ML19122A291

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Final Written Examination and Operating Test Outlines (Folder 3)
ML19122A291
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 04/29/2019
From:
Exelon Generation Co
To: Todd Fish
Operations Branch I
Shared Package
ML18124A201 List:
References
000500, CAC:000956
Download: ML19122A291 (26)
v • d • e


Text

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 02/25/2019 Examination Level: RO ~ SRO D Operating Test Number: 2019 NRC Administrative Topic Type Describe activity to be performed (See Note)

Code*

Conduct of Operations D,S G2.1.29( 4.1) - Lineup Standby Gas Treatment System For Automatic Operation (PLOR 337C)

Conduct of Operations N,R G2.1.25 (3.9) Perform AO 10.12-2 "Alternate Shutdown Cooling" (PLOR 384C)

G2.2.41 (3.5) - Determine Status of Instrument Nitrogen Equipment Control D,R, P Compressor Discharge Solenoid Valve Using Station Piping and Instrumentation Drawings (PLOR-220C) (2015 NRC)

Radiation Control N/A Not Required Emergency Plan D,R G2.4.43 (2.8) - Direct a Site Evacuation (PLOR-94C)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (~ 3 for ROs; ~ 4 for SROs & RO retakes)

(N)ew or (M}odified from bank (~ 1)

(P)revious 2 exams (~ 1; randomly selected)

ES 301, Page 22 of 27

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Peach Bottom Date of Examination: 02/25/2019 Examination Level: RO D SRO [gl Operating Test Number: 2019 NRC Administrative Topic Type Describe activity to be performed (See Note)

Code*

Conduct of Operations D,R G2.1.20 (4.6) - Review Daily Jet Pump Operability Surveillance (PLOR 282C)

Conduct of Operations D,R G2.1.32 (4.0) - Evaluation Of High CRD Temperature On Control Rod Scram Time (PLOR 347C)

Equipment Control D, R G2.2.40 (4.7) - Compensatory Actions for an Inoperable Fire Door (273C)

Radiation Control D,R G2.3.4 (3. 7) - Review and Authorize Two Emergency Exposures (287C)

Emergency Plan N,R G2.4.41 (3.6) Classification of Emergencies and PARS NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank(~ 3 for ROs; ~ 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (~ 1)

(P)revious 2 exams (~ 1; randomly selected)

ES 301, Page 22 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 02/25/2019 Exam Level: RO ~ SRO-I O SRO-U 0 Operating Test Number: 2019 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code*

Safety Function

a. 202002 A4.07 (3.3/3.2) - Recirculation Flow Control System I Reset the D,S 1

Recirculation System Upper Flow Limit f PLOR-007Cl

b. 206000 A2.09 (3.5/3.7) - High Pressure Coolant Injection/ Raise HPCI A,D, EN,S 2

Flow (Alternate Path - Suction Valves Fail to Swap on Low Condensate Storage Tank Level) (PLOR-333CA)

c. 239001 A4.01 (4.2/4.0) - Main Steam System/ Open Main Steam D, L, S 3

Isolation Valves After a Group-1 Isolation (PLOR-083C)

d. 209001 A4.03 (3. 7/3.6) - Manual Startup of CS for Injection (Alternate A, N, L, S 4

Path - CS Valve Trips on Thermal Overload) (PLOR-383CA)

e. 223002 A4.03 (3.6/3.5) - Primary Containment Isolation System/

D, L, S 5

Perform a Group 1 PCIS Isolation Reset (GP-SA) (PLOR-024C)

f. 262001 A4.04 (3.6/3.7) -AC Distribution/ Excite The Main Generator D,S 6

(PLOR-031C)

g. 212000 A4.01 (4.6/4.6) - Inputting RPS trip IAW GP-25 (Alternate Path -

A,N, EN,S 7

Initial Channel Fails to Input Trip) (PLOR-385CA)

h. 400000 A2.01 (3.3/3.4) Diesel Generator Quick Start from the Control A,D,S 8

Room (Alternate Path - ESW Pump Trips After Auto Start) (PLOR-284CA)

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. 217000 A1.04 (3.6/3.6)- Reactor Core Isolation Cooling/ Defeat RCIC D, E, R 4

Interlocks IAW T-251-2 (PLOR-156P)

j. 218000 K4.04 (3.5/3.6)- Bypass of SV-9130A IAW T-331-3 (PLOR-D, E, R 3

386P) (Unit 3)

k. 201001 A2.06 (2.9/2.9) - Loss of CRD Regulating Function (Outside D, E, R 1

Control Room Actions) (PLOR-073P)

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 I 4-6 I 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 I -

I

e: 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 I ~ 3 I ~ 2 (randomly selected)

(R)CA

~1 I ~ 1 I ~ 1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 02/25/2019 Exam Level: RO O SRO-I !SJ SRO-U 0 Operating Test Number: 2019 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code*

Safety Function

a. 202002 A4.07 (3.3/3.2) - Recirculation Flow Control System I Reset the D,S 1

Recirculation System Upper Flow Limit f PLOR-007Cl

b. 206000 A2.09 (3.5/3.7) - High Pressure Coolant Injection/ Raise HPCI A,D,EN,S 2

Flow (Alternate Path - Suction Valves Fail to Swap on Low Condensate Storage Tank Level) (PLOR-333CA)

c. 239001 A4.01 (4.2/4.0) - Main Steam System/ Open Main Steam D, L, S 3

Isolation Valves After a Group-1 Isolation (PLOR-083C)

d. 209001 A4.03 (3. 7/3.6) - Manual Startup of CS for Injection (Alternate A, N, L, S 4

Path - CS Valve Trips on Thermal Overload) (PLOR-383CA)

e. 223002 A4.03 (3.6/3.5) - Primary Containment Isolation System/

D, L, S 5

Perform a Group 1 PCIS Isolation Reset (GP-8A) (PLOR-024C)

g. 212000 A4.01 (4.6/4.6) - Inputting RPS trip IAW GP-25 (Alternate Path -

A,N, EN,S 7

Initial Channel Fails to Input Trip) (PLOR-385CA)

h. 400000 A2.01 (3.3/3.4) Diesel Generator Quick Start from the Control A,D,S 8

Room (Alternate Path - ESW Pump Trips After Auto Start) (PLOR-284CA)

In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. 217000 A1.04 (3.6/3.6) - Reactor Core Isolation Cooling/ Defeat RCIC D,E, R 4

Interlocks IAW T-251-2 (PLOR156P)

j. 218000 K4.04 (3.5/3.6)- Bypass of SV-9130A IAW T-331-3 (PLOR-D,E,R 3

386P) (Unit 3)

k. 201001 A2.06 (2.9/2.9) - Loss of CRD Regulating Function (Outside D,E,R 1

Control Room Actions) (PLOR-073P)

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 I SRO-U (A)lternate path 4-6 I 4-6 I 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 I -

I

~ 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 I ~ 1 I ~ 1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Peach Bottom Date of Examination: 02/25/2019 Exam Level: RO O SRO-I O SRO-U [:8J Operating Test Number: 2019 NRC Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code*

Safety Function

a.
b.

C.

d. 209001 A4.03 (3.7/3.6) - Manual Startup of CS for Injection (Alternate A, N, L, S 4

Path - CS Valve Trips on Thermal Overload) (PLOR-383CA)

e.
f.
g. 212000 A4.01 (4.6/4.6) - Inputting RPS trip IAW GP-25 (Alternate Path -

A,N,EN,S 7

Initial Channel Fails to Input Trip) (PLOR-385CA)

h. 400000 A2.01 (3.3/3.4) Diesel Generator Quick Start from the Control A,D,S 8

Room (Alternate Path - ESW Pump Trips After Auto Start) (PLOR-284CA)

In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. 217000 A 1.04 (3.6/3.6) - Reactor Core Isolation Cooling I Defeat RCIC D, E, R 4

Interlocks IAW T-251-2 (PLOR156P)

j. 218000 K4.04 (3.5/3.6) - Bypass of SV-9130A IAW T-331-3 (PLOR-D,E,R 3

386P) (Unit 3)

k.

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 I 4-6 I 2-3 (C)ontrol room (D)irect from bank 5.9/5.8/5.4 (E)mergency or abnormal in-plant

?. 1 /?_1/?_1 (EN)gineered safety feature I -

I

<?: 1 (control room system)

(L)ow-Power I Shutdown

?_1/?_1/?_1 (N)ew or (M)odified from bank including 1 (A)

?_2/?_2/?_1 (P)revious 2 exams

5. 3 / 5. 3 / 5. 2 (randomly selected)

(R)CA

?. 1 I ?. 1 I ?. 1 (S)imulator ES-301, Page 23 of 27

Appendix D Scenario Outline Form ES-D-1 Facility: Peach Bottom Scenario No.:

1 Op-Test No.: 2019 NRC Examiners: ------------ Operators:

Scenario Outline:

The scenario begins at 85% power with direction to lower power to 80% to allow the PRO to perform RT-0-001-408-2, "Cycling of Combined Intermediate Valves" for CIV 1 only.

When the RT is complete, Alarm 211 J-3, "Standby Liquid or P~e Hi-Lo Temp" will be received. The crew will receive a report that the SLC tank temperature is 125 F and rising slowly. The CRS should review Tech Specs and determine that Tech Spec 3.1. 7, Condition C applies requiring temperature to be restored in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The Crew should determine that the heater control has failed and that the breaker for the heaters should be opened. The CRS should direct that electrical power be removed to the SLC tank heaters.

When the Tech Spec determination had been made, the "A" Recirc pump speed will oscillate. The URO should notice the change in Reactor power. The Crew may enter and execute OT-104, "Positive Reactivity Insertion" base on the power changes. The Crew should enter and execute OT-112, "Unexpected Unexplained Change in Core Flow". The URO should place a speed hold on the "A" Recirc pump. The CRS should review Tech Specs for the Recirc flow mismatch (3.4.1 ).

After the speed hold is placed on the "A" Recirc pump, alarm "Logic Bus Power Lost" (222 A-5) will alarm. The Crew should enter and execute ARC 222 A-5. The PRO should recognize the Torus suction valves opening and close M0-2-13-18, "Condensate Tank Suction" valve to prevent draining the CST to the Torus. The PRO should also recognize that the RCIC Steam Supply valve is also isolated. The CRS should review Tech Specs for RCIC instrumentation (3.3.5.2 and 3.5.3). With the spurious RCICI system operation and the isolated Steam Supply valve the CRS should determine that RCIC is INOP and that Tech Spec 3.5.3.A applies and RCIC must be restored in 14 days.

When M0-2-13-18 is closed, The "B Service water pump will trip. The Crew should enter and execute ON-127, "Loss of Service Water". The URO should start the standby service water pump.

When the standby service water pump has been started, a trip of the 2R4 Transformer Breaker will occur. Power will be lost to the 2R4 bus. The PRO will be able to cross tie the 2R4 bus with the 1 R4 bus to recover power to loads lost on the loss of power.

The loss of power will also cause a loss of power to the "C" Instrument Air compressor. This will require the crew to enter and execute ON-119, "Loss of Instrument Air". The PRO should direct an Equipment operator to reset the under voltage trips to the "C" air compressor so the "C" air compressor can be restarted.

A Leak on the HPCI steam line will occur. Temperatures in the HPCI room will continue to rise. This will require the Crew to enter and execute T-103, "Secondary Containment Control". HPCI will fail to isolate when the Crew attempts to close M0-2-23-15, "HPCI Steam Isolation" valve. The leak will gradually worsen, requiring a reactor scram and entry into T-101, "RPV Control". (Critical Task: When a Primary System is discharging into Secondary Containment through an unisolable leak, scram the Reactor prior to performing an Emergency Slowdown).

Scenario 1 (1007L) Rev 1

Appendix D Scenario Outline Form ES-D-1 When depressurization using Bypass Valves is performed, Bypass valves will initially function normally but then fail closed, requiring use of the SRVs or alternate methods for depressurization.

Conditions will continue to deteriorate in the Reactor Building due to the HPCI steam leak. When the second Reactor Building area (Torus Room) exceeds its T-103 Action Level, the crew should perform a T-112 "Emergency Slowdown". (Critical Task: Perform an Emergency Blowdown when the second Reactor Building area Temperature exceeds an Action level).

The scenario will end when the RPV is depressurized and RPV level is being maintained with Condensate.

Initial Conditions: Unit 2 is a approximately 85% power with no equipment out of service.

Turnover: Lower power to approximately 80%, then perform RT-0-001-408-2, "Cycling of Combined Intermediate Valves" for the# 1 CIV only.

Critical Tasks:1. When a Primary System is discharging into Secondary Containment through an unisolable leak, scram the Reactor prior to performing an Emergency Slowdown. 2. Perform an Emergency Slowdown when the second Reactor Building area Tem12erature exceeds an Action level.

Event Malf. No.

Event Event No.

Type*

Description 1

See Scenario R URO Lower Reactor Power to approximately 80%

Guide CRS 2

See Scenario N PRO Perform RT-0-001-408-2, "Cycling of Combined Intermediate Guide CRS Valves".

3 See Scenario TSCRS High Temperature alarm on Standby Liquid Control Tank.

Guide Declare Standby Liquid Control !NOP 4

See Scenario CURO "A" Recirc pump speed oscillates, place speed hold on "A" Guide CRS Recirc pump 5

See Scenario C PRO RCIC Logic Bus Power Loss, Close RCIC CST suction Guide TSCRS 6

See Scenario CURO "B" Service water pump trips, enter ON-127 and start the Guide CRS standbv Service water pump 7

See Scenario C PRO Trip of 2R4 Transformer BKR. Cross tie 480 vac load centers.

Guide CRS 8

See Scenario MALL HPCI steam leak. Conditions will degrade requiring a Reactor Guide Scram 9

See Scenario C PRO Isolation fails and conditions degrade requiring a blowdown Guide CRS 10 See Scenario CURO Bypass valves fail closed. Use SRVs for depressurization Guide CRS 11 See Scenario MALL Slowdown when two areas in the Reactor Building exceed the Guide action level.

(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Scenario 1 (1007L) Rev 1

Appendix D Scenario Outline Form ES-D-1 Facility: Peach Bottom Scenario No.:

2 -----

Op-Test No.: 2019 NRC Examiners: ------------ Operators:

Scenario Summary:

The scenario begins with the reactor at 100% power. After taking the shift the crew is required to swap operating TBCCW pumps for inspection of a noisy bearing on the 'A' TBCCW pump.

When the "B" TBCCW pump is in service, an individual control rod drive scram accumulator will experience low nitrogen pressure and alarm in the main control room. The crew will initiate corrective action but the accumulator pressure will remain low requiring the crew to declare the control rod slow or inoperable per Technical Specifications.

Following the Tech Spec determination, the E-4 Diesel Generator will inadvertently start and one minute later 005 F-1 "E-4 Diesel Gen Differential and Ground" alarm will annunciate. The crew will take action in the alarm response to shutdown the E-4 diesel generator and place in Pull-to-Lock. The CRS will apply Technical Specifications for an inoperable Diesel Generator.

Following the Diesel Generator Tech Spec determination, the crew should then recognize and respond to lowering Main Condenser vacuum caused by a failure of the in service Steam Jet Air Ejector steam supply valve. The crew must enter OT-106 "Condenser Low Vacuum" and reduce reactor power in accordance with GP-9-2 "Fast Power Reduction". The PRO should place the alternate air supply in service to recover Main Condenser vacuum.

When Main Condenser vacuum has stabilized, the '"A' Clean-up Recirc Pump Motor Winding Temperature High" alarm will annunciate. The Crew should enter and execute ARC 215 A-1 and dispatch an Equipment Operator to investigate the high temperature condition. The Equipment Operator will report that temperature is 142°F and rising fast. The Crew may elect to Swap RWCU pumps or remove the "A" RWCU pump from service. If they elect to swap RWCU pumps, conditions will continue to deteriorate until the '"A' Clean-up Recirc Pump Motor Winding Temperature High-High" alarm is received. The "A" RWCU pump will not trip as expected and the URO will be required to remove the pump from service.

When the RWCU system is removed from service, the "B" Recirc pump will trip followed by a trip of the "A" Recirc pump one minute later. The URO will identify the loss of forced circulation and scram the Reactor. When the Reactor Mode switch is taken to Shutdown the control rods will begin to insert but a Hydraulic ATWS will occur. The loss of forced circulation will cause Thermal Hydraulic Instabilities.

The Crew should enter and execute T-101, "RPV Control" and T-117, "Level/Power Control." (Critical Task: Inhibit ADS initiation during an ATWS with Feedwater available within 10 minutes and 12 seconds)

When SBLC is initiated the SBLC pump will trip, requiring the URO to place the alternate SBLC pump in service. (Critical Task: Attempt to shut down the Reactor by performing one or more of the following: T-216, "Control Rod Insertion by Manual Scram of Individual Scram Test Switches",

T-220, "Driving Control Rods During a Failure to Scram", Injecting Standby Liquid Control Before Torus Temperature exceeds 110 degrees Fahrenheit.)

Scenario 2 (1008L) Rev 2

Appendix D Scenario Outline Form ES-D-1 The Crew will need to lower RPV level to below -60 inches to minimize the effects of THI. (Critical Task: Perform T-240, "Termination and Prevention of Injection into the RPV to minimize Thermal-hydraulic instabilities (THI) until RPV level is below -60 inches). When the Crew has stabilized RPV level below -60 inches, the "C" RFP will trip and require the PRO to stabilize RPV level with an alternate system (HPCI, RCIC or the "A" or "B" RFP).

The crew should perform T-220, "Driving Control Rods During Failure to Scram" and T-216, "Control Rod Insertion by Manual Scram or Individual Scram Test Switches" to insert control rods.

The scenario may be terminated when the crew has control of RPV power and level using T-240 "Termination and Prevention of Injection into the RPV" and the crew is inserting control rods.

Initial Conditions: IC-14, 100% i;1ower with no eguigment out of service Turnover: The "A" TBCCW gumg is making an unusual noise reguiring the PRO to swag TBCCW oumos.

Critical Tasks:1. Attemgt to shut down the Reactor by gerforming one or more of the following: T-216, "Control Rod Insertion by Manual Scram of Individual Scram Test Switches", T-220, "Driving Control Rods During a Failure to Scram", Injecting Standby Liguid Control Before Torus Temi;1erature exceeds 110 degrees Fahrenheit. (T-101-4) 2. Perform T-240, "Termination and Prevention of Injection into the RPV to minimize Thermal-hydraulic instabilities (THI) until RPV level is below-60 inches. (T-117-1)

3. Inhibit ADS initiation during an ATWS with Feedwater available within 10 minutes and 12 seconds.

(T-117-7 Event Malf. No.

Event Event No.

Type*

Description 1

See Scenario N PRO Swap operating TBCCW Pumps Guide CRS 2

See Scenario TS CRS Individual control rod drive scram accumulator low pressure Guide (Tech Spec) 3 See Scenario C PRO E4 Diesel Generator spurious start/ Diesel Generator Guide TSCRS shutdown (Tech Spec) 4 See Scenario C PRO Failure of Steam Jet Air Ejector Steam Supply valve/ re-open Guide CRS by placinq additional valve air supply in service 5

See Scenario R URO Fast Reactor power reduction (w/ Recirc) for lowering Main Guide CRS Condenser vacuum 6

See Scenario CURO "A" RWCU pump motor winding high temperature, remove Guide CRS the "A" RWCU pump from service and isolate the system 7

See Scenario CURO "B" and "A" Recirc pump trip. Mode switch to Shutdown Guide CRS 8

See Scenario MALL ATWS (hydraulic), lower RPV level to minimize THI Guide 9

See Scenario CURO Standby Liquid Control (SBLC) pump trips/ start second Guide CRS SBLC pump 10 See Scenario CURO "C" RFP trips, control RPV level with HPCI or another RFP Guide CRS (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Scenario 2 (1008L) Rev 2

Appendix D Scenario Outline Form ES-D-1 Facility: Peach Bottom Scenario No.:

3 Op-Test No.: 2019 NRC Examiners: ------------ Operators:

Scenario Outline:

Unit 2 is at approximately 5% power with the "D" HPSW pump blocked out of service for a motor inspection. SRV's "E" and "L" have leaking bellows and a TCCP is applied to clear the alarm. When the Crew has the shift the PRO will secure the Drywell Purge lineup. When the purge line up is secured the URO will begin to withdraw control rods with a goal of three bypass valves open.

When control rods have been withdrawn, the "A" Drywell Chiller will trip. The crew should recognize the trip and start the standby chiller.

When the standby chiller is running, the "Slowdown Relief Valve Bellows Leaking" alarm will be received. The Crew will determine that the "C" SRV has the leaking bellows. The CRS should review Tech Spec (3.4.3) and determine that the SRV is !NOP and condition A applies.

When the Tech Spec determination has been made, the Feedwater Water Level controller will fail causing the setpoint to fail to 12 inches. The Crew should enter and execute OT-100, "Reactor Low Level". The URO should take manual control of the startup level controller (AO 8091) and recover RPV level to 23 inches.

When RPV level is stable, control rod 02-31 will drift into the core. The Crew should enter and execute ON-121, "Drifting Control Rod" to insert and disarm the control rod. The control rod will not settle at position 00 and will need to be scrammed to get the rod to settle at position 00. The CRS should reference Tech Specs 3.1.3 for the INOP rod.

When the Tech Spec determination has been completed, a second control rod will drift into the core requiring the URO to scram the Reactor. (Critical Task: Shutdown the reactor when a second control rod drifts into the core)

Following the Reactor scram, a leak will deve!op in the Torus. The leak will require the Crew to align river water to refill the Torus. When the crew attempts to place the "B" HPSW pump in service it will trip. This will require the crew to place the "A" and "C" HPSW in service from the "A" loop.

When the crew attempts to bypass and restore instrument nitrogen the key lock bypass switch will fail to operate correctly and the Crew will be required to place the backup nitrogen bottles in service to provide a supply of nitrogen to the SRVs.

Torus level will continue to drop until an RPV blowdown is required. (Critical Task: Perform an Emergency Slowdown when Torus level cannot be maintained above 10.5 feet.)

The scenario may be terminated when the RPV is depressurizing and HPSW is injecting into the Torus.

Scenario 3 (1009L) Rev 1

Appendix D Scenario Outline Form ES-D-1 Initial Conditions: Unit 2 is oi;:1erating at a1212roximately 5% 12ower with the "D" HPSW 12um12 out of service for motor insi;:1ection. SRV's "E" and "L" have leaking bellows and a TCCP is ai;:1i;:1lied to clear the alarm.

Turnover: When the Crew takes the shift the PRO will be reguired to secure the DOO!Yell Purge lineui;:1.

The URO will begin withdrawing control rods until 3 byi;:1ass valves are oi;:1en.

Critical Tasks: 1. Shutdown the reactor when a second control rod drifts into the core. 2. Perform an Emergency Slowdown when Torus level cannot be maintained above 10.5 feet.

_.. t Malf. No.

Event Event I

No.

Type*

Description 1

See Scenario N PRO Secure the Drywell Purge lineup Guide CRS 2

See Scenario R URO Continue the Reactor Startup by withdrawing control rods Guide CRS 3

See Scenario C PRO "A" Drywell Chiller trips, start a Drywell chiller Guide CRS 4

See Scenario TSCRS

""C" SRV Bellows Leaking Guide 5

See Scenario CURO "Master Feedwater Controller Failure, recover level with the Guide CRS bvoass in manual 6

See Scenario CURO Control Rod 02-31 Drifts In followed by a second drifting Guide C PRO control rod, insert control rod.

TSCRS 7

See Scenario MALL Torus leak, Fill the Torus with river water Guide 8

See Scenario C PRO Instrument Nitrogen fails to bypass, place the backup bottles Guide CRS in service 9

See Scenario C PRO "B" HPSW pump trip, place the "A" loop of HPSW pumps in Guide CRS service 10 See Scenario MALL RPV Slowdown based on low Torus level Guide (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Scenario 3 (1009L) Rev 1

Appendix D Scenario Outline Form ES-D-1 Facility:

Peach Bottom Scenario No.: _4-'------

Op-Test No.: NRC 2019 Examiners: ------------ Operators:

Scenario Outline:

The scenario begins with Unit 2 at 100% with no equipment out of service. When the Crew has the shift the PRO will place the "B" loop of Torus cooling in service to support testing.

The Crew will have direction to lower Reactor power with control rods to 95% following the supplied ReMA. As the URO insert control rods, a control rod will initially be stuck. The URO will take actions per SO 62.1.A-2, "Withdrawing Inserting a Control Rod". The Control Rod will move when drive pressure is raised.

When the control rod is moving the "D" HPSW pump will experience a timed overcurrent condition.

Timed overcurrent will not trip the "D" HPSW pump and the PRO will need to remove the "D" HPSW pump from service and secure the Torus Cooling lineup. The CRS should review Tech Spec section 3.7.1, 3.6.2.3, 3.6.2.4 and 3.6.2.5 for the INOP HPSW pump.

When the Torus cooling lineup is secured, an RHR pipe break will cause flooding in the "D" RHR pump room. The CRS should direct the PRO to isolate the suction lineup for the "D" RHR pump to isolate the leak in the RHR room. The CRS should review Tech Spec section 3.5.1 for the INOP RHR pump.

When the RHR leak is isolated, a trip of the "A" Condensate pump will occur, the automatic Recirc runback will not occur and the URO will need to perform a manual runback to stabilize RPV level.

When RPV level is stable, a loss of the #2 bus will occur removing power from remaining Condensate pumps. The Crew should recognize the loss of Feedwater and scram the reactor. The Reactor Operators should take their scram actions. The CRS should enter and execute T-101, "RPV Control".

HPCI and RCIC should be used to maintain RPV level. HPCI and RCIC will trip when started. This combined with the loss of Feedwater is a loss of all high pressure feed. The RCIC trip can be reset and RCIC can be restarted. When RCIC is running, a Recirc line break larger than the capacity of RCIC will occur causing RPV level to drop and for Containment pressure to rise. (Critical Task:

Inhibit ADS before an automatic depressurization occurs.) (T-101-9)

When RPV level drops to -172 inches, the crew should perform an Emergency Slowdown to lower RPV pressure and allow low pressure ECCS pumps to inject into the RPV. (Critical Task: Perform an Emergency Slowdown when RPV level reaches -172 inches.) (T-111-4) When RPV pressure drops to 450 psig, M0-2-10-25A, "RHR Inboard Discharge" and M0-2-10-258, RHR Inboard Discharge" valves will trip on magnetics and M0-2-14-12A and M0-2-14-128 "Core Spray Inboard Discharge" valves will not automatically open and the Crew will need to open the Core Spray injection valves using the control switches to recover RPV level (Critical Task: Following an Emergency Slowdown, open a low pressure ECCS injection valve to restore RPV level above -172 inches before RPV pressure is less than 270 psig and RPV level is less than -205 inches) (T-111-6)

The scenario can end when RPV level has recovered above -172 inches.

Scenario 4 ( 1114L) Rev 0

Appendix D Scenario Outline Form ES-D-1 Initial Conditions: Unit 2 is operating at 100% power with no equipment out of service.

Turnover: Place the "D" RHR in Torus cooling for testing. Insert control rods in accordance with the REMa to support testing.

Critical Tasks: 1. Inhibit ADS before an automatic depressurization occurs. (T-101-9) 2. Perform an Emergenc~ Slowdown when RPV level reaches -172 inches. (T-111-4) 3. Following an Emergenc~

Slowdown, open a low gressure ECCS injection valve to restore RPV level above -172 inches before RPV pressure is less than 270 psig and RPV level is less than -205 (T-111-6) inches.

Event Malf. No.

Event Event No.

Type*

Description 1

See scenario N PRO Place Torus cooling in service guide CRS 2

See scenario R URO Insert control rods in accordance with the ReMA auide CRS 3

See scenario CURO Stuck Control Rod, control rod moves when drive pressure is quide CRS raised.

4 See scenario C PRO "D" HPSW pump Over Current, secure HPSW pump and the auide TSCRS Torus Cooling lineup 5

See scenario C PRO "D" RHR room flood, isolate the suction valves to stop the leak quide TSCRS 6

See scenario CURO "A" Condensate pump trip with Recirc Runback Failure, URO guide CRS reduces Recirc flow 7

See scenario MALL Loss of #2 Buss causing a loss of High Pressure Feed (Loss

!Wide of feedwater, HPCI trip, RCIC trio) 8 See scenario CURO RCIC trip, can be manually reset quide CRS 9

See scenario MALL Recirc leak greater than RCIC flow rate, requires an guide Emerqencv Blow down 10 See scenario C PRO ECCS Injection Valves Fail to open, manually align Core quide CRS Sorav for injection (N}ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Scenario 4 (1114L) Rev 0

ES-401 BWR Examination Outline FORM ES-401-1 Facility Name: Peach Bottom Date of Exam: 2/25/2019 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G

G*

1 2

3 4

5 6

1 2

3 4

Total A2 Total

1.

1 3

3 4

4 3

3 20 4

3 7

Emergency &

Abnormal 2

1 2

1 N/A 1

1 N/A 1

7 1

2 3

Plant Evolutions Tier Totals 4 5

5 5

4 4

27 5

5 10 1

3 2

2 2

2 2

2 3

2 3

3 26 2

3 5

2.

Plant 2

1 1

1 2

1 1

1 1

1 1

1 12 0

1 2

3 Systems Tier Totals 4 3

3 4

3 3

3 4

3 4

4 38 3

5 8

3. Generic Knowledge and Abilities 1

2 3

4 1

2 3

4 Categories 10 7

3 2

2 3

1 2

2 2

Note: 1.

Ensure that at least two topics from every applicable KIA 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 KIA category shall not be less than two). (One Tier 3 Radiation Control KIA is allowed if the KIA is replaced by a KIA from another Tier 3 Category).

2.

The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3.

Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to Section 0.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA 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 KIAs 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 KIA categories.

7.*

The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA 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 KIAs.

8.

On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (I Rs) 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 KIA catalog, and enter the KIA numbers, descriptions, I Rs, and point totals(#) on Form ES-401-3. Limit SRO selections to KIAs that are linked to 10 CFR 55.43.

G*

Generic KIAs

ES-401 2

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO) 11------------------..--..--....--....--

~---.-----11 E/APE #/Name I Safety Function K

K K A 2

3 1

295001 Partial or Complete Loss of Forced Core Flow Circulation I 1 & 4 295003 Partial or Complete Loss of AC 16 0

6 295004 Partial or Total Loss of DC Pwr / 6 0

295005 Main Turbine Generator Trip / 3 295006 SCRAM / 1 0

295016 Control Room Abandonment 17 95018 Partial or Total Loss of CCW I 8 0

295019 Partial or Total Loss of Inst. Air/ 8 0

2 95021 Loss of Shutdown Cooling / 4 295023 Refueling Ace / 8 0

3 295024 High Drywell Pressure I 5 295025 High Reactor Pressure I 3 0

9 295026 Suppression Pool High Water 0

Temp./ 5 2

295027 High Containment Temperature I 5 95028 High Drywell Temperature I 5 0

1 95030 Low Suppression Pool Wtr Lvl / 5 0

95031 Reactor Low Water Level/ 2 0

8 295037 SCRAM Condition Present and 0

Reactor Power Above APRM Downscale or Unknown/ 1 295038 High Off-site Release Rate I 9 600000 Plant Fire On Site I 8 700000 Generator Voltage and Electric Grid 0

Disturbances / 6 1

KIA Category Totals:

3 3

4 4

KIA Topic(s)

Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation: Neutron monitoring Knowledge of the reasons for the following responses as they apply to Partial or Complete Loss of AC: Containment isolation Knowledge of the interrelations between Partial or Total Loss of DC Pwr and the following: Battery charger Knowledge of EOP entry conditions and immediate action steps.

Knowledge of the operational implications of the following concepts as they apply to SCRAM: Decay heat generation and removal.

Ability to determine and/or interpret the following as they apply to Control Room Abandonment: Reactor power Knowledge of the reasons for the following responses as they apply to Partial or Total Loss of CCW: Isolation of non-essential heat loads: Plant-Specific Ability to operate and/or monitor the following as they apply to Partial or Total Loss of Inst. Air: Instrument air system valves: Plant-Specific Ability to use plant computers to evaluate system or component status.

Knowledge of the operational implications of the following concepts as they apply to Refueling Accidents: Inadvertent criticality Ability to operate and/or monitor the following as they apply to High Drywall Pressure:

Drywall spray: Mark-1&11 Knowledge of the interrelations between High Reactor Pressure and the following:

Reactor power Knowledge of the reasons for the following responses as they apply to Suppression Pool High Water Temp.. Suppression pool cooling Knowledge of the operational implications of the following concepts as they apply to High Drywall Temperature: Reactor water level measurement Ability to operate and/or monitor the following as they apply to Low Suppression Pool Wtr Lvl: ECCS systems (NPSH considerations): Plant-Specific Ability to operate and/or monitor the following as they apply to Reactor Low Water Level: Alternate injection systems: Plant-specific Knowledge of the reasons for the following responses as they apply to SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown:

Recirculation pump triplrunback: Plant-Specific Knowledge of the specific bases for EOPs.

Ability to determine and/or interpret the following as they apply to Plant Fire On Site:

Fire alarm Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Motors Group Point Total:

ES-401, Page 34 of 50 IR 3.1 3.7 3.1 4.6 3.7 4.1 2.9 3.3 3.9 3.7 4.2 3.9 3.9 0

3.5 3.6 3.8 4.1 3.3 2.8 3.1 20

ES-401 3

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO) 11-----------------..----r---.-----.---

.----r----11 E/APE # / Name I Safety Function 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure I 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure I 5 295011 High Containment Temp I 5 295012 High Drywell Temperature/ 5 95013 High Suppression Pool Temp./ 5 295014 Inadvertent Reactivity Addition I 1 295015 Incomplete SCRAM/ 1 295017 High Off-site Release Rate I 9 295020 Inadvertent Cont. Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Wtr Lvl / 5 95032 High Secondary Containment Area Temperature I 5 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure I 5 295036 Secondary Containment High Sump/Area Water Level / 5 500000 High CTMT Hydrogen Cone. I 5 KIA Category Totals:

K K

K A

0 2

2 3

0 4

0 5

2 0

4 0

KIA Topic(s)

IR 0

Knowledge of the reasons for the following responses as they apply to High Reactor

4.0 Pressure

Safety/relief valve operation: Plant-Specific 0

Ability to evaluate plant performance and make operational judgments based on 4.4 operating characteristics, reactor behavior, and instrument interpretation.

Knowledge of the interrelations between High Drywell Pressure and the following:

2.6 Nitrogen makeup system: Plant-Specific 0

0 Ability to determine and/or interpret the following as they apply to High Suppression 3.2 Pool Temp.. Localized heating/stratification 0

Knowledge of the interrelations between Incomplete SCRAM and the following: Rod 2.6 worth minimizer: Plant-Specific Ability to operate and/or monitor the following as they apply to High Off-site Release 2.7 Rate: Radwaste 0

0 0

0 Knowledge of the operational implications of the following concepts as they apply to 3.9 High Secondary Containment Area Radiation Levels: Personnel protection 0

0 0

0 Group Point Total:

7 ES-401, Page 35 of 50

ES-401 4

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)

System # / Name K K K K K K A A A KIA Topic(s)

IR 2

3 4 5

6 1

203000 RHR/LPCI: Injection Mode 0

Knowledge of the effect that a loss or malfunction of the following will have on 3.4 9

the RHR/LPCI: Injection Mode: Nuclear boiler instrumentation 205000 Shutdown Cooling 0

Knowledge of the operational implications of the following concepts as they apply 2.8 3

to Shutdown Cooling: Heat removal mechanisms Ability to monitor automatic operations of the HPCI including: Lights and alarms:

3.9; 206000 HPCI BWR-2, 3, 4; Ability to manually operate and/or monitor in the control room:

2 Condensate storage tank level: BWR-2, 3, 4 3.5 207000 Isolation (Emergency) 0 Condenser 209001 LPCS 0

2.5 2

Knowledge of electrical power supplies to the following: Valve power 209002 HPCS 0

211000 SLC those predictions, use procedures to correct, control, or mitigate the 3.1; 2

consequences of those abnormal conditions or operations: Inadequate system 3.7 1

Knowledge of the effect that a loss or malfunction of the RPS will have on 212000 RPS 0

following: The ability of the core cooling systems to provide adequate core 3.5 cooling during loss of coolant accidents 15003 IRM 0

Knov,edge of electrical power supplies to the following: IRM channels/detectors 2.5 215004 Source Range Monitor 0

1 Knowledge of the physical connections and/or cause-effect relationships 215005 APRM / LPRM 0

between APRM / LPRM and the following: Reactor manual control system: Plant-3.3 Specific 217000 RCIC Ability to manually operate and/or monitor in the control room: RCIC turbine 3.7 speed 218000 ADS 0

Knowledge of the physical connections and/or cause-effect relationships 3.9 5

between ADS and the following: Remote shutdown system: Plant-Specific 223002 PCIS/Nuclear Steam Supply 0

0 3.5; 2

Shutoff 4

8 3.3 239002 SRVs 3.6 259002 Reactor Water Level Control 3.2; 2

or mitigate the consequences of those abnormal conditions or operations: Loss 4.1 261000 SGTS 0

Knowledge of the effect that a loss or malfunction of the following will have on 3.1 5

the SGTS: Reactor protection system: Plant-Specific Ability to predict and/or monitor changes in parameters associated with 2.9; 262001 AC Electrical Distribution operating the AC Electrical Distribution controls including: Bus voltage; Ability to 2

interpret and execute procedure steps.

4.6 262002 UPS (AC/DC)

Ability to manually operate and/or monitor in the control room: Transfer from 2.8 alternative source to preferred source 263000 DC Electrical Distribution 0

Knowledge of DC Electrical Distribution design feature(s) and/or inte~ocks which 3.1 1

provide for the following: Manual/ automatic transfers of control: Plant-Specific 264000 EDGs 0

Knowledge of the operational implications of the following concepts as they apply 3.4 5

to EDGs: Paralleling AC. power sources 300000 Instrument Air 0

on following: Containment air system; Ability to (a) predict the impacts of the 2.7; 2

following on the Instrument Air; and (b) based on those predictions, use 2.9 00000 Component Cooling Water Ability to predict and/or monitor changes in parameters associated with 2.8 operating the Component Cooling Water controls including: CCW temperature otals:

3 2

2 2

2 2

roup Point Total:

26 ES-401, Page 36 of 50

ES-401-1 5

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO)

System # I Name K

K K

K K

K IR 2

3 4

5 6

01001 CRD Hydraulic 0

3.1 3

01002 RMCS 0

01003 Control Rod and Drive Mechanism 0

01004 RSCS 0

01005 RCIS 0

01006 RWM 0

02001 Recirculation 0

02002 Recirculation Flow Control 0

04000 RWCU 0

14000 RPIS 0

15001 Traversing In-core Probe 0

15002 RBM 0

Knowledge of the effect that a loss or malfunction of the following will have on the 2.8 4

RBM: APRM reference channel: BWR-3, 4, 5 16000 Nuclear Boiler Inst.

Ability to predict and/or monitor changes in parameters associated with operating 3.4 the Nuclear Boiler Inst. controls including: Recorders and meters 19000 RHR/LPCI: Torus/Pool Cooling 0

Mode 23001 Primary CTMT and Aux.

0 26001 RHR/LPCI: CTMT Spray Mode Ability to perform without reference to procedures those actions that require 4.6 immediate operation of system components and controls.

30000 RHR/LPCI Torus/Pool Spray Mode 0

33000 Fuel Pool Cooling/Cleanup Ability to manually operate and/or monitor in the control room: Pool temperature 2.7 34000 Fuel Handling Equipment 0

39001 Main and Reheat Steam 0

39003 MSIV Leakage Control 0

41000 Reactor/Turbine Pressure Regulator Ability to monitor automatic operations of the Reactor/Turbine Pressure 2.8 Regulator including: Turbine speed control: Plant-Specific 45000 Main Turbine Gen./ Aux.

0 56000 Reactor Condensate 0

2.7 59001 Reactor Feedwater 3.1 68000 Radwaste 0

2.7 71 000 Offgas 0

72000 Radiation Monitoring 0

3.6 3

86000 Fire Protection 0

Knowledge of the effect that a loss or malfunction of the Fire Protection will have 3.6 3

on following: Plant protection 88000 Plant Ventilation 0

Knowledge of Plant Ventilation design feature(s) and/or interlocks which provide 3.7 for the following: Automatic initiation of standby gas treatment system 90001 Secondary CTMT 0

90003 Control Room HVAC 0

90002 Reactor Vessel Internals 0

KIA Category Totals:

2 12 ES-401, Page 37 of 50

ES-401 2

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO) 11----------------~-..---,----,--

'--'----------------.---~---II E/APE # / Name I Safety Function 295001 Partial or Complete Loss of Forced Core Flow Circulation I 1 & 4 295003 Partial or Complete Loss of AC / 6 295004 Partial or Total Loss of DC Pwr / 6 295005 Main Turbine Generator Trip / 3 295006 SCRAM / 1 295016 Control Room Abandonment/ 7 295018 Partial or Total Loss of CCW / 8 295019 Partial or Total Loss of Inst. Air/ 8 295021 Loss of Shutdown Cooling/ 4 295023 Refueling Ace / 8 295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp./ 5 295027 High Containment Temperature/ 5 295028 High Drywell Temperature/ 5 295030 Low Suppression Pool Wtr Lvl / 5 295031 Reactor Low Water Level/ 2 295037 SCRAM Condition Present nd Reactor Power Above APRM Downscale or Unknown / 1 295038 High Off-site Release Rate I 9 600000 Plant Fire On Site I 8 700000 Generator Voltage and Electric Grid Disturbances / 6 ategory Totals:

K K

K A

1 2

3 0

0 0

0 KIA Topic(s)

IR Ability to determine and/or interpret the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation: Individual jet pump

3.1 flows

Not-BWR-1&2 0

0 Ability to prioritize and interpret the significance of each annunciator or 4.3 alarm.

0 0

0 Ability to determine and/or interpret the following as they apply to Partial 3.6 or Total Loss of Inst Air: Instrument air system pressure 0

Knowledge of how abnormal operating procedures are used in 4.5 conjunction with EOPs.

0 0

0 0

0 0

Ability to determine and/or interpret the following as they apply to Reactor 4.2 Low Water Level: Reactor pressure Ability to determine and/or interpret the following as they apply to SCRAM Condition Present and Reactor Power Above APRM Downscale or

4.1 Unknown

Reactor pressure Ability to locate control room switches, controls, and indications, and to 4.3 determine that they correctly reflect the desired plant lineup.

0 0

Group Point Total:

7 ES-401, Page 34 of 50

ES-401 3

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO) 11----------------""T'"--.---..---.--

.--------;__-'--...;.._------------.----.----n E/APE #/Name I Safety Function 295002 Loss of Main Condenser Vac / 3 95007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure/ 5 295011 High Containment Temp I 5 295012 High Drywell Temperature/ 5 295013 High Suppression Pool Temp./ 5 295014 Inadvertent Reactivity Addition/ 1 295015 Incomplete SCRAM/ 1 295017 High Off-site Release Rate I 9 295020 Inadvertent Cont. Isolation I 5 & 7 295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Wtr Lvl / 5 295032 High Secondary Containment Area Temperature I 5 95033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Level / 5 500000 High CTMT Hydrogen Cone. I 5 KIA Category Totals:

K K

K A

2 3

0 0

0 0

KIA Topic(s)

Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnonmal operating procedures.

Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Ability to detenmine and/or interpret the following as they apply to High Suppression Pool Wtr Lvl: Drywell/containment water level Group Point Total:

ES-401, Page 35 of 50 IR 4.7 4.2 3.5 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

3

ES-401 4

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (SRO) 11---------------~---~---~-

r---,----il System # I Name 203000 RHR/LPCI: Injection 205000 Shutdown Cooling Mode 206000 HPCI 207000 Isolation (Emergency)

Condenser 209001 LPCS 209002 HPCS 211000SLC 212000 RPS 215003 IRM 215004 Source Range Monitor 215005 APRM / LPRM 217000 RCIC 18000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff 239002 SRVs 259002 Reactor Water Level Control 261000 SGTS 262001 AC Electrical Distribution 262002 UPS (AC/DC) 263000 DC Electrical Distribution 264000 EDGs 300000 Instrument Air 00000 Component Cooling Water KIA Category Totals:

K K K K K K A A A 234561 0

0 0 0 0

0 0

KIA Topic(s)

Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

system, and understand how operator actions and directives affect plant and system conditions.

Ability to (a) predict the impacts of the following on the RCIC; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: D.C. power loss Knowledge of limiting conditions for operations and safety limits.

ES-401, Page 36 of 50 IR 0

0 0

0 4.6 0

0 3.9 4.4 0

0 3.3 0

0 0

0 0

0 0

4.7 0

0 0

ES-401 5

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (SRO)

System # / Name K

K K K K K KIA Topic(s)

IR 2

3 4

5 6

01001 CRD Hydraulic 0

01002 RMCS 0

01003 Control Rod and Drive Mechanism 0

01004 RSCS 0

01005 RCIS 0

01006 RWM 0

02001 Recirculation 0

02002 Recirculation Flow Control 0

04000 RWCU Ability to apply Technical Specifications for a system.

4.7 14000 RPIS 0

15001 Traversing In-core Probe 0

15002 RBM 0

16000 Nuclear Boiler Inst.

0 19000 RHR/LPCI: Torus/Pool Cooling 4.3 Mode 23001 Primary CTMT and Aux.

0 26001 RHR/LPCI: CTMT Spray Mode 0

30000 RHR/LPCI: Torus/Pool Spray Mode 0

33000 Fuel Pool Cooling/Cleanup 0

34000 Fuel Handling Equipment 0

39001 Main and Reheat Steam 0

39003 MSIV Leakage Control 0

41000 Reactor/Turbine Pressure Regulator 0

45000 Main Turbine Gen./ Aux.

0 56000 Reactor Condensate 0

59001 Reactor Feedwater 0

68000 Radwaste 0

71000 Olfgas Ability to interpret reference materials, such as graphs, curves, tables, etc.

4.2 72000 Radiation Monitoring 0

86000 Fire Protection 0

88000 Plant Ventilation 0

90001 Secondary CTMT 0

90003 Control Room HVAC 0

90002 Reactor Vessel Internals 0

KIA Category Totals:

0 0

0 0

0 0

3 ES-401, Page 37 of 50

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3

!Facility Name:Peach Bottom Date of Exam:2/25/2019 I

Category KIA#

Topic KU SRO-Only IR IR 2.1. 32 Ability to explain and apply system limits and precautions.

3.8 1

4.0 2.1. 08 Ability to coordinate personnel activities outside the control room.

3.4 1

4.1 2.1. 45 Ability to identify and interpret diverse indications to validate the response of another 4.3 1

4.3

1.

indicator.

Conduct of 2.1. 42 Knowledge of new and spent fuel movement procedures.

2.5 3.4 1

Operations 2.1.

2.1.

Subtotal 3

1 2.2. 01 Ability to perform pre-startup procedures for the facility, including operating those controls 4.5 1

4.4 associated with plant equipment that could affect reactivity.

2.2. 39 Knowledge of less than or equal to one hour Technical Specification action statements for 3.9 1

4.5 systems.

2.

2.2. 05 Knowledge of the process for making design or operating changes to the facility.

3.2 1

Equipment 2.2. 18 Knowledge of the process for managing maintenance activities during shutdown 2.6 3.9 1

Control operations, such as risk assessments, work prioritization, etc.

2.2.

2.2.

Subtotal 2

2 2.3. 14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, 3.4 1

3.8 or emergency conditions or activities.

2.3. 05 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, 2.9 1

2.9 portable survey instruments, personnel monitoring equipment, etc.

2.3. 15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, 2.9 3.1 1

3.

portable survey instruments, personnel monitoring equipment, etc.

Radiation Knowledge of radiological safety procedures pertaining to licensed operator duties, such 2.3. 13 as response to radiation monitor alarms, containment entry requirements, fuel handling 3.4 3.8 1

Control resoonsibilities access to locked hinh-radiation areas alinninn filters etc.

2.3.

2.3.

Subtotal 2

2 2.4. 29 Knowledge of the emergency plan.

3.1 1

4.4 Knowledge of events related to system operation/status that must be reported to internal 2.4. 30 organizations or external agencies, such as the State, the NRC, or the transmission 2.7 1

4.1 svstem ooerator

4.

2.4. 25 Knowledge of fire protection procedures.

3.3 1

3.7 Emergency Ability to take actions called for in the facility emergency plan, including supporting or Procedures 2.4. 38 acting as emergency coordinator if required.

4.4 1

/ Plan 2.4. 44 Knowledge of emergency plan protective action recommendations.

4.4 1

2.4.

Subtotal 3

2 Tier 3 Point Total 7

ES-401, Page 43 of 50

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