ML15148A347
| ML15148A347 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 04/13/2015 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | Arizona Public Service Co |
| References | |
| Download: ML15148A347 (78) | |
Text
ES-401 PWR Examination Outline (RO) Form ES-401-2 Facility: Palo Verde Nuclear Generating Station Date of Exam: April 2015 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 4 3 3 2 18 6 Emergency &
Abnormal 2 1 1 1 N/A 2 2 N/A 2 9 4 Plant Evolutions Tier Totals 4 4 5 5 5 4 27 10 1 3 3 3 2 2 2 3 2 2 3 3 28 5 2.
Plant 2 1 1 1 2 1 1 0 1 1 0 1 10 3 Systems Tier Totals 4 4 4 4 3 3 3 3 3 3 4 38 8
- 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 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 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).
- 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 and justified; 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 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.
Rev 0
ES-401 2 Form ES-401-2 E/APE # / Name / Safety K/A Topic(s) #
Function EK1.06 Knowledge of the operational implications 000007 (BW/E02&E10; 1 of the following concepts as they apply to the CE/E02) Reactor Trip -
Stabilization - Recovery / 1 reactor trip: Relationship of emergency feedwater flow to S/G and decay heat removal following reactor trip .
(CFR 41.8 / 41.10 / 45.3)
AK2.02 Knowledge of the interrelations between the 000008 Pressurizer Vapor 2 Pressurizer Vapor Space Accident and the Space Accident / 3 following: Sensors and detectors.
(CFR 41.7 / 45.7) 2.4.6 Knowledge of EOP mitigation strategies.
000009 Small Break LOCA / 3 3
(CFR: 41.10 / 43.5 / 45.13)
EA2.09 Ability to determine or interpret the 000011 Large Break LOCA 4 following as they apply to a Large Break LOCA:
/3 Existence of adequate natural circulation.
(CFR 43.5 / 45.13)
AA1.22 Ability to operate and / or monitor the 000015/17 RCP 5 following as they apply to the Reactor Coolant Malfunctions / 4 Pump Malfunctions (Loss of RC Flow): RCP seal failure/malfunction.
(CFR 41.7 / 45.5 / 45.6) 000022 Loss of Rx Coolant Makeup / 2 AK1.01 Knowledge of the operational implications 000025 Loss of RHR 6 of the following concepts as they apply to Loss of System / 4 Residual Heat Removal System: Loss of RHRS during all modes of operation.
(CFR 41.8 / 41.10 / 45.3)
Rev 0
AK3.01 Knowledge of the reasons for the following 000026 Loss of Component 7 responses as they apply to the Loss of Component Cooling Water / 8 Cooling Water:
The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the CCWS coolers.
(CFR 41.5,41.10 / 45.6 / 45.13)
AK3.04 Knowledge of the reasons for the following 000027 Pressurizer 8 responses as they apply to the Pressurizer Pressure Control System Malfunction / 3 Pressure Control Malfunctions: Why, if PZR level is lost and then restored, that pressure recovers much more slowly.
(CFR 41.5,41.10 / 45.6 / 45.13)
EA1.12 Ability to operate and monitor the following 000029 ATWS / 1 9 as they apply to a ATWS: M/G set power supply and reactor trip breakers.
(CFR 41.7 / 45.5 / 45.6)
EK1.01 Knowledge of the operational implications 000038 Steam Gen. Tube 10 of the following concepts as Rupture / 3 they apply to the SGTR: Use of steam tables (CFR 41.8 / 41.10 / 45.3)
AK2.02 Knowledge of the interrelations between the 000040 (BW/E05; CE/E05; 11 Steam Line Rupture and the following: Sensors and W/E12) Steam Line Rupture - Excessive Heat detectors.
Transfer / 4 (CFR 41.7 / 45.7) 2.4.21 Knowledge of the parameters and logic used 000054 (CE/E06) Loss 12 to assess the status of safety functions, such as of Main Feedwater / 4 reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
(CFR: 41.7 / 43.5 / 45.12)
EA1.06 Ability to operate and monitor the following 000055 Station Blackout / 6 13 as they apply to a Station Blackout: Restoration of power with one ED/G.
(CFR 41.7 / 45.5 / 45.6)
Rev 0
AK3.01 Knowledge of the reasons for the following 000056 Loss of Off-site 14 responses as they apply to the Loss of Offsite Power / 6 Power: Order and time to initiation of power for the load sequencer.
(CFR 41.5,41.10 / 45.6 / 45.13)
AA2.19 Ability to determine and interpret the 000057 Loss of Vital AC 15 following as they apply to the Loss of Vital AC Inst. Bus / 6 Instrument Bus: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus.
(CFR: 43.5 / 45.13) 000058 Loss of DC Power /
6 AA2.01 Ability to determine and interpret the 000062 Loss of Nuclear Svc 16 following as they apply to the Loss of Nuclear Water / 4 Service Water: Location of a leak in the SWS.
(CFR: 43.5 / 45.13)
AK3.04 Knowledge of the reasons for the following 000065 Loss of Instrument 17 responses as they apply to Air / 8 the Loss of Instrument Air: Cross-over to backup air supplies (CFR 41.5,41.10 / 45.6 / 45.13)
AK2.07 Knowledge of the interrelations between 000077 Generator Voltage 18 Generator Voltage and Electric Grid Disturbances and Electric Grid Disturbances / 6 and the following: Turbine / generator control.
(CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)
Rev 0
ES-401 3 Form ES-401-2 E/APE # / Name / Safety K/A Topic(s) #
Function 000001 Continuous Rod 2.1.28 Knowledge of the purpose and function of 19 Withdrawal / 1 major system components and controls.
(CFR: 41.7) 000003 Dropped Control AK1.07 Knowledge of the operational implications 20 Rod / 1 of the following concepts as they apply to Dropped Control Rod: Effect of dropped rod on insertion limits and SDM.
(CFR 41.8 / 41.10 / 45.3) 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level AK3.05 Knowledge of the reasons for the following 21 Malfunction / 2 responses as they apply to the Pressurizer Level Control Malfunctions: Actions contained in EOP for PZR level malfunction.
(CFR 41.5,41.10 / 45.6 / 45.13) 000032 Loss of Source AK2.01 Knowledge of the interrelations between the 22 Range NI / 7 Loss of Source Range Nuclear Instrumentation and the following: Power supplies, including proper switch positions.
(CFR 41.7 / 45.7) 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator 2.2.42 Ability to recognize system parameters that 23 Tube Leak / 3 are entry-level conditions for Technical Specifications.
(CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3) 000051 Loss of Condenser Vacuum / 4 Rev 0
000059 Accidental Liquid AA1.01 Ability to operate and / or monitor the 24 RadWaste Rel. / 9 following as they apply to the Accidental Liquid Radwaste Release: Radioactive-liquid monitor.
(CFR 41.7 / 45.5 / 45.6) 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site /
8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss AA2.02 Ability to determine and interpret the 25 of CTMT Integrity / 5 following as they apply to the Loss of Containment Integrity: Verification of automatic and manual means of restoring integrity.
(CFR: 43.5 / 45.13) 000074 (W/E06&E07) Inad.
Core Cooling / 4 000076 High Reactor Coolant Activity / 9 BW/E09; CE/A13; W/E09&E10 Natural Circ. /
4 CE/A11; W/E08 RCS AA1.1 Ability to operate and / or monitor the 26 Overcooling - PTS / 4 following as they apply to the (RCS Overcooling):
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(CFR: 41.7 / 45.5 / 45.6)
CE/A16 Excess RCS AA2.2 Ability to determine and interpret the 27 Leakage / 2 following as they apply to the (Excess RCS Leakage): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.
(CFR: 43.5 / 45.13)
CE/E09 Functional Recovery Rev 0
ES-401 4 Form ES-401-2 System # / Name K/A Topic(s) #
A1.02 Ability to predict and/or monitor changes in 003 Reactor Coolant Pump 28 parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: RCP pump and motor bearing temperatures .
(CFR: 41.5 / 45.5)
A1.09 Ability to predict and/or monitor changes in 004 Chemical and Volume parameters 29 Control (to prevent exceeding design limits) associated with operating the CVCS controls including: RCS pressure and temperature.
(CFR: 41.5 / 45.5)
K1.10 Knowledge of the physical connections 005 Residual Heat Removal and/or cause-effect 30 relationships between the RHRS and the following systems: CSS (CFR: 41.2 to 41.9 / 45.7 to 45.8)
A4.05 Ability to manually operate and/or monitor in 006 Emergency Core 31 the control room: Transfer of ECCS flowpaths prior to Cooling recirculation.
(CFR: 41.7 / 45.5 to 45.8)
A3.03 Ability to monitor automatic operation of the 006 Emergency Core ECCS, including: ESFAS-operated valves. 32 Cooling (CFR: 41.7 / 45.5)
K5.02 Knowledge of the operational implications of 007 Pressurizer the following concepts as the 33 Relief/Quench Tank apply to PRTS: Method of forming a steam bubble in the PZR.
(CFR: 41.5 / 45.7) 2.1.28 Knowledge of the purpose and function of 008 Component Cooling major system components and controls. 34 Water (CFR: 41.7)
Rev 0
K3.02 Knowledge of the effect that a loss or 008 Component Cooling malfunction of the CCWS will have on the 35 Water following: CRDS.
K6.03 Knowledge of the effect of a loss or 010 Pressurizer Pressure malfunction of the following will have on the PZR 36 Control PCS: PZR sprays and heaters.
(CFR: 41.7 / 45.7)
K2.01 Knowledge of bus power supplies to the 012 Reactor Protection following: RPS channels, components, and 37 interconnections (CFR: 41.7)
A2.01 Ability to (a) predict the impacts of the 012 Reactor Protection 38 following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulty bistable operation.
(CFR: 41.5 / 43.5 / 45.3 / 45.5) 2.4.31 Knowledge of annunciator alarms, 013 Engineered Safety indications, or response procedures. 39 Features Actuation (CFR: 41.10 / 45.3)
K1.04 Knowledge of the physical connections 022 Containment Cooling 40 and/or cause/effect relationships between the CCS and the following systems: Chilled water.
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K3.01 Knowledge of the effect that a loss or 022 Containment Cooling malfunction of the 41 CCS will have on the following: Containment equipment subject to damage by high or low temperature, humidity, and pressure.
(CFR: 41.7 / 45.6)
A4.05 Ability to manually operate and/or monitor in 026 Containment Spray the control room: Containment spray reset switches. 42 (CFR: 41.7 / 45.5 to 45.8)
Rev 0
K5.05 Knowledge of the operational implications of 039 Main and Reheat the following concepts as they apply to the MRSS: 43 Steam Bases for RCS cooldown limits.
(CFR: 441.5 / 45.7)
A2.11 Ability to (a) predict the impacts of the 059 Main Feedwater following malfunctions or operations on the MFW; 44 and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of feedwater control system.
(CFR: 41.5 / 43.5 / 45.3 / 45.13)
K1.07 Knowledge of the physical connections 061 Auxiliary/Emergency and/or cause/effect relationships between the AFW 45 Feedwater and the following systems: Emergency water source.
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K2.02 Knowledge of bus power supplies to the 061 Auxiliary/Emergency following: AFW electric drive pumps. 46 Feedwater (CFR: 41.7)
A3.05 Ability to monitor automatic operation of the 062 AC Electrical ac distribution system, including: Safety-related 47 Distribution indicators and controls.
(CFR: 41.7 / 45.5)
A1.03 Ability to predict and/or monitor changes in 062 AC Electrical 48 parameters (to prevent exceeding design limits)
Distribution associated with operating the ac distribution system controls including: Effect on instrumentation and controls of switching power supplies.
(CFR: 41.5 / 45.5)
K3.01 Knowledge of the effect that a loss or 063 DC Electrical 49 malfunction of the DC electrical system will have Distribution on the following: ED/G.
(CFR: 41.7 / 45.6)
K4.02 Knowledge of ED/G system design feature(s) 064 Emergency Diesel and/or interlock(s) which provide for the following: 50 Generator Trips for ED/G while operating (normal or emergency).
(CFR: 41.7)
Rev 0
K6.07 Knowledge of the effect of a loss or 064 Emergency Diesel malfunction of the following will have on the ED/G 51 Generator system: Air receivers.
(CFR: 41.7 / 45.7)
A4.02 Ability to manually operate and/or monitor in 073 Process Radiation the control room: Radiation monitoring system 52 Monitoring control panel.
(CFR: 41.7 / 45.5 to 45.8)
K2.08 Knowledge of bus power supplies to the 076 Service Water following: ESF-actuated MOVs. 53 (CFR: 41.7) 2.4.11 Knowledge of abnormal condition 078 Instrument Air procedures. 54 (CFR: 41.10 / 43.5 / 45.13)
K4.06 Knowledge of containment system design 103 Containment 55 feature(s) and/or interlock(s) which provide for the following: Containment isolation system.
(CFR: 41.7)
Rev 0
ES-401 5 Form ES-401-2 System # / Name K/A Topic(s) #
K3.02 Knowledge of the effect that a loss or 001 Control Rod Drive 56 malfunction of the CRDS will have on the following:
RCS.
(CFR: 41.7/45.6)
K1 07 Knowledge of the physical connections 002 Reactor Coolant and/or cause-effect relationships between the RCS 57 and the following systems: Reactor vessel level indication system.
(CFR: 41.2 to 41.9 / 45.7 to 45.8) 011 Pressurizer Level Control 2.2.12 Knowledge of surveillance procedures.
014 Rod Position Indication 58 (CFR: 41.10 / 45.13)
K2.01 Knowledge of bus power supplies to the 015 Nuclear Instrumentation following: NIS channels, components, and 59 interconnections.
(CFR: 41.7) 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge A3.02 Ability to monitor automatic operation of the 033 Spent Fuel Pool Spent Fuel Pool Cooling System including: Spent 60 Cooling fuel leak or rupture.
(CFR: 41.7 / 45.5)
Rev 0
034 Fuel Handling Equipment K4.03 Knowledge of S/GS design feature(s) and/or 035 Steam Generator 61 interlock(s) which provide for the following:
Automatic blowdown and sample line isolation and reset.
(CFR: 41.7)
K5.01 Knowledge of the operational implications of 041 Steam Dump/Turbine the following concepts as they apply to the SDS: 62 Bypass Control Relationship of no-load T-ave. to saturation pressure relief setting on Valves.
(CFR: 41.5 / 45.7)
A2.17 Ability to (a) predict the impacts of the 045 Main Turbine Generator 63 following malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Malfunction of electrohydraulic control.
(CFR: 41.5 / 43.5 / 45.3 / 45.5) 055 Condenser Air Removal 056 Condensate K6.10 Knowledge of the effect of a loss or 068 Liquid Radwaste 64 malfunction on the following will have on the Liquid Radwaste System: Radiation monitors.
(CFR: 41.7 / 45.7) 071 Waste Gas Disposal 072 Area Radiation Monitoring K4.01 Knowledge of circulating water system 075 Circulating Water design feature(s) and interlock(s) which provide for 65 the following: Heat sink.
(CFR: 41.7) 079 Station Air 086 Fire Protection Rev 0
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Palo Verde Nuclear Generating Station Date of Exam: April 2015 Category K/A # Topic RO SRO-Only IR # IR #
2.1.25 Ability to interpret reference materials, such as 2.1.25 3.9 66 graphs, curves, tables, etc.
(CFR: 41.10 / 43.5 / 45.12) 1.
2.1.34 Knowledge of primary and secondary plant Conduct 2.1.34 2.7 67 chemistry limits.
of Operations (CFR: 41.10 / 43.5 / 45.12)
Subtotal (multi-unit license) Knowledge of the design, 2.2.3 3.8 68 procedural, and operational differences between units.
(CFR: 41.5 / 41.6 / 41.7 / 41.10 / 45.12) 2.
2.2.43 Knowledge of the process used to track Equipment 2.2.43 3.0 69 inoperable alarms.
Control (CFR: 41.10 / 43.5 / 45.13) 2.2.13 Knowledge of tagging and clearance 2.2.13 4.1 70 procedures.
(CFR: 41.10 / 45.13)
Subtotal 2.3.12 Knowledge of radiological safety principles 2.3.12 3.2 71 pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation 3.
areas, aligning filters, etc.
Radiation (CFR: 41.12 / 45.9 / 45.10)
Control 2.3.15 Knowledge of radiation monitoring systems, 2.3.15 2.9 72 such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
(CFR: 41.12 / 43.4 / 45.9)
Subtotal 2.4.25 Knowledge of fire protection procedures.
2.4.25 3.3 73 (CFR: 41.10 / 43.5 / 45.13) 2.4.35 Knowledge of local auxiliary operator tasks
- 4. 2.4.35 3.8 74 during an emergency and the resultant operational Emergency effects.
Procedures /
Plan (CFR: 41.10 / 43.5 / 45.13) 2.4.43 Knowledge of emergency communications 2.4.43 3.2 75 systems and techniques.
(CFR: 41.10 / 45.13)
Subtotal Tier 3 Point Total 10 7 Rev 0
ES-401 Record of Rejected K/As (RO) Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A T2/G1 Not selected yet Questions 31 and 32 are both system 006, ECCS. In addition
Rev 0
ES-401 PWR Examination Outline (SRO) Form ES-401-2 Facility: Palo Verde Nuclear Generating Station Date of Exam: April 2015 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 18 3 3 6 Emergency &
Abnormal 2 N/A N/A 9 3 1 4 Plant Evolutions Tier Totals 27 10 1 28 2 3 5 2.
Plant 2 10 2 1 3 Systems Tier Totals 38 8
- 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only 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).
- 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 and justified; 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 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.
Rev 0
ES-401 2 Form ES-401-2 E/APE # / Name / Safety K/A Topic(s) #
Function 000007 (BW/E02&E10; CE/E02) Reactor Trip -
Stabilization - Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA /
3 000011 Large Break LOCA /
3 AA2.10 Ability to determine and interpret the 000015/17 RCP Malfunctions 1 following as they apply to the Reactor Coolant
/4 Pump Malfunctions (Loss of RC Flow): When to secure RCPs on loss of cooling or seal injection.
(CFR 43.5 / 45.13) 000022 Loss of Rx Coolant Makeup / 2 2.1.7 Ability to evaluate plant performance and 000025 Loss of RHR System 2 make operational judgments based on
/4 operating characteristics, reactor behavior, and instrument interpretation.
(CFR: 41.5 / 43.5 / 45.12 / 45.13) 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control System Malfunction /
3 000029 ATWS / 1 2.4.18 Knowledge of the specific bases for EOPs.
000038 Steam Gen. Tube 3 Rupture / 3 (CFR: 41.10 / 43.1 / 45.13) 2.4.30 Knowledge of events related to system 000040 (BW/E05; CE/E05; 6 operation/status that must be reported to internal W/E12) Steam Line Rupture
- Excessive Heat Transfer organizations or external agencies, such as the
/4 State, the NRC, or the transmission system operator.
(CFR: 41.10 / 43.5 / 45.11)
Rev 0
000054 (CE/E06) Loss of Main Feedwater / 4 000055 Station Blackout / 6 AA2.72 Ability to determine and interpret the 000056 Loss of Off-site 4 following as they apply to the Loss of Offsite Power / 6 Power: Auxiliary feed flow.
(CFR: 43.5 / 45.13) 000057 Loss of Vital AC Inst.
Bus / 6 AA2.03 Ability to determine and interpret the 000058 Loss of DC Power / 6 5 following as they apply to the Loss of DC Power:
DC loads lost; impact on ability to operate and monitor plant systems.
(CFR: 43.5 / 45.13) 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 Moved #6 to system 040 000077 Generator Voltage and Electric Grid Disturbances / 6 Rev 0
ES-401 3 Form ES-401-2 E/APE # / Name / Safety K/A Topic(s) #
Function 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod AA2.04 Ability to determine and interpret the 7
/1 following as they apply to the Dropped Control Rod: Rod motion stops due to dropped rod.
(CFR: 43.5 / 45.13) 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration
/1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate AA2.08 Ability to determine and interpret the 8 This has been rejected but have not yet selected a new KA Range NI / 7 following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:
Intermediate range channel operability.
(CFR: 43.5 / 45.13) 000036 (BW/A08) Fuel 2.4.41 Knowledge of the emergency action level 9 Handling Accident / 8 thresholds and classifications.
(CFR: 41.10 / 43.5 / 45.11) 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 Rev 0
000067 Plant Fire On-site / 8 AA2.16 Ability to determine and interpret the 10 following as they apply to the Plant Fire on Site:
Vital equipment and control systems to be maintained and operated during a fire.
(CFR: 43.5 / 45.13) 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad.
Core Cooling / 4 000076 High Reactor Coolant Activity / 9 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery Rev 0
ES-401 4 Form ES-401-2 System # / Name K/A Topic(s) #
003 Reactor Coolant Pump 2.4.41 Knowledge of the emergency action level 004 Chemical and Volume 11 thresholds and classifications.
Control (CFR: 41.10 / 43.5 / 45.11) 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water A2.02 Ability to (a) predict the impacts of the 010 Pressurizer Pressure following malfunctions or operations on the PZR 12 Control PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Spray valve failures.
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 012 Reactor Protection 013 Engineered Safety Features Actuation 022 Containment Cooling A2.07 Ability to (a) predict the impacts of the 026 Containment Spray following malfunctions or operations on the CSS; 13 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: 41.5 / 43.5 / 45.3 / 45.13)
Rev 0
2.2.44 Ability to interpret control room 039 Main and Reheat Steam indications to verify the status and operation of a 14 This has been rejected but have not yet selected a new KA system, and understand how operator actions and directives affect plant and system conditions.
(CFR: 41.5 / 43.5 / 45.12) 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator 2.2.22 Knowledge of limiting conditions for 073 Process Radiation operations and safety limits. 15 Monitoring (CFR: 41.5 / 43.2 / 45.2) 076 Service Water 078 Instrument Air 103 Containment Rev 0
ES-401 5 Form ES-401-2 System # / Name K/A Topic(s) #
001 Control Rod Drive 002 Reactor Coolant 2.2.12 Knowledge of surveillance procedures.
011 Pressurizer Level 16 Control (CFR: 41.10 / 45.13) 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling A2.03 Ability to (a) predict the impacts of the 034 Fuel Handling following malfunctions or operations on the Fuel 17 Equipment Handling System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Mispositioned fuel element.
(CFR: 41.5 / 43.5 / 45.3 / 45.13)
A2.01 Ability to (a) predict the impacts of the 035 Steam Generator following malfunctions or operations on the 18 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.
(CFR: 41.5 / 43.5 / 45.3 / 45.5)
Rev 0
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 Rev 0
ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Palo Verde Nuclear Generating Station Date of Exam: April 2015 Category K/A # Topic RO SRO-Only IR # IR #
2.1.1 2.1.1 Knowledge of conduct of operations 4.2 19 requirements.
1.
Conduct (CFR: 41.10 / 45.13) of Operations 2.1.5 2.1.5 Ability to use procedures related to shift 3.9 20 staffing, such as minimum crew complement, overtime limitations, etc.
(CFR: 41.10 / 43.5 / 45.12)
Subtotal 2.2.20 2.2.20 Knowledge of the process for managing 3.8 21 troubleshooting activities.
- 2. (CFR: 41.10 / 43.5 / 45.13)
Equipment 2.2.18 2.2.18 Knowledge of the process for managing 3.9 22 Control maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.
(CFR: 41.10 / 43.5 / 45.13)
Subtotal 2.3.13 2.3.13 Knowledge of radiological safety 3.8 23 procedures pertaining to licensed operator duties, such as response to radiation monitor
- 3. alarms, containment entry requirements, fuel Radiation Control handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
(CFR: 41.12 / 43.4 / 45.9 / 45.10)
Subtotal Rev 0
2.4.40 2.4.40 Knowledge of SRO responsibilities in 4.5 24 emergency plan implementation.
4.
Emergency (CFR: 41.10 / 43.5 / 45.11)
Procedures / Plan 2.4.23 2.4.23 Knowledge of the bases for prioritizing 4.4 25 emergency procedure implementation during emergency operations.
(CFR: 41.10 / 43.5 / 45.13)
Subtotal Tier 3 Point Total 10 7 Rev 0
ES-401 Record of Rejected K/As (SRO) Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A T1/G1 2.4.30 Retained 2.4.30 but in the case of the original system 077,
- 6 040 Steam Line Generator Voltage and Electric Grid Disturbances, most events that Rupture would relate would involve receiving a communication from an outside agency about grid conditions, not us calling them.
T1/G2 Not selected yet 4.2 033 AA2.08 - PVNGS does not have Intermediate Range
- 8 Nuclear Instruments T2/G1 Intend to keep same 039 2.2.44 - This generic appears to better suited as an RO KA,
- 14 system but select an Ability to interpret CR indications .
alternate 2.2 generic Rev 0
ES-301 Administrative Topics Outline Form ES-301-1 Facility: PVNGS Date of Examination: __April 2015__
Examination Level: RO X SRO Operating Test Number: 2015 NRC Administrative Topic Type Describe activity to be performed (see Note) Code*
RM A1 - Determine Ability to Stand Shift Conduct of Operations (2.1.4 3.3/3.8)
RM A2 - Shutdown Margin Calculation Conduct of Operations (2.1.37 4.3/4.6; 2.1.20 4.6/4.6)
RM A3 - Technical Review of a Tag Assignment Sheet Equipment Control (2.2.13 4.1/4.3)
RM A4 - Perform RO Radiological Tasks Radiation Control (2.3.13 3.4/3.8)
N/A N/A - This Topic not selected for ROs 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 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)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
Page 1 of 2 Rev. 1
2015 NRC Exam RO Admin JPM Summary A1, Determine ability to stand shift with training and license proficiency: This JPM requires the RO to determine if he/she meets the requirements for watchstanding proficiency, including participation in LOCT, in accordance with 40DP-9OP02, Conduct of Shift Operations. The JPM is modified from the original JPM (2012 NRC A-1) in that previous watchstanding dates and hours are changed and participation in training has also changed. In 2012, the reason for not being able to stand a watch was not meeting proficiency requirements. In 2015, the reason is not being current in training.
A2, Shutdown Margin Calculation: This JPM requires the RO to calculate a Shutdown Margin (SDM) in accordance with 72ST-9RX14, Shutdown Margin, Modes 3, 4 and 5; and the Unit 1 Core Data Book. This JPM is modified from the original JPM (from 2010 NRC Exam) in that parameters, such as current boron concentration and time in core life, have changed. Additionally, a planned cooldown to 500°F was added to the Initial Conditions, which now requires the applicant to determine, rather than be given, the Most Conservative Tcold. The number of stuck rods was increased from 1 to 2, which impacts the Acceptance Criteria for the Xenon Adjusted Required Boron Concentration from met to not met.
A3, Technical Review of a Tag Assignment Sheet: This JPM requires the RO to perform a technical review of a Tag Assignment Sheet and identify errors. This JPM is modified from the original JPM (2009 NRC Exam) in that the induced errors, such as positions of valves and required tags, have been changed.
A4, Perform RO Radiological Tasks: This JPM requires the RO to review given conditions and determine dose received for a task, required authorization for that dose, and posting requirements for the area where the task will be performed; in accordance with 75DP-9RP01, Radiation Exposure and Access Control, and 75DP-0RP01, Radiological Posting and Labeling. This JPM is modified from the original JPM (2013 NRC Exam) in that the exposures, the required approval, and the posting are all different.
Emergency Procedures/Plan Topic not selected for ROs Page 2 of 2 Rev. 1
ES-301 Administrative Topics Outline Form ES-301-1 Facility: PVNGS Date of Examination: __April 2015__
Examination Level: RO SRO X Operating Test Number: 2015 NRC Administrative Topic Type Describe activity to be performed (see Note) Code*
RDP A5 - Ensure Compliance with Fatigue Rule Program Conduct of Operations (2.1.5 2.9/3.9)
RM A6 - Review Shutdown Margin Calculation Conduct of Operations (2.1.37 4.3/4.6)
(2.1.20 4.6/4.6)
RM A7 - Review Surveillance Test Equipment Control (2.2.12 3.7/4.1)
RM A8 - Perform SRO Radiological Tasks Radiation Control (2.3.13 3.4/3.8)
RD A9 - Classify Event and Make PARs Emergency Procedures/Plan (2.4.41 2.9/4.6)
(2.4.44 2.4/4.4)
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)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
Page 1 of 2 Rev. 1
2015 NRC Exam SRO Admin JPM Summary A5 - Ensure Compliance with Fatigue Rule Program: This JPM requires the SRO to determine if crew members can assume the shift while meeting the fatigue requirements outlined in 01DP-0AP17, Managing Personal Fatigue. The JPM was randomly selected from the previous two NRC exams (2013) at PVNGS.
A6 - Review Shutdown Margin Calculation: This JPM requires the SRO to review a Shutdown Margin (SDM) calculation for accuracy in accordance with 72ST-9RX14, Shutdown Margin, Modes 3, 4 and 5; and the Unit 2 Core Data Book. This JPM is modified from the original JPM (2010 NRC A-5) in that induced errors have changed.
A7 - Review Surveillance Test: This JPM requires the SRO to perform a technical review of a surveillance, Inoperable Power Sources Action Statement, and identify errors. This JPM is modified from the original JPM (2008 NRC SA3) in that the induced errors, such as transmission lines capable of power transmission, acceptance criteria, and operable redundant equipment, have been changed.
A8 - Perform SRO Radiological Tasks: This JPM requires the SRO to review given conditions and determine dose received for a task, required authorization for that dose, and determine who makes the entry; in accordance with 75DP-9RP01, Radiation Exposure and Access Control, and 75DP-0RP01, Radiological Posting and Labeling. This JPM is modified from the original JPM (2012 NRC A8) in that all of the dose values have changed and the individual to perform the task is different.
A9 - Classify Event and Make PARs: This JPM requires the SRO to review given conditions and determine the Emergency Action Level in accordance with EP-0900, Appendix L, and EP-0901, Classifications. It also requires the SRO to make Protective Action Recommendations in accordance with EP-0905, Protective Actions. This is modified Bank JPM EP008-CR-009. The modifications include requiring the use of Met Tower Data, requiring identification of Potentially Affected Sectors, and adding the Emergency Exposure Limit for Life-Saving.
Page 2 of 2 Rev. 1
PVNGS License Examination PVNGS Form ES301-2 Control Room/In-Plant Systems Outline Facility: PVNGS_ Date of Examination: 4/13/2015 Exam Level: RO Operating Test No.: 2015 NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)
JPM System/JPM Title Type Code* Safety
- Function S1 CIAS Initiation and Verification S N A L EN 5 (40EP-9EO03, Loss of Coolant Accident, Steps 13 & 14))
3.5 103 A3.01 3.9/4.2 S2 Perform BDAS Alarm Check SLD 7 (40EP-9EO10, Standard Appendix 8) 3.7 015 A3.03 3.9/3.9 S3 Reenergize NBN-X02 (Non ESF Transformer) and NBN-S02 SN 6 (Non-classs 4160v bus) 3.6 062 A4.07 3.1/3.1 (40OP-9NB01, 4.16 Non-Class 1E Power (NB))
S4 Place Containment Refueling Purge Subsystem in Service SNL 8 (40OP-9CP01, Containment Purge System) 3.8 029 A2.03 2.7/3.1 S5 Respond to a Pressurizer Pressure Instrument Failure S MA 3 (40AL-9RK4A, Panel B04A Alarm Responses ) 4.2 027 A1.01 4.0/3.9 S6 Reset Inadvertent MSIS SDP 2 (40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations) 3.2 013 A4.01 4.5/4.8 S7 Reconnect and Reset the Steam Bypass Control System SD 4S (40OP-9SF05, Operation of Steam Bypass Control System) 3.4 041 A4.08 3.0/3.1 S8 Borate the RCS SAD 1 (40OP-9CH01, CVCS Normal Operations, Section 6.33) 3.1 004 A4.07 3.9/3.7 In-Plant Systems (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
P1 Energize PBB-S04 from EDG B ADE 8 4.2 068 AA1.10 3.7/3.9 (40AO-9ZZ19, Control Room Fire)
P3 Align Charging Pump Discharge to Hot Leg Injection Train A NER 1 HPSI 4.4 A16 AA1.1 3.4/3.6 (40EP-9EO10, Standard Appendix 208, Attachment 208-A)
P2 Reset Overspeed Trip on AFA-P01 ADE 4S (40EP-9EO10, Standard Appendix 38, Attachment 38-A) 3.4 061 A2.04 3.4/3.8 Rev 0 Page 1
PVNGS License Examination PVNGS Form ES301-2 Control Room/In-Plant Systems Outline
@ 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)ngineered 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 Rev 0 Page 2
PVNGS License Examination Control Room/In-Plant Systems Outline JPM Summary:
S1, CIAS Actuation/Verification: This JPM will be conducted simultaneously with S2. After a LOCA a CIAS will fail to initiate. A CIAS will be manually initiated and two sets of containment isolation valves will fail to automatically close. The RO is directed to perform steps 13 and 14 from the LOCA procedure. After verifying a CIAS should have actuated, the RO will manually initiate a CIAS. The next step is to ensure one isolation per containment penetration is closed. Two penetrations will have both valves fail to close and the RO will isolate at least one valve in each penetration. This is a NEW JPM. There are 3 critical steps in this JPM.
S2, BDAS Alarm Check: This JPM will be conducted simultaneously with S1. After an inadvertent Reactor Trip, an RCS leak will develop and a CSAS is manually initiated on trend. The BOP is directed to perform Appendix 8, Boron Dilution Alarm Check, of the Standard Appendices. This is a Time Critical JPM because Boron Dilution Alarms must be confirmed operable within one hour after neutron flux is within the start-up range following a reactor shutdown. There are 11 critical steps in this JPM.
S3, Reenergize NBN-X02 (Non ESF Transformer and NBN-S02 (Non-Class 4160 v bus): This JPM will be conducted simultaneously with S4. The plant is in Mode 5 when the repairs to the 4160 bus NBN-S02 are completed. The RO is directed to restore power to NBN-X02 and NBN-S02 in accordance with 40OP-9NB01, 4.16 kV Non-Class 1E Power (NB), Section 6.6, Energizing 4.16 kV NBN-S02, beginning with Step 6.6.2.10.
This is a NEW JPM. There are 3 critical steps in this JPM.
S4, Place Containment Refueling Purge Subsystem in Service: This JPM will be conducted simultaneously with S3. The plant is in Mode 5 when the BOP is directed to place the Containment Refueling Purge Subsystem in service in accordance with 40OP-9CP01, Containment Purge System, Section 7.0, Placing the Containment Refueling Purge Subsystem in Service with Power to CPA-2A/2B and CPB-3A/3B, beginning with Step 7.3.10.
This is a NEW JPM. There are 3 critical steps in this JPM.
S5, Respond to a Pressurizer Pressure Instrument Failure: This JPM will be conducted simultaneously with S6. The plant is at power when and inadvertent MSIS occurs andPT-100X, Pressurizer Pressure Control Transmitter, fails LOW. HS-100, PPCS Selector Switch, fails in the X position (fails to transfer to the Y position). This results in an actual high pressure condition, as the heaters energize and the spray valves close.
The RO is directed to restore Pressurizer pressure in accordance with 40AL-9RK4A, Panel B04A Alarm Responses, window 4A01B, Group B. This is an Alternate Path JPM because the main spray valves will not open and the RO must initiate auxiliary spray to reduce RCS pressure. This is a modified JPM because to the failure of the main spray valves. There are 3 critical steps in this JPM.
S6, Reset Inadvertent MSIS: This JPM will be conducted simultaneously with S5. It was randomly selected from among the 2012 and 2013 NRC Exam systems and controls JPMs using a random generator on an Excel spreadsheet. The plant is at power when and inadvertent MSIS occurs and PT-100X, Pressurizer Pressure Control Transmitter, fails LOW. The BOP is directed to reset the inadvertent MSIS in accordance with 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations, Appendix B, PPS-ESFAS Reset. There are 5 critical steps in this JPM.
S7, Reconnect and Reset the Steam Bypass Control System: This JPM will be conducted simultaneously with S8. It is an RO Only JPM. The plant is at power with the Steam Bypass Control System (SBCS) in Manual. The BOP is directed to reconnect and reset the SBCS in accordance with 40OP-9SF05, Operation of Rev 0 Page 3
PVNGS License Examination Control Room/In-Plant Systems Outline Steam Bypass Control System, Appendix C, Connecting and Resetting Steam Bypass Control System. There are 5 critical steps in this JPM.
S8, Borate the RCS: This JPM will be conducted simultaneously with S7. It is an RO Only JPM. The RO is directed to borate the RCS in accordance with 40OP-9CH01, CVCS Normal Operations, Section 6.33. This is an Alternate Path JPM because the Boric Acid Filter clogs and the RO must use a different method to borate in accordance with the alarm response procedure. There are 5 critical steps in this JPM.
P1, Energize PBB-S04 with EDG B: The Area Operator is directed to perform the actions of Appendix E of 40AO-9ZZ19, Control Room Fire. This Appendix separates the controls for the B DG and 4160 SWGR breakers from the control room and starts EDG B to supply PBB-S04. This is an Alternate Path JPM in that the DG breaker will not close electrically and the operator must go the contingency step and manually close the breaker. This is a Time Critical JPM because the operator must close the DG breaker and start the Spray Pond pump to supply cooling to the DG within a 15 minute period. This JPM has 13 critical steps.
P2, Align Charging Pump Discharge to Hot Leg Injection: The Area Operator is directed to align charging pump discharge to Hot Loeg Injection Train A HPSI in accordance with 40EP-9EO10, Standard Appendix 208, 08-A. This is a NEW JPM. There are 3 critical steps in this JPM.
P3, Reset Overspeed Trip on AFA-P01: The Area Operator is directed to reset an overspeed trip on the Turbine Driven Auxiliary Feedwater Pump in accordance with 40EP-9EO10, Standard Appendix 38, 8-A. This JPM is the 4th most significant Key Operator Action in the PVNGS PRA. This is an Alternate Path JPM because the Latch Lever and the Trip Hook are not aligned, requiring the Operator to close AFA-HV-54 (T&TV) in accordance with Contingency Action1.1. There are 2 critical steps in this JPM.
Rev 0 Page 4
PVNGS License Examination PVNGS Form ES301-2 Control Room/In-Plant Systems Outline Facility: PVNGS_ Date of Examination: 4/13/2015 Exam Level: SRO-I Operating Test No.: 2015 NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)
JPM System/JPM Title Type Code* Safety
- Function S1 CIAS Initiation and Verification S N A L EN 5 (40EP-9EO03, Loss of Coolant Accident, Steps 13 & 14))
3.5 103 A3.01 3.9/4.2 S2 Perform BDAS Alarm Check SLD 7 (40EP-9EO10, Standard Appendix 8) 3.7 015 A3.03 3.9/3.9 S3 Reenergize NBN-X02 (Non ESF Transformer) and NBN-S02 SN 6 (Non-classs 4160v bus) 3.6 062 A4.07 3.1/3.1 (40OP-9NB01, 4.16 Non-Class 1E Power (NB))
S4 Place Containment Refueling Purge Subsystem in Service SNL 8 (40OP-9CP01, Containment Purge System) 3.8 029 A2.03 2.7/3.1 S5 Respond to a Pressurizer Pressure Instrument Failure S MA 3 (40AL-9RK4A, Panel B04A Alarm Responses ) 4.2 027 A1.01 4.0/3.9 S6 Reset Inadvertent MSIS SDP 2 (40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations) 3.2 013 A4.01 4.5/4.8 S9 Isolate a Ruptured SG SDAL 4P (40EP-9EO10 Standard Appendix 114) 3.2 006 A2.02 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
P1 Energize PBB-S04 from EDG B ADE 8 4.2 068 AA1.10 3.7/3.9 (40AO-9ZZ19, Control Room Fire)
P2 Align Charging Pump Discharge to Hot Leg Injection Train A NER 1 HPSI 4.4 A16 AA1.1 3.4/3.6 (40EP-9EO10, Standard Appendix 208, Attachment 208-A)
P3 Reset Overspeed Trip on AFA-P01 ADE 4S (40EP-9EO10, Standard Appendix 38, Attachment 38-A) 3.4 061 A2.04 3.4/3.8
@ 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.
Rev 0 Page 1
PVNGS License Examination PVNGS Form ES301-2 Control Room/In-Plant Systems Outline
- 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)ngineered 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 Rev 0 Page 2
PVNGS License Examination Control Room/In-Plant Systems Outline JPM Summary:
S1, CIAS Actuation/Verification: This JPM will be conducted simultaneously with S2. After a LOCA a CIAS will fail to initiate. A CIAS will be manually initiated and two sets of containment isolation valves will fail to automatically close. The RO is directed to perform steps 13 and 14 from the LOCA procedure. After verifying a CIAS should have actuated, the RO will manually initiate a CIAS. The next step is to ensure one isolation per containment penetration is closed. Two penetrations will have both valves fail to close and the RO will isolate at least one valve in each penetration. This is a NEW JPM. There are 3 critical steps in this JPM.
S2, BDAS Alarm Check: This JPM will be conducted simultaneously with S1. After an inadvertent Reactor Trip, an RCS leak will develop and a CSAS is manually initiated on trend. The BOP is directed to perform Appendix 8, Boron Dilution Alarm Check, of the Standard Appendices. This is a Time Critical JPM because Boron Dilution Alarms must be confirmed operable within one hour after neutron flux is within the start-up range following a reactor shutdown. There are 11 critical steps in this JPM.
S3, Reenergize NBN-X02 (Non ESF Transformer and NBN-S02 (Non-Class 4160 v bus): This JPM will be conducted simultaneously with S4. The plant is in Mode 5 when the repairs to the 4160 bus NBN-S02 are completed. The RO is directed to restore power to NBN-X02 and NBN-S02 in accordance with 40OP-9NB01, 4.16 kV Non-Class 1E Power (NB), Section 6.6, Energizing 4.16 kV NBN-S02, beginning with Step 6.6.2.10.
This is a NEW JPM. There are 3 critical steps in this JPM.
S4, Place Containment Refueling Purge Subsystem in Service: This JPM will be conducted simultaneously with S3. The plant is in Mode 5 when the BOP is directed to place the Containment Refueling Purge Subsystem in service in accordance with 40OP-9CP01, Containment Purge System, Section 7.0, Placing the Containment Refueling Purge Subsystem in Service with Power to CPA-2A/2B and CPB-3A/3B, beginning with Step 7.3.10.
This is a NEW JPM. There are 3 critical steps in this JPM.
S5, Respond to a Pressurizer Pressure Instrument Failure: This JPM will be conducted simultaneously with S6. The plant is at power when and inadvertent MSIS occurs andPT-100X, Pressurizer Pressure Control Transmitter, fails LOW. HS-100, PPCS Selector Switch, fails in the X position (fails to transfer to the Y position). This results in an actual high pressure condition, as the heaters energize and the spray valves close.
The RO is directed to restore Pressurizer pressure in accordance with 40AL-9RK4A, Panel B04A Alarm Responses, window 4A01B, Group B. This is an Alternate Path JPM because the main spray valves will not open and the RO must initiate auxiliary spray to reduce RCS pressure. This is a modified JPM because to the failure of the main spray valves. There are 3 critical steps in this JPM.
S6, Reset Inadvertent MSIS: This JPM will be conducted simultaneously with S5. It was randomly selected from among the 2012 and 2013 NRC Exam systems and controls JPMs using a random generator on an Excel spreadsheet. The plant is at power when and inadvertent MSIS occurs and PT-100X, Pressurizer Pressure Control Transmitter, fails LOW. The BOP is directed to reset the inadvertent MSIS in accordance with 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations, Appendix B, PPS-ESFAS Reset. There are 5 critical steps in this JPM.
Rev 0 Page 3
PVNGS License Examination Control Room/In-Plant Systems Outline S9, Isolate a Rupted SG: This JPM is an SRO Only JPM. The RO is directed to isolate SG#2 due to a SG Tube Rupture. This is an Alternate Path JPM because the Downcomer Isolation valves will not close and the RO must perform the contingency to isolate other vavles. There are 4 critical steps in this JPM.
P1, Energize PBB-S04 with EDG B: The Area Operator is directed to perform the actions of Appendix E of 40AO-9ZZ19, Control Room Fire. This Appendix separates the controls for the B DG and 4160 SWGR breakers from the control room and starts EDG B to supply PBB-S04. This is an Alternate Path JPM in that the DG breaker will not close electrically and the operator must go the contingency step and manually close the breaker. This is a Time Critical JPM because the operator must close the DG breaker and start the Spray Pond pump to supply cooling to the DG within a 15 minute period. This JPM has 13 critical steps.
P2, Align Charging Pump Discharge to Hot Leg Injection: The Area Operator is directed to align charging pump discharge to Hot Loeg Injection Train A HPSI in accordance with 40EP-9EO10, Standard Appendix 208, 08-A. This is a NEW JPM. There are 3 critical steps in this JPM.
P3, Reset Overspeed Trip on AFA-P01: The Area Operator is directed to reset an overspeed trip on the Turbine Driven Auxiliary Feedwater Pump in accordance with 40EP-9EO10, Standard Appendix 38, 8-A. This JPM is the 4th most significant Key Operator Action in the PVNGS PRA. This is an Alternate Path JPM because the Latch Lever and the Trip Hook are not aligned, requiring the Operator to close AFA-HV-54 (T&TV) in accordance with Contingency Action1.1. There are 2 critical steps in this JPM.
Rev 0 Page 4
PVNGS License Examination PVNGS Form ES301-2 Control Room/In-Plant Systems Outline Facility: PVNGS_ Date of Examination: 4/13/2015 Exam Level: SRO-U Operating Test No.: 2015 NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)
JPM System/JPM Title Type Code* Safety
- Function S1 CIAS Initiation and Verification S N A L EN 5 (40EP-9EO03, Loss of Coolant Accident, Steps 13 & 14))
3.5 103 A3.01 3.9/4.2 S2 Perform BDAS Alarm Check SLD 7 (40EP-9EO10, Standard Appendix 8) 3.7 015 A3.03 3.9/3.9 S9 Isolate a Ruptured SG SDAL 4P (40EP-9EO10 Standard Appendix 114) 3.2 006 A2.02 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
P1 Energize PBB-S04 from EDG B ADE 8 4.2 068 AA1.10 3.7/3.9 (40AO-9ZZ19, Control Room Fire)
P2 Align Charging Pump Discharge to Hot Leg Injection Train A NER 1 HPSI 4.4 A16 AA1.1 3.4/3.6 (40EP-9EO10, Standard Appendix 208, Attachment 208-A)
@ 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)ngineered 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 Rev 0 Page 1
PVNGS License Examination Control Room/In-Plant Systems Outline JPM Summary:
S1, CIAS Actuation/Verification: This JPM will be conducted simultaneously with S2. After a LOCA a CIAS will fail to initiate. A CIAS will be manually initiated and two sets of containment isolation valves will fail to automatically close. The RO is directed to perform steps 13 and 14 from the LOCA procedure. After verifying a CIAS should have actuated, the RO will manually initiate a CIAS. The next step is to ensure one isolation per containment penetration is closed. Two penetrations will have both valves fail to close and the RO will isolate at least one valve in each penetration. This is a NEW JPM. There are 3 critical steps in this JPM.
S2, BDAS Alarm Check: This JPM will be conducted simultaneously with S1. After an inadvertent Reactor Trip, an RCS leak will develop and a CSAS is manually initiated on trend. The BOP is directed to perform Appendix 8, Boron Dilution Alarm Check, of the Standard Appendices. This is a Time Critical JPM because Boron Dilution Alarms must be confirmed operable within one hour after neutron flux is within the start-up range following a reactor shutdown. There are 11 critical steps in this JPM.
S9, Isolate a Rupted SG: This JPM is an SRO Only JPM. The RO is directed to isolate SG#2 due to a SG Tube Rupture. This is an Alternate Path JPM because the Downcomer Isolation valves will not close and the RO must perform the contingency to isolate other vavles. There are 4 critical steps in this JPM.
P1, Energize PBB-S04 with EDG B: The Area Operator is directed to perform the actions of Appendix E of 40AO-9ZZ19, Control Room Fire. This Appendix separates the controls for the B DG and 4160 SWGR breakers from the control room and starts EDG B to supply PBB-S04. This is an Alternate Path JPM in that the DG breaker will not close electrically and the operator must go the contingency step and manually close the breaker. This is a Time Critical JPM because the operator must close the DG breaker and start the Spray Pond pump to supply cooling to the DG within a 15 minute period. This JPM has 13 critical steps.
P2, Align Charging Pump Discharge to Hot Leg Injection: The Area Operator is directed to align charging pump discharge to Hot Loeg Injection Train A HPSI in accordance with 40EP-9EO10, Standard Appendix 208, 08-A. This is a NEW JPM. There are 3 critical steps in this JPM.
Rev 0 Page 2
ES-301 Transient and Event Checklist (Rev. 0) Form ES-301-5 Facility: PVNGS Date of Exam: April, 2015 Operating Test No.:
A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW CREW CREW T N I T POSITION POSITION POSITION POSITION A I
M C S A B S A B S A B S A B L U A T R T O R T O R T O R T O M(*)
N Y O C P O C P O C P O C P R I U T P E
RO RX 0 1 1 0 NOR 1 1 1 1 1 R1, R3, I/C 2,3,8 3,4,6 2,3,6, 10 4 4 2 R4, R6, 7 R7, R8 MAJ 7,9 5 7 4 2 2 1 TS 0 0 2 2 RX 0 1 1 0 RO NOR 0 1 1 1 I/C 2,3,8 3,4,6 6 4 4 2 MAJ 7,9 5 3 2 2 1 R2, R5 TS 0 0 2 2 RX 0 1 1 0 RO NOR 1 1 2 1 1 1 I/C 5,6 2,4, 5 4 4 2 6
R9, R11, R12, MAJ 7,9 5 3 2 2 1 R13 TS 0 0 2 2 RO RX 0 1 1 0 NOR 1 1 1 3 1 1 1 R10 I/C 5,6 2,4, 2,3,6, 9 4 4 2 6 7 MAJ 7,9 5 7 4 2 2 1 TS 0 0 2 2
Facility: PVNGS Date of Exam: April, 2015 Operating Test No.:
A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW CREW CREW T N I T POSITION POSITION POSITION POSITION A I
M C S A B S A B S A B S A B L U A T R T O R T O R T O R T O M(*)
N Y O C P O C P O C P O C P R I U T P E
SRO-I RX 0 1 1 0 NOR 1 1 2 1 1 1 I1, I2, I3, I/C 2,3,5, 2,3,4, 4,5,6, 13 4 4 2 I4, I5, I6, 6,8 6 8 I7 MAJ 7,9 5 7 4 2 2 1 TS 4,6 2,4 4 0 2 2 RX 0 1 1 0 NOR 1 1 1 3 1 1 1 SRO-I I/C 5,6 2,4, 2,3,4, 11 4 4 2 6 5,6,8 MAJ 7,9 5 7 4 2 2 1 I8, I9, I10 TS 3,4 2 0 2 2 RX 0 1 1 0 SRO-U NOR 1 1 2 1 1 1 I/C 2,3,5, 2,3,4, 11 4 4 2 6,8 5,6,8 U-2 MAJ 7,9 7 3 2 2 1 TS 4,6 3,4 4 0 2 2 SRO-U RX 1 1 0 NOR 1 1 2 1 1 1 U3 I/C 2,3,4, 2,3,4, 9 4 4 2 6 5,6,8 MAJ 5 7 2 2 2 1 TS 2,4 3,4 4 0 2 2 SRO-U RX 0 1 1 0 NOR 1 1 1 1 1 U1, U4 I/C 2,3,4, 6 4 4 2 5,6,8 MAJ 7 1 2 2 1 TS 3,4 2 0 2 2
Instructions:
- 1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
- 2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.
- 3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
Appendix D Scenario Outline Form ES-D-1 Facility: PVNGS Scenario No.: 1 (Rev. 0) Op-Test No: NRC - 2015 Examiners: Operators:
Initial Conditions: (100% power, MOC).
Turnover: See attached.
Event Malf. No. Event Event Description No. Type*
1 N/A N Place Reactor Power Cutback System in service in accordance with 40OP-BOP/SRO 9SF04, Operation of the Reactor Power Cutback System, Sections 6.1.1 and 6.1.2.
2 cmCNCV01CHEPDIC240_2 C CHN-PDIC-240, Charging Line to RC Loop 2A DP Control, fails LOW in the RO/SRO AUTO Mode. Crew responds in accordance with 40AL-9RK3A, window 3A8A (CHG HDR SYS TRBL) and window 3A11B (RCP SEAL INJ FLOW HI-HI OR LO).
3 mfAN_1B01A4 C Auxiliary Transformer High Temperature. Crew responds in accordance with RO/SRO 40AL-9RK1B, window 1B01A, UNIT AUX XFMR X02 PROT TRIP/TRBL and 40AL-9MA01, UNIT AUX TRANSFORMER MAN-X02, Group H, High Winding Temp. Crew transfers loads from the Unit Aux Transformer in accordance with 40OP-9NA03, 13.8 kV Electrical System (NA), Section 7.0 and 11.0.
4 cmTRMC04CTBLT36_4 SRO CTB-LT-36, CST Level Transmitter, fails LOW. Crew responds in accordance (TS) with 40AL-9RK6A, windows 6A15B (CST AT MINIMUM OPERATING LEVEL) and 6A15C (CST EMPTY).
[LCO 3.3.10 Condition A]
5 mfFW17B C B MFP trips, requiring the CRS to enter 40AO-9ZZ09, Reactor Power BOP/SRO Cutback (Loss of Feedpump).
(AOP) 6 mfRP06L1 C Inadvertent AFAS Train B (AFB-P01 86 Lockout). CRS implements 40AO-mfRP06L2 BOP/SRO 9ZZ17, Inadvertent PPS-ESFAS Actuations, Section 3.0, AFAS.
cmCPFW07AFBP01_6 (AOP/TS) [LCO 3.7.5 Condition C]
7 mfMC01A M Loss of condenser vacuum. (Trip Initiator) CRS implements 40AO-9ZZ07, ALL Loss of Vacuum. Crew should initiate a manual Reactor trip and enter 40EP-9EO01, Standard Post Trip Actions. CRS transitions to 40EP-9EO02, Reactor Trip.
8 cmCNRC03RCNPIC100_2 C RCN-PIC-100, Pressurizer Master Controller, fails to 100% output in the RO/SRO AUTO mode. RO responds in accordance with 40AL-9RK4B, window 4B1B (PZR PRESS HI-LO).
CRITICAL TASK - Close the Pressurizer Spray Valves before a SIAS occurs at 1837 psig.
9 cmCPFW07AFNP01_6 M Loss of Feedwater ALL 86 Lockout of AFN-P01 (at Step 9 of 40EP-9EO02). CRS ultimately transitions to 40EP-9EO09, Functional Recovery, to establish feedwater from the condensate pumps using Standard Appendix 44, Feeding with the Condensate Pumps. (May initially transition to 40EP-9EO06, Loss of All Feedwater).
End The scenario may be ended once the selected Steam Generator is being fed to at point a rate that raises SG level, and/or lowers/stabilizes RCS temperature, OR when deemed appropriate by the Lead Examiner.
CRITICAL TASK - Establish feedwater to at least one SG from the Condensate Pumps prior to dryout of the selected/depressurized SG.
NUREG-1021, Rev 9, Supp 1 1 of 8 Rev 0
Appendix D Scenario Outline Form ES-D-1
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes
- 1. Total malfunctions (5-8) 9
- 2. Malfunctions after EOP entry (1-2) 2
- 3. Abnormal events (2-4) 2
- 4. Major transients (1-2) 2
- 5. EOPs entered/requiring substantive actions (1-2) 1
- 6. EOP contingencies requiring substantive actions (0-2) 1
- 7. Critical tasks (2-3) 2 CRITICAL TASK JUSTIFICATION Close the Pressurizer Spray Valves before a SIAS occurs at Failure to close the Pressurizer Spray Valves prior to RCS pressure 1837 psig. lowering to less than 1837 psig will result in a loss of subcooling, which could require securing all RCPs, unnecessarily complicating recovery strategies. Additionally, allowing a SIAS to unnecessarily actuate will also complicate mitigation strategies, as the crew will be required to shutdown unneeded equipment while implementing the FRP.
Establish feedwater flow from the Condensate Pumps and Failure to prevent dry-out in a SG leads to unnecessary feed the selected/depressurized SG prior dryout of the complications in recovery strategy. When SG mass is reduced selected SG. below 5000 lbm (see FSAR Section 15.2.8.2.3, part of Decrease in Heat Removal By the Secondary System), feedwater flow to that SG must be limited to prevent thermal shock, slowing recovery efforts.
Standard Appendix 44, Feeding with the Condensate Pumps, Step 14.d (and 15.d), limits feed flow rate to 1000 gpm if a SG is dry.
Excessive feedwater flow to a hot, dry SG can lead to structural damage to SG components, limiting the ability of the SG to remove heat from the RCS. According to 40OP-9SG02, Operating the SGs, Precaution and Limitation 3.7, there are about 16,000 gallons of water in the SG at 0%WR level.
NUREG-1021, Rev 9, Supp 1 2 of 8 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 1 Overview Event 1 The BOP operator places the Reactor Power Cutback System (RPCS) in service in accordance with 40OP-9SF04, Operation of the Reactor Power Cutback System, Sections 6.1.1 and 6.1.2. This BOP, normal evolution involves testing the cutback circuits at the RPCS module and selecting the appropriate subgroups.
After the subgroups have been selected for LOSS OF FEED PUMP, or at the discretion of the Lead Examiner, the next event can be initiated.
Event 2 CHN-PDIC-240, Charging Line to RC Loop 2A DP Control, fails LOW in the AUTO Mode. The crew is alerted by the following:
o 3A08A (CHG HDR SYS TRBL) o 3A11B (RCP SEAL INJ FLOW HI-HI OR LO)
- This is a reverse-acting controller in that the actual DP goes low when the controller fails to 100% output.
Crew responds in accordance with 40AL-9RK3A for 3A08A and 3A11B. Window 8A, Group C (PT ID CHPDS240) directs the RO to take manual control of CHN-PDIC-240 and raise the DP to between 90 and 135 psid. The actions for window 11B require the RO to adjust affected RCP seal injection controllers and/or CHN-PDIC-240 to achieve charging header pressure between 2430 and 2500 psig and RCP seal injection flow between 6.0 and 7.5 gpm. The CRS may refer to 40AO-9ZZ04, Reactor Coolant Pump Emergencies.
When seal injection flow and charging header pressure are adjusted, or at the discretion of the Lead Examiner, the next event can be initiated.
Event 3 Unit Auxiliary Transformer High Temperature. The crew is alerted by the following:
- Annunciator 1B01A (UNIT AUX XFMR X02 PROT TRIP/TRBL)
- Computer Point ID MAYS57, Unit Aux Xfmr MAN-X02 Trouble Crew responds in accordance with 40AL-9RK1B, window 1B01A (Point ID MAYS57, Unit Aux Xfmr MAN-X02 Trouble). The RO directs an Area Operator to locally investigate the trouble alarm. The AO uses 40AL-9MA01, UNIT AUX TRANSFORMER MAN-X02, Group H, High Winding Temp. The AO reports that winding temperature is 125°C and rising slowly. The AO also reports that all fans and oil pumps are operating. In accordance with Operator Action 4, the AO recommends a reduction in Unit Aux Xfmr load or a transfer to the alternate power source. 40AL-9RK1B, Point ID MAYS57, Unit Aux Xfmr MAN-X02 Trouble, Operator Action 4 prompts the crew to transfer bus NAN-S01 to NAN-S03 and bus NAN-S02 to NAN-S04, then refer to 40OP-9NA03, 13.8 kV Electrical System (NA), Section 7.0 and 11.0. Transfer actions involve placing the Synchronizing Switch to ON, closing the associated tie breaker, ensuring the supply breaker opens, checking for proper voltage, and turning off the Synchronizing Switch. When loads have been transferred, the AO reports that winding temperature on the Auxiliary Transformer is NUREG-1021, Rev 9, Supp 1 3 of 8 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 1 Overview lowering slowly.
After the report from the AO, or at the discretion of the Lead Examiner, the next event can be initiated.
Event 4 CTB-LT-36, CST Level Transmitter, fails LOW. The crew is alerted by the following:
- Annunciator 6A15B (CST AT MINIMUM OPERATING LEVEL)
- Annunciator 6A15C (CST EMPTY)
- Level indication on CTB-LI-36A Crew responds in accordance with 40AL-9RK6A for windows 15B (CST AT MINIMUM OPERATING LEVEL) and 15C (CST EMPTY). This is a TS call for the CRS. First Priority Operator Action 1 of 40AL-9RK6A, window 6A15B (and window 6A15C) requires the crew to check for a valid alarm by comparing CTA-LI-35A and CTB-LI-36A. The crew diagnoses the LOW failure of CTB-LT-36B/LI-36A. If an Area Operator is dispatched, the AO will report that the local level reading is 40 feet on LI-22.
Since actual CST level is approximately 40 feet, the LCO for TS 3.7.6 ( 29.5 ft) is met.
Since 2 channels of CST level indication are the required for Function 12 of Table 3.3.10-1, the CRS applies Condition A, for One or more Functions with one required channel inoperable. TS Bases for LCO 3.3.10, Function 12, cites CTA-LT-35 and CTB-LT-36 as the instruments required for the LCO.
After the CRS briefs the crew on the entry into LCO 3.3.10, Condition A, or at the discretion of the Lead Examiner, the next event may be initiated.
Event 5 B MFP trips. Operators are alerted to the trip by the following:
- Annunciator 6A09A (FWPT B TRIP)
- Annunciator 4A11B (REAC PWR CUTBACK)
The CRS implements 40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump), Section 3.0, Loss of Feedpump. On the cutback, CEA Groups 4 and 5 fully insert into the core, the Main Turbine sets back to approximately 60%, the runback circuit lowers load to match the secondary plant to the primary, and the Reactor Regulating System inserts CEAs to respond to the initial increase in Tave. Section 3.0 requires the crew to verify that subgroups 4, 5, and 22 have inserted and that Main Turbine load is less than 65%. The STA (or a designated Operator) performs Appendix D, Status Check RPCB Loss of Feedwater Pump. The BOP raises the Speed Bias on the operating MFP to zero or more and the BOP or RO checks that the RRS is adjusting CEAs to restore Tave/Tref to within 3°F. The Steam Bypass Control System (SBCS) is checked to ensure that main steam pressure is being controlled at setpoint.
The RO or BOP takes the RPCS out of service and the BOP reduces the load limit potentiometer until the potentiometer has control of the Main Turbine control valves. The BOP/RO places CEDMCs in Manual Sequential. The RO starts boron equalization.
After the RO has started the boron equalization, or at the discretion of the Lead Examiner, the next event may be initiated.
NUREG-1021, Rev 9, Supp 1 4 of 8 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 1 Overview Event 6 An inadvertent AFAS Train B occurs and AFB-P01 fails due to an 86 lockout. Since AFA-P01 is out of service, it will not auto-start as designed. Crew responds in accordance with 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations, Section 3.0, AFAS. This section directs the crew to override and operate AFW valves as necessary to control SG levels. Chemistry is informed that Blowdown is isolated and Blowdown constants are updated. Once Blowdown constants have been updated in the CMC and PC, the next event may be initiated.
With the failure of AFB-P01, the CRS enters Condition C of TS 3.7.5, since two trains of AFW (AFA tagged out, AFB failed) are now inoperable.
After the CRS briefs the crew on entry into Condition C of LCO 3.7.5, or at the discretion of the Lead Examiner, the next event may be initiated.
Event 7 A loss of condenser vacuum occurs, requiring the crew to implement 40AO-9ZZ07, Loss of Vacuum. The CRS may refer to Appendix F, Reactor Trip Criteria. This is a significant loss of vacuum; hence, no substantial action will be taken in this AOP prior to initiation of a Reactor trip. Crew should initiate a manual Reactor trip and enter 40EP-9EO01, Standard Post Trip Actions. This is the entry procedure for the Emergency Operating (EOP) System. This procedure is used for any event which actuates or requires a reactor trip. The crew checks each Safety Function and performs the Contingency Actions if required. Once the SPTAs are complete, the CRS selects the appropriate recovery procedure using the Diagnostic flowchart in 40EP-9EO01. The most likely initial diagnosis results in a transition to 40EP-9EO02, Reactor Trip.
Event 8 RCN-PIC-100, Pressurizer Master Controller, fails to 100% output in the AUTO mode, which causes both Pressurizer Spray Valves to open 100%. RO responds in accordance with ARP for B04 window B401B (PZR PRESS HI-LO), Group A, Pressurizer Pressure Ch X(Y)
Lo. If the main spray valves are not closed, First Priority Operator Action 5 requires the operator to take manual control of RCN-PIK-100, Pressure Spray Control, and close the main spray valves. Recovery actions are also addressed in general terms in the SPTAs (Contingency Action 5.1) if the RCS Pressure Control acceptance criteria are not met.
CRITICAL TASK: Close the Pressurizer Spray Valves before a SIAS occurs at 1837 psig.
When the main spray valves are closed, or at the discretion of the Lead Examiner, the next event may be initiated.
NUREG-1021, Rev 9, Supp 1 5 of 8 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 1 Overview Event 9 The CRS progresses through 40EP-9EO02, Reactor Trip, until Step 9, when AFN-P01 trips on an 86 lockout. With the loss of vacuum disabling the MFPs, malfunctions of AFB and AFN, and AFA out of service; this results in a Loss of All Feedwater and the CRS rediagnoses the event. The CRS may initially transition to 40EP-9EO06, Loss of All Feedwater, and progress until Step 6. Since Step 6 cannot be accomplished, Contingency Action 6.1 directs a transition to 40EP-9EO09, Functional Recovery. If the CRS recognizes that the FRP is the only procedure with guidance for establishing feedwater flow from the Condensate Pumps, he/she may transition directly to 40EP-9EO09. The CRS then implements 40EP-9EO09, Functional Recovery, to establish feedwater from the Condensate Pumps using Standard Appendix 44, Feeding with the Condensate Pumps. This Appendix involves selecting a SG to depressurize, lining up that SGs downcomer to accept flow, isolating that SGs economizer, tripping the FWPs, lining up feedwater heaters, ensuring adequate RCS makeup flow, and depressurizing the selected SG using atmospheric dump valves (ADVs). The CRS may elect to conserve inventory in the unselected SG by isolating it.
CRITICAL TASK: Establish feedwater flow from the Condensate Pumps and feed the selected/depressurized SG prior to dryout of that SG.
The scenario may be ended once the selected Steam Generator is being fed to at a rate that raises SG level, and/or lowers/stabilizes RCS temperature, OR when deemed appropriate by the Lead Examiner.
NUREG-1021, Rev 9, Supp 1 6 of 8 Rev 0
Appendix D Scenario Outline Form ES-D-1 TURNOVER Plant conditions:
Unit 1 is at 100% power.
The core is presently at 250 EFPD.
Risk Management Action Level is ORANGE.
AFA-P01 is out of service for unscheduled maintenance.
Train B is protected.
PC is NOT recircing the RWT.
Unit 2 is supplying the Aux Steam cross-tie header.
Equipment out of service:
The Reactor Power Cutback System is out of service to replace overheating components.
AFA-P01 is under clearance for maintenance. LCO 3.7.5, Condition A and Condition B have been entered. The pump is expected to return to service in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Planned shift activities:
Restore the RPCB System to service in accordance with 40OP-9SF04, Operation of the Reactor Power Cutback System, Sections 6.1.1 and 6.1.2.
NUREG-1021, Rev 9, Supp 1 7 of 8 Rev 0
Appendix D Scenario Outline Form ES-D-1 CREW HANDOUT Plant conditions:
Unit 1 is at 100% power.
The core is presently at 250 EFPD.
Risk Management Action Level is ORANGE.
AFA-P01 is out of service for unscheduled maintenance.
Train B is protected.
PC is NOT recircing the RWT.
Unit 2 is supplying the Aux Steam cross-tie header.
Equipment out of service:
The Reactor Power Cutback System is out of service to replace overheating components.
AFA-P01 is under clearance for maintenance. LCO 3.7.5, Condition A and Condition B, have been entered. The pump is expected to return to service in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Planned shift activities:
Restore the RPCB System to service in accordance with 40OP-9SF04, Operation of the Reactor Power Cutback System, Sections 6.1.1 and 6.1.2.
NUREG-1021, Rev 9, Supp 1 8 of 8 Rev 0
Appendix D Scenario Outline Form ES-D-1 Facility: PVNGS Scenario No.: 2 (Rev. 0) Op-Test No: NRC - 2015 Examiners: Operators:
Initial Conditions: (100% power, MOC).
Turnover: See attached Event Malf. No. Event Event Description No. Type*
1 None N Remove the Pressurizer from boron equalization in accordance with 40OP-RO/SRO 9ZZ05, Power Operations, Appendix H.6.
2 mfSI03C C SIT-1A gas leak develops requiring the crew to address 40AL-9RK2B, Panel RO/SRO B02B Alarm Responses. Crew pressurizes SIT-1A in accordance with 40OP-(TS) 9SI03, Safety Injection Tank Operations, Section 6.1.
[LCO 3.5.1 CONDITION B]
3 cmTRRX05RCNTT111Y_1 I RCN-TT-111Y, Tcold Channel 1, fails LOW. Crew enters 40AO-9ZZ16, RRS BOP/SRO Malfunctions, and selects the unaffected channel.
(AOP) 4 mfCC02A C Nuclear Cooling Water is lost due to a leak in the discharge header. This RO/BOP/ requires the operators to crosstie EW to NC IAW 40AO-9ZZ03, Loss of SRO Cooling Water, Section 4.0. This also causes a loss of letdown and the crew takes actions per 40AO-9ZZ05, Loss of Letdown.
(AOP/TS)
[LCO 3.4.9, Condition A]
[LCO 3.7.7, Condition A]
5 mfFW12A M Feedwater Line Break Inside Containment (Economizer) (Trip Initiator). Crew may initiate a manual Reactor Trip and enter 40EP-9EO01, Standard Post ALL Trip Actions. CRS transitions to 40EP-9EO05, Excess Steam Demand.
6 mfRP07A C Train A BOP ESFAS Sequencer fails on the trip. The RO responds by mfRH01B RO/BOP/ manually starting A Train Safety Injection and Support Equipment.
cmCPFW07AFBP01_5 SRO SIB-P03, CS Pump B, trips after start.
(Critical Task: Start CS Pump A prior to exiting the SPTAs.)
AFB-P01, AF Pump B, fails to automatically start.
(Critical Task: Control primary and secondary systems to prevent lifting the primary safeties.)
End N/A ALL Scenario may be terminated when SG #2 level is being maintained 45-60% NR, point at the discretion of the Lead Examiner.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Rev 9, Supp 1 1 of 7 Rev 0
Appendix D Scenario Outline Form ES-D-1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes
- 1. Total malfunctions (5-8) 7
- 2. Malfunctions after EOP entry (1-2) 3
- 3. Abnormal events (2-4) 2
- 4. Major transients (1-2) 1
- 5. EOPs entered/requiring substantive actions (1-2) 1
- 6. EOP contingencies requiring substantive actions (0-2) 0
- 7. Critical tasks (2-3) 2 CRITICAL TASK JUSTIFICATION Start CS Pump A prior to exiting the SPTAs. Failure to initiate Containment Spray when the Containment Spray Actuation Setpoint is reached could unnecessarily complicate mitigation strategies. Without spray on a FWLB, Containment pressure and temperature will be higher than expected and could unnecessarily result in harsh conditions in Containment.
Step 9.d of 40EP-9EO01, Standard Post Trip Actions, requires the crew to ensure Containment Spray Actuation Signal (CSAS) is actuated if containment pressure exceeds 8.5 psig.
Control primary and secondary systems to prevent lifting the 40DP-9AP10, Excess Steam Demand Technical Guideline, states:
primary safeties. The second action is to stabilize RCS temperature and pressure. It is important to establish heat removal capability via the unaffected SG prior to the affected SG boiling dry. Failure to stabilize RCS temperature could lead to a solid Pressurizer, Pressurized Thermal Shock (PTS) of the RCS, or result in exceeding post accident Pressure/Temperature (P/T) limits. Either of these events will unnecessarily alter mitigation strategies.
NUREG-1021, Rev 9, Supp 1 2 of 7 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 2 Overview Event 1 The RO removes the Pressurizer from boron equalization in accordance with 40OP-9ZZ05, Power Operations, Appendix H.6. This RO, normal evolution involves deenergizing backup heaters, adjusting the Pressurizer Master Control, and placing the Spray Valve Selector in BOTH. After the Spray Valve Selector is placed in BOTH, the next event can be initiated.
Event 2 SIT-1A gas leak. The crew is alerted by the following:
- Annunciator SIT 1A-1B PRESS LOW (PZR INTLK)
- Lowering pressure indications for SIT-1A on B03 and ERFDADS The crew initially responds in accordance with the alarm procedure for 40AL-9RK2C, SIT 1A-1B PRESS LOW (PZR INTLK). The ARP directs the crew to check SIT vent valves, potential drain lineups, and RDT level. If pressure is low, the ARP directs the crew to raise pressure using 40OP-9SI03, Safety Injection Tank Operations. 40OP-9SI03 directs the operator to lineup nitrogen to the affected accumulator and raise pressure. Once pressure has been raised per the CRS direction, the nitrogen lineup is secured. Since pressure in SIT 1A drops below 600 psig, the CRS enters LCO 3.5.1, Safety Injection Tanks (SITs) -
Operating, Condition A. The crew has 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore the SIT to OPERABLE status.
Event 3 RCN-TT-111Y, Tcold Channel 1, fails LOW. The crew is alerted by the following:
- Annunciator 4A10B (AMI (AUTOMATIC MOTION INHIBIT))
- Annunciator 4A08B (TAVG-TREF HI-LO)
- Lowered setpoint indication on RCN-LIC-110, Level Setpoint Control Refer to Operator Information Manual, Page 60 of 88, RRS Functional. When TTY-111Y fails LOW, the input to the averaging circuit for Loop 1 Tave fails LOW. This causes the Loop 1 Tave input into the averaging circuit of both loops Tave to be low. Since the selector switch at the RRS panel is selected to AVERAGE, the Tave output the PLCS will be low, reducing the Pressurizer level setpoint to near minimum. This causes letdown flow to increase. The AMI (AUTOMATIC MOTION INHIBIT) alarm actuates because the Loop 1 and Loop 2 Tave signals deviate by more than 5°F. In the SBCS, the Quick Open function of the bypass valves is blocked. A turbine runback demand signal will be sent to the RPCS, but no automatic action will occur until an actual runback actuation signal is generated (TLI or MFP Trip). In the DFWCS, the low Tave signal results in no feedwater flow, as the Reactor Trip Override Refill Demand senses that Tave is always below 564°F.
B04A windows 8B (TAVG-TREF HI-LO) and 10B (AMI (AUTOMATIC MOTION INHIBIT)) are received and acknowledged. Alarm response procedure 40AL-9RK4A is referenced for operator response. 40AL-9RK4A directs the crew to determine if an instrument failure has occurred. If so, the ARP directs the crew to transition to 40AO-9ZZ16, RRS Malfunctions. The crews implements Section 3.0, Temperature Instrument Failures. The BOP first ensures that CEDMCS is NOT in Auto sequential. The RO takes control of the Pressurizer Level Controller to maintain level between 33 and 53% (may refer to Appendix A, Pressurizer Level Setpoint Program). The AOP also directs the crew to select the unaffected instrument Tave 2 (Loop 2) at the RRS Test Panel. Once the NUREG-1021, Rev 9, Supp 1 3 of 7 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 2 Overview unaffected instrument has been selected, CEDMCS is placed in the desired mode, and the PLCS is returned to Remote Auto, the next event may be initiated.
Event 4 Nuclear Cooling Water is lost due to a leak in the discharge header. The crew is alerted by the following:
- Annunciator 7A07A (NUC CLG WTR SYS TRBL)
- Annunciator 7A07B (NCWS PMPS DSCH HDR PRESS HI-LO)
- Reduced current on running NC Pump
- Automatic start of standby NC Pump, with amber light.
The crew implements 40AO-9ZZ03, Loss of Cooling Water, Section 4.0. Since seal injection is in service, the crew has 10 minutes to restore cooling water to the RCPs. The AOP initially directs the crew to ensure an NC Pump is running. Since the leak is on the common discharge header, a running pump will still not deliver cooling flow to the RCPs.
When the standby pump is started and discharge pressure is still low, the operators are directed to investigate for leaks. The Area 2 AO reports a significant leak on the common discharge header. The CRS should direct the BOP to secure any running pumps. The CRS refers to 40AO-9ZZ04, Reactor Coolant Pump Emergencies. The CRS should then direct the BOP perform Appendix A, Cross-connect EW to NC. Appendix A involves startup of a Spray Pond Pump and an Essential Cooling Water Pump. Nuclear cooling water is isolated from Containment and EW is aligned to NC. To limit heat load on EW and to ensure adequate cooling flow to the RCPs, flow to Normal Chilled Water is limited to 1 chiller. An Area Operator unlocks and throttles EWA-HCV-53, SDCHX A OUTLET ISOLATION, until all of the RCP low NC flow alarms are clear. Once the low flow alarms are clear, the BOP starts a Normal Chiller. When EW has been cross connected, the CRS enters LCO 3.7.7, Condition A, due to the inoperability of the cross-connected EW train.
During the event, letdown isolates and the RO performs 40AO-9ZZ05, Loss of Letdown. If Pressurizer level exceeds 56%, the RO secures all charging pumps and the CRS enters LCO 3.4.9, Condition A.
When the Normal Chiller is started, the next event can be initiated.
Event 5 Feedwater Line Break Inside Containment (Economizer) (Trip Initiator). The crew is alerted by the following:
o 6A06A (FWCS PROCESS TRBL) o 7B03A (CNTMT SUMPS TRBL) o 7B03B (CNTMT SUMPS EXCESS LEAKAGE)
- Containment pressure and temperature rising
- Automatic initiation of SIAS, CIAS and CSAS Various other alarms on B04, B05, and B06 are received and acknowledged. Alarm response procedures 40AL-9RK6A and 40AL-9RK7B may be referenced for operator response. Operators will have little time between receipt of the first alarm and an automatic Reactor trip to implement the alarm response.
NUREG-1021, Rev 9, Supp 1 4 of 7 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 2 Overview Crew may initiate a manual Reactor Trip and enter 40EP-9EO01, Standard Post Trip Actions.
Event 6 While implementing the SPTAs, the RO observes that the Train A BOP ESFAS Sequencer failed and manually starts Train A equipment. Since CS Pump B trips on an 86 lockout, the RO must manually start CS Pump A to ensure that containment spray flow is actuated following a CSAS.
(Critical Task: Start CS Pump A prior to exiting the SPTAs.)
While implementing the SPTAs, the BOP observes that AFB-P01, AF Pump B, failed to start and manually starts the pump. Since CSAS has actuated, either the RO or the BOP secures all RCPs and the RO uses auxiliary spray and heaters to control RCS pressure.
(Critical Task: Control primary and secondary systems to prevent lifting the primary safeties.)
When the SPTAs are complete, the CRS uses Section 4.0, Diagnostic Actions, to determine that an ESD is in progress and transitions to 40EP-9EO05, Excess Steam Demand.
In 40EP-9EO05, the RO ensures that all Train A BOP ESFAS equipment is running as required. MSIS is actuated and SG #1 is identified as the most affected SG. Standard Appendix 113 is used to isolate SG #1. The SG is isolated by closing ADVs, MSIVs, MSIV Bypass, Economizer FWIVs, Downcomer Isolation Valves, Blowdown Containment Isolation Valves, steam trap isolation valves, AFA Steam Supply Valves, and AFW Isolation Valves. RCS temperature is stabilized by steaming the least affected SG.
(Critical Task: Control primary and secondary systems to prevent lifting the primary safeties.)
End Scenario may be terminated when SG #2 level is being maintained 45-60% NR, at the Point discretion of the Chief Examiner.
NUREG-1021, Rev 9, Supp 1 5 of 7 Rev 0
Appendix D Scenario Outline Form ES-D-1 TURNOVER Plant conditions:
Unit 1 is at 100% power.
The core is presently at 250 EFPD.
Risk Management Action Level is ORANGE.
AFA-P01 is out of service for unscheduled maintenance.
Train B is protected.
PC is NOT recircing the RWT.
Unit 2 is supplying the Aux Steam cross-tie header.
At the request of Chemistry, the pressurizer is in boron equalization in accordance with 40OP-9ZZ05, Power Operations.
Equipment out of service:
AFA-P01 is under clearance for maintenance. LCO 3.7.5, Condition A and Condition B, have been entered. The pump is expected to return to service in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Planned shift activities:
Remove the Pressurizer from boron equalization in accordance with 40OP-9ZZ05, Power Operations, Appendix H.6.
NUREG-1021, Rev 9, Supp 1 6 of 7 Rev 0
Appendix D Scenario Outline Form ES-D-1 CREW HANDOUT Plant conditions:
Unit 1 is at 100% power.
The core is presently at 250 EFPD.
Risk Management Action Level is ORANGE.
AFA-P01 is out of service for unscheduled maintenance.
Train B is protected.
PC is NOT recircing the RWT.
Unit 2 is supplying the Aux Steam cross-tie header.
At the request of Chemistry, the pressurizer is in boron equalization in accordance with 40OP-9ZZ05, Power Operations.
Equipment out of service:
AFA-P01 is under clearance for maintenance. LCO 3.7.5, Condition A and Condition B, have been entered. The pump is expected to return to service in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Planned shift activities:
Remove the Pressurizer from boron equalization in accordance with 40OP-9ZZ05, Power Operations, Appendix H.6.
NUREG-1021, Rev 9, Supp 1 7 of 7 Rev 0
Appendix D Scenario Outline Form ES-D-1 Facility: PVNGS Scenario No.: 3 (Rev. 0) Op-Test No: NRC - 2015 Examiners: Operators:
Initial Conditions: (100% power, MOC).
Turnover: See attached.
Event Malf. No. Event Type* Event Description No.
1 N/A N Secure and isolate SG Blowdown from SG #1in accordance with 40OP-BOP/SRO 9SG03, Operating the Steam Generator Blowdown System, Section 5.3.
2 mfCH01A C CEDM Fans A and C Trip, standby fans (HCN-A02B and A02D) fail to mfCH01C BOP/SRO automatically start. Crew implements 40AO-9ZZ20, Loss of HVAC, cmCPCH03HCNA02B_5 Section 10.0, Loss of Containment Building HVAC - CEDM.
(AOP) cmCPCH03HCNA02D_5 3 doED_ZLS037271DS_W1 C The UV-1 LOV relay for PBA-S03 fails, requiring the crew to address the doRP_ZLSAAC02ALOP1_W1 BOP/SRO alarm response procedure 40AL-9RK1A, Panel B01A Alarm Responses, mfAN_1A03D1 (TS) window 3D, UNDV A CH TRIP. Failed relay channel is bypassed in accordance with 40OP-9SA01, Section 6.8, Placing BOP ESFAS Modules in Bypass.
[LCO 3.3.7, Condition A]
4 mfTH07 C A small RCS Leak (approximately 4 gpm) develops. Crew implements RO/SRO 40AO-9ZZ02, Excessive RCS Leakrate.
(AOP/TS) [LCO 3.4.14, Condition A]
5 mfRC03A C RCP 1A Thrust Bearing oil leak, resulting in low level. RO refers to the RO/SRO Alarm Response Procedure 40AL-9RJ01 for point RCL107 (Low), RCP 1A BRG OIL RESVR LEV. SRO directs restoring oil level above the alarm setpoint using 40OP-9RC01, Reactor Coolant Pump Operation.
CRS may implement 40AO-9ZZ04, Reactor Coolant Pump Emergencies, Section 3.0, Abnormal RCP Motor or Bearing Parameters.
6 mfED13A C Loss of NNN-D11. The CRS implements 40AO-9ZZ14, Loss of Non-RO/BOP/ Class Instrument or Control Power.
SRO (AOP) 7 mfTH08 M A Pressurizer Steam Space LOCA occurs. PPS fails to initiate a Reactor See scenario file ALL Trip. Pushbuttons are successful in tripping the Reactor. Crew implements C 40EP-9EO01, Standard Post Trip Actions, then transitions to 40EP-9EO03, Loss of Coolant Accident.
BOP (CRITICAL TASK: Trip the Reactor prior to exiting Step 1 of SPTAs) 8 cmCPSI01SIAP02_6 C HPSI Pump A trips and HPSI Pump B fails to automatically start.
cmCPSI01SIBP02_5 RO (CRITICAL TASK: Manually start HPSI Pump B prior to exiting SPTAs.)
End N/A ALL The scenario may be terminated once the RCS cooldown step in reached, at point the discretion of the Chief Examiner.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Rev 9, Supp 1 1 of 9 Rev 0
Appendix D Scenario Outline Form ES-D-1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes
- 1. Total malfunctions (5-8) 10
- 2. Malfunctions after EOP entry (1-2) 2
- 3. Abnormal events (2-4) 3
- 4. Major transients (1-2) 1
- 5. EOPs entered/requiring substantive actions (1-2) 1
- 6. EOP contingencies requiring substantive actions (0-2) 0
- 7. Critical tasks (2-3) 2 CRITICAL TASK JUSTIFICATION Trip the Reactor prior to exiting Step 1 of SPTAs Failure to ensure that the Reactivity Control Safety Function is met will result in excess heat input into the RCS and overheating of the nuclear fuel. 40DP-9AP06, Standard Post Trip Actions Technical Guidelines, explain that, to ensure the Reactor is shutdown, operators must take Contingency Actions if the Reactor is not automatically shut down by the Plant Protection System.
Manually start HPSI Pump B prior to exiting SPTAs. Inadequate Safety Injection flow may result in loss of subcooled margin and/or core uncovery. Additionally, failure to establish SI flow may lead to an inappropriate transition to the Functional Recovery Procedure, which would complicate mitigation strategies.
Failure to start HPSI will delay the point where SI throttle criteria are met and could result in extended operation of the LPSI pumps, which could, in turn, result in LPSI pump damage (degraded ECCS).
NUREG-1021, Rev 9, Supp 1 2 of 9 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 3 Overview Event 1 Secure and isolate SG Blowdown from SG #1 in accordance with 40OP-9SG03, Operating the Steam Generator Blowdown System, Section 5.3. This normal evolution involves entering new Blowdown Constants into the new Core Monitoring Computer per Appendix O. The BOP places SCN-HS-1, Steam Generator 1 Blowdown Path Selector, to the OFF position to stop flow. The BOP then verifies system response using Appendix G, Blowdown Verifications (per the Stopping B/D column). An AO is dispatched to perform a local lineup per Appendix I, Securing Steam Generator 1 Blowdown. The BOP will then close the SG 1 Blowdown Containment Isolation Valves (UV-500P/Q) and the 3 SG 1 Isolation Valves (HV-43, 41, 47). Cooling water to the Blowdown Heat Exchanger will remain in service. When SGE-HV-47 is closed, the next event may be initiated.
Event 2 CEDM Fans A and C trip and the standby fans (HCN-A02B and A02D) fail to automatically start. The standby fans normally start on a low DP after a 120 second time delay. The crew is alerted by the following:
- Brighter than green lights on the previously-running fans
- Annunciator 7A09B (CEDM ACU COOL SYS TRBL), Group E
- Computer alarm point HCYS49 (CEDM ACU A Fan A(C) Elect Prot)
- SEAS/SEIS alarms (21B, CEDM NORM; 6D1,Non-ESF Load Shed)
The BOP refers to 40AL-9RK7A for window 7A09B. CRS implements 40AO-9ZZ20, Loss of HVAC, Section 10.0, Loss of Containment Building HVAC - CEDM. Since RCS temperature is greater than 300°F, the crew has 40 minutes to restore CEDM cooling or trip the Reactor. The BOP waits for approximately two minutes, then start fans B and D.
If they do not start the standby fans within 10 minutes; they must perform 40OP-9ZZ05, Power Operations, Section 8.0, Rapid Shutdown, to ensure the Unit is shut down within 40 minutes of the loss of CEDM HVAC. NOTE: Both 40AL-9RK7A and 40AO-9ZZ20 provide direction to start the standby fans.
When the standby CEDM fans are started, on Board 1, alarm 1A5D (120 VAC 1E PNL 28 INVERTER D TRBL) actuates. This is accompanied by an alarm on Computer Point ID PNYS4, 120 VAC INV D AC/DC STATUS. 40AL-9RK1A directs the RO to dispatch an AO to check indications on the inverter control panel. When dispatched, the AO reports that there is a red LOSS OF SYNC light on PND-N14 Inverter D. In accordance with the table under Operator Action 9, the ARP directs the AO to check the availability of the alternate supply, then depress the SYNCHRONIZATION button to clear the alarm When the standby fans have been started, or at the discretion of the Lead Examiner, the next event may be initiated.
Event 3 The UV-1 LOV relay for PBA-S03 fails. The crew is alerted by the following:
- Annunciator 1A3D (UNDV A CH TRIP)
- On Panel B01, the white light PHASE AB 727-1 (for the 4.16KV BUS POTENTIAL INDICATION) is extinguished)
NUREG-1021, Rev 9, Supp 1 3 of 9 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 3 Overview
The RO refers to 41AL-1RK1A for window 1A03D. There are NO automatic actions for one channel UV trip. The ARP directs the operator to check the 4.16KV BUS POTENTIAL INDICATION lights and the RO observes that the PHASE AB 727-1 light is off. Operator Action 3 of the ARP provides direction for only 1UV relay failure. Once alarm validity has been checked and the relay identified, the ARP directs the operator to bypass the malfunctioned relay in accordance with 40OP-9SA01, BOP ESFAS Modules Operation. The BOP uses Section 6.8, Placing BOP ESFAS Modules in Bypass. After obtaining a key and verifying Perquisites and Initial Conditions are met, the BOP performs a lamp test (6.8.4), selects the proper relay channel (6.8.7), and checks that the opposite Train is NOT in Bypass (6.8.10). To complete the bypass, the BOP inserts the key, turns it clockwise 1/4 turn, and verifies that the Bypass light is ON (6.8.11).
When the BOP opens the BOP ESFAS Panel door, the CR will receive alarm 5A2D (BOP ESFAS IN TEST) and the alarm will clear when the door is closed. When the BOP turns the key to Bypass, the CR will receive alarm 5A3D (BOP ESFAS CH BYP), which is an expected alarm.
The CRS evaluates TSs 3.3.7, 3.8.1, and 3.8.2. LCO 3.3.7, Diesel Generator (DG) - Loss of Voltage Start (LOVS), Condition A is entered because only one LOVS channel is inoperable. Condition A requires the failed channel to be placed in bypass or trip within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. LCO 3.8.1, AC Sources - Operating, is still met because the failed relay channel does not make offsite sources, the associated DG, nor the load sequencer inoperable. LCO 3.8.1, AC Sources - Shutdown, is not applicable because the Unit is NOT in Mode 5 or 6.
Event 4 A small RCS Leak (approximately 4 gpm) develops. The crew is alerted by the following:
- Alarm on RU-1, Containment Atmosphere
- Rising Containment Sump levels on BO7, RDN LI-410, RDN LI-10
- Rising Containment Sump levels on BO7, Yokogawa recorder RMN-TRJ-1, Points 17, 18, 18, and 20 The crew initially responds using 74RM-9EF41, Radiation Monitoring System Alarm Response, for the RU-1 alarm. RP and the Radiological Monitoring Technician are informed. Operator Response 4 directs the crew to perform an RCS water inventory balance per 40ST-9RC02, ERFDADS (Preferred) Calculation of RCS Water Inventory.
When rising Containment sump levels and temperatures are observed, the CRS implements 40AO-9ZZ02, Excessive RCS Leakrate, Section 3.0, RCS Leakage. For this small leakage, Pressurizer level is relatively stable and letdown remains in service with the existing Charging Pump configuration. LCO 3.4.14, RCS Operational Leakage, is evaluated. Chemistry and RP are informed. The leakrate is quantified, most likely using Appendix B, ERFDADS Leak Rate Determination. This appendix directs the RO to secure Reactor Makeup and setup ERFDADS to run the calculation by selecting RCS LEAK RATE on the SPDS Overview screen and selecting TREND-1 on the Analog Point Attributes screen. The trend is run for at least 15 minutes or until VCT level lowers to 15%. Once the leak rate has been determined, VCT makeup is restored.
NUREG-1021, Rev 9, Supp 1 4 of 9 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 3 Overview Since the leak is UNIDENTIFIED leakage and the calculated leak rate is approximately 4 gpm, the CRS should enter LCO 3.4.14, Condition A. Once the CRS has determined that LCO 3.4.14, Condition A, must be entered, OR at the discretion of the Lead Examiner, the next event may be initiated.
Event 5 RCP 1A Thrust Bearing oil leak, resulting in low level. RO refers to the Alarm Response Procedure 40AL-9RJ01 for point RCL107 (Low), RCP 1A BRG OIL RESVR LEV. The alarm actuates at 64%. The ARP directs the crew to validate the alarm by calling up the PMS or ERFDADS point. It then directs filling the RCP thrust bearing reservoir per 40OP-9RC01, Reactor Coolant Pump Operation. The operator uses Section 6.14 of 40OP-9RC01 to raise reservoir level. Instruction 6.14.5 directs the operator to start (and hold) RCN-P02A, RCP Lift Oil Pump P02A, to begin filling the reservoir. When ERFDADS point RCL107 indicates level is between 64% and 85% (determined by CRS), the lift pump switch is allowed to spring-return to AUTO. Operator Action 2.5 of the ARP prompts evaluation of 40AO-9ZZ04, Reactor Coolant Pump Emergencies, and the CRS may implement Section 3.0, Abnormal RCP Motor or Bearing Parameters. Section 3.0 directs the crew to monitor Upper Thrust Bearing temperature (may use lift pump to slow the rate) and restore the reservoir level per Appendix C, Restoring RCP Oil Reservoir Levels.
Once the RCP oil lift pump is returned to AUTO and the reservoir is filled per the CRS direction, or at the discretion of the Lead Examiner, the next event may be initiated.
Examiner NOTE: To prevent repetitive alarms and fill operations, the malfunction will be deleted when the first alarm actuates.
Event 6 Loss of NNN-D11. The crew is alerted by the following:
- Annunciator 1C12B (120 VCA PNL D11/D12/15/16 TRBL)
- Numerous annunciators on B03, B04, and other panels
- Loss of power to recorders for Pressurizer level and VCT level/pressure
- Numerous Computer alarm points The RO refers to 40AL-9RK1C, point NNYS3 (Bkr Ovld Trip), which prompts the RO to direct an AO to investigate the alarm. The AO reports that there is a ground detection indicating light on the center panel of the switchgear and that the bus feeder breaker 52-D11 has tripped. Operator Action 2 of the ARP then directs performance of 40AO-9ZZ14, Loss of Non-Class Instrument or Control Power. The crew walks down the control boards to evaluate affected equipment. FIN/electrical maintenance is informed and PR&C is notified to locate the ground. 40AO-9ZZ14 directs the crew to operate ADVs to control SG pressures. The RO places the following handswitches in Channel X:
- RCN-HS-110, Level Control Selector Switch
- RCN-HS-100-3, Heater Control Selector Switch
- RCN-HS-100, Pressure Control Selector Switch The BOP ensures CEDMCS is NOT selected to Auto Sequential AS. The RO ensures that no more than one Charging Pump is running and implements 40AO-9ZZ05, Loss of Letdown. The RO initially ensures no more than 1 Charging Pumps is running. At the direction of the CRS, the RO performs Appendix C, Extended Operations Without NUREG-1021, Rev 9, Supp 1 5 of 9 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 3 Overview Letdown. The RO closes the Seal Injection Flow Control Valves and places all Charging Pumps in PULL TO LOCK.
The CRS evaluates the following TSs:
- LCO 3.2.1, Linear Heat Rate (LHR)
- LCO 3.2.2, Planer Radial Peaking Factors (Fxy)
- LCO 3.2.3, Azimuthal Power Tilt (Tq)
- LCO 3.2.5, Axial Shape Index (ASI)
The CRS will perform 40DP-9OP05, Control Room Data Sheet Instructions, due to the loss of JSCALOR. Since JSCALOR is not available and COLSS is functioning, 40DP-9OP05, Instruction 3.3.11 directs the crew to record the current NKBDELTC values and establish that value as the current steady state maximum power.
When the RO has completed the actions in 40AO-9ZZ05, Loss of Letdown, or at the discretion of the Lead Examiner, the next event may be initiated.
Event 7 A Pressurizer Steam Space LOCA occurs. PPS fails to initiate a Reactor Trip and the BOP uses the MANUAL REACTOR TRIP pushbuttons to trip the Reactor. Crew implements 40EP-9EO01, Standard Post Trip Actions.
(CRITICAL TASK: Trip the Reactor prior to exiting Step 1 of SPTAs)
While implementing the SPTAs, the RO observes that Pressurizer level is NOT trending to 33-53% and that RCS subcooling is less than 24°F, and then secures all RCPs. The RO also observes that Pressurizer pressure is less than 1837 psia and is NOT trending to 2225-2275 psia. The RO then ensures that SIAS is actuated. At this point, the RO may note that HPSI Pump A has tripped and HPSI Pump B has failed to automatically start. The RO may start HPSI Pump B at this time.
When the SPTAs are complete, the CRS uses the Diagnostics Actions to determine that there is a LOCA in progress and then transitions to 40EP-9EO03, Loss of Coolant Accident.
Event 8 HPSI Pump A trips and HPSI Pump B fails to automatically start. While implementing 40EP-9EO03, Loss of Coolant Accident, Instruction 5.a directs the crew to check the status of the HPSI and LPSI pumps. If not already noted in the SPTAs, the RO observes that HPSI Pump A has tripped and HPSI Pump B has failed to automatically start. The RO shall start HPSI Pump B at this time.
(CRITICAL TASK: Manually start HPSI Pump B prior to exiting SPTAs.)
The crew then attempts to locate and isolate the leak, place the Hydrogen Analyzers in service, and ensure CIAS has properly actuated. The RO will ensure that at least one CS header flow is greater than 4350 gpm and isolate RCP control bleedoff flow. The Hydrogen Recombiners are placed in service. Since Containment pressure is less than 50 psig, and SI flow is within the SI delivery curves, one CS Pump is stopped. The crew directs an AO to reenergize SIAS Load Shed Panels in accordance with Appendix 21. The crew cools down the Steam generators (and RCS) using the ADVs (since SBCS is NUREG-1021, Rev 9, Supp 1 6 of 9 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 3 Overview unavailable due to the loss of NNN-D11).
End Point The scenario may be terminated once the RCS cooldown step in reached, at the discretion of the Chief Examiner.
NUREG-1021, Rev 9, Supp 1 7 of 9 Rev 0
Appendix D Scenario Outline Form ES-D-1 TURNOVER Plant conditions:
Unit 1 is at 100% power.
The core is presently at 250 EFPD.
Risk Management Action Level is ORANGE.
AFA-P01 is out of service for unscheduled maintenance.
Train B is protected.
PC is NOT recircing the RWT.
Unit 2 is supplying the Aux Steam cross-tie header.
Equipment out of service:
AFA-P01 is under clearance for maintenance. LCO 3.7.5, Condition A and Condition B, have been entered. The pump is expected to return to service in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Planned shift activities:
To support maintenance on HCN-HV-1A, SG #1 Normal Rate Blowdown Flow Control Valve, secure and isolate Steam Generator Blowdown from SG #1 in accordance with 40OP-9SG03, Operating the Steam Generator Blowdown System, Section 5.0. Cooling water to the Blowdown Heat Exchanger will remain in service.
NUREG-1021, Rev 9, Supp 1 8 of 9 Rev 0
Appendix D Scenario Outline Form ES-D-1 CREW HANDOUT Plant conditions:
Unit 1 is at 100% power.
The core is presently at 250 EFPD.
Risk Management Action Level is ORANGE.
AFA-P01 is out of service for unscheduled maintenance.
Train B is protected.
PC is NOT recircing the RWT.
Unit 2 is supplying the Aux Steam cross-tie header.
Equipment out of service:
AFA-P01 is under clearance for maintenance. LCO 3.7.5, Condition A and Condition B, have been entered. The pump is expected to return to service in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Planned shift activities:
To support maintenance on HCN-HV-1A, SG #1 Normal Rate Blowdown Flow Control Valve, secure and isolate Steam Generator Blowdown from SG #1 in accordance with 40OP-9SG03, Operating the Steam Generator Blowdown System, Section 5.0. Cooling water to the Blowdown Heat Exchanger will remain in service.
NUREG-1021, Rev 9, Supp 1 9 of 9 Rev 0
Appendix D Scenario Outline Form ES-D-1 Facility: PVNGS Scenario No.: 4 (Rev. 0) Op-Test No: NRC - 2015 Examiners: Operators:
Initial Conditions: (75% power, MOC).
Turnover: See attached Event Malf. No. Event Event Description No. Type*
1 cmBSEG03DGBPSL4_2 N Crew unloads and shuts down DG B in accordance with 40OP-9DG02, cmBSEG03DGBPSL6_2 RO/SRO Emergency Diesel Generator B. When PEB-SC-G02, Diesel Generator B Speed handswitch is placed in LOWER for the second time, the DG trips on cmBSEG03DGBPSL8_2 (TS) low lube oil pressure.
cmBSEG03DGBPSL10_2 [LCO 3.8.1, Condition B]
2 mfEG04 C Main Generator AC Voltage Regulator fails. BOP addresses ARP for B06, BOP/SRO window 6B08D, EXCTN SYS TRBL, and switches to DC control per 40OP-9MB01, Main Generation and Excitation, Section 9.0, Transferring Between AC and DC Voltage Regulation.
3 cmBKED13NGNL06C4_5 C Loss of NHN-M50 (86 Lockout). Crew implements 40AO-9ZZ12, Degraded RO/SRO Electrical, Section 80.0, Loss of NHN-M50.
(AOP) 4 mfRD02B C CEA 15 (Reg Group 5) slips half way into the core. Crew implements 40AO-RO/BOP/ 9ZZ11, CEA Malfunctions. The crew begins a 15% power reduction.
SRO [LCO 3.1.5 CONDITION A]
(AOP/TS) (CRITICAL TASK: Begin power reduction within 10 minutes of slipped CEA.)
5 cmCPRC02RCEP01A_6 M RCP 1A 86 Lockout Trip. (Trip Initiator) Crew implements 40EP-9EO01, ALL Standard Post Trip Actions.
6 mfED02 C During implementation of the SPTAs, a Loss of Offsite Power (LOOP) occurs.
RO/BOP The CRS should transition to 40EP-9EO07, Loss of Offsite Power/Loss of M-All Forced Circulation.
7 mfEG06A C DG A trips due to a generator differential. This results in a loss of all AC RO power, requiring the crew to implement 40EP-9EO08, Blackout.
(CRITICAL TASK: Restore power to at least one vital AC bus and commence restoring essential equipment on that bus within one hour of the Blackout.)
End N/A ALL After the crew has restored power to at least one vital AC bus and essential point equipment on that bus has been started, the scenario may be terminated at the discretion of the Chief Examiner.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Rev 9, Supp 1 1 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 Target Quantitative Attributes (Per Scenario; See Section D.5.d) Actual Attributes
- 1. Total malfunctions (5-8) 7
- 2. Malfunctions after EOP entry (1-2) 2
- 3. Abnormal events (2-4) 2
- 4. Major transients (1-2) 1
- 5. EOPs entered/requiring substantive actions (1-2) 1
- 6. EOP contingencies requiring substantive actions (0-2) 0
- 7. Critical tasks (2-3) 2 CRITICAL TASK JUSTIFICATION Begin power reduction within 10 minutes of slipped CEA. Section 15.4.3.2 of the FSAR assumes the operators takes action within 900 seconds to reduce power. This assumption is used to ensure the core does not exceed DNBR or LPD limits. Although the FSAR states 900 seconds, Tech Specs requires a power reduction per the COLR, which requires a power reduction within 10 minutes.
Failure to reduce power could result in not meeting the Shutdown Margin (SDM) requirements of TS 3.1.2, SDM RTBs Closed.
Inadequate SDM at power could lead to exceeding fuel design limits for normal shutdown and anticipated operational occurrences.
Restore power to at least one vital AC bus and commence FSAR Section 9.5.9, Station Blackout Evaluation, explains that the restoring essential equipment on that bus within one hour of SBO 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> coping evaluation (based on NUMARC 87-00, the Blackout. Revision 1 criteria) assumes that an alternate AC power source is started and loaded within the first hour. Failure to restore alternate AC power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> will result in RCP seal leakage beyond that assumed in the SBO coping evaluation. This will, in turn, have an adverse impact on containment temperature and pressure (along with the loss of containment ventilation).
NUREG-1021, Rev 9, Supp 1 2 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 4 Overview Event 1 Crew unloads and shuts down DG B in accordance with 40OP-9DG02, Emergency Diesel Generator B.
The Turnover indicates that the DG is being run for a surveillance. For the Turnover, the crew will be provided a marked-up copy of 40ST-9DG02 (up to Step 7.5) and 40OP-9DG02 (up to Step 6.7.2). The surveillance is complete and Step 7.5 of the ST directs the crew to continue operation of the DG per 40OP-9DG02. The RO will use Section 6.7, Unloading Train B Diesel Generator and will follow the direction of Appendix G, Loading and Unloading Schedule. When PEB-SC-G02, Diesel Generator B Speed handswitch is placed in LOWER for the second time, the DG trips on low lube oil pressure.
When DG B trips, the crew is alerted by the following annunciators on B01:
- 1C16A (DG B TRIP)
- 1C16D (DG B HI PRIORITY TRBL)
A note at the beginning of Operator Actions for 40AL-9RK1C, window 1C16A, prompts the crew to evaluate LCOs 3.8.1, AC Sources - Operating, and 3.8.2, AC Sources -
Shutdown. The RO confirms the trip and directs an AO to investigate locally (These responses are common to all three annunciator windows). The AO will report the following indications:
- Annunciator DGB01A (LUBE OIL LOW PRESSURE ENGINE)
- Annunciator DGB02A (LUBE OIL LOW PRESSURE TURBO)
- Annunciator DGB01D (LUBE OIL LOW PRESSURE ENGINE)
- Annunciator DGB02D (LUBE OIL LOW PRESSURE TURBO)
- DGN-PI-2, Engine Lube Oil Pressure (DGB-B01), reads 22 psig
- DGN-PI-80, Lube Oil Pressure at Engine (Panel NW side of diesel), reads 18 psig.
The CR may direct the AO to locally secure the lube oil pumps and turn off lube oil heaters.
Since the Unit is in Mode 1, LCO 3.8.2 is not applicable. The CRS declares DG B inoperable and enters LCO 3.8.1, Condition B, since only 1 DG is inoperable. The crew has one hour to perform Surveillance Requirement 3.8.1.1 for the OPERABLE required offsite circuits. This SR verifies the breaker alignment and indicated power availability for each required offsite circuit.
Event 2 Main Generator AC Voltage Regulator fails LOW. Operators are alerted by the following:
- Field Amps and Field Volts decrease, armature amps increase.
- MVARS on MAN-JIV-G01 shift from BOOST to BUCK
- Flashing Point 11 (URAL) on the Generrex Control Panel (B06)
- Annunciator 6B08D (EXCTN SYS TRBL)
- Computer alarm point MBYS22, Excitation Voltage Regulator Mode Change NUREG-1021, Rev 9, Supp 1 3 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 4 Overview
- Computer alarm point MBYS8, AC VOLT ADJ IN MIN POS
- Annunciator 1A15C, 4.16KV UNIT 1 SWGR S01 TRBL
- Annunciator 1A17B, 13.8KV UNIT 1 SWGR S01 TRBL
- Annunciator 1B13B, 13.8KV UNIT 1 SWGR S02 TRBL The BOP addresses 40AL-9RK6B for B06, window 6B08D, EXCTN SYS TRBL, Group F, Excitation Voltage Regulator Mode Change. This ARP directs the BOP to check the LEDs on the Generrex mimic bus. If the Operator observes LED 11 flashing, then Operator Action 1 directs the BOP to increase the setpoint adjuster to allow the AC regulator to regain control. If the AC regulator is not available, Operator Action 2 directs the Operator to perform 40OP-9MB01, Main Generation and Excitation, Section 9.0, Transferring Between AC and DC Voltage Regulation. Instruction 9.3.2 of 40OP-9MB01 directs the BOP to depress the DC button on the AC/DC transfer switch and check for a Mode change alarm, LED #3 light out, and LED #4 light flashing. The BOP then adjusts MVARS as required by the ECC. If requested, the ECC directs the crew to establish a station total of 300 MVARS, 100 per Unit.
When the BOP has checked for the proper LED status, or at the discretion of the Lead Examiner, the next event may be initiated.
Event 3 Loss of NHN-M50 (86 Lockout). Crew is alerted by the following:
o 2A01A (CONT BLDG HVAC SYS TRBL) o 1B17D (480 VOLT LC L06 TRBL) o 1A02D (DG A LO PRIORITY TRBL) o 1C14D (DG RMS HVAC TRBL) o 1C17 D (DG B LO PRIORITY TRBL) o 1B06C (PMUX TRBL)
- Brighter than green indication on NON CLASS 1E MCC SUPPLY BREAKER STATUS for NHN-M50 INCM FDR NGN-L06C4
- Loss of indication on the following (most significant components):
o HJN-HS-5, CR NORM SPLY AHU FAN A02A o HJN-HS-86, BATTERY ROOM B NORM EXH FAN J01B o HJN-HS-88, BATTERY ROOM D NORM EXH FAN J01D o HJN-HS-139, ESF SWGR ROOM NORM AHU FAN N
- Numerous Computer alarm points, including the following (most significant components):
o HJPDS90 (Cont Bldg Battery Room B Exhaust Fan Diff Press Lo) o HJPDS92 (Cont Bldg Battery Room D Exhaust Fan Diff Press Lo) o HJYS5 (Control Room Normal Air Handling Unit Fan Elect Prot) o HJYS20 (ESF Swgr Rm Normal AHU Fan Electrical Protection) o HJYS2 (Control Bldg Battery Room B Normal Exhaust Fan Elect Prot o HJYS4 (Control Bldg Battery Room D Normal Exhaust Fan Elect Prot) o DGYS3 (Diesel Generator A Lo Priority Trouble) o DGYS4 (Diesel Generator B Lo Priority Trouble)
NUREG-1021, Rev 9, Supp 1 4 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 4 Overview o HDYS1 (Diesel Generator Room A/B HVAC Trouble)
Crew implements 40AO-9ZZ12, Degraded Electrical, Section 80.0, Loss of NHN-M50.
Section 80.0 directs the crew to ensure that the Control Building elevator is empty, then evaluate table NHN-M50 Loads to determine other required actions.
The RO implements 40AL-9RK2A to respond to the CONT BLDG HVAC SYS TRBL alarm. In response to Computer points HJYS2(4) the RO starts Essential Battery Room exhaust fans HJB-JO1B (Battery Room B) and HJB-J01A (Battery Room D). In response to Computer point HJYS5, the RO starts a Control Room Essential AHU per 40OP-9HJ01, Control Building HVAC. In response to Computer point HJYS20, the RO starts an ESF Swgr Rm Essential AHU per 40OP-9HJ01.
In response to annunciator 1C14D (DG RMS HVAC TRBL), the RO may direct an AO to investigate locally. The AO reports the following:
- DG A: Annunciator DGA2F (ANY SWITCH NOT IN AUTO)
- DG B: Annunciator DGB2F (ANY SWITCH NOT IN AUTO)
- DG B: Annunciator DGB1F (GEN HV CUBICLE SPACE HTR TROUBLE)
Based on the many indications related to loss of ventilation, the CRS may elect to implement 40AO-9ZZ20, Loss of HVAC, to start essential HVAC. In Section 11.0, if a train of Control Room Essential is started, operators will first ensure that the following associated equipment is actuated:
- SP Pump - running
- EW Pump - running
- Essential Chiller - running
- Control Room Normal AHU - stopped
- HJA-F04 - running Next, the operators secure the ESF Switchgear Room Norm Supply AHU and close the rooms supply damper. Control Building Essential Isolation dampers are aligned and the ESF Switchgear Room and ESF Equip Room Fans are started. The majority of the remaining Steps in Section 11.0 involve contingencies taken when temperature limits are exceeded.
When essential AHUs have been started for the Battery Rooms, ESF Switchgear Rooms, and Control Room; or at the discretion of the Lead Examiner, the next event may be initiated.
Event 4 CEA 15 (Reg Group 5) slips half way into the core. The crew is alerted by the following:
- Annunciator 4A08A (CEDMCS TRBL)
- Annunciator 4A09B (CWP (CEA WITHDRAWAL PROHIBIT))
- Annunciator 5A13B (CPC/CEAC TRBL)
NUREG-1021, Rev 9, Supp 1 5 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 4 Overview
- Annunciator 5B01D (COLSS PC ALARM)
- Annunciator 5B01C (COLSS CMC ALARM)
- CEA DEV alarms on the DNBR/LPD Calculator Panels
- Computer alarm point SBYS76 (CEAC 1A DEVIATION (HI)) (several other similar alarms)
- Computer alarm point SBYS20 (CROSS CH COMPARISON FAIL)
Crew implements 40AO-9ZZ11, CEA Malfunctions, Section 3.0, Dropped or Slipped CEA Mode 1 or 2. Section 3.0 directs the BOP to place CEDMCS in STANDBY and perform Appendix E, Initial Actions. In Appendix E, an AO is dispatched to investigate at alarm panel J-SFN-C01D. AO reports that there is a CWP alarm and no breakers are open.
I&C and Reactor Engineering are informed. The RO initiates Pressurizer boron equalization. Within 10 minutes, the crew begins a power reduction.
(CRITICAL TASK: Crew begins power reduction within 10 minutes of slipped CEA.)
The BOP initially lowers turbine load to raise Tave 3°F greater than Tref. The CRS determines that the initial power reduction is 15% (as directed by Instruction 14, Bullet 2) and calculates the amount of boron required. The BOP lowers turbine load to maintain Tave 3°F above Tref and the RO begins a boration at a minimum of 35 gpm. The power reduction follows the requirements of Appendix B, Core Power Reduction After a CEA Deviation. This Appendix establishes the minimum times allowed to complete the required downpower, based on the pre-event power level.
The CRS may initiate Appendix J, LCO Required Action Tracker (normally SM or STA duty). During the downpower, the CRS refers to Appendix H, Required Power Ramp with a CEA Misalignment Greater than 6.6. Notes in the upper right corner explain that these curves reflect the initial power reduction required by LCO 3.1.5, CEA Alignment, Condition A.
The CRS enters LCO 3.1.5 Condition A, due to one CEA trippable and misaligned from it group by > 9.9 inches. The Required Action is to reduce THERMAL POWER in accordance with the limits in the COLR within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND restore CEA alignment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Once the power reduction has started, or at the discretion of the Lead Examiner, the next event may be initiated.
Event 5 RCP 1A 86 Lockout Trip. (Trip Initiator) The crew is alerted by the following:
- Brighter than green indication on RCN-HS-1, RCP1A P01A
- Rod bottom lights on Core Mimic Panel on B04, except 4 stuck rods
- Annunciator 4A07A (REAC TRIP (CEDMCS BUS UNDV))
NUREG-1021, Rev 9, Supp 1 6 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 4 Overview
- Annunciator 4A07B REAC CKT BKR OPEN
- Reactor Trip 1ST Out - DNBR LO
- CEA CRT indicates all rods fully inserted, except 4 stuck rods Crew implements 40EP-9EO01, Standard Post Trip Actions.
Event 6 4 CEAs (14, 16, 43, and 38) fail to fully insert on the trip. While implementing the SPTAs, the crew observes that 4 rods have failed to insert. The crew is alerted by the following:
- Lack of rod bottom lights on Core Mimic Panel on B04 for 4 rods
- CEA CRT indicates all rods fully inserted, except 4 rods Contingency Action c.1 of 40EP-9EO01 directs the RO to borate the RCS, using Standard Appendix 103, 10, or 11, until adequate SDM is established. The most likely boration method is using Appendix 103, RCS Makeup/Emergency Boration, Attachment 103-A, Normal Boration Path. This Attachment directs the RO to setup CHN-FIC-201Y, Boric Acid Makeup to VCT Flow Control to less than 40 gpm and the target makeup volume to a minimum of 5000 gallons. The RO places CHN-HS-210, Makeup Mode Select Switch, to BORATE, ensures a BAMP is running, and ensures CHN-UV-527, Makeup to CHRG PUMPS (VCT Bypass), is open. The boration begins when the RO presses the Start button on CHN-FQIS-210Y. The RO ensures that only boric acid flow has been established, and then raises flow to a minimum of 44 gpm.
(CRITICAL TASK: Establish >26 gpm boration flow within 15 minutes of the stuck CEAs.)
When the crew completes Step 8 of the SPTAs, or at the discretion of the Lead Examiner, the next event may be initiated.
Event 7 During implementation of the SPTAs, a Loss of Offsite Power (LOOP) occurs. The crew is alerted by the following:
- Observation that PBB-S04 is deenergized (DG B previously tripped).
- Observation that only DG A is carrying PBA-S03.
- Observation that non-class buses are deenergized.
- Observation that no RCPs are running.
The CRS may elect to start over with the SPTAs. The RO observes that no Charging Pumps are running and manually starts Charging Pump A..
When SPTAs are complete, the CRS refers to Section 4.0, Diagnostic Actions, to diagnose the event and determine the appropriate recovery procedure. The CRS transitions to 40EP-9EO07, Loss of Offsite Power/Loss of Forced Circulation. 40EP-9EO07 directs the crew to check that Safety Function Status Check acceptance criteria are met, inform Chemistry, and classify the event. Since a LOOP has occurred, the crew verifies that loads have sequenced onto PBA-S03. No charging pumps are running (Charging Pump A trips on the LOOP and is not automatically restarted), so the RO isolates seal injection and seal bleedoff and then resets the anti-pump condition on the always running Charging Pump NUREG-1021, Rev 9, Supp 1 7 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 4 Overview (A/1) by placing the handswitch in STOP. Since CW flow to the Main Condenser is lost, so the BOP actuates MSIS. After the MSIS has actuated, an AO is dispatched to check Condenser Reheat Tray levels. When the AO reports levels are normal, the BOP overrides and open trap isolation valves SGA-HS-1133 and 1134. The BOP controls Tc less than 570°F using the ADVs.
Once the BOP establishes control of Tc with the ADVs and has established feed with AFA, or at the discretion of the Lead Examiner, the next event may be initiated.
Event 8 DG A trips due to a generator differential. The crew is alerted by the following:
- Annunciator 1A01A (DG A TRIP)
- PBA-S03 deenergized
- Running class equipment on PBA-S03 no longer running This results in a loss of all AC power, requiring the CRS to transition to 40EP-9EO08, Blackout. In 40EP-9EO08, the crew actuates MSIS, informs the Energy Control Center of the Blackout. Security is dispatched to allow an AO access to the SBOGs and an AO is dispatched to start an SBOG using 40EP-9EO10, Appendix 111, Station Blackout Generator Operation. When the SBOG is running, the AO energizes NAN-S07. The RO places all Charging Pumps in PULL TO LOCK and minimizes RCS leakage by isolating letdown, RCP controlled bleedoff, and RCS sample flowpaths. The BOP uses ADVs to control RCS Tc less than 570°F and maintains SG levels between 45-60% NR.
An AO is dispatched to perform Attachment 80-A, Disable PBA-S03 Breakers. This Appendix disables breakers on PBA-S03 and ensures the bus feeder breakers are open.
The RO performs Appendix 80, Align SBOG to PBA-S03 (BO). When the AO has completed Attachment 80-A and the RO has opened feeders to PBA-S03, the RO directs an AO to close NAN-S03AB, 13.8KV Supply from GTG. An AO is then directed to close NAN-S07D. When NAN-S07D is closed, the RO energizes PBA-S03 through the normal supply breaker. The RO also performs Appendix 53, Align Deenergized Buses. This Appendix is similar to Appendix 80 in that it ensures all feeder breakers are open and all breakers to major loads (RCPs, Circ Water Pumps) are open. When the AO reports that Attachment 80-A (81-A) is complete, essential equipment is then started in a controlled manner to ensure SBOG limitations are not exceeded.
(CRITICAL TASK: Restore power to at least one vital AC bus and commence restoring essential equipment on that bus within one hour of the Blackout.)
NUREG-1021, Rev 9, Supp 1 8 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 2015 NRC Scenario 4 Overview End Point After the crew has restored power to at least one vital AC bus and the Step to start essential equipment on that bus has been evaluated, the scenario may be terminated at the discretion of the Chief Examiner.
NUREG-1021, Rev 9, Supp 1 9 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 TURNOVER Plant conditions:
Unit 1 is at 75% power.
The core is presently at 250 EFPD.
Risk Management Action Level is GREEN.
Train A is protected.
PC is NOT recircing the RWT.
Unit 2 is supplying the Aux Steam cross-tie header.
DG B is running in accordance with 40ST-9DG02, Diesel Generator B Test, and 40OP-9DG02, Emergency Diesel Generator B.
Equipment out of service:
Diesel Generator B Planned shift activities:
Shut down DG B in accordance with 40OP-9DG02, Emergency Diesel Generator B.
NUREG-1021, Rev 9, Supp 1 10 of 11 Rev 0
Appendix D Scenario Outline Form ES-D-1 CREW HANDOUT Plant conditions:
Unit 1 is at 75% power.
The core is presently at 250 EFPD.
Risk Management Action Level is GREEN.
Train A is protected.
PC is NOT recircing the RWT.
Unit 2 is supplying the Aux Steam cross-tie header.
DG B is running in accordance with 40ST-9DG02, Diesel Generator B Test, and 40OP-9DG02, Emergency Diesel Generator B.
Equipment out of service:
Diesel Generator B Planned shift activities:
Shut down DG B in accordance with 40OP-9DG02, Emergency Diesel Generator B.
NUREG-1021, Rev 9, Supp 1 11 of 11 Rev 0