ML091050208
| ML091050208 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/03/2009 |
| From: | Ryan Lantz Operations Branch IV |
| To: | Blevins M Luminant Generation Co |
| References | |
| 50-445/09-301, 50-446/09-301 | |
| Download: ML091050208 (27) | |
Text
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
CPNPP Units 1 & 2 Date of Examination:
03/30/09 Examination Level RO
Operating Test Number:
NRC Administrative Topic (see Note)
Type Code*
Describe Activity to be Performed Conduct of Operations M, R 2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc. (3.9).
JPM: Calculate Shutdown Margin During a Cooldown From the Remote Shutdown Panel (RO1010).
Conduct of Operations M, R 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation (4.3).
JPM: Determine Time to Core Uncovery on a Loss of RHR (RO5153).
Equipment Control N, S 2.2.37 Ability to determine operability and/or availability of safety related equipment (3.6).
JPM:
Verify Pressurizer Heater Bank operability (New).
Radiation Control M, R 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions (3.5).
JPM:
Determine RWP Entry Requirements and Identify Low Dose Waiting Area (RO7533).
Emergency 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; for 4 for SROs & RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
NUREG-1021, Revision 9.1 Page 1 of 2 CPNPP Mar 2009 RO ES-301-1 Rev d
Administrative Topics Outline Task Summary NUREG-1021, Revision 9.1 Page 2 of 2 CPNPP Mar 2009 RO ES-301-1 Rev d A.1.a The candidate will perform a Shutdown Margin Calculation per OPT-301, Shutdown Margin Calculation, during a cooldown from the Remote Shutdown Panel in accordance with ABN-803A, Response to a Fire in the Control Room or Cable Spreading Room. Critical tasks include identifying individual parameters and determining the Shutdown Margin. This is a modified bank JPM.
A.1.b The candidate will determine the time to boil, time to core uncovery, and time to achieve Containment Closure on a total loss of Residual Heat Removal cooling per ABN-104, RHR System Malfunction. Critical tasks include determining the time for each evolution. This is a modified bank JPM.
A.2 The candidate will verify Pressurizer Heater Bank OPERABILITY per IPO-001A, Plant Heatup From Cold Shutdown to Hot Standby, Section 5.3, Heatup and Pressurization for MODE 3 Entry. The critical tasks include verifying Pressurizer Heater Bank capacities. This is a new JPM.
A.3 The candidate will determine the radiological requirements for entering an area in the RCA for implementing a Clearance per STA-656, Radiation Work Control. Critical tasks include identifying proper RWP, identify dress-out requirements, determine dosimetry requirements, coverage requirements, and determine what briefings are required. This is modified bank JPM.
A.4 N/A
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
CPNPP Units 1 and 2 Date of Examination:
03/30/09 Examination Level SRO
Operating Test Number:
NRC Administrative Topic (see Note)
Type Code*
Describe Activity to be Performed Conduct of Operations M, R 2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc. (4.2).
JPM: Calculate Shutdown Margin During a Cooldown From the Remote Shutdown Panel (RO1010).
Conduct of Operations M, R 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation (4.4).
JPM: Determine Time to Core Uncovery on a Loss of RHR (RO5153).
Equipment Control N, R 2.2.12 Knowledge of surveillance procedures (4.1).
JPM:
Review a completed surveillance (New).
Radiation Control N, R 2.3.6 Ability to approve release permits (3.8).
JPM:
Approve a Liquid Waste Release Permit (New).
Emergency Plan M, R 2.4.29 Knowledge of the emergency plan (4.4).
JPM:
Review a Notification Message Form (SRO 8004).
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; for 4 for SROs & RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
NUREG-1021, Revision 9.1 Page 1 of 2 CPNPP Mar 2009 SRO ES-301-1 Rev d
Administrative Topics Outline Task Summary NUREG-1021, Revision 9.1 Page 2 of 2 CPNPP Mar 2009 SRO ES-301-1 Rev d A.1.a The candidate will perform a Shutdown Margin Calculation per OPT-301, Shutdown Margin Calculation, during a cooldown from the Remote Shutdown Panel in accordance with ABN-803A, Response to a Fire in the Control Room or Cable Spreading Room. Critical tasks include identifying individual parameters and determining the Shutdown Margin. This is a modified bank JPM.
A.1.b The candidate will determine the time to boil, time to core uncovery, and time to achieve Containment Closure on a total loss of Residual Heat Removal cooling per ABN-104, RHR System Malfunction. Critical tasks include determining the time for each evolution. This is a modified bank JPM.
A.2 The candidate will perform the review of a completed Centrifugal Charging Pump surveillance per OPT-201A, Charging System. Critical tasks include identifying that unsatisfactory data is obtained during testing. This is a new JPM.
A.3 The candidate will be given a Liquid Release Permit for approval prior to release per STA-603, Control of Station Radioactive Effluents. Critical tasks include determining that release conditions will have changed and not approve the release until proper conditions are established. This is a new JPM.
A.4 The candidate will review a Notification Message Form per EPP-203-8, Notification Message Form. The critical task is to determine if the proper notifications and classification are being transmitted. This is a modified bank JPM.
ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 NUREG-1021, Revision 9.1 Page 1 of 3 CPNPP Mar 2009 ES-301-2 Rev d Facility:
CPNPP Units 1 and 2 Date of Examination:
03/30/09 Exam Level:
Operating Test No.:
NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)
System / JPM Title Type Code*
Safety Function S-1 001 - Control Rod Drive System (RO1022)
Respond to a Continuous Rod Withdrawal (RO ONLY)
A, D, S 1
S-2 004 - Chemical and Volume Control System (RO1307)
Perform Manual Makeup to the RWST D, S 2
S-3 010 - Pressurizer Pressure Control System (RO8082)
Respond to Pressurizer Spray Valve Failure A, M, S 3
S-4 005 - Residual Heat Removal System (New)
Respond to a Shutdown Loss of Coolant L, N, S 4-P S-5 026 - Containment Spray System (RO2008B)
Operate Containment Spray with Inadvertent Spray Flow A, D, EN, S 5
S-6 062 - AC Electrical Distribution System (New)
Transfer Buses from Transformer XST2 to XST1 N, S 6
S-7 016 - Non-Nuclear Instrumentation System (New)
Respond to Turbine Impulse Pressure Instrument Malfunction A, N, S 7
S-8 029 - Containment Purge System (New)
Operate the Containment Purge System A, N, S 8
In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)
P-1 062 - AC Electrical Distribution System (AO4204)
Energize a Protection System Inverter D
6 P-2 068 - Liquid Radwaste System (RO8063)
Respond to Accidental Release of Radioactive Liquid A, D, E, R 9
P-3 061 - Auxiliary/Emergency Feedwater System (AO6405)
Reset and Open TDAFW Pump Trip & Throttle Valve D, E, S 4-S
Control Room / In-Plant Systems Outline Task Summary NUREG-1021, Revision 9.1 Page 2 of 3 CPNPP Mar 2009 ES-301-2 Rev d All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (EN)gineered safety feature
- / - / 1 (control room system)
(L)ow Power / Shutdown 1 / 1 / 1 (N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)
(R)CA 1 / 1 / 1 (S)imulator NRC JPM Examination Summary Description S-1 The candidate will perform actions per ABN-712, Rod Control System Malfunction, in response to improper Control Rod motion. The alternate path includes actions to deenergize the Rod Drive Motor Generator Sets when the Control Rods fails to insert. This is a bank JPM under the Control Rod Drive System-Reactivity Control Safety Function.
S-2 The candidate will perform actions to makeup to the Refueling Water Storage Tank per SOP-104A, Reactor Makeup and Chemical Control System. This is a bank JPM under Chemical and Volume Control System-Reactor Coolant System Inventory Control Safety Function.
S-3 The candidate will perform the actions for a Pressurizer Spray Valve failing open per ABN-705, Pressurizer Pressure Malfunction. The alternate path includes actions to trip the Reactor and Reactor Coolant Pump when the Pressurizer Spray Valve fails to close. This is a modified bank JPM under the Pressurizer Pressure Control System-Reactor Pressure Control Safety Function. This is a PRA significant action.
S-4 The candidate will respond to a leak in the Reactor Coolant System per ABN-108, Shutdown Loss of Coolant. This is a new JPM under the Residual Heat Removal System-Primary System Heat Removal From Reactor Core Safety Function. This is a PRA significant action.
Control Room / In-Plant Systems Outline Task Summary NUREG-1021, Revision 9.1 Page 3 of 3 CPNPP Mar 2009 ES-301-2 Rev d S-5 The candidate will align Refueling Water Storage Tank recirculation with the Containment Spray Pump per the Alarm Response Procedures. The alternate path is to respond to an inadvertent Containment Spray flow actuation. This is a bank JPM under Containment Spray System-Containment Integrity Safety Function.
S-6 The candidate will perform the actions of SOP-603A, 6900 V Switchgear to support maintenance on Transformer XST2. This is a new JPM under AC Electrical Distribution System-Electrical Safety Function.
S-7 The candidate will respond to a Turbine Impulse Pressure Instrument Malfunction per ABN-709, Steam Line, Steam Header, Turbine 1st Stage Pressure, and Feed Header Pressure Instrument Malfunction. This is a new JPM under Non-Nuclear Instrumentation System-Instrumentation Safety Function.
S-8 The candidate will perform the actions specified in SOP-801A, Containment Ventilation System, Section 5.6.1, Containment Purge and Exhaust System Startup. The alternate path occurs when the Containment Purge System fails to isolate on a high airborne activity. This is a new JPM under Containment Purge System-Plant Service Systems Safety Function.
P-1 The candidate will perform the actions to return a Protection System Inverter to service per SOP-607A, 118V AC Distribution System and Inverters. This is a bank JPM under the AC Electrical Distribution System-Electrical Safety Function.
P-2 The candidate will perform actions for a Liquid Waste Processing Discharge high radiation alarm per ABN-903, Accidental Release of Radioactive Liquid. The alternate path requires closing of valves when it is determined that a Liquid Waste Processing Panel alarm is not dark. This is a bank JPM under the Liquid Radwaste System-Radioactivity Release Safety Function.
P-3 The candidate will locally reset the Trip and Throttle Valve on the Turbine Driven Auxiliary Feedwater Pump per ABN-305, Auxiliary Feedwater System Malfunction. This is a bank JPM under the Auxiliary Feedwater System-Secondary System Heat Removal from Reactor Core Safety Function. This is a PRA significant action.
ES-401 PWR Examination Outline Form ES-401-2 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d 1
Facility:
CPNPP 1 & 2 Date of Exam:
03/30/09 RO K/A Category Points SRO-Only Points Tier Group K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G
Total A2 G*
Total 1
3 1
4 3
4 3
18 3
3 6
2 1
1 2
3 0
2 9
3 1
4
- 1. Emergency
& Abnormal Plant Evolutions Tier Totals 4
2 6
6 4
5 27 6
4 10 1
3 2
2 4
1 1
2 3
3 3
4 28 3
2 5
2 2
0 0
1 2
1 2
1 0
0 1
10 0
2 1
3
- 2. Plant Systems Tier Totals 5
2 2
5 3
2 4
4 3
4 5
38 5
3 8
1 2
3 4
1 2
3 4
- 3. Generic Knowledge and Abilities Categories 2
2 3
3 10 2
2 1
2 7
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.
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 Form ES-401-2 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
Number K/A Topic(s)
Imp.
Q#
026 / Loss of Component Cooling Water / 8 X
2.4.41 Emergency Procedures / Plan: Knowledge of the emergency action level thresholds and classifications 4.6 76 007 / Reactor Trip - Stabilization - Recovery / 1 X
EA2.05 Ability to determine and interpret the following as they apply to a reactor trip: Reactor trip first-out indication 3.9 77 057 / Loss of Vital AC Inst. Bus / 6 X
2.4.1 Emergency Procedures / Plan: Knowledge of EOP entry conditions and immediate action steps 4.8 78 W/E04 / LOCA Outside Containment / 3 X
EA2.1 Ability to determine and interpret the following as they apply to the LOCA Outside Containment:
Facility conditions and selection of appropriate procedures during abnormal and emergency operations 4.3 79 008 / Pressurizer Vapor Space Accident / 3 X
2.1.23 Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation 4.4 80 062 / Loss of Nuclear Service Water / 4 X
AA2.03 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:
The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition 2.9 81 007 / Reactor Trip - Stabilization - Recovery / 1 X
EK1.05 Knowledge of the operational implications of the following concepts as they apply to the reactor trip: Decay power as a function of time 3.3 39 058 / Loss of DC Power / 6 X
AA2.02 Ability to determine and interpret the following as they apply to the Loss of DC Power: 125 VDC bus voltage, low/critical low, alarm 3.3 40 W/E11 / Loss of Emergency Coolant Recirculation
/ 4 X
EK2.1 Knowledge of the interrelations between the Loss of Emergency Coolant Recirculation and the following: Components, and functions of control and safety systems, including instrumentation signals, interlocks, failure modes, and automatic and manual features 3.6 41 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d 2
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 Form ES-401-2 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
Number K/A Topic(s)
Imp.
Q#
062 / Loss of Nuclear Service Water / 4 X
AA1.07 Ability to operate and/or monitor the following as they apply to the Loss of Nuclear Service Water:
Flow rates to the components and systems that are serviced by the SWS; interactions among the components 2.9 42 065 / Loss of Instrument Air / 8 X
AA2.05 Ability to determine and interpret the following as they apply to the Loss of Instrument Air: When to commence plant shutdown if instrument air pressure is decreasing 3.4 43 038 / Steam Generator Tube Rupture / 3 X
2.1.19 Conduct of Operations: Ability to use plant computers to evaluate system or component status 3.9 44 W/E05 / Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 X
EA2.2 Ability to determine and interpret the following as they apply to the Loss of Secondary Heat Sink:
Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments 3.7 45 077 / Generator Voltage and Electric Grid Disturbances / 6 X
AA1.03 Ability to operate and/or monitor the following as they apply to the Generator Voltage and Electric Grid Disturbances: Voltage regulator controls 3.8 46 026 / Loss of Component Cooling Water / 8 X
AK3.03 Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: Guidance actions contained in EOP for Loss of CCW 4.0 47 027 / Pressurizer Pressure Control System Malfunction / 3 X
AK3.04 Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunction: Why, if pressurizer level is lost and then restored, that pressure recovers much more slowly 2.8 48 009 / Small Break LOCA / 3 X
EK1.01 Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Natural circulation and cooling, including reflux boiling 4.2 49 008 / Pressurizer Vapor Space Accident / 3 X
AK1.02 Knowledge of the operational implications of the following concepts as they apply to the Pressurizer Vapor Space Accident: Change in the leak rate with change in pressure 3.1 50 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d 3
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 Form ES-401-2 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
Number K/A Topic(s)
Imp.
Q#
4 054 / Loss of Main Feedwater / 4 X
2.4.11 Emergency Procedures/Plan: Knowledge of abnormal condition procedures 4.0 51 W/E04 / LOCA Outside Containment / 3 X
EK3.3 Knowledge of the reasons for the following responses as they apply to the LOCA Outside Containment: Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations 3.8 52 056 / Loss of Offsite Power / 6 X
AK3.02 Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Actions contained in EOP for loss of offsite power 4.4 53 057 / Loss of Vital AC Instrument Bus / 6 X
2.2.37 Equipment Control: Ability to determine operability and/or availability of safety related equipment 3.6 54 015/17 / RCP Malfunctions / 4 X
AA2.02 Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunction (Loss of RC Flow): Abnormalities in RCP air vent flow paths and/or cooling oil system 2.8 55 011 / Large Break LOCA / 3 X
EA1.04 Ability to operate and monitor the following as they apply to a Large Break LOCA: ESF actuation system in manual 4.4 56 K/A Category Point Totals:
3 1
4 3
4 / 3 3 / 3 Group Point Total:
18 / 6
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 Form ES-401-2 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
Number K/A Topic(s)
Imp.
Q#
076 / High Reactor Coolant Activity / 9 X
2.4.8 Emergency Procedures / Plan: Knowledge of how abnormal operating procedures are used in conjunction with EOPs 4.5 82 W/E08 / RCS Overcooling - PTS / 4 X
EA2.1 Ability to determine and interpret the following as they apply to the Pressurized Thermal Shock:
Facility conditions and selection of appropriate procedures during abnormal and emergency operations 4.2 83 003 / Dropped Control Rod / 1 X
AA2.03 Ability to determine and interpret the following as they apply to the Dropped Control Rod: Dropped rod, using in-core/ex-core instrumentation, in-core or loop temperature measurements 3.8 84 032 / Loss of Source Range NI / 7 X
AA2.09 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: Effect of improper HV setting 2.9 85 059 / Accidental Liquid Radwaste Release / 9 X
AA1.02 Ability to operate and/or monitor the following as they apply to the Accidental Liquid Radwaste Release: ARM system 3.3 57 W/E15 / Containment Flooding / 5 X
EK3.2 Knowledge of the reasons for the following responses as they apply to the Containment Flooding: Normal, abnormal, and emergency procedures associated with Containment Flooding 2.8 58 W/E16 / High Containment Radiation / 9 X
2.4.9 Emergency Procedures/Plan: Knowledge of low power/shutdown implications in accident (e.g.,
loss of coolant accident or loss of residual heat removal) mitigation strategies 3.8 59 024 / Emergency Boration / 1 X
2.1.23 Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation 4.3 60 067 / Plant Fire on Site / 8 X
AK3.04 Knowledge of the reasons for the following responses as they apply to the Plant Fire on Site: Actions contained in EOP for plant fire on site 3.3 61 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d 5
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 Form ES-401-2 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
Number K/A Topic(s)
Imp.
Q#
6 W/E03 / LOCA Cooldown - Depressurization / 4 X
EK1.2 Knowledge of the operational implications of the following concepts as they apply to the LOCA Cooldown and Depressurization: Normal, abnormal, and emergency operating procedures associated with LOCA Cooldown and Depressurization 3.6 62 W/E08 / RCS Overcooling - PTS / 4 X
EA1.3 Ability to operate and/or monitor the following as they apply to the Pressurized Thermal Shock:
Desired operating results during abnormal and emergency situations 3.6 63 W/E06 & E07 / Inadequate Core Cooling / 4 X
EK2.2 Knowledge of the interrelations between the Saturated Core Cooling and the following:
Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of the these systems to the operation of the facility 3.5 64 036 / Fuel Handling Accident / 8 X
AA1.04 Ability to operate and/or monitor the following as they apply to the Fuel Handling Incidents: Fuel handling equipment during an incident 3.1 65 K/A Category Point Totals:
1 1
2 3
0 / 3 2 / 1 Group Point Total:
9 / 4
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 Form ES-401-2 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
Number K/A Topics Imp.
Q#
103 / Containment X
3 A2.0 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation 3.8 86 061 / Auxiliary/Emergency Feedwater X
2 A2.0 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of air to steam supply valve 3.6 87 010 / Pressurizer Pressure Control X
2.2.44 Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 4.4 88 004 / Chemical and Volume Control X
2.1.7 Conduct of Operations: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation 4.7 89 013 / Engineered Safety Features Actuation X
5 A2.0 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of DC control power 4.2 90 003 / Reactor Coolant Pump X
4 K5.0 Knowledge of the operational implications of the following concepts as they apply to the RCPs: Effects of RCP shutdown on secondary parameters, such as steam pressure, steam flow, and feed flow 3.2 1
004 / Chemical and Volume Control X
0 A1.1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: Reactor power 3.7 2
005 / Residual Heat Removal X
2.4.3 Emergency Procedures/Plan: Ability to identify post-accident instrumentation 3.7 3
NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d 7
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 Form ES-401-2 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
Number K/A Topics Imp.
Q#
006 / Emergency Core Cooling X
7 K4.2 Knowledge of ECCS design features and/or interlocks which provide for the following:
Alarm for misalignment of the accumulator isolation valve 2.9 4
006 / Emergency Core Cooling X
3 K6.1 Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Pumps 2.6 5
007 / Pressurizer Relief /
Quench Tank X
2.4.31 Emergency Procedures/Plan: Knowledge of annunciator alarms, indications, or response procedures 4.2 6
007 / Pressurizer Relief /
Quench Tank X
2.4.49 Emergency Procedures/Plan: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls 4.6 7
008 / Component Cooling Water X
9 K4.0 Knowledge of CCWS design features and/or interlocks which provide for the following: The standby feature for the CCW pumps 2.7 8
008 / Component Cooling Water X
8 A4.0 Ability to manually operate and/or monitor in the control room: CCW pump control switch 3.1 9
010 / Pressurizer Pressure Control X
1 K1.0 Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the following systems: RPS 3.9 10 012 / Reactor Protection X
8 K4.0 Knowledge of RPS design features and/or interlocks which provide for the following: Logic matrix testing 2.8 11 013 / Engineered Safety Features Actuation X
3 K3.0 Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following: Containment 4.3 12 022 / Containment Cooling X
1 A1.0 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Containment temperature 3.6 13 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d 8
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 Form ES-401-2 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
Number K/A Topics Imp.
Q#
026 / Containment Spray X
4 A2.0 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of spray pump 3.9 14 039 / Main and Reheat Steam X
2 A3.0 Ability to monitor automatic operation of the MRSS, including: Isolation of the MRSS 3.1 15 039 / Main and Reheat Steam X
1 K1.0 Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems: SG 3.1 16 059 / Main Feedwater X
3 K3.0 Knowledge of the effect that a loss or malfunction of the MFW will have on the following: SGs 3.5 17 059 / Main Feedwater X
6 A3.0 Ability to monitor automatic operation of the MFW, including: Feedwater isolation 3.2 18 061 / Auxiliary/Emergency Feedwater X
2 K2.0 Knowledge of bus power supplies to the following: AFW electric drive pumps 3.7 19 062 / AC Electrical Distribution X
4 A4.0 Ability to manually operate and/or monitor in the control room: Local operation of breakers 2.6 20 063 / DC Electrical Distribution X
1 K4.0 Knowledge of DC electrical system design features and/or interlocks that provide for the following: Manual/automatic transfers of control 2.7 21 063 / DC Electrical Distribution X
1 K2.0 Knowledge of bus power supplies to the following: Major DC loads 2.9 22 064 / Emergency Diesel Generator X
8 A2.0 Ability to (a) predict the impacts of the following malfunctions or operations on the EDG system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Consequences of opening/closing breakers between buses (VARS, out-of-phase, voltage) 2.7 23 064 / Emergency Diesel Generator X
2.2.38 Equipment Control: Knowledge of conditions and limitations in the facility license 3.6 24 073 / Process Radiation Monitoring X
1 A4.0 Ability to manually operate and/or monitor in the control room: Effluent release 3.9 25 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d 9
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 Form ES-401-2 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
Number K/A Topics Imp.
Q#
10 X
1 076 / Service Water K1.0 Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems: CCW system 3.4 26 078 / Instrument Air X
1 A3.0 Ability to monitor automatic operation of the IAS, including: Air pressure 3.1 27 103 / Containment X
4 A2.0 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Containment evacuation (including recognition of the alarm) 3.5 28 K/A Category Point Totals:
3 2
2 4
1 1
2 3 / 3 3
3 4 / 2 Group Point Total:
28 / 5
ES-401 CPNPP 1 & 2 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
Number K/A Topics Imp.
Q#
11 X
1 015 / Nuclear Instrumentation A2.0 Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Power supply loss or erratic operation 3.9 91 071 / Waste Gas Disposal X
2.4.46 Emergency Procedures/Plan: Ability to verify that the alarms are consistent with the plant conditions 4.2 92 035 / Steam Generator X
1 A2.0 Ability to (a) predict the impacts of the following malfunctions or operations on the SG; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulted or ruptured SGs 4.6 93 027 / Containment Iodine Removal X
1 K5.0 Knowledge of the operational implications of the following concepts as they apply to the CIRS: Purpose of charcoal filters 3.1 29 055 / Condenser Air Removal X
6 K1.0 Knowledge of the physical connections and/or cause-effect relationships between the CARS and the following systems: PRM system 2.6 30 068 / Liquid Radwaste X
1 K4.0 Knowledge of design features and/or interlocks that provide for the following:
Safety and environmental precautions for handling hot, acidic and radioactive liquids 3.4 31 041 / Steam Dump/Turbine Bypass Control X
1 A1.0 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SDS controls including: Tave, verification above low/low setpoint 2.9 32 002 / Reactor Coolant X
5 K1.0 Knowledge of the physical connections and/or cause-effect relationships between the RCS and the following systems: PRT 3.2 33 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d
ES-401 CPNPP 1 & 2 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 Form ES-401-2 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
Number K/A Topics Imp.
Q#
12 X
4 014 / Rod Position Indication A1.0 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RPIS controls, including: Axial and radial power distribution 3.5 34 086 / Fire Protection X
4 K6.0 Knowledge of the effect of a loss or malfunction on the Fire Protection System will have on the: Fire, smoke, and heat detectors 2.6 35 001 / Control Rod Drive X
4 A2.0 Ability to (a) predict the impacts of the following malfunctions or operations on the CRDS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Positioning of axial shaping rods and their effect on SDM 3.2 36 071 / Waste Gas Disposal X
2.1.28 Conduct of Operations: Knowledge of the purpose and function of major system components and controls 4.1 37 072 / Area Radiation Monitoring X
1 K5.0 Knowledge of the operational implications of the following concepts as they apply to the ARM system: Radiation theory, including sources, types, units, and effects 2.7 38 K/A Category Point Totals:
2 0
0 1
2 1
2 1 / 2 0
0 1 / 1 Group Point Total:
10 / 3
ES-401 Generic Knowledge and Abilities Outline (Tier3)
Form ES-401-3 Facility: CPNPP 1 & 2 Date of Exam: 03/30/09 RO SRO-Only Category K/A #
Topic IR IR 2.1.6 Ability to manage the control room crew during plant transients 4.8 94 2.1.34 Knowledge of primary and secondary plant chemistry limits 3.5 95 2.1.3 Knowledge of shift or short-term relief turnover practices 3.7 66 2.1.17 Ability to make accurate, clear, and concise verbal reports 3.9 67
- 1.
Conduct of Operations Subtotal 2
2 2.2.35 Ability to determine Technical Specification Mode of Operation 4.5 96 2.2.7 Knowledge of the process for conducting special or infrequent tests 3.6 97 2.2.1 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant and equipment that could affect reactivity 4.5 68 2.2.40 Ability to apply Technical Specifications for a system 3.4 69
- 2.
Equipment Control Subtotal 2
2 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
3.1 98 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions 3.5 70 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities 3.4 71 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.
3.4 72
- 3.
Radiation Control Subtotal 3
1 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator 4.1 99 2.4.11 Knowledge of the abnormal condition procedures 4.2 100 2.4.43 Knowledge of emergency communications systems and techniques 3.2 73 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions 3.7 74 2.4.31 Knowledge of annunciator alarms, indications, or response procedures 4.2 75
- 4.
Emergency Procedures /
Plan Subtotal 3
2 Tier 3 Point Total 10 7
NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d 13
ES-401 Record of Rejected K/As Form ES-401-4 NUREG-1021 Rev 9.1 CPNPP Mar 2009 NRC Outline Rev d 14 Tier /
Group Randomly Selected K/A Reason for Rejection 1 / 1 062 AA2.05 Q #81 - Randomly reselected 062 AA2.03. Unable to develop an appropriate SRO level question with the selected K/A. Similar to Question #42.
1 / 1 026 AK3.01 Q #47 - Randomly reselected 026 AK3.03. There are no automatic Service Water System valves that service the CCW Heat Exchangers.
1 / 2 W/E16 G 2.2.3 Q #59 - Randomly reselected W/E16 G 2.4.9. There is no operational difference between Units for the Containment High Radiation Monitors.
2 / 1 073 K4.02 Q #24 - There is no automatic letdown isolation on high RCS activity at CPNPP. Randomly reselected 006 K6.13 (Q #05) to meet Tier 2 K6 skyscraper requirements.
2 / 2 055 A3.03 Q #35 - There is no automatic diversion of CARS exhaust at CPNPP.
Randomly reselected 086 K6.04 to meet Tier 2 K6 skyscraper requirements.
2 / 2 072 A3.01 Q #38 - Randomly reselected 072 K5.01. There are no changes in ventilation alignments associated with the Area Radiation Monitoring System.
2 / 1 005 G 2.2.25 Q #03 - Randomly reselected 005 G 2.4.3. Unable to develop an appropriate RO level question with the selected K/A.
2 / 2 014 K3.02 Q #34 - Randomly reselected 014 A1.04. The Rod Position Indicating System does not interface with the Plant Computer at CPNPP.
2 / 2 075 A4.01 Q #30 - Randomly reselected 075 K1.02. There is no relationship between the Station Service Water System Pumps and the Circulating Water System at CPNPP.
2 / 2 075 K1.02 Q #30 - Randomly reselected 055 K1.06. Unable to develop an appropriate RO level question with the selected K/A. Similar to Question #25.
1 / 1 027 AK3.02 Q #48 - Randomly reselected 027 AK3.04. Unable to develop an appropriate RO level question with the selected K/A.
3 / 3 G 2.3.11 Q #72 - Randomly reselected G 2.3.13. Original topic adequately covered on exam Q #25.
3 / 4 G 2.4.46 Q #99 - Reselected G 2.4.30. Original topic adequately covered on exam.
Appendix D Scenario Outline Form ES-D-1 CPNPP March 2009 NRC Sim Scenario ES-D-1 Rev f.doc Facility:
CPNPP 1 & 2 Scenario No.:
1 Op Test No.:
March 2009 NRC Examiners:
Operators:
Initial Conditions:
67% power MOC - RCS Boron is 922 ppm (by sample).
Motor Driven Auxiliary Feedwater Pump 1-02 OOS for coupling repair. RTS in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Turnover:
Maintain steady-state power conditions while Heater Drain Pump repairs are made.
Critical Tasks:
Emergency borate due to two (2) stuck Control Rods.
Perform actions to identify and isolate the ruptured Steam Generator.
Perform actions to cooldown and depressurize the Reactor Coolant System.
Event No.
Malf. No.
Event Type*
Event Description 1
+ min TC09D FW03B C (BOP, SRO)
Main Feedwater Pump B trip with Auto Turbine Runback failure.
2
Letdown Isolation Valve (HV-8160) fails closed.
Place Excess Letdown in service.
3
TS (SRO)
Station Service Water Pump (1-02) trip.
4
TS SRO Steam Generator #1 Tube Leak at 175 gpd.
5
+ min SG01A M (ALL)
Steam Generator #1 Tube Rupture at 650 gpm.
6
+ min RD04M14 RD04P8 C (RO)
Two (2) Control Rods fail to insert on Reactor trip. Emergency boration required.
7
+ min RH01C C (BOP)
Residual Heat Removal Pump 1-01 fails to auto start on Safety Injection signal.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications
Appendix D Scenario Outline Form ES-D-1 CPNPP March 2009 NRC Sim Scenario ES-D-1 Rev f.doc SCENARIO
SUMMARY
NRC #1 The crew will assume the watch with power reduced due to a Heater Drain Pump Seal failure. Power is maintained per IPO-003A, Power Operations until repairs are made.
The first event is a trip of Main Feedwater Pump B with an automatic Turbine Runback failure. The crew responds per ABN-302, Feedwater, Condensate, Heater Drain System Malfunction. When it is determined that automatic plant response has not activated, a manual Turbine Runback will be initiated. The crew will stabilize load at 700 MWe.
When ABN-302 actions are complete, a loss of Letdown occurs due to a Letdown Isolation Valve failing closed. Actions are per ABN-105, Chemical and Volume Control System Malfunction, and require controlling Charging and Seal Injection flows until Letdown can be restored. The RO will be directed to place Excess Letdown in service.
This is followed by a trip of Station Service Water Pump 1-02. The crew will enter ABN-501, Station Service Water System Malfunction. Initial operator actions include placing the Train B Emergency Diesel Generator in PULL-OUT. The SRO will refer to Technical Specifications.
The next event is a Steam Generator #1 tube leak of 175 gpd. Crew actions are per ABN-106, High Secondary Activity. The SRO will refer to Technical Specifications.
When Technical Specifications are referenced, a Steam Generator #1 tube rupture occurs and leakage rises to 650 gpm. An uncontrolled loss of Pressurizer level will require a Reactor trip, initiation of Safety Injection and entry into EOP-0.0A, Reactor Trip or Safety Injection. At Step 13, a transition to EOP-3.0A, Steam Generator Tube Rupture will occur to isolate the ruptured Steam Generator. The event is complicated when two Control Rods fail to insert. Additionally, Residual Heat Removal Pump 1-01 fails to auto start on the Safety Injection signal.
The scenario is terminated when the ruptured Steam Generator is isolated, feedwater flow is properly aligned and the Reactor Coolant System cooldown and depressurization is commenced.
Risk Significance:
Risk important components out of service:
Auxiliary Feedwater Pump 1-02 Failure of risk important system prior to trip:
Station Service Water Pump 1-01 Risk significant core damage sequence:
Steam Generator Tube Rupture Risk significant operator actions:
Emergency borate due to 2 stuck rods Start Residual Heat Removal Pump 1-01 Identify and isolate the ruptured SG Cooldown and depressurize the RCS
Appendix D Scenario Outline Form ES-D-1 CPNPP March 2009 NRC Sim Scenario ES-D-1 Rev f.doc Facility:
CPNPP 1 & 2 Scenario No.:
2 Op Test No.:
March 2009 NRC Examiners:
Operators:
Initial Conditions:
~1X10-8 amps BOC - RCS Boron is 1545 ppm (by sample).
Steam Dump System in service for RCS Temperature Control.
Turnover:
Raise Power to 2% in preparation for plant startup to 100% power.
Critical Tasks:
Restore feedwater flow from the Turbine Driven Auxiliary Feedwater Pump.
Restore power to at least one 6900 VAC Safeguards Bus.
Initiate emergency boration due to loss of Digital Rod Position Indication.
Event No.
Malf. No.
Event Type*
Event Description 1
+ min R (RO)
Raise Reactor power to 2%.
2
TS (SRO)
Motor Driven Auxiliary Feedwater Pump (1-01) trip.
3
Volume Control Tank Level Transmitter (LT-112) fails low.
4
TS (SRO)
Pressurizer Pressure Channel (PT-455) fails low.
5
+ min Tornado Warning from the National Weather Service.
6
+ min ED01 M (ALL)
Loss of All AC Power due to Loss of Offsite Power.
7
+ min EG06A C (BOP)
Emergency Diesel Generator (1-01) fails to start.
8
+ min EG15B C (BOP)
Emergency Diesel Generator (1-02) fails to auto and emergency start; EDG (1-02) starts upon Normal Start Switch actuation.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications
Appendix D Scenario Outline Form ES-D-1 CPNPP March 2009 NRC Sim Scenario ES-D-1 Rev f.doc SCENARIO
SUMMARY
NRC #2 The crew takes the shift with a Plant Startup in progress and will continue raising power to approximately 2% per IPO-002A, Plant Startup from Hot Standby.
When conditions are stable, Motor Driven Auxiliary Feedwater Pump 1-01 will trip. The crew will refer to ABN-305, Auxiliary Feedwater System Malfunction and determine that Steam Generator levels are slowly decreasing and start the Turbine Driven Auxiliary Feedwater Pump. The SRO will refer to Technical Specifications.
The next event is the failure of Volume Control Tank level transmitter LT-112. Actions per ABN-105, Chemical and Volume Control System Malfunction will be performed. This event will disable automatic makeup flow to the Volume Control Tank.
When ABN-105 actions are complete, Pressurizer Pressure Channel PT-455 will fail low. The crew will respond per ABN-705, Pressurizer Pressure Malfunction, and ensure Pressurizer Heaters are controlled and Power Operated Relief Valves remain closed. The SRO will refer to Technical Specifications.
When plant conditions are stable, a tornado warning from the National Weather Service will require entry into ABN-907, Acts of Nature. This is the precursor event to a Loss of Offsite Power.
The next event is a Loss of Offsite Power with a failure on each Emergency Diesel Generator resulting in a Total Loss of All AC. The crew enters EOP-0.0A, Reactor Trip or Safety Injection and then exits to ECA-0.0A, Loss of All AC Power, at Step 3. As an alternative the Unit Supervisor may directly enter ECA-0.0A. While in ECA-0.0A, the crew will take actions that start Emergency Diesel Generator 1-02.
The RO may initiate an emergency boration due to loss of Digital Rod Position Indication (DRPI) while in ECA-0.0A if rod position is not observed prior to loss of DRPI.
Once Emergency Diesel Generator 1-02 is started and the breaker is closed in ECA-0.0A, a return to the procedure and step in effect is required. The crew will return to EOP-0.0A and then transition to EOS-0.1A, Reactor Trip Response when it is determined that Safety Injection is not required.
The scenario is terminated after EOS-0.1A, Reactor Trip Response is entered and the actions to stabilize the Unit are performed.
Risk Significance:
Risk important components out of service:
None Failure of risk important system prior to trip:
Loss of MDAFW Pump 1-01 Risk significant core damage sequence:
Loss of all AC Power Risk significant operator actions:
Restore Auxiliary Feedwater flow Restore power to 6.9 KV Safeguards Bus Initiate emergency boration
Appendix D Scenario Outline Form ES-D-1 CPNPP March 2009 NRC Sim Scenario ES-D-1 Rev f.doc Facility:
CPNPP 1 & 2 Scenario No.:
3 Op Test No.:
March 2009 NRC Examiners:
Operators:
Initial Conditions:
100% power MOC - RCS Boron is 910 ppm (by sample).
Motor Driven Auxiliary Feedwater Pump 1-02 OOS for coupling repair. RTS in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Turnover:
Maintain steady-state power conditions.
Critical Tasks:
Perform actions to identify and isolate faulted Steam Generators.
Manually actuate both trains of Phase A Containment Isolation.
Secure Reactor Coolant Pumps due to a loss of subcooling.
Event No.
Malf. No.
Event Type*
Event Description 1
TS (SRO)
Reactor Coolant System Loop 2 Tcold Transmitter (TI-421B) fails high.
2
TS (SRO)
Steam Generator #4 Level Transmitter (LT-554) fails low.
3
- 1 Seal failure on Reactor Coolant Pump #2 between 6 and 8 gpm.
4
+ min R (RO)
Unit Power reduction.
5
- 2 Seal failure on Reactor Coolant Pump #2 requiring manual trip.
6
+ min MS02 M (ALL)
Main Steam header leak.
Small Break Loss of Coolant Accident.
7
+ min MS08 A/B/C/D C (RO)
Main Steam Isolation Valves fail to automatically close.
8
+ min RP09A RP09B C (BOP)
Train A Containment Isolation Phase A actuation failure.
Train B Containment Isolation Phase A actuation failure.
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications
Appendix D Scenario Outline Form ES-D-1 CPNPP March 2009 NRC Sim Scenario ES-D-1 Rev f.doc SCENARIO
SUMMARY
NRC #3 The crew initially takes the shift at 100% power with no scheduled activities per IPO-003A, Power Operations. Auxiliary Feedwater Pump 1-02 is out-of-service for coupling repair.
The first event it is a failure of Tcold instrument, TI-421B. Operator actions are per ABN-704, TC/N-16 Instrumentation Malfunction and require stopping Control Rod motion and stabilizing RCS temperature and Pressurizer level. The SRO will refer to Technical Specifications.
Once systems are stable, a Feedwater Flow instrument fails low. Operator response is per ABN-708, Feedwater Flow Instrument Malfunction. The operator must take manual control of the affected Feed Control Valve to prevent a Unit trip on high Steam Generator water level. After manual control is established, an alternate channel is selected and automatic control restored.
The next event is a failure of #1 Seal on Reactor Coolant Pump #2. Crew actions are per ABN-101, Reactor Coolant Pump Trip/Malfunction. The crew should determine that based on leakoff flow and stable seal temperatures an orderly Unit shutdown to have the RCP secured within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is required.
At this point the crew will commence a power reduction per IPO-003A, Power Operations. When a sufficient down power has been achieved, the #2 Seal on Reactor Coolant Pump #2 will fail requiring a Reactor trip, stopping of the RCP and isolation of seal return from the affected pump per ABN-101, Reactor Coolant Pump Trip/Malfunction.
On the Unit trip, a Main Steam header leak will occur downstream of the Main Steam Isolation Valves.
Additionally, the Main Steam Isolation Valves (MSIV) fail to automatically or manually close along with a Train A and Train B Containment Isolation Phase A failure.
The crew enters EOP-0.0A, Reactor Trip or Safety Injection and then transitions to EOP-2.0A, Faulted Steam Generator Isolation. Once entry into EOP-2.0A is made, a transition to ECA-2.1A, Uncontrolled Depressurization of All Steam Generators is required. When Auxiliary Feedwater flow is throttled in ECA-2.1A, the MSIVs on all Steam Generators will be closed. FRH-0.1A, Response to Loss of Secondary Heat Sink will be entered and exited due to an operator induced reduction in feedwater flow.
The crew will transition from ECA-2.1A, Uncontrolled Depressurization of All Steam Generators back to EOP-2.0A, Faulted Steam Generator Isolation and then to EOP-1.0A, Loss of Reactor or Secondary Coolant.
This scenario is terminated when a Reactor Coolant System cooldown and depressurization is commenced.
Risk Significance:
Risk important components out of service:
Auxiliary Feedwater Pump 1-02 Risk significant core damage sequence:
Faulted SG with SBLOCA Risk significant operator actions:
Initiate Phase A Containment Isolation Manually Initiate MSIV Closure Isolate all Faulted SGs