ML20274A105

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Draft Written Examination and Operating Test Outlines (Folder 2)
ML20274A105
Person / Time
Site: Calvert Cliffs  Constellation icon.png
Issue date: 02/07/2020
From:
Constellation Energy Group
To: Thomas Setzer
Operations Branch I
Shared Package
ML19105A101 List:
References
CAC 000500
Download: ML20274A105 (26)
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ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 1/27/2020 Exam Level: RO Operating Test #: 2020 Administrative Topic Type Describe activity to be performed (see Note) Code*

Estimate Time to Boil Conduct of Operations R, D G2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. (RO-3.9)

Determine Shutdown Margin Conduct of Operations R, D G2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management. (RO-4.3)

Print Reading Equipment Control R, N G2.2.41 Ability to obtain and interpret station electrical and mechanical drawings. (RO-3.5)

Emergency Plan R, N Respond to a Plant Fire G2.4.27 Knowledge of fire in the plant procedures (RO-3.4)

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) [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 Date of Examination: 1/27/2020 Exam Level: RO Operating Test #: 2020 Control Room Systems:

  • 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U Safety System / JPM Title Type Code*

Function

a. SIM-1 Respond to a CEDS failure 001 Control Rod Drive System A, N 1 A4.03 CRDS mode control (RO-4.0)
b. SIM-2 Verify Validity of CIS Actuation 013 Engineered Safety Features Actuation System D, EN 2 A3.01 Input channels and logic (RO-3.7)
c. SIM-3 Verify HPSI/LPSI Flow 006 Emergency Core Cooling System A, EN, N, L 3 A4.07 ECCS pumps and valves (RO-4.4)
d. SIM-4 Attempt to Correct the Abnormal SDC Condition 005 Residual Heat Removal System A, D, L 4P A4.01 Controls and indications for RHR pumps (RO-3.6)
e. SIM-5 Align Saltwater Emergency Overboard 076 Service Water System N, L 4S A4.04 Emergency heat loads (RO-3.5)
f. SIM-6 Pressurizer Vent Valve Testing 007 Pressurizer Relief Tank/Quench Tank System N, L 5 A4.04 PZR Vent Valve (RO-2.6)
g. SIM-7 Perform Vital Auxiliaries 062 AC Electrical Distribution System N, L 6 A4.01 All Breakers (including available switchyard) (RO-3.3)
h. SIM-8 Respond to RCS leakage into Component Cooling System 008 Component Cooling Water System A, N 8 A4.01 CCW indications and controls (RO-3.3)

In-Plant Systems

  • 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U
i. PLT-1 Align AFW flow control to 1C43 061 Auxiliary/Emergency Feedwater System D, E 4S 2.1.30 Ability to locate and operate components (RO-4.4)
j. PLT-2 Operate MCC Load Breakers 068 Control Room Evacuation N, E, R 6 AA1.10 Power Distribution: ac and dc (RO-3.7)
k. PLT-3 Respond to a Loss of Instrument Air 065 Loss of Instrument Air D, E 8 AA1.04 Emergency air compressor (RO-3.5)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 [4]

(C)ontrol room (D)irect from bank 9 / 8 / 4 [4]

(E)mergency or abnormal in-plant 1 / 1 / 1 [3]

(EN)gineered safety feature 1 / 1 / 1 (control room system) [2]

(L)ow-Power / Shutdown 1 / 1 / 1 [5]

(N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 [7 including 3(A)]

(P)revious 2 exams 3 / 3 / 2 (randomly selected) [0]

(R)CA 1 / 1 / 1 [1]

(S)imulator

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 1/27/2020 Exam Level: SRO-I Operating Test #: 2020 Administrative Topic Type Describe activity to be performed (see Note) Code*

Evaluate Overtime Work Request Conduct of Operations R, N G2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (SRO-3.9)

Supervisory Review of Shutdown Margin Conduct of Operations R, D G2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management. (SRO-4.6)

Print Reading and Application of Tech Specs Equipment Control R, N G2.2.41 Ability to obtain and interpret station electrical and mechanical drawings. (SRO-3.9)

Administrative Requirements for RMS Inoperability Radiation Control R, N G2.3.15 - Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (SRO-3.1)

Perform Shift Emergency Director Functions Emergency Plan R, D G2.4.41 Knowledge of emergency action level thresholds and classifications. (SRO-4.6)

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) [3]

(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 Date of Examination: 1/27/2020 Exam Level: SRO-I Operating Test #: 2020 Control Room Systems:

  • 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U Safety System / JPM Title Type Code*

Function

a. SIM-1 Shutdown CEA Free Movement Test 001 Control Rod Drive System A, N 1 A4.03 CRDS mode control (SRO-3.7)
b. SIM-2 Verify Validity of CIS Actuation 013 Engineered Safety Features Actuation System D, EN 2 A3.01 Input channels and logic (SRO-3.9)
c. SIM-3 Verify HPSI/LPSI Flow 006 Emergency Core Cooling System A, EN, N, L 3 A4.07 ECCS pumps and valves (SRO-4.4)
d. SIM-4 Attempt to Correct the Abnormal SDC Condition 005 Residual Heat Removal System A, D, L 4P A4.01 Controls and indications for RHR pumps (SRO-3.4)
e. SIM-5 Align Saltwater Emergency Overboard 076 Service Water System N, L 4S A4.04 Emergency heat loads (SRO-3.5)
f. SIM-7 Perform Vital Auxiliaries 062 AC Electrical Distribution System N, L 6 A4.01 All Breakers (including available switchyard) (SRO-3.1)
g. SIM-8 Respond to RCS leakage into Component Cooling System 008 Component Cooling Water System A, N 8 A4.01 CCW indications and controls (SRO-3.1)

In-Plant Systems

  • 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U
h. PLT-1 Align AFW flow control to 1C43 061 Auxiliary/Emergency Feedwater System D, E 4S 2.1.30 Ability to locate and operate components (SRO-4.4)
i. PLT-2 Operate MCC Load Breakers 068 Control Room Evacuation N, E, R 6 AA1.10 Power Distribution: ac and dc (SRO-3.9)
j. PLT-3 Respond to a Loss of Instrument Air 065 Loss of Instrument Air D, E 8 AA1.04 Emergency air compressor (SRO-3.4)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 [4]

(C)ontrol room (D)irect from bank 9 / 8 / 4 [4]

(E)mergency or abnormal in-plant 1 / 1 / 1 [3]

(EN)gineered safety feature 1 / 1 / 1 (control room system) [2]

(L)ow-Power / Shutdown 1 / 1 / 1 [4]

(N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 [6 including 3(A)]

(P)revious 2 exams 3 / 3 / 2 (randomly selected) [0]

(R)CA 1 / 1 / 1 [1]

(S)imulator

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 1/27/2020 Exam Level: SRO-U Operating Test #: 2020 Administrative Topic Type Describe activity to be performed (see Note) Code*

Evaluate Overtime Work Request Conduct of Operations R, N G2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (SRO-3.9)

Supervisory Review of Shutdown Margin Conduct of Operations R, D G2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management. (SRO-4.6)

Print Reading and Application of Tech Specs Equipment Control R, N G2.2.41 Ability to obtain and interpret station electrical and mechanical drawings. (SRO-3.9)

Administrative Requirements for RMS Inoperability Radiation Control R, N G2.3.15 - Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (SRO-3.1)

Perform Shift Emergency Director Functions Emergency Plan R, D G2.4.41 Knowledge of emergency action level thresholds and classifications. (SRO-4.6)

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) [3]

(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 Date of Examination: 1/27/2020 Exam Level: SRO-U Operating Test #: 2020 Control Room Systems:

  • 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U Safety System / JPM Title Type Code*

Function

a. SIM-2 Verify Validity of CIS Actuation 013 Engineered Safety Features Actuation System D, EN 2 A3.01 Input channels and logic (SRO-3.9)
b. SIM-3 Verify HPSI/LPSI Flow 006 Emergency Core Cooling System A, EN, N, L 3 A4.07 ECCS pumps and valves (SRO-4.4)
c. SIM-8 Respond to RCS leakage into Component Cooling System 008 Component Cooling Water System A, N 8 A4.01 CCW indications and controls (SRO-3.1)

In-Plant Systems

  • 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U
d. PLT-1 Align AFW flow control to 1C43 061 Auxiliary/Emergency Feedwater System D, E 4S 2.1.30 Ability to locate and operate components (SRO-4.4)
e. PLT-2 Operate MCC Load Breakers 068 Control Room Evacuation N, E, R 6 AA1.10 Power Distribution: ac and dc (SRO-3.9)
  • 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 [2]

(C)ontrol room (D)irect from bank 9 / 8 / 4 [2]

(E)mergency or abnormal in-plant 1 / 1 / 1 [2]

(EN)gineered safety feature 1 / 1 / 1 (control room system) [2]

(L)ow-Power / Shutdown 1 / 1 / 1 [1]

(N)ew or (M)odified from bank including 1(A) 2 / 2 / 1 [3 including 2(A)]

(P)revious 2 exams 3 / 3 / 2 (randomly selected) [0]

(R)CA 1 / 1 / 1 [1]

(S)imulator

Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #1 OP-Test # 2020 Examiners: Operators:

Initial Conditions: Unit-1 is at 100% power, MOC. Unit-2 is at 100% power.

Turnover: 13 AFW Pump is OOS.

Instructions to the crew: Maintain 100% power.

Critical Tasks

1. Identifies 12 S/G as the ruptured SG and isolates 12 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.
2. Recognizes SFSC in EOP-6 not met and transitions to EOP-8 within 30 minutes of Loss of All Feedwater initiation.
3. Establishes AFW flow to at least one S/G prior to S/G levels going below (-)350 inches OR Commences OTCC when both S/G levels are below (-)350 inches or Tcold rises uncontrollably by 5°F or greater and prior to CET temperatures reaching 560°F.

Event # Malfunction # Event Type* Event Description I-RO/SRO 1 rcs026_01 Pressurizer Level Transmitter, LT110X, Fails Low T-SRO 2 hdv001_02 C-BOP/SRO 12 Heater Drain Tank Pump trips/AOP-3G 3 ms009_02 C-ALL TBV-3942 Fails Open/AOP-7K ms001_02 C-BOP/SRO 4 R-ATC 12 S/G Tube Leak (100 gpm)/AOP-2A Downpower T-SRO 5 13kv001_01 C-ALL P-13000-1 Transformer Loss/EOP-0 6 ms002_02 M-ALL SGTR (1 tube) in 12 SG/EOP-6 7 afw001_01 M-ALL 11 AFW Pump Trip (Loss of All Feed)/EOP-8 8 2-CV-4550 C-BOP/SRO 23 AFW Pump Trip

  • (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 #1 OP-Test # 2020 Scenario Overview Initial Conditions:

Unit-1 at 100% power, MOC, Unit-2 at 100% power Equipment OOS: 13 AFW Pump.

Abnormal Conditions: None Instructions for shift: Maintain 100% power.

Event 1 - Pressurizer Level Transmitter, LT110X Fails Low, causing the process variable of the in-service PZR level controller to fail low. Alarm Response Manual 1C06 actions will direct the crew to swap Pressurizer Level Controller channels and the Pressurizer Low Level Cut-off switch, and then reset the Proportional Heaters. The crew should reference OP-CA-103-102-0200, enter TS LCO 3.3.10.A, and determine the required action is to restore the indication channel to operable status within 30 days.

Event 2 - 12 Heater Drain Tank Pump trips requiring the crew to enter AOP-3G, Malfunction of Main Feedwater System. The crew will start the third Condensate Booster Pump and ensure adequate Steam Generator Feed Pump suction pressure is maintained.

Event 3 - TBV-3942 Fails Open causing RCS temperatures, MWe, and Pressurizer Level to quickly lower. The crew will enter AOP-7K and take prompt actions to control Reactor Power by inserting CEAs and/or performing an RCS boration as well as lowering turbine load. The crew will then direct the local isolation of the failed open TBV.

Event 4 - 12 S/G will begin to experience tube leakage at 100 GPM. The crew will respond per AOP-2A, Excessive Reactor Coolant Leakage, and will reduce power to reduce TAVE to less than 537F while monitoring for reactor trip criteria. Determines TS LCO 3.4.13.B is applicable with required actions to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to be in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Event 5 - During the rapid downpower per AOP-2A, a fault will occur on P-13000-1 13KV Transformer causing an automatic reactor trip. The crew will enter EOP-0, Post Trip Immediate Actions.

Event 6 - During EOP-0, the S/G tube leak will increase to a rupture of 1 S/G tube. The crew is expected to implement EOP-6. The crew will commence a cooldown, at a rate not to exceed 100°F/hr, to prepare to isolate 12 S/G. Once Thot <515°F, the crew will perform the Critical Task to isolate 11 S/G. If plant conditions degrade or the crew is unsure of the diagnosis it is acceptable for them to enter EOP-8 directly from EOP-0. If EOP-8 is entered, all critical tasks still apply unless individual tasks are invalidated by the exam team.

Events 7/8 - After 12 S/G has been isolated, 11 AFW Pump will trip resulting in a Loss of All Feedwater. 23 AFW Pump will fail to start when attempted by the crew. 12 AFW will be unable to be reset when initially requested. The crew is expected to perform the Critical Task to transition to EOP-8 based on the combined SGTR and Loss of All Feedwater in progress. The crew should identify the success paths in EOP-8 (RC-1 Met, VA-1 Met, PIC-4 Met, HR-2 Not Met (Met if SG Levels are in the band), CE-1 Met, RLEC-2 Met) and priority (HR-2 if not met, PIC-4, RLEC-2, RC-1, VA-1, CE-1). Crew will commence HR-2 and PIC-4. As part of HR-2, the crew should attempt to establish AFW flow when available. As part of PIC-4, the crew should pursue lowering RCS pressure to attempt to lower the SGTR leak rate. The scenario will end once AFW flow has been established to 11 Steam Generator after 12 AFW Pump is repaired.

Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #2 OP-Test # 2020 Examiners: Operators:

Initial Conditions: Unit-1 is at 100% power, MOC. Unit-2 is at 100% power.

Turnover: 13 CAC is OOS.

Instructions to the crew: The Shift Manager directs the crew to commence RCS Boron Equalization per OI-1H Section 6.6 to support a scheduled reactor downpower next shift.

Critical Tasks

1. Establishes AFW flow to at least one S/G prior to S/G levels going below (-)350 inches.
2. Trips all RCPs within 15 minutes after receiving CIS actuation.
3. Identifies 12 Steam Generator as faulted and isolates 12 S/G.
4. Establishes at least one train of Containment Spray flow to Containment.

Event # Malfunction # Event Type* Event Description 1 N/A N-RO/SRO Equalize Boron I-BOP/SRO 2 rcs004_03 11A Loop TCOLD 1-TT-112CC Fails High T-SRO C-ALL 3 480v001_08 Loss of 14B 480V Bus/AOP-7I T-SRO U25-11-LCL-C-BOP/SRO 4 BKR Main Transformer Loss of Cooling/AOP-7E R-ATC Downpower 11 SGFP Trip/AOP-3G 5 fw004_01 C-ALL EOP-0 6 fw004_02 C-BOP/SRO 12 SGFP Trip 7 ms010_02 M-ALL Steam Line Rupture in Containment/EOP-4 esfa004_01/02 8 I-ATC/SRO CSAS Auto and Manual Failure esfa005_01/02

  • (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 # 2020 Scenario Overview Initial Conditions:

Unit-1 at 100% power, MOC, Unit-2 at 100% power Equipment OOS: 13 CAC Abnormal Conditions: None Instructions for shift: The Shift Manager directs the crew to commence RCS Boron Equalization per OI-1H Section 6.6 to support a scheduled reactor downpower next shift.

Event 1 - The crew will commence RCS Boron Equalization per OI-1H Section 6.6 to support a scheduled reactor downpower next shift.

Event 2 - 11A Loop TCOLD 1-TT-112CC Fails High. 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 C trip units 1,7, & 10 and enter Tech Spec 3.3.1.A with required actions to place trip units 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 then restore trip units 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 14B 480V Bus will occur. The crew will implement AOP-7I Section XXVII, Loss of 14B 480V Bus, which will direct their actions in protecting plant equipment and placing a Charging Pump and 11 Main Vent Fan in service.

Determines Tech Spec LCO 3.8.9.A is applicable with a required action to restore the AC subsystem to operable status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Event 4 - A Loss of Cooling to Main Transformer U-25000-11 will occur. The crew will respond using AOP-7E Section XII, Total Loss of Unit Transformer Cooling, and will commence a reactor downpower per OP-3.

Event 5 - At 90% reactor power during the downpower, 11 SGFP will trip. The crew is expected to implement AOP-3G and monitor for reactor trip criteria. The crew will be unable to reset 11 SGFP and will then trip the reactor.

Event 6 - In EOP-0, 12 SGFP will trip 1 minute after the reactor trip requiring the crew to perform the Critical Task to initiate AFW flow to the Steam Generators based on the loss of Main Feedwater flow.

Event 7 - The major event will be a 12 SG Steam Line Rupture inside Containment that will occur 6 minutes after the reactor trip. In EOP-0, the crew will perform the Critical Task of securing all RCPs after the CIS actuation. In EOP-4, the crew will perform the Critical Task of identifying 12 SG as being faulted and isolating 12 SG.

Event 8 - Both CSAS channels will fail to automatically actuate and manually actuate with the pushbuttons requiring the crew to perform the Critical Task to take manual actions to initiate at least one train of Containment Spray flow.

Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #3 OP-Test # 2020 Examiners: Operators:

Initial Conditions: Unit-1 is at 3% power, MOC. Unit-2 is at 100% power.

Turnover: 13 CBP is OOS.

Instructions to the crew: Raise reactor power and stabilize at 8% per OP-3.

Critical Tasks

1. Commences an RCS Cooldown not to exceed 100°F in any one hour.
2. Trip 11A & 12B RCPs or 11B & 12A RCPs when RCS pressure decreases to <1725 PSIA prior to RCS subcooling being less than 20°F for 4 minutes.

Event # Malfunction # Event Type* Event Description N-BOP/SRO 1 N/A Raise Reactor Power R-ATC 2 rcs023_02 I-ATC/SRO 1-PT-100Y PZR Pressure PV Fails High C-BOP/SRO 3 sw002_01 11 Saltwater Pump Trip/AOP-7A T-SRO 4 ia001 C-BOP/SRO Instrument Air Leak/AOP-7D C-ALL RCS Leak Unisolated/AOP-2A 5 rcs003 T-SRO EOP-0 6 rcs002 M-ALL LOCA (800 gpm)/EOP-5 7 esfa010_01/02 C-ATC/SRO CIS Auto Failure

  • (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 # 2020 Scenario Overview Initial Conditions:

Unit-1 at 3% power, MOC, Unit-2 at 100% power Equipment OOS: 13 Condensate Booster Pump.

Abnormal Conditions: None Instructions for shift: Raise reactor power and stabilize at 8% per OP-3.

Event 1 - The crew will perform a normal evolution of raising reactor power.

Event 2 PT-100Y, Pressurizer Pressure Controller process variable will fail high. Alarm Response Manual 1C06 actions will have the crew swap Pressurizer Pressure controllers to control RCS pressure.

Event 3 - 11 SW Pump Breaker will trip. The crew will implement AOP-7A, Loss of Saltwater Cooling, which will direct their actions in protecting plant equipment and aligning and starting 13 SW Pump. The crew will determine the following Tech Spec LCOs are applicable: 3.5.2.A with a required action to restore ECCS train to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (this LCO was already entered); 3.6.6.A with a required action to restore containment spray train to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; 3.6.6.B with a required action to restore containment cooling train to operable status within 7 days; 3.7.5.A with a required action to restore Component Cooling loop to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; 3.7.6.B with a required action to restore SRW subsystem to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and 3.7.7.A with a required action to restore SW subsystem to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Event 4 - An Instrument Air Leak will occur. The crew will implement AOP-7D, Loss of Instrument Air, which will direct the crew to start the Saltwater Air Compressors and ensure backup Instrument Air systems respond. When operators are dispatched to investigate the issue, the air leak will be quickly isolated.

Event 5 - An RCS Leak inside Containment of 70 gpm will occur. The crew will respond using AOP-2A, Excessive Reactor Coolant Leakage, and will not be able to isolate the leak at which point the reactor will be tripped. EOP-0, Post-Trip Immediate Actions will be implemented.

Determines Tech Spec LCO 3.4.13.A is applicable with a required action to reduce leakage to within limits of the LCO 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 />. It is acceptable to enter LCO 3.4.13.B with the assumption that pressure boundary leakage exists.

Event 6 - After the reactor trip, an 800 gpm RCS LOCA inside Containment will occur. The crew is expected to implement EOP-5 based on the LOCA in progress. The actions in EOP-5 will be to commence an RCS cooldown and depressurization to minimize the leakage and control RCS subcooling.

Event 7 - A failure of CIS Channels A and B to automatically actuate will occur. The crew is expected to recognize the failure and manually actuate CIS using either the pushbuttons or manually operate the components necessary.

Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #4 OP-Test # 2020 Examiners: Operators:

Initial Conditions: Unit-1 is at 100% power, MOC. Unit-2 is at 100% power.

Turnover: 12 Boric Acid Pump is OOS, 0C DG is OOS.

Instructions to the crew: The Shift Manager has directed the crew to shift disconnects for 13 IRU to the 14B 480V Bus per OI-5B Section 6.2.

Critical Tasks

1. Establishes AFW flow to at least one S/G prior to S/G levels going below (-)350 inches.
2. Recognizes SFSC in EOP-2 not met and transitions to EOP-7 within 30 minutes of Station Blackout conditions.
3. Restores power to 11 or 14 4KV Bus prior to 11 or 22 DC Bus voltages going below 106V.

Event # Malfunction # Event Type* Event Description N-BOP/SRO 1 N/A 13 IRU Shift Disconnects T-SRO C-ATC/SRO 2 rcs027_02 PORV-404 Leakage T-SRO 3 480v001_04 C-BOP/SRO Loss of 12B 480V Bus/AOP-7I tg017 C-BOP/SRO 4 High Turbine Vibrations/AOP-7E Downpower R-ATC 5 cvcs014_01 C-ATC/SRO 11 Boric Acid Pump Trips EOP-0 6 swyd002 M-ALL Loss of Offsite Power/EOP-2 7 dg001_02 C-ALL 1B EDG Trips 8 dg002_02 M-ALL Station Blackout/EOP-7

  • (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 # 2020 Scenario Overview Initial Conditions:

Unit-1 at 100% power, MOC, Unit-2 at 100% power Equipment OOS: 12 Boric Acid Pump. 0C DG.

Abnormal Conditions: None Instructions for shift: The Shift Manager has directed the crew to shift disconnects for 13 IRU to the 14B 480V Bus per OI-5B Section 6.2.

Event 1 - The crew will shift disconnects for 13 IRU to the 14B 480V Bus per OI-5B. During the evolution, the crew will enter Tech Spec 3.6.8.A with a required action to restore the Iodine Removal System train to operable status with a completion time of 7 days.

Event 2 - Leakage from PORV-404 into the Quench Tank will occur. Alarm Response Manual 1C06 actions will have crew shut the Block Valve for PORV-404 and verify the leakage has been isolated. The crew is expected to enter Tech Spec 3.4.11.A for PORV-404 being declared inoperable with a required action to close and maintain power to associated block valve with a completion time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Event 3 - After PORV-404 is isolated, a loss of 12B 480V Bus will occur. The crew will implement AOP-7I Section XV, Loss of 12B 480V Bus, which will direct their actions in protecting plant equipment and pursue tying MCCs 106 and 116.

Event 4 - The Main Turbine will begin to experience high vibrations on Bearing #4 ramping in from 6.5 mils to 11 mils over a 5 minute period. The crew will respond using AOP-7E, Main Turbine Malfunction, and will commence a reactor downpower in an attempt to lower turbine vibrations. The turbine vibrations will eventually reach trip criteria and EOP-0, Post-Trip Immediate Actions will be implemented.

Event 5 - When a rapid downpower is commenced, 11 Boric Acid Pump will immediately trip upon starting. The crew is expected to continue the downpower using one of the alternate boration steps of OP-3.

Event 6 - After the reactor trip in EOP-0, a Loss of Offsite Power will occur. The crew will verify the safety related buses are energized by their Emergency Diesel Generators and is expected to restart a Component Cooling Pump.

Event 7 - After the implementation of EOP-2, the 1B EDG will trip. The crew will pursue tying underlying MCCs and ensure proper operation of remaining equipment available.

Event 8 - Then, 8 minutes after the loss of the 1B EDG, the 1A EDG will trip causing Station Blackout conditions on Unit-1. The crew is expected to perform the Critical Task of transitioning to EOP-7, Station Blackout. When the 1A EDG is repaired, the crew will perform the Critical Task to energize 11 4KV Bus.

ES-401 PWR Examination Outline Form ES-401-2 Facility: Calvert Cliffs Date of Exam: January 2020 RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 3 3 3 3 3 3 18 3 3 6 Emergency and Abnormal Plant 2 2 2 2 N/A 1 1 N/A 1 9 2 2 4 Evolutions Tier Totals 5 5 5 4 4 4 27 5 5 10 1 3 2 3 2 2 2 3 3 2 3 3 28 3 2 5 2.

Plant 2 1 0 1 1 1 1 1 1 1 1 1 10 0 2 1 3 Systems Tier Totals 4 2 4 3 3 3 4 4 3 4 4 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 3 2 1 2 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
3. Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

  • These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
    • These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

ES-401 2 Form ES-401-2 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 #

Relationship of emergency feedwater flow to S/G 000007 (EPE 7; BW E02&E10; CE E02) 6 and decay heat removal following reactor trip 3.7 1 Reactor Trip, Stabilization, Recovery / 1 000008 (APE 8) Pressurizer Vapor Space 6 Control of PZR level 3.6 2 Accident / 3 000009 (EPE 9) Small Break LOCA / 3 24 ECCS throttling or termination criteria 4.1 3 000011 (EPE 11) Large Break LOCA / 3 2 Pumps 2.6 4 000015 (APE 15) Reactor Coolant Pump 1 Natural circulation in a nuclear reactor power 4.4 5 Malfunctions / 4 plant Relationship of charging flow to pressure 000022 (APE 22) Loss of Reactor Coolant 2 differential between charging and RCS 2.7 6 Makeup / 2 000025 (APE 25) Loss of Residual Heat 4 Location and isolability of leaks 3.3 7 Removal System / 4 000026 (APE 26) Loss of Component 1 Location of a leak in the CCWS 2.9 8 Cooling Water / 8 000027 (APE 27) Pressurizer Pressure 2 SCR-controlled heaters in manual mode 3.1 9 Control System Malfunction / 3 000029 (EPE 29) Anticipated Transient 2.4.6 Knowledge symptom based EOP mitigation 3.7 10 Without Scram / 1 strategies.

Actions contained in EOP for RCS water 000038 (EPE 38) Steam Generator Tube 6 inventory balance, S/G tube rupture, and plant 4.2 11 Rupture / 3 shutdown procedures 000040 (APE 40; BW E05; CE E05; W E12) 2 Sensors and detectors 2.6 12 Steam Line RuptureExcessive Heat Transfer / 4 000055 (EPE 55) Station Blackout / 6 2 RCS core cooling through natural circulation 4.4 13 cooling to S/G cooling Knowledge of annunciators alarms, indications or 000056 (APE 56) Loss of Offsite Power / 6 2.4.31 response procedures 4.2 14 Feedwater pump speed to control pressure and 000057 (APE 57) Loss of Vital AC 3 level in S/G 3.6 15 Instrument Bus / 6 Ability to identify post-accident instrumentation.

000058 (APE 58) Loss of DC Power / 6 2.4.3 3.7 16 Loss of Salt Water 2 The automatic actions (alignments) within the 3.6 17 salt water system resulting from the actuation of the ESFAS.

Facility s heat removal systems, including 000054 (CE E06) Loss of Main Feedwater 2 primary coolant, emergency coolant, the decay 3.5 18

/4 heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Ability to verify that the alarms are consistent 000007 (EPE 7; BW E02&E10; CE E02) 2.4.46 with the plant conditions. 4.2 76 Reactor Trip, Stabilization, Recovery / 1 Whether charging line leak exists 000022 (APE 22) Loss of Reactor Coolant 1 3.8 77 Makeup / 2 Knowledge of limiting conditions for operations 000027 (APE 27) Pressurizer Pressure 2.2.22 and safety limits. 4.7 78 Control System Malfunction / 3 125V dc bus voltage, low/critical low, alarm 000058 (APE 58) Loss of DC Power / 6 2 3.6 79

ES-401 3 Form ES-401-2 Ability to locate and operate components, 000062 (APE 62) Loss of Nuclear Service 2.1.30 including local controls. 4.0 80 Water / 4 Relationship of flow readings to system operation 000065 (APE 65) Loss of Instrument Air / 8 2 2.6 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 #

000001 (APE 1) Continuous Rod Withdrawal / 1 21 Integral rod worth 2.9 19 000003 (APE 3) Dropped Control Rod / 1 17 Fuel temperature coefficient 2.9 20 000032 (APE 32) Loss of Source Range Nuclear 1 Power supplies, including 2.7 21 Instrumentation / 7 proper switch positions 000051 (APE 51) Loss of Condenser Vacuum / 4 2.2.44 Ability to interpret control room 4.2 22 indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 000061 (APE 61) Area Radiation Monitoring System Alarms 1 Automatic actuation 3.6 23

/7 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling / 7 The difference between a 4.1 24 4 LOCA and inadequate core cooling from trends and indicators (CE A11**; W E08) RCS OvercoolingPressurized Thermal 1 Components, and functions of 3.2 25 Shock / 4 control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(BW E09; CE A13**; W E09 & E10) Natural Circulation/4 1 Facility operating 3.4 26 characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

(CE E09) Functional Recovery 3 Manipulation of controls 3.7 27 required to obtain desired operating results during abnormal, and emergency situations.

000005 (APE 5) Inoperable/Stuck Control Rod / 1 2.2.40 Ability to apply Technical 4.7 82 Specifications for a system.

000028 (APE 28) Pressurizer (PZR) Level Control 14 The effect on indicated PZR 2.8 83 Malfunction / 2 levels, given a change in ambient pressure and temperature of reflux boiling 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8 2.4.34 Knowledge of RO tasks 4.1 84 performed outside the main control room during an emergency and the resultant operational effects.

(CE A16) Excess RCS Leakage / 2 1 Facility conditions and 3.5 85 selection of appropriate procedures during abnormal and emergency operations.

K/A Category Point Totals: 2 2 2 1 1/2 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 3 RCP seal system 3.3 28 Pump Effects of RCP shutdown on secondary 003 (SF4P RCP) Reactor Coolant 4 parameters, such as steam pressure, 3.2 29 Pump steam flow and feed flow 004 (SF1; SF2 CVCS) Chemical and 1 Importance of oxygen control in RCS 2.7 30 Volume Control 005 (SF4P RHR) Residual Heat 3 RHR heat exchanger 2.5 31 Removal 006 (SF2; SF3 ECCS) Emergency 6 Water hammer 3.3 32 Core Cooling 007 (SF5 PRTS) Pressurizer 3 RCS 3.0 33 Relief/Quench Tank 007 (SF5 PRTS) Pressurizer 1 Containment 3.3 34 Relief/Quench Tank 008 (SF8 CCW) Component Cooling 1 Loads cooled by CCWS 3.4 35 Water 010 (SF3 PZR PCS) Pressurizer 3 Over pressure control 3.8 36 Pressure Control 012 (SF7 RPS) Reactor Protection 7 M/G set breakers 3.9 37 Knowledge of operational implications of 013 (SF2 ESFAS) Engineered 2.4.20 EOP warnings, cautions and notes. 3.8 38 Safety Features Actuation Containment equipment subject to 022 (SF5 CCS) Containment Cooling 1 damage by high or low temperature, 2.9 39 humidity and pressure 022 (SF5 CCS) Containment Cooling 5 Containment cooling after LOCA destroys 2.6 40 ventilation ducts 026 (SF5 CSS) Containment Spray 4 Failure of spray pump 3.9 41 039 (SF4S MSS) Main and Reheat 4 Malfunctioning steam dump 3.4 42 Steam Ability to explain and apply all system 039 (SF4S MSS) Main and Reheat 2.1.32 limits and precautions. 3.8 43 Steam Power level restrictions for operation of 059 (SF4S MFW) Main Feedwater 3 MFW pumps and valves. 2.7 44 061 (SF4S AFW) 3 Interactions when multi-unit systems are 3.1 45 Auxiliary/Emergency Feedwater cross tied 061 (SF4S AFW) 2 AFW electric drive pumps 3.7 46 Auxiliary/Emergency Feedwater Operation of inverter (e.g. precharging 062 (SF6 ED AC) AC Electrical 4 synchronizing light,static transfer) 2.7 47 Distribution 063 (SF6 ED DC) DC Electrical 1 Meters, annunciators, dials, recorders 2.7 48 Distribution and indicating lights 064 (SF6 EDG) Emergency Diesel 1 Local and remote operation of the ED/G 4.0 49 Generator 064 (SF6 EDG) Emergency Diesel 8 Fuel oil storage tanks 3.2 50 Generator

ES-401 6 Form ES-401-2 073 (SF7 PRM) Process Radiation 1 Radiation levels 3.2 51 Monitoring 076 (SF4S SW) Service Water 17 PRMS 3.6 52 078 (SF8 IAS) Instrument Air 1 Instrument air compressor 2.7 53 103 (SF5 CNT) Containment 4 Phase A and phase B resets 3.5 54 Ability to perform specific system and 103 (SF5 CNT) Containment 2.1.23 integrated plant procedures during all 4.3 55 modes of plant operation.

004 (SF1; SF2 CVCS) Chemical and 13 Low RWST 3.9 86 Volume Control Ability to recognize system parameters 008 (SF8 CCW) Component Cooling 2.2.42 that are entry-level conditions for 4.6 87 Water Technical Specifications 010 (SF3 PZR PCS) Pressurizer 2.4.46 Ability to verify that the alarms are 4.2 88 Pressure Control consistent with the plant conditions.

013 (SF2 ESFAS) Engineered 4 Loss of instrument bus 4.2 89 Safety Features Actuation 012 (SF7 RPS) Reactor Protection 3 Incorrect channel bypassing 3.7 90 K/A Category Point Totals: 3 2 3 2 2 2 3 3/3 2 3 3/2 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 #

011 (SF2 PZR LCS) Pressurizer 6 Correlation of demand signal indication 2.5 56 Level Control on charging pump flow valve controller to the valve position 014 (SF1 RPI) Rod Position 1 Rod selection control 3.3 57 Indication 016 (SF7 NNI) Nonnuclear 1 Separation of control and protection 2.7 58 Instrumentation circuits 029 (SF8 CPS) Containment Purge 2 Containment entry 2.9 59 033 (SF8 SFPCS) Spent Fuel Pool 2 RHRS 2.5 60 Cooling 035 (SF 4P SG) Steam Generator 1 S/ G water level control 4.0 61 045 (SF 4S MTG) Main Turbine 5 Expected response of primary plant 3.8 62 Generator parameters (temperature and pressure) following T/G trip 068 (SF9 LRS) Liquid Radwaste 3 Insufficient sampling frequency of the 2.5 63 boric acid in the evaporator bottoms 017 (SF7 ITM) In-Core Temperature 2.4.21 Knowledge of the parameters and logic 4.0 64 Monitor used to assess the status of safety functions 086 Fire Protection 2 Maintenance of fire header pressure 3.0 65 016 (SF7 NNI) Nonnuclear 2 Loss of power supply 3.2 91 Instrumentation 045 (SF 4S MTG) Main Turbine 2.1.20 Ability to execute procedure steps. 4.6 92 Generator 075 (SF8 CW) Circulating Water 2 Loss of circulating water pumps 2.7 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: January 2020 Category K/A # Topic RO SRO-only IR # IR #

2.1.42 Knowledge of new and spent fuel movement procedures 2.5 66 Ability to identify and interpret diverse indications to validate 2.1.45 the response of another indication 4.3 67 2.1.44 Knowledge of RO duties in the control room during fuel 3.9 68 handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems

1. Conduct of operated from the control room in support of fueling operations, Operations and supporting instrumentation.

Knowledge of industrial safety procedures (such as 2.1.26 rotating equipment, electrical, high temperature, high pressure, 3.6 94 caustic, chlorine, oxygen and hydrogen).

Subtotal 3 1 2.2.13 Knowledge of tagging and clearance procedures. 4.1 69 (multi-unit) Ability to explain the variations in control board 2.2.4 layouts, systems, instrumentation and procedural actions 3.6 70 between units at a facility.

2. Equipment Control Knowledge of the process for controlling temporary design 2.2.11 changes. 3.3 95 (multi-unit license) Knowledge of the design, procedural and 2.2.3 operational differences between units. 3.9 96 Subtotal 2 2 2.3.11 Ability to control radiation releases. 3.8 71 Knowledge of radiation or contamination hazards that 2.3.14 may arise during normal, abnormal, or emergency conditions or 3.4 72 activities 2.3.5 Ability to use radiation monitoring systems 2.9 73
3. Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or 3.7 97 emergency conditions.

2.3.6 Ability to aprove release permits 3.8 98 Subtotal 3 2 Knowledge of facility protection requirements including fire 2.4.26 brigade and portable fire fighting equipment usage. 3.1 74 2.4.27 Knowledge of "fire in the plant" procedures. 3.4 75

4. Emergency Procedures/Plan 2.4.14 Knowledge of general guidelines for EOP usage. 4.5 99 Knowledge of how abnormal operating procedures are used in 2.4.8 conjunction with EOPs. 4.5 100 Subtotal 2 2 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A Rejected Emergency and Abnormal Plant Evolutions associated 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 RO Question #17: 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 17.

1/1 000065 (APE 65) Cannot write an operationally valid question regarding flow Loss of readings in the instrument air system Instrument Air A2.02 Randomly reselected 065 Loss of Instrument Air, A2.05 SRO Question

  1. 81 2/2 011 (SF2 PZR Cannot write an operationally valid question regarding LCS) Pressurizer demand signal indication on charging pump flow valve Level Control controller to the valve position.

K6.06 Randomly reselected 011 Pressurizer Level Control, K6.04 RO Question #56 2/2 068 (SF9 LRS) Cannot write an operationally valid question regarding Liquid Radwaste insufficient sampling frequency of the boric acid in the evaporator bottoms.

A2.03 Randomly reselected 068 Liquid Radwaste, A2.04 RO Question #63

ES-401 Record of Rejected K/As Form ES-401-4

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