ML21334A017

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GG-2021-10 Final Outlines
ML21334A017
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 10/25/2021
From: Heather Gepford
Operations Branch IV
To:
Entergy Operations
References
Download: ML21334A017 (59)
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Text

ES-401 BWR Examination Outline Form ES-401-1

Facility: Grand Gulf Station Date of Exam: 2021 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 4 4 4 3 2 20 4 3 7 Emergency & 2 1 1 1 N/A 1 1 N/A 2 7 1 2 3 Abnormal Plant Evolutions Tier Totals 4 5 5 5 4 4 27 5 5 10
2. 1 2 2 2 3 2 3 3 2 3 2 2 26 2 3 5 Plant 2 1 1 1 1 1 1 1 1 1 2 1 12 1 1 1 3 Systems Tier Totals 3 4 4 4 4 3 3 3 3 3 4 38 4 4 8
3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 3 2 2 2 1 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO -only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category.)
2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO -only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES -401 for guidance regarding the elimination of inappropriate K/A statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO -only portions, respectively.
6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES -401 for the applicable K/As.
8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO -only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO -only exams.
9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES -401 -3. Limit SRO selections to K/As that are linked to 10 C FR 55.43.

G* Generic K/As

ES-401 2 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function KKKAA2 G* K/A Topic(s) IR #

1 2 3 1

295001 Partial or Complete Loss of Forced X G2.1.19 Ability to use plant computers to evaluate 1 Core Flow Circulation / 1 & 4 system or component status. (CFR: 41.10 / 45.12) 3.9

Ability to determine and/or interpret the following as 295003 Partial or Complete Loss of AC / 6 they apply to partial or complete loss of AC :

X AA2.02 Reactor power / pressure / and level. 4.2* 2 (CFR: 41.10 / 43.5 / 45.13)

Ability to operate and/or monitor the following as 295004 Partial or Total Loss of DC Pwr / 6 they apply to partial or total loss of DC Pwr:

X AA1.01 D.C. electrical distribution systems. 3.3 3 (CFR: 41.7 / 45.6)

Knowledge of the reasons for the following 295005 Main Turbine Generator Trip / 3 responses as they apply to Main Turbine Generator Trip: AK3.02 Recirculation pump downshift/trip: 3.4 4 X Plant Specific.

(CFR: 41.5 / 45.6)

Knowledge of the interrelations between SCRAM 3.1 295006 SCRAM / 1 X and the following: AK2.05 CRD mechanism 5 (CFR: 41.7 / 45.8)

Knowledge of the interrelations between Control 4.0*

295016 Control Room Abandonment / 7 Room Abandonment and the following: AK2.02 X Local control stations: Plant Specific. 6 (CFR: 41.7 / 45.8)

Knowledge of the operational implications of the 3.5 295018 Partial or Total Loss of CCW / 8 following concepts as they apply to partial or Total loss of CCW: AK1.01 Effects on component/system 7 X operations (CFR: 41.8 to 41.10)

Knowledge of the reasons for the following 3.2 295019 Partial or Total Loss of Inst. Air / 8 responses as they apply to partial or complete loss X of instrument air: AK3.0 3 Service air isolations 8 (CFR: 41.5 / 45.6)

Ability to operate and/or monitor the following as 3.7 9 295021 Loss of Shutdown Cooling / 4 they apply to Loss of shutdown cooling:

X AA1.04 Alternate heat removal methods (CFR: 41.7 / 45.6)

Ability to determine and/or interpret the following as X they apply to loss of shutdown cooling: 3.5 76 AA2.03 Reactor water level

3.4 295023 Refueling Acc / 8 Ability to determine and/or interpret the following as X they apply to refueling accidents: AA2.02 Fuel Pool 10 Level (CFR: 41.10 / 43.5 / 45.13)

ES-401 3 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function KKKAA2 G* K/A Topic(s) IR #

1 2 3 1 G2.2.44 Ability to interpret control room indications 4.2 11 295024 High Drywell Pressure / 5 to verify status and operation of a system and understand how operator actions and directives X affect plant and system conditions. (CFR: 41.5 / 43.5

/ 45.12)

Ability to determine and/or interpret the following as3.777 X they apply to High Drywell Pressure: EA2.10 Containment temperature: Mark III.

(CFR: 41.10 / 43.5 / 45.13)

295025 High Reactor Pressure / 3 X 4.2 12 Ability to determine and/or interpret the following as they apply to high reactor pressure: EA2.02 Reactor Power (CFR: 41.10 / 43.5 / 45.13)

Ability to operate and/or monitor the following as 4.1 13 295026 Suppression Pool High Water Temp. they apply to Suppression Pool High Water Temp:

/ 5 X EA1.01 Suppression pool cooling (CFR: 41.7 / 45.6)

G2.2.40 Ability to apply Technical Specifications for X a system. (CFR: 41.10 / 43.2 / 43.5 / 45.3) 4.7 78

Knowledge of the reasons for the following 3.7 14 295027 High Containment Temperature / 5 responses as they apply to high containment temperature (Mark III containment only):

X EK3.01 Emergency depressurization: Mark-III (CFR: 41.5 / 45.6)

Knowledge of the interrelations between High 3.6 15 295028 High Drywell Temperature / 5 Drywell Temperature and the following:

X EK2.04 Drywell ventilation (CFR: 41.7 / 45.8)

Knowledge of the operational implications of the 3.8* 16 295030 Low Suppression Pool Wtr Lvl / 5 following concepts as they apply to Low Supp Pool X Wtr Lvl: EK1.01 Steam condensation.

(CFR: 41.8 to 41.10)

G2.2.25 Knowledge of the bases in Technical X Specifications for limiting conditions for operations 4.2 79 and safety limits. (CFR: 41.5 / 41.7 / 43.2)

Knowledge of the interrelations between Reactor 4.1 17 295031 Reactor Low Water Level / 2 Low Water Level and the following: E K2.13 X ARI/RPT/ATWS: Plant Specific (CFR: 41.7 / 45.8)

ES-401 4 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function KKKAA2 G* K/A Topic(s) IR #

1 2 3 1 Knowledge of the reasons for the following 3.6* 18 295037 SCRAM Condition Present and responses as they apply to SCRAM Condition Reactor Power Above APRM Downscale or X Present and R eactor Power Above APRM Unknown / 1 Downscale or Unknown: EK3.08 ATWS circuitry:

Plant Specific.

(CFR: 41.5 / 45.6)

Ability to determine and/or interpret the following as

X they apply to SCRAM condition present and reactor power above ARPRM downscale or unknown: 4.4* 80 EA2.03 SBLC tank level.

(CFR: 41.10 / 43.5 / 45.13)

Ability to operate and/or monitor the following as 3.5 19 295038 High Off-site Release Rate / 9 X they apply to High offsite release rate:

EA1.06 Plant ventilation Ability to determine and/or interpret the following as 4.3* 81 they apply to High offsite release rate: EA2.03 X Radiation levels.

(CFR: 41.10 / 43.5 / 45.13)

600000 Plant Fire On Site / 8 X Knowledge of the operation applications of the 2.9 20 following concepts as they apply to Plant Fire On Site: AK1.02 Fire Fighting

G2.2.37 Ability to determine operability and/or 4.6 82 700000 Generator Voltage and Electric Grid X availability of safety related equipment. (CFR: 41.7 /

Disturbances / 6 43.5 / 45.12)

K/A Category Totals: 3 4 4 4 3/4 2/3 Group Point Total: 20/

7

ES-401 5 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function KKKAAG* 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 2.9 21 X as they apply to Loss of Main Condenser Vac:

AA2.01 Condenser Vacuum/absolute pressure

295007 High Reactor Pressure / 3 Not sampled 295008 High Reactor Water Level / 2 X G2.1.20 Ability to interpret and execute procedure 4.6 22 steps. (CFR: 41.10 / 43.5 / 45.12) 295009 Low Reactor Water Level / 2 Not sampled 295010 High Drywell Pressure / 5 Not sampled 295011 High Containment Temp / 5 G2.2.22 Knowledge of limiting conditions for 4.7 83 X operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2) 295012 High Drywell Temperature / 5 Not sampled 295013 High Suppression Pool Temp. / 5 Not sampled 295014 Inadvertent Reactivity Addition / 1 Ability to operate and/or monitor the following as 3.3 23 they apply to Inadvertent Reactivity Addition:

X AA1.06 Reactor/turbine pressure regulating system (CFR: 41.7 / 45.6)

295015 Incomplete SCRAM / 1 X G2.4.3 Ability to identify post-accident 3.7 24 instrumentation. (CFR: 41.6 / 45.4) 295017 High Off -site Release Rate / 9 Not sampled 295020 Inadvertent Cont. Isolation / 5 & 7 Knowledge of the interrelations between 3.1 25 X Inadvertent Containment Isolation and the following:

AK2.12 Instrument air/nitrogen: Plant Specific 295022 Loss of CRD Pumps / 1 Knowledge of the reasons for the following X responses as they apply to Loss of CRD Pumps:

AK3.01 Reactor SCRAM 3.7 26 295029 High Suppression Pool Wtr Lvl / 5 Not sampled 295032 High Secondary Containment Area Not sampled Temperature / 5 295033 High Secondary Containment Area X Ability to determine and/or interpret the following Radiation Levels / 9 as they apply to high Secondary Containment Area 3.9 84 Radiation Levels: EA2.01 Area radiation levels (CFR: 41.10 / 43.5 / 45.13) 295034 Secondary Containment Not sampled Ventilation High Radiation / 9 295035 Secondary Containment High Not sampled Differential Pressure / 5 295036 Secondary Containment High Knowledge of the operational implications of the Sump/Area Water Level / 5 following concepts as they apply to Secondary 2.6* 27 X Containment High Sump / Area Water Level:

EK1.02 Electrical ground/ circuit malfunction (CFR: 41.8 to 41.10) 500000 High CTMT Hydrogen Conc. / 5 X 2.4.6 Knowledge of EOP mitigation strategies. 4.7 85 (CFR: 41.10 / 43.5 / 45.13)

K/A Category Point Totals: 1 1 1 1 1/1 2/2 Group Point Total: 7/3

ES-401 6 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A2 AA G* K/A Topic(s) IR #

1 2 3 4 5 6 1 3 4 Knowledge of RHR/LPCI: Injection Mode design feature(s) and/or interlocks which 3.2 28 X provide for the following: K4.03 Pump minimum flow protection 203000 RHR/LPCI: Injection (CFR: 41.7)

Mode X 2.2.42 Ability to recognize system parameters that are entry-level conditions 4.6 86 for Technical Specifications.

Knowledge of the effect that a loss or malfunction of the following will have on the 3.5 29 Shutdown Cooling System (RHR SDC 205000 Shutdown Cooling X Mode) : K6.08 RHR service water: Plant-Specific (CFR: 41.7 / 45.7)

Knowledge of electrical power supplies to X the following: K2.03 Initiation logic 2.9* 30 (CFR: 41.7)

209001 LPCS X Ability to predict and/or monitor changes in parameters associated with operating the 3.8 31 LPCS controls including: A1.03 Reactor water level (CFR: 41.5 / 45.5)

Knowledge of the operational implications of the following concepts as they apply to 2.6 32 X HPCS: K5.02 Heat removal (transfer) mechanism: BWR-5,6 (CFR: 41.5 / 45.3)

Ability to (a) predict the impacts of the 209002 HPCS X following on the HPCS SYSTEM; and (b) 3.4 33 based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.13 Low condensate storage tank level BWR-5,6 (CFR: 41.5 / 45.6)

Ability to monitor automatic operations of the SLC SYSTEM including: A3.02 Tank 3.9 34 211000 SLC X level: Plant-Specific (CFR: 41.7 / 45.7)

Ability to manually operate and/or monitor in the control room: A4.02 Perform system 4.6 35 212000 RPS X functional test(s)

(CFR: 41.7 / 45.5 to 45.8)

G2.4.31 Knowledge of annunciator alarms, 215003 IRM X indications, or response procedures. 4.2 36 (CFR: 41.10 / 45.3)

Ability to manually operate and/or monitor in the control room: A4.05 SRM back panel 3.1 37 215004 Source Range Monitor X switches, meters, and indicating lights (CFR: 41.7 / 45.5 to 45.8)

ES-401 7 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A2 AAG* K/A Topic(s) IR #

1 2 3 4 5 6 1 3 4 X Ability to monitor automatic operations of the APRM/LPRMs including: A3.06 3.0 38 Maximum disagreement between flow comparator channels: Plant-Specific (CFR: 41.7 / 45.7)

215005 APRM / LPRM X Ability to (a) predict the impacts of the 3.6 87 following on the APRM/LPRM SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.02 Upscale or downscale trips

Knowledge of the physical connections X and/or cause/effect relationships between 3.5 39 REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following: K1.02 Nuclear boiler system (CFR: 41.2 to 41.9 /

45.7 to 45.8) 217000 RCIC Ability to (a) predict the impacts of the 3.1 40 X following on the RCIC; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A2.07 Loss of lube oil (CFR: 41.5 / 45.6)

Ability to predict and/or monitor changes in parameters associated with operating the 3.2 41 218000 ADS X ADS controls including: A1.03 ADS valve Air supply pressure: Plant-Specific (CFR: 41.5 / 45.5)

Knowledge of PCIS/NSSS design X feature(s) and/or interlocks which provide 3.0 42 for the following: K4.01 Redundancy (CFR: 41.7) 223002 PCIS/Nuclear Steam Supply Shutoff X Knowledge of the effect that a loss or malfunction of the following will have on the 3.5 43 PCIS/NSSS: K6.08 Reactor protection system (CFR: 41.7 / 45.7)

Knowledge of electrical power supplies to X the following: K2.01 SRV solenoids 2.8* 44 (CFR: 41.7)

239002 SRVs Knowledge of the operational implications X of the following concepts as they apply to 3.7 45 RELIEF/SAFETY VALVES:

K5.02 Safety function of SRV operation

Knowledge of the effect that a loss or malfunction of the Reactor Water level 3.7 46 259002 Reactor Water Level X Control will have on following:

Control K3.02 Reactor feedwater system (CFR: 41.7 / 45.4)

ES-401 8 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A2 AA G* K/A Topic(s) IR #

1 2 3 4 5 6 1 3 4 Knowledge of the physical connections 261000 SGTS X and/or cause/effect relationships between SGTS and the following: K1.08 Process 2.8 47 radiation monitoring system.

G2.4.21 Knowledge of the parameters X and logic used to assess the status of 4.6 88 safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 / 45.12)

Knowledge of the effect that a loss or 262001 AC Electrical malfunction of the AC Elect Distribution will 3.2 48 Distribution X have on following: K3.05 Off site power system (CFR: 41.7 / 45.4)

Knowledge of UPS design feature(s) and/or interlocks which provide for the following: 3.1 49 262002 UPS (AC/DC) X K4.01 Transfer from preferred power to alternate power supplies (CFR: 41.7)

X G2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7) 3.9 50

Ability to (a) predict the impacts of the 263000 DC Electrical following on the DC Electrical Dist; and (b)

Distribution X based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal 3.2 89 conditions or operations: A2.01 Grounds (CFR: 41.5 / 45.6)

Knowledge of the effect that a loss or malfunction of the following will have on the 3.6 51 EDG System: K6.02 Fuel oil pumps X (CFR: 41.7 / 45.7) 264000 EDGs 2.2.25 Knowledge of bases for Tech Specs X for LCOs and safety limits (CFR: 41.5 / 4.2 90 41.7 / 43.2)

Ability to monitor automatic operations of the Instrument Air system including: A3.02 3.1 52 300000 Instrument Air X Air temperature (CFR: 41.7 / 45.7)

Ability to predict and/or monitor changes in 400000 Component Cooling parameters associated with operating the 2.7 53 Water X CCW controls including: A1.03 CCW pressure (CFR: 41.5 / 45.5)

K/A Category Point Totals: 2 2 2 3 2 3 3 2/2 3 2 2/3 Group Point Total: 26/5

ES-401 9 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K K K K K K A A2 A A4 G* K/A Topic(s) IR #

1 2 3 4 5 6 1 3 201001 CRD Hydraulic X Knowledge of the physical 3.0 54 connections and/or cause effect relationships between CRD Hydraulic System and the following:

K1.02 Condensate System (CFR: 41.2 to 41.9 / 45.7 to 45.8) 201002 RMCS X Knowledge of the effect that a 2.9 loss or malfunction of the Reactor Manual Control System will have onfollowing: K3.03 Ability to REPLACE with process rod block signals 271000 (CFR: 41.7 / 45.4)

201003 Control Rod and Drive X Ability to (a) predict the impacts of 3.4 55 Mechanism the following on the Control Rod and Drive Mechanism; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A2.01 Stuck Rod (CFR: 41.5 / 45.6) 201004 RSCS Not sampled 201005 RCIS X Knowledge of the operational 3.5 56 implications of the following concepts as they apply to ROD CONTROL AND INFORMATION SYSTEM (RCIS):

K5.09 High power setpoints BWR-6 (CFR: 41.5 / 45.3) 201006 RWM Not sampled 202001 Recirculation X Knowledge of the effect that a loss or 3.5 57 malfunction of the following will have on the Recirculation System: K6.01 Jet Pumps (CFR: 41.7 / 45.7) 202002 Recirculation Flow Control Not sampled 204000 RWCU X Knowledge of REACTOR WATER 2.9 58 CLEANUP SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.03 Over temperature protection for system components 214000 RPIS Not sampled 215001 Traversing In-Core Probe Not sampled 215002 RBM Not sampled 216000 Nuclear Boiler Inst. X G2.2.38 Knowledge of conditions 4.5 91 and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1 / 45.13) 219000 RHR/LPCI: Torus/Pool X Knowledge of electrical power 2.5* 59 Cooling Mode supplies to the following: (CFR: 41.7)

K2.01 Valves 223001 Primary CTMT and Aux. X Ability to (a) predict the impacts of 4.1 92 the following on the Primary CTMT and Aux; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A2.02 Steam bypass of suppression pool (CFR: 41.5 / 45.6)

ES-401 10 Form ES-401-1

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K K K K K K A A2 A A4 G* K/A Topic(s) IR #

1 2 3 4 5 6 1 3 226001 RHR/LPCI: CTMT Spray X Ability to (a) predict the impacts of 3.6 60 Mode the following on the RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE ; and (b) based on those From last KA, 204000 RWCU predictions, use procedures to already sampled correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR:

41.5 / 45.6) A2.15 High containment / drywell pressure 230000 RHR/LPCI: Torus/Pool Not sampled Spray Mode 233000 Fuel Pool Cooling/Cleanup Not sampled 234000 Fuel Handling Equipment X Knowledge of the effect that a loss or 3.6 93 malfunction of the following will have on the Fuel handling Equipment:

K6.06 Fuel transfer tube interlocks:

Mark-III (CFR: 41.7 / 45.7) 239001 Main and Reheat Steam Not sampled 239003 MSIV Leakage Control Not sampled 241000 Reactor/Turbine Pressure X Ability to manually operate and/or 3.2 61 Regulator monitor in the control room: A4.09 Combined intermediate valves (operation): Plant-Specific.

(CFR: 41.7 / 45.5 to 45.8) 245000 Main Turbine Gen. / Aux. X Ability to predict and/or monitor 2.7 62 changes in parameters associated with operating the Main Turbine Generator and Aux System controls including: A1.03 Turbine valve position (CFR: 41.5 / 45.5) 256000 Reactor Condensate X G2.1.32 Ability to explain and apply 3.8 63 system limits and precautions.

(CFR: 41.10 / 43.2 / 45.12) 259001 Reactor Feedwater X Ability to monitor automatic 3.4 64 operations of the Reactor Feedwater System including: A3.10 Pump trips 268000 Radwaste Not sampled 271000 Offgas X Knowledge of the effect that a loss or 3.5 65 malfunction of the Offgas System will have on the following: K3.01 Condenser vacuum (CFR: 41.5 /

45.3)

272000 Radiation Monitoring Not sampled 286000 Fire Protection Not sampled 288000 Plant Ventilation Not sampled 290001 Secondary CTMT Not sampled 290003 Control Room HVAC Not sampled 290002 Reactor Vessel Internals Not sampled

K/A Category Point Totals: 1 1 1 1 1 1/1 2/1 1 1 1/1 Group Point Total: 12/3 1

ES-401 11 Form ES-401-3

Facility: Grand Gulf Nuclear Station Date of Exam: 10/2021

Category K/A # Topic RO SRO-Only IR # IR #

2.1.1 Knowledge of conduct of operations requirements. 3.8 66 (CFR: 41.10 / 45.13) 2.1.8 Ability to coordinate personnel activities outside the 3.4 67 control room. (CFR: 41.10 / 45.5 / 45.12 / 45.13)

1. 2.1.21 Ability to verify the controlled procedure copy. 3.5* 68 Conduct of (CFR 41.10 / 45.10 / 45.13)

Operations 2.1.2 Knowledge of operator responsibilities during all modes of 4.4 94 plant operation (CFR 41.10 / 45.13) 2.1.34 Knowledge of primary and secondary plant chemistry 3.5 95 limits. (CFR 41.10 / 43.5 / 45.12)

Subtotal 3 2 2.2.2 Ability to manipulate the controls as required to operate 4.6 69 the facility between shutdown and power levels 2.2.12 Knowledge of surveillance procedures. (CFR 41.10 / 3.7 70

2. 45.13)

Equipment 2.2.14 Knowledge of the process for controll ing equipment 3.9 71 Control configuration or status. (CFR: 41.10 / 43.3 / 45.13)

2.2.7 Knowledge of the process for conducting special or 3.6 96 infrequent tests. (CFR 41.10 / 43.3 / 45.13)

Subtotal 3 1 2.3.4 Knowledge of radiation exposure limits under normal or 3.2 72 emergency conditions. (CFR: 41.12 / 43.4 / 45.10) 2.3.12 Knowledge of radiological safety principles pertaining to 3.2 73 licensed operator duties, such as containment entry

3. requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Radiation 2.3.11 Ability to control radiation releases. 4.3 97 Control (CFR 41.11 / 43.4 / 45.10)

2.3.14 Knowledge of radiation or contamination hazards that 3.8 98 may arise during normal, abnormal, or emergency conditions or activities. (CFR 41.12 / 43.4 / 45.10)

Subtotal 2 2 2.4.12 Knowledge of general operating crew responsibilities 4.0 74 during emergency operations 2.4.19 Knowledge of EOP layout, symbols, and icons. 3.4 75

4. (CFR 41.10 / 45.13)

Emergency 2.4.23 Knowledge of the bases for prioritizing emergency 4.4 99 Procedures / procedure implementation during emergency operations.

Plan (CFR 41.10 / 43.5 / 45.13)

2.4.37 Knowledge of the lines of authority during implementation 4.1 100 of the emergency plan. (CFR 41.10 / 45.13)

Subtotal 2 2 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4

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

RO T1 - G1 Original KA: 295038 High Off-site Release Rate, EA1 - Ability to 295038 295038 operate and/or monitor the following as they apply to High offsite EA1.01 EA1.06 release rate:

Q# 19 EA1.01 Stack gas monitoring system: Plant Specific (CFR: 41.7 / 45.6)

GGNS does not have a Stack. No credible tie for this K/A exists for the system.

Randomly selected new K/A EA1.06 Plant ventilation

Page 1 point totals not affected by this change.

RO T1 - G2 Original KA: 295002 Loss of Main Condenser Vac, AA2 - Ability to 295002 295004 determine and/or interpret the following as they apply to Loss of AA2.04 AA2.01 Main Condenser Vac:

Q# 21 AA2.04 Offgas system flow

This KA was rejected due to low operational value for discriminatory RO level question, the Offgas flow will go up or down depending on condenser vacuum.

Randomly selected new K/A: AA2.01 Condenser Vacuum/absolute pressure

Knowledge of the automatic actions during a loss of vacuum is more important than offgas flow going up or down.

Page 1 point totals not affected by this change.

RO T1 - G2 Original KA: 295020 Inadvertent Cont. Isolation, AK2, Knowledge 295020 295020 of the interrelations between Inadvertent Containment Isolation and AK2.11 AK2.12 the following:

Q# 25 AK2.11 Standby gas treatment system/FRVS (plant specific)

At GGNS, Standby Gas Treatment has no tie with Containment Ventilation or Isolation. No credible tie for this K/A exists for the system.

Randomly selected New K/A: AK2.12, Instrument air/nitrogen.

Page 1 point totals not affected by this change.

ES-401 Record of Rejected K/As Form ES-401-4

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

RO T1 - G2 Original KA: AK2, Knowledge of the reasons for the following 295022 295022 responses as they apply to Loss of CRD Pumps:

AK3.02 AK3.01 AK3.02 CRDM high temperature Q# 26 This KA was rejected due to low operational value for discriminatory RO level question.

Randomly selected New K/A: A3.01, Reactor SCRAM.

Page 1 point totals not affected by this change.

RO T2 - G1 Original KA: A2, Ability to (a) predict the impacts of the following on 209002 209002 the HPCS SYSTEM; and (b) based on those predictions, use A2.06 A2.13 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Q# 33 A2.06 Core spray line break: BWR-5,6

This KA was rejected due to low operational value for discriminatory RO level question.

Randomly selected New K/A: A2.13. Low condensate storage tank level BWR-5,6 (CFR: 41.5 / 45.6)

Page 1 point totals not affected by this change.

RO T2 - G1 Original KA: 239002 SRVs, K2, Knowledge of electrical power 239002 239001 supplies to the following: K2.01 SRV solenoids (CFR: 41.7)

K5.06 K5.02 Knowledge of the operational implications of the following concepts Q# 45 as they apply to RELIEF/SAFETY VALVES:

K5.06 Vacuum breaker operation (CFR: 41.5 / 45.3)

This KA was rejected due to low operational value for discriminatory RO level question.

Randomly selected New K/A: K5.02 Safety function of SRV operation

Page 1 point totals not affected by this change.

ES-401 Record of Rejected K/As Form ES-401-4

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

RO T2 - G1 Original KA: 261000 SGTS, Knowledge of the physical connections 261000 261000 and/or cause/effect relationships between SGTS and the following:

K1.12 K1.08 K1.12 Primary containment purge system: Plant-Specific (CFR:

Q# 47 41.2 to 41.9 / 45.7 to 45.8

This KA was rejected due to at GGNS there is no connection between SBGT and Primary Containment Purge System. No credible tie for this K/A exists for the system

Randomly selected New K/A: K 1.08 Process radiation monitoring system.

Page 1 point totals not affected by this change.

RO T2 - G1 Original KA: 400000 Component Cooling Water, A1 Ability to 400000 400000 predict and/or monitor changes in parameters associated with A1.01 A1.03 operating the CCW controls including:

Q# 53 A1.01 CCW flowrate (CFR: 41.5 / 45.5)

The flowrate changes was already tested within question 7.

Rejected due to over sampling.

Randomly selected A1.03 CCW Pressure

Page 1 point totals not affected by this change.

RO T2 - G2 Original KA: 204000 RWCU, A2, Ability to (a) predict the impacts of 204 000 226001 the following on the RWCU System; and (b) based on those A2.01 A2.15 predictions, use procedures to correct, control,or mitigate the consequences of those abnormal conditions or operations:

Q# 60 A2.01 Loss of Component Cooling Water (CFR: 41.5 / 45.6)

This KA was already sampled. Rejected due to over sampling.

Randomly selected New K/A: 226001 RHR/LPCI: CTMT Spray Mode, A2, Ability to (a) predict the impacts of the following on the RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.15 High containment / drywell pressure

Page 1 point totals not affected by this change.

ES-401 Record of Rejected K/As Form ES-401-4

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

RO T2 - G2 Original KA: 259001 Reactor Feedwater, A3, Ability to monitor 259001 259001 automatic operations of the Reactor Feedwater System including:

A3.05 A3.10 A3.05 Feedwater inlet temperature (CFR: 41.7 / 45.7)

Q# 64 Feedwater inlet temperature doesnt have any automatic actions to monitor. No credible tie for this K/A exists for the system.

Randomly Selected new K/A: A3.10 Pump trips

Page 1 point totals not affected by this change.

RO T2 - G2 Original KA: 201002 RMCS, K3, Knowledge of the effect that a loss 201002 271000 or malfunction of the Reactor Manual Control System will have on K3.03 K3.01 following:

Q# 65 K3.03 Ability to process rod block signals(CFR: 41.7 / 45.4)

GGNS does NOT have a RMCS system. No credible tie for this K/A exists for the system.

Randomly Selected new K/A: 271000 Offgas: K3, Knowledge of the effect that a loss or malfunction of the Offgas System will have on the following:

K3.01 Condenser vacuum (CFR: 41.5 / 45.3)

Page 1 point totals not affected by this change.

ES-401 Record of Rejected K/As Form ES-401-4

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

SRO T1-G1 Original KA: 295021 Loss of Shutdown Cooling A2, Ability to 295021 295021 determine and/or interpret the following as they apply to loss of AA2.05 AA2.03 shutdown cooling:

Q# 76 AA2.05 Reactor vessel metal temperature (CFR: 41.10 / 43.5 / 45.13)

This KA was rejected due to low operational value for discriminatory SRO level question, the SRO will control coolant temp to control metal temp. No credible tie for this K/A exists for the system.

Randomly selected AA2.03, Reactor water level.

Page 1 point totals not affected by this change.

SRO T1-G2 Original KA: 295033 High Secondary Containment Area Radiation 295033 295033 Levels / 9, EA2 Ability to determine and/or interpret the following as EA2.03 EA2.01 they apply to high Secondary Containment Area Radiation Levels:

Q# 84 EA2.03 Cause of high area radiation (CFR: 41.10 / 43.5 / 45.13)

This KA was used on previous NRC exam.

Randomly selected EA2.01, Area radiation levels

Page 1 point totals not affected by this change.

SRO T2-G1 Original KA: 203000 RHR/LPCI: Injection Mode 203000 203000 2.2.40 2.2.42 2.2.40 Ability to apply Technical Specifications for a system.

(CFR: 41.10 / 43.2 / 43.5 / 45.3)

Q# 86 This KA is also used in another selection on this exam. Rejected due to similarities with another question.

Randomly selected New K/A: 2.2.42, Ability to recognize system parameters that are entry-level conditions for Technical Specifications

Page 1 point totals not affected by this change.

ES-401 Record of Rejected K/As Form ES-401-4

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

SRO T2-G1 Original KA: 215005 APRM / LPRM A2, Ability to (a) predict the 215005 215005 impacts of the following on the APRM/LPRM SYSTEM; and (b)

A2.07 A2.02 based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or Q# 87 operations: (CFR: 41.5 / 45.6)

A2.07 Recirculation flow channels flow mismatch

This KA was oversampled due to multiple questions dealing with Recirc Flow channels, RO KA was disagreement with flow channels. Rejected due to over sampling

Randomly selected New K/A: 2. 02, Upscale or downscale trips

Page 1 point totals not affected by this change.

ES-301 Administrative Topics Outline Form ES-301-1

Facility: Grand Gulf Nuclear Station Date of Examination: 10/18/2021 Examination Level: RO O Operit Nber: GGNS0 -2021

Ainistriveics Type Dcribe activityo perfmed Code*

Loss of Shutdown Cooling, Time to 200°F Determination, This task has the student to estimate time to reach 200°F following loss of Condt ofpions R M shutdown cooling using 05-1 III-1, Inadequate Decay Heat Removal ONEP.

Modified from 2015 NRC Exam

GJPM-OPS-10-2021AR1, K&A: 2.1.25 (3.9)

Determine Tagging Requirements, The applicant will determine the proper electrical Equipment Control R N isolations for the A Containment Cooler and complete the Tagout Tags Sheet per procedure.

GJPM -OPS-10-2021AR2, K&A: 2.2.13 (4.1),

2.2.41 (3.5)

Determine Primary Containment Water Level, Using EP Attachment 29 and the attached images to obtain RCIC Suction Pressure and Containment Pressure, the applicant will determine Primary Containment Conduct of Operations R M Water Level from the Delta Pressure to Ctmt Level Conversion Table, EP Attachment 29 Table 1.

Modified from 12-2017 NRC Exam

GJPM-OPS-10-2021AR3, K&A: 2.1.25 (3.9) 2.4.21 (4.0)

Determine Exposure, The applicant will determine the correct dose limit due to Alert being declared (5000 mr) and correct dose to be received before the exposure limit is Radiation Control R M exceeded. Also Determines the correct stay time for this task.

Modified from 2-2020 NRC Exam

GJPM -OPS-10-2021AR4, K&A: 2.3.4 (3.2)

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

  • Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom 4 (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes) 0 (N)ew or (M)odified from bank ( 1) 4 (P)revious 2 exams ( 1, randomly selected) 0

ES-301 Administrative Topics Outline Form ES-301-1

JPM

Description:

AR1 - Loss of Shutdown Cooling, Time to 200°F Determination,

A loss of Shutdown cooling has occurred. The CRS has entered Inadequate Decay Heat Removal ONEP and requires the time to 200°F. This task has the applicant to estimate time to reach 200°F following loss of shutdown cooling using 05-1 III-1, Inadequate Decay H eat Removal ONEP.

A similar JPM was given on the 2015 NRC Exam, however, the RPV level was changed and therefore the time also changed, which makes this JPM modified.

AR2 - Determine Tagging Requirements

The applicant will determine the proper isolations for the A Containment Cooler and complete the Tagout Tags Sheet per procedure.

AR3 - Determine Primary Containment Water Level

Using EP Attachment 29 and the attached images to obtain RCIC Suction Pressure and Containment Pressure, the applicant will determine Primary Containment Water Level from the Delta Pressure to Ctmt Level Conversion Table, EP Attachment 29 Table 1.

A similar JPM was given on the 12-2017 NRC Exam, however, the suction pressure for RCIC and Containment pressure is different, which calculates to a different Containment Water Level.

AR4 - Emergency Exposure Limits

The applicant will evaluate a condition involving abnormally high radiological conditions and determine final exposure following the task to maintain below current limit. Then calculate the max stay time before exceeding the limit.

Emergency Response personnel are adm inistratively extended to the Federal limits of 10CFR20 at the declaration of an Alert, Site Area Emergency, or General Emergency. Therefore the limit during an emergency is the federal 10CFR limit

A similar JPM was given on the 2/2020 NRC Exam, however, the radiation level is different and NO task to determine stay time.

ES-301 Administrative Topics Outline Form ES-301-1

Facility: Grand Gulf Nuclear Station Date of Examination: 10/18/2021 Examination Level: RO O Operit Nber: GGNS0 -2021

Ainistriveics Type Dcribe activityo perfmed Code*

Reactor Water Chemistry Required Actions, Applicant will determine the required actions Condt ofpions R N and the time at which those actions are required per ONEP.

GJPM-OPS-10-2021AS1, K&A: 2.1.34 (3.5)

Manual On-Line Risk Assessment, Applicant performs a manual analysis in accordance with 01-S 6 (Risk Assessment of Maintenance Conduct of Operations R, M Activities).

Modified from 12-2017 NRC Exam

GJPM -OPS-10-2021AS2, K&A: 2.1. 25 ( 4.2 )

Perform Review of Completed Surveillance, This task is to review a completed surveillance Equipment Control R, N and complete the Deficiencies and Approval sections. The applicant will find 4 discrepancies with the completed document.

GJPM-OPS-10-2021AS3, K&A: 2.2.12 (4.1)

Direct Actions for High Rad During Fuel Handling. Applicant will evaluate high radiation Radiation Control R, N conditions on the Refueling floor during an outage and using correct procedure determine correct mitigating actions.

GJPM-OPS-10-2021AS4, K&A: 2.3.13 (3.8)

Emergency Event Classification, This task is an event classification in accordance with EALs and is required of all licensed SROs not Emergency Plan R, D qualified as an Emergency Director per the Emergency Preparedness Plan.

NOT used on NRC ILT Exam

GJPM -OPS-10-2021AS5, 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).

ES-301 Administrative Topics Outline Form ES-301-1

  • Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom 5 (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes) 1 (N)ew or (M)odified from bank ( 1) 4 (P)revious 2 exams ( 1, randomly selected) 0

ES-301 Administrative Topics Outline Form ES-301-1

JPM

Description:

AS1 - Reactor Water Chemistry Required Actions,

Applicant will determine the required actions and the time at which those actions are required per ONEP.

New JPM, NOT used on any previous NRC Exam.

AS2 - Manual On-Line Risk Assessment

Applicant performs a manual analysis in accordance with 01-S 6 (Risk Assessment of Maintenance Activities).

A similar JPM was given on the 12-2017 NRC Exam, however, a different system and components are used, which results in a different Risk statement.

Also, this is a manual analysis the previous was using the computer.

A similar JPM was given on the 2015 NRC Exam, however, different components are used

AS3 - Perform Review of Completed Surveillance

This task is to review a completed surveillance and complete the Deficiencies and Approval sections. The applicant will find 4 discrepancies with the completed document

A similar JPM was given on the 12-2017 NRC Exam, however, a different surveillance is used with different discrepancies.

AS4 - Direct Actions for High Rad During Fuel Handling.

This task has the applicant evaluate high radiation conditions on the Refueling floor during an outage and using correct procedure determine correct mitigating actions.

New JPM, NOT used on any previous NRC Exam.

AS5 - Emergency Event Classification,

This task is an event classification in accordance with EALs and is required of all licensed SROs not qualified as an Emergency Director per the Emergency Preparedness Plan.

A similar JPM was given on the 10- 2020 and 12-2017 NRC Exam, however, different plant conditions causes a different EAL and addition of completion of notification form ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

Facility: Grand Gulf Nuclear Station Date of Examination: 10/18/2021 Exam Level: RO S -I SRO -U Operating Test Number: GGNS 10-2021

Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U

System/JPM Title Type Code* Safety Function S1 - CRD Pump Rotation A N S 1 GJPM -OPS-10-2021S1, 201001 A4.01 (3.1); A4.03 (2.9 ); A2.01 (3.3)

S2 - Perform RCIC System Quarterly Pump Operability Verification A N S 2 GJPM -OPS-10-2021S2, 217000: A2.01 (3.8); A4.03 (3.4)

S3 - Perform Actions for Generator Trip WITHOUT Turbine Trip A N S 3 GJPM -OPS-10-2021S3, 241000 A1.13 (2.7), A2.18 (3.5), A4.14 (3.8)

S4 - Start RHR A in Shutdown Cooling GJPM -OPS-10-2021S4, 205000 A2.10 (2.9), A3.01 (3.2), A D L S 4 A4.02 (3.6), A4.01 (3.7), A4.03 (3.6), A4.06 (3.8 ), A4.09 (3.1)

S5 - Manually Initiate Suppression Pool Makeup D EN L S 5 GJPM -OPS-10-2021S5, 223001 A2.11 (3.6), A3.01 (3.4)

S6 - Restore CCW to FPCCU Heat Exchangers N S 8 GJPM -OPS-10-2021S6, 400000: K1.02 (3.2), A1.01 (2.8), A4.01 (3.1)

S7 - Start the Fuel Pool Cooling and Cleanup System N S 9 GJPM -OPS-10-2021S7, 261000 A4.03 (3.0), A4.0 2 (2.7)

S8 - Control Rod Operability Surveillance, with RPS M/G Failure (RO and SROU ONLY) N S 7 GJPM-OPS-10-2021S8, 212000 A2.01 (3.7); 201005 A4,01 (3.7)

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1 - Monitor Div 1/2 Diesel Generator, Emergency Shutdown A E EN M 6 GJPM -OPS-10-2021P1, 264000 A4.04 (3.7)

P2 - Align Fire Water to LPCS per EP Attachment 26 E D R L 8 GJPM -OPS-10-2021P2, 286000 A1.05 (3.2)

P3 - Scram the Reactor and Close MSIVs Per Shutdown From the Remote Shutdown Panel ONEP D E L 7 GJPM-OPS-10-2021P3, 212000 A4.17 (4.1)

  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for R /SRO-I/SRO -U

(A)lternate path 4-6 5 (C)ontrol room (D)irect from bank 9 4 (E)mergency or abnormal in-plant 1 9 (EN)gineered safety feature 1 (control room system) 2 (L)ow-Power/Shutdown 1 4 (N)ew or (M)odified from bank including 1(A) 2 7 (P)revious 2 exams 3 (randomly selected) 0 (R)CA 1 1 (S)imulator

JPM

Description:

S1 - CRD Pump Rotation.

Alternate Path GJPM-OPS 2021S1 202001 A4.01 (3.1); A4.03 (2.9) ; A2.01 (3.2)

  • This task is to rotate CRD pumps per SOI.
  • During the rotation, after B CRD pump is shutdown, 04-1-02-1H13-P601-22A-G3, CRD PMP B OIL PRESS LO, alarm will come in and the pump will fail to trip.
  • The applicant will manually trip the B CRD pump per EN -OP-120.
  • The CRS will enter the ONEP and direct the RO to start the standby pump per ONEP.
  • New JPM, not used previously.

S2 - Perform RCIC System Quarterly Pump Operability Verification Alternate Path GJPM-OPS 2021S2, 217000: A2.01 (3.8); A4.03 (3.4)

  • This task requires the ability to complete the surveillance on the RCIC system.
  • As RCIC is placed in the STANDBY mode the plant will experience a total loss of feedwater with a failure of the HPCS system. The applicant must recognize RCIC failure to auto initiate and attempt to manually initiate by arm and depress the initiation pushbutton. This will also fail to start RCIC, a manual start per Hardcard is required to start the RCIC system and restore Reactor water level.
  • New JPM, not used previously.

S3 - Perform Actions for Generator Trip WITHOUT Turbine Trip Alternate Path GJPM-OPS 2021S2, 241000 A1.13 (2.7), A2.18 (3.5), A4.14 (3.8)

  • This task is to perform actions of the Turbine and Generator Trips ONEP for a Generator trip without Turbine trip.
  • Step A2 directs the candidate to use the Ovation HMI to trip the Main Turbine.
  • New JPM, not used previously.

S4 - Start RHR A in Shutdown Cooling Alternate Path GJPM-OPS 2021S4, 205000 A2.10 (2.9), A3.01 (3.2), A4.02 (3.6), A4.01 (3.7),

A4.03 (3.6), A4.06 (3.8), A4.09 (3.1)

  • During the evolution, Hi room temperature switch will actuate.
  • This requires the recognition of a failure of Group 3 isolation.
  • Manual isolation is required.
  • A similar JPM was used during the 12/2017 NRC Exam, however, that JPMs failure was on the Min Flow Valve opening due to a failed instrument, this JPM develops a high temperature within the RHR A room and fails to isolate and requires manual isolation.

S5 - Manually Initiate Suppression Pool Makeup GJPM-OPS 2021S5, 223001 A2.11 (3.6), A3.01 (3.4)

  • This JPM will evaluate the applicant's ability to manually initiate Suppression Pool Makeup (SPMU) without an extremely low Suppression Pool Level. Using the Hard Card.
  • A similar JPM was used during the 5/2017 NRC Exam, however, that JPM was Alternate Path, this JPM is NOT Alternate Path.

S6 - Restore CCW to FPCCU Heat Exchangers GJPM-OPS 2021S6, 400000: K1.02 (3.2), A1.01 (2.8), A4.01 (3.1)

  • This task is to restore CCW flow to the FPCCU heat exchangers following a CCW low flow isolation per the SOI.
  • New JPM, not used previously.

S7 - Start the Fuel Pool Cooling and Cleanup System GJPM-OPS 2021S7, 261000 A4.03 (3.0), A4.02 (2.7)

  • This task is to startup the FPCCU system after a system trip at power.
  • New JPM, not used previously.

S8 - Control Rod Operability Surveillance with RPS M/G Failure (RO and SROU only)

GJPM-OPS 2021S8, 201005 A4,01 (3.7).

  • This task is to perform the monthly control rod operability surveillance by moving control rods one notch and return to original position. During the test, one RPS Motor Generator (M/G) will fail. There are NO immediate actions and the Applicant is directed to perform steps within an ONEP, therefore, this JPM is NOT Alternate Path.
  • A similar JPM was used during the 10/2020 NRC Exam, however, that JPM was had 2 Control Rods Scram, this JPM has applicant recover from an RPS M/G Failure by being directed by the CRS to transfer to alternate power and reset the 1/2 scram per ONEP.

P1 - Monitor Div 1/2 Diesel Generator, Emergency Shutdown GJPM-OPS 2021P1, 264000 A4.04 (3.7)

  • This task is to monitor operation of the Emergency Diesel Generators and prepare for parallel operation. An alternate path will occur when a low oil pressure alarm is received, and the diesel is required to be shutdown.
  • A similar JPM was used during the 2 /2020 NRC Exam, however, that JPM was on the HPCS Div 3 EDG, this JPM is on Div 1/2 EDG, which is totally different Diesel Generators.

P2 - Align Fire Water to LPCS per EP Attachment 26 GJPM-OPS 2021P2, 286000 A1.05 (3.2)

  • This task simulates routing and connecting fire hoses from hose stations to test connections on ECCS injection piping in the Auxiliary Building.
  • A similar JPM was used during the 2 /2020 NRC Exam, however, this JPM is on a different system within Attachment 26.

P3 - Scram the Reactor and Close MSIVs Per Shutdown From the Remote Shutdown Panel ONEP GJPM-OPS 2021P3, 212000 A4.17 (4.1)

  • This task is to perform the operator actions after evacuating the control room to scram the reactor and close MSIVs from the RPS MG set rooms prior to manning the Remote Shutdown Panels per 05-1 II-1.
  • New JPM, not used previously.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

Facility: Grand Gulf Nuclear Station Date of Examination: 10/18/2021 Exam Level: RO S -I O -U Operit Nber: GGNS 10-2021

Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U

System/JPM Title Type Code* Safety Function S1 - CRD Pump Rotation A N S 1 GJPM -OPS-10-2021S1, 201001 A4.01 (3. 1); A4.03 ( 2.9 ); A2.01 (3. 3)

S2 - Perform RCIC System Quarterly Pump Operability Verification A N S 2 GJPM -OPS-10-2021S2, 217000: A2.01 (3.8); A4.03 (3.4)

S3 - Perform Actions for Generator Trip WITHOUT Turbine Trip A N S 3 GJPM -OPS-10-2021S3, 241000 A1.13 (2.7), A2.18 (3.5), A4.14 (3.8)

S4 - Start RHR A in Shutdown Cooling GJPM -OPS-10-2021S4, 205000 A2.10 (2.9), A3.01 (3.2), A D L S 4 A4.02 (3.6), A4.01 (3.7), A4.03 (3.6), A4.06 (3.8 ), A4.09 (3.1)

S5 - Manually Initiate Suppression Pool Makeup D EN L S 5 GJPM -OPS-10-2021S5, 223001 A2.11 (3.6), A3.01 (3.4)

S6 - Restore CCW to FPCCU Heat Exchangers N S 8 GJPM -OPS-10-2021S6, 400000: K1.02 (3.2), A1.01 (2.8), A4.01 (3.1)

S7 - Start the Fuel Pool Cooling and Cleanup System N S 9 GJPM -OPS-10-2021S7, 261000 A4.03 (3.0), A4.02 (2.7)

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1 - Monitor Div 1/2 Diesel Generator, Emergency Shutdown A E EN M 6 GJPM -OPS-10-2021P1, 264000 A4.04 (3.7)

P2 - Align Fire Water to LPCS per EP Attachment 26 D E R L 8 GJPM -OPS-10-2021P2, 286000 A1.05 (3.2)

P3 - Scram the Reactor and Close MSIVs Per Shutdown From the Remote Shutdown Panel ONEP D E L 7 GJPM-OPS-10-2021P3, 212000 A4.17 (4.1)

  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for R /SRO-I /SRO-U

(A)lternate path 4-6 5 (C)ontrol room (D)irect from bank 8 4 (E)mergency or abnormal in-plant 1 8 (EN)gineered safety feature 1 (control room system) 2 (L)ow-Power/Shutdown 1 4 (N)ew or (M)odified from bank including 1(A) 2 6 (P)revious 2 exams 3 (randomly selected) 0 (R)CA 1 1 (S)imulator

JPM

Description:

S1 - CRD Pump Rotation.

Alternate Path GJPM-OPS 2021S1 202001 A4.01 (3.1); A4.03 (2.9) ; A2.01 (3.2)

  • This task is to rotate CRD pumps per SOI.
  • During the rotation, after B CRD pump is shutdown, 04-1-02-1H13-P601-22A-G3, CRD PMP B OIL PRESS LO, alarm will come in and the pump will fail to trip.
  • The applicant will manually trip the B CRD pump per EN-OP-120.
  • The CRS will enter the ONEP and direct the RO to start the standby pump per ONEP.
  • New JPM, not used previously.

S2 - Perform RCIC System Quarterly Pump Operability Verification Alternate Path GJPM-OPS 2021S2, 217000: A2.01 (3.8); A4.03 (3.4)

  • This task requires the ability to complete the surveillance on the RCIC system.
  • As RCIC is placed in the STANDBY mode the plant will experience a total loss of feedwater with a failure of the HPCS syst em. The applicant must recognize RCIC failure to auto initiate and attempt to manually initiate by arm and depress the initiation pushbutton. This will also fail to start RCIC, a manual start per Hardcard is required to start the RCIC system and restore Reactor water level.
  • New JPM, not used previously.

S3 - Perform Actions for Generator Trip WITHOUT Turbine Trip Alternate Path GJPM-OPS 2021S2, 241000 A1.13 (2.7), A2.18 (3.5), A4.14 (3.8)

  • This task is to perform actions of the Turbine and Generator Trips ONEP for a Generator trip without Turbine trip.
  • Step A2 directs the candidate to use the Ovation HMI to trip the Main Turbine.
  • New JPM, not used previously.

S4 - Start RHR A in Shutdown Cooling Alternate Path GJPM-OPS 2021S4, 205000 A2.10 (2.9), A3.01 (3.2), A4.02 (3.6), A4.01 (3.7),

A4.03 (3.6), A4.06 (3.8), A4.09 (3.1)

  • During the evolution, Hi room temperature switch will actuate.
  • This requires the recognition of a failure of Group 3 isolation.
  • Manual isolation is required.
  • A similar JPM was used during the 12/2017 NRC Exam, however, that JPMs failure was on the Min Flow Valve opening due to a failed instrument, this JPM develops a high temperature within the RHR A room and fails to isolate and requires manual isolation.

S5 - Manually Initiate Suppression Pool Makeup GJPM-OPS 2021S5, 223001 A2.11 (3.6), A3.01 (3.4)

  • This JPM will evaluate the applicant's ability to manually initiate Suppression Pool Makeup (SPMU) without an extremely low Suppression Pool Level. Using the Hard Card.
  • A similar JPM was used during the 5/2017 NRC Exam, however, that JPM was Alternate Path, this JPM is NOT Alternate Path.

S6 - Restore CCW to FPCCU Heat Exchangers GJPM-OPS 2021S6, 400000: K1.02 (3.2), A1.01 (2.8), A4.01 ( 3.1)

  • This task is to restore CCW flow to the FPCCU heat exchangers following a CCW low flow isolation per the SOI.
  • New JPM, not used previously.

S7 - Start the Fuel Pool Cooling and Cleanup System GJPM-OPS 2021S7, 261000 A4.03 (3.0), A4.02 (2.7)

  • This task is to startup the FPCCU system after a system trip at power.
  • New JPM, not used previously.

P1 - Monitor Div 1/2 Diesel Generator, Emergency Shutdown GJPM-OPS 2021P1, 264000 A4.04 (3.7)

  • This task is to monitor operation of the Emer gency Diesel Generators and prepare for parallel operation. An alternate path will occur when a low oil pressure alarm is received, and the diesel is required to be shutdown.
  • A similar JPM was used during the 2 /2020 NRC Exam, however, that JPM was on the HPCS Div 3 EDG, this JPM is on Div 1/2 EDG, which is totally different Diesel Generators. Modified due to facility bank has existing JPM used in Requal.

P2 - Align Fire Water to LPCS per EP Attachment 26 GJPM-OPS 2021P2, 286000 A1.05 (3.2)

  • This task simulates routing and connecting fire hoses from hose stations to test connections on ECCS injection piping in the Auxiliary Building.
  • A similar JPM was used during the 2 /2020 NRC Exam, however, this JPM is on a different system within Attachment 26.

P3 - Scram the Reactor and Close MSIVs Per Shutdown From the Remote Shutdown Panel ONEP GJPM-OPS 2021P3, 212000 A4.17 (4.1)

  • This task is to perform the operator actions after evacuating the control room to scram the reactor and close MSIVs from th e RPS MG set rooms prior to manning the Remote Shutdown Panels per 05-1 II-1.
  • New JPM, not used previously.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

Facility: Grand Gulf Nuclear Station Date of Examination: 10/18/2021 Exam Level: RO S -I O -U Operit Nber: GGNS 10-2021

Control Room Systems:* 8 for RO, 7 for SRO -I, and 2 or 3 for SRO-U

System/JPM Title Type Code* Safety Function S1 - CRD Pump Rotation A N S 1 GJPM -OPS-10-2021S1, 201001 A4.01 (3.1); A4.03 (2.9 ); A2.01 (3.3)

S3 - Perform Actions for Generator Trip WITHOUT Turbine Trip A N S 3 GJPM -OPS-10-2021S3, 241000 A1.13 (2.7), A2.18 (3.5), A4.14 (3.8)

S8 - Control Rod Operability Surveillance, with RPS M/G Failure (RO and SROU ONLY) D S 7 GJPM-OPS-10-2021S8, 212000 A2.01 (3.7); 201005 A4,01 (3.7)

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1 - Monitor Div 1/2 Diesel Generator, Emergency Shutdown A E EN M 6 GJPM -OPS-10-2021P1, 264000 A4.04 (3.7)

P2 - Align Fire Water to LPCS per EP Attachment 26 D E R L 8 GJPM -OPS-10-2021P2, 286000 A1.05 (3.2)

  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for R /SRO-I/SRO -U

(A)lternate path 2-3 2 (C)ontrol room (D)irect from bank 4 2 (E)mergency or abnormal in-plant 1 4 (EN)gineered safety feature 1 (control room system) 1 (L)ow-Power/Shutdown 1 1 (N)ew or (M)odified from bank including 1(A) 1 3 (P)revious 2 exams 2 (randomly selected) 0 (R)CA 1 1 (S)imulator

JPM

Description:

S1 - CRD Pump Rotation.

Alternate Path GJPM-OPS 2021S1 202001 A4.01 (3.1); A4.03 (2.9) ; A2.01 (3.2)

  • This task is to rotate CRD pumps per SOI.
  • During the rotation, after B CRD pump is shutdown, 04-1-02-1H13-P601-22A-G3, CRD PMP B OIL PRESS LO, alarm will come in and the pump will fail to trip.
  • The applicant will manually trip the B CRD pump per EN -OP-120.
  • The CRS will enter the ONEP and direct the RO to start the standby pump per ONEP.
  • New JPM, not used previously.

S3 - Perform Actions for Generator Trip WITHOUT Turbine Trip Alternate Path GJPM-OPS 2021S2, 241000 A1.13 (2.7), A2.18 (3.5), A4.14 (3.8)

  • This task is to perform actions of the Turbine and Generator Trips ONEP for a Generator trip without Turbine trip.
  • Step A2 directs the candidate to use the Ovation HMI to trip the Main Turbine.
  • New JPM, not used previously.

S8 - Control Rod Operability Surveillance with RPS M/G Failure (RO and SROU only)

GJPM-OPS 2021S8, 201005 A4,01 (3.7).

  • This task is to perform the monthly control rod operability surveillance by moving control rods one notch and return to original position. During the test, one RPS Motor Generator (M/G) will fail. There are NO immediate actions and the Applicant is direc ted to perform steps within an ONEP, therefore, this JPM is NOT Alternate Path.
  • A similar JPM was used during the 10/2020 NRC Exam, however, that JPM was had 2 Control Rods Scram, this JPM has applicant recover from an RPS M/G Failure by being directed by the CRS to transfer to alternate power and reset the 1/2 scram per ONEP.

P1 - Monitor Div 1/2 Diesel Generator, Emergency Shutdown GJPM-OPS 2021P1, 264000 A4.04 (3.7)

  • This task is to monitor operation of the Emergency Diesel Generators and prepare for parallel operation. An alternate path will occur when a low oil pressure alarm is received and the diesel is required to be shutdown.

P2 - Align Fire Water to LPCS per EP Attachment 26 GJPM-OPS 2021P2, 286000 A1.05 (3.2)

  • This task simulates routing and connecting fire hoses from hose stations to test connections on ECCS injection piping in the Auxiliary Building.
  • A similar JPM was used during the 2 /2020 NRC Exam, however, this JPM is on a different system within Attachment 26.

Appendix D Required Operator Actions Form ES-D-1 GGNS 10-2021 NRC Scenario 1

Facility: Grand Gulf Nuclear Station Scenario No.: 1 Op-Test No.: GGNS 10 -2021 -1

Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100% power, EOC Inoperable equipment: 115 kV / ESF 12 OOSVC

  • PLCO 3.8.1, ESF 12 tagged OOSVC Turnover:

Event No. Malf. No. Event Type Event Description N,C (BOP) Parallel / Load Div 3 DG. 17AC lockout r21139g A (Crew) after loaded.

1 LCO 3.8.7, condition D TS (CRS) LCO 3.5.1, condition B

2 fw232i C (BOP,ATC ) Low Pressure Feedwater Heater 3C tube fw270c A (Crew) leak / failure to auto isolate

3 O/R C (ATC) RWCU high filter strainer DP requiring p680_11a_d_5 removal from service.

C (BOP) ESF 21 lockout 4 r21180 A (Crew) LCO 3.8.1, condition A TS (CRS) ov_194 5 ov_195 M (Crew) Both EHC control fluid pumps trip ov_274 ov_275 ATWS > 5%

CT-1, Terminates and prevents injection 6 c11164 M (Crew) to lower RPV level to < -70 WR CT-2, Terminates and prevents feedwater injection to lower Rx power CT-3, Inserts control rods 7 c41263 C (BOP) Standby Liquid Control pipe rupture

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

Revision 0 2/24/2021 1 Appendix D Required Operator Actions Form ES-D-1 GGNS 10-2021 NRC Scenario 1

Quantitative Attributes Table

Attribute E3-301-4 Actual Description Target Malfunctions after 1-2 1

  • 17ACko
  • Feedwater heater 3C tube leak with Abnormal Events 2-4 4 failure to auto isolate
  • RWCU F/D strainer high DP
  • ESF 21 lkout Majransies 1 -2 2
  • Loss of EHC pumps

EOP entries requiring 1-2 2

  • EP-3, Containment Control

Entry into a contingency EOP 1 1

  • CT-1, Terminate and prevent feedwater injection to lower RPV level to <-70 WR
  • CT-2, when suppression pool temperature exceeds 110°F, terminate Preidentified critical and prevent feedwater injection to lower tasks 2-3 3 Rx power

Revision 0 2/24/2021 2 Appendix D Required Operator Actions Form ES-D-1 GGNS 10-2021 NRC Scenario 1

Initial Conditions:

  • Plant is operating at 100% power, end of cycle

Inoperable Equipment:

Planned activities:

  • Parallel and load Div 3 DG to 17AC and load to 1800 kW

Scenario Notes:

  • This scenario is a NEW Scenario
  • Validation Time: 65 minutes

Revision 0 2/24/2021 3 Appendix D Required Operator Actions Form ES-D-1 GGNS 10-2021 NRC Scenario 1

SCENARIO ACTIVITIES:

Event 1 - Parallel and load Div 3 DG to 17AC (normal evolution / auto trigger)

When the crew accepts the shift, the Div 3 DG will be running and ready to be paralleled and loaded.

Using 04-1 1P81-1, High Pressure Core Spray Diesel Generator (step 5.2), the BOP will parallel Div 3 DG to 17AC and commence raising load to achieve1800 kW.

When Div 3 DG is loaded to >800 kW, a 17AC lockout will occur. Annunciator P601-16A-B1, 4.16KV BUS 17AC INCM FDRS LOCKOUT TRIP will alarm. 05 02-I-4, Loss of ESF AC POWER ONEP should be entered. The BOP should recognize Div 3 DG is running without cooling water and emergency trip the Div 3 DG and place it in the maintenance mode per the hardcard.

152-1702, HPCS pump breaker should be racked out.

Tech Specs should be referred to and the following identified:

Event 2 - Low pressure feedwater heater tube leak with failure to auto isolate (Triggered by Lead Examiner)

FW heater 3C level will rise and will result in the following annunciators to alarm.

  • P680-2A-B8, FW HTR 3C LVL HI
  • P870-6A-B3, FW HTR 3C LVL HI -HI The ARIs should be referred to and the BOP should identify the C heater string inlet and outlet isolation valves on the P870 panel failed to automatically close and should close them.

05-1 V-5, Loss of Feedwater Heating ONEP should be entered and core flow should be lowered to obtain < 90% reactor power. Since this is an EOC scenario, control rods will not have to be inserted since operation is already below the 100% load line.

When core flow is lowered, 05-1 III-3, Reduction in Recirculation System Flow Rate ONEP should be entered to ensure the plant parameters have stabilized. Power / Flow should be determined and plotted on Attachment 2 Power / Flow map. The OPRM armed region will not be entered.

The tube leak / isolation will not result in a > 10°F feedwater temperature reduction.

Event 3 - RWCU filter demin outlet strainer clogging (Triggered by Lead Examiner)

Annunciator P680- 11A-D5, RWCU FILTR DMIN CONT TROUBLE will alarm. The ARI should be referred to and an NLO dispatched to G36-P002, RWCU Filter Demin Panel to investigate the alarm. The NLO will report the RWCU A filter demin strainer DP is indicating 5 psid and slowly rising. Using the ARI, the ATC should identify the A filter demin should be removed from service and the strainer should be flushed.

04-1 G33-1, Reactor Water Cleanup SOI (step 4.6) should be referred to and the filter should be removed from service for strainer backflush. When directed, the booth operator will remove the filter over a 90 second period. The ATC will be required to manipulate 1G33-F044, RWCU FLTR DMIN BYP VLV on 1H13-P680 to maintain 450- 500 gpm system flow as F/D flow is lowered.

Once the filter/demin is removed from service, no further action will be performed.

Revision 0 2/24/2021 4 Appendix D Required Operator Actions Form ES-D-1 GGNS 10-2021 NRC Scenario 1

Event 4 - ESF 21 lockout (Triggered by Lead Examiner)

(The 2/2020 scenario 2 exam contained an ESF 21 lockout that occurred after the major events. It too was an ATWS scenario. In that scenario, the focus was to challenge the crew to re -terminate and prevent low pressure ECCS and to restore CRD pump B to service for driving and scrammin g control rods. Also, the isolations had already occurred due to the lowered RPV level. In this scenario, multiple other actions are required. Also, this event will also exercise an LCO that has not been previously challenged.)

ESF Transformer 21 will trip. The Div 2 DG will auto start and re -energize bus 16AB.

05-1 I-4, Loss of ESF AC Power should be entered. The following should be performed:

  • Power to Division 2 Drywell chiller skid should be reclosed
  • Control power to the switchyard should be reclosed
  • Standby Gas Treatment should be initiated
  • NSSSS logic reset.

The CRS should also enter 05-1 III-5, Automatic Isolations ONEP due to multiple primary and secondary containment isolation valves failing closed due to loss of power. Using Attachment I of the ONEP, the isolation valves should be restored except for those systems where SOI restoration is required.

RWCU will trip due to the momentary loss of power to leak detection temperature switches. No action will be performed.

Fuel pool cooling and cleanup will trip on low flow due to one of the filter demin inlet valves closing.

05-1 III-1, Inadequate Decay Heat Removal ONEP should be entered and spent fuel pool and upper containment pool temperatures should be monitored. No further action will be performed.

An RC&IS inop will occur due to the power loss. When I&C is notified, it will be reset by the booth operator.

2 offsite sources are now inop (1 less than the minimum required).

Tech Specs should be referred to and the following LCOs identified:

Performance of SR 3.8.1.1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required for operable offsite circuits. Once identified, the CRS will be notified that the work control center will dispatch an operator to perform this surveillance. No action will be performed.

Event 5 - Trip of both EHC pumps (Triggered by Lead Examiner)

EHC pump A will trip. 30 seconds later, EHC pump B will trip. With no EHC pumps available the main turbine will trip on low trip oil pressure and the reactor will automatically scram. When the scram occurs the ATC should observe a hydraulic block ATWS with Rx power > 5%.

Revision 0 2/24/2021 5 Appendix D Required Operator Actions Form ES-D-1 GGNS 10-2021 NRC Scenario 1

Event 6 - ATWS (Automatically triggered)

The ATC should perform immediate actions for an ATWS > 5% power:

  • Initiation of ARI/RPT
  • Termination / prevention of HPCS and LP ECCS (Critical task-2)
  • Verification of SLC injection
  • Placing Suppression pool cooling into service The BOP should perform immediate actions for an ATWS > 5% power:
  • Terminate / prevent feedwater and align for startup level control when directed and then control RPV level within a band of - 70 to - 130 WR. (CT-1)

Note: this scenario is not designed for RPV pressure to be lowered. Therefore, terminating and preventing of low pressure ECCS is not critical. If, for whatever reason, RPV pressure is lowered below the shutoff head of low pressure ECCS, then the terminate/prevent for low pressure ECCS becomes a critical task.

This scenario is designed to combat the ATWS and perform a second terminate/prevent due to containment challenge. The 110°F terminate / prevent has not been challenged on the previous 2 sets of exams.

With the loss of the EHC pumps, the main turbine and bypass stop/control valves will close. Even though the bypass valves close, steam will still be available to the reactor feed pumps and other systems in the turbine building. This will also make some steam line drains available (minimal) to minimize energy input into suppression pool.

When SLC is initiated it will fail to inject. A pipe rupture should be identified and reported.

Reactor pressure will rise and will result in SRVs cycling to maintain RPV pressure. As a result, suppression pool temperature will rise and will eventually exceed 110°F. With reactor power >5%

concurrent with s uppression pool temperature > 110°F, a second terminate / prevent is required to establish a new RPV level band of - 167 to -191 CFZ to lower reactor power. (CT-2)

Failure to direct EP attachment 8 installation (defeating level 1 MSIV isolation) will result in an MSIV isolation when the second terminate/prevent is executed and adding additional energy to the suppression pool from SRVs. An MSIV isolation will necessitate an RPV pressure reduction to establish a pressure band of 450- 600 psig to allow for condensate booster pump injection. This is a voidable.

The CRS should direct installation of the following EP attachments. The booth operator will install these attachments when directed:

  • Attachment 8, defeats MSIV isolation on RPV level 1 (-150.3 WR)
  • Attachment 18, defeats ARI/RPT logic
  • Attachment 19, defeats RPS logic
  • Attachment 20, defeats RC&IS interlocks
  • Attachment 26, alternate SLC injection Once EP attachments 18, 19, 20 are installed, the CRS should direct rod insertion by scramming and driving rods (CT-3). It takes 22 minutes from the scram to exceed HCTL limit for this scenario. Once directed, it takes 10 minutes for installation of EP attachments to insert control rods.

After the second terminate/prevent is completed and control rods are being inserted, this scenario can be terminated.

Termination:

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Revision 0 2/24/2021 6 Appendix D Required Operator Actions Form ES-D-1 GGNS 10-2021 NRC Scenario 1

(CT-1) During failure to scram conditions with power > 5%,

  • terminate feedwater injection to lower RPV level to below - 70 wide range. The Critical Task level band of -70 to - 130 WR should be initially established.
  • maintains control of RPV level such that an automatic MSIV isolation due to low RPV level (-150 WR) does not occur.

Event 6 Regarding lowering level below -70 wide range, to prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities. RPV water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the 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 Safety upper bound of the RPV water level control band. This water level is sufficiently low Significance 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.

With reactor power > 5%, an MSIV isolation (when avoidable) would unnecessarily add heat to the suppression pool at a rate greater than that capable of being removed by RHR A and B. This could result in exceeding the Heat Capacity Temperature limit and subsequent loss of the primary containment due to over pressurization.

A scram is initiated (either automatically or manually) and numerous control rods Cueing indicate beyond position 02 and reactor power is > 5% on panel P680 indications and SPDS and RPV level is > -70 wide range on SPDS and PDS.

To terminate/prevent Feedwater:

Reduces feedwater discharge pressure to below RPV pressure.

Closes startup level control valve.

Measurable Closes N21-F009A, FW HTR 6A OUTL VLV.

Performance Closes N21-F009B, FW HTR 6B OUTL VLV.

Indicators Closes N21-F040, FW SU BYP VLV.

To control RPV level once in band:

Selects Speed Auto on RFP used and raises RFP discharge pressure > reactor pressure.

Opens startup level control valve as required to control RPV level.

To terminate/prevent Feedwater:

1C34-LK-R600, Feedwater level master controller output indicates - 5.00%.

1C34-LK-R602, Startup level controller output indicates - 5.00%.

N21-F009A/B, and N21-F040 indicate green light on and red light out.

Performance Feedwater flow to the RPV indicates 0.

Feedback To control RPV level once in band:

Speed Auto pushbutton is backlit.

RFP discharge pressure indicates > reactor pressure on indicator N21-R612A or B.

Raises output of 1C34-LK-R602, Startup level controller output indicates > 0.

Observes feedwater flow to the RPV.

Revision 0 2/24/2021 7 Appendix D Required Operator Actions Form ES-D-1 GGNS 10-2021 NRC Scenario 1

Critical Task (CT-2) All injection except boron, CRD, and RCIC are terminated and prevented when criteria specified in EP2A step L5 are met before Suppression Pool temperature reaches 120°F.

Event 6 Safety 02-S-01-40, EP Technical Bases, Attachment V Step L8 Significance A combination of high reactor power, high suppression pool temperature, and an open SRV or high drywell pressure indicates that heat is being added to the suppression pool faster than it is being removed by available suppression pool cooling. This condition could ultimately result in overpressurization and loss of primary containment integrity.

Loss of containment integrity could, in turn, lead to a loss of adequate core cooling and uncontrolled release of radioactivity to the environment. RPV water level is therefore lowered to reduce reactor power while efforts to shut down the reactor continue.

A suppression pool temperature of 110°F is the temperature at which Technical Specifications requires a reactor scram and the most limiting value of the Boron Injection Initiation Temperature. A temperature above this value indicates that suppression pool heatup is occurring. If power is not reduced, emergency depressurization may be required before the reactor can be shut down with boron.

120°F Suppression pool temperature is used due to it is the initial temperature used on the HCTL graph and Tech Specs Bases 3.6.2.1 states Continued addition of heat to the suppression pool with pool temperature >120°F could result in exceeding the design basis maximum allowable values for primary containment temperature or pressure.

Cueing Rx power indicating > 5% on P680 and SPDS.

Suppression pool temperature indicating > 110°F on SPDS.

Performance Operator verifies Startup Level Controller in MANUAL on panel P680 and lowers output Indicator to -5.00% to stop feedwater injection until RPV water level lowers below - 70 wide range.

Feedwater flow indicates 0 P680/SPDS.

When RPV level is < - 167, restores feedwater flow and controls level -167 to - 191 CFZ.

Performance Startup Level Controller output indicates -5.00% on panel P680.

Feedback Observes Feedwater flow indicates 0 P680/SPDS.

Observes Feedwater flow restored on P680/SPDS.

Observes RPV level within band of -167 to -191 CFZ on PDS/SPDS.

Revision 0 2/24/2021 8 Appendix D Required Operator Actions Form ES-D-1 GGNS 10-2021 NRC Scenario 1

Critical Task CT -3, When control rods fail to scram, crew begins to insert all control rods to position 02 or beyond before exceeding HCTL limit.

Event 6 Safety Failure to effect shutdown of the reactor when a RPS setpoint has been exceeded Significance would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. The crew is authorized by Conduct of Operations to take mitigating actions when automatic safety systems fail to perform their intended function. Action to shut down the reactor is required when RPS and control rod drive systems fail IAW EP-2A. Failure to insert control rods in a timely manner (once EP attachments are inserted) will result in the continual addition of energy to the suppression pool resulting in exceeding the HCTL limit. Prior to exceeding the HCTL limit, it is expected the crew will lower RPV pressure in an attempt to gain margin to the HCTL limit. If Control Rods are not inserted in a timely manner and HCTL is challenged, lowering Reactor pressure to prevent exceeding HCTL limit could become a new Critical Task. Lowering RPV pr essure will result in lowering moderator temperature and adding positive reactivity to the reactor and end up with a higher reactor power once conditions stabilize. The insertion of control rods as soon as possible will minimize the challenge to the HCTL limit.

Cueing A scram is initiated (either automatically or manually) and numerous control rods indicate beyond position 02.

Scram is reset by turning both RPS keylock handswitches (in at least one division) to reset on P680.

Drive water DP is raised to maximum value by closing C11-F003, CRD DRIVE WTR Measurable PRESS CONT VLV on P601.

Performance With scram reset, drives control rods in by selecting control rods and depressing IN Indicators TIMER SKIP pushbutton.

With scram reset, depresses and holds all rods pushbutton and then depresses manual scram pushbutton(s).

At least one division of RPS white lights on P680 panel are illuminated.

Performance On P601, C11-F003 indicates green light only and C11-R602, CRD DRIVE WTR DP Feedback indicates upscale.

Observes control rod insertion.

Observes reactor power lowering.

  • 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). An unintentional or unnecessary RPS or ESF actuation may result in the creation of a post-scenario Critical Task, if that actuation results in a significant plant degradation or significantly alters a mitigation strategy.

Revision 0 2/24/2021 9 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 2

Facility: Grand Gulf Nuclear Station Scenario No.: 2 Op-Test No.: GGNS 10-2021-2

Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 90% power, BOC Inoperable equipment:

  • LPCS due to test return valve E21-F012 open. LCO 3.5.1, Condition A.

Turnover:

  • A sequence exchange has been completed and Rx power is being restored. Presently Rx power is 90% for RE to verify thermal limits. They will inform Operations when complete.
  • LPCS pump is operating IAW 06 -OP-1E21-Q-0006, LPCS QUARTERLY FUNCTIONAL TEST to obtain vibration data only for system engineering.

Event No. Malf. No. Event Type Event Description 1 O/R p60 1_21a_a_7 C (BOP) LPCS pump overload alarm A (Crew) 2 rm157b TS (CRS) Offgas Pre-Treat rad mon downscale ODCM 6.3.10 3 fw123b C (ATC) RFPT B trip / Recirc FCV A fails to rr217a A (Crew) automatically runback.

RPS / NSSSS level transmitter LOP 4 ltb21n080b_e I,TS (CRS) LCO 3.3.1.1, Cond A LCO 3.3.6.1, Cond A TR 3.1.5, Cond A & B 5 O/R p680_10a_e13 C (ATC,BOP) Circulating water pump high vibration /

n71114b A (Crew) trip 6 r21139e M (CREW) 15AA lockout 7 fw070b M (CREW) Feedwater rupture in turbine building 8 e22052 C (BOP) HPCS pump trip 9 rr063a M (CREW) Recirc leak RHR B/C injection valves fail to auto e12k023b open 10 e12k023c C (BOP) CT-1, E.D. prior to - 191 CFZ CT-2, open RHR B/C injection valve(s) prior to 206 psig RPV pressure (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

Revision 0 2/28/2021 1 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 2

Quantitative Attributes Table Attribute E3-301-4 Actual Description Target Malfunctions after

  • RHR B/C injection valves fail to auto open
  • LPCS overload alarm Abnormal Events 2-4 3
  • RFPT trip / Recirc FCV fails to runback
  • Recirc leak EOP entries requiring 1-2 2
  • EP-2, RPV Control substantive action
  • EP-3, Containment Control

Entry into a

  • EP-2, Alternate level control contingency EOP with 1 2
  • EP-2, Emergency depressurization substantive actions
  • CT-1, Emergency depressurize the RPV before RPV water level drops to the Minimum Steam Cooling RPV Water Preidentified critical 2-3 2 Level (-191).

tasks

  • CT-2, after emergency depressurization with RPV level < - 191 CFZ, aligns RHR B or C for injection prior to RPV pressure lowering below MSCP of 206 psig.

Revision 0 2/28/2021 2 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 2

Initial Conditions:

  • Plant is operating at 90% power, BOC following a sequence exchange
  • LPCS pump is operating IAW 06-OP-1E21-Q-0006, LPCS QUARTERLY FUNCTIONAL TEST to obtain vibration data only for system engineering.

Inoperable Equipment:

  • LPCS due to E12-F012, LPCS TEST RTN TO SUPP POOL being open. LCO 3.5.1, Condition A is active

Planned activities:

  • When notified by reactor engineering, continue power ascension

Scenario Notes:

  • New Scenario
  • Validation Time: 6 0 minutes

Revision 0 2/28/2021 3 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 2 SCENARIO ACTIVITIES:

Event 1 - LPCS pump overload alarm (Triggered by Lead Examiner)

With LPCS running through its test return to the suppression pool to obtain vibration data only. The test return valve, E21-F012, being open requires LPCS to be declared INOP. LCO 3.5.1, Condition A is active.

Annunciator P601-21A-A7, LPCS PMP OVERLD will alarm. After verifying flow is < 9,100 gpm, flow should be lowered by throttling closed E21F012, LPCS TEST RTN TO SUPP POOL while current is being monitored at the pump breaker. The alarm is received at 320 amps and a pump trip occurs at 480 amps.

If observed, AC current will be slowly rising on ESF 11 feed to 15AA (1R21-R617A, P864 -1B). If requested, the booth will provide feedback that current is approximately 350 amps and slowly rising. The ARI directs securing the LPCS pump if unable to reduce motor amps. The booth will not report a lowering of motor amps until the pump is secured. When the pump is secured, the overload alarm will clear.

Event 2 - Offgas pre-treatment radiation monitor downscale (Triggered by Lead Examiner)

This is a Tech Spec exercise only.

When triggered, annunciator P601 -19A-F7, OG PRE-TREAT RAD MON DNSC will alarm. The ARI directs checking the main steam line radiation monitors and Offgas post-treatment radiation monitors.

Using V-Panel, selects the following and reports all radiation monitors are indicating normal values:

Panel Instrument Name P670 D17-RITS-K610B MSL radiation monitor B P672 D17-RITS-K610D MSL radiation monitor D P669 D17-RITS-K610A MSL radiation monitor A P671 D17-RITS-K610C MSL radiation monitor C P600 D17-RR-R601 OFFGAS POSTTREAT radiation recorder

The CRS should refer to Tech Specs / ODCM 6.3.10:

A. One or more required channels nonfunctional.

A.1 Suspend release of radioactive effluent via affected pathway immediately.

OR A.2 Enter the Condition referenced in Table 6.3.10-1 for the channel immediately. (F)

F. As required by Required Action A.2 and referenced in Table 6.3.10-1 F.1 Verify the Offgas system is not bypassed immediately.

F.2 Verify by administrative means that the charcoal vault radiation monitor and the main steam line radiation monitors are FUNCTIONAL.

AND F.3.1 Take grab samples and analyze within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

OR F.3.2 Verify an installed portable radiation monitor on the Offgas pre-treatment line is capable of detecting a 50% change in radiation level and record the value.

AND F.4 Restore the channel to FUNCTIONAL within 30 days.

The Offgas pre-treatment radiation monitor has a backup radiation monitor that will satisfy F.3.2.

Revision 0 2/28/2021 4 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 2

Event 3 - RFPT B trip / A Recirc FCV fails to auto runback (Triggered by Lead Examiner)

When triggered, annunciator P680- 2A-A12, RFPT B TRIP will alarm and reactor feed pump "B " will trip.

With the recirc pumps operating in fast speed combined with only 1 RFPT in service and RPV level lowering to < 32NR, a Recirc flow control runback should occur to lower power to within the capability of 1 RFPT ( 20% valve position). The recirc FCV runback will occur on the B loop only. The ATC should observe the A FCV did not respond and lower A loop flow to within TS limit for loop mismatch. If no action is taken, RPV level will lower to 20 NR and will then start slowly trending up.

05-1 V-7, Feedwater System Malfunctions ONEP should be entered that ensures the runback has occurred.

05-1 III-3, Reduction in Recirculation System Flow Rate ONEP should be entered to ensure:

  • Recirc loop flow mismatch is within limits
  • Perform plot on power/flow map
  • THI watch established The A RFPT will speed up to maintain RPV level and will result in receipt of alarm P680- 2A-E2, RFP A VIBR HI. An NLO should be dispatched to investigate. When dispatched, the booth operator will inform the control room the vibration levels are below alarm values and will reset the alarm.

Event 4 - B21N080B transmitter loss of power (Triggered by Lead Examiner)

This is a Tech Spec exercise only.

B21N080B transmitter feeds trip units that feed into the RPS level 3 and level 8 trip functions. This transmitter also inputs into the RHR shutdown cooling isolation for the low RPV level 3 function. When triggered, numerous alarms will be received along with a division 2 half scram. The half scram will not be capable of being reset. When I&C is asked to investigate, the booth will inform the CRS the following P692 trip unit indications:

  • B21-N680B is indicating downscale, tripped and a gross fail
  • B21-N683B (slave trip unit) will indicate tripped with a gross fail 17-S 5, Technical Specification Instrumentation Loop Logic should be referenced to identify the effects.

Tech Specs should be referenced and the following LCOs identified:

3.3.1.1 RPS instrumentation A.1 Place channel in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 3.3.6.1 Primary Containment isolation instrumentation A.1 Place channel in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> TR3.1.5 Control Rod Scram Accumulator Alarms A.1 Verify the affected accumulator pressure 1520 psig once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B.1 Verify the affected accumulator water drained once per 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Revision 0 2/28/2021 5 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 2

Event 5 - Circulating Water pump trip (Triggered by Lead Examiner)

This scenario is performed at a reduced power due to the certainty of a scram if performed at 100%

power. The intent is to provide reasonable time to ensure the ONEP can be properly exercised for the given condition. The success path s are to either align the operating circulating water pump in a single pump/dual train alignment or insert control rods. Once aligned, the operating circulating water pump will be providing flow to all circulating water tubes in the condensers. To transfer to single pump/dual train, the auxiliary cooling tower must be secured, otherwise a pump runout condition will continue. The booth operator will remove the auxiliary cooling tower from service when directed. When the pump trip occurs, 05-1 V-8, Loss of Condenser Vacuum ONEP should be entered. The ONEP directs Rx power to be lowered to 54% by lowering core flow and inserting control rods. The tripping of the RFPT with the Recirc FCV runback earlier resulted in core flow being < 70 mlbm/hr. Many control rods will be required to be inserted individually to lower power to 54%. Therefore, once in single pump/dual train configuration and control rods are being inserted, the next event can be directed (15AA lockout).

Event 6 - 15AA bus lockout (Triggered by Lead Examiner)

The 15AA bus lockout will provide 2 things:

(1) P53F001, INST AIR SPLY HDR TO CTMT will lose power and close.

(2) Power will not be available to Div 1 ECCS pumps (LPCS/RHR A).

Without instrument air available to the containment, control rods will start drifting in after 2 minutes due to scram valves opening. A manual scram should be inserted.

05-1 I-4, Loss of ESF AC Power / Loss of Offsite Power / Station Blackout ONEP should be entered which provides the following guidance:

RETAINMENT OVERRIDE CONDITION ACTION I. ESF Bus 15AA CANNOT be RE-ENERGIZED PLACE Reactor Mode Switch in SHUTDOWN AND Reactor Mode Switch in RUN OR STARTUP II. ESF Bus 15AA is in LOCKOUT DO NOT RE -ENERGIZE Bus UNTIL condition is OR FAULTED Condition repaired OR isolated

05-1 V-9, Loss of Instrument Air ONEP may also be entered and provides the following guidance:

CONDITION ACTION I. DEGRADED Instrument Air Header PLACE Reactor Mode Switch in SHUTDOWN Pressure AND ANY Control Rod Drift

Revision 0 2/28/2021 6 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 2

Events 7, 8, 9 - Feedwater break in Turb Bldg / HPCS pump trip / Recirc suction rupture (Automatically triggered)

(The 10/2020 scenario 1 contained malfunctions that prevented both an auto and manual ECCS initiation. Without an initiation signal present, the RHR B/C pumps do not auto start and the E12F042B/C injection valves do not receive an automatic open signal. In this scenario, an initiation signal is received and pumps will start, but the injection valves E12F042B&C will fail to auto open when RPV pressure lowers to < 476 psig.)

When the Rx scram occurs, the following will automatically occur:

  • 5 seconds after the scram, a feedwater line break in the turbine building will occur
  • 2 minutes after the scram, a recirc leak will occur
  • HPCS pump will trip on start signal When feedwater is lost, RCIC should be initiated for injection. RCIC would normally be capable o f maintaining RPV level provided no LOCA exists. However, a LOCA will occur and initiation signals will occur on ECCS systems. The HPCS pump will trip on start signal. The leak rate will be greater than the makeup capability of RCIC, CRD B, and SLC B. The lockout on 15AA will also result in loss of CRD pump A and SLC pump A. The isolation of P53F001 will prevent maximizing CRD flow due to loss of air to the CRD FCV (fails closed). RPV level will lower below - 160 WR where an emergency depressurization is required. Failure to commence opening of ADS/SRVs prior to RPV level lowering to <-191 CFZ will result in a failed critical task (CT-1).

Event 10 - RHR B/C injection valves fail to automatically open (Automatically triggered)

During emergency depressurization, RPV pressure will lower. The RHR B and C injection valves (E12F042B&C) will fail to automatically open when RPV pressure lowers to < 476 psig. When RPV pressure is < 476 psig, the BOP should observe neither valve opens an d opens both valves using their respective control room handswitches. Failure to open either E12F042B or F042C prior to RPV pressure lowering below the MSCP pressure of 206 psig (for 8 ADS/SRVs) concurrent with RPV level < - 191 CFZ will result in a failed critical task (CT -2).

After RPV level has been restored and stabilized, EP-3 should be entered and RHR B should be placed into suppression pool cooling using its hard card. Also, the division 2 Ctmt/DW hydrogen igniters should be turned on.

Termination:

When directed by Lead Evaluator:

Take the simulator to Freeze and turn horns off.

Stop and save the SBT report and any other recording devices.

Instruct the crew to not erase any markings or talk about the scenario until after follow -up questions are asked.

Revision 0 2/28/2021 7 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 2

(CT-1) 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 Critical Task 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 7,8,9 The MSCWL is the lowest RPV water level at which the covered portion of the reactor Safety core will generate sufficient steam to preclude any clad temperature in the uncovered Significance portion of the core from exceeding 1500°F. When water level decreases below MSCWL with injection, clad temperatures may exceed 1500°F.

Wide range indication (SPDS and PDS) falls to - 160 and lowering trend continues, Cueing 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 RPV level falls below - 191 CFZ.

Measurable Operator manually places handswitches for at least 7 SRVs to open using Performance handswitches on P601.

Indicators Performance Red light indication on at least 7 SRVs on P601.

Feedback RPV pressure lowering.

Revision 0 2/28/2021 8 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 2

Critical Task (CT-2) During emergency depressurization and concurrent with RPV level below

-191 CFZ, crew manually aligns at least 1 ECCS subsystem for injection prior to RPV pressure lowering below the MSCP pressure of 206 psig (with 8 SRVs open).

Event 10 Safety The pressures in Table P-3 are the Minimum Steam Cooling Pressures, defined to Significance be the lowest RPV pressures at which steam flow through open SRVs is sufficient to preclude the clad temperature of the hottest fuel rod from exceeding 1500°F even if the reactor core is not completely covered.

Cueing Absence of open indication for E12F042B or C with reactor pressure < 476 psig on P601-17B and reactor level below - 191 CFZ as observed on SPDS.

Annunciator P601-17A-C1, LPCI B/C INJ VLV RPV PRESS LO illuminated.

Measurable Wide range RPV pressure recorder B21-R623B indicate < 476 psig on P601-17B.

Performance E12-F042B/C, RHR B/C INJ SHUTOFF VLV both indicate green light only on P601-Indicators 17C.

Handswitches for E12-F042B/C are taken to the open position.

Performance Observes red lights for E12-F042B/C illuminated.

Feedback Observes P601/17B RHR PMP B/C DISCH FLO indicators rising after RPV pressure lowers below shutoff head of RHR B/C pumps ( 300 psig).

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Revision 0 2/28/2021 9 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 3

Facility: Grand Gulf Nuclear Station Scenario No.: 3 Op-Test No.: GGNS 10-2021-3

Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 71 % power, E OC Inoperable equipment:

  • ADS valve B21F051C is inoperative with fuses pulled
  • RHR A, LPCI mode due to operating in suppression pool cooling
  • LCO 3.5.1, Cond F active, (1 ADS valve inop concurrent with 1 LPCI subsystem inop)
  • CRD pump A Turnover:
  • An earthquake has occurred. A controlled shutdown is in progress
  • RHR A is in Suppression pool cooling. Suppression pool temp is 87°F due to ADS valve B21F051C lifting last shift.

Event No. Malf. No. Event Type Event Description z021021_04_41 C (ATC) Control rod drift in / sticks 1 z022022_04_41 A (Crew) LCO 3.1.3, Cond C TS (CRS)

C (BOP) RCIC room steam leak remaining <

2 e51190 A (Crew) isolation setpoint.

TS (CRS) LCO 3.5.3, Cond A 3 fw125d I, A (ATC) FWLC Steam flow signal failure C (BOP) Supp pool leak RHR A (isolable) 4 ct218a A (Crew) LCO 3.6.1.7, Cond A (Ctmt Spray)

TS (CRS) LCO 3.6.1.8, Cond A (FWLC)

LCO 3.6.2.3, Cond A (Supp Pool Clg) 5 ttt41n048_d M (CREW) Steam leak in MST / fail to auto isolate rf Att 9 CT-1, isolate the MSIVs 6 rr193a M (CREW) Recirc pump A discharge leak 7 r21135 M (CREW) LOP O/R C (ATC) Div 3 DG auto start failure 8 Div 3 DG auto start A (Crew) CT-2, Start Div 3 DG failure (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

Revision 0 3/11/2021 1 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 3

Quantitative Attributes Table Attribute E3-301-4 Actual Description Target Malfunctions after 1-2 1

  • Div 3 DG fails to auto start EOP entry
  • RCIC leak remaining < isolation setpt
  • C34 steam flow signal failure
  • Supp pool leak RHR A
  • Recirc pump A discharge leak
  • EP-2, RPV Control substantive action 1-2 3
  • EP-3, Containment Control
  • EP-4, Auxiliary Building Control Entry into a contingency EOP with 1 1
  • EP-2, Alternate level control substantive actions

Preidentified critical 2-3 2

  • CT-1, Isolate MSIVs within 15 minutes tasks
  • CT-2, Start Div 3 DG before -191 CFZ

Revision 0 3/11/2021 2 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 3

Initial Conditions:

  • Plant is operating at 71% power, EOC
  • Plant shutdown in progress. Awaiting Operations management decision to continue with shutdown.
  • RHR A operating in suppression pool cooling

Inoperable Equipment:

  • ADS valve B21F051C is inoperative with fuses pulled
  • RHR A, LPCI mode due to operating in suppression pool cooling
  • LCO 3.1.5, Cond F active, (1 ADS valve inop concurrent with 1 LPCI subsystem inop)
  • CRD pump A tagged OOSVC

Planned activities:

  • When directed by Operations Management, continue plant shutdown per IOI -2.

Scenario Notes:

  • New Scenario
  • Validation Time: 65 minutes

Revision 0 3/11/2021 3 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 3 SCENARIO ACTIVITIES:

Event 1 - Control rod drift in / sticks (Triggered by Lead Examiner)

(The 10/20 scenario 1 exam contained a control rod that drifts out. The actions for a control rod drifting out is different as it requires you to apply a continuous insert signal on the control rod that is drifting out until it can be isolated fr om inside the containment. This is not required for a control rod drifting in. Also, when driving the drifting rod in, it will stick. This will require raising drive water pressure to insert the control rod. )

Control rod 04 -41 will drift in. 05-1 IV-1, Control Rod/Drive Malfunctions ONEP should be entered. Per the ONEP, direction should be given to fully insert 04 -41 by using the IN TIMER SKIP pushbutton. When rod position is < position 12, it will stick. Drive water DP should be raised IAW 04-1 C11-1, Control Rod Drive Hydraulic System SOI in 25 psid increments to insert the rod. Once drive water DP exceeds 290 psid, the control rod can be fully inserted.

Once fully inserted, the CRS should direct the HCU to be isolated IAW 04-1 C11-1, Control Rod Drive Hydraulic System SOI (step 5.7.2 NOTE), by closing C11-103 and C11-105 located inside containment to comply with Tech Specs for an inop control rod.

Tech Specs should be referenced and the following LCO identified:

  • 3.1.3, Condition C.1 and C.2

Event 2 - RCIC steam leak with room temp remaining < isolation setpoint (Triggered by Lead Examiner)

RCIC will develop a steam leak that remains below its isolation setpoint. RCIC room temperature will rise to 157°F and stabilize. The CRS should direct isolating RCIC by closing RCIC steam isolation valves E51-F063 and F064. RCIC room temperature will remain below its isolation setpoint of 185°F. Per guidance from 02-S 01-43, Transient Mitigation Strategy, (step 6.4.1) states leaks should be isolated without CRS direction.

Tech Specs should be referred to and LCO 3.5.3, Condition A identified.

Event 3 - FWLC steam flow transmitter failure (Triggered by Lead Examiner)

When feedwater is operating in 3-element control, the feedwater level control receives one of its inputs from the summed values of the 4 steam line flow transmitters.

The D transmitter will start fluctuating over the next 10 minutes and will then stabilize at a slightly higher value. As a result, RPV level will fluctuate also. Annunciator P680- 3A-A3, RX LVL 40"/32" HI/LO will alarm. The ARI should be referred to. ARI Immediate Operator Action 3.3 directs transfer of Feedwater control from 3-element control to single-element control per 04-1-01-N21-1, Feedwater System SOI. SOI step 5.5.2b provides instruction to depress th e SINGLE ELEMENT PB (680- 2C). Once depressed, RPV level will return to normal value.

Revision 0 3/11/2021 4 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 3

Event 4 - Suppression pool leak RHR A (Triggered by Lead Examiner)

RHR A will develop a leak on the pump side of the suppression pool suction isolation valve E12F004A.

The following alarms / indications will be received:

  • Lowering suppression pool level (P870, 4B and 10B)

The CRS should determine a leak has developed into the RHR A room and direct the trip of RHR pump A and closure of E12F004A, RHR PMP A SUCT FM SUPP POOL. Per guidance from 02 -S 43, Transient Mitigation Strategy, (step 6.4.1) states leaks should be isolated without CRS direction. EP-4, Auxiliary Building Control (step 5) also provides direction to isolate the leak.

The CRS should direct racking out RHR pump A breaker 152-1509 and securing of RHR A jockey pump.

He may also direct opening RHR A jockey pump breaker 52-153132.

The CRS should direct raising suppression pool level IAW 04-1-01-P11-2, Refueling Water Storage and Transfer System SOI section 5.9.2b. Time will not be provided to complete this action.

Tech Specs should be referred to and the following LCOs identified:

LCO 3.5.1, Cond A will remain active (LPCI).

Event 5, 6, 7, 8 - Spurious main turbine trip / Recirc pump A discharge leak / main steam tunnel steam leak / Loss of offsite power / Div 3 DG fails to automatically start (The turbine trip is trigged by the leak examiner, all other events are a utomatically triggered)

A spurious main turbine trip will occur resulting in a reactor scram.

2 seconds after the scram, a recirc pump discharge leak will occur resulting in drywell pressure exceeding automatic isolations and ECCS initiation setpoints. All ECCS initiations will occur. The Div 1 and 2 DGs will auto start, Div 3 DG will fail to auto start. HPCS will inject. HPCS injection should be secured prior to receiving an automatic closure of its injection valve, E22F004 due to high RPV level.

After isolations, ECCS initiations, and diesel generators are verified, the ATC should report Div 3 DG failed to start and start Div 3 diesel generator.

The CRS should direct placing startup level control in service with a level band of 11.4 to 53.5 NR.

The CRS should direct control of reactor pressure within a band of 800- 1060 psig using bypass valves.

After RPV parameters are stabilized, the CRS should direct the prevention of low pressure ECCS injection and directing a reduced pressure band of 450- 600 psig using bypass valves.

6 minutes after the scram occurs, a steam leak will occur in the main steam tunnel resulting in receipt of the following alarms:

  • 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 These alarms indicate steam tunnel temperature has exceeded its automatic isolation setpoint and is also an EP-4, Auxiliary Building Control entry condition. The CRS should enter and direct action from EP-4, Auxiliary Building Control. The ATC should report a failure of the MSIVs to automatically close and close them. (CT-1), Isolate the MSIVs within 15 minutes from receipt of above alarms. With the MSIVs closed, the CRS should direct RPV pressure control with SRVs.

Revision 0 3/11/2021 5 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 3

20 seconds after the MSIVs are closed, a loss of offsite power will occur. This will result in a loss of power to condensate and feedwater and all safety busses. Div 1 and 2 diesel generators will automatically energize their respective busses. Div 3 diesel will energize 17AC if it was previously started. If it wasnt previously started, the CRS should enter 05 02-I-4, Loss of ESF AC Power ONEP, and direct starting Div 3 diesel generator IAW attachment 4 hardcard. Failure to restore power to 17AC will result in RPV level lowering below - 191 CFZ. (CT-2), Energize 17AC by starting Div 3 DG prior to RPV level lowering below -191 CFZ.

The CRS should enter EP-3, Containment Control and direct maximizing suppression pool cooling and starting Div 1 and 2 hydrogen igniters.

Termination:

When directed by Lead Evaluator:

Take the simulator to Freeze and turn horns off.

Stop and save the SBT report and any other recording devices.

Instruct the crew to not erase any markings or talk about the scenario until after follow -up questions are asked.

Revision 0 3/11/2021 6 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 3

(CT-1) After Group 1 auto isolation signal fails to automatically isolate, manually close the (below) Group 1 isolations valves within 15 minutes following receipt of the Critical Task following annunciators:

P601-18A-C3/4, MSL PIPE TNL CH-B/C TEMP HI P601-19A-C3/4, MSL PIPE TNL CH-A/D TEMP HI Event 8 EAL CNB9 allows 15 minutes to isolate the leak. Failure to initiate actions to close Safety Group 1 isolation valves within this time frame will result in a continued discharge of Significance RPV steam into the environment and unwarranted personnel exposure. This would also result in the declaration of a site area emergency and site evacuation.

P601-18A-C3/4, MSL PIPE TNL CH-B/C TEMP HI illuminated.

Cueing P601-19A-C3/4, MSL PIPE TNL CH-A/D TEMP HI illuminated.

Main steam tunnel temperature indications displayed on PDS.

MSIVs handswitches indicate red light only. (P601 18C/19C)

At a minimum, places the handswitch (for at least) one MSIV in each main steam line to close on P601-18C/19C:

Measurable MSL A DRWL INBD ISOL, B21-F022A and/or MSL A CTMT OTBD ISOL, B21-F028A Performance MSL B DRWL INBD ISOL, B21-F022B and/or MSL B CTMT OTBD ISOL, B21-F028B Indicators MSL C DRWL INBD ISOL, B21-F022C and/or MSL C CTMT OTBD ISOL, B21-F028C

MSL D DRWL INBD ISOL, B21-F022D and/or MSL D CTMT OTBD ISOL, B21-F028D Performance Each of the above main steams listed above contains at least one MSIV indicating Feedback green light only.

Revision 0 3/11/2021 7 Appendix D Simulator Outline Form ES-D-1 GGNS 10 -202 1 NRC Scenario 3

(CT-2) Following a loss of offsite power event concurrent with a Div 3 DG failure to start Critical Task and energize 17AC, the crew starts Div 3 DG and ensures 17AC is re-energized prior to RPV level lowering to - 191 CFZ.

Event 7,8 Failure to start Div 3 DG will result in the inability to use HPCS system to restore and maintain RPV level above the value where emergency depressurization is required.

-191 CFZ (MSCWL), is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the Safety uncovered portion of the core from exceeding 1500°F. When water level decreases below MSCWL with injection (SLC), clad temperatures may exceed 1500°F. An Significance avoidable emergency depressurization will result is exceeding cooldown limits as addressed in Tech Specs. Per TS Bases 3.4.11, - Failure to maintain heatup/cooldown limits can result in brittle fracture of the reactor coolant pressure boundary and possibly lead to an unisolable leak or loss of coolant accident. In addition, if the leak were allowed to continue, an unnecessary loss of Rx coolant inventory will occur which is avoidable.

P601/16B HPCS INIT RESET pushbutton white light on.

Cueing P601/16C, all 3 17AC incoming feeder breakers indicates green light.

P601/16C HPCS DSL ENG RUNNING annunciator not illuminated.

CFZ level indications displayed on PDS and SPDS indicates RPV level > - 191CFZ.

Measurable Starts Div 3 DG by placing P601/16C HPCS DSL ENG CONT handswitch to start.

Performance Indicators Observes P601/16C HPCS DSL ENG RUNNING annunciator illuminated.

Performance Observes P601/16C 17AC FDR FM DG 13, 152-1701 red light illuminated.

Feedback Observes P601/16C HPCS INJ SHUTOFF VLV, E22-F004 red light only.

Observes P601/16B HPCS PUMP DISCH FLO indicates rising flow.

Observes rising RPV level trend on PDS/SPDS and/or B21R623A/B on P601-20B/17B.

  • 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). An unintentional or unnecessary RPS or ESF actuation may result in the creation of a post-scenario Critical Task, if that actuation results in a significant plant degradation or significantly alters a mitigation strategy.

Revision 0 3/11/2021 8

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