ML20297A279

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Draft Written Examination and Operating Test Outlines (Folder 2)
ML20297A279
Person / Time
Site: Salem  PSEG icon.png
Issue date: 10/23/2020
From:
Public Service Enterprise Group
To:
Operations Branch I
Shared Package
ML19105A168 List:
References
CAC 000500
Download: ML20297A279 (44)
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Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: SALEM Date of Exam: JULY 2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

3 3

3 N/A 3

3 N/A 3

18 6

2 2

2 1

1 1

2 9

4 Tier Totals 5

5 4

4 4

5 27 10

2.

Plant Systems 1

4 3

2 2

2 1

2 4

3 2

3 28 5

2 1

1 2

1 1

1 1

1 1

0 0

10 3

Tier Totals 5

4 4

3 3

2 3

5 4

2 3

38 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

3 3

2 2

Note: 1.

Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

2.

The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3.

Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4.

Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.

5.

Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher 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 IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.

9.

For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 2

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR 000007 (EPE 7; BW E02&E10; CE E02)

Reactor Trip, Stabilization, Recovery / 1 X

EK 2.03 - Knowledge of the interrelations between a reactor trip and the following: Reactor trip status panel 3.5 1

000008 (APE 8) Pressurizer Vapor Space Accident / 3 X

AK 2.02 - Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Sensors and detectors 2.7 2

000009 (EPE 9) Small Break LOCA / 3 X

EK 1.02 - Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Use of steam tables 3.5 3

000011 (EPE 11) Large Break LOCA / 3 N/A - Not randomly selected 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 X

AK 3.01 - Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow) :

Potential damage from high winding and/or bearing temperatures 2.5 4

000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X

AA 1.07 - Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup: Excess letdown containment isolation valve switches and indicators 2.8 5

000025 (APE 25) Loss of Residual Heat Removal System / 4 X

AA 2.04 - Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Location and isolability of leaks 3.3 6

000026 (APE 26) Loss of Component Cooling Water / 8 N/A - Not randomly selected 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X

AK 3.03 - Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: Actions contained in EOP for PZR PCS malfunction 3.7 7

000029 (EPE 29) Anticipated Transient Without Scram / 1 X

AK 3.12-Knowledge of the reasons for the following responses as the apply to the ATWS: Actions contained in EOP for ATWS 4.4 8

000038 (EPE 38) Steam Generator Tube Rupture / 3 N/A - Not randomly selected 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer / 4 X G 2.4.46 - Ability to verify that the alarms are consistent with the plant conditions.

4.2 9

000054 (APE 54; CE E06) Loss of Main Feedwater /4 X

AK 1.02 - Effects of feedwater introduction on dry S/G 3.6 10 000055 (EPE 55) Station Blackout / 6 N/A - Not randomly selected 000056 (APE 56) Loss of Offsite Power / 6 X

AA 1.03 - Ability to operate and / or monitor the following as they apply to the Loss of Offsite Power:

Adjustment of ED/G load by selectively energizing PZR backup heaters 3.2 11 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X

AA 2.13 - Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: VCT level and pressure indicators and recorders 3.0 12 000058 (APE 58) Loss of DC Power / 6 X G 2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc.

3.9 13 000062 (APE 62) Loss of Nuclear Service Water / 4 X G 2.2.40 - Ability to apply Technical Specifications for a system 3.4 14 000065 (APE 65) Loss of Instrument Air / 8 X

AA 1.03 - Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air: Restoration of systems served by instrument air when pressure is regained 2.9 15 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 N/A - Not randomly selected

ES-401 3

Form ES-401-2 (W E04) LOCA Outside Containment / 3 X

EK 2.1 - Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

3.5 16 (W E11) Loss of Emergency Coolant Recirculation / 4 X

EK 1.2 - Knowledge of the operational implications of the following concepts as they apply to the (Loss of Emergency Coolant Recirculation): Normal, abnormal and emergency operating procedures associated with (Loss of Emergency Coolant Recirculation) 3.6 17 (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 X

EA 2.1 - Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink): Facility conditions and selection of appropriate procedures during abnormal and emergency operations 3.4 18 K/A Category Totals:

3 3

3 3

3 3

Group Point Total:

18/6

ES-401 4

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR 000001 (APE 1) Continuous Rod Withdrawal / 1 N/A - Not randomly selected 000003 (APE 3) Dropped Control Rod / 1 X

AK 1.22 - Knowledge of the operational implications of the following concepts as they apply to Dropped Control Rod:

Calculation of power defect:

algebraic sum of moderator temperature and fuel temperature defects 2.5 19 000005 (APE 5) Inoperable/Stuck Control Rod / 1 X

AA 1.04 - Ability to operate and /

or monitor the following as they apply to the Inoperable / Stuck Control Rod: Reactor and turbine power 3.9 20 000024 (APE 24) Emergency Boration / 1 X

AK 2.01 - Knowledge of the interrelations between Emergency Boration and the following: Valves 2.7 21 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 N/A - Not randomly selected 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 N/A - Not randomly selected 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 X

AA 2.02 - Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:

Indications of unreliable intermediate-range channel operation 3.3 22 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 N/A - Not randomly selected 000037 (APE 37) Steam Generator Tube Leak / 3 N/A - Not randomly selected 000051 (APE 51) Loss of Condenser Vacuum / 4 N/A - Not randomly selected 000059 (APE 59) Accidental Liquid Radwaste Release / 9 N/A - Not randomly selected 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 N/A - Not randomly selected 000061 (APE 61) Area Radiation Monitoring System Alarms

/ 7 X

AK 3.02 - Knowledge of the reasons for the following responses as they apply to the Area Radiation Monitoring (ARM) System Alarms: Guidance contained in alarm response for ARM system 3.4 23 000067 (APE 67) Plant Fire On Site / 8 N/A - Not randomly selected 000068 (APE 68; BW A06) Control Room Evacuation / 8 N/A - Not randomly selected 000069 (APE 69; W E14) Loss of Containment Integrity / 5 N/A - Not randomly selected 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /

4 X G 2.1.19 - Ability to use plant computers to evaluate system or component status 3.9 24 000076 (APE 76) High Reactor Coolant Activity / 9 N/A - Not randomly selected 000078 (APE 78*) RCS Leak / 3 N/A - Not randomly selected (W E01 & E02) Rediagnosis & SI Termination / 3 X G 2.4.45 - Ability to prioritize and interpret the significance of each annunciator or alarm 4.1 25 (W E13) Steam Generator Overpressure / 4 N/A - Not randomly selected (W E15) Containment Flooding / 5 N/A - Not randomly selected

ES-401 5

Form ES-401-2 (W E16) High Containment Radiation /9 X

EK 2.1 - Knowledge of the interrelations between the (High Containment Radiation) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 3.0 26 (BW A01) Plant Runback / 1 N/A - Not randomly selected (BW A02 & A03) Loss of NNI-X/Y/7 N/A - Not randomly selected (BW A04) Turbine Trip / 4 N/A - Not randomly selected (BW A05) Emergency Diesel Actuation / 6 N/A - Not randomly selected (BW A07) Flooding / 8 N/A - Not randomly selected (BW E03) Inadequate Subcooling Margin / 4 N/A - Not randomly selected (BW E08; W E03) LOCA CooldownDepressurization / 4 X

EK 1.3 - Knowledge of the operational implications of the following concepts as they apply to the (LOCA Cooldown and Depressurization): Annunciators and conditions indicating signals, and remedial actions associated with the (LOCA Cooldown and Depressurization) 3.5 27 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 N/A - Not randomly selected (BW E13 & E14) EOP Rules and Enclosures N/A - Not randomly selected (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 N/A - Not randomly selected (CE A16) Excess RCS Leakage / 2 N/A - Not randomly selected (CE E09) Functional Recovery N/A - Not randomly selected (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 N/A - Not randomly selected K/A Category Point Totals:

2 2

1 1

1 2

Group Point Total:

9/4

ES-401 6

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 003 (SF4P RCP) Reactor Coolant Pump X

A 1.04 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: RCP oil reservoir levels 2.6 28 004 (SF1; SF2 CVCS) Chemical and Volume Control X

K 2.06 - Knowledge of bus power supplies to the following: Control instrumentation 2.6 29 004 (SF1; SF2 CVCS) Chemical and Volume Control X

G 2.1.28 - Knowledge of the purpose and function of major system components and controls 4.1 30 005 (SF4P RHR) Residual Heat Removal X

K 6.03 - Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: RHR heat exchanger 2.5 31 006 (SF2; SF3 ECCS) Emergency Core Cooling X

K 5.09 - Knowledge of the operational implications of the following concepts as they apply to ECCS: Thermodynamics of water and steam, including subcooled margin, superheat,and saturation 3.3 32 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X

K 3.01 - Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: Containment 3.3 33 008 (SF8 CCW) Component Cooling Water X

K 1.04 - Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems:

RCS, in order to determine source(s) of RCS leakage into the CCWS 3.3 34 008 (SF8 CCW) Component Cooling Water X

K 4.09 - Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following: The "standby" feature for the CCW pumps 2.7 35 010 (SF3 PZR PCS) Pressurizer Pressure Control X

A 3.01 - Ability to monitor automatic operation of the PZR PCS, including: PRT temperature and pressure during PORV testing 3.0 36 012 (SF7 RPS) Reactor Protection X

K 5.01 - Knowledge of the operational implications of the following concepts as the apply to the RPS: DNB 3.3 37 012 (SF7 RPS) Reactor Protection X G 2.4.31 - Knowledge of annunciator alarms, indications, or response procedures 4.2 38 013 (SF2 ESFAS) Engineered Safety Features Actuation X

A 2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Rapid depressurization 4.4 39 022 (SF5 CCS) Containment Cooling X

A 1.01 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Containment temperature 3.6 40 025 (SF5 ICE) Ice Condenser N/A - Not randomly selected, not applicable to Salem 026 (SF5 CSS) Containment Spray X

K 2.01 - Knowledge of bus power supplies to the following: Containment spray pumps 3.4 41 026 (SF5 CSS) Containment Spray X

G 2.4.8 - Knowledge of how abnormal operating procedures are used in conjunction with EOPs 3.8 42

ES-401 7

Form ES-401-2 039 (SF4S MSS) Main and Reheat Steam X

K 1.02 - Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems:

Atmospheric relief dump valves 3.3 43 059 (SF4S MFW) Main Feedwater X

K 1.02 - Knowledge of the physical connections and/or cause-effect relationships between the MFW and the following systems:

AFW system 3.4 44 059 (SF4S MFW) Main Feedwater X

A 2.12 - Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of feedwater regulating valves 3.1 45 061 (SF4S AFW)

Auxiliary/Emergency Feedwater X

A 3.03 - Ability to monitor automatic operation of the AFW, including: AFW S/G level control on automatic start 3.9 46 062 (SF6 ED AC) AC Electrical Distribution X

K 1.02 - Knowledge of the physical connections and/or cause-effect relationships between the ac distribution system and the following systems: ED/G 4.1 47 062 (SF6 ED AC) AC Electrical Distribution X

A 4.07 - Ability to manually operate and/or monitor in the control room: Synchronizing and paralleling of different ac supplies 3.1 48 063 (SF6 ED DC) DC Electrical Distribution X

A 4.03 - Ability to manually operate and/or monitor in the control room: Battery discharge rate 3.0 49 064 (SF6 EDG) Emergency Diesel Generator X

A 2.16 - Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of offsite power during full-load testing of ED/G 3.3 50 073 (SF7 PRM) Process Radiation Monitoring X

K 4.01 - Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following: Release termination when radiation exceeds setpoint 4.0 51 076 (SF4S SW) Service Water X

K 3.07 - Knowledge of the effect that a loss or malfunction of the SWS will have on the following: ESF loads 3.7 52 076 (SF4S SW) Service Water X

A 2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of SWS 3.5 53 078 (SF8 IAS) Instrument Air X

K 2.02 - Knowledge of bus power supplies to the following: Emergency air compressor 3.3 54 103 (SF5 CNT) Containment X

A 3.01 - Ability to monitor automatic operation of the containment system, including:

Containment isolation 3.9 55 053 (SF1; SF4P ICS*) Integrated Control N/A for NUREG-1122, Rev. 2 K/A Category Point Totals:

4 3

2 2

2 1

2 4

3 2

3 Group Point Total:

28/5

ES-401 8

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 001 (SF1 CRDS) Control Rod Drive X

K 2.05 - Knowledge of bus power supplies to the following: M/G sets 3.1 56 002 (SF2; SF4P RCS) Reactor Coolant N/A - Not randomly selected 011 (SF2 PZR LCS) Pressurizer Level Control N/A - Not randomly selected 014 (SF1 RPI) Rod Position Indication N/A - Not randomly selected 015 (SF7 NI) Nuclear Instrumentation N/A - Not randomly selected 016 (SF7 NNI) Nonnuclear Instrumentation X

K3.02 - Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: PZR LCS 3.4 57 017 (SF7 ITM) In-Core Temperature Monitor X

K 4.01 - Knowledge of ITM system design feature(s) and/or interlock(s) which provide for the following: Input to subcooling monitors 3.4 58 027 (SF5 CIRS) Containment Iodine Removal N/A - Not randomly selected 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control N/A - Not randomly selected 029 (SF8 CPS) Containment Purge X

K 1.03 - Knowledge of the physical connections and/or cause-effect relationships between the Containment Purge System and the following systems: Engineered safeguards 3.6 59 033 (SF8 SFPCS) Spent Fuel Pool Cooling N/A - Not randomly selected 034 (SF8 FHS) Fuel-Handling Equipment N/A - Not randomly selected 035 (SF 4P SG) Steam Generator X

K 6.02 - Knowledge of the effect of a loss or malfunction on the following will have on the S/GS: Secondary PORV 3.1 60 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X

A 3.05 - Ability to monitor automatic operation of the SDS, including: Main steam pressure 2.9 61 045 (SF 4S MTG) Main Turbine Generator X

K 5.17 - Knowledge of the operational implications of the following concepts as the apply to the MT/B System: Relationship between moderator temperature coefficient and boron concentration in RCS as T/G load increases 2.5 62 055 (SF4S CARS) Condenser Air Removal X

K 3.01 - Knowledge of the effect that a loss or malfunction of the CARS will have on the following: Main condenser 2.5 63 056 (SF4S CDS) Condensate N/A - Not randomly selected 068 (SF9 LRS) Liquid Radwaste N/A - Not randomly selected 071 (SF9 WGS) Waste Gas Disposal N/A - Not randomly selected 072 (SF7 ARM) Area Radiation Monitoring N/A - Not randomly selected 075 (SF8 CW) Circulating Water X

A 2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of circulating water pumps 2.5 64 079 (SF8 SAS**) Station Air N/A - Not randomly selected

ES-401 9

Form ES-401-2 086 Fire Protection X

A 1.05 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Fire Protection System operating the controls including: FPS lineups 2.9 65 050 (SF 9 CRV*) Control Room Ventilation N/A for NUREG-1122, Rev. 2 K/A Category Point Totals:

1 1

2 1

1 1

1 1

1 0

0 Group Point Total:

10/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility: SALEM Date of Exam: JULY 2020 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.

G 2.1.15 - Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc.

2.7 66 2.1.

G 2.1.30 - Ability to locate and operate components, including local controls.

4.4 67 2.1.

G 2.1.37 - Knowledge of procedures, guidelines, or limitations associated with reactivity management.

4.3 68 2.1.

2.1.

2.1.

Subtotal

2. Equipment Control 2.2.

G 2.2.14 - Knowledge of the process for controlling equipment configuration or status.

3.9 69 2.2.

G 2.2.13 - Knowledge of tagging and clearance procedures.

4.1 70 2.2.

G 2.2.12 - Knowledge of surveillance procedures.

3.7 71 2.2.

2.2.

2.2.

Subtotal

3. Radiation Control 2.3.

G 2.3.14 - Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

3.4 72 2.3.

G 2.3.4 - Knowledge of radiation exposure limits under normal or emergency conditions.

3.2 73 2.3.

2.3.

2.3.

2.3.

Subtotal

4. Emergency Procedures/Plan 2.4.

G 2.4.39 - Knowledge of RO responsibilities in emergency plan implementation.

3.9 74 2.4.

G 2.4.25 - Knowledge of fire protection procedures.

3.3 75 2.4.

2.4.

2.4.

2.4.

Subtotal Tier 3 Point Total 10 7

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

Group Randomly Selected K/A Reason for Rejection 1/1 000058G2.1.25 Unable to write a discriminating question utilizing graphs, curves, tables, etc. Replaced with K/A 000058G2.4.6.

1/1 000058G2.4.6.

Unable to write a fair and discriminating RO level question.

Replaced with K/A 000058AA2.03 1/1 000065AA1.03 Unable to write a discriminating question with plausible distractors. Replaced with K/A 000065AA1.04.

1/2 000003AK1.22 Unable to write a discriminating question with plausible distractors. Replaced with K/A 000003AK3.04.

1/2 000033AA2.02 Unable to write a fair and discriminating RO level question.

Replaced with K/A 000033AA2.09.

1/2 WE01G2.4.45 Unable to write a discriminating question with plausible distractors. Replaced with K/A WE02G2.4.45.

2/1 003A1.04 Potential overlap with Question #4. Replaced with K/A 003A1.10.

2/1 004K2.06 Potential overlap with Question #12. Replaced with K/A 004K2.03.

2/1 006K5.09 Potential overlap with LOCA-2 questions (2,3,25,27).

Replaced with K/A 006K5.04.

2/1 010A3.01 Potential overlap with Question # 33. Replaced with K/A 010A3.02.

2/1 013A2.03 Unable to write a fair and discriminating RO level question.

Replaced with K/A 013K4.12.

2/1 026K2.01 Potential overlap with Question #40. Replaced with K/A 026K4.08.

2/1 026G2.4.8 Unable to write a discriminating question with plausible distractors. Replaced with K/A 026G2.1.25.

2/1 064A2.16 Unable to write a fair and discriminating RO level question.

Replaced with K/A 064K6.07.

2/2 035K6.02 Potential overlap with Question # 43. Replaced with K/A 035K6.03.

2/2 045K5.17 Unable to write a fair and discriminating RO level question.

Replaced with K/A 033K4.03.

3/1 G2.1.15 Unable to write a fair and discriminating RO level question.

Replaced with K/A G2.1.1.

3/1 G2.1.30 Unable to write discriminating question that maintained its focus on being a plantwide generic question. Replaced with K/A G2.1.3.

ES-401 Record of Rejected K/As Form ES-401-4 3/4 G2.4.39 Unable to write discriminating question that maintained its focus on being a plantwide generic question. Replaced with K/A G2.4.17.

ES-401 PWR Examination Outline Form ES-401-2 Facility: SALEM Date of Exam: JULY 2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

N/A N/A 18 4

2 6

2 9

3 1

4 Tier Totals 27 7

3 10

2.

Plant Systems 1

28 3

2 5

2 10 2

1 3

Tier Totals 38 5

3 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

2 2

1 2

Note: 1.

Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

2.

The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3.

Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4.

Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.

5.

Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher 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 IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.

9.

For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 2

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR 000007 (EPE 7; BW E02&E10; CE E02)

Reactor Trip, Stabilization, Recovery / 1 N/A - Not randomly selected 000008 (APE 8) Pressurizer Vapor Space Accident / 3 N/A - Not randomly selected 000009 (EPE 9) Small Break LOCA / 3 N/A - Not randomly selected 000011 (EPE 11) Large Break LOCA / 3 X

EA 2.07 - Ability to determine or interpret the following as they apply to a Large Break LOCA:

That equipment necessary for functioning of critical pump water seals is operable 3.4 1/76 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 N/A - Not randomly selected 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X G 2.4.18 - Knowledge of the specific bases for EOPs 4.0 2/77 000025 (APE 25) Loss of Residual Heat Removal System / 4 N/A - Not randomly selected 000026 (APE 26) Loss of Component Cooling Water / 8 X

A 2.06 - Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The length of time after the loss of CCW flow to a component before that component may be damaged 3.1 3/78 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 N/A - Not randomly selected 000029 (EPE 29) Anticipated Transient Without Scram / 1 N/A - Not randomly selected 000038 (EPE 38) Steam Generator Tube Rupture / 3 X

EA 2.04 - Ability to determine or interpret the following as they apply to a SGTR: Radiation levels (MREM/hr) 4.2 4/79 000040 (APE 40; BW E05; CE E05; W E12)

Steam Line RuptureExcessive Heat Transfer / 4 N/A - Not randomly selected 000054 (APE 54; CE E06) Loss of Main Feedwater /4 N/A - Not randomly selected 000055 (EPE 55) Station Blackout / 6 X

EA 2.05 - Ability to determine or interpret the following as they apply to a Station Blackout: When battery is approaching fully discharged 3.7 5/80 000056 (APE 56) Loss of Offsite Power / 6 N/A - Not randomly selected 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 N/A - Not randomly selected 000058 (APE 58) Loss of DC Power / 6 N/A - Not randomly selected 000062 (APE 62) Loss of Nuclear Service Water / 4 N/A - Not randomly selected 000065 (APE 65) Loss of Instrument Air / 8 N/A - Not randomly selected 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X

G 2.2.25 - Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits 4.2 6/81 (W E04) LOCA Outside Containment / 3 N/A - Not randomly selected (W E11) Loss of Emergency Coolant Recirculation / 4 N/A - Not randomly selected (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 N/A - Not randomly selected

ES-401 3

Form ES-401-2 K/A Category Totals:

4 2

Group Point Total:

18/6

ES-401 4

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR 000001 (APE 1) Continuous Rod Withdrawal / 1 X G 2.2.44 - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 4.4 7/82 000003 (APE 3) Dropped Control Rod / 1 N/A - Not randomly selected 000005 (APE 5) Inoperable/Stuck Control Rod / 1 N/A - Not randomly selected 000024 (APE 24) Emergency Boration / 1 N/A - Not randomly selected 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 N/A - Not randomly selected 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 N/A - Not randomly selected 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 N/A - Not randomly selected 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 N/A - Not randomly selected 000037 (APE 37) Steam Generator Tube Leak / 3 N/A - Not randomly selected 000051 (APE 51) Loss of Condenser Vacuum / 4 N/A - Not randomly selected 000059 (APE 59) Accidental Liquid Radwaste Release / 9 N/A - Not randomly selected 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 N/A - Not randomly selected 000061 (APE 61) Area Radiation Monitoring System Alarms

/ 7 N/A - Not randomly selected 000067 (APE 67) Plant Fire On Site / 8 X

AA 2.04 - Ability to determine and interpret the following as they apply to the Plant Fire on Site:

The fire's extent of potential operational damage to plant equipment 4.3 8/83 000068 (APE 68; BW A06) Control Room Evacuation / 8 N/A - Not randomly selected 000069 (APE 69; W E14) Loss of Containment Integrity / 5 N/A - Not randomly selected 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /

4 N/A - Not randomly selected 000076 (APE 76) High Reactor Coolant Activity / 9 N/A - Not randomly selected 000078 (APE 78*) RCS Leak / 3 N/A - Not randomly selected (W E01 & E02) Rediagnosis & SI Termination / 3 N/A - Not randomly selected (W E13) Steam Generator Overpressure / 4 N/A - Not randomly selected (W E15) Containment Flooding / 5 N/A - Not randomly selected (W E16) High Containment Radiation /9 N/A - Not randomly selected (BW A01) Plant Runback / 1 N/A - Not randomly selected (BW A02 & A03) Loss of NNI-X/Y/7 N/A - Not randomly selected (BW A04) Turbine Trip / 4 N/A - Not randomly selected (BW A05) Emergency Diesel Actuation / 6 N/A - Not randomly selected (BW A07) Flooding / 8 N/A - Not randomly selected (BW E03) Inadequate Subcooling Margin / 4 N/A - Not randomly selected (BW E08; W E03) LOCA CooldownDepressurization / 4 N/A - Not randomly selected (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 X

EA 2.2 - Ability to determine and interpret the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS): Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments 3.9 9/84

ES-401 5

Form ES-401-2 (BW E13 & E14) EOP Rules and Enclosures N/A - Not randomly selected (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 X

EA 2.2 - Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock): Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments 4.1 10/

85 (CE A16) Excess RCS Leakage / 2 N/A - Not randomly selected (CE E09) Functional Recovery N/A - Not randomly selected (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 N/A - Not randomly selected K/A Category Point Totals:

3 1

Group Point Total:

9/4

ES-401 6

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 003 (SF4P RCP) Reactor Coolant Pump X

G 2.4.34 - Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects 4.1 11/

86 004 (SF1; SF2 CVCS) Chemical and Volume Control N/A - Not randomly selected 005 (SF4P RHR) Residual Heat Removal N/A - Not randomly selected 006 (SF2; SF3 ECCS) Emergency Core Cooling X

G 2.4.47 - Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material 4.2 12/

87 007 (SF5 PRTS) Pressurizer Relief/Quench Tank N/A - Not randomly selected 008 (SF8 CCW) Component Cooling Water N/A - Not randomly selected 010 (SF3 PZR PCS) Pressurizer Pressure Control N/A - Not randomly selected 012 (SF7 RPS) Reactor Protection N/A - Not randomly selected 013 (SF2 ESFAS) Engineered Safety Features Actuation X

A 2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Loss of instrument bus 4.2 13/

88 022 (SF5 CCS) Containment Cooling X

A 2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of service water 3.2 14/

89 025 (SF5 ICE) Ice Condenser N/A - Not randomly selected, not applicable to Salem 026 (SF5 CSS) Containment Spray N/A - Not randomly selected 039 (SF4S MSS) Main and Reheat Steam N/A - Not randomly selected 059 (SF4S MFW) Main Feedwater N/A - Not randomly selected 061 (SF4S AFW)

Auxiliary/Emergency Feedwater N/A - Not randomly selected 062 (SF6 ED AC) AC Electrical Distribution N/A - Not randomly selected 063 (SF6 ED DC) DC Electrical Distribution N/A - Not randomly selected 064 (SF6 EDG) Emergency Diesel Generator N/A - Not randomly selected 073 (SF7 PRM) Process Radiation Monitoring X

A 2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure 3.2 15/

90

ES-401 7

Form ES-401-2 076 (SF4S SW) Service Water N/A - Not randomly selected 078 (SF8 IAS) Instrument Air N/A - Not randomly selected 103 (SF5 CNT) Containment N/A - Not randomly selected 053 (SF1; SF4P ICS*) Integrated Control N/A for NUREG 1122, Rev. 2 K/A Category Point Totals:

3 2 Group Point Total:

28/5

ES-401 8

Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 001 (SF1 CRDS) Control Rod Drive N/A - Not randomly selected 002 (SF2; SF4P RCS) Reactor Coolant N/A - Not randomly selected 011 (SF2 PZR LCS) Pressurizer Level Control X

A 2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of one, two or three charging pumps 3.7 16/

91 014 (SF1 RPI) Rod Position Indication N/A - Not randomly selected 015 (SF7 NI) Nuclear Instrumentation N/A - Not randomly selected 016 (SF7 NNI) Nonnuclear Instrumentation N/A - Not randomly selected 017 (SF7 ITM) In-Core Temperature Monitor N/A - Not randomly selected 027 (SF5 CIRS) Containment Iodine Removal N/A - Not randomly selected 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control X G 2.1.20 - Ability to interpret and execute procedure steps 4.6 17/

92 029 (SF8 CPS) Containment Purge N/A - Not randomly selected 033 (SF8 SFPCS) Spent Fuel Pool Cooling N/A - Not randomly selected 034 (SF8 FHS) Fuel-Handling Equipment N/A - Not randomly selected 035 (SF 4P SG) Steam Generator N/A - Not randomly selected 041 (SF4S SDS) Steam Dump/Turbine Bypass Control N/A - Not randomly selected 045 (SF 4S MTG) Main Turbine Generator N/A - Not randomly selected 055 (SF4S CARS) Condenser Air Removal N/A - Not randomly selected 056 (SF4S CDS) Condensate X

A 2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of condensate pumps 2.8 18/

93 068 (SF9 LRS) Liquid Radwaste N/A - Not randomly selected 071 (SF9 WGS) Waste Gas Disposal N/A - Not randomly selected 072 (SF7 ARM) Area Radiation Monitoring N/A - Not randomly selected 075 (SF8 CW) Circulating Water N/A - Not randomly selected 079 (SF8 SAS**) Station Air N/A - Not randomly selected 086 Fire Protection N/A - Not randomly selected 050 (SF 9 CRV*) Control Room Ventilation N/A for NUREG 1122, Rev. 2 K/A Category Point Totals:

2 1 Group Point Total:

10/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility:SALEM SRO Date of Exam:JULY 2020 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.

G 2.1.40 - Knowledge of refueling administrative requirements 3.9 19/94 2.1.

G 2.1.34 - Knowledge of primary and secondary plant chemistry limits 3.5 20/95 2.1.

2.1.

2.1.

2.1.

Subtotal 2

2. Equipment Control 2.2.

G 2.2.7 - Knowledge of the process for conducting special or infrequent tests 3.6 21/96 2.2.

G 2.2.21 - Knowledge of pre-and post-maintenance operability requirements 4.1 22/97 2.2.

2.2.

2.2.

2.2.

Subtotal 2

3. Radiation Control 2.3.

G 2.3.15 - Knowledge of radiation monitoring systems, such as fixed radiation monitors and l alarms, portable survey instruments, personnel monitoring equipment, etc.

3.1 23/98 2.3.

2.3.

2.3.

2.3.

2.3.

Subtotal 1

4. Emergency Procedures/Plan 2.4.

G 2.4.37 - Knowledge of the lines of authority during implementation of the emergency plan.

4.1 24/99 2.4.

G 2.4.14 - Knowledge of general guidelines for EOP usage 4.5 25/

100 2.4.

2.4.

2.4.

2.4.

Subtotal 2

Tier 3 Point Total 10 7

7

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

Group Randomly Selected K/A Reason for Rejection 1/1 000022G2.4.18 Unable to write an SRO level question related the specific bases for EOPs. Replaced with K/A 000022G2.2.22.

1/1 000038EA2.04 Unable to write a discriminating SRO level question about interpreting MREM/HR relative to SGTR. Replaced with K/A 000038EA2.07.

1/1 000077G2.2.25 Unable to write a discriminating SRO level question about Tech Spec bases specifically. Replaced with K/A 000077G2.1.7.

1/2 000001G2.2.44 Unable to write question on continuous rod withdrawal as a recent DCP has removed all Auto Rod Withdrawal capability.

Resampled, replaced with 000036G2.1.32.

3/RC G2.3.15 Unable to write discriminating SRO level question about fixed radiation monitors, alarms, etc. Replaced with K/A G2.3.11.

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

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

SALEM Date of Examination:

JULY 2020 Examination Level: RO SRO Operating Test Number:

Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations D, R Calculate Shutdown Margin IAW SC.RE-ST.ZZ-0002(Q)

(2.1.43, RO-4.1, Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.)

Conduct of Operations N, R Verification of Active License Status, ability to relieve RO IAW OP-AA-105-101, Administrative Process for NRC License and Medical Requirements and OP-AA-105-102, NRC Active License Maintenance.

(2.1.4, RO-3.3, Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.)

Equipment Control M, R Isolate a Component Cooling System Leak.

(2.2.41, RO-3.5, Ability to obtain and interpret station electrical and mechanical drawings.)

Radiation Control N/A N/A Emergency Plan D, S, P Activate ERDS during an ALERT.

(2.4.43, RO-3.2, Knowledge of emergency communications systems and techniques.)

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

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

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

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

SALEM Date of Examination:

JULY 2020 Examination Level: RO SRO Operating Test Number:

Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations N, R Determine acceptable spent fuel pool storage locations for selected spent fuel assemblies IAW SC.RE-FR.ZZ-0001(Q),

Fuel Handling, Attachment 4 and Technical Specification 3.7.12.

(2.1.42, SRO-3.4; Knowledge of new and spent fuel movement procedures.)

Conduct of Operations N, R Evaluate a shift staffing situation and take corrective actions IAW administrative procedures.

(2.1.5, SRO-3.9; Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.)

Equipment Control M, R Determine Component Operability and Technical Specification Action Statement(s) following failure of the 1A EDG with 13 SW Pump tagged for motor replacement, and then any additional Technical Specification Action Statement(s) and required actions with a subsequent failure of 12 SW Pump Strainer. Explain why?

(2.2.37, SRO-4.6; Ability to determine operability and/or availability of safety related equipment.)

Radiation Control D, R Select Release Path for Radioactive Liquid Waste Release IAW S1.OP-SO.WL-0002, Release of Radioactive Liquid Waste from 12 CVCS Monitor Tank.

(2.3.6, SRO-3.8; Ability to approve release permits.)

Emergency Plan R, D, P Classify Emergency Event and complete the ICMF IAW EP-SA-111-101 (Time Critical)

(2.4.41, SRO-4.6: Knowledge of the emergency action level thresholds and classifications.)

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

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

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

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

SALEM Date of Examination:

JULY 2020 Exam Level: RO SRO-I SRO-U Operating Test Number:

Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function

a. Pressurizer Pressure Control Channel Fails High, Spray Valve fails to close (need to trip RCPs).

A, D, E 3

b. Respond to Loss of RHR in Mode 6, Rx Head tensioning in progress.

A, E, L, N 4P

c. Control Rod Exercise Test.

A, D 1

d. Raise Safety Injection Accumulator Level.

D 2

e. Manually initiate Containment Spray (TRIP-1, step 11) (NaOH valves & Thermal Barrier Valves remain open)

A, E, EN, N 5

f. Remove a Power Range Channel from Service (N41) (Failed Low).

(RO Only)

N 7

g. Perform RCS Cooldown using steam dumps IAW step 10 of SGTR-1.

E, N 4S

h. Transfer 4 KV Group Buses from SPT to APT IAW S2.OP-SO.4KV-0008.

A, D, P 6

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. Locally Control Charging Flow (CV-55).

R, D, E 2

j. Start SBO Diesel Air Compressor D, E 8
k. Place 11 CVCS MT in Recirc.

R, N 9

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

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 9/ 8/ 4 1/ 1/ 1 1/ 1/ 1 (control room system) 1/ 1/ 1 2/ 2/ 1 3/ 3/ 2 (randomly selected) 1/ 1/ 1

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

SALEM Date of Examination:

JULY 2020 Exam Level: RO SRO-I SRO-U Operating Test Number:

Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function

a. Pressurizer Pressure Control Channel Fails High, Spray Valve fails to close (need to trip RCPs).

A, D, E 3

b. Respond to Loss of RHR in Mode 6, Rx Head tensioning in progress.

A, E, L, N 4P

c. Control Rod Exercise Test.

A, D 1

d. Raise Safety Injection Accumulator Level.

D 2

e. Manually initiate Containment Spray (TRIP-1, step 11) (NaOH valves & Thermal Barrier Valves remain open)

A, E, EN, N 5

f. N/A - RO Only N/A N/A
g. Perform RCS Cooldown using steam dumps IAW step 10 of SGTR-1.

E, N 4S

h. Transfer 4 KV Group Buses from SPT to APT IAW S2.OP-SO.4KV-0008.

A, D, P 6

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. Locally Control Charging Flow (CV-55).

R, D, E 2

j. Start SBO Diesel Air Compressor D, E 8
k. Place 11 CVCS MT in Recirc.

R, N 9

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

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 9/ 8/ 4 1/ 1/ 1 1/ 1/ 1 (control room system) 1/ 1/ 1 2/ 2/ 1 3/ 3/ 2 (randomly selected) 1/ 1/ 1

Appendix D Scenario Outline Form ES-D-1 Facility: SALEM ___________

Scenario No.: #1__________ Op-Test No.: JULY 2020_________

Examiners: ___________________________ Operators:

Initial Conditions: 90% Power MOL______________________________

Turnover: The following equipment is out of service: 2C EDG C/T for governor replacement. 68 hours7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br /> remaining in LCO, S2.OP-ST.500-0001 Line Surveillance completed 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ago. Raise power to 100%, IAW IOP-4, Fuel is conditioned, Severe Thunderstorm alert in the area for the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, SC.OP-AB.ZZ-0001, Adverse Environmental Conditions, in effect and all required actions completed.________________

Critical Tasks: Critical Tasks:

1.Energize at least one AC Emergency Bus (see WOG CT-24) 2.Manually start SW pump for EDG cooling (see WOG CT-25)_

Event No.

Malf.

No.

Event Type*

Event Description 1

N/A ATC (R)

BOP (N)

Raise power to 100% IAW IOP-4 (pre-brief activity) 2 AN0025 BOP (C),

CRS (TS)

OHA B-47, SW ACCUM Tank Trouble alarms. Investigation reveals #21 SW Accumulator Tank Temperature is low out of spec. Requires declaring accumulator inoperable IAW ARP, and isolating 21 and 22 CFCUs. Tech Specs 3.6.1.1 for Primary Containment Integrity and Tech Spec 3.6.2.3 for CFCUs are applicable.

3 PR0017A ATC (I)

CRS (TS)

PZR Level Channel I (Controlling) fails low. Crew enters AB.CVC-0001, takes manual control of charging, swaps controlling channel, and restores heaters &

letdown. Tech Specs 3.3.1.1, action 6 is applicable.

4 EL0048 BOP (C),

CRS (TS)

Loss of #4 Station Power Transformer (500 KV Breakers 1-5, 1-8, and 1-9 open).

Requires entry into AB.LOOP-0003, Partial Loss of Off-Site Power due to loss of

  1. 23 Station Power Transformer. Will lose 4KV CW Bus Section 23 (21A, 22A, 23A Circulators). BOP will utilize attachment 4 section 1.0 to energize 2CW Bus Section 23 from 2CW Bus Section 24. BOP will restart circulators as needed. Crew may enter AB.CW-001, Circulating Water System Malfunction and monitor condenser backpressure. CRS may review off-site power sources surveillance requirements -

Tech Spec 3.8.1.1 action c will be applicable due to all vital buses aligned to 24 Station Power Transformer. (If not reviewed, ask follow-up question after scenario termination) 5 EL0134 ALL (M)

Loss of Offsite Power Results in Reactor Trip/Turbine Trip, Entry into EOP-TRIP-1 6

EL0146 EL0162 ALL (C)

Time-delayed by 30 seconds Loss of 2A Vital Bus (Diff Protection) 2B EDG trips on overspeed Entry into LOPA-1, Loss of ALL AC (Maintenance will return 2C EDG to available status (tags released & ready to run) after step 26 of EOP-LOPA-1.)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Scenario No.: #1 Target Quantitative Attributes per Scenario (See Section D.5.d)

Actual Attributes Event No.

1. Total malfunctions (5-8) 6 2-6
2. Malfunctions after EOP entry (1-2) 2 6
3. Abnormal events (2-4) 3 2,3,4
4. Major transients (1-2) 1 5
5. EOPs entered/requiring substantive actions (1-2) 1 LOPA-1
6. Entry into a contingency EOP with substantive actions ( 1 per scenario set) 0 N/A
7. Preidentified critical tasks (2) 2 CT-24, CT-25
8. Tech Specs exercised ( 2) 3 2,3,4

Simulator Exam Scenario # 1 Summary The evaluation begins with the plant at 90% power and power ascension in progress to 100%

power IAW IOP-4, Power Operation. During the pre-briefed power ascension, OHA B-47, SW ACCUM tank Trouble alarms. Investigation reveals #21 SW Accumulator Tank Temperature is low out of spec. Field operator (if dispatched) will find that the #21 SW Accumulator Tank Temperature is 50°F. All other parameters are normal. If asked to investigate cause, state that the associated heater breaker is tripped with an acrid odor. The crew will follow the OHA alarm response procedure, declaring the accumulator inoperable and isolating 21 and 22 CFCUs.

Tech Spec 3.6.1.1 for Primary Containment Integrity and Tech Spec 3.6.2.3 for CFCUs are applicable.

Following identifying and implementing the required actions for the inoperable SW Accumulator Tank, the controlling Pressurizer Level Channel (Channel I) fails low causing a loss of letdown.

The crew inters AB.CVC-0001, Loss of Charging, takes manual control of charging, swaps the controlling channel, and restores pressurizer heaters and letdown. Tech Spec 3.3.1.1, action 6 is applicable.

Once the crew has restored charging and letdown to normal, a Loss of #4 Station Power Transformer (500 KV Breakers 1-5, 1-8, and 1-9 open) will occur. This requires entry into AB.LOOP-0003, Partial Loss of Off-Site Power due to loss of #23 Station Power Transformer.

The crew will lose 4KV CW Bus Section 23 (21A, 22A, 23A Circulators). BOP will utilize attachment 4 section 1.0 of AB.LOOP-0003 to energize 2CW Bus Section 23 from 2CW Bus Section 24. BOP will restart circulators as needed. Crew may enter AB.CW-001, Circulating Water System Malfunction and monitor condenser backpressure. CRS may review off-site power sources surveillance requirements - Tech Spec 3.8.1.1 action c will be applicable due to all vital buses aligned to 24 Station Power Transformer. (If not reviewed, ask follow-up question after scenario termination)

After restoring circulators, a loss of off-site power will occur. This will result in a Reactor Trip /

Turbine Trip and entry into EOP-TRIP-1, Reactor Trip or Safety Injection. Timed delayed by 30 seconds is a loss of 2A Vital Bus (Diff Protection) and the 2B EDG tripping on overspeed. The crew will transition to EOP-LOPA-1, Loss of All AC Power, based on all three 4Kv Vital Buses de-energized. The crew will recognize that the 2A Diesel Generator is running with no service water pumps and will need to be stopped.

Maintenance will return the 2C EDG to available status (tags released & ready to run) after step 26 of EOP-LOPA-1. Once 2C EDG has been restored, the crew will start 2C EDG, close its output breaker and re-energize 2C 4KV Vital Bus (CT). Following the starting of the 2C EDG and energizing 2C 4KV Bus, the crew will start one Service Water Pump on C Bus (CT).

The scenario will terminate when the 2C Bus is re-energized by the 2C EDG and one service water pump is providing EDG cooling.

Procedure flow path: IOP-4, AR.ZZ-2, AB-CVC-001, AB-LOOP-0003, EOP-TRIP-1, LOPA-1

Appendix D Scenario Outline Form ES-D-1 Facility: SALEM ___________

Scenario No.: #2__________ Op-Test No.: JULY 2020_________

Examiners: ___________________________ Operators:

Initial Conditions: 100% Power EOL______________________________

Turnover: The following equipment is out of service: #23 Charging Pump C/T for packing replacement. 21 Charging Pump is in service. Severe Thunderstorm alert in the area for the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, SC.OP-AB.ZZ-0001, Adverse Environmental Conditions, in effect and all required actions completed. ________________

Critical Tasks: Critical Tasks:

CT #1 - Manually actuate at least one train of Safety Injection before existing E-0.

CT #2 - Depressurize SGs to atmospheric pressure (at < 100°F/hr) to inject ECCS accumulators and establish low-head injection flow before a Core Cooling Red Path develops, CT #3 - Manually start an RHR pump prior to depressurizing SGs to atmospheric pressure.

Event No.

Malf.

No.

Event Type*

Event Description 1

N/A ATC (R),

BOP (N)

Reduce power to 90% IAW IOP-4 (pre-brief activity) in preparation for performing Main Turbine Valve Testing later in the shift.

2 SG0100C BOP (I),

CRS (TS) 23 SG Pressure Transmitter Channel III (536) fails low.

Tech Spec 3.3.2.1, action 19 is applicable. (Functional units 1.e & 1.f) (Bezel &

OHA ARPs) 3 RC0015D ATC (I),

CRS (TS)

Loop 24 Cold Leg RTD Fails High. (24 Loop Tavg fails high, 24 Loop T fails low)

Crew enters AB.ROD-0003, Continuous Rod Motion if rods in AUTO. Bezel ARP if not.

4 EL0144 ATC (C),

CRS (TS)

Loss of 2C Vital Bus (Diff Protection)

Crew enters AB.4KV-0003 & AB.460-0003.

5 RC0002 RP0108 ALL (M)

RCS Leak (4 equivalent / 18000gpm) into Containment (Either instantaneous or ramp quickly)

Failure of Automatic SI Results in Rx Trip, SI must be initiated manually.

6 EL0134 ALL (C)

Loss of Off-site Power, simultaneous with Rx Trip.

Crew enters EOP-TRIP-1, transitions to EOP-LOCA-1 7

CV0208B RP318A2 ALL (C) 21 Charging Pump Trip (motor / breaker damage - acrid odor) 22 RHR Pump fails to start on SEC 8

N/A ALL FRCC-2, Degraded Core Cooling (need >700°F CETs or RVLIS < 39%) May have to adjust decay heat / break size, need decay heat > than break flow & SI.

Crew expected to depressurize SGs to inject accumulators.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Scenario No.: #2 Target Quantitative Attributes per Scenario (See Section D.5.d)

Actual Attributes Event No.

1. Total malfunctions (5-8) 7 2-7
2. Malfunctions after EOP entry (1-2) 2 7
3. Abnormal events (2-4) 3 2-4
4. Major transients (1-2) 1 5
5. EOPs entered/requiring substantive actions (1-2) 3 TRIP-1, LOCA-1, FR.CC-2 (5-8)
6. Entry into a contingency EOP with substantive actions ( 1 per scenario set) 1 FRCC-2
7. Preidentified critical tasks (2) 3 8
8. Tech Specs exercised ( 2) 3 2-4

Simulator Exam Scenario # 2 Summary The evaluation begins with the plant at 100% power and the crew has been pre-briefed to reduce power to 90% power IAW IOP-4, Power Operation. During the power reduction, #23 Steam Generator Pressure Transmitter Channel III (536) fails low. Technical Specification 3.3.2.1, action 19 is applicable (Functional units 1.e & 1.f). BOP will review both bezel 4-9 alarm response and OHA G-33 alarm response IAW respective alarm response procedures.

Following completion of the power reduction to 90% (or at examiners direction), Loop 24 Cold Leg RTD fails high. It will result in 24 Loop Tavg failing high and 24 Loop T failing low. If rods are in AUTO, the crew will enter AB.ROD-0003, Continuous Rod Motion to stabilize the plant.

The crew will enter the bezel alarm response if the rods are in manual. Tech Spec 3.3.1.1, action 6 (Functional units 7 & 8) and Tech Spec 3.3.2.1, action b, Functional unit 1.f, action 19 &

Functional unit 4.d, action 19 are applicable. Tech Spec requirement being to place the inoperable channel in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Once the plant is stable and the previous technical specifications have been documented by the CRS (or at Examiners direction), a loss of 2C Vital Bus will occur (Diff Protection). The crew will enter AB.4KV-0003 & AB.460-0003.

Once again, after the plant is stable, an RCS Leak into Containment (4 equivalent / 18,000 gpm) will occur. The leak results in a Reactor Trip, but an AUTO Safety Injection will fail to initiate, the crew will need to manually initiate Safety Injection.

A loss of Off-Site Power will simultaneously occur with the reactor trip, the crew will enter EOP-TRIP-1 and subsequently transition to EOP-LOCA-1. During TRIP-1 safeguards verification, the crew will recognize that 21 Charging Pump Tripped (motor/breaker damage - acrid odor) and 22 RHR Pump fails to start on the SEC sequence.

Crew will transition to FRCC-2, Degraded Core Cooling (>700°F CETS or RVLIS < 39%). Crew will be expected to depressurize SGs to inject the ECCS Accumulators.

The scenario will terminate when the accumulators have injected.

Procedure flow path: IOP-4, AR.ZZ-12, AR.ZZ-07, AB.ROD-0003, AR.ZZ-12, AB.4KV-0003 &

AB.460-0003, EOP-TRIP-1, LOCA-1, FRCC-2.

Appendix D Scenario Outline Form ES-D-1 Facility: SALEM_________

Scenario No.: #3_____________

Op-Test No.: JULY 2020__

Examiners: ___________________________ Operators:

Initial Conditions: 100% Power ___________________________________________________________________

Turnover: Maintain 100% Power, 21 SW Pump C/T for motor oil change. ( for simulator setup only - 22, 23, and 25 SW Pumps I/S, 26 SW Pump in auto)___________________________

Critical Tasks:

CT #1: Manually Trip the Reactor from the Control Room before transition to FRSM-1.

CT #2: Isolate the Faulted SG prior to transition out of LOSC-1.

Event No.

Malf.

No.

Event Type*

Event Description 1

RC0022A ATC (I),

CRS (TS)

RC Wide Range Pressure (PT-405) fails low. OHA alarm D-40 "SUBCLG CH A MARGIN LO". Tech Spec 3.3.3.7, action 1 (30 days) is applicable.

2 SW0215C SW0339F ATC/BOP (I), CRS (TS) 23 SW Pump Trip and 26 SW Pump fails to auto start due to failed high pressure transmitter. Crew will start 26 or another pump in manual. Tech Spec 3.7.4 (72 hrs) is applicable due to 21 & 23 SW pumps inoperable in one bay.

3 CN0117C ATC (R),

BOP (C) 23 Condensate Pump Trip Load Reduction to 85%

4 CV0036 ATC (I)

VCT Level Channel 114 fails high, causes high level divert to CVCS HUT via CV-

35. (Only indication will be lowering VCT level and possible auto makeup along with CV-35 position full divert) 5 BF0111A ALL (M) 21 FW Line Break inside containment.

6 RP0058 RP0059A ATC (C)

ATWT, requires opening the 2E6D and 2G6D breakers from the control room.

7 VL0023 VL0446 BOP (C) 21BF-19 and 21BF-13 fail to close. Both can be closed from the control console.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Scenario No.: #3 Target Quantitative Attributes per Scenario (See Section D.5.d)

Actual Attributes Event No.

1. Total malfunctions (5-8) 8 1-4, 6,7
2. Malfunctions after EOP entry (1-2) 2 6,7
3. Abnormal events (2-4) 3 2-4
4. Major transients (1-2) 1 5
5. EOPs entered/requiring substantive actions (1-2) 4 TRIP-1, LOSC-1, LOCA-1, TRIP-3
6. Entry into a contingency EOP with substantive actions ( 1 per scenario set) 0 N/A
7. Preidentified critical tasks (2) 2 6,7
8. Tech Specs exercised ( 2) 2 1,2

Simulator Exam Scenario # 3 Summary The evaluation begins with the plant at 100% power. While maintaining 100% power, OHA alarm D-40, SUBCLG CH A MARGIN LO alarms due to RC Wide Range Pressure (PT-405) failing low. Tech Spec 3.3.3.7, action 1 (30 days) is applicable.

Shortly after the previous tech spec call is made by the CRS (or at Examiners direction) #23 Service Water Pump trips and #26 Service Water Pump fails to start. The crew will start 26 or another pump in manual. Tech Spec 3.7.4 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) is applicable due to 21 & 23 SW Pumps inoperable in one bay. The crew may use alarm response procedure or AB.SW-0001 to mitigate.

At the examiners direction, #23 Condensate Pump trips requiring a load reduction to 85% IAW AB.CN-0001; Main Feedwater / Condensate System Abnormality.

Once the plant is stable (or at the Examiners direction), VCT Level Channel 114 fails high, causing high level divert to the CVCS HUT via CV-35. (Only indication will be lowering VCT level and possible auto makeup along with CV-35 position full divert) Crew may utilize AB.CVC-0001 to mitigate failure.

After the VCT level channel failure has been corrected, (CV-35 in manual to VCT), then a 21 SG Feed Water Line Break inside containment will occur. This will require a reactor trip, but an ATWS will result requiring the opening of the 2E6D and 2G6D breakers from the control room IAW EOP-TRIP-1.

21BF-19 and 21BF-13 will fail to close on Safety Injection signal, both can be closed from the control room. The crew will transition to LOSC-1 to isolate the faulted SG.

The scenario can be terminated once the transition to TRIP-3 is made and SI is terminated.

Procedure flow path: IOP-4, AR.ZZ-04, AR.ZZ-02, AB.SW-0001, AB.CVC-0001, EOP-TRIP-1, LOSC-1, LOCA-1, TRIP-3.

Appendix D Scenario Outline Form ES-D-1 Facility: SALEM_________

Scenario No.: #4_____________

Op-Test No.: JULY 2020__

Examiners: ___________________________ Operators:

Initial Conditions: 4% Reactor Power, BOL___________________________________________________________________

Turnover: IOP-3, Hot Standby to Minimum Load has been completed through step 4.3.15, all requirements for Mode 1 Entry are met and Mode Change is authorized. 21 SGFP is in service, steam dumps are in pressure control mode. Raise power to >15%

and place the Main Turbine online IAW S2.O-SO,TRB-0001(Q), Turbine Generator Startup Operations. (pre-brief evolutions)

Critical Tasks:

CT #1: Isolate feed flow into and steam flow from the ruptured SG before a transition to SGTR-3 occurs..

CT #2: Establish/maintain an RCS temperature so that transition from SGTR-1 does not occur because the RCS temperature is too high to maintain minimum required subcooling or below the RCS temperature that causes a RED path or a Purple path challenge to the subcriticality and/or the integrity CSF.

CT#3: Close the PZR PORV Block Valve before completion of RCS Pressure Check in SGTR-1.

Event No.

Malf.

No.

Event Type*

Event Description 1

N/A ATC (R),

BOP (N)

Raise power to >15% and place the Main Turbine online IAW S2.O-SO,TRB-0001(Q), Turbine Generator Startup Operations. (pre-brief evolutions) 2 RM0207A ATC (I),

CRS (TS)

Process Rad Monitor 2R1B fails high. Tech Spec 3.3.3.1, action 27 is applicable.

3 VC0137E BOP (C),

CRS (TS) 25 CFCU Trips. Tech Spec 3.6.2.3, action a, is applicable.

4 BF0105A ALL (C)

(Insert prior to exceeding P-10) 21 SGFP Trip, crew enters AB.CN-0001(Q), Main Feedwater/Condensate System Abnormality. Crew reduces power to less than 5%

and initiates Aux Feed to control SG levels. Stabilizes plant.

5 RC0012C ATC (C) 23 RCP High Vibration, crew enters AB.RCP-0001(Q), Reactor Coolant Pump Abnormality due to high shaft vibrations > 20 mils requiring stopping 23 RCP IAW. Trip Reactor, Confirm Reactor Trip, then stop RCP. Then go to TRIP-1. (Note if crew was unsuccessful in stabilizing plant <5% and remaining critical from loss of 21 SGFP, then inert this malfunction during immediate actions read through.) Crew transitions from TRIP-1 to TRIP-2, no SI required.

6 SG0078A ALL (M)

Once crew has transitioned to TRIP-2, insert 21 SGTR starting at 50 gpm and ramping to 650 gpm over 5 minutes. Expect crew to manually initiate SI and return to TRIP-1.

7 EL0134 ALL (C)

Loss of Off-site power occurs simultaneously with SI signal (either manual of auto).

8 SJ0184A ALL (C) 21 SI Pump fails to start on SEC 9

VL0297 VL0298 ALL (C)

PZR PORV (2PR1 or 2PR2) fails to close during RCS depressurization in EOP-SGTR-1. Operator can close PZR Block Valve.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Scenario No.: #4 Target Quantitative Attributes per Scenario (See Section D.5.d)

Actual Attributes Event No.

1. Total malfunctions (5-8) 7 2-5, 7-9
2. Malfunctions after EOP entry (1-2) 3 7-9
3. Abnormal events (2-4) 2 4,5
4. Major transients (1-2) 1 6
5. EOPs entered/requiring substantive actions (1-2) 3 TRIP-1, TRIP-2, SGTR-1
6. Entry into a contingency EOP with substantive actions ( 1 per scenario set) 0 N/A
7. Preidentified critical tasks (2) 3 6,7,9
8. Tech Specs exercised ( 2) 2 2,3

Simulator Exam Scenario # 4 Summary The evaluation begins with the plant at 4% power, BOL. IOP-3, Hot Standby to Minimum Load has been completed through step 4.3.15, all requirements for Mode 1 Entry are met and the Mode Change is authorized. 21 SGFP is in service, steam dumps are in pressure control mode.

The crew has been pre-briefed to raise power to >15% and place the Main Turbine online IAW S2.O-SO,TRB-0001(Q), Turbine Generator Startup Operations.

After taking the watch, 2R1B (Control Room Ventilation Rad Monitor) fails high. Tech Spec 3.3.3.1, action 27 is applicable. The crew may use OHA ARPs or AB.RAD-0001 to respond to failure.

Subsequently during the power ascension / after mode change, 25 CFCU trips. Tech Spec 3.6.2.3, action a, is applicable.

Prior to the crew exceeding 10% (P-10), 21 SGFP will trip. The crew will enter AB.CN-0001(Q),

Main Feedwater / Condensate Abnormality. The crew will reduce power to less than 5%, initiate Aux Feed to control SG Levels and stabilize the plant.

After the plant is stabilized on auxiliary feed, 23 RCP High Vibration alarm occurs. The crew enters AB.RCP-0001(Q), Reactor Coolant Pump Abnormality due to high shaft vibrations > 20 mils requiring stopping 23 RCP IAW Attachment 2. They Trip Reactor, Confirm Reactor Trip, then stop RCP. Then they go to TRIP-1. (Note if crew was unsuccessful in stabilizing plant

<5% and remaining critical from loss of 21 SGFP, then inert this malfunction during immediate actions read through.) Crew transitions from TRIP-1 to TRIP-2, no SI required.

Once crew has transitioned to TRIP-2, a 21 SGTR will start at 50 gpm and ramp up to 650 gpm over 5 minutes. The crew will manually initiate SI and return to TRIP-1. A loss of Off-Site Power occurs simultaneously with the SI signal (either manual or auto). 21 SI Pump fails to start on SEC. The crew will block the A SEC and start 21 SI Pump per EOP-TRIP-1.

The crew will transition to EOP-SGTR-1 and cooldown and depressurize to stop / reduce the primary to secondary leak.

PZR PORV (2PR1 or 2PR2) fails to close during the RCS depressurization in EOP-SGTR-1.

The crew can close the PZR Block Valve. The scenario may be terminated after the crew closes the PZR Block Valve. Examiner may continue scenario through SI termination.

Procedure flow path: IOP-3, AR.ZZ-01, AB.RAD-0001, AR.ZZ-02, AR.ZZ-11, AB.CN-0001, EOP-TRIP-1, TRIP-2, TRIP-1, SGTR-1.

Appendix D Scenario Outline Form ES-D-1 Facility: SALEM_________

Scenario No.:

  1. 5_____________ Op-Test No.: JULY 2020__

Examiners: ___________________________ Operators:

Initial Conditions: 75% EOC___________________________________________________________________

Turnover: Start 23 Condensate Pump and continue to raise reactor power to 90% IAW IOP-4, Power Operation.

Critical Tasks:

CT #1: Manually Trip the Turbine (in this case manually initiate MSLI) before a severe challenge develops to either the sub-criticality or the integrity CSF or before transition to LOSC-2, whichever happens first.

CT #2: Initiate RCS bleed and feed so that the RCS depressurizes sufficiently for SI pump injection into the RCS. This requires one PORV and RCS head vents open.

Event No.

Malf.

No.

Event Type*

Event Description 1

N/A ATC (R),

BOP (N)

Start 23 Condensate Pump and continue to raise reactor power to 90% IAW IOP-4, Power Operation.

2 TU0081H ALL (C) 24MS29 Turbine Control Valve fails closed. Per SO.TURB-0001, reduce turbine load to less than 75% at 5%/min (step 2.2.7).

3 TU0056 BOP (I),

CRS (TS)

PT-506, 1ST STG PRESS XMTR 506 fails Low. Will ARM the steam dump load rejection controller and ACTUATE the Channel II High Steam Line Flow bistables for all SGs. Tech Spec 3.3.2.1, action 19 is applicable, place in tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. (Functional Units 1.f and 4.d)

See S2.OP-SO.RPS-0006(Q), Main Turbine Channel Trip / Restoration section 5.2 4

ELO145 ALL (C),

CRS (TS)

Loss of 2B Vital Bus (Diff Protection)

Crew enters AB.4KV-0002 & AB.460-0002 5

VL0448 ALL (M) 23BF19 fails closed, Rx Trip on 23 SG Lo-Lo Level. (Manual or Auto)

EOP-TRIP-1 entered.

6 RP0073 RP0279A RP0279B ALL (C)

Main Turbine fails to trip. Auto Main Steam Line Isolation fails also, manual works.

Crew enters EOP-TRIP-1, probable SI signal due to all Channel II high steam flow channels already tripped and lo-lo Tavg (2/4) from cooldown (failure of turbine to initially trip). If SI signal occurs, Main Steam Line Isolation (MSLI) signal will also occur on same cooldown? Either way, a Manual MSLI will be required.

7 RP0275A RP0275B ALL (C)

Auto Phase A Fails to actuate on both trains. (Can be manually actuated). This will happen if auto SI occurs or if SI alone is manually actuated later in FRHS-1?

8 AF0181A ALL(C) 21 AFW Pump Trips 4 minutes after Rx Trip signal. If PO dispatched - (motor /

breaker damage - acrid odor) 9 AF0183 ALL (C) 23 AFW Pump Overspeed Trip. (Trip 23 AFW Pump when Phase A manual initiation has occurred or after TRIP-2 transition.) If PO dispatched - significant damage to trip / throttle valve and linkage. Cant be reset.

10 N/A ALL (M)

Crew transitions to FRHS-1 at step 20 of EOP-TRIP-1 or after above malfunction in TRIP-2.

11 VL0297 ALL (C) 2PR1 fails closed.

When 3/4 SG WR levels are < 32%, crew initiates bleed and feed at step 23 of FRHS-1. 2PR1 will not open, head vents must be opened along with 2PR2.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Scenario No.: #5 Target Quantitative Attributes per Scenario (See Section D.5.d)

Actual Attributes Event No.

1. Total malfunctions (5-8) 9 2-4, 6-11
2. Malfunctions after EOP entry (1-2) 6 7-9,11
3. Abnormal events (2-4) 3 2-4
4. Major transients (1-2) 1 5
5. EOPs entered/requiring substantive actions (1-2) 2 TRIP-1/TRIP-2, FRHS-1
6. Entry into a contingency EOP with substantive actions ( 1 per scenario set) 1 FRHS-1
7. Preidentified critical tasks (2) 2 6,11
8. Tech Specs exercised ( 2) 2 3,4

Simulator Exam Scenario # 5 Summary The evaluation begins with the plant at 75% power, EOC. The crew has been pre-briefed to start 23 Condensate Pump and continue to raise power to 90% IAW IOP-4, Power Operation.

During the power ascension, 24MS29 Turbine Control Valve fails closed. Per SO.TURB-0001, reduce turbine load to less than 75% at 5%/min (step 2.2.7).

PT-506, 1ST STG PRESS XMTR 506 fails Low. Will ARM the steam dump load rejection controller and ACTUATE the Channel II High Steam Line Flow bistables for all SGs. Tech Spec 3.3.2.1, action 19 is applicable, place in tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. (Functional Units 1.f and 4.d) See S2.OP-SO.RPS-0006(Q), Main Turbine Channel Trip / Restoration section 5.2 Loss of 2B Vital Bus (Diff Protection)

Crew enters AB.4KV-0001 & AB.460-0001 23BF19 fails closed, Rx Trip on 23 SG Lo-Lo Level. (Manual or Auto)

EOP-TRIP-1 entered.

Main Turbine fails to trip. Auto Main Steam Line Isolation fails also, manual works.

Crew enters EOP-TRIP-1, probable SI signal due to all Channel II high steam flow channels already tripped and lo-lo Tavg (2/4) from cooldown (failure of turbine to initially trip). If SI signal occurs, Main Steam Line Isolation (MSLI) signal will also occur on same cooldown? Either way, a Manual MSLI will be required.

Auto Phase A Fails to actuate on both trains. (Can be manually actuated). This will happen if auto SI occurs or if SI alone is manually actuated later in FRHS-1?

21 AFW Pump Trips 4 minutes after Rx Trip signal. If PO dispatched - (motor / breaker damage - acrid odor) 23 AFW Pump Overspeed Trip. (Trip 23 AFW Pump when Phase A manual initiation has occurred or after TRIP-2 transition.) If PO dispatched - significant damage to trip / throttle valve and linkage. Cant be reset.

Crew transitions to FRHS-1 at step 20 of EOP-TRIP-1 or after above malfunction in TRIP-2.

2PR1 fails closed. When 3/4 SG WR levels are < 32%, crew initiates bleed and feed at step 23 of FRHS-1. 2PR1 will not open, head vents must be opened along with 2PR2.

The scenario may be terminated after bleed and feed is successfully initiated.

Procedure flow path: IOP-4, S0.TRB-0001, AR.ZZ-13, SO.RPS-0006, AB.4KV-0002, AB.460-0002, EOP-TRIP-1, TRIP-2?, FRHS-1.

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