IR 05000382/2020008
| ML20315A149 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 08/26/2020 |
| From: | Greg Werner Operations Branch IV |
| To: | Entergy Operations |
| References | |
| Download: ML20315A149 (65) | |
Text
ES-401 PWR Examination Outline (RO) Form ES-401-2 Facility: Waterford 3 Date of Exam: August 26, 2020 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total Emergency and Abnormal Plant 2 1 2 2 N/A 1 1 N/A 2 9 4 Evolutions Tier Totals 4 4 5 5 5 4 27 10 1 2 2 2 5 2 2 4 2 3 2 2 28 5 Plant 2 1 0 2 1 1 0 1 1 1 1 1 10 3 Systems Tier Totals 3 2 4 6 3 2 5 3 4 3 3 38 8 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 3 2 3 Note: 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.) 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 point . 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 statement . 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 evolutio . 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, respectivel . Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categorie . 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/A . 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 exam . 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 catalo ** 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 pla Rev. 11
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 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #
000007 (EPE 7) Reactor Trip / 1 Knowledge of the operational implications of the following concepts as they apply to the Reactor Trip (CFR: 41.8 /
X 41.10, 45.3) EK1.02 Shutdown Margin 000008 (APE 8) Pressurizer Vapor Space AK2. Knowledge of the interrelations Accident / 3 between the Pressurizer Vapor Space Accident and the following: (CFR 41.7 /
X .7)
AK2.03 Controllers and positioners 000009 (EPE 9) Small Break LOCA / 3 EK3 Knowledge of the reasons for the following responses as the apply to the small break LOCA: (CFR 41.5 / 41.10 / 45.6 /
X 45.13) EK3.16 Containment temperature, pressure, humidity and level limits 000011 (EPE 11) Large Break LOCA / 3 EA1 Ability to operate and monitor the following as they apply to a Large Break LOCA:
X (CFR 41.7 / 45.5 / 45.6) EA1.16 Balancing of HPI loop flows 000015 (APE 15) Reactor Coolant Pump AA2. Ability to determine and interpret the Malfunctions / 4 following as they apply to the Reactor Coolant Pump Malfunctions X (Loss of RC Flow): (CFR 43.5/ 45.13) AA2.01 Cause of RCP failure 000022 (APE 22) Loss of Reactor Coolant 2.2.44 Ability to interpret control room Makeup / 2 indications to verify the status and operation X of system conditions. (CFR: 41.5 / 43.5 /
45.12)
000025 (APE 25) Loss of Residual Heat Removal System / 4 Not Sampled 000026 (APE 26) Loss of Component AA2 Ability to determine and/or interpret the Cooling Water / 8 following as they apply to Loss of Component Cooling Water: (CFR: 41.10 /
X 43.5 / 45.13) AA2.06 The length of time after the loss of CCW flow to a component before that component may be damaged 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 Not Sampled Rev. 11
ES-401 3 Form ES-401-2 000029 (EPE 29) Anticipated Transient EA1 Ability to operate and/or monitor the Without Scram / 1 following as they apply to an Anticipated Transient Without Scram: (CFR: 41.7 / 45.5 /
X 45.6) EA1.12 M/G set power supply and reactor trip breakers 000038 (EPE 38) Steam Generator Tube EK3 Knowledge of the reasons for the Rupture / 3 following responses as the apply to the X SGTR: (CFR 41.5 / 41.10 / 45.6 / 45.13) EK3.09 Criteria for securing/throttling ECCS 000040 (APE 40; BW E05; CE E05; W E12)
EK2. Knowledge of the interrelations Steam Line RuptureExcessive Heat between the (Excess Steam Demand)
Transfer / 4 and the following: CFR: 41.7 / 45.7)
X EK2.1 Components, and functions of control 3.3 10 and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
000054 (APE 54; CE E06) Loss of Main EK1. Knowledge of the operational Feedwater /4 implications of the following concepts as they apply to the (Loss of Feedwater) (CFR:
41.8 / 41.10 / 45.3)
X 3.2 11 EK1.2 Normal, abnormal and emergency operating procedures associated with (Loss of Feedwater).
000055 (EPE 55) Station Blackout / 6 EK1 Knowledge of the operational implications of the following concepts as they apply to the Station Blackout :
X X (CFR 41.8 / 41.10 / 45.3) 3.3 12 EK1.01 Effect of battery discharge rates on capacity 000056 (APE 56) Loss of Offsite Power / 6 AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: (CFR 41.5,41.10 /
X 45.6 / 45.13) 4.4 13 AK3.02 Actions contained in EOPs for loss of offsite power 000057 (APE 57) Loss of Vital AC AA1 Ability to operate and/or monitor the Instrument Bus / 6 following as they apply to Loss of Vital AC Electrical Instrument Bus: (CFR: 41.7 / 45.5 /
X 45.6) 3.5 14 AA1.06 Manual control of components for which automatic control is lost 000058 (APE 58) Loss of DC Power / 6 AA2 Ability to determine and/or interpret the following as they apply to Loss of DC Power:
X (CFR: 43.5 / 45.13) 3.4 15 AA2.02 125V dc bus voltage, low/critical low, alarm Rev. 11
ES-401 4 Form ES-401-2 000062 (APE 62) Loss of Nuclear Service 2.2.39 Knowledge of less than or equal to one Water / 4 X hour technical action statements for system .9 16 (CFR: 41.7 / 41.10 / 43.5 / 45.3)
000065 (APE 65) Loss of Instrument Air / 8 AA2 Ability to determine and/or interpret the following as they apply to Loss of Instrument X Air: (CFR: 41.10 / 43.5 / 45.13) 2.9 17 AA2.01 Cause and effect of low-pressure instrument air alarm 000077 (APE 77) Generator Voltage and AA2 Ability to determine and/or interpret the Electric Grid Disturbances / 6 following as they apply to Generator Voltage X and Electric Grid Disturbances: (CFR: 41.5 / 3.6 18 43.5 / 45.5 / 45.7 / 45.8)
AA2.04 VARs outside the capability curve (W E04) LOCA Outside Containment / 3 Not Applicable (W E11) Loss of Emergency Coolant Recirculation / 4 Not Applicable (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 Not Applicable K/A Category Totals: 3 2 3 4 4 2 Group Point Total: 18 Rev. 11
ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 RO E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #
000001 (APE 1) Continuous Rod Withdrawal / 1 000003 (APE 3) Dropped Control Rod / 1 AA2 Ability to determine and/or interpret the following as they apply to a Dropped Control Rod: (CFR: 43.5 /
45.13)
X AA2.03 Dropped rod, using in-core/ex-core instrumentation, in-core or loop temperature measurements 000005 (APE 5) Inoperable/Stuck Control Rod / 1 2.4.6 Knowledge of X EOP mitigation strategie .7 20 (CFR: 41.10 / 43.5 / 45.13)
000024 (APE 24) Emergency Boration / 1 AK2 Knowledge of the relationship between the Emergency Boration and the X following systems or components: (CFR: 41.7 /
45.7)
AK2.04 Pumps 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand X how operator actions and directives affect plant and system conditions. (CFR:
41.5 / 43.5 / 45.12)
000037 (APE 37) Steam Generator Tube Leak / 3 AK3 Knowledge of the reasons for the following responses as they apply to a Steam Generator Tube Leak:
X (CFR: 41.5 / 41.7 /41.10 /
45.6 / 45.13)
AK3.07 Actions contained in EOP for S/G tube leak 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms
/7 Rev. 11
ES-401 6 Form ES-401-2 000067 (APE 67) Plant Fire On Site / 8 AK1. Knowledge of the operational implications of the following concepts as they apply to Plant Fire On X 3.1 24 Site. (CFR 41.8 / 41.10 /
45.3)
AK1.02 Fire Fighting 000068 (APE 68; BW A06) Control Room Evacuation / 8 AA1 Ability to operate and/or monitor the following as they apply to Control Room X Evacuation: (CFR: 41.7 / 4 .4 25
/ 45.6)
AA1.12 Auxiliary Shutdown panel controls and indicators 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 Not Applicable (W E13) Steam Generator Overpressure / 4 Not Applicable (W E15) Containment Flooding / 5 Not Applicable (W E16) High Containment Radiation /9 Not Applicable (BW A01) Plant Runback / 1 Not Applicable (BW A02 & A03) Loss of NNI-X/Y/7 Not Applicable (BW A04) Turbine Trip / 4 Not Applicable (BW A05) Emergency Diesel Actuation / 6 Not Applicable (BW A07) Flooding / 8 Not Applicable (BW E03) Inadequate Subcooling Margin / 4 Not Applicable (BW E08; W E03) LOCA CooldownDepressurization / 4 Not Applicable (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 Not Applicable (BW E13 & E14) EOP Rules and Enclosures Not Applicable (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 AK2 Knowledge of the relationship between the Excess RCS Leakage and the following systems or components: (CFR: 41.7 /
45.7)
X 3.2 26 AK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual feature Rev. 11
ES-401 7 Form ES-401-2 (CE E09) Functional Recovery EK3 Knowledge of the reasons for the following responses as they apply to Functional Recovery: (CFR:
41.5 / 41.10, 45.6, 45.13)
EK3.1 Facility operating characteristics during X 3.5 27 transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.
(CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 K/A Category Point Totals: 1 2 2 1 1 2 Group Point Total: 9 Rev. 11
ES-401 8 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 RO System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #
003 (SF4P RCP) Reactor Coolant X K1 Knowledge of the physical Pump connections and/or cause and effect relationships between the Reactor Coolant Pump System and the following systems: (CFR: 41.2 to 41.9 / 45.7 / 45.8)
K1.01 RCP lube oil 004 (SF1; SF2 CVCS) Chemical and X K2 Knowledge of bus power supplies to Volume Control the following: (CFR: 41.7)
K2.06 Control instrumentation 005 (SF4P RHR) Residual Heat X X K3 Knowledge of the effect that a loss or Removal malfunction of the Residual Heat Removal System will have on the following systems or system parameters:
(CFR: 41.7 / 45.6)
K3.01 RCS K6 Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: (CFR: 41.7 /
45.7)
K6.03 RHR heat exchanger 006 (SF2; SF3 ECCS) Emergency X K4 Knowledge of Emergency Core Core Cooling Cooling System design feature(s) and/or interlock(s), which provide for the following: (CFR: 41.7 / 41.8)
K4.17 Safety Injection valve interlocks K5 Knowledge of the operational 007 (SF5 PRTS) Pressurizer X Relief/Quench Tank implications of the following concepts as the apply to PRTS: (CFR: 41.5 / 45.7)
K5.02 Method of forming a steam bubble in the PZR K4 Knowledge of CCWS design 008 (SF8 CCW) Component Cooling X X Water feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
K4.02 Operation of the surge tank, including the associated valves and controls A1 Ability to predict and/or monitor 010 (SF3 PZR PCS) Pressurizer X Pressure Control changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including: (CFR: 41.5 / 45.5)
A1.06 RCS heatup and cooldown effect on pressure Rev. 11
ES-401 9 Form ES-401-2 A2 Ability to (a) predict the impacts of 012 (SF7 RPS) Reactor Protection X X 3.1 36 the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR:
41.5 / 43.5 / 45.3 / 45.5)
A2.01 Faulty bistable operation K4 Knowledge of Reactor Protection System design feature(s) and/or 3.9 37 interlock(s), which provide for the following: (CFR: 41.7)
K4.02 Automatic reactor trip when RPS setpoints are exceeded for each RPS function; functional basis for each A3 Ability to monitor automatic 013 (SF2 ESFAS) Engineered X X 3.7 38 Safety Features Actuation operation of the ESFAS including: (CFR:
41.7 / 45.5)
A3.01 Input channels and logic 2.4.31 Knowledge of annunciator alarms, 4.2 55 indication, OR response procedure (CFR: 41.10 / 45.3 022 (SF5 CCS) Containment Cooling X A4 Ability to manually operate and/or 3.6 39 monitor in the control room: (CFR: 41.7 /
45.5 to 45.8)
A4.02 CCS pumps 025 (SF5 ICE) Ice Condenser Not Applicable 026 (SF5 CSS) Containment Spray X 2.4.46 Ability to verify that the alarms 4.2 40 are consistent with the plant condition (CFR: 41.10 / 43.5 / 45.3 / 45.12)
039 (SF4S MSS) Main and Reheat X X A4 Ability to manually operate and/or 3.8 41 Steam monitor in the control room: (CFR: 41.7 /
45.5 to 45.8)
A4.04 Emergency feedwater pump turbines A1 Ability to predict and/or monitor changes in parameters (to prevent 3.2 42 exceeding design limits) associated with operating the MRSS controls including:
(CFR: 41.5 / 45.5)
A1.05 RCS T-ave 059 (SF4S MFW) Main Feedwater X A3 Ability to monitor automatic features 2.5 43 of the Main Feedwater System, including: (CFR: 41.7 / 45.5)
A3.04 Turbine driven feed pump Rev. 11
ES-401 10 Form ES-401-2 A2 Ability to (a) predict the impacts of 061 (SF4S AFW) X X 3.1 44 Auxiliary/Emergency Feedwater the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR:
41.5 / 43.5 / 45.3 / 45.13)
A2.05 Automatic control malfunction K5 Knowledge of the operational implications of the following concepts as the apply to the AFW: (CFR: 41.5 / 2.7 45 45.7)
K5.05 Feed line voiding and water hammer A1 Ability to predict and/or monitor 062 (SF6 ED AC) AC Electrical X 2.5 46 Distribution changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: (CFR: 41.5 / 45.5)
A1.03 Effect on instrumentation and controls of switching power supplies K4 Knowledge of DC electrical system 063 (SF6 ED DC) DC Electrical X X 2.9 47 Distribution design feature(s) and/or interlock(s)
which provide for the following: (CFR:
41.7)
K4.02 Breaker interlocks, permissives, bypasses and cross-ties K6 Knowledge of the effect of a loss or 064 (SF6 EDG) Emergency Diesel X X 2.7 48 Generator malfunction of the following will have on the ED/G system: (CFR: 41.7 / 45.7)
K6.07 Air receivers K4 Knowledge of PRM system design 073 (SF7 PRM) Process Radiation X X 4.0 49 Monitoring feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)
K4.01 Release termination when radiation exceeds setpoint A1 Ability to predict and/or monitor changes in parameters (to prevent 3.2 50 exceeding design limits) associated with operating the PRM system controls including: (CFR: 41.5 / 45.7)
A1.01 Radiation levels Rev. 11
ES-401 11 Form ES-401-2 076 (SF4S SW) Service Water X K3 Knowledge of the effect that a loss or malfunction of the Service Water System will have on the following systems or system parameters: (CFR: 41.7 / 45.6)
K3.02 Secondary closed cooling water 078 (SF8 IAS) Instrument Air X X X K2 Knowledge of bus power supplies to the following: (CFR: 41.7)
K2.01 Instrument Air Compressor A3 Ability to monitor automatic operation of the IAS, including: (CFR:
41.7 / 45.5) A3.01 Air pressure 103 (SF5 CNT) Containment X K1 Knowledge of the physical connections and/or cause and effect relationships between the Containment System and the following systems: (CFR:
41.9 / 45.7 / 45.8)
K1.01 CCS 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals: 2 2 2 5 2 2 4 2 3 2 2 Group Point Total: 28 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 RO System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #
001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor K3 Knowledge of the effect that a loss or Coolant malfunction of the Reactor Coolant System will have on the following X systems or system parameters: (CFR:
41.7)
K3.03 Containment 011 (SF2 PZR LCS) Pressurizer K1 Knowledge of the physical Level Control connections and/or cause and effect relationships between the Fire Protection X System and the following systems: (CFR: .4 / 41.7 / 41.8 / 45.7 / 45.8)
K1.05 Reactor regulating system 014 (SF1 RPI) Rod Position Indication Rev. 11
ES-401 12 Form ES-401-2 015 (SF7 NI) Nuclear A3 Ability to monitor automatic Instrumentation operation of the NIS, including: (CFR:
X 41.7 / 45.5) 3.9 62 A3.03 Verification of proper functioning/operability 016 (SF7 NNI) Nonnuclear K3 Knowledge of the effect that a loss or Instrumentation malfunction of the Nonnuclear Instrumentation System will have on the X 2.6 57 following systems or system parameters:
(CFR: 41.7 / 45.6)
K3.04 MFW system 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen K2 Knowledge of bus power supplies to Recombiner and Purge Control X X the following: (CFR: 41.7) 2.5 58 K2.01 Hydrogen recombiners 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator K5 Knowledge of the operational implications or cause and effect relationships of the following concepts as X they apply to the Steam Generator 3.4 59 System: (CFR: 41.5 / 45.7)
K5.01 Effect of secondary parameters, pressure, and temperature on reactivity 041 (SF4S SDS) Steam K6 Knowledge of the effect of a loss or Dump/Turbine Bypass Control malfunction on the following will have on the SDS: (CFR: 41.7 / 45.7)
X 2.7 60 K6.03 Controller and positioners, including ICS, S/G, CRDS 045 (SF 4S MTG) Main Turbine A1 Ability to predict and/or monitor Generator changes in parameters associated with operation of the Main Turbine Generator X 3.3 61 System, including: (CFR: 41.5 / 45.5)
A1.06 Expected response of secondary plant parameters following T/G trip 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate A2 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 X X 2.5 63 consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 /
45.13)
A2.04 Loss of condensate pumps Rev. 11
ES-401 13 Form ES-401-2 068 (SF9 LRS) Liquid Radwaste 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and X 3.8 64 the resultant operational effects. (CFR:
41.10 / 43.5 / 45.13 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals: 1 1 2 0 1 0 1 1 1 1 1 Group Point Total: 10 Rev. 11
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Waterford 3 Date of Exam: August 26, 2020 Category K/A # Topic RO SRO-only IR # IR #
2.1.20 Ability to interpret and execute procedure steps. (CFR: .10 / 43.5 / 45.12)
1. Conduct of Operations 2.1.41 2.1.41 Knowledge of the refueling proces .8 67 Subtotal 2 2.2.15 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, lineups, or tag-outs. (Reference Potential) (CFR: 41.10 / 43.3 / 45.13)
2. Equipment Control 2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.3 / 45.2)
2.2.35 Ability to determine TS for mode of operation. (CFR: .7 / 41.10 / 43.2 / 45.13)
Subtotal 3 2.3.11 Ability to control radiation releases. (CFR: 41.11 / 4 .8 71
/ 45.10)
2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry 3. Radiation Control requirements, fuel-handling responsibilities, access to locked high-radiation areas, or alignment of filter (CFR: 41.12 / 45.9 / 45.10)
Subtotal 2 2. Knowledge of EOP entry conditions and immediate action steps. (CFR: 41.10 / 43.5 / 45.13)
2. Knowledge of low power/shutdown implications in accident (e.g., loss of coolant / accident or loss of residual heat removal) mitigation strategies. (CFR: . Emergency Procedures/Plan 41.10 / 43.5 / 45.13)
2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: .10 / 43.5 / 45.13)
Subtotal 3 Tier 3 Point Total 10 Rev. 11
ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 Date of Exam: August 26, 2020 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total Emergency and Abnormal Plant 2 N/A N/A 9 2 2 4 Evolutions Tier Totals 27 5 5 10 1 28 3 2 5 Plant 2 10 2 1 3 Systems Tier Totals 38 5 3 8 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 1 2 Note: 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.) 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 point . 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 statement . 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 evolutio . 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, respectivel . Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categorie . 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/A . 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 exam . 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 catalo ** 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 pla Rev. 11
ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 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 000008 (APE 8) Pressurizer Vapor Space Accident / 3 000009 (EPE 9) Small Break LOCA / 3 000011 (EPE 11) Large Break LOCA / 3 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 000022 (APE 22) Loss of Reactor Coolant AA2 Ability to determine and interpret the Makeup / 2 following as they apply to Loss of Reactor X Coolant Makeup: (CFR: 43.5 / 45.13) AA2.02 Charging pump problems 000025 (APE 25) Loss of Residual Heat 2.2.25 Knowledge of the bases in TS for Removal System / 4 X limiting conditions for operations and safety limits. (SRO Only) (CFR: 43.2)
000026 (APE 26) Loss of Component Cooling Water / 8 000027 (APE 27) Pressurizer Pressure AA2 Ability to determine and interpret the Control System Malfunction / 3 following as they apply to a Pressurizer Pressure Control System Malfunction: (CFR:
X 43.5 / 45.13) AA2.02 Normal values for RCS pressure 000029 (EPE 29) Anticipated Transient Without Scram / 1 000038 (EPE 38) Steam Generator Tube 2.1.20 Ability to interpret and execute Rupture / 3 X procedure steps. (CFR: 41.10 / 43.5/ 45.12/)
000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 000054 (APE 54; CE E06) Loss of Main Feedwater /4 000055 (EPE 55) Station Blackout / 6 000056 (APE 56) Loss of Offsite Power / 6 000057 (APE 57) Loss of Vital AC 2.4.20 Knowledge of the operational Instrument Bus / 6 X implications of EOP warnings, cautions, and notes. (CFR: 41.10 / 41.5 / 45.13)
000058 (APE 58) Loss of DC Power / 6 AA2.03 Ability to determine and interpret the following as they apply to Loss of DC Power:
X DC loads lost; impact on ability to operate and monitor plant systems (43.5 / 45.13)
000062 (APE 62) Loss of Nuclear Service Water / 4 000065 (APE 65) Loss of Instrument Air / 8 Rev. 11
ES-401 3 Form ES-401-2 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 (W E04) LOCA Outside Containment / 3 Not Applicable (W E11) Loss of Emergency Coolant Recirculation / 4 Not Applicable (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 Not Applicable K/A Category Totals: 3 3 Group Point Total: 6 Rev. 11
ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 SRO E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #
000001 (APE 1) Continuous Rod Withdrawal / 1 2.1.28 Knowledge of the purpose and function of major X system components and controls (CFR: 41.7)
000003 (APE 3) Dropped Control Rod / 1 000005 (APE 5) Inoperable/Stuck Control Rod / 1 000024 (APE 24) Emergency Boration / 1 000028 (APE 28) Pressurizer (PZR) Level Control AA2 Ability to determine and Malfunction / 2 interpret the following as they apply to Pressurizer Level X Control Malfunctions: (CFR:
43.5 / 45.13)
AA2.11 Leak in PZR 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 000037 (APE 37) Steam Generator Tube Leak / 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 2.1.31 Ability to locate control room switches, controls, and indications, and X to determine that they correctly reflect the desired plant lineup. (CFR: 41.10 /
45.12)
000060 (APE 60) Accidental Gaseous Radwaste Release / 9 000061 (APE 61) Area Radiation Monitoring System Alarms
/7 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity / 5 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
000076 (APE 76) High Reactor Coolant Activity / 9 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination / 3 Not Applicable (W E13) Steam Generator Overpressure / 4 Not Applicable (W E15) Containment Flooding / 5 Not Applicable (W E16) High Containment Radiation /9 Not Applicable (BW A01) Plant Runback / 1 Not Applicable (BW A02 & A03) Loss of NNI-X/Y/7 Not Applicable (BW A04) Turbine Trip / 4 Not Applicable (BW A05) Emergency Diesel Actuation / 6 Not Applicable (BW A07) Flooding / 8 Not Applicable (BW E03) Inadequate Subcooling Margin / 4 Not Applicable (BW E08; W E03) LOCA CooldownDepressurization / 4 Not Applicable (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 Not Applicable (BW E13 & E14) EOP Rules and Enclosures Not Applicable Rev. 11
ES-401 5 Form ES-401-2 (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 (CE A16) Excess RCS Leakage / 2 (CE E09) Functional Recovery EA2. Ability to determine and interpret the following as they apply to the (Functional Recovery) (CFR: 43.5 /
45.13)
X 4.4 85 EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
(CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 EA2 Ability to determine and/or interpret the following as they apply to Loss of Forced Circulation and/or X LOOP and/or a Blackout:
(CFR: 41.10 / 43.5 / 45.13)
EA2.7 RCS subcooling K/A Category Point Totals: 2 2 Group Point Total: 4 Rev. 11
ES-401 6 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 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 004 (SF1; SF2 CVCS) Chemical and X A2 Ability to (a) predict the impacts of Volume Control the following malfunctions on the Chemical and Volume Control System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 to 41.7 / 43.5 /
45.3 / 45.5)
A2.18 High VCT level 005 (SF4P RHR) Residual Heat Removal 006 (SF2; SF3 ECCS) Emergency Core Cooling 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 008 (SF8 CCW) Component Cooling X 2.2.22 Knowledge of limiting conditions Water for operations and safety limits (CFR:
41.5 / 43.2 / 45.2)
010 (SF3 PZR PCS) Pressurizer Pressure Control 012 (SF7 RPS) Reactor Protection 013 (SF2 ESFAS) Engineered Safety Features Actuation 022 (SF5 CCS) Containment Cooling 025 (SF5 ICE) Ice Condenser Not Applicable A2 Ability to (a) predict the impacts of 026 (SF5 CSS) Containment Spray X the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR:
41.5 / 43.5 / 45.3 / 45.13)
A2.07 Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding), or sump level below cutoff (interlock) limit 039 (SF4S MSS) Main and Reheat Steam 059 (SF4S MFW) Main Feedwater 061 (SF4S AFW)
Auxiliary/Emergency Feedwater Rev. 11
ES-401 7 Form ES-401-2 062 (SF6 ED AC) AC Electrical X 2.2.36 Ability to analyze the effect of Distribution maintenance activities, such as degraded power sources, on the status of limiting conditions of operations. (CFR: 41.10 /
43.2 / 45.13)
063 (SF6 ED DC) DC Electrical Distribution A2 Ability to (a) predict the impacts of 064 (SF6 EDG) Emergency Diesel X Generator 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: (CFR: 41.5 / 43.5 / 45.3 /
45.13)
A2.22 Potential automatic sequences (CO2/Water) and electrical damage (loose wires)
073 (SF7 PRM) Process Radiation Monitoring 076 (SF4S SW) Service Water 078 (SF8 IAS) Instrument Air 103 (SF5 CNT) Containment 053 (SF1; SF4P ICS*) Integrated Control K/A Category Point Totals: 3 2 Group Point Total: 5 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 002 (SF2; SF4P RCS) Reactor Coolant 011 (SF2 PZR LCS) Pressurizer A2 Ability to (a) predict the impacts of Level Control the following malfunctions or operations on the PZR LCS; and (b)
based on those predictions, use procedures to correct, control, or mitigate X the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 /
45.13)
A2.04 Loss of one, two, or three charging pumps 014 (SF1 RPI) Rod Position Indication 015 (SF7 NI) Nuclear Instrumentation Rev. 11
ES-401 8 Form ES-401-2 016 (SF7 NNI) Nonnuclear Instrumentation 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control 029 (SF8 CPS) Containment Purge 033 (SF8 SFPCS) Spent Fuel Pool 2.4.8 Knowledge of how abnormal Cooling operating procedures are used in X 4.5 92 conjunction with EOPs. (CFR: 41.10 /
43.5/ 45.13)
034 (SF8 FHS) Fuel-Handling Equipment 035 (SF 4P SG) Steam Generator 041 (SF4S SDS) Steam Dump/Turbine Bypass Control 045 (SF 4S MTG) Main Turbine Generator 055 (SF4S CARS) Condenser Air Removal 056 (SF4S CDS) Condensate 068 (SF9 LRS) Liquid Radwaste A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System ; and (b)
based on those predictions, use procedures to correct, control, or mitigate X 3.3 93 the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 /
45.13)
A2.04 Failure of automatic isolation 071 (SF9 WGS) Waste Gas Disposal 072 (SF7 ARM) Area Radiation Monitoring 075 (SF8 CW) Circulating Water 079 (SF8 SAS**) Station Air 086 Fire Protection 050 (SF 9 CRV*) Control Room Ventilation K/A Category Point Totals: 2 1 Group Point Total: 3 Rev. 11
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Date of Exam:
Category K/A # Topic RO SRO-only IR # IR #
2.1.36 Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 / 43.6 / 45.7)
1. Conduct of 2.1.39 Knowledge of conservative decision making practices.
Operations (CFR: 41.10 / 43.5 / 45.12)
Subtotal 2 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / .2 / 45.13)
2. Equipment 2.2.43 Knowledge of the process used to track inoperable Control alarms. (CFR: 41.10 / 43.5 / 45.13)
Subtotal 2 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12 / 43.4 / 45.10)
3. Radiation Control Subtotal 1 2. Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR:
41.10 / 43.2 / 45.6)
4. Emergency Procedures/Plan 2.4.42 Knowledge of emergency response facilities. (CFR: .10 / 45.11)
Subtotal 2 Tier 3 Point Total 7 Rev. 11
ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Selected Reason for Rejection Group K/A 1/1 000007 EK1.02 EK1.02 is not listed in rev 2 of K/A catalog. Randomly resampled Reactor Trip,/ 1 RO and selected EK1.2 for procedure question 1 Changed K/A from 007 (CE02) EK 1.2 Reactor Trip recovery because it was hard to write an RO question on this subject that encompassed normal, abnormal and EOP procedure Randomly resampled and selected 007 EK1.02 Reactor Trip.
1/1 000008 (APE 8) AK2.09 is not listed in rev 2 of K/A catalog. Randomly Pressurizer Vapor resampled and selected AK2.03 Controllers and positioners Space Accident / 3 RO question 2 1/1 000015 (APE 15) AA2.14 is not listed in rev 2 of K/A catalog. Randomly Reactor Coolant resampled and selected AA2.01 Cause of RCP failur Pump Malfunctions
/ 4 RO question 5 1/1 022 G2.2.44 RO NRC Chief examiner identified 022 G2.3.12 does not meet question 6 section ES-401 D.1.b of NUREG 1021 and was randomly replaced with 022 G2.44.
1/1 000029 (EPE 29) EA1.17 not listed in rev 2 of K/A catalog. Randomly resampled and Anticipated selected EA1.03 Charging pump suction valves from VCT operating Transient Without switc Scram / 1 RO question 8 1/1 0029 EA1.12 RO 029 EA1.03 rejected. Could not find a relationship between an question 8 ATWS and the Charging Pump suction from the VCT operating switch for W3. Randomly selected another EA1 item for 02 Randomly selected 029 EA1.12.
1/1 000038 (EPE 38) EK3.11 not listed in rev 2 of K/A catalog. Randomly resampled and Steam Generator selected EK3.09 Criteria for securing/throttling ECCS Tube Rupture / 3 RO question 9 1/1 000040 (APE 40; AK2.17 is a Westinghouse APE K/A. Resampled and replaced with BW E05; CE E05; CE05 EK W E12) Steam Line Rupture Excessive Heat Transfer / 4 RO question 10 0040 CE05 RO CE05 EK2.2 rejected. Could not develop a question referring to question 10 emergency coolant and heat removal systems that did not overlap a question presented in the last 2 exams. See 2018 RO11, 2017 RO11, 2017 RO27, 2017 RO33 and RO54 of the 2020 exam. Randomly selected another K/A from CE05 EK2.2 fiel The only other number in this field is CE05 EK Rev. 11
ES-401 Record of Rejected K/As Form ES-401-4 1/1 000054 (APE 54; AK1.04 is a Westinghouse APE K/A. Resampled and replaced CE E06) Loss of with CE06 EK Main Feedwater /4 RO question 11 1/1 000055 (EPE 55) EK2.09 is not listed in rev 2 of the K/A catalog. No EK 2 K/As Station Blackout / 6 have an IR of 2.5 or higher. Randomly resampled and selected RO question 12 EK1.01.
1/1 062 2.2.39 RO 062 2.2.17 rejected. K/A 2.2.17 was already selected for RO69 of question 16 this exam. A question is already written for RO69 and finding another question in the same risk assessment procedure (OI-037-000) on a RO level was not possible. Randomly selected another generic K/A from section 2.0. Randomly selected 2.2.39.
1/2 000003 (APE 3) AA2.11 is not listed in rev 2 of the K/A catalog. Randomly Dropped Control resampled and selected AA2.0 Rod / 1 RO question 19 1/2 005 2.4.6 question 005 2.4.26 rejected. There is no tie at W3 for a stuck/inoperable 20 control to knowledge of facility protection requirements, including fire brigade and portable firefighting equipment usage. Randomly selected another 2.4 generic item for 00 Randomly selected 005 2.4.6.
1/2 000024 (APE 24) AK2.07 is not listed in rev 2 of the K/A catalog. Randomly Emergency resampled and selected AK2.04 Pump Boration / 1 RO question 21 1/2 036 2.2.44 RO 036 2.2.3 rejected. This K/A is specifically for a multi-unit question 22 license. Waterford 3 Is not a multi-unit license. Randomly selected another 2.2 generic item for 036. Randomly selected 036 2.2.44.
1/2 00067 AK1.02 RO 060 AK1.04 rejected. Calculation of offsite dose is an SRO task question 24 and an RO question could not be developed for this K/ Randomly selected another system in Tier 1/Group 2 that had not been used and then randomly selected between AK1.01 and AK1.02. Randomly selected 067 AK1.02.
1/2 000068 (APE 68; AA1.33 is not listed in rev 2 of the K/A catalog. Randomly BW A06) Control resampled and selected AA1.14 Reactor Trip Breakers and Room Evacuation / Switches 8 RO question 25 1/2 068 AA1.12 RO 068 AA1.14 rejected. No direction is provided in the W3 control question 25 room evacuation procedure, OP-901-502, for operation of reactor trip breakers or switches outside of tripping the reactor before leaving the control room. This is only one step and an RO question could not be developed from it. Randomly selected another AA1 item for 068. Randomly selected 068 AA1.1 Rev. 11
ES-401 Record of Rejected K/As Form ES-401-4 1/2 (CE A16) Excess AK2.8 is not listed in rev 2 of the K/A catalog. Randomly RCS Leakage / 2 resampled and selected AK RO question 26 1/2 (CE E09) EK3.30 is not listed in rev 2 of the K/A catalog. Randomly Functional resampled and selected EK Recovery RO question 27 2/1 005 (SF4P RHR) K3.02 was less than 2.5 IR on rev 2 of K/A catalog. Randomly Residual Heat selected new K/A K3.01 RC Removal RO question 30 2/1 005 (SF4P RHR) K6.11 was less than 2.5 IR on rev 2 of K/A catalog. Randomly Residual Heat resampled and replaced with K/A K6.03 RHR heat exchange Removal RO question 31 2/1 008 (SF8 CCW) None of the K6 topics were 2.5 or higher IR on rev 2 of k/a Component Cooling catalog. Randomly resampled and replaced with K4.02 Water RO question Operation of the surge tank, including the associated valves 34 and controls.
2/1 010 (SF3 PZR A1.10 PZR liquid temperature not listed in rev 2. Randomly PCS) Pressurizer resampled and replaced with A1.06 RCS heatup and cooldown Pressure Control effect on pressur RO question 35 2/1 039 (SF4S MSS) A4.05 MSR startup has an IR of 1.8 on rev 2, randomly Main and Reheat resampled and replaced with A4.04 Emergency feedwater Steam RO question pump turbines
2/1 039 (SF4S MSS) A1.07 only has IR of 2.4 on rev 2, randomly resampled and Main and Reheat replaced with A1.06 Main steam pressure Steam RO question
2/1 039 A1.05 RO 039 A1.06 Main Steam Pressure rejected. Could not develop an Question 42 RO question to predict or monitor Main Steam pressure due to overlap with this exam and the previous two exams. See RO61 on this exam and 2017 RO43. Randomly selected another K/A from system 039 in the A1 field. Randomly selected 039 A1.05.
2/1 059 (SF4S MFW) A3.09 MFW pump trips is not listed in rev 2. Randomly Main Feedwater RO resampled and replaced with A3.04 Turbine driven feed pum question 43 2/1 061 (SF4S AFW) A3.03 Automatic AFW S/G level control should have been Auxiliary/Emergenc selected from an A2 K/A. Replaced with A2.05 Automatic y Feedwater RO control malfunction question 44 2/1 061 (SF4S AFW) K5.07 Back leakage through discharge check valves is not Auxiliary/Emergenc listed in rev 2; randomly resampled and replaced with K5.05 y Feedwater RO Feed line voiding and water hamme question 45 Rev. 11
ES-401 Record of Rejected K/As Form ES-401-4 2/1 062 (SF6 ED AC) A1.07 Inverter outputs has IR rating of 2.4 in rev 2; randomly AC Electrical resampled and replaced with A1.03 Effect on instrumentation Distribution RO and controls of switching power supplie question 46 2/1 063 (SF6 ED DC) All of the K6 topics have less than 2.5 IR in rev 2, randomly DC Electrical resampled and replaced with K/A K4.02 Breaker interlocks, Distribution RO permissives, bypasses and cross-tie question 47 2/1 064 (SF6 EDG) All the K5 topics are less than 2.5 IR in rev 2, randomly Emergency Diesel resampled and replaced with K/A K6.07 Air receiver Generator RO question 48 2/1 076 K3.02 RO 076 K3.09 has a K/A rating of 1.9. Must be 2.5 or highe question 51 Resampled K3 K/As for system 076. Randomly selected 076 A3.02 Secondary closed cooling water.
2/1 078 K2.01 question 078 K2.02 rejected. Waterford 3 does not have an Emergency 52 Air Compressor. Waterford 3 has 2 IA compressors and 3 SA compressors. Randomly selected another K/A for system 078 in the K2 section.
2/1 078 (SF8 IAS) A2.02 is not listed in rev 2, no other A2 K/As are 2.5 or higher Instrument Air RO in IRs. Randomly resampled and replaced with A3.01 Air question 53 Pressure.
2/1 103 (SF5 CNT) K1.10 CSS is not listed in rev 2. Randomly resampled and Containment RO replaced with K1.01 CC question 54 2/1 013 2.4.31 RO 053 2.4.23 rejected. Waterford 3 does not have an Integrated question 55 Control System. Randomly selected another Tier 2/ Group 1 system that did not already have two questions selected to it. All other systems were used at least once. Randomly selected another Generic Item. Randomly selected 013 2.4.1 NRC Chief examiner identified 013 2.4.11 does not meet section ES-401 D.1.b of NUREG 1021 and was randomly replaced with 013 2.4.31.
2/2 002 (SF2; SF4P K3.04 RMS is not listed in rev 2. Randomly resampled and RCS) Reactor replaced with K3.03 Containmen Coolant RO question 56 2/2 016 (SF7 NNI) K3.05 Condensate is less than 2.5 IR in rev 2. Randomly Nonnuclear resampled and replaced with K3.04 MFW syste Instrumentation RO question 57 No K4 K/As are listed in rev 2. Randomly resampled and 2/2 028 (SF5 HRPS) replaced with K2.01 Hydrogen recombiner Hydrogen Recombiner and Purge Control RO question 58 Rev. 11
ES-401 Record of Rejected K/As Form ES-401-4 2/2 035 K5.01 (SF 4P K5.06 S/G tube leakage detection not listed in rev 2. Randomly SG) Steam resampled and replaced with K5.01 Effect of secondary Generator RO parameters, pressure, and temperature on reactivit question 59 2/2 041 K6.13 (SF4S K6.13 IAS not listed in rev 2. Randomly resampled and SDS) Steam replaced with K6.03 Controller and positioners, including ICS, Dump/Turbine S/G, CRD Bypass Control RO question 60 2/2 045 A1.06 (SF 4S A1.07 Lights and Alarms is not listed in rev 2. Randomly MTG) Main Turbine resampled and replaced with A1.06 Expected response of Generator RO secondary plant parameters following T/G tri question 61 A2.01 Loss of circulating/cooling water system is less than 2.5 IR in 2/2 055 (SF4S CARS) rev 2. All A2 K/As are less than 2.5 IR. Randomly resampled to Condenser Air K4 - all K4s less than 2.5 IR in rev 2. Randomly resampled again Removal RO to K6 - all K6s less than 2.5 IR in rev 2. Randomly resampled question 62 again to A3: A3 Ability to monitor automatic operation of the CARS, including: A3.03 Automatic diversion of CARS exhaus A3.03 was rejected. Waterford 3 does not have an 015 A3.03 RO automatic diversion of CARS exhaust. There were no other question 62 K/As in the A3 field for CARS to randomly select that were greater than 2.5. Selected another system in Tier 2/Group Randomly selected 015 Nuclear Instrument System. Randomly selected a K/A in the A3 field for system 015. Randomly selected A3.03 Verification of proper functioning/operability.
2/2 056 (SF4S CDS) A3.07 has an IR of 1.7 in rev 2. All A3 K/As have IRs less than Condensate RO 2.5. Randomly resampled to A2.04 Loss of condensate pump question 63 2/2 079 G2.4.35 RO NRC Chief examiner identified 079 2.4.12 does not meet question 64 section ES-401 D.1.b of NUREG 1021 and was randomly replaced with 079 2.4.35.
2/2 011 K1.05 question 086 K1.03 rejected. There is not enough written guidance in W3 65 procedures that reference fire protection and its association with the EFW system that would allow for an RO question to be written. Randomly selected a different Tier 2 Group 2 system (011 PLC) and then randomly selected a K/A number from the K1 field. Randomly selected 011 K1.05.
3/2 G2.2.22 RO 2.2.17 rejected. The knowledge of the process for managing question 69 maintenance activities as described in this K/A is a planning department and SRO function (WMC) at W3. Could not develop a question on the RO level. Randomly selected another generic K/A from the 2.2 field. Randomly selected G2.2.22.
1/1 00022 AA2.02 00022 AA2.03 rejected. W3 does not have any flow control question 76 (SRO1) valves or controllers in the charging system (only solenoid operated isolation valves). Randomly selected another K/A for system 00022 in the AA2 section. Randomly selected 00022 AA2.0 Rev. 11
ES-401 Record of Rejected K/As Form ES-401-4 1/1 00027 AA2.02 00027 AA2.01 rejected. This K/A is very similar to SRO question 78 (SRO3) question 16. Also, the W3 Pressurizer Pressure Control System has no features or failures that will result in an increase in Pressurizer level. Randomly selected another K/A for system 00027 in the AA2 section. Randomly selected 00027 AA2.02.
1/1 057 G2.4.20 (SRO NRC Chief examiner identified 057 2.4.17 does not meet Question 79) section ES-401 D.1.b of NUREG 1021 and was randomly replaced with 057 2.4.20 1/1 000065 (APE 65) AA2.09 not listed in rev 2. Randomly resampled and replaced with Loss of Instrument AA2.01 Cause and effect of low-pressure instrument air alar Air / 8 SRO question 80 058 AA2.03 SRO 065 AA2.01 rejected. 065 A2.01 is already a selected K/A for the Question 80 2020 RO Exam. See RO17 on the 2020 Exam. There is also an abundance of low instrument air K/As on the previous two exams and it was not possible to create a new question here that did not overlap in part any of those questions. Randomly selected another system in Tier 1/Group 1 (selected 058 Loss of DC Power) and another K/A in the A2 field. (Randomly selected AA2.02 which is already used on the RO exam). So, randomly selected again and selected 058 AA2.03.
1/1 00038 G2.1.20 NRC Chief examiner identified 077 2.1.15 does not meet SRO Question 81 section ES-401 D.1.b of NUREG 1021 and was randomly replaced with 077 2.1.20 Could not develop a question that did not overlap RO18 on the 2020 Exam or SRO81 on the 2018 exam (Both questions are related to Generator voltage and electrical grid disturbanc Randomly selected another Tier 1/Group 1 evolution and retained the same Generic K/A. Randomly selected 038 2.1.20 1/2 028 AA2.11 SRO 033 AA2.07 rejected. Intermediate nuclear instrumentation at question 83 Waterford 3 is called the control channels. These channels only go to Reactor Regulating System and Steam Bypass Control. The failure of this channel is not covered in any Waterford 3 procedure. Randomly selected another system from Tier 1 Group 2 and another K/A for that system in the A2 field. Randomly selected 028 AA2.11.
1/2 059 G2.3.6 SRO 059 2.3.14 rejected. There is no guidance in OP-007-004, Liquid question 84 Waste Management, OP-901-412, Liquid Waste Discharge High Radiation or any other W3 operations procedure pertaining to radiation or contamination hazards while discharging a Waste Condensate Tank or Boric Acid Tank. Randomly selected another section 3 Generic K/A while for 059 Accidental Liquid Rad Waste Release. Randomly selected 059 2. NRC Chief examiner identified 059 2.3.6 does not meet section ES-401 D.1.b of NUREG 1021 and was randomly replaced with 059 2.1.3 (CE E09) EA This K/A not listed in rev 2. Randomly resampled to CE E09 1/2 Functional Recovery, K/A EA2.1 Facility conditions and Functional Recovery / 4 SRO selection of appropriate procedures during abnormal and question 85 emergency operation Rev. 11
ES-401 Record of Rejected K/As Form ES-401-4 004 A2.18 (SF1; 004 A2.28 rejected. Could not develop an SRO question for 2/1 depressurizing the RCS when it is hot. This would occur when SF2 CVCS) SRO question 86 a pressurizer spray valve fails open are depressurizing the RCS following a SGTR. Neither are related to the CVCS System in a way that a question could be developed. Randomly selected another K/A for CVC system in the A2 field. Randomly selected A2.18 High VCT level.
2/1 007 (SF5 PRTS) Generic Knowledge K/a 2.1.47 is not listed in rev 2 of the K/A Pressurizer catalog. Randomly resampled and replaced with 2.1.4 Relief/Quench Tank SRO question 87 2/1 008 (SF8 CCW) 007 2.1.44 rejected. Could not develop an SRO question that Component Cooling relates PRTS to RO duties during Fuel Handling. The PRTS has no Water SRO interrelation with any fuel handling activities at W3. Randomly question 87 selected another system in Tier 2/ Group 1 that was not previously G2.2.22 used. Then randomly selected another Generic knowledge K/A in group 1. Randomly selected 008 2. NRC Chief examiner identified 008 2.1.9 does not meet section ES-401 D.1.b of NUREG 1021 and was randomly replaced with 008 2.2.22 2/1 062 G2.2.36 SRO NRC Chief examiner identified 062 2.2.21 does not meet question 89 section ES-401 D.1.b of NUREG 1021 and was randomly replaced with 062 2.2.36.
2/1 064 A2.2.22 SRO 064 A2.15 rejected. Could not develop an SRO question for Question 90 water buildup in cylinders. No procedure guidance in W3 operations procedures on this topic. Randomly selected another K/A in the 064 A2 field. Randomly selected 064 A2.22.
2/2 033 G2.4.8 (SF8 029 2.4.39 rejected. Could not develop an SRO question SFPCS) Spent Fuel relating to RO duties in the Emergency Plan procedures and Pool Cooling SRO Containment Purge. RO duties in the Emergency Plan question 92 procedures are minimal and there is no duties associated with Containment Purge in E-Plan procedures. Randomly selected another Tier2/Group 2 system not previously used and then randomly selected another Generic Knowledge K/A in group 4 not previously used. Randomly selected 033 2. Rev. 11
ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11
h a n g e Pro h a n g e Pro XC d XC d F- F-PD ! uc PD ! uc W t W t O O N N y y bu bu to to ww om ww om k k lic lic C .c C .c w tr w tr re re
. .
ac ac k e r- s o ft w a k e r- s o ft w a ES-301 Administrative Topics Outline Form ES-301-1 Facility: Waterford 3 Date of Examination: 08/17/2020 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Code* Describe activity to be performed (see Note)
A1 N,R 2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant Conduct of Operations system, temperature, secondary plant, fuel depletion, et K/A Importance: Complete a calculation for projecting the final RCS boron concentration due to draining a Steam Generator to the RCS in accordance with OP-902-009 Appendix 39, SG Backflow Lo A2 2.1.25, Ability to interpret reference materials, such as graphs, curves, tables, et M,R Conduct of Operations Determine Spent Fuel Pool (SFP) level by alternate K/A Importance: monitoring and calculate time for level to reach the top of the fuel assemblies upon a loss of all cooling to the SFP per OP-901-513, SFP Cooling Malfunctio A3 2.2.12, Knowledge of Surveillance Procedure Equipment Control Perform Keff Calculation in accordance with OP-903-090, P,R Shutdown Margin, Section 7.5, Keff Calculatio K/A Importance: A4 M,R 2.3.4, Knowledge of radiation exposure limits under normal and emergency condition Radiation Control Calculate stay time to perform a tagout verification. Room K/A Importance: dose rate and operators yearly dose provided. The stay time will be based on the W3 administrative yearly dose limit (2000 mr/year).
Emergency Plan Not Selected 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)
h a n g e Pro h a n g e Pro XC d XC d F- F-PD ! uc PD ! uc W t W t O O N N y y bu bu to to ww om ww om k k lic lic C .c C .c w tr w tr re re
. .
ac ac k e r- s o ft w a k e r- s o ft w a ES-301 Administrative Topics Outline Form ES-301-1 Facility: Waterford 3 Date of Examination: 8/17/2020 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Code* Describe activity to be performed (see Note)
A5 D, R 2.1.20, Ability to interpret and execute procedure steps Conduct of Operations Review a completed Containment Pressure calculation in accordance with OP-903-001, Technical K/A Importance: Specification Surveillance Logs, Attachment 11.15, Containment Pressure Calculatio A6 D, R 2.1.23, Ability to perform specific system and integrated plant procedures during all modes of plant Conduct of Operations operatio K/A Importance: Perform SM/CRS review of OP-901-501, PMC or Core Operating Limit Supervisory System Malfunction, Attachments 1, 2 and 3 following a PMC failur A7 2.2.12, Knowledge of Surveillance Procedures Equipment Control Review Keff Calculation in accordance with OP-903-P, R 090, Shutdown Margin, Section 7.5, Keff Calculatio K/A Importance: Applicant determines Keff does not meet Tech Spec 3.1.2.9 requirements and identifies required corrective action A8 M, R 2.3.4, Knowledge of radiation exposure limits under normal or emergency condition Radiation Control Authorize Emergency Exposure as the Emergency Director in accordance with EP-002-030, Emergency K/A Importance: Radiation Exposure Guidelines and Control A9 2.4.41, Knowledge of the emergency action level thresholds and classification Emergency Plan N,R Determine appropriate Emergency Plan action level in K/A Importance: accordance with EP-001-001, Recognition and Classification of Emergency Condition 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).
h a n g e Pro h a n g e Pro XC d XC d F- F-PD ! uc PD ! uc W t W t O O N N y y bu bu to to ww om ww om k k lic lic C .c C .c w tr w tr re re
. .
ac ac k e r- s o ft w a k e r- s o ft w a ES-301 Administrative Topics Outline Form ES-301-1
- 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: Waterford 3 Date of Examination: 08/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 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 004 Chemical and Volume Control D, S 1 Align Charging to HPSI Header A in accordance with OP-901-112, Charging or Letdown Malfunction 004 A4.08 Charging RO-3.8, SRO-3.4 S2 006 Emergency Core Cooling System A, D, L, S 2 Mitigate the release of radioactivity through RWSP following RAS Actuation in accordance with OP-902-002, Loss of Coolant accident Recover Fault: LPSI pump continues to run after RAS actuates, requiring the applicant to manually stop the running LPSI pump after additional valve manipulation EA1.12 Long term containment of radioactivity RO-4.1, SRO-4.4 S3 005 Shutdown Cooling System A, EN, N, L, 4P S
Place Shutdown Cooling Train B to service in accordance with OP-009-005, Shutdown Coolin Fault: When 4100 gpm SDC flow is verified through SI-135, an Inadvertent RAS occurs that will secure the running LPSI pumps. The applicant will be required to restart LPSI Pump A4.01 Controls and indications for RHR Pumps RO - 3.6, SRO - ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 08/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function 0BS4 039 Main and Reheat Steam System; BOP operator A, D, S 4S immediate operator actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuatio Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressur A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 S5 026 Containment Spray System D, EN, L, S 5 Reset CSAS in accordance with OP-902-009, Standard Appendices, Section 5 - A4.01 CSS Controls RO - 4.5, SRO - 4.3 S6 062 A.C. Electrical Distribution System D, P, S 6 Transfer the 3AB Bus to the A train and Start Component Cooling Water Pump AB as the second CCW Pump.in accordance with OP-901-311, Loss of Train B Safety Bu (From 2018 NRC Exam)
062 A4.01 All breakers RO-3.3, SRO-3.1 S7 015 Nuclear Instrumentation System L, D, S 7 Perform Range Check functional test of startup Channels in accordance with OP-903-101, Startup Channel Functional Test 015 A3.03, Verification of proper functioning/operability RO - 3.9, SRO - 3.9 S8 034 Fuel Handling Equipment System D, S 8 Place the FHB Emergency Filtration Unit A in service in accordance with OP-002-009, Fuel Handling Building HVAC 034 A4.01 Radiation levels RO - 3.3, SRO - 3.7 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 08/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function P1 013 Engineered Safety Features Actuation System (ESFAS) A, D, E, L, P 2 Actuate a Recirculation Actuation Signal (RAS) manually in accordance with OP-902-009 Appendix 34, RAS Manual Actuatio Fault: The RAS will not actuate using the manual pushbuttons and will actuate by opening the breaker (From 2018 NRC Exam)
A4.03 ESFAS Initiation RO-4.5, SRO-4.7 P2 076 Service Water System L, N 4S Align Potable Water to Instrument Air Compressors in accordance with OP-902-009, EOP Standard Appendices, App. 1 A4.04 Emergency Heat Loads RO-3.5, SRO-3.5 P3 064 Emergency Diesel Generator (ED/G) System; A, D, E, R 6 Trip Emergency Diesel Generator B locall Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG K4.02 Trips for ED/G while operating (normal or emergency)
- 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 roo * Type Codes Criteria for R /SRO-I/SRO-U
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 08/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 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)lternate path 4-6/4-6 /2-3 5 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 9 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature 1/ 1/ 1 (control room system) 2 (L)ow-Power/Shutdown 1/ 1/ 1 6 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 2 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 2 (R)CA 1/ 1/ 1 1 (S)imulator 8
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 8/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 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 004 Chemical and Volume Control D, S 1 Align Charging to HPSI Header A in accordance with OP-901-112, Charging or Letdown Malfunction A4.08 Charging RO-3.8, SRO-3.4 S2 006 Emergency Core Cooling System A, D, L, S 2 Mitigate the release of radioactivity through RWSP following RAS Actuation in accordance with OP-902-002, Loss of Coolant accident Recover Fault: LPSI pump continues to run after RAS actuates, requiring the applicant to manually stop the running LPSI pump after additional valve manipulation EA1.12 Long term containment of radioactivity RO-4.1, SRO-4.4 S3 005 Shutdown Cooling System A, EN, N, L, 4P S
Place Shutdown Cooling Train A to service in accordance with OP-009-005, Shutdown Coolin Fault: When 4100 gpm SDC flow is verified through SI-135, an Inadvertent RAS occurs that will secure the running LPSI pumps. The applicant will be required to restart LPSI Pump A4.01 Controls and indications for RHR Pumps RO - 3.6, SRO - ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 8/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function 0BS4 039 Main and Reheat Steam System; BOP operator A, D, S 4S immediate operator actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuatio Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressur A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 S5 026 Containment Spray System D, EN, L, S 5 Reset CSAS in accordance with OP-902-009, Standard Appendices, Section 5 - A4.01 CSS Controls RO-4.5, SRO-4.3 S6 062 A.C. Electrical Distribution System D, P, S 6 Transfer the 3AB Bus to the A train and Start Component Cooling Water Pump AB as the second CCW Pump.in accordance with OP-901-311, Loss of Train B Safety Bu A4.01 All breakers RO-3.3, SRO-3.1 S7 S8 034 Fuel Handling Equipment D, S 8 Place the FHB Emergency Filtration Unit in service in accordance with OP-002-009, Fuel Handling Building HVAC A4.01 Radiation levels RO - 3.3, SRO - 3.7 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 8/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function P1 013 Engineered Safety Features Actuation System (ESFAS) A, D, E, L, P 2 Actuate a Recirculation Actuation Signal (RAS) manually in accordance with OP-902-009 Appendix 34, RAS Manual Actuatio Fault: The RAS will not actuate using the manual pushbuttons and will actuate by opening the breaker (From 2018 NRC Exam)
A4.03 ESFAS Initiation RO-4.5, SRO-4.7 P2 076 Service Water System L, N 4S Align Potable Water to Instrument Air Compressors in accordance with OP-902-009, EOP Standard Appendices, App. 1 A4.04 Emergency Heat Loads RO-3.5, SRO-3.5 P3 064 Emergency Diesel Generator (ED/G) System; A, D, E, R 6 Trip Emergency Diesel Generator B locall Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG K4.02 Trips for ED/G while operating (normal or emergency)
- 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 roo * Type Codes Criteria for R /SRO-I/SRO-U
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 8/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 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)lternate path 4-6/4-6 /2-3 5 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 8 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature 1/ 1/ 1 (control room system) 2 (L)ow-Power/Shutdown 1/ 1/ 1 5 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 2 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 2 (R)CA 1/ 1/ 1 1 (S)imulator 7
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 8/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 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 004 Chemical and Volume Control D, S 1 Align Charging to HPSI Header A in accordance with OP-901-112, Charging or Letdown Malfunction A4.08 Charging RO-3.8, SRO-3.4 S3 005 Shutdown Cooling System A, EN, N, L, 4P S
Place Shutdown Cooling Train A to service in accordance with OP-009-005, Shutdown Coolin Fault: When 4100 gpm SDC flow is verified through SI-135, an Inadvertent RAS occurs that will secure the running LPSI pumps. The applicant will be required to restart LPSI Pump A4.01 Controls and indication for RHR Pumps RO - 3.6, SRO - 3.4 S8 034 Fuel Handling Equipment D, S 8 Place the FHB Emergency Filtration Unit in service in accordance with OP-002-009, Fuel Handling Building HVAC 034 A4.01 Radiation levels RO - 3.3, SRO - 3.7 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 8/17/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function P1 013 Engineered Safety Features Actuation System (ESFAS) A, D, E, L, P 2 Actuate a Recirculation Actuation Signal (RAS) manually in accordance with OP-902-009 Appendix 34, RAS Manual Actuatio Fault: The RAS will not actuate using the manual pushbuttons and will actuate by opening the breaker (From 2018 NRC Exam)
A4.03 ESFAS Initiation RO-4.5, SRO-4.7 P3 064 Emergency Diesel Generator (ED/G) System; A, D, E, R 6 Trip Emergency Diesel Generator B locall Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG K4.02 Trips for ED/G while operating (normal or emergency)
- 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 roo * Type Codes Criteria for R /SRO-I/SRO-U (A)lternate path 4-6/4-6 /2-3 3 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 4 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature 1/ 1/ 1 (control room system) 1 (L)ow-Power/Shutdown 1/ 1/ 1 2 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 1 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 1 (R)CA 1/ 1/ 1 1 (S)imulator 3
Appendix D Scenario Outline Form ES-D-1 Facility: Waterford 3 Scenario No.: 1 Op Test No.: 1 Examiners: Operators:
Initial Conditions: Mode 2, Reactor Power ~1%. Two Charging Pumps in operation. AB Buses are aligned to Train B.
Turnover: Protected Train is B. Dilute to 5-10% power.
Critical Tasks: (1) Establish Reactivity Control (2) Establish Containment Temperature and Pressure Control (3) Trip any RCP exceeding operating limits Event Mal Event Event N N Type* Description R - ATC 1 N/A N - SRO Dilute to 5-10% power, perform 100 gallon PMU additio I-BOP 2 RC20B I-SRO RCS Loop 1 Hot leg transmitter failure high (RC-ITI-102HB).
TS-SRO C - ATC 3 CV35A C - BOP During dilution, PMU counter fails to secure flo CVR101 C - SRO I-ATC Pressurizer Level Control Channel Level Transmitter, RC-ILT-4 RC15A2 I-SRO TS-SRO 0110X, fails lo C-BOP 5 RC08B C-SRO Reactor Coolant Pump 1B Lower Seal fail RC03B RP02A C-ATC RCP 1B Trip with no automatic Reactor Trip (Critical Task 1, 6 RP02B C-SRO Establish Reactivity Control)
RP02C RP02D 7 MS11A M-All Excess Steam Demand Event will occur on #1 Steam Generator inside Containmen RP05A3 No automatic Containment Spray Actuation Signal (CSAS)
RP05B3 C-BOP (Critical Task 2, Establish Containment Temperature and
RP05C3 C-SRO Pressure Control) (Critical Task 3, Trip any RCP RP05D3 exceeding Operating limits)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Quantitative Attributes 1. Malfunctions after EOP entry (1-2) 2 2. Abnormal events (2-4) [Events 3,4,5 credited] 3 3. Major transients (1-2) 1 4. EOPs entered/requiring substantive actions (1-2) 1 5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 0 6. Preidentified critical tasks (> 2) 3-1-2020 NRC Exam Scenario 1 D-1 Rev 0
NRC Scenario 1 - Narrative The crew assumes the shift with the reactor at 1% power following a forced outage. The turnover will include instructions to perform RCS dilution to 5 - 10% power.
Event 1: The reactivity plan will include instructions to dilute in multiple PMU batches. The initial batch will be 100 gallons of PMU. Each subsequent batch will be 50 gallons of PMU. This will allow for an observable power rise without concern for a reactor trip on the PMU failure.
Event 2: After the first 100 gallons of PMU are added, RCS THOT instrument RC-ITI-102HB fails high.
The SRO should review and enter Technical Specification 3.3.1 action 2 and bypass Hi LPD and Lo DNBR bistables (3 & 4) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with OP-009-007, Plant Protection System. The SRO should review Technical Specifications 3.3.3.5 and 3.3.3.6 and OP-903-013, Monthly Channel Checks.
Tech Spec 3.3.3.5 and 3.3.3.6 should not be applicable.
Event 3: During the second dilution, the Primary Water counter will fail to secure dilution. The ATC should attempt to secure Primary Water Flow by operating PMU-144 and CVC-510. Neither of these actions will secure flow. The CRS should enter OP-901-104, Inadvertent Positive Reactivity Addition, and secure Primary Makeup Pump A.
Event 4: After the crew has stopped the inadvertent dilution, pressurizer Level Control Channel Level Transmitter, RC-ILT-0110X, fails low. The SRO should enter OP-901-110, Pressurizer Level Control Malfunction and implement Section E1. The crew should take manual control of the Pressurizer Level Controller and/or operate Charging Pumps to restore Pressurizer level, swap control to the Channel Y level channel, and return the Pressurizer Level Controller back to AUTO. The SRO should review Technical Specifications 3.3.3.5 and 3.3.3.6 and OP-903-013, Monthly Channel Checks. Tech Spec 3.3.3.5 action a should be entered with the action to restore the inoperable channel within 7 days. The SRO should determine that TS 3.3.3.6 requirements are met after verifying QSPDS value for PZR level is within the channel check limit of 5%.
Event 5: Reactor Coolant Pump 1B Lower Seal fails. The crew should enter OP-901-130, Reactor Coolant Pump Malfunction and implement Section E1, Seal Failure. The SRO should direct the BOP to lower Component Cooling Water Temperature by operating Dry Cooling Tower fans or by adjusting Auxiliary Component Cooling Water flow.
Event 6: After the crew is in Section E1 of OP-901-130 AND the BOP has adjusted Component Cooling Water Temperature, RCP 1B will trip and the reactor will not automatically trip. The ATC will manually trip the reactor from the RTGB (CRITICAL TASK 1). The crew will perform Standard Post Trip Actions using OP-902-000, SPTAs and diagnose to OP-902-001, Reactor Trip Recovery.
Event 7: After the crew has completed the Maintenance of Vital Auxiliaries Safety Function in OP-902-000, Standard Post Trip Actions, an Excess Steam Demand Event will occur on Steam generator #1 inside containment. This will result in containment pressure exceeding 17.7 psia (the setpoint for automatic Containment Spray Actuation).
Event 8: The CSAS will not occur and the crew will be required to manually initiate containment spray (CRITICAL TASK 2). The actuation of CSAS will cause CCW to isolate to the RCPs and the crew will need to stop any running RCP prior to exceeding 3 minutes without CCW flow (CRITICAL TASK 3). The crew will diagnose to OP-902-004 and stabilize RCS temperature and pressure using the least affected SG and auxiliary spray valves.
The scenario can be terminated after the crew has stabilized RCS temperature and pressure or at the lead examiners discretio NRC Exam Scenario 1 D-1 Rev 0
NRC Scenario 1 - Critical Task Determination Measurable Performance Critical Task Safety Significance Cueing Performance Indicator Feedback CT-1: Establish Reactivity Control Failure to trip the Reactor when an RCP off light illuminated automatic PPS signal has failed to Reactor Trip This task is satisfied by manually actuate can lead to a degradation of Trips and pre-trips on SG lo breakers open tripping the Reactor by using the fission product barriers. 1 minute is flow on CP-7 Depresses two Reactor manual trip pushbuttons, Diverse determined to be a reasonable time All CEA rod bottom limit to identify and take action for Trip pushbuttons on CP-Reactor trip pushbuttons, or by de- All CEA rod bottom lights lights illuminated satisfactory performance. OPS 2 or CP-8 energizing busses 32A and 32B within extinguished 1 minute of exceeding a Plant management standard documented in Reactor power Protection System (PPS) limit. This TM-OP-100-0 Procedurally driven from lowering task becomes applicable following the OP-902-000 step 1.a.1.1)
trip of RCP 1 (TM-OP-100-03, CT-1)
CT-2: Establish Containment Containment pressure >
Temperature and Pressure Control 17.7 psia CSAS annunciators Depresses 2 CSAS actuated This task is satisfied by manually Failure to take action to establish CS Pumps stop light pushbuttons on CP-7 initiating CSAS, or manually aligning CS Containment Temperature and illuminated CS Pumps run light Pumps and valves to establish at least 1 Pressure Control would lead to a or illuminated train of Containment Spray prior to degradation of a fission product CS-125 indicates closed exiting OP-902-000, Standard Post Trip barrier. OPS management standard Starts CS Pump and CS-125 indicates Actions. This action satisfies the documented in TM-OP-100-0 CS Header flow not open associated CS-125 open Containment Temperature and (TM-OP-100-03, CT-15) indicated valve using control Pressure Control Safety Function in OP-switch CS Header flow 902-000. This task becomes applicable Procedurally driven from indicated once 17.7 PSIA has been exceeded OP-902-000 step 9.3 inside containment following the ESD NRC Exam Scenario 1 D-1 Rev 0
NRC Scenario 1 - Critical Task Determination Measurable Performance Critical Task Safety Significance Cueing Performance Indicator Feedback CCW flow low/lost to RCPs Based on EOP required actions for an alarms on CP-2 and CP-18 CT-3: Trip any RCP exceeding Excess Steam Demand Event. The operating limits RCPs are stopped to prevent RCP RCP off light CCW valve status CP-2 seal damage. 3 minutes is the illuminated Stops RCPs using This task is satisfied by manually analyzed time for a RCP to run without CCW cooling to the RCP sea CSAS initiated CP-8 control switch stopping Reactor Coolant Pumps within RCP indicated flow 3 minutes of a loss of CCW flow to the OPS management standard lowering RCPs. This task becomes applicable documented in TM-OP-100-0 Procedurally driven from following the actuation of CSA OP-902-000 step 3.b.1 and (TM-OP-100-03, CT-23) 9.3 Critical Task (NUREG-1021, Rev. 11 Appendix D)
- 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.
- Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-102 NRC Exam Scenario 1 D-1 Rev 0
NRC Scenario 1 REFERENCES Event Procedures 1 OP-010-003, Plant Startup, Rev. 351 OP-002-005, Chemical and Volume Control, Rev. 64 2 OP-009-007, Plant Protection System, Rev. 20 OP-903-013, Monthly Channel Checks, Rev. 19 Technical Specifications 3.3.1, 3.3.3.5, 3.3. OP-901-104, Inadvertent Positive Reactivity Addition, Rev. 304 4 OP-901-110, Pressurizer Level Control Malfunction, Rev. 11 Technical Specifications 3.3.3.5, 3.3. OP-901-130, Reactor Coolant Pump Malfunction, Rev. 11 6 OP-902-000, Standard Post Trip Actions, Rev. 16 OP-902-009, Standard Appendices, Rev. 319 7/8 OP-902-004, Excess Steam Demand Event Recovery, Rev. 17 OP-902-009, Standard Appendices, Rev. 319 GEN EN-OP-115, Conduct of Operations, Rev. 26 EN-OP-115-08, Annunciator Response, Rev. 5 EN-OP-200, Plant Transient Response Rules, Rev. 6 OI-038-000, EOP Operations Expectations / Guidance, Rev. 19 OP-100-017, EOP Implementation Guide, Rev 5 TM-OP-100-03, Simulator Training, Rev. 20-5-2020 NRC Exam Scenario 1 D-1 Rev 0
Appendix D Scenario Outline Form ES-D-1 Facility:
Waterford 3 Scenario No.: 2 Op Test No.: 1 0B Examiners: Operators:
Initial Conditions: Reactor power is 100%. AB Buses are aligned to Train B.
Turnover: Protected Train is B; Maintain 100%. CS Pump A (TS) is out of service.
Critical Tasks: (1) Trip any RCP exceeding operating limits (2) Critical Task, align LPSI pump to replace CS pump Event Mal Event Event N N Type* Description N - BOP During performance of OP-903-052, CVAS Fan A will fail to 1 DI-18A4s27-1 N - SRO star TS - SRO C - BOP Plant Protection Sys Ch. D, Cont. Press (CIAS), CB-IPI-2 CH08E1 C - SRO 6701SMD, fails hig TS - SRO I - ATC 3 RX14A PZR pressure control channel, RC-IPIC-0100X, failure lo I - SRO R - ATC Feedwater Heater 5B tube leak, Rapid Plant Power 4 FW35B C - BOP Reduction C - SRO TP01A C - BOP 5 Turbine Cooling Water Pump A trips TP08B C - SRO Loss of Coolant Accident will occur. (Critical Task 1, Trip 6 RC23A M - All any RCP exceeding operating limits)
I - ATC Relay K202A fails, CVC-401, CVC-109, IA-909, and FP-7 RP08C I - BOP 601A fail to close automatically I - SRO Containment Spray Pump B trips and cannot be restarted requiring entry into OP-902-008, Functional Recovery and DI-08A04S22-1 C - BOP action taken to align Low Pressure Safety Injection pump to
CS01B C - SRO provide Containment Spray (Critical Task 2, align LPSI pump to replace CS pump prior to exiting Appendix 28 of OP-902-009)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor-1-2020 NRC Exam Scenario 2 D-1 Rev 1
Appendix D Scenario Outline Form ES-D-1 Scenario Quantitative Attributes 1. Malfunctions after EOP entry (1-2) 2 2. Abnormal events (2-4) 3 3. Major transients (1-2) 1 4. EOPs entered/requiring substantive actions (1-2) 1 5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 1 6. Preidentified critical tasks (> 2) 2-2-2020 NRC Exam Scenario 2 D-1 Rev 1
NRC Scenario 2 - Narrative The crew assumes the shift at 100% power with instructions to maintain 100% power. Containment Spray Pump A is out of service.
Event 1: Last shift, it was discovered that OP-903-052, CVAS Operability Test, will exceed its Tech Spec late date this shift. The crew will be directed to start CVAS Train A in accordance with OP-903-052. This surveillance will have the BOP operator secure RAB Normal Supply and Normal Exhaust Fans A and start CVAS Fan A. After securing both normal ventilation fans, CVAS Fan A will fail to start. This will require entering Tech Spec 3.7.7, a 7 day action requirement. RAB Normal Supply and Normal Exhaust Fans A will have to be re-started.
Event 2: After RAB Normal Ventilation is running and Tech Specs have been addressed, CB-IPI-6701SMD, Containment Pressure (CIAS) fails high. The SRO should review Technical Specifications 3.3.1 and 3.3.2. Per Table 3.3-1 under Containment Pressure - High (Functional Unit 6) the SRO should enter Technical Specification 3.3.1 action 2. Per Table 3.3-3 under Functional Units 1b (Safety Injection, Containment Pressure-High), 3b (Containment Isolation, Containment Pressure-High), and 4c (Main Steam Line Isolation, Containment Pressure High) the SRO should enter Tech 3.3.2 action 13. The SRO should direct the BOP to bypass the Containment Pressure High (RPS) and Containment Pressure High (ESF) trip bistables (13&16) in PPS Channel D within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The BOP should bypass the trip bistables in accordance with OP-009-007, Plant Protection System.
Event 3: After the BOP has bypassed bistables at CP-10 and the SRO has addressed Tech Specs, Pressurizer Pressure Control Channel RC-IPIC-0100X will fail low. The crew will observe pressurizer pressure alarms and that all PZR heaters are energized. The SRO will implement OP-901-120, Pressurizer Pressure Control Malfunction, Section E1, Pressurizer Pressure Control Channel Instrument Failure. The SRO should direct the ATC to align the alternate pressurizer pressure channel and verify correct Pressurizer pressure control response.
Event 4:. After the alternate pressurizer pressure channel has been aligned, a tube leak occurs in Feedwater Heater 5B, causing Condensate flow to isolate through Low Pressure Feedwater Heaters 5B and 6B. The crew will enter OP-901-221, Secondary System Transient, Section E1, Loss of Feedwater Preheating. This also requires a power reduction in accordance with OP-901-212, Rapid Plant Power Reduction, which will prompt a reactivity manipulation.
Event 5: The running Turbine Cooling Water Pump A will trip and the standby pump will fail to start automatically. The CRS should direct the BOP to start the standby pump in accordance with OP-901-512, Loss of Turbine Cooling Water and OP-500-005, Control Room Cabinet E.
Event 6: RCS leak occurs on RCS Cold Leg 1A that rapidly progresses to a Large Break Loss of Coolant Accident and reactor trip. The ATC should manually stop all RCPs following a loss of subcooling or loss of CCW to the RCPs (CRITICAL TASK 1). The crew should re-diagnose to OP-902-002, Loss of Coolant Accident Recovery Procedure.
Event 7: Relay K202A will not actuate and CVC-401, CVC-109, IA-909, and FP-601A fail to close automatically. The ATC and BOP should position these valves to ensure Containment Isolation.
Event 8: Once the crew has entered OP-902-002, Containment Spray Pump B will trip and will not be able to be restarted. The crew should determine that Containment isolation and Containment Pressure and Temperature Control Safety Functions are not being met and diagnose into OP-902-008, Functional Recovery. The SRO should prioritize Containment Isolation first due to CS-125B being open and Containment Pressure and Temperature Control second. The crew will perform steps in OP-902-008, Functional Recovery and align the LPSI pump B to replace CS Pump B to establish Containment Temperature and Pressure Control (CRITICAL TASK 2).
The scenario can be terminated after the established Containment Spray flow from the Low Pressure Safety Injection Pump or at the lead examiners discretio NRC Exam Scenario 2 D-1 Rev 1
NRC Scenario 2 - Critical Task Determination Measurable Performance Critical Task Safety Significance Cueing Performance Indicator Feedback Based on EOP required actions for a Loss of Coolant Accident, the RCPs CCW flow low/lost to RCPs are stopped for numerous reason alarms on CP-2 and CP-18 CT-1: Trip any RCP exceeding The potential for RCP seal damage operating limits RCS pressure and and pumping more mass out of the break can result in uncovering fue temperature on CP-2 Stops RCPs using Pump light indication This task is satisfied by manually In this case, the RCPs are stopped to control switch stopping Reactor Coolant Pumps prior prevent more mass being lost out of CSAS initiated CP-8 to exiting Standard Post Trip Action the RCS which jeopardizes the fuel This task becomes applicable following clad barrier and to prevent RCP seal a loss of subcooling in the RCS or a Procedurally driven from:
damage. OPS management standard OP-902-000 steps 5.3, loss of CCW to the RCP seal coole documented in TM-OP-100-0 OP-902-002 steps 8, 9 (TM-OP-100-03, CT-23)
Containment Spray flow on CP_8 CT-2: Establish Containment Failure to take action to establish containment pressure and CS pump light status on Temperature and Pressure Control temperature control may result in CP-8 The crew takes actions This task is satisfied by manually containment pressure exceeding its to manually align an Containment Spray aligning an available LPSI pump to maximum design and therefore Containment pressure exceeding design leakage. This available LPSI pump to flow indication replace a CS pump prior to exiting the replace a CS pump would result in degradation of a OP-902-008, Functional step to align a LPSI pump to replace a fission product barrier.
CS pump in Appendix 2 Recovery CPTC actions (TM-OP-100-03, CT-15)
OP-902-009, Standard Appendices Critical Task (NUREG-1021, Rev. 11 Appendix D)
- 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.
- Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-102 NRC Exam Scenario 2 D-1 Rev 1
NRC Scenario 2 REFERENCES Event Procedures 1 OP-903-052, CVAS Operability Test, Rev 13 Technical Specification 3. OP-009-007, Plant Protection System, Rev. 20 Technical Specification 3.3.1, 3. OP-901-120, Pressurizer Pressure Control Malfunction, Rev. 302 4 OP-901-221, Secondary System Transient, Rev. 5 OP-901-212, Rapid Plant Power Reduction, Rev. 14 5 OP-901-512, Loss of Turbine Cooling Water, Rev 3 OP-500-005, Control Room Cabinet E, Rev 28 6 OP-902-000, Standard Post Trip Actions, Rev. 16 7 OI-038-000, EOP Operations Expectations/Guidance, Rev 19 8 OP-902-002, Loss of Coolant Accident Recovery, Rev 21 OP-902-008, Function Recovery Procedure, Rev. 30 OP-902-009, Standard Appendices, Rev. 319 GEN EN-OP-115, Conduct of Operations, Rev. 26 EN-OP-115-08, Annunciator Response, Rev. 5 EN-OP-200, Plant Transient Response Rules, Rev. 6 OI-038-000, EOP Operations Expectations / Guidance, Rev. 19 TM-OP-100-03, Simulator Training, Rev. 21 OP-100-017, Emergency Operating Procedures Implementation Guide, Rev 5-5-2020 NRC Exam Scenario 2 D-1 Rev 1
Appendix D Scenario Outline Form ES-D-1 Facility:
Waterford 3 Scenario No.: 3 Op Test No.: 1 0B Examiners: Operators:
Initial Conditions: Reactor power is 100%. AB Buses are aligned to Train B. No equipment OOS Turnover: Protected Train is B; Maintain 100%.
Critical Tasks: (1) Trip any RCP exceeding operating limits, (2) Energize at least 1 vital bus Event Mal Event Event N N Type* Description C - ATC C - SRO 1 CV01B Charging Pump B trips TS - SRO I - BOP Steam Generator 1 Pressure Instrument, 2 SG04G I - SRO SG-IPT-1013C, fails low requiring Technical Specification entry and TS - SRO bypass of multiple Plant Protection System C trip bistable MS05A1 C - BOP Atmospheric Dump Valve on Steam Generator #1 fails open 3 C - SRO requiring entry into OP-901-221, Secondary System Transient, TS - SRO turbine load reduction and local isolation of the ADV. (TS 3.7.1.7)
R-ATC FW21A C-BOP A lowering of Main Condenser vacuum. Requires a rapid plant
FW21AA C-SRO downpower IAW OP-901-212. Vacuum continues to lower until a manual or automatic tri RD11 A07 C - ATC 5 RD11 A37 3 Control Element Assemblies fail to insert into the core following RD11 A39 the reactor trip, Emergency Boration (Critical Task 1)
ED01A ED01B M - All 6 Following entry into OP-902-001, a loss of offsite power occur ED01C ED01D EDG B overspeed trip on start EG10B C-BOP 7 EDG A auto voltage regulator fails low, manually raise voltage to EG13A C-SRO allow breaker auto closure (Critical Task 2)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor-1-2020 NRC Exam Scenario 3 D-1 Rev 0
Appendix D Scenario Outline Form ES-D-1 Scenario Quantitative Attributes 1. Malfunctions after EOP entry (1-2) 2 2. Abnormal events (2-4) 3 3. Major transients (1-2) 1 4. EOPs entered/requiring substantive actions (1-2) 1 5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 0 6. Preidentified critical tasks (> 2) 2-2-2020 NRC Exam Scenario 3 D-1 Rev 0
NRC Scenario 3 - Narrative The crew assumes the shift at 100% power with instructions to maintain 100% power. No equipment OOS.
Event 1: Charging Pump B trips. Per the Annunciator Response Procedure, the SRO should direct the ATC to start a standby charging pump after verifying a suction path available or isolate Letdown using CVC-101, Letdown Stop Valve. The SRO will implement OP-901-112, Charging or Letdown Malfunction, Section E1, Charging Malfunction. If Letdown is isolated, Charging and Letdown will be re-initiated using Attachment 2 of OP-901-112. The SRO should review and enter Technical Specification 3.1.2.4.
Technical Specification 3.1.2.4 may be exited after aligning Charging Pump AB to replace Charging Pump B.
Event 2: Steam Generator 1 Pressure Instrument, SG-IPT-1013C, fails low. The SRO should review and enter Technical Specifications 3.3.1 action 2 and 3.3.2 actions 13 and 19. The SRO will direct the BOP to bypass the Steam Generator 1 Pressure Lo, Steam Generator 1 P, and Steam Generator 2 P trip bistables in Plant Protection System Channel C within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in accordance with OP-009-007, Plant Protection System. The SRO should review Technical Specifications 3.3.3.5 and 3.3.3.6 and OP-903-013, Monthly Channel Checks, and determine that Technical Specification entry for 3.3.3.5 and 3.3.3.6 is not required.
Event 3: #1 Steam Generator Atmospheric Dump Valve (ADV) fails open and cannot be closed using the CP-8 controller (MS-IPIC-0303-A1). The SRO should enter into procedure OP-901-221, Secondary System Transient and OP-500-012, Control Room Cabinet N, and direct the BOP to reduce Main Turbine load. The crew should direct an NAO to locally isolate or close the ADV. The SRO should enter Technical Specification 3.7.1.7.
Event 4:. A leak in the Main Condenser develops and Main Condenser vacuum begins to drop. The SRO will enter OP-901-220, Loss of Condenser Vacuum. Main Condenser vacuum will drop below 25 inches, requiring a rapid plant power reduction. The SRO will enter OP-901-212, Rapid Plant Power Reduction. .
For the power reduction, the ATC will perform direct boration to the RCS as well as ASI control with CEAs and Pressurizer boron equalization. The BOP will manipulate the controls to reduce Main Turbine load.
Vacuum will stabilize during the power reduction then the vacuum leak will grow until a manual reactor trip is directed by the CRS or an automatic trip occurs.
Event 5: During the trip 3 CEAs remain fully withdrawn and the ATC will commence emergency boration to the RCS in accordance with OP-902-000, Standard Post Trip Actions (Critical Task 1).
Event 6: After the crew transitions to OP-902-001, Reactor Trip Recovery, a loss of offsite power will occur. The SRO will diagnose and enter OP-902-003, Loss of Offsite Power / Loss of Forced Circulation to stabilize the plant and to protect the Main Condenser.
Event 7: Emergency Diesel Generator B will trip immediately on overspeed and Emergency Diesel Generator A will require the BOP to manually raise voltage which will allow its output breaker to automatically close (Critical Task 2)
The scenario can be terminated after the actions are taken to protect the main condensers located in OP-902-003, Loss of Offsite Power / Loss of Forced Circulation-3-2020 NRC Exam Scenario 3 D-1 Rev 0
NRC Scenario 3 - Critical Task Determination Measurable Performance Critical Task Safety Significance Cueing Performance Indicator Feedback CT-1: Establish Reactivity Control Based on Emergency Operating Procedure Required actions for This task is satisfied by manually Reactivity Control. Failure to initiate Rod bottom lights The crew takes action to initiating emergency boration using the emergency boration would result in a extinguished manually initiate gravity feed flowpath prior to entering condition that is not allowed by the emergency boration Required valve and OP-902-003, Loss of Offsite Power / facility license as analysis assumes CEA indicates withdrawn using the gravity drain pump indications Loss of Forced Circulation. This is that all CEAs are fully inserted during on CEAC flowpat and adequate accomplished by opening BAM-113A or a reactor trip with the exception of the BAM-113B and closing CVC-183. This charging flow most reactive rod. OPS management Procedurally driven from OP-903-103, Emer task becomes applicable following the Standard documented in TM-OP-100-Reactor Trip and the Loss of Offsite OP-902-000 Boration 0 Powe (TM-OP-100-03, CT-1)
Breaker indication on CP-1 The crew takes actions CT-2: Energize at least 1 vital AC bus Failure to energize at least one to manually start an emergency bus will result in the plant Control Room Lighting available EDG and/or EDG status and This task is satisfied by manually remaining in a configuration that will not support protection if a subsequent adjusting EDG voltage output breaker adjusting EDG output voltage to the OP-902-000, Standard Post event would occu to within the required indication required band for the output breaker to Trip Actions band for output breaker automatically clos (TM-OP-100-03, CT-03) closure.
Critical Task (NUREG-1021, Rev. 11 Appendix D)
- 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.
- Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-102 NRC Exam Scenario 3 D-1 Rev 0
REFERENCES Event Procedures 1 OP-500-007, Control Cabinet G, Rev. 023 OP-901-112, Charging or Letdown Malfunction, Rev 8 Technical Specification 3.1. OP-009-007, Plant Protection System, Rev. 20 OP-903-013, Monthly Channel Checks, Rev. 19 Technical Specification 3. Technical Specification 3. OP-901-221, Secondary System Transient, Rev. 5 OP-500-012, Control Room Cabinet N, Att. 4.61, Rev 29 Technical Specification 3.7. OP-901-220, Loss of Condenser Vacuum, Rev. 304 OP-901-212, Rapid Plant Power Reduction, Rev. 14 5 OP-902-000, Standard Post Trip Actions, Rev. 16 OP-901-103, Emergency Boration, Rev 4 6 OP-902-003, Loss of Offsite Power / Loss of Forced Circulation, Rev. 11 OP-902-009, Standard Appendices, Rev. 319 7 OP-902-000, Standard Post Trip Actions, Rev. 16 GEN EN-OP-115, Conduct of Operations, Rev. 26 EN-OP-115-08, Annunciator Response, Rev. 5 EN-OP-200, Plant Transient Response Rules, Rev. 6 OI-038-000, EOP Operations Expectations / Guidance, Rev. 19 OP-100-017, EOP Implementation Guide, Rev 5 TM-OP-100-03, Simulator Training, Rev. 20-5-2020 NRC Exam Scenario 3 D-1 Rev 0
Appendix D Scenario Outline Form ES-D-1 Facility:
Waterford 3 Scenario No.: 4 Op Test No.: 1 0B Examiners: Operators:
Initial Conditions: Reactor power is 100%. AB Buses are aligned to Train B.
Turnover: Protected Train is B; Maintain 100%.
Critical Tasks: (1) Isolate Most Affected Steam Generator (2) Cool and Depressurize RCS to Prevent Lifting Affected SG Safety Valves Event Mal Event Event N N Type* Description C - BOP Secure Auxiliary Component Cooling Water Pump A 1 CC02A C - SRO following chemical mixing. Component Cooling Water TS - SRO Header A Controller fails to manua I - ATC Reactor Coolant pressure instrument RC-IPT-9120 B 2 RC36B fails high, requiring removal of Diverse Reactor Trip I - SRO from servic C - ATC Letdown Flow Control Valve, CVC-113A, fails closed 3 CV30A2 requiring entry into OP-901-112, Charging or Letdown C - SRO Malfunctio R-ATC C-BOP Steam Generator #2 Tube Leak. Requires a rapid plant 4 SG01B C-SRO power reduction IAW OP-901-21 TS-SRO I - BOP Steam Generator #1 Feed Flow Instrument, FW-IFR-5 FW26A 1111, fails low requiring implementation of OP-901-201, I - SRO Steam Generator Level Control Malfunctio Steam Generator tube leakage leading to Reactor Trip and Safety Injection 6 SG01B M - All (Critical Task 1, Isolate Most Affected Steam Generator)
(Critical Task 2, Cool and Depressurize RCS to Prevent Lifting Affected SG Safety Valves C - BOP 7 EG05 Main Generator Exciter Field Breaker fails to trip C - SRO C - BOP SIAS push buttons on CP-7 fail to actuate Safety 8 XXXXX C - SRO Injection signal.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor-1-2020 NRC Exam Scenario 4 D-1 Rev 1
Appendix D Scenario Outline Form ES-D-1 Scenario Quantitative Attributes 1. Malfunctions after EOP entry (1-2) 2 2. Abnormal events (2-4) 3 3. Major transients (1-2) 1 4. EOPs entered/requiring substantive actions (1-2) 1 5. Entry into a contingency EOP with substantive actions (> 1 per scenario set) 0 6. Preidentified critical tasks (> 2) 2-2-2020 NRC Exam Scenario 4 D-1 Rev 1
NRC Scenario 4 - Narrative The crew assumes the shift at 100% power with instructions to maintain 100% power. No equipment is out of service.
Event 1: Following crew turnover, the crew is directed to secure Auxiliary Component Cooling Water Pump A following basin chemical mixing in accordance with OP-002-001, Auxiliary Component Cooling Water. After ACCW Pump A is off, controller CC-ITIC-7070A for ACCW temperature control will fail in manual. The SRO should declare ACCW Train A inoperable and enter a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action for Tech Spec 3.7.3 as well as cascading Tech Specs. The SRO should address the need to accomplish surveillance OP-903-066, Electrical Breaker Alignment Checks, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to comply with Tech Spec 3.8.1.1.b.
They must also address the need to accomplish the requirements of Tech Spec 3.8.1.1.d within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Event 2: When the Tech Spec review is complete, Reactor Coolant pressure instrument RC-IPT-9120 B will fail high. Based on the direction in the annunciator response procedure, the SRO should direct the ATC to remove Diverse Reactor Trip from service using OP-004-021, Anticipated Transient System.
Event 3: After Diverse Reactor Trip has been removed from service, the in-service Letdown flow control valve, CVC-113A, fails closed. The SRO should enter OP-901-112, Charging or Letdown Malfunction and implement Section E2, Letdown Malfunction, and place the backup flow control valve, CVC-113B, in-service. The SRO may implement EN-OP-200, Transient Response Rules.
Event 4:. After the crew has restored Letdown to automatic with CVC-113B in service, Steam Generator 2 develops a tube leak at ~ 600 gpd. The SRO should implement OP-901-202, Steam Generator Tube Leakage or High Activity. The CRS should determine that based on leak indications, the limits of Technical Specification 3.4.5.2 have been exceeded for Primary-to-Secondary Leakage and enter Tech Spec 3.4.5.2 Action a. The SRO should also determine that the current leakage requires implementation of OP-901-212, Rapid Plant Power Reduction. The crew should commence a rapid power reduction to lower power to < 50% power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Event 5: During the power reduction, after Main Turbine load reduction has commenced, Steam generator #1 Feed Flow Instrument FW-IFR-1111 fails low requiring manual control of Feedwater Control and restoring of Steam Generator water level. The SRO should enter OP-901-201, Steam Generator Level Control Malfunction and evaluate and enter Technical Requirements Manual 3.3.5 action a for the Ultrasonic Flowmeter being inoperable. The BOP will control Steam Generator #1 level in manual for the remainder of the power reduction.
Event 6: Once the crew has taken action to stabilize steam generator level and with the power reduction still in progress, the steam generator tube leak will get worse and will turn into a Steam Generator Tube Rupture. The ATC will note that Pressurizer level is lowering and all available Charging pumps will be operating. When it is determined that the Charging system cannot maintain RCS inventory, the SRO should direct the ATC to perform a manual Reactor Trip and to initiate a manual Safety Injection (SIAS)
and Containment Isolation (CIAS). The SRO should enter OP-902-007, Steam Generator Tube Rupture Recovery Procedure. The crew should perform a rapid RCS cooldown to < 520 °F and isolate Steam Generator #2.
Event 7: When the reactor is tripped, the Main Generator Exciter Field Breaker will fail to open, requiring the BOP operator open the breaker manually.
Event 8: When the CRS directs the actuation of Safety Injection Actuation Signal, the buttons on CP-7 will fail to initiate the signal. The Operators will have to initiate the SIAS from CP-8.
The scenario can be terminated after the crew has completed the Reactor Coolant System rapid cooldown and isolated Steam Generator 2, and commenced an RCS depressurization or at the lead examiners discretio NRC Exam Scenario 4 D-1 Rev 1
NRC Scenario 4 - Critical Task Determination Measurable Performance Performance Critical Task Safety Significance Cueing Indicator Feedback A ruptured SG presents a flowpath Places MS-116B setpoint to 980 for radioactive release. Isolating psig and verifies controller in the SG will terminate the release AUTO lowering the dose to plant personnel and the publi Verifies MS-124B is closed CT-1: Isolate Most Affected Steam RCS THOT < 520°F is based on Verifies FW-184B is closed Generator Emergency Operating Procedure IF EFAS-2 is NOT initiated, THEN This task is satisfied by manually standard. Isolating the SG at a temperature higher than this value closes EFW Isolation valves EFW-closing Main Steam Isolation valve #2, may complicate the response later 228B and EFW-229B Main Feedwater Isolation Valve #2, and MS-401B after RCS THOT has reduced in the scenario by not allowing a Procedurally driven Places EFW Flow Control valves Proper indication below 520°F and prior to exiting the full depressurization of the RCS to from OP-902-007 in MAN and THEN closes EFW- for listed below the lowest Main Steam step 17 components step (Step 17) to isolate the most Safety Valve setpoint. This higher 224B and EFW-223B affected Steam generator in OP-902- RCS pressure would allow the Closes MS-401B 007, Steam Generator Tube Rupture RCS to continue to flow into the Recovery. This task becomes affected SG ultimately raising the Close Main Steam Line 2 Drains applicable once the SGTR has SG level and pressure to the MS-120B and MS-119B commenced. (OP-902-007, 17) safety valve setpoint and cause a release that could have been Closes Steam Generator prevented. OPS management Blowdown Isolation valves BD-Standard documented in TM-OP- 103B and BD-102B 100-0 Checks SG2 East Side Main (TM-OP-100-03, CT-14) Steam Safety valves are close NRC Exam Scenario 4 D-1 Rev 1
NRC Scenario 4 - Critical Task Determination Measurable Performance Performance Critical Task Safety Significance Cueing Indicator Feedback A ruptured SG presents a flowpath for radioactive release. Cooling and depressurizing the RCS will prevent lifting an Atmospheric Dump Valve or Main Steam Safety CT-2: Cool and Depressurize RCS to Valve and thus prevent an Prevent Lifting Affected SG Safety unnecessary release and lowering Valves the dose to plant personnel and the publi Opens Steam Bypass Control This task is satisfied by performing a RCS THOT > 520°F valve cooldown to THOT less than 520°F using RCS THOT < 520°F and RCS pressure < 930 psia are based on RCS THOT <
the Steam Bypass Control System prior RCS pressure > Opens either Main Spray valve(s)
Emergency Operating Procedure 520°F to closing the affected MSIV and 930 psia or Auxiliary Spray valve(s)
Standard. RCS temperature must commencing RCS depressurization be reduced below the temperature RCS pressure towards <930 psia prior to lifting an corresponding to the saturation Procedurally driven Depressurization does not lowering Atmospheric Dump Valve or Main pressure for the ADV/MSSV set from OP-902-007 necessarily need to be completed Steam Safety Valve on the affected S point prior to closing the affected step 11 and step 12 but must be commenced to This task becomes applicable once the SG MSIV. RCS must be receive credit for the task.
SGTR has commence depressurized to below the lift setpoint of the MSSVs to prevent (OP-902-007, 11, 12) overfilling the SG and lifting the MSSV OPS management Standard documented in TM-OP-100-0 (TM-OP-100-03, CT-21)
Critical Task (NUREG-1021, Rev. 11 Appendix D)
- 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.
- Causing an unnecessary plant trip or ESF actuation may constitute a CT failure. Actions taken by the applicant(s) will be validated using the methodology for critical tasks in Appendix D to NUREG-102 NRC Exam Scenario 4 D-1 Rev 1
NRC Scenario 4 REFERENCES Event Procedures 1 OP-002-001, Auxiliary Component Cooling Water, Rev. 315 Technical Specification 3.7.3 and cascading Tech Specs Technical Requirements Manual 3.7.15 OP-100-014, Technical Specification and Technical Requirements Compliance, Rev. 354 2 OP-500-011, Control Room Cabinet M, Rev. 41 OP-004-021, Anticipated Transient System, Rev. 302 3 OP-901-112, Charging or Letdown Malfunction, Rev. 8 4 OP-901-202, Steam Generator Tube Leakage or High Activity, Rev. 15 OP-901-212, Rapid Plant Power Reduction, Rev. 14 Technical Specification 3.4. OP-901-201, Steam Generator Level Control Malfunction, Rev. 6 Technical Requirements Manual 3. OP-902-000, Standard Post Trip Actions, Rev. 16 OP-902-007, Steam Generator Tube Rupture Recovery Procedure, Rev. 18 OP-902-009, Standard Appendices, Rev. 319 7 OP-902-000, Standard Post Trip Actions, Rev. 16 GEN EN-OP-115, Conduct of Operations, Rev. 26 EN-OP-115-08, Annunciator Response, Rev. 5 EN-OP-200, Plant Transient Response Rules, Rev. 6 OI-038-000, EOP Operations Expectations / Guidance, Rev. 19 OP-100-017, Emergency Operating Procedure Implementation Guide, Rev 5 TM-OP-100-03, Simulator Training, Rev. 21-6-2020 NRC Exam Scenario 4 D-1 Rev 1