ML20042C044
| ML20042C044 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 10/12/2019 |
| From: | Greg Werner Operations Branch IV |
| To: | Arizona Public Service Co |
| References | |
| Download: ML20042C044 (51) | |
Text
ES-401 PWR Examination Outline Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 Facility: Palo Verde Date of Exam: October 4, 2019 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
3 2
4 N/A 3
3 N/A 3
18 6
2 2
2 1
1 1
2 9
4 Tier Totals 5
4 5
4 4
5 27 10
- 2.
Plant Systems 1
3 3
3 3
2 2
2 3
2 2
3 28 5
2 1
0 1
1 1
1 1
1 1
1 1
10 3
Tier Totals 4
3 4
4 3
3 3
4 3
3 4
38 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
2 3
2 3
Note: 1.
Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4.
Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 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; BW E02&E10; CE E02)
Reactor Trip, Stabilization, Recovery / 1 X
EA1.02 Ability to operate and monitor the following as they apply to a reactor trip: MFW System 3.8 19 000008 (APE 8) Pressurizer Vapor Space Accident / 3 X
AK1.01 Knowledge of the operational implications of the following concepts as they apply to a pressurizer vapor space accident: Thermodynamics and flow characteristics of open or leaking valves 3.2 12 000009 (EPE 9) Small Break LOCA / 3 X
EK3.21 Knowledge of the reasons for the following responses as they apply to the small break LOCA and the following: Actions contained in the EOP for SBLOCA / leak 4.2 13 000011 (EPE 11) Large Break LOCA / 3 X
EA2.13 Ability to determine or interpret the following as they apply to a Large Break LOCA: Difference between overcooling and LOCA indications 3.7 14 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 X
AA1.07 Ability to operate and / or monitor the following as they apply to the reactor coolant pump malfunctions (Loss of RC Flow): RCP seal water injection subsystem 3.5 15 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 X
G.2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation 4.3 16 000025 (APE 25) Loss of Residual Heat Removal System / 4 X
AK2.03 Knowledge of the interrelations between the loss of Residual Heat Removal System and the following: service water or closed cooling water pumps 2.7 17 000026 (APE 26) Loss of Component Cooling Water / 8 X
AA2.06 Ability to determine and interpret the following as they apply to the loss of Component Cooling Water: the length of time after the loss of CCW flow to a component before that component may be damaged 2.8 18 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 Not sampled 000029 (EPE 29) Anticipated Transient Without Scram / 1 X
G2.4.2 Knowledge of system set points, interlocks, and automatic actions associated with EOP entry conditions 4.5 20 000038 (EPE 38) Steam Generator Tube Rupture / 3 X
EK3.08 Knowledge of the reasons for the following responses as they apply to the SGTR: Criteria for securing RCP 4.1 21 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 X
CE E05 Excess Steam Demand:
EK1.1 Knowledge of the operational implications of the following concepts as they apply to an ESD:
components, capacities, and functions of emergency systems 3.0 22 000054 (APE 54; CE E06) Loss of Main Feedwater /4 X
AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of main Feedwater (MFW): Reactor and/or turbine trip, manual and automatic 4.1 23 000055 (EPE 55) Station Blackout / 6 X
EA2.06 Ability to determine and interpret the following as they apply to the Station Blackout:
Faults and lockouts that must be cleared prior to re-energizing buses 3.7 24
ES-401 3
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 000056 (APE 56) Loss of Offsite Power / 6 X
AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer 3.5 25 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 X
AA1.01 Ability to operate and / or monitor the following as they apply to the Loss of Vital AC Instrument Bus: Manual inverter swapping 3.7 26 000058 (APE 58) Loss of DC Power / 6 X
AK1.01 Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation 2.8 27 000062 (APE 62) Loss of Nuclear Service Water / 4 Not sampled 000065 (APE 65) Loss of Instrument Air / 8 X
G2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 4.2 28 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 X
AK2.07 Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Turbine / Generator Control 3.6 29 (W E04) LOCA Outside Containment / 3 N/A for CE design (W E11) Loss of Emergency Coolant Recirculation / 4 N/A for CE design (BW E04; W E05) Inadequate Heat TransferLoss of Secondary Heat Sink / 4 N/A for CE design K/A Category Totals:
3 2
4 3
3 3
Group Point Total:
18/6
ES-401 4
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 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 Not sampled 000003 (APE 3) Dropped Control Rod / 1 Not sampled 000005 (APE 5) Inoperable/Stuck Control Rod / 1 Not sampled 000024 (APE 24) Emergency Boration / 1 X
AK1.01 Knowledge of the operational implications of the following concepts as they apply to emergency boration:
Relationship between boron addition and change in Tave 3.4 47 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 X
AA1.04 Ability to predict and/or monitor the following as they apply to the Pressurizer Level Control Malfunction: Regenerative heat exchanger and temperature limits 2.7 50 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 Not sampled 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 X
AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Termination of startup following loss of intermediate range instrumentation 3.2 48 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 Not sampled 000037 (APE 37) Steam Generator Tube Leak / 3 Not sampled 000051 (APE 51) Loss of Condenser Vacuum / 4 Not sampled 000059 (APE 59) Accidental Liquid Radwaste Release / 9 X
AA2.05 Ability to determine and interpret the following as they apply to the Accidental Liquid Radwaste Release: The occurrence of automatic safety actions as a result of a high PRM signal 3.6 49 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 Not sampled 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 Not sampled 000067 (APE 67) Plant Fire On Site / 8 Not sampled 000068 (APE 68; BW A06) Control Room Evacuation / 8 X
AK2.01 Knowledge of the interrelations between the Control Room Evacuation and the following: Auxiliary shutdown panel layout 3.9 53 000069 (APE 69; W E14) Loss of Containment Integrity / 5 Not sampled 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
4 Not sampled 000076 (APE 76) High Reactor Coolant Activity / 9 X
AK2.01 Knowledge of the interrelations between the High Reactor Coolant Activity and the following: Process radiation monitors 2.6 51 000078 (APE 78*) RCS Leak / 3 Not sampled (W E01 & E02) Rediagnosis & SI Termination / 3 N/A for CE design (W E13) Steam Generator Overpressure / 4 N/A for CE design (W E15) Containment Flooding / 5 N/A for CE design (W E16) High Containment Radiation /9 N/A for CE design
ES-401 5
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 (BW A01) Plant Runback / 1 N/A for CE design (BW A02 & A03) Loss of NNI-X/Y/7 N/A for CE design (BW A04) Turbine Trip / 4 N/A for CE design (BW A05) Emergency Diesel Actuation / 6 N/A for CE design (BW A07) Flooding / 8 N/A for CE design (BW E03) Inadequate Subcooling Margin / 4 N/A for CE design (BW E08; W E03) LOCA CooldownDepressurization / 4 N/A for CE design (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 Not sampled (BW E13 & E14) EOP Rules and Enclosures N/A for CE design (CE A11**; W E08) RCS OvercoolingPressurized Thermal Shock / 4 X
G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation 4.4 46 (CE A16) Excess RCS Leakage / 2 X
AK1.3 Knowledge of the operational implications of the following concepts as they apply to the Excess RCS Leakage:
Annunciators and conditions, indicating signals, and remedial actions associated with the Excess RCS Leakage 3.2 52 (CE E09) Functional Recovery Not sampled (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 X
G2.2.4 (multi-unit license) Ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility 3.6 54 K/A Category Point Totals:
2 2
1 1
1 2
Group Point Total:
9/4
ES-401 6
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 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 Pump X
X K1.04 Knowledge of the physical connections and/or cause effect relationship between the RCPS and the following system: CVCS A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: problems with RCP seals, especially rates of seal leak-off 2.6 3.5 1
55 004 (SF1; SF2 CVCS) Chemical and Volume Control X
X K3.06 Knowledge of the effect that a loss or malfunction of the CVCS will have on the following: RCS temperature and pressure A4.07 Ability to manually operate and/or monitor in the control room: Boration/dilution 3.4 3.9 2
3 005 (SF4P RHR) Residual Heat Removal X
K2.03 Knowledge of bus power supplies to the following: RCS pressure boundary motor-operated valves 2.7 4
006 (SF2; SF3 ECCS) Emergency Core Cooling X
X A1.14 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: Reactor vessel level A3.04 Ability to monitor automatic operation of the ECCS, including: Cooling water systems 3.6 3.8 6
7 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X
K3.01 Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: Containment 3.3 8
008 (SF8 CCW) Component Cooling Water X
K2.02 Knowledge of bus power supplies to the following: CCW pump, including backup 3.0 9
010 (SF3 PZR PCS) Pressurizer Pressure Control X
K4.03 Knowledge of the PZR PCS design feature(s) and/or interlocks which provide for the following: over pressure control 3.8 10 012 (SF7 RPS) Reactor Protection X G 2.4.46 Ability to verify that alarms are consistent with the plant conditions 4.2 31 013 (SF2 ESFAS) Engineered Safety Features Actuation X
X K6.01 Knowledge of the effect of a loss or malfunction of the following will have on the ESFAS system: Sensors and detectors A4.02 Ability to manually operate and or monitor in the control room: Reset of ESFAS channels 2.7 4.3 32 33 022 (SF5 CCS) Containment Cooling X
K3.01 Knowledge of the effect that a loss or malfunction of the CCS will have on the following: Containment equipment subject to damage by high or low temperature, humidity, and pressure 2.9 34 025 (SF5 ICE) Ice Condenser N/A for PV
ES-401 7
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 026 (SF5 CSS) Containment Spray X
A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: failure of spray pump 3.9 35 039 (SF4S MSS) Main and Reheat Steam X
X K5.05 Knowledge of the operational implications of the following concepts as they apply to MRSS: Bases for RCS cooldown limits A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Indications and alarms for main steam and area radiation monitors (during SGTR) 2.9 3.4 36 30 059 (SF4S MFW) Main Feedwater X
K4.16 Knowledge of the design feature(s) and/or interlocks which provide for the following: Automatic trips for MFW pumps 3.1 37 061 (SF4S AFW)
Auxiliary/Emergency Feedwater X
X G 2.2.38 Knowledge of conditions and limitations in the facility license K5.01 Knowledge of the operational implications of the following concepts as they apply to the AFW: Relationship between AFW flow and RCS heat transfer 3.6 3.6 38 11 062 (SF6 ED AC) AC Electrical Distribution X
A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls, including:
Significance of D/G load limits 3.4 39 063 (SF6 ED DC) DC Electrical Distribution X
K4.01 Knowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: Manual/automatic transfers of control 2.7 40 064 (SF6 EDG) Emergency Diesel Generator X
X K6.08 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Fuel oil storage tanks G.2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
3.2 4.2 41 5
073 (SF7 PRM) Process Radiation Monitoring X
K1.01 Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems: Those systems served by PRMs 3.6 42 076 (SF4S SW) Service Water X
K2.01 Knowledge of bus power supplies to the following: Service water pump 2.7 43 078 (SF8 IAS) Instrument Air X
K1.05 Knowledge of the physical connections and/or cause-effect relationships between the IAS system and the following systems: MSIV air 3.4 44
ES-401 8
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 103 (SF5 CNT) Containment X
A3.01 Ability to monitor automatic operation of the containment system, including:
Containment isolation 3.9 45 053 (SF1; SF4P ICS*) Int. Control N/A for CE design K/A Category Point Totals:
3 3
3 3
2 2
2 3
2 2
3 Group Point Total:
28/5
ES-401 9
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 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 Not sampled 002 (SF2; SF4P RCS) Reactor Coolant X
K5.08 Knowledge of the operational implications of the following concepts as they apply to the RCS: Why PZR level should be kept within the programmed band 3.4 56 011 (SF2 PZR LCS) Pressurizer Level Control X
K6.05 Knowledge of the effect of a loss or malfunction of the following will have on the PZR LCS: Function of PZR level gauges as post-accident monitors 3.1 57 014 (SF1 RPI) Rod Position Indication X
A1.04 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RPIS controls, including: Axial and radial power distribution 3.5 58 015 (SF7 NI) Nuclear Instrumentation Not sampled 016 (SF7 NNI) Nonnuclear Instrumentation Not sampled 017 (SF7 ITM) In-Core Temperature Monitor Not sampled 027 (SF5 CIRS) Containment Iodine Removal X
K1.01 Knowledge of the physical connections and/or cause-effect relationships between the CIRS system and the following systems: CSS 3.4 60 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control Not sampled 029 (SF8 CPS) Containment Purge X
A3.01 Ability to monitor automatic operation of the Containment Purge System including:
CPS isolation 3.8 64 033 (SF8 SFPCS) Spent Fuel Pool Cooling X
K3.02 Knowledge of the effect that a loss or malfunction of the SFPCS will have on the following: Area and ventilation radiation monitoring systems 2.8 61 034 (SF8 FHS) Fuel-Handling Equipment Not sampled 035 (SF 4P SG) Steam Generator X G2.2.3 Knowledge of the design, procedural, and operational differences between units 3.8 62 041 (SF4S SDS) Steam Dump/Turbine Bypass Control X
K4.17 Knowledge of SDS design feature(s) and/or interlock(s) which provide for the following: Reactor trip 3.7 59 045 (SF 4S MTG) Main Turbine Generator X
A2.17 Ability to (a) predict the impacts of the following malfunctions or operations on the MTG; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Malfunction of electrohydraulic control 2.7 63 055 (SF4S CARS) Condenser Air Removal Not sampled 056 (SF4S CDS) Condensate Not sampled 068 (SF9 LRS) Liquid Radwaste Not sampled 071 (SF9 WGS) Waste Gas Disposal Not sampled 072 (SF7 ARM) Area Radiation Monitoring Not sampled 075 (SF8 CW) Circulating Water Not sampled
ES-401 10 Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 079 (SF8 SAS**) Station Air Not sampled 086 Fire Protection X
A4.02 Ability to manually operate and/or monitor in the control room: Fire detection panels 3.5 65 K/A Category Point Totals:
1 0
1 1
1 1
1 1
1 1
1 Group Point Total:
10/3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 Facility: Palo Verde Date of Exam: October 4, 2019 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of license status, 10CFR55, etc.
3.3 66 2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations Memos, etc 2.7 67 2.1.
2.1.
2.1.
Subtotal 2
- 2. Equipment Control 2.2.7 Knowledge of the process for conducting special or infrequent tests 2.9 68 2.2.21 Knowledge of pre-and post-maintenance operability requirements 2.9 69 2.2.35 Ability to determine Technical Specification Mode of Operation 3.6 70 2.2.
2.2.
2.2.
Subtotal 3
- 3. Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions 3.2 71 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
3.2 72 2.3.
2.3.
Subtotal 2
- 4. Emergency Procedures/Plan 2.4.17 Knowledge of EOP terms and definitions 3.9 73 2.4.19 Knowledge of EOP layout, symbols, and icons 3.4 74 2.4.31 Knowledge of annunciator alarms, indications, or response procedures 4.2 75 2.4.
2.4.
2.4.
Subtotal 3
Tier 3 Point Total 10 10
ES-401 Record of Rejected K/As Form ES-401-4 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 Tier /
Group Randomly Selected K/A Reason for Rejection 2 / 1 (Q1) 003 K1.02 003 K1.02 asks about RCP cooling and ventilation, however there are other questions which also ask about RCP cooling and there is no direct ventilation cooling for RCPs at PVNGS so we could not write a question without significantly overlapping with other questions and/or portions of the exam. Reselected 003 K1.04.
2 / 1 (Q5) 005 G 2.2.44 Due to multiple systems being selected three times between the RO and SRO exams with other Tier 2 Group 1 systems only selected once, replaced KA to ensure all systems are tested at least twice on either the RO and SRO exams prior to any system being tested three times. Residual Heat Removal System was already selected three times and Emergency Diesel Generators was only selected once, so we replaced Residual Heat Removal System with Emergency Diesel Generators but kept the generic KA (G 2.2.44). Reselected 064 K4.17.
2 / 1 (Q11) 010 K5.01 Unable to write a clear question about the condition of the fluid in the Pressurizer due to the mix of saturated and subcooled mixture with no clear delineation between when it transitions from one state to the other, as well as to ask about the operational implications of the status of the fluid. All attempts to write the question to this KA resulted in a question where the distractors had very low plausibility or the question lacked operational validity. The only other K5 KA in 010 asks about constant enthalpy through a valve and we already have a question about Pressurizer Relief Valves.
Reselected 061 K5.01 1 / 1 (Q14) 011 EA2.08 011 EA2.08 asks about conditions for recovery from a LBLOCA when conditions reach stable phase. We had written a question for this KA that was operationally relevant and appropriate for the RO level, however one of the scenarios is a LBLOCA which has a failure of containment suction valves to automatically open, which is exactly what the question was asking about. We tried to create another question actions taken when RAS occurs however we could not come up with an operationally relevant question at the RO level that would not significantly overlap with the scenario event.
Reselected 011 EA2.13.
1 / 1 (Q19) 027 AA1.04 Due to multiple systems being selected for both the RO and SRO exams with other Tier 1 Group 1 systems not selected on either exam, replaced KA to ensure all systems tested at least once on either the RO and SRO exams prior to any system being tested twice. PPCS Malfunction was already selected for the SRO exam and Reactor trip was not selected for either, so we replaced PPCS Malfunctions with Reactor Trip but kept the topic as an A1. Reselected 007 EA1.02.
1 / 1 (Q21) 038 EK3.09 038 EK3.09 asks about the criteria for securing/throttling ECCS and Question 6 asks about predicting/monitoring changes in parameters associated with operating the ECCS controls including reactor vessel level -
which is only really used when determining if SI flow can be stopped/throttled, so questions on both K/As kept coming back to SI throttle criteria. Reselected 038 EK3.08.
ES-401 Record of Rejected K/As Form ES-401-4 2019 PVNGS NRC Initial Written Exam Outline - RO - ES-401-2/3/4 Rev 4 2 / 1 (Q30) 012 A2.07 Due to multiple systems being selected three times between the RO and SRO exams with other Tier 2 Group 1 systems only selected once, replaced KA to ensure all systems tested at least twice on either the RO and SRO exams prior to any system being tested three times. Reactor Protection was selected three times and Main and Reheat Steam was only selected once, so we replaced Reactor Protection with Main and Reheat Steam System, but kept the topic as an A2. Reselected 039 A2.03.
1 / 2 (Q46) 001 G 2.2.3 PVNGS does not have any design, procedural, or operational differences between the units for continuous rod withdrawal situations, nor any other non-sampled Tier 1 Group 2 systems. Additionally, due to multiple systems being selected for both the RO and SRO exams with other Tier 1 Group 1 systems not selected on either exam, replaced KA to ensure all systems tested at least once on either the RO and SRO exams prior to any system being tested twice. Continuous Rod Withdrawal was already selected for the SRO exam and RCS Overcooling was not selected for either, so we replaced Continuous Rod Withdrawal with RCS Overcooling and changed from G 2.2.3 to G 2.1.7. Reselected CE A11 G 2.1.7.
1 / 2 (Q50) 074 EK2.04 After multiple unsuccessful attempts to write an operationally valid question at the RO level, reselected 028 AA1.04.
1 / 2 (Q51) 076 AA1.04 Due to the KA reselections in order to ensure no system/evolution was selected twice before all systems/evolutions were selected once and for skyscraper balancing, reselected 076 AK2.01. Original question written the original KA remained virtually unchanged.
1 / 2 (Q53)
CE E09 EK2.1 Due to multiple systems being selected for both the RO and SRO exams with other Tier 1 Group 2 systems not selected on either exam, replaced KA to ensure all systems tested at least once on either the RO and SRO exams prior to any system being tested twice. Functional Recovery was already selected for the SRO exam and Control Room Evacuation was not selected for either, so we replaced Functional Recovery with Control Room Evacuation but kept the topic as a K2. Reselected 068 AK2.01.
1 / 2 (Q54)
CE E13 G 2.4.8 There are no AOPs which would be used in conjunction with the LOFC/LOOP or Blackout EOPs at PVNGS. Reselected CE E13 G 2.2.4 2 / 2 (Q59) 017 K4.03 Due to multiple systems being selected for both the RO and SRO exams with other Tier 2 Group 2 systems not selected on either exam, replaced KA to ensure all systems tested at least once on either the RO and SRO exams prior to any system being tested twice. In-Core Temperature Monitor System was already selected for the SRO exam and Steam Dump/Turbine Bypass Control was not selected for either, so we replaced In-Core Temperature Monitor System with Steam Dump/Turbine Bypass Control but kept the topic as a K4. Reselected 041 K4.17.
2 / 2 (Q64) 068 A3.02 068 A3.02 is a liquid radwaste KA and PVNGS is a zero liquid release plant.
Based on already having an accidental liquid release question and PVNGS being a zero release plant, reselected 029 A3.01
ES-401 PWR Examination Outline Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - SRO - ES-401-2/3/4 Rev 4 Facility: Palo Verde Date of Exam: October 4, 2019 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
N/A N/A 18 3
3 6
2 9
2 2
4 Tier Totals 27 5
5 10
- 2.
Plant Systems 1
28 2
3 5
2 10 1
1 1
3 Tier Totals 38 4
4 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
2 1
2 2
Note: 1.
Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4.
Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - SRO - ES-401-2/3/4 Rev 4 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 Not sampled 000008 (APE 8) Pressurizer Vapor Space Accident / 3 Not sampled 000009 (EPE 9) Small Break LOCA / 3 X
G2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation (CFR 43.5) 4.4 76 000011 (EPE 11) Large Break LOCA / 3 X
EA2.10 Ability to determine or interpret the following as they apply to a Large Break LOCA: Verification of adequate core cooling 4.7 77 000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 Not sampled 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 Not sampled 000025 (APE 25) Loss of Residual Heat Removal System / 4 Not sampled 000026 (APE 26) Loss of Component Cooling Water / 8 Not sampled 000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 X
AA2.06 Ability to determine and interpret the following as they apply to the Pressurizer pressure Control Malfunction: Conditions which require a plant shutdown (CFR 43.5) 3.9 78 000029 (EPE 29) Anticipated Transient Without Scram / 1 Not sampled 000038 (EPE 38) Steam Generator Tube Rupture / 3 X G2.4.41 Knowledge of the emergency action level thresholds and classifications (CFR 43.5) 4.6 79 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line Rupture-Excessive Heat Transfer / 4 Not sampled 000054 (APE 54; CE E06) Loss of Main Feedwater /4 Not sampled 000055 (EPE 55) Station Blackout / 6 X
EA2.03 Ability to determine and interpret the following as they apply to the Station Blackout:
Actions necessary to restore power (CFR 43.5) 4.7 80 000056 (APE 56) Loss of Offsite Power / 6 Not sampled 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 Not sampled 000058 (APE 58) Loss of DC Power / 6 Not sampled 000062 (APE 62) Loss of Nuclear Service Water / 4 X G2.2.40 Ability to apply Technical Specifications for a system (43.2 / 43.5 / 45.3) 4.7 81 000065 (APE 65) Loss of Instrument Air / 8 Not sampled 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 Not sampled (W E04) LOCA Outside Containment / 3 N/A for CE designs (W E11) Loss of Emergency Coolant Recirculation / 4 N/A for CE designs (BW E04; W E05) Inadequate Heat Transfer-Loss of Secondary Heat Sink / 4 N/A for CE designs K/A Category Totals:
3 3
Group Point Total:
6
ES-401 3
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - SRO - ES-401-2/3/4 Rev 4 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 X
G2.2.22 Knowledge of limiting conditions for operations and safety limits (CFR 43.2) 4.7 82 000003 (APE 3) Dropped Control Rod / 1 Not sampled 000005 (APE 5) Inoperable/Stuck Control Rod / 1 Not sampled 000024 (APE 24) Emergency Boration / 1 Not sampled 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 Not sampled 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 Not sampled 000033 (APE 33) Loss of Intermediate Range Nuclear Instrumentation / 7 Not sampled 000036 (APE 36; BW/A08) Fuel Handling Incidents / 8 Not sampled 000037 (APE 37) Steam Generator Tube Leak / 3 X
AA2.10 Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: Tech-Spec limits for RCS leakage 4.1 83 000051 (APE 51) Loss of Condenser Vacuum / 4 Not sampled 000059 (APE 59) Accidental Liquid Radwaste Release / 9 Not sampled 000060 (APE 60) Accidental Gaseous Radwaste Release / 9 Not sampled 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 Not sampled 000067 (APE 67) Plant Fire On Site / 8 Not sampled 000068 (APE 68; BW A06) Control Room Evacuation / 8 Not sampled 000069 (APE 69;) Loss of Containment Integrity / 5 X
G2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits CFR (41.5 / 41.7 / 43.2) 4.2 84 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling /
4 Not sampled 000076 (APE 76) High Reactor Coolant Activity / 9 Not sampled 000078 (APE 78*) RCS Leak / 3 Not sampled (WE01 & E02) Rediagnosis & SI Termination / 3 N/A for CE designs (W E13) Steam Generator Overpressure / 4 N/A for CE designs (W E15) Containment Flooding / 5 N/A for CE designs (W E16) High Containment Radiation /9 N/A for CE designs (BW A01) Plant Runback / 1 N/A for CE designs (BW A02 & A03) Loss of NNI X/Y/7 N/A for CE designs (BW A04) Turbine Trip / 4 N/A for CE designs (BW A05) Emergency Diesel Actuation / 6 N/A for CE designs (BW A07) Flooding / 8 N/A for CE designs (BW E03) Inadequate Subcooling Margin / 4 N/A for CE designs (BW E08; W E03) LOCA Cooldown-Depressurization / 4 N/A for CE designs (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 Not sampled (BW E13 & E14) EOP Rules and Enclosures N/A for CE designs (CE A11 W E08) RCS Overcooling-Pressurized Thermal Shock / 4 Not sampled
ES-401 4
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - SRO - ES-401-2/3/4 Rev 4 (CE A16) Excess RCS Leakage / 2 Not sampled (CE E09) Functional Recovery X
EA2.1 Ability to determine and interpret the following as they apply to the Functional Recovery:
Facility conditions and selection of appropriate procedures during abnormal and emergency operations (CFR 43.5) 4.4 85 (CE E13*) Loss of Forced Circulation/LOOP/Blackout / 4 Not sampled K/A Category Point Totals:
2 2
Group Point Total:
4
ES-401 5
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - SRO - ES-401-2/3/4 Rev 4 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 Not sampled 004 (SF1; SF2 CVCS) Chemical and Volume Control Not sampled 005 (SF4P RHR) Residual Heat Removal X
A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Pressure transient protection during cold shutdown (CFR 43.5) 3.7 86 006 (SF2; SF3 ECCS) Emergency Core Cooling Not sampled 007 (SF5 PRTS) Pressurizer Relief/Quench Tank X G2.2.40 Ability to apply Technical Specifications for a system (CFR 43.5) 4.7 88 008 (SF8 CCW) Component Cooling Water Not sampled 010 (SF3 PZR PCS) Pressurizer Pressure Control Not sampled 012 (SF7 RPS) Reactor Protection X
G2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits (CFR 43.2) 4.2 87 013 (SF2 ESFAS) Engineered Safety Features Actuation 022 (SF5 CCS) Containment Cooling Not sampled 025 (SF5 ICE) Ice Condenser Not sampled 026 (SF5 CSS) Containment Spray X
A2.07 Ability to (a) predict the impacts of 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: 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 (CFR: 43.5) 2.6 89 039 (SF4S MSS) Main and Reheat Steam Not sampled 059 (SF4S MFW) Main Feedwater Not sampled 061 (SF4S AFW)
Auxiliary/Emergency Feedwater Not sampled 062 (SF6 ED AC) AC Electrical Distribution X
G2.4.6 Knowledge of EOP mitigation strategies (CFR 43.5) 4.7 90 063 (SF6 ED DC) DC Electrical Distribution Not sampled 064 (SF6 EDG) Emergency Diesel Generator Not sampled 073 (SF7 PRM) Process Radiation Monitoring Not sampled 076 (SF4S SW) Service Water Not sampled 078 (SF8 IAS) Instrument Air Not sampled Not Sampled
ES-401 6
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - SRO - ES-401-2/3/4 Rev 4 053 (SF1; SF4P ICS*) Integrated Control N/A for CE designs K/A Category Point Totals:
2 3 Group Point Total:
5 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 001 (SF1 CRDS) Control Rod Drive X
A2.12 Ability to a) predict the impacts of the following malfunctions or operations on the CRDS; and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Erroneous ECP calculation (CFR 43.5) 4.2 91 002 (SF2; SF4P RCS) Reactor Coolant Not sampled 011 (SF2 PZR LCS) Pressurizer Level Control Not sampled 014 (SF1 RPI) Rod Position Indication Not sampled 015 (SF7 NI) Nuclear Instrumentation Not sampled 016 (SF7 NNI) Nonnuclear Instrumentation Not sampled 017 (SF7 ITM) InCore Temperature Monitor X
G2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits (CFR 43.2) 4.2 92 027 (SF5 CIRS) Containment Iodine Removal Not sampled 028 (SF5 HRPS) Hydrogen Recombiner and Purge Control Not sampled 029 (SF8 CPS) Containment Purge Not sampled 033 (SF8 SFPCS) Spent Fuel Pool Cooling Not sampled 034 (SF8 FHS) FuelHandling Equipment X
K4.01 Knowledge of design feature(s) and/or interlocks(s) which provide for the following:
Fuel protection from binding and dropping (CFR 43.5 / 43.7) 3.4 93 035 (SF 4P SG) Steam Generator Not sampled 041 (SF4S SDS) Steam Dump/Turbine Bypass Control Not sampled 045 (SF 4S MTG) Main Turbine Generator Not sampled 055 (SF4S CARS) Condenser Air Removal Not sampled 056 (SF4S CDS) Condensate Not sampled 068 (SF9 LRS) Liquid Radwaste Not sampled 071 (SF9 WGS) Waste Gas Disposal Not sampled 072 (SF7 ARM) Area Radiation Monitoring Not sampled 075 (SF8 CW) Circulating Water Not sampled 079 (SF8 SAS**) Station Air Not sampled 086 Fire Protection Not sampled 050 (SF 9 CRV*) Control Room Ventilation Not sampled K/A Category Point Totals:
1 1
1 Group Point Total:
3
ES-401 7
Form ES-401-2 2019 PVNGS NRC Initial Written Exam Outline - SRO - ES-401-2/3/4 Rev 4
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 2019 PVNGS NRC Initial Written Exam Outline - SRO - ES-401-2/3/4 Rev 4 Facility: Palo Verde Date of Exam: Oct 4, 2019 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.34 Knowledge of primary and secondary plant chemistry limits (CFR: 41.10 / 43.5 / 45.12) 3.5 94 2.1.36 Knowledge of procedures and limitations involved in core alterations (CFR: 41.10 / 43.6 / 45.7) 4.0 95 Subtotal 2
- 2. Equipment Control 2.2.5 Knowledge of the process for making design or operating changes to the facility (CFR: 41.10 / 43.3 /
45.13) 3.2 96 Subtotal 1
- 3. Radiation Control 2.3.11 Ability to control radiation releases (CFR: 41.11 / 43.4 /
45.10) 4.3 97 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc. (CFR 43.4, 43.5) 3.8 98 Subtotal 2
2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc (CFR: 41.7 / 43.5 / 45.12) 4.6 99 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator (CFR: 41.10 / 43.5 /
45.11) 4.1 100 Subtotal 2
Tier 3 Point Total 7
7
ES-401 Record of Rejected K/As Form ES-401-4 2019 PVNGS NRC Initial Written Exam Outline - SRO - ES-401-2/3/4 Rev 4 Tier /
Group Randomly Selected K/A Reason for Rejection 1 / 1 (Q77) 011 EA2.03 The loss of CCW would be significant during a small break LOCA since the long term cooling strategy for a small break LOCA is Shutdown Cooling -
which is cooled via CCW. However during a large break LOCA, the long term cooling is provided by Recirculation Actuation Signal - which is cooled by a combination of the volume of the Refueling Water Tank and Containment Spray flow. Reselected 011 EA2.10.
1 / 1 (Q79) 038 G 2.4.8 The only situation in which an AOP may be used in conjunction with an EOP for a SGTR is if the rupture was first a leak then degraded into a rupture. Even so, the AOP says to GO TO SPTAs if the leak degrades into a rupture (meaning to exit the AOP) and the actions in the optimal EOP (SGTR) contain the same actions as the AOP for supplemental actions such as minimizing release to the environment. The two procedures would not be used concurrently, and the AOP would not be used if a SGTR was in progress as the supplemental actions are contained in the EOPs.
Reselected 038 G 2.4.41.
2 / 1 (Q89) 026 A2.09 Mitigating actions taken in response to a malfunction that sends radioactive water back into the RWT (BWST) are not specifically addressed in the EOPs as these actions would be at the discretion of the EC, which would open the question we wrote up to a potential post-exam challenge. As a result, reselected 026 A2.07.
1 / 1 (Q81) 058 G 2.2.40 Due to multiple systems being selected for both the RO and SRO exams with other Tier 1 Group 1 systems not selected on either exam, replaced KA to ensure all systems tested at least once on either the RO and SRO exams prior to any system being tested twice. Loss of DC power was already selected for the RO exam and Loss of Nuclear Service Water was not selected for either, so we replaced Loss of DC Power with Loss of Nuclear Service Water but kept the generic KA. Reselected 062 G 2.2.40.
1 / 2 (Q83) 024 AA2.05 Due to multiple systems being selected for both the RO and SRO exams with other Tier 1 Group 2 systems not selected on either exam, replaced KA to ensure all systems tested at least once on either the RO and SRO exams prior to any system being tested twice. Emergency Boration was already selected for the RO exam and Steam Generator Tube Leak was not selected for either, so we replaced Emergency Boration with Steam Generator Tube Rupture and kept the topic as an A2 K/A. Re-selected 037 AA2.10.
2 / 1 (Q88) 013 G 2.2.40 Due to multiple systems being selected for both the RO and SRO exams with other Tier 1 Group 1 systems not selected on either exam, replaced KA to ensure all systems tested at least twice on either the RO and SRO exams prior to any system being tested three times. ESFAS was selected three times between the RO and SRO exams and Pressurizer Relief / Quench Tank was only selected once, so we replaced ESFAS with Pressurizer Relief / Quench Tank but kept the generic KA. Reselected 007 G 2.2.40.
Administrative Topics Outline PVNGS 2019 NRC RO Admin JPM Outline ES-301-1 Rev 2 Facility:
PVNGS Date of Examination:
10/4/19 Examination Level RO Operating Test Number:
2019 NRC Administrative Topic (see Note)
Type Code*
Describe Activity to be Performed (A1)
N, R JPM: Determine medication reporting and No Solo operator applicability KA:
2.1.4 IR:
3.3 (A2)
D, R JPM:
Manual RCS Water Inventory Balance Calculation KA:
2.2.12 IR:
3.7 (A3)
N, R JPM: Determine which Air Removal Pump suctions are potentially flooded KA:
2.3.14 IR:
3.4 (A4)
M, R JPM: Determine minimum required HPSI makeup flow to prevent boiling while in Mid-Loop and maximum required EW temperature to lower SDC flow KA:
2.1.25 IR:
3.9 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (1)
(N)ew or (M)odified from bank ( 1) (3)
(P)revious 2 exams ( 1; randomly selected) (0)
Administrative Topics Outline Task Summary PVNGS 2019 NRC RO Admin JPM Outline ES-301-1 Rev 2 A1 The applicant will be directed to determine what changes in their medication requires PV Health Services notification and the time requirement to notify the NRC of a change in an operators medical condition. They will also determine if the prescribed medication requires a No Solo designation for the operator. This is a new JPM covering the Conduct of Operations K/A category.
A2 The applicant will be directed to perform a manual RCS water inventory balance calculation per 40ST-9RC05, Manual Calculation of RCS Water Inventory Balance. This is a bank JPM covering the Equipment Control K/A category.
A3 The applicant will be directed to determine which, if any, Air Removal Pump suction lines are potentially flooded following a SGTR per 40DP-9ZZ14, Containment Water Management. This is a new JPM covering the Radiation Control K/A category.
A4 The applicant will be determine the minimum required makeup flow to prevent boiling will during drain evolution to Mid-Loop conditions per 40OP-9ZZ16, RCS Drain Operations.
They will also determine the maximum indicated Essential Cooling Water temperature required before SDC flow can be lowered to 3780 gpm, per the Unit 2 Safety Analysis Operational Data Manual. This is a modified bank JPM covering the Conduct of Operations K/A category.
Administrative Topics Outline PVNGS 2019 NRC SRO Admin JPM Outline ES-301-1 Rev 1 Facility:
PVNGS Date of Examination:
10/4/19 Examination Level SRO Operating Test Number:
2019 NRC Administrative Topic (see Note)
Type Code*
Describe Activity to be Performed (A5)
N, R JPM:
Perform an online risk evaluation KA:
2.1.39 IR:
4.3 (A6)
M, R JPM:
Assess Fatigue Rule requirements KA:
2.1.5 IR:
3.9 (A7)
D, R JPM:
MSSV Technical Specification Evaluation KA:
2.2.40 IR:
4.7 (A8)
D, R JPM: Determine if release can continue and ODCM actions for loss of RU-143 and RU-144 KA:
2.3.11 IR:
4.3 (A9)
N, R JPM:
Complete PVNGS NAN Emergency Message Form KA:
2.4.38 IR:
4.4 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (2)
(N)ew or (M)odified from bank ( 1) (3)
(P)revious 2 exams ( 1; randomly selected) (0)
Administrative Topics Outline Task Summary PVNGS 2019 NRC SRO Admin JPM Outline ES-301-1 Rev 1 A5 The applicant will be directed to determine the nuclear risk classification and mandatory risk mitigating strategies for upcoming on-line troubleshooting per 02DP-0RS01, Online Integrated Risk. This is a new JPM covering the Conduct of Operations K/A category.
A6 The applicant will be directed to evaluate the fatigue status of four Reactor Operators and determine which Reactor Operator(s) are eligible to take the shift per 01DP-0AP17, Managing Personnel Fatigue. This is a modified bank JPM covering the Conduct of Operations K/A category.
A7 The applicant will be directed to evaluate a list of Main Steam Safety Valve lift settings provided from engineering and determine evaluate Technical Specifications requirements and, if applicable, document which LCO condition(s) are not met and the associated required actions and operational limitations. This is a bank JPM covering the Equipment Control K/A category.
A8 The applicant will be directed to determine if a release through the Plant Vent may continue following a loss of power and what actions are required to continue (or re-initiate) the release per the PVNGS Offsite Dose Calculation Manual. This is a bank JPM covering the Radiation Control K/A category.
A9 The applicant will be directed to complete the Palo Verde NAN Emergency Message Form, EP-0541, following an Alert classification. The applicant will be provided with plant conditions, meteorological data, and radiation monitor trends, and use these documents/conditions to determine and document the correct wind speed, wind direction, and release status. This is a new JPM covering the Emergency Procedures / Plan K/A category.
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 1 of 4 2019 PVNGS NRC Initial Exam ES-301-2 RO and SRO JPM Outline Rev 6 Facility:
PVNGS Date of Examination:
10/4/19 Exam Level:
Operating Test No.:
2019 NRC Control Room Systems (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Code*
Safety Function S1 029 - Commence filling the 1A RCP Bearing Oil Reservoir Level and trip the Reactor when RCP 1A trips A, N, S 1
S2 013 - Perform actions to establish adequate SI flow per 40EP-9EO03, Loss of Coolant Accident A, D, EN, L, S 2
S3 005 - Parallel SDC Pump Operation to Single SDC Pump Operation L, N, S 4P S4 076 - Respond to a loss of Turbine Cooling Water per 40AO-9ZZ03, Loss of Cooling Water, Appendix B, Minimize Cooling Load on TC D, L, S 4S S5 026 - Respond to an inadvertent Train A Containment Spray Actuation Signal steps 3 - 8 A, EN, N, S 5
S6 062 - Perform contingency actions for verification of Vital Auxiliaries in SPTAs A, D, L, S 6
S7 012 - RTCB Trip Test (RO Only)
N, S 7
S8 060 - Verify Control Room Ventilation is isolated per 40AO-9ZZ26, Toxic Gas EN, N, S 9
In-Plant Systems (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
P1 065 - Instrument Air Dryer Relief Valve Pop Test A, N 8
P2 064 - Place the A EDG in Standby following corrective maintenance D
6 P3 022 - Vent Charging Pumps and Charging Pump Header D, E, R 2
RO: Will perform all simulator and in-plant JPMs SRO (I): Will perform all simulator and in-plant JPMs with the exception of S3 SRO (U): Will perform S1 / S3 / S5 / P1 / P3 NOTE: S4 and S6, and S7 and S8 can be run in parallel
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 2 of 4 2019 PVNGS NRC Initial Exam ES-301-2 RO and SRO JPM Outline Rev 6 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 (5) / 4-6 (5) / 2-3 (3)
(C)ontrol room (D)irect from bank 9 (5) / 8 (5) / 4 (1)
(E)mergency or abnormal in-plant 1 (1) / 1 (1) / 1 (1)
(EN)gineered safety feature 1 (3) / 1 (3) / 1 (1) (control room system)
(L)ow Power / Shutdown 1 (4) / 1 (4) / 1 (1)
(N)ew or (M)odified from bank including 1(A) 2 (6 - 3A) / 2 (5 - 3A) / 1 (4 - 3A)
(P)revious 2 exams 3 (0) / 3 (0) / 2 (0) (randomly selected)
(R)CA 1 (1) / 1 (1) / 1 (1)
(S)imulator NRC JPM Examination Summary Description S1 The applicant will be directed to fill the 1A RCP Bearing Oil Reservoir per 40OP-9RC01, Reactor Coolant Pump Operation, due to a low Bearing Oil Level alarm. When level has been raised to clear the alarm, the 1A RCP will trip and the Reactor will fail to auto trip.
The applicant will have to recognize the ATWS condition and manually trip the Reactor within three minutes. This is a new alternate path JPM covering Safety Function 1.
S2 The applicant will be directed to perform steps 4-7 of 40EP-9EO03, Loss of Coolant Accident. SIAS will have actuated due to the LOCA, however the A HPSI Pump failed to start due to a Train A Sequencer failure and the B HPSI Pump will have degraded discharge pressure causing it to fail to inject. The applicant will manually start the A HPSI Pump and verify injection flow is meeting the requirements of Appendix 2, Figures.
The applicant should determine that two loops are not meeting the minimum flow requirements due to two HPSI Injection Valves failing to auto open and take action to manually open the injection valves. The contingency actions for this step will direct the applicant to ensure operation of ESF auxiliary equipment. The applicant will manually start the Train A Spray Pond Pump, Essential Chilled Water Pump, and the Essential Cooling Water Pump. This is a bank alternate path JPM covering Safety Function 2.
S3 The applicant will be directed to transition from Train B LPSI and Train B CS parallel SDC operation to only Train B LPSI SDC operation per 40OP-9SI01, Shutdown Cooling Initiation, Section 6.25, Removing the Train B Containment Spray Pump From Parallel Operation With the Train B Low Pressure Safety Injection Pump. The applicant will adjust SDC flows to meet the target flow range prior to stopping the CS Pump, then adjust flow after the CS Pump is stopped to reestablish the proper SDC flowrate. This is a new JPM coving Safety Function 4P.
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 3 of 4 2019 PVNGS NRC Initial Exam ES-301-2 RO and SRO JPM Outline Rev 6 S4 The applicant will be directed to perform 40AO-9ZZ03, Loss of Cooling Water, Appendix B, Minimize Heat Loads on TC, following a manual Reactor trip due to a complete loss of Turbine Cooling Water. The applicant will secure SG Blowdown, establish auxiliary feed flow using AFN-P01, stop both Main Feedwater Pumps, stop all three Condensate Pumps, and stop both Heater Drain Pumps. This is a bank JPM covering Safety Function 4S.
S5 The applicant will be directed to take action in response to an inadvertent Train A CSAS actuation per 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations. The applicant will stop the A CS Pump, attempt to close the A CS Header Isolation Valve (valve is seized so alternate valves will have to be closed), override NC CIV to restore cooling flow to the RCPs, and restore Instrument Air to Containment. This is a new alternate path JPM covering Safety Function 5.
S6 The applicant will be directed to perform the Vital Auxiliaries verification per 40EP-9EO01, Standard Post Trip Actions. The applicant will determine that the Train B Class 4kV bus is de-energized with the B EDG running and EDG output breaker open. The applicant will utilize the Standard Appendix 115 hard card to take contingency actions to re-energize the Train B Class 4kV bus. Upon restoration of the bus, the applicant will then have to recognize that the associated Spray Pond Pump has not started automatically and start the pump manually within 2.6 minutes (time critical action). This is a bank alternate path JPM covering Safety Function 6.
S7 The applicant will be directed to perform a Reactor Trip Circuit Breaker functional test per 40ST-9SB01, Reactor Trip Breaker Functional Test. The applicant will open the RTCB from the control room then reset the trip and reclose the breaker at the RPS Cabinet behind the control room. This is a new JPM covering Safety Function 7.
S8 The applicant will be directed to respond to a toxic gas condition in the Control Building per 40AO-9ZZ26, Toxic Gas. The applicant will manually initiate a Control Room Ventilation Isolation Actuation Signal and ensure all components are in their actuated condition. A K-Relay failure will prevent some of the dampers from auto closing and will take action to manually isolate the Control Room Ventilation System. This is a new JPM covering Safety Function 9.
P1 The applicant will be directed to perform Instrument Air Dryer Relief Valve Pop Test per 40OP-9IA01, Instrument Air. When simulating actuating the pop valve on the Left Chamber, the relief valve will open but not reseat. The applicant will take contingency actions to isolate IAN-M01C, Instrument Air Dryer C. This is a new alternate path JPM covering Safety Function 8.
P2 The applicant will be directed to place the A EDG in standby following corrective maintenance per 40ST-9DG01, Emergency Diesel Generator A, Section 6.1, Placing Train A Diesel Generator in Standby. This is a bank JPM covering Safety Function 6.
P3 The applicant will be directed to vent the Charging Pumps and Charging Header following the loss of all Charging due to gas binding per 40AO-9ZZ05, Loss of Charging or
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 4 of 4 2019 PVNGS NRC Initial Exam ES-301-2 RO and SRO JPM Outline Rev 6 Letdown, Appendix I, Venting Charging Pumps and Header to the Recycle Drain Header.
This is a bank JPM covering Safety Function 2.
Appendix D Scenario Outline Form ES-D-1 PVNGS 2019 NRC Scenario # 1 Rev 4 Facility:
Palo Verde Scenario: 1 Test:
2019 NRC Exam Examiners:
Operators:
Initial Conditions: 100% power, BOC, B EDG OOS Turnover: Maintain power at 100%
CT-1: Restore Feedwater to at least one SG prior to reaching the AFAS setpoint of 25.8% WR in either SG CT-2: Restore power to PBB-S04 and ensure adequate SI flow within 30 minutes of entry into the Functional Recovery Procedure Event Number Event Type*
Event Description 1
C (CRS, BOP) 7A FW Heater Switch Fails High 2
TS (CRS)
Containment Wide Range Pressure Transmitter, HCD-PI-352D, Fails High 3
I (CRS, OATC)
Charging DP Controller PDIC-240 Output Fails High in Auto 4
R (ALL)
ECC Directed Turbine Unloading (100 MW) 5 C (CRS, OATC)
TS (CRS)
Inadvertent Train B 2-4 Leg SIAS 6
M (ALL)
Small Break LOCA (~ 400 gpm) 7 C (OATC)
Train B 4kV Class Bus Supply Transformer NBN-X04 Fault 8
C (OATC)
A HPSI Pump Trip (Rx Trip + 5 min) 9 C (BOP)
AFN-P01 Suction Valve HV-1 Seized Closed / AFAS Fails to Auto Actuate
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Technical Specification Actual Target Quantitative Attributes 9
Total malfunctions (5-8) 3 Malfunctions after EOP entry (1-2) 5 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-2) 2 Critical tasks (2-3)
Appendix D Scenario Event Summary Form ES-D-1 NRC Exam Scenario # 1 PVNGS 2019 NRC Scenario # 1 Rev 4 2019 NRC Exam Scenario # 1 Overview Event 1 The 7A FW Heater Switch will fail high resulting in a decrease in feedwater heating and causing Reactor power to exceed 100%. The crew will address the ARP and take action to lower turbine load as needed to maintain Reactor power < 100%.
Event 2 Containment Wide Range Pressure Transmitter, HCD-PI-352D, will fail high. The crew will address the alarm response procedure to identify the failed transmitter and the CRS will address Technical Specifications. The CRS will direct the crew to bypass the associated bistable at the PPS cabinet within one hour to comply with Technical Specifications.
TS: LCO 3.3.5 Condition A (Function 2)
Event 3 When the associated bistable has been bypassed, Charging Header Backpressure Controller, PDIC-240, output fails high in automatic. The crew will address the ARP and take manual control of PDIC-240 and adjust controller output to restore proper Charging line D/P.
Event 4 The ECC will call and report grid disturbances are occurring and PVNGS is required to lower MW output by 100 MW in the next 15 minutes to aid in grid stabilization. The crew will lower the lift setting of the SBCS valves, initiate boron equalization (force pressurizer spray), and lower turbine load by 100 MW.
Event 5 When the crew has reduced turbine load, and inadvertent Train B Leg 2-4 SIAS will occur.
The CRS will enter 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations, and direct the crew to override and stop the B HPSI, LPSI and CS Pumps, and the CRS will address Technical Specifications for the overridden pumps and the inadvertent actuation.
TS: LCO 3.3.6 Condition D (Function 1c), LCO 3.5.3 Condition B (when HPSI Pump is stopped), LCO 3.5.3 Condition A (when LPSI Pump is stopped), and LCO 3.6.6 Condition A (when CS Pump is stopped)
Event 6 After the CRS has addressed Technical Specifications, a small break LOCA will develop.
The crew will start the standby Charging Pump, and trip the reactor when the leak rate exceeds Charging Pump capacity.
Event 7 On the trip, the Train B 4kV Supply Transformer, NBN-X04, will fault resulting in a loss of the Train B 4kV Class Bus.
Event 8 The Train A HPSI Pump will trip 5 minutes after the Reactor trip, resulting in a loss of all HPSI flow. This will force the crew to transition to the Functional Recovery procedure following SPTAs in order to restore power to the Train B Bus and reinitiate HPSI flow using the Train B HPSI Pump.
Event 9 AFAS will fail to auto actuate (if SG level lowers to the AFAS setpoint) and AFN-P01 Suction Valve, CTN-HV-1, will be seized closed, requiring the crew to manually start and align AFA-P01 to establish a feed path to the non-faulted SG.
Appendix D Critical Task Summary Form ES-D-1 NRC Exam Scenario # 1 PVNGS 2019 NRC Scenario # 1 Rev 4 Critical Task # 1: Restore Feedwater to at least one SG prior to reaching the AFAS setpoint of 25.8% WR in either SG Basis for CT bounding criteria: Since AFAS fails to auto actuate, the crew will need to take manual action to ensure feedwater is restored to maintain the RCS heat removal safety function. The AFAS setpoint is the bounding criteria for the CT since manual operator action is required to replace the automatic actuation which would otherwise restore feed flow.
Safety Significance: The crew will have to take manual action to restore feed to at least one SG to ensure adequate inventory in the SG to remove decay heat from the core.
Cueing: The crew will have indication of a complete loss of feed water due to MSIS isolating Main Feedwater, AFAS failing to auto actuate, and the loss of Train B 4kV power (loss of AFB-P01). There is also indication provided by all feed water flow indicators indicating 0 gpm to each SG.
Measurable Performance Indicator: The crew will restore feed to at least one SG using AFA-P01 by manually starting and aligning feedwater valves from the control room. These actions must be completed prior to either SG lowering to 25.8% WR (AFAS setpoint).
Performance Feedback: When the crew has started AFA-P01 and aligned a feed path to at least one SG, the crew will have indication of feed flow, a rising trend on SG level(s), and depending on feed flow rate, a lowering trend on RCS temperature.
Appendix D Critical Task Summary Form ES-D-1 NRC Exam Scenario # 1 PVNGS 2019 NRC Scenario # 1 Rev 4 Critical Task # 2: Restore power to PBB-S04 and ensure adequate SI flow within 30 minutes of entry into the Functional Recovery Procedure Basis for CT bounding criteria: At PVNGS, safety function status checks are required to be performed once every 15 minutes when operating in a post-SPTA EOP. One performance of SFSCs is allotted to identify failed safety functions, and the next performance is allotted to restore safety functions which were identified as not being met in the previous performance, for a total of 30 minutes from entry into the EOP to restore the failed safety function.
Safety Significance: This is based on a degraded core cooling system. Inadequate SI flow may result in loss of subcooled margin and/or core uncovery, increasing the risk of core damage.
Cueing: ERFDADS indication of insufficient SI flow, SI flow not meeting the minimum flow requirements of Standard Appendix 2, Figures, and indications of a loss of power to PBB-S04 Measurable Performance Indicator: The crew will restore power to PBB-S04 by placing synchronizing switch PBB-SS-S04L, 4.16 KV Bus S04 Alternate Supply, to ON, and closing breaker PBB-S04L, 4.16 KV Bus S04 Alternate Supply. When power is restored to PBB-S04, the crew will initiate SI flow by taking the B HPSI Pump handswitch to START. SI flow must be restored within 30 minutes of entry into the Functional Recovery procedure.
Performance Feedback: Power restoration can be verified by observing PBB-S04 voltage on B01 or ERFDADS, and adequate SI flow can be verified by comparing current RCS pressure to the SI flowrate using Standard Appendix 2, Figures (located on B02) or any of the ERFDADS teminals.
NOTE: (Per NUREG-1021 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 Critical Task identified in the post-scenario review
Appendix D Driver Set-Up Instructions Form ES-D-1 NRC Exam Scenario # 1 PVNGS 2019 NRC Scenario # 1 Rev 4 Driver Setup Instructions Reset to IC-19 Run scenario file NRC Scenario # 1 Hang OOS tags on the B EDG
Appendix D Scenario File Description Form ES-D-1 NRC Exam Scenario # 1 PVNGS 2019 NRC Scenario # 1 Rev 4 Event Type Malf #
Description Final Initiator rfEG29 B EDG OOS OFF rfEG38 CLOSE rfEG39 CLOSE rfEG30 SHUTOFF crB2EG03PBBS04B_2 Rackout doEG_ZLDGBHS2_G1 OFF RF rfIP08 RWT Level 88%
88 MF cmBSFW10EDNLSHH711_2 7A FW Heater Level Switch Fails High Key 1 2
MF cmTRCH05HCDPT352D_1 C WR Containment Pressure Transmitter Fails High 85 Key 2 3
MF cmCNCV01CHEPDIC240_2 PDIC-240 Output Fails High in Auto 100 Key 3 4
ECC Directed Turbine Unloading 5
MF mfRP06D2 Inadvertent B Leg 2-4 SIAS Key 5 6
MF mfTH01C Small Break LOCA (10 min ramp) 3 Key 6 7
MF mfED10B Train B ESF Transformer Trip Rx Trip 8
MF mfSI01A A HPSI Pump Trip Rx Trip
+ 5 min 9
CM cmMVMC04CTAHV1_6 AFN Suction CTN-HV-1 Seized Closed 9
cmBSRP01BSSG1LVLLAT_1 AFAS Fails to Auto Actuate cmBSRP01BSSG1LVLLBT_1 cmBSRP01BSSG1LVLLCT_1 cmBSRP01BSSG1LVLLDT_1 cmBSRP01BSSG2LVLLAT_1 cmBSRP01BSSG2LVLLBT_1 cmBSRP01BSSG2LVLLCT_1 cmBSRP01BSSG2LVLLDT_1
Appendix D NRC Exam Scenario # 1 Plant Conditions Form ES-D-1 PVNGS 2019 NRC Scenario # 1 Rev 4 Plant Conditions:
Unit 1 is operating at 100% power, BOC Unit 2 is operating at 100% power Unit 3 is in a refueling outage Equipment Out of Service:
The B EDG was taken out of service last shift for preventive maintenance o LCO 3.8.1, Condition B has been entered o SR 3.8.1.1 was last completed one hour ago and is not due again for seven hours Planned Shift Activities:
Maintain Reactor power stable at 100%
Appendix D Scenario Outline Form ES-D-1 PVNGS 2019 NRC Scenario # 2 Rev 4 Facility:
Palo Verde Scenario: 2 Test:
2019 NRC Exam Examiners:
Operators:
Initial Conditions: 50% power, MOC, B EDG OOS Turnover: Maintain power at 50%
CT-1: Restore power to PBA-S03 within 14 minutes and 35 seconds of the loss of offsite power CT-2: Restore power to Train B Class 4kV Bus PBB-S04 by cross-connecting the A EDG to PBB-S04, and restore feed to at least one SG within 30 minutes of entry into the Function Recovery Procedure Event Number Event Type*
Event Description 1
C (OATC, CRS)
Running RMW Pump Trips / Standby Fails to Auto Start 2
TS (CRS)
SG Level Transmitter LT-1114A Fails High 3
Degraded Condenser Vacuum 4
C(ALL), TS (CRS)
CEA 43 Slips to 30 Withdrawn 5
C Second CEA (CEA 4) Drops - Automatic Reactor Trip 6
M (ALL)
Loss of Offsite Power (Rx trip + 1:30) 7 C (OATC)
A EDG Output Breaker Fails to Auto Close 8
C (BOP)
AFA-P01 Degraded Discharge Head / AFN-P01 Trip (Loss of lube oil)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Technical Specification Actual Target Quantitative Attributes 9
Total malfunctions (5-8) 3 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-2) 2 Critical tasks (2-3)
Appendix D Scenario Event Summary Form ES-D-1 NRC Exam Scenario # 2 PVNGS 2019 NRC Scenario # 2 Rev 4 2019 NRC Exam Scenario # 2 Overview Event 1 The running RMW Pump will trip and the standby pump will fail to auto start. The crew will address the ARP and manually start the standby RMW Pump.
Event 2 SG #1 Level Transmitter LT-1114A will fail high. The crew will address the ARP and the CRS will address Technical Specifications to determine the affected bistable(s) and direct the BOP to bypass the affected bistable at the PPS cabinet.
TS: LCO 3.3.1 Condition A (Function 10 - SG #1 Level High), LCO 3.3.5 Condition A (Function 4c - SG #1 Level High)
Event 3 Main Condenser vacuum will begin to degrade. The CRS will enter 40AO-9ZZ07, Loss of Condenser Vacuum, and direct the crew to commence a turbine load reduction to stabilize vacuum. When turbine load has been lowered ~ 5%, an AO will report an empty loop seal and commence filling the loop seal to restore vacuum.
Event 4 When the cause of the loss of vacuum has been corrected, CEA 43 will slip to 30 withdrawn. The CRS will enter 40AO-9ZZ11, CEA Malfunctions, and direct the crew to reduce power to comply with the Core Operating Limits Report.
TS: LCO 3.1.5 Condition A Event 5 When the crew has commenced a power reduction, another CEA in the same quadrant (CEA 4) will drop and the Reactor will automatically trip.
Event 6 One minute after the Reactor is tripped, a loss of offsite power will occur, resulting in a temporary blackout condition.
Event 7 The A EDG Output Breaker will fail to auto close, placing the unit in a blackout condition.
The crew will take action to manually close the A EDG output breaker and restore power to the Train A Class 4kV Bus.
Event 8 Turbine-driven AFW Pump AFA-P01 will have degraded discharge head when started, making it unable to feed the SGs.
Event 9 AFN-P01 suction valve, CTA-HV-4 will fail to open due to a valve seizure, making the pump unable to be used. The CRS will have to transition to the Functional Recovery procedure to align the Train A EDG to the Train B bus and restore feed using AFB-P01.
Appendix D Critical Task Summary Form ES-D-1 NRC Exam Scenario # 2 PVNGS 2019 NRC Scenario # 2 Rev 4 Critical Task # 1: Restore power to PBA-S03 within 14 minutes and 35 seconds of the loss of offsite power Basis for CT bounding criteria: The A EDG will automatically start when power is lost to PBA-S03.
The EDG will be running unloaded with no cooling water flow from the A Spray Pond Pump due to no power being available on PBA-S03. The EDG can run for 15 minutes unloaded before damage may occur. When the EDG output breaker is closed, the A Spray Pond Pump will auto start 25 seconds later, restoring cooling flow to the EDG. Therefore, the A EDG must be loaded onto the PBA-S03 bus within 14 minutes and 35 seconds to prevent damaging to the only available Unit 1 EDG (the B EDG is OOS in this scenario).
Safety Significance: Failure to manually close the A EDG output breaker within the allotted time would run the risk of damage to the only available EDG (EDG can only for 15 minutes unloaded without cooling water before damage may occur and the Spray Pond Pump is sequenced on 25 seconds after the bus is re-energized). Additionally, failure to restore power to at least one class 4kV bus within 15 minutes would result in a Site Area Emergency EAL declaration, which would result in assembly and accountability as well as the evacuation of all non-essential personnel.
Cueing: The crew will have indication of the loss of all AC power as indicated by 0V readings on all AC instrumentation as well as alarms on B01 and the reduction of normal lighting in the control room.
Measurable Performance Indicator: The crew will determine the bus is not faulted and manually close the A EDG output breaker on B01 by placing the synchronizing switch to ON and closing the A EDG output breaker. This must be completed within 14 minutes and 35 seconds of the loss of offsite power.
Performance Feedback: The crew will have indication of re-energizing PBA-S03 by half of the normal control room lights coming back on and indication of normal voltage being restored to Train A electrical buses.
Appendix D Critical Task Summary Form ES-D-1 NRC Exam Scenario # 2 PVNGS 2019 NRC Scenario # 2 Rev 4 Critical Task # 2: Restore power to Train B Class 4kV Bus PBB-S04 by cross-connecting the A EDG to PBB-S04, and restore feed to at least one SG within 30 minutes of entry into the Function Recovery Procedure Basis for CT bounding criteria: At PVNGS, safety function status checks are required to be performed once every 15 minutes when operating in a post-SPTA EOP. One performance of SFSCs is allotted to identify failed safety functions, and the next performance is allotted to restore safety functions which were identified as not being met in the previous performance, for a total of 30 minutes from entry into the EOP to restore the failed safety function.
Safety Significance: The crew will have to restore feed water to at least one SG to ensure adequate inventory in the SG(s) to remove decay heat from the core.
Cueing: The crew will have indication of a complete loss of feed water due to the loss of offsite power tripping both Main Feedwater Pumps, the loss of power to PBB-S04 (loss of AFB-P01), the seized suction valve on AFN-P01, and the degraded discharge head on AFA-P01. There will also be indication provided by all feed water flow indicators indicating 0 gpm to each SG.
Measurable Performance Indicator: The crew will have to close the normal feeder breaker for PBA-S03 and the alternate feeder breaker for PBB-S04 to align the A EDG to PBB-S04, start AFB-P01, and open feed block valves and flow control valves to commence feeding at least one SG. These actions must be completed prior to the CRS exiting MVAC-2.
Performance Feedback: When the crew has restored power to PBB-S04, started AFB-P01, and aligned a feed path to at least one SG, the crew will have indication of feed flow to at least one SG as well as a rising trend on SG level(s), and depending on feed flow rate, a lowering trend on RCS temperature.
NOTE: (Per NUREG-1021 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 Critical Task identified in the post-scenario review
Appendix D Driver Set-Up Instructions Form ES-D-1 NRC Exam Scenario # 2 PVNGS 2019 NRC Scenario # 2 Rev 4 Driver Setup Instructions Reset to IC-16 (50% MOC)
Run scenario file NRC Scenario # 2 Hang OOS tags on the B EDG Ensure CEDMCS in Auto Sequential Ensure ALL Cooling Tower Fans are running Ensure DFWCS alarms are cleared on the DFWCS screen and 6A06A alarm is reset
Appendix D Scenario File Description Form ES-D-1 NRC Exam Scenario # 2 PVNGS 2019 NRC Scenario # 2 Rev 4 Event Type Malf #
Description Final Initiator rfEG29 B EDG OOS OFF rfEG38 CLOSE rfEG39 CLOSE rfEG30 SHUTOFF crB2EG03PBBS04B_2 RACKOUT doEG_ZLDGBHS2_G1 OFF RF rfIP08 RWT Level 88%
88 1
MF cmCPCV10CHNP03A_2 A RMW Pump Trip Key 1 MF cmCPCV10CHNP03B_5 B RMW Pump FTAS 2
MF cmTRRX12SGALT1114A_1 LT-1114 Fails High 100 Key 2 3
MF mfMC01C Degraded Condenser Vacuum C Shell 6
Key 3 4
MF mfRD02G CEA 43 Slips to 30 Withdrawn 80 Key 4 5
MF mfRD02H CEA 4 Drops 100 Key 5 6
MF mfED02 Loss of Offsite Power Rx Trip
+1:30 7
MF cmBKEG02PBAS03B_2 A EDG Output Breaker Fails to Auto Close 8
MF cmCPFW07AFAP01_3 AFA-P01 Degraded Discharge Head 90 MF mfFW21A AFN-P01 Trip
Appendix D Scenario File Description Form ES-D-1 NRC Exam Scenario # 2 PVNGS 2019 NRC Scenario # 2 Rev 4 Plant Conditions:
Unit 1 is operating at 50% power, MOC o The B MFP is in service o The A MFP is in hot standby Unit 2 is operating at 100% power Unit 3 is in a refueling outage Equipment Out of Service:
The B EDG was taken out of service last shift for preventive maintenance o LCO 3.8.1, Condition B has been entered o SR 3.8.1.1 was last completed one hour ago and is not due again for seven hours Planned Shift Activities:
o Maintain 50% power
Appendix D Scenario Outline Form ES-D-1 PVNGS 2019 NRC Scenario # 3 Rev 4 Facility:
Palo Verde Scenario: 3 Test:
2019 NRC Exam Examiners:
Operators:
Initial Conditions: 75% power, MOC, B EDG OOS Turnover: Maintain 75% power CT-1: Place both Hydrogen Analyzers in service within 30 minutes of the LOCA CT-2: Ensure both trains of HPSI and CS have their pump suctions aligned to the Containment Sump by opening SIA-UV-673 and SIB-UV-676 and closing CHA-HV-531 and CHB-HV-530 within 5 minutes of the RAS actuation Event Number Event Type*
Event Description 1
TS (CRS)
SG DP Low Instrument, RCC-PDT-115C, Fails Low 2
I (OATC, CRS)
Channel X Pressurizer Pressure Transmitter PT-100X Fails High 3
C (ALL)
Loss of Non-Class Instrument Bus NNN-D15 4
C (ALL)
TS (CRS)
A NC Pump Trip, B NC Pump Shaft Shear - Cross Tie NC-EW / Extended Loss of Letdown 5
M (ALL)
Large Break LOCA 6
C (OATC, CRS)
Containment Suction Valves Fail to Auto Open on RAS
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Technical Specification Actual Target Quantitative Attributes 7
Total malfunctions (5-8) 2 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 2 Critical tasks (2-3)
Appendix D Scenario Event Summary Form ES-D-1 NRC Exam Scenario # 3 PVNGS 2019 NRC Scenario # 3 Rev 4 2019 NRC Exam Scenario # 3 Overview Event 1 SG DP Low Transmitter, RCC-PDT-115C, will fail low. The crew will address the ARP and the CRS will address Technical Specifications. The crew will bypass the affected bistable at the PPS cabinet.
TS: LCO 3.3.1 Condition A (Function 12 - RCS Flow, SG #1 Low)
Event 2 After the SG DP bistable has been bypassed at the PPS cabinet, the controlling Pressurizer Pressure transmitter, PT-100X, will fail high resulting in both Main Spray Valves opening and Pressurizer pressure lowering. The crew will address the ARP and select the unaffected pressurizer pressure channel and restore Pressurizer pressure and Pressurizer heaters.
Event 3 After Pressurizer pressure has been restored, a loss of Non-Class Instrument Bus NNN-D15 will occur. This will cause a loss of feed flow from the A MFP, however the pump will not automatically trip on the loss of the bus. The CRS will enter 40AO-9ZZ15, Loss of Non-Class Instrument Bus or Control Power, and direct manually tripping the A MFP to cause a Reactor Power Cutback. When the cutback actuates, the CRS will also enter 40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump), and direct the crew to take action to stabilize the unit.
Event 4 When the unit has been stabilized, the A NC Pump will trip and the B NC Pump shaft will shear, resulting in a complete loss of NC. The CRS will enter 40AO-9ZZ03, Loss of Cooling Water, and direct the crew to cross-tie NC to EW to supply cooling water to the RCPs. Additionally, letdown will be lost on the loss of NC and the CRS will enter 40AO-9ZZ05, Loss of Charging or Letdown, and direct the crew to establish conditions for extended operations with letdown isolated.
TS: LCO 3.7.7 Conditions A, LCO 3.7.10 Condition A Event 5 When the crew has restored cooling water to the RCPs, a large break LOCA will occur.
The crew will perform SPTAs, transition to the LOCA EOP, place Hydrogen Analyzers in service, and monitor for RAS.
Event 6 On the RAS actuation, the containment suction valves for the HPSI and CS Pumps will fail to auto open. The crew will take action to manually re-align the pump suctions to containment to prevent damage to the HPSI and CS Pumps.
Appendix D Critical Task Summary Form ES-D-1 NRC Exam Scenario # 3 PVNGS 2019 NRC Scenario # 3 Rev 4 Critical Task # 1: Place both Hydrogen Analyzers in service within 30 minutes of the LOCA Basis for CT bounding criteria: Placing all available Hydrogen Analyzers in service within 30 minutes of the start of a LOCA is listed in the PVNGS Time Critical Action Program (TCA-55) and is based on the PVNGS UFSAR section 6.2.5.2.1.
Safety Significance: Per the PVNGS UFSAR, Hydrogen Analyzers must be placed in service within 30 minutes of a LOCA. The crew must be aware of hydrogen concentration inside containment to ensure the Containment Temperature and Pressure Control safety function is met, to determine when hydrogen recombiners or hydrogen purge must be placed in service, and to monitor potential EAL escalation criteria based on containment hydrogen levels.
Cueing: The crew will have procedural direction to place Hydrogen Analyzers in service per 40EP-9EO03, LOCA.
Measurable Performance Indicator: The crew will open the inside and outside containment isolation valve for the Hydrogen Analyzers and place the Power/Control handswitch for each analyzer to the ANALYZE position. The H2 analyzers must be in service within 30 minutes of the LOCA.
Performance Feedback: The crew will have indication of the CIVs being open as indicated by a red light on each valve and the red ANALYZE light being illuminated on each Hydrogen Analyzer.
Appendix D Critical Task Summary Form ES-D-1 NRC Exam Scenario # 3 PVNGS 2019 NRC Scenario # 3 Rev 4 Critical Task # 2: Ensure both trains of HPSI and CS have their pump suctions aligned to the Containment Sump by opening SIA-UV-673 and SIB-UV-676 and closing CHA-HV-531 and CHB-HV-530 within 5 minutes of the RAS actuation Basis for CT bounding criteria: Closing the RWT outlet valves, CHA-HV-531 and CHB-HV-530, within 5 minutes of a RAS actuation is listed in the PVNGS Time Critical Action Program (TCA-52) and is based on the PVNGS UFSAR, table 6.3.2-3 item 21. Before the RWT outlet valves can be closed, the HPSI and CS Pump suctions must be realigned to the containment sump to ensure a suction source for the pumps, which is why SIA-UV-673 and SIB-UV-676 must be opened as well (they fail to auto open in the scenario).
Safety Significance: Failure of the crew to realign the HPSI and CS Pumps to the containment sump could result in air binding of the pumps, resulting in a potential loss of core head removal.
Cueing: The crew will have procedural direction to ensure HPSI and CS pump suctions have shifted to the containment sump as well as board indications (green light on the valves and SESS alarms indicating a failure of the valves to reposition as required) that the Train A inside containment sump suction valve, SIA-UV-673 and the Train B outside containment sump suction valve, SIB-UV-676, failed to auto open.
Measurable Performance Indicator: The crew will manually open SIA-UV-673 and SIB-UV-676 and close CHA-HV-531 and CHB-HV-530 to align the suction of the HPSI and CS pumps to the containment sump within 5 minutes of the RAS actuation.
Performance Feedback: The crew will have indication that the HPSI and CS pump suctions have been aligned to the containment sump based on the VPIs on the control board, the SESS alarms extinguishing, and steady amps and flow on the HPSI and CS pumps.
NOTE: (Per NUREG-1021 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 Critical Task identified in the post-scenario review
Appendix D Driver Set-Up Instructions Form ES-D-1 NRC Exam Scenario # 3 PVNGS 2019 NRC Scenario # 3 Rev 4 Driver Setup Instructions Reset to IC-18 Run scenario file NRC Scenario # 3 Hang OOS tags on the B EDG
Appendix D Scenario File Description Form ES-D-1 NRC Exam Scenario # 3 PVNGS 2019 NRC Scenario # 3 Rev 4 Event Type Malf #
Description Final Initiator Setup RF rfEG29 B EDG OOS OFF N/A RF rfEG38 CLOSE RF rfEG39 CLOSE RF rfEG30 SHUTOFF RF crB2EG03PBBS04B_2 RACKOUT OR doEG_ZLDGBHS2_G1 OFF RF rfIP08 RWT Level 88%
88 1
MF cmTRRX09RCCPDT115C_1 PDT-115 Fails Low 0
Key 1 2
MF cmTRRC03RCNPT100X_1 PT-100X Fails High 2500 Key 2 3
MF mfED13C Loss of NNN-D15 Key 3 4
MF cmCPCC01NCNP01A_6 A NC Pump Trip Key 4 MF cmCPCC01NCNP01B_1 B NC Pump Shaft Shear N/A Key 4 RF rfCC34 EWA-HCV-53 Throttle 90 Key 14 5
MF mfTH03A LB LOCA 100 Key 5 6
MF cmMVRH01SIAUV673_4 SIA-UV-673 Fails to Auto Open on RAS MF cmMVRH04SIBUV676_4 SIB-UV-676 Fails to Auto Open on RAS
Appendix D Scenario File Description Form ES-D-1 NRC Exam Scenario # 3 PVNGS 2019 NRC Scenario # 3 Rev 4 Plant Conditions:
Unit 1 is operating at 75% power, MOC Unit 2 is operating at 100% power Unit 3 is in a refueling outage Equipment Out of Service:
The B EDG was taken out of service last shift for preventive maintenance o LCO 3.8.1, Condition B has been entered o SR 3.8.1.1 was last completed one hour ago and is not due again for seven hours Planned Shift Activities:
Maintain power at 75%