ML21193A321
| ML21193A321 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/16/2021 |
| From: | Thomas Setzer Operations Branch I |
| To: | Exelon Generation Co |
| Shared Package | |
| ML20261H327 | List: |
| References | |
| EPID L-2021-OLL-0028 | |
| Download: ML21193A321 (24) | |
Text
Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #2 OP-Test # 2021 Examiners:
Operators:
Initial Conditions: Unit-1 is at 100% power, MOC. Unit-2 is at 100% power.
Turnover: 13 Component Cooling Pump is OOS.
Instructions to the crew: The Shift Manager directs the crew to commence equalizing RCS boron per OI-1H Section 6.6 to support activities scheduled next shift.
Critical Tasks:
- 1. Trips the reactor within 1 minute of the PROT CH TRIP alarm.
- 2. Commences a rapid RCS Cooldown not to exceed 100°F in any one hour.
- 3. Commences OTCC when both S/G levels are below (-)350 inches and prior to CET temperatures reaching 560°F after Heat Removal capability has been lost.
Event #
Malfunction #
Event Type*
Event Description 1
N/A N-ATC/SRO Commence Boron Equalization / OI-1H 2
rcs026_01 I-ATC/SRO T-SRO LT-110X Fails Low 3
1-CC-273 C-BOP/SRO T-SRO 11 CCW Pump Leak / AOP-7C 4
cd001 C-BOP/SRO R-ATC Loss of Condenser Vacuum / AOP-7G 5
rps005 rps006 C-ATC/SRO ATWS EOP-0 6
swyd002 C-ALL Loss of Offsite Power 7
1-AFW-161 M-ALL AFW Suction Line Block Loss of All Feedwater / EOP-3 (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor (T)ech Spec
Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #2 OP-Test # 2021 Page 2 of 18 Scenario Overview Initial Conditions:
Unit-1 is at 100% power, MOC, Unit-2 at 100% power.
Equipment OOS: 13 Component Cooling Pump.
Abnormal Conditions: None.
Instructions for shift: The Shift Manager directs the crew to commence equalizing RCS boron per OI-1H Section 6.6 to support activities scheduled next shift.
Event 1 - The ATC Operator will commence equalizing RCS boron per OI-1H Section 6.6 by starting all Backup Heaters and lowering the setpoint of the Pressurizer Pressure Controllers.
Event 2 - LT-110X, the PZR Level Controller process variable transmitter, will fail low. The ATC Operator will perform the alarm manual actions to shift the PZR Pressure and Low-Level Cutoff controls to Channel Y. The crew is expected to enter TS 3.3.10.A and determine the required action is to restore the indication channel to operable status with a required completion time of 30 days.
Event 3 - Next, a leak from 11 Component Cooling Pump will occur. The crew will implement AOP-7C, start 12 CCW Pump, and secure 11 CCW Pump. The crew will determine Tech Spec LCO 3.7.5.A applies with a required action to restore the CC loop to operable status with a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Event 4 - A Loss of Main Condenser Vacuum will occur requiring the crew to recognize the failure by determining Main Generator electrical MW and/or condenser vacuum is lowering. The crew will respond using AOP-7G, start the available Condenser Air Removal unit, and then commence a rapid downpower. After reactor power is lowered to 90%, condenser vacuum will rapidly lower to less than the Main Turbine trip criteria.
Event 5 - When the Main Turbine trip criteria is reached, an ATWS occurs when RPS fails to automatically trip the reactor. The ATC Operator is expected to recognize the malfunction, determine the reactor trip push buttons also fail, and take manual actions to perform the Critical Task of tripping the reactor within 1 minute of the PROT CH TRIP alarm by deenergizing the CEDM MG set electrical buses. Then, the remainder of EOP-0, Post-Trip Immediate Actions will be implemented.
Event 6 - After the Critical Task of tripping the reactor is performed, a Loss of Offsite Power will occur. The crew will be required to restore vital auxiliary functions such as Component Cooling and Charging Pump flow manually.
Event 7 - A total blockage of the AFW Suction Line occurs causing a loss of all sources of feedwater. The crew will implement EOP-3, Loss of All Feedwater, and will perform the Critical Task of commencing a rapid RCS Cooldown not to exceed 100°F in any one hour. When Once Through Core Cooling entry conditions are met, the crew will perform the Critical Task of commencing OTCC when both S/G levels are below (-)350 inches and prior to CET temperatures reaching 560°F after Heat Removal capability has been lost. The scenario ends after this final Critical Task is complete.
Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #3 OP-Test # 2021 Examiners:
Operators:
Initial Conditions: Unit-1 is at 100% power, MOC. Unit-2 is at 100% power.
Turnover: 13 Condensate Booster Pump is OOS.
Instructions to the crew: The Shift Manager directs the crew to shift Charging Pumps by starting 12 and securing 11 Charging Pump per OI-2A Section 6.2 to support activities scheduled next shift.
Critical Tasks:
- 1. Trip 11A & 12B RCPs or 11B & 12A RCPs when RCS pressure decreases to <1725 PSIA prior to RCS pressure going below 1300 PSIA.
- 2. Establishes at least one train of High-Pressure Safety Injection flow to the RCS within 15 minutes of the SIAS auto failure.
- 3. Identifies 12 Steam Generator as faulted and isolates 12 S/G.
Event #
Malfunction #
Event Type*
Event Description 1
N/A N-ATC/SRO Shift Charging Pumps / OI-2A 2
ms019_03 I-BOP/SRO T-SRO 12 S/G LT1123C Fails Low 3
rcs003 C-ALL T-SRO RCS Leak Inside Containment / AOP-2A 4
U25-11-LCL-BKR C-BOP/SRO R-ATC Transformer Loss of Cooling / AOP-7E 5
esfa001_01 esfa001_02 C-ATC/SRO SIAS Auto Failure 7
ms016_09 M-ALL 12 S/G Steam Line Rupture Outside Containment EOP-8 (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor (T)ech Spec
Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #3 OP-Test # 2021 Page 2 of 18 Scenario Overview Initial Conditions:
Unit-1 is at 100% power, MOC, Unit-2 at 100% power.
Equipment OOS: 13 Condensate Booster Pump.
Abnormal Conditions: None.
Instructions for shift: The Shift Manager directs the crew to shift Charging Pumps by starting 12 and securing 11 Charging Pump per OI-2A Section 6.2 to support activities scheduled next shift.
Event 1 - The ATC Operator will shift Charging Pumps by starting 12 and securing 11 Charging Pump per OI-2A Section 6.2.
Event 2 - 12 S/G Level Transmitter, LT-1123C, will fail low. The crew will respond per the 1C03 and 1C05 Alarm Manuals and determine the LT failure impacts both RPS and ESFAS.
They may use the OP-CA-103-102-0200 (Common Tap Analysis attachment 5) to determine that the failure is isolated to LT-1123C only. The crew will bypass the affected RPS Channel C Trip unit using OI-6 (T/U-4 SG Level) and direct the bypassing of the affected ESFAS Sensor ZF modules per OI-34 (SG Hi Level Turbine Trip). The crew is expected to enter TS 3.3.1.A and determine the required actions are to place the affected bistable trip unit 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 /> and to restore the affected bistable trip unit and measurement channel to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or to place the affected bistable trip unit in trip within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Event 3 - Next, an RCS leak inside Containment will occur. The crew will implement AOP-2A Section V, RCS Leakage Within the Capacity of One Charging Pump, which will direct their actions to control PZR level. The crew will determine Tech Spec LCO 3.4.13.A applies with a required action to reduce RCS leakage to within limits with a completion time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Event 4 - Then, a Loss of Main Transformer Cooling will occur requiring the crew to implement AOP-7E. The BOP will reduce Main Generator MVARs to zero and dispatch Equipment Operators to investigate. The crew will determine the extent of the issue and commence a rapid downpower per OP-3.
Event 5 - When reactor power reaches 90%, a LOCA inside Containment commences requiring the crew to trip the reactor or verify an automatic RPS trip occurs. Then, the remainder of EOP-0, Post-Trip Immediate Actions will be implemented.
Events 6/7 - The crew will recognize that SIAS fails to automatically actuate and will perform the Critical Task of establishing at least one train of High-Pressure Safety Injection flow to the RCS within 15 minutes of the SIAS auto failure. The crew will also perform the Critical Task of tripping 11A & 12B RCPs or 11B & 12A RCPs when RCS pressure decreases to <1725 PSIA prior to RCS pressure going below 1300 PSIA. After these Critical Tasks are performed and during EOP-0, a 12 S/G Steam Line Rupture Outside Containment will occur. The crew will implement the functional recovery procedure EOP-8 and will perform the final Critical Task to identify 12 Steam Generator as faulted and isolate 12 S/G.
Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #4 OP-Test # 2021 Examiners:
Operators:
Initial Conditions: Unit-1 is at 50% power, MOC. Unit-2 is at 100% power.
Turnover: 12 Boric Acid Pump and 13 Component Cooling Pump are OOS.
Instructions to the crew: The Shift Manager directs the crew to lower reactor power and stabilize at 45% per OP-3 to support activities scheduled next shift.
Critical Tasks
- 1. Establishes AFW flow to at least one S/G prior to S/G levels going below (-)350 inches.
- 2. Identifies 11 S/G as the ruptured SG and isolates 11 S/G within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the reactor trip.
Event #
Malfunction #
Event Type*
Event Description 1
N/A N-BOP/SRO R-ATC Lower Reactor Power 2
rcs024_04 I-BOP/SRO T-SRO RCS Pressure PT-102D Fails Low 3
480v002_01 C-ALL T-SRO Loss of MCC-104 / AOP-7I 4
cd008 C-ALL Condensate Header Rupture / AOP-3G 5
ms002_01 M-ALL EOP-0 SGTR (2 tubes) in 11 SG 6
P1C01 EXDISC C-BOP/SRO Generator Field/Exciter Breakers Fail to Open 7
125v001_02 M-ALL Loss of 12 DC Bus / EOP-8 (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor (T)ech Spec
Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #4 OP-Test # 2021 Page 2 of 19 Scenario Overview Initial Conditions:
Unit-1 at 50% power, MOC, Unit-2 at 100% power.
Equipment OOS: 12 Boric Acid Pump and 13 Component Cooling Pump.
Abnormal Conditions: None.
Instructions for shift: The Shift Manager directs the crew to lower reactor power and stabilize at 45% per OP-3 to support activities scheduled next shift.
Event 1 - The crew will lower reactor power to 45% per OP-3.
Event 2 - RCS Pressure transmitter, 1-PT-102D, will fail low. Alarm Response Manual 1C05 actions will have crew investigate failure at 1C15. When the failure is recognized, the crew should reference OP-CA-103-102-0200 and bypass RPS Channel D trip units 6 & 7, and ESFAS Channel ZG sensor modules for SIAS PP, SIAS PPB, and DSS. The crew is expected to enter TS 3.3.1.A and 3.3.4.A and determine the required actions are to place the trip units or sensor modules 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 /> and to restore the trip units or sensor modules to operable status or in trip status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Event 3 - Once the trip units are bypassed, a loss of MCC-104 will occur. The crew will implement AOP-7I Section XXV, Loss of Reactor MCC 104R, which will direct their actions to secure the running Charging Pumps and realign the suction back to the VCT. The crew will determine Tech Spec LCO 3.8.9.A applies with a required action to restore MCC-104 to operable status with a completion time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Event 4 - A Condensate Header Rupture will occur. The crew will respond using AOP-3G, Malfunction of Main Feedwater System, and will trip the reactor, perform Reactivity Control of EOP-0, trip both SGFPs, secure secondary feedwater pumps, shut the SG Feedwater MOVs, and start an AFW Pump. Then, the remainder of EOP-0, Post-Trip Immediate Actions will be implemented.
Event 5/6 - On the reactor trip, a 2 tube SGTR on 11 SG will occur. Also, on the reactor trip, the Main Generator Field/Exciter Breakers will fail to Open. The Balance of Plant Operator is expected to recognize the malfunction and take manual action to open the Field and Exciter Breakers.
Event 7 - 5 minutes after the reactor trip, a Loss of 12 125V DC Bus will occur. The crew is expected to implement EOP-8. The crew should identify the success paths in EOP-8 (RC-1 Met, VA-1 Not Met, PIC-4 Met, HR-2 Met, CE-1 Met, RLEC-2 Not Met) and priority (VA-1, RLEC-2, PIC-4, RC-1, HR-2, CE-1). Crew will commence VA-1 and RLEC-2 and then the remaining success paths as required. As part of VA-1, the crew will identify that concurrent implementation of AOP-7J is required due to the loss of 12 125V DC Bus. As part of RLEC-2, the crew will commence a cooldown, at a rate not to exceed 100F/hr, to prepare to isolate 11 S/G. Once Th
<515°F, the crew will perform the Critical Task to isolate 11 S/G. When the RCP Net Positive Suction Head pump curve is violated, the crew will perform the Critical Task of securing any running RCPs. The scenario will end after this Critical Task is complete.
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Exam Level: RO Date of Examination: 6/6/2021 Operating Test #: 2021 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations R, M Calculate BAST volume required to raise RWT to refueling boron concentration G2.1.20 Ability to interpret and execute procedure steps. (RO-4.6)
Conduct of Operations R, N Determine Power Ratio Recorder Potentiometer setting with the plant computer failed G2.1.7 Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation. (RO-4.4)
Radiation Control R, D Determine requirements and dose limits associated with performance of a task in the RCA G2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions. (RO-3.5)
Equipment Control R, D Respond to Condenser Vacuum reduction (Evaluate U-2 Load versus Vacuum)
G2.2.44 Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions. (RO-4.2)
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) [2]
(N)ew or (M)odified from bank ( 1) [2]
(P)revious 2 exams ( 1; randomly selected) [0]
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Exam Level: RO Date of Examination: 6/6/2021 Operating Test #: 2021 Control Room Systems:
Safety Function
- a.
SIM-1 Respond to a charging header rupture 004 Chemical and Volume Control System A4.08 Charging (RO-3.8/SRO-3.4)
A, N 1
- b. SIM-2 Align a LPSI Pump for Core Flush 006 Emergency Core Cooling System A4.07 ECCS pumps and valves (RO-4.4/SRO-4.4)
A, D, EN, L 2
- c.
SIM-3 Override Shut a PORV 010 Pressurizer Pressure Control System A4.03 PORV and block valves (RO-4.0/SRO-3.8)
A, D, L 3
- d. SIM-4 Respond to a loss of Secondary Pumps 056 Condensate System A2.04 Loss of condensate pumps (RO-2.6/SRO-2.8)
A, N 4S
- e.
SIM-5 Shift 13 IRU Power Supply 027 Containment Iodine Removal System (CIRS)
A4.01 CIRS controls (RO-3.3/SRO-3.3)
D, EN 5
- f.
SIM-6 Perform Vital Auxiliaries 062 AC Electrical Distribution System A4.01 All Breakers (RO-3.3/SRO-3.1)
N, L 6
- g. SIM-7 Reactor Protection System Test 012 Reactor Protection System A4.04 Bistables, trips, reset and test switches (RO-3.3/SRO-3.3)
N 7
- h. SIM-8 Shift Component Cooling Heat Exchangers 008 Component Cooling Water System A4.01 CCW indications and controls (RO-3.3/SRO-3.1)
D 8
In-Plant Systems
- 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U
- a.
PLT-1 Start 11 & 12 Containment Air Coolers 022 Containment Cooling System A4.01 CCS Fans (RO-3.6/SRO-3.6)
D, E, R 5
- b. PLT-2 Respond to a fire on site 086 Fire Protection System (FPS)
AA4.05 Deluge Valves (RO-3.0/SRO-3.5)
N 8
- c. PLT-3 Control RCS and S/G Inventory from 2C43 035 Steam Generator System A3.01 S/G Water Level Control (RO-4.0/SRO-3.9)
D, E 4P
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
- All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6 / 4-6 / 2-3 [4]
9 / 8 / 4 [6]
1 / 1 / 1 [2]
1 / 1 / 1 (control room system) [2]
1 / 1 / 1 [3]
2 / 2 / 1 [5 including 2(A)]
3 / 3 / 2 (randomly selected) [0]
1 / 1 / 1 [1]
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Exam Level: SRO-I Date of Examination: 6/6/2021 Operating Test #: 2021 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations R, N Determine Operator Qualifications G2.1.8 Ability to coordinate personnel activities outside the control room. (SRO-4.1)
Conduct of Operations R, D Review and approve a completed Surveillance Test Procedure G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (SRO-4.7)
Equipment Control R, D Establish Initial Conditions for STP O-8A-1 G2.2.12 Knowledge of surveillance procedures. (SRO-4.1)
Radiation Control R, N Review and Authorize KI G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (SRO-3.8)
Emergency Plan R, D Recommend Protective Action Guidelines to Public Officials G2.4.44 Knowledge of emergency plan protective action recommendations (SRO-4.4)
NOTE: All items (five total) are required for SROs. RO applicants require only four 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) [3]
(N)ew or (M)odified from bank ( 1) [2]
(P)revious 2 exams ( 1; randomly selected) [0]
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Exam Level: SRO-I Date of Examination: 6/6/2021 Operating Test #: 2021 Control Room Systems:
Safety Function
- a.
SIM-1 Respond to a charging header rupture 004 Chemical and Volume Control System A4.08 Charging (RO-3.8/SRO-3.4)
A, N 1
- b. SIM-2 Align a LPSI Pump for Core Flush 006 Emergency Core Cooling System A4.07 ECCS pumps and valves (RO-4.4/SRO-4.4)
A, D, EN, L 2
- c.
SIM-3 Override Shut a PORV 010 Pressurizer Pressure Control System A4.03 PORV and block valves (RO-4.0/SRO-3.8)
A, D, L 3
- d. SIM-4 Respond to a loss of Secondary Pumps 056 Condensate System A2.04 Loss of condensate pumps (RO-2.6/SRO-2.8)
A, N 4S
- e.
SIM-5 Shift 13 IRU Power Supply 027 Containment Iodine Removal System (CIRS)
A4.01 CIRS controls (RO-3.3/SRO-3.3)
D, EN 5
- f.
SIM-6 Perform Vital Auxiliaries 062 AC Electrical Distribution System A4.01 All Breakers (RO-3.3/SRO-3.1)
N, L 6
- g. SIM-8 Shift Component Cooling Heat Exchangers 008 Component Cooling Water System A4.01 CCW indications and controls (RO-3.3/SRO-3.1)
D 8
In-Plant Systems
- 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U
- a.
PLT-1 Start 11 & 12 Containment Air Coolers 022 Containment Cooling System A4.01 CCS Fans (RO-3.6/SRO-3.6)
D, E, R 5
- b. PLT-2 Respond to a fire on site 086 Fire Protection System (FPS)
AA4.05 Deluge Valves (RO-3.0/SRO-3.5)
N 8
- c. PLT-3 Control RCS and S/G Inventory from 2C43 035 Steam Generator System A3.01 S/G Water Level Control (RO-4.0/SRO-3.9)
D, E 4P
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
- All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6 / 4-6 / 2-3 [4]
9 / 8 / 4 [6]
1 / 1 / 1 [2]
1 / 1 / 1 (control room system) [2]
1 / 1 / 1 [3]
2 / 2 / 1 [4 including 2(A)]
3 / 3 / 2 (randomly selected) [0]
1 / 1 / 1 [1]
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Exam Level: SRO-U Date of Examination: 6/6/2021 Operating Test #: 2021 Administrative Topic (see Note)
Type Code*
Describe activity to be performed Conduct of Operations R, N Determine Operator Qualifications G2.1.8 Ability to coordinate personnel activities outside the control room. (SRO-4.1)
Conduct of Operations R, D Review and approve a completed Surveillance Test Procedure G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (SRO-4.7)
Equipment Control R, D Establish Initial Conditions for STP O-8A-1 G2.2.12 Knowledge of surveillance procedures. (SRO-4.1)
Radiation Control R, N Review and Authorize KI G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (SRO-3.8)
Emergency Plan R, D Recommend Protective Action Guidelines to Public Officials G2.4.44 Knowledge of emergency plan protective action recommendations (SRO-4.4)
NOTE: All items (five total) are required for SROs. RO applicants require only four 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) [3]
(N)ew or (M)odified from bank ( 1) [2]
(P)revious 2 exams ( 1; randomly selected) [0]
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Exam Level: SRO-U Date of Examination: 6/6/2021 Operating Test #: 2021 Control Room Systems:
Safety Function
- a.
SIM-1 Respond to a charging header rupture 004 Chemical and Volume Control System A4.08 Charging (RO-3.8/SRO-3.4)
A, N 1
- b. SIM-2 Align a LPSI Pump for Core Flush 006 Emergency Core Cooling System A4.07 ECCS pumps and valves (RO-4.4/SRO-4.4)
A, D, EN, L 2
- c.
SIM-6 Perform Vital Auxiliaries 062 AC Electrical Distribution System A4.01 All Breakers (RO-3.3/SRO-3.1)
N, L 6
In-Plant Systems
- 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U
- a.
PLT-1 Start 11 & 12 Containment Air Coolers 022 Containment Cooling System A4.01 CCS Fans (RO-3.6/SRO-3.6)
D, E, R 5
- b. PLT-2 Respond to a fire on site 086 Fire Protection System (FPS)
AA4.05 Deluge Valves (RO-3.0/SRO-3.5)
N 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.
- Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6 / 4-6 / 2-3 [2]
9 / 8 / 4 [2]
1 / 1 / 1 [1]
1 / 1 / 1 (control room system) [1]
1 / 1 / 1 [2]
2 / 2 / 1 [3 including 1(A)]
3 / 3 / 2 (randomly selected) [0]
1 / 1 / 1 [1]
ES-401 PWR Examination Outline Form ES-401-2 Facility: Calvert Cliffs Date of Exam: June 2021 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
3 3
3 N/A 3
3 N/A 3
18 3
3 6
2 1
1 2
2 2
1 9
2 2
4 Tier Totals 4
4 5
5 5
4 27 5
5 10
- 2.
Plant Systems 1
3 3
3 2
3 2
3 3
2 2
2 28 3
2 5
2 1
0 1
1 1
1 1
1 1
1 1
10 0
2 1
3 Tier Totals 4
3 4
3 4
3 4
4 3
3 3
38 5
3 8
- 3. Generic Knowledge and Abilities Categories 1
2 3
4 10 1
2 3
4 7
2 2
3 3
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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4.
Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000007 (EPE 7; BW E02&E10; CE E02)
Reactor Trip, Stabilization, Recovery / 1 6
Occurrence of a reactor trip 4.3 1
000008 (APE 8) Pressurizer Vapor Space Accident / 3 2
Sensors and detectors 2.7 2
000009 (EPE 9) Small Break LOCA / 3 2.4.20 Knowledge of operational implications of EOP warnings, cautions and notes.
3.8 3
000015 (APE 15) Reactor Coolant Pump Malfunctions / 4 7
RCP seals 2.9 4
000025 (APE 25) Loss of Residual Heat Removal System / 4 1
Loss of RHRS during all modes of operation 3.9 5
000026 (APE 26) Loss of Component Cooling Water / 8 2
Loads on the CCWS in the control room 3.2 6
000027 (APE 27) Pressurizer Pressure Control System Malfunction / 3 3
Controllers and positioners 2.6 7
000029 (EPE 29) Anticipated Transient Without Scram / 1 1
Reactor nuclear instrumentation 4.4 8
000038 (EPE 38) Steam Generator Tube Rupture / 3 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
4.5 9
000055 (EPE 55) Station Blackout / 6 1
Effect of battery discharge rates on capacity 3.3 10 000056 (APE 56) Loss of Offsite Power / 6 6
Safety injection pump 3.6 11 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 1
Actions contained in EOP for loss of vital ac electrical instrument bus 4.1 12 000058 (APE 58) Loss of DC Power / 6 3
Vital and battery bus components 3.1 13 000062 (APE 62) Loss of Nuclear Service Water / 4 1
Location of a leak in the SWS 2.9 14 000065 (APE 65) Loss of Instrument Air / 8 2.2.44 Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions.
4.2 15 000077 (APE 77) Generator Voltage and Electric Grid Disturbances / 6 2
Over-excitation 3.3 16 000040 (APE 40; BW E05; CE E05; W E12)
Steam Line RuptureExcessive Heat Transfer / 4 3
Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations.
3.8 17 000054 (APE 54; CE E06) Loss of Main Feedwater /4 2
Normal, abnormal and emergency operating procedures associated with Loss of Feedwater 3.2 18 000011 (EPE 11) Large Break LOCA / 3 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.
4.7 76 000022 (APE 22) Loss of Reactor Coolant Makeup / 2 2
Charging pump problems 3.7 77 000026 (APE 26) Loss of Component Cooling Water / 8 2.1.19 Ability to use plant computer to evaluate system or component status.
3.8 78 Loss of Salt Water 2
Ability to determine and interpret the following as they apply to the Loss of Salt Water: Salt Water pump problems 3.0 79
ES-401 3
Form ES-401-2 000057 (APE 57) Loss of Vital AC Instrument Bus / 6 2.1.28 Knowledge of the purpose and function of major system components and controls.
4.1 80 000029 (EPE 29) Anticipated Transient Without Scram / 1 7
Reactor trip breaker indicating lights 4.3 81 K/A Category Totals:
3 3
3 3
3/3 3/3 Group Point Total:
18/6
ES-401 4
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)
E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR 000005 (APE 5) Inoperable/Stuck Control Rod / 1 3
Required actions if more than one rod is stuck or inoperable 3.5 19 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 2
Relationships between PZR pressure increase and reactor makeup/letdown imbalance 2.9 20 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 1
Radiation exposure hazards 3.5 21 00051 (APE 51) Loss of Condenser Vacuum / 4 1
Loss of steam dump capability upon loss of condenser vacuum 2.8 22 000059 (APE 59) Accidental Liquid Radwaste Release / 9 2
The permit for liquid radioactive-waste release 2.9 23 000067 (APE 67) Plant Fire On Site / 8 5
Plant and control room ventilation systems 3.0 24 000069 (APE 69; W E14) Loss of Containment Integrity / 5 1
Isolation valves, dampers and electropneumatic devices.
3.5 25 000076 (APE 76) High Reactor Coolant Activity / 9 1
Process radiation monitors 2.6 26 (CE A16) Excess RCS Leakage / 2 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies 3.8 27 000003 (APE 3) Dropped Control Rod / 1 2.2.22 Knowledge of limiting conditions for operations and safety limits.
4.7 82 000032 (APE 32) Loss of Source Range Nuclear Instrumentation / 7 6
Confirmation of reactor trip 4.1 83 000061 (APE 61) Area Radiation Monitoring System Alarms
/ 7 6
Required actions if alarm channel is out of service 4.1 84 (CE E09) Functional Recovery 2.4.18 Knowledge of the specific bases for EOPs.
4.0 85 K/A Category Point Totals:
1 1
2 2
2/2 1/2 Group Point Total:
9/4
ES-401 5
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 1 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 003 (SF4P RCP) Reactor Coolant Pump 5
RCP seal leakage detection instrumentation 3.1 28 004 (SF1; SF2 CVCS) Chemical and Volume Control 6
RCS temperature and pressure 3.4 29 005 (SF4P RHR) Residual Heat Removal 1
Failure modes for pressure, flow, pump motor amps, motor temperature and tank level instrumentation 2.7 30 005 (SF4P RHR) Residual Heat Removal 3
RCS pressure boundary motor-operated valves 2.7 31 006 (SF2; SF3 ECCS) Emergency Core Cooling 1
Avoidance of thermal and pressure stresses due to pump startup 3.1 32 007 (SF5 PRTS) Pressurizer Relief/Quench Tank 2.4.46 Ability to verify that the alarms are consistent with the plant conditions.
4.2 33 008 (SF8 CCW) Component Cooling Water 2
CCW pump, including emergency backup 3.0 34 010 (SF3 PZR PCS) Pressurizer Pressure Control 1
Pressure detection systems 2.7 35 012 (SF7 RPS) Reactor Protection 1
Trip setpoint adjustment 2.9 36 012 (SF7 RPS) Reactor Protection 1
CRDS 3.9 37 013 (SF2 ESFAS) Engineered Safety Features Actuation 1
Definitions of safety train and ESF channel 2.8 38 013 (SF2 ESFAS) Engineered Safety Features Actuation 2
Safety system logic and reliability 2.9 39 022 (SF5 CCS) Containment Cooling 1
Containment cooling fans 3.0 40 026 (SF5 CSS) Containment Spray 5
Containment spray reset switches 3.5 41 026 (SF5 CSS) Containment Spray 2
Recirculation spray system 4.2 42 039 (SF4S MSS) Main and Reheat Steam 4
Malfunctioning steam dump 3.4 43 059 (SF4S MFW) Main Feedwater 3
Power level restrictions for operation of MFW pumps and valves 2.7 44 059 (SF4S MFW) Main Feedwater 2
AFW system 3.4 45 061 (SF4S AFW)
Auxiliary/Emergency Feedwater 1
Relationship between AFW flow and RCS heat transfer 3.6 46 062 (SF6 ED AC) AC Electrical Distribution 5
Methods for energizing a dead bus 2.9 47 062 (SF6 ED AC) AC Electrical Distribution 3
Interlocks between automatic bus transfer and breakers 2.8 48 063 (SF6 ED DC) DC Electrical Distribution 3
Battery charger and battery 2.9 49 064 (SF6 EDG) Emergency Diesel Generator 8
Fuel oil storage tanks 3.2 50 073 (SF7 PRM) Process Radiation Monitoring 2
Letdown isolation on high-RCS activity 3.3 51 076 (SF4S SW) Service Water 2.1.27 Knowledge of system purpose and or function.
3.9 52
ES-401 6
Form ES-401-2 078 (SF8 IAS) Instrument Air 1
Air pressure 3.1 53 078 (SF8 IAS) Instrument Air 4
Cooling water to compressor 2.6 54 103 (SF5 CNT) Containment 1
Containment isolation 3.9 55 004 (SF1; SF2 CVCS) Chemical and Volume Control 12 CIAS, SIAS 4.3 86 013 (SF2 ESFAS) Engineered Safety Features Actuation 6
Inadvertent ESFAS actuation 4.0 87 063 (SF6 ED DC) DC Electrical Distribution 2.4.41 Knowledge of the emergency action level thresholds and classifications.
4.6 88 076 (SF4S SW) Service Water 2.1.32 Ability to explain and apply all system limits and precautions.
4.0 89 103 (SF5 CNT) Containment 2
Necessary plant conditions for work in containment 3.2 90 K/A Category Point Totals:
3 3
3 2
3 2
3 3/3 2 2
2/2 Group Point Total:
28/5
ES-401 7
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant SystemsTier 2/Group 2 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR 002 (SF2; SF4P RCS) Reactor Coolant 12 Radioactivity level when venting CRDS 2.9 56 015 (SF7 NI) Nuclear Instrumentation 3
Component interconnections 2.6 57 017 (SF7 ITM) In-Core Temperature Monitor 1
Natural circulation indications 3.5 58 033 (SF8 SFPCS) Spent Fuel Pool Cooling 1
Maintenance of spent fuel level 2.9 59 034 (SF8 FHS) Fuel-Handling Equipment 2
Dropped cask 3.4 60 035 (SF 4P SG) Steam Generator 1
S/G water level control 4.0 61 055 (SF4S CARS) Condenser Air Removal 2.1.30 Ability to locate and operate components, including local controls.
4.4 62 071 (SF9 WGS) Waste Gas Disposal 4
Relationship of hydrogen/oxygen concentrations to flammability 2.5 63 072 (SF7 ARM) Area Radiation Monitoring 3
Fuel building isolation 3.6 64 011 (SF2 PZR LCS) Pressurizer Level Control 5
Letdown flow controller 3.2 65 014 (SF1 RPI) Rod Position Indication 4
Misaligned Rod 3.9 91 045 (SF 4S MTG) Main Turbine Generator 15 Turbine overspeed 2.6 92 075 (SF8 CW) Circulating Water 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.
4.3 93 K/A Category Point Totals:
1 0
1 1
1 1
1 1/2 1 1
1/1 Group Point Total:
10/3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility: Calvert Cliffs Date of Exam: 2021 Category K/A #
Topic RO SRO-only IR IR
- 1. Conduct of Operations 2.1.15 Knowledge of administrative requirements for temporary management directives such as standing orders, night orders, Operations memos, etc.
2.7 66 2.1.5 Ability to locate and use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.
2.9 67 2.1.41 Knowledge of the refueling processes 3.7 94 2.1.8 Ability to coordinate personnel activities outside the control room.
4.1 95 Subtotal 2
2
- 2. Equipment Control 2.2.37 Ability to determine operability and/or availability of safety related equipment 3.6 68 2.2.6 Knowledge of the process for making changes to procedures 3.0 69 2.2.20 Knowledge of the process for managing troubleshooting activities.
3.8 96 2.2.38 Knowledge of conditions and limitations in the facility license.
4.5 97 Subtotal 2
2
- 3. Radiation Control 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties 3.2 70 2.3.15 Knowledge of radiation monitoring systems 2.9 71 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.
3.2 72 2.3.6 Ability to approve release permits.
3.8 98 Subtotal 3
1
- 4. Emergency Procedures/Plan 2.4.19 Knowledge of EOP layout, symbols and icons.
3.4 73 2.4.3 Ability to identify post-accident instrumentation.
3.7 74 2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
4.2 75 2.4.12 Knowledge of general operating crew responsibilities during emergency operations.
4.3 99 2.4.11 Knowledge of abnormal condition procedures 4.2 100 Subtotal 3
2 Tier 3 Point Total 10 7
ES-401 Record of Rejected K/As Form ES-401-4 Tier /
Group Randomly Selected K/A Reason for Rejection Rejected Emergency and Abnormal Plant Evolutions associated solely with Babcock and Wilcox (BW) reactors -
Calvert Cliffs is a CE design.
Rejected 025 (SF5 ICE) Ice Condenser - Calvert Cliffs does not have an ice condenser installed.
Rejected 033 Loss of Intermediate Range Nuclear Instrumentation. Calvert Cliffs does not have Intermediate Range NIs 1/1 SRO Question #79: Added Loss of Salt Water to the Tier 1, Group 1 list of E/APEs to be available for random selection in accordance with ES-401 Attachment 1. Loss of Salt Water was randomly selected for Question 79.
2/1 073, K4.02 RO Question #51 Calvert Cliffs does not have a letdown isolation on a high RCS activity. Randomly reselected K4.01 2/2 002, A1.12 RO Question #56 Calvert Cliffs does not vent CRDS. Randomly reselected A1.13 3
2.4.11 SRO Question
- 100 Knowledge of abnormal condition procedures is tested extensively throughout the operating test. This K/A is considered oversampled.
Randomly reselected 2.4.29