ML18061A165

2017-12 Final Outlines

ML18061A165
Person / Time
Site:	Grand Gulf
Issue date:	12/11/2017
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
References
Download: ML18061A165 (72)

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 2

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 2

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 2

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 WATER LEVEL and the following:

X 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 apply to INADVERTENT CONTAINMENT ISOLATION:

X 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 2

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 2

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 2

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 2

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 2

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 2

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 2

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 2

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 2.2.25, Knowledge of the bases in Technical 4.2 78 295006 SCRAM / 1 X Specifications for limiting conditions for operations and safety limits.

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 2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry 4.5 80 295023 Refueling Acc / 8 X conditions.

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 2

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 as they apply to INCOMPLETE SCRAM X

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 2

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 2

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 2.4.8 Knowledge of how abnormal 259001 Reactor Feedwater X operating procedures are used in 4.5 93 conjunction with EOPs.

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 2

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 2

Rev 2 ES-401 Record of Rejected K/As Form ES-401-4 Randomly Tier / Group Selected K/A Reason for Rejection (Original)

(New)

RO T1/G1 Original KA: AA2. Ability to determine and/or interpret the following as 295016 295016 they apply to CONTROL ROOM ABANDONMENT: AA2.01 Reactor AA2.01 AA2.02 power At GGNS we are unable to determine Reactor Power at the Remote Shutdown Panel or external from the main control room. The ability to determine or interpret reactor power for a control room abandonment would infer ATWS which is outside our design bases.

Randomly selected New K/A - 295016, AA2.02, Reactor water level.

Page 1 point totals not affected by this change. (Rev 1)

RO T1/G1 Original KA: 2.1.19 Ability to use plant computers to evaluate system or 295027 295027 component status.

2.1.19 2.1.25 At GGNS the use of plant computers to evaluate Containment Temperatures is limited due to the computer indication is an average of several instruments. Individual instrument indications are on the main Control Room panels. These indications are used more readily by ROs to determine the validity of RPV water level instrumentation by using Caution 1 of EOPs. Caution 1 uses a table and specific containment temperature instruments to determine RPV water level instrumentation validity.

Randomly selected New K/A - 2.1.25, Ability to interpret reference materials, such as graphs, curves, tables, etc.

Page 1 point totals not affected by this change. (Rev 1)

RO T1/G1 Original KA: EK3. Knowledge of the reasons for the following responses 295037 295037 as they apply to SCRAM CONDITION PRESENT AND REACTOR EK3.08 EK3.01 POWER ABOVE APRM DOWNSCALE OR UNKNOWN : EK3.08 ATWS circuitry: Plant-Specific GGNS doesnt have an ATWS Circuitry. Randomly selected New K/A -

EK3.01, Recirculation pump trip/runback.

Page 1 point totals not affected by this change. (Rev 1)

Rev 2 Randomly Tier / Group Selected K/A Reason for Rejection (Original)

(New)

RO T2/G1 215004 215004 Original K/A: K4. Knowledge of SOURCE RANGE MONITOR (SRM)

SYSTEM design feature(s) and/or interlocks which provide for the K4.02 K4.01 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 Page 1 point totals not affected by this change. (Rev 1)

RO T2/G1 223002 223002 Original K/A: K3. Knowledge of the effect that a loss or malfunction of the K3.20 K3.10 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 Page 1 point totals not affected by this change. (Rev 1)

RO T2/G1 Original K/A: A1. Ability to predict and/or monitor changes in parameters 262002 262002 associated with operating the UNINTERRUPTABLE POWER SUPPLY A1.02 A4.01 (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 that has a 2.4 importance factor.

Operations Representative reviewed the other KA and determined that the importance factor is below a 2.5, therefore A4 was randomly selected.

Page 1 point totals were affected - A1 decreased by 1 in T2G1 and A4 increased by 1 in T2G1. (Rev 1)

Rev 2 Randomly Tier / Group Selected K/A Reason for Rejection (Original)

(New)

RO T2/G2 234000 234000 Original K/A: A4. Ability to manually operate and/or monitor in the control room:

A4.02 A4.01 A4.02 Control rod drive system At GGNS the CRD system is considered to be the hydraulic and mechanism portion. The Control Rod blade is part of the Reactor Vessel Internal system, therefore, there is no tie between Fuel Handling Equipment and the CRD system at GGNS.

Randomly selected K/A - A4.01, Neutron Monitoring system.

SRO T1/G1 295006 295006 Original K/A: 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for 2.2.36 2.2.39 operations.

Unable to tie SRO only direction with the parent KA of SCRAM.

Randomly selected K/A - 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems.

SRO T1/G1 295023 295023 Original K/A: 2.4.3 Ability to identify post-accident instrumentation.

2.1.19 2.4.2 Unable to tie SRO only direction with the parent KA of Refueling Accidents. There is no direct connection with post-accident instrumentation and a refuel accident.

Randomly selected K/A - 2.4.2, Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

SRO T1/G1 295006 295006 Original K/A: 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems.

2.2.39 2.2.25 This K/A is RO knowledge, discussed with Chief during review.

Randomly selected K/A - 2.2.25, Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Rev 2 Randomly Tier / Group Selected K/A Reason for Rejection (Original)

(New)

SRO T2/G2 295001 295001 Original K/A: 2.4.9 Knowledge of low power/shutdown implications in accident (e.g. loss of cooling coolant accident or loss of residual heat 2.4.9 2.4.8 removal) mitigation strategies.

This K/A doesnt fit the mitigation strategies for GGNS. GGNS uses the same EOPs and Abnormal procedures regardless of power level.

Randomly selected K/A - 2.4.8, Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Grand Gulf Nuclear Station Date of Examination: 12/04/2017 Examination Level: RO SRO Operating Test Number: LOT 12-2017 Administrative Topic Type Describe activity to be performed Code*

(see Note)

AR1 - Determine Primary Containment Water Level (GJPM-OPS-2017IAR1)

Conduct of Operations R; D K/A 2.1.25: (3.9); 2.1.20: (4.6); 2.4.21: (4.0)

AR2 - Perform AC Lineup Surveillance (GJPM-OPS-2017IAR2)

Conduct of Operations S; M K/A 2.1.31: (4.6); 2.2.12: (3.7); 2.1.20: (4.6)

AR3 - Determine Tagging Requirements (GJPM-OPS-2017IAR3)

Equipment Control R; D K/A 2.2.41: (3.5); 2.2.13: (4.1)

Radiation Control AR4 - Perform Emergency Notifications (GJPM-OPS-2017IAR4)

Emergency Plan R; N K/A 2.4.43: (3.2); 2.4.39: (3.9)

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (4)

(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (2)

(N)ew or (M)odified from bank ( 1) (2)

(P)revious 2 exams ( 1; randomly selected) (0)

ES-301, Page 22 of 33 Rev 1 08/02/2017

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Grand Gulf Nuclear Station Date of Examination: 12/04/2017 Examination Level: RO SRO Operating Test Number: LOT 12-2017 Administrative Topic Type Describe activity to be performed Code*

(see Note)

AS1 - Perform EOOS Risk Assessment (GJPM-OPS-2017IAS1)

Conduct of Operations R; M K/A 2.1.39: (4.3)

AS2 - Review Completed Surveillance (GJPM-OPS-2017IAS2)

Conduct of Operations R; N K/A 2.1.2: (4.4); 2.1.7: (4.7); 2.2.12: (4.1);

2.2.22: (4.7)

AS3 - Determine Impact on Plant Operations for Failed Relay Equipment Control R; D; P (GJPM-OPS-2017IAS3)

K/A 2.2.41: (3.9); 2.2.22: (4.7); 2.2.36: (4.2)

AS4 - Authorize Emergency Exposure (GJPM-OPS-2017IAS4)

Radiation Control R; M K/A 2.3.4: (3.7)

AS5 - Perform Emergency Classification (GJPM-OPS-2017IAS5)

Emergency Plan R; N K/A 2.4.41: (4.6)

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (5)

(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (1)

(N)ew or (M)odified from bank ( 1) (4)

(P)revious 2 exams ( 1; randomly selected) (1)

ES-301, Page 22 of 33 Rev. 1 08/02/2017

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.: LOT 12-2017 Control Room Systems* (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title Type Code*

Function S1 - Manually Startup RCIC (GJPM-OPS-2017IS1)

A-N-S-L 2 217000 A4.04 (3.6)

S2 - Retest MSIV Slow Closure (GJPM-OPS-2017IS2)

P-A-D-S 3 239001 A2.11 (4.1)

S3 - Startup Shutdown Cooling (GJPM-OPS-2017IS3)

A-D-L-S 4 205000 A4.01 (3.7)

S4 - Start, Parallel and Load EDG (GJPM-OPS-2017IS4)

A-D-S 6 264000 A4.04 (3.7)

S5 - Startup H2 Recombiner (GJPM-OPS-2017IS5)

EN-D-S 5 223001 A4.13 (3.4)

S6 - Secure Standby Gas Treatment (GJPM-OPS-2017IS6)

EN-N-S 9 261000 A4.02 (3.1)

C1 - Bypass Control Rod in RACS (GJPM-OPS-2017ICR1)

D-C-L 7 201005 A2.04 (3.2)

S7 - Shift RR Pump B to Fast Speed (GJPM-OPS-2017IS7)

A-D-S 1 202001 A4.01 (3.7) (RO ONLY)

In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1 - Align SP Cooling from RSP (GJPM-OPS-2017IP1)

A-E-D-L 5 219000 A4.01 (3.8)

P2 - Install N2 Bottles on ADS Air Supply (GJPM-OPS-2017IP2)

P-E-D-R-L 3 218000 A2.03 (3.4)

PB2 - Return Fire Water Pumps to Stby (GJPM-OPS-2017IPB2)

D 8 286000 4.05 (3.3)

• 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 4-6 / 4-6 / 2-3 (6)

(A)lternate path (C)ontrol room C ----- (1)

(D)irect from bank D <9 / <8 / <4 (9)

(E)mergency or abnormal in-plant E >1 / >1 / >1 (2)

(EN)gineered safety feature EN >1 / >1 / > 1 (control room sys) (2)

(L)ow-Power / Shutdown L >1 / >1 / >1 (5)

(N)ew or (M)odified from bank including 1(A) N-M >2 / >2 / >1 (2)

(P)revious 2 exams P <3 / <3 / < 2 (randomly selected) (2)

(R)CA (S)imulator R >1 / >1 / >1 (1)

S (7)

Rev 4 11/29/2017

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.: LOT 12-2017 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 S1 - Manually Startup RCIC (GJPM-OPS-2017IS1)

A-N-S-L 2 217000 A4.04 (3.6)

S2 - Retest MSIV Slow Closure (GJPM-OPS-2017IS2)

P-A-D-S 3 239001 A2.11 (4.3)

S3 - Startup Shutdown Cooling (GJPM-OPS-2017IS3)

A-D-L-S 4 205000 A4.01 (3.7)

S4 - Start, Parallel and Load EDG (GJPM-OPS-2017IS4)

A-D-S 6 264000 A4.04 (3.7)

S5 - Startup H2 Recombiner (GJPM-OPS-2017IS5)

EN-D-S 5 223001 A4.13 (3.4)

S6 - Secure Standby Gas Treatment (GJPM-OPS-2017IS6)

EN-N-S 9 261000 A4.02 (3.1)

CR1 - Bypass Control Rod in RACS (GJPM-OPS-2017ICR1)

D-C-L 7 201005 A2.04 (3.2)

In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1 - Align SP Cooling from RSP (GJPM-OPS-2017IP1)

A-E-D-L 5 219000 A4.01 (3.7)

P2 - Install N2 Bottles on ADS Air Supply (GJPM-OPS-2017IP2)

E-D-R-L 3 218000 A2.03 (3.6)

PB2 - Return Fire Water Pumps to Stby (GJPM-OPS-2017IPB2)

D 8 286000 4.05 (3.3)

• 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 4-6 / 4-6 / 2-3 (5)

(A)lternate path (C)ontrol room C ----- (1)

(D)irect from bank D <9 / <8 / <4 (8)

(E)mergency or abnormal in-plant E >1 / >1 / >1 (2)

(EN)gineered safety feature EN >1 / >1 / > 1 (control room sys) (2)

(L)ow-Power / Shutdown L >1 / >1 / >1 (5)

(N)ew or (M)odified from bank including 1(A) N-M >2 / >2 / >1 (2)

(P)revious 2 exams P <3 / <3 / < 2 (randomly selected) (1)

(R)CA (S)imulator R >1 / >1 / >1 (1)

S (6)

Rev 4 11/29/2017

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 1 of 47 Facility: Grand Gulf Nuclear Station Scenario No.: 2 Op-Test No.: GGNS 12-2017 Examiners: ____________________________ Operators: _____________________________

Event Malf. No. Event Type Event No. Description 1 N/A N (BOP,CRS) Transfer ESF Bus 17AC from ESF Transformer 21 to ESF Transformer 12 C (ATC,CRS) 2 (r) fw211 Reactor Narrow Range Level C instrument oscillations TS (CRS) r21134h TS (CRS) ESF Transformer 12 Lockout with HPCS Diesel Generator 3 C (BOP,CRS) n41141c auto start failures A (BOP,CRS) 4 z022021_24_53 C(ATC,CRS) Control Rod 24-53FN drifting in A(CREW)

R (ATC)

C(BOP,CRS) 5 z022022_24_53 Control Rod 24-53FN stuck at position 32.

A(CREW)

TS (CRS)

Control Rod 32-27HJ drifting in. Reactor scram due to two 6 z021021_32_37 M(CREW) controls drifting in Hydraulic Block ATWS > 5% RTP with SLC failure (EP-2, 2A)

• (CT-1) When control rods fail to scram, crew inserts all control rods to position 02 or beyond before exiting EP-2A c11164 M(CREW) * (CT-2) Inhibit ADS prior to automatic ADS valves 7 opening during ATWS

• (CT-3) 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 prior to exiting EP-2A Reactor Feedwater Pump trip.

8 fw123a(b) C(BOP,CRS) * (CT-4) Restores injection using Condensate/Feedwater to restore/maintain RPV level above -191 CFZ before exiting EP-2A 9 c41263 C(ATC,CRS) ESF Bus 15AA power loss (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.

Reason for Revision: Editorial changes to enhance Technical Specification actions.

Revision 4

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 2 of 47 Objectives: To evaluate the applicants ability to operate the facility in response to the following evolutions:

1. Transfer ESF Bus 17AC from ESF Transformer 21 to ESF Transformer 12.

2. Respond to Narrow Range C level transmitter oscillations.

3. Respond to an ESF Transformer 12 Lockout with failure of HPCS Diesel Generator to automatically start.

4. Respond to a control rod drifting in.

5. Respond to a stuck control rod.

6. Respond to a second control rod drifting in, resulting in a manual reactor scram.

7. Respond to a Hydraulic Block ATWS with power > 5% RTP.

8. Respond to a Reactor Feed Pump trip.

9. Respond to an ESF Bus 15AA power loss.

Initial Conditions: Plant is operating at 100% power.

Inoperable Equipment: None 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 FDR FM ESF 21, for preventative maintenance Scenario Notes:

This scenario is a NEW Scenario.

Validation Time: 80 minutes Revision 4

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 3 of 47 Quantitative Attributes Table Attribute E3-304-1 Actual Description Target Malfunctions after

• Reactor Feed Pump trip (E8) 1-2 2 EOP entry

• ESF Bus 15AA power loss (E9)

• ESF Transformer 12 lockout with a failure of HPCS Diesel Generator to automatically start (Loss of AC Power ONEP) (E3)

Abnormal Events 2-4

• Control Rod 24-53FN drifting in (Control Rod/Drive Malfunctions 3 ONEP) (E4)

• Control Rod 24-53FN stuck at position 32 (Control Rod/Drive Malfunctions ONEP) (E5)

• Control Rod 32-37HJ drifting in - Reactor Scram (E6)

Major Transients 1-2 2

• Hydraulic block ATWS with power > 5% RTP- SLC failure (E7)

EOP entries requiring

• EP-2 1-2 2 substantive action

• EP-3 EOP contingencies requiring substantive 0-2 1

• EP-2A action

• (CT-1) When control rods fail to scram, crew inserts all control rods to position 02 or beyond before exiting EP-2A

• (CT-2) Inhibit ADS prior to automatic ADS valve opening during ATWS EOP based Critical * (CT-3) During failure to scram conditions with power > 5%

2-3 4 RTP, terminate and prevent all injection from all sources Tasks (except boron, CRD, and RCIC) as necessary to lower RPV level to below -70 wide range prior to exiting EP-2A

• (CT-4) 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 (E1)

Reactivity

• Lower core flow to 70 mlbm/hr using Reactor Recirc Flow N/A 1 Manipulations Control Valves (E5)

• Narrow Range C level transmitter oscillations (E2)

• ESF Transformer 12 lockout with a failure of HPCS Diesel Instrument / Generator to automatically start (E3)

N/A 6

• Control Rod 24-53FN drifting in (E4)

Component failures

• Control Rod 24-53FN stuck (E5)

• Reactor Feed Water Pump A(B) trip (E8)

• ESF Bus 15AA power loss (E9)

• Narrow Range C level transmitter oscillations (E2)

• ESF Transformer 12 lockout with a failure of HPCS Diesel Generator to automatically start (E3)

• Control Rod 24-53FN drifting in (E4)

Total Malfunctions N/A 8

• Control Rod 24-53FN stuck (E5)

• Control Rod 32-37HJ drifting in - Reactor Scram(E6)

• Hydraulic block ATWS with power > 5% RTP-SLC Failure (E7)

• Reactor Feed Water Pump A(B) trip (E8)

• ESF Bus 15AA power loss (E9)

Revision 4

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 4 of 47 Top 10 systems and operator actions important to risk that are tested:

RPS (Event 6)

ESF Power (Event 3)

Condensate (Event 8)

Failure to align alternate power to 4.16 KV or 6.9 KV buses (Event 3)

Revision 4

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 5 of 47 SCENARIO ACTIVITIES:

The plant is operating at 100% power.

Event 1 - Transfer ESF Bus 17AC from ESF Transformer 21 to ESF Transformer 12 After crew assumes the shift, 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 - Narrow Range C Level instrument oscillations (Triggered by Lead Examiner)

When ESF Bus 17AC has been transferred to ESF Transformer 12, Narrow Range C Level transmitter will begin oscillating. Crew will respond using ARI 04-1-02-1H13-P680-4A2-A2, RX WTR LVL SIG FAIL HI/LO, and manually select Narrow Range Level A or B. CRS will enter LCO TRM 6.3.7.

Event 3 - ESF Transformer 12 Lockout with HPCS Diesel Generator auto start failure (Triggered by Lead Examiner)

After Narrow Range Level A or B channel is selected and Tech Specs addressed, ESF Transformer 12 will lockout due to sudden pressure, causing a loss of power to ESF Bus 17AC. HPCS Diesel Generator will fail to auto start. BOP will recognize the failure of HPCS Diesel Generator to auto start and restore ESF Bus 17AC power from ESF Transformer 21 per 05-1-02-I-4, Loss of AC Power ONEP. CRS will enter TS 3.8.1.B for HPCS Diesel Generator inoperable.

Event 4 - Control Rod 24-53 drifting in (Triggered by Lead Examiner)

After ESF Bus 17AC power has been restored and Tech Specs addressed, Control Rod 24-53FN will begin drifting in. ATC will select Control Rod 24-53FN and apply a continuous insert signal per 05-1-02-IV-1, Control Rod/Drive Malfunctions ONEP.

Event 5 - Control Rod 24-53 stuck at position 32 (automatically triggered)

When Control Rod 24-53FN reaches position 32, it will become stuck. ATC will recognize and report Control Rod 24-53FN has stopped inserting. CRS will direct ATC to lower core flow to 70 mlbm/hr IAW Control Rod/Drive Malfunctions ONEP. ATC will lower core flow to 70 mlbm/hr using Recirc Flow Control Valves in fast detent. CRS will enter and direct actions from Reduction in Recirculation System Flow Rate ONEP, 05-1-02-III-

3. After actions of Reduction in Recirculation System Flow Rate ONEP have been completed, CRS will direct actions IAW Control Rod/Drive Malfunctions ONEP for BOP to raise CRD drive water pressure in 25 psid increments and for ATC to attempt to insert Control Rod 24-53FN after each drive water pressure adjustment.

When CRD Drive water pressure is raised to greater than 325 psid, 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 - Control Rod 32-27 drifting in (automatically triggered)

When CRD Drive Water pressure is raised above 325 psid, Control Rod 32-37HJ will begin to drift in. ATC will insert a manual scram per Control Rod/Drive Malfunctions ONEP. CRS will enter Reactor Scram ONEP, 05 02-I-1, and Turbine Generator Trip ONEP, 05-1-02-I-2.

NOTE: Event 6 can be triggered before CRD Drive Water pressure is raised to greater than 325 psid at the direction of the Lead Examiner.

Revision 4

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 6 of 47 Event 7 - Hydraulic Block ATWS > 5% RTP (No trigger required)

When reactor is scrammed, an ATWS occurs due to a hydraulic block of both scram discharge volumes. ATC will verify Reactor Recirc Pumps transfer to LFMGs, initiate ARI/RPT, inhibit ADS to prevent automatic operation (CT-2) and initiate and override HPCS IAW Reactor Scram ONEP, 05-1-02-I-1, immediate operator actions. CRS will enter EP-2A via EP-2. Reactor power will be above 5% RTP. ATC will initiate SLC which will fail to inject and initiate and override low pressure ECCS IAW Reactor Scram ONEP, 05-1-02-I-1, immediate operator actions. Terminate and Prevent of Feedwater 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-3). Crew will install the necessary attachments to bypass RPS and RC&IS interlocks and insert controls rods via manual scrams and RC&IS (CT-1). Suppression Pool Cooling will be maximized using RHR A and RHR B. Bypass valves will control reactor pressure during this event. Feedwater is available for RPV level control.

Event 8 - Reactor Feedwater Pump trip (Triggered by the Lead Examiner after reactor water level is stabilized below -70 inches wide range)

When reactor level lowers below -70 inches wide range, the in-service Reactor Feed Pump will trip. BOP will restore Feedwater injection to the RPV by starting the standby Reactor Feed Pump (CT-4) IAW 04-1-01-E12-1, Attachment 6, per 02-S-01-43, Transient Mitigation Strategy. An alternate success path would be CRS directing ATC to lower reactor pressure to 450 to 600 psig to allow RPV injection with Condensate Booster Pumps (CT-4) IAW 02-S-01-43, Transient Mitigation Strategy.

Event 9 - ESF Bus 15AA power loss (Triggered by Lead Examiner before controls rods are inserted)

After the running Reactor Feed Pump has tripped and RPV level has been stabilized, breaker 152-1514, ESF BUS 15AA FDR FM XFMR 11, will trip. Division 1 Diesel Generator will automatically restore power to ESF Bus 15AA. The ATC will recognize the loss of override function for LPCS and RHR A and override the associated pumps and injection valves IAW EP-2A and 02-S-01-43, Transient Mitigation Strategy. CRS will direct the ATC to restore Instrument Air to Containment IAW 05-1-02-I-4, Loss of AC Power ONEP. ATC will restore Instrument Air to Containment by opening P53-F001.

NOTE: While crew is responding to Event 9, report Attachments needed for scramming and driving control rods (18, 19, and 20) are complete to allow crew to prioritize actions.

After crew has begun inserting control rods, at direction of the Lead Examiner, the control rods are allowed to be fully inserted with the next scram. CRS transitions from EP-2A to EP-2 and RPV level restoration is directed.

NOTE: Examiner watching ATC should assist the Lead Examiner in determining when to allow all control rods to be inserted.

The exercise ends when controls rods are inserted, EP-2A has been exited and RPV water level band has been changed to 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 7 of 47 Critical Task (CT-1) When control rods fail to scram, crew inserts all (CT-2) Inhibit ADS prior to automatic ADS valve control rods to position 02 or beyond before exiting opening during ATWS EP-2A EVENT 7 7 Safety Failure to effect shutdown of the reactor when a RPS Steps in EP-2A may intentionally lower RPV water level Significance setpoint has been exceeded would unnecessarily extend below the ADS setpoint to reduce reactor power.

the level of degradation of the safety of the plant. This Permitting automatic ADS initiation may be undesirable could further degrade into damage to the principle fission for the following reasons:

product barriers if left unmitigated. The crew is authorized

• ADS actuation can impose a severe thermal transient by Conduct of Operations to take mitigating actions when on the RPV and may complicate efforts to control automatic safety systems fail to perform their intended RPV water level function. Action to shut down the reactor is required when RPS and control rod drive systems fail IAW EP-2A.

• If only RCIC is 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 may not exist when the specified actions are being carried out

• The operating crew can draw on much more information than available to ADS logic and can better judge, based on instructions contained in procedure, when and how to depressurize the RPV

• Subsequent steps provide explicit and detailed instructions for RPV water level control and identify the specific conditions when RPV blowdown is required

• Rapid, uncontrolled injection of relatively cold, unborated water could occur as RPV pressure decreases. If reactor is not shutdown or if shutdown margin is small, this could add sufficient positive reactivity to cause power excursion large enough to damage the core Automatic initiation of ADS is therefore inhibited upon entry of EP-2A.

Cueing Manual scram is initiated and numerous control rods ADS Timer initiated alarm on P601 indicate beyond position 02.

Reactor power indicating > 5% RTP on APRMs on panel P680.

APRM downscale lights on panel P680 extinguished.

Performance Operator selects control rod gangs by depressing the Manipulation of ADS A and ADS B MANUAL INHIBIT Indicator respective pushbuttons on panel P680 and inserts the switches on panel P601 vertical section.

rods by depressing the IN-TIMER SKIP pushbutton.

Operator resets reactor scram signal with key-locked switches on panel P680 and inserts manual reactor scram using scram pushbuttons on panel P680.

Performance Operator selecting and inserting control rods indicated by Inhibit switches click into INHIBIT position on panel P601 Feedback rod position decreasing to 00 for selected rods on panel vertical section.

P680. White indicating light on ADS A and ADS B MANUAL Control rod movement on subsequent reactor scrams. INHIBIT switches illuminate.

Reactor power lowering. Receipt of ADS/SRV A and ADS/SRV B OOSVC alarms on panel P601/18A-H2 and P601/19A-H2.

Justification There is no time limit for effecting complete reactor The 105 second ADS timer allows sufficient time for the for the shutdown via control rod insertion. For the timeframe of crew to recognize and override automatic operation of the chosen this scenario, containment limits are not challenged and system. As long as ADS is inhibited before ADS valves performance power oscillations are not experienced. However, if the open, reactor pressure will not be reduced to the shutoff limit failure to scram EP were to be exited, other procedures heads of high volume, cold water systems.

would not provide the guidance necessary to achieve reactor shutdown. Before exiting EP-2A ensures guidance to effect reactor shutdown is not removed.

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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 8 of 47 BWR Owners App. B, step RC/Q6, RC/Q-7 App. B, step RC/Q-6 Group Appendix Licensed 02-S-01-40, EP Technical Bases, Attachment V, Step Q- 02-S-01-40, EP Technical Bases, Attachment V, Step 1 Bases 1 UFSAR Chapter 15.8 Documents UFSAR Chapter 15.8

• If an operator or the crew significantly deviates from, or fails to, follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review (NUREG 1021, Appendix D)

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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 2 Page 9 of 47 Critical Task (CT-3) During failure to scram conditions with power (CT-4) Restores injection using

> 5%, terminate and prevent all injection from all Condensate/Feedwater to restore/maintain RPV level sources (except boron, CRD, and RCIC) as necessary above -191 CFZ before exiting EP-2A to lower RPV level to below -70 wide range prior to exiting EP-2A EVENT 7 8 Safety Regarding lowering level below -70 wide range, to If RPV water level cannot be restored and maintained Significance prevent or mitigate the consequences of any large above the Minimum Steam Cooling RPV Water Level (-

irregular neutron flux oscillations induced by 191 CFZ), emergency RPV depressurization is neutronic/thermal-hydraulic instabilities. RPV water level performed to maximize injection flow. Emergency is lowered sufficiently below the elevation of the feedwater depressurization is undesirable under ATWS conditions sparger nozzles. This places the feedwater spargers in since the core response is difficult to predict and the risk the steam space providing effective heating of the of power excursions is increased.

relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and 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 rods Reactor Feed Pump trip annunciators and Feedwater indicate beyond position 02 and reactor power is > 5% on flow and RPV level lowering on indicators on panel P680 panel P680 indications and SPDS and RPV level is > -70 and PDS and SPDS.

wide range on SPDS and PDS.

Performance Operator initiates HPCS using HPCS manual initiation Operator manipulates switches on panel P680 panel to Indicator switch, then secures HPCS pump and overrides the start the standby Reactor Feed Pump HPCS injection valve closed. Alternately, operator lowers RPV pressure using Bypass Operator initiates Div 1 and Div 2 ECCS with their Valves or SRVs to allow injection with Condensate respective manual initiation switches and overrides the Booster Pumps.

associated injection valves closed and secures LPCS and RHR C pumps. RHR A and B pumps may be left running for Suppression Pool Cooling.

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 Feedwater flow indication on panel P680 and SPDS Feedwater flow and RPV level rising on panel P680 and Feedback indicate zero. PDS and SPDS.

Valves N21-F009A and B and N21-F040 green lights illuminated.

Master Level Controller output indicates 0% on panel P680.

High Pressure Core Spray, Low Pressure Core Spray and RHR C pump and injection valve override annunciators illuminated on panel P601. RHR A and RHR B 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 10 of 47 Justification Applicability for this CT is during EP-2A conditions where The Minimum Steam Cooling RPV Water Level (-191 for the it is necessary to lower level to control power with no high CFZ) is the lowest RPV water level at which the covered chosen energy input into the primary containment. There is no portion of the reactor core will generate sufficient steam performance time limit for this lowering level, but it establishes margin to preclude any clad temperature in the uncovered limit to conditions where fuel damaging power oscillations may portion of the core from exceeding 1500°F. Maintaining theoretically occur. Before exiting EP-2A was chosen RPV water level above the Minimum Steam Cooling RPV because other procedures would not provide the guidance Water Level thus ensures that the core remains necessary to establish margin for power oscillation adequately cooled.

mitigation. Before exiting 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, Contingency #5 Step C5-4 App. B, Contingency #5 Step C5-4 Group Appendix Licensed 02-S-01-40, EP Technical Bases, Attachment V, Step L-7 02-S-01-40, EP Technical Bases, Attachment V, Step L-9 Bases UFSAR Chapter 15.8 UFSAR Chapter 15.8 Documents

• If an operator or the crew significantly deviates from, or fails to, follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review (NUREG 1021, Appendix D)

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 2 Page 11 of 47 Simulator Setup:

A. Initialization

1. Log off all simulator PDS and SPDS computers (PDS and SPDS must come up after the simulator load for proper operation).

2. Startup the simulator using Simulator Instructors Job Aid section 7.3.

Note:

Prior to running the Schedule File, ensure no Event Files are Open. If an existing Event File is Open prior to running the Schedule File, then any associated Event Files will not automatically load.

3. Open Schedule.exe and Director.exe by clicking on the Icon in the Thunder Bar.

4. Set the Simulator to IC-101 and perform switch check (Using Quick Reset in Director).

5. Click on Open in the Schedule window and Open Schedule File 12-2017 NRC Exam Scenario 2.sch (in the Schedule Directory)

6. In Schedule window, click on the Stopped red block. The red block will change to a green arrow and indicate the scenario is active (Running).

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 2 Page 12 of 47

7. Click the Summary tab in the Director window. Verify the schedule files are loaded and opened per Section B below. (Note: Any actions in the schedule file without a specific time will not load into the director until triggered.)

8. Take the simulator out of freeze.

9. Log on to all simulator PDS and SPDS computers.

10. Verify or perform the following:

• IC-101

• Ensure all procedures are marked as indicated for turnover conditions

• Advance all chart recorders and ensure all pens are inking properly

• Clear any graphs and trends off PDS and SPDS

11. Run through any alarms and ensure alarms are on. (Note: On T-Rex, to verify alarms are ON, the indicator will indicate Alarms On).

12. Place the simulator in Freeze.

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 2 Page 13 of 47 File loaded verification:

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 2 Page 14 of 47 Revision 4

Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 2 Page 15 of 47 Procedures that may be used in this scenario:

Procedure No. Rev Procedure Title 04-1-01-C11-1 154 Control Rod Drive Hydraulic System 04-1-01-C41-1 123 Standby Liquid Control System 04-1-01-E12-1 147 Residual Heat Removal System 04-1-01-N21-1 74 Feedwater System 04-1-01-N32-2 33 Turbine Generator Control 04-1-01-R21-17 10 ESF Bus 17AC 04-1-02-1H13-P601 161 Alarm Response Instruction Panel No.: 1H13-P601 04-1-02-1H13-P680 233 Alarm Response Instruction Panel No.:1H13-P680 04-1-02-1H13-P864 31 Alarm Response Instruction Panel No.: 1H13-P864 04-1-02-1H13-P870 154 Alarm Response Instruction Panel No.: 1H13-P870 04-S-02-SH13-P807 32 Alarm Response Instruction Panel No.: SH13-P807 05-1-02-I-1 130 Reactor Scram 05-1-02-I-2 37 Turbine and Generator Trips 05-1-02-I-4 51 Loss of AC Power 05-1-02-III-3 115 Reduction In Recirculation System Flow Rate 05-1-02-III-5 49 Automatic Isolations 05-1-02-IV-1 117 Control Rod / Drive Malfunctions 05-S-01-EP-1 36 Emergency / Severe Accident Procedure Support Documents 05-S-01-EP-2 45 RPV Control Tech Spec 3.1.3 Tech Spec 3.8.1 Tech Spec TR6.3.7 Revision 4

Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 2 Page 16 of 47 Expected Alarms:

P601-16A-E1, 4.16 KV BUS 17AC INCM FDR 152-1705 P680-2A-C9, DFCS TROUBLE P680-4A2-A2, RX WTR LVL SIG FAIL HI/LO P680-3A-A3, RX LVL 40/32 HI/LO P807-4A-B3, ESF XFMR 12 LOCKOUT TRIP P807-4A-F4, ESF XFMR 12 TROUBLE P807-1A-B1, SWYD XFMR T3 INCM FDR 152-1905 TRIP P807-1A-B2, ESF DIST BUSES INCM FDR 152-1903 TRIP P807-1A-B3, ESF DIST BUSES INCM FDR 152-1904 TRIP P601-16A-A1, HPCS GEN TRIP/LOCKOUT P601-16A-A2, HPCS DSL ENG TRIP P601-16A-D3, HPCS DSL ENG TROUBLE P601-16A-C1, 4.16KV BUS 17AC INCM FDR 152-1704 TRIP P601-16A-F2, HPCS SYS UNDERVOLT P601-16A-G1, 480V MCC 17B01 UNDERVOLT P601-16A-H1, HPCS SYS NOT READY FOR AUTO START P601-16A-E4, HPCS JKY PMP DISCH PRESS LO P870-9A-3B, SSW DIV 3 OOSVC P870-9A-F1, DG 13 TRIP UNIT TROUBLE P870-9A-F2, SSW LOOP C LEAK HI P870-9A-G1, DG 13 FUEL OIL XFER PMP CONT PWR FAIL P680-4A2-E4 CONT ROD DRIFT Revision 4

INITIAL CONDITIONS A. Plant Status: 100% power, middle of cycle B. Tech. Spec. Limitations in effect: None C. Significant problems/abnormalities: None D. Integrated Risk: Green E. Division Work Week: Division 3 F. Evolutions/maintenance for the up-coming shift :

• Transfer ESF Bus 17AC from ESF Transformer 21 to ESF Transformer 12 in preparation for red-tagging breaker 152-1705, 17AC FDR FM ESF 21, for preventative maintenance.

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 1 of 40 Facility: Grand Gulf Nuclear Station Scenario No.: 3 Op-Test No.: GGNS 12-2017 Examiners: ____________________________ Operators: _____________________________

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) Division 2 Diesel Generator lube oil leak C (BOP,CRS)

I (ATC,BOP,CRS) ltb21n091b 3 A(CREW) Division 2 ECCS initiation on spurious RPV low level signal ltb21n091f TS (CRS) 4 fw163c R (ATC,CRS) Loss of condenser vacuum A(CREW) r21135 5 M(CREW) LOP/LOCA (EP-2, 3) rr063b HPCS Pump trip

• (CT-1) Inhibit ADS prior to automatic ADS valve opening during a LOCA

• (CT-2) When RPV level lowers to -160 wide range and e22052 C(ATC,BOP,CRS) cannot be maintained above -191 CFZ (MSCWL) and 6 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)

Failure of Division 1 ECCS to automatically initiate

• (CT-3) When operating injection systems cannot rr040a maintain RPV level and ECCS systems fail to 7 C(ATC,CRS) rr041a 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.

Reason for Revision: Editorial changes to enhance Technical Specification actions.

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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 2 of 40 Objectives: To evaluate the applicants 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.

3. Respond to a Division 2 ECCS initiation on spurious RPV low level signal.

4. Respond to a loss of condenser vacuum.

5. Respond to a loss of Offsite Power / LOCA

6. Respond to a HPCS Pump trip.

7. 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.

• CRD Pump B is tagged out of service for oil replacement in the speed increaser.

Turnover:

Planned activities for this shift are:

• Transfer RPS Bus B from normal to alternate power supply IAW SOI 04-1-01-C71-1, Section 5.1, in preparation for preventative maintenance on the RPS B Motor Generator.

• The Motor Generator will be tagged out on the next shift.

• No scram or isolation surveillances are in progress or planned for this shift.

Scenario Notes:

This scenario is a NEW Scenario.

Validation Time: 75 minutes Revision 3

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 3 of 40 Quantitative Attributes Table Attribute E3-304-1 Actual Description Target Malfunctions after

• HPCS Pump trip (E6) 1-2 2 EOP entry

• Failure of Division 1 ECCS to automatically initiate (E7)

• Spurious Division 2 ECCS initiation (Loss of One or Both RPS Abnormal Events 2-4 Buses ONEP and Automatic Isolations ONEP) (E3) 2

• Loss of condenser vacuum (Loss of Condenser Vacuum ONEP)

(E4)

• 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 opening at least seven SRVs before RPV level lowers below Tasks

-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 (E1)

Reactivity

• Lower core flow to 70 mlbm using Reactor Recirc Flow Control N/A 1 Manipulations Valves (E4)

• Division 2 Diesel Generator lube oil leak (E2)

Instrument /

• Spurious Division 2 ECCS initiation (E3)

N/A 5

• Loss of vacuum (E4)

Component failures

• HPCS Pump trip (E6)

• Failure of Division 1 ECCS to automatically initiate (E7)

• Division 2 Diesel Generator lube oil leak (E2)

• Spurious Division 2 ECCS initiation (E3)

Total Malfunctions N/A

• Loss of vacuum (E4) 6

• LOP/LOCA (E5)

• HPCS Pump trip (E6)

• Failure of Division 1 ECCS to automatically initiate (E7)

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 40 SCENARIO ACTIVITIES:

The plant is operating at 100% power. TBCW Pump C is tagged out of service. CRD Pump B is tagged out of service.

Event 1 - Transfer RPS Bus B from normal to alternate power supply After the crew assumes the shift, the BOP will transfer RPS Bus B from normal to alternate power supply per 04-1-01-C71-1, Reactor Protection System SOI, Section 5.1. The ATC will reset the half-scram.

Event 2 - Division 2 Diesel Generator lube oil leak (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 2 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 IAW SOI 04-1-01-P75-1, Standby Diesel Generator System, Attachment VI. The CRS will enter LCO 3.8.3.E and LCO 3.8.1.B.

Event 3 - Division 2 ECCS initiation on spurious RPV low level signal (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 IX.

CRS will enter 05-1-02-I-4, Loss of AC Power. 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 ONEP. 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, 3.3.5.1.F, 3.3.6.1.A, B, and F, 3.3.6.3.B and 3.3.6.4.B.

Event 4 - Loss of condenser vacuum (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 5/6 - LOP/LOCA/HPCS Pump trip (Automatically triggered)

When the reactor is scrammed, a total loss of offsite power occurs, followed by a small recirculation pipe break after 5 minutes. HPCS pump will trip when it is initiated (Event 6). 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. ATC 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 7 - Failure of Division 1 ECCS to automatically initiate (Automatically triggered)

Division 1 ECCS will fail to automatically initiate on either high drywell pressure or low RPV level. ATC will manually initiate Division 1 ECCS using the lock-collared pushbutton (CT-3) IAW EN-OP-200, Plant Transient Response Rules.

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 40 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 6 6 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 Step L-5 of EP-2, RPV CONTROL, Alternate Level Wide range indication (SPDS and PDS) falls to -160 Control Contingency 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 40 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 License 02-S-01-40, EP Technical Bases, Attachment IV, 02-S-01-40, EP Technical Bases, Attachment IV, Step Bases Step L-5 L through L-13 Documents UFSAR Chapter 15A.6.3.1 UFSAR Chapter 15A.6.3.1

• If an operator or the crew significantly deviates from, or fails to, follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review (NUREG 1021, Appendix D)

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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 3 Page 7 of 40 Critical Task (CT-3) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically initiate, crew manually initiates or aligns ECCS systems for injection prior to RPV pressure lowering below 300 psig EVENT 7 Safety Failure to recognize the auto initiation not occurring, Significance 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 License 02-S-01-40, EP Technical Bases, Attachment IV, Bases Step L-14 Documents UFSAR Chapter 15A.6.3.1

• If an operator or the crew significantly deviates from, or fails to, follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review (NUREG 1021, Appendix D)

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 3 Page 8 of 40 Simulator Setup:

A. Initialization

1. Log off all simulator PDS and SPDS computers (PDS and SPDS must come up after the simulator load for proper operation).

2. Startup the simulator using Simulator Instructors Job Aid section 7.3.

Note:

Prior to running the Schedule File, ensure no Event Files are Open. If an existing Event File is Open prior to running the Schedule File, then any associated Event Files will not automatically load.

3. Open Schedule.exe and Director.exe by clicking on the Icon in the Thunder Bar.

4. Set the Simulator to IC-102 and perform switch check (Using Quick Reset in Director).

5. Click on Open in the Schedule window and Open Schedule File 12-2017 NRC Exam Scenario 3.sch (in the Schedule Directory)

6. In Schedule window, click on the Stopped red block. The red block will change to a green arrow and indicate the scenario is active (Running).

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 3 Page 9 of 40

7. Click the Summary tab in the Director window. Verify the schedule files are loaded and opened per Section B below. (Note: Any actions in the schedule file without a specific time will not load into the director until triggered.)

8. Take the simulator out of freeze.

9. Log on to all simulator PDS and SPDS computers.

10. Verify or perform the following:

• IC-102

• Place red tag on TBCW C pump handswitch

• Place red tag on CRD PMP B and CRD PUMP B AUX OIL PUMP handswitches

• Ensure all procedures are marked as indicated for turnover conditions

• Advance all chart recorders and ensure all pens are inking properly

• Clear any graphs and trends off PDS and SPDC

11. Run through any alarms and ensure alarms are on. (Note: On T-Rex, to verify alarms are ON, the indicator will indicate Alarms On).

12. Place the simulator in Freeze.

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 3 Page 10 of 40 File loaded verification:

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 3 Page 11 of 40 Revision 3

Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 3 Page 12 of 40 Procedures that may be used in this scenario:

Procedure No. Rev Procedure Title 04-1-01-C11-1 154 Control Rod Drive Hydraulic System 04-1-01-C41-1 123 Standby Liquid Control System 04-1-01-C71-1 35 Reactor Protection System 04-1-01-E12-1 147 Residual Heat Removal System 04-1-01-E30-1 25 Suppression Pool Makeup System 04-1-01-E51-1 136 Reactor Core Isolation Cooling System 04-1-01-E61-1 41 Combustible Gas Control System 04-1-01-P75-1 106 Standby Diesel Generator System 04-1-02-1H13-P601 161 Alarm Response Instruction Panel No.: 1H13-P601 04-1-02-1H13-P680 233 Alarm Response Instruction Panel No.:1H13-P680 04-1-02-1H13-P864 31 Alarm Response Instruction Panel No.: 1H13-P864 04-1-02-1H13-P870 154 Alarm Response Instruction Panel No.: 1H13-P870 04-S-02-SH13-P807 32 Alarm Response Instruction Panel No.: SH13-P807 04-1-02-1H22-P401 118 Alarm Response Instruction Panel No.: 1H22-P401 05-1-02-I-1 130 Reactor Scram 05-1-02-I-2 37 Turbine and Generator Trips 05-1-02-I-4 51 Loss of AC Power 05-1-02-III-2 26 Loss of One or Both RPS Buses 05-1-02-III-3 115 Reduction In Recirculation System Flow Rate 05-1-02-III-5 49 Automatic Isolations 05-1-02-V-1 24 Loss of Component Cooling Water 05-1-02-V-8 24 Loss of Condenser Vacuum 05-S-01-EP-1 36 Emergency / Severe Accident Procedure Support Documents 05-S-01-EP-2 45 RPV Control 05-S-01-EP-3 29 Containment Control Tech Spec 3.3.5.1 Tech Spec 3.3.6.1 Tech Spec 3.3.6.3 Tech Spec 3.3.6.4 Tech Spec 3.8.1 Tech Spec 3.8.3 Revision 3

Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 3 Page 13 of 40 Expected Alarms:

P680-7A-A2, RX SCRAM TRIP P864-2A-D2, DIV 2 DSL GEN TROUBLE P864-2A-B1, DIV 2 DSL GEN TRIP P864-2A-D1, DG 12 AUTO START NOT AVAIL P601-17A-D2 RHR PMP B AUTO START P601-17A-H3 RHR C SYS OOSVC P601-17A-B3 RHR B-RHR C ACTUATED P807-3A-H4 STATIC INVRTR 1Y97 TROUBLE P807-3A-H3 STATIC INVRTR 1Y82 TROUBLE P807-3A-H2 STATIC INVRTR 1Y81 TROUBLE P601-17A-D5 RHR PMP C AUTO START P870-8A-E1 CCW PMP B DISCH PRESS LO P870-8A-A1 CCW PMP B TRIP P870-5A-C2 CCW PMP A-C DISCH PRESS LO P680-4A2-B6 FPCC FLTR DMIN SYS TROUBLE P845-1A-A4 ADSORBER TRAIN A FLOW HIGH-LOW P845-1A-B4 ADSORBER TRAIN B FLOW HIGH-LOW P680-4A2-E3 OG PNL P845 TROUBLE P680-10A-C9 LP CNDSR SHELL PRESS HI P680-10A-A8 TURB VAC LO Revision 3

INITIAL CONDITIONS A. Plant Status: 100% power, middle of cycle B. Tech. Spec. Limitations in effect: None C. Significant problems/abnormalities:

1. TBCW Pump C is tagged out for motor oil replacement.

2. CRD Pump B is tagged out of service for oil replacement in the speed increaser.

D. Integrated Risk: Green E. Division Work Week: Division 2 F. Evolutions/maintenance for the up-coming shift :

1. Transfer RPS Bus B from normal to alternate power supply in preparation for maintenance on the RPS B Motor Generator.

2. The Motor Generator will be tagged out on the next shift.

3. No scram or isolation surveillances are in progress or planned for this shift.

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 1 of 43 Facility: Grand Gulf Nuclear Station Scenario No.: 4 Op-Test No.: GGNS 12-2017 Examiners: ____________________________ Operators: _____________________________

Event Malf. No. Event Type Event No. Description 1 N/A N (ATC,CRS) Withdraw control rods to 10% Bypass Valve position C (ATC,CRS) Condensate Pump C trip 2 fw115c A (CREW) 3 pte22n654c_a TS (CRS) HPCS CST Level Lo trip unit failing high I (BOP,CRS)

C(ATC,CRS) Startup Level Control controller failing downscale 4 fw274 A(CREW) (Feedwater Malfunctions ONEP) r21142u TS (CRS) ESF Transformer 11 trip with failure of Div 1 DG to start and 5 r21134g C(BOP,CRS) failure 15BA4 to re-energize (Loss of AC Power ONEP) n41140a A (CREW)

Main Steam Tunnel steam leak with failure of one steam ms066a 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) b21f065b_i Feedwater Line B line break inside Drywell with B21-F065B 7 M (CREW) fw171b power loss HPCS Pump Trip

• (CT-2) Inhibit ADS prior to automatic ADS valve opening during a LOCA

• (CT-3) When RPV level lowers to -160 wide range and e22052 C (ATC,BOP,CRS) cannot be maintained above -191 CFZ (MSCWL) and 8 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)

LPCS logic power failure

• (CT-4) When operating injection systems cannot r21219 C (BOP,CRS) maintain RPV level and ECCS systems fail to 9

automatically initiate, crew manually aligns 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.

Reason for Revision: Editorial changes to enhance Technical Specification actions.

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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 2 of 43 Objectives: To evaluate the applicants ability to operate the facility in response to the following evolutions:

1. Withdraw control rods to 10% Bypass Valve position

2. Respond to Condensate Pump B trip.

3. Respond to a HPCS CST Level LO trip unit failing high.

4. Respond to a Startup Level Control controller failing downscale.

5. Respond to an ESF Transformer 11 trip with a failure of Div 1 DG and failure of 15BA4 to re-energize.

6. Respond to an Main 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 HPCS Pump trip.

9. Respond to a LPCS logic power failure.

Initial Conditions:

• Reactor power is approximately 4% power

• Reactor pressure is 750 psig Inoperable Equipment: None Turnover:

• Reactor startup is in progress:

o Step 45 of IOI 03-1-01-1, Attachment XV o Step 96 of Control Rod Movement Sequence is complete o SJAE B is in service

• Condensate System is lined up as follows:

o Condensate Pumps A and C in service o Condensate Booster Pump C 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 3

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 3 of 43 Quantitative Attributes Table Attribute E3-304-1 Actual Description Target Malfunctions after

• HPCS Pump Trip E8) 1-2 2 EOP entry

• LPCS Logic power failure (E9)

• Condensate Pump C trip (Feedwater Malfunctions ONEP)

(E2)

Abnormal Events 2-4

• Startup Level Control Controller fails downscale (Feedwater 3

Malfunctions ONEP) (E4)

• ESF Transformer 11 trip with failure of Div 1 DG to start and failure 15BA4 to re-energize (Loss of AC Power ONEP) (E5)

• Main Steam Tunnel steam leak with failure of one steam line to Major Transients 1-2 isolate (E6) 2

• Feedwater Line B line break inside the Drywell with B21-F065B power loss (E7)

EOP entries requiring

• EP-4 1-2 3

• EP-2 substantive action

• 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) Inhibit ADS prior to automatic ADS valve opening during a LOCA

• (CT-3) When RPV level lowers to -160 wide range and cannot be maintained above -191 CFZ (MSCWL) and EOP based Critical insufficient high pressure injection systems are available to 2-3 4 Tasks 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.)

• (CT-4) 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 Normal Events N/A 1

• Withdraw control rods to 10% Bypass Valve position (E1)

Reactivity N/A 0

• N/A Manipulations

• Condensate Pump C trip (E2)

• HPCS CST Level LO trip unit failing high (E3)

• Startup Level Control controller failing downscale (E4)

Instrument /

N/A 6

• ESF Transformer 11 trip with failure of Div 1 DG to start and Component failures failure 15BA4 to re-energize (Loss of AC Power ONEP) (E5)

• HPCS Pump trip (E8)

• LPCS logic power failure (E9)

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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 4 of 43

• Condensate Pump C trip (E2)

• HPCS CST Level LO trip unit failing high (E3)

• Startup Level Control controller failing downscale (E4)

• ESF Transformer 11 trip with failure of Div 1 DG to start and failure 15BA4 to re-energize (Loss of AC Power ONEP) (E5)

Total Malfunctions N/A 8

• Main Steam Tunnel steam leak with failure of one steam line to isolate (E6)

• Feedwater Line B line break inside the Drywell with a B21-F065B power loss (E7)

• HPCS pump trip (E8)

• LPCS logic power failure (E9)

Top 10 systems and operator actions important to risk that are tested:

ADS (Event 7)

RHR (Event 9)

ESF Power (R20) (Event 5)

Condensate (Event 2)

Failure to manually depressurize with ADS/SRVs (Event 8)

Failure to align alternate power to 4.16 KV or 6.9KV buses (Event 5)

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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 5 of 43 SCENARIO ACTIVITIES:

The plant is operating at 4% power during a reactor startup.

Event 1 - Withdraw control rods to 10% Bypass Valve position After the crew assumes the shift, the ATC will withdraw control rods in accordance with Control Rod Movement Sequence IAW SOI 04-1-01-C11-2, Rod Control and Information System.

Event 2 - Condensate Pump C trip (Triggered by Lead Examiner)

At the direction of the Lead Examiner, Condensate Pump C will trip. CRS will enter 05-1-02-V-7, Feedwater System Malfunctions ONEP, and direct ATC to start Condensate Pump B IAW SOI 04-1 N19-1.

Event 3 - HPCS CST Level Lo trip unit failing upscale (Triggered by Lead Examiner)

After actions of Condensate Pump C trip are complete, HPCS CST Level LO trip unit, E22-N654C, will fail high. CRS will enter LCO 3.3.5.1 Condition A and, using Table 3.3.5.1-1, enter LCO 3.3.5.1 Condition D.

CRS will direct BOP to transfer HPCS suction from CST to Suppression Pool IAW SOI 04-1-01-E22-1.

Event 4 - Startup Level Control controller failing downscale (Triggered by Lead Examiner)

After HPCS Pump suctions are swapped and Tech Specs 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 IAW 05-1-02-V-7, Feedwater System Malfunctions ONEP, immediate actions. CRS will enter 05-1-02-V-7, Feedwater System Malfunctions ONEP.

Event 5 - ESF Transformer 11 trip with failure of 15BA4 to re-energize (Triggered by Lead Examiner)

After RPV level is stabilized, ESF Transformer 11 will trip. Division 1 Diesel Generator will fail to start.

CRS will enter 05-1-02-I-4, Loss of AC Power ONEP, 05-1-02-III-5, Automatic Isolations ONEP and 05 02-III-1, Inadequate Decay Heat Removal ONEP. BOP will re-energize Bus 15AA from ESF Transformer 12 and restore Instrument Air to CTMT by opening P53-F001. BOP will recognize the failure of 15BA4 to re-energize. CRS will enter LCO 3.8.7, Condition A, for LCC 15BA4 failure and LCO 3.8.1.B for failure of Division 1 Diesel Generator.

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 6 - Main Steam Tunnel steam leak with failure of one steam line to isolate (Triggered by Lead Examiner)

When ESF Bus 15AA has been re-energized and Tech Specs addressed, a steam leak in the Auxiliary Building Main 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.

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Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 6 of 43 Event 7 - Feedwater Line B line break inside Drywell with B21-F065B power loss (Triggered automatically)

When 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.

Event 8 - HPCS Pump Trip (Triggered automatically)

When Drywell pressure reaches 1.39 psig, HPCS Pump will trip and ESF Bus 16AB will lockout after 5 minutes causing a loss of all Division 2 ECCS. 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. Crew will inhibit ADS to prevent automatic operation (CT-2). When RPV level lowers to -160 wide range, the crew will emergency depressurize the RPV using ADS/SRVs (CT-3) and restore RPV level with Division 1 ECCS systems.

Event 9 - LPCS logic power failure (Triggered automatically)

When Drywell pressure rises to 1.39 psig, a LPCS logic power failure will occur. BOP 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-4).

The exercise ends when emergency depressurization is complete and RPV level restoration is being controlled.

Revision 3

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 7 of 43 Critical Task (CT-1) When MSIVs fail to isolate, manually (CT-2) Inhibit ADS prior to automatic ADS valve scram the reactor and close the MSIVs prior to opening during a LOCA Steam Tunnel temperature exceeding 250°F (Max Safe Temperature)

EVENT 6 8 Safety If a primary system is discharging into the Permitting automatic ADS initiation may be undesirable Significance secondary containment when this step of the for the following reasons:

procedure is reached, one of three conditions must

• ADS actuation can impose a severe thermal exist: transient on the RPV and may complicate efforts

• A primary system break cannot be isolated to control RPV water level.

because system operation is required to assure

• If only steam-driven systems are available for adequate core cooling or to shut down the injection, ADS actuation may directly lead to loss reactor. of adequate core cooling and subsequent core

• No isolation valves exist upstream of a primary damage.

system break, or if isolation valves do exist,

• The conditions assumed in the design of the ADS they cannot be closed because of some actuation logic (e.g., no operator action for 115 mechanical/ electrical/pneumatic failure. seconds after event initiation) may not exist when

• The source of the discharge cannot be the actions specified in this step are being determined. performed.

Since the RPV is the only significant source of

• The operating crew can draw on much more heat, other than a fire, which might cause area information than is available to the ADS logic temperatures to increase to their maximum safe (e.g., equipment out of service for maintenance, operating values, the action of manually operating experience with certain systems, scramming the reactor should terminate increasing probability of restoration of off-site power, etc.)

secondary containment temperatures. and can better judge, based on instructions contained in the EPGs/SAGs, when and how to If temperatures in any one of the areas listed in depressurize the RPV.

Table SC-1 of the Secondary Containment Control guideline approach their maximum safe operating Defeating the logic relieves the operating crew of the value, adequate core cooling, containment task of detecting timer initiation during execution of the integrity, safety of personnel, or continued more complex steps of Contingency #1 and precludes operability of equipment required to perform EPG unnecessary and unwanted automatic initiations.

actions can no longer be assured. Subsequent steps provide explicit and detailed instructions for controlling RPV water level and specify when emergency depressurization is appropriate.

Cueing Main Steam Tunnel temperature rising on PDS. Step L-5 of EP-2, RPV CONTROL, Alternate Level Main Steam Tunnel temperature alarms on panel Control Contingency P601.

MSIV open position indication on panel P601 and panel P858.

Performance Operator places the Reactor Mode Switch to Manipulation of ADS A and ADS B MANUAL INHIBIT Indicator SHUTDOWN on panel P680. switches on panel P601 vertical section.

Operator manipulates switches for MSIVs for Steam Line A to CLOSE on panel P601.

Performance RPS Group lights de-energized on panel P680. Inhibit switches click into INHIBIT position on panel Feedback P601 vertical section.

Control Rod full -in indication on panel P680.

Reactor power trend on nuclear instrumentation White indicating light on ADS A and ADS B MANUAL on panel P680. INHIBIT switches illuminate.

Green light indication energized and red light Receipt of ADS/SRV A and ADS/SRV B OOSVC indication off for MSIVs for Steam Line A on alarms on panel P601/18A-H2 and P601/19A-H2.

panel P601 and P858.

Justification If temperatures in any one of the areas listed in The 115 second ADS timer allows sufficient time for the for the Table SC-1 of the Secondary Containment Control crew to recognize and override automatic operation of chosen guideline approach their maximum safe operating the system. As long as ADS is inhibited before ADS performance value, adequate core cooling, containment valves open, reactor pressure will not be reduced.

limit integrity, safety of personnel, or continued operability of equipment required to perform EPG actions can no longer be assured.

BWR Owners App. B, step SC/T-4 and SC/T-4.1 App. B, step C1-1 Group Appendix Revision 3

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 8 of 43 Licensed 02-S-01-40, EP Technical Bases, Attachment VII, 02-S-01-40, EP Technical Bases, Attachment IV, Step Bases Step 8 through 10 L-5 Documents UFSAR Chapter 15.6.4 UFSAR Chapter 15A.6.5.3

• If an operator or the crew significantly deviates from, or fails to, follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review (NUREG 1021, Appendix D)

Revision 3

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 9 of 43 Critical Task (CT-3) When RPV level lowers to -160 wide range (CT-4) When operating injection systems cannot and cannot be maintained above -191 CFZ maintain RPV level and ECCS systems fail to (MSCWL) and insufficient high pressure injection automatically initiate, crew manually aligns ECCS systems are available to restore level, crew begins systems for injection prior to RPV pressure to Emergency Depressurize by opening at least lowering below 300 psig 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 8 9 Safety The MSCWL is the lowest RPV water level at which the Failure to recognize the auto initiation not occurring, and Significance covered portion of the reactor core will generate failure to take manual action per Conduct of Ops will sufficient steam to preclude any clad temperature in the result in unavailability of safety-related equipment uncovered portion of the core from exceeding 1500°F. necessary to provide adequate core cooling, otherwise When water level decreases below MSCWL with resulting in core damage and a large offsite release.

injection, clad temperatures may exceed 1500°F.

Cueing Wide range indication (SPDS and PDS) falls to -160 Indication of ECCS systems not initiating with initiation and lowering trend continues, and, before -160 wide conditions present:

range is reached, initial conditions, field reports, and

• Indication of Drywell pressure 1.39 psig or RPV control room indications convey that adequate high level -150.3 wide range pressure injection cannot be restored before level falls

• White light on LPCS/RHR A INIT RESET below -191 CFZ.

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 Manipulation of seven of the eight ADS/SRVs on panel Operator manually manipulates switches for Div 1 Indicator P601: ECCS pumps and directs operators to manually open B21-F041K Div 1 ECCS injection valves from Division 1 Remote B21-F047L Shutdown Panel (RHR A) and locally (LPCS).

B21-F041F B21-F047A B21-F051C B21-F041D B21-F051A B21-F051B Performance Crew will observe ADS/SRV light indication go from Red light on and green light extinguished on LPCS Feedback green to red, reactor pressure lowering on SPDS and and/or RHR A pump and valve handswitches on panel panel P601 indications. 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 The MSCWL (-191 CFZ) is the lowest RPV water level Attempting to align high pressure ECCS systems must for the at which the covered portion of the reactor core will be performed to determine their availability by the time chosen generate sufficient steam to preclude any clad TAF is reached in order to properly implement EP-2 performance temperature in the uncovered portion of the core from decision steps regarding restoring and maintaining RPV limit exceeding 1500°F. Emergency depressurization is level. Attempting to align low pressure ECCS systems allowed when level goes below TAF (-160 wide range) can only be done once RPV pressure falls below the and should be performed, if in the judgment of the CRS, injection valve RPV pressure permissive and will only level cannot be maintained above -191 CFZ. Since it is be effective once RPV pressure falls below the shutoff intended for the scenario supporting this CT to, early in head of the respective ECCS pump. The reduction in the event, clearly indicate no high pressure injection RPV pressure will normally be via Emergency systems can be made available to reverse the lowering Depressurization, which is a separate critical task level trend, the crew will have time to communicate and bounded by a minimum RPV level.

open 7 of 8 ADS/SRVs before -191 CFZ.

BWR Owners App. B, Contingency #1 Step C1-4 App. B, Contingency 1, step C1-3 Group Appendix Revision 3

Appendix D Scenario Outline Form ES-D-1 NRC GGNS 12-2017 Scenario 4 Page 10 of 43 Licensed 02-S-01-40, EP Technical Bases, Attachment IV, Step 02-S-01-40, EP Technical Bases, Attachment IV, Step Bases L through L-13 L-14 Documents UFSAR Chapter 15A.6.5.3 UFSAR Chapter 15A.6.5.3

• If an operator or the crew significantly deviates from, or fails to, follow procedures that affect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review (NUREG 1021, Appendix D)

• • For Crew 1, consisting of I1, R1, and R2, a Post Scenario CT was created because the crew failed to recognize that the ESF 11 trip caused a loss of power to the bus that powers the F001 valve, which is the instrument air to containment isolation valve. The bus loss to this valve was not recognized and the valve was not reopened before control rods started to drift into the core. By procedure when multiple control rods start to drift the crew is required to SCRAM the reactor. The crew recognized that multiple control rods were drifting in and completed the SCRAM of the reactor, however they created this critical task because they unnecessarily SCRAMMED when they were expected to reopen the F0001 valve and restore instrument air, which would have prevented any rod drift conditions and keep the reactor at power. This was an unnecessary challenge on the RPS system when it was not needed.

The Chief Examiner consulted with HQ staff to confirm it was a post-scenario CT before the examination team left the site that day and HQ staff concurred that it was a post-scenario CT.

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 4 Page 11 of 43 Simulator Setup:

A. Initialization

1. Log off all simulator PDS and SPDS computers (PDS and SPDS must come up after the simulator load for proper operation).

2. Startup the simulator using Simulator Instructors Job Aid section 7.3.

Note:

Prior to running the Schedule File, ensure no Event Files are Open. If an existing Event File is Open prior to running the Schedule File, then any associated Event Files will not automatically load.

3. Open Schedule.exe and Director.exe by clicking on the Icon in the Thunder Bar.

4. Set the Simulator to IC-103 and perform switch check (Using Quick Reset in Director).

5. Click on Open in the Schedule window and Open Schedule File 12-2017 NRC Exam Scenario 4.sch (in the Schedule Directory)

6. In Schedule window, click on the Stopped red block. The red block will change to a green arrow and indicate the scenario is active (Running).

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 4 Page 12 of 43

7. Click the Summary tab in the Director window. Verify the schedule files are loaded and opened per Section B below. (Note: Any actions in the schedule file without a specific time will not load into the director until triggered.)

8. Take the simulator out of freeze.

9. Log on to all simulator PDS and SPDS computers.

10. Verify or perform the following:

• IC-103

• Ensure the correct rod movement sequence available at the P680 and marked up through Step 96 complete

• Provide applicants with copy of 02-S-01-27, Operations Philosophy, Attachment 1, Control Rod Movement Expectation

• Provide applicants with copy of 04-1-01-C11-2, Rod Control and Information System, Section 3, Precautions and Limitations, and Section 4.3, RC&IS Rod/Rod Gang Notch Out.

• Ensure all procedures are marked as indicated for turnover conditions

• Set IRM recorder scales

• Advance all chart recorders and ensure all pens inking properly

• Clear any graphs and trends off of SPDS

• Place a tape flag on annunciator P680-4A2-C5, CONT ROD WITHDRAWL BLOCK

• Place a tape flag on annunciator P680-4A1-A7, CRD DRIVE WTR TO RX P HI

• Place red tag on Generator Disconnect pushbutton

11. Run through any alarms and ensure alarms are on. (Note: On T-Rex, to verify alarms are ON, the indicator will indicate Alarms On).

12. Place the simulator in Freeze.

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 4 Page 13 of 43 File loaded verification:

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Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 4 Page 14 of 43 Revision 3

Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 4 Page 15 of 43 Procedures that may be used in this scenario:

Procedure No. Rev Procedure Title 03-1-01-IOI-1 172 Cold Shutdown To Generator Carrying Minimum Load 04-1-01-C11-1 154 Control Rod Drive Hydraulic System 04-1-01-C41-1 123 Standby Liquid Control System 04-1-01-E12-1 147 Residual Heat Removal System 04-1-01-N21-1 74 Feedwater System 04-1-01-N32-2 33 Turbine Generator Control 04-1-01-R21-17 10 ESF Bus 17AC 04-1-02-1H13-P601 161 Alarm Response Instruction Panel No.: 1H13-P601 04-1-02-1H13-P680 233 Alarm Response Instruction Panel No.:1H13-P680 04-1-02-1H13-P864 31 Alarm Response Instruction Panel No.: 1H13-P864 04-1-02-1H13-P870 154 Alarm Response Instruction Panel No.: 1H13-P870 04-S-02-SH13-P807 32 Alarm Response Instruction Panel No.: SH13-P807 05-1-02-I-1 130 Reactor Scram 05-1-02-I-2 37 Turbine and Generator Trips 05-1-02-I-4 51 Loss of AC Power 05-1-02-III-3 115 Reduction In Recirculation System Flow Rate 05-1-02-III-5 49 Automatic Isolations 05-1-02-IV-1 117 Control Rod / Drive Malfunctions 05-S-01-EP-1 36 Emergency / Severe Accident Procedure Support Documents 05-S-01-EP-2 45 RPV Control Tech Spec 3.1.3 Tech Spec 3.8.1 Revision 3

Appendix D Scenario Outline Form ES-D-2 NRC GGNS 12-2017 Scenario 4 Page 16 of 43 Expected Alarms:

P680-4A2-C5, CONT ROD WITHDRAWL BLOCK P680-1A-A3, CNDS PMP C TRIP P601-16A-H5, HPCS SYS OOSVC P680-3A-A3, RX LVL 40/32 HI/LO P680-2A-C9, DFCS TROUBLE P807-4A-E6, ESF XFMR 11 TROUBLE P807-4A-B2, ESF XFMR 11 LOCKOUT TRIP P807-1A-B5, ESF DIST BUSES INCM FDR 152-1902 P807-1A-B4, ESF DIST BUSES INCM FDR 152-1901 P864-1A-A3, 4.16KV BUS 15AA UNDERVOLTAGE P864-1A-D1, DG 11 AUTO START NOT AVAIL P864-1A-D2, DIV 1 DSL GEN TROUBLE P864-1A-D3, 480V LCC 15BA1 UNDERVOLT P864-1A-D4, 480V LCC 15BA2 UNDERVOLT P864-1A-F4, 480V LCC 15BA6 UNDERVOLT P864-1A-F3, 480V LCC 15BA5 UNDERVOLT P864-1A-E4, 480V LCC 15BA4 UNDERVOLT P864-1A-E3, 480V LCC 15BA3 UNDERVOLT P680-5A-C4, SCRAM PILOT VLV AIR HDR PRESS LO P680-4A2-A4, RC&IS INOP P864-1A-A4, 4.16KV BUS 15AA INCM FDRS TRIP P807-3A-H1, STATIC INVRTR 1Y80 TROUBLE P807-3A-G4, STATIC INVRTR 1Y79 TROUBLE P807-3A-G3, STATIC INVRTR 1Y98 TROUBLE P807-1A-A6, ESF XFMR 11 INCM FDR 552-1104 TRIP P601-18A-A3, MSL PIPE TNL CH-B TEMP HI P601-18A-A4, MSL PIPE TNL CH-C TEMP HI P601-19A-A3, MSL PIPE TNL CH-A TEMP HI P601-19A-A4, MSL PIPE TNL CH-D TEMP HI Revision 3

INITIAL CONDITIONS A. Plant Status:

• Reactor startup is in progress with power at approximately 4%:

o Step 45 of IOI 03-1-01-1, Attachment XV o Step 96 of Control Rod Movement Sequence is complete o SJAE B is in service

• Condensate System is lined up as follows:

o Condensate Pumps A and C in service o Condensate Booster Pump C 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 B. Tech. Spec. Limitations in effect: None C. Significant problems/abnormalities: None D. Integrated Risk: High E. Division Work Week: Division 2 F. Evolutions/maintenance for the up-coming shift :

1. 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.