1A:
A. Upon the loss of offsite power, Event 8 will automatically trigger.
B. The Division 1 Diesel Generator will fail to reach voltage required to automatically energize ENS-SWGR1A.
C. The crew will manually raise diesel voltage and energize ENS-SWGR1A.
RCIC manual initiation fails:
A. Manual initiation of RCIC using the manual pushbutton will not work, Event 9 will automatically trigger.
B. The crew will manually align RCIC per the hardcard.
Appendix D Scenario Outline Form ES-D-1 NRC 2016 Scenario 4 Page 5 of 3 Critical Task Number Description Basis Adjust division 1 voltage to energize ENS SWGR A within 10 minutes or if shutdown, then once fill and vent of LP 1
ECCS systems are complete, restart and power ENS SWGR A prior to auto trip on high temperature.
Manually start RCIC prior to RPV level 2 reaching -187 inches compensated fuel zone.
- Critical Task (As defined in NUREG 1021 Appendix D)
Simulator Notes:
Provide and mark up GOP 1 and reactivity plan.
Rod drift inserted on first rod of plan. The rod selected should be one that requires being at 48 inches.
This will prevent an issue with the rod pattern controller. Insert the rod drift on the second notch withdraw.
CRD006 Reset CRDM high Temperature ECCS003 LPCS pump breaker (control power fuses)
ECCS004 RHR A pump breaker (control power fuses)
ES-301 Transient and Event Checklist Form ES-301-5 FACILITY: RBS DATE OF EXAM: 09/12/2016 OPERATING TEST NO.: NRC 301-1 Scenarios A E P V 1 2 3 (Spare) 4 T M P E O I L N T N I T C
A I A T L M N Y U T P Tues Wed Thurs M(*)
E CREW POSITION CREW CREW CREW POSITION POSITION POSITION R I U S A B S A B S A B S A B R T O R T O R T O R T O O C P O C P O C P O C P 0 1 1 0 RX R1 0 1 1 1 NOR 3,4,9 3,4, 7 4 4 2 I/C 7,8 6,8 5,6 4 2 2 1 MAJ TS 0 0 2 2 3 1 1 1 0 RX R2 1 1 1 1 1 NOR 2 4,5 5 4 4 2 I/C 8,9 5,6 7 3 2 2 1 MAJ TS 0 0 2 2 2 1 1 1 0 RX R3 1 1 1 1 1 NOR 3,4, 3 5 4 4 2 I/C 7,8 5,6 7 3 2 2 1 MAJ TS 0 0 2 2 3 1 1 1 0 I1 RX 1 1 1 1 1 NOR 1-4, 2 7 4 4 2 I/C 7,9 6,8 5,6 4 2 2 1 MAJ TS 1,5 2 0 2 2 Instructions:
- 1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
- 2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.
- 3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
- 4. For licenses that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating controls.
ES-301 Transient and Event Checklist Form ES-301-5 FACILITY: RBS DATE OF EXAM: 09/12/2016 OPERATING TEST NO.: NRC 301-1 Scenarios A E P V 1 2 3 (Spare) 4 T M P E O I L N T N I T C
A I A T L M N Y U T P Tues Wed Thurs M(*)
E CREW POSITION CREW CREW CREW POSITION POSITION POSITION S A B S A B S A B S A B R I U R T O R T O R T O R T O O C P O C P O C P O C P 3 2 2 1 1 0 RX I2 1 1 2 1 1 1 NOR 1-4, 2 3-6, 7 4 4 2 I/C 7,9 8,9 6,8 5,6 7 5 2 2 1 MAJ TS 1,5 2,6 4 0 2 2 3 1 1 1 0 RX U1 1 1 1 1 1 NOR 1,2,7 2-4, 8 4 4 2 I/C 7,8 6,8 5,6 4 2 2 1 MAJ TS 2,4 2 0 2 2 3 1 1 1 0 RX U2 1 1 1 1 1 NOR 1,2,7 2-4, 8 4 4 2 I/C 7,8 6,8 5,6 4 2 2 1 MAJ TS 2,4 2 0 2 2 1 1 0 RX 1 1 1 1 NOR 1-4, 3,4, 2,3 4 4 2 I/C 8,9 8,9 6,7 6,7 6,7 2 2 1 MAJ TS 2,3 0 2 2 Instructions:
- 1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
- 2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.
- 3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.