ML16293A976
ML16293A976 | |
Person / Time | |
---|---|
Site: | Comanche Peak |
Issue date: | 07/11/2016 |
From: | Vincent Gaddy Operations Branch IV |
To: | |
References | |
Download: ML16293A976 (46) | |
Text
ES-401 PWR Examination Outline Form ES-401-2 Facility: CPNPP Date of Exam: July 18, 2016 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total
- 1. 1 2 3 4 3 3 3 18 Emergency &
Abnormal 2 2 1 2 N/A 1 2 N/A 1 9 Plant Evolutions Tier Totals 4 4 6 4 5 4 27 1 3 3 3 2 2 2 2 3 3 3 2 28 2.
Plant 2 1 0 1 1 1 1 1 1 1 1 1 10 Systems Tier Totals 4 3 4 3 3 3 3 4 4 4 3 38
- 3. Generic Knowledge and Abilities 1 2 3 4 10 Categories 3 2 2 3 Notes:
- 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.
Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G*
on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As Rev 5
ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 000007 Reactor Trip - Stabilization - X EA1.10 Ability to operate and monitor the 3.7 39 Recovery / 1 following as they apply to a reactor trip: S/G (1) pressure 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 X 2.4.6 Knowledge of EOP mitigation strategies 3.7 40 (2) 000011 Large Break LOCA / 3 X EK2.02 Knowledge of the interrelations between 2.6* 41 the Large Break LOCA and the following: Pumps (3) 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 X AK3.03 Knowledge of the reasons for the 3.1 42 following responses as they apply to the Loss of (4)
Reactor Coolant Makeup: Performance of lineup to establish excess letdown after determining need 000025 Loss of RHR System / 4 000026 Loss of Component Cooling X AA2.03 Ability to determine and interpret the 2.6 43 Water / 8 following as they apply to the Loss of Component (5)
Cooling Water: The valve lineups necessary to restart the CCWS while bypassing the portions of the system causing the abnormal condition 000027 Pressurizer Pressure Control X AK1.02 Knowledge of the operational implications 2.8 44 System Malfunction / 3 of the following concepts as they apply to (6)
Pressurizer Pressure Control Malfunctions:
Expansion of liquids as temperature increases.
000029 ATWS / 1 X EK2.06 Knowledge of interrelations between the 2.9* 45 following and an ATWS: Breakers, relays and (7) disconnects.
000038 Steam Gen. Tube Rupture / 3 X 2.4.4 Ability to recognize abnormal indications for 4.5 46 system operating parameters that are entry-level (8) conditions for emergency and abnormal operating procedures.
000040 (W/E12) Steam Line Rupture - X AK3.04 Knowledge of the reasons for the 4.5 54 Excessive Heat Transfer / 4 following responses as they apply to the Steam (16)
Line Rupture: Actions contained in EOPs for steam line rupture 000054 Loss of Main Feedwater / 4 X AK3.01 Knowledge of the reason for the following 4.1 47 responses as they apply to the Loss of Main (9)
Feedwater (MFW): Reactor and/or turbine trip, manual or automatic 000055 Station Blackout / 6 X EA2.04 Ability to determine or interpret the 3.7 48 following as they apply to a Station Blackout: (10)
Instruments and controls operable with only dc battery power available.
000056 Loss of Off-site Power / 6 X AK1.03 Knowledge of the operational implications 3.1* 49 of the following concepts as they apply to Loss of (11)
Offsite Power: Definition of subcooling: use of the steam tables to determine it.
Rev 5
ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 000057 Loss of Vital AC Inst. Bus / 6 X AA1.03 Ability to operate and/or monitor the 3.6* 50 following as they apply to the Loss of Vital AC (12)
Instrument Bus: Feedwater pump speed to control pressure and level in S/G.
000058 Loss of DC Power / 6 X 2.1.20 Ability to interpret and execute procedure 4.6 51 steps. (13) 000062 Loss of Nuclear Svc Water / 4 X AK3.04 Knowledge of the reasons for the 3.5 52 following responses as they apply to the Loss of (14)
Nuclear Service Water: Effect on the nuclear service water discharge flow header on a loss of CCW 000065 Loss of Instrument Air / 8 X AA1.04 Ability to operate and/or monitor the 3.5 53 following as they apply to the Loss of Instrument (15)
Air: Emergency air compressor.
W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 W/E05 Inadequate Heat Transfer - Loss X EK2.2 Knowledge of the interrelations between 3.9 55 of Secondary Heat Sink / 4 the Loss of Secondary Heat Sink and the (17) following: Facility*s heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
000077 Generator Voltage and Electric X AA2.04 Ability to determine and interpret the 3.6 56 Grid Disturbances / 6 following as they apply to Generator Voltage and (18)
Electric Grid Disturbances: VARs outside the capability curve.
K/A Category Totals: 2 3 4 3 3 3 Group Point Total: 18 Rev 5
ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 X AK1.21 Knowledge of the operational 2.9 57 implications of the following concepts as (19) they apply to Continuous Rod Withdrawal:
Integral rod worth.
000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 X EK3.05 Knowledge of the reasons for the 3.7 58 following responses as they apply to the (20)
Pressurizer level Control malfunctions:
Actions contained in EOP for PZR level malfunctions.
000032 Loss of Source Range NI / 7 X 2.4.20 Knowledge of the operational 3.8 59 implications of EOP warnings, cautions, (21) and notes.
000033 Loss of Intermediate Range NI / 7 X AA1.01 Ability to operate and / or monitor 2.9 60 the following as they apply to the Loss of (22)
Intermediate Range Nuclear Instrumentation: Power-available indicators in cabinets or equipment drawers 000036 Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 X AA2.11 Ability to determine and interpret 3.8 61 the following as they apply to the Steam (23)
Generator Tube Leak: When to isolate one or more S/Gs 000051 Loss of Condenser Vacuum / 4 X AK3.01 Knowledge of the reasons for the 2.8* 62 following responses as they apply to the (24)
Loss of Condenser Vacuum: Loss of steam dump capacity upon loss of condenser vacuum 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 Control Room Evac. / 8 000069 Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 X AA2.02 Ability to determine and interpret 2.8 64 the following as they apply to the High (26)
Reactor Coolant Activity: Corrective actions required for high fission product activity in RCS W/E01 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 X EK2.1 Knowledge of the interrelations 2.8 65 between the (Containment Flooding) and (27) the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Rev 5
ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 W/E16 High Containment Radiation / 9 W/E09 & 10 Natural Circulation with Steam Void in X EK1.2 Knowledge of the operational 3.4 63 Vessel with/without RVLIS/4 implications of the following concepts as (25) they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS):
Normal, abnormal and emergency operating procedures associated with (Natural Circulation with Steam Void in Vessel with/without RVLIS).
K/A Category Point Totals: 2 1 2 1 2 1 Group Point Total: 9 Rev 5
ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 1 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump X A1.03 Ability to predict and/or monitor 2.6 1 changes in parameters (to prevent (28) exceeding design limits) associated with operating the RCPS controls including: RCP motor stator winding temperatures 004 Chemical and Volume X 2.4.45 Ability to prioritize and interpret 4.1 2 Control the significance of each annunciator (29) or alarm.
005 Residual Heat Removal X K1.06 Knowledge of the physical 3.5 3 connections and/or cause effect (30) relationships between the RHRS and the following systems: ECCS 006 Emergency Core Cooling X K5.11 Knowledge of the operational 2.5 4 implications of the following concepts (31) as they apply to ECCS: Basic heat transfer equation 007 Pressurizer Relief/Quench X A2.06 Ability to (a) predict the impacts 2.6 5 Tank of the following malfunctions or (32) operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Bubble formation in PZR 007 Pressurizer Relief/Quench X A4.04 Ability to manually operate 2.6* 6 Tank and/or monitor in the control room: (33)
PZR vent valve 008 Component Cooling Water X K4.09 Knowledge of CCWS design 2.7 7 feature(s) and/or interlock(s) which (34) provide for the following: The "standby" feature for the CCW pumps 010 Pressurizer Pressure Control X K1.05 Knowledge of the physical 3.4 8 connections and/or cause-effect (36) relationships between the PZR PCS and the following systems: PRTS 010 Pressurizer Pressure Control X K2.03 Knowledge of bus power 2.8* 9 supplies to the following: Indicator for (35)
PORV position 012 Reactor Protection X A1.01 Ability to predict and/or monitor 2.9 10 Changes in parameters (to prevent (38) exceeding design limits) associated with operating the RPS controls including: Trip setpoint adjustment 012 Reactor Protection X A3.07 Ability to monitor automatic 4.0 11 operation of the RPS, including: Trip (37) breakers 013 Engineered Safety Features X K1.18 Knowledge of the physical 3.7 12 Actuation connections and/or cause effect (39) relationships between the ESFAS and the following systems: Premature reset of ESF actuation Rev 5
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 1 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 022 Containment Cooling X K4.04 Knowledge of CCS design 2.8 14 feature(s) and/or interlock(s) which (40) provide for the following: Cooling of control rod drive motors 026 Containment Spray X K2.02 Knowledge of bus power 2.7* 15 supplies to the following: MOVs (42) 039 Main and Reheat Steam X K5.08 Knowledge of the operational 3.6 16 implications of the following concepts (43) as the apply to the MRSS: Effect of steam removal on reactivity 059 Main Feedwater X A3.02 Ability to monitor automatic 2.9 17 operation of the MFW, including: (44)
Programmed levels of the S/G 061 Auxiliary/Emergency X K6.02 Knowledge of the effect of a 2.6 24 Feedwater loss or malfunction of the following will (52) have on the AFW components:
Pumps 061 Auxiliary/Emergency X A2.06 Ability to (a) predict the impacts 2.7 18 Feedwater of the following malfunctions or (45) operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Back leakage of MFW 062 AC Electrical Distribution X A4.03 Ability to manually operate 2.8 19 and/or monitor in the control room: (46)
Synchroscope, including an understanding of running and incoming voltages 063 DC Electrical Distribution X 2.2.44 Ability to interpret control room 4.2 20 indications to verify the status and (47) operation of a system, and understand how operator actions and directives affect plant and system conditions.
064 Emergency Diesel Generator X K3.01 Knowledge of the effect that a 3.8* 21 loss or malfunction of the ED/G (48)
Systems controlled by automatic loader 064 Emergency Diesel Generator X K6.07 Knowledge of the effect of a 2.7 22 loss or malfunction of the following will (49) have on the ED/G system: Air receivers 073 Process Radiation Monitoring X A2.02 Ability to (a) predict the impacts 2.7 23 of the following malfunctions or (50) operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure 073 Process Radiation Monitoring X K3.01 Knowledge of the effect that a 3.6 13 loss or malfunction of the PRM (41) system will have on the following:
Radioactive effluent release Rev 5
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 1 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 076 Service Water X A4.02 Ability to manually operate 2.6 25 and/or monitor in the control room: (51)
SWS valves 078 Instrument Air X K2.01 Knowledge of bus power 2.7 26 supplies to the following: Instrument (53) air compressor 078 Instrument Air X A3.01 Ability to monitor automatic 3.1 27 operation of the IAS, including: Air (54) pressure 103 Containment X K3.01 Knowledge of the effect that a 3.3* 28 loss or malfunction of the containment (55) system will have on the following:
Loss of containment integrity under shutdown conditions K/A Category Point Totals: 3 3 3 2 2 2 2 3 3 3 2 Group Point Total: 28 Rev 5
ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 2 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive X A3.07 Ability to monitor automatic 4.1 38 operation of the CRDS, including: (65)
Boration/dilution.
002 Reactor Coolant X K5.08 Knowledge of the operational 3.4 31 implications of the following concepts as (58) they apply to the RCS: Why PRZ level should be kept within the programmed level.
011 Pressurizer Level Control X 2.1.23 Ability to perform specific system 4.3 29 and integrated plant procedures during all (56) modes of plant operation.
014 Rod Position Indication X K4.06 Knowledge of RPIS design 3.4 37 feature(s) and/or interlock(s) which (64) provide for the following: Individual and group misalignment 015 Nuclear Instrumentation X A1.02 Ability to predict and/or monitor 3.5 30 changes in parameters to prevent (57) exceeding design limits) associated with operating the NIS controls including: SUR 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and X A4.01 Ability to manually operate and/or 4.0* 32 Purge Control monitor in the control room: HRPS (59) controls 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator X K6.01 Knowledge of the effect of a loss or 3.2 33 malfunction on the following will have on (60) the S/GS: MSIVs 041 Steam Dump/Turbine Bypass X K3.02 Knowledge of the effect that a loss 3.8 34 Control or malfunction of the SDS will have on (61) the following: RCS 045 Main Turbine Generator X K1.20 Knowledge of the physical 3.4 35 connections and/or cause effect (62) relationships between the MT/G system and the following systems: Protection system 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring Rev 5
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 2 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 075 Circulating Water X A2.02 Ability to (a) predict the impacts of 2.5 36 the following malfunctions or operations (63) on the Circulating Water System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of circulating water pumps 079 Station Air 086 Fire Protection K/A Category Point Totals: 1 0 1 1 1 1 1 1 1 1 1 Group Point Total: 10 Rev 5
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: CPNPP Date of Exam: July 18, 2016 Category K/A # Topic RO SRO-Only IR # IR #
Knowledge of administrative requirements for temporary 2.1.15 2.7 66 management directives, such as standing orders, night orders, Operations memos, etc.
2.1.38 Knowledge of the station's requirements for verbal 3.7 67 communications when implementing procedures.
1.
Conduct of 2.1.41 Knowledge of the refueling process. 2.8 68 Operations Subtotal 3 2.2.13 Knowledge of tagging and clearance procedures. 4.1 69 2.2.38 Knowledge of conditions and limitations in the facility 3.6 70 license.
2.
Equipment Control Subtotal 2 2.3.4 Knowledge of radiation exposure limits under normal or 3.2 71 emergency conditions.
Knowledge of radiological safety principles pertaining to 2.3.12 3.2 72 licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to
- 3. locked high-radiation areas, aligning filters, etc Radiation Control Subtotal 2 2.4.39 Knowledge of RO responsibilities in emergency plan 3.9 73 implementation.
2.4.46 Ability to verify that the alarms are consistent with the 4.2 74 plant conditions.
4.
Emergency 2.4.50 Ability to verify system alarm setpoints and operate 4.2 75 Procedures / controls identified in the alarm response manual.
Plan Subtotal 3 Tier 3 Point Total 10 Rev 5
ES-401 PWR Examination Outline Form ES-401-2 Facility: CPNPP Date of Exam: July 18, 2016 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total
- 1. 1 3 3 6 Emergency &
Abnormal 2 N/A N/A 2 2 4 Plant Evolutions Tier Totals 5 5 10 1 3 2 5 2.
Plant 2 0 2 1 3 Systems Tier Totals 5 3 8
- 3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 7 Categories 2 1 2 2 Notes:
- 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.
Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As Rev 5
ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 000007 Reactor Trip - Stabilization - X EA2.05 Ability to determine or interpret the 3.9 76 Recovery / 1 following as they apply to a reactor trip: Reactor trip first-out indication 000008 Pressurizer Vapor Space X 2.1.7 Ability to evaluate plant performance and 4.7 77 Accident / 3 make operational judgments based on operating characteristics, reactor behavior, and interpretation.
000009 Small Break LOCA / 3 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 X 2.4.9 Knowledge of low power/shutdown 4.2 78 implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.
000038 Steam Gen. Tube Rupture / 3 000040 Steam Line Rupture - Excessive X AA2.05 Ability to determine and interpret the 4.5 79 Heat Transfer / 4 following as they apply to the Steam Line Rupture: When ESFAS systems may be secured 000054 Loss of Main Feedwater / 4 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 X 2.4.11 Knowledge of abnormal condition 4.2 80 procedures.
000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 W/E05 Inadequate Heat Transfer - Loss X EA2.1 Ability to determine and interpret the 4.4 81 of Secondary Heat Sink / 4 following as they apply to the (Excessive Heat Transfer): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 3 3 Group Point Total: 6 Rev 5
ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 (RO / SRO)
E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #
1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 X AA2.04 Ability to determine and interpret 3.6 82 the following as they apply to the Dropped Control Rod: Rod motion stops due to dropped rod 000005 Inoperable/Stuck Control Rod / 1 X 2.1.32 Ability to explain and apply system 4.0 83 limits and precautions.
000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 X 2.2.38 Knowledge of conditions and 4.5 84 limitations in the facility license.
000068 Control Room Evac. / 8 000069 Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 W/E 08 RCS Overcooling - PTS / 4 X EA2.1 Facility conditions and selection of 4.2 85 appropriate procedures during abnormal and emergency operations K/A Category Point Totals: 2 2 Group Point Total: 4 Rev 5
ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 1 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal 006 Emergency Core Cooling X A2.13 Ability to (a) predict the impacts 4.2 86 of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Inadvertent SIS actuation 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water X 2.2.3 (multi-unit license) Knowledge of 3.9 87 the design, procedural, and operational differences between units.
010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features X A2.05 Ability to (a) predict the impacts 4.2 88 Actuation of the following malfunctions or operations on the ESFAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of dc control power 022 Containment Cooling 026 Containment Spray 039 Main and Reheat Steam X A2.04 Ability to (a) predict the impacts 3.7 89 of the following malfunctions or operations on the MRSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Malfunctioning steam dump 059 Main Feedwater X 2.2.44 Ability to interpret control room 4.4 90 indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator Rev 5
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 1 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals: 3 2 Group Point Total: 5 Rev 5
ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 2 (RO / SRO)
System # / Name K K K K K K A A A A G* K/A Topic(s) IR #
1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor X 2.2.25 Knowledge of the bases in 4.2 91 Technical Specifications for limiting conditions for operations and safety limits 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment X A2.01 Ability to (a) predict the impacts of 4.4 93 the following malfunctions or operations on the Fuel Handling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Dropped fuel element 035 Steam Generator X A2.01 Ability to (a) predict the impacts of 4.6 92 the following malfunctions or operations on the S/GS ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulted or ruptured S/Gs 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 2 1 Group Point Total: 3 Rev 5
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: CPNPP Date of Exam: July 18, 2016 Category K/A # Topic RO SRO-Only IR # IR #
Knowledge of operator responsibilities during all modes of 2.1.2 4.4 94 plant operation.
Knowledge of individual licensed operator responsibilities 2.1.4 3.8 95 related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status,
- 1. 10CFR55, etc.
Conduct of Operations Subtotal 2 2.2.42 Ability to recognize system parameters that are entry- 4.6 96 level conditions for Technical Specifications.
2.
Equipment Control Subtotal 1 2.3.11 Ability to control radiation releases. 4.3 97 Knowledge of radiological safety principles pertaining to 2.3.12 3.7 98 licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
3.
Radiation Control Subtotal 2 2.4.29 Knowledge of the emergency plan 4.4 99 2.4.40 Knowledge of SRO responsibilities in emergency plan 4.5 100 implementation.
4.
Emergency Procedures /
Plan Subtotal 2 Tier 3 Point Total 7 Rev 5
ES-401 Record of Rejected K/As Form ES-401-4 Tier/Group Randomly Reason for Rejection Selected K/A 1/1 009 G.2.4.6 Question 40 Small Break LOCA: Unable to write an operationally valid question to RO tasks performed outside the control room as the recovery procedures do not have any actions which are performed by the ROs outside of the control room. For the few instances that field actions are taken these would not be performed by the ROs. Randomly/Systematically Replaced K/A 009 G.2.4.34 with K/A 009 G.2.4.6.
1/1 022 AK3.03 Question 42 Loss of Rx Coolant Makeup: Unable to write an operationally valid question that could not be challenged as CPNPP procedure guidance does not exist for reasons to avoid plant transients, thus no plausible distractors could be excluded.
Randomly/Systematically Replaced K/A 022 AK3.05 with K/A 022 AK3.03.
1/1 027 AK1.02 Question 44 Pressurizer Pressure Control Malfunctions: Could not write a discriminating question based on the content of the original K/A. Randomly/Systematically Replaced K/A 027 AK1.01 with K/A 027 AK1.02 1/1 038 G.2.4.4 Question 46 Steam Generator Tube Rupture: Could not write a question to the original K/A as no immediate action steps exist for a SGTR. Randomly/Systematically Replaced K/A 038 G.2.4.1 with K/A 038 G.2.4.4.
1/1 054 AK3.01 Question 47 Loss of Main Feedwater: Could not write an operationally valid RO level question on PORVs cycling as this requires operational errors in performing the actions of FRH-0.1A.
Additionally, this is a Loss of Heat Sink which is beyond the Loss of Main Feedwater. Randomly/Systematically Replaced K/A 054 AK3.05 with AK3.01.
1/1 057 AA1.03 Question 50 Loss of Vital AC Inst. Bus: Unable to an develop operationally valid question as no guidance is provided for use of backup instrument indications. Randomly/Systematically Replaced K/A 057 AA1.05 with 057 AA1.03.
1/1 G.2.1.20 Question 51 Loss of DC Power: Unable to write operationally valid question for original K/A. Randomly/Systematically Replaced K/A 058 G.2.2.12 with 058 G.2.1.20.
1/1 065 AA1.04 Question 53 Loss of Instrument Air: CPNPP design does not include manual loaders. Randomly/Systematically Replaced K/A 065 AA1.01 with K/A 065 AA1.04.
Rev. 5
ES-401 Record of Rejected K/As Form ES-401-4 1/1 040 AK3.04 Question 54 LOCA Outside Containment: This ECA is three steps in length and the details of this procedure are beyond RO knowledge.
Randomly/Systematically Replaced K/A W/E 04 EK3.02 with 040 AK3.04 1/2 033 AA1.01 Question 60 Loss of Intermediate Range NI: Unable to write an operationally valid question with plausible distractors based on CPNPP design for the original K/A. Randomly/Systematically Replaced K/A 033 AA1.02 with K/A 033 AA1.01.
1/2 W/E10 EK1.2 Question 63 Loss of Containment Integrity: Unable to write an operationally valid question on the effect of pressure on leak rate as licensed operators are not required to perform tasks during normal or emergency operations of calculating containment leakage rates. Randomly/Systematically Replaced K/A 069 AK1.01 with W/E10 EK1.2.
2/1 004 Question 2 G.2.4.45 Chemical and Volume Control: Original K/A was not RO level knowledge. Randomly/Systematically Replaced K/A 004 G.2.4.41 with K/A 004 G.2.4.45.
2/1 073 K3.01 Question 13 Containment Cooling: Original K/A concerning equipment damage to containment equipment by temperature, humidity and pressure was beyond RO knowledge level as these concerns are addressed via the Technical Specifications and Technical Requirement Manual below the line of RO knowledge. Randomly/Systematically Replaced K/A 022 K3.01 with K/A 073 K3.01.
2/1 026 K2.02 Question 15 Containment Spray: Original K/A was on power supply to the pumps. As the pumps are powered directly from the respective train related 6.9 kV Safeguards Bus an operationally discriminating question could not be written.
Randomly/Systematically Replaced K/A 026 K2.01 with K/A 026 K2.02.
2/1 061 K6.02 Question 24 Service Water: Original K/A was intended to be a K6 per the random selection process and K4.03 was listed as the topic.
This placed the Tier Totals out of balance.
Randomly/Systematically Replaced K/A 076 K4.03 with K/A 061 K6.02.
2/1 078 K2.01 Question 26 Instrument Air: Original K/A involved securing Service Air upon loss of cooling water. The instrument and service air systems at CPNPP are not connected and thus this K/A is not applicable to the plant. Randomly/Systematically Replaced K/A 078 K4.03 with K/A 078 K2.01.
Rev. 5
ES-401 Record of Rejected K/As Form ES-401-4 2/2 002 K5.08 Question 31 In-core Temperature Monitor: Could not write an operationally valid question on the operational implications of the ITM system with respect to saturation and subcooling of water without overlapping other questions which existed on the exam. Randomly/Systematically Replaced K/A 017 K5.02 with K/A 002 K5.08.
2/2 075 A2.02 Question 36 Waste Gas Disposal: Replaced K/A as CPNPP design has any stuck open relief valve either discharging to another Gas Decay Tank or discharging via an unisolable path to the plant vent thus no procedural actions exist for the failure.
Randomly/Systematically Replaced K/A 071 A2.09 with K/A 075 A2.02.
2/2 014 K4.06 Question 37 Area Radiation Monitoring: Replace K/A as CPNPP Area Radiation Monitoring system does not provide isolations or interlocks to operations. Randomly/Systematically Replaced K/A 072 K4.02 with K/A 014 K4.06.
2/2 001 A3.07 Question 38 Circulating Water: Replaced K/A as CPNPP design does not have emergency essential SSW pumps in the circulating water system. Randomly/Systematically Replaced K/A 075 K2.03 with K/A 001 A3.07.
3 G. 2.2.13 Question 69 ROs do not manage maintenance activities and therefore this generic was not applicable to the RO position.
Randomly/Systematically Replaced Generic G.2.2.18 with G.2.2.13.
3 G.2.3.12 Question 72 Question was unsat based on overlap with Operating Test and LOD was rated by NRC as 1 based on looking up on an RWP.
Randomly/Systematically Replaced Generic G.2.3.7 with G.2.3.12.
1/1 008 G.2.1.7 Question 77 Pressurizer Vapor Space Accident: Original K/A was an oversampling of G.2.1.32 with Question 83.
Randomly/Systematically Replaced K/A 008 G.2.1.32 with K/A 008 G.2.1.7.
1/1 040 AA2.05 Question 79 Steam Line Rupture - Excessive Heat Transfer: Original K/A asked for the ability to determine and interpret the difference between a Steam Line Rupture and LOCA. Was unable to write a question which met the NUREG-1021 criteria for SRO only as the determination aligned itself with identification and entry into Major EOPs as delineated in NUREG-1021.
Randomly/Systematically Replaced K/A 040 AA2.03 with K/A 040 AA2.05.
Rev. 5
ES-401 Record of Rejected K/As Form ES-401-4 1/1 057 G.2.4.11 Question 80 Loss of Main Feedwater: Original K/A did not have a corresponding 10CFR 55.43(b) tie. Randomly/Systematically Replaced K/A 054 G.2.4.31 with K/A 054 G.2.4.11. Unable to write acceptable SRO level question to selected K/A.
Randomly/Systematically Replaced K/A 054 G.2.4.11 with K/A 057 G.2.4.11.
1/2 W/E 08 A2.01 Question 85 Steam Generator Over-pressure: Original K/A was based on a Westinghouse Yellow path FRG. This procedure has few steps and does not offer discriminating value for an SRO only question. Randomly/Systematically Replaced K/A W/E 13 A2.02 with W/E 08 A2.01.
2/2 017 G.2.2.25 Question 91 Containment Iodine Removal: As the Preaccess filtration system is the only approved Iodine Removal System at CPNPP, was unable to write an operationally valid SRO level question to a non-safety system with limited administrative requirements. Randomly/Systematically Replaced K/A 027 G.2.4.50 with 017 G.2.2.25.
2/2 035 A2.01 Question 92 Spent Fuel Pool Cooling: Unable to write an SRO discriminatory question to the K/A. Randomly/Systematically Replaced K/A 033 A2.02 with K/A 035 A2.01.
3 G.2.1.4 Question 95 Unable to write a question at the SRO level for locating and operating components that is not at the RO level or is not tied to a specific system or event which is contrary to the Tier 3 requirements. Randomly/Systematically Replaced K/A G.2.1.30 with K/A G.2.1.4.
3 G.2.4.29 Question 99 Unable to write an SRO level question to the original K/A.
Randomly/Systematically Replaced K/A G.2.4.17 with K/A G.2.4.18. Unable to write an acceptable Tier 3 question for the replaced K/A. Randomly/Systematically Replaced K/A G.2.4.18 with K/A G.2.4.29.
Rev. 5
ES-301 Administrative Topics Outline Form ES-301-1 Task Summary Facility: CPNPP Units 1 and 2 Date of Examination: July 2016 Examination Level: RO SRO Operating Test Number: NRC Administrative Topic Type Describe activity to be performed (See Note) Code*
2.1.23 Ability to perform specific system and integrated plant procedures during all Conduct of Operations modes of plant operation. (4.3)
(RA1) D,R JPM: Calculate BOL Boration for Long Term Use.
(RO1307D) 2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant temperature, secondary Conduct of Operations plant, fuel depletion etc. (4.1)
(RA2) M,R JPM: Determine Reactivity Effects When Starting Positive Displacement Charging Pump. (RO1310E) 2.2.1 Ability to perform pre-startup procedures including operating those controls Equipment Control associated with plant equipment that could (RA3) D,R affect reactivity. (4.5)
JPM: Perform a 1/M Plot and Predict Critical Conditions. (RO1003A) 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal Radiation Control conditions. (3.5)
(RA4) D,R JPM: Determine Entry Conditions for Radiation Area Clearance. (RWT056B)
Emergency Procedures/Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.
Rev. 1
ES-301 Administrative Topics Outline Form ES-301-1 Task Summary
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs; < 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (> 1)
(P)revious 2 exams (< 1; randomly selected)
RA1 The applicant will calculate BOL Boration for long term use per SOP-104A, Reactor Makeup and Chemical Control System, Attachment 2, BOL Boration for Long-Term Use. Critical steps include determining Reactor Coolant System corrected boron, gallons of Reactor Makeup Water to offset boron, and potentiometer settings for the Chemical and Volume Control System. This is a direct from bank JPM. (K/A 2.1.23 - IR 4.3)
RA2 The applicant is presented with information pertaining to the boron concentration in the suction piping to the Positive Displacement Charging Pump and the Reactor Coolant System. The applicant will use SOP-103A, Chemical and Volume Control System to determine the reactivity effects of the planned evolution. The critical steps will be to calculate the change in RCS boron concentration and RCS temperature. The JPM is modified from a previous version by changing the RCS boron concentration and the concentration in the PDP suction line. The prior version was a resulting boration. The modified version is a resulting dilution. This is a modified JPM. (K/A 2.1.43 - IR 4.1)
RA3 The applicant will perform a 1/M plot for a Reactor Startup per IPO-002A, Plant Startup From Hot Standby, Attachment 2, Inverse Count Rate Ratio Calculation The critical steps include the critical steps include calculating and plotting 1/M, predicting critical conditions, and identifying action for criticality above the power dependent insertion limit. This is a direct from bank JPM. (K/A 2.2.1 - IR 4.5)
RA4 The applicant will determine the radiological requirements for implementing a Clearance in a Radiological Controlled Area per STA-656, Radiation Work Control, RPI-602, Radiological Surveillance and Posting, and RPI-606, Radiation Work and General Access Permits. Critical tasks include identifying Dose Monitoring Requirements, Protective Clothing Requirements, highest contamination level, and highest dose rate. This is a direct from bank JPM.
(K/A 2.2.37 - IR 3.5)
Rev. 1
ES-301 Administrative Topics Outline Form ES-301-1 Task Summary Facility: CPNPP Units 1 and 2 Date of Examination: July 2016 Examination Level: RO SRO Operating Test Number: NRC Administrative Topic Type Describe activity to be performed (See Note) Code*
2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high Conduct of Operations temperatures, high pressure, caustic, (SA1) N,R chlorine, oxygen and hydrogen). (3.6)
JPM: Determine Electrical Safe Work Practices Requirements. (SO1028) 2.1.37 Knowledge of procedures, guidelines or limitations associated with reactivity Conduct of Operations management. (4.6)
(SA2) N,R JPM: Determine Reactivity Management Severity and Notifications. (SO1017B) 2.2.14 Knowledge of the process for controlling Equipment Control equipment configuration or status. (4.3)
(SA3) D,R JPM: Determine Fire Compensatory Measures for an Emergent Condition. (SO1048) 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions. (3.6) 2.4.30 Knowledge of events related to system operation/status that must be reported to Radiation Control D,R internal organizations or external agencies, (SA4) such as the State, the NRC, or the transmission system operator. (4.1)
JPM: Determine Entry Conditions for Radiation Area Clearance and Reporting Requirements. (SO1112B) 2.4.41 Knowledge of emergency action level Emergency thresholds and classifications. (4.6)
Procedures/Plan D,R (SA5) JPM: Classify an Emergency Plan Event.
(SO1136I)
Rev. 1
ES-301 Administrative Topics Outline Form ES-301-1 Task Summary NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 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)
(N)ew or (M)odified from bank (> 1)
(P)revious 2 exams (< 1; randomly selected)
SA1 The applicant is presented with a task to determine as the Unit Supervisor, the Personnel Protective Equipment and Safety Boundaries for emergent work of racking the Rx trip breaker from disconnect to remove in accordance with STA-124, Electrical Safe Work Practices. The critical steps will be to identify the Hazard/Risk Category, Clothing requirements and Boundaries. In addition, the applicant will be required to determine if their position has approval authority for the task. This is a new JPM. (K/A 2.1.26 - IR 3.6)
SA2 The applicant is presented with a plant transient event and response. As Unit Supervisor, the applicant is required to take necessary actions for a reactivity management event in accordance with STA-102, Reactivity Management Program. The critical steps will be to make a determination of the Severity Level and determine the written and verbal notifications. This is a new JPM.
(K/A 2.1.37 - IR 4.6)
SA3 The applicant will evaluate a Fire Protection Impairment per STA-738, Fire Protection Systems/Equipment Impairments. The critical steps are to determine Fire Watch Implementation and other Compensatory Measures. This is a direct from bank JPM. (K/A 2.2.14 - IR 4.3)
SA4 The applicant will determine the radiological requirements for implementing a Clearance in a Radiological Controlled Area per STA-656, Radiation Work Control, RPI-602, Radiological Surveillance and Posting, RPI-606, Radiation Work and General Access Permits and STA-501, Nonroutine Reporting. Critical steps include identifying Dose Monitoring Requirements, Protective Clothing Requirements, highest contamination level, highest dose rate and determination of proper oral and written notifications due to an overexposure event. This is a direct from bank JPM. (K/A 2.3.7 - IR 3.6 & K/A 2.4.30 - IR 4.1)
SA5 The applicant will determine the appropriate Emergency Plan Classification in accordance with EPP-201, Assessment of Emergency Action Levels, Emergency Classification, and Plan Activation. The critical step will be the determination of the correct classification. This is a direct from bank JPM. (K/A 2.4.41 - IR 4.6)
Rev. 1
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Facility: CPNPP Units 1 and 2 Date of Examination: July 2016 Exam Level: RO SRO(I) SRO (U) Operating Test Number: NRC Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
Safety System / JPM Title Type Code*
Function 003 - Dropped Control Rod (RO1024A)
S-1 M,S 1 Respond to Control Rod Misalignment 010 - Pressurizer Pressure Control System (RO1205)
S-2 A,D,S 3 PORV Block Valve Operability Test 002 - Reactor Coolant System (RO1412C) L,M,S 4P S-3 Respond to a Shutdown Loss of Coolant 045 - Main Turbine Generator System (RO3113)
S-4 A,L,N,S 4S Perform Pre-Startup Turbine Trip Checks 026 - Containment Spray System (RO2002C)
S-5 Transfer Containment Spray to Recirculation with A,D,EN,L,S 5 Cavitation 064 - Emergency Diesel Generator System (RO4215B)
S-6 A,D,P,S 6 Restore Safeguards Bus 1EA1 to Offsite Power 015 - Nuclear Instrumentation System (RO1820)
S-7 D,S 7 Respond to a Power Range Channel Malfunction 067 - Plant Fire On-site (RO4405)
S-8 D,S 8 Respond to Fire in the Safeguards Building In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) 004 - Chemical and Volume Control (AO5202A) A,D,E,R 2 P-1 Perform Local Actions to Restart the Positive Displacement Pump 055 - Loss of All AC Power (RO4217H) N,E,L 6 P-2 Perform Attachment 2A DC Load Shedding 068 - Control Room Evacuation (AO5115B) D,E,L,R 8 P-3 Emergency Borate from the Remote Shutdown Panel Page 1 of 4CPNPP 2016 NRC ES-301-2 RO & SRO SYSTEM JPM OUTLINE REV. 3.DOCX
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2
@ 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 / 4-6 / 2-3 (C)ontrol room (D)irect from bank <9 / <8 / <4 (E)mergency or abnormal in-plant >1 / >1 / >1 (EN)gineered safety feature - / - / > 1(control room system)
(L)ow-Power / Shutdown >1 / >1 / >1 (N)ew or (M)odified from bank including 1(A) >2 / >2 / >1 (P)revious 2 exams <3 <3 / < 2 (randomly selected)
(R)CA >1 / >1 / >1 (S)imulator NRC JPM Examination Summary Description S-1 Control Rod H-8 which is part of Control Bank D is misaligned from its bank.
Control Rod H-8 is at 204 steps as indicated on DRPI and Control Bank D indicates 216 steps. The applicant is provided ABN-712, Rod Control System Malfunction and is required to realign Control Rod H-8 using the DRPI Method.
The critical steps include selecting the proper bank, withdrawing the entire bank to a known position, deselecting the non-misaligned rods from moving, aligning Control Rod H-8, resetting the Rod Control Urgent Failure alarm, returning the entire bank to its pre-malfunction position and restoring the Control Rod system for continued operation. This is a modified from bank JPM as a recent procedural change added the directions which are to be used for clearing the Rod Control Urgent Failure alarm if present and which this JPM now exercises. This JPM is under the Control Rod Drive System - Reactivity Control Safety Function. (K/A 003.AA1.02 - IR 3.6 / 3.4)
S-2 The applicant will be provided with OPT-109A, PORV Block Valve Test and will be required to perform the Operability Test. This is an Alternate Path JPM because when PORV Block Valve 1/1-8000B is reopened as part of the test, the PORV partially opens requiring the applicant to take action to isolate the open PORV.
The critical steps include closing each PORV Block Valve, performing the stroke test of each PORV and restoring the original configuration. An additional critical step of isolating the stuck open PORV follows the malfunction. PORV Block Valves are provided to isolate a PORV if excessive leakage develops and are discussed in FSAR 15.4.13.2. This is a direct from bank JPM under the Pressurizer Pressure Control System - Reactor Pressure Control Safety Function.
(K/A 010.A4.03 - IR 4.0 / 3.8)
Page 2 of 4CPNPP 2016 NRC ES-301-2 RO & SRO SYSTEM JPM OUTLINE REV. 3.DOCX
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 S-3 The applicant will respond to a lowering Pressurizer level with the Residual Heat Removal System in service per ABN-108, Shutdown Loss of Coolant, Section 2.0, Shutdown Loss of Coolant. This is a modified JPM under the Residual Heat Removal System - Primary System Heat Removal from Reactor Core Safety Function. The modification consists of a different plant configuration as the Initial Conditions which do not require performance of an Alternate Path.
(K/A 025.AA1.02- IR 3.8 / 3.9)
S-4 The applicant will use OPT-410A, Pre-Startup Turbine Trip Checks to perform the task. This is an Alternate Path JPM as the Turbine speed will increase above the allowable procedural guidance while the HP Stop Valves are opening. This speed increase requires that the turbine be tripped in accordance with OPT-410A. The critical steps will include resetting the turbine trip, latching the turbine, opening the HP Stop Valves and tripping the turbine when speed increases. This is a new JPM under the Main Turbine Generator System - Heat Removal from Reactor Core Secondary Systems Safety Function.
(K/A 045.A4.01 - IR 3.1 / 2.9)
S-5 Following a LBLOCA, the applicant will transfer the Containment Spray System from the Injection mode to Recirculation in accordance with EOS-1.3A, Transfer to Cold Leg Recirculation. This is an Alternate Path JPM as the applicant will not be able to open the containment sump valves to the Train B Containment Spray Pumps. This will require the applicant to secure Train B. Critical steps will include transferring Train A suction to the containment sump and securing both Train B pumps when suction cannot be realigned. Transferring Containment Spray to Recirculation Mode is considered a Time Significant Action. STI-214.01, Control of Timed Operator Actions, TSA-2.8 requires Containment Spray transferred to Recirculation Mode within 70 seconds of RWST level reaching 6%. This Time Significant Action is performed to avoid the requirement to secure Containment Spray Pumps due to losing suction supply when RWST level reaches 0%. This is a direct from bank JPM under the Containment Spray System - Containment Integrity Safety Function. (K/A 026.A4.01 - IR 4.5 / 4.3)
S-6 The applicant will restore Offsite Power to 6.9 kV Safeguards Bus 1EA1 from Transformer XST1 in accordance with SOP-609A, Diesel Generator System, Section 5.7, Transferring From DG Supplying Alone to Normal or Alternate Supply.
The alternate path occurs when a lowering frequency requires separating the Emergency Diesel Generator from the grid. This is a bank JPM, previously used on the 2014 NRC operating test, under the Emergency Diesel Generator System -
Electrical Safety Function. (K/A 064.A4.07 - IR 3.4 / 3.4)
S-7 Following a Power Range Instrument failure. The applicant is required to perform the actions of ABN-703, Power Range Instrument Malfunction. Critical steps include several repositions on the NI Detector cabinets to defeat the failed instrument, defeating the N-16 Channel on CB-05 and the TAVE channel on CB-07.
This is a direct from bank JPM under the Nuclear Instrumentation System -
Instrumentation Safety Function. (K/A 015.A2.01 - IR 3.5 / 3.9)
Page 3 of 4CPNPP 2016 NRC ES-301-2 RO & SRO SYSTEM JPM OUTLINE REV. 3.DOCX
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 S-8 A fire has been identified in the Safeguards Building. The applicant is directed to respond to the fire in accordance with ABN-804A, Response to a Fire in the Safeguards Building. Critical steps include performing an emergency start of Diesel Generator 1-02, performing CVCS realignments and starting CCP 1-02.
Comanche Peak has commitments within ABN-804A, Response to a Fire in the Safeguards Building, to maintain CCP suction due to possible Gas Intrusion as noted in SOER 97-01, Loss of HP Injection & Charging from Gas Intrusion. This is a direct from bank JPM under the Plant Fire On-site - Plant Service Systems Safety Function. (K/A 067.AA2.16 - IR 3.3 / 4.0)
P-1 Following a loss of instrument air, the applicant is required to reset control air to the Positive Displacement Charging Pump in accordance with ABN-301, Instrument Air System Malfunction and restore the PDP to operation in accordance with SOP-103A, Chemical and Volume Control System. This JPM is Alternate Path as the Stuffing Box Coolant Tank level is out of specification during the pump restart and requires filling. Critical steps include resetting the air to the hydraulic speed changer, repositioning the fill valve to the coolant tank and opening the pump discharge valve. This is a direct from bank JPM under the Chemical and Volume Control System - Reactivity Control System Inventory Control Safety Function. (K/A 004.A4.08 - IR 3.8 / 3.4)
P-2 During a complete loss of All AC Power, the applicant is required to perform ECA-0.0A, Loss of All AC Power Attachment 2A which is Initial DC Load Shed.
Critical steps include performing several operations on Distribution Panels to properly align equipment from Unit 2 where possible and shed loads where required. This is a new JPM as DC Load Shedding has been redeveloped following BDBEE considerations. (K/A 055.EA1.04 3.5/3.9)
P-3 During a Control Room evacuation due to a security threat, the applicant is required to take action to place the plant in control of the operators from outside the control room. Actions will be performed using ABN-905B, Loss of Control Room Habitability. The critical steps include transferring control of equipment from the Control Room to the Hot Shutdown Panel, starting a Boric Acid Transfer Pump and opening the emergency borate valve. This is a direct from bank JPM under the Control Room Evacuation System - Plant Service Systems Safety Function.
(K/A 068.AA1.11 - IR 3.9 / 4.1)
Page 4 of 4CPNPP 2016 NRC ES-301-2 RO & SRO SYSTEM JPM OUTLINE REV. 3.DOCX
Appendix D Scenario Outline Form ES-D-1 Facility: CPNPP 1 & 2 Scenario No.: 2 Op Test No.: July 2016 NRC Examiners: Operators:
Initial Conditions: 53% power MOL - RCS Boron is 1054 ppm Turnover: 600 MWe due to a B MFP trip. 1-02 MDAFWP in Pull Out (DANGER tagged) with breaker de-energized for scheduled maintenance. Restored B MFP following the trip but 1-PV-2286 was damaged and is DANGER tagged out for repairs. Hold power per load dispatch.
Critical Tasks: CT-1 Initiate Emergency Boration prior to exiting EOS-0.1A, Reactor Trip Response within a maximum of 15 minutes following the loss of DRPI.
CT-2 Establish an RCS bleed and feed path prior to exiting FRH-0.1A, Response to a Loss of Secondary Heat Sink.
Event Malf. Event Type* Event Description No. No.
I (RO, SRO) 1 RX05A PRZR level instrument LT-459 fails low TS (SRO) 2 RX18 I (BOP, SRO) Feed Header Pressure Transmitter (PT-508) Fails High I (RO, BOP, SRO) 3 RX09A Main Turbine 1st Stage Pressure (PT-505A) Fails Low TS (SRO) 4 CH03 C (BOP, SRO) Neutron Detector Well Fan 9 trips on motor overload 5 FW06A C (BOP, SRO) Main Feed Pump A Recirc valve fails open 6 ED02 TS (SRO) Loss of XST1 Transformer ED01 Loss of offsite power 7 M (RO,BOP,SRO)
EG06A Failure of the DG 1-01 to start (air start failure) 8 Emergency Boration due to loss of DRPI Loss of all AFW 9 FW09A M (RO,BOP,SRO)
TDAFWP Overspeed Trip
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 8 Total malfunctions (5-8) 2 Malfunctions after EOP entry (1-2) 6 Abnormal events (2-4) 2 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2)
EOP contingencies requiring substantive actions 1
(0-2) 2 Critical tasks (2-3)
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 1.DOCX
Scenario Event Description NRC Scenario 2 SCENARIO
SUMMARY
NRC 2 Turnover:
The plant is at 600 MW following a B MFP trip. Reactor power is being held stable per instruction of the Load Dispatcher. MDAFWP 1-02 is Danger tagged for planned maintenance. When the B MFP tripped it caused damage to 1-PV-2286, Low Pressure Feedwater Heater Bypass Valve. The MFP has been restored to operation but 1-PV-2286 is Danger tagged to complete repairs.
Event 1 (Key 1)
The first event will be a PRZR level channel (LT-459) failing low. Entry into ABN-706, PRZR Level Instrumentation Malfunction, section 2.0, will be required. Letdown will isolate, charging will be placed in manual to control PRZR level. Actions will include selecting an operable channel, restoring letdown, and then restoring PRZR level to program and placing controls back in automatic. The SRO will determine the loss of this channel is a TS entry for LCO 3.3.1, Reactor Trip System Instrumentation; function 9, Condition M.
Event 2 (Key 2)
The next event is Main Feedwater (MFW) Header Pressure Transmitter (PT-508) will fail high. Entry into ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st Stage Pressure, and Feed Header Pressure Instrument Malfunction Section 5.0, is required. Section 5.0 is designated for Feed Header Pressure Malfunction. Actions include placing the MFW Pump Turbine Master Speed Controller in MANUAL. This controller will remain in MANUAL for the duration of the scenario and require monitoring/adjustment. If Pressurizer Pressure falls below 2220 psig, the DNB TS 3.4.1 should be entered.
Event 3 (Key 3)
Once the plant is stabilized, the next event is a Turbine 1st Stage Pressure Transmitter (PT-505) will fail low. Crew actions are per ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st Stage Pressure, and Feed Header Pressure Instrument Malfunction, Section 4.0 is required. Section 4.0 is designated for Turbine 1st Stage Pressure Malfunction. Actions include placing Rod Control in Manual and bypassing the failed Turbine 1st Stage Pressure channel. The SRO will refer to Technical Specification LCO 3.3.1, Reactor Trip System Instrumentation; Function 18f, Condition T.
Event 4 (Key 4)
The next event will be a trip of the running Neutron Detector Well Fan #9. This will alarm 2.1 CNTMT FN MASTER TRIP. The ALM will direct the crew to determine which fan has tripped and start the other fan as required using SOP-801A, Containment Ventilation System. The crew will place the tripped fan handswitch in Pull Out or Stop as applicable.
Event 5 (Key 5)
The next event is MFP A Recirculation Valve, 1-FCV-2289, opening. ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, section 11.0 will be entered. Since earlier in this scenario the main feedwater header pressure transmitter failed MFP speed control is in manual.
Manual speed control is required to restore S/G levels and stabilize the plant. The RO must ensure rods are in auto for this event. The crew will dispatch an operator to isolate the failed open recirculation valve. Once the failed valve is isolated, the BOP will adjust MFP speed again for the current plant configuration.
Event 6 (Key 6)
The next event is a loss of XST1 which is the alternate offsite power source for Unit 1. The ALM will have the crew enter ABN-601, Response to a 138/345 KV System Malfunction, as well as a TS entry for 3.8.1, Electrical Power Systems, AC Sources - Operating, Condition A.
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 1.DOCX
Scenario Event Description NRC Scenario 2 Event 7 (Key 7)
The major event is a loss of all offsite power causing a reactor trip. The crew will enter EOP-0.0A, Reactor Trip or Safety Injection. Coincident with the loss of offsite power DG 1-01 will fail to auto start and cannot be manually started due to an air start failure. This will cause a complete loss of all safeguards train A power. The crew will transition to EOS-0.1A, Reactor Trip Response, to continue with recovery efforts.
Event 8 - CT-1 (Auto)
Due to the Loss of Offsite Power, DRPI is lost and per EOP-0.0A and EOS-0.1A, Reactor Trip Response, Attachments 1.A, an Emergency Boration will be required. The crew will then perform CT-1; Initiate Emergency Boration prior to exiting EOS-0.1A, Reactor Trip Response. This will be completed by entering ABN-107, Emergency Boration.
Event 9 (Auto Triggered when 1/1-8104 is placed in open per ABN-107)
After the crew has commenced the emergency boration the TDAFWP will trip on Overspeed. This combined with the loss of all Safeguards Train A power as well as the inoperability of MDAFWP 1-02 will place the crew in a loss of heat sink event. The crew will enter FRH-0.1A, Response to Loss of Secondary Heat Sink, and actuate SI. The crew will then perform CT-2, Establish an RCS bleed and feed path prior to exiting FRH-0.1A, Response to a Loss of Secondary Heat Sink.
Termination Criteria This scenario is terminated after the crew establishes a bleed and feed path per FRH-0.1A. One CCP and one SI pump running with both PRZR PORVs open.
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 1.DOCX
Scenario Event Description NRC Scenario 2 Risk Significance Determination Risk Significance Event Guidance Failure of risk important systems Loss of Transformer XST1 FSAR 8.2.1.2.1 - Two prior to Reactor Trip independent power sources are available on an immediate basis following a DBA to ensure operation of the vital safety functions. The second offsite power source will no longer be available on loss of XST1.
Risk significant core damage FSAR 15.2.6.3 Loss of Non- EOS-0.2A, Natural Circulation sequence emergency AC power to the Cooldown - For Units 1 and 2, station auxiliaries the analysis of the natural circulation capability of the RCS has demonstrated that sufficient heat removal capability exists following reactor coolant pump coastdown to prevent fuel or clad damage.
Loss of Secondary Heat Sink CPNPP Accident Sequence Quantification, R&R-PN-022 -
Loss of secondary heat removal, not related to ventilation failures, accounts for about 9% of CDF.
Risk significant operator actions Initiation of Boration to Add STI 214.01; ABN-107, Negative Reactivity to the Core Emergency Boration; (TSA 2.14) WCAP-1687 1-P, Section 6.3.5; TRM Bases 13.1.31 - Within 15 minutes, when local alignment is required to establish boration flow. Boration is initiated within the prescribed time. When local manual control credited, admin controls are utilized to ensure personnel are aware/designated to perform alignment to establish boration flow.
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 1.DOCX
Scenario Event Description NRC Scenario 2 Critical Task Determination Measurable Safety Performance Critical Task Cueing Performance Significance Feedback Indicators Initiate Shutdown Margin After the loss of offsite Started CCP Boration flow will Emergency must be power and the failure 1-02, started be indicated on Boration prior to maintained. of the DG 1-01, DRPI Boric Acid 1-FI-183A, EMER exiting EOS-0.1A, Since there are will be dark and no Transfer Pump BORATE FLO.
Reactor Trip NO DRPI lights lit CCP will be running. 1-02, and Response within the bases states Per attachment 1.A of opened 1/1-a maximum of 15 to borate at least EOS-0.1A, ABN-107 8104, EMER minutes following 3600 gallons of will be performed. BORATE VLV.
the loss of DRPI. 7000 ppm borated water to ensure shutdown margin is maintained. This gallon value corresponds to 2 of the most reactive rods stuck out.
Establish an RCS Actuating SI will AFW flow will not be Actuated SI, Flow indicated on bleed and feed ensure a feed indicated on any AFW ensured at least both a CCP and path prior to path of cool flow meter. Also no one CCP and SI an SI pump.
exiting FRH-0.1A, water to the RCS AFW pumps will be pump is running PRZR PORVs Response to a (core) and isolate running. A RED path with flow open with block Loss of the containment showing on CSFST for indicated valves open.
Secondary Heat to confine any heat sink. The need providing a feed RCS pressure Sink. RCS releases for a heat sink as path for the and temperature from the bleed indicated by RCS RCS. Both lowering.
flow. The bleed temperature and PRZR PORVs flow through both pressure. open providing a PORVs will bleed path for ensure that the RCS.
enough cool water will feed from the ECCS flow path to remove sufficient decay heat.
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 1.DOCX
Appendix D Scenario Outline Form ES-D-1 Facility: CPNPP 1 & 2 Scenario No.: 2 Op Test No.: July 2016 NRC Examiners: Operators:
Initial Conditions: 53% power MOL - RCS Boron is 1054 ppm Turnover: 600 MWe due to a B MFP trip. 1-02 MDAFWP in Pull Out (DANGER tagged) with breaker de-energized for scheduled maintenance. Restored B MFP following the trip but 1-PV-2286 was damaged and is DANGER tagged out for repairs. Hold power per load dispatch.
Critical Tasks: CT-1 Initiate Emergency Boration prior to exiting EOS-0.1A, Reactor Trip Response within a maximum of 15 minutes following the loss of DRPI.
CT-1 Manually control the Main Feedwater Master Speed Controller to prevent receiving an automatic reactor trip due to low steam generator levels, or a trip of Main Feed Pumps due to low suction pressure, and subsequent manual reactor trip.
CT-2 Establish an RCS bleed and feed path prior to exiting FRH-0.1A, Response to a Loss of Secondary Heat Sink.
Event Malf. Event Type* Event Description No. No.
I (RO, SRO) 1 RX05A PRZR level instrument LT-459 fails low TS (SRO) 2 RX18 I (BOP, SRO) Feed Header Pressure Transmitter (PT-508) Fails High I (RO, BOP, SRO) 3 RX09A Main Turbine 1st Stage Pressure (PT-505A) Fails Low TS (SRO) 4 CH03 C (BOP, SRO) Neutron Detector Well Fan 9 trips on motor overload 5 FW06A C (BOP, SRO) Main Feed Pump A Recirc valve fails open 6 ED02 TS (SRO) Loss of XST1 Transformer ED01 Loss of offsite power 7 M (RO,BOP,SRO)
EG06A Failure of the DG 1-01 to start (air start failure) 8 Emergency Boration due to loss of DRPI Loss of all AFW 9 FW09A M (RO,BOP,SRO)
TDAFWP Overspeed Trip
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 87 Total malfunctions (5-8) 2 Malfunctions after EOP entry (1-2) 65 Abnormal events (2-4) 2 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2)
EOP contingencies requiring substantive actions 1
(0-2) 2 Critical tasks (2-3)
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 2.DOCX
Scenario Event Description NRC Scenario 2 SCENARIO
SUMMARY
NRC 2 Turnover:
The plant is at 600 MW following a B MFP trip. Reactor power is being held stable per instruction of the Load Dispatcher. MDAFWP 1-02 is Danger tagged for planned maintenance. When the B MFP tripped it caused damage to 1-PV-2286, Low Pressure Feedwater Heater Bypass Valve. The MFP has been restored to operation but 1-PV-2286 is Danger tagged to complete repairs.
Event 1 (Key 1)
The first event will be a PRZR level channel (LT-459) failing low. Entry into ABN-706, PRZR Level Instrumentation Malfunction, section 2.0, will be required. Letdown will isolate, charging will be placed in manual to control PRZR level. Actions will include selecting an operable channel, restoring letdown, and then restoring PRZR level to program and placing controls back in automatic. The SRO will determine the loss of this channel is a TS entry for LCO 3.3.1, Reactor Trip System Instrumentation; function 9, Condition M.
Event 2 (Key 2)
The next event is Main Feedwater (MFW) Header Pressure Transmitter (PT-508) will fail high. Entry into ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st Stage Pressure, and Feed Header Pressure Instrument Malfunction Section 5.0, is required. Section 5.0 is designated for Feed Header Pressure Malfunction. Actions include placing the MFW Pump Turbine Master Speed Controller in MANUAL. This controller will remain in MANUAL for the duration of the scenario and require monitoring/adjustment. If Pressurizer Pressure falls below 2220 psig, the DNB TS 3.4.1 should be entered.
Event 3 (Key 3)
Once the plant is stabilized, the next event is a Turbine 1st Stage Pressure Transmitter (PT-505) will fail low. Crew actions are per ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st Stage Pressure, and Feed Header Pressure Instrument Malfunction, Section 4.0 is required. Section 4.0 is designated for Turbine 1st Stage Pressure Malfunction. Actions include placing Rod Control in Manual and bypassing the failed Turbine 1st Stage Pressure channel. The SRO will refer to Technical Specification LCO 3.3.1, Reactor Trip System Instrumentation; Function 18f, Condition T.
Event 4 (Key 4)
The next event will be a trip of the running Neutron Detector Well Fan #9. This will alarm 2.1 CNTMT FN MASTER TRIP. The ALM will direct the crew to determine which fan has tripped and start the other fan as required using SOP-801A, Containment Ventilation System. The crew will place the tripped fan handswitch in Pull Out or Stop as applicable.
Event 5 (Key 5)
The next event is MFP A Recirculation Valve, 1-FCV-2289, opening. ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, section 11.0 will be entered. Since earlier in this scenario the main feedwater header pressure transmitter failed MFP speed control is in manual.
Manual speed control is required to restore S/G levels and stabilize the plant. The RO must ensure rods are in auto for this event. The crew will dispatch an operator to isolate the failed open recirculation valve. Once the failed valve is isolated, the BOP will adjust MFP speed again for the current plant configuration.
Event 6 (Key 6)
The next event is a loss of XST1 which is the alternate offsite power source for Unit 1. The ALM will have the crew enter ABN-601, Response to a 138/345 KV System Malfunction, as well as a TS entry for 3.8.1, Electrical Power Systems, AC Sources - Operating, Condition A.
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 2.DOCX
Scenario Event Description NRC Scenario 2 Event 7 (Key 7)
The major event is a loss of all offsite power causing a reactor trip. The crew will enter EOP-0.0A, Reactor Trip or Safety Injection. Coincident with the loss of offsite power DG 1-01 will fail to auto start and cannot be manually started due to an air start failure. This will cause a complete loss of all safeguards train A power. The crew will transition to EOS-0.1A, Reactor Trip Response, to continue with recovery efforts.
Event 8 - CT-1 (Auto)
Due to the Loss of Offsite Power, DRPI is lost and per EOP-0.0A and EOS-0.1A, Reactor Trip Response, Attachments 1.A, an Emergency Boration will be required. The crew will then perform CT-1; Initiate Emergency Boration prior to exiting EOS-0.1A, Reactor Trip Response. This will be completed by entering ABN-107, Emergency Boration.
Event 9 (Auto Triggered when 1/1-8104 is placed in open per ABN-107)
After the crew has commenced the emergency boration the TDAFWP will trip on Overspeed. This combined with the loss of all Safeguards Train A power as well as the inoperability of MDAFWP 1-02 will place the crew in a loss of heat sink event. The crew will enter FRH-0.1A, Response to Loss of Secondary Heat Sink, and actuate SI. The crew will then perform CT-2, Establish an RCS bleed and feed path prior to exiting FRH-0.1A, Response to a Loss of Secondary Heat Sink.
Termination Criteria This scenario is terminated after the crew establishes a bleed and feed path per FRH-0.1A. One CCP and one SI pump running with both PRZR PORVs open.
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 2.DOCX
Scenario Event Description NRC Scenario 2 Risk Significance Determination Risk Significance Event Guidance Failure of risk important Loss of Transformer XST1 FSAR 8.2.1.2.1 - Two systems prior to Reactor Trip independent power sources are available on an immediate basis following a DBA to ensure operation of the vital safety functions. The second offsite power source will no longer be available on loss of XST1.
Risk significant core damage FSAR 15.2.6.3 Loss of Non- EOS-0.2A, Natural Circulation sequence emergency AC power to the Cooldown - For Units 1 and 2, station auxiliaries the analysis of the natural circulation capability of the RCS has demonstrated that sufficient heat removal capability exists following reactor coolant pump coastdown to prevent fuel or clad damage.
Loss of Secondary Heat Sink CPNPP Accident Sequence Quantification, R&R-PN-022 -
Loss of secondary heat removal, not related to ventilation failures, accounts for about 9% of CDF.
Risk significant operator actions Initiation of Boration to Add STI 214.01; ABN-107, Negative Reactivity to the Core Emergency Boration; (TSA 2.14) WCAP-1687 1-P, Section 6.3.5; TRM Bases 13.1.31 - Within 15 minutes, when local alignment is required to establish boration flow. Boration is initiated within the prescribed time. When local manual control credited, admin controls are utilized to ensure personnel are aware/designated to perform alignment to establish boration flow.
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 2.DOCX
Scenario Event Description NRC Scenario 2 Critical Task Determination Measurable Safety Performance Critical Task Cueing Performance Significance Feedback Indicators Initiate Shutdown After the loss of Started CCP Boration flow will Emergency Margin must be offsite power and the 1-02, started be indicated on Boration prior to maintained. failure of the DG 1- Boric Acid 1-FI-183A, EMER exiting EOS- Since there are 01, DRPI will be dark Transfer Pump BORATE FLO.
0.1A, Reactor NO DRPI lights and no CCP will be 1-02, and Trip Response lit the bases running. Per opened 1/1-within a states to borate attachment 1.A of 8104, EMER maximum of 15 at least 3600 EOS-0.1A, ABN-107 BORATE VLV.
minutes following gallons of 7000 will be performed.
the loss of DRPI. ppm borated water to ensure shutdown margin is maintained.
This gallon value corresponds to 2 of the most reactive rods stuck out.
Manually control Result of After the Main Feed S/G levels Neither reactor the Main improper Pump A recirc valve maintained on nor Main Feed Feedwater operator action fails open, S/G levels program without Pumps do not Master Speed or inaction, i.e., will begin decreasing. tripping the trip.
Controller to such as an Manual control of the reactor or prevent receiving unintentional feed pump speed will tripping the Main an automatic RPS or ESF maintain S/G levels Feed Pumps on reactor trip due to actuation. on program. low suction low steam Pressure, generator levels, followed by a or a trip of Main manual reactor Feed Pumps due trip (ABN-302, to low suction immediate pressure, and operator action).
subsequent manual reactor trip.
Establish an RCS Actuating SI will AFW flow will not be Actuated SI, Flow indicated on bleed and feed ensure a feed indicated on any AFW ensured at least both a CCP and path prior to path of cool flow meter. Also no one CCP and SI an SI pump.
exiting FRH- water to the RCS AFW pumps will be pump is running PRZR PORVs 0.1A, Response (core) and running. A RED path with flow open with block to a Loss of isolate the showing on CSFST indicated valves open.
Secondary Heat containment to for heat sink. The providing a feed RCS pressure Sink. confine any RCS need for a heat sink path for the and temperature releases from as indicated by RCS RCS. Both lowering.
the bleed flow. PRZR PORVs CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 2.DOCX
Scenario Event Description NRC Scenario 2 The bleed flow temperature and open providing a through both pressure. bleed path for PORVs will the RCS.
ensure that enough cool water will feed from the ECCS flow path to remove sufficient decay heat.
CPNPP 2016 NRC SIMULATOR SCENARIO 2 REV. 3_AS RUN 2.DOCX
Appendix D Scenario Outline Form ES-D-1 Facility: CPNPP 1 & 2 Scenario No.: 3 Op Test No.: July 2016 NRC Examiners: Operators:
Initial Conditions: 100% power MOL - RCS Boron is 924 ppm Turnover: Maintain steady-state power conditions.
Critical Tasks: CT-1 Manually Trip Reactor due to Failure to Automatically Trip prior to exiting EOP-0.0A, Reactor Trip or Safety Injection.
CT-2 Identify and Isolate the Ruptured Steam Generator Prior to Commencing an Operator Induced Cooldown per EOP-3.0A, Steam Generator Tube Rupture.
Event No. Malf. No. Event Type* Event Description I (RO, SRO) Cold Leg Loop 4 NR Temperature Transmitter Failure (TE-441B) 1 RP05D TS (SRO) Fails High I (BOP, SRO) Steam Generator (1-01) Steam Line Pressure Instrument (PT-514) 2 RP03A TS (SRO) Fails Low.
3 Override C (RO, SRO) Letdown HX Outlet Flow Controller Failure (TK-130) Fails Low C (BOP, SRO) 4 FW22 Station Service Water Pump 1-01 Trip TS (SRO) 5 TC08C C (BOP, SRO) High Pressure Turbine Stop Valve #3 (UV-2430A) Fails Closed 6 SG02C M (RO, BOP, SRO) Steam Generator 1-03 Tube Rupture 7 RP15E C (BOP, SRO) Automatic Reactor Trip Failure, Manual Trip from 1B3 and 1B4 8 MS08C C (RO, SRO) Steam Generator 1-03 MSIV Fails to Close
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 8 Total malfunctions (5-8) 2 Malfunctions after EOP entry (1-2) 5 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)
CPNPP 2016 NRC SIMULATOR SCENARIO 3 REV. 2.DOCX
Scenario Event Description NRC Scenario 3 SCENARIO
SUMMARY
NRC 3 Event 1 (Key 1)
The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations.
The first event is a failure high of a Reactor Coolant System Loop 4 Narrow Range Temperature (TE=441B) element. Crew actions are per ABN-704, Tc/N-16 Instrumentation Malfunction, Section 2.0. Section 2.0 is designated for Tc/N-16 Instrumentation Malfunction. Actions include placing the Control Rods in MANUAL and defeating the failed channel. Control Rods will be restored in Manual to their pre-failure position and remain in Manual until restored to Operable per ABN-704. The SRO will refer to Technical Specification LCO 3.3.1, Reactor Trip System Instrumentation (Functions 6 & 7); Condition E, One channel inoperable.
Event 2 (Key 2)
The next event is a failure low of Main Steam Line 1 Pressure Instrument (PT-514). Crew actions are per ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st-Stage Pressure and Feed Header Pressure Instrument Malfunction, Section 2.0. Section 2.0 is designated for Steam Line Pressure Malfunction. The crew must manually control Steam Generator level, transfer to an Alternate Steam Flow Channel, and restore Steam Generator (SG) Feedwater Flow Control to AUTO. The SRO will refer to Technical Specification LCO 3.3.2, Engineered Safety Feature System (ESFAS) Instrumentation (Functions 1.e & 4.d); Condition D, One channel inoperable.
Event 3 (Key 3)
The next event is a failure of the Letdown Heat Exchanger Outlet Flow Controller, TK-130. The controller output will fail to zero demand and cause TCV-4646, LTDN HX OUT TEMP CTRL valve to close. This will result in Letdown Heat Exchanger High temperature alarms and Letdown flow to divert to the VCT on high temperature.
The crew will respond per the ALM, take manual control of TK-130 and raise demand to establish Letdown Heat Exchanger Outlet temperature to approximately 95°F.
Event 4 (Key 4)
The next event is a trip of Station Service Water Pump 1-01. Crew actions are per ABN-501, Station Service Water System Malfunction, Section 2.0. Section 2.0 is designated for Station Service Water Pump Trip. Various equipment controls, as directed by ABN-501, are placed in PULL-OUT to prevent starting with no cooling water available. The SRO will refer to Technical Specification LCO 3.7.8, Station Service Water System; Condition B, One SSWS Train inoperable. The SRO will also refer to Technical Specification LCO 3.8.1, AC Sources -
Operating; Condition B, One DG inoperable as DG 1-01 must be placed in PULL-OUT upon the loss of Train A Station Service Water.
Event 5 (Key 5)
The next event is High Pressure Turbine Stop Valve #3 fails closed. The crew will enter ABN-401, Main Turbine Malfunction, Section 9.0. Section 9.0 is designated for Inadvertent Closure of an HP or LP Stop or Control Valve.
Actions include placing rod control in Auto to allow the rod control system to respond to the plant transient and reducing turbine load to allow all operable HP Control Valves to come off their full open seat.
Event 6 - (Key 6)
The major event is a Tube Rupture on SG 1-03. The Crew will diagnose the Tube Rupture due to multiple Radiation alarms and lowering Pressurizer Pressure and Level. The crew will enter EOP-0.0A, Reactor Trip or Safety Injection and transition to EOP-3.0A, Steam Generator Tube Rupture. A maximum rate RCS cooldown to a target CET temperature as determined in EOP-3.0A will be conducted.
Event 7 - CT-1 (Auto)
The Reactor will be manually tripped and Safety Injection manually initiated. The Reactor will fail to trip from both handswitches at CB-07 and CB-10. The crew will then perform CT-1, Manually Trip Reactor due to Failure to Automatically Trip prior to exiting EOP-0.0A, Reactor Trip or Safety Injection. The Reactor must be manually tripped by momentarily de-energizing 480V Normal Switchgear 1B3 and 1B4 to de-energize the Rod Drive MG Sets. The critical task is considered not met if the crew is not successful in tripping the reactor during EOP-0.0A and transitions to FRS-0.1A.
CPNPP 2016 NRC SIMULATOR SCENARIO 3 REV. 2.DOCX
Scenario Event Description NRC Scenario 3 Event 8 (Auto) CT-2 During performance of CT-2, Identify and Isolate the Ruptured Steam Generator Prior to Commencing an Operator Induced Cooldown per EOP-3.0A, Steam Generator Tube Rupture. SG 1-03 MSIV will fail to close. The crew will close all remaining MSIVs, disable the Steam Dumps, and close the Main Steam to Auxiliary Steam Supply Valve.
The RCS cooldown will then be conducted via the intact SG ARVs to atmosphere.
Termination Criteria This scenario is terminated when the target CET Temperature is reached during the RCS cooldown in accordance with EOP-3.0A, Steam Generator Tube Rupture.
CPNPP 2016 NRC SIMULATOR SCENARIO 3 REV. 2.DOCX
Scenario Event Description NRC Scenario 3 Risk Significance Determination Risk Significance Event Guidance Failure of risk important system Event 4 - Station Service Water ABN-501; DBD-ME-011 - Initial prior to Reactor Trip Pump Trip operator action to place the affected DG Emergency Stop/Start handswitch in PULL-OUT to remove the DG from service as it will ONLY operate for 1 minute under load, without service water cooling flow, before damage will occur.
Risk significant operator actions Event 7 - Manually tripping the FSAR 15.8 - The worst common Reactor by momentarily de- mode failure which is postulated energizing 1B3 and 1B4 to occur is the failure to trip the reactor after an anticipated transient has occurred.
Event 8 - Closing all intact SG FSAR 15.6.3.2 - The closing of MSIVs upon failure of the all intact SG MSIVs falls in line ruptured SG MSIV to close with the conservative analysis of the postulated SGTR which assumes a loss of offsite power.
Thus, a release of steam from the secondary system occurs due to the loss of steam dump capability and the subsequent venting to the atmosphere through the ARVs.
Risk significant core damage Events 5 - Steam Generator (1)
STI-214.01 TCA-1.9 - Manual sequence Tube Rupture Actions to Mitigate Effects of a Steam Generator Tube Rupture:
- 1) TDAFWP flow stopped (excessive AFW flow) within 3 minutes of reactor trip. 2) Identify and Isolate ruptured SG within 13 minutes after initiation of SGTR. 3) Initiate maximum rate cooldown within 5 minutes after isolation of ruptured SG. 4)
Initiate RCS depressurization with PORVs within 2 minutes after completion of RCS cooldown. 5) Secure ECCS within 2 minutes after completion of RCS depressurization.
(1) Crew manning for Initial License Examination less than Timed Operator Action validation constraints CPNPP 2016 NRC SIMULATOR SCENARIO 3 REV. 2.DOCX
Scenario Event Description NRC Scenario 3 Critical Task Determination Measurable Safety Performance Critical Task Cueing Performance Significance Feedback Indicators CT Manually Recognize a failure Procedural direction at The operator will De-energizing the Trip Reactor due to or an incorrect EOP-0.0A Step 1 to attempt to Rod Drive MG sets Failure to automatic determine if a reactor trip manually trip the will result in a loss Automatically Trip actuation of an has occurred. Position Reactor with the of power to the Rod prior to exiting ESF system or indication of the Reactor handswitches on Drive Mechanisms EOP-0.0A, Reactor component. FSAR Trip breakers and both CB-07 and and the Control Trip or Safety 7.1.2.1 Reactor Power, CB-10; however, Rods will insert into Injection Annunciator First out the Reactor will the core. Reactor alarms. fail to trip. The Trip Breakers will operator will then remain closed, momentarily neutron flux will deenergize the lower and rod 480V normal bottom lights will be switchgear 1B3 lit.
and 1B4 to secure power to the Rod Drive MG sets.
CT Identify and Take one or more Procedurally driven from The operator will SG pressure Isolate the actions that would EOP-3.0A, to identify attempt close the increasing, AFW Ruptured Steam prevent a and isolate a ruptured ruptured SG MSIV flow reduced to Generator Prior to challenge to plant SG. Indications include from the control zero and valve Commencing an safety. STI-214.01, MSL Radiation alarms room, however, position indications.
Operator Induced TCA-1.9; FSAR and SG level. the MSIV will fail Cooldown per EOP- 15.6.3.1.1; WCAP- to close and all 3.0A, Steam 16871-P, Section other MSIVs must Generator Tube 6.4; DBD-ME-027 be closed. The Rupture. MSIV will be locally closed in the field. The operator will stop feeding the SG once sufficient level to cover the tubes is available.
CPNPP 2016 NRC SIMULATOR SCENARIO 3 REV. 2.DOCX