ML14017A043
| ML14017A043 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/15/2013 |
| From: | Garchow S Operations Branch IV |
| To: | Entergy Nuclear Operations |
| laura hurley | |
| References | |
| ES-401, ES-401-1 | |
| Download: ML14017A043 (92) | |
Text
ES-401 BWR Examination Outline - RO Form ES-401-1 Facility: Grand Gulf Date of Exam: November, 2013 RO K/A Category Points SRO-Only Points Tier Group 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 4 5 2 3 3 20 7 Emergency &
Abnormal Plant 2 2 2 1 N/A 1 0 N/A 1 7 3 Evolutions Tier Totals 5 6 6 3 3 4 27 10 1 3 2 2 2 2 3 2 2 2 4 2 26 5 2.
Plant 2 2 0 1 2 2 1 0 1 1 1 1 12 3 Systems Tier Totals 5 2 3 4 4 4 2 3 3 5 3 38 8
- 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 4 2 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).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.
The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.
Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO)
E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR Q#
1 2 3 1 2 AK3.01 - Knowledge of the reasons for the 295001 Partial or Complete Loss of Forced X following responses as they apply to PARTIAL OR 1 Core Flow Circulation / 1 & 4 3.3 COMPLETE LOSS OF A.C. POWER.
2.1.7 Ability to evaluate plant performance and 295003 Partial or Complete Loss of AC / 6 X make operational judgments based on operating characteristics / reactor behavior / and instrument 3.7 2 interpretation.
AK2.02 -. Knowledge of the interrelations between 295004 Partial or Total Loss of DC Pwr / 6 X PARTIAL OR COMPLETE LOSS OF D.C. POWER 3.0 3 and the following: Batteries AA1.04 - Ability to operate and/or monitor the 295005 Main Turbine Generator Trip / 3 X following as they apply to MAIN TURBINE 2.7 4 GENERATOR TRIP : Main generator controls AK1.02 - Knowledge of the operational implications 295006 SCRAM / 1 X of the following concepts as they apply to SCRAM : 3.4 5 Shutdown margin AA2.05 - Ability to determine and/or interpret the 295016 Control Room Abandonment / 7 X following as they apply to CONTROL ROOM 3.8 6 ABANDONMENT : Drywell pressure AA2.02 - Ability to determine and/or interpret the 295018 Partial or Total Loss of CCW / 8 X following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING 3.1 7 WATER : Cooling water temperature AK2.16 - Knowledge of the interrelations between 295019 Partial or Total Loss of Inst. Air / 8 X PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Reactor core 2.8 8 isolation cooling AK3.01 - Knowledge of the reasons for the 295021 Loss of Shutdown Cooling / 4 X following responses as they apply to LOSS OF SHUTDOWN COOLING : Raising reactor water 3.3 9 level AA1.04 - Ability to operate and/or monitor the 295023 Refueling Acc / 8 X following as they apply to REFUELING 3.4 10 ACCIDENTS : Radiation monitoring equipment EK1.02 - Knowledge of the operational implications 295024 High Drywell Pressure / 5 X of the following concepts as they apply to HIGH DRYWELL PRESSURE : Containment building 3.9 11 integrity: Mark-III 2.4.50 - Ability to verify system alarm setpoints and 295025 High Reactor Pressure / 3 X operate controls identified in the alarm response 4.2 12 manual.
EK3.05 - Knowledge of the reasons for the 295026 Suppression Pool High Water X following responses as they apply to Temp. / 5 SUPPRESSION POOL HIGH WATER 3.9 13 TEMPERATURE: Reactor SCRAM EK1.03 - Knowledge of the operational implications 295027 High Containment Temperature / 5 X of the following concepts as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III 3.8 14 CONTAINMENT ONLY) : Containment integrity:
Mark-III.
EK2.02 - Knowledge of the interrelations between 295028 High Drywell Temperature / 5 X HIGH DRYWELL TEMPERATURE and the following: Components 3.2 15 internal to the drywell EA2.03 - Ability to determine and/or interpret the 295030 Low Suppression Pool Wtr Lvl / 5 X following as they apply to LOW SUPPRESSION 3.7 16 POOL WATER LEVEL : Reactor pressure 2.1.31 Ability to locate control room switches /
295031 Reactor Low Water Level / 2 X controls and indications and to determine that they 4.2 17 are correctly reflecting the desired plant lineup.
295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 EK3.02 - Knowledge of the reasons for the 295038 High Off-site Release Rate / 9 X following responses as they apply to HIGH OFF- 3.9 18 SITE RELEASE RATE: System isolations AK3.04 - Knowledge of the reasons for the 600000 Plant Fire On Site / 8 X following responses as they apply to PLANT FIRE ON SITE: Actions contained in the abnormal 2.8 19 procedure for plant fire on site AK2.03 - Knowledge of the interrelations between 700000 Generator Voltage and Electric Grid X GENERATOR VOLTAGE AND lELECTRIC GRID Disturbances / 6 DISTURBANCES and the following: Sensors, 3.0 20 detectors, indicators K/A Category Totals: 3 4 5 2 3 3 Group Point Total: 20
ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO)
E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #
1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 AK1..05 - Knowledge of the operational implications of 295009 Low Reactor Water Level / 2 X the following concepts as they apply to LOW 3.3 21 REACTOR WATER LEVEL : Natural circulation 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 2.2.44 - Ability to interpret control room indications to 295012 High Drywell Temperature / 5 X verify the status and operation of a system, and understand how operator actions and directives affect 4.2 22 plant and system conditions.
AA1.02 - Ability to operate and/or monitor the following 295013 High Suppression Pool Temp. / 5 X as they apply to HIGH SUPPRESSION POOL TEMPERATURE : Systems that add heat to the 3.9 23 suppression pool 295014 Inadvertent Reactivity Addition / 1 AK2. Knowledge of the interrelations between 295015 Incomplete SCRAM / 1 X INCOMPLETE SCRAM and the following: Rod control 3.2 24 and information system: Plant-Specific 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Wtr Lvl / 5 EK3.02 - Knowledge of the reasons for the following 295032 High Secondary Containment X responses as they apply to HIGH SECONDARY Area Temperature / 5 CONTAINMENT AREA TEMPERATURE : Reactor 3.6 25 SCRAM 295033 High Secondary Containment Area Radiation Levels / 9 EK2.02 - Knowledge of the interrelations between 295034 Secondary Containment X SECONDARY CONTAINMENT VENTILATION HIGH Ventilation High Radiation / 9 RADIATION and the following: Area radiation 3.8 26 monitoring system 295035 Secondary Containment High Differential Pressure / 5 EK1.01 - Knowledge of the operational implications of 295036 Secondary Containment High X the following concepts as they apply to SECONDARY Sump/Area Water Level / 5 CONTAINMENT HIGH SUMP/AREA WATER LEVEL : 2.9 27 Radiation releases 500000 High CTMT Hydrogen Conc. / 5 K/A Category Point Totals: 2 2 1 1 0 1 Group Point Total: 7
ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)
System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 K6.10 - Knowledge of the effect that a loss or 203000 RHR/LPCI: Injection X malfunction of the following will have on the Mode RHR/LPCI: INJECTION MODE (PLANT 3.0 28 SPECIFIC) : Component cooling water systems A2. Ability to (a) predict the impacts of the 205000 Shutdown Cooling X following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) ; and (b) based on those predictions, use procedures to correct, control, or mitigate 2.6 29 the consequences of those abnormal conditions or operations: Low shutdown cooling suction pressure: Plant-Specific K2.01 - Knowledge of electrical power X supplies to the following: Pump motors 3.1 30 206000 HPCI NA GGNS 207000 Isolation (Emergency) NA GGNS Condenser K3.03 - Knowledge of the effect that a loss or 209001 LPCS X malfunction of the LOW PRESSURE CORE SPRAY SYSTEM will have on following: 2.9 31 Emergency generators A4.08 - Ability to manually operate and/or X monitor in the control room: Reactor water 3.9 32 level 2.4.18 - Knowledge of the specific bases for 209002 HPCS X EOPs. 3.3 33 K4.09 - Knowledge of STANDBY LIQUID 211000 SLC X CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:
Dampening of positive displacement pump 2.5 34 discharge oscillations (accumulators): Plant-Specific K1. Knowledge of the physical connections 212000 RPS X and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the 3.4 35 following: A.C. electrical distribution K1.05 - Knowledge of the physical 215003 IRM X connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR 3.3 36 (IRM) SYSTEM and the following: Display control system: Plant-Specific K6.02 - Knowledge of the effect that a loss or X malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM) 3.6 37 SYSTEM : 24/48 volt D.C. power: Plant-Specific A1.05 - Ability to predict and/or monitor 215004 Source Range Monitor X changes in parameters associated with operating the SOURCE RANGE MONITOR 3.6 38 (SRM) SYSTEM controls including: SCRAM, rod block, and period alarm trip setpoints A3.04 - Ability to monitor automatic 215005 APRM / LPRM X operations of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE 3.2 39 MONITOR SYSTEM including: Annunciator and alarm signals
A2.15 - Ability to (a) predict the impacts of the 217000 RCIC X following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ;
and (b) based on those predictions, use 3.8 40 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Steam line break K5.01 - Knowledge of the operational 218000 ADS X implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION 3.8 41 SYSTEM : ADS logic operation A4.03 - Ability to manually operate and/or 223002 PCIS/Nuclear Steam X monitor in the control room: Reset system Supply Shutoff 3.6 42 isolations K2.01 - Knowledge of electrical power 239002 SRVs X supplies to the following: SRV solenoids 2.8 43 A3.01 - Ability to monitor automatic 259002 Reactor Water Level X operations of the REACTOR WATER LEVEL Control CONTROL SYSTEM including: Runout flow 3.0 44 control: Plant-Specific K1.08 - Knowledge of the physical 261000 SGTS X connections and/or cause-effect relationships between STANDBY GAS TREATMENT 2.8 45 SYSTEM and the following: Process radiation monitoring system K4.04 - Knowledge of A.C. ELECTRICAL 262001 AC Electrical X DISTRIBUTION design feature(s) and/or Distribution interlocks which provide for the following: 2.8 46 Protective relaying A4.04 Ability to manually operate and/or X monitor in the control room: Synchronizing 3.6 47 and paralleling of different A.C. supplies A1.02 - Ability to predict and/or monitor 262002 UPS (AC/DC) X changes in parameters associated with operating the UNINTERRUPTABLE POWER 2.5 48 SUPPLY (A.C./D.C.) controls including: Motor generator outputs 2.4.6 - Knowledge of EOP mitigation 263000 DC Electrical X strategies. 3.7 49 Distribution K3.02 - Knowledge of the effect that a loss or X malfunction of the D.C. ELECTRICAL DISTRIBUTION will have on the following: 3.5 50 Components using D.C. control power (i.e.
breakers)
K5.06 - Knowledge of the operational 264000 EDGs X implications of the following concepts as they apply to EMERGENCY GENERATORS 3.4 51 (DIESEL/JET) : Load sequencing K6.03 - Knowledge of the effect that a loss or 300000 Instrument Air X malfunction of the following will have on the INSTRUMENT 2.7 52 AIR SYSTEM: Temperature indicators A4.01 - Ability to manually operate and / or 400000 Component Cooling X monitor in the control room: CCW indications Water 3.1 53 and control K/A Category Point Totals: 3 2 2 2 2 3 2 2 2 4 2 Group Point Total: 26
ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO)
System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS K1.03 - Knowledge of the physical 201003 Control Rod and Drive X connections and/or cause-effect Mechanism relationships between CONTROL ROD 3.1 54 AND DRIVE MECHANISM and the following: RPIS 201004 RSCS 201005 RCIS 201006 RWM 2.4.9 - Knowledge of low 202001 Recirculation X power/shutdown implications in accident (e.g., loss of coolant accident or loss of 3.8 55 residual heat removal) mitigation strategies.
202002 Recirculation Flow Control K4.08 - Knowledge of REACTOR 204000 RWCU X WATER CLEANUP SYSTEM design feature(s) and/or interlocks which provide for the following: Reducing 3.3 56 reactor pressure upstream of low pressure piping: LP-RWCU 214000 RPIS 215001 Traversing In-core Probe 215002 RBM K5.13 - Knowledge of the operational 216000 Nuclear Boiler Inst. X implications of the following concepts as they apply to NUCLEAR BOILER 3.5 57 INSTRUMENTATION : Reference leg flashing: Design-Specific 219000 RHR/LPCI: Torus/Pool Cooling Mode K4.06 - Knowledge of PRIMARY 223001 Primary CTMT and Aux. X CONTAINMENT SYSTEM AND AUXILIARIES design feature(s) and/or interlocks which provide for the following: 3.1 58 Maintains proper containment/secondary containment to drywell differential pressure 226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray Mode A2.19 - Ability to (a) predict the impacts 233000 Fuel Pool Cooling/Cleanup X of the following on the FUEL POOL COOLING AND CLEAN-UP ; and (b) based on those predictions, use procedures to correct, control, or 2.5 59 mitigate the consequences of those abnormal conditions or operations:
Inadequate system/pool chemistry 234000 Fuel Handling Equipment
K6.08 - Knowledge of the effect that a 239001 Main and Reheat Steam X loss or malfunction of the following will have on the MAIN AND REHEAT 3.3 60 STEAM SYSTEM: Main condenser vacuum 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Regulator K3.08 - Knowledge of the effect that a 245000 Main Turbine Gen. / Aux. X loss or malfunction of the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS will have on 3.7 61 following: Reactor/turbine pressure control system: Plant-Specific 256000 Reactor Condensate K5.03 - Knowledge of the operational 259001 Reactor Feedwater X implications of the following concepts as they apply to REACTOR FEEDWATER 2.8 62 SYSTEM : Turbine operation: TDRFP's-Only 268000 Radwaste A4.01 - Ability to manually operate 271000 Offgas X and/or monitor in the control room: Reset 2.8 63 system isolations 272000 Radiation Monitoring A3.06 - Ability to monitor automatic 286000 Fire Protection X operations of the FIRE PROTECTION 3.0 64 SYSTEM including: Fire dampers 288000 Plant Ventilation 290001 Secondary CTMT K4.01 - Knowledge of CONTROL ROOM 290003 Control Room HVAC X HVAC design feature(s) and/or interlocks which provide for the following: System 3.1 65 initiations/reconfiguration: Plant-Specific 290002 Reactor Vessel Internals K/A Category Point Totals: 2 0 1 2 2 1 0 1 1 1 1 Group Point Total: 12
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Grand Gulf Date of Exam: November, 2013 Category K/A # Topic RO SRO-Only IR Q# IR #
2.1.21 Ability to verify the controlled procedure copy. 3.5 66 1.
Conduct 2.1.27 Knowledge of system purpose and/or function. 3.9 67 of Operations Ability to locate and operate components, including local 2.1.30 4.4 68 controls.
2.1.40 Knowledge of refueling administrative requirements. 2.8 69 Ability to obtain and interpret station electrical and mechanical
- 2. 2.2.41 drawings. 3.5 70 Equipment Knowledge of the process used to track inoperable alarms.
Control 2.2.43 3.0 71 Knowledge of radiation exposure limits under normal or
- 3. 2.3.4 emergency conditions. 3.2 72 Radiation Control Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor 2.3.13 alarms, containment entry requirements, fuel handling 3.4 73 responsibilities, access to locked high-radiation areas, aligning filters, etc.
Ability to prioritize and interpret the significance of each
- 4. 2.4.45 annunciator or alarm. 4.1 74 Emergency Ability to diagnose and recognize trends in an accurate and Procedures / Plan 2.4.47 timely manner utilizing the appropriate control room reference 4.2 75 material.
Tier 3 Point Total 10
ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A SYSTEMS DELETED 201002 Reactor Manual Control System - This system is not incorporated into the BWR-6 design. The functions of this system are incorporated into the Rod Control and Information System.
201004 Rod Sequence Control System - This system is not incorporated into the BWR-6 design. The functions of this system are incorporated into the Rod Control and Information System.
201006 Rod Worth Minimizer System - This system is not incorporated into the BWR-6 design. The functions of this system are incorporated into the Rod Control and Information System.
214000 Rod Position Information System - This system is not incorporated into the BWR-6 design. The functions of this system are incorporated into the Rod Control and Information System.
215002 Rod Block Monitor System - This system is not incorporated into the BWR-6 design.
The functions of this system are incorporated into the Rod Control and Information System.
206000 High Pressure Core Injection (HPCI) - This system is not incorporated into the BWR 6 design.
207000 Isolation (Emergency) Condenser - This system is not incorporated into the BWR 6 design. This was replaced by the Mark III Containment Suppression Pool.
230000 RHR/LPCI: Torus/Pool Spray Mode - This system is not incorporated into the BWR 6 Mark III Containment design.
Written Exam Sample Plan: This sampling plan was developed using the manual sampling method described in ES401,Attachment 1 of NUREG1021, Revision 9, Supplement 1. It applies this same manual method throughout the sampling process, including the sampling of the Generic KAs listed on page 4 of ES401, section D.1.b, as well as any resampling that is required for rejected KAs (i.e., KA swaps).
Instead of tokens, the plan was developed using the web site random.org to generate the random number associated with each decision. Instead of bounding the possible selections by the number of tokens, the web site allows the user to specify the range of possible numbers for each choice.
ES-401 BWR Examination Outline - SRO Form ES-401-1 Facility: Grand Gulf Date of Exam: November, 2013 RO K/A Category Points SRO-Only Points Tier Group 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 20 4 3 7 Emergency &
Abnormal Plant 2 N/A N/A 7 2 1 3 Evolutions Tier Totals 27 6 4 10 1 26 2 3 5 2.
Plant 2 12 2 1 3 Systems Tier Totals 38 4 4 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 outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.
The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.
Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)
E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #
1 2 3 1 2 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 AA2.03 - Ability to determine and/or interpret the 295003 Partial or Complete Loss of AC / 6 X following as they apply to PARTIAL OR 3.5 76 COMPLETE LOSS OF A.C. POWER : Battery status: Plant-Specific 295004 Partial or Total Loss of DC Pwr / 6 295005 Main Turbine Generator Trip / 3 2.2.39 Knowledge of less than or equal to one hour 295006 SCRAM / 1 X Technical Specification action statements for 4.5 77 systems.
295016 Control Room Abandonment / 7 AA2.01 - Ability to determine and/or interpret the 295018 Partial or Total Loss of CCW / 8 X following as they apply to PARTIAL OR 3.4 78 COMPLETE LOSS OF COMPONENT COOLING WATER : Component temperatures 295019 Partial or Total Loss of Inst. Air / 8 295021 Loss of Shutdown Cooling / 4 295023 Refueling Acc / 8 295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 2.1.31 Ability to locate control room switches, 295026 Suppression Pool High Water X controls, and indications, and to determine that 4.3 79 Temp. / 5 they correctly reflect the desired plant lineup.
295027 High Containment Temperature / 5 EA2.03 - Ability to determine and/or interpret the 295028 High Drywell Temperature / 5 X following as they apply to HIGH DRYWELL 3.9 80 TEMPERATURE : Reactor water level 295030 Low Suppression Pool Wtr Lvl / 5 295031 Reactor Low Water Level / 2 EA2.04 - Ability to determine and/or interpret the 295037 SCRAM Condition Present X following as 4.1 81 and Reactor Power Above APRM they apply to SCRAM CONDITION PRESENT Downscale or Unknown / 1 AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Suppression pool temperature 295038 High Off-site Release Rate / 9 600000 Plant Fire On Site / 8 2.1.20 Ability to interpret and execute procedure 700000 Generator Voltage and Electric Grid X steps. 4.6 82 Disturbances / 6 K/A Category Totals: 4 3 Group Point Total: 7
ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO)
E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #
1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 295015 Incomplete SCRAM / 1 295017 High Off-site Release Rate / 9 AA2.04 - Ability to determine and/or interpret the 295020 Inadvertent Cont. Isolation / 5 & 7 X following as they apply to INADVERTENT 3.4 83 CONTAINMENT ISOLATION :Drywell/containment temperature 295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Wtr Lvl / 5 295032 High Secondary Containment Area Temperature / 5 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 EA2.01- Ability to determine and/or interpret the 295035 Secondary Containment High X following as they apply to SECONDARY 3.9 84 Differential Pressure / 5 CONTAINMENT HIGH DIFFERENTIAL PRESSURE:
2.2.42 - Ability to recognize system parameters that 295036 Secondary Containment High X are entry-level conditions for Technical Specifications. 4.6 85 Sump/Area Water Level / 5 500000 High CTMT Hydrogen Conc. / 5 K/A Category Point Totals: 2 1 Group Point Total: 3
ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (SRO)
System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 2.1.19 - Ability to use plant computers to 203000 RHR/LPCI: Injection X evaluate system or component status. 3.8 86 Mode 205000 Shutdown Cooling 206000 HPCI 207000 Isolation (Emergency)
Condenser 209001 LPCS 209002 HPCS A2.07 - Ability to (a) predict the impacts of the 211000 SLC X following on the STANDBY LIQUID 3.2 87 CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Valve closures 212000 RPS 215003 IRM 215004 Source Range Monitor A2.02 - Ability to (a) predict the impacts of the 215005 APRM / LPRM X following on the AVERAGE POWER RANGE 3.7 88 MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Upscale or downscale trips 217000 RCIC 218000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff 239002 SRVs 2.4.34 - Knowledge of RO tasks performed 259002 Reactor Water Level X outside the main control room during an 4.1 89 Control emergency and the resultant operational effects.
261000 SGTS 262001 AC Electrical Distribution 262002 UPS (AC/DC) 263000 DC Electrical Distribution 264000 EDGs 300000 Instrument Air
2.4.2 - Knowledge of system set points, 400000 Component Cooling X interlocks and automatic actions associated 4.6 90 Water with EOP entry conditions.
K/A Category Point Totals: 2 3 Group Point Total: 5
ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (SRO)
System # / Name K K K K K K A A A A G K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS A2.06 - Ability to (a) predict the impacts 201003 Control Rod and Drive X of the following on the CONTROL ROD 3.1 91 Mechanism AND DRIVE MECHANISM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of CRD cooling water flow 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-core Probe 215002 RBM 216000 Nuclear Boiler Inst.
A2.03 - Ability to (a) predict the impacts 219000 RHR/LPCI: Torus/Pool X of the following on the RHR/LPCI: 3.2 92 Cooling Mode TORUS/SUPPRESSION POOL COOLING MODE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve closures 223001 Primary CTMT and Aux.
226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Regulator 245000 Main Turbine Gen. / Aux.
256000 Reactor Condensate X 2.4.1 Knowledge of EOP entry 4.8 93 conditions and immediate action steps.
259001 Reactor Feedwater 268000 Radwaste
271000 Offgas 272000 Radiation Monitoring 286000 Fire Protection 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel Internals K/A Category Point Totals: 2 1 Group Point Total: 3
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Grand Gulf Date of Exam: November, 2013 Category K/A # Topic RO SRO-Only IR # IR #
Ability to perform specific system and integrated plant 2.1.23 4.4 94 procedures during all modes of plant operation.
Knowledge of the station's requirements for verbal
- 1. 2.1.38 3.8 95 communications when implementing procedures.
Conduct of Operations Subtotal 2 2.2.35 Ability to determine Technical Specification Mode of Operation. 4.5 96 2.
Equipment Ability to analyze the effect of maintenance activities, such as 2.2.36 4.2 97 Control degraded power sources, on the status of limiting conditions for operations.
Subtotal 2 Knowledge of radiation or contamination hazards that may arise
- 3. 2.3.14 during normal, abnormal, or emergency conditions or activities. 3.8 98 Radiation Control Subtotal 1 Knowledge of abnormal condition procedures.
- 4. 2.4.11 4.2 99 Emergency Knowledge of fire protection procedures.
2.4.25 3.7 100 Procedures / Plan Subtotal 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 Written Exam Sample Plan: This sampling plan was developed using the manual sampling method described in ES401,Attachment 1 of NUREG1021, Revision 9, Supplement
- 1. It applies this same manual method throughout the sampling process, including the sampling of the Generic KAs listed on page 4 of ES401, section D.1.b, as well as any resampling that is required for rejected KAs (i.e., KA swaps). Instead of tokens, the plan was developed using the web site random.org to generate the random number associated with each decision. Instead of bounding the possible selections by the number of tokens, the web site allows the user to specify the range of possible numbers for each choice.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Grand Gulf Nuclear Station Date of Examination: 11/11/2013 Examination Level: RO SRO Operating Test Number: LOT 2013 Administrative Topic Type Describe activity to be performed (see Note) Code*
Determine Surveillance Requirements for OPDRV Conduct of Operations GJPM-OPS-2013-AS1 R-N 2.1.23 (4.3/4.4), 2.1.25 (3.9/4.2), 2.1.40 (2.8/3.9)
Manual On-Line Risk Assessment Conduct of Operations R-B GJPM-OPS-2013-AS2 2.1.20 (4.6/4.6), 2.1.25 (3.9/4.2), 2.2.17 (2.6/3.8)
Determine Penetration Isolation Requirements Equipment Control R-N GJPM-OPS-2013-AS3 2.2.15 (3.9/4.3), 2.2.40 (3.4/4.7), 2.2.41 (3.5/3.9)
Authorize Emergency Exposure Radiation Control R-N GJPM-OPS-2013-AS4 2.3.4 (3.2/3.7)1 Emergency Event Classification Emergency Procedures/Plan S/R-N GJPM-OPS-2013-AS5 2.4.41 (2.9/4.6)
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: GRAND GULF NUCLEAR STATION Date of Examination: 11/8/2013 Exam Level: RO SRO-I SRO-U Operating Test No.: LOT-2013 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Code* Safety Function
- a. 202001 A2.04 (3.7/3.8), A2.26 (2.9/3.1), A3.07 (3.3/3.3) / A-N-S 1 Returning a Recirculation Loop to Service at Power, GJPM-OPS-2013CR1
- b. 217000 A4.04 (3.6/3.6) / RCIC Manual Startup, A-M-S 2 GJPM-OPS-2013CR2
- c. 241000 A2.06 (3.1/3.2) / Open Main Steam Isolation Valves, M-L-S 3 GJPM-OPS-2013CR3
- d. 205000 A4.03 (3.6/3.5) / Defeat SDC Injection Valve Isolation C-D-E-L-EN 4 Interlocks, GJPM-OPS-2013CR4
- e. 219000 A4.01 (3.8/3.7), A4.02 (3.7/3.5), A4.05 (3.4/3.4 / A-M-S 5 Startup Suppression Pool Cooling B GJPM-OPS-2013CR5
- f. 264000 A4.04 (3.7/3.7), / Start,Parallel and Load Div 1 D/G A-D-S 6 GJPM-OPS-2013CR6
- g. 201005 A2.03 (3.2/3.2), A2.04 (3.2/3.2), A2.06 (3.2/3.2), A2.07 C-D-L 7 3.2/3.2, / Bypass a Control Rod in RACS, GJPM-OPS-2013CR7 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- h. 212000 A2.01, (3.7/3.9), / RPS Motor Generator Startup D 7 GJPM-OPS-2013PS1
- i. 286000 A1.05 (3.2/3.2), / Perform Attachment IV of Shutdown D-E-L-R 4 from Remote Shutdown Panel ONEP, GJPM-OPS-2013PS2
- j. 262001 2.1.30 (4.4/4.0), 2.1.20 (4.6/4.6), / Placing LSS in D 6 Standby, GJPM-OPS-2013PS3
@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/ 8/ 4 (E)mergency or abnormal in-plant 1/ 1/ 1 (EN)gineered safety feature - / - / 1 (control room system)
(L)ow-Power / Shutdown 1/ 1/ 1 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)
(R)CA 1/ 1/ 1 (S)imulator
Appendix D Scenario Outline Form ES-D-1 Scenario 1 Page 1 of 2 Facility: Grand Gulf Nuclear Station Scenario No.: 1 Op-Test No.: LOT-2013 Examiners: ____________________________ Operators: _____________________________
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
Withdraw control rods per the start up control rod movement sequence.
Control rod data fault requiring entry of substitute position Control rod drift in.
Failure of ESF Transformer 11.
Main Steam Line leak outside containment with inability to isolate.
Scram Air Header failure to vent - ATWS below 5% power.
RCIC failure to automatically isolate.
Standby Liquid Control System piping failure.
Initial Conditions: Plant is operating at 5% power during start up. Reactor pressure is 400 psig.
Inoperable Equipment: none Turnover:
Plant is operating at 5% power during start up.
Reactor pressure is 400 psig.
Step 142of the start up control rod movement sequence has been completed.
Condensate Pumps A and C are in service.
Condensate Booster Pump A is in service.
Continue start up by performing steps 143 through 146 of the start up control rod movement sequence. Then, place Reactor Feed Pump A in service per IOI-1, step 6.2.13d.
o Reactor Feed Pump minimum flow controllers on 1H22-P171 have been verified to be in AUTO per IOI-1 step 6.2.13d.
o The walkdown of Reactor Feed Pumps per IOI-1 step 6.2.13d(1) has been completed satisfactorily.
This is a Division 1 work day.
Scenario Notes:
This is a new scenario.
Validation Time (60-90 min): 65 min Revision 2 7/08/13
Appendix D Scenario Outline Form ES-D-1 Scenario 1 Page 2 of 2 Event Malf. No. Event Type Event No. Description Withdraw control rods per the start up control rod 1 N (ATC) movement sequence. (03-1-01-1, Cold Shutdown to Generator Carrying Minimum Load; 04-1-01-C11-2, Rod Control and Information System; Control Rod Movement Sequence).
Control Rod 44-05 data fault resulting in a control rod 2 rci036 C (ATC) withdrawal block (ARI 04-1-02-1H13-P680-4A-C5; 04-1-01-C11-2, Rod Control and Information System, section 4.11).
C (ATC) z021021_60 Control Rod 60-45 drift in (05-1-02-IV-1, Control Rod/Drive 3 A (CREW)
-45 Malfunctions; ARI 04-1-02-1H13-P680-4A2-E4).
TS (CRS)
C (BOP) ESF Transformer 11 trip followed by trip of Division 1 Diesel r21134g 4 A (CREW) Generator (05-1-02-I-4, Loss of AC Power; 05-1-02-III-5, n41141a TS (CRS) Automatic Isolations, 02-S-01-27, Operations Philosophy).
Main Steam Line B leak in Auxiliary Building Steam Tunnel with failure/inability to isolate, ATWS - failure of scram air header to vent, power below 5%(EP-4, Auxiliary Building Control; EP-2A, ATWS RPV Control; 05-1-02-III-5, Automatic Isolations; 05 ms066b 02-I-1, Reactor Scram; 02-S-01-27 Operations Philosophy).
ms183b When the second area reaches its maximum safe ms184b temperature, the crew performs Emergency 5 c11167 M(CREW) Depressurization by opening 8 ADS/SRVs tte31n004a_d tte31n004b_d Before Emergency Depressurization, the crew terminates and prevents all injection except boron, CRD, and RCIC per EP-2A.
Following the ATWS, crew directs installation of EP Attachment 20 and inserts control rods by normal rod insertion per EP-2A step Q-1.
rf ATT03 RCIC failure to automatically isolate (ARI 04-1-02-1H13-P601-6 e51187a C (BOP) 21A-G3).
e51187b c41263 Standby Liquid Control System piping failure (SOI 04-1-01-C41-7 C (BOP) 1, Standby Liquid Control System, Attachment VI; EP-2, ATWS RPV Control, Attachment 28).
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec
- Critical Task (As defined in NUREG 1021 Appendix D)
Quantitative Attributes Table Normal Events 1 Abnormal Events 2 Reactivity Manipulations 0 Total Malfunctions 8 Instrument/Component Failures 4 EP Entries (Requiring substantive action) 2 Major Transients 1 EP Contingencies 1 Tech Spec Calls 2 Critical Tasks 3 Revision 2 7/08/13
Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 1 of 23 Simulator Setup:
A. Initialization
- 1. Log off all simulator PDS and SPDS computers (PDS and SPDS must come up after the simulator load for proper operation).
- 2. Startup the simulator using Simulator Instructors Job Aid section 6.3.
Note:
Prior to running the Schedule File, ensure no Event Files are Open. If an existing Event File is Open prior to running the Schedule File, then any associated Event Files will not automatically load.
- 3. Open Schedule.exe and Director.exe by clicking on the Icon in the Thunder Bar.
- 4. Set the Simulator to IC-193 and perform switch check (Using Quick Reset in Director).
- 5. Click on Open in the Schedule window and Open Schedule File 2013 NRC scenario 1.sch (in the Schedule Directory)
- 6. In Schedule window, click on the Stopped red block. The red block will change to a green arrow and indicate the scenario is active (Running).
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 2 of 23
- 7. Click the Summary tab in the Director window. Verify the schedule files are loaded and opened per Section B below. (Note: Any actions in the schedule file without a specific time will not load into the director until triggered.)
- 8. Take the simulator out of freeze.
- 10. Verify or perform the following:
IC-193 Power approximately 5%
Ensure CRD Pump B is in service.
Ensure the BOC startup movement sequence available at the P680 and marked up through step 142 complete.
Advance all chart recorders and ensure all pens inking properly.
Ensure APRM/IRM recorders are displaying IRM trend mode.
Clear any graphs and trends off of SPDS.
Markup IOI-1 through step 6.2.13c.
- 11. Run through any alarms and ensure alarms are on. (Note: On T-Rex, to verify alarms are ON, the indicator will indicate Alarms On).
- 12. Place the simulator in Freeze.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 3 of 23 B. File loaded verification:
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 4 of 23 Revision 2 7/08/13
Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 5 of 23 Crew Turnover:
B. Assign the candidates crew positions.
C. Turnover the following conditions:
Power 5%
Pressure 400 psig BOC EOOS GREEN Work Day Division 1 Plant is operating at 5% power during start up.
Reactor pressure is 400 psig.
Step 142 of the start up control rod movement sequence has been completed.
At IOI-1 step 6.2.13d.
Condensate Pumps A and C are in service.
Condensate Booster Pump A is in service.
Planned Evolutions this shift:
Immediately following turnover:
Continue start up by the ATC performing steps 143 through 146 of the start up control rod movement sequence.
o Note that an independent Reactivity Management SRO per Operations Philosophy 6.8.1 will not be provided for this scenario.
Then, place Reactor Feed Pump A in service per IOI-1, step 6.2.13d.
o Reactor Feed Pump minimum flow controllers on 1H22-P171 have been verified to be in AUTO per IOI-1 step 6.2.13d.
o The walkdown of Reactor Feed Pumps per IOI-1 step 6.2.13d(1) has been completed satisfactorily.
This is a Division 1 work day.
B. Allow the crew to perform pre-shift brief and review procedures for planned evolutions.
C. Bring the crew into the Simulator, place the simulator is in RUN.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 6 of 23 D. Allow the crew to walk down panels.
E. When the crew assumes the shift begin Scenario Activities.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 7 of 23 SCENARIO ACTIVITIES:
Start SBT report and any other required recording devices (Video recording not allowed for NRC exams).
Withdraw control rods per the start up control rod movement sequence:
The crew will withdraw control rods per the startup rod sequence. When control rod 44-05, in the second step of the pull sheets to be performed, is withdrawn past position 04, a control rod block due to a simulated failed rod position reed switch will occur (Auto Event 2).
Control Rod 44-05 data fault causing control rod withdrawal block The crew will respond using ARI 04-1-02-1H13-P680-4A-C5, CONT ROD WITHDRAWL BLOCK and 04-1-01-C11-2, Rod Control and Information System, section 4.11.
If contacted as I&C to investigate the data fault, respond there is nothing that can be determined from the Control Room and that a WO will be needed for any troubleshooting from cabinets inside Containment.
The data fault will automatically delete when rod 44-05 is withdrawn to position 08 (Auto Event 12).
The crew will resume control rod withdrawal. When rod 44-61 in the next pull step is withdrawn to position 12, then rod 60-45 will begin to drift in (Auto Event 3).
Control Rod 60-45 drift in (05-1-02-IV-1, Control Rod/Drive Malfunctions; ARI 04 02-1H13-P680-4A2-E4):
The crew will respond using ARI 04-1-02-1H13-P680-4A2-E4 and 05-1-02-IV-1, Control Rod/Drive Malfunctions and insert the control rod.
When requested to valves 103QL and 105 QL to isolate control rod 60-45(QL) at its HCU, wait 5 minutes, then delete drift malfunction z061061_60_45 by triggering Event 13, and report valves 103 QL and 105 QL have been closed.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 8 of 23 When ONEP actions have been addressed, the CRS has entered TS LCO 3.1.3 for rod 60-45, and any briefs are complete, at the Lead Evaluators discretion, insert malfunctions r21134g and n41141b by triggering Event 4 to cause ESF Transformer 11 to trip and DG11 to fail.
ESF Transformer 11 trip followed by trip of Division 1 Diesel Generator:
The crew will respond using 05-1-02-I-4, Loss of AC Power, and 05-1-02-III-5, Automatic Isolations.
If directed to investigate, as the building operator after approximately 3 minutes report DG11 is tripped and alarm 1H22P400-1A-E4, TRIP GENERATOR FAULT is sealed in, but there are no visible signs of a problem.
If requested to investigate, as Electrical Maintenance after approximately 5 minutes report a work order will be required for troubleshooting.
If directed to place DG11 in MAINTENANCE, wait 30 seconds, then, as the building operator at DG11 report you are ready to place DG11 in MAINTENANCE. When the control room operator reports the remote MAINTENANCE pushbutton is depressed, trigger Event 14 to insert remote function p75057 to simulate the local MAINTENANCE pushbutton on 1H22P400 being depressed and released.
If directed to rack out DG11 output breaker 152-1508, wait 5 minutes, then trigger Event 24 to insert remote function p75062, and report as the building operator you have racked out breaker 152-1508.
When the CRS has addressed LCO TS 3.8.1, and any transient briefs are complete, at the Lead Evaluators discretion, insert malfunction ms066b by triggering Event 5 to cause an unisolable steam leak from MSL B in the Aux Building steam Tunnel.
Main Steam Line B leak in Auxiliary Building Steam Tunnel with failure/inability to isolate, ATWS - failure of scram air header to vent, power below 5%, Standby Liquid Control System piping failure:
Main Steam Tunnel Temperature will rise to above the Max Safe value, lagged by RCIC Room Temperature.
The crew will respond using EP-4, Auxiliary Building Control; EP-2A, ATWS RPV Control; 05-1-02-III-5, Automatic Isolations; and 05-1-02-I-1, Reactor Scram.
Do NOT install EP Attachment 23.
Do NOT install EP Attachment 19 or 20 until the CRS has entered the Emergency Depressurization leg of EP-2A Revision 2 7/08/13
Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 9 of 23 Install other EP Attachments as requested.
Termination:
Once the crew has begun control rod insertion and is maintaining the reactor within the established level band, and as directed by the lead evaluator:
Place the simulator in Freeze and turn horns off.
Stop and save the SBT report and any other recording devices.
Instruct the crew to not erase any markings or talk about the scenario until after follow-up questions are asked.
Critical Tasks:
When the second area reaches its maximum safe temperature, the crew performs Emergency Depressurization by opening 8 ADS/SRVs Before Emergency Depressurization, the crew terminates and prevents all injection except boron, CRD, and RCIC per EP-2A.
Following the ATWS, crew directs installation of EP Attachment 20 and inserts control rods by normal rod insertion per EP-2A step Q-1.
Emergency Classification:
Site Area Emergency on FS1-RC3,PC3.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 10 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 1 / 2 Event
Description:
Withdraw control rods per the start up control rod movement sequence.
Control Rod 44-05 data fault causing control rod withdrawal block.
TIME Position Applicants Actions or Behavior Directs ATC to continue startup by performing steps 143 through 146 of the control rod movement sequence.
CRS Performs reactivity brief (This may have been done immediately before the crew entered the simulator.)
BOP Peer Check Control Rod Selection while monitoring reactor parameters.
Selects Control Rods per step 143 of the Control Rod Movement Sequence sheets. (Gang mode is allowed but is not used in the plant. Using either continuous or notch withdrawal is allowed.)
Step 143 Rods: 44-05, 20-61, 04-21, 60-45 from 00 to 12 Step 144 Rods: 20-05, 44-61, 60-21, 04-45 from 00 to 12 ATC For continuous withdraw, simultaneously DEPRESS and HOLD WITHDRAW and CONT WITHDRAW pushbuttons.
For notch withdraw MOMENTARILY DEPRESS WITHDRAW pushbutton and observe proper response.
(When step 143 is performed and 44-05 is withdrawn beyond position 04, a data fault will occur, producing a control rod withdrawal block)
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 11 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 1 / 2 (Cont.)
Event
Description:
Withdraw control rods per the start up control rod movement sequence.
Control Rod 44-05 data fault causing control rod withdrawal block.
TIME Position Applicants Actions or Behavior Recognizes and reports a control rod withdrawal block indicated by:
alarm 1H13-P680-4A-C5, CONT ROD WITHDRAWL BLOCK status light WITHDRAW BLOCK blinking on 1H13P680-6C status light WITHDRAWL INHIBIT blinking on 1H13P680-6C Refers to ARI and determines rod block is due to a data fault indicated by:
status light CH DISAGREE backlit on 1H13P680-6C Determines affected control rod is 44-05:
Recognizes in ALTERNATE data mode, display for 44-05 will ATC alternately flash 08 and FF If in gang mode, deselects gang mode by depressing DRIVE MODE on 1H13P680-6C Individually selects each control rod in that gang (the rods listed in pull step 144) and identifying WITHDRAW BLOCK status light is backlit when rod 44-05 is selected.
Determines affected channel is channel 2:
Selects individual data mode by depressing DATA MODE pushbutton on 1H13P680-6C, then depressing adjacent DATA SOURCE pushbutton and noting CH 2 DATA status light is backlit when WITHDRAW BLOCK status light is backlit. Informs CRS.
CRS Directs ATC to substitute data for rod 44-05 at position 08.
Substitutes data for rod 44-05 at position 08 in accordance with 04-1-01-C11-2, Rod Control and Information System, section 4.11, on 1H13P680-6C:
Ensures individual drive mode is selected Deselects raw data by depressing RAW DATA pushbutton Ensures rod 44-05 is selected ATC Presses and releases ENTER SUBST pushbutton Presses SUBST POS and verifies red LED is lit beside rod 44-05 on RC&IS display on 1H13P680-6D Notifies I&C of data fault and entry of substitute position Selects alternating data by pressing DATA MODE pushbutton Selects raw data by pressing RAW DATA pushbutton.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 12 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 1 / 2 (Cont.)
Event
Description:
Withdraw control rods per the start up control rod movement sequence.
Control Rod 44-05 data fault causing control rod withdrawal block.
TIME Position Applicants Actions or Behavior Directs continuing control rod withdrawal per the startup sequence through step CRS 146.
Resumes control rod withdrawal step 144.
ATC (When rod 44-61 is withdrawn to position 12 in step 144, rod 60-45 will begin to drift in.)
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 13 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 3 Event
Description:
Control Rod 60-45 drift in TIME Position Applicants Actions or Behavior Recognizes and reports Control Rod 60-45 is drifting in as indicated by:
Annunciator P680-4A2-E4, CONT ROD DRIFT Determine Control Rod(s) that have drifted by depressing the ROD DRIFT pushbutton and observe the red LEDs of the drifted Control Rod(s) on the Control Rod Display Module. (from annunciator ARI).
Direction of drift may be noted by observing control rod position compared to as-left position in the previous pull sheet step but is not necessary.
ATC Applies continuous insert signal:
Selects rod 60-45 and Selects individual drive mode by pressing DRIVE MODE pushbutton Presses and holds IN TIMER SKIP pushbutton to fully insert rod 60-45.
This pushbutton should be held until rod 60-45 is hydraulically isolated if the direction of drift was NOT noted.
Acknowledges drift alarm using the RESET DRIFT pushbutton on P680 RC&IS (from annunciator ARI)
Enter Control Rod/Drive Malfunctions, 05-1-02-IV-1:
Directs the ATC to apply continuous insert signal until Control Rod reaches position 00.
CRS Directs the building operator to hydraulically isolate rod 60-45 by closing valves 103QL and 105QL at its HCU.
Directs the ATC to acknowledge the ROD DRIFT in accordance with the ARI 1H13P680-4A2-E4)
When building operator reports HCU 60-45 has been hydraulically isolated, releases IN TIMER SKIP, if held due to not knowing the direction of drift, and ATC resets drift alarm using the RESET DRIFT pushbutton on P680-6C (from annunciator ARI P680-4A2-E4)
Notifies Work Management and Reactor Engineering of event and requests assistance formulating troubleshooting plan and new reactivity management plan.
CRS Enters LCO 3.1.3 Condition C for control rod 60-45.
Conducts transient brief.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 14 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 4 Event
Description:
ESF Transformer 11 trip followed by trip of Division 1 Diesel Generator TIME Position Applicants Actions or Behavior Recognize and report ESF Transformer 11 Lockout as indicated by:
Annunciator P807-4A-B5, ESF XFMR 11 LOCKOUT TRIP BOP Annunciator P807-4A-F2, ESF XFMR 11 TROUBLE There are also other alarms associated with the loss of power.
Enter 05-1-02-I-4, Loss of AC Power, and 05-1-02-III-5, Automatic Isolations.
CRS Direct the BOP to Reenergize 15AA using an alternate feeder breaker and Reenergize 15B42.
Recognize and report Division 1 Diesel Generator automatically re-energizes bus 15AA after approximately 7 seconds:
Indicating lights above handswitches for Division 1 AC loads on 1H13P601/P870 lit.
alarm 1H13P864-1A-A3, 4.16KV BUS 15AA UNDERVOLTAGE clears when acknowledged.
bus 15AA voltmeter 1R21R615A on 1H13P864-1B indicates ~4.16kv.
status light DG 11 READY TO LOAD LIT ON 1H13P864-1B lit
( ~ 5 seconds after DG11 re-energizes bus 15AA, DG11 will trip.)
Recognizes and reports trip of DG11 indicated by:
alarm 1H13P864-1A-A3, 4.16KV BUS 15AA UNDERVOLTAGE and BOP Div 1 LCC undervoltage alarms P864-1A-D3/4,E3/4, F3/4 alarm 1H13P864-1A-B1, DIV 1 DSL GEN TRIP Indicating lights above handswitches for Division 1 AC loads on 1H13P601/P870 extinguish after approximately 12 seconds, alarm 1H13-P864-1A-H1, DIV 1 LSS SYS FAIL Reenergizes 15AA with an alternate feeder breaker using one of the following sources:
ESF 12 via 152-1511 ESF 21 via 152-1501 Solicits cause of DG11 trip from local building operator. Reports to CRS.
CRS Direct the BOP restore instrument air to containment.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 15 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 4 (Cont.)
Event
Description:
ESF Transformer 11 trip followed by trip of Division 1 Diesel Generator TIME Position Applicants Actions or Behavior BOP Opens 1P53F001, INST AIR SPLY HDR TO CTMT, on 1H13P870-3C.
Direct the BOP to perform applicable subsequent actions of 05-1-02-I-4, Loss CRS of AC Power.
Performs applicable subsequent actions of 05-1-02-I-4, Loss of AC Power.
Reenergizes 15B42 via 52-15405 using handswitch MCC 15B42 FDR FM LCC 15BA4 on 1H13P864-1C.
Resets NSSSS logic by depressing NSSSS INBD(OTBD) ISOL RESET pushbuttons on 1H13P601-18B(19B).
BOP Opens 1P41F239, PSW INL TO ESF RM CLRS A on 1H13P870-1C.
Opens 1P41F240, PSW OUTL TO ESF RM CLRS A on 1H13P870-1C.
Notifies Plant Chemistry to ensure SGTS Sping and AXM Rad Monitors are functioning correctly.
Solicits cause of DG11 trip from local building operator. Reports to CRS.
Direct the BOP to restore systems from isolations per 05-1-02-III-5, Automatic Isolations subsequent actions. May elect to start Standby Gas Treatment CRS System A to maintain Aux Building negative pressure pending restoration of normal ventilation systems.
Restores systems from isolations due to loss of Div 1 power per 05-1-02-III-5, Automatic Isolations Att. I (hard card) and subsequent actions and respective system SOIs:
P45, Floor and Equipment Drains System P11, Condensate and Refueling Water Transfer System T41, Auxiliary Building Ventilation System BOP T42, Fuel Pool Ventilation System P52, Service Air System P21, Makeup Water Treatment System P66, Domestic Water System If directed, starts SGTS A by depressing SGTS DIV 1 MAN INIT LOGIC A and LOGIC C pushbuttons on 1H13P870-2B.
(Not all systems will be restored during the timeframe of this scenario.)
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 16 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 4 (Cont.)
Event
Description:
ESF Transformer 11 trip followed by trip of Division 1 Diesel Generator TIME Position Applicants Actions or Behavior Recognizes and reports loss of Fuel Pool Cooling system:
FPCCU trouble alarms 1H13P680-4A2-B6, C7, D6, D7 ATC Recognizes and reports loss of Reactor Water Clean-Up system:
Multiple RWCU trouble alarms 1H13P680-11A and handswitch indication of both RWCU pumps tripped on 1H13P680-11C.
Enters 05-1-02-III-1, Inadequate Decay Heat Removal for loss of spent fuel pool cooling.
CRS (The CRS may direct frequent monitoring of spent fuel pool temperature but is not expected to formulate plans for recovery of FPCCU or RWCU within the time frame of this scenario.)
Notifies Work Management of event and requests assistance formulating troubleshooting plan for ESF Transformer 11 and DG11.
Enters LCO 3.8.1 Condition B for DG11. If scenario timing allows, enters potential LCO 3.8.1 condition A for one offsite AC supply, ESF11, to bus CRS 15AA, and enters potential LCO for TR SR 6.4.1.5 for loss of RWCU influent conductivity monitor, one of two continuous conductivity monitors.
Conducts transient brief.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 17 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 5 Event
Description:
Main Steam Line B leak in Auxiliary Building Steam Tunnel with failure/inability to isolate, ATWS - failure of scram air header to vent, power below 5%, Standby Liquid Control System piping failure TIME Position Applicants Actions or Behavior Recognizes and reports Main Steam Line Tunnel Temperature rising::
Alarm 1H13P601-19A-E3, MN STM TNL AMBIENT TEMP HI.
BOP Alarm 1H13P601-19A-F3, MN STM TNL dT HI Directs checking Aux Building Main Steam Tunnel temperature on PDS or CRS backpanel 1H13P642.
Retrieves and reports MSL Tunnel temperature and trend from PDS EP-4 ATC guide/display.
Retrieves and reports MSL Tunnel temperature and trend from NUS BOP temperature switch 1E31N604B or N604D on 1H13P642.
Recognizes and reports Main Steam Line Tunnel Temperature exceeding isolation setpoint and EP-4 entry condition (185°F):
Alarm 1H13P601-19A-A3(A4), MSL PIPE TNL CH-A(D) TEMP HI/INOP.
Alarm 1H13P601-18A-A3(A4), MSL PIPE TNL CH-B(C) TEMP HI/INOP Recognizes and reports Main Steam Line B failed to automatically isolate by observing MSIVs 1B21F022B and F028B full open indication on 1H13P601-BOP 18C and 19C:
Places handswitches MSL B DRWL INBD ISOL for 1B21F022B to CLOSE on 1H13P601-18C and observes valve remains open.
Places handswitches MSL B DRWL OTBD ISOL for 1B21F028B to CLOSE on 1H13P601-19C and observes valve remains open.
Reports inability to close 1B21F022B and F028B.
Enters EP-4 on MSL Tunnel Temperature high:
Announces evacuation of the Auxiliary Building on PA system.
CRS Enters EP-2 from EP-4 due to unisolable steam discharge outside primary containment.
Directs manual scram.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 18 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 5 (Cont.)
Event
Description:
Main Steam Line B leak in Auxiliary Building Steam Tunnel with failure/inability to isolate, ATWS - failure of scram air header to vent, power below 5%, Standby Liquid Control System piping failure TIME Position Applicants Actions or Behavior BOP Recognizes and reports loss of switchyard/offsite power.
Provides a scram report:
Reactor Mode SW in SHUTDOWN.
All Rods are NOT Inserted (scram air header failure to vent ATWS).
Reactor power (~5%, Value depends when data taken).
ATC Reactor water level and trend.
Reactor pressure and trend.
Condensate/Feedwater is available.
Bypass valves and MSIVs are available (due to MSL B failure to isolate).
Enter EP-2A, ATWS RPV CONTROL, from EP-2 and 05-1-02-I-1, Reactor Scram.
Direct actions of EP-2A, ATWS RPV CONTROL steps 1 - 4:
CRS Directs ATC to initiate ATWS ARI/RPT.
Directs BOP to initiate and Override HPCS injection.
ATC Initiates ATWS ARI/RPT by arming and depressing ATWS/ARI CHANNEL 1 and CHANNEL 2 initiation pushbuttons on 1H13P680-3C.
Inhibits ADS on 1H13P601-19B.
Place ADS A and ADS B keylock switches to INHIBIT Override HPCS injection on 1H13P601-16C.
BOP Holds the HPCS pump handswitch in the STOP position.
Arms and depresses HPCS MAN INIT pushbutton on 1H13P601-16B Place the E22-F004, HPCS injection valve, handswitch to the CLOSE position.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 19 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 5 / 6 (Cont.)
Event
Description:
Main Steam Line B leak in Auxiliary Building Steam Tunnel with failure/inability to isolate, ATWS - failure of scram air header to vent, power below 5%, Standby Liquid Control System piping failure TIME Position Applicants Actions or Behavior Enter EP-2A step L-6 and direct the ATC to maintain level band 11.4 to 53.5 on Startup Level Control.
Direct BOP to initiate and override low pressure ECCS.
Enter EP-2A step Q-4 and direct SLC injection before Suppression Pool CRS temperature reaches 110°F. (SLC will not inject, Attachment 28 is required)
Enter EP-2A step P-4 and direct the BOP to stabilize pressure (should maintain pressure ~400 psig) using IPC and BCV Manual Jack.
Call for EP Attachments 12, 20, and 23.
Initiates Div 1 and 2 ECCS and overrides LPCS and LPCI A/B/C on 1H13P601:
Arms and depresses LPCS/RHR A MAN INIT and RHR B/RHR C MAN INIT pushbuttons.
Places handswitches for LPCS, RHR A, RHR B, and RHR C pumps to stop.
BOP Places handswitches for injection MOVs 1E21F005, 1E12F042A, 1E12F042B, and 1E12F042C to close.
Verifies amber override alarms for LPCS, RHR A, RHR B, and RHR C pump and injection MOV annunciate and seal in.
(This is optional for EP-2A step L-6 and may not be performed until the Emergency Depressurization leg of EP-2A is entered.)
Recognizes and reports RCIC room temperature high and failure of RCIC to automatically isolate:
(Event 6) 1H13P601-21A-G3, RCIC EQUIP AREA TEMP HI/INOP BOP Manually closes 1E51F063, RCIC STM SPLY DRWL INBD ISOL, and 1E51F064, RCIC STM SUPLY DRWL OTBD ISOL, on 1H13P601-21C.
Recognizes and reports E51F063 and F064 loss of power upon stroke.
CRS Directs monitoring EP-4 parameters as a critical parameter.
Monitors EP-4 area temperatures, radiation levels, and water levels using PDS and or backpanel NUS temperature switches and area radiation monitors.
BOP/
Reports Main Steam Tunnel temperature above Max Safe limit, 250°F.
ATC Recognizes and reports RCIC room temperature above Max Safe limit, 212°F.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 20 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 5 (Cont.)
Event
Description:
Main Steam Line B leak in Auxiliary Building Steam Tunnel with failure/inability to isolate, ATWS - failure of scram air header to vent, power below 5%, Standby Liquid Control System piping failure TIME Position Applicants Actions or Behavior Enters Emergency Depressurization leg of EP-2A from EP-4:
Directs termination and prevention of all injection except RCIC, SLC, CRS and CRD.
Before Emergency Depressurization, the crew terminates and prevents all injection except boron, CRD, and RCIC per EP-2A.
Terminates and prevents Feedwater by Closing/ verifying closed the Startup ATC Level Control valve, N21-F040, N21-F009A, and N21-F009B on 1H13P680-1C.
Verifies HPCS overridden, and if previously overridden, verifies LPCS, RHR A, RHR B, and RHR C pumps overridden off and injection valves overridden closed on 1H13P601-16C, 21C, 20C, 17C; Or if not previously performed, initiates Div 1 and 2 ECCS and overrides LPCS and LPCI A/B/C on 1H13P601:
Arms and depresses LPCS/RHR A MAN INIT and RHR B/RHR C MAN INIT pushbuttons.
BOP Places handswitches for LPCS, RHR A, RHR B, and RHR C pumps to stop.
Places handswitches for injection MOVs 1E21F005, 1E12F042A, 1E12F042B, and 1E12F042C to close.
Verifies amber override alarms for LPCS, RHR A, RHR B, and RHR C pump and injection MOV annunciate and seal in.
CRS Directs opening 8 ADS/SRVs.
Opens 8 ADS/SRVs at 1H13P601-19C.
When the second area reaches its maximum safe temperature, the BOP crew performs Emergency Depressurization by opening 8 ADS/SRVs Revision 2 7/08/13
Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 21 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 5 / 7 (Cont.)
Event
Description:
Main Steam Line B leak in Auxiliary Building Steam Tunnel with failure/inability to isolate, ATWS - failure of scram air header to vent, power below 5%, Standby Liquid Control System piping failure TIME Position Applicants Actions or Behavior When reactor pressure decreases to MSCP (206 psig), directs ATC to slowly inject using Condensate to re-establish level band 11.4-53.5.
CRS (Initial injection rate should not exceed 4000 gpm, with incremental increases not to exceed 100 gpm each. However, 4000 gpm may not be needed initially due to low power ATWS.)
When reactor pressure decreases to MSCP (206 psig), injects using Startup ATC Level Control to restore and maintain level band 11.4-53.5.
If directed, initiates SBLC A and B using 04-1-01-C41-1, Standby Liquid Control System, Attachment VI (Initiation of Standby Liquid Control hard card) on 1H13P601-19B(18B):
Insert keys and turn SBLC Pmp A(B) pump key switch to START.
Verify system initiation by observing the following:
o F004A(B) SQUIB valves fired:
White SQUIB valve ready light OFF Annunciator SLC SYS A OOSVC (P601-19A-H1)
Amber status light SQUIB A LOSS CONT or PWR LOSS is ON.
o C41-F001A(B) tank outlet valves are open.
Event 7 o SBLC Pump A(B) running.
BOP o RWCU Isolated G33-F004, F001, F251 Closed o Verify SBLC A(B) is injecting into the RPV by observing the following:
SBLC pump discharge pressure exceeds reactor pressure. (SBLC piping failure will manifest in the form of low discharge pressure on meter 1C41R600 on 1H13P601-19B.)
SBLC tank level lowering.
After placing SLC A/B key switches to start, reports failure of SLC to develop adequate discharge pressure.
CRS Calls for EP Attachment 28.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 1 Page 22 of 23 Op-Test No: 2013 Scenario No: 1 Event No: 5 (Cont.)
Event
Description:
Main Steam Line B leak in Auxiliary Building Steam Tunnel with failure/inability to isolate, ATWS - failure of scram air header to vent, power below 5%, Standby Liquid Control System piping failure TIME Position Applicants Actions or Behavior CRS Directs BOP to maximize CRD flow.
BOP Maximizes CRD for flow: 04-1-01-C11-1 Att.
Places CRD SYS FLO CONT C11-R600 in MANUAL.
Using CRD SYS FLOW CONT C11-R600, fully opens C11-F002A(B),
Fully opens C11-F003, CRD DRIVE WTR PRESS CONT VLV When directed, maximizes CRD A for flow.
Re-energizes 15B42 on 1H13P864-1C Starts CRD PUMP A AUX OIL PMP on 1H13P601-22C ATC Starts CRD PMP A on 1H13P601-22C Although there are other actions per the procedure, only these listed will accomplish anything with scram sealed in and loss of offsite power.
CRS When Attachment 20 is reported installed, directs ATC to insert control rods and BOP to maximize CRD drive water pressure.
BOP Maximizes CRD drive water pressure by fully closing C11-F003, CRD DRIVE WTR PRESS CONT VLV on 1H13P601-22C BOP/ Inserts control rods by selecting rods and depressing IN TIMER SKIP or INSERT pushbutton on 1H13P680-6C.
ATC Following the ATWS, crew directs installation of EP Attachment 20 and inserts control rods by normal rod insertion per EP-2A step Q-1.
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Give this page to the CRS Turnover the following conditions:
Power 5%
Pressure 400 psig BOC EOOS GREEN Work Day Division 1 Plant is operating at 4.5% power during start up.
Reactor pressure is 400 psig.
Step 142 of the start up control rod movement sequence has been completed.
At IOI-1 step 6.2.13d.
Condensate Pumps A and C are in service.
Condensate Booster Pump A is in service.
Planned Evolutions this shift:
Immediately following turnover:
Continue start up by the ATC performing steps 143 through 146 of the start up control rod movement sequence.
o Note that an independent Reactivity Management SRO per Operations Philosophy 6.8.1 will not be provided for this scenario.
Then, place Reactor Feed Pump A in service per IOI-1, step 6.2.13d.
o Reactor Feed Pump minimum flow controllers on 1H22-P171 have been verified to be in AUTO per IOI-1 step 6.2.13d.
o The walkdown of Reactor Feed Pumps per IOI-1 step 6.2.13d(1) has been completed satisfactorily.
This is a Division 1 work day.
Revision 1 4/22/12 Page 23 of 23
Appendix D Scenario Outline Form ES-D-1 Scenario 2 Page 1 of 2 Facility: Grand Gulf Nuclear Station Scenario No.: 2 Op-Test No.: LOT-2013 Examiners: ____________________________ Operators: _____________________________
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
Return Condensate Pump B and Condensate Booster Pump B to service.
Control Rod Drive Pump A trip resulting in CRD HCU pressure fault.
RHR pump A trip while operating in Suppression Pool Cooling.
Feedwater Master Level Controller output failure high.
Feedwater Line B rupture in Turbine Building.
ATWS - hydraulic block power above 5%.
Standby Liquid Control System squib valves fail to immediately actuate (actuate after a delay, when Emergency Depressurization is commenced).
RCIC failure to automatically start on Level 2 Initial Conditions: Plant is operating at 75% power. Condensate Pump B and Condensate Booster Pump B are in standby following maintenance on Condensate Pump B. RHR A is operating in Suppression Pool Cooling mode. Standby Service Water A is in operation for weekly chemical addition.
Inoperable Equipment: LPCI A is inoperable due to RHR A is operating in Suppression Pool Cooling.
Turnover:
Plant is operating at 75% power.
Core flow is 70 mlbm/hr with operation in the OPRM Trip Enabled region of the Power-Flow Map.
Condensate Pump B and Condensate Booster Pump B are in standby following maintenance on Condensate Pump B.
RHR A is operating in Suppression Pool Cooling.
This is a Division 1 work day.
Immediately following the brief The ATC will return Condensate Pump B and Condensate Booster Pump B to service in accordance with 04-1-01-N19-1 section 4.3 and 03-1-01-2 step 6.6.
All prerequisites of step 4.3.1 have been completed.
Steps 4.3.2a(1), (2), and (3) have been completed.
After Condensate Pump B and Condensate Booster Pump B have been returned to service, and before Suppression Pool temperature decreases below 75°F, shut down RHR A Suppression Pool Cooling.
Scenario Notes:
This is a new scenario.
Validation Time (60-90 min): 70 min Revision 2
Appendix D Scenario Outline Form ES-D-1 Scenario 2 Page 2 of 2 Event Malf. No. Event Type Event No. Description Return Condensate Pump B and Condensate Booster Pump B 1 N (ATC) to service. (03-1-01-2, Power Operations, step 6.6; 04-1 N19-1, Condensate System, section 4.3.2).
Control Rod Drive Pump A trip resulting in CRD HCU 32-17 low C (BOP/ATC) 2 c11028a A (CREW) pressure fault that will not clear (05-1-02-IV-7, Control z024_024_32_17 Rod/Drive Malfunctions; ARI 04-1-02-1H13-P680-4A2-D4; TS TS (CRS) 3.1.5, TR 3.1.5)
RHR pump A trip while operating in Suppression Pool Cooling.
C (BOP) (ARI 04-1-02-1H13-P601-20A-A4,C4,H6; SOI 04-1-01-E12-1, 3 e12050a TS (CRS) Residual Heat Removal System; TS 3.5.1, TS 3.6.1.7, TS 3.6.2.3).
4 fw127 C (ATC) Feedwater Master Level Controller output failure high (05-1 A (CREW) V-7, Feedwater System Malfunctions).
Feedwater Line B rupture in Turbine Building, ATWS - hydraulic block power above 5%(EP-2A, ATWS RPV Control; 05-1-02-I-1, Reactor Scram; 05-1-02-I-2, Turbine Generator Trips; 02-S 27, Operations Philosophy).
When reactor water level cannot be restored and maintained above -191 cfz using Table 1 systems, the crew performs Emergency Depressurization by opening 8 ADS/SRVs.
fw070b 5 M(CREW) Before Emergency Depressurization, the crew terminates c11164 and prevents all injection except boron, CRD, and RCIC per EP-2A.
When reactor pressure decreases to MSCP, 219 psig, the crew commences and slowly raises injection using RHR B via E12F053B, SDC Return to Feedwater, to restore and maintain RPV level to greater than -191". Criterion is to give the highest priority to restore RPV level greater than -191".
Standby Liquid Control System squib valves fail to immediately actuate (actuate after a delay, when Emergency c41f004a_a C (BOP) Depressurization is commenced) (SOI 04-1-01-C41-1, Standby c41f004b_a Liquid Control System, Attachment VI; EP-2, ATWS RPV Control, Attachment 28).
6 e51043 I (ATC) RCIC failure to automatically start (04-1-01-E51-1 Att. VI)
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec
- Critical Task (As defined in NUREG 1021 Appendix D)
Quantitative Attributes Table Normal Events 1 Abnormal Events 2 Reactivity Manipulations 0 Total Malfunctions 8 Instrument/Component Failures 4 EP Entries (Requiring substantive action) 2 Major Transients 1 EP Contingencies 1 Tech Spec Calls 2 Critical Tasks 3 Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 1 of 20 Simulator Setup:
A. Initialization
- 1. Log off all simulator PDS and SPDS computers (PDS and SPDS must come up after the simulator load for proper operation).
- 2. Startup the simulator using Simulator Instructors Job Aid section 6.3.
Note:
Prior to running the Schedule File, ensure no Event Files are Open. If an existing Event File is Open prior to running the Schedule File, then any associated Event Files will not automatically load.
- 3. Open Schedule.exe and Director.exe by clicking on the Icon in the Thunder Bar.
- 4. Set the Simulator to IC-194 and perform switch check (Using Quick Reset in Director).
- 5. Click on Open in the Schedule window and Open Schedule File 2013 NRC Scenario 2.sch (in the Schedule Directory)
- 6. In Schedule window, click on the Stopped red block. The red block will change to a green arrow and indicate the scenario is active (Running).
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 2 of 20
- 7. Click the Summary tab in the Director window. Verify the schedule files are loaded and opened per Section B below. (Note: Any actions in the schedule file without a specific time will not load into the director until triggered.)
- 8. Take the simulator out of freeze.
- 10. Verify or perform the following:
IC-194 APRMs are turned on (4,1,2,3)
Ensure the BOC rod movement sequence available at the P680.
Advance all chart recorders and ensure all pens inking properly.
Clear any graphs and trends off of SPDS.
Ensure RHR A is aligned in Suppression Pool Cooling.
Ensure SSW A is aligned for chemical addition through all loads (through step 7.6 of 04-1-03-P41-1).
Marked up copy of 03-1-01-2 Att. VIII, Temporary Downpower, is placed on CRS desk.
Mark up a copy of 04-1-01-P41-1, SSW A Chemical Addition Run, through step 7.6 and place on CRS desk.
Mark up a copy of 04-1-01-E12-1 through step 5.2.2a(7) and place on CRS desk.
- 11. Run through any alarms and ensure alarms are on. (Note: On T-Rex, to verify alarms are ON, the indicator will indicate Alarms On).
- 12. Place the simulator in Freeze.
B. File loaded verification:
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 3 of 20 Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 4 of 20 Crew Turnover:
B. Assign the candidates crew positions.
C. Turnover the following conditions:
Power 75%
BOC EOOS GREEN Core flow is 70 mlbm/hr with operation in the OPRM Trip Enabled region of the Power-Flow Map.
Condensate Pump B and Condensate Booster Pump B are in standby following maintenance on Condensate Pump B.
RHR A is operating in Suppression Pool Cooling.
LCO 3.5.1 was entered one hour ago for LPCI A due to RHR A Test Return to Suppression Pool valve 1E12F024A open for Suppression Pool Cooling.
SSW A is aligned per 04-1-03-P41-1, SSW A Chemical Addition Run, and is supporting RHR A Suppression Pool Cooling.
This is a Division 1 work day.
Immediately following the brief:
The ATC will return Condensate Pump B and Condensate Booster Pump B to service in accordance with 04-1-01-N19-1 section 4.3 and 03-1-01-2 step 6.6.
All prerequisites of step 4.3.1 have been completed.
Steps 4.3.2a(1), (2), and (3) have been completed.
After Condensate Pump B and Condensate Booster Pump B have been returned to service, and before Suppression Pool temperature decreases below 75°F, shut down RHR A Suppression Pool Cooling.
Reactor Engineering is revising the Reactivity Management Plan for power ascension to 100%, which is scheduled to begin on this shift.
Note that an independent Reactivity Management SRO per Operations Philosophy 6.8.1 will not be provided for this scenario.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 5 of 20 D. Allow the crew to perform pre-shift brief and review procedures for planned evolutions.
E. Bring the crew into the Simulator, place the simulator is in RUN.
F. Allow the crew to walk down panels.
G. When the crew assumes the shift begin Scenario Activities.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 6 of 20 SCENARIO ACTIVITIES:
Start SBT report and any other required recording devices (Video recording not allowed for NRC exams).
Return Condensate Pump B and Condensate Booster Pump B to service.
The crew will place Condensate Pump B and Condensate Booster Pump B in service in accordance with 04-1-01-N19-1, Condensate System, section 4.3.2 and 03-1-01-2, Power Operations, step 6.6.
If requested to first isolate and later unisolate the local discharge pressure gauges for Condensate Pump B (gauge 1N19R002B) and Condensate Booster Pump B (gauge 1N19R005B), as the building operator wait one minute, then report the associated gauge is isolated/unisolated.
When the crew has completed all required steps 04-1-01-N19-1, Condensate System, section 4.3.2 and at the direction of the lead evaluator, trigger Event 2 to cause CRD Pump A trip and HCU 32-17 fault.
Control Rod Drive Pump A trip resulting in CRD HCU 32-17 low pressure fault that will not clear.
The crew will start CRD pump B in accordance with 05-1-02-IV-7, Control Rod/Drive Malfunctions.
If requested to investigate CRD Pump A trip and check CRD Pump B for proper operation, as the building operator wait 5 minutes, then report there is no obvious reason locally why CRD Pump A tripped, and CRD B is running normally.
If directed to check CRD Pump A breaker 152-1505 for trip indication locally, as the building operator wait 5 minutes, then report breaker 152-1505 is tripped and instantaneous overcurrent flags are present.
If requested to investigate, as Electrical Maintenance wait 5 minutes, then report you will not be able to diagnose anything without a work order, since troubleshooting will have to be intrusive.
The crew will respond to alarm 1H13-P680-4A2-D4, HCU TROUBLE.
When directed to investigate HCU 32-17(HD) trouble, as the building operator wait 5 minutes, then report by PA system HCU accumulator pressure is 1500 psig. If directed to recharge HCU 32-17, acknowledge the direction, but this will not be simulated within the time frame of this scenario.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 7 of 20 If asked, as Reactor Engineering state control rod 32-17 passed its last scram time test, with a scram time that met all limits of TS Table 3.1.4-1.
When the crew has completed all required steps of 05-1-02-IV-7, Control Rod/Drive Malfunctions and ARI 04-1-02-1H13-P680-4A2-D4, and the CRS has addressed TS 3.1.5 for HCU 32-17, and at the direction of the lead evaluator, trigger Event 3 to cause RHR pump A to trip.
RHR Pump A Trip:
RHR Pump A will trip while RHR A Test Return to Suppression Pool valve 1E12F024A is open, causing RHR A piping to drain to the Suppression Pool.
The crew will respond per 04-1-02-1H13-P601-20A-A4,C4,H6 and will realign RHR A valves to standby per SOI 04-1-01-E12-1, Residual Heat Removal System, to stop RHR A piping from draining.
If requested to inspect RHR pump A, as the building operator wait 5 minutes and report there are no obvious signs locally at the pump of why it tripped.
If requested to check RHR pump A breaker 152-1509, as the building operator wait 5 minutes and report the lockout relay is tripped and time-overcurrent flags are indicated at the breaker.
If requested to check RHR pump A breaker 152-1509, as Electrical Maintenance wait 5 minutes and say RHR pump A motor should be meggared, which will require a tagout and work order.
When the crew has closed 1E12F024A and the CRS has addressed TS 3.6.1.7 and TS 3.6.2.3, and/or at the direction of the lead evaluator, trigger Event 4 to cause the Feedwater Master Level Controller output to slowly fail high, raising level.
Feedwater Master Level Controller output failure high:
Feedwater Master Level Controller output will begin to rise, causing reactor water level to slowly rise. The crew will respond by taking manual control of the Master Level Controller in accordance with 05-1-02-V-7, Feedwater System Malfunctions.
If asked to investigate, as I&C wait 5 minutes, then report to the control room.
Respond that a work order will be required for any troubleshooting.
After the crew has stabilized level at ~36 in manual control and/or at the direction of the lead evaluator, trigger Event 5 to cause Feedwater line B in the Turbine Building to rupture.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 8 of 20 Feedwater Line B rupture in Turbine Building, ATWS - hydraulic block power above 5%:
Reactor water level will lower rapidly. A scram will occur but control rods will not fully insert due to hydraulic block. High Pressure Core Spray and RCIC will automatically initiate on low reactor water level, Level 2.
When the CRS directs Attachment 12 to be installed, install attachment 12 as Done. Before reactor level drops below -160, take the attachment paperwork to the CRS and report that jumpers for attachment 12 are installed.
Install other EP Attachments as requested, and take a copy of the completed Attachments to the CRS.
RCIC will fail to automatically initiate on reactor water level low, but it can be manually initiated.
SLC squib injection valves 1C41F004A/B will lose continuity as though fired but will not open when first initiated (auto events 6/7). This is to maintain power sufficiently high to cause water level to fall to the point emergency Depressurization is required.
The squib valves will open during Emergency Depressurization if SLC has been initiated and at least 7 ADS/SRVs have been opened (auto events 16 and 17).
Allowing SLC to inject at this point ensures water level can be restored with only RHR B injection via its return to Feedwater.
Termination:
Once emergency depressurization has been conducted and reactor water level is stabilized above TAF, using RHR B E12-F053B, or as directed by Lead Evaluator:
Take the simulator to Freeze and turn horns off.
Stop and save the SBT report and any other recording devices.
Instruct the crew to not erase any markings or talk about the scenario until after follow-up questions are asked.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 9 of 20 Critical Tasks:
When reactor water level cannot be restored and maintained above -191 cfz using Table 4 systems, the crew performs Emergency Depressurization by opening 8 ADS/SRVs.
Before Emergency Depressurization, the crew terminates and prevents all injection except boron, CRD, and RCIC per EP-2A.
When reactor pressure decreases to MSCP, 219 psig, the crew commences and slowly raises injection using RHR B via E12F053B, SDC Return to Feedwater, to restore and maintain RPV level to greater than -191". Criterion is to give the highest priority to restore RPV level greater than -191".
Emergency Classification:
Site Area Emergency on SS3.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 10 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 1 Event
Description:
Return Condensate Pump B and Condensate Booster Pump B to service.
TIME Position Applicants Actions or Behavior Conducts pre-job brief. (This may have been done before the students enter the simulator.)
CRS Directs the ATC to place Condensate Pump B in operation per 04-1-01-N19-1 section 4.3, beginning at step 4.3.2a(4).
Places Condensate Pump B in operation:
Starts Condensate Pump B by pressing CNDS PMP B START pushbutton on 1H13P680-1C.
Verifies Condensate Pump B Discharge Valve 1N19F024B opens as indicated at its handswitch on 1H13P680-1C.
ATC Checks discharge pressure is approximately 250 psig on CNDS PMP DISCH HDR PRESS indicator 1N19R607 on 1H13P680-1B.
Checks total condensate flow 7.5 mlbm/hr on CNDS PMPS MIN FLO indicator 1N19R621 on 1H13P680-1B.
Directs building operator to locally open the Condensate Pump B discharge pressure indicator 1N19R002B instrument isolation valve.
BOP Provides peer check to ATC.
Directs the ATC to place Condensate Booster Pump B in operation per 04 CRS 01-N19-1 section 4.3.2b.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 11 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 1 (Cont.)
Event
Description:
Return Condensate Pump B and Condensate Booster Pump B to service.
TIME Position Applicants Actions or Behavior Places Condensate Booster Pump B in operation:
Directs building operator to locally close Condensate Booster Pump B discharge pressure indicator 1N19R005B isolation valve 1N19FX212.
Starts Condensate Booster Pump B by pressing CNDS BSTR PUMP B START pushbutton on 1H13P680-1C.
Verifies Condensate Booster Pump B Discharge Valve 1N19F046B opens as indicated at its handswitch on 1H13P680-1C.
ATC Directs building operator to locally open Condensate Booster Pump B discharge pressure indicator 1N19R005B isolation valve 1N19FX212.
Checks discharge pressure is approximately 450 psig on CNDS BOOSTER PUMP DISCH HDR PRESS indicator 1N19R610 on 1H13P680-1B.
Checks total condensate booster flow 4.2 mlbm/hr on CNDS BST PMP MIN FLO indicator 1N19R659 on 1H13P680-1B.
BOP Provides peer check to ATC Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 12 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 2 Event
Description:
Control Rod Drive Pump A trip resulting in CRD HCU 32-17 low pressure fault that will not clear TIME Position Applicants Actions or Behavior Recognizes and reports that CRD Pump A has tripped off as indicated by:
BOP Alarm 1H13P601-22A-C-3, CRD PMP A/B AUTO TRIP Alarm 1H13P601-22A-A-3, CRD CHRG WTR PRESS LO.
Enters 05-1-02-IV-7, Control Rod/Drive Malfunctions.
CRS Ensures immediate actions for CRD pump trip are performed.
Take immediate operator actions per CRD malfunctions ONEP (from memory).
Place CRD SYS FLO CONT (C11-R600 on P601-22B) in MANUAL and REDUCE output to zero.
Start CRD pump B.
BOP Slowly adjust CRD SYS FLO CONT to 54-66 gpm after charging pressure returns to normal. (~1700 psig)
Return CRD SYS FLO CONT to AUTO with tapeset at 54-66 gpm.
(THERE ARE NO REQUIRED SUBSEQUENT ACTIONS FOR THIS EVENT)
Recognizes and reports CRD HCU fault for rod 32-17:
Alarm 1H13-P680-4A2-D4, HCU TROUBLE ATC Presses HCU FAULT pushbutton on 1H13P680-6C and identifies rod 32-17(HD) by red LED on RC&IS Full Core Display.
Presses ACKN HCU FAULT pushbutton on 1H13P680-6C to reset the HCU TROUBLE alarm.
CREW Ensures that an operator and electrician are sent to investigate the cause of the pump trip, and sends operator to investigate HCU 32-17 trouble.
Notifies Work Management of CRD A pump trip and requests assistance formulating troubleshooting plan.
Directs building operator to charge HCU Accumulator 32-17 (HD) per 04 01-C11-1 section 4.4.
CRS Enters LCO 3.1.5 Condition A for control rod 32-17 (HD) HCU accumulator. The CRS may declare rod 32-17 slow per TS 3.1.5 Action A.1 or he may declare rod 32-17 inoperable per TS 3.1.5 Action A.2, either is acceptable.
Conducts crew transient briefing Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 13 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 3 Event
Description:
RHR Pump A Trip TIME Position Applicants Actions or Behavior Recognizes and reports RHR Pump A trip as indicated by.
Alarm 1H13P601-20A-A4, RHR PMP A OVERLD Alarm 1H13P601-20A-H6, RHR A OOSVC Status lights RHR PMP OVERLD/PWRLOSS and RHR PMP A AUTO BOP TRIP on 1H13P601-20B.
Recognizes and reports RHR A piping is not filled due to Test Return to Suppression Pool valve 1E12F024A open without RHR Pump A running:
Alarm 1H13P601-20A-C4, RHR PMP A DISCH PRESS ABNORMAL Directs BOP to secure RHR A lineup from Suppression Pool Cooling in CRS accordance with 04-1-01-E12-1, Residual Heat Removal System, step 5.2.2c.
Realigns RHR A MOVs per 04-1-01-E12-1, step 5.2.2c, using handswitches on 1H13P601-20C:
Closes 1E12F024A, RHR A TEST RTN TO SUPP POOL Opens 1E12F048A, RHR HX A BYP VLV BOP Checks open 1E12F003A, RHR HX A OUTL VLV Opens 1E12F064A, RHR A MIN FLO TO SUPP POOL (Only closing 1E12F024A is required to stop RHR A from draining. Other valve manipulations to place RHR A in its standby alignment, which would include filling and venting RHR A, is not expected to be performed in the timeframe of this scenario.)
Sends operators to RHR Pump A and to breaker 152-1509 to investigate RHR CREW Pump A trip.
Notifies Work Management of RHR A pump trip and requests assistance formulating troubleshooting plan.
CRS Determines that in addition to existing LCO TS 3.5.1, TS 3.6.1.7 Condition A and TS 3.6.2.3 Condition A apply.
Conducts crew briefing.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 14 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 4 Event
Description:
Feedwater Master Level Controller output failure high TIME Position Applicants Actions or Behavior Recognizes and reports reactor water level rising as indicated by:
Level rising on 1H13P680-2B RX WTR LVL NARROW RANGE meters 1C34R605A,B,C and/or Master Level Controller and/or Startup Level Controller digital indication (if selected to VARIABLE)
Diagnoses and reports Feedwater Master Level Controller has failed as indicated by:
Master Level Controller output rising while above the normal setpoint, ATC 36 inches.
RFPT A and B following Master Level Controller output.
Reactor water level following Master Level Controller output.
Places Master Level Controller in manual by pressing MAN on the controller.
May press (down) manual output pushbutton on the controller as necessary to stabilize and control level below the scram setpoint, 53.5 inches.
Enters 05-1-02-V-7, Feedwater System Malfunctions:
CRS Directs the ATC to take manual control of the Master Level Controller.
Directs the ATC to control level 32 to 42 inches narrow range.
ATC Using (down) and (up) manual output pushbuttons on the controller as necessary to stabilize and control level 32 to 42 inches.
BOP Provides peer check to the ATC.
Notifies Work Management/I&C of Master Level Controller problem and requests assistance formulating troubleshooting plan.
CRS Conducts crew transient briefing, including contingencies for operation with the Master Level Controller in manual.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 15 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 5 Event
Description:
Feedwater Line B rupture in Turbine Building, ATWS - hydraulic block, power above 5%
TIME Position Applicants Actions or Behavior Recognize and report reactor water level falling rapidly.
Places the Reactor Mode Switch to shutdown per EN-OP-115, Conduct of Operations, section 5.2[1] and 02-S-01-27, Operations Philosophy, section 6.1.1.
Recognize symptoms of a FW Line rupture in the Turbine Building as indicated by:
Feedwater flows rapidly dropping to zero on FW FLO A/B meters on ATC 1H13P680-2B (sensed from flow elements in the Aux Building Steam Tunnel). (Because feedwater lines are cross-tied upstream of the flow elements and flows in both feedwater lines fall so quickly, the operator is not expected to diagnose which line is broken.)
Alarms 1H13-P680-8A1-C1(D1), TURB BLDG E(W) FLOOR DR SMP LVL HI-HI.
Report the Feed water leak to the CRS and BOP.
BOP/
Trip all condensate pumps.
Reactor Mode Switch in SHUTDOWN.
All Rods are NOT Inserted (Hydraulic Block ATWS).
Reactor power is initially above 5%.
ATC Reactor water level and trend.
Reactor pressure and trend.
Feedwater is unavailable due to feedwater line break in the Turbine Building.
MSIVs/Bypass valves are available.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 16 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 5 (Cont.) / 6 Event
Description:
Feedwater Line B rupture in Turbine Building, ATWS - hydraulic block, power above 5%
TIME Position Applicants Actions or Behavior Enters the SCRAM ONEP and the Turbine and Generator Trips ONEP.
CRS Enters EP-2, when ATWS is discovered Enter EP-2A.
Directs actions of EP-2A steps 1 - 4:
Directs ATC to Verify Recirc Pumps transferred to LFMG.
CRS Directs ATC to Verify ARI/RPT initiation.
Directs BOP to Override HPCS injection.
Verifies Recirc Pumps transferred to LFMG (not in fast speed).
ATC Verifies/Initiates ARI/RPT.
Inhibits ADS.
Place ADS A and ADS B keylock switches to INHIBIT on 1H13P601-19B.
BOP Override HPCS injection on 1H13P601-16C:
Place the HPCS pump handswitch to the STOP position.
Place the E22-F004, HPCS injection valve, handswitch to the CLOSE position.
Direct the ATC or BOP to verify initiations and isolations for Reactor Level 2 CRS are completed.
Verify Division 3 Diesel Generator is running (1H13P601-16C) with cooling water (1H13P870-5C).
BOP Isolations for Reactor Level 2 are completed by observation of Isolation Status Board and/or 1H13P870.
Verifies RCIC initiation on 1H13P601-21B/C.
Event 6 Recognizes RCIC failure to start on Level 2.
ATC Manually initiates RCIC by arming and depressing RCIC MAN INIT pushbutton on 1H13P601-21B per hard card 04-1-01-E51-1 Att. VI.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 17 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 5 (Cont.)
Event
Description:
Feedwater Line B rupture in Turbine Building, ATWS - hydraulic block, power above 5% / Failure of SLC to inject Enters EP-2A step L-6 or L-7 and direct the ATC to establish level band -70 to -
130.
Directs maximizing CRD flow.
Enters EP-2A step P-4 and direct the BOP to establish a pressure band 800 -
1060 psig using IPC and BCV Manual Jack.
CRS Enters EP-2A step Q-4, directs SLC injection with both SLC pumps.
Calls for EP Attachments 8, 12, 18, 19, 20. (The CRS should state Att. 12 has the highest priority.)
Directs restoration of the Aux Building (restore Instrument Air and Drywell Chilled Water isolations)
Initiates SLC B as directed using hard card by placing keylock switch to START on 1H13P601-18B.
Initiates SLC A as directed using hard card by placing keylock switch to START BOP on 1H13P601-19B.
Recognizes excessive SLC discharge pressure on meter 1C41R600, and reports SLC failure to inject.
CRS Calls for Attachment 28.
BOP Maximizes CRD for flow per 04-1-01-C11-1 Att. VIII hard card:
Places CRD SYS FLO CONT C11-R600 in MANUAL on 1H13P601-22B.
Using CRD SYS FLOW CONT C11-R600, fully opens C11-F002A(B),
Fully opens C11-F003, CRD DRIVE WTR PRESS CONT VLV on 1H13P601-22C.
Restores isolations as directed:
Opens INST AIR SPLY HDR TO CTMT 1P53F001 on 1H13P870-3C.
Opens DWCW SPLY HDR TO CTMT 1P72F121 on 1H13P870-3C Opens DWCW RTN HDR FM CTMT 1P72F122 on 1H13P870-3C BOP Opens DWCW RTN HDR FM DRWL 1P72F125 on 1H13P870-3C Opens DWCW SPLY HDR TO CTMT 1P72F124 on 1H13P870-9C.
Opens DWCW RTN HDR FM CTMT 1P72F123 on 1H13P870-9C Opens DWCW RTN HDR FM DRWL 1P72F126 on 1H13P870-9C Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 18 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 5 (Cont.)
Event
Description:
Feedwater Line B rupture in Turbine Building, ATWS - hydraulic block, power above 5%
TIME Position Applicants Actions or Behavior ATC Keep CRS updated with reactor level as it trends down.
When reactor level falls below -160, as soon as EP Attachment 12 is installed or reactor level reaches -191, exits Level and Pressure Legs of EP-2A and Enter Emergency Depressurization.
When EP-2A requires Emergency Depressurization, Crew terminates CRS and prevents all injection except boron, CRD, and RCIC per 02-S-01-27 Operations Philosophy. Feedwater and ECCS system alignments prevent injection into the RPV as evidenced by available instrumentation. Criterion is to give the highest priority to prevent all injection except boron, CRD, and RCIC until reaching MSCP.
Verify SP level is above 10.5 ft.
Direct the BOP or ATC operator to verify/perform Terminate and Prevent CRS injection into the RPV by overriding low pressure systems (LPCS/LPCI)
Direct BOP to Open 8 ADS valves.
When directed, Terminate and Prevent injection into the RPV.
Verify HPCS is initiated with annunciators P601-16A-B5, HPCS MTR CONT MAN OVERRD, and P601-16A-D5, HPCS INJ VLV F004 MAN OVERRD in.
Verify/perform low pressure ECCS systems overridden by ensuring division 1 ECCS initiation signal is present and placing the LPCS and RHR A pump hand switches to off and placing the E21F005 and E12F042A handswitches to CLOSE. (RHR A pump is tripped.)
ATC / o This is verified by annunciators P601-21A-B7, LPCS INJ VLV BOP F005 MAN OVERRD, P601-21A-C8, LPCS PMP MAN OVERRD, P601-20A-B2, RHR INJ VLV F042A MAN OVERRD, and P601-20A-C5, RHR PMP A MAN OVERRD in.
Verify/perform low pressure ECCS systems overridden by ensuring division 2 ECCS initiation signal is present and placing the LPCI B and C pump hand switches to off and placing the E12-F042B and C handswitches to CLOSE.
o This is verified by annunciators P601-17A-B1, RHR INJ VLV F042B MAN OVERRD, P601-17A-C2, RHR PMP B MAN OVERRD, P601-17A-B4, RHR INJ VLV F042C MAN OVERRD, and P601-17A-C5, RHR PMP C MAN OVERRD in.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 2 Page 19 of 20 Op-Test No: LOT-2013 Scenario No: 2 Event No: 5 (Cont.)
Event
Description:
Feedwater Line B rupture in Turbine Building, ATWS - hydraulic block, power above 5%
When directed, opens at 8 ADS valves on 1H13P601-19C for Emergency BOP Depressurization.
Enter EP-2A step L-10 (following the Emergency Depressurization).
CRS Establish reactor pressure (MSCP) as critical parameter for the ATC.
ATC Keep CRS updated with reactor pressure as it trends down.
When reactor pressure is below MSCP (206 psig) direct ATC to feed the reactor using RHR B via the E12-F053B with 4000 gpm and then at intervals of 1000 gpm until reactor level begins to trend up.
- Reactor pressure decreases to MSCP. Crew commences and slowly CRS raises injection using RHR B via E12F053B, SDC Return to Feedwater, to restore and maintain RPV level to greater than -191". Criterion is to give the highest priority to restore RPV level greater than -191".
Enter EP-2A step L-6 or L-7 and establish level band -70 to -130.
CRS When Attachments 18, 19, and 20 are reported installed, directs ATC to reset scram and BOP to maximize CRD drive water pressure.
ATC Resets RPS A by placing RPS Div 1 & 3 or 2 & 4 Reset switches to RESET on P680.
BOP Maximizes CRD drive water pressure by fully closing C11-F003, CRD DRIVE WTR PRESS CONT VLV on 1H13P601-22C.
BOP/ Inserts Control Rods by scramming rods using RPS scram arm/depress pushbuttons and/or by selecting control rods on RC&IS and depressing IN ATC TIMER SKIP or INSERT pushbutton on P680.
Following the ATWS, crew directs installation of EP Attachments 18, 19, and 20 and inserts control rods by manual scram and/or normal rod insertion per EP-2A step Q-1.
Enter EP-3 on Suppression Pool Temperature and Drywell pressure; however, CRS no substantial operator actions are expected for this entry during the scenario.
Revision 2
Give this page to the CRS Turnover the following conditions:
Power 75%
BOC EOOS GREEN Core flow is 70 mlbm/hr with operation in the OPRM Trip Enabled region of the Power-Flow Map.
Condensate Pump B and Condensate Booster Pump B are in standby following maintenance on Condensate Pump B.
RHR A is operating in Suppression Pool Cooling.
LCO 3.5.1 was entered one hour ago for LPCI A due to RHR A Test Return to Suppression Pool valve 1E12F024A open for Suppression Pool Cooling.
SSW A is aligned per 04-1-03-P41-1, SSW A Chemical Addition Run, and is supporting RHR A Suppression Pool Cooling.
This is a Division 1 work day.
Immediately following the brief:
The ATC will return Condensate Pump B and Condensate Booster Pump B to service in accordance with 04-1-01-N19-1 section 4.3 and 03-1-01-2 step 6.6.
All prerequisites of step 4.3.1 have been completed.
Steps 4.3.2a(1), (2), and (3) have been completed.
After Condensate Pump B and Condensate Booster Pump B have been returned to service, and before Suppression Pool temperature decreases below 75°F, shut down RHR A Suppression Pool Cooling.
Reactor Engineering is revising the Reactivity Management Plan for power ascension to 100%, which is scheduled to begin on this shift.
Note that an independent Reactivity Management SRO per Operations Philosophy 6.8.1 will not be provided for this scenario.
Revision 1 Page 20 of 20
Appendix D Scenario Outline Form ES-D-1 Scenario 3 Page 1 of 2 Facility: Grand Gulf Nuclear Station Scenario No.: 3 Op-Test No.: LOT-2013 Examiners: ____________________________ Operators: _____________________________
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
Place Reactor Water Clean-Up filters in HOLD and secure RWCU pump A.
RWCU Heat Exchanger leak inside Containment with failure of Division 1 RWCU valves to automatically isolate.
Drywell Chilled Water Pump B trip with failure of standby pump to automatically start.
Jet Pumps 11/12 failure.
Service Transformers 11 and 21 lockout.
Reactor Recirc piping rupture with HPCS failure.
RHR B/C logic power failure.
Initial Conditions: Plant is operating at 100% power.
Inoperable Equipment: Division 1 Diesel Generator is tagged out of service.
Turnover:
Plant is operating at 100% power.
Division 1 Diesel Generator is tagged out of service to replace an optical isolator in 1H22-P400.
The plant is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into 14 day LCO 3.8.1b.
The next AC/DC lineup is due in 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Maintenance on DG11 should be complete within the next hour.
Immediately following turnover:
Place RWCU filters in hold per 04-1-01-G33-1 section 4.6.
Then, secure RWCU pump A per 04-1-01-G33-1 section 5.9 for planned maintenance on its discharge check valve, 1G33F012A.
This is a Division 1 work week.
Scenario Notes:
This is a new scenario.
Validation Time (60-90 min): 65 min Revision 2
Appendix D Scenario Outline Form ES-D-1 Scenario 3 Page 2 of 2 Event Malf. No. Event Type Event No. Description Place Reactor Water Clean-Up filters in HOLD and secure 1 N (BOP) RWCU pump A. (04-1-01-G33-1, Reactor Water Clean-Up System, sections 4.6 and 5.9).
g33033 C (ATC) RWCU Heat Exchanger leak inside Containment with failure of 2 (rf) e31210 A (CREW) Division 1 RWCU valves to automatically isolate (05-1-02-III-5, Automatic Isolations; ARI 04-1-02-1H13-P680-11A-A3, A4,E3; TS (CRS) 04-1-01-G33-1 Att. VII; TS 3.3.6.1) 3 di_1p72m602b C (BOP) Drywell Chilled Water Pump B trip (ARIs 04-1-02-1H13-P870-4A-G1/G2, 9A-E2, 10A-G3/H3)
C (ATC) Failure of Jet Pumps 11/12 rams head. (05-1-02-III-6, Jet 4 rr011f A (CREW) Pump Anomalies; 05-1-02-III-3, Reduction in Recirculation TS (CRS) System Flow Rate; TS 3.4.3).
Service Transformers 11 and 21 lockout / Recirc loop B rupture
/ HPCS L8 relay failure (EP-2, RPV Control; EP-3, Containment Control; 05-1-02-I-1, Reactor Scram; 05-1-02-I-2, Turbine Generator Trips; 05-1-02-I-4, Loss of AC Power; ARI 04-1 1H13-P601-16A-A3).
When reactor water level decreases to -160 (TAF), the crew r21133a opens 8 SRVs per EP-2 step ED-4. (At least seven SRVs must 5 r21133b be open before RPV level drops to -217.)
M (CREW) rr063b e22159a The crew manually restores power to Division 1 ECCS bus 15AA, AND/OR manually aligns RHR B for injection and restores reactor water level to above -160 following emergency depressurization per EP-2 step L-14. (At least one ECCS must be aligned for injection prior to reactor pressure decreasing below 125 psig during Emergency Depressurization.)
6 (rf) r21221 C (BOP) RHR B/C logic power failure (ARI 04-1-02-1H13-P601-H2).
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec
- Critical Task (As defined in NUREG 1021 Appendix D)
Quantitative Attributes Table Normal Events 1 Abnormal Events 2 Reactivity Manipulations 0 Total Malfunctions 8 Instrument/Component Failures 4 EP Entries (Requiring substantive action) 1 Major Transients 1 EP Contingencies 1 Tech Spec Calls 2 Critical Tasks 2 Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 1 of 21 Simulator Setup:
A. Initialization
- 1. Log off all simulator PDS and SPDS computers (PDS and SPDS must come up after the simulator load for proper operation).
- 2. Startup the simulator using Simulator Instructors Job Aid section 6.3.
Note:
Prior to running the Schedule File, ensure no Event Files are Open. If an existing Event File is Open prior to running the Schedule File, then any associated Event Files will not automatically load.
- 3. Open Schedule.exe and Director.exe by clicking on the Icon in the Thunder Bar.
- 4. Set the Simulator to IC-195 and perform switch check (Using Quick Reset in Director).
- 5. Click on Open in the Schedule window and Open Schedule File 2013 NRC Scenario 3.sch (in the Schedule Directory)
- 6. In Schedule window, click on the Stopped red block. The red block will change to a green arrow and indicate the scenario is active (Running).
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 2 of 21
- 7. Click the Summary tab in the Director window. Verify the schedule files are loaded and opened per Section B below. (Note: Any actions in the schedule file without a specific time will not load into the director until triggered.)
- 8. Take the simulator out of freeze.
- 10. Verify or perform the following:
IC-195 APRMs are turned on (4,1,2,3)
Ensure the BOC rod movement sequence available at the P680.
Advance all chart recorders and ensure all pens inking properly.
Clear any graphs and trends off of SPDS.
Ensure 1G33F034 is closed on P680-11C.
Ensure Div 1 DG is in MAINTENANCE with 152-1508 racked out:
Hang tags on DG11 start pushbutton and 152-1508 on 1H13P864-1C.
- 11. Run through any alarms and ensure alarms are on. (Note: On T-Rex, to verify alarms are ON, the indicator will indicate Alarms On).
- 12. Place the simulator in Freeze.
B. File loaded verification:
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 3 of 21 Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 4 of 21 Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 5 of 21 Crew Turnover:
B. Assign the candidates crew positions.
C. Turnover the following conditions:
Power 100%
BOC EOOS GREEN Division 1 Diesel Generator is tagged out of service to replace an optical isolator in 1H22-P400.
The plant is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into 14 day LCO 3.8.1b.
The next AC/DC lineup is due in 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Maintenance on DG11 should be complete and the tagout released within the next hour.
Immediately following turnover:
Place RWCU filters in hold per 04-1-01-G33-1 section 4.6.
Then, secure RWCU pump A per 04-1-01-G33-1 section 5.9 for planned maintenance on its discharge check valve, 1G33F012A.
RWCU filters are to remain in HOLD until RWCU pump A is returned to operation.
This is a Division 1 work week.
D. Allow the crew to perform pre-shift brief and review procedures for planned evolutions.
E. Bring the crew into the Simulator, place the simulator is in RUN.
F. Allow the crew to walk down panels.
G. When the crew assumes the shift begin Scenario Activities.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 6 of 21 SCENARIO ACTIVITIES:
Start SBT report and any other required recording devices (Video recording not allowed for NRC exams).
Place Reactor Water Clean-Up filters in HOLD and secure RWCU pump A.
The crew will remove RWCU filters A and B from service in accordance with 04 01-G33-1, Reactor Water Clean-Up System, section 4.6.
When requested as the building operator at the RWCU F/D control panel to remove RWCU filter A from service, trigger Event 1 to simulate lowering flow through filter A.
When requested as the building operator at the RWCU F/D control panel to remove RWCU filter B from service, trigger Event 11 to simulate lowering flow through filter B.
When both filters have been removed from service, wait 2 minutes and as the building operator who had been at the RWCU panel, report you have exited containment.
The crew will then secure RWCU pump A in accordance with 04-1-01-G33-1, Reactor Water Clean-Up System, section 5.9.
When the crew has completed all required steps of 04-1-01-G33-1, Reactor Water Clean-Up System, section 5.9 and at the direction of the lead evaluator, trigger Event 2 to cause RWCU Heat Exchanger leak inside Containment with failure of Division 1 RWCU valves to automatically isolate.
RWCU Heat Exchanger leak inside Containment with failure of Division 1 RWCU valves to automatically isolate.
The crew will respond per ARIs 04-1-02-1H13-P680-11A-A3, -A4, and -E3 and 05 02-III-5, Automatic Isolation.
If asked the status of Division 1 NUS RWCU leak detection temperature switches, as the booth operator report RWCU Heat Exchanger Area NUS switches are indicating tripped, and that readings are slowly lowering. (If requested, an exact temperature reading for RWCU heat exchanger room Div 1 NUS switch 1E31N620A can be obtained by viewing the tracking value for malfunction tte31n034a_d.)
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 7 of 21 If asked for indications from upper control room, after the Div 2 RWCU isolation is received, report RWCU Heat Exchanger area temperature on recorder 1E31R608 and NUS switch 1E31N620A is 125°F, with no trend yet, and the NUS switch is tripped. Report RWCU Hx room T is 80°F, if asked.
If requested to investigate the Division failure to isolate, as I&C report required troubleshooting will be intrusive, so a work order and possibly a tagout will be required.
If directed to perform surveys of containment, as Radiation Protection acknowledge the request.
If containment evacuation is ordered, as Security wait 15 minutes and report accountability is complete and no personnel are inside containment.
When the crew has completed all required steps of 05-1-02-III-5 and the CRS has addressed TS 3.3.6.1 for the Division 1 RWCU isolation failure, and at the direction of the lead evaluator, trigger Event 3 to cause Drywell Chilled Water Pump B trip.
Drywell Chilled Water Pump B trip with failure of the standby pump to automatically start:
The crew will respond using ARIs 1H13-P870-4A-G1/G2, 9A-E2, 10A-G3/H3 and manually start Drywell Chilled Water Pump A. Drywell chillers with automatically restart when pump A is started.
If requested to investigate, as the building operator wait 2 minutes, then report local breaker 52-1BP60303 is tripped on overcurrent, but you can see no apparent cause locally at DW Chilled Water Pump B. If asked, report DW Chilled Water Pump A is running normally (if the RO has started it).
When the crew has restored Drywell Chilled Water system to operation, and at the direction of the lead evaluator, trigger Event 4 to cause failure Jet Pumps 11/12.
Failure of Jet Pumps 11 and 12:
The crew will observe a reduction in generator output.
The crew will respond per 05-1-02-III-6, Jet Pump Anomalies, and 05-1-02-III-3, Reduction in Recirculation System Flow Rate.
If requested to perform a Jet Pump surveillance 06-RE-1B33-D-0001, as Reactor Engineering wait 5 minutes then report Jet Pumps 11 and 12 failed 2 of 3 criteria per the surveillance.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 8 of 21 When the crew has plotted the point of operation on the Power/Flow Map and the CRS has addressed TS 3.4.3, and at the direction of the lead evaluator, trigger Event 5 to cause Service Transformers 11 and 21 lockout.
Service Transformers 11 and 21 lockout followed by Recirc Loop B rupture and HPCS Level 8 relay failure:
The crew will respond using 05-1-02-I-1, Reactor Scram; 05-1-02-I-2, Turbine and Generator Trips; 05-1-02-I-4, Loss of AC Power, EP-2, RPV Control, EP-3, Containment Control.
A false HPCS Level 8 signal will seal in (Auto Event 25), causing injection valve 1E22F004 to close.
If sent to manually open 1E22F004, as the building operator wait 10 minutes, then report the handwheel for 1E22F004 cannot be rotated.
If requested to install EP Attachment 5, acknowledge the request but DO NOT INSTALL Att. 5.
Install other EP Attachments as directed.
RHR B/C Logic Power failure:
When Drywell pressure reaches 1.35 psig, RHR B/C logic power will be lost (Auto Event 6).
If requested to investigate RHR B/C Logic Power Failure, as the building operator or Electrical Maintenance wait 5 minutes, then report breaker 72-11B14 is tripped. If requested to reset the breaker, report the breaker will not reset.
If sent to manually open 1E12F042C, as the building operator wait 10 minutes, then report the handwheel for 1E12F042C cannot be rotated.
If sent to manually reset breaker 72-11B14, as the building operator wait 5 minutes, then report the breaker appears damaged.
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Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 9 of 21 Termination:
Once emergency depressurization has been conducted and reactor water level is stabilized above TAF, or as directed by Lead Evaluator:
Take the simulator to Freeze and turn horns off.
Stop and save the SBT report and any other recording devices.
Instruct the crew to not erase any markings or talk about the scenario until after follow-up questions are asked.
Critical Tasks:
When reactor water level decreases to -160 (TAF), the crew opens 8 SRVs per EP-2 step ED-4. (At least seven SRVs must be open before RPV level drops to
-217.)
The crew manually restores power to Division 1 ECCS bus 15AA, AND/OR manually aligns RHR B for injection and restores reactor water level to above -
160 following emergency depressurization per EP-2 step L-14. (At least one ECCS must be aligned for injection prior to reactor pressure decreasing below 125 psig during Emergency Depressurization.)
Emergency Classification:
ALERT on FA1-RC1, RC3 Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 10 of 21 Op-Test No: LOT-2013 Scenario No: 3 Event No: 1 Event
Description:
Place Reactor Water Clean-Up filters in HOLD and secure RWCU pump A.
TIME Position Applicants Actions or Behavior Conducts pre-job brief. (This may have been done before the students enter the simulator.)
CRS Directs the BOP to coordinate with the building operator to remove RWCU filters A and B from service, one at a time, and place them in HOLD per 04 01-G33-1 section 4.6, and then remove RWCU pump A from service per 04 01-G33-1 section 5.9 Notifies Plant Chemistry both RWCU filters and RWCU pump A are being CRS removed from service.
Coordinates with the building operator to remove RWCU filters A and B from service:
Placing RWCU filter A in HOLD:
Slowly opens 1G33-F044, RWCU FLTR DMIN BYP VLV on 1H13-P680-11C while building operator reduces F/D flow with flow controller 1G36-FC-R022A on1G36-P002.
Maintains a nearly constant system flow rate, (450-500 gpm is recommended), as indicated on 1G33-FI-R609, RWCU INL FLO, on1H13-P680-11B.
Placing RWCU filter B in HOLD:
Slowly opens 1G33-F044, RWCU FLTR DMIN BYP VLV on 1H13-P680-11C while building operator reduces F/D flow with flow controller BOP 1G36-FC-R022B on1G36-P002.
Maintains a nearly constant system flow rate, (450-500 gpm is recommended), as indicated on 1G33-FI-R609, RWCU INL FLO, on1H13-P680-11B.
Removing RWCU pump A from service:
Lowers system flow rate to < 280 gpm by throttling 1G33F044 as indicated on 1G33FI-R609, RWCU INL FLO, on 1H13-P680-11B.
Trips RWCU pump A on 1H13-P680-11C.
Establishes 90 to 300 gpm flow as indicated on 1G33-FI-R609, RWCU INL FLO, on 1H13-P680-11B by throttling the Bypass Valve 1G33F044.
ATC Provides peer check to BOP.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 11 of 21 Op-Test No: LOT-2013 Scenario No: 3 Event No: 2 Event
Description:
RWCU Heat Exchanger leak inside Containment with failure of Division 1 RWCU valves to automatically isolate.
TIME Position Applicants Actions or Behavior Recognizes and reports indications of a leak from RWCU system indicated by:
Alarm 1H13-P680-11A-E3, RWCU AREA AMBIENT TEMP HI ATC Alarm 1H13-P680-11A-E4, RWCU AREA dT HI (The RWCU isolation setpoint will be reached approximately 80 seconds after the first high area temperature alarm is received.)
Calls up RWCU area temperature data on PDS and/or checks leak detection temperature NUS switch 1E31N620B indication for RWCU areas on backpanel BOP 1H13P642. Checks Div 1 NUS switch 1E31N620A and/or recorder 1E31R608 (upper control room data provided by booth operator).
Recognizes and reports isolation setpoint has been exceeded indicated by:
ATC Alarm 1H13-P680-11A-A3, RWCU HX RM TEMP HI Division 2 RWCU Group 8 valves close Enters 05-1-02-III-5, Automatic Isolations.
CRS Directs ATC / BOP to verify (group 8) isolations.
Determine RWCU Division 2 isolation valves have isolated.
Recognize and report failure of Division 1 RWCU isolation valves to automatically isolate. May use hard card 04-1-01-G33-1 Att. VII, Rapid Isolation of RWCU or Automatic Isolation Checklist from 05-1-02-III-5.
Manually close Division 1 RWCU isolation valves:
ATC/ 1G33F004 on 1H13-P680-11C BOP 1G33F039 on 1H13-P680-11C 1G33F054 on 1H13-P680-11C 1G33F250 on 1H13-P870-3C 1G33F253 on 1H13-P870-3C Directs ATC/BOP to close Division 1 RWCU isolation valves. Observes CRS Isolation Status Board to verify valves have been closed.
May direct limited evacuation of containment due to RWCU leak in the heat exchanger room.
CRS May direct Radiation Protection to perform area surveys and air monitoring.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 12 of 21 Op-Test No: LOT-2013 Scenario No: 3 Event No: 2 (Cont.)
Event
Description:
RWCU Heat Exchanger leak inside Containment with failure of Division 1 RWCU valves to automatically isolate.
TIME Position Applicants Actions or Behavior Notifies Work Management of event. Requests assistance determining cause of RWCU leak and failure of Division 1 to isolate.
Enters TS 3.3.6.1 Condition A for Division 1 RWCU isolation instrumentation.
CRS (The CRS may elect to instead enter TS 3.6.1.3 or TS 3.6.5.3, as appropriate, for each affected Division 1 RWCU isolation valve pending discovery of what caused the failure to isolate.)
Provides transient brief to the crew.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 13 of 21 Op-Test No: LOT-2013 Scenario No: 3 Event No: 3 Event
Description:
Drywell Chilled Water Pump B trip with failure standby pump A to automatically start.
TIME Position Applicants Actions or Behavior Recognizes and reports Drywell Chilled Water Pump B trip resulting in loss of all system flow and Drywell Chillers trip indicated by:
Alarm 1H13-P870-9A-E2, DRWL CHILL WTR IN-CTMT TROUBLE Alarm 1H13-P870-4A-G1 (10A-G3), DRWL CHLD WTR PMP DISCH FLO LO BOP Alarm 1H13-P870-4A-G2 (10A-H3), DRWL CHLD WTR TEMP HI Recognizes failure of Drywell Chilled Water Pump 1P72C001A to automatically start, and manually starts it using handswitch 1P72C001A on 1H13-P870-4C.
Verifies and reports DW chilled water flow is re-established, Drywell Chillers automatically restart, and alarms reset.
Sends building operator to investigate trip of DW Chilled Water Pump B.
Verifies average Drywell Temperature is maintained within limits (below 135°F) on SPDS or PDS.
CRS Notifies Work Management of event. Requests assistance determining cause of Drywell Chilled Water Pump B trip.
Provides transient brief to the crew.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 14 of 21 Op-Test No: 2013 Scenario No: 3 Event No: 4 Event
Description:
Jet pumps 11/12 rams head failure TIME Position Applicants Actions or Behavior Recognizes and reports step change reduction in power indicated on 1H13P680:
MWe loss ATC APRM trends Feedwater flow reduction Reactor water level perturbation Diagnoses jet pump failure as indicated by on 1H13P680:
Reduced total core flow Recirc Loop A and B elevated drive flows ATC Calibrated Jet Pump B flow less than calibrated Jet Pump A, C, and D flows Loop A total jet pump flow lower than loop B total jet pump flow Reduced Core dp Checks individual Jet Pump dp on 1H13P619.
Recognizes and reports Jet Pumps 11 and 12 dps less than other Jet Pump BOP dps.
Enters 05-1-02-III-6, Jet Pump Anomalies:
Directs ATC to determine actual total core flow and plot point of operation on Power/Flow map CRS Directs Reactor Engineering to perform 06-RE-1B33-D-0001, Jet Pump Functional Test, and to verify thermal limits are met.
Enters 05-1-02-III-3 Reduction in Recirculation System Flow Rate:
Determines actual core flow per 05-1-02-III-6, Jet Pump Anomalies (by doubling the indicated flow for calibrated jet pump A and subtracting that from indicated core flow twice).
ATC Plots and reports point of operation on the Power/Flow Map Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 15 of 21 Op-Test No: 2013 Scenario No: 3 Event No: 4 (Cont.)
Event
Description:
Jet pumps 11/12 rams head failure TIME Position Applicants Actions or Behavior Enters TS 3.4.3 Condition A.
Makes notifications for entering TS 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shutdown statement.
Begins reviewing 03-1-01-2, Power Operations for commencing plant CRS shutdown.
Notifies Work Management of Jet Pump failure and requests assistance formulating forced outage plan Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 16 of 21 Op-Test No: LOT-2013 Scenario No: 3 Event No: 5 Event
Description:
Service Transformers 11 and 21 lockout followed by Recirc Loop B rupture TIME Position Applicants Actions or Behavior Recognizes and reports reactor scram:
Places Reactor Mode Switch to shutdown Verifies all control rods fully inserted (indicated on RPIS channel 2 after ATC Div 2 DG restores power to bus 16AB) and power approximately 0%
Recognizes and reports reactor water level reduced due to loss of Condensate/Feedwater Recognizes and reports Main Turbine tripped and MSIVs closed BOP Recognizes and reports loss of switchyard/offsite power.
Enters the following procedures (simultaneously):
EP-2, RPV Control 05-1-02-I-1, Reactor Scram CRS 05-1-02-I-2, Turbine and Generator Trips 05-1-02-I-4, Loss of AC Power Check Recognizes and reports Reactor Water Level Low, Level 2.
ATC Verifies HPCS and RCIC initiated and injecting on 1H13P601-16, 21 Checks Division 2 and 3 Diesel Generators have energized their respective buses (Div 1 DG is tagged out of service).
Verifies SSW B aligned to Div 2 DG on 1H13P870-7C.
BOP Verifes SSW C aligned to Div 3 DG on 1H13P870-5C Verifies isolations for Level 2 on 1H13P870-3C, 9C.
BOP Manually energizes bus 15AA via feeder 152-1511 from ESF Transformer 12 on 1H13P864-1C.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 17 of 21 Op-Test No: LOT-2013 Scenario No: 3 Event No: 5 (Cont.) / 6 Event
Description:
Service Transformers 11 and 21 lockout followed by Recirc Loop B rupture TIME Position Applicants Actions or Behavior Recognize and report indications of a LOCA in the Drywell as indicated by rising drywell pressure and temperature (P680-5A-C3, DRWL PRESS HI/LO; CREW P680-5A-C2, DRWL PRESS HI HI; Low Pressure ECCS initiations due to 1.39 psig in the drywell).
Directs verify initiations and isolations for Reactor Level 2 Directs ATC establish a level band of +30 to -30 in using HPCS and CRS RCIC Directs BOP establish a pressure band of 800 -1060 psig using SRVs Verifies Low Pressure ECCS initiations for Drywell Pressure High:
Recognizes Div 1 ECCS initiation occurred on 1H13P601(white light above LPCS/RHR A INIT RESET push button lit on 1H13P601-21B)
ATC/
Recognizes and reports RHR B/C Logic Power Failure as indicated by BOP RHR B OOSVC alarm and logic power failure status light on (Event 6) 1H13P601-17B.
Recognizes and reports RHR B available for injection from 1H13P601 via 1E12F053B, RHR B SHUTDN CLG RTN TO FW Recognizes and reports failure of HPCS Injection MOV 1E22F004 to reopen after it was closed the first time to maintain level within the band.
ATC/
Recognizes HPCS Level 8 signal sealed in preventing opening BOP 1E22F004.
Sends building operator to manually open 1E22F004 Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 18 of 21 Op-Test No: LOT-2013 Scenario No: 3 Event No: 5 / 6 (Cont.)
Event
Description:
Service Transformers 11 and 21 lockout followed by Recirc Loop B rupture RHR B/C Logic Power failure and HPCS Injection MOV 1E22F004 failure TIME Position Applicants Actions or Behavior Enter EP-3 and re-enter EP-2 Direct actions of EP-2:
Verify initiations and isolations for Drywell Pressure 1.39 psig.
Enter the Alternate Level Control procedure of EP-2.
Inhibit ADS CRS Maximize CRD for flow.
Inject SLC A and B for level control Addresses all legs of EP-3 Directs verifying Div 2 Drywell and Containment hydrogen analyzers are operating and starting Div 2 Drywell/Containment Hydrogen Igniters (Div 1 if power is manually restored to bus 15AA).
When directed, maximize CRD A and B for flow.
Re-energize 15B42 on P864-1C Re-energize 16B42 on P864-2C Start CRD A Aux Oil Pump on P601-22C Start CRD A Pump on P601-22C ATC Start CRD B Aux Oil Pump on P601-22C Start CRD B Pump on P601-22C (Although there are other actions per the procedure, only these listed will accomplish anything with scram sealed in and loss of power. CRD A if power is manually restored to bus 15AA.)
Update the CRS with reactor water level as it continues to lower.
When directed, Inhibit ADS by placing both ADS A and B inhibit switches to INHIBIT on 1H13P601-19B.
Dispatch operators to recover out of service water injection sources (may BOP request manually opening 1E22F004, 1E12F042C and/or resetting 72-11B14.)
Evaluate and deliver ECCS Status Report to the CRS (hard card).
Provides frequent reports of reactor water level and trend.
ATC Reports level at Top of Active Fuel when level trends offscale low on wide range recorders on 1H13P601-20B, 17B.
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 19 of 21 Op-Test No: LOT-2013 Scenario No: 3 Event No: 5 / 6 (Cont.)
Event
Description:
Service Transformers 11 and 21 lockout followed by Recirc Loop B rupture RHR B/C Logic Power failure and HPCS Injection MOV 1E22F004 failure TIME Position Applicants Actions or Behavior Verifies Div 1 and 2 Drywell/Containment Hydrogen Analyzers are operating BOP and starts Div 1 and 2 Drywell/Containment Hydrogen Igniters on 1H13P870-3C, 10C. (Div 1 if power is manually restored to bus 15AA)
Directs aligning EP-2 Table 2 systems for injection when determined level cannot be restored to nominal band, 30 to +30:
CRS Directs ATC to initiate Standby Liquid Control System A and B Calls Maintenance/OSC for Fire Water to be connected for injection per Attachment 26 via RHR C pathway.
When directed to initiate SBLC A(B), use 04-1-01-C41-1, Standby Liquid Control System, Attachment VI (Initiation of Standby Liquid Control hard card) on 1H13P601-19B (18B):
Insert keys and turn SBLC Pmp A(B) pump key switch to START.
Verify system initiation by observing the following:
o F004A(B) SQUIB valves fired:
White SQUIB valve ready light OFF Annunciator SLC SYS A(B) OOSVC (P601-19A-H1)
Amber status light SQUIB A(B) LOSS CONT or PWR LOSS is ON.
ATC/ o C41-F001A(B) tank outlet valve is open.
o SBLC Pump A(B) running.
o RWCU Isolated G33-F004 (F001, F251) Closed (already closed due to RWCU isolation earlier) o Verify SBLC A(B) is injecting into the RPV by observing the following:
SBLC pump discharge pressure exceeds reactor pressure.
SBLC tank level lowering.
(SLC A is available only if power is manually restored to bus 15AA)
Revision 2
Appendix D Required Operator Actions Form ES-D-2 Scenario 3 Page 20 of 21 Op-Test No: LOT-2013 Scenario No: 3 Event No: 5 / 6 (Cont.)
Event
Description:
Service Transformers 11 and 21 lockout followed by Recirc Loop B rupture RHR B/C Logic Power failure and HPCS Injection MOV 1E22F004 failure TIME Position Applicants Actions or Behavior When reactor water level reaches -160 and before -191, enter Emergency Depressurization leg of EP-2, RPV Control CRS Verify Suppression Pool level is above 10.5 Direct the BOP to open 8 ADS/SRVs When directed, Opens 8 ADS valves on 1H13P601. (At least seven BOP SRVs must be open before RPV level drops to -217.)
When reactor pressure lowers to <500 psig, directs injection with RHR B by starting RHR B pump and opening 1E12F053B, RHR B SHUTDN CLG RTN CRS TO FW from 1H13P601.
As directed, injects to restore level above -160 wide range with:
RHR A and LPCS (if power has been restored to bus 15AA),
and/or RHR B by starting RHR B pump and opening 1E12F053B, ATC/ RHR B SHUTDN CLG RTN TO FW on 1H13P601-17C.
(1E12F053Bmust be opening prior to reactor pressure going BOP below 125 psig during Emergency Depressurization.)
(1E12F053B handswitch must be held in the open or closed position since it is a throttle valve.)
(Either LPCS and RHR A, OR RHR B, OR all may be used to restore reactor water level above -160 to meet the overall critical task.)
When reactor water level begins rising, directs restoring and maintaining level -
CRS 30 to +30.
Monitors for LPCS and/or RHR A injection and ensures level is re-established above TAF.
ATC Controls LPCS and/or RHR A and/or RHR B injection by operating injection valves or cycling pumps to restore and control reactor water level -30 to +30 on wide range level instrumentation.
Revision 2
Give this page to the CRS Turnover the following conditions:
Power 100%
BOC EOOS GREEN Division 1 Diesel Generator is tagged out of service to replace an optical isolator in 1H22-P400.
The plant is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into 14 day LCO 3.8.1b.
The next AC/DC lineup is due in 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Maintenance on DG11 should be complete and the tagout released within the next hour.
Immediately following turnover:
Place RWCU filters in hold per 04-1-01-G33-1 section 4.6.
Then, secure RWCU pump A per 04-1-01-G33-1 section 5.9 for planned maintenance on its discharge check valve, 1G33F012A.
This is a Division 1 work week.
Revision 1 Page 21 of 21