ML111860008

2011-06 - Draft Outlines

ML111860008
Person / Time
Site:	Cooper
Issue date:	02/28/2011
From:	Kelly Clayton Operations Branch IV
To:	Nebraska Public Power District (NPPD)
References
50-298/11-06, ES-301, ES-301-1
Download: ML111860008 (38)

Text

ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 2/28/11 Exam Level: RO SRO-I SRO-U Operating Test No.:

Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title Type Code*

Function

a. SKL0342XXXR00-J-Respond to a Trip of a RR Pump S,D,E 1

b. Perform quick restart of RFPT B (Hard Card) (Alt. Path) A,D,P 2

c. SKL034-21-XXX R00 Restart Reactor Building Ventilation S,N 8

d. SKL034-20-107 Lowering DEH pressure setpoint S,L,D 3

e. SKL0342XXXXR00-J-Respond to a HPCI System Automatic A,N,En 4 Initiation (Alternte Path)

f. SKL03420XXR00-J-Monitor SGT System Following Automatic A,S,N 9 Initiation(Alt Path)

g. Align RPS to Alternate Power from the Control Room N,En 6

h. SKL03420XXXR00-J-Verify Group 1 Isolation (Alt Path) A,N,S 5 In-Plant Systems @ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. SKL03410106R00- J - Emergency Shutdown a Diesel A,N,En 6 Generator (Alternate Path)

j. SKL0341058R10-J-Startup the RPS Motor Generator Set D 7

k. SKL0341085R04-J-Place standby CRD Flow Control Valve in D,R 1 Service When In service Valve Fails Closed

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank <9/<8/<4 (E)mergency or abnormal in-plant >1/>1/>1 (EN)gineered safety feature - / - / >1 (control room system)

(L)ow-Power / Shutdown >1/>1/>1 (N)ew or (M)odified from bank including 1(A) >2/>2/>1 (P)revious 2 exams < 3 / < 3 / < 2 (randomly selected)

(R)CA >1/>1/>1 (S)imulator ES-301, Page 23 of 27

ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 2/28/11 Exam Level: RO SRO-I SRO-U Operating Test No.:

Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title Type Code*

Function

a. SKL0342XXXR00-J-Respond to a Trip of a RR Pump S,D,E 1

b. Perform quick restart of RFPT B (Hard Card) (Alt. Path) A,D,P 2

c. N/A

d. SKL034-20-107 Lowering DEH pressure setpoint S,L,D 3

e. SKL0342XXXXR00-J-Respond to a HPCI System Automatic A,N,En 4 Initiation (Alternte Path)

f. SKL03420XXR00-J-Monitor SGT System Following Automatic A,S,N 9 Initiation(Alt Path)

g. Align RPS to Alternate Power from the Control Room N,En 6

h. SKL03420XXXR00-J-Verify Group 1 Isolation (Alt Path) A,N,S 5 In-Plant Systems @ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. SKL03410106R00- J - Emergency Shutdown a Diesel A,N,En 6 Generator (Alternate Path)

j. SKL0341058R10-J-Startup the RPS Motor Generator Set D 7

k. SKL0341085R04-J-Place standby CRD Flow Control Valve in D,R 1 Service When In service Valve Fails Closed

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank <9/<8/<4 (E)mergency or abnormal in-plant >1/>1/>1 (EN)gineered safety feature - / - / >1 (control room system)

(L)ow-Power / Shutdown >1/>1/>1 (N)ew or (M)odified from bank including 1(A) >2/>2/>1 (P)revious 2 exams < 3 / < 3 / < 2 (randomly selected)

(R)CA >1/>1/>1 (S)imulator ES-301, Page 23 of 27

Appendix D Scenario Outline Form ES-D-1 Facility: Cooper Nuclear Station Scenario No.: NRC 1 Op-Test No.: 1 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 80% power in single valve control with the Main Turbine Governor Valves. After turnover, the crew is to placed the governor valves into Sequential GV control and continue the power ascension to 100% at XXXXX Rate.

Event Event Event No. Malf. No.

Type* Description 1 N/A N Place MT Gov Valves into sequential valve control 2 N/A R Raise Power with Flow 3 3 I APRM A fails upscale 4 4 C SRV Fails Open 5 5 N,C HPCI Aux Oil Pump failure 6 6 C RR pump trip 7 N/A C Operation in the Exclusion Region and THI - Manual Scram 8 8 M Earthquake; Suppression Pool Rupture: ED on Low SP/L

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 1

Appendix D Scenario Outline Attachment 2 Scenario Objective Evaluate the crews ability to perform normal operations and raise power during non-emergency operations Evaluate the crews response to a lowering Suppression Pool Level and their ability to emergency depressurize the Reactor to suppress the steam in the torus before level gets too low and its energy is released into the Reactor Building.

Scenario Summary Initial Conditions:

The plant is operating at approximately 80% power No equipment out of service.

It is a red light day because record grid loads are expected.

Events:

Shift Main Turbine Governor Valves to Single Valve Control Raise power to 100%

APRM A fails upscale SRV Partial opening HPCI Aux Oil Pump failure RR Pump Trip Thermal Hydraulic Instability - Manual Scram; Earthquake; Suppression Pool leak: ED on low SP/L Scenario Sequence Normal evolution - Place MT into Sequential Gov. Valve Control Reactivity manipulation - Raise Power with RR Flow Instrument Failure - APRM A fails upscale Component Failure before the EOPs - SRV Partial opening Component Failure before the EOPs - HPCI Aux Oil Pump failure Component Failure before the EOPs - RR Pump Trip Thermal Hydraulic Instability - Manual Scram; Major Event - Earthquake; Component Failure after the EOPs - Suppression Pool leak, below the water line:

Emergency Depressurization on low Suppression Pool Level.

2

Appendix D Scenario Outline Attachment 2 Event One: Place MT Gov Valves into sequential valve control Malfunction Required:

No malfunction required; this is a normal manipulation for the BOP.

Objective:

Evaluate the crews ability to select the Digital Electro Hydraulic Controller and perform the required manipulations on the touch screen to shift the main turbine valve governor control to a Sequential alignment.

Success Path:

The Operator, in accordance with Procedure 2.2.77.1, performs the steps necessary to make the Turbine Governor Valves transition smoothly from single valve control to sequential valve control which causes turbine efficiency improvement (MWatts increase).

Event Two: Raise Power to 100%

Malfunction Required:

No malfunction required, this is a normal manipulation for the BOP.

Objective:

Evaluate the RO ability to adjust Reactor Recirculation Pump flows to make power increase from ~80% to ~100%, by alternately raising flow on one RR Pump, letting the plant respond then making a similar adjustment on the other pumps controller.

Success Path:

Both Reactor Recirculation Pumps speeds are raised from ~65 to ~80, maintaining them within the required 5% allowed by Tech Specs, and it is done in accordance with Procedure 2.1.10.

Event Three: APRM A fails upscale Malfunction Required:

Malfunction NM09A APRM Signal Failure Channel A; Event 4 Final value 100.

Objective:

Evaluate the crews response to APRM A fails upscale resulting in a half scram.

Evaluate the RO bypassing the failed APRM and resetting the half scram in accordance with the annunciator card.

Evaluate the CRS addressing Technical Specifications and TRM for the failed APRM.

Success Path:

APRM A is bypassed, the half scram is reset. The CRS initiates a potential LCO on APRM A in accordance with Technical Specifications 3.3.1.1 (RPS Instrumentation) 3

Appendix D Scenario Outline Attachment 2 Table 3.3.1.1-1 Function 2 and TRM 3.3.1 (Rod Block Instrumentation) Table 3.3.1-1 Function 3.

Event Four: SRV Fails Open Malfunction Required:

AD06b set at 50% to start the relief valve leaking then it is modified to 20% to minimize the heat addition to the torus.

Objective:

Evaluate the crews performance of Abnormal Procedure 2.4SRV for a leaking SRV which has the Operator cycle the leaking valve to reseat it.

Evaluate the CRS addressing Technical Specification 3.6.2.1 for Suppression Pool Average Temperature, if the SRV is not closed within a short period of time.

Success Path:

The Operator notices and responds to the leaking SRV in accordance with Annunciator Procedures and 2.4SRV. The SRV will be cycled open then taken back to the closed position to reseat the valve. Once the valve has been cycled the tail pipe temperature starts lowering and the heat addition into the Torus is secured. The CRS determines that Technical Specification 3.6.2.1 for Suppression Pool Average Temperature applies or not.

Event Five: HPCI Aux Oil Pump failure Malfunction Required:

HP12 Active set at 100%, at the start of the scenario. This failure will not start until the HPCI Aux Oil Pump is started.

Objective:

Evaluate the crews response to an oil leak in the HPCI control oil system while performing the section of the Operating Procedure for the HPCI system 2.2.33 to start the Aux Oil Pump for maintenance. During this event the Operator will be required to secure the pump.

Evaluate the CRSs ability to determine that HPCI is Inoperable.

Success Path:

The HPCI system is declared Inoperable per TS 3.5.1. Condition C 14 day LCO. RCIC is checked to be operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the Aux Oil Pump is secured to prevent operation.

4

Appendix D Scenario Outline Attachment 2 Event Six: RR Pump Trip Malfunction Required:

RR05a Trip RR MG Set field breaker.

Objective:

Evaluate the crews response in accordance with Abnormal Procedure 2.4RR and determine Operation in the Exclusion Region of the Power to Flow Map.

Evaluate the CRS addressing Tech Specs for operation in the stability exclusion region.

Success Path:

The Operator determines the Recirc Pump tripped and enters Abn Procedure 2.4RR and notes operation in the Stability Exclusion Region of the Power to Flow Map and makes preparations to exit the region. CRS will declare the pump not in operation in accordance with Tech Specs 3.4.1 Condition A.

Event Seven: Thermal Hydraulic Instability - Manual Scram Malfunction Required:

CR04b Core Thermal Hydraulic Instability Out of Phase 20%

RR05a RR Pump Field Breaker Trip IOR ZAIRRFCDEMD(2) = 60% to generate a runback on the other Recirc pump.

Objective:

Evaluate the crews recognition of abnormal neutron flux oscillations are occurring while operating in the Stability Exclusion Region and Scrams the reactor.

Success Path:

The Operator scrams the plant, in accordance with 2.4RR when core thermal instability is noticed.

Event Ten: Earthquake; Suppression Pool leak: ED on low SP/L Malfunction Required:

HV02b Major Earthquake set to 25%

PC08 Suppression Pool Water Leak 25% level lowers at -0.2/min Objective:

Evaluate the crews response to a major earthquake in accordance with Emergency Procedure 5.1Quake.

Evaluate the crews ability to monitor and control the consequences of an Unisolable leak in the torus below the normal water level.

Evaluate the crews ability to anticipate emergency depressurization and transfer as much energy to the condenser prior to emergency depressurizing the Reactor.

5

Appendix D Scenario Outline Attachment 2 Success Path:

The crew emergency depressurizes the reactor when suppression pool level lowers to 9.6.

Scenario Termination:

When the reactor is depressurized (50 psig above Torus pressure) and level is being maintained between +3 to +54 and the lead examiner has seen enough the scenario may be terminated.

6

Appendix D Scenario Outline Form ES-D-1 Facility: Cooper Nuclear Station Scenario No.: NRC 2 Op-Test No.: 1 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 100% power. After turnover, the crew is to secure DG#1 following its monthly load test.

Event Event Event No. Malf. No.

Type* Description 1 N/A N DG-1 Monthly Surveillance 2 2 I LPRM Fails Upscale 3 3 I Turbine Building Vent Radiation Monitor fails 4 4 C FWH-5B Low Level, Loss of FW heating 5 5 C DEH Leak Requiring Manual Scram 6 6 C Loss of Startup Transformer 7 7 C DG-1/2 Fails to Auto Start 8 8 M Small Break LOCA, Containment Sprays

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 1

Appendix D Scenario Outline Attachment 2 Scenario Objective Evaluate the crews ability to perform normal surveillances and to respond to instrument and component failures during non-emergency conditions Evaluate the crews response to a rising reactor power level and a lowering Turbine High Pressure Fluid Reservoir.

Evaluate the crews response to a small break LOCA which causes Drywell temperatures and pressures to rise and before 280F in the Drywell the crew initiates Drywell Sprays to maintain temperature below 280F.

Scenario Summary Initial Conditions:

The plant is operating at approximately 100% power No equipment out of service.

6.1DG.101 is in progress.

Events:

DG-1 Monthly Surveillance LPRM Fails Upscale Turbine Building Vent Radiation Monitor fails FWH-5B Low Level Loss of FW heating DEH Leak Requiring Manual Scram Loss of Startup Transformer DG-1/2 Fails to Auto Start Small break LOCA Containment Spray Scenario Sequence Normal activity - DG-1 Monthly Surveillance Instrument Failure - LPRM Fails Upscale Instrument Failure - Turbine Building Vent Radiation Monitor fails Component Failure before EOPs - FWH-5B Low Level Loss of FW heating Component Failure before EOPs - DEH Leak Requiring Manual Scram Component Failure after EOPs - Loss of Startup Transformer Component Failure after EOPs - DG-1/2 Fails to Auto Start Major Failure - Small break LOCA, with raising DW temperature and pressure Accident mitigation strategy - Containment Sprays 2

Appendix D Scenario Outline Attachment 2 Event One: DG-1 Monthly Surveillance Malfunction Required:

No malfunction required; this is a normal manipulation for the BOP.

Objective:

Evaluate the crew during normal surveillance activities.

Evaluate the BOP Operator unloading and securing DG1 in accordance with 6.1DG.101, 31 day load test.

Success Path:

The #1 DG is unloaded from 3500 KW and the engine is secured in accordance with the steps in the surveillance. Time compression for hold points between load changes is needed.

Event Two: LPRM Fails Upscale Malfunction Required:

Malfunction NM082813c LPRM 28-13C failure set to 100%.

Objective:

Evaluate the crews response to a failed Local Power Range Monitor (LPRM).

Evaluate the At the Controls (ATC) Operators actions to determine which LPRM and APRM is affected and bypass the failed LPRM.

Evaluate the CRS addressing Technical Specifications for the failed LPRM.

Success Path:

LPRM 28-13C is bypassed, and the CRS initiates a potential LCO on APRM A in accordance with Technical Specifications 3.3.1.1 (RPS Instrumentation) Table 3.3.1.1-1 Function 2 and TRM 3.3.1 (Rod Block Instrumentation) Table 3.3.1-1 Function 3.

Event Three: Turbine Building Vent Radiation Monitor fails Malfunction Required:

RM02J Gas Radiation Monitor Turbine Building Normal Range Kaman RMV-RM-20A in at 100% fails the instrument upscale.

Objective:

Evaluate the crews response to the failure of the normal range KAMAN and takes appropriate action in accordance with the Annunciator Procedure.

Evaluate the CRS addressing Technical Specifications, TRM and ODAM.

Success Path:

Alternate sampling of this release path is requested of Radiation Protection to be installed to provide compensatory actions in accordance with TRM T3.3.3; ODAM 3

Appendix D Scenario Outline Attachment 2 D3.2.1, 3.2.2, and 3.2.3 Event Four: FWH-5B High Level Trip Malfunction Required:

FW20b 1B-5 FW heater level controller failure high. 20%

Objective:

Evaluate the crews response to Annunciator A-2/C-5 HEATER LOW LEVEL.

Evaluate the BOP Operators entry into Abnormal Procedure 2.4EX-STM.

Evaluate the ATC Operator lowering power in response to a lowering feedwater temperature.

Success Path:

Reactor Power has been lowered to the value less than it was prior to the feedwater heater level problem.

Event Five: DEH Leak Requiring Manual Scram Malfunction Required:

TC10 Turbine High Press Fluid leak set at 15%.

TC09d Governor Valve #4 Failure to 0% over 2.5 minutes.

Objective:

Evaluate the crews response to a lowering Turbine High Pressure Fluid reservoir level and Fluid leak on the #4 Turbine Governor Valve.

Evaluate the pre-staging and conservative decision making prior to the need to Scram the Reactor prior to losing Turbine High Pressure Fluid Pumps and control of Turbine GVs, Stop Valves and Bypass Valves.

Success Path:

Reactor is scrammed and pressure control is transferred to HPCI and SRVs.

Event Six: Loss of Startup Transformer Malfunction Required:

ED05 Loss of Power (Startup Transformer).

Objective:

Evaluate the crews response to the loss of the Startup Transformer during the Scram recovery.

Evaluate the BOPs ability to enter Emergency Procedure 5.3EM-PWR and ensure that 4

Appendix D Scenario Outline Attachment 2 the Critical Busses are powered by an emergency power source.

Evaluate the crews ability to shift RPV level control to the High Pressure ECCS and RCIC systems due to a loss of all Condensate and Booster pumps.

Success Path:

Critical Busses are supplied by the Emergency Transformer.

RPV Level is being controlled within the +3 to 54 inch range with RCIC, CRD and HPCI.

Event Seven: DG-1/2 Fails to Auto Start Malfunction Required:

DG06A Diesel Generator #1 Fails to Auto Start DG06B Diesel Generator #2 Fails to Auto Start Objective:

Evaluate the crew recognition that both Diesels failed to auto start when required and to perform the necessary steps to start both Diesel and make them available to load.

Success Path:

Both Diesel Generators are started.

Event Eight: LOCA Containment Sprays Malfunction Required:

RR20A Coolant Leakage Inside Primary Containment Objective:

Evaluate the crew response to a slow increase in Drywell Temperature and pressure and to vent Primary Containment in an attempt to control the pressure rise.

Evaluate the crews ability to spray the Drywell in accordance with the EOPs to control pressure and temperature, as the LOCA gradually worsens.

Success Path:

Torus and Drywell Sprays are initiated prior to DW temperature reaching 280F.

Scenario Termination:

When Reactor water level is being controlled between +3 and +54 inches and Drywell Sprays are controlling Drywell Pressure between 2 and 10 psig.

5

Appendix D Scenario Outline Form ES-D-1 Facility: Cooper Nuclear Station Scenario No.: NRC 3 Op-Test No.: 1 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 1.5% power. After turnover, the crew is to shift CRD Pumps.

Event Event Event No. Malf. No.

Type* Description 1 N/A N Shift CRD Pumps 2 2 C REC Pump B trip 3 3 I IRM fails downscale 4 4 C Rod Drop 5 5 C ATWS non-EOP Rod driving 6 6 M RCIC Steam Line Leak 7 7 C Group 6 Failure 8 8 ED on Secondary Containment 2 Areas

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 1

Appendix D Scenario Outline Attachment 2 Scenario Objective Evaluate the Crews ability to operate at low power levels.

Evaluate the Crews response to a rod drop accident and fuel failure.

Evaluate the crews actions when RCIC develops a steam line leak that will not fully isolate and when two areas in Secondary Containment exceed their Max Safe values the RPV is Emergency Depressurized to limit the release of highly radioactive steam through the RCIC steam line break.

Scenario Summary Initial Conditions:

The plant is operating at approximately 1.5% power IRM G is fully withdrawn and will not drive in It has been declared inoperable.

Shift CRD Pumps following turnover.

Events:

Shift CRD Pumps REC Pump B trip IRM fails downscale Rod Drop ATWS non-EOP Rod driving RCIC Steam Line Leak Group 6 Failure ED on Secondary Containment 2 Areas Scenario Sequence Normal activity - Shift CRD Pumps B to A Component Failure before EOPs - REC Pump B Trip Instrument Failure - IRM C Failure Component Failure before EOPs - Rod Drops causing fuel failure Major Event - ATWS Component Failure after EOPs - RCIC Steam Line Leak Component Failure after EOPs - Group 6 Failure to isolate containment.

Accident mitigation strategy - Emergency Depressurize the RPV 2

Appendix D Scenario Outline Attachment 2 Event One: Shift CRD Pumps Malfunction Required:

No malfunction required; this is a normal manipulation for the RO.

Objective:

Evaluate the crew during normal equipment shifting.

Evaluate the Reactor Operator shifting from the B CRD Pump running to the A CRD Pump running and securing the B Pump.

Success Path:

The A CRD Pump is running and the B CRD Pump is secured. All CRD parameters indicated on Panel 9-5 restored to within their normal band.

Event Two: REC Pump B trip Malfunction Required:

SW11B - REC Pump Trip 1B.

Objective:

Evaluate the crews response to the tripping of one of the three Reactor Equipment Cooling Pumps and takes appropriate action in accordance with the Annunciator Procedure to restart another pump prior to receiving an REC Isolation.

Evaluate the CRS addressing Technical Specifications.

Success Path:

The BOP Operator either responds quickly enough (within 1 minute) to the tripping of the pump and starts an additional REC Pump in accordance with the Annunciator Card.

Or, the REC system isolation is reset following the restart of the third REC Pump and system flows and pressures are returned to normal.

The SRO will address Tech Specs and determine that LCO 3.7.3 Condition B, a 30 day LCO on one sub system.

Event Three: IRM fails downscale Malfunction Required:

Malfunction NM13C - IRM INOP Channel-C.

Objective:

Evaluate the crews response to a failed Intermediate Range Monitor (IRM).

Evaluate the At the Controls (ATC) Operators actions to determine the cause of the 1/2 Scram, and bypasses the failed IRM. This allows resetting the half-scram.

Evaluate the CRS addressing Technical Specifications for the failed IRM.

Success Path:

3

Appendix D Scenario Outline Attachment 2 IRM - C is bypassed, and the CRS initiates an LCO on IRM C in accordance with Technical Specifications 3.3.1.1 (RPS Instrumentation) Table 3.3.1.1-1 Function 1. Also TRM T3.3.1 Function 2 potential LCO with these two INOP IRMS there remains the minimum number required of 6.

Event Four: Rod Drop Malfunction Required:

CR023431 Increased Rod Worth on rod 34-31 set at 25%

RD133431 Rod Uncoupled RD123431 Rod Stuck CR01 Fuel Failure at 5%

Objective:

Evaluate the crews response to a reactivity addition and rise in reactor power.

Evaluate the crews entry into Abnormal Procedure 2.4RX-PWR.

Evaluate the crews entry into Abnormal Procedure 5.1FUEL.

Evaluate the crews entry into Abnormal Procedure 5.1RAD.

Success Path:

The Reactor is scrammed and actions are in place to drive the control rods into the core to achieve a shutdown reactor.

Event Five: ATWS non-EOP Rod driving Malfunction Required:

RD02A ATWS North Bank set at 100%

RD02B ATWS South Bank set at 75%

Objective:

Evaluate the crews response to minimal control rod insertion on a reactor scram signal.

Evaluate the ROs ability to perform 2.4CRD and drive control rods using RMCS.

Evaluate the CRS implementing strategy for reactivity controls outside the EOPs.

Evaluate the crews teamwork in installing jumpers and controlling Reactor Pressure and level.

Success Path:

Control Rods are inserted using RMCS.

4

Appendix D Scenario Outline Attachment 2 Event Six: RCIC Steam Line Leak Malfunction Required:

RC06 RCIC Steam Line Break in at 100%

RC12A RCIC Steam Isolation Valve Leakage RCIC-MO-15 in at 100%

OR RC06A RCIC-MO-16 Control Power De-energized Objective:

Evaluate the crews response to a failure of RCIC to fully isolate during a RCIC Steam Line Break.

Evaluate the BOPs ability to monitor and report Secondary Containment Temperatures and Radiation Levels to the CRS.

Evaluate the crews ability to continue Control Rod insertion in accordance with 2.4CRD and Emergency Depressurize the RPV when 2 Areas in Secondary Containment exceed Max Safe values.

Success Path:

RPV Level is being controlled within the +3 to 54 inch range with CRD and HPCI and Condensate. The Reactor is depressurized to <50 psig above Torus Pressure when two areas in Secondary Containment reach and exceed Max Safe values.

Event Seven: Group 6 Failure Malfunction Required:

RP15 Group 6 failure Objective:

Evaluate the crew recognition that the Group 6 isolation group valves have failed to isolate when Secondary Containment Vent Exhaust Rad Monitors exceed their setpoints.

Evaluate the crews ability to insert a Group 6 isolation to initiate SBGTs and isolate normal ventilation to prevent releases from the reactor building.

Success Path:

All valves in the Group 6 isolation set are closed and SGTs are started to support reactor building atmosphere control.

Event Eight: ED on Secondary Containment 2 Areas Malfunction Required:

None Objective:

Evaluate the crew response to a slow increase in Reactor Building Temperatures and 5

Appendix D Scenario Outline Attachment 2 Radiation levels to the point where the RPV must be Emergency Depressurized.

Evaluate the crews ability to manually open 6 SRVs and reduce RPV Pressure to less than 50 psig above Torus pressure, in accordance with the EOPs.

Success Path:

RPV is depressurized to 50 psig above Torus pressure.

Scenario Termination:

When Reactor water level is being controlled between +3 and +54 inches and the RPV has been Emergency Depressurized and All but one Control Rod have been inserted.

6

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: 2-28-2011 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4 *

1. 1 4 3 3 3 4 3 20 3 4 7 Emergency &

Abnormal Plant 2 1 1 1 N/A 1 1 N/A 2 7 2 1 3 Evolutions Tier Totals 5 4 4 4 5 5 27 6 4 10 1 2 2 2 3 2 2 2 3 3 3 2 26 2 3 5 2.

Plant 2 1 1 2 1 1 1 1 1 1 1 1 12 1 1 1 3 Systems Tier Totals 3 3 4 4 3 3 3 4 4 4 3 38 4 4 8

3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 Categories 3 2 2 3 10 2 2 1 2 7 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 RO E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 AA2.05 Ability to determine and/or interpret Jet 295001 Partial or Complete Loss of Forced pump operability as they apply to PARTIAL OR Core Flow Circulation / 1 & 4 X COMPLETE LOSS OF FORCED CORE FLOW 3.1 53 CIRCULATION:

AK3.06 Knowledge of the reasons for Containment 295003 Partial or Complete Loss of AC / 6 isolation as they apply to PARTIAL OR COMPLETE X 3.7 46 LOSS OF A.C.POWER 2.1.27 Knowledge of system purpose and/or 295004 Partial or Total Loss of DC Pwr / 6 X function. 3.9 56 AK2.01 Knowledge of the interrelations between 295005 Main Turbine Generator Trip / 3 X MAIN TURBINE GENERATOR TRIP and RPS 3.8 43 2.4.45 Ability to prioritize and interpret the 295006 SCRAM / 1 X significance of each annunciator or alarm. 4.1 54 AA1.02 Ability to operate and/or monitor the 295016 Control Room Abandonment / 7 reactor/turbine pressure regulating system as they X 2.9 49 apply to CONTROL ROOM ABANDONMENT:

AK2.01 Knowledge of the interrelations between 295018 Partial or Total Loss of CCW / 8 PARTIAL OR COMPLETE LOSS OF COMPONENT X 3.3 42 COOLING WATER and system loads AK2.01 Knowledge of the interrelations between 295019 Partial or Total Loss of Inst. Air / 8 PARTIAL OR COMPLETE LOSS OF INSTRUMENT X 3.8 44 AIR and CRD hydraulics AK3.03 Knowledge of the reasons for increasing 295021 Loss of Shutdown Cooling / 4 drywell cooling as they apply to LOSS OF X 2.9 45 SHUTDOWN COOLING AK1.03 Knowledge of the operational implications 295023 Refueling Acc / 8 of inadvertent criticality as they apply to X 3.7 41 REFUELING ACCIDENTS EK1.01 Knowledge of the operational implications of 295024 High Drywell Pressure / 5 drywell integrity as they apply to HIGH DRYWELL X 4.1 40 PRESSURE EA2.03 Ability to determine and/or interpret 295025 High Reactor Pressure / 3 Suppression pool temperature as they apply to X 3.9 57 HIGH REACTOR PRESSURE:

EA2.02 Ability to determine and/or interpret 295026 Suppression Pool High Water Temp. Suppression pool level as they apply to

/5 X 3.8 51 SUPPRESSION POOL HI WTR TEMPERATURE:

295027 High Containment Temperature / 5 EK1.01 Knowledge of the operational implications of 295028 High Drywell Temperature / 5 the reactor water level measurement as they apply X 3.5 39 to HIGH DRYWELL TEMPERATURE EA1.05 Ability to operate and/or monitor HPCI as 295030 Low Suppression Pool Wtr Lvl / 5 they apply to LOW SUPPRESSION POOL WATER X 3.5 50 LEVEL:

EK1.03 Knowledge of the operational implications of 295031 Reactor Low Water Level / 2 water level effects on reactor power as they apply to X 3.7 58 REACTOR LOW WATER LEVEL:

EK3.03 Knowledge of the reasons for lowering 295037 SCRAM Condition Present reactor water level as they apply to SCRAM and Reactor Power Above APRM CONDITION PRESENT AND REACTOR Downscale or Unknown / 1 X 4.1 47 POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

2.2.44 Ability to interpret control room indications to 295038 High Off-site Release Rate / 9 verify the status and operation of a system, and X understand how operator actions and directives 4.2 55 affect plant and system conditions.

AA2.05 Ability to determine and interpret Ventilation 600000 Plant Fire On Site / 8 alignment necessary to secure affected area as they X 2.9 52 apply to PLANT FIRE ON SITE:

AA1.01 Ability to operate and/or monitor Grid 700000 Generator Voltage and Electric Grid frequency and voltage as they apply to Disturbances / 6 X GENERATOR VOLTAGE AND ELECTRIC GRID 3.6 48 DISTURBANCES:

K/A Category Totals: 4 3 3 3 4 3 Group Point Total: 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 E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 AK2.08 Knowledge of the interrelations between LOSS 295002 Loss of Main Condenser Vac / 3 OF MAIN CONDENSER VACUUM and the Condenser X 3.1 60 circulating water system 295007 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 / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 AK3.01 Knowledge of the reasons for bypassing rod 295015 Incomplete SCRAM / 1 insertion blocks as they apply to INCOMPLETE X 3.4 61 SCRAM 295017 High Off-site Release Rate / 9 AK1.04 Knowledge of the operational implications of 295020 Inadvertent Cont. Isolation / 5 & 7 Bottom head thermal stratification as they apply to X 2.5 59 INADVERTENT CONTAINMENT ISOLATION:

AA2.01 Ability to determine and/or interpret 295022 Loss of CRD Pumps / 1 accumulator pressure as they apply to LOSS OF CRD X 3.5 63 PUMPS:

2.4.1 Knowledge of EOP entry conditions and 295029 High Suppression Pool Wtr Lvl / 5 X immediate action steps. 4.6 65 EA1.05 Ability to operate and/or monitor affected 295032 High Secondary Containment systems so as to isolate damaged portions as they Area Temperature / 5 X apply to HIGH SECONDARY CONTAINMENT AREA 3.7 62 TEMPERATURE:

2.4.9 Knowledge of low power/shutdown implications 295033 High Secondary Containment in accident (e.g., loss of coolant accident or loss of Area Radiation Levels / 9 X 3.8 64 residual heat removal) mitigation strategies.

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 Conc. / 5 K/A Category Point Totals: 1 1 1 1 1 2 Group Point Total: 7

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 A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 K5.02 Knowledge of the operational 203000 RHR/LPCI: Injection implications of core cooling methods as they Mode X 3.5 9 apply to RHR/LPCI INJECTION MODE.

K6.04 Knowledge of the effect that a loss or malfunction of reactor water level will have on 205000 Shutdown Cooling X the SHUTDOWN COOLING SYSTEM (RHR 3.6 11 SHUTDOWN COOLING MODE):

A1.08 Ability to predict and/or monitor 206000 HPCI changes in parameters associated with X operating the HIGH PRESSURE 4.1 14 COOLANT INJECTION SYSTEM controls including system lineup 207000 Isolation (Emergency)

Condenser 207000 Isolation (Emergency)

Condenser K4.01 Knowledge of LOW PRESSURE 209001 LPCS CORE SPRAY SYSTEM design feature(s)

X and/or interlocks which provide for prevention 3.2 7 of over pressurization of core spray piping 209002 HPCS A4.07 Ability to manually operate and/or 211000 SLC X monitor in the control room: Lights and alarms 3.6 23 A2.08 Ability to (a) predict the impacts of Low reactor level on the REACTOR PROTECTION SYSTEM ; and (b) based on 212000 RPS X those predictions, use procedures to correct, 4.1 15 control, or mitigate the consequences of those abnormal conditions or operations:

K5.01 Knowledge of the operational 215003 IRM implications of detector operation as they X apply to INTERMEDIATE RANGE MONITOR 2.6 10 (IRM) SYSTEM:

K2.01 Knowledge of electrical power supplies 215004 Source Range Monitor X to the SRM channels/detectors 2.6 4 K3.07 Knowledge of the effect that a loss or malfunction of the AVERAGE POWER 215005 APRM / LPRM X RANGE MONITOR/LOCAL POWER RANGE 3.2 5 MONITOR SYSTEM will have on the rod block monitor:

K3.02 Knowledge of the effect that a loss or 217000 RCIC malfunction of the REACTOR CORE X ISOLATION COOLING SYSTEM (RCIC) will 3.6 6 have on Reactor vessel pressure A3.04 Ability to monitor automatic operations 217000 RCIC of the REACTOR CORE ISOL COOLING X 3.6 18 SYSTEM (RCIC) including, system flow.

2.1.19 Ability to use plant computers to 218000 ADS X evaluate system or component status. 3.9 22 A4.02 Ability to manually operate and/or 223002 PCIS/Nuclear Steam monitor in the control room: Manually initiate Supply Shutoff X 3.9 19 the system 2.4.21 Knowledge of the parameters and logic 239002 SRVs used to assess the status of safety functions, such as reactivity control, core cooling and X heat removal, reactor coolant system 4.0 21 integrity, containment conditions, radioactivity release control, etc.

A2.03 Ability to (a) predict the impacts of 239002 SRVs Stuck open SRV on the RELIEF/SAFETY VALVES; and (b) based on those predictions, X use procedures to correct, control, or mitigate 4.1 26 the consequences of those abnormal conditions or operations:

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 A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 A4.07 Ability to manually operate and/or 259002 Reactor Water Level monitor in the control room: All individual Control X component controllers when transferring from 3.8 20 automatic to manual mode K1.12 Knowledge of the physical connections 261000 SGTS and/or cause/effect relationships between X STANDBY GAS TREATMENT SYSTEM and 3.1 2 the primary containment purge system:

A2.04 Ability to (a) predict the impacts of 261000 SGTS High train moisture content on the STANDBY GAS TREATMENT SYSTEM; and (b) based X on those predictions, use procedures to 2.5 16 correct, control, or mitigate the consequences of those abnormal conditions or operations:

262001 AC Electrical K2.01 Knowledge of electrical power supplies X to the following: Off-site sources of power 3.3 3 Distribution K4.03 Knowledge of A.C. ELECTRICAL 262001 AC Electrical DISTRIBUTION design feature(s) and/or Distribution X interlocks which provide for the interlocks 3.1 25 between automatic bus transfer and breakers K4.01 Knowledge of UNINTERRUPTABLE 262002 UPS (AC/DC) POWER SUPPLY (A.C./D.C.) design X feature(s) and/or interlocks which provide for 3.1 8 transfer from preferred power to alternate power supplies.

K1.03 Knowledge of the physical connections 263000 DC Electrical and/or cause/effect relationships between Distribution X D.C. ELECTRICAL DISTRIBUTION and 2.6 1 battery ventilation:

A1.04 Ability to predict and/or monitor 264000 EDGs changes in parameters associated with X operating the EMERGENCY GENERATORS 2.6 13 (DIESEL/JET) controls including crank case temperature and pressure A3.02 Ability to monitor automatic operations 300000 Instrument Air of the INSTRUMENT AIR SYSTEM including X 2.9 17 air temperature K6. 07 Knowledge of the effect that a loss or 400000 Component Cooling malfunction of Breakers, relays, and Water X 2.7 12 disconnects will have on the CCWS A3.01 Ability to monitor automatic operations 400000 Component Cooling of the CCWS including: Setpoints on Water instrument signal levels for normal operations, X 3.0 24 warnings, and trips that are applicable to the CCWS A2.04 400000 Component Cooling X 2.9 16 Water K/A Category Point Totals: 2 2 2 3 2 2 2 3 3 3 2 Group Point Total: 26

ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS K4.08 Knowledge of CONTROL ROD 201003 Control Rod and Drive 2.6 30 Mechanism AND DRIVE MECHANISM design X feature(s) and/or interlocks which provide for the following: Monitoring CRD mechanism temperature 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation K2.02 Knowledge of electrical power 202002 Recirculation Flow Control 2.6 28 X supplies to the following: Hydraulic power unit:

204000 RWCU K6.01 Knowledge of the effect that a loss 214000 RPIS 2.5 32 or malfunction of the A.C. electrical X power will have on the ROD POSITION INFORMATION SYSTEM:

A2.07 Ability to (a) predict the impacts of 215001 Traversing In-core Probe 3.4 34 failure to retract during accident conditions on the TRAVERSING IN-X CORE PROBE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations 215002 RBM K3.29 Knowledge of the effect that a loss 216000 Nuclear Boiler Inst. 3.1 38 or malfunction of the NUCLEAR BOILER X Instrumentation will have on Jet pump flow monitoring:

K3.01 Knowledge of the effect that a loss or malfunction of the RHR/LPCI: TORUS 219000 RHR/LPCI: Torus/Pool Cooling / SUPPRESSION POOL COOLING X 3.9 29 Mode MODE will have on Suppression pool temperature control 223001 Primary CTMT and Aux.

226001 RHR/LPCI: CTMT Spray Mode K1.01 Knowledge of the physical 230000 RHR/LPCI: Torus/Pool Spray 3.6 27 Mode connections and/or cause/effect relationships between RHR/LPCI:

X TORUS/SUPPRESSION POOL SPRAY MODE and the following: Suppression pool K5.07 Knowledge of the operational 233000 Fuel Pool Cooling/Cleanup 2.5 31 implications of the Maximum (abnormal)

X heat 102d load as they apply to FUEL POOL COOLING AND CLEAN-UP:

234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control A1.06 Ability to predict and/or monitor 241000 Reactor/Turbine Pressure 3.2 33 Regulator changes in parameters associated with operating the REACTOR/TURBINE X PRESSURE REGULATING SYSTEM controls including: Main turbine steam flow 3.2 245000 Main Turbine Gen. / Aux.

A3.05 Ability to monitor automatic 256000 Reactor Condensate 3.0 35 operations of the REACTOR X CONDENSATE SYSTEM including:

Lights and alarms

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 RO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 259001 Reactor Feedwater 268000 Radwaste 271000 Offgas 272000 Radiation Monitoring 286000 Fire Protection 288000 Plant Ventilation A4.04 Ability to manually operate and/or 290001 Secondary CTMT 2.6 36 X monitor in the control room: Auxiliary building area temperature:

290003 Control Room HVAC 2.2.22 Knowledge of limiting conditions 4.0 37 X

for operations and safety limits.

290002 Reactor Vessel Internals K/A Category Point Totals: 1 1 2 1 1 1 1 1 1 1 1 Group Point Total: 12

S-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 E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 295001 Partial or Complete Loss of Forced Core Flow Circulation / 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 AA2.05 Ability to determine and/or interpret the 295023 Refueling Acc / 8 entry conditions of emergency plan as they apply to X 4.6 77 REFUELING ACCIDENTS.

2.4.6 Knowledge of EOP mitigation strategies.

295024 High Drywell Pressure / 5 X 4.7 80 EA2.03 Ability to determine and/or interpret 295025 High Reactor Pressure / 3 suppression pool temperature as they apply to X 4.1 78 HIGH REACTOR PRESSURE:

EA2.03 Ability to determine and/or interpret the 295026 Suppression Pool High Water Reactor Pressure as they apply to SUPPRESSION Temp. / 5 X 4.0 76 POOL HIGH WATER TEMPERATURE 2.1.27 Knowledge of system purpose and/or 295027 High Containment Temperature / 5 X function 4.0 81 295028 High Drywell Temperature / 5 295030 Low Suppression Pool Wtr Lvl / 5 2.4.3 Ability to identify post-accident 295031 Reactor Low Water Level / 2 X instrumentation. 3.9 82 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 2.2.42 Ability to recognize system parameters that 295038 High Off-site Release Rate / 9 are entry-level conditions for Technical X 4.6 79 Specifications.

600000 Plant Fire On Site / 8 700000 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 3 4 Group Point Total: 7

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 E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 295007 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 / 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 / 9 295020 Inadvertent Cont. Isolation / 5 & 7 2.2.22 Knowledge of limiting conditions for operations X 4.7 84 and safety limits.

295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Wtr Lvl / 5 EA2.01 Ability to determine and/or interpret X Suppression pool water level as they apply to HIGH 3.9 85 SUPPRESSION POOL WATER LEVEL. 3.9 295032 High Secondary Containment Area Temperature / 5 EA2.01 Ability to determine and/or interpret Area 295033 High Secondary Containment radiation levels as they apply to HIGH SECONDARY Area Radiation Levels / 9 X 3.9 83 CONTAINMENT AREA RADIATION LEVELS.

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 Conc. / 5 K/A Category Point Totals: 2 1 3

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 SRO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 203000 RHR/LPCI: Injection Mode A2.09 Ability to (a) predict the impacts of Reactor low water level on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN 205000 Shutdown Cooling X COOLING MODE); and (b) based on those 3.8 86 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

206000 HPCI 207000 Isolation (Emergency)

Condenser 209001 LPCS 209002 HPCS 211000 SLC A2.19 Ability to (a) predict the impacts of Partial system activation (half-SCRAM) on the REACTOR PROTECTION SYSTEM ; and (b) 212000 RPS X based on those predictions, use procedures 3.9 87 to correct, control, or mitigate the consequences of those abnormal conditions or operations.

215003 IRM 215004 Source Range Monitor 2.4.20 Knowledge of the operational 215005 APRM / LPRM X implications of EOP warnings, cautions, and 4.3 90 notes.

217000 RCIC 218000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff 239002 SRVs 259002 Reactor Water Level Control 261000 SGTS 262001 AC Electrical 2.2.7 Knowledge of the process for X conducting special or infrequent tests. 3.6 89 Distribution 262002 UPS (AC/DC) 263000 DC Electrical Distribution 2.2.36 Ability to analyze the effect of 264000 EDGs maintenance activities, such as degraded X power sources, on the status of limiting 4.2 88 conditions for operations.

K/A Category Point Totals: 2 3 Group Point Total: 5

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 A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-core Probe 215002 RBM 216000 Nuclear Boiler Inst.

A2.16 Ability to (a) predict the impacts of high suppression pool level on the RHR/LPCI: TORUS/SUPPRESSION 219000 RHR/LPCI: Torus/Pool Cooling POOL COOLING MODE; and (b) based X on those predictions, use procedures to 3.2 93 Mode correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.07 Ability to (a) predict the impacts of High drywell pressure on the PRIMARY CONTAINMENT SYSTEM AND 223001 Primary CTMT and Aux. X AUXILIARIES; and (b) based on those 4.3 91 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Regulator 245000 Main Turbine Gen. / Aux.

256000 Reactor Condensate 259001 Reactor Feedwater 2.1.23 Ability to perform specific system 268000 Radwaste X and integrated plant procedures during 4.4 92 all modes of plant operation. 4.4 271000 Offgas 272000 Radiation Monitoring 286000 Fire Protection 288000 Plant Ventilation

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 SRO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 290001 Secondary CTMT K/A Category Point Totals: 2 1 Group Point Total: 3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam: 2-28-2011 Category K/A # Topic RO SRO-Only IR # IR #

2.1.5 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

2.9 66

1. 2.1.17 Ability to make accurate, clear, and concise verbal 2.1.17 reports. 3.9 67 Conduct of Operations 2.1.27 2.1.27 Knowledge of system purpose and/or function. 3.9 75 Knowledge of individual licensed operator responsibilities related 2.1.4 to shift staffing, such as medical requirements, no-solo 3.8 94 operation, maintenance of active license status, 10CFR55, etc.

2.1.32 Ability to explain and apply system limits and precautions. 4.0 99 Subtotal 3 2 2.2.14 Knowledge of the process for controlling equipment 2.2.14 configuration or status. 3.9 69 Ability to perform pre-startup procedures for the facility, including 2.2.1 operating those controls associated with plant equipment that 4.4 95

2. could affect reactivity.

Equipment Control 2.2.43 Knowledge of the process used to track inoperable alarms. 3.3 98 2.2.36 Ability to analyze the effect of maintenance activities, such 2.2.36 as degraded power sources, on the status of limiting conditions 3.1 68 for operations.

Subtotal 1 2 2.3.11 Ability to control radiation releases.

2.3.11 3.8 70 2.3.4 Knowledge of radiation exposure limits under normal or 2.3.4 emergency conditions. 3.2 71

3. Knowledge of radiological safety procedures pertaining to Radiation licensed operator duties, such as response to radiation monitor Control 2.3.13 alarms, containment entry requirements, fuel handling 3.8 96 responsibilities, access to locked high-radiation areas, aligning filters, etc.

Subtotal 2 1 2.4.43 Knowledge of emergency communications systems and 2.4.43 techniques. 3.2 72 2.4.22 Knowledge of the bases for prioritizing safety functions

4. 2.4.22 during abnormal/emergency operations. 3.6 73 Emergency 2.4.32 Knowledge of operator response to loss of all Procedures / 2.4.32 annunciators. 3.6 74 Plan Knowledge of procedures relating to a security event (non-2.4.28 safeguards information).

4.1 97 Knowledge of the organization of the operating procedures 2.4.5 network for normal, abnormal, and emergency evolutions. 4.3 100 Subtotal 3 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 2/1 207000.A3.02 Cooper does not have an isolation condenser 2/1 207000.K3.02 Cooper does not have an isolation condenser 2/1 209002.A2.08 Cooper does not have a high pressure core spray system 2/1 2.2.3 Cooper is not a multi-unit facility 2/2 201004.A2.01 Cooper no longer uses a Rod Sequence Control System 1/1 295027.EK1.02 Cooper has a Mark I containment design not a Mark III.

This Generic K/A deals with the effects of maintenance on the 295038 Generic 1/1 status of LCOs - A psychometrically valid question could not 2.2.36 be developed based on 10CFR55.41.

3 2.2.4 Cooper is not a multi-unit facility Replaced Generic K/A with a randomly selected A2 K/A to fill 2/2 2.4.45 all blocks on form 401-1.

Could not come up with a psychometrically valid question for 2/1 400000.A2.04 this K/A, Also CCW was sampled three times in this group.

Randomly selected another system.