ML18061A169
| ML18061A169 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 12/11/2017 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | Entergy Operations |
| References | |
| Download: ML18061A169 (58) | |
Text
ES-401 BWR Examination Outline Form ES-401-1 Facility: Grand Gulf Nuclear Station Date of Exam: December 2017 Tier Group RO K/A Category Points SRO-Only Points K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total
- 1. 1 3 3 4 3 4 3 20 Emergency & N/A N/A 2 1 2 1 1 1 1 7 Abnormal Plant Evolutions Tier Totals 4 5 5 4 5 4 27 1 2 3 3 3 3 2 1 2 2 3 2 26 2.
Plant 2 2 1 1 1 1 1 1 1 1 1 1 12 Systems Tier Totals 4 4 4 4 4 3 2 3 3 4 3 38
- 3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 Categories 10 3 3 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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply.) Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As
- These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
- These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
Rev 1
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 A2 G* K/A Topic(s) IR #
1 2 3 1 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS 3.3 1 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X OF FORCED CORE FLOW CIRCULATION :
AK1.02 Power/flow distribution Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF A.C. 3.3 2 295003 Partial or Complete Loss of AC / 6 X POWER :
AK3.01 Manual and auto bus transfer Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following: 3.3 3 295004 Partial or Total Loss of DC Pwr / 6 X
AK2.03 D.C. bus loads Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP : 3.6 4 295005 Main Turbine Generator Trip / 3 X
AA1.05 Reactor/turbine pressure regulating system Knowledge of the reasons for the following responses as they apply to SCRAM : 3.1 5 295006 SCRAM / 1 X
AK3.04 Reactor water level setpoint setdown: Plant-Specific Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT : 4.2 6 295016 Control Room Abandonment / 7 X
AA2.02 Reactor water level Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT 3.3 7 295018 Partial or Total Loss of CCW / 8 X COOLING WATER :
AA1.02 System loads 2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how 4.2 8 295019 Partial or Total Loss of Inst. Air / 8 X operator actions and directives affect plant and system conditions.
Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING : 3.4 9 295021 Loss of Shutdown Cooling / 4 X
AA2.02 RHR/shutdown cooling system flow Knowledge of the interrelations between REFUELING ACCIDENTS and the following: 3.4 10 295023 Refueling Acc / 8 X
AK2.06 Containment ventilation: Mark-III 2.2.37 Ability to determine operability and/or availability of safety related equipment. 3.6 11 295024 High Drywell Pressure / 5 X Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR 3.6 12 295025 High Reactor Pressure / 3 PRESSURE :
X EK1.03 Safety/relief valve tailpipe temperature/pressure relationships Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER 3.9 13 295026 Suppression Pool High Water Temp. / 5 X TEMPERATURE:
EA2.03 Reactor pressure 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. 3.9 14 295027 High Containment Temperature / 5 X Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: 3.6 15 295028 High Drywell Temperature / 5 X
EK2.03 Reactor water level indication Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: 3.4 16 295030 Low Suppression Pool Wtr Lvl / 5 X
EA1.03 HPCS: Plant-Specific Rev 1
ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO)
E/APE # / Name / Safety Function K K K A A2 G* K/A Topic(s) IR #
1 2 3 1 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL : 4.6* 17 295031 Reactor Low Water Level / 2 X
EA2.04 Adequate core cooling Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND 4.1 18 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 REACTOR POWER ABOVE APRM DOWNSCALE OR X UNKNOWN :
EK3.01 Recirculation pump trip/runback: Plant-Specific NOT SAMPLED --
295038 High Off-site Release Rate / 9 Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: 2.9 19 600000 Plant Fire On Site / 8 X
AK1.02 Fire Fighting Knowledge of the reasons for the following responses as they apply to GENERATOR VOLTAGE AND ELECTRIC 3.6 20 700000 Generator Voltage and Electric Grid Disturbances / 6 GRID DISTURBANCES:
X AK3.02 Actions contained in abnormal operating procedure for voltage and grid disturbances K/A Category Totals: 3 3 4 3 4 3 Group Point Total: 20 Rev 1
ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 Knowledge of the operational implications of the following 295002 Loss of Main Condenser Vac / 3 3.6 21 concepts as they apply to LOSS OF MAIN CONDENSER X VACUUM:
AK1.03 Loss of heat sink 295007 High Reactor Pressure / 3 NOT SAMPLED 295008 High Reactor Water Level / 2 NOT SAMPLED Knowledge of the interrelations between LOW REACTOR 295009 Low Reactor Water Level / 2 3.9 22 X WATER LEVEL and the following:
AK2.02 Reactor water level control 295010 High Drywell Pressure / 5 NOT SAMPLED Knowledge of the reasons for the following responses as 295011 High Containment Temp / 5 3.6 23 they apply to HIGH CONTAINMENT TEMPERATURE X (MARK III CONTAINMENT ONLY):
AK3.01 Increased containment cooling: Mark-III 295012 High Drywell Temperature / 5 NOT SAMPLED 295013 High Suppression Pool Temp. / 5 NOT SAMPLED Ability to operate and/or monitor the following as they apply 295014 Inadvertent Reactivity Addition / 1 3.3 24 to INADVERTENT REACTIVITY ADDITION:
X AA1.06 Reactor/turbine pressure regulating system 295015 Incomplete SCRAM / 1 NOT SAMPLED 295017 High Off-site Release Rate / 9 NOT SAMPLED Ability to determine and/or interpret the following as they 295020 Inadvertent Cont. Isolation / 5 & 7 3.7 25 X apply to INADVERTENT CONTAINMENT ISOLATION:
AA2.03 Reactor power 295022 Loss of CRD Pumps / 1 NOT SAMPLED 2.4.4 Ability to recognize abnormal indications for system 295029 High Suppression Pool Wtr Lvl / 5 4.5 26 X operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
295032 High Secondary Containment Area NOT SAMPLED Temperature / 5 Knowledge of the interrelations between HIGH 295033 High Secondary Containment Area 3.8 27 SECONDARY CONTAINMENT AREA RADIATION LEVELS Radiation Levels / 9 X and the following:
EK2.01 Area radiation monitoring system 295034 Secondary Containment Ventilation NOT SAMPLED High Radiation / 9 295035 Secondary Containment High NOT SAMPLED Differential Pressure / 5 295036 Secondary Containment High NOT SAMPLED Sump/Area Water Level / 5 500000 High CTMT Hydrogen Conc. / 5 NOT SAMPLED K/A Category Point Totals: 1 2 1 1 1 1 Group Point Total: 7 Rev 1
ES-401 5 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 A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 Knowledge of RHR/LPCI: INJECTION MODE (PLANT 3.7 28 SPECIFIC) design feature(s) and/or interlocks which provide for the following:
203000 RHR/LPCI: Injection Mode X X K4.07 Emergency generator load sequencing 2.4.31 Knowledge of annunciator alarms, indications, 4.2 29 or response procedures.
Ability to manually operate and/or monitor in the 3.7 30 205000 Shutdown Cooling X control room:
A4.01 SDC/RHR pumps 206000 HPCI N/A for GGNS 20700 Isol Condenser N/A for GGNS Knowledge of the operational implications of the 2.6 31 following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM :
K5.01 Indications of pump cavitation 209001 LPCS X X --- ---
Ability to monitor automatic operations of the LOW PRESSURE CORE SPRAY SYSTEM including: 3.6 32 A3.06 Lights and alarms Ability to (a) predict the impacts of the following on the 3.4 33 HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) ;
and (b) based on those predictions, use procedures to 209002 HPCS X correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.13 Low condensate storage tank level BWR-5,6 Knowledge of the physical connections and/or cause- 3.2* 34 effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following:
K1.09 Core spray system: Plant-Specific 211000 SLC X X -------------------------------------- --- ---
Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID 3.6 35 CONTROL SYSTEM controls including:
A1.04 Valve operations Knowledge of the effect that a loss or malfunction of 3.6 36 the following will have on the REACTOR 212000 RPS X PROTECTION SYSTEM :
K6.01 A.C. electrical distribution Knowledge of the operational implications of the 3.0 37 following concepts as they apply to INTERMEDIATE 215003 IRM X RANGE MONITOR (IRM) SYSTEM :
K5.03 Changing detector position Knowledge of SOURCE RANGE MONITOR (SRM) 3.7 38 SYSTEM design feature(s) and/or interlocks which 215004 Source Range Monitor X provide for the following:
K4.01 Rod withdrawal blocks Rev 1
ES-401 6 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 A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 Knowledge of the effect that a loss or malfunction of 3.8 39 the AVERAGE POWER RANGE MONITOR/LOCAL 215005 APRM / LPRM X POWER RANGE MONITOR SYSTEM will have on following:
K3.05 Reactor power indication Knowledge of electrical power supplies to the 2.8* 40 217000 RCIC X following:
K2.02 RCIC initiation signals (logic)
Knowledge of the physical connections and/or cause- 3.9 41 effect relationships between AUTOMATIC 218000 ADS X DEPRESSURIZATION SYSTEM and the following:
K1.05 Remote shutdown system: Plant-Specific Knowledge of the effect that a loss or malfunction of 2.9 42 the PRIMARY CONTAINMENT ISOLATION 223002 PCIS/Nuclear Steam SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will X
Supply Shutoff have on following:
K3.10 Reactor water cleanup Knowledge of electrical power supplies to the 2.8* 43 239002 SRVs X following:
K2.01 SRV solenoids Knowledge of REACTOR WATER LEVEL CONTROL 3.5 44 259002 Reactor Water Level SYSTEM design feature(s) and/or interlocks which X provide for the following:
Control K4.12 Manual and automatic control of the system Knowledge of the effect that a loss or malfunction of 2.9 45 the following will have on the STANDBY GAS 261000 SGTS X TREATMENT SYSTEM :
K6.04 Process radiation monitoring Knowledge of the effect that a loss or malfunction of 3.8 46 the A.C. ELECTRICAL DISTRIBUTION will have on following:
K3.02 Emergency generators 262001 AC Electrical Distribution X X Knowledge of the operational implications of the following concepts as they apply to A.C. ELECTRICAL 3.1 47 DISTRIBUTION:
K5.01 Principle involved with paralleling two A.C.
sources 2.8 48 Ability to manually operate and/or monitor in the 262002 UPS (AC/DC) X control room:
A4.01 Transfer from alternative source to preferred source Ability to monitor automatic operations of the D.C. 3.2 49 ELECTRICAL DISTRIBUTION including:
263000 DC Electrical Distribution X A3.01 Meters, dials, recorders, alarms, and indicating lights Rev 1
ES-401 7 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 A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 Ability to (a) predict the impacts of the following on the 3.4 50 EMERGENCY GENERATORS (DIESEL/JET) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.03 Operating unloaded, lightly loaded, and highly 264000 EDGs X X loaded Ability to manually operate and/or monitor in the control room: 3.6 51 A4.05 Transfer of emergency generator (with load) to grid Knowledge of electrical power supplies to the 2.8 52 300000 Instrument Air X following:
K2.01 Instrument air compressor 2.1.28 Knowledge of the purpose and function of major 4.1 53 400000 Component Cooling Water X system components and controls.
K/A Category Point Totals: 2 3 3 3 3 2 1 2 2 3 2 Group Point Total: 26 Rev 1
ES-401 8 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 NOT SAMPLED 201002 RMCS N/A for GGNS 201003 Control Rod and Drive NOT SAMPLED Mechanism 201004 RSCS N/A for GGNS 201005 RCIS X Ability to (a) predict the impacts of the following 3.2 54 on the ROD CONTROL AND INFORMATION SYSTEM (RCIS) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.04 Withdraw block: BWR-6 201006 RWM N/A for GGNS 202001 Recirculation X Knowledge of the physical connections and/or 3.2 55 cause-effect relationships between RECIRCULATION SYSTEM and the following:
K1.19 Feedwater flow 202002 Recirculation Flow Control NOT SAMPLED 204000 RWCU X Knowledge of the effect that a loss or malfunction 2.6 56 of the REACTOR WATER CLEANUP SYSTEM will have on following:
K3.06 Area radiation levels 214000 RPIS N/A for GGNS 215001 Traversing In-Core Probe NOT SAMPLED 215002 RBM N/A for GGNS 216000 Nuclear Boiler Inst. NOT SAMPLED 219000 RHR/LPCI: Torus/Pool Cooling X Knowledge of the operational implications of the 2.9 57 Mode following concepts as they apply to RHR/LPCI:
TORUS / SUPPRESSION POOL COOLING MODE:
K5.04 Heat exchanger operation 223001 Primary CTMT and Aux. NOT SAMPLED 226001 RHR/LPCI: CTMT Spray Mode X Ability to predict and/or monitor changes in 3.0 58 parameters associated with operating the RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE controls including:
A1.10 Emergency generator loading 230000 RHR/LPCI: Torus/Pool Spray N/A for GGNS Mode 233000 Fuel Pool Cooling/Cleanup NOT SAMPLED 234000 Fuel Handling Equipment X Ability to manually operate and/or monitor in the 3.7 59 control room:
A4.01 Neutron monitoring system 239001 Main and Reheat Steam NOT SAMPLED 239003 MSIV Leakage Control NOT SAMPLED Rev 1
ES-401 9 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 241000 Reactor/Turbine Pressure X Ability to monitor automatic operations of the 3.8 60 Regulator REACTOR/TURBINE PRESSURE REGULATING SYSTEM including:
A3.08 Steam bypass valve operation 245000 Main Turbine Gen. / Aux. X Knowledge of the effect that a loss or malfunction 3.0 61 of the following will have on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS:
K6.06 Electrical distribution 256000 Reactor Condensate X Knowledge of electrical power supplies to the 2.7* 62 following:
K2.01 System pumps 259001 Reactor Feedwater X Knowledge of the physical connections and/or 3.6 63 cause-effect relationships between REACTOR FEEDWATER SYSTEM and the following:
K1.08 Reactor water level control system 268000 Radwaste NOT SAMPLED 271000 Offgas NOT SAMPLED 272000 Radiation Monitoring NOT SAMPLED 286000 Fire Protection X 2.4.50 Ability to verify system alarm setpoints and 4.2 64 operate controls identified in the alarm response manual.
288000 Plant Ventilation NOT SAMPLED 290001 Secondary CTMT NOT SAMPLED 290003 Control Room HVAC X Knowledge of CONTROL ROOM HVAC design 3.1 65 feature(s) and/or interlocks which provide for the following:
K4.01 System initiations/reconfiguration: Plant-Specific 290002 Reactor Vessel Internals NOT SAMPLED 204000 RWCU NOT SAMPLED K/A Category Point Totals: 2 1 1 1 1 1 1 1 1 1 1 Group Point Total: 12 Rev 1
ES-401 Generic Knowledge and Abilities Outline (Tier 3 - RO) Form ES-401-3 Facility: Grand Gulf Nuclear Station Date of Exam: December 2017 Category K/A # Topic RO SRO-Only IR # IR #
2.1.1 Knowledge of conduct of operations requirements. 3.8 66 2.1.5 Ability to use procedures related to shift staffing, such as minimum 2.9* 67 crew complement, overtime limitations, etc.
1.
Conduct of 2.1.45 Ability to identify and interpret diverse indications to validate the 4.1 68 Operations response of another indication.
Subtotal 3 2.2.12 Knowledge of surveillance procedures. 3.7 69 2.2.14 Knowledge of the process for controlling equipment configuration or 3.9 70
- 2. status.
Equipment 2.2.43 Knowledge of the process used to track inoperable alarms. 3.0 71 Control Subtotal 3 2.3.13 Knowledge of radiological safety procedures pertaining to licensed 3.4 72 operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
- 3. 2.3.14 Knowledge of radiation or contamination hazards that may arise during 3.4 73 Radiation normal, abnormal, or emergency conditions or activities.
Control Subtotal 2 2.4.12 Knowledge of general operating crew responsibilities during 4.0 74 emergency operations.
2.4.34 Knowledge of RO tasks performed outside the main control room 4.2 75
- 4. during an emergency and the resultant operational effects.
Emergency Procedures /
Plan Subtotal 2 Tier 3 Point Total 10 Rev 1
ES-401 BWR Examination Outline Form ES-401-1 Facility: Grand Gulf Nuclear Station Date of Exam: December 2017 Tier Group RO K/A Category Points SRO-Only Points K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total
- 1. 1 4 3 7 Emergency &
2 2 1 3 Abnormal Plant Evolutions Tier Totals 6 4 10 1 3 2 5 2.
Plant 2 1 1 1 3 Systems Tier Totals 5 3 8
- 3. Generic Knowledge and Abilities Categories 1 2 3 4 1 2 3 4 7 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.
The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply.) Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As
- These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
- These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
Rev 0
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)
K/A Topic(s)
E/APE # / Name / Safety Function K K K A A G* IR #
1 2 3 1 2 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF 3.7 76 295003 Partial or Complete Loss of AC / 6 X A.C. POWER:
AA2.04 System lineups 295004 Partial or Total Loss of DC Pwr / 6 Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP : 3.8 77 295005 Main Turbine Generator Trip / 3 X
AA2.04 Reactor pressure 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the 4.2 78 295006 SCRAM / 1 X status of limiting conditions for operations.
295016 Control Room Abandonment / 7 295018 Partial or Total Loss of CCW / 8 295019 Partial or Total Loss of Inst. Air / 8 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING 3.5 79 295021 Loss of Shutdown Cooling / 4 X AA2.03 Reactor water level 2.4.3 Ability to identify post-accident 3.9 80 295023 Refueling Acc / 8 X instrumentation.
295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp. / 5 Ability to determine and/or interpret the following as they apply to HIGH CONTAINMENT 3.7 81 295027 High Containment Temperature / 5 X TEMPERATURE (MARK III CONTAINMENT ONLY)
EA2.01 Containment temperature: Mark-III 295028 High Drywell Temperature / 5 295030 Low Suppression Pool Wtr Lvl / 5 295031 Reactor Low Water Level / 2 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 295038 High Off-site Release Rate / 9 2.4.30 Knowledge of events related to system operation/status that must be reported to internal 4.1 82 600000 Plant Fire On Site / 8 organizations or external agencies, such as the X State, the NRC, or the transmission system operator.
700000 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 0 0 0 0 4 3 Group Point Total: 7 Rev 0
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 Ability to determine and/or interpret the following 4.1 83 295007 High Reactor Pressure / 3 X as they apply to HIGH REACTOR PRESSURE AA2.02 Reactor power 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 Ability to determine and/or interpret the following 4.2 84 295015 Incomplete SCRAM / 1 X as they apply to INCOMPLETE SCRAM AA2.02 Control rod position 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 295032 High Secondary Containment Area Temperature / 5 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Level / 5 2.4.41 Knowledge of the emergency action level 4.6 85 500000 High CTMT Hydrogen Conc. / 5 X thresholds and classifications.
K/A Category Point Totals: 0 0 0 0 2 1 Group Point Total: 3 Rev 0
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 A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 G2.1.23: Ability to perform specific system 4.4 86 203000 RHR/LPCI: Injection X and integrated plant procedures during all Mode modes of plant operations 205000 Shutdown Cooling 206000 HPCI N/A for GGNS 20700 Isol Condenser N/A for GGNS 209001 LPCS 209002 HPCS 211000 SLC Ability to (a) predict the impacts of the 4.1 87 following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, 212000 RPS X control, or mitigate the consequences of those abnormal conditions or operations:
A2.11 Main steamline isolation valve closure 215003 IRM 215004 Source Range Monitor 215005 APRM / LPRM 217000 RCIC Ability to (a) predict the impacts of the following on the AUTOMATIC 3.6 88 DEPRESSURIZATION SYSTEM ; and (b) based on those predictions, use 218000 ADS X procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.05 Loss of A.C. or D.C. power to ADS valves 223002 PCIS/Nuclear Steam Supply Shutoff G2.2.12 Knowledge of Surveillance 89 239002 SRVs X 4.1 procedures 259002 Reactor Water Level Control 261000 SGTS 262001 AC Electrical Distribution 262002 UPS (AC/DC) 263000 DC Electrical Distribution Ability to (a) predict the impacts of the following on the EMERGENCY 4.1 90 GENERATORS (DIESEL/JET) ; and (b) based on those predictions, use 264000 EDGs X procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.09 Loss of A.C. power 300000 Instrument Air 400000 Component Cooling Water K/A Category Point Totals: 0 0 0 0 0 0 0 3 0 0 2 Group Point Total: 5 Rev 0
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 A2 A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 3 4 201001 CRD Hydraulic 201002 RMCS N/A for GGNS 201003 Control Rod and Drive Mechanism 201004 RSCS N/A for GGNS 201005 RCIS 201006 RWM N/A for GGNS Ability to (a) predict the impacts of the 202001 Recirculation X following on the RECIRCULATION 3.8 91 SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.12 Loss of reactor feedwater 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS N/A for GGNS 215001 Traversing In-Core Probe 215002 RBM N/A for GGNS 216000 Nuclear Boiler Inst.
219000 RHR/LPCI: Torus/Pool Cooling Mode 223001 Primary CTMT and Aux.
226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray N/A for GGNS Mode 233000 Fuel Pool Cooling/Cleanup Knowledge of FUEL HANDLING 234000 Fuel Handling Equipment X EQUIPMENT design feature(s) and/or 4.1 92 interlocks which provide for the following:
K4.02 Prevention of control rod movement during core alterations 239001 Main and Reheat Steam 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Regulator 245000 Main Turbine Gen. / Aux.
256000 Reactor Condensate 2.4.9 Knowledge of low power/shutdown 259001 Reactor Feedwater X implications in accident (e.g., loss of 4.2 93 coolant accident or loss of residual heat removal) mitigation strategies.
268000 Radwaste 271000 Offgas 272000 Radiation Monitoring 286000 Fire Protection 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel Internals 204000 RWCU K/A Category Point Totals: 0 0 0 1 0 0 0 1 0 0 1 Group Point Total: 3 Rev 0
ES-401 Generic Knowledge and Abilities Outline (Tier 3-SRO) Form ES-401-3 Facility: Grand Gulf Nuclear Station Date of Exam: December 2017 Category K/A # Topic RO SRO-Only IR # IR #
2.1.2 Knowledge of operator responsibilities during all modes 4.4 94 of plant operation.
2.1.37 Knowledge of procedures, guidelines, or limitations 4.6 95
- 1. associated with reactivity management.
Conduct of Operations Subtotal 2 2.2.7 Knowledge of the process for conducting special or 3.6 96 infrequent tests.
2.2.21 Knowledge of pre- and post-maintenance operability 4.1 97
- 2. requirements.
Equipment Control Subtotal 2 2.3.4 Knowledge of radiation exposure limits under normal or 3.7 98 emergency conditions.
3.
Radiation Control Subtotal 1 2.4.29 Knowledge of the emergency plan. 4.4 99 2.4.42 Knowledge of emergency response facilities. 3.8 100 4.
Emergency Procedures / Plan Subtotal 2 Tier 3 Point Total 7 Rev 0
ES-401 Record of Rejected K/As Form ES-401-4 RO EXAM Tier/Group Randomly Reason for Rejection Selected K/A 1/1 295016 - AA2.01 AA2. Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT:
AA2.01 Reactor power At GGNS we are unable to determine Reactor Power external from the main control room. Also, having an ATWS during a Control Room Abandonment is outside our design bases.
Randomly selected K/A - AA2.02, Reactor water level.
1/1 295027 - 2.1.19 2.1.19 Ability to use plant computers to evaluate system or component status.
At GGNS the use of plant computers to evaluate Containment Temperatures is limited, Control Room panel indication is used.
Randomly selected K/A - 2.1.25, Ability to interpret reference materials, such as graphs, curves, tables, etc.
1/1 295037 - EK3.08 EK3. Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :
EK3.08 ATWS circuitry: Plant-Specific GGNS doesnt have an ATWS Circuitry.
Randomly selected K/A - EK3.01, Recirculation pump trip/runback 2/1 215004 - K4.02 K4. Knowledge of SOURCE RANGE MONITOR (SRM) SYSTEM design feature(s) and/or interlocks which provide for the following:
K4.02 Reactor SCRAM signals The SRMs do not provide Scram signals unless the shorting links are removed. There are no current procedures that will allow the shorting links to be removed at GGNS.
Randomly selected K/A - K4.01, Withdrawal Blocks
2/1 223002 - K3.20 K3. Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following:
K3.20 Standby gas treatment system At GGNS the Primary Containment Isolation system has no affect on the Standby gas treatment system.
Randomly selected K/A - K3.10, RWCU 2/1 262002 - A1.02 A1. Ability to predict and/or monitor changes in parameters associated with operating the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) controls including:
A1.02 Motor generator outputs Motor Generators (MGs) are not used for Uninterruptable power at GGNS.
Randomly selected K/A - A4.01 A4. Ability to manually operate and/or monitor in the control room:
A4.01 Transfer from alternative source to preferred source.
A1 had only one other selection, it has a 2.4 importance factor, therefore A4 was randomly selected.
2/2 234000 - A4.02 A4. Ability to manually operate and/or monitor in the control room:
A4.02 Control rod drive system There is no tie between Fuel Handling Equipment and the CRD system at GGNS.
Randomly selected K/A - A4.01, Neutron Monitoring system.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 12/4/2017 Examination Level: RO SRO Operating Test Number: LOT12-2017 Administrative Topic (see Note) Type Describe activity to be performed Code*
Primary Containment Water Level Determination GJPMOPS2017IAR1 Conduct of Operations K/A 2.1.25: 3.9; 2.1.20: 4.6; 2.4.21: 4.0 R; D Perform AC/DC Lineup Surveillance GJPMOPS2017IAR2 Conduct of Operations K/A 2.1.31: 4.6; 2.2.12: 3.7; 2.1.20: 4.6 R; M Determine Tagging Requirements GJPMOPS2017IAR3 Equipment Control K/A 2.2.41: 3.5; 2.2.13: 4.1 R; D Radiation Control Emergency Notifications GJPMOPS2017IAR4 Emergency Plan K/A 2.4.43: 3.2; 2.4.39: 3.9 R; N NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes & 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)
Rev 0 6/28/2017
ES-301 Administrative Topics Outline Form ES-301-1 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 12/4/2017 Examination Level: RO SRO Operating Test Number: LOT12-2017 Administrative Topic (see Note) Type Describe activity to be performed Code*
Perform EOOS Risk Assessment GJPM-OPS-2017IAS1 K/A 2.1.39: 4.3 Conduct of Operations R; M Review Completed Surveillance GJPM-OPS-2017IAS2 K/A 2.1.2: 4.4; 2.1.7: 4.7; 2.2.12: 4.1; Conduct of Operations R; N 2.2.22: 4.7 Determine Impact on Plant Operations for Failed Relay GJPM-OPS-2017IAS3 Equipment Control R; D; P K/A 2.2.41: 3.9; 2.2.22: 4.7; 2.2.36: 4.2 Authorize Emergency Exposure GJPM-OPS-2017IAS4 K/A 2.3.4: 3.7 Radiation Control R; M Emergency Classification GJPM-OPS-2017IAS5 K/A 2.4.41: 4.6 Emergency Plan R; N NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes & 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)
Rev. 0 6/28/2017
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 12/04/2017 Exam Level: RO SRO-I SRO-U Operating Test No.: LOT12-2017 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. 217000 A4.04 (3.6), RCIC Manual Startup, A-D-S 2 GJPM-OPS-2017IS1
- b. 239001 A2.11 (4.1), Slow Closing MSIVs, P-A-D-S 3 GJPM-OPS-2017IS2
- c. 205000 A4.01 (3.7), Startup Shutdown Cooling A, A-D-L-S 4 GJPM-OPS-2017IS3
- d. 264000 A4.04 (3.7), Start, Parallel and Load Div. 1 DG, A-D-S 6 GJPM-OPS-2017IS4
- e. 223001 A4.13 (3.4), Startup Hydrogen Recombiner, EN-D-S 5 GJPM-OPS-2017IS5
- f. 261000 A4.02 (3.1), Secure Standby Gas Treatment following EN-N-S 9 Automatic Initiation, GJPM-OPS-2017IS6
- g. 201005 A2.04 (3.2), Bypass a Control Rod in RACS, D-C 7 GJPM-OPS-2017ICR1
- h. 202001 A4.01 (3.7), Shift Reactor Recirc Pumps to Fast Speed, A-D-S 1 GJPM-OPS-2017IS7 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. 219000 A4.01 (3.8), Align Suppression Pool Cooling from P-A-E-D 5 Remote Shutdown Panel, GJPM-OPS-2017IP1
@ 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 Rev 0 6/28/2017
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: GRAND GULF NUCLEAR STATION Date of Examination: 12/04/2017 Exam Level: RO SRO-I SRO-U Operating Test No.: LOT12-2017 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. 217000 A4.04 (3.6), RCIC Manual Startup, A-D-S 2 GJPM-OPS-2017IS1
- b. 239001 A2.11 (4.3), Slow Closing MSIVs, P-A-D-S 3 GJPM-OPS-2017IS2
- c. 205000 A4.01 (3.7), Startup Shutdown Cooling A, A-D-L-S 4 GJPM-OPS-2017IS3
- d. 264000 A4.04 (3.7), Start, Parallel and Load Div. 1 DG, A-D-S 6 GJPM-OPS-2017IS4
- e. 223001 A4.13 (3.4), Startup Hydrogen Recombiner, EN-D-S 5 GJPM-OPS-2017IS5
- f. 261000 A4.02 (3.1), Secure Standby Gas Treatment following EN-N-S 9 Automatic Initiation, GJPM-OPS-2017IS6
- g. 201005 A2.04 (3.2), Bypass a Control Rod in RACS, D-C 7 GJPM-OPS-2017ICR1
- h. N/A In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i. 219000 A4.01 (3.7), Align Suppression Pool Cooling from P-A-E-D 5 Remote Shutdown Panel, GJPM-OPS-2017IP1
@ 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 Rev 0 6/28/2017
ES-301 Transient and Event Checklist Form ES-301-5 Facility: Grand Gulf Nuclear Station Date of Exam: 12/2017 Operating Test No.: LOT12/2017 A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I
C A S A B S A B S A B S A B M A T L R T O R T O R T O R T O U N Y O C P O C P O C P O C P M(*)
T P E R I U RX 0 0 1 1 1 1 0 NOR 1 1 1 3 1 1 1 I1 & I2 I/C 5 5 1 11 4 4 2 MAJ 1 1 2 4 2 2 1 TS 3 2 5 0 2 2 RX 0 0 1 1 1 1 0 NOR 1 1 1 3 1 1 1 I3 I/C 5 4 1 10 4 4 2 MAJ 1 1 2 4 2 2 1 TS 3 3 0 2 2 RX 1 0 0 1 1 1 0 NOR 0 1 1 2 1 1 1 I4 I/C 2 5 3 10 4 4 2 MAJ 1 1 2 4 2 2 1 TS 2 2 4 0 2 2 RX 1 0 1 1 1 0 NOR 0 1 1 1 1 1 R1 I/C 2 4 6 4 4 2 R3 MAJ 1 1 2 2 2 1 TS 0 2 2 RX 0 1 0 1 1 1 0 R2 NOR 1 0 1 2 1 1 1 R4 I/C 3 2 3 8 4 4 2 R5 MAJ 1 1 2 4 2 2 1 TS 0 2 2 RX 0 0 0 1 1 0 NOR 1 1 0 1 1 1 B/U I/C 6 3 3 4 4 2 MAJ 1 1 1 2 2 1 TS 2 0 2 2
Instructions:
- 1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
- 2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional I/C malfunctions on a one-for-one basis.
- 3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
- 4. For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 1 Page 1 of 7 Facility: Grand Gulf Nuclear Station Scenario No.: 1 Op-Test No.: GGNS 12-2017 Examiners: ____________________________ Operators: _____________________________
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
- 1. Start RHR A in Suppression Pool Cooling.
- 2. Respond to a LPCS Jockey Pump trip.
- 3. Respond to a RWCU Division 1 isolation signal with failure of G33-F250 to isolate due to power loss.
- 4. Respond to a Reactor Feed Pump A Speed Controller failure.
- 5. Respond to a Low Pressure Feedwater 4A tube leak with a failure to isolate.
- 6. Respond to a RCIC Steam Line break with a failure to isolate.
- 7. Respond to a low power ATWS (<5%) with failed fuel.
- 8. Respond to a HPCS Service Water Pump trip.
Initial Conditions: Plant is operating at approximately 100% power. Average Suppression Pool temperature is 85°F due to weeping SRVs.
Inoperable Equipment: None Turnover:
The plant is at 100% power.
Planned activities for this shift are:
Start RHR A in Suppression Pool Cooling to lower Suppression Pool temperature to 78°F to 80°F.
Scenario Notes:
This scenario is a NEW Scenario.
Validation Time: 60 minutes Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 1 Page 2 of 7 Event Malf. No. Event Type Event No. Description 1 N/A TS (CRS) Start RHR A in Suppression Pool Cooling.
N (BOP,CRS) 2 e21645 TS (CRS) Respond to a LPCS Jockey Pump trip.
C (BOP,CRS) tte31n043a TS (CRS) Respond to a Division 1 RWCU Isolation signal with G33-3 I(BOP,CRS) g33f250 F250 power loss.
A (CREW)
C(ATC,CRS) Respond to a Reactor Feed Pump A Speed Controller 4 fw272a A (CREW) failure.
n19f042a_f C (BOP,CRS) n19f040a_f R (ATC) Respond to a LP FW HTR 4A tube leak with a failure to 5 isolate.
fw232j A (CREW) e51050 6 e51187a M (CREW) Respond to RCIC steam line break with a failure to isolate.
e51187b ATWS < 5% power with Fuel Failure
- (CT-1) When control rods fail to scram, crew inserts control rods before exiting EP-2A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at c11164 least 5 control rod gangs using RC&IS.)
M (CREW) 7 rr071 * (CT-2) When two areas exceed their max safe radiation levels, emergency depressurize the RPV before exiting EP-4.
- (CT-3) When emergency depressurization is required, crew terminates and prevents all RPV injection, except RCIC, CRD, and Boron, prior to emergency depressurizing the RPV.
r21133b C (ATC) Respond to a HPCS Service Water Pump trip.
8 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec
- Critical Task (As defined in NUREG 1021 Appendix D)
CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.
Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 1 Page 3 of 7 Quantitative Attributes Table Attribute E3-304-1 Actual Description Target Malfunctions after EOP entry 1-2 1 HPCS Service Water Pump trip Division 1 RWCU Isolation signal with G33-F250 power loss Abnormal Events 2-4 3 Reactor Feed Pump A Speed Controller failure LP FW HTR 4A leak with a failure to isolate 2 RCIC steam line break with a failure to isolate Major Transients 1-2 ATWS < 5% power/Fuel Failure EOP entries requiring EP-4 1-2 2 substantive action EP-2 EOP contingencies EP-2A ATWS requiring substantive 0-2 2 action EP-2A Emergency Depressurization (CT-1) When control rods fail to scram, crew inserts control rods before exiting EP-2A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rod gangs using EOP based Critical RC&IS.)
2-3 3 Tasks (CT-2) When two areas exceed their max safe radiation levels, emergency depressurize the RPV before exiting EP-4 (CT-3) When emergency depressurization is required, crew terminates and prevents all RPV injection, except RCIC, CRD, and Boron, prior to emergency depressurizing the RPV Normal Events N/A 1 Start RHR A in Suppression Pool Cooling Reactivity N/A Lower core flow to 70 mlbm/hr using Recirculation Flow Control Manipulations 1 Valves LPCS Jockey Pump trip Division 1 RWCU Isolation signal with G33-F250 power loss Instrument / Reactor Feed Pump A Speed Controller failure N/A 5 Component failures LP FW HRT 4A tube leak with failure to isolate HPCS Service Water Pump trip LPCS Jockey Pump trip Division 1 RWCU Isolation signal with G33-F250 power loss Reactor Feed Pump A Speed Controller failure Total Malfunctions N/A 7 LP FW HRT 4A tube leak with failure to isolate RCIC steam line break with a failure to isolate ATWS < 5% power/Fuel Failure HPCS Service Water Pump trip Top 10 systems and operator actions important to risk that are tested:
ADS (Event 7)
RPS (Event 7)
Condensate (Event 4)
Failure to manually depressurize with ADS/SRVs (Event 7)
Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 1 Page 4 of 7 SCENARIO ACTIVITIES:
Plant is operating at 100% power. Average Suppression Pool temperature is 85°F due to weeping SRVs.
Event 1 After the crew assumes the shift, the BOP will start RHR A in Suppression Pool Cooling mode. The CRS will enter Tech Spec 3.5.1 Condition A when E12-F024A, RHR A Test Return valve, is opened.
Event 2 (Triggered by Lead Examiner)
When RHR A is aligned in Suppression Pool Cooling mode and Tech Specs are addressed, the LPCS Jockey Pump will trip. The CRS will enter TS 3.5.1 Condition C and may direct BOP to place RHR A in standby. If E12-F024A, RHR A Test Return valve, is closed, CRS will exit Tech Spec 3.5.1 Condition C and determine Tech Spec 3.5.1 Condition A is still applicable. CRS will enter Tech Spec 3.3.6.4, Condition A and C, 3.3.3.1, Condition A, 3.3.3.2, Condition A and TR3.6.2.2, Condition A.
Event 3 (Triggered by Lead Examiner)
After RHR A has been secured from Suppression Pool Cooling and Tech Specs have been addressed, a Division 1 isolation of RWCU will occur due to a failed temperature instrument. G33-F250, RWCU SPLY TO RWCU HXS, will lose power and fail to isolate. CRS will enter 05-1-02-III-5, Automatic Isolations ONEP and direct BOP to close G33-F251, RWCU SPLY TO RWCU HXS. The CRS will enter Tech Specs 3.3.6.1 Condition A and 3.6.5.3 Condition A.
Event 4 (Triggered by Lead Examiner)
After G33-F251 has been closed and Tech Specs are addressed, the Reactor Feed Pump A Speed Controller will begin failing low. ATC will take manual control of the controller and balance Reactor Feed Pump controller outputs and stabilize RPV level. CRS will enter 05-1-02-V-7, Feedwater System Malfunctions ONEP.
Event 5 (Triggered by Lead Examiner)
When Reactor Feed Pump controller outputs have been balanced and RPV level has stabilized, a tube leak in Low Pressure Feedwater Heater 4A will occur. The heater isolation valves, N19-F040A and N19-F042A will fail to automatically close on the HI-HI level. The BOP will manually close the isolation valves to isolate the heater. The CRS enter 05-1-02-V-5, Loss of Feedwater Heating ONEP, and direct the ATC to lower core flow to 70 mlbm/hr using Recirc Flow Control Valves in slow detent. The CRS will enter 05-1-02-III-3, Reduction in Recirculation System Flow Rate ONEP.
Event 6 (Triggered by Lead Examiner)
After core flow has been lowered and RPV power, pressure and level are stable, a RCIC steam line break will occur. BOP will attempt to isolate RCIC by closing the RCIC Steam Isolation valves. RCIC steam isolation valves will lose power. CRS will enter EP-4 and direct the ATC to manually scram the reactor.
Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 1 Page 5 of 7 Event 7 (Triggered automatically)
When the reactor is scrammed, an ATWS occurs due to a hydraulic block of both scram discharge volumes with failed fuel, and EP-2A is entered via EP-2. Reactor power is below 5% rated thermal power. The crew will install the necessary attachments to bypass RPS and RC&IS interlocks and insert controls rods manually via RC&IS (CT-1). RPV level will be maintained in the normal band of +11.4 to
+53.5 narrow range. Bypass valves will control reactor pressure during this event. Feedwater is available for RPV level control. The crew may decide to lower RPV pressure to reduce the driving head of the leak using manual Bypass Valves control.
When two areas (RCIC Room and SBGTS) exceed the max safe radiation levels of EP-4, the CRS will direct the ATC and BOP to terminate and prevent all injection into the RPV (except RCIC, CRD and BORON) (CT-3) and emergency depressurize the RPV (CT-2). When RVP pressure has lowered to Minimum Steam Cooling Pressure of 206 psig, the CRS will direct the crew to slowly commence injection into the RPV with available systems to restore and maintain RPV level between 11.4 and 53.5 narrow range.
After the crew have emergency depressurized the RPV and inserted at least 5 gangs of control rods, or at the discretion of the Lead Examiner, the control rods are allowed to be fully inserted with the next scram. The CRS transitions from EP-2A to EP-2 and RPV level restoration is directed.
Event 8 (Triggered automatically)
When High Pressure Core Spray is initiated, the HPCS Service Water Pump will trip. The crew will secure the High Pressure Core Spray Diesel Generator due to lack of cooling water.
The exercise ends when controls rods are inserted and RPV water level is being maintained between
+11.4 inches and +53.5 inches narrow range.
Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 1 Page 6 of 7 Critical Task (CT-1) When control rods fail to scram, crew (CT-2) When two areas exceed their max safe inserts control rods before exiting EP-2A. (All radiation levels, emergency depressurize the RPV control rods do not have to be fully inserted to before exiting EP-4.
satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rod gangs using RC&IS.)
EVENT 7 7 Safety Failure to effect shutdown of the reactor when a If secondary containment radiation levels continue to Significance RPS setting has been exceeded would increase and exceed their maximum safe operating unnecessarily extend the level of degradation of values in more than one area, the RPV must be the safety of the plant. This could further degrade depressurized. RPV depressurization places the into damage to the principle fission product primary system in its lowest possible energy state, barriers if left unmitigated. The crew is authorized rejects heat to the suppression pool in preference to by Conduct of Operations to take mitigating outside the containment, and reduces the driving head actions when automatic safety systems fail to and flow of primary systems that are unisolated and perform their intended function. Action to shut discharging into the secondary containment.
down the reactor is required when RPS and control rod drive systems fail.
Cueing Manual scram is initiated and numerous control EP-4 max safe indication on PDS computer points.
rods indicate beyond position 02. Radiation levels exceeding max safe values on area Reactor power indicating > 0% rated thermal radiation instrumentation on panel P844.
power.
Performance Operator selects control rod gangs by depressing Manipulation of seven of the eight ADS/SRVs on panel Indicator the respective pushbuttons on panel P680 and P601:
inserts the rods by depressing the IN-TIMER SKIP B21-F041K pushbutton. B21-F047L B21-F041F B21-F047A B21-F051C B21-F041D B21-F051A B21-F051B Performance Operator selecting and inserting control rods Crew will observe ADS/SRV light indication go from Feedback indicated by rod position decreasing to 00 for green to red and reactor pressure lowering on SPDS selected rods on panel P680 and panel P601 indications.
Justification There is no time limit for effecting complete reactor There is no time frame for performing the emergency for the shutdown via control rod insertion. For the depressurization of the RPV when two area radiation chosen timeframe of this scenario, containment limits are levels exceed their max safe values. However, if the performance not challenged and power oscillations are not emergency depressurization is not performed before limit experienced. However, if the failure to scram EP EP-4 is exited, other procedures would not provide the were to be exited, other procedures would not guidance necessary to direct the depressurization.
provide the guidance necessary to achieve reactor Before exiting EP-4 ensures this guidance to shutdown. Before exiting EP-2A ensures emergency depressurize the RPV is not removed.
guidance to effect reactor shutdown is not removed.
BWR Owners App. B, step RC/Q-7 App. B, step SC/R-2.2 Group Appendix Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 1 Page 7 of 7 Critical Task (CT-3) When emergency depressurization is required, crew terminates and prevents all RPV injection, except RCIC, CRD, and Boron, prior to emergency depressurizing the RPV.
EVENT 7 Safety Injection into the RPV is terminated and prevented Significance while emergency RPV depressurization proceeds, in order to prevent uncontrolled injection of large amounts of cold water as RPV pressure decreases below the shutoff head of operating system pumps.
Injection from boron injection systems and CRD need not be terminated since the flowrates are relatively small and the systems may be needed to shut down the reactor. RCIC injection need not be terminated since its flowrate is also relatively small, turbine operation helps to depressurize the RPV, and RPV depressurization is not expected to result in significant flow variations.
Cueing EP-4 indications on PDS computer points in more than one area.
Radiation levels exceeding max safe values on more than one area radiation instruments on panel P844.
Manual scram is initiated and numerous control rods indicate beyond position 02.
Performance Operator manipulates Div 1 and Div 2 ECCS and Indicator HPCS manual initiation switches and associated pump and injection valve handswitches on panel P601.
Operator manipulates Master Level Controller or Startup Level Controller in MANUAL and lowers output to 0%.
Operator ensures N21-F009A and B and N21-F040 closed.
Performance Green light on and red light extinguished on ECCS Feedback pump and injection valve handswitches on panel P601.
ECCS pump and valve overridden annunciators on panel P601.
Feedwater flow indicating 0 mlbm/hr on panel P680 instruments.
Master Level Controller/Startup Level Controller output indicating 0%.
Green light on and red light extinguished on N21-F009A, F009B and F040 handswitches and CLOSE pushbuttons depressed on panel P680.
Justification Injection into the RPV is terminated and prevented for the while emergency RPV depressurization proceeds, chosen in order to prevent uncontrolled injection of large performance amounts of cold water as RPV pressure decreases limit below the shutoff head of operating system pumps.
Performance of this task before emergency depressurizing the RPV ensures that RPV injection sources are secured prior to RPV pressure lowers below the shutoff head of the associated pumps.
BWR Owners App. B, Contingency 5, step C5-5.1 Group Appendix Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 1 of 8 Facility: Grand Gulf Nuclear Station Scenario No.: 2 Op-Test No.: GGNS 12-2017 Examiners: ____________________________ Operators: _____________________________
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
- 2. Respond to an ESF Transformer 12 Lockout with failure of HPCS Diesel Generator to automatically start.
- 3. Respond to a CRD Pump B trip.
- 4. Respond to a control rod drifting in.
- 5. Respond to a stuck control rod.
- 6. Respond to a second control rod drifting in.
- 8. Respond to a Reactor Feed Pump trip.
Initial Conditions: Plant is operating at 100% power.
Inoperable Equipment: None Turnover:
The plant is at 100% power.
Planned activities for this shift are:
Transfer Bus 17AC from ESF Transformer 21 to ESF Transformer 12 in preparation for red-tagging breaker 152-1705, 17AC FFDR FM ESF 21, for preventative maintenance Scenario Notes:
This scenario is a NEW Scenario.
Validation Time: 60 minutes Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 2 of 8 Event Malf. No. Event Type Event No. Description N/A Transfer ESF Bus 17AC from ESF Transformer 21 to ESF 1 N (BOP,CRS)
Transformer 12.
r21134h TS (CRS) Respond to an ESF Transformer 12 Lockout with a failure of 2 C (BOP,CRS) n41141c HPCS Diesel Generator to auto start.
A (CREW)
C(BOP,CRS) 3 c11028b Respond to a CRD Pump B trip.
A (CREW)
C(ATC,CRS) 4 z022021_24_53 Respond to Control Rod 24-53 drifting in.
A(CREW)
R (ATC)
C(BOP,CRS) 5 z022022_24_53 Respond to Control Rod 24-53 stuck at position 32.
A(CREW)
TS (CRS)
Respond to Control Rod 32-27 drifting in. Reactor scram 6 z021021_32_37 C (ATC,CRS) due to two controls drifting in.
Respond to a Hydraulic Block ATWS > 5% RTP (EP-2, 2A)
- (CT-1) When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EP-2A.
(All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by c11164 M(CREW) fully inserting at least 5 control rod gangs using 7
RC&IS.)
- (CT-2) During failure to scram conditions with power
> 5% RTP, terminate and prevent all injection from all sources (except boron, CRD, and RCIC) as necessary to lower RPV level to below -70 wide range and control between -70 wide range to -191 CFZ (MSCWL) prior to exiting EP-2A.
Respond to a Reactor Feedwater Pump trip.
8 fw123a(b) C(BOP,CRS) *(CT-3) Restores injection using Condensate/Feedwater to restore/maintain RPV level above -191 CFZ before exiting EP-2A.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec
- Critical Task (As defined in NUREG 1021 Appendix D)
CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 3 of 8 Quantitative Attributes Table Attribute E3-304-1 Actual Description Target Malfunctions after EOP entry 1-2 1 Reactor Feed Water Pump trip ESF Transformer 12 lockout with a failure of HPCS Diesel Generator to automatically start Abnormal Events 2-4 4 CRD Pump B trip Control Rod 24-53 drifting in Control Rod 24-53 stuck at position 32 Major Transients 1-2 1 Hydraulic block ATWS with power > 5% RTP EOP entries requiring EP-2 1-2 2 substantive action EP-3 EOP contingencies requiring substantive 0-2 1 EP-2A ATWS action (CT-1) When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EP-2A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rod gangs using RC&IS.)
EOP based Critical 2-3 3 (CT-2) During failure to scram conditions with power > 5%
Tasks RTP, terminate and prevent all injection from all sources (except boron, CRD, and RCIC) as necessary to lower RPV level to below -70 wide range and control between -70 wide range to -191 CFZ (MSCWL) prior to exiting EP-2A.
(CT-3) Restores injection using Condensate/Feedwater to restore/maintain RPV level above -191 CFZ before exiting EP-2A.
Normal Events N/A Transfer ESF Bus 17AC from ESF Transformer 21 to ESF 1
Transformer 12 Reactivity N/A Lower core flow to 70 mlbm/hr using Reactor Recirc Flow Manipulations 1 Control Valves ESF Transformer 12 lockout with a failure of HPCS Diesel Generator to automatically start Instrument / CRD Pump B trip Component failures N/A 6 Control Rod 24-53 drifting in Control Rod 24-53 stuck Control Rod 32-37 drifting in Reactor Feed Water Pump A(B) trip ESF Transformer 12 lockout with a failure of HPCS Diesel Generator to automatically start CRD Pump B trip Control Rod 24-53 drifting in Total Malfunctions N/A 7 Control Rod 24-53 stuck Control Rod 32-37 drifting in Hydraulic block ATWS with power > 5% RTP Reactor Feed Water Pump A(B) trip Top 10 systems and operator actions important to risk that are tested:
RPS (Event 6)
ESF Power (Event 2)
Condensate (Event 8)
Failure to align alternate power to 4.16 KV or 6.9 KV buses (Event 2)
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 4 of 8 SCENARIO ACTIVITIES:
The plant is operating at 100% power.
Event 1 After the crew assumes the shift, the BOP will transfer ESF Bus 17AC from ESF Transformer 21 to ESF Transformer 12 per System Operating Instruction 04-1-01-R21-17, ESF BUS 17AC, Section 4.2.
Event 2 One minute after ESF Bus 17AC has been transferred, ESF Transformer 12 will lockout due to sudden pressure, causing a loss of power to ESF Bus 17AC. The HPCS Diesel Generator will fail to start. The BOP will recognize the failure of the HPCS Diesel Generator to auto start and restore Bus 17AC power from ESF Transformer 21 per 05-1-02-I-4, Loss of AC Power ONEP. The CRS will enter 05-1-02-I-4, Loss of AC Power ONEP, and TS 3.8.1.B for the HPCS Diesel Generator inoperable.
Event 3 (Triggered by Lead Examiner)
After Bus 17AC power has been restored and Tech Specs have been addressed, the B CRD pump will trip. The BOP will start the A CRD pump per 05-1-02-IV-1, CRD Malfunctions ONEP. The CRS will enter 05-1-02-IV-1, CRD Malfunctions ONEP.
Event 4 (automatically triggered)
When the A CRD pump is started, Control Rod 24-53 will begin drifting in. The ATC will select Control Rod 24-53 and apply a continuous insert signal per 05-1-02-IV-1, CRD Malfunctions ONEP. The CRS will re-enter 05-1-02-IV-1, CRD Malfunctions ONEP.
Event 5 (automatically triggered)
When Control Rod 24-53 reaches position 32, it will become stuck. The ATC will recognize and report Control Rod 28-53 has stopped inserting. The ATC will lower core flow to 70 mlbm/hr using Recirc Flow Control Valves in fast detent per 05-1-02-IV-1, CRD Malfunctions ONEP. The CRS will direct actions to raise CRD drive water pressure and attempt to insert Control Rod 24-53. When CRD Drive water pressure is raised to greater than 325 psig, Event 6 will automatically be triggered.
NOTE: Due to the amount of time required to complete the actions associated with the stuck control rod, a follow up question should be asked concerning the stuck rod and Technical Specification requirements (TS 3.1.3, Condition A).
Event 6 (automatically triggered)
When CRD Drive Water pressure is raised above 325 psig, Control Rod 32-37 will begin to drift in. The ATC will insert a manual scram per 05-1-02-IV-1, CRD Malfunctions ONEP.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 5 of 8 Event 7 (No trigger required)
When the reactor is scrammed, an ATWS occurs due to a hydraulic block of both scram discharge volumes, and EP-2A is entered via EP-2. Reactor power will be above 5% RTP. The crew will inject SLC and install the necessary attachments to bypass RPS and RC&IS interlocks and insert controls rods manually via RC&IS (CT-1). Terminate and Prevent is required because reactor power is above 5% RTP. RPV level is intentionally lowered below -70 inches wide range in order to lower core inlet subcooling and lower reactor power (CT-2). Bypass valves will control reactor pressure during this event. Feedwater is available for RPV level control.
Event 8 (Triggered when reactor level lowers below -70 inches wide range)
When reactor level lowers below -70 inches wide range, the running Reactor Feed Pump will trip. The BOP will restore Feedwater flow by starting the standby Reactor Feed Pump (CT-3). An alternate success path would be the CRS the ATC to lower reactor pressure to 450 to 600 psig to allow RPV injection with the Condensate Booster Pumps (CT-3).
After the crew has fully inserted 5 control rod gangs or at the direction of the Lead Examiner, the control rods are allowed to be fully inserted with the next scram. The CRS transitions from EP-2A to EP-2, SLC injection is stopped and RPV level restoration is directed.
The exercise ends when controls rods are inserted and RPV water level is being maintained between
+11.4 inches and +53.5 inches narrow range.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 6 of 8 Critical Task (CT-1) When control rods fail to scram, crew (CT-2) During failure to scram conditions with injects SLC and/or inserts control rods before power > 5%, terminate and prevent all injection exiting EP-2A. (All control rods do not have to from all sources (except boron, CRD, and RCIC) as be fully inserted to satisfy this critical task; this necessary to lower RPV level to below -70 wide only requires that the crew is making progress range to -191 CFZ (MSCWL) prior to exiting EP-2A.
to achieving all rods in by fully inserting at least 5 control rod gangs using RC&IS.)
EVENT 7 7 Safety Failure to effect shutdown of the reactor when a Regarding lowering level below -70 wide range, to Significance RPS setting has been exceeded would prevent or mitigate the consequences of any large unnecessarily extend the level of degradation of irregular neutron flux oscillations induced by the safety of the plant. This could further degrade neutronic/thermal-hydraulic instabilities. RPV water into damage to the principle fission product level is lowered sufficiently below the elevation of the barriers if left unmitigated. The crew is authorized feedwater sparger nozzles. This places the feedwater by Conduct of Operations to take mitigating spargers in the steam space providing effective heating actions when automatic safety systems fail to of the relatively cold feedwater and eliminating the perform their intended function. Action to shut potential for high core inlet subcooling. For conditions down the reactor is required when RPS and that are susceptible to oscillations, the initiation and control rod drive systems fail. growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.
24 below the lowest nozzle in the feedwater sparger has been selected as the upper bound of the RPV water level control band. This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment). This water level is sufficiently high that most plants without the capability to readily defeat the low RPV water level MSIV isolation should be able to control RPV water level with feedwater pumps to preclude the isolation.
Cueing Manual scram is initiated and numerous control Manual scram is initiated and numerous control rods rods indicate beyond position 02. indicate beyond position 02 and reactor power is > 5%
Reactor power indicating > 5% RTP on APRMs on on panel P680 indications and SPDS and RPV level is >
panel P680. -70 wide range on SPDS and PDS.
APRM downscale lights on panel P680APRM extinguished.
Performance Operator manipulates key-locked switches for SLC Operator manipulates the Master Level Controller in Indicator Pump A and B to START on panel P601. MANUAL on panel P680 and lowers output to 0% to Operator selects control rod gangs by depressing stop feedwater injection until RPV water level lowers the respective pushbuttons on panel P680 and below -70 wide range.
inserts the rods by depressing the IN-TIMER SKIP Operator manually initiates High Pressure Core Spray pushbutton. and Division 1 and Division 2 ECCS on panel P601, Operator resets reactor scram signal with key- then stops the respective pumps and overrides the locked switches on panel P680 and inserts manual associated injection valves closed using their respective reactor scram using scram pushbuttons on panel handswitches on panel P601.
P680.
Performance SLC A and B red lights illuminate, SLC discharge Feedwater flow indication on panel P680 and SPDS Feedback pressure rising, and SLC tank level lowering on indicate zero.
panel P601. Master Level Controller output indicates 0% on panel Operator selecting and inserting control rods P680.
indicated by rod position decreasing to 00 for High Pressure Core Spray, RHR A, RHR B, and RHR C selected rods on panel P680. pump and injection valve override annunciators illuminated on panel P601.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 7 of 8 Justification There is no time limit for effecting complete reactor Applicability for this CT is during EP-2A conditions for the shutdown via boron injection or control rod where it is necessary to lower level to control power with chosen insertion. For the timeframe of this scenario, no high energy input into the primary containment.
performance containment limits are not challenged and power There is no time limit for this lowering level, but it limit oscillations are not experienced. However, if the establishes margin to conditions where fuel damaging failure to scram EP were to be exited, other power oscillations may theoretically occur. Before procedures would not provide the guidance exiting EP-2A was chosen because other procedures necessary to achieve reactor shutdown. Before would not provide the guidance necessary to establish exiting EP-2A ensures guidance to effect reactor margin for power oscillation mitigation. Before exiting shutdown is not removed. EP-2A ensures guidance to effect this control is not removed.
NOTE - This critical task must be evaluated carefully based on the level changes. If power is reduced significantly below 5%, reactor water level may continue to rise above -70 with only CRD and SLC while driving control rods. This would not result in an UNSAT on this critical task.
BWR Owners App. B, step RC/Q6, RC/Q-7 App. B, Contingency #5 Step C5-4 Group Appendix Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 8 of 8 Critical Task (CT-3) Restores injection using Condensate/Feedwater to restore/maintain RPV level above -191 CFZ before exiting EP-2A.
EVENT 8 Safety If RPV water level cannot be restored and Significance maintained above the Minimum Steam Cooling RPV Water Level (-191 CFZ), emergency RPV depressurization is performed to maximize injection flow. Emergency depressurization is undesirable under ATWS conditions since the core response is difficult to predict and the risk of power excursions is increased.
Cueing Reactor Feed Pump trip annunciators and Feedwater flow and RPV level lowering on indicators on panel P680 and on PDS and SPDS.
Performance Operator manipulates switches on panel P680 Indicator panel to start the standby Reactor Feed Pump Alternately, operator lowers RPV pressure using Bypass Valves or SRVs to allow injection with Condensate Booster Pumps.
Performance Feedwater flow and RPV level rising on panel Feedback P680 and PDS and SPDS.
Justification The Minimum Steam Cooling RPV Water Level for the (-191 CFZ) is the lowest RPV water level at which chosen the covered portion of the reactor core will performance generate sufficient steam to preclude any clad limit temperature in the uncovered portion of the core from exceeding 1500°F. Maintaining RPV water level above the Minimum Steam Cooling RPV Water Level thus ensures that the core remains adequately cooled.
BWR Owners App. B, Contingency #5 Step C5-4 Group Appendix Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 1 of 7 Facility: Grand Gulf Nuclear Station Scenario No.: 3 Op-Test No.: GGNS 12-2017 Examiners: ____________________________ Operators: _____________________________
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
- 1. Transfer RPS Bus B from normal to alternate power supply.
- 2. Respond to a Division 2 Diesel Generator lube oil leak.
- 4. Respond to a loss of condenser vacuum.
- 5. Respond to a loss of Offsite Power / LOCA
- 6. Respond to a failure of Division 1 ECCS to automatically initiate.
Initial Conditions: Plant is operating at 100% power.
Inoperable Equipment: TBCW Pump C is tagged out for motor oil replacement.
Turnover:
The plant is at 100% power.
Planned activities for this shift are:
Transfer RPS Bus B from normal to alternate power supply Scenario Notes:
This scenario is a NEW Scenario.
Validation Time: 60 minutes Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 2 of 7 Event Malf. No. Event Type Event No. Description 1 N/A N (BOP,ATC,CRS) Transfer RPS Bus B from normal to alternate power supply.
2 p864_2a_d_2 TS (CRS) Respond to a Division 2 Diesel Generator lube oil leak.
I (ATC,BOP,CRS) ltb21n091b Respond to Division 2 ECCS on spurious RPV low level 3 A(CREW) ltb21n091f signal.
TS (CRS) 4 fw163c R (ATC,CRS) Respond to loss of condenser vacuum.
A(CREW)
Respond to a LOP/LOCA (EP-2, 3)
- (CT-1) Inhibit ADS prior to automatic ADS valve opening during a LOCA r21135 * (CT-2) When RPV level lowers to -160 wide range and M(CREW) cannot be maintained above -191 CFZ (MSCWL) and 5 rr063b insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening at least seven SRVs before RPV level lowers below -191 CFZ. (Momentary shrink below -191 due to automatic SRV closure does not constitute failure of this critical task).
Respond to a failure of Division 1 ECCS to automatically initiate.
rr040a * (CT-3) When operating injection systems cannot 6 C(BOP,CRS) rr041a maintain RPV level and ECCS systems fail to automatically initiate, crew manually initiates ECCS systems for injection prior to RPV pressure lowering below 300 psig.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec
- Critical Task (As defined in NUREG 1021 Appendix D)
CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 3 of 7 Quantitative Attributes Table Attribute E3-304-1 Actual Description Target Malfunctions after EOP entry 1-2 1 Failure of Division 1 ECCS to automatically initiate Spurious Division 2 ECCS initiation Abnormal Events 2-4 2 Loss of condenser vacuum LOP Major Transients 1-2 2 LOCA EOP entries requiring EP-2 1-2 2 substantive action EP-3 EOP contingencies EP-2 Alternate Level Control requiring substantive 0-2 2 action EP-2 Emergency Depressurization (CT-1) Inhibit ADS prior to automatic ADS valve opening during a LOCA (CT-2) When RPV level lowers to -160 wide range and cannot be maintained above -191 CFZ (MSCWL) and insufficient high pressure injection systems are available to EOP based Critical restore level, crew begins to Emergency Depressurize by 2-3 3 Tasks opening at least seven SRVs before RPV level lowers below
-191 CFZ. (Momentary shrink below -191 due to automatic SRV closure does not constitute failure of this critical task)
(CT-3) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically initiate, crew manually initiates ECCS systems for injection prior to RPV pressure lowering below 300 psig Normal Events N/A 1 Transfer RPS Bus B from normal to alternate power supply Reactivity Lower core flow to 70 mlbm using Reactor Recirc Flow Control N/A 1 Manipulations Valves Division 2 Diesel Generator lube oil leak Instrument / Spurious Division 2 ECCS initiation N/A 4 Component failures Loss of vacuum Failure of Division 1 ECCS to automatically initiate Division 2 Diesel Generator lube oil leak Spurious Division 2 ECCS initiation Total Malfunctions N/A 5 Loss of vacuum LOP/LOCA Failure of Division 1 ECCS to automatically initiate Top 10 systems and operator actions important to risk that are tested:
Div 1 & 2 EDGs (Event 2)
ADS (Event 5)
Offsite Power (Event 5)
Failure to manually depressurize with ADS/SRVs (Event 5)
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 4 of 7 SCENARIO ACTIVITIES:
The plant is operating at 100% power. TBCW Pump C is tagged out of service.
Event 1 After the crew assumes the shift, the BOP will transfer RPS Bus B from normal to alternate power supply per System Operating Instruction 04-1-01-C71-1, Reactor Protection System, Section 5.1. The ATC will reset the half-scram.
Event 2 (Triggered by Lead Examiner)
After RPS Bus B is transferred to alternate power, annunciator DIV 2 DSL GEN TROUBLE will alarm.
BOP will dispatch plant operator to investigate. After 3 minutes, the plant operator will report lube oil spraying out from the Division 2 Lube Oil Circulating Pump discharge piping and lube oil sump level is 20 below the top of the sump, which is less than 350 gallons. The BOP will place Division 2 Diesel Generator in the MAINTENANCE Mode. The CRS will enter LCO 3.8.1.B and LCO 3.8.3.E.
Event 3 (Triggered by Lead Examiner)
When Tech Specs have been addressed, a spurious Division 2 ECCS initiation on low RPV level will occur. The BOP will verify the initiation is spurious by two independent means and recover from the Division 2 ECCS initiation using 04-1-01-E12-1, Residual Heat Removal System SOI, Attachment 9 hardcard. The ATC will recognize the Division 2 half-scram due to RPS Bus B loss of power. CRS will enter 05-1-02-III-2, Loss of One or Both RPS Buses. BOP will restore RPS Bus B to normal power supply and the ATC will reset the Division 2 half-scram. The CRS will enter LCO 3.3.5.1.B.
Event 4 (Triggered by Lead Examiner)
When Division 2 ECCS initiation has been reset, systems have been secured and Tech Specs have been addressed, a main condenser leak will result in a slow loss of condenser vacuum. The CRS will enter 05-1-02-V-8, Loss of Condenser Vacuum ONEP. The ATC will lower core flow to 70 mlbm/hr using Recirc Flow Control Valves in fast detent. When condenser vacuum continues to lower, the ATC will insert a manual scram.
Event 5 (No trigger required)
When the reactor is scrammed, a total loss of offsite power occurs, followed by a small recirculation pipe break after 5 minutes. HPCS will trip when it is initiated. The CRS enters EP-2 and EP-3. RPV level will lower due to the leak being greater than the capacity of RCIC. When the CRS determines there are insufficient high pressure injection sources to maintain RPV level above -160 wide range, enters Alternate Level Control contingency of EP-2. Crew will inhibit ADS to prevent automatic operation (CT-1). When RPV level lowers to -160 wide range, the crew will emergency depressurize the RPV using ADS/SRVs (CT-2) and restore RPV level with Division 1 ECCS systems.
Event 6 (No Trigger required)
Division 1 ECCS will fail to automatically initiate on either high drywell pressure or low RPV level. The crew will manually initiate Division 1 ECCS using the lock-collared pushbutton (CT-3).
The exercise ends when emergency depressurization is complete and RPV level restoration is being controlled.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 5 of 7 Critical Task (CT-1) Inhibit ADS prior to automatic ADS (CT-2) When RPV level lowers to -160 wide range valve opening during a LOCA. and cannot be maintained above -191 CFZ (MSCWL) and insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening at least seven SRVs before RPV level lowers below -191 CFZ. (Momentary shrink below -191 due to automatic SRV closure does not constitute failure of this critical task.)
EVENT 5 5 Safety Permitting automatic ADS initiation may be The MSCWL is the lowest RPV water level at which the Significance undesirable for the following reasons: covered portion of the reactor core will generate ADS actuation can impose a severe thermal sufficient steam to preclude any clad temperature in the transient on the RPV and may complicate uncovered portion of the core from exceeding 1500°F.
efforts to control RPV water level. When water level decreases below MSCWL with injection, clad temperatures may exceed 1500°F.
If only steam-driven systems are available for injection, ADS actuation may directly lead to loss of adequate core cooling and subsequent core damage.
The conditions assumed in the design of the ADS actuation logic (e.g., no operator action for 115 seconds after event initiation) may not exist when the actions specified in this step are being performed.
The operating crew can draw on much more information than is available to the ADS logic (e.g., equipment out of service for maintenance, operating experience with certain systems, probability of restoration of off-site power, etc.) and can better judge, based on instructions contained in the EPGs/SAGs, when and how to depressurize the RPV.
Defeating the logic relieves the operating crew of the task of detecting timer initiation during execution of the more complex steps of Contingency #1 and precludes unnecessary and unwanted automatic initiations. Subsequent steps provide explicit and detailed instructions for controlling RPV water level and specify when emergency depressurization is appropriate.
Cueing ADS Timer initiated alarm on panel P601/19A-A1 Wide range indication (SPDS and PDS) falls to -160 and lowering trend continues, and, before -160 wide range is reached, initial conditions, field reports, and control room indications convey that adequate high pressure injection cannot be restored before level falls below -191 CFZ.
Performance Manipulation of ADS A and ADS B MANUAL Manipulation of seven of the eight ADS/SRVs on panel Indicator INHIBIT switches on panel P601 vertical section. P601:
B21-F041K B21-F047L B21-F041F B21-F047A B21-F051C B21-F041D B21-F051A B21-F051B Performance Inhibit switches click into INHIBIT position on Crew will observe ADS/SRV light indication go from Feedback panel P601 vertical section. green to red, reactor pressure lowering on SPDS and White indicating light on ADS A and ADS B panel P601 indications.
MANUAL INHIBIT switches illuminate.
Receipt of ADS/SRV A and ADS/SRV B OOSVC alarms on panel P601/18A-H2 and P601/19A-H2.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 6 of 7 Justification The 115 second ADS timer allows sufficient time The MSCWL (-191 CFZ) is the lowest RPV water level for the for the crew to recognize and override automatic at which the covered portion of the reactor core will chosen operation of the system. As long as ADS is generate sufficient steam to preclude any clad performance inhibited before ADS valves open, reactor temperature in the uncovered portion of the core from limit pressure will not be reduced. exceeding 1500°F. Emergency depressurization is allowed when level goes below TAF (-160 wide range) and should be performed, if in the judgment of the CRS, level cannot be maintained above -191 CFZ. Since it is intended for the scenario supporting this CT to, early in the event, clearly indicate no high pressure injection systems can be made available to reverse the lowering level trend, the crew will have time to communicate and opens at least seven ADS/SRVs before -191 CFZ.
BWR Owners App. B, step C1-1 App. B, Contingency #1 Step C1-4 Group Appendix Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 7 of 7 Critical Task (CT-3) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically initiate, crew manually initiates ECCS systems for injection prior to RPV pressure lowering below 300 psig.
EVENT 6 Safety Failure to recognize the auto initiation not Significance occurring, and failure to take manual action per Conduct of Ops will result in unavailability of safety-related equipment necessary to provide adequate core cooling, otherwise resulting in core damage and a large offsite release.
Cueing Indication of ECCS systems not initiating with initiation conditions present:
Indication of Drywell pressure 1.39 psig or RPV level -150.3 wide range White light on LPCS/RHR A INIT RESET pushbutton extinguished on panel P601 Green light on and red light extinguished on LPCS and RHR A pump handswitches on panel P601 Performance Operator manually initiates Division 1 ECCS by Indicator rotating the arming collar and depressing the LPCS/RHR A MAN INIT pushbutton on panel P601.
Performance Red light on and green light extinguished on LPCS Feedback and RHR A pump handswitches on panel P601.
Rising level trend on indications on panel P601, PDS and SPDS.
Rising flow rate on LPCS and/or RHR A flow indicators on panel P601, PDS, and SPDS.
Justification Attempting to align high pressure ECCS systems for the must be performed to determine their availability by chosen the time TAF is reached in order to properly performance implement EP-2 decision steps regarding restoring limit and maintaining RPV level. Attempting to align low pressure ECCS systems can only be done one RPV pressure falls below the injection valve RPV pressure permissive and will only be effective once RPV pressure falls below the shutoff head of the respective ECCS pump. The reduction in RPV pressure will normally be via Emergency Depressurization, which is a separate critical task bounded by a minimum RPV level.
BWR Owners App. B, Contingency 1, step C1-3 Group Appendix Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 1 of 9 Facility: Grand Gulf Nuclear Station Scenario No.: 4 Op-Test No.: GGNS 12-2017 Examiners: ____________________________ Operators: _____________________________
Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:
- 1. Withdraw control rods to 10% Bypass Valve position
- 3. Respond to Condensate Pump B trip.
- 4. Respond to a Bus 15AA Feeder Breaker trip with 15BA4 failure to re-energize.
- 5. Respond to a Startup Level Control controller failure.
- 6. Respond to an Aux Steam Tunnel steam leak with failure of one steam line to isolate.
- 7. Respond to a Feedwater Line B line break inside the Drywell with B21-F065B power loss.
- 8. Respond to a LPCS logic power failure.
Initial Conditions:
Reactor startup is in progress Reactor power is approximately 4% power Reactor pressure is 750 psig Inoperable Equipment: None Turnover:
Reactor startup is in progress o Step 90 of Control Rod Movement Sequence is complete o SJAE B is in service Condensate System is lined up as follows:
o Condensate Pumps B and C in service o Condensate Booster Pump A in service o Reactor Feed Pump A in service at approximately 950 psig discharge pressure o CFFF is in service o 4 Deep-Bed Condensate Demineralizers are in service Annunciators P680-4A2-C5, CONT ROD WITHDRAWAL BLOCK, and P680-4A1-A7, CRD DRIVE WTR TO RX P HI, are flagged as expected annunciators Planned activities for this shift are:
Withdraw control rods until 10% Bypass Valve position on the lagging valve, then continue raising TURB STM PRESSURE DEMAND setpoint to 935 psig per step 45 of IOI 03-1-01-1, Attachment XV Scenario Notes:
This scenario is a NEW Scenario.
Validation Time: 60 minutes Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 2 of 9 Event Malf. No. Event Type Event No. Description 1 N/A N (ATC,CRS) Withdraw control rods to 10% Bypass Valve position.
2 pte22n654c_a TS (CRS) Respond to a HPCS CST Level Lo trip unit failure.
C (ATC,CRS) Respond to a Condensate Pump B trip.
3 fw115a A (CREW)
(or) TS (CRS) Respond to a Bus 15AA feeder breaker trip with a failure of 4 di_r21m606a C(BOP,CRS) 15BA4 to re-energize.
r21142u A (CREW) 5 fw124 C(ATC,CRS) Respond to a Startup Level Control controller failure.
Respond to an Aux Steam Tunnel steam leak with failure of ms066a one steam line to isolate.
6 ms183a M (CREW) * (CT-1) When MSIVs fail to isolate, manually scram the ms184a reactor and close the MSIVs prior to Steam Tunnel temperature exceeding 250°F (Max Safe Temperature)
Respond to a Feedwater Line B line break inside Drywell with a B21-F065B power loss.
- (CT-2) When RPV level lowers to -160 wide range and cannot be maintained above -191 CFZ (MSCWL) and b21f065b_i 7 C (BOP,CRS) insufficient high pressure injection systems are fw171b available to restore level, crew begins to Emergency Depressurize by opening at least seven SRVs before RPV level lowers below -191 CFZ. (Momentary shrink below -191 due to automatic SRV closure does not constitute failure of this critical task.)
Respond to a LPCS logic power failure.
- (CT-3) When operating injection systems cannot r21219 C (ATC,CRS) maintain RPV level and ECCS systems fail to 8
automatically initiate, crew manually initiates ECCS systems for injection prior to RPV pressure lowering below 300 psig (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec
- Critical Task (As defined in NUREG 1021 Appendix D)
CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 3 of 9 Quantitative Attributes Table Attribute E3-304-1 Actual Description Target Respond to a Feedwater Line B line break inside the Drywell Malfunctions after 1-2 2 with a B21-F065B power loss EOP entry Respond to a LPCS Logic power failure Respond to a 15AA Feeder Breaker trip with a failure of 15BA4 Abnormal Events 2-4 2 to re-energize Respond to a Condensate Pump B trip Major Transients 1-2 Respond to an Aux Steam Tunnel steam leak with a failure of 1
one steam line to isolate EP-4 EOP entries requiring substantive action 1-2 3 EP-2 EP-3 EOP contingencies EP-2 Alternate Level Control requiring substantive 0-2 2 action EP-2 Emergency Depressurization (CT-1) When MSIVs fail to isolate, manually scram the reactor and close the MSIVs prior to Steam Tunnel temperature exceeding 250°F (Max Safe Temperature)
(CT-2) When RPV level lowers to -160 wide range and cannot be maintained above -191 CFZ (MSCWL) and insufficient high pressure injection systems are available to EOP based Critical restore level, crew begins to Emergency Depressurize by 2-3 3 Tasks opening at least seven SRVs before RPV level lowers below
-191 CFZ. (Momentary shrink below -191 due to automatic SRV closure does not constitute failure of this critical task.)
(CT-3) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically initiate, crew manually initiates ECCS systems for injection prior to RPV pressure lowering below 300 psig Normal Events N/A 1 Withdraw control rods to 10% Bypass Valve position Reactivity N/A 0 N/A Manipulations Respond to a HPCS CST Level LO trip unit failure Respond to Condensate Pump B trip Respond to a 15AA Feeder Breaker trip with a failure of 15BA4 Instrument / to re-energize N/A 6 Respond to a Startup Level Control controller failure Component failures Respond to a Feedwater Line B line break inside the Drywell with a B21-F065B power loss Respond to a LPCS logic power failure Respond to a HPCS CST Level LO trip unit failure Respond to Condensate Pump B trip Respond to a 15AA Feeder Breaker trip with a failure of 15BA4 to re-energize Respond to a Startup Level Control controller failure Total Malfunctions N/A 7 Respond to an Aux Steam Tunnel steam leak with failure of one steam line to isolate Respond to a Feedwater Line B line break inside the Drywell with a B21-F065B power loss Respond to a LPCS logic power failure Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 4 of 9 Top 10 systems and operator actions important to risk that are tested:
ADS (Event 7)
RHR (Event 8)
ESF Power (R20) (Event 4)
Condensate (Event 3)
Failure to manually depressurize with ADS/SRVs (Event 7)
Failure to align alternate power to 4.16 KV or 6.9KV buses (Event 4)
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 5 of 9 SCENARIO ACTIVITIES:
The plant is operating at 4% power.
Event 1 After the crew assumes the shift, the ATC will withdraw control rods in accordance with Control Rod Movement Sequence.
Event 2 (Triggered automatically)
When Control Rod 28-53GN is withdrawn to position 08, HPCS CST Level LO trip unit will fail high.
CRS will direct BOP to transfer HPCS suction from CST to Suppression Pool. CRS will enter TS 3.3.5.1 Condition A applies and, using Table 3.3.5.1-1, enter TS 3.3.5.1 Condition D.
Event 3 (Triggered by Lead Examiner)
After Tech Specs are addressed, Condensate Pump B will trip. CRS will enter 05-1-02-V-7, Feedwater System Malfunctions ONEP, and direct ATC to start Condensate Pump A.
Event 4 (Triggered by Lead Examiner)
After actions of Condensate Pump B trip are complete, Bus 15AA Feeder Breaker from ESF Transformer 11, 152-1514, will trip. Division 1 Diesel Generator will automatically start and supply Bus 15AA. BOP will recognize the failure of 15BA4 to re-energize. CRS will enter 05-1-02-III-5, Automatic Isolations ONEP and 05-1-02-III-1, Inadequate Decay Heat Removal ONEP. BOP will restore Instrument Air to CTMT by opening P53-F001. CRS will enter TS 3.8.7 Condition A.
NOTE: CRS is not expected to formulate plans for recovery of Fuel Pool Cooling and Cleanup or Reactor Water Cleanup systems within the time frame of this scenario.
Event 5 (Triggered by Lead Examiner)
After Tech Specs are addressed, the Startup Level Control Controller will begin failing low, resulting in RPV level lowering. ATC will take manual control of the Startup Level Controller and restore RPV level to normal band. CRS will enter 05-1-02-V-7, Feedwater System Malfunctions ONEP.
Event 6 (Triggered by Lead Examiner)
When RPV level has been returned to normal and stabilized, a steam leak in the Auxiliary Building Steam Tunnel will occur. The A Steam Line will fail to isolate. The CRS will enter EP-4 and direct the ATC to manually scram the reactor and the BOP to manually close B21-F022A, INBD MSIV, and B21-F028A, OTBD MSIV (CT-1). When the reactor is scrammed, the CRS will enter EP-2.
Event 7 (Triggered automatically)
Five minutes after the reactor is scrammed, an unisolable Feedwater Line B break in the Drywell will occur. The BOP will secure all Condensate Pumps and close B21-F065B, FW INL SHUTOFF VLV.
B21-F065B will not close due to a power loss when its CLOSE handswitch is depressed. ESF Bus 16AB will lockout and HPCS will trip when Drywell pressure rises to 1.39 psig. When CRS determines there are insufficient high pressure injection sources to maintain RPV level above -160 wide, enters Alternate Level Control contingency of EP-2. When RPV level lowers to -160 wide range, the crew will emergency depressurize the RPV using ADS/SRVs (CT-2) and restore RPV level with Division 1 ECCS systems.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 6 of 9 Event 8 (Triggered automatically)
When Drywell pressure rises to 1.39 psig, a LPCS logic power failure will occur. Crew will respond using ARI 04-1-02-1H13-P601-21A-H8, LPCS SYS OOSVC, and manually align Div 1 ECCS systems for injection to the RPV (CT-3).
The exercise ends when emergency depressurization is complete and RPV level restoration is being controlled.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 7 of 9 Critical Task (CT-1) When MSIVs fail to isolate, manually (CT-2) When RPV level lowers to -160 wide range scram the reactor and close the MSIVs prior to and cannot be maintained above -191 CFZ Steam Tunnel temperature exceeding 250°F (MSCWL) and insufficient high pressure injection (Max Safe Temperature). systems are available to restore level, crew begins to Emergency Depressurize by opening at least seven SRVs before RPV level lowers below -191 CFZ. (Momentary shrink below -191 due to automatic SRV closure does not constitute failure of this critical task.)
EVENT 6 7 Safety If a primary system is discharging into the The MSCWL is the lowest RPV water level at which the Significance secondary containment when this step of the covered portion of the reactor core will generate procedure is reached, one of three conditions must sufficient steam to preclude any clad temperature in the exist: uncovered portion of the core from exceeding 1500°F.
A primary system break cannot be isolated When water level decreases below MSCWL with because system operation is required to assure injection, clad temperatures may exceed 1500°F.
adequate core cooling or to shut down the reactor.
No isolation valves exist upstream of a primary system break, or if isolation valves do exist, they cannot be closed because of some mechanical/ electrical/pneumatic failure.
The source of the discharge cannot be determined.
Since the RPV is the only significant source of heat, other than a fire, which might cause area temperatures to increase to their maximum safe operating values, the action of manually scramming the reactor should terminate increasing secondary containment temperatures.
If temperatures in any one of the areas listed in Table SC-1 of the Secondary Containment Control guideline approach their maximum safe operating value, adequate core cooling, containment integrity, safety of personnel, or continued operability of equipment required to perform EPG actions can no longer be assured.
Cueing Main Steam Tunnel temperature rising on PDS. Wide range indication (SPDS and PDS) falls to -160 Main Steam Tunnel temperature alarms on panel and lowering trend continues, and, before -160 wide P601. range is reached, initial conditions, field reports, and control room indications convey that adequate high MSIV open position indication on panel P601 and pressure injection cannot be restored before level falls panel P858. below -191 CFZ.
Performance Operator places the Reactor Mode Switch to Manipulation of seven of the eight ADS/SRVs on panel Indicator SHUTDOWN on panel P680. P601:
Operator manipulates switches for MSIVs for B21-F041K Steam Line A to CLOSE on panel P601. B21-F047L B21-F041F B21-F047A B21-F051C B21-F041D B21-F051A B21-F051B Performance RPS Group lights de-energized on panel P680. Crew will observe ADS/SRV light indication go from Feedback green to red, reactor pressure lowering on SPDS and Control Rod full -in indication on panel P680.
panel P601 indications.
Reactor power trend on nuclear instrumentation on panel P680.
Green light indication energized and red light indication off for MSIVs for Steam Line A on panel P601 and P858.
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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 8 of 9 Justification If temperatures in any one of the areas listed in The MSCWL (-191 CFZ) is the lowest RPV water level for the Table SC-1 of the Secondary Containment Control at which the covered portion of the reactor core will chosen guideline approach their maximum safe operating generate sufficient steam to preclude any clad performance value, adequate core cooling, containment temperature in the uncovered portion of the core from limit integrity, safety of personnel, or continued exceeding 1500°F. Emergency depressurization is operability of equipment required to perform EPG allowed when level goes below TAF (-160 wide range) actions can no longer be assured. and should be performed, if in the judgment of the CRS, level cannot be maintained above -191 CFZ. Since it is intended for the scenario supporting this CT to, early in the event, clearly indicate no high pressure injection systems can be made available to reverse the lowering level trend, the crew will have time to communicate and open 7 of 8 ADS/SRVs before -191 CFZ.
BWR Owners App. B, step SC/T-4 and SC/T-4.1 App. B, Contingency #1 Step C1-4 Group Appendix Revision 0
Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 9 of 9 Critical Task (CT-3) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically initiate, crew manually aligns ECCS systems for injection prior to RPV pressure lowering below 300 psig.
EVENT 8 Safety Failure to recognize the auto initiation not occurring, and Significance failure to take manual action per Conduct of Ops will result in unavailability of safety-related equipment necessary to provide adequate core cooling, otherwise resulting in core damage and a large offsite release.
Cueing Indication of ECCS systems not initiating with initiation conditions present:
Indication of Drywell pressure 1.39 psig or RPV level -150.3 wide range White light on LPCS/RHR A INIT RESET pushbutton extinguished on panel P601 Green light on and red light extinguished on LPCS and RHR A pump handswitches on panel P601 LPCS SYS OOSVC annunciator on panel P601 Performance Operator manually manipulates switches for Div 1 Indicator ECCS pumps and directs operators to manually open Div 1 ECCS injection valves from Division 1 Remote Shutdown Panel (RHR A) and locally (LPCS).
Performance Red light on and green light extinguished on LPCS Feedback and/or RHR A pump and valve handswitches on panel P601.
Rising level trend on indications on panel P601, PDS and SPDS.
Rising flow rate on LPCS and/or RHR A flow indicators on panel P601, PDS, and SPDS Justification Attempting to align high pressure ECCS systems must for the be performed to determine their availability by the time chosen TAF is reached in order to properly implement EP-2 performance decision steps regarding restoring and maintaining RPV limit level. Attempting to align low pressure ECCS systems can only be done once RPV pressure falls below the injection valve RPV pressure permissive and will only be effective once RPV pressure falls below the shutoff head of the respective ECCS pump. The reduction in RPV pressure will normally be via Emergency Depressurization, which is a separate critical task bounded by a minimum RPV level.
BWR Owners App. B, Contingency 1, step C1-3 Group Appendix Revision 0